ML20214T374
| ML20214T374 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/01/1987 |
| From: | Fredrickson P, Garner L, Mellen L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214T346 | List: |
| References | |
| TASK-2.K.3.27, TASK-TM 50-324-87-11, 50-325-87-11, IEB-79-27, NUDOCS 8706100228 | |
| Download: ML20214T374 (15) | |
See also: IR 05000324/1987011
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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 STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos. 50-325/87-11 and 50-324/87-11
Licensee: Carolina Power and Light Company
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P. O. Box 1551
Raleigh, NC 27602
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Docket Nos.: 50-325 and 50-324
License Nos.: OPR-71 and DPR-62
Facility Name: Brunswick 1 and 2
Inspection Conducted: April 4 - May 4, 1987
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Inspectors:
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SlM M
W. H. RDland
Date Signed
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,L. W. GW ner
Date Signed
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L. S. MeVlen
Date Sfgned
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Visit by:
G.
well
Senior Resident Inspector, Harris
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Approved by:
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G
P.~E. Fredricksor., Sect 16n Chief
Date' Signed
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Division of Reactor Projects
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SUMMARY
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Scope:
This routine safety inspection involved the areas of followup on
previous enforcement matters, maintenance observation, surveillance observa-
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tion, operational safety verification, onsite Licensee Event Reports (LER)
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review, followup on inspector identified and unresolved items, IE Bulletin
followup, onsite followup of events, refueling activities, four hour reports,
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environmental qualifications of electrical _ equipment, visit by back-up resident
inspector, HPCI min-flow valve gearing, and abnormal liquid effluent release.
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Results:
One violation was identified - failure to make a timely four hour
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report, (Paragraph 12).
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0706100220 870601
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ADOCK 05000324
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REPORT DETAILS
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
G. Oliver, Manager - Site Planning and Control
J. Holder, Manager - Outages
R. Eckstein, Manager - Technical Support
E. Bishop, Manager - Operations
L. Jones, Director - Quality Assurance (0A)/ 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
K. Enzor, Director - Regulatory Compliance
R. Groover, Manager - Project Construction
A. Hegler, Superintendent - Operations
W. Hogle, Engineering Supervisor
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B. Wilson, Engineering Supervisor
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B. Parks, Engineering Supervisor
R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)
W. Dorman, Supervisor - QA
W. Hatcher, Supervisor - Security
R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
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C. Treubel, Mechanical Maintenance Supervisor (Unit 1)
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R. Poulk, Senior NRC Regulatory Specialist
D. Novotny, Senior Regulatory Specialist
W. Murray, Senior Engineer - Nuclear Licensing Unit
Other licensee employees contacted included construction craftsmen,
engineers, technicians, operators, office personnel, and security force
members.
2.
Exit Interview (3070 )
The inspection scope and findings were summarized on May 4, 1987, with the
general manager.
The licensee acknowledged the findings without
exception.
On May 28,1987, regional management informed P. Howe by
telephone that the failure to make a timely four-hour report (see
paragraph 12) had been upgraded to a violation after management review of
the issue.
The licensee acknowledged the change without dissenting
comments.
The licensee did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during the inspection.
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3.
Followup on Previous Enforcement Matters (92702)
(CLOSED) Violation 325/83-26-02 and 324/83-26-02, Failure to Adequately
Address the Position of All Valve Alignments in Operating Procedure OP-41.
The inspector reviewed the revision to OP-41 to determine if the specific
valve in question had been addressed.
(CLOSED) Violations 325/85-22-02 and 324/85-22-02, Hydraulic Control Units
(HCU) Not Installed Per Drawing G.E. 9190615; and 324/85-27-01, Bolts
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Replaced on Hydraulic Control Units with Type Other Than the Type
Specified on General Electric Drawing G.E. 9190615.
The inspector
reviewed the work packages associated with these violations, conducted a
walkdown inspection of the HCU, and interviewed selected personnel
associated with the HCU and fasteners.
The inspector did not find any
discrepancies between Drawing G.E. 919D615 and the installed HCU.
