ML17263A303
| ML17263A303 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/24/1993 |
| From: | Andrea Johnson Office of Nuclear Reactor Regulation |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TAC-M84947, NUDOCS 9306030377 | |
| Download: ML17263A303 (9) | |
See also: IR 05000244/1991201
Text
Docket No. 50-244
Dr. Robert
C. Hecredy
Vice President,
Nuclear Production
Rochester
Gas
& Electric Corporation
89 East
Avenue
Rochester,
New York 14649
Dear Dr. Hecredy:
May 24,
199=
DISTRIBUTION:
Docket File
NRC
Il Local
PDI-3 Reading
SVarga
JCalvo
WButler
TClark
AJohnson
ACRS(10)
JLinville,RI
CMcCracken
JTatum
WLazarus,RI
SUBJECT:
REQUEST
FOR ADDITIONAL INFORMATION RE:
R.
E.
GINNA NUCLEAR POWER
PLANT SERVICE WATER SYSTEM ROCHESTER
GAS AND ELECTRIC
CORPORATION'S
RESPONSES
TO DEFICIENCIES IDENTIFIED IN NRC
REPORT 50-244/91-201,
JANUARY 30,
1992
(TAC NO. H84947)
Rochester
Gas
and Electric Corporation
(RGEE)
has
responded
to the concerns
expressed
in the subject
NRC inspection report
by letters
dated January
31,
April 6, April 9,
May 4, July 2, September
1,
and September
30,
1992.
RG8 E
also submitted information pertaining to heat
exchanger
performance testing in
a letter dated
June
1,
1992.
We have completed
our review of the responses
cited above
and find that additional
information is needed
to complete our
review.
While the staff agrees
with your position pertaining to the check
valve single failure issue
and licensing basis,
the staff is asking for
additional
information to better
understand
the bases for your determination
that these
issues
are not safety significant.
Please
respond
to the enclosed
request for additional
information within
60 days of receipt of this letter.
This requirement affects fewer than ten respondents
and, therefore,
is not
subject to Office of Management
and Budget review under
P.L.96-511.
Sincerely,
/s/
Allen R. Johnson,
Project
Manager
Project Directorate I-3
Division of Reactor Projects - I/II
Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional
Information
cc w/enclosure:
See next
pa
e
OFFICE
LA:
H: PD I-3
D: PDI-3
NAME
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AJohnson:mw
WButler
OFFICIAL RECORD
COPY
FILENAME: A:iGIH84947. RAI
9306030377
930524
ADOCK 05000244
8
Dr. Robert
C. Mecredy
R.E.
Ginna Nuclear
Power Plant
CC:
Thomas
A. Moslak, Senior Resident
Inspector
R.E. Ginna Plant
U.S. Nuclear Regulatory
Commission
1503
Lake Road
Ontario,
14519
Regional Administrator,
Region I
U.S. Nuclear Regulatory
Commission
475 Allendale
Road
King of Prussia,
19406
Hs.
Donna
Ross
Division of Policy Analysis
5 Planning
New York State
Energy Office
Agency Building 2
Empire State
Plaza
Albany,
12223
Charlie Donaldson,
Esq.
Assistant Attorney General
New York Department of Law
120 Broadway
10271
- Nicholas
S.
Reynolds
Winston
8 Strawn
1400
L St.
N.W.
20005-3502
Hs.
Thelma
Wideman
Director,
Wayne County Emergency
Management Office
Wayne County Emergency Operations
Center
7370 Route
31
Lyons,
14489
Hs. Mary Louise Meisenzahl
Administrator,
Monroe County
Office of Emergency
Preparedness
ill West Fall
Road,
Room ll
Rochester,
14620
RE(VEST
FOR ADDITIONAL INFORMATION RE:
GINNA SERVICE
WATER SYSTEM ISSUES
Enclosure
The inspection report for the Ginna service water system operational
performance
inspection
(SWSOPI),
NRC Report 50-244/91-201,
dated January
30,
1992, raised
a number of issues
that were identified for further staff review.
One issue of particular concern is related to the staff's review of the Ginna
service water system discussed
in the safety evaluation for Systematic
Evaluation
Program
(SEP) Topic IX-3,
."Station Service
and Cooling Water
Systems,"
dated
November 3,
1981.
