ML20135G016
| ML20135G016 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/29/1985 |
| From: | Panciera V, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20135G002 | List: |
| References | |
| 50-302-85-33, IEIN-85-058, IEIN-85-58, NUDOCS 8509180061 | |
| Download: ML20135G016 (13) | |
See also: IR 05000302/1985033
Text
.
Q RfC
UNITED STATES
f
NUCLEAR REGULATORY COMMISSION
%' \\
REGION ll
'
[ ,I [
101 MARIETTA STREET.N.W.
- I
" c
ATL ANTA, GEORGI A 30323
i,
,gv...../
Report No.:
50-302/85-33
Licensee:
Florida Power Corporation
3201 34th Street, South
St. Petersburg, FL 33733
Docket No.:
50-302
License No.
Facility Name: Crystal River 3
Inspection Conducted: July 27 - August 16, 1985
T. F. Stetka," SenioFT<esident Insp . wor
~,
$ AY0$
Inspector:
'
~
ate Signed
Accompanying Personne
E. Tedrow, Resident Inspector
. Sasser, Resident Inspector (0conee) (August 9-16)
5)/2')/ f'S
j
/
Approved by:
.
],
n , , , j fA ,,
V'. W . Pa nbiWa','th fe f , P roj e c t S ec t i o n 2 B
" l)a tE' Si gned
Division of Reactor Projects
SUMMARY
Scope:
This routine inspection involved 112 inspector-hours on site by two
resident inspectors in the areas of plant operations, security, radiological
controls, Licensee Event Reports and Nonconforming Operations Reports, plant
startup preparations, and licensee action on previous inspection items. Numerous
facility tours were conducted and facility operations observed.
Some of these
tours and observations were conducted on backshifts.
Results:
Three violations were identified (Failure to take and analyze grab
samples each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by Technical Specification (TS) 3.3.3.1 when
RM-A12 is out of service, paragraph 5.b.(1); failure to have at least one
operable EDG as required by TS 3.8.1.2.b, paragraph 6.b.(3); and failure to have
an adequate equipment control procedure for the performance of valve lineups as
required by TS 6.8.1.a, paragraph 7.)
8509180061 850829
ADOCK 0500
2
0
-
.
.. -
_ _ -
-
_.
__ .
.
-
--
-
.
.
>
,
REPORT DETAILS
4
.'
1.
Persons Contacted
i
Licensee Employees
.
!
- S. Baggett, Nuclear Principal Mechanical / Structural Engineer
t
G. Boldt, Nuclear Plant Operations Manager
- P. Bleedlove, Nuclear Records Management Supervisor
2
.(
- C. Brown, Assistant Nuclear Outage and Modifications Manager
!
- J. Bufe, Nuclear Compliance Specialist
D. Fields, Nuclear Reliability Supervisor
i
E. Howard, Director, Site Nuclear Operations
'
- W. Johnson, Nuclear Plant Engineering Superintendent
- K. Lancaster, Manager, Site Nuclear Quality Assurance
i
- J. Lander, Nuclear Outage and Modifications Manager
- C. Long, Senior Quality Auditor
j
- J. May, Welding Engineer
- P. McKee, Nuclear Plant Manager
- T. Miller, Nuclear Shift Supervisor
,
W. Neuman, Senior Nuclear Inservice Inspection Specialist
- P. Patel, Nuclear Quality Engineering Supervisor
- W. Rossfeld, Nuclear Compliance Manager
,
- P. Skramstad, Nuclear Chemistry and Radiation Protection Superintendent
,
l
F. Sullivan, Nuclear Electrical /I&C Engineer
K. Vogel, Nuclear Senior Electrical /I&C Engineer
- G. Westafer, Manager, Nuclear Operations Licensing and Fuel Management
4
- K. Wilson, Supervisor, Site Nuclear Licensing
R. Whittman, Nuclear Operations Superintendent
- J. Wright, Nuclear Support Specialist, Chem / Rad Protection Services
1
Other personnel contacted included office, operations, engineering,
4
maintenance, chem / rad and corporate personnel.
