ML20247K477
| ML20247K477 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/17/1989 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20247K434 | List: |
| References | |
| 50-285-89-13, NUDOCS 8906010308 | |
| Download: ML20247K477 (34) | |
See also: IR 05000285/1989013
Text
-_--
t.
r-
1i
(.
j
,
i
I
- t
APPENDIX B
.
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV ,
,
,
,
NRC Inspection. Report:
50-285/89-13-
I.
License:
Docket:
50-285
Licensee: Omaha Public Power District (OPPD) .
~F
1623 Harney Street
Omaha, Nebraska 68102
3
Facility Name:
Fort Calhoun Station (FCS)
.
Inspection At:
Inspection Conducted: March 1-31, 1989
Inspectors:
P. H. Harrell, Senior Resident Inspector
T. Reis, Resident Inspector
'
Approved:
T. F. Westerman, Chief, Project Section B
Date
Division of Reactor Projects
Inspection Summary
Inspection Conducted March 1-31, 1989 (Report 50-285/89-13)
Areas Insoected:
Routine, unannounced inspection including review of
previously identified items; licensee event report followup; operational safety
verification; plant tours; safety-related system walkdown; monthly maintenance
observations; monthly surveillance observations; security observations;
radiological protection observations; in-office review of periodic, special,
and nonroutine event reports; and followup of onsite events.
Results: During tours of the plant, the NRC inspector noted a variety of items
that were identified to the licensee as concerns. After reviewing the
concerns, two items were identified es violations and two items were identified
as unresolved items in this inspection report. During the previous inspection
period, as documented in NRC Inspection Report 50-285/89-09, nine items were
identified during tours of the plant by the NRC inspector. Based on the number
of items being identified by the NRC inspector, it appears that the licensee
needs to concentrate on performing sufficient plant tours of an appropriate
quality to ensure that nonconforming conditions are being readily identified.
During this inspection period, the NRC inspector found that the licensee had
taken nonconservative actions with respect to potentially degraded fire
barriers. The NRC inspector noted that it appeared that the fire doors in the
auxiliary building were nonfunctional due to a hole in the door frame. At the
time the NRC inspector identified this concern to the licensee, it was not
8906010308 890519
ADOCK 05000285
4
Q
{
-
_ _ _ _ _ _ _ _ _ _ - - _ _ _ _ - _ - _ _ _ _ _ _ - _ - - _ - _ _
_ _ - _ - _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
-
_ - _ _ _ _ - _ _
_________-__Q
_
_ _ _ _ - - .
.
.
"
^....
2
'
,.
,
,
'known whether or not the fire doors were functional. The licensee subsequently '
4
performed an evaluation to verify that the fire barriers were adequate as ..
.-
, 3.'zinstalled based on the limited amounts of combustible material located in the
'
fire areas separated by the barriers.
-
*
,
3w
l.:.
The licensee' experienced three problems in the area of security.
The., problems
~ included issuance of security badges and the transmittal of safeguards
,
-
1
- Linformatio'n'., In each case it appeared that the licensee took timely and
,
,
? appropriate actions to correct the identified problems.
'
S
- Four plant events were addressed during this inspection period by the licensee.
-
The' events included the loss of, and a near loss of, the offsite power
supplies, and plant equipment problems. During review of each event by the NRC'
inspectors, it appeared that the licensee took proactive, timely, and
comprehensive actions to resolve each problem. The actions taken by the
licensee were prudent and appeared to be directed toward the safe operation of
.
,
the plant. During a wind storm that caused voltage fluctuations on the 345-kV
L.
offsite power supply, the NRC inspector observed the actions of the enshift
e.
,
operations crew in the control room. The NRC inspector noted that the
operations crew was responsive to control room alarms, performed their tasks in
a professional manner, and took actions that were based on safe plant
operation.
_l
,
,
_ _ - - _ _ _ _ - - _ - . _ _ _ _ . _ _ _ _ - _ - . - . - - - - _ _ - - - - - -
- _ _ _ _ _ _ _ _ _ _
'
-
,
,,
,
i
3
l
'
-
DETAILS
1.
Persons Contacted
'
- K. Morris, Division Manager, Nuclear Operations
- W. Gates,11anager, Fort Calhoun Station
J. Adams, Reactor Engineer
'
- J. Bobba, Supervisor, Radiation Protection
C. Brunnert, Supervisor, Operations Quality Assurance
- J. Fisicaro,llanager, Nuclear Licensing and Industry Affairs
- R. Fussell, Supervisor, Materials
'*J. Gasper, Managar, Training
- J. MacKinnon, Acting Division Manager, Production Engineering Division
- H. Hodges, Specialist, Safety Review Group
- R. Jaworski, Manager, Station Engineering.
(J. Kecy, Supervisor, System Engineering
- D. Lieber, Supervisor, Security Operations
D. Matthews, Supervisor, Station Licensing
- K. Miller, Supervisor, Maintenance
- G. Peterson, Assistant Manager, Fort Calhoun Station
.
- R. Phelps, Manager, Design Engineering Nuclear
- D. Ritter, Security Services
- C. Simmons, Station Liccnsing Engineer
- F. Smith,. Plant Chemist
l
-
'
'D. Trausch, Supervisor, Operations
S. W111 rett, Supervisor, Administrative Services
- Denotes attendance at the monthly exit interview.
The NRC inspectors also cont mted other plant personnel, including
operators, technicians, and administrative personnel.
I
2.
Plant Status
During this inspection period, the FCS operated at full power.
On
March 4, 1989, the licensee lost one source of offsite power, the
161-kV line. The licensee's handling of this situation is discussed in
paragraph 13 of this report.
No other challenges of safeguards equipment
,
occurred.
3.
Review of Previously Identified Items
(92701 and 92702)
a.
(Closed) Open Item 285/8811-07:
Potential loss of pressurizer
pressure transmitters following a postulated pressurizer spray line
break.
This item involved the licensee's activities related to the
reconstitution of the design basis for the FCS. Durir.g these
activities, the licensee identified a potential problem where a break
in the pressurizer spray line could cause the loss of two channels of
i
-___._m...-_
__.__________ . _ _
_
_ _ - .
. _ _ _ _ _ _
t
u
~,
-
.
4
pressurizer pressure indication.
The existence of this potential
problem created a condition where the plant did not meet the design
basis requirements as stated in Section 7.2.9 of the Updated Safety
Analysis Re,Trt (USAR).
In response to this problem, the licensee performed an analysis of
the potential for a break in the pressurizer spray line. The
analysis concluded that a. leak would occur prior to a break,
operations personnel would be able to detect a leak from the -
presently installed instrumentation, and operations personnel would
have sufficient time to take corrective actions.
The analysis completed by the licensee was forwarded to the NRC's
Office of Nuclear Reactor Regulation (NRR) for review.
In a letter
dated February 27, 1989, NRR stated that a review had been completed
of the licensee's analysis and that the analysis satisfactorily
established that the postulation of circumferential and longitudinal
i
pipe breaks on this section of piping were not required.
Based on the completion of a review of the licensee's analysis by
NRR, it appeared that the presently installed configuration for the
pressurizer spray line is acceptable. No further licensee action is
required on this item.
b.
(Closed) Unresolved Item 285/8315-07: Design basis for a component
cooling water (CCW) line break concurrent with a loss-of-coolant
accidett(LOCA),
i
This item was related to verification that the licensee's design
basis did not require the licensee to design the plant with a break
in a CCW line inside containment concurrent with a LOCA. This design
basis consideration was forwarded to NRR for review.
In a memorandum
dated December 20, 1988, NRR stated that the licensee's design basis
did not require design considerations of a CCW line break inside
containment concurrent with a LOCA. Based on this determination,
this item is considered closed.
c.
(Closed) Open Item 285/8832-06:
Verification of containment
temperatures.
The NRC inspector performed Temporary Instruction (TI) 2515/98,
"Information on High Temperatures Inside Containment /Drywell," during
September 1988. The rescits of this inspection are documented in
paragraph 13 of NRC Inspection Report 50-285/88-32. Open
Item 285/8832-06 was generated in the inspection report pending field
verification by the NRC inspector of the location of the temperature
sensors inside containment. This verification was completed in
October 1988. During this inspection period, the NRC inspector
formally compiled the results of tne analysis and forwarded them to
NRC Region IV management.
i
tl.
_ _ _
_ .
.______--__._-______-__-___-_____-_-_____.)
_-_
_.
.
a
"
.,
5
.
.
d.
(0 pen) Open Item 285/8846-06:
Inadvertent rise in the reactor
coolant system (RCS) water level.
This item was gererated to track licensee corrective action of a
plant event that occurred on December 5, 1988, when the licensee
experienced an unplanned rise in the.RCS water level. The plant was
in a refueling shutdown at the time with no fuel in the reactor
vessel. The NRC inspector reported in NRC Inspection
Report 50-285/88-46 that the source of the water was undetermined
from approximately 4:30-p.m. to 8:30 p.m.
At approximately
8:30 p.m., it was determined that the influent was demineralized
water caused oy an open manual valve (CH-363) used to flush Charging
Pump CH-1A. The change in RCS water level over this period was
,'
approximately 10 inches in the reactor vessel as measured by Level
Indicator LI-197.
