ML20149K703
| ML20149K703 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/30/1988 |
| From: | Fredrickson P, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148M128 | List: |
| References | |
| 50-324-88-05, 50-324-88-5, 50-325-88-05, 50-325-88-5, NUDOCS 8804050354 | |
| Download: ML20149K703 (14) | |
See also: IR 05000324/1988005
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STRE ET, N.W.
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ATLAN TA, GEORGI A 30323
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Report Nos.: 50-325/88-05 and 50-324/88 05
Licensee:
Carolina Power and Light Company
P. O. Box 1551
Raleigh, NC 27602
Docket Nos.:
50-325 and 50-324
License Nos.:
OPR-71 and DPR-62
Facility Name:
Brunswick 1 and 2
Inspection Conducted:
February 1-29, 1988
Inspector:.
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W. H. RJ&nd
Date Signed
Accompanying Personnel:
S. M. Shaeffer
Approved b :
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/ P. E. Fredrickson, Section Chief
Date Signed
Division of Reactor Safety
SUMMARY
Scope: This routine safety inspection by the resident inspector involved the
areas of followup on previous enforcement matters, maintenance observation,
surveillance observation, operational safety verification, onsite Licensee
Event Reports (LER) review, Q-List Review concerns, Diesel Generator air system
seismic qualification, silicon bronze bus bar bolts, hydrogen leak / unusual
event, ESF System walkdown, inerting line failure, and plant modifications.
Results:
In the areas inspected, no programmatic weaknesses, violations or
deviations were identified.
Two Unresolved Items were identified:
service
water system operating mode concerns, and containment atmosphere dilution
system discrepancies.
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8004050354 880330
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ADOCK 05000324
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REPORT DETAILS
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1.
Persons Contacted
Licensee Employees
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- B. Altman, Principal Engineer
- E. Bishop, Manager - Operations
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- J. Brown, Res. Engineer - Engineering
- S. Callis, Onsite Licensing Engineer - Licensing & Nuclear Fuel
- G. Cheatham, Manager - Environmental & Radiation Control
R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
- C. Dietz, General Manager - Brunswick Nuclear Project
W. Dorman, Supervisor - QA
- R. Eckstein, Manager - Technical Support
- K. Enzor, Director - Regulatory Compliance
- W. Hatcher, Supervisor - Security
A. Hegler, Superintendent - Operations
- R. Helme, Director - Onsite Nuclear Safety - BSEP
J. Holder, Manager - Outages
- P. Howe, Vice President - Brunswick Nuclear Project
"L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)
R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
- G. Oliver, Manager - Site Planning and Control
- J. O'Sullivan, Man >ger - Maintenance
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B. Parks, Engineering Supervisor
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- R. Poulk, Senior NRC Regulatory Specialist
- J. Smith, Manager - Administrative Support
R. Warden I&C/ Electrical Maintenance Supervisor (Unit 1)
D. Warren, Acting Engineering Supervisor
B. Wilson, Engineering Supervisor
- T. Wyllie, Manager - Engineering and Construction
Other licensee employees contacted included construction craftsmen,
engineers, technicians, operators, of fice personnel, and security force
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members.
- Attended the exit interview
Joy Industrial Equipment Company
Ralph L. Susey, Manager, Administration and Development
2.
Exit Interview (30703)
The inspection scope and findings were summarized on March 4,1988, with
those persons indicated in paragraph 1.
The inspectors described the
areas inspected and discussed in detail the inspection findings listed
below.
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Dissenting comments were not received from the licensee.
Proprietary
information is not contained in this report.
Item Number
Description / Reference Paragraph
325/88-05-01 &
"URI - Service Water System Operating Mode
324/88-05-01
Concerns (paragraph 8.b).
325/88-05-05 &
URI - CAD System Discrepancies (paragraph 12).
324/88-05-05
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325/88-05-02 &
IFI - DG Air Compressor Seismic Testing Performed
324/88-05-02
Without Compressor Running (paragraph 9).
325/88-05-03 &
IFI - Sand Introduced Into Torus / Vacuum Breakers
324/88-05-03
from Failed Inerting Line (paragraph 13).
325/88-05-04 &
IFI - Hydrogen Leak in Turbine Building Pipe
324/88-05-04
Tunnel (paragraph 11).
Note: Acronyms and abbreviations used in the report are listed in para-
graph 15.
3.