During that review, it was the staff's
understanding 'that the Ginna service water system
(SWS)
was designed
and
configured
such that:
a) non-safety-related
loads would be automatically
isolated
upon initiation of a reactor
safeguards
actuation signal,
b)
potential single failure problems
were overcome since
each safety-related
load
had
a redundant
counterpart
cooled
by the other
and c) the service
water supply loop is normally split into two separate
with no sizeable
cross connections.
The staff's understanding
with regard to these
features
was
based
on the system description that was contained
in the Final Safety
Analysis Report
(FSAR) for the facility which was in error.
As a result of the
SWSOPI,
the staff learned that the Ginna
SWS was in fact
normally operated
in
a cross-connected
manner
and non-safety-related
loads
were not automatically isolated
upon initiation of a reactor safeguards
actuation signal.
Other issues
related to the
SWS design basis that were identified for staff
review during the
SWSOPI included:
(1) application of the hydraulic model to
the
SWS;
(2) assessment
and resolution of the
pump discharge
single failure vulnerability; (3) basis/criteria for the
SWS low pressure
setpoint;
(4) identification and resolution of potential single failure
vulnerabilities that
may exist due to the cross
connected
configuration of the
SWS loop;
(5) assessment
and resolution of preoperational
test disc} epancies;
and
(6) assessment/validation
of the minimum number of SWS
pumps required to
be operable to mitigate the consequences
of an accident.
Rochester
Gas
and Electric Corporation
(the licensee)
h'as responded
to the
concerns
expressed
in the
NRC ihspection report by letters dated January
31,
April 6, April 9,
May 4, July 2, September
1,
and September
30,
1992.
The
licensee
also submitted information pertaining to heat
exchanger
performance
testing in a letter dated
June
1,
1992, which is also relevant.
We have
completed
our review of the
above cited submittals
and find that
some
additional
information is needed
to facilitate completion of the staff's
review of the Ginna
SWS.
Also, while the staff agrees with the licensee's
position pertaining to the check valve single failure and licensing basis
issue,
additional
information has
been requested
to better understand
the
bases
for the licensee's
determination that these
issues
are not safety
significant.
The licensee
is requested
to provide complete
responses
to each of the staff's
questions,
including assumptions
and justifications, in order to expedite
review of the Ginna
SWSOPI issues.
2.
3.
In order for the staff to judge whether
one service water
pump can
provide adequate
flow for the system configuration,
the specific
flow requirements for each safety-related
component
must
be identified.
In making this determination,
the worst-case
conditions must
be
assumed
for each
component
(not necessarily
the design basis
accident
conditions).
Describe specifically how the flow requirements
were
determined for each
component,
including assumptions
that were made.
Also, indicate to what extent vendor concurrences
have
been obtained for
judgements
and evaluations
pertaining to equipment
performance
capabilities.
Also, given the flow requirements
established
for the
Component
Cooling Water
(CCW) heat exchangers,
discuss
any changes
in
the ability and times required to shutdown
and cooldown the plant.
Based
on recent test results
using the most limiting service water
pump
(corrected for worst-case
lake level, water temperature,
instrument
inaccuracies,
and
pump degradation
allowed unde} the inservice testing
program),
annotate
on
a simplified diagram of the
SWS the
pump dischar'ge
pressure,
flow rates
through
each safety-related
component,
flow rates
through non-safety-related
supply lines
(as applicable),
and the system
configuration/alignment that is necessary
to satisfy the flow
requirements
for safety-related
equipment.
Also, indicate the
electrical division (A, B, or non-safety-related)
where
each
component
receives
power.
Describe
how this system configuration/alignment will
be achieved during
an event
assuming off-site power is available
and
also for the condition where off-site power is not available,
and
identify specific time constraints
that must
be met in order to satisfy
all accident
analyses
assumptions
for heat removal.
Provide
conservative
estimates
of how long it will take to complete
necessary
actions for each
case (i.e., off-site power available
and off-site power
not available).,
The basis for these estimations
should
be provided,
which may include reference
to previous plant experience,
plant
simulations
and exercise walk-throughs.
