ML17262A655
| ML17262A655 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/04/1991 |
| From: | Eapen P, Moy D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17262A654 | List: |
| References | |
| 50-244-91-18, GL-88-17, NUDOCS 9111180072 | |
| Download: ML17262A655 (14) | |
See also: IR 05000244/1991018
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report
No.
50-244/91-18
Docket No.
50-244
License
No.
Licensee:
Rochester
Gas
and Electric Cor oration
49 East Avenue
Rochester
14649
Facility Name:
Ginna Nuclear Power Plant
Inspection At:
Ontario
Inspection
Conducted:
Au ust
-16
1991
Inspector:
D. T. Moy,
eacto
En ineer
Systems
Section
E
ov,4 ice
date
Approved by:
Dr.
P.
K. Eapen,
Chief.
Systems
Section,
EB,
date
Ins ection Summar:
Routine
announced
safet
ins ection
on Au ust
12-16
1991.
Ins ection
Re ort No. 50-244/91-18
.
L "
7. "L
of Decay Heat
Removal" during non-power operation.
The inspector
reviewed
mid-loop operating
procedures,
instrumentation,
plant hardware modification
and thermal hydraulic analysis
as related to reduced reactor coolant
system
inventory operation.
Ins ection Results:
The Generic Letter 88-17 recommendations
as described
in
the licensee
s response
were adequately
implemented.
No violations or additional
unresolved
items were identified.
'Vf11i80072 9iif04
ADOCK 05000244
6
'
1.0
Lon
Term Pro
rammed
Enhancement
"Loss of Deca
Heat Removal'on-Power
0 eration
Loss of Decay Heat
Removal
(DHR) during non-power operation
and the
consequences
of such
a loss are of significant safety concern.
Many
events of loss of DHR have occurred at nuclear
power plants while the
has
been drained
down for such
a mid-loop activities
as
inspection
and repair of a reactor coolant
pump.
These activities are
often in progress
when both the reactor coolant
system
and
the'ontainment
are less
than adequately
secured.
"Loss of Residual
Heat
Removal
(RHR) while the Reactor Coolant
System
(RCS) is Partially Filled," was i ssued to all licensees
of
operating
and holders of construction
permits
on July 9,
1987.
Responses
to the
NRC indicated that
some licensees
did not understand
the
identified problems,
and the problems
have continued,
as evidenced
by
occurrence
of events
since the generic letter was issued.
The seriousness
and continuation of these
problems resulted
in issuance
of GL 88-. 17.
It requires
the recipients to respond with two plans of
actions:
a.
A short-term
program entitled "expeditious actions" that was
essentially limited to reduced
inventory conditions.
b.
A longer-term
program entitled
programmed
enhancements.
The Generic Letter stated that the
programmed
enhancements
consisting of
hardware installation and/or modification,
and
programmed
enhancements
that depend
upon hardware installation
and modification, should
be
implemented:
a.
by the
end of the first refueling outage that is initiated
18 months
or later following receipt of the GL, or
'.
by the
end of the
second refueling outage following the receipt of
the GL, which ever occurs first. If a shutdown for refueling has
been initiated as of the date of receipt of this letter, that is to
be counted
as the first refueling outage.
Programmed
enhancements
that do not depend
upon hardware
changes
were to
be implemented within 18 months of receipt of the GL.
The licensee
provided the responses
for "expeditious actions"
and
"programmed
enhancements"
respectively
in letters dated January
4,
1989
and February
1,
1989.
NRC reviewed the
above expeditious
actions
response
and documented
the conclusion in
NRC letter dated
February
22,
1989.
The
licensee's
short term or expeditious actions
program
was previously
reviewed
as detailed in Inspection
Report 50-244/89-05.
The purpose of this inspection
was to assess
the adequacy
of the licensee's
long term progr'am for Generic Letter 88-17 as detailed
below:
1. 1
-Instrumentation for Mid Loo
0 eration
Generic Letter 88-17 provided the following recommendations
for
instrumentation for Mid Loop operation:
Provide reliable indications of parameters
that describe
the state of
the
RCS and the performance
of systems
normally used to cool the
for both normal
and accident conditions.
At -a minimum, provide the
following i'n the Control
Room:
b.
two independent
RCS level indications
at least
two independent
temperatures
measurements
representative
of the core exit whenever
the Reactor
Vessel
(RV) head is located
on top of the
RV
the capability of continuously monitoring
DHR system
performance
whenever
a
DHR system is being
used for cooling the
RCS.
d.
visible and audible indications of abnormal
conditions in
temperature,
level,
and
DHR system performance.
