ML20154P978
| ML20154P978 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/20/1988 |
| From: | Fredrickson P, Levis W, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20154P965 | List: |
| References | |
| 50-324-88-15, 50-325-88-15, NUDOCS 8806060144 | |
| Download: ML20154P978 (20) | |
See also: IR 05000324/1988015
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report No. 50-325/88-15 and_50-324/88-15
' Licensee: Carolina Pcwer and Light Company
P. O. Box 1551
Raleigh, NC 27602
' Docket No. 50-325 and 50-324
License No. DPR-71 and=DPR-62
Facility Name: Brunswick 1 and 2
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Inspection Co uc ed: April 1 - 30, 19 8
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Inspectors:
g k.
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~ ~ bate S'igned
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W M / W ~~
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Accompanying Pe son el:. S. Schaeffer
Approved by:
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P. E. Fredrickson,3ec~tWChWf
D6te 5TgneT~
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Division of Reactor Projects
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SUMARY
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 review
comittee, in office Licensee Event Report (LER) review, followup cn inspector
identified and unresolved items, information meeting with local officials,
plant startup from refueling, and onsite followup of events.
]
Results:
One violation. was identified - Improper change in operational
condition during startup of Unit 2.
Several examples of failure to follow
procedure were included as part of the overall violation. Two unresolved items
were identified:
questions concerning procurement of comercial grade items
for safety-related applications; questions concerning qualification of
Victoreen Radiation Detector Cable.
A significant safety issue concerning cracking of silicon bronze bus bar bolts,
first identified in Inspection Report 324,325/88-05, was inspected further.
RBP98W 8888L
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REPORT DETAILS
1.
Persons Contacted
' Licensee Employees
-W. Biggs, Principal Engineer
- E. Bishop, Manager - Operations
T. Cantebury, Mechanical Maintenance Supervisor (Unit 1)
G. Cheatham, Manager - Environmental & Radiation Control
R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
C. Dietz, General Manager - Brunswick Nuclear Project
P. Dorosko, Administrative Supervisor
W. Dorman, Supervisor - QA
- R. Eckstein, Manager - Technical Support
K. Enzor, Director - Regulatory Compliance
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R. Groover, Manager - Project Construction
W. Hatcher, Supervisor - Security
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A. Hegler, Superintendent - Operations
- R. Helme, Director - Onsite Nuclear Safety - BSEP
- J. Holder, Manager - Outages
P. Howe, Vice President - Brunswick Nuclear Project
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L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)
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R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
J. McKee, Supervisor - QC
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J. Moyer, Manager - Training
G. Oliver, Manager - Site Planning and Control
- J. O'Sullivan, Manager - Maintenance
- B. Parks, Engineering Supervisor
- R. Poulk, Senior NRC Regulatory Specialist
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- A. Richards, Project Engineer - QA
- S. Scharff, Operations Engineer
- J. Simon, Operations Engineer
J. Smith, Manager - Administrative Support
V. Wagoner, Director - IPBS/Long Range Planning
R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)
D. Warren, Acting Engineering Supervisor
B. Wilson, Engineering Supervisor
- T. Wyllie, Manager - Engineering and Construction
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Other licensee employees contacted included construction craftsmen,
engineers, technicians, operators, office personnel, and security force
members.
- Attended the exit interview
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2.
ExitInterview.(30703)
The inspection scope-and findings were summarized on April 29, 1988, with
those persons indicated in paragraph 1.
The inspectors described the
areas- inspected and discussed in detail the inspection findings listed
below.
Dissenting coments were not received from the licensee.
Proprietary information is not contained in this report.
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Item Number
D e s cL p_t i o n1R e fe r_e_n c e_ Pa ra g ra p h
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324/88-15-01
VIOLATION - Improper Change In Operational
Condition (paragraph 11.b.(2)
325/88-15-02 &
- URI - Procurement of Conrnercial Grade Items
324/88-15-02
Intended for Safety-Related Applications
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(paragraph 4.a)
324/88-15-03
URI - Qualification of Victoreen Radiation
Detector Cable (paragraph 11.b.(1)
325/88-15-04 &
IFI - Failures of GE 305 Auxiliary Contact Adder
324/88-15-04
Blocks (paragraph 12.a)
325/88-15-05
IFI - Normal Position for SW-V117, Nuclear Header
to Vital Header Isolation Valve'(paragraph 4.a)
325/88-15-06 &
IFI - Silicon Bronze Bolts in Safety-Related
324/88-15-06
Switchgear(paragraph 12.b)
Note:
Acronyms and abbreviations used in the report are listed in
,
paragraph 13.
3.
Followup on Previous Enforcement Matters (92702)
(CLOSED)
Violation 325/87-20-04 and 324/87-20-04, Failure to Implement
Procedures.