(CLOSED) Violation 324/85-33-01, Failure to Maintain Valves RNA-V202 and
RNA-V203 in the Position Required by the Operating Procedure Valve Line
Up.
The inspector reviewed the documentation of the real time training,
the associated LER (2-85-05) and the revision to Operating Procedure 46.
(CLOSED) Deviation 325/84-31-03 and 324/84-31-03, Failure to Upgrade the
Chlorine Detection System as Described in Response to the Original FSAR
Appendix M, M14.5 Questions. The inspectors verified the installation and
operability of the 2 chlorine detectors mounted inside the service water
building. Additionally, the inspectors reviewed completed work package PM
86-072.
No additional violations or deviations were identified.
4.
MaintenanceObservation(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.
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The inspectors observed / reviewed portions of the following maintenance
activities:
86-HE0451
Performance of Maintenance Instruction MI-10-21,125/250
VDC Motor Control Center Breaker Functional Test, on Valve
2-E41-F007.
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87-AGUP1
Performance of MI-03-1T, Johnson Control Model P-5217-1 and
P-5217-2 Differential Pressure Transmitter, for Calibration
of Reactor Building Differential Pressure Instrument.
87-HGH151
Performance of MI-10-213, 125/250 VDC Motor Control Center
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Breaker Compartment Checkout, on Valve 2-E41-F041.
87-LTV124
Performance of OPM-BKR003, Preventive Maintenance of
General Electric 480 VAC Motor Control Center Compartment
for Valve 1-E11-F006A.
During performance of MI-10-213 on April 22, 1987, the inspector observed
that the DC compartment check did not include the same type of check of
the contactor's auxiliary contacts as being performed on those associated
with AC motor control centers.
Enhanced attention is being given to the
auxiliary contacts in the AC compartments because of a number of recent
problems which rendered valves inoperable.
The auxiliary contacts in the
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DC compartments were observed to be of a different model and type than
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those in the AC compartments. The difference between the checks performed
on the DC and AC compartment auxiliary contacts was discussed with the
manager of maintenance who indicated that the item would be reviewed.
No violations or deviations were identified.
5.
SurveillanceObservation(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; instru-
mentation 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:
Average Power Range Monitor ( APRM) Rod Block Functional
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Test.
Main Steam Line High Radiation Channel Functional Test.
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Residual Heat Removal (RHR), low Pressure Coolant Injection
(LPCI), Core Spray System (CSS), High Pressure Coolant
Injection (HPCI), and High Drywell Pressure Trip Unit
Channel Calibration.
RHR Reactor Vessel Shroud Level Trip Unit Channel Calibra-
tion.
PT-12.2C
No. 3 Diesel Generator (DG) Monthly Load Test.
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PT-12.8
Electrical Power Systems Operability Test.
On April 17, 1987, the inspector observed that procedure 2MST-RHR24R
required applying some pressures to the instrument transmitter with a
specified accuracy, e.g., more decimal places, than the digital monitoring
equipment could display.
The I&C technician performing the calibration
indicated that this had been discussed with his supervisor prior to
performing the procedure.
Apparently, calculations had shown that the
required accuracy could be obtained using less than the stated number of
significant figures.
No procedure change had been made to 2MST-RHR24R
prior to performing it on this date. A change was being considered. The
administrative handling of this type of discrepancy was discussed with the
manager of maintenance, who indicated that he would review the matter and
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take action as necessary.
No violations or deviations were identified.
6.
Operational Safety Verification (71707)
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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.
The inspectors verified that control room manning requirements of 10 CFR 50.54 and the technical specifications were met.
Control room, shift
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supervisor, clearance and jumper / bypass logs were reviewed to obtain
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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.
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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 identi-
fied by Technical Specifications, was completed; there was no leakage of
major components; there was proper lubrication and cooling 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 inspector 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 identi-
fication 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 a clearance, and
verified the operability of onsite and offsite emergency power sources.
No violations or deviations were identified.
7.
Onsite Review of Licensee Event Reports (92700)
The listed Licensee Event Reports (LERs) were reviewed to verify that the
information provided met NRC reporting requirements,
The verification
included adequacy of event description and corrective action taken or
planned, existence of potential generic problems and the relative safety
significance of the event.