- Attended exit interview
2.
Exit Interview
j
The inspector met with licensee representatives (denoted in paragraph 1) at
j
the conclusion of the inspection on August 26, 1985. During this meeting,
j
the inspector summarized the scope and findings- of the inspection as they
l
are detailed in this report with particular emphasis on the violations,
j
unresolved items, and inspector followup items.
5
!
Also during this meeting, the inspectors discussed the observations from
1
walkdowns conducted on the High Pressure Injection (HPI) and Emergency
'
!
i,
4
i
!
i
._.
.
._
-
_,
_ _ . _ .
._
..
.
.~
. .
. -
.
.
.
.
2
..
The following items were identified:
The following valves in the HPI system were missing identification
-
tags: MUV-413, MUV-414, MUV-415, and MUV-335;
Modifications to the EFW system have removed valves EFW-73 and EFW-74
-
but the system drawing, FD-302-082, has not yet been revised; and
Modifications to the EFW system have also removed valve COV-104 and
-
taken the internals from valve CDV-103 but both these valves are still
shown on drawing FD-302-101 and listed in procedure SP-381, Locked
Valve List.
~
The licensee representatives acknowledged the inspector's comments -and did
not identify as proprietary any of the materials provided to or reviewed by
the inspectors during this inspection.
3.
Licensee Action on Previous Inspection Items
(Closed) Inspector Followup Item (302/85-02-01): The licensee has completed
a detailed review of their dose assessment which includes the RADDOSE II
computer program. As a result of this review, it was determined that the
errors made during the emergency plan exercise were caused by personnel
error and not the computer program.
The licensee is taking corrective
actions to minimize a repetition of these occurrences.
(0 pen) Inspector Followup Item (302/85-29-03): The licensee has received
verbal notification from their architect-engineer, Gilbert Associates (GAI),
that at least one of the AHF-22 fans per diesel must be running during
diesel operation. It appears that these fans are necessary to assure proper
,
generator cooling. The licensee has written a Short Term Instruction (STI)
to assure operators are aware of the importance of these fans.
The
'
licensee's diesel
operating procedure, SP-354, already contains the
requirement that these fans must be running for diesel operation. This item
remains open pending receipt of GAI's written evaluation and further review
of this item by the inspectors.
(Closed) Inspector Followup Item (302/85-08-05): The licensee has replaced
valves CHV-68 and CHV-69 with the new Fisher valves.
4.
Unresolved Items
Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or devia-
tions.
A new unresolved item is identified in paragraph 5.b.(2) of this
report.
t
!
.
.
3
5.
Review of Plant Operations
At the beginning of this inspection period, the plant was in the cold
shutdown mode (Mode 5).
Plant heatup was commenced and the plant reached
the hot shutdown mode (Mode 4) at 0543 on August 7,1985, followed by the
hot standby mode (Mode 3) at 2359 on August 8,1985, where it remained for
the duration of this inspection period.
a.
Shift Logs and Facility Records
The inspector reviewed records and discussed various entries with
operations personnel to verify compliance with the Technical Specifica-
tions (TS) and the licensee's administrative procedures.
The following records were reviewed:
Shift Supervisor's Log; Reactor Operator's Log; Shift Relief Checklist;
Auxiliary Building Operator's Log; Active' Clearance Log; Daily Opera-
ting Surveillance Log; Short Term Instructions (STIs); selected
Chemistry / Radiation Protection Logs; and Outage Jhift Manager's (OSM)
Log.
In addition to these record reviews, the inspector independently
verified clearance order tagouts.
No violations or deviations were identified.
b.
Facility Tours and Observations
Throughout the inspection period, facility tours were conducted to
observe operations and maintenance activities in progress.
Some
operations and maintenance activity observations were conducted during
backshifts.
Also, during this inspection period, licensee meetings
were attended by the inspector to observe planning and management
activities.