Followup on Previous Enforcement Matters (92702)
(CLOSED) Violation 324/87-03-01, Failure to Maintain Fuel Pool Storage
Capacity Within Technical Specification Limits.
The inspector revievied the licen:ee's response dated April 29, 1987. The
licensee has completed modification work (PM-83-004) involved with the
installation of the Unit 2 high density fuel storage racks. A request for
license amendment was submitted on June 23, 1987, regarding number of
assemblies allowed in the spent fuel pool for both units.
No significant safety matters, violations, or deviations were identified.
4.
Maintenance Observation (62703)
The inspectors observed maintenance activities, interviewed personnel, 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 compon-
ents were operable; administrative controls were followed; tagouts were
'An Unresolved Item is a matter about which more information is required to
determine whether it is acceptable or may involve a violation or deviation.
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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
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implemented. The inspectors independently verified that selected equip-
ment 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:
88-ABJN1
Installation of Snubber 2-E51-40SS84.
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88-BAPH1
Installation and Removal of MSIV Pit HVAC Discharge
Check Valve 1-VA-CV-RB.
SP-87-082, Rev. O
Temporary Power Package for 2-832-F0238 Suction Valve.
SP-88-006, Rev. O'
Reactor Vessel Water Level Adjustment for Repair of 12
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Inch Recirculation Nozzle.
No significant safety matters, violations, or deviations were identified.
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5.
Surveillance Observation (61726)
The inspectors observed surveillance testing required by Technical
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Specifications.
Through observation, interviews, and record review, the
inspectors verified that:
tests conformed to Technical Specification
requirements; administrative controls were followed; personnel were
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qualified; instrumentation was calibrated; and data was accurate and
complete. The inspectors independently verified selected test results and
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proper return to service of equipment.
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The inspectors witnessed / reviewed portions of the following test activi-
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ties:
01-3.1
Unit 1 Control Operator Daily Surveillance.
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01-3.2
Unit 2 Control Operator Daily Surveillance.
Other surveillance activities were observed or reviewed throughout the
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reporting period.
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No significant safety matters, violations, or deviations were identified.
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6.
Operational Safety Verification (71707)
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The inspectors verified that Unit 1 and Unit 2 were operated in compliance
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with Technical Specifications and other regulatory requirements by direct
observations of activities, facility tours, discussions with personnel,
reviewing of records and independent verification of safety system status.
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The inspectors verified that centrol room manning requirements of 10 CFR 50.54 and the Technical Specifications were met. Control operator, shif t
supervisor, clearance, STA, daily and standing instructions, 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 Opera-
tions. Direct observations were conducted of control room panels, instru-
mentation and recorder traces important to safety to verify operability
and that operating parameters were within Technical Specif uation 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 division was verified
weekly by insuring that:
each accessible valve in the flow path was in
its correct position; each power supply and breaker was closed for compon-
ents that must activate upon initiation signal; the RHR subsystem cross-
tie valve for each unit was closed with the power removed from the valve
operator; 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 observation of HP practices and 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
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protected area; vehicles were properly authorized, searched and escorted
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within the PA; persons within the PA displayed photo identification
badges; personnel in vital areas were authorized and effective compen-
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satory measures were employed when required.
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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.
The inspector toured the Unit 1 torus and reviewed the results of the
newly implemented AI-96, Rev. O, Drywell Closecut, for Unit 1 drywell
prior to Unit I startup after a maintenance outage. The licensee resolved
all problems identified during the inspection prior to startup.
Problem
resolution, per procedure, was presented to PNSC for approval .
The
problems found included a misaligned RWCU snubber pipe clamp (1-G31-1553)
and sand in the torus inerting penetration (see paragraph 8). Maintenance
replaced the snubber and included the problem as caused by a system
transient in the RWCU system. The system had been placed back in service
during unit operation and the gasket on the 1A F/0 had failed. This event
was reported in LER 88-04. Any followup on the event will be inspected as
part of the LER closecut,
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No significant safety matters, violations, or deviations were identified.
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7.
Onsite Review of Licensee Event Reports Unit 2(92700)
The listed LERs were reviewed to verify that the information provided met
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NRC reporting requirements.
The verification included adequacy of event
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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 concluded that necessary corrective
actions have been taken in accordance with existing requirements, licensee-
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conditions and commitments.
The following reports are considered closed.
(CLOSED) LER 2-86-08, Automatic Closure of Reactor Water Cleanup System
Due to Erroneous Area High Temperature Signal. The licensee replaced the
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failed module and calibrated the new module as per MI-03-03.