Onsite inspections were performed and
cnneluded that necessary corrective actions have been taken in accordance
with existing requirements, licensee conditions and commitments.
The
following reports are considered closed.
(CLOSED) LER 1-85-56, Inadequate Surveillance Tests to Verify Slope of
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Average Power Rar.ge Monitor (APRM) Flow Bias Tnermal Trip and Upscale
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Alarm.
The inspector reviewed revision 18 of APRM channel calibration
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procedure test, PT-01.1.7PC, to insure the correct steps had been incor-
porated to verify slope of the flow bias thermal trip and upscale alarm.
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(CLOSED) LER 1-85-67, Loss of Emergency Bus E-1 During Channel Functional
Testing of the Unit 1 and 2 Common AC Emergency Buses'
Degraded Voltage
Relays.
This event was attributed to a procedural weakness with
OMST-0G24M.
The inspector reviewed revision 1 to OMST-DG24M.
This
revision provided appropriate corrective actions to preclude recurrence of
this event.
(CLOSED) LER 1-86-01, Isolation of Reactor Building Fire Hose Stations,
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On January 2,1986, at 9:00 p.m., the Unit I reactor building fire hose
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stations were rendered inoperable when the downstream isolation valve of
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the hose stations' supply deluge valve was shut to isolate a gasket leak.
Technical Specification (TS) 3.7.7.4-1 required action within one hour.
The system was inoperable for 17-1/2 hours.
The inspector has concluded
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that due to the nature of the event, in that it was licensee identified,
the corrective actions taken have precluded recurrence, and due to the
limited safety significance, no Notice of Violation should be issued.
(CLOSED) LER 2-86-10, Primary Containment Group 6 Isolation During Plant
Modification Work Due to Personnel Error. While Unit 2 was in a refueling
outage, Plant Modification PM-82-066 was in progress.
Part of PM-82-066
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involved changing the limit switch wiring configuration of the drywell
equipment drain sump outboard primary containment isolation valve.
When
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the technician lifted the lug from the switch terminal, the washer fell
off and contacted the switch terminal causing the fuse in the circuit to
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blow, which in turn caused the Group 6 isolation. The inspector reviewed
the licensee's LER information sheet that detailed actions to prevent
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recurrence, the shift foreman's log and the completed LER package.
(CLOSED) LER 2-86-12, Automatic isolation of Reactor Water Cleanup (RWCU)
Due to a Loose Electrical Contact.
The electrical contact for the RWCU
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room ambient temperature isolation instrumentation was disturbed during
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the installation of the Emergency Response Facility Information System
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(ERFIS). The inspector reviewed the completed work package, the completed
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LER package and conducted selected interviews. The information contained
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in the packages and obtained from the interviews answered all the
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inspector's questions.
No violations or deviations were identified.
8.
FollowuponInspectorIdentifiedandUnresolvedItems(92701)
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(CLOSED) Inspector Followup Item 325/82-01-01, The Apparent Inability of
the Condensing Pot to Make Up Small Leaks.
There were apparently two
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potential causes for the problems associated with the condensing pot. The
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first problem was debris in the instrument lines, installed prior
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to the problem's occurrence.
This problem was corrected by flushing the
lines.
The second problem was air entrapment in the instrument lines
after maintenance. This problem was corrected by a procedural enhancement
that ensures that the lines are full prior to the instrument line being
returned to service.
The lack of additional problems in this area is
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evidence that this has corrected the problems noted.
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(CLOSED) Inspector Followup Item 325/82-08-09
Completion of TMI Item
II.K.3.27
Common Reference Level for Vessel Level Instrumentation.
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inspector reviewed the completed Modification Package 80-180.
This
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adequately addressed the completion of THI Item II.K.3.27,
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(CLOSED) Inspector Followup Item 325/82-18-02 and 324/82-18-02, Installa-
tion of Parts Without Required Level of Documentation.
The inspector
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reviewed the correspondence between the licensee and NRC, licensee
internal investigation documentation, and interviewed selected personnel.
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The inspector has observed significant improvements in this area.
