ML20137X384
| ML20137X384 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/20/1985 |
| From: | Fredrickson P, Garner L, Hicks T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20137X354 | List: |
| References | |
| 50-324-85-27, 50-325-85-27, NUDOCS 8510040510 | |
| Download: ML20137X384 (11) | |
See also: IR 05000324/1985027
Text
--
-
-
-
_
-
_.
.
.
pn Rico
UNITED STATES
f
'o
NUCLEAR REGULATORY COMMISSION
et' *$ k
/ ,^S
REGION 11
E I
I
101 MARIETTA STREET. N.W.
-
f
ATLANTA, GEORGI A 30323
% .' ,',#
SEP 2 31985
,.
Report Nos. 50-325/85-27 and 50-324/85-27
Licensee: Carolina Power and Light Company
P. O. Box 1551
Raleigh, NC 27602
1
Docket Nos.:
50-325 and 50-324
License Nos.
Facility Name: Brunswick 1 and 2
Inspection Co
ed:
ugust 1 ; 31,1985
\\
W [
Inspectors:
-
h L.
arner
Vate figned
_
j M
$4
h T. E.
'cks
D6te Sfgned
"
/
7
[fd'
Approved By:
-
P. E. Fredrickson, Section Chief
Dite Signed
Division of Reactor Projects
SUMMARY
Scope: This routine safety inspection involved 177 inspector-hours on site in
the areas of maintenance observation, surveillance observation, operational
safety verification, onsite review committee, ESF System walkdown, Licensee Event
Reports review, follewup on inspector identified items, refueling activities and
plant modifications.
Results:
One violation was identified:
Bolts Replaced on Hydraulic Control
Units with Type Other Than That Specified on Drawings. One unresolved item was
identified:
Seismic Qualification of Hydraulic Control Unit Frame.
I
h0040510850923
G
ADOCK 05000324
__
_
..
__
_
__
- .
.
_ ,_ _ _
.
._
..._ ___. .
._.
-
. - - - .
._ __ _
- _.
_ __
r-
.
.
1
i
REPORT DETAILS
!
1.
Licensee Employees
Persons Contacted
,
P. Howe, Vice President - Brunswick Nuclear Project
.
C. Dietz, General Manager - Brunswick Nuclear Project
T. Wyllie, Manager - Engineering and Construction
,
G. Oliver, Manager - Site Planning and Control
!
J. Holder, Manager - Outages
l
E. Bishop, Assistant to General Manager
I
i
L. Jones, Director - QA/QC
r
M. Shealy, Acting Director - Training
,
M. Jones, Acting Director - Onsite Nuclear Safety - BSEP
J. Chase, Manager - Operations
'
J. O'Sullivan, Manager - Maintenance
G. Cheatham, Manager - Environmental & Radiation Control
,
4
K. Enzor, Director - Regulatory Compliance
B. Hinkley, Manager - Technical Support
L. Boyer, Director - Administrative Support
V. Wagoner, Director - IPBS/Long Range Planning
L
C. Blackmon, Superintendent - Operations
j
J. Wilcox, Principle Engineer - Operations
W. Hogle, Engineering Supervisor
W. Tucker, Engineering Supervisor
,
B. Wilson, Engineering Supervisor
R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
J. Moyer, I&C/ Electrical Maintenance Supervisor (Unit 1)
R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
R. Poulk, Senior.NRC Regulatory Specialist
D. Novotny, Senior Regulatory Specialist
W. Dorman, QA - Supervisor
4
W. Hatcher, Security Supervisor
.
W. Murray, Senior Engineer - Nuclear Licensing Unit
Other licensee employees contacted included construction craftsmen,
engineers, technicians, operators, office personnel, and security force
members.
,
!
2.
Exit Interview (30703)
i
The inspection scope and findings were summarized on September 5, 1985 with
the general manager.
The licensee acknowledged the findings without
i
exception.
The licensee did not identify as proprietary any of the
j
materials provided to or reviewed by the inspectors during the inspection.
3.
Followup on' Previous Enforcement Matters (92702)
Not inspected.
.
. - _ . . . - ,
- ..
.
- - . - - -
. . _ - . , ,
. . - . -
- _ - , , . - - - - . ,- ,,.-.-,.
_
..
.
a
2
i
4.
Followup on Inspector Followup Items
a.
