ML20127E517
| ML20127E517 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/30/1985 |
| From: | Fredrickson P, Garner L, Hicks T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127E462 | List: |
| References | |
| 50-324-85-05, 50-324-85-5, 50-325-85-05, 50-325-85-5, NUDOCS 8505200105 | |
| Download: ML20127E517 (9) | |
See also: IR 05000324/1985005
Text
.
p rt;
UNITED S FATES
o
NUCLEAR REGULATORY COMMISSION
y
~,
REGloN il
g
j
101 MARIETTA STREET, N.W.
's
ATLANTA, GEORGI A 30323
%,...../
Report Nos.: 50-325/85-05 and 50-324/85-05
Licensee:
Carolina Power and Light Company
411 Fayetteville Street
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: March 1-31, 1985
M
[/Jd[87
Inspectors:
m[
%
L.W. Garner ~,ActingSeniorResidentInsptor
Date Signed
1
Y
bi &
dh[
T.
ira;, Resident nspector
g
Date Signed
.
Approved by:
-
-
4[30/l'6
P. E. Fredrickson, Section Chief
Da'te Si'gned
Division of Reactor Projects
SUMMARY
'
Scope: This routine safety inspection entailed 280 inspector-hours on site in
the areas of. surveillance, maintenance, operational safety verification, ESF
System walkdown, in-office and on-site Licensee Event Report review, independent
inspection and modification review.
Results:
One violation was identified in one area - " Failure To Follow
Surveillance Procedure PT 1.17PC" (paragraph 6).
l
.
!
850506
05000324
0505200105ADOCK
PDR
G
'
_ _.
-
.
.
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
C. Blackmon, Superintendent - Operations
- L. Boyer, Director - Administrative Support
- J. Chase, Manager - Operations
- G. Cheatham, Manager - Environmental & Radiation Control
R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
- C. Dietz, General Manager - Brunswick Nuclear Project
W. Dorman, QA - Supervisor
-
- K. Enzor, Director - Regulatory Compliance
W. Hatcher, Security Specialist
- R. Helme, Director - Onsite Nuclear Safety - BSEP
- B. Hinkley, Manager - Technical Support
W. Hogle, Engineering Supervisor
J. Holder, Manager - Technical Support
P. Hopkins, Director - Training
- P. Howe, Vice President - Brunswick Nuclear Project
L. Jones, Director - QA/QC
' R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
J. Moyer, I&C/ Electrical Maintenance Supervisor (Unit 1)
D. Novotny, Senior Regulatory Specialist
G. Oliver, Manager - Site Planning & Control
- J. O'Sullivan, Manager - Maintenance
R. Poulk, Senior NRC Regulatory Specialist
L. Tripp, Radiation Control Supervisor
V. Wagoner, Director - IPBS/Long range Planning
- J. Wilcox, Principle Engineer - Operations
B. Wilson, Engineering Supervisor
Other licensee employees contacted included technicians, operators, and
engineering staff personnel.
- Attended exit interview
2.
Exit Interview-
The inspection scope and findings were summarized on April 2,1985, with
those persons indicated in paragraph one above.
Meetings were also held
with senior facility management periodically during the course of this
inspection to discuss the inspection scope and findings. The licensee did
not identify as proprietary any of the materials provided to or reviewed by
the inspectors during this inspection.
o
.
.-
2
3.
Licensee Action on Previous Enforcement Matters
(Closed) Unresolved Item 325/84-31-02.
This item involved two concerns
relative to the operation of the Unit 1 Standby Gas Treatment System (SGTS).
Further investigation was necessary in order to verify the original design
requirements (see Inspection Report 84-35).
The first concern dealt with the operation of the Unit 1 SGTS train A and B
inlet and outlet dampers (B,
C,
E and G-BFV-RB).
It was initially
understood that the Final Safety Analysis Report (FSAR), described the
dampers as having automatic open capability. After reviewing correspondence
between the licensee and the A/E (designers of the SGTS), along with
original startup data, it can be verified that the original design was to
have these dampers normally open during operation and that no requirement
existed to have automatic open capability.
The dampers serve only as
maintenance isolation valves. The licensee does intend to clarify both FSAR
and the system descriptions regarding the operation of these dampers. This
concern is resolved.
The second concern was relative to the disparity between the SGTS damper
operation for each unit. Unit 2 SGTS train A and B inlet and outlet dampers
do have automatic open capability. The FSAR makes no statement regarding
this difference. The A/E explained that during construction of Unit 2 (Unit
2 was built before Unit 1), a modification was made to the Unit 2 SGTS in
order to allow the system to automatically isolate itself from the drywell
during an accident.