The facility tours and observations encompassed the following areas:
Security Perimeter Fence; Control Room; Emergency Diesel Generator
Room; Auxiliary Building; Intermediate Building; Battery Rooms;
Electrical Switchgear Rooms; and Reactor Building.
During these tours, the following observations were made:
(1) Monitoring Instrumentation - The following instrumentation was
observed to verify that indicated parameters were in accordance
with the TS for the current operational mode:
Equipment operating status; area, atmospheric, and liquid
radiation monitors; electrical
system
lineup;
reactor
operating parameters; and auxiliary equipment operating
parameters.
l
l
L
_ _
_
__
_
_
_
.
.
4
On August -9,
while examining the area radiation monitors, the
inspector noted that monitor RM-A12, which monitors the main
r
'
condenser vacuum pump, was removed from the panel and therefore
out of service. TS 3.3.3.1, item 2.C in table 3.3-6 requires that
,
grab samples be taken and analyzed for gross activity at least
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> whenever this monitor is out of service.
The inspector discussed the missing instrument with -licensee
personnel and questioned whether grab samples were being obtained.
The inspector 'also reviewed work requests to determine when the
instrument was removed from service.
As a result of this review, it appears that RM-A12 was removed
from service on the evening of August 7.
A grab sample was taken
and analyzed at 6:27 p.m.
on August 7, but only because this
sample - is required to be taken weekly and this time period
happened to coincide with the due date. Additional grab samples
4
'
were not taken until approximately 10:30 a.m., on August 9, when
the inspector asked about the missing instrument.
Therefore, the-
time period between samples was approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
l
Failure to take and analyze these grab samples at least once every.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is contrary to the requirements of TS 3.3.3.1 and is
considered to be a violation.
'
Violation (302/85-33-01):
Failure to take and analyze grab
samples each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as required by TS 3.3.3.1 when RM-A12 is
out-of-service.
(2) Safety Systems Walkdown - The inspector conducted a walkdown of
the high pressure injection (HPI), emergency feedwater (EFW), and
control rod drive (CRD) systems to verify that the linaups were in
accordance with license requirements for system oper oility and
i
that the system drawings and procedures correctly reflect
"as-built" plant conditions.
On August 2, 1985, the inspectors completed a walkdown of the EFW
,
system.
During this walkdown, which utilized feedwater system
drawing FD-302-081, the inspectors were unable to determine the
.'
location of valve FWV-159, a vent valve on the "B" Once Through
Steam Generator (OTSG) EFW line. The inspectors requested help
'
from the licensee but, even with the aid of an Auxiliary Nuclear
Operator (AN0), were unable to find the valve. The EFW system has
recently undergone major modifications to install a new Emergency
.Feedwater Initiation and Control (EFIC) system and these modifica-
tions have replaced the section of pipe in which this valve is
supposed to be located.
'After discussions with licensee
personnel, it appears that the EFIC modifications package still
requires the installation of vent valve FWV-159.
The licensee
d
i
!
!
,
-- -- ,
y
,
,
+
.-
- ,
-
-
-
-~,,w,
_ , . , -
9
&m,,,-
,
,--,.-yem
- -
. . -
.,
.
-
.
_-
--
.-
---
-
.
t
5
-
,
i
will provide information to the inspector to demonstrate that
'
either valve FW-159 actually does exist in the field, or that the
modification package 1for the EFW system has deleted this valve
because it is no longer required.
,
Unresolved Item (302/85-33-02):
Review the licensee's determina-
tion that valve FWV-159 actually exists or that EFIC modifications
have deleted this valve.
(3) Shift Staffing - The inspector verified that operating shift
staffing was in accordance with TS requirements and that control
4
room operations were being conducted in an orderly and profes-
sional manner.
In addition, the inspector observed shift turn-
overs on various occasiens to verify the continuity of plant
status, operational problems, and other pertinent plant informa-
tion during these turnovers.
No violations or deviations were . identified.
(4) Plant Housekeeping Conditions - Storage of material.and components
"
and cleanliness conditions of various areas throughout the
facility were observed _to determine whether safety and/or fire
hazards existed.
