ML20062G394
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| ML20062G394 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/19/1990 |
| From: | Carroll R, Prevatte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20062G388 | List: |
| References | |
| 50-324-90-41, 50-325-90-41, NUDOCS 9011290162 | |
| Download: ML20062G394 (17) | |
See also: IR 05000324/1990041
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UNITED STATES
NUCLEAR nEGULATORY COMMISslON
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REGION 11
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101 MARIETTA STREET, N.W.
* 'r ATLANT A, GEORGI A 30323
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Report Nos.: 50-325/90-41 and 50-324/90-41
Licensee: Carolina Power and Light Company
P. O. Box 1551
Raleigh, NC 27602
Docket Nos.: 50-325 and 50-324 License Nos. DPR-71 and DPR-62
Facility Name: Brunswick 1 and 2
Inspection Conducted: October 2 - November 4, 1990
Lead Inspector: . IdxM[M I ///O/fC
R. L. Prevat C ~ g g Date Signed
Other Inspectors: W. Levis
D. J. Nelson
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Approved Bh: . E. C/orroll, Acting Section Chief Date 5figned
Reactor Projects Branch 1
Division of Reactor Projects
SUMMARY
Scope:
This routine safety inspection by the resident inspectors involved the areas of
maintenance observation, surveillance observation, operational safety verifica -
tion,-initial response to onsite events, onsite review committee, onsite
, , followup of. licensee event reports, review of 10 CFR Part 21 items, and action
on previous inspection findings.
Results:-
In the areas inspected, no programmatic weaknesses or significant safety
matters were identified.
-A non-cited violation for the failure to place a channel of the reactor
protection' system scram discharge volume water level high trip system in the
trip condition after exceeding the Technical Specification two hour time limit'
for having this equipment in test was identified, paragraph 7.a.
Two minor deficiencies involving the lack of attention to detail by the contrml
room operators were noted.during routine plant tours, paragraph 4.
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:A-review of the unit trip-and degraded voltage event that occurred on
September- 27,1990,' identified four items where the inspector did not have
sufficient information and/or. the necessary resources to fully evaluate. These "
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- items will be referred to NRR for further review, paragraph 7.c.
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Unit 1 was in a refueling outage during the reporting period. Unit 2 experi- '
encedian automatic trip on October 12 as the result of a blown fuse in-the
'- feedwater level' control' system, paragraph 5. The unit was restarted on
0ctober- 18,:1990. The~ licensee appeared very conservative-in their approach to
unit restart and reduced- power twice during equipment malfunctions _ that may
have placed the unit at risk.
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REPORT DETAILS
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1. Persons Contacted
Licensee Employees
*K. Altman, Manager - Regulatory Compliance
F. Blackmon, Manager - Radwaste/ Fire Protection
S. Callis, On-Site Licensing Engineer
T. Cantebury, Manager -- Unit 1 Mechanical Maintenance
*G. Cheatham, Manager - Environmental-& Radiation Control
M. Ciemnicki, Security
R. Creech, Manager - Unit 2 I&C Maintenance-
J. Cribb, Manager - Quality Control (QC)
*W. Dorman, Manager - Quality Assurance (QA)/(QC)
*M.-Foss, Supervisor - Regulatory-Compliance
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V. Grouse, Employee Relations
'J. Harness, General' Manager - Brunswick Steam Electric Plant
W ' Hatcher, Supervisor - Security
R. Helme, Manager - Technical Support
J.-Holder, Manager-OutageManagement& Modifications:(OM&M)
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'*M. Jones,. Manager - On-Site Nuclear Safety - BSEP
R. Kitchen, Manager:- Unit 2 Mechanical' Maintenance
t*B. Leonard, Manager - Training
*J. Leviner, Manager - Engineering Projects
J.E McKee, Manager' -QA
', (*J. Moyer, Technical. Assistant to Plant. General' Manager ,
*P. Musser, Manager - Maintenance Staff
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n B.'Poteat,? Administrative: Assistant to Plant General Manager
R.-Poulk,-Manager -. License Training-
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' *J. Simon, Manager L- Operations Unit 1 .
W. Simpson,-Manager - Site Planning and Control-
S. Smith; Manager -Unit 1 1&C Maintenance.
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' R.L Starkey, Vice President -' Brunswick Nuclear Project
R. Tart,-Manager ,0perations; Unit-2 .
J.'Titrington,. Manager ,0perations, Staff
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*R. Warden, Manager Maintenance
B. Wilson,: Manager - Nuclear Systems Engineering -l
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L0ther licensee employees contacted included construction craftsmen,
h : engineers, . technicians, operators,toffice personnel, and security force
+ members.-
* Attended:the' exit interview' g
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Acronyms:and initialisms used in the report are listed in the last:
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paragraph.