Included in this modification was the installation of
automatic open circuits for the train's inlet and outlet dampers.
Subsequent to the modification, an Engineering Review altered the Unit 1
SGTS design from that of Unit 2 during initial construction. This review
determined that the automatic open capability for these two dampers (per
train) was not necessary. Consequently, the Unit 1 SGTS was built with the
new automatic isolation capability but, the automatic opening function of
the train inlet and outlet dampers was deleted. Again, the proposed FSAR
change will clarify the disparity between units.
This concern and the
unresolved item are considered closed.
No violations or deviations were identified.
4.
Review of Licensee Even't Reports (92700)
The below-listed Licensee Event Reports (LER) 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 the
inspectors concluded that necessary corrective actions have been taken in
accordance with existing requirements, licensee conditions and commitments.
These reports are considered closed.
-
,
..
3
LER .1-83-03 - An under reactor vessel inspection revealed that detector
cables were_ separated from their associated detectors.
LER 1-83-06 - Insert / withdrawal positions and drive power indication for
SRM's and IRM's ~were not working. Control power supply fuse blown.
LER 1-83-07 -- One fuel bundle was located around each of the withdrawn
LER 1-83-08 - HVAC System exhaust inboard isolation damper open limit switch-
sticking.
LER 1-83-15 - Well water isolation valves to both Standby Gas Treatment
System Train's Deluge Systems were closed rendering both deluge systems
i.
LER 1-83-19 - Standby Liquid Control System Relief Valves lifted at 1321 psi
and 1592 psig, respectively.
LER 1-83-20 - IRM
"A" was showing- instrument upscale indication from
moisture accumulation.
No violation or deviation was identified in this area.
5.
Maintenance Observations (62703)
Maintenance activities were observed and reviewed throughout the inspection
period to verify that activities were accomplished using approved procedures
or the activity was within the skill of the trade and that the work was done
'by qualified personnel.
Where appropriate, limiting conditions for
operation were examined to ensure that, while equipment was removed from
_
service, the Technical Specification requirements were satisfied.
Also,
work activities, procedures, and work requests were reviewed to ensure
adequate fire, cleanliness and radiation protection precautions were
observed, and that equipment was tested and properly returned to service.
Acceptance criteria used for this review were maintenance procedures and
Technical Specifications.
Outstanding work requests that were initiated by the operations group for
Units 1 and '2 were reviewed to verify the licensee is giving priority to
safety-related maintenance and not allowing a backlog of work items - to
permit a degradation of system performance.
1No violations or deviations were identified.
6.
- Surveillance Testing (61726)
Selected surveillance tests were analyzed and/or witnessed by the inspector
to ascertain procedural and performance adequacy.
the completed test
i
procedures examined were analyzed for embodiment of the necessary test
prerequisites,
preparations,
instructions,
acceptance
criteria
and
'
.
.
~
4
sufficiency of technical content.
The selected tests witnessed . were
examined to ascertain -that current, written approved procedures were
available and in use, .that. test equipment in use was calibrated, that test
prerequisites were met, system restoration was completed and test results
were adequate. The selected procedures attested conformance with applicable
Technical' Specifications, they appeared to have received the required
administrative review and they were performed within the surveillance
frequency prescribed.
- Acceptance criteria for evaluating surveillance tests were 10 CFR, ANSI
N18.7Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..and Technical Specifications.
During the performance of Surveillance Test PT-1.1.7PC, Average Power Range
Monitor (APRM's) Channel Calibration, techniciahs' caused individual APRMs to
be inoperable without placing them in bypass as required by the procedure.
This action caused a half Reactor Protection System trip (half scram) when
one APRM sensed only 10 Local Power Range Monitor (LPRM) inputs vice the
minimum 11. Unit I was operating at approximately 60% of power.
,
'
Technical Specification 3.3.1, requires that each ApRM channel have at least
- two LPRM inputs - per level and eleven total LPRM inputs in order .to be
considered operable.
Less than 11 inputs will cause an APRM channel to
trip.
There are six APRM channels divided into two Reactor Protection
System (RPS) channels which have three APRM's each. Technical Specification 3.3.1, also requires that at least two operable APRM's be in service for
each RPS channel.
l
PT-1.1.7PC requires that LPRM's, be calibrated prior to calibration of the
4
associated APRM.- Each LPRM is calibrated individually by placing the LPRM
card selector switch to "By" (Bypass), which then permits the technician to
perform the necessary adjustments.