Starting with event initiation, provide bounding
assessments
(i.e.,
normal
system alignment supplying service water to safety-related
components,
including the
CCW heat exchangers
and the spent fuel pool
(SFP)
heat exchangers,
and to non-safety-related
components
with off-
site
power available
and also for the condition where off-site power is
not available) of SWS flow rates
over time for single
pump operation.
The basis for this assessment
should
be provided,
which may include
reference
to previous plant experience,
plant simulations,
and exercise
walk-throughs.
Provide confirmation
and vendor concurrence
(as
appropriate)
that
pump performance will not be jeopardized
by these
postulated
SWS flow conditions.
Describe periodic surveillances,
tests,
inspections,
and training that
will be performed to ensure that the
SWS will be capable of satisfying
its function during an event
and to ensure that all assumptions
are
valid.
Confirm that all existing accident
analyses,
including safe
shutdown/Appendix
R analyses,
have
been
updated
pursuant to
requirements
to account for single service water
pump operation,
assuming worst-case
conditions of lake level, water temperature,
instrument inaccuracies
and
pump degradation.
Similarly, confirm that
the current submittal
on boron dilution requirements
only credits
a
single service water
pump
(assuming worst-case
conditions)
in its
supporting accident
analyses.
Also, discuss specifically the effect of
having only one service water
pump operable
has
on the ability to
mitigate
tube rupture event.
Provide
a detailed description of the assumptions
and inputs
used in the
previous/original limiting containment
analysis
versus
assumptions
and
inputs
used in the current (single service water
pump) containment
analysis.
Provide detailed explanations for any differences of inputs
and assumptions
between
the two analyses.
Describe what the service
water flow rates
are for all components
(safety-related
and non-safety-
related)
over time and describe
in detail
how the containment
heat
removal rates
were determined
(including justification for all
assumptions)
for each of these
analyses.
Confirm that for a steam line break event at hot zero power with off-
site
power available, that one service water
pump is sufficient to
mitigate the event.
Similar to item 6 (above),
provide
a detailed
explanation (justification) of any differences
in assumptions
and inputs
used in the previous/original limiting analysis
versus
assumptions
and
inputs
used in the current (single service water
pump) analysis.
Also,
describe
what the service water flow rates
are for all components
(safety-related
and non-safety-related)
over time and describe
in detail
how the containment
heat
removal rates
were determined
(including
justification for all assumptions)
for each of these
analyses.
Describe actions that have
been taken to validate the
SWS flow model for
various configurations of split and cross-connected
system operation
that may be encountered
and, in particular,
discuss validation of the
flow model with respect
to single
pump operation for various
system
operating configurations.
Provide the results of the check valve single failure analysis
using the
validated flow model, including
pressure
and
SWS flow rates
through the failed check valve and all other components
assuming that
the two most limiting service water
pumps
are operating.
Justify all
assumptions.
Also, describe
the results of the Ginna probabilistic risk
assessment
(PRA) with regard to failure of a
pump discharge
check
valve, including justification for check valve failure frequencies
that
were
assumed.
Describe, periodic maintenance,
surveillance,
test
and inspection
activities that are performed that provide assurance
that
a
pump
discharge
check valve will function properly.
Also, describe
how the
failure of a
pump discharge
check valve to function will be identified
and addressed
by the operators,
including
a description of training that
has
been provided
and periodic training that will be provided in the
future.
Provide confirmation that operator action is sufficient to
ensure that
a
pump discharge
check valve failure will remain
bounded
by the accident
analyses
that have
been performed.
Provide the conclusions
and recommendations
of the revised single
failure analysis.
Discuss corrective actions that are being taken to
address
vulnerabilities.
Also, confirm that the revised single failure
analysis
encompasses
all operating configurations of the
SWS that are
allowed
and include
a description of the specific configurations that,
were considered
to be applicable in this regard.
Identify which safety-related
components
do not need service water
cooling during the injection phase of an accident
and provide
justification for this position, including vendor concurrence'he
justification should include consideration for small
break loss of
coolant accident
(SBLOCA) conditions,
where
pumps
may be required to
operate
at or near shut-off head conditions.
Provide
a conservative
estimate for when service water cooling must
be restored to this
equipment
and describe
how this will be accomplished.