'
It should be noted that operators had the capability to stop the
,
influent by closing'the loop charging valves.
It was decided to
leave them open to assist in discovering the leak path.
Incident Report 880480 was generated.on December 7, 1988, describing
the event. The licensee stated the resson for the valve
mispositioning could not be determined.
As corrective action, the licensee revised Procedure OI-CH-1,
" Chemical and Volume Control System - Normal Operation," to change
the status of all demi 1eralized water inlets to charging pumps from a
closed position to a locked-shut valve status. Additionally, the
licensee provided calculations to the NRC inspector demonstrating
that the resultant boron dilution did not exceed the Technical
S) edification (TS) limit of 1800 ppm. The NRC inspector determined
tlat the inadvertent dilution was bounded by the safety analysis of
Section 14 of the USAR.
Further, licensee management stated that all systems which interface
with the RCS will be canvassed to determine if other possibic
influent paths exist that should be locked closed in lieu of closed.
l
This item will remain open pending a response by the licensee
L
indicating all potential influent paths have been surveyed and
addressed, and the appropriate documentation has been revised.
e.
(Closed) tinresolved Item 285/88201-03 (Violation B.2):
Use of
p
operations memoranda (OM) to make procedure changes.
This item involved the licensee's use of oms to provide instructions
h
to operations personnel. The instructions provided in the oms should
'
have been provided to operations personnel by processing a change to
,
the appropriate procedure (s).
The licensee failed to use the
l
appropriate method.
b
L
-
-
-
.
t
a
.,
6
-
As discussed in paragraph 12 of NRC Inspection Report 50-285/89-03,
the NRC inspectors performed an extensive review of the actions taken
by the . licensee to resolve this problem.
Based-on the review
documented in the inspection report, this item is considered closed.
f.
(Closed) Unresolved Item 285/88201-07: Control of drawings for
temporary modifications (TM).
This item was related to the licensee's failure to adequately control
drawings that were affected by the installation of a TM. Examples
were identified where a TM was installed in the plant and the
. appropriate drawing (s) was not updated in a timely manner to reflect
installation of the TH.
<
In NRC Inspection Report 50-285/89-03, the NRC inspectors issued
'
Violation 285/8903-01 to address the apparent problem related to
drawing control. This unresolved item is considered closed as the
licensee's corrective actions will be verif'ed during closecut of the
I
"'
violation.
g.,
(Closed) Open Item 285/8903-02: Approval of a TS amendment for the
'
temperature of the safety injection and refueling water tank (SIRWT).
4
-
This item involved the issuance of a TS amendment request by the
'
,
licensee to raise the minimum al bwable temperature of the SIRWT from
i
40'F to 50'F.
The amendment request was submitted based on concerns
_
related to the capability of the safety injection pumps to pump water
at 40'F.
,
NRR reviewed the amendment request submitted by the licensee.
On
r
February 14, 1989, NRP, issued TS Amendment 119 to approve the
licensee's request for raising the minimum allowable temperature in
the SIRWT; therefore, this item is considered closed.
h.
(Closed) Open Item 285/8903-06: Overpressurization of the charging
This open item originated when an operator ran a positive
displacement charging pump against closed loop-injection valves. This
'
occurred during plant startup while performing Procedure 01-RC-3,
'
l
" Reactor Coolant System Startup." The operator was working per
l
procedure; however, the verification of valve lineup sheets was 6
days old.
The lineup sheets indicated that the discherge valves were
!
caen as required for the establishment of letdown and charging flow.
T1e actual position of the valves was closed indicating they had been
changed since the lineup was verified.
No damage was done to the charging system.
The immediate corrective
)
'
actions taken by the licensee and reviewed by the NRC inspector are
discussed in paragraph 14.c of NRC Inspection Report 50-285/89-03.
_ . _ _ - - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - - _ - - - - _ - _ _ -
, _ _
. _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
-
. _ _ _ _ _
.-
.
.
7
l
'
'
1
To prevent recurrence of this event, the licensee revised
i
Procedure OI-CH-1, " Chemical and Volume Control System - Normal
J
Operation." The revision included a statement in the precautions
l
section of the procedure and specific instructions within the
j
procedure addressing operator assurance that suction and discharge
flow paths exist prior to energizing a charging pump.
The NRC inspector reviewed the revised procedure. Based on the
review, it appeared that the precautionary step and instructions were
adequate to prevent recurrence of this event.
4.
Licensee Event Report (LER) Followup (92700)
Through direct observation, discussions with licensee personnel, and
review of records, the following event report was reviewed to determine
that deportability requirements were fulfilled, immediate corrective
actions were accomplished, and corrective action to prevent recurrence had
been accomplished in accordance with the TS.
During this inspection period, LER 89-004, " Inadvertent Actuation of
Ventilation Isolatiot Actuation Signal," was closed. LER 89-004 reported
an unplanned actuation of the ventilation isolation actuation
signal (VIAS) which occurred during plant startup from refueling. The
licensee reported that the actuation occurred because the operational
setpoints, as opposed to the shutdown setpoints, had not yet been
programmed into Radiation Monitor RM-050. This monitor measured gaseous
radioactivity inside containment.
Procedure UI-RC-3, " Reactor Coolant
System Startup," required that the operational setpoints be programmed
into RM-050 prior to the RCS temperature reaching 395*F. The plant was at
approximately 360*F at the time of the actuation.
Operations personnel reported they were aware of an upward trend in
containment radiation levels and had dispatched an instrumentation and
control (I&C) technician to program the operational setpoints into RM-050.
However, before the I&C technician was able to install the setpoints, the
radiation monitor received a spiking signal that exceeded the setpoint and
the VIAS occurred.
All equipment functioned as designed. No releases occurred and no
abnormally high radiation conditions existed. The setpoints in use were
too conservative for the plant condition at the time of the VIAS.
LER 87-020 reported a similar VIAS actuation which occurred during startup
from the previous refueling outage. At that time, the licensee changed
Procedure 01-RC-3 to require that the operational setpoints be programmed
into RM-050 prior to reaching 395 F instead of after reaching 395 F.
This
procedural adjustment was apperently inadequate based on the radiological
status of the RCS.
To preclude recurrence of a VIAS, the licensee revised Procedure 01-RC-3.
The procedure now directs that the operational setpoints be programmed
_
_ _ _
..
- _ _ .
._ __-___-_ _____-__________-______________
_______- _
.
'
..
.
,
-
.
into RM-050 prior to exceeding 210 F.
The NRC inspector reviewed this
procedure change and finds that, if properly implemented, the change
should prevent recurrence of this event.
Based on the reviews performed by the NRC inspector, as described above,
it appears that the licensee took appropriate actions in response to the
identified events to provide timely corrective actions and implementation
of controls to p. event recurrence of the event.
No violations or deviations were identified.
5.
0,yerational Safety Verification (71707)
The NRC inspectors conducted reviews and observations of selected
activities to verify that facility operations were performed in
conformance with the requirements established under 10 CFR, the licensee's
administrative procedures, and the TS. The NRC inspectors made several
control room observations to verify the following:
Proper shift staffing was maintained and conduct of control room
personnel was appropriate.
Operator adherence to approved procedures and TS requirements was
evident.
Operability of reactor protective system, engineered safeguards
equipment, and the safety parameter display system was maintained. If
not, the appropriate TS limiting condition for operations (LCOs) were
net.
Logs, records, recorder traces, annunciators, panel indications, and
switch positions complied with the appropriate requirements.
Proper return to service of components was performed.
Maintenance orders (MO) were initiated for equipment in need of
'
maintenance.
Management personnel toured the control room on a regular basis.
Control room access was properly controlled.
Control room annunciator status was reviewed to verify operator
awareness of plant conditions.
Mechanical and electricel temporary modification logs were properly
'
maintained.
Engineered safeguards systems were properly aligned for the specific
plant condition.
_ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ - _ _ _ - _ _ _ -
-
-
_
l
(
l
'
..
.
l
9
'
'
During observation of the above items, the NRC inspector noted the
problems listed below:
a.
On March 27, 1989, the NRC inspector noted that a danger tag was
installed on Valve MS-100, the steam supply valve to the main steam
line radiation monitor (RM-064). The danger t g. stated that the ,
valve was required to be shut; however, the NRC inspector noted that
the valve appeared to be open. The NRC inspector immediately
notified the shift supervisor who verified that the valve was open.
The valve was immediately shut.
'TS 5.8.1 states, in part, that written procedures shall be
implemented that meet the minimum requirements of Appendix A to
Paragraph 1.c of Appendix A to Regulatory Guide 1.33 requires that
equipment control (e.g., tagging and locking) be addressed by written
procedures.
Paragraph 4.1.6 of Procedure 50-0-20, " Equipment Tagging Procedure,"
states, in part, that [ danger] tags shall be hung in accordance with
the tag-out sheet and the components shall be in their required
position.
Contrary to the above, the licensee failed to properly implement
Procedure 50-0-20 in that a danger tag was installed on Valve MS-100
and the valve was not in the required positior. The danger tag
stated the valve position was shut; however, the valve was found to
be open. This is the first example of the failure to follow
procedures.