1
No violations or deviations were identified.
T
l
(5) Radiation Areas - Radiation Control Areas (RCAs) were observed to
,
verify proper identification and implementation. These observa-
'
tions included selected licensee conducted surveys, review of
step-off pad conditions, disposal of contaminated clothing, and
area posting.
Area postings were independently verified for
,
accuracy through the use of the inspector's own radiation moni-
toring instrument. The inspector also reviewed selected radiation
.
work permits and observed personnel use' of protecthe clothing,
respirators, and - personnel monitoring devices to assure that the
!
licensee's radiation monitoring policies were being followed.
.
l
.
No violations or deviations were identified.
l
}
(6) Security Control - Security controls were observed to verify that
security barriers are intact, guard forces are on duty, and access
to Protected Area (PA) is controlled in accordance with the
,
facility security plan. -Personnel within the PA were observed to
verify proper display of badges and that personnel requiring
,
i
escort were properly escorted. ' Personnel within vital areas were
i
observed to ensure proper authorization for the area.
No violations or deviations were identified.
k
'
+
r
m-
w
v.
,
-,
,y3-+
e
p-
-.-----.m
m
,.
- ----- --
..-g-
- -
-
-.m-.v-
.
.
6
-(7) Fire Protection - Fire protection activities, staffing, and
equipment were observed to verify that fire brigade staffing was
appropriate and that fire alarms, extinguishing equipment,
actuating controls, fire fighting equipment, emergency equipment,
and fire barriers were operable.
No violations or deviations were identified.
(8) Surveillance
. Surveillance tests were observed to verify that
approved procedures were being used; qualified personnel were
conducting the tests; tests were adequate to verify equipment
operability; calibrated equipment, as required, was utilized; and
TS requirements were followed.
The following tests were observed and/or data reviewed:
-
SP-102, Control Rod Drop Time Tests;
'
SP-204, Class 1 System Leakage Test for Inservice Inspection;
-
SP-210, ASME Class 3 Hydrostatic Testing;
-
SP-295, Surveillance Capsule Insertion / Removal (and procedure
-
FP-502 for the supporting data);
SP-323, Evacuation and Fire Alarm Demonstration;
-
SP-332, Monthly Feedwater Isolation Functional Tests;
-
SP-340, ECCS Pump Operability;
-
-
SP-349, Emergency Feedwater System Operability Demonstration;
SP-381, Locked Valve List (Position Verification of Locked
-
Valves);
SP-414, High Pressure Injection Flow Verification Test;
-
SP-417, Refueling Interval Integrated Plant Response to
-
Engineered Safeguards Actuation;
SP-422. RC System Heatup and Cooldown Surveillance;
-
- '
SP-440, Unit Startup Surveillance Plan; and
-
SP-457, Refueling Interval ECCS Response to a Safety Injec-
tion Test Signal.
l
No violations or deviations were identified.
I
I
. . .
. . .
_-
.
- - -
-
_
. ._. - .. . _ _ . - _
. - , -
'
..
..
7
i
j.
,
(9) Maintenance Activities - The inspector observed maintenance
l
"
activities to verify that correct equipment clearances were in
'
I
effect; Work Requests and Fire Prevention Work Permits, as
required, were issued and being followed; Quality Control
i
personnel were available for inspection activities as required;
,
i
j
and TS requirements were being followed.
!
Maintenance was observed and work packages were reviewed for the
following maintenance activities:
1
Replacement of engineered safeguards actuation relay for
-
MUV-41, DWV-160, and CFV-26;
,
Replacement . of a torque switch for valve EFW-1, emergency
-
feedwater pump #2 (EFP-2) main condenser hotwell suction
isolation valve;
)
!
Troubleshooting of the " A'.'
-
(EDG-3A) for relay problems and governor oil change;
Control Rod Drive Mechanism (CRDM) breaker replacement for
-
rod #7 in rod group #7;
_1
Replacement of EDG-3A in accordance with procedure MP-117;
-
j
and
l
Verification of CRDM power patch panels, group patch panels,
-
!
and position indicator panels to. incorporate cycle 6 rod
positions in accordance with maintenance procedure MP-108-A.