The inspector reviewed the licensee's response dated
September 25, 1987.
The inspector verified that the licensee implemented
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the required corrective actions.
All procedure documentation changes as
well as procedure adequacy reviews were completed.
In addition, for
example b, of this violation, the inspector _ also interviewed selected
engineering and training personnel and concluded that appropriate emphasis
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was placed on increased procedural training in the area of OP-19, High
Pressure Coolant Injection System Operating Procedure, Section 5.3, Manual
Startup.
- An UnresoWd'Ttem 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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4.
MaintenanceObservation(62703)
The inspectors observed maintenance activities, interviewed personnel, and
reviewed records to verify that work. was conducted in accordance with
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approved procedures, Technical Specifications, and applicable industry.
codes and standards.
The inspectors also verified that:
redundant
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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.
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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
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activities
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Oil changeout on HPCI booster pump - ticket 88-QMK205.
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Troubleshooting efforts of F004,1-SW-V117, F041 valves.
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Troubleshooting efforts of Unit 1 Recirc. MG set field break'er.
a.
While watching oil changeout of Unit 1 HPCI booster pump on April 21,
1988, the inspector questioned licensee personnel as to which
documents specified the lubricant type and what controls were in
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place to ensure that the proper lubricant was used.
Th'e inspector
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found that the oil type for this component is specified in OPM -
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LUB500ATT211.
This document states that DTE 797 is the proper
lubricant with an annual changeout' requirement.
The' inspector then
verified that DTE 797 was used and traced down the purchase order of
the lubricant to determine how the lubricant was procured.
The
inspector noted that purchase order 3720588A, dated February 2, 1987,
specified that the item be procured as an "Off the Shelf" item, which
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rreant that the vendor was not required to be on the approved
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supplier's list.
The receipt inspection performed en this purchase
verified only that the part ordered matched the part received.
No
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other quality requiremer ts were specified.
ENP-42.2, Revision 1, dated January 6, 1988, Purchase Requis'ition and
Data Base Review Procedure, defines a Q-0TS (Off the Shelf) item as a
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comnercial grsde item used in safety-related applications.
ENP-42.3,
Revision 1,
aud January 6,1988, Material Engineering Evaluation
Procedure, 'tep / 2.3.4, requires that a documented evaluation, which
includes idt< *it, ing the components critical characteristics, be
perfonned, when it is decided to procure an item comercial grade
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that is. intended for safety-related applications. Step 7.2.9 of the
same procedure requires that special receipt instructions,- which
should verify the component's critical characteristics, be specified.
The inspector then questioned licensee personnel concerning why these
requirements were not followed with respect to the DTE 797 oil as no
documented engineering evaluation was found and no special receipt
inspection requirements were noted.
The licensee responded that
these requirements only went into effect in October 1987, and that
any items procured prior to this time would not have been subject to
these procedures.
However, the licensee .was able to provide
information to the inspector which showed the lubricant used in the
HPCI booster pump to be of the proper type.
The possibility exists
though, that there are commercial grade items presently in stores
intended for safety-related applications in which a proper
engineering evaluation has not been performed.
The licensee had
previously -identified programmatic procurement deficiencies in NCR
No. A-87-023 which is still ope 1.
This item is Unresolved pending-
NRC review of the licensee's corrective action for the NCR:
Procurement of Commercial Grade Items Intended for Safety-Related
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Applications,(325/88-15-02and324/88-15-02),
b.
On April 20, 1988, during maintenance surveillance testing, the
licensee experienced a failure of 1-E41-F041, the HPCI Outboard
Suppression Pool Suction Valve. Under accident conditions this valve
is required to open on low CST level or high suppression pool level
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to provide an alternate suction path for HPCI.
Troubleshooting
revealed that the series field had been shorted to ground.
No
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evidence of actuator or valve problems were found that may have
caused the failure.
The motor was then sent to Harris E&E for
analysis.
It should be noted that the licensee has experienced other DC motor
failures (see LER 1-87-23) and has committed to a followup report on
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the above LER by July 1,1988, to explain any root cause of these
motor failures. The inspectors will continue to followup on licensee
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tctivities in this area.
c.
On April 23, 1988, while performing maintenance activities on the
Nuclear Service Water Vital Header, the Vital Header Motor Operated
Isolation Valve, 1-SW-V117, failed to open when manually activated by
the control switch.
Under accident conditions this valve opens to
provide Nuclear Service Water to the Core Spray Pump room cooling
units, the RHR Pump room cooling units and to each RHR Pump Seal
Cooling Heat Exchanger.
Troubleshooting efforts by the licensee
which included motor checks and control logic checks showed a limit
switch problem.
