ML13224A246
| ML13224A246 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/02/2013 |
| From: | Gebbie J P Indiana Michigan Power Co |
| To: | Document Control Desk, NRC/RGN-III |
| References | |
| AEP-NRC-2013-53 IR-13-010 | |
| Download: ML13224A246 (25) | |
See also: IR 05000315/2013010
Text
INDIANAMICHIGANPOWERA unit of American
Electric
PowerAugust 2, 2013Docket Nos.: 50-31550-316Indiana Michigan
PowerCook Nuclear PlantOne Cook PlaceBridgman,
MI 49106Indiana Michigan
Power.com
AEP-NRC-2013-53
10 CFR 2.201U.S. Nuclear Regulatory
Commission
Attn: Document
Control DeskWashington,
DC, 20555-0001
Donald C. Cook Nuclear Plant Units 1 and 2Response
to the Non-Cited
Violations
Resulting
from Component
Design Bases Inspection
References:
1. Letter from W. Hodge, Indiana Michigan
Power Company (I&M), to C. Tilton, U.S. NuclearRegulatory
Commission
(NRC), "D. C. Cook CDBI Response
to Question
2012-CDBI-298,"
dated November
15, 2012, (ADAMS Accession
No. ML12320A544).
2. Letter from K. O'Brien,
NRC, to S. Bahadur,
NRC, "Task Interface
Agreement
-Licensing
Basis for Donald C. Cook Nuclear Power Plant, Units 1 and 2, During a Steam Generator
Tube Rupture Event Coincident
with a Loss of Offsite Power (TIA 2012-11),"
datedDecember
7, 2012, (ADAMS Accession
No. ML13011A382).
3. Letter from A. M. Stone, NRC, to L. J. Weber, I&M, "D. C. Cook Nuclear Power Plant, Units1 and 2, Component
Design Bases Inspection
dated January 11, 2013 (ADAMS Accession
No. ML13011A401).
4. Letter from G. Shear, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Power Plant,Units 1 and 2, Component
Design Bases Inspection
dated July 8, 2013, (ADAMS Accession
No. ML13189A243).
This letter provides
Power Company's
(l&M's),Nuclear Plant (CNP) Units 1 and 2, response
contesting
thedocumented
by Reference
4, Component
Design Bases05000315/2013010;
licensee
for Donald C. CookNon-Cited
Violations
(NCVs)Inspection
(CDBI) ReportIn Reference
1, I&M identified
docketed
correspondence
supporting
I&M's understanding
of CNP'slicensing
basis to assume only a single-unit
loss of offsite power (LOOP) coincident
with a designbasis Steam Generator
Tube Rupture (SGTR) accident.
In Reference
2, the Nuclear Regulatory
Commission
(NRC) Region III Staff issued a Task Interface
Agreement
Report documenting
U.S. Nuclear Regulatory
Commission
AEP-NRC-2013-53
Page 2the results of its consultation
with the NRC Office of Nuclear Reactor Regulation
regarding
the NRCStaff's understanding
of CNP's licensing
basis to assume a multi-unit
LOOP as an initial condition
of a design basis SGTR accident.
In Reference
3, the NRC Staff notified
I&M that two potential
findings
relating
to the operability
power operated
relief valves (SG PORVs)during a design basis SGTR accident
identified
by the NRC Staff during a CDBI performed
at CNPbetween July 23, 2012, and December
31, 2012, would remain unresolved
items (URIs) pendingthe NRC Staffs resolution
of questions
regarding
the scope of a LOOP assumed within CNP'sSGTR accident
analysis.
In Reference
4, the NRC Staff resolved
the URIs issued by Reference
3and issued NCVs of CNP Technical
Specifications
5.4.1 (prescribing
emergency
operating
procedures
(EOPs) to mitigate
the consequences
of a design basis SGTR accident)
and 3.7.4(governing
the operability
4 states that I&M had violated
Technical Specification 5.4.1 because CNP EOPs could not ensure that personnel
would be able to operateSG PORVs as required
by CNP's licensing
basis during an SGTR accident
accompanied
by aLOOP affecting
both units at CNP. Reference
4 also states that I&M had violated
Technical Specification 3.7.4 because it had failed on several occasions
to declare the SG PORVsunavailable
after taking a control air compressor
out of service for maintenance.