(285/8913-01)
The licensee performed a review to determine why the valve was open
when the tag stated that the valve should be shut. The licensee
determined that a temporary clearance tag had beer installed, on
March 23, 1989, so'the valve could be' opened to allow testing of
RM-064. The temporary clearance tag system is used by the licensee
as a method of temporarily changing the position of valves and
components without removing the danger tag. When the temporary
clearance tag is removed, the component is returned to the position
_
specified on the danger. tag.
In this particular instance, an operator removed the temporary
'
clearance tag on March 24, 1989. When the tag was removed from
Valve MS-100, the operator failed to shut the valve as specified on
the danger tag.
No work was performed on RP-064 from March 24, 1989,
until the valve was found open by the NRC inspector; therefore, no
personnel safety hazard existed.
In re<ponse to this problem, the licensee performed a review of all
danger and caution tags installed in the plant. The review was
performed to verify that all components were in the pcsition
_
--
l
7
-
_
=-
--
-
_ -_ _ _
_ _ _
L
n
n
..
'
3_
,e
- ,n
'
10
'
i
specified by the instructions provided on the tag. No additional
,
problems were identified during the licensee's review.
The NRC. inspector also toured additional plant areas and verified
that other components were in the position specified by the danger or.
caution tag. No additional problems were_noted.
On March 30, 1989, the licensee issued a hot line (HL 89-015) to all
licensed operators to notify the operators of the problem identified
with equipment ~ tagging. The hot-line system is used by the licensee
as a means of notifying, operators of information in a timely manner.
The hot line stressed the absolute need for verification of the
position of components when danger tags are installed or when
temporary clearance tags are removed.
b.
On March 28, 1989, the NRC ir.spector reviewed.the temporary
modification (TM) log to verify that the log was being maintained in
accordance with the appropriate procedure. During review of the log,
the NRC inspector identified two problems with the log.
(1) The TM control form was not properly completed after
installation of TM 89-M-019, "Feedwater Regulation Valves
.
(FCV-1101 and FCV-1102) Pipe Supports." The modification was
'
-installed but the control form did not indicate completion of:
the installation. The pipe supports were installed on March 27,
1989.
+
'
4
,
.
,
.
l(2) TM 89-M-013, "CCW Duratek System," was verified to be installed
" '
'
(J
.
but the control form was not properly com)1eted because the
'
'
control form was not signed to indicate t1at verification of the
-
~
'
' installation had been completed.
o
Although the safety significance of the above two examples is minor,
vthe NRC inspector is' concerned about the apparent lack of attention
,
t
to details when completing the TM log. During previ]us inspections,
,
sNRC inspectors have noted other problems related to control of TMs.
>
- iThe previously identified problems are discussed in NRC Inspection
<
Reports 50-285/88-201 and 50-285/89-03.
It appears that the licensee
,
,
!
continues to encounter problems with the control of TMs.
Upon notification of the problems with the TM log, the licensee took
i
actions to correct the problems. The licensee stated that all
>
problems had been corrected prior to the end of this inspection
r
<
period.
6.
Plant Tours (71707)
The NRC inspectors conducted plant tours a+ various times to assess plant
and equipment conditions. The following i wms were observed during the
tours:
_ -_ - __ _ ___________ _ __ -
___
_-
. -
_ _ _ _ _ _ _ _
t
'
p
.
p
,:.
.,
,
g
y
.
.
General plant conditions, including operability of standby equipment,
were satisfactory.
Equipment was being maintained in proper condition, without fluid
leaks and excessive vibration.
Valves and/or switches for safety-related systems were in the proper
'
position.
Plant housekeeping and cleanliness practices were observed, including
no fire hazards and the control of combustible material.
Performance of work activities was in accordance with approved
'
procedures.
Portable gas cylinders were properly stored to prevent possible
missile hazards.
Tag out of equipment was performed properly.
Management personnel toured the operating spaces on a regular basis.
During tours of the plant, the NRC inspectors noted the items listed
below:
a.
During tours of the plant throughout this inspection period, the NRC
inspector identified four concerns related to the condition of the
plant. The concerns are discussed below:
(1) On March 13 and 22, 1989, the NRC inspector noted two guillotine
fire curtains that needed additional licensee attention. The
curtains separated different safety-related fire areas.
In one
case, the sheet metal track for the curtain was badly bent and,
in the other case, the sheet metal track contained numerous
strips of tape.
The licensee tested both curtains in the as-found condition and
verified that the curtains functioned properly.
After
completion of the testing, the sheet metal track was
straightened and the tape removed from the other track. The
dampers were retested and functioned satisfactorily.
(2) On March 13, 1989, the NRC inspector noted that the splash guard
for the air blower for the generator on Emergency Diesel
~
Generator (EDG) 1 had come loose on one side. The cover not
being secured did not appear to affect the operability of EDG 1.
The licensee issued an M0 to reinstall tne splash ccver. The
,
cover had not been replaced prior to the end of this inspection
j
period.
!
l
'
i
,
_____-______________________-__-_A
e.
4
.
,
.
- ,.
12
"
-
(3) On March 13, 1989, the NRC inspector noted that trash was left
in a safety-related cable tray in Room 81. The trash consisted'
of a portion of a plywood sheet, rags, and a bucket half full of
water. .It appe'Arcd that the trash had been left by a cleaning
crew. The concern of trash in safety-related cable trays will
be discussed in more detail in the Maintenance Team Inspection
Report 50-285/89-01.
,
'
Upon notification by the"NRC inspector, the licensee removed the
.
'.
trash.
l
(4) On March 22 and 24, 1989, the NRC inspector noted that portions
of plywood sheets had been left or,the cable trays in Room 19.
The licensee had painted . Room 19 and the plywood sheets were
,
'
being used by the painting crew to prevent walking in the cable
.-
trays.
Once the painting crew had completed their pairtinge the
'
plywood was not removed.
Upon notification by the NRC inspector, the licensee removed the
plywood.
b.
During tours of the plant on flarch 13~and 22, 1989, the NRC inspector
noted that five cable tray covers in the east.and west switchgear
rooms, Room 19, and the upper-level electrical penetration room had-
been removed and not reinstalled.
It appeared that the tray covers
were removed during outage activities and were not replaced when the
activity was completed.
Criterion V of Appendix B to 10 CFR Part 50 and the licensee's
NRC-approved quality assurance program state, in part, that
activities affecting quality shall be >rescribed by documented
instruction of a type appropriate to t1e circumstances and shall be
!
accomplished in accordance with these instructions.
Drawing 11405-E-60, " Reactor Auxiliary Cuilding Tray Conduit Layout
Plan," requires in Note 17 that solid ccvers be installed on cable
trays.
Contrary to the above, cable tray installations have not been main-
tained in accordance with design documents in that cable tray covers
were not properly installed on trays in Room 19, the upper-level
electrical penetration room, and in the east-and west switchgear
rooms. This is the second example of the failure to follow
procedures.
(285/8913-01)
The NRC inspector previously identified this same violation during an
inspection parformed in llay 1986.
It appears that the licensee has
not taken appropriate corrective actions to address this problem.
In response to this problem, the licensee reinstalled the tray covers
identified by the NRC inspector.
In addition, the licensee initiated
4
_.__-m_
_.m
..___ __ _ _ _ _ ________j
_ _________
~
-
s
a
14 N
,y
-
,*
13
-
,
,_
p
l
",
a walkdown of cable trays to verify installation of all covers. At
the end of this inspection period, the licensee had not completed the
walkdown.
c.
On liarch -15,19C9, the NRC inspector noted that the air-operated
valve operator for Valve HCV-1388B did not aave a seismic support
y' ,
installed on it. Tha valve is instelled in a 2-inch line and the
valve operator is approximately 3-feet high. The installation
appeared to be nonseismically installed.
.
The NRC inspector also noted that the piping for installation of
l
Valve HCV-1387B was almost identical to the piping configuration for
Valve HCV-1388B.
The operator, identical to the operator on
HCV-1388B, had a seismic support installed. The NRC inspector could
,
not determine why one valve would have a seismic support installed
and the other had no support when the piping configuration for both
valves was essentially identical.
'
In response to this concern, the licenste performed a seismic
analysis for both valves. The analysis indicated that both were
seismically installed even.though one did not have a support
installed.
The licensee provided the analysis at the end of.the
inspection period; consequently, the analysis was not reviewed during
this inspection period. This item remains open pending review of the
analysis performed by the licensee.
(285/8913-02)
d.
On March 20, 1989, the NRC inspector noted that screens on the motors
for the CCW pumps were not in place. On CCW Pump AC-3A, one screen
was missing, and on CCW Pump AC-3C, no screens were installed. The
motor for CCW Pump AC-3C was replaced during the 1988 outage. During
an inspection performed from April 29 through May 3, 1985, an NRC
inspection team reviewed the licensee's environmental equipment
qualification (EEQ) program. During the inspection, the team noted
that the motor manufacturer recommended that the motor screens be
maintained in place to sustain the EEQ status of the motor;. The
licensee responded to the violation issued by the NRC team based on
the licensee not cleaning the screens.