(Closed) Inspector Followup Item 325, 324/84-35-02; Post Trip Reviews
01-22. An open item was generated after a reactor scram in December
1984 because of a questionable post trip review.
The item was to
follow the licensee's progress toward enhancing the scram review
process.
Revisions 8-10 to the Post Trip Review Procedure 01-22 were
the result of this effort.
These revisions clarify the responsibili-
ties of the operations engineer and other associated requirements.
This item is considered closed.
b.
(Closed)
Inspector
Followup
Item
324/84-31-03;
Standby Air
Compressors.
This item was opened because of a concern over the
inoperability of an automatic start pressure switch for a standby air
compressor.
The licensee has failed to find a suitable replacement,
I
but is presently undertaking a plant modification on Unit 1 (and is
planned on Unit 2) which will alleviate the need for these air
compressors during post accident conditions.
The modification will
install a nitrogen backup system which will supply the necessary
.
pneumatic pressure during accident conditions. This item is considered
!
closed.
I
c.
(Closed) Inspector Followup Item 325, 324/85-03-01; Radwas*e Shipping.
This item was generated to track improvements in the radwaste shipping
,
quality control program and the auditing of this program. This item is
considered closed because a notice of violation 325, 324/85-17-01 was
written in this area and will track corrective actions.
5.
Maintenance Observation (62703)
The inspectors observed maintenance activities and reviewed records to
verify that work was conducted in accordance with approved procedures,
Technical Specifications, and applicable industry codes and standards. The
inspectors also verified that:
redundant components were operable;
j
administrative controls were followed; tagouts were adequate; personnel were
qualified; correct replacement parts were used; radiological controls were
,
proper; fire protection was adequate; QC 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.
4
Outstanding work requests and authorizations (WR&A) were reviewed to ensure
that the licensee gave priority to safety-related maintenance.
a.
Bolting Replacement and Seismic Qualification of Hydraulic Contr^1
Units
Inspection Report 325/85-22- issued a notice of violation for loose
<
and/or missing rack-support-to-foundation bolting for the control rod
!
hydraulic control units (HCU).
During followup of the repair and
replacement, it was observed by the inspector that all the replaced
i
i
_
,
_
_ _ - , _ _ , _ _ _
_____
,_. _
.
.
3
!
!
bolts (five) had been replaced with bolts which were not cadmium plated
as required by plant design drawing G.E. 919D615.
The apparent root
cause was that the maintenance planner verified that Q bolting was
,
required but failed to check the specifications for any special
requirements.
Discussion with licensee personnel indicates that this
may have been a common practice when replacing bolting. The licensee
is currently evaluating the impact of this on plant equipment.
Failure to install the type of bolting specified on G.E. 9190615
drawing is a violation of 10 CFR 50, Appendix
B,
Criterion V
(324/85-27-01):
Bolts Replaced on Hydraulic Control Units with Type
Other than that Specified on Drawings.
The licensee has attempted to establish the seismic qualification of
the as-found conditions documented in inspection report 325/85-22,
i.e., one HCU had 2 out of the 4 rack-support-to-foundation bolts
missing. Calculations were performed on the HCU's with missing bolts,
on HCU's which stand alone and on HCU's installed back-to-back with all
fasteners properly installed. The calculations show the as found HCU's
with loose or missing bolts met at least short term criteria (IEB 79-14),
i.e.,
bolts might deform but would not break.
The same
conclusion was deterrrined for the stand alone HCU's. However, in all
cases including back-to-back installation, stresses were calculated
which exceeded the allowable stress in the tubular frame. The original
i
seismic qualification was performed by the vendor (G.E.) based upon
results of field tests with the units tested back-to-back. This field
test data is not available to the licensee at this time. Without field
test data, the complexity of the installed configuration requires
several conservative assumptions to be made to allow analytic modeling.
The licensee believes that their calculated results are conservative.
Therefore, since the frame was qualified by the vendor from experi-
mental data, the licensee believes the frames to be qualified and the
HCU's are operable, i.e. seismically qualified. However, the licensee
expects to resolve this apparent discrepancy between their analytic
model and the original seismic qualification based on field data.
Resolution of this apparent discrepancy is an unresolved item
1
(325/85-27-01 and 324/85-27-02):
Seismic Qualification of HCU Frame,
b.