Although NRC will continue to inspect in this area, this item is closed
for tracking purposes.
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(CLOSED) Inspector Followup Item 325/82-33-01, Review of Quality Assurance
(QA) Audit of TMI Related Modifications.
The inspector reviewed the
current status of TMI related modifications and the associated QA audit.
Both of these appeared to be adequate.
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(CLOSED) Inspector Followup Item 325/84-13-03 and 324/84-13-03, Track
Licensee's Progress to Upgrade Annunciation in the Control Room to
Eliminate Unnecessary Alarms and Correct Nuisance Alarms Which Result from
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Poor Design.
The inspector reviewed portions of the Operating Instruc-
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tions and Administrative Instructions to determine if adequate corrective
actions are in place.
The inspector interviewed selected personnel, and
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reviewed the control board of Unit 2.
While there still may be room for
some improvement, the inspector found that considerable effort has been
made to eliminate excessive annunciation.
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(CLOSED) Inspector Followup Item 325/84-15-02 and 324/84-15-02, Remote
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Manual Operation of the Core Spray Minimum Flow Valves from the Control
Room.
The inspector reviewed the work package associated with the remote
manual operation of the core spray minimum flow valves and noted that
there is no documentation of the interim measures taken that were noted in
Inspection Report No. 84-15. However, the inspector verified the modifica-
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tions were completed and the balance of the closecut package was accept-
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able.
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(OPEN) Inspector Followup Item 325/84-31-01 and 324/84-31-01, Submitted
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for a TS Change for Rod Sequence Control System (RSCS).
The inspector
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reviewed the licensee's submittal for a TS change for the RSCS.
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subsequently withdrew the submittal for a TS change to revise the
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submittal.
This item will remain open pending the resolution of the TS
amendment,
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(CLOSED) Unresolved Item 325/83-42-02 and 324/83-42-02, Incorporate an
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Acceptable Means of Testing Thermocouples Within the Bounds of the
Definition of Channel Calibration.
The inspector reviewed the licensee's
" Position on instrumentation channel sensors and their testing require-
ments", which essentially states that thermocouples are inherently not
subject to drift.
The inspector discussed this issue with the manufac-
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turer, the licensee and Region 11 personnel.
Based upon the information
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reviewed for the particular thermocouples addressed by these open items,
there are no further questions at this time.
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(CLOSED)UnresolvedItem 325/85-27-01 and 324/85-27-02, Seismic Qualifica-
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tion of HCU frame.
The inspector reviewed the licensee's internal
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documentation for the interim and long term qualification of the HCU
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frames. The inspector reviewed the long term upgrade work in progress for
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Unit I and the interim qualification in place for Unit 2.
The inspector
found the work in progress and the program that has been developed
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acceptable.
The inspector will verify completion of the Unit 2 modifica-
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tions in accordance with the routine inspection program,
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(CLOSED) Special Bulletin 325/78-S8-09 and 324/78-58-09, Verify Expansion
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of Spent Fuel Storage Capacity Is Done Per Regulatory Requirements.
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inspector reviewed the TS section on spent fuel storage capacity and noted
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that, with the exception of the violation noted in Inspection Report 87-03
for Unit 2, the licensee meets the TS requirements.
The inspector
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reviewed the completed modification package.
Further inspection of spent
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fuel pool expansions will be done according to the routine inspection
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program.
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No violations or deviations were identified.
9.
IEBulletinFollowup(92703)
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(CLOSED) IE Bulletin 325/79-80-27 and 324/79-BU-27, Loss of Non-Class 1E
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Instrumentation and Control Power System Bus During Operation.
The
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inspector reviewed the licensee's response to this bulletin.
It
adequately addressed the concerns expressed in the bulletin.
In addition,
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the inspector reviewed the closcout package for this item.
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No violations or deviations were identified.
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10. Onsite followup of Events (93702)
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On April 6, 1987, at 5:20 p.m., the licensee began lowering Unit 2 reactor
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power from 100% power to allow a drywell entry to determine the source of
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increasing floor drain leakage. Recorded leak rates during the day were:
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Leak Rates in
Time, April 6
Gallons Per Minute (GPM)
12:01 a.m.