!
No violations or deviations were identified.
(10) Radioactive Waste Controls - Selected liquid releases and solid
waste compacting activities were observed to verify that approved
,
I
procedures were utilized, that appropriate release approvals were
obtained, and that required surveys were taken.
No violations or deviations were identified.
i
(11) Pipe Hangers and Seismic Restraints - Several pipe hangers and
seismic restraints (snubbers) on safety-related systems were
observed to ensure that-fluid levels were adequate and no leakage
was evident, that restraint settings were appropriate, and that
anchoring points were not binding.
No violations or deviations were identified.
l
7
.
.
.
.
. .
.
.
.
.
. .
.
.
i
.
-
i
.
,
'
.
i
'
8
i
i
6.
Review of Licensee Event Reports and Nonconforming Operations Reports
,
a.
Licensee Event Reports (LERs) were reviewed for potential generic
,
impact, to detect trends, and to determine whether corrective actions
'
appeared appropriate.
Events, which were reported immediately, were
!
reviewed as they occurred to determine if the TS were satisfied.
LER 84-15 reported that inadvertent Emergency Diesel Generator (EDG)
starting was caused by failure of an air filter petcock on the air
'
start system and that replacement of the filter was being evaluated.
This evaluation resulted in a modification (MAR 85-02-01) that replaced
1
the air filters with a more durable filter.
This modification has been installed and tested.
The inspectors have
reviewed the modification package and examined the installation and
'
note that this installation should prevent recurrence of these events.
This LER is considered to be closed.
i
,
)
b.
The inspector reviewed Nonconforming Operations Reports (NCORs) to
j
verify the following:
compliance with the TS, corrective actions as
identified in the reports or during subsequent reviews have been
,
j
accomplished or are being pursued for completion, generic items are
identified and reported as required by 10 CFR Part 21, and items are
j
reported as required by TSs.
.
I
All NCORs were reviewed in accordance with the current NRC enforcement
i
policy.
As a result of this review, the following items were identified:
I
(1) NCOR 85-142 reported that several anchor supports for three cable
trays (trays 134, 135, and 212) are pulling away from ~ the walls
inside the reactor building.
The licensee has investigated the
'
condition of the cable trays and a determination as to whether the
<
trays are safety-related and/or seismically qualified has been
<
.
made, but this determination was not available for the inspector's
4
review.
,
)
Inspector Followup Item (305/85-33-03):
Review the licensee's
investigation into the condition of cable trays 134,135, and 212
l
and the determination as to whether the trays are safety related
,
and/or seismically qualified.
(2) NCOR 85-118 reported a non-terminated circuit in the Non-Nuclear
i
Instrument (NNI) cabinet for the third mechanical seal cavity
!
bleedoff pressure transmitter (RC-19A-PT2) for Reactor Coolant
{
The licensee is presently evaluating the cause
and safety significance of this item.
i
i
{
'
-
-
-
,
.-
. .. - . - - . -
.
.. - ..
. . ~ . - . _ - . - - - - .
- ~ - -
,
..
.
!
!
9
,
t
,
'
!
!
t
j
Inspector Followup Item (302/85-33-04):
Review the licensee's
!
evaluation of the non-terminated circuit in the NNI cabinet for
.
RC-19A-PT2 for cause and safety significance.
4
(3) NCOR 85-137 reported that a portion of surveillance procedure
!
SP-135, Engineered Safeguards Actuation System Response Time Test,
j
which tests the B emergency diesel generator (B-EDG) automatic
l
l
load sequence timers as required by TS 4.81.2, was not completed
until July 25 at which time the automatic load sequence timers
were found to be operable.
This surveillance is necessary to
,
j
assure that the B-EDG is operational. The licensee's assessment
1
of this event as documented in this report was only that "No core
1
alterations or positive reactivity additions in progress."
.