Post Maintenance Test Requirement Test Sheet Fails to Specify Pressure
Test /VT-2 Inspection
During a routine inspection of post maintenance surveillance testing
,
'
for the Unit I standby liquid control injection check valve C41-F007,
the inspector noticed that no pressure test /VT-2 inspection was
specified on the Post Maintenance Test Requirement (PMTR) sheet even
though the pressure boundary for the valve had been broken during
maintenance. The valve is Class 1.
Two similar maiatenance activities
4
involving Class 1 valves B21-F028B and B21-F019 were also reviewed and
found to not contain the pressure test /VT-2 inspection requirement on
i
'
4
- - - . ,
-. -
-
,-, - _ - - -
-- -- ,- - - - - -
_ -
_ _ _
.-
-- . _ = - -
-
. _ _ - - _ _
- - - .
_.
.
.
>
.
4
the PMTR sheet. It became clear that the maintenance planners were not
using ENP-16, Inservice Inspection Requirements, properly.
After
scoping the jobs and realizing that the work involved disassembly of
the valves, the planners should have realized that the pressure
boundaries of each of these valves was to be broken and that entry into
Section VI, Visual Inspection, of ENP-16 was necessary to determine
additional post maintenance test requirements.
This process was not
done.
,
i
The problem was discussed with plant management and the following
!
immediate corrective actions were taken:
(1) Training was conducted for all mechanical maintenance planners in
'
the proper use of ENP-16.
(2) A review was conducted of PMTR's for Class 1 valves worked during
the Unit 1 outage.
For those which required pressure test /VT-2
inspections, an additional test / inspection was included on the
PMTR sheet. Those activities already closed out, were reopened by
an additional PMTR sheet stating the required test.
At the completion of the present Unit 1 outage, a vessel
hydrostatic test and inspection including all Class 1 piping and
components is to be performed (PT-80.1, 10 year Inservice
Inspection Reactor Vessel Hydrostatic Test). This test would have
satisfied the inservice inspection requirements for the valves
identified. Followup of long term corrective actions will be an
-
inspector followup
item (IFI 325/85-27-02): Administrative
!.
Control Changes to Ensure Pressure Test /VT-2 Inspections are
Identified as Post Maintenance Requirements.
f
6.
Surveillance Observation (61726)
l
)
The inspectors observed surveillance testing required by Technical
Specifications.
Through observation and record review, the inspectors
verified that:
tests conformed to Technical Specification requirements;
administrative
controls were
followed;
personnel
were
qualified;
instrumentation was calibrated; and data was accurate and complete.
The
inspectors independently verified selected test results and proper return to
4
service of equipment.
A special review was performed of the following Licensee finding:
On August 21, 1985, the Maintenance Surveillance Test (MST) rewrite group
discovered that Periodic Tests, PT-A22.2-1,
PT-22.2-2,
PT-A24.2 and
PT-45.2.4, covering Secondary Containment Isolation Response Time Testing,
did not adequately test all the relays in the associated logic circuit.
,
Technical Specification Surveillance 4.3.2.3 requires that this be done
'
every 18 months.
i
,
-
, , - .
,_
-, _ _ . _ . _
_
_ _ _ _
~ - .
.
..-..__.
._,
,
.
-
_
. - - _
.
,
.
_
_
_
.
_- -
4
.
i
- J
5
,
Specifically, relays K66, K67, 3AA, 3AB, 3BA, 3BB, 3BD, A-CRMX and B-CRMX
,/
were not being included in the response time test. At the time, Unit I was
shutdown for a refueling outage (no core alterations were in progress) and
Unit 2 was at 100% power.
The surveillance test problem involved both
>
units.
4
Licensee management conducted a Plant Nuclear Safety Committee (PNSC)
meeting concerning this problem and concluded that there was no technical
reason to consider this instrumentation inoperable based on the following:
,
a.
Relays not presently being timed have been verified operable in logic
4
system functional tests which were performed in October 1984 (Unit 2).
b.
The manufacturer's expected response time for these relays is less than
85 milliseconds.
c.
The allowed response time for the instrumentation is less than or equal
to 13 seconds.
Adding the relay's expected response time to the
existing instrumentation response time still results in a response time
of less than or equal to 1 second.
A special test procedure was generated to test the relays (SP-85-086) and
'was satisfactorily performed for Units i and 2 on August 25, 1985.