The contacts on limit switch rotor No.1 were
cleaned after which the valve was satisfactorily cycled from the RTGB
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and also with the remote keylock switch.
The inspectors will
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followup licensee activities withlespect to their determinatdin if
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the valve should be open rather than closed, duping nonnal oMration.
141s is an Inspector Followup Item:
Nonnal Positinn for 5If117, /
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Nuclear Header to Vital Header Isolation Val 77(ry/%-154W.
No signif cant.s'afety matters, violations or devit dgns dre identified.
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5.
Surveillarge Jhservation (61726).
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The inspectors observed surveillance testing required by Technical
Specifications.
Through observation, intfrviews, and record review, the
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inspectors verified that:
tests conformed ^.w Tych leal Specification
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requirements; adQistrative1 controls were'
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qualified; lipstruuqtation was calibrated; 'an
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complete.
lhe inspectors independently verified he ected pf
esults and-
proper return to service of equipment.
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The inspectors witnessed / reviewed portions of the' following ^ t'est
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activities:
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HP'CI Operability Test
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RCIC' System Operability Tes:' N Flow Rates at 150 PSIG
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Control Rod Coupling Check and CRD Testing .'
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', .(?gquence Critical Shutdown MarginMlculation, Rev. 4
PT-14.3.1
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PT-20.8
AI,N.rosen' Backup System Operability Test
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PT-71.0
' General Atomic Stack Radiatija Monitor Channel dlibration ,,(f s
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No significant safety matters, violations, or deviations were 1/jptified.
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6.
OperationalSNetyVerification(71707)
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The inspectors perified that Unit 1 and Unit 2 were operated in ,cabYiance
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withTechnical,.SpecificationsandotherreguJatoryrequirementsbydirect
observations of lactivities, ucility tours.; discussions with personnel.
reviewingofrecordstrv}independentverificationofsafetysystemstatus,
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The inspectoh verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met.
Control operator, shift
supervisor, clearance, STA, daily and
landing instructions, and
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jumper / bypass logs were reviewed to obtait information concerning
operating trends and out of service safety systems to ensure that there
~ were no conflicts with Technical Specifications Limiti79 Conditions foh
Operat' ions .
Direct observations were conducted of co rol room panels,
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rin trtdientation and recorder traces important to safety to verify
operability and that operating parameters were within Technical
Specification limits.
The inspectors observed shift turnovers to verify
that continuity of system status was maintained. The inspectors verified
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the status of selected control room annunciators.
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Operabilit
weekly by'y of a selected Engineered Safety Featur'e division was verified-
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insuring that:
each accessible valve in the flow path was in
its correct- position; each power supply and breaker was closed' for
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components that must activate upon an initiation signal; the RHR subsystem
cross-tie valve- for cach unit was closed with the power removed from the
valve operator; there was no leakage of major components; there was proper-
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lubrication and cooling water available; and a condition did not exist
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which might prevent fulfillment of the system's functional requirements.
Instrumentation essential to system actuation or performance was verified
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operable by observing ori-scale indication and proper instrument valve
lineup, if accessible.
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The
insrctors -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.
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The inspectors votified 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 tN
protected area; vehicles were properly authorized, searched and escorted
within the PA; persons within the PA displayed photo identification
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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.
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The inspector found the limit switch cover not completely screwed down for
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valve 1-CAC-V7, the Inboard Suppression Pool Purge Exhaust Valve.
This
installation is not in accordance with the design drawings or normal
installation practices which require the cover to be tightened
sufficiently to ensure the
"0" ring is seated. The licensee has generated
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NCR A-88-012 to track and resolve this issue.
The inspector will continue
to followup in this area in future routine inspections.
No significant safety matters, violations, or deviations were identified.
7.
Onsite Review Comittee (40700)
The inspector attended -pre-startup Plant Nuclear Safety Comittee meeting
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88-047 conducted on Apdl 13, 1988.
The inspectors verified that the
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meeting was conducted in accordance with Technical Specification
requirements regarding quorum membership, review process, and personnel
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qualifications.
lieeting minutes were reviewed to confirm that
decisions /reconmendations were reflected in the minutes and followup of
-corrective actions was documented.
No significant safety matters, violations, or deviations were identified.
8.
In Office Licensee Event Report Retiew (90712).
The listed LER was 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 cf the event.
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(CLOSED) LER 1-88-09, Accidental Deenergization of Unit 2 Process Off-Gas
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Radiation Monitor and Reactor Building Ventilation Exhaust Radiation
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Monitors During Maintenance.
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No significant safety matters, violations, or deviations were identified.
9.
Followup on Inspector Identified and Unresolved Items (92701)
a.