The
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inspector reviewed the licensee's TS compliance and the completed work
package.
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(CLOSED)
LER 2-86-15, Autoactuation of Control Building Emergency Air
' Filtration Trains 2A and 2B.
The cause for the intermittent system
actuation was due to failure of an area radiation monitor / indicator trip
unit for the Control Building common air intake supply duct. The licensee
replaced the failed part and returned the unit to service. The licensee
is also submitting a revision to the LER to include specific component
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information on the failed part as pertaining to NPRDS reportability
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requirements.
The inspector reviewed the Instrument and Calibration
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Failure Analysis and the completed work package.
No significant safety matters, violations, or deviations were identified.
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8.
Q-List Review Concerns (71707)
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Sargent & Lundy raised questions by letter dated February 3,
1988,
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concerning the classification of certain relays in safety related
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circuits. S&L was under contract with the licensee to develop a Q-list to
the component level.
Two potentially significant issues were raised:
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a.
80P-27-1 Undervoltage Relay
S&L postulated a failure of a normally energized undervoltage relay
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(27-1) on each non-safety related bus feeding each emergency bus.
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Should the relay fail to drop out or certain contacts fail to open,
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the core spray pump and RHR pump on the associated bus would start
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simultaneously (block start) vice sequential start during a LOCA with
a loss of off-site power.
The licensee concluded that startup of Unit I was allowed because
their preliminary review of the design showed no common mode failure
and that the design met the originci design basis.
The inspector
concluded that no immediate safety issue existed since the licensee
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had tested the diesel generators with block starting the LOCA loads
during startup testing.
The inspector reviewed Pre-Operational Test
PO-47, Integrated ECCS Test, performed September 4,
1976, that
-documented that voltage and frequency on the E busses met the Regula-
tory Guide 1.9, March 10,1971, voltage and frequency requirements
during the starting transient. Also, the inspector verified that the
27-1 relays in the BOP busses appeared installed in-a satisfactory
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manner and that the DG overcurrent relay with voltage restraint
appeared correctly set.
By letter dated February 19, 1988, UE&C
informed the licensee that:
(1)
the 27-1 relays were considered safety related in the original
plant design,
(2)
relays were purchased to the same specification as other safety
related relays,
(3) both the safety related and BOP switch gear were purchased to
the same specification,
(4)
these relays were listed in FSAh table 8.3.1-17, including their
location, showing review by NRC.
Based on the documents provided to the inspector and further inter-
views with plant personnel, the inspector has no further questions at
this time. A Licensee Identified Item:
BOP 27-1 Relays Treated as
Safety Related (325/88-05-06 and 324/88-05-06), will be opened and
closed for documentation.
See paragraph 14 for additional inspection performed in the area.
b.
Service Water V106 Valve
S&L also identified four "non-safety" relays whose failure would
prevent 1-SW-V106 or 2-SW-V106, SW Header to Reactor Building Closed
Cooling Water Heat Exchangers Primary Isolation Valve, from closing
during an accident. The four relays fed the loss of off-site power
signal to the V106 closing circuit. Normally, on a LOCA with a LOSP,
V10( receives a close signal, isolating the SW to the RB CCW HXs.
However, since V106 does not have a redundant valve, its failure must
already be assumed during a LOCA with LOSP.
Therefore, the logic
failure modes assumed by S&L are bounded by a DG failure.
For
example, if DG No. 3 failed during a LOCA/LOSP, the E-3 bus would be
dead, resulting in no power to Unit 2 V106 and Nuclear Service Water
Pump 2A and Conventional Service Water Pump 2A.
This results in a
failed open V106 and only NSW pump 2B running on the Unit 2 NSW
header. The licensee has documented how they were operating the SW
pumps in Enclosure 2 to the report.
Per Enclosure 2, the licensee
would permit operation of the SW system, for example with NSW pumps
2A and 2B operable together with CSW pumps IB,1C, 2C and 28. Thus
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with a failed DG No. 3, Unit 2 V106 and NSW pump 2A, NSW pump 2B
would be required to supply SW to all four diesels (3,000 GPM) the
vital header loads (470 GPM), SW pump bearing lubrication (150 GPM)
and the RB CCW HXs (7200 GPM). This total (10,820 GPM) exceeds the
capacity of the single NSW pump 2A (8,000 GPM). All flow numbers are
taken from Table 9.2.1-1 of the FSAR.