Ins ection Findin
(a)
Reactor
Coolant
S stem Water Level Indications
As stated
in the response
the Licensee
upgraded
and installed
two reactor
water level indications in
the control
room.
Engineering
Work Request
(EWR) 4671,
(Rev.
0,
December
19,
1989) Reactor
Loop Level Upgrade
replaced
the
pr'esent
system
(RCS) "B" loop level indication
hardware
including pressure
transmitter,
local level indicator
and the control board indicator with modern
equipment.
A
functionally independent
and reliable level indication instrument
system
was also installed
on the
RCS "A" loop.
The inspector
reviewed "Calibration and Uncertainly Analysis" -of
Loop Level
Upgrade Modification Package
(EWR 4671),
Rev 1,
April 18,
1990.
This Calculation
was based
on Westinghouse's
report "Report
on Calculation of Instrument Uncertainties for
the
R.
E. Ginna dated
November
1983.
The calculated
instrument
uncertainties
for the upgraded
level indication system
were 2.31
inches
and 1.73 inches for loops
A and
B respectively.
Both are
within their respective
maximums of 3.0 inches
and 2.5 inches
specified in the Design Criteria Section of "Loop Level Upgrade"
package
EWR-4671.
These uncertainties
are
based
on the assumption
'4
'hat the transmitters
were calibrated to
a 0.5% accuracy
and
Main Control board was'calibrated
to
a .1.50% accuracy.
The
inspector
reviewed this calculation
and found it to be
technically adequate.
(b)
The inspector
concluded that the'evel
instrument uncertainties
are appropriate for use during mid-loop operation.
The inspector
reviewed the "Level Correction for LIT-432A due to
RHR Flow,"
EWR 4671-ME-001,
Rev.
0, February
8,- 1990.
The licensee
incorporated corrections to account for the location of the
'ensing
line tap
and the pressure
loss
due to the Flow in the
RHR suction line.
The inspector verified that the flow correction
=
factors were incorporated
in the plant process
computer
system.
The inspector
had
no further concerns
in this regard.
Core Exit Thermocou les
Tem erature
Indications
The licensee
assures
the availability of two core exit
thermocouples
during reduced
inventory operations
(Operating
procedures
No. 0-2.3. 1,
Rev.
43,
Step
3. 14. 1 to 3. 14.3.2).
The
thermocouples
are digitally displayed
in the control
room.
The
range of this indication is from 0 to 2300'F.
The temperature
indications are fed to the Plant Process
Computer
System
(PPSC)
to monitor and alarm
as necessary.
The temperatures
are required
to be manually logged
once per hour when the
PPCS is'perating;
otherwise,
these
temperatures
are required to be logged at
15
minute intervals.
(c)
Monitorin
Deca
Heat
Removal
S stem Performance
The licensee
monitors
pump motor current to detect
RHR pump
vortexing.
The
PPCS continually monitors the following
parameters
in the control
room:
Point ID
Descri tion of Parameter
F0626
L0432A/B
L0432ACF
L0432ACL
I0685A/B
P0682A/B
F0683A/B
T0684A/B
NPSHRHRA/B
RHR l oop flow
Reactor Coolant loop A/B level
Level correction for
RHR flow
Corrected
level
RHR pump A/B motor current
RHR pump A/B suction pressure
RHR discharge
flow loop A/B flow
RHR pump A/B suction temperature
RHR pump A/B margin to
NPSH loss
The inspector
reviewed
each of the above
parameters
5 alarm
setpoint against
the "Analog Input Point Data
Base Values"
and
verified that the alarm setpoints
were adequately
established
for mid-loop operation.
The inspector also noted that, the
RHR system is routinely
monitored in the main control
room utilizing RHR system flow
( FT 626),
system pressure
system temperature
(FT-630)
and loop level
(PT 432A & B).
Based
on the above,
the inspector
concluded that the .licensee
implemented visible and audible indications of temperature,
level
and
RHR system
performance
per Generic Letter 88-17.
1.2
Review of Mid Loo
0 eratin
Procedures
recommended
the development
and implementation of
the following to cover reduced
inventory operation.