(CLOSED)
Unresolved Item 325/87-11-02, Mispositioned Equalizing
The inspector reviewed OER-87-20,
approved on June 26, 1987, which adequately detailed the events
concerning the mispositioned valve.
The licensee has revised AI-58
and ENP-03 in order to improve clearance tag sheet verification as
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well as requiring independent verification in the acceptance test
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section of the plant modification, as opposed to BESU's past practice
of taking credit for the clearance and/or system lineup at the
completion of the unit outage. The inspector concluded that the root
causes of the event have been identified and corrected, including
action to prevent recurrence.
Since the licensee has met all the
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conditions of 10 CFR 2, Appendix C, regarding licensee identified
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violations, no notice of violation will be issued,
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b.
(CLOSED)
Inspector Followup Item 325/86-24-05 and 324/86-24-05,
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Review RHR Room Cooler Operation.
The inspector reviewed EER No.
86-0460, completed on July 24, 1987, which included detailed RHR Room
Cooler Operation Analysis.
The licensee has issued TSI-87-02, dated
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August 28,
1987, which provides
operator
instruction for
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administrative equivalent TS LC0 implementation.
The inspector has
no further concerns with this issue.
c.
(OPEN)
Inspector Followup Iten 325/86-33-01 and 324/86-33-01, IRM
Fuse Testing and Subsequent Required Modifications.
The inspector
reviewed the content and results of SP-86-068, Revision 2 (Unit 1),
which was performed on January 19, 1987.
The test results
SP-86.073 (Unit 2) proved the
concerns in GE SIL No. 445 were valid.
, was not
run because the licensee concluded that plant modifications were
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required.. To assure pre-modification operability, SP-87-014 was
performed on February 12, 1987, and verified that applicable portions
of surveillance procedures MST-IRM11W and 12W would-detect a blown -
-24 volt DC fuse (F2) in the IRM.
EWR-04527 and PID No. 5400A have
been -issued and completed to- develop the plant modification.
The
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licensee plans to modify the system by December 8, _1989.
The
inspector concluded that the _ licensee's actions to address the SIL
issues have been appropriate.
This item will remain open pending
completion of the modifications,
d.
(CLOSED)
Inspector Followup Item 325/87-03-03, Review 'of IMST-DG12R
Procedure Violation OER.
The' inspector reviewed- the licensee's
OER-87-09, dated -' April 28, 1987, which addressed the root cause of-
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'the communication failure between the I&C personnel in charge of the
test and the control operator.
The licensee concluded that the test
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director had been given too many tasks.
through 14R for Units 1 and 2 were revised on May 4,1987, to
incorporate lessons learned and a ~ redistribution of the-procedural
steps to allow verifications to be done in other locations separate
from the Control Room.
The inspector concluded that the licensee's
corrective actions were appropriate,
e.
(OPEN)
Inspector Followup Item 325/87-03-04, Inadequate Board
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Walkdown and Review.
The licensee has issued Standing Instruction
87-014, dated February 13, 1987, which requires the shift foreman or
the shift operating supervisor to walk the Control Board with the
respective control operator in order to double check RTGB indication
and enhance SR0 awareness of plant status.
The licensee has also
secured general business activity at the shift foremar, window for an
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hour during shift turnover, to reduce distractions.-
The initial
cause of the mispositioned valve has been corrected via procedure
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revision OP-17, Revisions 12 and 68, respectively. for Units 1 and 2.
This item will remain open until the requirements specified in
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SI-87-014 are incorporated into the licensee's permanent procedures.
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(CLOSED)
Inspector Followup Item 325/87-06-01. Review Motor-Driven
Fire Pump Breaker Misalignment OER.
The inspector reviewed the
completed OER-87-10 and concluded that the root cause~ determination
and corrective actions taken were appropriate.
The licensee issued
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an LC0 declaring the motor-driven fire pump inoperable for over 38
hours.
No LC0 Action Statement time limitation was exceeded during
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the event.
The 4160 breaker misalignment occurred because OP-41,
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Fite Protection and Well Water System, was unclear.
The positions in
OP-41 have been revised as of May 19, 1987, to include a more
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specific description of nonnal breaker position.
This problem
appears to be unique at Brunswick to the motor-driven fire pump.
g.
(OPEN)
Inspector Followup Item 325/87-06-02, Repair of Diesel
Generator Exhaust Silencers.
The inspector examined the rusted
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bottom of the exhaust silencers and does not consider the condition
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to adversely affect the operability of the diesels at this time. The
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licensee plans to replace the silencers by December 1,1988, to
prevent any potential problems.
This item remains open pending
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installation of the new silencers.
h.
(CLOSED)
Inspector Followup Item 325/87-06-03, Documentation of
Welding Associated with LER 1-84-02.