Reference
4characterized
the NCVs as representing
a more-than-minor
performance
deficiency
with cross-cutting aspects.I&M contests
the NCVs identified
in Reference
4 because those NCVs lack technical
justification
and are inconsistent
with NRC regulations
and guidance.
Specific
bases for I&M's contest of theNCVs include the following:
- The NCVs are based on an erroneous
understanding
of CNP's licensing
basis. Contrary
basis assumptions
regarding
the initial conditions
for a SGTRaccident
have never considered
a coincident
LOOP involving
both units. Further,
the NRCStaff's understanding
of CNP's licensing
basis underlying
the NCVs does not acknowledge
docketed
correspondence
between I&M and NRC Staff supporting
I&M's position,
does notrepresent
a fair reading of CNP's Updated Final Safety Analysis
Report (UFSAR),
and isinconsistent
with the NRC's current regulatory
position
regarding
the loss of offsite power tonon-safety
related auxiliary
systems at other multi-unit
sites.* The NRC Staff has not demonstrated
that I&M's understanding
of CNP's licensing
basis failsto provide adequate
protection
of public health and safety from either design basis events orbeyond-design
basis external
events. Further,
the NRC Staff has not demonstrated
that itsown position
would provide a meaningful
improvement
in the protection
of public health andsafety.* The NRC Staff's determination
that the NCVs represent
a more-than-minor
performance
deficiency
with cross-cutting
aspects is based on an erroneous
understanding
of the scopeof a LOOP assumed within CNP's design basis SGTR accident
analysis,
is inconsistent
withthe NRC Staffs statements
in docketed
correspondence,
and is unrepresentative
of presentlicensee
performance.
Enclosure
1 to this letter contains
an affirmation
statement.
Enclosure
2 to this letter lays out indetail the regulatory
and factual support for I&M's response
contesting
the NCVs.
U.S. Nuclear Regulatory
Commission
AEP-NRC-2013-53
Page 3Regardless
of the outcome of I&M's contest of the NCVs, I&M will continue
to evaluate
cost-effective
measures
for the improvement
of safety margins against SGTR accidents.
Following
the 2012 CDBI, I&M revised CNP procedures
and implemented
plant modifications
toprovide additional
defense-in-depth
and improved
safety margins during an SGTR accident.
InMarch 2013, I&M completed
installation
of a plant modification
and revised CNP operating
procedures
to ensure that backup nitrogen
tanks are immediately
and automatically
available
duringan SGTR accident
for operation
of SG PORVs without the need for manual valve manipulation
outside the control room. I&M has also revised CNP Work Control processes
to provide additional
defense-in-depth
from a loss of control air pressure
by restricting
removal for maintenance
of theoperating
unit's control air compressor
when the opposite
unit is shutdown
and the shutdown
unit'splant air compressor
is aligned to preferred
offsite power.This letter contains
no new or revised commitments.
If you have any questions,
please contactMr. Michael K. Scarpello,
Regulatory
Affairs Manager,
at (269) 466-2649.