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2. 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
components were operable; administrative controls were followed; tagouts
were adequate; personnel were qualified; correct replacement parts were
used; radiological controls were proper; fire protection was adequate;
quality control hold points were adequate and observed; adequate
post-maintenance testing was performed; and independent verification
requirements were implemented. The inspectors independently verified that
selected equipment was properly returned to service. ,
Outstanding work requests were reviewed to ensure that the licensee gave
priority to safety-related maintenance. The inspectors observed / reviewed
portions of the following maintenance activities:
90-ALUC1 2C RBCCW Pump Rebuild
90-ASYH1 MSL Rad Monitor B Power Supply Changeout
90-PME411 Route on 2-CAC-1262
90-PZI395- Sample Pump Replacement for 2-CAC-4409
90-WFI414 Rosemount ATTU Output Voltage Check
Violations and deviations were not identified.
3. SurveillanceObservation(61726)
The inspectors observed surveillance testing required by Technical Speci-
fications. .Through observation, interviews, and record review, the
.. inspectors verified that: tests conformed to Technical Specification
requirements; administrative controls were followed; personnel were
qualified; instrumentation was-calibrated; and data was accurate and
complete. The inspectors independently verified selected test results and
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proper return to service of equipment.
The inspectors witnessed / reviewed portions of-the following test
activities:
1-MST-IRM12W IRM Channels B, D, F, H Functional Test
1-MST-SRM22R SRM Channels A and C Channel Calibration
PT-7.1.1.a Core Spray Injection Check Valve Operability Test
PT-20.7.2 1-E11-F050B LLRT
Violations and deviations were not identified.
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4. Operational Safety Verification (71707)
-The inspectors verified that Unit 1 and Unit 2 were operated in compliance
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.
The inspectors verified that control room manning requirements of 10 CFR
50.54 and the Technical Specifications were met. Control operator, shift
sup?rvisor, clearance, STA, daily and standing instructions, and
jumper / bypass logs were reviewed to obtain information concerni .g operating
trends and out of service safety systems to ensure that there were no
conflicts with Technical Specification Limiting Conditions for Operations.
Direct observations of control room panels and instrumentation and recorder
traces important to safety were conducted to verify operability and that -
operating parameters were within Technical Specification limits. The
inspectors observed shif t turnovers to verify that system status continuity
was maintained. The inspectors verified the status of selected control-
room annunciators.
Operability of a selected Engineered Safety Feature division was verified
weekly by ensuring that: each accessible valve in the flow path was in
its correct position; each power supply and breaker was closed for
components 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 conditions did not exist
which could 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
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policies / procedures were'followed. This included observation of HP
practices and a review of area. surveys, radiation work permits, postings,
and instrument calibration.
The inspectors verified by general observations that: the secarity
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organization was. properly manned and security personnel were capable of
performing- their assigned functions; persons and packages were checked
prior to entry into the PA; vehicles were properly authorized, searched
and escorted within the PA; persons within the PA displand photo identi-
fication badges; personnel in vital areas were authorized; effective
compensatory measures were employed when required; and security's response
to threats or alarms was adequate..
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Ine inspectors also observed plant housekeering controls, verified
position of certain containment isolation ' alves, checked clearances, ard ,
verified the operability of onsite and of' site emergency power sources.
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An inspector found two minor control room deficiencies on Unit 2 not
discovered by operators. On October 19, 1990, at approximately 7:30 a.m.,
the control switch key was found inserted in the control switch for Core
Spray Valve F001A, Su)pression Pool Suction Valve. Important component
control switches on tie control board are equipped with locks that require
keys for operation to preeent inadvertent manipulation. Dummy keys are
normally inserted with the color coded real keys attached. The valve was
in its correct position, but had been manipulated for the performance of a
surveillance test on the previous shift. Following the test, the key was
not switched back. Prior to the inspector's discovery, three control
operators and one trainee had performed their board walkdown without
detecting the error. The inspector informed the operators of the problem
and the key was promptly switched.
On October 24, 1990, the inspector observed that one of six APRM GAFs
indicated 0.00 on the 7:00 a.m. process computer P-1 printout. Normal GAF
values are 0.98 to 1.00. The printout also flagged APRM B as a failed
sensor. The process computer calculates the GAF from a ratio of calculated
power and APRM indicated power. The 0.00 value observed by the inspector
indicates that a meaningful GAF could not be calculated due to a problem
with calculated or indicated power. The inspector reviewed previous
hourly P-1s, which also indicated GAFs of 0.00 back to 3:00 a.m., which
had a 4.17 value. When questioned, the on-duty-operators were unaware of
and could not explain the discrepancy. Et 'ntually the operators determined
that a surveillance test prior to 3:00 a.m. for APRM Channel B caused the
computer to cease its " scan" of APRM B due to inconsistent values calcu-
lated when the APRM output was manipulated during the test. The APRM-was
placed back in service following the surveillance test, but the computer
" scan" was not reset. This did not affect the operability of the APRM,
and no other indications in the control room were affected.