,
These inadequate procedures constitute a violation of Technical Specifi-
i.
cation Surveillance 4.3.2.3,.in that they failed to adequately response time
1,
test all the necessary relays. However,10 CFR i Appendix C,Section V,
paragraph A, states that a notice of violation will generally not be issued
4
if a violation meets 5 stated criteria. This violation meets these criteria
and no notice of violation will be issued.
A permanent procedure to conduct the testing will also be written and
,
i
implemented by the MST rewrite group prior to the end of the next
surveillance interval.
No violations or deviations were identified.
,
7.
Operational Safety Verification (71707) (71710)
The inspectors verified conformance with regulatory requirements by direct
observations of activities, facility tours, discussions with personnel,
reviewing of records and independent verification of safety system status.
The inspectors verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met.
Control room, shift
_
supervisor, clearance and jumper / bypass logs were reviewed to obtain
i
information concerning operating trends and out of service safety systems to
ensure that there were no conflicts with Technical Specifications Limiting
j
Conditions for Operations.
Direct observations wera conducted of control
!
!
,
- - - .
- - . +
- . ~ . , , - - , - - .
..
s-.
-.
,,
,a,.
. ~ . . ,
,
. - , . , - , .
,.m
,,-,r-+-.
-, - - , , - , , -
. - - -
-
_
_
._
.
.
.
.
.
6
.
room panels, instrumentation and recorder traces important to safety to
verify operability and that parameters were within Technical Specification
limits. The inspectors observed shift turnovers to verify that continuity
of system status was maintained.
The inspectors verified the status of
selected control room annunciators.
J
Operability of a selected ESF train was verified by insuring that: each
accessible valve in the flow path was in its correct position; each power
supply and breaker, including control room fuses, were aligned for
components that must activate upon initiation signal; removal of power from
,
those ESF motor-operated valves, so identified by Technical Specifications,
'
was completed; there was no leakage of major components; there was proper
lubrication and cooling water available; and a condition did not exist which
might prevent fulfillment of the system's functional
requirements.
Instrumentation essential to system actuation or performance was verified
operable by observing on-scale indication and proper instrument valve
lineup, if accessible.
The inspectors verified that the licensee's health physics policies / pro-
cedures were followed.
This included a review of area surveys, radiation
l
work permits, posting, and instrument calibration.
The inspectors verified that: the security organization was properly manned
and that security personnel were cepable of performing their assigned
functions; persons and packages were checked prior to entry into the
protected area (PA); vehicles were properly authorized, searched and
escorted within the PA; persons within the PA displayed photo identification
badges; personnel in vital areas were authorized; effective compensatory
measures were employed when required; and security's response to threats or
alarms was adequate.
The inspectors also observed plant housekeeping controls, verified position
of certain containment isolation valves, checked clearances, and verified
the operability of onsite and offsite emergency power sources.
No violations or deviations were identified.
8.
Onsite Review Committee (40700)
'
The inspectors attended selected Plant Nuclear Safety Committee 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.
No violations or deviations were identified.
,
i
!
_
_
..
_.
_ _ . _ _
_
_.
.-
-.
..
-
. _ - .
.
.
7
9.
Onsite Review of Licensee Event Reports (92700)
The listed Licensee Event Reports (LER's) 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, licensee conditions and commitments.
The following
reports are considered closed:
(Closed) LER 1-80-20; Containment monitoring system isolated due -to
personnel error.
(Closed) LER 1-81-34; Four supports were found damaged due to water hammer.
(Closed) LER 1-83-10; Fire barrier / secondary containment seal degradation
allows water to leak into reactor building.
,
(Closed) LER 1-83-23; Fire in 4160/480 volt E-6 transformer.
(Closed) LER 1-83-26; Diesel generator trips due to operator failing to
follow procedure.
(Closed) LER 1-83-32; Control rods have no position indication.
(Closed) LER 1-83-36; Control power fuse to motor operator blew due to
ground in circuit.
(Closed) LER 1-83-40; Inadequate surveillance procedure and personnel error
cause HPCI to isolate.
,
(Closed) LER 1-83-62; Reactor building exhaust ventilation radiation monitor
actuated outside technical specification limit.
(Closed) LER 1-84-01; Air entrapped in suction header caused residual heat
removal service water pumps to trip.
!
(Closed) LER 1-84-29; Spurious actuation of control building emergency air
filtration system.
i
t
(Closed) LER 1-84-30; Spurious actuation of control building emergency air
filtration system.
(Closed) LER 1-84-31; Spurious actuation of control building emergency air
i
filtration system.