(a)
procedures
that cover
normal operation of the
NSSS,
containment,
and supporting
systems
under conditions for which cooling would
normally be provided by
DHR systems.
(b)
procedures
that cover emergency,
abnormal,
off-normal, or the
equivalent operation of the
NSSS,
the containment,
and supporting
systems if an off-normal condition occurs while operating
under
, conditions for which cooling would normally be provided by
systems,
k
(c)
administrati've controls that support
and supplement
the procedures
in items above,
and all other actions identified in this
communication,
as appropriate.
The .inspector
selected
the following mid-loop operating
procedures
for review to determine
whether
the Licensee
has
completed
the Generic Letter 88-17 recommendations:
Operating
procedure
No. 0-2.3. 1.
Rev.
43, June
18,
1991, "Draining
and Operation at Reduced
Inventory of the reactor coolant system".
Operating
procedure
No. 0-2.3. 1A, Rev.
9,
May 3,
1991,
"Containment
closure capabilities
in two hours during
RCS Reduced
Inventory
Operation".
Operating
procedure
No.
RHR.2,
Rev.
3, April 20,
1990,
"Loss of
RHR while operating at
RCS reduced
inventory conditions.
These
procedures
provide instructions for operation and'urveillance
whenever there is fuel in the reactor vessel
and the reactor coolant
system is in
a reduced
inventory condition (3 feet below the top of
the reactor flange or
RCS less
than
64 inches
on the level
B indicator
in the control room)..
0'
A majority of the licensee's
mid loop operating
procedures
were
developed
using the plant and utility experi.ence.
For each 'of the
above procedures,
the inspector verified that the precautions,
limitation and entry conditions to mid-loop operation
were techni'cally
adequate.
The inspector
found that the licensee's
actions
were
consistent
with the
GL recommendation
in this regard.
1.3
Review of RCS Inventor
Addition
In this area,
recommended
to:
(a)
Provide adequate
operating,
and/or available
equipment
of high reliability for cooling the
RCS and for avoiding
a loss
of
RCS cooling.
(b)
Maintain sufficient existing equipment
in an operable
or available
status
so as to mit'igate loss of DHR or loss of
RCS inventory
should they occur.
This should include at least
one high pressure
injection
pump and
one other system.
The water addition rate
provided by each
equipment
should
be at least sufficient to keep
the core covered.
(c)
Provide adequate
equipment for personnel
communications that
involve activities related to the
RCS of systems
necessary
to
maintain the
RCS in a stable
and controlled condition.
The inspector verified that the Licensee
has procedures
and
administrative control (Operating procedure
No. 0-2.3. 1,
Rev.
43,
steps
3. 17 to 3. 19) to provide at least
two adequate
means of adding
inventory to the
RCS, in addition to the
pumps of the normal
systems.
One source of inventory makeup is the gravity flow from
refueling water storage
tank
(RWST)
(DWG. 33013-1261,
Rev.
19 and
-1247,
Rev.
16).
(The injection flow into
RCS via the gravity feed
system is approximately
2500
GPM.)
The second
source of the makeup
is the High Pressure
charging
system path from the
RWST through the
charging
pump and heat exchangers.
(The injection flow into
RCS via
this high pressure
system is approximately
60
GPM)
(DWG 33013-1265
sheets
1
& 2,
Rev. 4).
The third source of the makeup is from RWST
through the safety injection pumps.
(The safety injection rate is
approximately
500
GPM)
(DWG. 33013-1262
sheet
1&2 Rev.
7 & 3
respectively).
The inspector verified that these
flow paths are
available
as stated
in the licensee's
response
letter to GL.
Inspector further verified that injection flow is higher than the
core boil off rate
(51
GPM) and sufficient to keep the core covered
during mid-loop operation.
The inspector
concluded that the licensee's
actions
were acceptable
as they were consistent with the licensee's
commitments in this regard.
0'
1.4
Review of Thermal
H draulic Anal sis
The Generic letter
recommended
that licensees
conduct analyses
to
supplement existing information and develop
a bases for procedures,
instrumentation installation
and response,
and equipment
interactions
and 'response.
The inspector
reviewed Westinghouse's
Mid Loop'Calculation
Report
NS-OPLS-OPL-I-89-111,
dated
February
20,
1989.
This report provides
best estimate calculations of the time to reach saturation,
the boil
off rate,
the minimum vent size required to prevent pressurization
and the estimated
time to core uncovery
as
a function of time after
shutdown.