Sincerely,
Joel P. GebbieSite Vice President
DMB/kmhEnclosures:
1. Affirmation
Power Company's
Response
to "Donald C. Cook Nuclear Power Plant,Units 1 and 2, Component
Design Bases Inspection
dated July 8,2013c: C. A. Casto, NRC Region IIIJ.T. King, MPSCS. M. Krawec, AEP Ft. Wayne, w/o enclosure
E. Leeds, NRC NRRMDEQ-RMD/RPS
NRC Resident
Inspector
A. M. Stone, NRC Region IIIC. Tilton, NRC Region IIIT. J. Wengert,
NRC Washington,
DCR.P. Zimmerman,
NRC Washington,
ENCLOSURE
I TO AEP-NRC-2013-53
AFFI RMATIONI, Joel P. Gebbie, being duly sworn, state that I am Site Vice President
PowerCompany (I&M), that I am authorized
to sign and file this request with the Nuclear Regulatory
Commission
on behalf of I&M, and that the statements
made and the matters set forth hereinpertaining
to I&M are true and correct to the best of my knowledge,
information,
Power CompanyJoel P. GebbieSite Vice President
SWORN TO AND SUBSCRIBED
BEFORE METHIS____
DAY OF ,A)ws 2013My Commission
Expires ( I 2 IO{
ENCLOSURE
2 TO AEP-NRC-2013-53
Power Company's
Response
to "Donald C. Cook Nuclear PowerPlant, Units 1 and 2, Component
Design Bases Inspection
dated July 8, 20131. Introduction
The Non-Cited
Violations
(NCVs) within the Nuclear Regulatory
Commission
(NRC) StaffsJuly 8, 2013, letter (Reference
Power Company (I&M) are based on anerroneous
understanding
of the licensing
basis of Donald C. Cook Nuclear Plant (CNP). TheNRC Staff's position
that CNP's design basis Steam Generator
Tube Rupture (SGTR) accidentassumes a coincident
loss of offsite power (LOOP) that can involve both units at CNP isinconsistent
with pertinent,
docketed
correspondence
between the NRC Staff and I&M. Further,the NRC Staff's position
is unsupported
by a fair reading of CNP's Updated Final SafetyAnalysis
Report (UFSAR),
and is likewise
inconsistent
with relevant
historical
and currentregulatory
positions
of the NRC. Additionally,
the NRC Staff has not demonstrated
that I&M'sunderstanding
of CNP's licensing
basis fails to provide adequate
protection
of public health andsafety from either design basis events or beyond-design
basis external
events. Lastly, the NRCStaff's determination
that the NCVs represent
a more-than-minor
performance
deficiency
withcross-cutting
aspects relies on an erroneous
understanding
of the scope of a LOOP assumedwithin CNP's design basis SGTR accident
analysis,
is inconsistent
with the NRC Staff'sstatements
in docketed
correspondence,
and is unrepresentative
of present licenseeperformance.
Documents
referenced
herein are listed as references
at the end of this Enclosure.
2. History of the Non-Cited
Violations
The NCVs contested
by I&M result from findings
by the NRC Staff during the Component
Design Bases Inspection
(CDBI) conducted
at CNP between July 23, 2012, andDecember
31, 2012. As described
in Reference
2, the CDBI entailed
a review of licensing
basisdocumentation
and drawings
of the CNP compressed
air system to verify that support functions
provided
to the steam generator
power operated
relief valves (SG PORVs) were consistent
withCNP's licensing
basis requirements
for SGTR accidents.
As stated in Reference
2, the NRC Staff contended
during the CDBI that CNP was not inconformance
with Technical
Specifications
5.4.1 (prescribing
emergency
operating
procedures
(EOPs) to mitigate
the consequences
of a design basis SGTR accident)
and 3.7.4 (governing
the operability
of SG PORVs). Based on its belief that CNP's licensing
basis assumptions
for aSGTR accident
included
a coincident
LOOP affecting
both units at CNP, the NRC Staffreasoned
that the only available
source of control air pressure
during the most limiting
SGTRaccident
would be the affected
unit's dedicated
control air compressor
(CAC) receiving
powerfrom one of the two emergency
diesel generators
(EDG). However,
if the affected
unit's CACwere unavailable
as a result of emergent
or planned maintenance,
then the NRC Staff reasonedthat control air pressure
would be unavailable
to operate the affected
CNP operating
records,
the NRC Staff identified
several occasions
in which CACs at
Enclosure
2 to AEP-NRC-2013-53
Page 2CNP would have been unavailable
due to maintenance,
but I&M had not declared
the SGPORVs inoperable.
I&M disagreed
with the NRC Staff's characterization
of CNP's licensing
basis assumptions
for aSGTR event. Noting that the CNP licensing
basis for an SGTR event did not consider
acoincident
multi-unit
LOOP, I&M contended
that the NRC Staffs finding was based on a beyonddesign basis accident
scenario.
The NRC Staff requested
assistance
from the NRC Office ofNuclear Reactor Regulation
(NRR) in resolving
the disagreement
regarding
CNP's licensing
basis assumptions.