In the response, the licensee
did not specifically address whether or not the motor screens had to
be installed to maintain the EEQ status.
Upon notification to the licensee of this concern, the licensee
i
contacted the motor manufacturer. The manufacturer stated that the
screens were not required to maintain the EEQ status. The
manufacturer did state that the screens should be maintained in place
to ensure that large pieces cf trash do not enter the motor.
During discussions with the licensee, management stated that a
response would be provided to this issue. The response will
establish the status of the screens with respect to the EEQ program
and actions that will be taken.
Based on this licensee commitment,
,
this item remains unresolved pending receipt of the licensee's
j
response and a review of the response by the NRC.
(285/8913-03)
,
'
!
>
!
.-
-
__ __--__
_ - - _ _ -
-
t.-
[. (
- 8
- . ,
. ..
g.
V
,
~
'
l
The' licensee' installed the missing screens on the CCW pumps.
Prior?
!
'
.
b
to installing the screens, the interior was inspected to verify no
f
. foreign material had entered the motor. No material was found..
e.
On March 21, 1989, the.NRC inspector toured the auxiliary building
and noted that the fire doors inttalled.on each room appeared to be '
!
, nonfunctional. The licensee recently, replaced'the fire doors in the.
'
auxiliary building with doors that were manufactured with a hole in
tha door frame. The hole was designed to accept a conduit: so that
'
/.
>
wiring.could be ccnnected to the door latch mechanism.
1
The fire' doors appeared to be nonfunctional due to the. existence of'
the open hole in the door frame. The NRC inspector checked to verify
that an hourly. fire watch patrol had been established for.the doors
<
and noted that no hourly patrol had been established. The:NRC
'
inspector notified the licensee of the apparent failure to comply
l
with TS 2.19 that- requires an hourly patrol oe established for all
!
safety-related, nonfunctional fire barriers. .On March 22, 1989, the
'
licensee established an hourly fire watch because the licensee did
not know whether or not the fire barriers were functional.
'
~
On March 31, 1989, the licensee provided an evaluation of the'
as-installed conditions of the fire doors.
This evaluation concluded
1
that the doors may not meet the-3-hour-fire. barrier rating; however,
1
due to the sma11' amounts of combustible material located in the fire
areas separated,by. the doors, no problems: existed.
j
The NRC inspector forwarded the evaluation perfonned by the licensee
l
to a fire protection specialist in the NRC's. Region IV office for
i
review. This item is considered unresolved pending the completion of
.
a review of the licensee's evaluation. -(285/8913-04)
!
The items discussed above' were identified by the NRC inspector and not by
licensee personnel. Based on the type and number of items, it appeared
i
that licensee personnel are not touring the plant as frequently as would
,
be expected and are not identifying conditions that potentially affect the
!
safety or safe operation of the plant. The NRC inspector noted during the
previous inspection period that it appeared that the licensee was not
performing adequate tours of the plant.
In the previous inspection
i
period, the NRC inspector identified nine items during plant tours. The
observations from the previous inspection period are documented in NRC
Inspection Report 50-289/89-09.
7.
Safety-Related System Walkdown (71710)
The NRC inspectors walked down accessible portions of- the raw water system
to verify system operability. Operability was determined by verification
of selected valve and switch positions. The system was walked down using
Drawing 11405-M-100, Revision 33, and Procedure 01-RW-1, " Raw Water System
.
!
- Normal Operation," Revision 23.
During the walkdown, the NRC inspectors
\\
_ - _ _ _ .
__
-
- _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ -
_ _ _ _ _ _ _ _ _
- _ - _ _ _
_ - -
.
.
.
..
.,
'
'
15
identified numerous problems related to maintaining the raw water system
and support systems in a configuration specified by the appropriate design
documentation.
During the walkdown, the NRC inspectors noted the following items:
a.
Conduit 6598-6597 for the circulating water, screen-wash, pressure
switches was supported by a light fixture using bare wire.
b.
A pipe support, adjacent to Valve SW-229, for raw water Pump AC-10A
seal-water supply line was missing.
c.
A tubing support for Pressure Controllers PC-2864 and PC-2865 was not
installed.
d.
The conduit for the electrical supply to the motor for raw water
Pump AC-10B was not connected to its support.
e.
The conduit for Flow Indicator FIA-1982B, installed to indicate
seal-water flow to raw water Pump AC-108, was not connected to its
'
support.
f.
The raw water pump bay sump pump discharge line was supported by a
conduit using bare wire.
In addition to the hardware discrepancies listed above, the following
drawing errors were noted:
a.
Five valves (IV-2854, IV-2855, IV-2856, IV-2857, and RW-114) were
shown on Drawing 11405-M-100, Revision 33, but were not lebeled on
the drawing. The five valves were installed in the plant.
b.
The isolation valves for Pressure Indicators PI-2854, PI-2855,
PI-2856, and PI-2857 were shown on Drawing 11405-M-100 and were
instelled in the plant, but were not labeled on the drawing or
included in the valve lineup list provided in Procedure 01-RW-1.
The NRC inspectors notified the licensee of the items identified during
the system walkdown. The licensee issued M0s to correct the discrepancies
and was in the process of correcting the discrepancies related to system
hardware installations at the end of this inspection period. The licensee
stated that the discrepancies noted rn Drawing M-100 and in
Procedure OI-RW-1 would be corrected in the near future.
Based on evaluations performed by the NRC inspectors, it appeared that
none of the hardware deficiencies affected the safety or safe operation of
the raw water system. The NRC inspectors also felt that the items noted
in the raw water system were not indications of a broader, generic
problem. However, the NRC inspectors were concerned with the relatively
large number of material discrepancies identified during the system
walkdown,
i
_- _
_
,
.
-
.,
16
'
-
In addition to the items listed above, the NRC inspector noted that
Radiation Monitor RM-056B was out of service and had not been in service
since April 9, 1988. RM-056B is installed in one of the two raw water
overboard discharge headers and monitors effluent discharge from a number
of plant components. RM-056B acts as a backup for the steam generator
blowdown radiation monitors and also monitors the effluent from the
shutdown cooling heat exchangers in the event that componer.t cooling water
is lost and the plant must be cooled down using raw water as the cooling
medium.
In the event that raw water had to be used as the cooling medium,
no neans would be available to monitor the effluent raw water from the
shutdown cooling heat exchanger for radioactivity.
If a leak occurred in
the heat exchanger during this configuration, potentially contaminated
effluent would be discharged to the overboard raw water discharge header
and into the Missouri River without the knowledge of the licensee.
Procedure A0P-ll, " Loss of Component Cooling Water," was issued by the
licensee ~to provide instructions to personnel on what actions to take in
the. event that CCW flow is lost. The procedure discussed the actions to
take to ensure that raw water flow is initiated to the components normally
supplied by CCW.
However, Procedure A0P-11 did not address the actions to
be taken to ensure that the raw water affluent is monitored for potential
radioactive releases.
>
TS 5.8.1 states, in part, that written procedures shall be established
"
that meet or exceed the minimum requirements of Appendix A to Regulatory
Guide.l.33. Section 6.y of Appendix A to Regulatory Guide 1.33 states-
that procedures shall be written to address the abnormal releases of
.
radioactivity during emergencies or other significant events.
Contrary to the above, the licensee failed to adequately establish a
procedure to address the abnormal release of radioactivity during a
significant event in that Procedure A0P-ll did not state that RM-056B was
required to be operable to monitor the potential for the release of
radioactive material in the event that raw water was supplied as a cooling
medium to components normally cooled by CCW. This is an apparent
violation.
(285/8913-05)
Upon notification to the licensee of the concerns identified by the NRC
inspector, the licensee attempted to restore the operability of RM-056B.
The monitor would not operate properly and, prior to the end of this
inspection period, the licensee did not determine the reason for
inoperability.
In addition, the licensee revised Procedure A0P-ll to
specify that any time raw water is supplied to any component normally
supplied by CCW, RM-0568 shall be placed in service. The change to
Procedure A0P-11 also specified that if RM-056B was inoperable, grab
samples of the raw water effluent would be taken as directed by the plant
i
chemist.
It appeared that the change made to Procedure A0P-ll was
i
incomplete because the licensee did not include what actions would be
taken if RM-056A, the radiation monitor for the other raw water overboard
discharge header, became inoperable. The licensee did not revise
Procedure A0P-11 to include RM-056A prior to the end of this inspection
1
1
_ ----------
.
.
-
.,
l
'
'
17
period. This discrepancy was discussed with the licensee, who agreed to
reference RM-056A in Procedure A0P-11.
This item will remain open pending
inspector review of the revised Procedure A0P-11.
(285/8913-06)
8.
Monthly Maintenance Observations (62703)
During this inspection period, a comprehensive and indepth review of
maintenance activities was performed by the maintenance inspection team.
The team was comprised of members from NRC's Region IV Office, NRC
Headquarters, and consultants. The results of the team inspection are
documented in NRC Inspection Report 50-285/89-01.
Due to the performance of the maintenance team inspection, the NRC
inspectors did not perform maintenance observations during this inspection
period.
No violations or deviations were identified.
9.
Monthly Surveillance Observations (61726)
The NRC inspector observed selected portions of the performance of
TS-required surveillance testing on safety-related systems and components.