I
The B-EDG was declared inoperable on June 27 at which time the
I
'
A-EDG was made inoperable for planned maintenance. As of July 25,
,
the A-EDG was still inoperable. A review of logs by the inspector
s
indicates that a core alteration (the reactor vessel . head was
4
j
installed) occurred on July 10 and that the plant changed modes
!
from the refueling mode (Mode 6) to the cold shutdown mode
j
(Mode 5) on July 18.
i
j
Failure to have at least one operable EDG is contrary to the
requirements of TS 3.8.1.2.b.
Also, since an operational mode
i
change was made without an operable EDG, this event is contrary to
j
the requirements of TS 3.0.4 which requires that the Limiting
l
Conditions for Operation (LCO) be met prior to changing opera-
j
tional modes,
j
Violation (302/85-33-05):
Failure to have an operable EDG as
!
j
required by TS 3.8.1.2.b and to meet the LCO requirement prior to
changing operational modes as required by TS 3.0.4.
1
7.
Design, Design Changes and Modifications
!
,
4
,
j
Installation of new or modified systems were reviewed to verify that the
I
j
changes were reviewed and approved in accordance with 10 CFR 50.59, that the
j
changes were performed in accordance with technically adequate and approved.
i
procedures, that subsequent testing and test results met acceptance criteria
}
or deviations were resolved in an acceptable manner, and that appropriate
i
drawings' and facility procedures were revised as necessary.
This review
l
l
included selected observations of modifications and/or testing in progress.
i
1
L
4
L
l
The following modification approval records (MARS) were reviewed and/or
'
associated testing observed:
'
t
I
Replacement of High Pressure Injection Valves MUV-23, MUV-24, MUV-25,
i
!
-
and MUV-26 in accordance with MAR 83-09-22-01; and functional stroke
'
l
test in accordance with procedure TP-1-
'
i
!
t
-
t
1
6
j
.
.
,
.
4
3
!
10
i
'
l
s
Replacement of control complex chiller cooling water flow control
-
valves CHV-68 and CHV-69 in accordance with MAR 81-03-09-01;
i
Integrated Test of the Remote Shutdown Panel in accordance with
-
procedure TP-18 of MAR 77-07-01;
i
!
Functional testing of the Emergency Feedwater Initiation and Control
-
(EFIC) system in accordance with the following test procedures of
i
i
MAR 80-10-66:
,
j
Tp-6, EFIC System Test
j
TP-14, Operability Test of EFV-55,56,57,58
i
t
i
Installation of a fuel oil relief valve on the "A" Emergency Diesel
~
-
j
Generator (EDG-3A) in accordance with MAR 85-03-05-01; and
I
Replacement of diesel generator air filters in accordance with
-
i
MAR 85-02-02-01.
!
]
On August 8, while observing the performance of TP-14 on the EFIC system,
'
4
the inspector noted that TP-14 required the completion of the valve check
!
list in procedure OP-605, Feedwater System.
This valve check list was
completed and signed off within about one hour of the commencement of TP-14.
When test TP-14 was run, the test did not work because there was no flow
j
from the emergency feedwater pump (EFP-1).
Investigation by the licensee
j
determined that valve EFV-7, the discharge stop check valve for EFP-1 was
,
j
shut. The valve check list for OP-605 requires this valve to be opened and
]
it was signed off on this checklist as open.
l
The licensee's Operations Section Implementation Manual (OSIM) requires
'
valve position indication to be verified by either hands-on operation of the
l
valve, visual observance of valve stem position, or by the fact that the
l
system is operating properly (e.g., the system is pressurized and fluid is
!
not coming out of vent and/or drain valves).
In this-case, the operator
apparently verified the valve position by stem observation only.
1
J
Based upon these findings, the OSIM valve verification procedure is
2
considered to be inadequate to provide positive equipment control.
An
!
inadequate equipment control procedure is contrary to the requirements of
TS 6.8.1.a. as listed in Regulatory Guide 1.33 and is considered to be a
violation.
1
i
Violation (302/85-33-06):
Failure to have an adequate equipment control
j
procedure for the performance of valve lineups as required by TS 6.8.1.a.