(Closed) LER 1-85-02; Automatic isolation of the control building heating,
ventilation and air conditioning system due to spurious chlorine signal.
i
I
__
-
. _ _ . . _ ,
.
- .
8
(Closed) LER 1-85-03; Inadequate logic system functional testing of degraded
and under-voltage relays of emergency buses.
(Closed) LER 1-85-05; Inadequate functional testing of rod block monitor.
(Closed) LER 1-85-07; Spurious actuation of control building emergency air
filtration system.
(Closed) LER 1-85-14; Steam leak causes reactor core isolation cooling
system isolation.
(Closed) LER 1-85-17; Loss of emergency bus E-1 normal feed.
(Closed) LER 1-35-19; Control building emergency air filtration system
actuation due to accidental shorting of detector circuit.
(Closed) LER 1-85-22; Standby gas treatment train IA relay over heats and
fails.
(Closed) LER 2-83-41; A " rod drift" annunciation was received due to the rod
position indication probe.
(Closed) LER 2-83-46; Wires to terminals in the terminal box were reversed.
(Closed) LER 2-83-52; Transmitter shorted to ground when alligator clip test
leads were accidentally bumped.
(Closed) LER 2-83-62; Prima ry containment temperature exceeds technical
specification limit as result of seasonal ambient temperatures.
(Closed) LER 2-83-65; Condensate storage tank level switches for High
Pressure Coolant Injection (HPCI) system improperly installed.
(Closed) LER 2-83-71; Suppression pool temperature exceeds limit as result
of HPCI run.
(Closed) LER 2-83-82; Suppression pool level exceeds limit due to personnel
error.
(Closed) LER 2-83-84.
This item was voided by the Licensee.
(Closed) LER 2-83-94; Reed switch problems cause incorrect position
indication.
(Closed) LER 2-84-04; Reactor scram initiated by HFA relay replacement and
surveillance testing.
[
'
(Closed) LER 2-84-06; Procedure failed to identify the need for jumping the
low low level signal.
.
__
. - _ _ _ .
.__
_ .
.
- __.
_._
.
.
.
4
3
i
9
4
(Closed) LER 2-85-01; Surveillance test not performed within allowable time
due to inadvertent deletion from scheduling system.
(Closed) LER 2-85-03; Inadequate surveillance procedure results in an
unexpected group 1 isolation.
No violations or deviations were identified.
10.
Refueling Activities (60710)
During the licensee's refueling operations, the inspectors verified that
i
selected surveillance testing required by Technical Specifications was
!
current and that the licensee's fuel- handling procedure was implemented.
'
The following additional items were verified:
a.
Selected fuel bundle movements,
b.
Core monitcring during refuel operations was in accordance with
Technical Specifications.
c.
Vessel water level was maintained in accordance with Technical
Specification.
1
i
,
'
d.
Reactor mode switch position was as required by Technical Specifi-
cation.
f
e.
Continuous communications were maintained between the refueling
,
platform and the control room and that . control room operators were
1
cognizant of the applicable procedure steps.
f.
Health-Physics personnel maintained constant coverage of all fuel
j
moving activities, ensuring area dose rates, contamination levels and
airborne samples were within required tolerances.
No violations or deviations were identified.
11. Modification Process and Masonry Walls (37700)
i
)
i
'
During review of design activities, the inspector reviewed the licensee's
method for ensuring that safety related equipment would not be installed on
-
j-
or in close proximity to masonry walls whose failure could affect the
equipment (Bulletin 80-11 concern).
The method currently employed relies
+
upon -notes on the wall drawings to inform the user of the wall status
(analyzed or unanalyzed) and their general engineering practice of. not
installing new supports onto masonry walls.
These appear to adequately
address installation onto walls but do not provide a positive means to
j
ensure that safety related equipment is not placed close to an unanalyzed
!
i
l
I
,, + - ~ - - -
.,
, , - - - -
,w.,
,-n--
, , - - - . .
n,.
~e--
,
.,,,,.,..p
o-m-,,--m.v
w w.
,.
mg-m m.-
--.,--,,,. ,,e
.
.
10
wall (walls which did not have safety related equipment around them were not
analyzed per Bulletin 80-11).
This is an inspector followup item
(324/85-27-03 and 325/85-27-03): Enhancements of controls to preclude
installation of safety related equipment in proximity to unanalyzed masonry
walls.
No violations or deviations were identified.