On November
15, 2012, I&M submitted
Reference
3 to NRC Staff,containing
information
identifying
the technical
and regulatory
bases supporting
I&M's positionand providing
docketed
correspondence.
Reference
3 in particular
identified
a SafetyEvaluation
Report (SER, Reference
4) dated October 24, 2001, explicitly
discussing
CNP'sassumptions
for SGTR accident
initial conditions,
and revealing
the NRC Staff's evaluation
andendorsement
of I&M's understanding
of the CNP licensing
basis assumptions
for an SGTRaccident.
On December
7, 2012, NRC Region III Staff issued Reference
5 after consulting
with NRR,contradicting
I&M's understanding
of CNP's licensing
basis assumptions
for SGTR accidents.
Reference
5 cited only three passages
6) in support of itsposition,
interpreting
a handful of references
to the terms "LOOP" and "station"
in descriptions
ofCNP electrical
systems to mean that CNP's licensing
basis assumed a LOOP would affect bothunits at CNP in an SGTR accident.
Reference
5 suggests
that it did not examine the technical
and regulatory
bases and docketed
correspondence
supporting
a contrary
position
referenced
within Reference
3 submitted
by I&M.On January 11, 2013, the NRC Staff issued Reference
2, identifying
the CDBI findings
at issueas unresolved
items (URIs) pending submission
of additional
information
from I&M regarding
CNP's licensing
basis assumptions
for SGTR accidents.
Reference
2 repeated
Reference
5'sconclusions
regarding
CNP's licensing
basis assumptions
for SGTR accidents
without furtherexplanation
or analysis;
further,
Reference
2 again did not address the technical
and regulatory
bases and docketed
correspondence
identified
in Reference
3 forwarded
by I&M. OnFebruary
8, 2013, I&M provided
Reference
7 to the NRC Staff, refuting
Reference
5'sinterpretation
additional
detail regarding
the technical
andregulatory
bases supporting
I&M's understanding
of the CNP licensing
basis assumptions
for anSGTR accident.
During a May 20, 2013, technical
debrief of the CDBI findings,
the NRC Staffrepeated
its understanding
of the scope of the LOOP assumed within SGTR's accident
analysis,
again without addressing
the technical
and regulatory
bases and docketed
correspondence
supporting
I&M's position.
In a re-exit teleconference
for the URIs conducted
on May 24, 2013,the NRC Staff informed
I&M that the NRC Staff planned to issue an NCV for violation
ofTechnical
Specification
3.7.4 requirements
regarding
the operability
of SG PORVs.On July 8, 2013, the NRC Staff issued Reference
1. In Reference
1, the NRC Staff identified
Specifications
5.4.1 (prescribing
EOPs to mitigate
the consequences
ofa design basis SGTR accident)
and 3.7.4 (governing
the operability
1 states that I&M had violated
Technical
Specification
5.4.1 because CNP EOPs could notensure that personnel
would be able to operate SG PORVs as required
by CNP's licensing
basis during an SGTR accident
accompanied
by a LOOP affecting
both units at CNP.Reference
1 also states that I&M had violated
Technical
Specification
3.7.4 because it had
Enclosure
2 to AEP-NRC-2013-53
Page 3failed on several occasions
to declare the SG PORVs unavailable
after taking a CAC out ofservice for maintenance.
Reference
1 characterized
the NCVs as representing
a more-than-minor,
cross-cutting
performance
deficiency
involving
areas of human performance,
the component
ofdecisionmaking,
and the aspect of conservative
assumptions
because I&M had incorrectly
assumed that control air pressure
to the SG PORVs of a unit experiencing
an SGTR accidentaccompanied
by a LOOP would remain available
from the unaffected
unit's plant air compressor
(PAC).Reference
1 also attempted
to refute I&M's explanation
within Reference
7 of its understanding
of CNP's licensing
basis assumptions
for SGTR accidents.
Acknowledging
I&M's position
thatCNP's licensing
basis did not assume a single failure of a non-safety-related
component
(inparticular,
the unaffected
unit's PAC), during an SGTR event, Reference
1 contends
that I&Mhad nevertheless
failed to demonstrate
that control air would reasonably
be available
during anSGTR event accompanied
by a multi-unit
LOOP. Similarly,
Reference
1 asserts that even if theunaffected
unit's PAC would be available
during a design basis SGTR accident,
I&M had failedto identify
that assumption
within its SGTR accident
analysis,
and the NRC Staff had neverexplicitly
approved
that assumption.