!
l
1
'
i
!
i
.
-
.
__
..
.
-
_
.
- - -
._
.
.
11
,
8.
Review of IE Notices (IENs)
The inspector discussed IEN 85-58, Failure of General Electric Type AK-2-25
Reactor Trip Breakers, with licensee representatives and has been informed
that appropriate checks of the critical parameters of the breakers has been
accomplished in accordance with procedure PM-118.
The inspectors will
verify implementation of the guidelines listed in IEN 85-58.
Inspector Followup Item (302/85-33-07):
Verify the implementatior, of the
guidelines discussed in IEN 85-58 in the reactor trip breaker maintenance
procedure PM-118.
1
1
9.
Loss of Non-Nuclear Instrumentation (NNI) Emergency Procedure Review
The inspector reviewed emergency procedures (EPs), abnormal procedures
l
(APs), and verification procedures (VPs) to verify that the loss of NNI
j
event was properly covered.
!!
In 1983, the licensee revised their EPs and APs from " event" oriented
j
procedures to " symptom" oriented procedures as recommended in NUREG 0737.
j
These revisions also established the verification' procedures primarily for
use of the shift technical advisors (STAS) to assist operators in the
,
!
analysis of an accident.
!
Following their event of February 26, 1980, the licensee committed to
i
develop and implement a procedure that addressed a loss of NNI.
This
]
commitment was confirmed by an NRC Confirmatory Order. The transition from
j
event oriented to symptom oriented procedures made coverage of a loss of NNI
event doubtful.
!
The inspector compared the old loss of NNI procedure (EP-114) with proce-
'
dures AP-580, Reactor Protection System (RPS) Actuation, and VP-580, Plant
Safety Verification Procedure.
l
The major result of a loss of NNI is a reactor and turbine trip. The old
EP-114 directed the operators to perform actions that would restore the
i
plant to a stable condition following these trips.
The only apparent
t
i
difference between EP-114 and the reactor / turbine trip procedures was an
i
additional statement to " Assess plant condition using redundant or alterna-
'
tive instrumentation."
i
The new AP-580 directs the actions to be taken to recover from a reactor /
This procedure provides more direction for response to the
I
reactor / turbine trip. The VP-580 provides an additional check to verify
that conditions are as expected.
4
l
Following review of these procedures, the inspector stated that it may be
!
appropriate to add a caution or note to the AP or VP to remind operators to
!
assure their instrumentation operable.
The licensee acknowledged the
.
- - _ - , - . - - ,
-
. - - - . - _ - - .
- - -
., -
_ _ . - _ . _ , -
,_
, . . . ~ , . . . _
. , - . _ , .
.- . - -
.
-
C.
- __ -
.
.
12
inspector's comment and stated that operator training has taught operators
to verify instrument status by observation of other plant parameters. The
licensee representative also stated that the inspector's comments were
appropriate and that the procedures would be evaluated to determine if such
a note would enhance the procedure.
Inspector Followup Item (302/85-33-08): Review the licensee's evaluation to
add loss of instrumentation caution notes to procedures AP-580 and/or
VP-580.
10. Licensed Operator Training Observations
The inspector observed licensee conducted licensed operator training
activities to evaluate the quality of the training. The following training
activities were observed:
Oral boards and system walkthroughs for three candidates;
-
Operation of the newly installed remote shutdown panel (RSP); and
-
Operation of the turbine trip valve on the turbine driven emergency
-
feedwater pump (EFP-2).
The inspector noted that the training on the RSP was being conducted due to
a commitment made to the Nuclear General Review Committee (NGRC), the
licensee's offsite safety review committee.
This training was judged
effective since it provided each operating shift with hands-on training on
the new panel.
Due to the problems at the Davis-Besse plant, as reported in NUREG-1154, the
licensee decided to conduct training on the EFP-2 turbine trip valve to
assure operators (both licensed and non-licensed) were knowledgeable in the
trip valve operation.
No violations or deviations were identified.