The NRC inspector verified the following items during the testing:
Testing was performed by qualified personnel using approved
procedures.
Test instrumentation was calibrated.
The TS limiting conditions for operation were met.
Removal and restoration of the affected system and/or component were
accomplished.
Test results conformed with TS and procedure requirements.
Test results were reviewed by personnel other than the individual
directing the test.
Deficiencies identified during the testing were properly reviewed and
resolved by appropriate management personnel.
Test was performed on schedule and complied with the TS-required
frequency.
The
inspector sbserved the following surveillance test activities.
The pivcedures us
for the test activities are noted in parenthesis:
Monthly check of the auxiliary feedwater flow transmitter channels
(ST-FW-2)
Monthly fire protection test and inspection (ST-FP-1)
. . _ .
_ _ _ _ _ _ _ _ _ _ -
_ _
_ _
.
'
d
- .
5',7
'
t
,
7.
'
18
-
-
'
1
m
,
Monthly check of the containment wide-range' pressure indication
.
'
channels (ST-CWP-1)
'
'
l
Monthly check of containment water level narrow-rance channels
',
(ST-CWL-1)
j
,
i
Monthly check of containment water level wide-range channels
$(ST-CWL-2)
Quarterly test of the main steam to auxiliary feedwater pump valves
'
,
(ST-ISI-MS-1)
A discussion of each surveillance observed is provided below:
'
a.
On March 8, 1989, the NRC inspector witnessed a licensed operator
perform Procedure ST-FW-2, " Auxiliary Feedwater Flow Transmitter
Check." Statements in the procedure required the operator to verify
that_ flow indicating gages were within a specified tolerance. The
operator found that he could not discern the meter reading within the
tolerances specified by the procedure.
He recorded what he n uld
determine with certainty and annotated in the remarks .section of the
procedure that the gages needed calibration. He stated that the
presently installed instruments made reading the gages within the
tolerances specified in the procedure very difficult.
The completed procedure was then forwarded to the operations
supervisor for review, as per standard policy.
Following the review,
M0s 891701 and 891702 were issued to calibrate the flow indicators.
The indicators were recalibrates and the surveillance was
successfully performed on March 16, 1989. The NRC inspector reviewed
the completed retest. Additionally, site licensing has assigned an
action item on the Site Licensing Action Log, to have engineering
reevaluate the acceptance criteria of ST-FW-2.
No additional problems were noted by the NRC inspector.
b.
On March 8, 1989, the NRC inspector witnessed a portion of the
performance of OI-FF-1, " Monthly Fire Protection Test and
Inspection." Specifically witnessed was the running of the
diesel-driven fire pump and the recording of its operating
parameters. The test was conducted by a water plant operator in
accordance with the procedure. No abnormalities were noted within
the portion of the test witnessed by the NRC inspector.
c.
On March 13, 1989, the NRC inspector witnessed the performance of
Surveillance Tests ST-CWP-1, ST-CWL-1, and ST-CWL-2. These tests
verified the operability of the containment wide-range pressure,
l
narrow-range water level, and wide-range water level instrumentation.
The tests were performed in accordance with the procedures and the
,
equipment operated as specified. No problems were noted by the NRC
inspector.
.
!
4
g
-_
_ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
_ _ - _ _ _ _ - -
___
h,
, ..
"
om
,,
,
. . .
J'
19
d.
.0n'Harch 7,~1989, the NRC inspector witnessed the performance of-
ST-ISI-MS-1. While witnessing this test, the main steam supply valve
to the steam-driven auxiliary feed pump (Valve YCV-1045A)
inadvertently cycled open on a loss of air pressure while operators
followed the procedure, as written. The NRC inspector had observed
performance of this test'on numerous other occasions with no unusual
,
,
occurrences.
The NRC inspector found that a procedure change had been made to
install a test gage on the instrument air.(IA) accumulator in order.
to monitor its pressure decay. The' procedure change failed'to
consider that loss of air would occur when the technicians were
instructed'to install the test gage.
3
The opening of; Valve YCV 1045A caused a momentary activation of the
steam-driven auxiliary feed purp. The turbine pump did not come up
to speed and pressuret sThe licensee'did not report the activation.
'
The NRC' inspector found the procedure change review to be inadequate
L
in this instance. Although the licensee-instituted
,
Procedure S0-G-30, "Setpoint/ Procedure Changes," which contains
provisions for ensuring system operationiis not affected by the
change, the engineers making the change apparently overlooked the
connection between installation of'tlye' test gage and loss of air
pressure.'
-
After the inadvertent cycling'of th'e valve occurred, the operator
leading the. surveillance appropriately reestablished initial
conditions and stopped'the test. The operator contacted the shift
technical advisor.
The procedure was revised to require the
performance of appropriate valve lineups prior to installing the test
gage, reissued, and performed without incident on March 14, 1989.
Upon further investigation, the NRC inspector found that the same
test, including the erroneous change, had been performed on
January 28, 1989. The NRC inspector reviewed the completed test
document and found that the. technician conducting the test noted no
abnormalities.
If the procedure was followed as written,
Valve YCV-1045A would have had to cycle open inadvertently on a loss
of air pressure. The lack of noting this anomaly is indicative of a
lack of understanding of the procedure by the technician or
p
inattention to detail.
Although the NRC inspector noted a weakness in design and test
control relative to this occurrence, no specific violations or
deviations of NRC requirements were found.
,
During review of surveillance activities, the NRC inspectors noted that it
appeared that all personnel involved with the performance of testing
performed ~the testing in accordance with the appropriate requirements.
However, it appears that a higher degree of review of procedure
preparation is warranted.
._
. _ _
_
_ _ _ - - - _ _ _ _ _ _
_ _. ._
_ _ _
_ _ _ _ _ _ _ _ _
'
,
o
.,
20
No violations or deviations were identified.
10. Security Observations (71707)
The NRC inspectors verified that the physical security plan was being
implemented by selected observation of the following items:
The security organization was properly manned.
,
Personnelwithinthe.protectedarea(PA)displayedtheir
identification badges.
Vehicles were properly authorized, searched, and escorted or
controlled within the PA.
Persons and packages were properly cleared and checked before entry
into the PA was permitted.
The effectiveness of the security program was maintained when
security equipment failure or impairment required compensatory
measures to be employed.
The PA barrier was maintained and the isolation zone kept free of
transient material.
The vital area barriers were maintained and not compromised by
breaches or weaknesses.
Illumination in the PA was adequate to observe the appropriate areas
at night.
Security monitors at the secondary and central alarm stations were
functioning properly for assessment of possible intrusions.
During this inspection period, the NRC inspector identified the following
items:
a.
On March 8, 1989, the licensee notified the NRC inspector that blue
badges had been issued to two individuals who had not satisfactorily
completed radiation worker training.
In addition, the two
individuals were also issued thermoluminescent dosimeters (TLD) and
l
allowed to enter the RCA. This problem was identified during the
!
review of TLD records by the licensee.
The licensee uses a colored badge system and the color of the badge
designates which areas in the plant the badged individual is allowed
to enter. A blue badge is issued to an individual that is allowed to
enter vital areas and the radiologically controlled area (RCA). To
obtain a blue badge, the individual must successfully pass the
general employee and radiation worker training programs.
- _ _
_
- _ _ - _ - -
.
,
1
>
<
,
- .
.
.
.
,
'
'
21
In reviewing this problem, the licensee identified the root cause as
- the method of verifying the completion of training utilized by the
general employee training organization. The computer printout
'
contained the name of the individual taking the radiation worker
training and the date the training was taken. The computer printout
i
did not indicate whether or not the individual passed the training
1
class. The. printout was provided'to security personnel who
incorrectly assumed that if the individual's name was on the
printout, then the individual had completed all the necessary
training to receive a blue badge. Therefore, blue badges were
incorrectly issued to the two individuals.
- The licensee processed the TLD for each individual. One individual
received 0 millirem and the other individual received an exposure of
16 millirem. Neither of the individuals currently work at the FCS.
On March 14, 1989, the NRC inspector noted that security personnel-
had issued the wrong type of badge to an. individual.
In this case,
the security organization prepared and issued a blue badge to an
individual that had not passed radiation worker training.
The NRC inspector notified the licensee of the incident and a review
of the incident was performed. .The licensee's review indicated that
a clerical error in the security organization _ resulted in issuance of
the wrong badge color.
Tho licensee has taken action to resolve the weakness in their badge
issuance program.- The computer data maintained by the training
department was reformatted such that only the names of those
individuals that pass all training classes will appear on the
computer printout. This action will allow security to readily verify
the qualifications of. the individual prior to issuance of a security
badge.
The licensee is currently in the process of reviewing additional
training and security records to verify other individuals were not
incorrectly issued the improper security badge. This effort was
still in process at the end of this inspection period. No other
problems had been identified.
In addition, the licensee issued a niemorandum (FC-SC-207-89) on
March 15, 1989, to security personnel to establish a requirement for
independent verification that the proper badge is being issued. The
independent verification requires a review of all documentation to
verify that the individual completed all requirements necessary to
'
receive a badge of a particular color. However, the licensee has not
revised Procedure S0-G-66, " Control of Key Card Badges," to
institutionalize the independent verification of the badge issuance
program. This procedure revision is considered an open item.
i
(285/8913-07)
l
4
i
_-.