0.94
4:00 a.m.
0.74
8:00 a.m.
0.91
12:01 p.m.
1.16
4:00 p.m.
2.54
6:15 p.m.
2.87
The 6:15 p.m. reading was the highest value calculated.
Subsequent
readings decreased as reactor power decreased.
In addition, all three
drywell particulate air filters showed increasing activity levels.
At
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3:25 a.m. on April 7, the licensee made a drywell entry and found the
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following leaks:
the 2A reactor recirculation pump discharge valve,
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B32-F031A, had approximately a one gpm packing leak; both feedwater check
valves, B21-F010A and B, as well as the reactor isolation cooling
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injection system inboard containment isolation valve, E51-F007, had small
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gasket or packing leaks totaling approximately I gpm.
The licensee
decided to place the unit in cold shutdown and repair the leaks.
Upon
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completion of the repairs, reactor startup commenced at 6:28 p.m. on
April 9.
However, due to an uncoupled control rod indication on Control
Rod Drive (CRD) 30-11, rods were re-inserted. Upon successful coupling of
CRD 30-11, reactor startup commenced at 1:55 a.m. on April 10. The unit
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was returned to service that day.
No violations or deviations were identified.
11. Refueling Activities (60710)
The inspector verified that attachment No. 2 of Operating Instruction
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01-03.1, Control Operator Daily Surveillance Report, Revision 4, contained
the following appropriate TS surveillance requirements:
4.9.1.1.a. .b and
.c; 4.9.2.a.2, .3 and .4; 4.9.3; 4.9.4; 4.9.5; 4.9.8 and 4.9.9.
The
inspector verified that 01-03.1 was being implemented as required.
The
inspector independently verified from the refueling bridge that selected
fuel moves were being completed in accordance with fuel move sheets.
No violations or deviations were identified.
12. FourHourReports(93702)
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The licensee failed to make a four hour red phone report in the required
time limit.
On April 24,1987, at 3:50 a.m., while performing post-
maintenance testing on the Unit 2 HPCI system subsequent to maintenance,
the system flow controller failed downscale in the automatic mode. Thus,
the HPCI '.,ystem was unable to perform its safety function.
HPCI is
required to mitigate the consequences of certain postulated accidents,
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since HPCI is classified as an Engineered Safety Feature per Chapter 6 of
the FSAR.
Thus, failure of the system's capability to automatically
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provide flow to the reactor requires the licensee to make a four hour
report.
The operations staff failed to recognize that the report was
required.
Regulatory Compliance recognized that the event was reportable
and the report was made at 8:38 a.m. on April 24, 48 minutes late.
The above failure to meet reportability requirements is similar to an
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event described in Inspection Report No. 87-02, dated March 3,1987,
concerning failure of a HPCI inboard steam isolation valve.
In that
report, no violation was issued based on the licensee's issuance of
Non-Conformance Report NCR S-87-002.
The corrective action specified in
the NCR required a revision of RCI-06.5, NRC Reporting Requirements, and
training for operators after issuance of the procedure change.
RCI-06.5
was revised effective April 13, 1987. The training had not been completed
at the time of the April 24 event.
QA did issue NCR S-87-002A on May 5,
1987 to require the licensee to address additional corrective action for
the reportability issue.
10 CFR 50.72(b)(2)(iii)(D) requires that the licensee shall notify the NRC
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within four hours of the occurrence of any event or condition that alone
could have prevented the fulfillment of the safety function of systems
that are needed to mitigate the consequences of an accident.
Based on the similar nature of this repeat event, the failure to report
HPCI flow controller failure is a Violation:
HPCI Flow Controller Failure
Not Reported Within Four Hours (324/87-11-01).
One violation and no deviations were identified.
13. Environmental Qualification of Electrical Equipment
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Failure to Replace EQ Breaker
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During a review of similar maintenance, the licensee discovered that
the breaker for valve 2-E11-F008, RHR Shutdc
voling Outboard
Suction Valve had not been replaced af ter inwrruption of a short
circuit current, as required by the Qualification Data Package (QDP).