The licensee issued a standing instruction to operations to restrict
RB CCW HX flow to 5,000 GPM during normal operation,
This flow,
together with the other flow requirements, would limit the flow on
one NSW pump to 8,620 GPM, or 7.2% above the design flow of the pump.
The licensee reported to the inspector that one NSW pump could handle
the excess flow requirements.
The licensee .has issued engineering work order (PID) to have BESU
evaluate the impact of these questions. Further, the licensee is in
the process of accepting design basis information from UE&C. Once
that information is in hand, it should be easier to explain the
design rational for the current design.
The licensee will continue
to limit flow to the RB CCW HXs to 5,000 GPM until the issue is
resolved further.
The inspector interviewed the system engineer and discussed the issue
with plant management to ascertain whether the licensee had adequ-
ately addressed the issue.
Results of the document review will be
covered in next month's Inspection Report Nos. 325,324/88-14.
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inspector also verified that the RB CCW flow transmitter had been
replaced on January 24, 1988, and it was successfully calibrated
under WR/JO 87-ANPL1. This matter remains unresolved pending further
inspector review and NRR review of the acceptability of the design
and current CP&L SW pump operating modes:
Service Water System
Operating Mode Concerns (325/88-05-01 and 324/88-05-01).
No significant safety matters, violations, or deviations were identified.
9.
Diesel Generator Air System Seismic Qualification (71707)
During a review of NUREG-1275, Volume 2, Operating Experience Feedback
Report - Air Systems Problems, the inspector came across a statement
regarding a problem identified in the report referring to Brunswick. On
pages 36 and 37, the report stated that, "as of March 1987, the
licensee... for Brunswick station could not confirm that the EDG pneumatic
control systems would continue to operate following a design basis event."
The licensee showed the inspector an internal memorandum where the
licensee agreed to supply the information to the author of NUREG-1275.
However, the licensee failed to supply the information to AE00. When
asked by the inspector, the licensee had still not obtained the informa-
tion.
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Since the question was re-asked, the licensee has completed a JC0 on the
starting air system (EER-88-0071). DG No. 4 three inch control air line
has been qualified with the other 3 DGs systems deemed qualified by
similarity until the verification can be completed by BESU.
Documentation was found on seismic testing for the air compressors but
only with the air compressors not running. No seismic qualification data
exists with the air compressors running.
The licensee rational for
continued operation included the following points:
Failure of the air compressors would not prevent DG start.
The
seismic Class I air receivers would provide the starting air.
The long term control air requirements are necessary to keep the
run/stop cylinders pressurized. Loss of air to these cylinders would
shut tne fuel racks, stopping the diesel.
Assuming a 5 psig pressure loss in the air receiver and only 1 of 2
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air receivers operable, the licensee calculated that the DG could
operate for 20.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after a simultaneous DG start and Operating
Basis Earthquake.
Damage assessment teams are directed by emergency
procedure to inspect plant equipment subsequent to an OBE (0.08 ).
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Measures could then be taken to supply air to the DG Control Air
system.
Similar air compressors have been qualified with the compressor
running.
The licensee has established an internal action item to permanently
resolve the DG air compressor issue within one year.
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The inspector will follow the licensee's complete resolution of this issue
in an Inspector Followup Item:
DG Air Compressor Seismic Testing Per-
formed Without Compressor Running (325/88-05-02 and 324/88-05-02).
A significant safety matter regarding qualification of the DG air
compressors was identified; no violations or deviations were identified.
10.
Silicon Bronze Bus Bar Bolts (93702)
The licensee founc a potential common mode failure of DC switchgear.
Multiple failures of 5/16 inch silicon bronze bolts were found in the four
site DC switchboards.
The bolts are used to make the electrical connec-
tions for the switchgear bus bars. An incorrect torque specification (18
f t.-lbs.) had been supplied by GE early in construction of the plant (in
1974).
The problem had been identified by UE&C at that time and GE
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revised the drawing, specifying 9 f t.-lbs.
Re-inspection of the switch-
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gear was done at that time. In mid 1986, the licensee discovered 1 or 2
silicon bronze bolts with cracked or detached heads in 480V and DC MCCs.
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Each bus bar connection is made with two silicon bronze bolts.
The
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licensee found only 1 or 2 bolts broken per MCC, and never found both
bolts broken at the same connection. At that time, the licensee performed
a record review of previous problems, and discovered the construction
problem reported above. They also found that the Maintenance Instruction
used to perform preventative maintenance on the MCCs did not include any
torque valve for the 5/16 inch silicon bronze bolts.