_ - - _
_ _ _ _ _ - _ _
_ _ _ _ - - _
_ _ _ - _ _ - _ .
]
m%, ; g'
T
-
~ - - -
- - - - - - - - - ~
,
.
,
i
"
'
'-
.,
,
..:
' jg .
,
_-
I
22
,
,
a:
'
b.
On' March 23,L1989, the' NRC inspector received safeguards infomation
transmitted through the U.S. Postal Service from the licensee. .The.
information was properly transmitted in that the information was:
placed inside two sealed envelopes'. However, when the outer envelope-
was opened, the NRC inspector noted that the inner envelope ldid not
contain the full address of the NRC inspector. This is the third
time:that this. problem has occurred. On the'two previous occasions,
e
the'NRC _ inspector notified the licensee and the licensee stated that-
actions would be taken.
,
The licensee's procedure requires that the address of the recipient
'of safeguards infomation be placed on the inner envelope.
In this-
-
case, this was not done.
.The-NRC-inspector-notified the licensee of this. apparent discrepancy.
The: licensee stated that the identified concern would be
,
appropriately resolved. .The NRC inspector will continue to monitor
the transmission of _ safeguards information to verify adequate
resolution of the concern.
c.
During an NRC inspection performed February 27 through March 3,1989,
security specialists from the NRC's Region IV office identified
problems with personnel, not wearing- their security badges. The
details of the inspection are provided in NRC Inspection
Report 50-285/89-10.
During this inspection period, the NRC inspector noted, during tours
of the' plant, that all personnel .were wearing their badges. However,
the'NRC inspector noted on six occasions that. personnel were not
properly displaying their badges in that the photo side of the badge'
was not facing outward.
The NRC inspector notified licensee management of the observation.
Licensee management stated that actions would be taken to ensure that
personnel' properly display their security badges.
d.
During an inspection performed by security specialists from the NRC's
[
!
Region IV Office from February 27 through March 3,1989, numerous
prob'lems were identified with the implementation of the licensee's
security program. The problem areas included items such as access
control of personnel, properly displaying security badges, escorting
of visitors, and. control of safeguards information. Details of the
areas of weakness identified by the NRC inspectors are provided in
NRC Inspection Report 50-285/89-10.
In response to the identified problems, the licensee established a
program where all personnel with unescorted access to the plant would
be required to attend a special training session. Attendance at the
training session was required if an individual was to maintain plant
access.
p
e :-
- _
_
_
_
_ _
_ _ _
_ _ _ _ _ _ _
_
w
..
.
--
,
"
,,
.
j
p
{
'
,c ,
'
- )
u
.:
1
r
,.
,<
'
23'
l
a
,.
-
,
'
.
i
'
- -
On March 9 -1989, the NRC inspector' attended the training. The NRC
1
'
~ inspector noted that the material was well~ presented, covered all the
)
appropriate subject areas, and the.information provided was accurate.
1
y
During ob.servations of the performance of the security force, the NRC
-inspectors noted that it appeared that security personnel performed their
duties'in.^a-professional manner,
j
,
,
}
,
No-violations or' deviations were identified.
1
I
11.
Radiological Protection Observations
(71707)
'
>
'
The NRC inspectors verified that selected activities of the licensee's
.
radiological protection program were implemented in conformance with the
, facility policies and procedures and in. compliance with regulatory
requirements. The activities listed below were observed and/or reviewed:
Health physics (HP) supervisory personnel conducted plant tours to
check on activities in progress.
'HP technicians were using calibrated instrumentation.
- " C Radiatio'n work 3ermits contained the appropriate information to
ensure-that worc was performed in a safe and controlled manner.
c
Personnel,in RCAs were wearing the required personnel monitoring
'~
equipment.and protective clothing and were properly frisked prior to
1'
'
exiting an RCA..
Radiation and/or contaminated' areas were properly posted and
controlled based on the activity levels within the area.
No violations or deviations were identified.
12.
In-Office Review of Periodic, Special, and Nonroutine Event Reports
(90712 and 90713)
e
,
'
In-office review of periodici special, and nonroutine event reports was
performed by the NRC inspectors to verify the following, as appropriate:
"'
Correspondence included the information required by appropriate NRC
requirements.
Test results and supporting information were consistent with design
predictions and specifications.
Planned corrective actions were adequate for resolution of identified
problems.
..
_.
_ _.
_ _ _
_ _ _ - .
.
.
. -
.,
24
Whether or not any information contained in the correspondence report
should be classified as an abnormal occurrence or additional reactive
. inspection is warranted.
Correspondence did not contain incorrect, inadequate, or incomplete
information.
The NRC inspectors reviewed the following correspondence:
Schedule For Completing Remaining Actions in NRC Bulletin 88-11,
dated March 6, 1989
Design Engineer Training, dated March 2, 1989
February Monthly Operating Report, dated March 15, 1989
11onthly Operations Report for February 1989, undated
Additional Information on Raw Water Pump Facility License Change,
dated March 15, 1989
Failure to Perform ST-FD-2 Within Allowable Interval to Meet
i
Technical Specifications (LER 89-006), dated March 13, 1989
Monthly Status Report of the Safety Enhancement Program, dated'
March 21, 1989
Additional Information Concerning the Loss of the 161-kV Power
'
e
Supply, dated March 23, 1989
,
Inservice Testing Program for Pumps and Valves, Revision 3,
Fort Calhoun Station, dated March 24, 1989
,
No viciations or deviations were identified.
13.
Followup on Onsite Events
(93702)
,
During this inspection period, the licensee experienced a number of onsite
events. Each event is discussed below:
a.
On March 4, 1989, the licensee lost one of the two available offsite
p
power sources. The 161-kV supply was lost due to an ice storm in the
vicinity of the plant which caused four support structures to
l.
l
.
collapse. The loss of the offsite supply caused the electrical loads
'
normally supplied by the 161-kV source to automatically shift to the
remaining 345-kV offsite power supply. The loads shifted power
sources without any problems.
,
l
l
The loss of the 161-kV offsite source placed the plant in an abnormal
I
situation.
If the second offsite source, the 345-kV supply, was
lost, the plant would be placed in the natural-circulation mode of
l
. _ - . _ _ . - - - - -
F
-
-
m
-
,
'
I.
'
'
'
'
,:=
-
m
,
,
i 25
}.
-
>
-
,
i
lU
>
,
,
k
a
operation. The loss of th'e 345-kV' supply would cause power to be
lost to the reactor coolant" pumps since the pumps are normally
supplied by'offsite
3ower.
Duegto the configuration of the plant
electricalisystem, tle station main generator can not be removed from
'
the power grid since opening the generator output breaker causes a
loss of offsite power to the plant loa'ds.
Due to the configuration
- of the electrical system, any occurrence that could cause a turbine
trip would place the' plant in a natural-circulation mode.
.To' minimize the possibility of an occurrence that could affect the
. station electrical system, the licensee instituted the following
' items:
(1) Specified that'no maintenance or testing would be performed on
any plant electrical system or component.
(2) No_ maintenance or testing would be performed on the EDGs. This
. item was specified to ensure that the EDGs would be available
for' automatic loading in the event the second offsite power
source was lost.
(3) _ Personnel were~ alerted 'o not perform any unnecessary activities
in:or around electrical systems or components that had the
potential of affecting the stability.of the electrical system.
(4) The plant electrical loads were shifted to equalize the amount
of-loads on the two 345-kV/4.16-kV transformers.
The licensee also took actions to ensure that the necessary
equipment, if required for natural circulation when the second
offsite power supply was' lost, would be available. The~ actions
included delaying the surveillance testing of equipment and
components to ensure continued operability or to prevent an
inadvertent plant trip during testing activities and verified that
all necessary equipment was available for. operation.
The control room operators also readied themselves for the potential
of entering the natural-circulation mode of operation. The operators
reviewed and performed walkthroughs of the appropriate emergency
procedures to ensure they fully understood all tie necessary actions
to be taken. The actions also included a review by the turbine
building' operator of the method to be used to manually disconnect the
main generator from the grid so that the 345-kV supply could be
'
restored to the station distribution system.
The licensee's electrical operations crew replaced the 161-kV support
7tructures and restored powerfto the plant on March 10, 1989.
Since
power has been restored, no further problems have occurred with the
161-kV grid.
,
I
'
.
>
t'
"
-.
- --.
_ _ - - _
__
_ _ _
.______________________________________________y
'.
.-
- _ - _
_
.
t
- -
.,
26
-
-
The NRC inspector reviewed the actions taken by the licensee.
It
appeared that the licensee took proactive actions to address the
, unusual status of the plant.
'
In discussions with the licensee, management stated that a
,
reliability analysis would be performed to determine what actions, if
any, could be taken to increase the reliability of the offsite power.
'
sources.
In addition, the licensee stated that TS 2.7 would be
<
,'
reviewed te determine if changes should be made to clarify the
i
actions to be taken when the 161-kV source was lost. The licensee
'
stated that a letter would be sent to the NRC by April 21, 1989, to
"
specify what actions would be taken with respect to increasing the
'
reliability of the 161-kV ofisite source.