00P-67, Revision 2, 4/6/87, for GE ICC 7700 Series Motor Control
Center.
Qualification Maintenance Requirement Activity 2 requires
the replacement of the breaker after interrupting short circuit
current.
On December 4, 1985, the F008 motor failed and the breaker
tripped while interrupting a short circuit as documented in work
request 85-AHCD1.
The licensee determined that this was the only
example where a breaker opened on a short circuit and failed to be
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Failure to replace the breaker has little safety significance.
The
valve is shut and the breaker open during operational conditions 1,
2, and 3 and is verified such by PT 2.2.4.a every 31 days. Operating
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procedures and isolation instrumentation require that reactor coolant
system pressure be less than 140 psig prior to opening the valve.
Thus, an accident occurring while the F008 was open wts remote. The
breaker was replaced on April 10, 1987, under work request 87-ALHC1.
The licensee has taken issue with GE's position that tha brea' er be
replaced after one short circuit interruption. GE verbally sL'.ed to
the licensee that the GE required breaker replacement, was based on
the uncertain effect of the fault current on the breaker.
The
licensee evaluated and documented this recommendation in Engineering
,
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Evaluation Report (EER) 87-0152, approved April 22, 1987.
In the
EER, the licensee concluded that the breaker replacement requirement
was unjustified and without technical basis. This conclusion by the
licensee was based on:
the requirement is not an industry standard
and the breakers are Underwriters Laboratory (UL) approved and
tested.
The licensee now plans to revise QDP-67 to require inspec-
tions and test of the circuit breakers after interruption of a short
circuit current.
Failure to replace the breaker for valve 2-E11-F008 is a failure to
accomplish work in accordance with documented instructions and,
therefore, is a violation ~
However, the problem was licensee
identified and met the other requirements of 10 CFR 2, Appendix C,
Section V.A; therefore, no Notice of Violation will be issued.
The
inspector reviewed EERs 87-0152 and 87-0140 that were issued
documenting the event and the licensee's corrective actions and
,
resolutions. The inspector has no further questions et this time,
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b.
SBGT System Accident Radiation Levels
The licensee discovered that the radiation levels assumed for tb
SBGT system post-accident failed to account for the radiation from
the gas and particulate within the train itself 5 QDP-68 for the SBGT
system assumed only an integrated dose of 1X10
rads whif e FSAR
section 15.6.4.5.3 specified an integrated dose of 4x10
rads
post-LOCA for the fan motor.
In EER-87-0133, the licensee performed
additional calculations of integrated dose and reviewed
existing
qualification data to show that the SBGT systems were environmentally
qualified. The inspector performed a preliminary review of the EER
and had no questions; however, this item will remain unresolved
pending review by regional based inspectors for technical adequacy
and potential violations of 10 CFR 50.49:
3
Basis Rad Levels
Incorrectly Assembled
(325/87-11-05 and
l
324/87-11-05).
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c.
Unit 1 and Unit 2 HPCI Condensate Float Switches
The licensee identified that the HPCI condensate float switches had
no Qualification Data Package.
The switches are connected to
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the same circuit as the HPCI start circuit. The licensee states that
the switches provide no safety function (annunication and start of
gland steam exhaust condensate pump), but the switches must be
qualified since their failure (ground) may cause a failure of the
HPCI start circuit.
The device can be qualified since it was part of the HPCI skid, which
underwent successful EQ testing by GE.
The licensee just failed to
purchase the documentation.
The licensee plans to either purchase
the documentation or separate the switches from the HPCI start
circuit.
Since this item had existed prior to the effective date of
10 CFR 50.49, this item is Unresolved pending future resolutions by
the NRC of the EQ enforcement issue:
HPCI Condensate Float Switches
Not EQ Documented (325/87-11-04 and 324/87-11-04).
14. Visit by Back-up Resident Inspector (71707)
During the week of April 21, 1987, the inspectors toured the on-site
emergency response facilities and reviewed portions of the licensee's
Plant Emergency Procedures and program.