The inspector was unable to
obtain any information regarding additional. welding records
associated with the repair of the EHC.
The inspector reviewed the
as-is LER package which contained the EWRs, trouble tickets, design
drawings, and certain portion; of the welding documentation.
Appendix'B does mot apply to the non-safety EHC system,
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(CLOSED)
Inspector Followup. Item -325/87-20-01 and 324/87-20-01,
Enhancement of PID Tracking.
The licensee has developed a .PID
tracking program which includes identification through acceptance -
milestones and the accompanying research of- late itens.
The
inspe'ctor reviewed DI-LRP-21, Revision 0, dated December 21, 1987,
which provides the procedural documentation of the administrative
controls relative to the PID tracking system, and found them to be
adequate.
The new system also identifies and tracks PIDs that have
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been accepted, but remain open for completion of the defined scope.
Site procedure BSP-14, BSEP Project Identification, is currently
being revised to incorporate the changes to the PID process.
This
revision is to be completed by June 30, 1988.
No significant safety matters, violations or deviations were identified.
10.
Information Meeting With Local Officials (94600)
The inspectors and the project section chief explained the NRC's role and
inspection program to local officials. Meetings were held with members of
the New Hanover County Commission on April 14, 1988, and with the Chairman
of the Brunswick County Commissioners on April 15, 1988.
The project
section chief described the NRC inspection progrco to the City of
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Southport board of aldermen at their regularly scheduled monthly meeting
on April 14, 1988.
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The inspectors and section chief visited the local public document room at
the University of North Carolina at Wilmington.
The collection was well
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maintained and in order.
The cognizant librarian stated that collection
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use averaged 1 person per month. The inspectors had no further questions.
11. Plant Startup from Refueling - Unit 2 (71711)
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The inspectors reviewed / observed activities associated with Unit 2 startup
after the refueling outage to determine if activities were conducted in
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accordance with approved procedures.
The inspection included a review of
the changes to GP-01, Revision 105, Startup Checklist; direct observation
of startup and approach to criticality; observation of selected
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surveillance -tests; . review of licensee's drywell closecut inspection '
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results with independent drywell inspection; examination of selected
systems to assure startup readiness; and review of selected modification
training packages.
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a.
Specific inspection items included:
(1) Training
During the refueling outage for Unit 2, the Alternate Rod
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Injection system was installed under PM-86-035 to comply with 10 CFR 50.62, the:ATWS rule.
The purpose of the ARI system is to
initiate a reactor _ scram by' a means independent of 'the Reactor
Protection System.
Four_ additional vent paths have been
~ installed on the scram outlet valve air header, with each of the
four vent paths consisting of two solenoid valves in series.
The system is automatically initiated from ARI logic or can be
manually initiated, if required.
The inspector reviewed the modification-training package, sample-
examinations and examination test results to determine the
adequacy and completeness of the training conducted for those
operators licensed on Unit 2.
(2) Valve Lineups
The completed valve lineup sheets for the Control Rod Drive and
the Standby Liquid Control System were reviewed to assess their
completeness.
In addition, the inspector physically verified
the valve positions of the SLC system valves.
(3)
Inspection
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Several areas were inspected that are normally high radiation
areas to look for housekeeping and general material conditions.
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The areas inspected included the HPCI roof, RCIC steam tunnel
and the 66 foot penetration room.
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(4) Drywell Closeout
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The inspector reviewed licensee preparations for closing out the-
drywell. Administrative Procedure, AP-96, Drywell Closecut, was
the governing procedure.
The inspector reviewed those
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discrepancies identified by the licensee along with their
corrective action.
In addition, a physical inspection of the
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drywell was performed to verify that licensee actions were
adequate.
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b.
Inspection findings were as follows:
(1) Victoreen Radiation Detector Cable
During the drywell inspection closecut, the inspector noted that
i.he installed configuration of the Victoreen high range
radiation monitor differed from the tested configuration as
documented in Victoreen report 950.301.
Specifically, the
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method of terminating the cable to the detector, the cable
manufacturer and lack of a sealed conduit system for the
detector cable, all differ from that which was tested.
The
termination method and use of a sealed conduit system by
Victoreen in their test program was crucial to the successful
completion of the test as evidenced by the numerous - test
failures experienced in previous testing without -this
configuration.
The licensee did provide some information to the
inspector describing the differences between their termination
means and that which was tested and the specific measures taken
to preclude moisture entry to the connector area.
No
information has been provided which demonstrates that the
installed cable which is supplied by a different manufacturer
and is not installed in sealed conduit, is acceptable for this
application.
This item is unresolved and is listed as
Unresolved Item: Qualification of Victoreen Radiation Detection
Cable,(324/88-15-03).