This item remains open
g
pending receipt of the letter and a review of the submittal by the
'
'NRC. -(285/8913-08)
-
'
,.
>
, i.
b.
On Itarch 14, 1989, the licensee experienced voltage fluctuations on'
'
the 345-kV offsite power grid due to high winds in the area of the
plant. The high winds caused the high-voltage lines to sway and
'
create an impedance interaction causing the supply voltage to surge
up 6 kV and dip downward 4 kV. As a result of the numerous
'
fluctuations, the following occurrences were noted with respect to
plant equipment:
-
"
(1) Breaker 345-4 (one of the two station generator output breakers
feeding the 345-kV.line) tripped open.
The licensee
successfully reclosed the breaker right after it tripped open.
(2) Numerous low-voltage alarms were received on various 4160-V
busses. .The alarms immediately cleared.
(3) The security system emergency diesel generator automatically
started due to voltage fluctuations. The power to the security
equipment was not lost. The diesel was secured.
(4) On two occasions, the technical support center (TSC) emergency
diesel automatically started.
The power to the TSC was not
lost.
The diesel was secured after each start.
The voltage fluctuations did not affect any plant safety-related
equipment. All plant equipment functioned normally. The plant EDGs
did not start nor did any plant equipment trip off the line.
Initially, the voltage fluctuations were occurring every 4 to
5 minutes. The electrical system dispatcher was able to isolate the
electrical surges from the grid being supplied by the plant so, after
approximately 30 minutes, the voltage fluctuations were occurring
every 30 to 45 minutes.
At the time the fluctuations started, the licensee had one
low-pressure safety injection and one auxiliary feedwater pump out of
service for preventive maintenance activities. The shift supervisor
1
- _ _ - - -
_
__----__-_-____a
-
.
-
q.
,
J'
..
.
e
( -
j
,
L
,,
'
'
,
,
27
-
-
immediately returned the pumps to an operable status in the event the
offsite power sources were lost.
In addition, the operations crew
'
verified that all equipment necessary for natural circulation was
available.
Management and crafts personnel remained onsite in the event that
additional assistance for the operations crew was required. All
personnel left after the fluctuations stopped.
The NRC inspector was in the control room and observed the actions of
the operations crew during this event. The NRC inspector noted that
the crew quickly reacted to all annunciator alarms and diagnosed the
affects the alarms may have on plant equipment. Based on the
,
observations made by the NRC inspector in the control room, it
l
appeared that the actions taken by the operations crew were timely
and performed in a professional manner and that decisions were made
that focused on the safety of the plant.
potential for a high-energy line break (HELB) problem related to the
On March 23, 1989, the licensee identified a
c.
in the main feedwater
system. The problem was identified during reconstitution of the
design basis.
In 1983, the licensee installed a modification in the main feedwater
system to change the size of the feedwater regulating valve (FRV)
from a 12-inch to an 8-inch valve. The modification was installed to-
. provide better control of feedwater flow to the steam generators. At
the time the modification was installed, an HELB analysis was not
performed because none was required for modifications installed in
nonsafety-related' systems.
The modification installed the 8-inch FRV
in a 16-inch pipe and required the installation of a 16-inch to
.
8-inch pipe reducer. The analysis performed during reconstitution of
I
the design basis indicated that an HELB point should have been
assumed at the pipe reducer. The data indicated that the stresses in
the piping at the point where the FRVs were installed exceeded the
piping Code USAS B31.1 allowables for stresses developed for the
L
total of pressure, dead weight, and thermal loads.
i
The licensee reviewed the details of the actual piping installation
and determined that no safety problem existed. The basis for this
!.
determination is provided below:
(1) A survey of the piping was performed using a transit to verify
)'
that the piping was not sagging due to dead weight loads. The
licensee did not detect any sagging in the piping.
(2)~ The FRV is located upstream of the main feedwater isolation
valves and, should a break occur, the FRVs could be isolated
from the steam generators.
.
l
l
u__--__---
W
/
-
p
i
_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _
-
..
.
.
'
'-
c,
i
29
\\
'
position is changed to increase the RM-061 setpoints. When the
temperature inversion had cleared, the switch is returned to its-
original position.
Based on the potential failure of the box during a seismic event,
continued operability of RM-061 could not be_ ensured. The signal
generated by RM-061 is used as one input.into the containment
radiation high signal (CRHS) circuitry. Four other radiation
monitors also input into the CRHS circuitry; however, failure of
RM-061 could potentially cause the CRHS circuitry to fail to function
even though valid signals were being generated by the other radiation
monitors. Based on the potential that the CRHS circuitry would not
function during a seismic event, the licensee declared both r.hannels
of CRHS inoperable. The CRHS circuitry provided actuation ir. puts
into the VIAS circuitry. A VIAS is generated to ensure isolation of
containment when a high radiation condition is sensed in the
containment.
When the licensee declared both CRHS channels inoperable, the
licenseeenteredtheLC0ofTS2.15(3). The LCO states that plant
operation may continue if the containment ventilation isolation
valves are closed. Shortly after entry into the LCO, the licensee
shut all containment ventilation isolation valves.
On March 30, 1989, the licensee removed the switch box from inside
the control room cabinet and restored operability of RM-061. Based
ontheseactions,thelicenseeexitedtheLC0ofTS2.15(3).
Additional followup on this item will be perfomed when review of the
LER issued by the licensee to address this problem is performed in
the near future.
No violations or deviations were identified.
14. Unresolved Items
An unresolved item is a matter about which more information is required in
order to determine whether it is acceptable, a violation, or a deviation.
Two unresolved items discussed in this inspection report are listed below:
Item
Paragraph
Subject
285/8913-03
6.d
CCW Pump flotor
Screens
285/8913-04
6.e
Operability of
Fire Doors
L--_______-__-_______-___-_-________-___
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
____ __
1
'
.
.,
.
.
.
,
,
30
-
15. Exit Interview
The NRC inspectors met with Mr. K. Morris (Division Manager, Nuclear
Operations) and other members of the licensee staff on April 7, 1989. The
meeting attendees are listed in paragraph 1 of this inspection report. -At
. this meeting, the NRC inspector sunnarized the scope of the inspection and
the findings.
.
,
b
d
f
1 *
m.
s3
I
r,
&
'
'
.,
, .
'
'
>
,
4
t
$W
/
l
i
l
lL
- - - - -
_ __ - ___- - _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,
--
_ _
-
- _ _ _ - _ _ _
--_--_- ---- _ - ----_
. _ - - _ - - - - _ _ . _ _
Nnc* m s66
U.S. NUCLEAR REGULATORY COMMISSION
PRINC. PAL lNSPECTOR(Nsme lasf hrst anarnee naar##87
, k
INSPECTOR'S REPORT
@
RE VIE WE 34
I
Of fice of Inspection and Enforcement
N9TmM M 'TM
wiCv0sy
gg}g
~ . . . . - .
_.~
_,
/
,
. ~ ~
. . - .
T
$c
'
UCE N561 VENM.R
DOCK ET NO 88 o geslOR LICENSE
yp
NO IBY PRODUCTill) digits)
NO -
MO.
VR
'
.'"5'"
O 5 0 0 o 2. 6 6
0 9 / S
h(,
% k. ,
__
^
__
"Di W c-r
_ .,,_,
_
" - " '"
D - DELETg
C
- * ~ " * "
- ~ - . . . ' .
~""
R
RE PL ACE
O
--
"E mdOD OF INVE S T sG AllON > lN SP5 C TION
INSPECTION PERf 0RMED Bv
ORGAN!2ATON CODE OF REGON/HQ CONDUCT-
lNG ACTlVITY t$ee IEMC 05~; m..g, power Report.
an
FROM
TO
'11 - REG'3N AL Of FICE ST AFF
OT HE' R
y
g
y,
,p
,
MO -
DevlVR
MO
D4(
YP
2 -. RESIDEf'T INSPECTOg
'
REGON
DiviftON
BRANCH
ol'b 01il%l9
015
511 ti9
i ' - a'ORMANCE APPRAl$ AL 1 E AM
h
20
25
26
31
32
33
34
35
REGIONAL ACTION
f VPE OF ACTIV Tv CONDUCTEDICheck one tms ontyi
~
"'"
02 - SAFETY
06 - MGMT VK'T
10 - PLANT SEC,
14. INOUIRY
1 - NRC FOPM 691
OJ INCIDENT
01 SPECIAL
11 - INVENT. VER
15 - INVESilGATION
2 - REGIONAL OFFICI LETTER
04 - 'ENF0F *EMENT
08 - VENDOft
12 - SHIPMENTIEXPORT
06 - MGMT. AUDIT
09 - MA f. ACCT.
13- BM POR T
36
37 3B
IN 5PE C T s
sh w t h .
I N INWNG5
TOT AL NUMOF.R
ENFORCEMENT CONFERENCE
REPORT CONT AIN 2 790
LETTER OR REPORT TRANSMITTAL DATE
08' VOLATION4 AND
HELD
INFORMATION
.IA
9
C
D
DEy:AicNS
REPORTSENT
I - Ct1LR
ORkEG
TO WO FOR
LETTER ISSUED
ACTION
h.