Further,
Reference
1 endorsed
Reference
5'sinterpretation
of the UFSAR's use of the term LOOP to refer to multi-unit
events, adding that theabsence of CNP operating
procedures
preventing
alignment
of the same offsite power sourcesto both units made a multi-unit
LOOP a credible
event within CNP's licensing
basis.3. Overview
of Pertinent
CNP Systems and Operatinq
Procedures
Power Operated
Relief ValvesIn accordance
with Reference
6 (at Sections
10.2.2 and 14.2.4),
the SG PORVs preventoverpressure
conditions
in the steam generators
by releasing
secondary
system steam toatmosphere
following
a loss of condenser
vacuum. The SG PORVs form part of the mainsteam system pressure
boundary,
and thus are safety-related
equipment
for main steam systempressure
retention.
CNP operating
procedures
prescribe
operator
actions in the event of a SGTR accident.
CNPoperating
procedures
using motive force provided
by controlair supplied
by either the compressed
air system shared between the two units, control airpressure
supplied
by a unit-specific
CAC, or installed
backup nitrogen
tanks that can be alignedto the SG PORVs. In March 2013, I&M completed
installation
of a plant modification
andrevised its operating
procedures
to ensure that the backup nitrogen
tanks are immediately
andautomatically
available
during an SGTR accident
without the need for manual valvemanipulation
outside the control room.b. CNP Compressed
Air SystemSection 9.8.2 of Reference
6 describes
the control air provided
by CNP's compressed
airsystem as the ordinary
source of motive force for operation
of SG PORVs for both units at CNP.Per Reference
6, Section 1.3.9.h,
CNP's compressed
air system is a single system sharedbetween both units at CNP. Each unit at CNP contains
one CAC capable of providing
control
Enclosure
2 to AEP-NRC-2013-53
Page 4air only within that unit, as well as a PAC capable of providing
control air to both units via ashared header. Both units share a single backup air compressor
capable of providing
control airto loads within either unit.During normal operations,
control air pressure
for operating
both units' SG PORVs is providedby one of the two PACs. Low pressure
in the shared plant compressed
air header will result inthe automatic
start and loading of the other unit's PAC. Low control air header pressure
in oneof the unit-specific
control air headers will cause that unit's CAC to start.During normal operations,
the operating
PAC receives
power from its unit's auxiliary
transformers,
which are in turn powered by that unit's main generator
or preferred
offsite powertransformers.
The CAC associated
with each unit at CNP can be powered by either offsitepower source in normal operations,
but can only receive power from its unit's CD EDG afteroffsite power has been lost to that unit. The CACs and PACs are both non-safety
relatedequipment
governed
by the Maintenance
Rule at 10 CFR 50.65.CNP Work Control processes
impose a series of administrative
controls
to maximize
availability
of control air pressure
when a CAC or PAC is taken out of service for maintenance:
- In the event a CAC is taken out of service for maintenance,
bothPACs and the installed
backup nitrogen
tanks must be guarded;
and* In the event that a PAC is taken out of service,
the following
equipment
is guarded:
(1) the opposite
unit's PAC, (2) both CACs, (3)the opposite
unit's CD EDG, and (4) the backup air compressor.
Following
the 2012 CDBI, I&M revised CNP Work Control processes
to provide additional
defense-in-depth
from a loss of control air pressure
by restricting
removal for maintenance
ofthe operating
unit's CAC when the opposite
unit is shutdown
and the shutdown
unit's PAC isaligned to preferred
offsite power.4. Regulatory
Basis for the Assumption
of Only a Single-Unit
LOOP within CNP's SGTRAccident
Analysisa. CNP's Licensing
Basis Has from the Beginning
Assumed that an SGTR AccidentWould Involve a Coincident,
Single-Unit
LOOPCNP's original
licensing
basis explicitly
assumed that SG PORVs would remain available
throughout
an SGTR accident.