The inspector
reviewed the following sections of the report:
Time for RCS in the core to reach saturation
temperature
Time to core uncovery with a large vent area
Makeup flow rate for boil off
Vent area
required to reduce
RCS pressurization
The inspector verified that the results of the
above calculations
were incorporated
in the mid loop operating
procedures.
For example,
the time to core uncovery (hrs) vs.
time after shutdown curve is used
in Operating
procedure 0.2.3. 1: 18 step
5. 14.2. 1 to 5. 14.3.4.
For
another
example,
core boiling vs. time after shutdown
data
were
used
in drain
down operating
procedure 0.2.3. 1:5, steps
3. 17 to 3. 19.
The
inspector
found that the licensee's
actions
were acceptable
as they
were, consistent with the licensee's
commitments
in this regard.
1.5
Technical
S ecification
The Generic letter
recommended
that Technical specifications that
restrict or limit the safety benefit of the actions identified in the
GL should
be identified and appropriate
changes
should
be submitted to
the
NRC.
The inspector verified that the facility has
no technical
specification
that would limit the safety benefits of the action in the GL, as stated
in'he licensee
response.
1.6
Reactor
Coolant
S
stem
erturbation
has
recommended. that the licensee
should consider
training, procedures,
and controls that reasonably
avoid perturbing
activities
when
RCS inventory is low and decay heat is high.
The inspector verified that the Licensee
has
implemented
procedures
and administrative controls to preclude operations that would lead to
perturbations
in the
RCS.
Operating
procedure
No. 0-2.3. 1,
for'raining
down in the notice after
step 5.0, establishes
administrative
control for requesting,
reviewing, working or authorizing activities
that
may affect or perturb the
RCS water level while the
RCS is
operating at reduced
inventory mid-loop conditions.
This procedure
also provides
a detail list of- components
which may perturb the
level during the
low loop level operation.
The inspector
concluded that the procedures
and administrative controls
established
to minimize reactor perturbation during mid-loop operation
were consistent with the programmed
enhancements
described
in GL 88-17.
2.0
A/
C Involvement in Mid Loo
0
eration'he
inspector
reviewed the licensee's
Quality Assurance
Surveillance
Report
91-024.
The purpose of the licensee's
audit was to provide
an assessment
of control
room activities for entering,
maintaining
and leaving the reduced
inventory condition.
The highlights of this assessment
were:
Operator
performance
during low level operation
was appropriate
and timely.
The responses
to unanticipated
were good.
Conditions
such
as
Board modification
The inspector
concluded
addressed
the technical
3.0
Conclusion
the adverse
weather alert
and the ongoing Control
were addressed
in a safe
manner.
that the licensee's
QA surveillance effectively
adequacy
of the mid-loop operation.
Management
support
was evident for the activities related to mid loop
operation..
The evaluations
were detailed
and technically sound.
The line
supervisor
and cognizant engineer
were knowledgeable
in the safety concerns
and the regulatory positions discussed
in GL 88-17.
The engineer
recommendations
were adequately
implemented.
QA surveillance for mid-loop
operation
was technical
and thorough.
'
4.0
Plant Tours
The inspector
made tours of the Ginna plant including the control
room,
charging
and
RHR pump
rooms to observe
any work in progress,
housekeeping
and cleanliness.
No unacceptable
conditions were found.
5.0
Exit Interview
On August 16,
1991,
an exit interview was conducted with the licensee's
senior site representative
(denoted
in attachment)
to summarize
the
observations
and conclusions
of this inspection.
ATTACHMENT
Mid Loo
0 eration
Ins ection
Person
contacted
Rochester
Gas
and Electric
Com
an
W. Backus
J. Bitter
R. Eliaz
T.
Kaza
F. Maciuska
R. Marchionda
T. Marlow
D. Markowski
J. St. Martin
C. Rioch
T. Schuler
Operation
Engineer
Electrical
Engineer
Nuclear Safety 5 Licensing Engineer
'omputer System, Engineer
Supervisor-Licensee
Training
Superintendant
Support Svc.
Superintendant
Ginna production
Mechanical
Engineer
Corrective Action-Engipeer
Liaison Engineer
Operation
Engineer
W. Stiewe
J.
Widay
Electrical
gC Engineer
Plant Manager
U.S. Nuclear
Re viator
Commission
T. Moslak
Senior
Resident
Inspector