Licensee review of
old construction records during the mid 1986 time did not reveal any
correlation between construction documented torque valves with bolt
failures found.
Based on a lack of correlation and the small number of
bolt failures per MCC, the licensee elected to use MI-10-2K1 and MI-10-211
performed over the next outages to inspect and replace the silicon bronze
bolts.
The DC switchboards,1A,18, 2A, 28, were inspected this month
with over 50 damaged bolts found, including both bolts broken for the
safety related battery connections for battery 1A-2 and 2B-2.
The licensee had replaced all silicon bronze bolts on safety related
switchgear, on Unit 1, prior to startup. All Unit 2 MCCs except 2xDA and
2xDB have had the silicon bronze bolts inspected and replaced.
The two remaining switchgear's bolts will be replaced prior to startup of
Unit 2.
The inspector will continue followup on this item after the LER
is issued. NRR Vendor Branch has been informed of the event.
They have
been in contact with GE concerning the bolting material and wfil continue
to address generic aspects of the event.
The inspection was conducted through interviews with plant engineers,
review of recent and construction documentation, and in plant examination
of hardware.
A significant safety matter related to silicon bronze bolts was identi-
fied; no violations or deviations were identified.
11.
Hydrogen Leak / Unusual Event (93702)
The licensee declared an Unusual Event at 4:49 p. m on January 4,1988,
when hydrogen gas concentrations of 8% to 10% were reported in the
electrical cable tunnel and breezeway. Unit I was defueled in a refueling
outage while Unit 2 was in cold shutdown for maintenance.
The licensee
had unisolated the hydrogen gas banks about 3:00 p. m. the same day after
completion of a modification for hydrogen gas water chemistry addition.
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At 4:00 p. m. , during a hydrogen gas system walkdown in the electrical
cable tunnel between the reactor building and turbine building, a techni-
cian discovered a packing leak on a newly installed hydrogen gas valve,
0-HWCH-V012.
The hydrogen gas banks were isolated at about 4:20 p. m.
The area was evacuated and all sources of ignition were secured. Normal
ventilation was maintained to remove the hydrogen gas. HP measurements of
the hydrogen gas concentration in the breezeway were reported to the
control room as 5% hydrogen gas in the breezeway and 8% hydrogen gas in
the electrical tunnel general area, with local hydrogen gas concentrations
near the leak as 16% hydrogen gas.
After the UE declaration, the Emer-
gency Director (Shif t Operating Supervisor) found that the reports of
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hydrogen gas concentration in the breezeway and electrical tunnel were
actually in percent of the lower explosive limit of hydrogen gas (4'4) and
not actual concentrations. The leaking valve was tightened at 5:45 p. m.
The UE was terminated when no detectable hydrogen gas was found.
The inspector verified through observation during the event, subsequent
interviews and record review, that the licensee implemented the applicable
portions of their emergency plan.
Initial licensee actions included
evacuation of the area, isolation of the hydrogen tanks, stopping all hot
work, and pre-positioning the fire brigade in the Unit I cable spreading
room. Two mechanics equipped with spark-resistant tools tightened the
valve oacking nut, stopping the leak.
The licensee used PEP-010, Fires
Involving Flammable Gases, Rev. O, to respond to the event.
The inspector will review the licensee's permanent corrective action and
root cause when the OER is issued.
This is an Inspector Followup Item:
Hydrogen Leak in Turbine Building Pipe Tunnel (325/88-05-04 and 324/
88-05-04).
No significant safety matters, violations, or deviations were identified.
12.
Engineered Safety Features System Walkdown (71710)
The inspector performed an inspection of the accessible components of the
Unit 1 and 2 Containment Atmospheric Dilution System to verify system
operability.
The inspection included verification of:
integrity of the
CAD vessel and system piping up to the drywell penetrations, proper valve
alignment, condition of seismic supports and hangers, power availability
to major components and instrumentation, and proper component labeling.
Nitrogen lines through the HPCI rooms of both units were not observed
because of accessibility. The inspector reviewed the monthly surveillance
of Periodic Test Procedure PT-16.1, Rev.12, System Checklists, in OP-24,
Rev. 22, Containment Atmosphere Control System Ooerating Procedure, and
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numerous plant drawings.
The inspector found several problems that were identified late in the
reporting period.
These items will be addressed in Inspection Report
Nos. 325,324/88-14.