GAF values are the most consistent indication of APRM reliability.
Although the APRM trend chart recorders would also indicate an actual APRM
problem, the recorders cannot trend all six APRM channels simultaneously.
Therefore, it is important for operators to monitor the GAFs.
These two examples, while not safety significant, indicate that continued
diligence by operators-is needed in monitoring control room indications.
Violations and deviations were not identified.
5. Initial Response to Onsite Events (93702)
Unit 2-Scram
Unit 2 was operating at 100 percent power on October 12, 1990, when a fuse
blew in the FWLCS resulting in a loss of power to a . number of components
in the control circuit. The loss of power gave the appearance of low
reactor water level which caused an increased demand signal to the RFPs
and a trip' signal to the reactor recirculation pump runback circuits. The
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"A" RFP also locked up due to the loss of power resulting in the "B" RFP
responding to the master controller demand for increased feed flow.
Twenty-six seconds after the fuse blew an actual high water level
condition was reached which caused a turbine trip and reactor scram.
The lowest reactor water level reached was approximately 117 inches.
Group 2, partial Group 3 and Group 6 isolations were received. Reactor
recirculation pumps tripped due to the low level condition. RCIC auto
started and injected and HPCI auto started but did not inject due to the
short duration of the level transient. HPCI was subsequently used to
raise water level to the normal band.
During post trip recovery, the operators had difficulty in placing a RFP
in service. Because of the loss of power to the FWLC circuit, the master
controller was still demanding 100 percent output. When 2B RFP was placed
back in service, in automatic, the pump increased speed to 5700 RPM and
discharge pressure increased to 1700 psig. With the flowpath isolated due
to feed pump trip recovery actions, the 4 and 5 feedwater heater relief
valves lifted and began releasing steam into the feedwater heater room.
When steam was reported as being released in these rooms, the feedwater
heater inlet valves were shut securing the leak. The FWLC system was also
placed in single element and selected to channel B, which then restored
level feedback to the controller.
:The fuse that blew in the' FWLC circuit was a Gould Shawmut A25Z? fuse and
was similar to the fuse which blew in this circuit on August 16, 1990,
that also resulted in a reactor scram. The licensee has subsequently
replaced the Gould Shawmut fuses in the FWLC circuitry with Bussman MIN
fuses which are the type installed in Unit 1. The Gould Shawmut fuses
were installed as an Appendix R modification to provide separation between
safe shutdown circuits and other associated circuits of concern.
Subsequent review by the licensee documented in EER 90-0262, October 14,
-1990, determined that separation of this circuit was not required.
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The inspector will review the licensee's corrective actions taken with
respect to.the feed pump operation and the fuse failure when the LER is
. issued.
6. Onsite Review Committee-(40700)
The inspectors attended selected Plant Nuclear Safety Conrittee meetings
conducted during the period. The inspectors verified that the meetings
were conducted in accordance with Technical Specification requirements
regarding quorum membership, review process, frequency, and personnel
qualifications. Meeting minutes-were reviewed to confirm that
decisions /recomendations were reflected in the minutes and followup of-
corrective actions was completed.
Violations and deviations were not identified.
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7. Onsite Followup of Licensee Eveat Reports (92700)
The below listed 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
concluded that necessary corrective actions have been taken in accordance
with existing requirements, license conditions, and commitments, unless
otherwise stated,
a. (Closed) LER 1-90-16, Operation Prohibited by Plant Technical
Specifications During SDV Maintenance and Surveillance Activities.
On September 17, 1990, the Unit 1 SF authorized the performance of
IMST-RPS27R, RPS Scram Discharge Volume High Water level Channel
Functional Test and Channel Calibration. In conjunction with the
test, the SF also authorized work to be performed in accordance with
the instructions of WR/JO 90-AMCT1, which replaced the electronic
printed circuit board for level switch 1-C11-LSH-4516C. The SF
entered a tracking LC0 for this work believing that only one input to
channel A2 would be disabled.
Subsequent review of this work on September 20, 1990, by a different
shift foreman, revealed that the MST disables the trip function of
channel A2 during a portion of the test. A two hour time limit is
allowed by Technical Specification 3.3.1 for a channel to be disabled
during surveillance testing provided that the other channel in the
same trip system is operable. After the two hour period, the channel
must be returned to operable status or placed in the tripped
condition. Because of the maintenance actions performed on
1-C11-LSH-4516C, the two hour time limit was exceeded during
performance of the MST on September 17, 1990. .The SF did not place
the channel in the tripped condition, as required by Technical
Specification 3.3.1, because he did not realize that the A2 channel
was disabled during the testing. The channel was inope aole for 2
hours and 20 minutes,
in their investigation..the licensee noted several contributing
conditions. First, it is not appropriate to perform corrective
maintenance while performing surveillance tests. The two hour time
period allowed by Technical Specifications for ' inoperable channels
applies. to surveillance testing and not corrective maintenance. A
Standing Instruction was put in place until permanent procedure
revisions are made that prohibit corrective maintenance to be
performed during surveillance testing under the two hour time
constraint. In addition, the MSTs will be revised to enhance the
specific instruments that will be made inoperable by jumper, and
training for the appropriate people will be performed.