The tours included the technical
support center, emergency operations facility, control room, emergency
diesel generator building, Unit 2 reactor building, radwaste building,
service water intake structure, and shop work areas.
The inspectors
reviewed the licensee's letters of agreement with the local town fire
departments, hospitals and designated medical doctors which make commit-
ments to provide their services as required to the Brunswick Plant.
!
No violations or deviations were identified.
<
15. HPCI Min - Flow Valve Gearing
The licensee informed the inspector on April 27, 1987 that the Unit 1 HPCI
minimum flow valve,1-E41-F012, had an undersized motor.
The licensee
discovered the problem while preparing to perform valve diagnostic testing
and motor replacement document review.
The licensee found that during
construction, plant modification 75-502 had changed gear ratios to
decrease the valve stroke time.
The motor size was not changed to
accommodate the increase torque requirements.
A similar problem exists
with the Unit 2 valve.
The valve was being tested as part of the
licensee's motor operated valve improvement program.
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The licensee performed an Engineering Evaluation to justify continued
operation with the current motor due to excessive lead time in obtaining a
new motor.
EER 87-0168 states that continued operations of Unit 2 is
based on:
HPCI will deliver rated flow to vessel with
F012 fully open
at a reactor pressure of 1105 psig (1020 psig assumed in FSAR accident
analysis); the differential pressure across the valve disk will aid valve
operation opening, while changing from injection mode to pressure control;
F012 valve closing in pressure control can be aided through HPIC flow
decrease;
and that any isolation valve functions during any postulated
HPCI line break were not affected. The inspector has no further questions
about valve operability at this time.
Failure to adequately modify the valve back during construction is a
violation of quality assurance requirements regarding design control,
However, since the licensee identified the problem through an aggressive
valve program; the failure did not render the system inoperable; the error
was reported to the resident inspectors; a detailed review showed no other
gear ration changes since construction; the problem will be corrected when
a new motor is received; and the problem could not have been prevented by
corrective action from a previous violation; therefore, no Notice of
Violation is being issued.
16. Abnormal Liquid Effluent Release (84723)
On April 26, 1987, at 10:40 a.m., the licensee released contaminated water
that entered into the service water side of the IB RHR heat exchanger.
The water had entered into the service water side of the heat exchanger
due to a mispositioned valve.
The licensee left the equalizing valve,
BX-5, of differential pressure instrument, 1-E11-PDT-N002, open upon
return to service of the heat exchanger.
The instrument measures the
differential pressure across the tubes of the heat exchanger, from the
service water side (inside tubes) to the RHR side (shell side). With the
equalizing valve open and no flow or pressure on either side of the heat
exchanger, RHR water from the pipe above the heat exchanger drained into
the tubes through the equalizing valve. At 10:30 a.m., on April 26, 1987,
the IB RHR Service Water (SW) booster pump was started. The plant was in
the refueling mode with the core refueled. The start of the IB pump sent
the RHR slug of water through the service water system and out the
discharge canal.
The service water radiation monitors on both units
spiked.
The monitors measure the service water gross activity prior to
discharge to the canal.
Unit l's rod monitor spiked from about 22 counts
per second (cps) to about 500 cps. The Unit 1 flow and spike was through
a partially secured system.
The Unit 1 service water discharge flow was
cross connected to the Unit 2 side. The Unit 2 SW rod monitor spiked from
about 20 cps to about 1300 cps.
The licensee initiated an Operations Experience Report (0ER) on the event
concerning the mispositioned equalizing valve. The inspectors will review
the event, specifically the mispositioned valve issue, upon completion of
the OER.
The item is Unresolved:
Mispositioned Equalizing Valve for IB
.
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In addition, the inspector has requested that the regional inspectors in
the radiation effluents section to followup on the licensee's calculations
and characterizations of the abnormal release.
Based on preliminary
discussions with licensee personnel, no TS or 10 CFR discharge limits were
exceeded.
The licensee claims that approximately 3 millicuries were
released to the canal with considerable dilution.
This item will be
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followed by the radiation effluents section as an Inspector Followup Item:
Abnormal Radiation Effluent Release from IB RHR HX (325/87-11-03).
No violations of deviations were identified.
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