This item remains Unresolved:
CAD System Discre-
pancies (325/88-05-05 and 324/88-05-05).
No significant safety matters, violations or deviations were identified.
13.
Inerting Pipe Failure (93702)
The licensee found sand in both units' torus inerting lines. Unit 1 was
in cold shutdown and Unit 2 was defueled. On January 4, 1988, the under-
ground inerting line (non-safety) between the liquid nitrogen tank and the
reactor buildings cracked, removing four feet of earth that covered the
pipe. The licensee performed a temporary repair on the 8 inch pipe the
same day using a pipe clamp. On January 10, both Unit 2 reactor building-
to-torus vacuum breakers failed their local
leak rate tests.
On
February 7. maintenance personnel found sand in the bottom of the vacuum
i
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breaker pipe in Unit 2.
The sand had apparently gotten into the inerting
line and been carried by the nitrogen gas into the torus and on one side
of the butterfly valve vacuum breakers. The licensee removed the sand and
!
performed additional inspections to determine the scope of the problem.
The licensee had adjusted the stops on the vacuum breakers, re performed
the Unit 2 LLRTs, with one penetration passing and the other f ailing.
Unit 1 LLRTs were performed successfully on February 11, 1988, without
removing any sand.
The licensee replaced the carbon steel underground
inerting line with Plexco, a polyethelyene pipe that is resistant to
corrosion and brittle fracture.
The licensee is writing an OER on the
event. The inspector will review the OER after its issuance to verify the
adequacy of the licensee's corrective actions.
A regional inspector
reviewed the licensee's documentation of repair activities; that review
will be included in Inspection Report Nos. 325,324/88-14.
This is an Inspector Followup Item:
Sand Introduced Into Torus / Vacuum
Breakers f rom Failed Inerting Line (325/88-05-03 and 324/88-05-03).
No violations or deviations were identified. A safety concern regarding a
non-safety system failure affecting safety related components was identi-
fied and resolved short-term since all valves were tested satisfactorily
prior to startup.
14.
Plant Modifications - Unit 2 (37700)
Reviewed plant modification PM-86-025 which concern eliminates block
loading of emergency loads on the emergency buses if a LOCA signal is
received and offsite power is available. As of March 4, 1988, all modifi-
cation work has been completed. Acceptance testing has been completed for
core spray loops A & B and RHR loop A, and interim operability estab-
lished. Acceptance test for RHR loop B is scheduled to be completed week
of March 7, 1988.
See paragraph 8.a for a related issue.
The inspector reviewed and observed the acceptance test for PM-84-017,
)
(Rev. 7), Local Control and Circuit Enhancement for 4SOV Switchgear.
The
test was conducted in accordance with procedure, QC sign-offs were
completed as necessary, and the acceptance criteria was met.
The
responsible engineer supervised the test with operations personnel
a
operating the equipment.
No significant safety matters, violations, or deviations were identified.
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15,
List of Abbreviations for Unit 1 and 2
AE00
Office of Analysis and Evaluation of Operational Data
AI
Administrative Instruction
BESU
Brunswick Engineering Sub Unit
Balance of Plant
Brunswick Steam Electric Plant
Containment Atmospheric Dilution
Carolina Power & Light Company
CSW
Conventional Service Water
-
OC
Direct Current
Diesel Generator
Engineering Evaluation Report
Engineered Safety Feature
F/D
Filter Demineralizer
Final Safety Analysis Report
g
Acceleration of Gravity
GPM
Gallons Per Minute
Health Physics
.,VAC
Heating, Ventilating, Air Conditioning System
Heat Exchanger
Instrumentation and Control
NRC Office of Inspection and Enforcement
IFI
Inspector Followup Item
JC0
Justification for Continued Operation
LER
Licensee Event Report
Local Leak Rate Test
Loss of Coolant Accident
Loss of Off-Site Power
,
Motor Control Center
MI
Maintenance Instruction
Nuclear Plant Reliability Data System
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
Nuclear Service Water
Nuclear Regulation
Operating Experience Report
OP
Operating Procedure
Protected Area
PEP
Plant Emergency Procedure
Project Identification
Plant Modification
PNSC
Plant Nuclear Safety Committee
P0
Pre-Operational
Procedure Test
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Quality Assurance
Quality Control
Reactor Building Closed Cooling Water
S&L
Sargent & Lundy
TS
Technical Specification
Unusual Event
UE&C
United Engineers & Constructors
Unresolved Item
V
Volt
WR/JO
Work Request / Job Order