The failure to place channel A2 of the RPS SDV water level high trip
scram signal in the tripped condition is a violation of TS 3.3.1:-
Failure to Place Channel- A2 In the Tripped Condition, (325/90-41-01).
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This licensee identified violation is not being cited because
criteria specified in Section V.G.1 of the NRC Enforcement Policy
were satisfied.
b. (Closed)LER 1-90-017, Unit 1 High Pressure Scram During Performance
of Turbine Control /Stop Valve Tightness Test. A unit 1 scram, from
high pressure, during the ierformance of Turbine Control /Stop Valve
m Tightness Test, occurred oi September 27, 1990, and was discussed in
inspection report 90-37. At the time of that report, the licensee
was still investigating tne event and preparing a LER. The investi-
gation was subsequently completed and LFR 90-017 was issued on
October 26, 1990. The LER provided a description ci the events
surrounding this-occurrence and contained recommendations to prevent
recurrence. This event occurred with the unit at approximately c2
percent po;:er while in the process of shutting down for a refueling
outage. The licensee had started the planned Periodic Test (PT)
40-2-10, Turbine Control /Stop Valves (TCV/TSV) Leak Tightness Testing.
The event occurred due to erroneous procedural guidance provided by
the vendor, General Electric, in GEK-25406A and defective switches on
the TSVs, which allowed the TCVs to open when the TSVs were closing.
The turbine BPV's open demand signal was limited by the maximum
combined flow circuitry of the turbine control system. The closure
of the TSVs without the BPVs being able to open caused reactor
pressure to increase to the SCRAM setpoint. A detailed licensee
review of this event has determined that this proceaure contained
weaknesses which are applicable to both nuclear and fossil plants
which have used this procedure / guideline to develop their plant
tests. The licensee had made this information available to other
utilities through " Network" and have indicated that GE may issue an
information letter on this item,
c. (Closed) LER 2-90-15, Unit 2 Reactor Scram-Due to Loss of Excitation
on Main Generator. A' scram occurred on Unit 2 on September 27, as a
. result of the loss of excitation on the main generator. This item
was also discussed in inspection report 90-37, and the licensee was
investigating this event at the close.of the previous inspection
period. The licensee subsequently completed their investigation into
the event and provided the details in LER 90-015, dated October 26,
1990. The inspectors have reviewed the LER and other licensee docu-
mentation associated with the event and discussed this matter.with
NRR. Based on this review and discussion with the licensee and NRR,
the inspectors are unable to determine the following: -(1) was the
capacity and. stability of the offsite power system prior to and
immediately after the Unit 2 trip adequate to provide acceptable
voltages to handle a design basis event in Unit' 2 and a safe shutdown
on Unit 1; (2) does the licensee's loss of voltage protection system
(relays). which only sheds the safety-related equipment off the "E"
buses, provide- adequate protection for system voltages less1 than the
setpoint of the degraded voltage relays; (3) if the offsite power
voltage drops to a level that is inadequate to start safety-related
loads, will the 10.5 second degraded voltage relay time delay result
in a delaying safety bus transfer to the diesel generator such that
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the plant may be outside the boundary of this safety analysis; and
(4) does CP&L provide adequate direction and guidance to system load
dispatchers to ensure that adequate voltage is maintained at all
nuclear units? These items will be referred to NRR for review.
Pending the outcome of the above, these items will be tracked as an
inspector followup item: Adequacy of Offsite Power, (325,324/90-41-02).
One e-cited violation was identified.
8. 10CFRPart21 Items (36100)
(OPEN) 325,324/P2188-01 - Worn Shaft Gear Failures in Size 2 Limitorque
Actuators and Also in Fisher Supplied H3BC Actuators. The inspector
discussed this item with the licensee. A review of licensee records could
not determine that this report had ever been received. They contacted
Limitorque who verified that the information had been provided, so it was
apparently lost. The licensee has entered this item into FACTS and has
assigned responsibility for investigation and resolution. This item will
be reviewed further as information becomes available.