'The basis for
these
estimations
and actions
should
be provided,
which may include
reference
to test data,
previous plant experience,
plant simulations
and
exercise walk-throughs.
Similar to item 12 (above),
provide justification and vendor concurrence
for the position that component cooling is not required during the
injection phase of an accident for the
RHR pump mechanical
seal
coolers
and bearing water jackets,
the safety injection
pump mechanical
seal
coolers,
and the containment
spray
pump mechanical
seal
coolers.
Describe
how long the emergency diesel
generator
(EDG) coolers
have
been
in service
and what periodic inspections
and preventive maintenance
activities are typically performed relative to these coolers,
including
frequencies.
Assuming the maximum decay heat load in the spent fuel pool, describe
how long service water to the
SFP heat exchanger
can
be isolated,
including supporting justification.
Also, describe alternative
measures
that can
be taken that are included in emergency
procedures
for cooling
the spent fuel pool during
an event.
Define the
low pressure
setpoints for all system
configurations
and discuss
the basis for the setpoints that have
been
established.
Confirm that, for various
modes of SWS operation,
with valves
4760
and
4669 in the open position (diesel generator
service water cross-connect
valves),
including both split and cross-connected
header configuration,
that
a passive failure will not jeopardize operability of both diesel
generators.
Describe the specific
SWS configurations that were
considered
in this regard.
18.
Given the system constraints
and limitations that must
be satisfied
during
an event with only one service water
pump available, identify any
changes
that should
be made to the existing Technical Specification
LCOs
and surveillance
requirements.
For example,
the current Technical
Specifications
only require
one loop header to be operable
and does
not
specifically require that
pump
be operable
in each loop
which is not consistent
with the
FSAR description that credits
both loop headers
as being available for redundancy of cooling
capability.
Also, requirements for split vs. cross-connected
loop
header operation
are not stipulated.
19.
Supplemental
response
to
GL 89-13 dated
June
1,
1992, only lists one
heat exchanger
as "critical."
This is not consistent
with the
description of,critical heat loads which lists two heat exchangers.
Explain.
20.
Identify specifically what service water flow rates
are required to be
established
through
each
heat
exchanger
being tested during heat
exchanger
performance testing that is periodically performed.
Also,
state
the worst-case
fouling factors that are
assumed for each of these
heat
exchangers
and the bases for the values
being credited.
gA~ REC0
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Docket No. 50-244
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555-0$ 1
May 24,
1993
Dr. Robert
C. Mecredy
Vice President,
Nuclear Production
Rochester
Gas
5 Electric Corporation
89 East Avenue
Rochester,
New York 14649
Dear Dr. Hecredy:
SUBJECT:
REQUEST
FOR ADDITIONAL INFORMATION RE:
R.
E.
GINNA NUCLEAR POWER
PLANT SERVICE WATER SYSTEM ROCHESTER
GAS AND ELECTRIC
CORPORATION'S
RESPONSES
TO DEFICIENCIES IDENTIFIED IN NRC
REPORT 50-244/91-201,
JANUARY 30,
1992
Rochester
Gas
and Electric Corporation
(RGEE)
has
responded
to the concerns
expressed
in the subject
NRC inspection report by letters dated January
31,
April 6, April 9,
Hay 4, July 2,
September
1,
and September
30,
1992.
RG&E
also submitted information pertaining to heat
exchanger
performance testing in
a letter dated
June
1,
1992.
We have completed
our review of the responses
cited above
and find that additional
information is needed to complete our
review.
While the staff agrees
with your position pertaining to the check
valve single failure issue
and licensing basis,
the staff is asking for
additional
information to better understand
the bases for your determination
that these
issues
are not safety significant.
Please
respond to the enclosed
request for additional information within
60 days of receipt of this letter.
This requirement affects fewer than ten respondents
and, therefore,
is not
subject to Office of Management
and Budget review under P.L.96-511.
Sincerely,
Enclosure:
Request for Additional
Information
cc w/enclosure:
See next page
en
R. Jo nson
Project Manager
Project Directo
te I-3
'vision of R
tor Projects - I/II
Of >c
uclear Reactor Regulation
~
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