However, this action also removes that
LPRM from its' associated APRM. To account for this, the procedure includes
a step to bypass each APRM'while its associated LPRMs are being calibrated.
- Failure to have _ an APRM . bypassed when the eleventh LPRM was bypassed -
. resulted in the half scram.
The root cause of the problem is the way in which the testing crews handle
the turnover of surveillance tests which carry over from one shift to the
next.
PT-1.1.7PC was begun on March 12, 1985, during day shift.
It had
been continued through the swing shift but stopped before midshift.
The
test was then continued on the following day shift. Howaver, one of the
prerequisites for continuing the test (which had previously been met) was to
ensure Lthat ~ operators bypassed the applicable APRM at the Reactor Turbine
Generator Board-in the Control Room prior to the LPRM calibrations. This
step was not re performed when the test was restarted. Although the shift
operators had given permission to continue the test, the _ technicians
informed them that no APRM's would be made inoperable. This information was
incorrect.
The consequences of this action was to place APRMs out of service without
shift operating personnel permission or knowledge. Each APRM already had at
.
.
5
l
1 east one LPRM level -with only two inputs remaining (due to various
unrelated problems). Consequently, when the LPRM calibration was performed,
each APRM was sequentially made inoperable and then, subsequently, returned
to service. At no time was more than one APRM inoperable in an RPS channel.
Technical Specification 6.8.1(c) requires that written procedures be
implemented for surveillance and test activities of safety related
equipment. Contrary to TS 6.8.1(c), requirements described in PT 1.1.7PC
step VII D.3.(a) were not met upon resumption of the surveillance test after
an eight-hour postponement in that the APRMs were not bypassed during the
individual LPRM calibration (violation 325/85-05-01).
The immediate
corrective actions were - to: (1) Revise PT 1.1.7PC to add precautions
concerning minimum. numbers of LPRM channels required and to better define
steps which require APRMs to be bypassed; (2) Conduct training of testing
crews as to the relationship between APRMs and LPRMs as well as general APRM
operation; (3) Counsel the technicians involved and testing crews regarding
t.
the necessity to review previously performed sections of any surveillance
test they start which had been temporarily interrupted. These actions have
been completed.
One violation was identified in this area.
7.
Operational Safety Verification (71707, 71710)
The inspector verified conformance with regulatory requirements throughout
the reporting period by direct observations of activities, tours of
facilities, discussions with personnel, reviewing of records and independent
verification of safety systems status.
The following verifications were
.made:
-
Control Room Observations - The inspectors verified that control room
manning requirements of 10 CFR 50.54, and the Technical Specifications
were being met.
Control room, shift supervisor, clearance and
jumper / bypass logs were _ reviewed to obtain information concerning
operating trends and out of service safety systems to insure that there
were no conflicts with Technical Specifications Limiting Conditions for
Operations. Direct observations were conducted of control room panels,
instrumentation and recorder traces important to safety to verify
operability and that parameters were within Technical Specification
limits. In addition, the inspectors observed shift turnovers to verify
that continuity of system status was maintained and questioned shift
personnel relative to their awareness of plant conditions.
The
inspectors verified the status of selected control room annunciators
and were assured that the control room operators understood the reasons
why impo-tant annunciators were lit.
In addition, periodic verifi-
cations were conducted to insure that corrective actions, if appro-
priate, were initiated and completed in a timely manner.
ESF Train Operability - Operability of selected ESF trains 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
i
y
.
2
6
fuses, are 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; a condition did not exist which might prevent
fulfillment of the train's functional requirements.
In addition,
-
instrumentation essential to system actuation or performance was
verified operable by. observing on scale indication and proper
instrument valve lineup, if accessible.
The High Pressure Coolant
Injection System-(Unit 1) was verified operable.
Radiation- Protection Controls - The inspectors' verified that the
-
licensee's health physics policies / procedures are being followed,
including . area surveys, RWP's, posting a'nd calibration of selected
radiation protection instruments in use.
Physical Security Plan - The inspectors verified that the security
'
--
organization is properly manned and that_ security personnel are capable
.of performing their assigned functions, that persons and packages are
,
checked prior to . entry into the Protected Area (PA), vehicles are-
properly authorized, searched and escorted within the PA, persons
within the PA display. photo identification badges, . personnel in vital
areas are authorized, that effective compensatory measures are employed
when required, and that security's response to threats or alarms
appears adequate.