(CLOSED) 325,324/P2188-04 - Reinstalling Foxboro Controller Circuit Cards
May Cause 100 Percent Output and Subsequent Transient to Occur. This
controller has been identified in BWR Recirculation Flow Control System
-DCS-88080301. This Part 21 was sent by the Foxboro Company to the Perry
Nuclear Power Plant, it related to SPEC 200, Model 2AC-D+44 controller
card with its associated 2AX+RM removable manual card. This notice was
not sent to Brunswick since they do not use these units in their
recirculation flow control system. A review of parts and installation by
the licensee has verified that these units are not installed or used in
spare parts at Brunswick.
(CLOSED) 325,324/P2189-01 - Brown Boveri K-Line, K-225 Through K-2000
Circuit Breakers Delivered Prior to 1974 Had Rebound Spring Added to Slow
Close In. This item identified that testing had discovered that the above
breakers may fail to function properly in that persistent sine dwell
vibration could occasionally cause the sicw close bar-to move into-a
position such that the breaker, when called on to close, could slow close
rather than closing normally. Adding a rebound spring to the slow close
' lever will prevent the slow close bar from vibrating to the undesired
position. Brunswick was identified as a plant that had received shipment
of these breakers. A survey by the licensee found that 10 electrically
operated breakers of this type were installed in the plant. Substations
3L and 4L located in the hot machine shop have K-1600 incoming main
breakers. These units were manufactured in 1972 and do not contain
rebound springs but are used in non-safety applications to provide power
.to the warehouse, maintenance shops and office buildings. Emergency
substations ES, E6, E7, and E8 were found to have K-3000 breakers and the
Part 21 does not apply to the breakers. The four remaining breakers are
installed as crosstie breakers on substations ES, E6, E7, and E8. These
.are K-1600 breakers manufactured before 1972 and do not contain rebound
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spri ngs ~. However, these breakers are administratively controlled in the
racked out position and verified racked out as part of normal system
surveillance testing. The breakers may be closed at the discretion of
the shift supervisors when both units are in mode 4 or 5 and other TS
requirements are met. No credit is taken in the accident analysis for use
of these tie breakers. Based upon the above, the licensee discovered that
the use of these components at Brunswick does not pose a substantial
safety hazard and, therefore, is not reportable under 10 CFR Part 21.
Since this defect would only occur under seismic event conditions and the
components are not used for safety applications, the licensee does not
have any current plans to install recommended springs in the substations
used for the warehouse, maintenance shop, and office buildings. The
licensee plans to purchase and install the springs in the crosstie
breakers for substations ES, E6, E7, and E8 at the next scheduled
maintenance period after receipt of the springs.
(CLOSED) 325,324/P2189-05 - PT-21/ Germane to Safety from GE:
Susceptibility of Weld Between Core Spray Line and Thermal Sleeve to
IGSCC. GE recommends that welds be included in IE Bulletin 80-13
surveillance. This item was evaluated by GE and the licensee and
determined to be germane to safety, but not reportable. The licensee has
implemented the vendor recommendations and included this item in their
surveillance program to be tested each refueling outage under Periodic
Test, Core Spray /Feedwater Visual Examination, PT-90.1, and reported under
IEB 80-13. The iaspector verified that these reports had been submitted
as required.
(CLOSED) 325,324/P2189-06 - PT-21/ Germane to Safety from GE: Concerns
with Core Neutron Flux Monitoring and Reactor Protection During Refueling.
This item was the result of NRC questioning the conservatism cf
center-spiral reloading because the SRMs were not on scale and, therefore,
not monitoring neutron flux changes during a refueling at the Brown's
Ferry Plant. GE performed an evaluation of this event and concluded that
this event did not-constitute a substantial safety hazard and was not
reportable under the context of 10 CFR Part 21. However, they oid conclude
that this issue was germane to safety. Based on the above, RICSIL No. 039
was issued by_GE on February 10,-1989, to alert BWR owners that interim
recommendations would be provided. These-interim recommendations were:
(1) that during refueling, the neutron monitoring system should be capable
of continuously monitoring changes in neutron flux in the region of the
core'where fuel is being loaded or control rods are being removed to
provide operators with indications of an approach to criticality; (2)'that
the
scramRPS should
based on be inputs capable at all
from the times of
neutron flux reliably. initiating)a
detectors; reactor
and (3 that the
recommendations of SIL 372 and SIL 68 should be followed when refueling
interlocks are bypassed. The inspector reviewed the licensee's Engineering
Evaluation Report (EER) 89-0022, dated January 16, 1989, that evaluated
this event and the vendor recommendations, the licensee's Refueling Fuel
Handling Procedure FH-11, Volume IX, Revision 41, and Engineering Procedure
Guidelines for Prepara+1on of Core Component Sequence Sheets, ENP-24.12,
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Volume XX, Revision 8, to verify that the vendor recommendations had been
implemented. This review indicates that Procedure FH-11 provides
Administrative Control and direction over core reloads, that all control
rods will be fully inserted during fuel movement, that the source range
monitors will be operable and providing on scale indications, that
refueling interlocks will .be operable and not bypassed or jumpered out,
and that the core loading will progress in a sequence which ensures that
the SRMs have accurate indication of changes in neutron levels. These
procedures, therefore, implement the vendor recommendations and appear to
be satisfactory. EPRI is currently conducting a study on this issue. It
is anticipated that this study will be completed in late 1990. The
licensee has indicated that they will review and implement the
recommendations of that study as appropriate to the Brunswick Plant.