(2)
Improper Change In Operational Condition
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While observing startup preparations being made for Unit 2
startup on April 26, 1988, the inspector first noted at about
2:45 p.m. that the Reactor Mode Switch was in the Startup/ Hot
Standby position.
Licensee personnel explained that the switch
was in this position because of previous testing performed on
the Rod Sequence Control System and the Rod Worth Minimizer and
that it was left in this position because their procedures
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allowed this if reactor startup would commence shortly.
The
testing was completed at 9:45 a.m.
Startup commenced at 4:00
p.m.
A detailed sequence of events for both the control room
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personnel and the NRC inspector is contained in enclosure 2.
This enclosure also addresses appropriate procedures and steps
utilized during the evolution.
The Rod Worth Minimizer Periodic Test 1.6.2-2, Step 7.1.23,
states to place the mode switch in SHUTDOWN unless the shif t
foreman verifies that the prerequisites are met for leaving the
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mode switch in START /H0T STANDBY pending reactor startup.
The
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Rod Sequence Control System Operability Periodic Test 1.6.1,
which was performed after the Rod Worth Minimizer Test, did not
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contain this step. - At 9:45 a.m., following completion of the
testing, the following prerequisites had not been met for
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etartup:
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RHR Division II not lined up for automatic LPCI initiation.
The "B" RHR loop was in the shutdown cooling mode at this
time and not lined up for automatic LPCI initiation.
At
2:03 p.m., the licensee stopped the RHR "B" pump to secure
the shutdown cooling mode of RHR and line it up for LPCI
initiation in accordance with Section 7.2 of OP-17.
The
licensee completed steps through B.4 in this procedure
(starting recirculation pumps) at 2:53 p.m.
Steps 5
through 16, which would have completed the restoration and
lined up the system for automatic LPCI initiation, were not
continued at this time.
At 3:03 p.m., GP-1 was signed off
by the SF and SOS stating that all prerequisites for
startup had been met.
A PA announcement was made that
primary and secondary containment was in effect and that
reactor startup was commencing. GP-2 was entered and steps
5.2.4, 5.2.5 and 5.2.6 were completed. The next step, step
5.2.7, states to withdraw control rods.
At this time the
NRC inspector noted that the RHR system had still not been
restored.
Specifically, he noted that the suppression pool
suction valves F0208, F004B and F004D were shut.
These
valves do not receive an open signal during LPCI
initiation.
When questioned about this configuration, the
SF directed his people to restore the lineup.
Steps
7.2.B.5 through 16 of OP-17 were completed at 3:46 p.m.
GP-1 was then signed off at 3:46 p.m. and reactor startup
commenced at 4:00 p.m.
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Nitrogen backup system inoperable.
The nitrogen backup system supplies a pneumatic source to
selected safety-related loads.
Following a LOCA and
subsequent containment isolation, the normal air supply to
the drywell will be isolated and supply will be from the
nitrogen backup system to the suppression pool to reactor
building vacuum breaker and the SRVs.
This system was
tagged out in accordance with the requirements of OG-3,
Primary Containment Access Control, which describes the
requirements for allowing personnel entry into the drywell.
This procedure requires that RNA-SV-V5251 and RNA-SV-V5253
be closed and placed under SF clearanc- if the plant is in
Condition 1, 2, or 3 or that the nitrereu backup isolation
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valves, RNA-V347 and RNA-V348 be tagged closed if the plant
is in Condition 4 or 5.
Clearance 2-13578 was hung at 4:25
a.m. on April 25, 1988, placing a tag on the control
switches (closed) to 2-RNA-SV-V5251 and 2-RNA-SV-V5253.
These valves supply air to the non-interruptible air header
which supplies the SRVs.
These valves open on loss of
power.
The plant was in Condition 1 at the time the
clearance was hung.
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Although the safety significance of this clearani.e being in
effect during the mode change is small, since the plant was
not at pressure and the'SRVs were not required, the system
was part of an active LC0 (A2-88-0716) and should have:been
cleared prior to any mode-change .
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Primary Containment - Airlock Door Surveillance .Not
Completed.
PT-2.6.6 was completed at 11:11 a.m.
This test verifies
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that the drywell airlock is operable.
It was necessary to
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perform this test since the airlock was previously made-
inoperable to support welding work . performed in the
drywell.
During this mode change period, therefore,
primary: containment was not in effect as required by the
plant's Technical Specifications.
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This startup showed several examples of failure to follow
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procedures, e.g., signing off OP-1 before it was complete, not
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finishing 0P-17, not following the' instructions of PT-1,6.2-2.
In addition, it showed an inccrrect interpretation of their
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Technical Specifications regarding testing the Rod Worth
Minimizer and the Rod Sequence Control System.