Plant . Housekeeping - Observations relative to plant housekeeping
-
identified no unsatisfactory conditions.
Containment Isolation - Selected containment isolation valves were
H
--
verified to be in their correct positions.
Radioactive Releases - The inspectors verified that selected liquid and
'
gaseous releases were made in conformance with 10 CFR 20 Appendix B and
Technical Specification requirements.
,
No violations or deviations were identified.
- 8.
' ADS Valve'Not Connected to Accumulator per Design (37700)
On March .12, 1985, with Unit 2 shutdown, the licensee discovered that
~ Automatic Depressurization System (ADS), valve B21-F013C solenoid, was being
'
- supplied 'with instrument air. from a line which did not contain an
accumulator. The- condition was found when maintenance personnel attempted
to- isolate the solenoid on the
"D"
ADS valve to repair an air leak.
Isolation of valves specified in the procedure, failed to remove air f rom
lthe
"D" _ ADS valve.
Walkdown of the instrument air tubing between the
-accumulators and the ADS valve's revealed the following:
a.
"C" ADS valve was receiving air from a line which had no accumulator.
.
.
..
7
b.
"D" ADS valve was receiving air from the accumulator tagged for "C" ADS
valve.
c.
"E" manual _ Safety Relief Valve (SRV), was re.ceiving air from the
accumulator tagged for "D" ADS valve.
d.
ADS valve "H"
and "J" had their air tubing connected to accumulators
for "J" and "H"
respectively.
e.
- Manual SRV "G" and "F" had their air tubing interchanged.
Of>these, only item a. has any safety significance in that an ADS valve did
-not have an accumulator to supply capability'to open the valve and-hold the
valve open upon' failure of the normal air supply,'as described in the Final
Safety Analysis Report, paragraph 5.2.2.4.
As documented in the safety
evaluation entitled ' Verify Qualification of Accumulator on ADS Valves'
.
dated June 15, 1984, the NRC found acceptable that (1) the accumulators are
used only as snubbers in the system and are not relied upon to maintain
pressure to the ADS valve actuators and (2) the standby compressor system
will supply the required air pressure under postulated accident conditions.
Hence, item a. has only. minor safety significance.
Apparently these problems were caused by installation of modification 80-086
in the summer of 1982. Mod 80-086 changed the SRV's from three stage.to two
stage valves. The modification procedure 80-086 did not address the removal
or _ reinstallation of the air tubing to the solenoids which allow remote
actuation of the SRV's. Specifically, the tubing had been disconnected at a
point at which the tubing had been bunched into groups to allow penetration
through the floor grating. Vhen they were reconnected, several of the tubes'
were mismatched.
.
The root cause of the event was attributed to personnel failing to follow
ENP-03, Plant Modification Procedure, in that work'was performed as part of
the modification but was outside the scope as defined in Plant Modification
80-086, a condition not authorized by ENP-03.
This is a violation of
Technical Specification 6.8.1.(a) which requires that procedures be
implemented.
However this event occurred prior to the -enhancements
-associated with the Brunswick Improvement Program which resulted in
clarification and upgrading of management expections in the areas of-
adherence to administrative controls and attention to details while
-performing tasks. These were communicated to all staff members and have
.been incorporated into the daily conduct of business at the site. As part-
of.the general employee training, each employee sees a film' emphasizing
corporate commitment to quality and procedural compliance. As a result of
- the discovery- of this event, a review of current training . practices was
. conducted of the affected organizations. No changes to.the current programs
were deemed necessary.
Therefore because corrective action has been-
accomplished as part of the Brunswick Improvement Program 'and because the
event had minor safety significance in that only a redundant design feature
was unavailable which did not render an ADS valve inoperable, the event is
!
.
.-
- .a_-..~,__
-. -
- . .
. .
. .
.
. - . . .
-
~ .
.
. . .
. -
.,
.
8
classified as a licensee identified violation per 10 CFR 2 Appendix C
paragraph IV.A.
At the time of discovery Unit I was operating; ha aver, the event was
considered as not applicable to Unit 1 in that all SRV's, both manual and
ADS, have accumulators.
No violation or deviation was issued in this area.
9.
'Onsite Review Committee (40700).
The inspectors attended several special Plant Nuclear Safety Committee
meetings conducted during the report period.
The inspectors verified the following items:
i
Meetings were conducted in accordance with Technical Specification
-
requirements regarding quorum, membership, review process and personnel
qualifications;
Corrective actions, recommendations and decisions were completed as
-
assigned.
No violations or deviations were identified.