(CLOSED) 325,324/P2189-18 - SMB Actuators Found to Have Melamine Torque
Switches That Undergo Post Mold Shrinkage and Cause Cam Binding. Melamine
Torque Switches Found to be Not Qualified. The licensee reviewed this
item and determined it to be applicable to BSEP. A decision was made to
replace the applicable torque switches on PCIS valves during the 1989-90
refueling outage for Unit 2 and on Unit I during its refueling outage in
1990-91.- The remaining torque switches were scheduled for replacement
when routine maintenance is performed on the remaining valve actuators
with all replacement work completed by July 3,1992. The inspector
reviewed the completed work request for PCIS replacement accomplished on
Unit 2 during the past outage, and the work scheduled for Unit 1 during
the current outage. The listing of work scheduled under the routine
maintenance program for the remaining torque switches was also reviewed.
It appears that the licensee has determined which torque switches require
replacement and have the program needed to complete these activities
underway with an established completion date.
(CLOSED) 325,324/P2189-12 - PRE-1981 SMB-000 and PRE-1976 SMB-00 Cam Type
Torque Switches Can Fail as a Result of Stationary Contact Screws
Loosening on Side of Torque Switches That Had Fiber Spacers. Two failure
modes were reported. The first type failure resulted when one of the
screws in the contact bridge came loose resulting in premature tripping of
the torque switch. The second failure resulted when both stationary
contact screws loosened and the contact bridge raised with the contact
fingers maintaining continuity on the torque switch. The licensee's
evaluation concluded that no events of this nature have occurred at BSEP
and, due to the low number of failures reported, concluded that the_ safety
significance was low. A licensee's review of safety-related SMB-00 and
SMB-000 valves at BSEP determined that these switches, which were made of
melamine or ahenolic materials, had been or were:in the process of being
replaced wit 1 new fiberite torque switches as a result of 10 CFR 21.89-18.
The remaining SMB-000 and SMB-00 torque switches-that were made of
fiberite, were: verified as having been replaced since 1985; with the
exception of one switch,1-E21-F001A-M0, that is planned to be worked
under Work Request 89-AZIA1. The new torque switches and any SMB-000
purchased since 1980 and SMB-00 switches purchased since 1976, do not
include fiber spacers and do not have this problem. Based on the above,
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it appears that the licensee has identified all equipment affected by this
L item and has either completed or established documentation to inspect and
replace the remaining items.
(CLOSED) 325,324/P-2190-04 - Rosemount Resistance Bridges Can Exhibit
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Premature Long Term Degradation Under Certain Combinations of Humidity.
- Power and Temperature. Two of the units referenced in the above, serial
numbers 0067897 and 0067898, were provided to BSEP. Neither of these
units have been installed. They have been placed under administrative
hold in stock under CP&L Part Number 731-758-12. They will be returned to
"
the vendor for replacemen+ when the new units on order are received. This
item is closed.
Violations and deviations were not identified,
f 9. Action on Previous Inspection Findings (92701) (92702)
-- (CLOSED) Violation 325,324/88-18 05, HPCI/RCIC High Steam Line Flow
Instruments Inoperable. The inspector reviewed the licensee's response to
the Notice of Violation and Civil Penalty dated January 27, 1989. The
specific corrective actions taken to resolve the instrument's setpoints
g
was detailed in LER'l-88-14, which was inspected and closed in Report No.
90-37. The violation was issued because of the inadequate corrective
actions taken to resolve the issue when it was first identified. In their
response, the licensee coninitted to revising their corrective action
program along with establishing a program to effectively implement the
BSEP system engineering concept.
Weaknesses in the licensee's corrective action and system engineering
programs were also noted in the Diagnostic Evaluation Team report. As a
result, the licensee included.these specific areas into their IAP program
which resulted in further NRC inspections to followup on the program's
implementation. The licensee's new corrective action program was
inspected in Report No. 90-31. The report stated that the actions
committed to in the IAP were completed. The effectiveness of the new
7
program was not assessed and will be evaluated in future inspections.
1 System engineering program improvements were inspected in Report No.
90-16. The report stated that the appropriate procedures and programs
were in place to correct deficiencies in this area. Based on the
inspections. completed to date, which address the required corrective
actions of the violations and the additional inspections planned to assess
_
the effectiveness of the licensee's corrective action, this item is
closed.