The Technical Specifications, Section 3.1.41 and 3.1.4.2, state that "Entry
into Cor.dition 2 and withdredal of selected control rods is
permitted for the purpose of determining operability of the RWM
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(RSCS) prior to withdrawal of control rods for the purpose of
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bringing the reactor to criticality."
The licensee interpreted
this statement as allowing the mode switch to be placed in
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"startup" for testing purposes.
This requirement addresses
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entry into Condition 2 which means that the prerequisites for
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meeting that mode change must be satisfied.
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The licensee's position was that, although procedurally
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permitted, they would not have pulled control rods with' the
Division II RHR system not lined up for automatic LPCI
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initiation.
During startup evolutions, the licensee stations an
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additional SR0 in the control room to oversee startup
activities.
Although his tasks are not administratively
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defined, the licensee states that this individual was aware that
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the LPCI lineup had not been restored and he would have
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prevented pulling control rods without the system lined up for
automatic initiation.
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This matter is a Violation of the licensee's Technical
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Specifications: Improper Change In Operational Condition,
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(324/88-15-01).
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One violation and no deviations were identified.
A failure in the
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licensee's administrative control of startup prerequisites was identified.
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12. Onsite Followup of Events (93702)
a.
Auxiliary Contacts in Motor Control Centers Breakers
GE _has found a potential-failure +. ode of an auxiliary contact block
assembly used with contractors 'in safety-related motor. control
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centers.
The licensee had previously -identified a problem with
auxiliary contAt blocks (CR-205 device) in early _1987 (see LER
,
1-87-01).
The CR-205 ' devices were sticking,-preventing
. motor-operated valve operation.
The licensee replaced all CR-205
-devices in important safety-related applications with a newer design
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CR-305 device which was not susceptible to the CR-205 failure mode.
However, .the licensee has found three sticking--305 devices since
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April 6,1988; two failed during replacement installation.
GE -
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Bloomington's initial assessment indicates that- the manufacturing
process was creating small burrs on the moving parts, resulting in
the failures.
GE has modified the process and slightly re-designed
the part to fix the problem.
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The licensee immediately inspected .133 important safety-related
breaker compartments and verified that no installed 305 device was
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sticking.
On April 29, 1988, in a conference call.with Region Il
management, the licensee reported that they had again inspected (a
week from previous inspection),133 important safety-related breaker
compartments and verified that no installed 305 device was sticking.
In addition, they described their test program of the- 305 devices
being conducted jointly with General Electric. The testing began on
April 28 to verify the failure rate of the old and new 305 devices.
The testing results will determine if the licensee's intention to
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replace the old 305 devices with new 305 devices on an "as failed"
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basis or during preventative maintenance should be modified.
The inspectors will continue to follow the licensee's actions
regarding auxiliary contractors.
This is an Inspector Followup Item:
Failures of GE 305 Auxiliary Contact Adder Blocks (325/88-15-04 and
324/88-15-04).
b.
Silicon Bronze Bolts in Safety-Related Switchgear
The licensee infomed the inspector on April 19, 1988, that the
broken and cracked bolts (see report No. 325,324/88-05) had failed'
due to intergranular stress corrosion cracking instead of excessive
This determination was made through metallurgical analysis
conducted by the Harris E&E Center.
The licensee, in a conference
call with NRC personnel on April 29, 1988, stated that they would be
replacing the 5/16 inch silicon bronza carriage head bolts with mild
steel bolts in all safety-related electrical panels except the DC
switchbeards by May 17, 1988; the DC switchboard plan would be ready
in 30 days.
This is an Inspector Followup Item:
Silicon Bronze
BoltsinSafety-RelatedSwitchgear(325/88-15-06and324/88-15-06).
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No significant safety matters, violations or' deviations were identified.
13. List of Abbreviations for Unit 1 and 2
AI
Administrative Instruction
A0.