2 - VOLATON
3 - DEVIATION
A
B lC
D
A
B
C
D
A
B
C
D
g
%
Mk MO,
DAY
YR.
1 - YES
1 . YES
' " l"
'l~
'] -' l
l
l
4e VIOLATION & DEVIATON
gg
j ,
,
,
39
40 41
42
43
44
49
50 -
Si
'
f 40DULE INf 0RV iTION
MODut f IN80RM ATION
MODULE REO FOLLOWUP
$I0
MODULE NUM8ER INSP
gg
MODULE REO FOLLOWUP
p,h
MODJLE NUM8Ep INSP.
yg. g I
n0
,d5 i i s s
,
Y
$o
'
l
- -
wa
w
w
w
bh
- si
!!
=
st
!! M
si
i
grl
g6BE. j i i s
n
s
,
E } =z<
51
8e ,5
E c f
e ,1
E.': E, =ns
f.t
s-
e
r gM
Eg
g
8
e
5
.
.
.
d. a g s
gg
s
n
s 3.g
.:
<
5
.
i
e,
rM -
-
-
.M
_
.
.
..
. -
.
m
.
al 5 1 i l1,o,1l p)
o,5,e
l, o, o c,
li,l # s3%ol1icibl
oilifwwo k
liiI
'
^
.
i
,
i
~
liil
'
liil
'
'
ope
,i
i i
i
i,
i s
i
EcTe
Mtm
liil
i i
i i
i
i i,
C
C
) gir
i,
i i
i
Vsw
i
i
_
-
o
i
ii
i 1
i 1 i i 1
I i
i i
i
1
0 ,i
0
i
M83 Gi ,l1;o 6l
IiiI # 6 0,olii i31
o,l i s l ais
C-
l
iI
2
Dioio tio.o c,
,
t
A
i
i
liiI
I'il
p.()pp-
& h q-
'
'i
i i
i
ie
i i
i
l
,l
Tyr
liil
DeAsev
c
c
i i
i i
i
I,,1
h ed
_s,
, i
i
liil
o
o
e i
, ,
,
i ,
, ,
,
oioi4 I cio e
liil W 5 7il l7d iol f) Oili r lioio C
i
4
836illliL61
liil
'
^
i
' '
' '
'
' '
' '
' '
'
' '
Mp hEV
li,l
A$Kbct0Al
liiI
c
c
i i
e i
i
i i
i i
i
gg
liil
liil
o
i i
i i
i
i i
, ,
i
o i) il
Ii p, o c,
i liil
I
ds; 9,2 M ioil I
oil 6 7vemt 5 ml7,opl 4 6 7,Rl7ioiol
^
^
e
liil
liiI
7p g 9
'
g
i i
e i
i
i i
i i
i
liil
liil
'
c
i i
i i
i
i i
i ,
i
l,,1
f****F
m m a r e,' A,ON
, ,
, ,
, I,,i
o
o
, ,
, ,
,
ilni.IS
20
o
,,
,$
,.
,.
,,
,
,,
,, ,p .i . i S
,0
,,
,,
,,
, . -
,,
,,
20
,,
,E
__,___ -_
~-------.--------r------------
r-----
- - - - - - ~ - - - - - - - - - " - - - -
-
-
- ~ - ' - '
- ' - -
4
MC Foaw 7,4 A..
DOCKEi NO (8 dig +ts10R UCE NSE
REPORT
'
r i
6
NO. t B Y PRODUCT)(13 cipital
NO.
SE Q-
INSPECTOR'S REPORT
D T o o 0 2 B 5
B 9- 1
S
^
~
(Continuation)
'
8
' Office of Irispection ana Enforcement
c
.
j
{
D
MC'DUL E INF OR M A f TON
MODut t tN80RMAfiON
' MODULE NUM0i# INS'-
Og y
MODULE REO FOLLOWUP fh
MODULE NUMBER INSP
dg
VODutt REO FOLLOWUA
3o
r
-
-
5
88
l l !
rh
$5
b$
! ! !
! !
!
!
!!
!
l
r
=
v 5 0217,o,1l
o,i ,7 boio- c
Ii,I d d f r717ioiol
oio $ oil io o
i liiI
^
o
^
,
i
1
,1
y
a
4
I,,1
a
,,
i ,
i
i i
e i
,
fe w up
liil
CourWD
c
l'il
'
ii
, i
,
' i
i '
i
liil
o
l i .,1
o
,,
, ,
i
, i
, ,
i
W'5 9,317icizl
oizi3 t o,o C
IiiI 4 6 7, Il9,ol
@ io /,vi o c
liiI
>
^
i
,
i
l i- i l
pg
li,1
,,
, ,
i
y
, i
i ,
i
I,,I
-ga p poo
c
li,1
p;,y
e
i,
, ,
,
, ,
i ,
i
I,,1
o
I,,1
o
i,
, ,
,
, ,
, ,
,
W 5 9,ol1,l i 21
oio a
l,o,0 c.
liiI
iI,i
'
Ii,I
'
i
e
i
, i
i i
i
RE$ggv* igg
I i ,' I
I'
l
' '
' '
' Il
'
r
, '
'
6
-Ii,I
c
,,
, ,
,
, ,
, i
i
liil
i,iI
o
,,
, ,
i
i ,
i i
i
a H Ti5loioi l
l
0,0,1
1,oi oC
li,I
iliiI
^
i
e
i
i i
i i
i iiiI
^
liil
pyA
a
liil
i,
, ,
,
, ,
i ,
e
NWifd
liil
c
c
J
le il
i,
, ,
,
i i
, ,
,
Ml"fu s s"# tar.o~
i,l
l
il
Iiil
o
o
,,
, ,
,
, ,
i
1"l4 Is - ,"
o
o r =a
dma
a = w* w r ia ilol.1 - m cs-" * =1
u w-
a, a
a i~a
2,
>>
2.
.
__
_ _ _ _ _ _ , _ _ _ _ - _ - _ - - - , - - - - - - - -
- - - - - - - - - - - - - - - - - " - - - - - - - - - ' - - - ' - - - " ' ' ^ ~ ~ ~ ~ ^ ~
-~
~
_ _ _ _ _ - - _ _ _ _ - - - _ _ _ - _ _ _ _ _ _ _
. _ _ .
_ _ - -
.
percar
%.t.e
o c;oA m a ,
u n t o .. .., ,e o. t a. e.,a
~o , noovci,.m
,
,,,)
!Sto 7 IOl.O I
-
., ,gg , , ,; . s
INSPEC10R'd hEPORT
R6 0 o 6 *k B C
99
( l*
' g ., ; c, ,,,
,,t ' j , 0
~
5m
e
c
,
s
.
o A,{
(Contir uetion)
,
2
3
,
-
.
Of fice of Inspectior, and Enf orcement
y
_
og
+
l
v
n
r,
vsous teom aa ps ves'ck wnre uc swx cnava:te s tor en:% erem et tne rent escenas ters number. st wa:be <+scesse p ro careonrase Lomst omes tc ha cherevers eech t
,
TK [.N;
o S LG_1.Tt
5L-
__1
.,
,
..
.* ' -e
Jo9 '
-
>
a,o ;1jn
Ls
-a
i
%
K,
aw
.
c,x 1
-az.
1
u
_r
-
u-r
.
-
6
7
8
9
10
11
12
13
14
16
'6
17
,.
18
l
59
20
-
at
22
23
24
7
76
77
29
29
30
31
37
33
Se
E
at
38
_ _ _ _ . _ _
6'
r.
NRCfonu
COCKCI NO. (S ainnslOR UCE NSE
I'Ef* ORT
MODULE NUMBER
- fspq: gv
Je.j$,y,f A
-
)l l l)lQl 7l l
d
NO. (BY PRODUCT)(13 0'0Hs)
NO.
,
.-
INSPECTOR'S REPORT.
O 5 o o c 'I O E
B 9
1
3
^
v eg'g*,*,vg'"
d'h
'
sure
,
e
(Continuation)
,
,
,
,
,
,
, y, ,c
- Office of Incpection and Enforcement
g
-
c
aD
a
o
,,
,.
,,
vo' AisON OR DE vnATON wnter up to 2400 characte rs tor each tsem. N tno test enceeds tnia number, rt wnt be necessary to personesse. Limn unes to so charactern esen.)
.
,.
2-
.
,
~% licemue
(%drm +o Ts 6.7.1 twd KG l.%.
,
Jew 1%> .4.+dme 4c rwd
-Iru M
.
hwowhw Y h +ksk a vdive (New 4and
do
..
~
IAT M & W .
'
' ?.
a.
9.
10
11.
12.
13.
14.
15.
18,
'
17.
is.
19.
20.
21.
22.
23
24.
29
28
27
2
29
- n
31.
.
32.
ss
!
sa
35
'
se
-
at
h
2
si
- '"*"'
U.S. NUCLEAR REGULATORY COMMISSION
- - - -
- - - - - ~ - - - ~
-________x____--_-__
. - _ - - _ _ _
___ _ -_
_ _ _ . _ _ _ _