(CLOSED) Violation 325,324/88-24-03, Silicon Bronze Bolts Corrective
. Actions. .The inspector reviewed the licensee's response to the Notice of
Violation and Civil Penalty dated January 27, 1989. This violation, along
with 325,324/88-18-05, were included in EA-149 and constituted the 2
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examples of inadequate corrective action. The specific actions taken with
regard to the silicon bronze bolt issue is detailed in LER 1-88-06, which
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was inspected and closed out in Report No. 90-37. As discussed in the
closcout of Violation 88-15-05, the licensee's actions to resolve
corrective action program deficiencies have been and are continuing to be
inspected as part of the followup to the licensee's IAP. Based on these
inspections, this item is closed.
(CLOSED) Violation 325,324/88-34-01, Inadequate Design Control Related to
RCIC Steam Exhaust Check Valve. This violation concerned the Unit 1 and 2
RCIC Steam Exhaust Check Valve, E51-F040. Plant Modifications 81-274 and
81-275, replaced the Unit 1 and 2 valves which included discs with design
pressures of 25 psig. Based on the calculated peak containment pressure
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during a DBA of 49 psig, and containment design pressure of 62 psig, the
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discs were under rated. The inspector reviewed the licensee's response to
the violation and supporting documentation.. tER-88-0461 was written to
evaluate the design pressure discrensocy and concluded that the discs were
sufficient. This was based on successful local leak rate testing at 49
psid and documentation from the valve manufacturer stating that the discs
would " withstand a pressure of 62 psig at 248 degrees F for a sustained
period of time". The EER concluded that these values are greater than the
containment system requirements for the DBA and higher than credible
turbine exhaust operating pressures. The cause of the design error was
that the containment accident pressure and LLRT pressure were not
evaluated; only operational exhaust pressures were considered during the
design process. Revisions to applicable administrative and engineering
procedures have added formalized checklists and training requirements for
safety reviewers.
(CLOSED) Violation 325,324/88-38-01, Failure to Control Combustibles in
Restricted Plant Areas. The inspector reviewed the licensee's response to
the Notice of Violation and supporting dxumentation. The licensee stated
that the reason for the violation was t' n personnel not f amiliar with
fire-retardant wood requirements obtaired wood from outside the protected
area for use as forms to install a plant modification. A contributing
factor was that the modification instructions did not specify the type of
material to be used for the forms. The-licensee enhanced the existing
controls to allow only fireproof material and/or fire retardant wood to be
used within the plant protected area. An improvement in identification
markings for fire retardant wood to ensure this type of wood, including
cut up pieces, is readily identifiable for use in the protected area. The
training lesson plan for the plant construction support group annual
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training was revised to address this topic.
(CLOSED) Unresolved Item 325,324/88-05-01, Service Water System Operating
Mode Concerns. Other inspections of this issue are documented in Inspec-
tion Reports 89-09, 89-12 and 89-14. -As a result of these inspections and
I
others perfonned on the service water system, a Notice of Violation and.
Imposition of Civil Penalty was issued on January 26, 1990, for failure to
take adequate corrective actions for identified service water deficiencies.
Example A of- the violation describes the inadequate evaluation performed
when determining system operability with the single failure concerns
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associated with the SW-V106 valve. Further inspection of this issue will
be performed in the closeout of Violation 325,324/89-34-47.
(CLOSED) Unresolved item 325/88-34-02, Reactor Vessel Water Level Wide
o Range Indication Anomolies. This item was also discussed in inspection
report 325,324/90-02. Currently the licensee is replacing Rosemount
transmitters in Unit 1 and is delaying any further action on the wide
range indicators pending completion of the replacements. The licensee
expects some indication difference with the new transmitters. The
inspector determined that no regulatory issues existed. Therefore, this
item is closed.
(CLOSED) Unresolved item 325,324/88-38-02, Failure to include All LPCI
and Suppression Pool Cooling Flow Path Boundary Valves in Surveillance
Program. This item was also discussed in inspection report 89-05 and was
expanded to include core spray system valves. This item concerned the pts
that meet the monthly TS surveillance requirements of ECCS systems for
verification that each valve.in the flow path that is not locked, sealed,
or otherwise secured in position is in its correct position. The licensee
originally disagreed with the inspector on what was a " flow path" valve. -
Specifically, minimum flow valves, vent isolation valves, valves normally
outofpositioninstandbylineup(i.e.,E11-F027A(B),RHRSuppression
Pool Spray Isolation), or flow path boundary valves were not included in
the surveill'ances. The licensee has subsequently revised the pts to
include these valves. No occurrences are known where these valves have '
been found out of position due to their omission in the pts. This item is
-closed.