Auxiliary Operator
ARI-
Alternate Rod Injection
Anticipated Transient Without Scram
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BESU
Brunswick Engineering Sub Unit
Brunswick Steam Electric Plant
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BSP
Brunswick Site Procedure
C0
Control Operator
Carolina Power and Light Company
Control Rod Drive
Condensate Storage Tank
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Direct Current
Diesel Generator
Differential Pressure Test
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E8E
Energy and Environmental
Engineering Evaluation Report
Electro Hydraulic Control System
ENP
Engineering Procedure
ERFIS
Emergency Response Facility Information System
Engineered Safety Feature
Engineering Work Request
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F
Degrees Fahrenhtit
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General Procedure
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HP-
Health Physics
High Pressure Coolant Injection
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Heat Exchanger
Instrumentation and Control
NRC Office of Inspection and Enforcement
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IFI
Inspector Followup Item
IPBS
Integrated Planning Budget System
Interrediate Range Monitor
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LC0
Limiting Condition for Operation
LER
Licensee Event Report
Loss of Coolant Accident
Low Pressure Coolant Injection
Motor Generator
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Maintenance Surveillance Test
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Non-Conformance Report
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NRC
Nuclear Regulatory Conrnission
Operating Experience Report
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OP
Operating Procedure
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Operating Procedure Manual
Protected Area
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Project Identification
Plant Modification
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PNSC
Plant Nuclear Safety Committee
Pounds per Square Inch Gauge
Periodic Test
Quality Assurance
Quality Control
Reactor Core Isolation Cooling
Reactor Turbine Gauge Board
RX
Reactor
SF
Shift Foreman
Standing Instruction
Service Information Letter
Sequence of Events
SOS
Shift Operating Supervisor
Special Procedure
SR0
Senior Reactor Operator
S/U
Startup
TS
Technical Specification
Technical Specification Interpretation
Unresolved Item
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ENCLOSURE _2
U-2 Startup Sequence of Events for 4/26/88
Initial Conditions: All rods in, Reactor temperature 180 degrees F Reactor
Coolant System depressurized with manual head vents open.
Procedure _ Step
Time
Action
Mode switch in S/U for testing) pts
0435
1.6.1 and 1.6.1-2 (from C0's log
2041
PT-1.6.1 complete (RSCS - from SF log)
(0945 from GP-1, Attachment 1)
1046
PT-1.6.2-2 complete (RWM - from SF log
(0912 from GP-1, Attachment 1)
1406
Secured Shutdown Cooling (from C0's
log)
1419
Drywell LC0 cancelled (SF log)
1430
NRC inspector asks U-2 SF about status
of RHR Loop
"B" and why it was not
lined up for auto LPC1 initiati n.
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Informed by SF that procedure was in
progress and that following start of
"B" Recirc. Pump the lineup would be
restored.
(Note:
OP-17, Step 7.2.b.4
shows this to be the correct sequence.)
1445
NRC inspector asks U-2 SF why mode
switch is in S/U.
Informed that it was
placed in that position earlier for
testing and that their procedures allow
them to keep it there if S/U expected
shortly.
(Note:
Step 7.1.23 of
PT-1.6.2-2 states to place mode switch
in shutdown, unless the SF verifies
that the prerequisites are met for
leaving the mode switch in START-HOT
STBY pending reactor startup.)
Step 7.2.B.4 of OP-17
1453
Started 2B Rx Recirc. MG set (from
Step 5.1.16 of GP-1
C0's log)
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Enclosure 2
2
Procedure _S,tep
Time
Action
1503
Primary and Secondary Containment in
effect(CO' slog)
1503
GP-01 Startup Checklist complete (time
later changed to 1546)
Step 5.1.13 of GP-1
1505
PA announcement that Primary and
in effect,
commencingRxS/U(fromChemistrylog)
Step 5.2.4 of GP-2
1507
Verified Rx Vessel Shell temperature to
5.2.5
right of criticality line (C0's log)
1510
(After PA announcement)
NRC inspector
again asks about status of RHR
"B"
Loop.
U-2 SF talks with U-2 operators
and then flush begins on RHR "B" Loop
per Step 7.2.B.5 of OP-17,
1516
Commenced Reactor Startup (SF log, time
later changed to 1600)
Step 5.2.6 of GP-2
1519
Chemistry informed of U-2 mode change
to mode 2 at 1516 (time later changed
to 1600 - Chemistry log)
1530
When leaving Control Room, NRC
inspector overhears SOS asking U-2
operators why rods are not being
pulled. He was told that RHR flush was
in progress, to which he asked, why
can't one operator do the flush while
the other pulls rods?
The operators explained that the flush
had to be done in order to restore the
"B" Loop for auto LPCI initiation
and rod pull would begin after LPCI "B"
was lined up.
Step 7.2.B.13 of OP-17
1546
RHR B Loop in standby per OP-17 (CO's
Step 5.1.15 of GP-1
log)
Step 5.2.4 of GP-2
1559
Verified Rx Yessel Shell temperature to
right of criticality line (C0's log)
Step 5.2.7 of GP-2
1600
Cormenced Rx S/U, first Rod out (SF,
C0's log)
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Enclosure 2
3
APPLICAB_LE _PROCEDU_R_ES_
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PT-1.6.2-2
Rod Worth Minimizer System Operability Test
PT-1.6.1
Rod Sequence Control System Operability.
Residual Heat Removal System Operating Procedure
GP-01
Startup Checklist
GP-02
Approach to Criticality and Pressurization of the
Reactor
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