(CLOSED). Unresolved item 325,324/90-17-02, Potential Inoperability of
CBEAF System. Based on inspector's quertions regarding the differential
pressure measurement technique used to determine the CBEAF system
operability, the licensee changed the test procedure to more accurately
measure the differential pressure (see inspection report 90-02).
Subsequent tests verified that previous test results may have been s
inaccurate in cietermining that a positive pressure existed in the control
. building relative to the outside atmosphere. Based on the licensee's
previous analysis and documentation from 1985,.the erroneous differential
pressure measurement technique has minimum safety significance, therefore,
this item is closed.
(CLOSED) IFI 324/88-15-05, Normal Position for SW-V117, Nuclear Header to
Vital Header Isolation Valve. As a result of extensive review and
analysis of the service water system design, the normal position of the
SW-V117 was changed from closed to open. The valve position was changed
to allow a.RHR room cooler to be placed in service affording the service
water pumps minimum flow protection under worst case single failure ,
scenarios. The Service Water System Operating Procedures, 1-0P-43,
' Revision 31, and 2-0P-43, Revision 68, require that the valve be open.
The licensee plans to keep the valves open until the service water pump
l thrust bearing modifications are completed. This work is currently
scheduled to begin in 1992. ,
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(CLOSED) IFl 325,324/88-38-05, Licensee Activities Related to Correcting
Keepfill System Discrepancies. The licensee has determined that the
primary cause of keepfill system problems are bent stems on pressure
control valves due to over-pressurization downstream. Several causes of
over-pressurization have been identified. Currently three of twelve
keepfill systems are inoperable. Modifications are scheduled for the
current Unit 1 outage and the next Unit 2 outage to replace pressure
control valves. This item is closed based on the licensee's
identification of the causes of the keepfill system discrepancies and
planned corrective actions.
Violations and deviations were not identified.
10. ExitInterview(30703)
The inspection scope and findings were summarized on November 5, 1990,
with those persons indicated in paragraph 1. The inspectors described the
areas inspected and discussed in detail the inspection findings listed
below. Dissenting comments were not received from the licensee.
Proprietary information is not contained in this report.
Item Number Description / Reference Paragraph
325/90-41-01 h0N-CITED VIOLATION - Failure to Place Channel A2
in the Tripped Position (paragraph 7.a).
325,324/90-41-02- IFl - Adequacy of Offcite Power (paragraph 7.c).
11. Acronyms and Initialisms
A0- . Auxiliary Operator
APRM Average' Power Range Monitor
ATTU Analog Transmitter Trip Unit
BPV Bypass Valve
BSEP Brunswick Steam Electric Plant
-BWR Boiling Water Reactor
CBEAF Control Building Emergency Air Filtration
CP&L Carolina Power & Light Company
DBA Design Basis Accident
ECCS Emergency Core Cooling System
EER Engineering Evaluation Report
ENP Engineering Procedure
EPRI Electric Power Research Institute-
ESF Engineered Safety Feature
F Decrees Fahrenheit
FACT" Facili y Automated Conmitment Tracking System
FWLCS Feedwater Level Control System
GAF Gain Adjustment Factor
GE General Electric
HP Health Physics
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HPCI High; Pressure Coolant Injection
IAP- Integrated Action Plan
I&C- Instrumentation and Control
~1E Inspection- and ' Enforcement
IEB- . Inspection and Enforcement Bulletin
IFl -Inspector Followup Item
IGSCC' Intergranular Stress Corrosion Cracking ;
IPBS Integrated Planning, Budget'ng and Scheduling
IRM Intermediate Range Monitor '
LC0~ Limiting Condition for Opetation
-LER Licensee Event-Report'
LLRT -Local Leak Rate Test
LPCI Low Pressure Coolant Injection
MSL Main Steamline
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MST Maintenance Surveillance Test -
NRC Nuclear Regulatory Commission i
NRR Nuclear Reactor Regulation
PA Protected Area
.PCIS Primary Containment Isolation System !
PNSC- ' Plant Nuclear Safety Committee
PSID Pounds per Square Inch Differential i
PSIG, Pounds per-Square Inch Gauge
PT ' Periodic Test
QA Quality Assurance
QC- Quality Control. -
t
RBCCW Reactor Building Closed' Cooling Water
RCIC- Reactor-Core Isolation Cooling ,
RFP Reactor Feed Pump
. RHR = Residual Heat Removal
,
,RICSIL Rapid Information: Communication Service Information Letter i
RPM Revolutions Per Minute
'RPS: ' Reactor Protection System
, .SDVt Scram Discharge Volume'- :
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~SF Shift Foreman
iSRM Source Range Monitor i
STA Shift Technical-Advisor- - t
, TCV/TSV Turbine Control-Valve / Turbine Stop Valve .;
TS >-Technical Specification
' URI? Unresolved. Item
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WR/J0: Work. Request / Job Order
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