As described
in the Preliminary
Safety Analysis
Report (PSAR,Reference
9) for Units 1 and 2 submitted
on December
18, 1967, and repeated
in Sections14.2.4 and 14.2.7 of the FSAR for Units 1 and 2 dated February
2, 1971 (Reference
10), CNP'soriginal
licensing
basis evaluated
the radiological
consequences
of an SGTR accident
byconservatively
estimating
the mass release of radioactivity
to the environment
over the30-minute
time span between SGTR accident
initiation
and subsequent
termination
of primaryto secondary
mass transfer
from the completion
of mitigation
measures
taken by operators.
I&M's analytical
assumption
of 30 minutes'
mass release before termination
of the event wasconsidered
inherently
conservative
because it neglected
the reduction
in mass flow that wouldoccur during this same time period.
Enclosure
2 to AEP-NRC-2013-53
Page 5Inherent
in that postulated
30-minute
mass release was an assumption
of the success ofoperator
actions such as the operation
the event. Section 14.2.4 ofReference
10 in several places explicitly
credited
the availability
of SG PORVs during a designbasis SGTR regardless
of conditions.
Reference
10's evaluation
of SGTR accidents
omits any mention of the possibility
thatcompressed
air system components
could be unavailable
as a result of a single failure ormaintenance,
as it prefaced
its elaboration
of the sequence
of events initiated
by an SGTRevent by stating that its analysis
had "assum[ed]
normal operation
of the various plant controlsystems ....... Reference
10 at Section 14.2.4. Further,
Reference
10 assumed that SGPORVs would remain available
regardless
of the status of offsite power, stating that when a unitwas "without
offsite power":Condenser
bypass valves will automatically
close and the steamgenerator
pressure
will rapidly increase
resulting
in steam discharge
tothe atmosphere
through the steam generator
safety valves and/or thepower operated
relief valves.Reference
10 at Section 14.2.4. Elsewhere,
Reference
10 noted that:In the event of a co-incident
station blackout,
the steam dump valveswould automatically
close to protect the condenser.
The steam generator
pressure
would rapidly increase
resulting
in steam discharge
to theatmosphere
through the steam generator
safety and/or power operatedrelief valves.Reference
10 at Section 14.2.4 (emphasis
added).I&M's assumption
available
for mitigation
of an SGTR accident
isconsistent
with the description
of the compressed
air system elsewhere
within CNP's originalFSAR. Among the design bases for CNP's compressed
air system within Reference
10 is arequirement
for continued
availability
of control air:The [compressed
air system] must provide a continuous
supply ofcompressed
air to vital systems under both normal and abnormalconditions.
Reference
10 at Section 9.8.2 (emphasis
added). With this in mind, each of CNP's PACs weredesigned
to be "capable
of supplying
the entire demand of both plant and control-instrument
airrequirements
for both units," as the offline PAC automatically
started on low pressure
in the(shared)
plant air header. Reference
10 at Section 9.8.2.3.Although
CNP's original
FSAR accounted
for the availability
of compressed
air systemcomponents
within the opposite
plant, the staggered
construction
and licensing
of CNP Units 1and 2 resulted
in a more unit-specific
design and function
for other CNP systems.
For example,Unit l's construction
and licensing
(1974) several years before Unit 2 (1977) meant that thedesign bases of the electrical
systems for each of the two units at CNP were, as a practical
matter, unit-specific.
For example,
although
each EDG shares a fuel oil tank with an EDG in the
Enclosure
2 to AEP-NRC-2013-53
Page 6other unit, the fuel oil tank's capacity
is based on the design operational
requirements
of asingle EDG. Reference
6 at Section 8.4. Consequently,
references
within Reference
10'sSGTR accident
analysis
to a "loss of offsite power" or a "station
blackout"
referred
to an eventinvolving
only a single unit.The analysis
of a design basis SGTR accident
in the revised FSAR evaluating
Unit 2 as-built(Reference
11) used nearly identical
language
to that used within the SGTR accident
analysis
inthe original
Units 1 and 2 FSAR (Reference
10). Further,
subsequent
versions
of both units'UFSAR analyses
for SGTR accidents
retained
the CNP's original
assumptions
regarding
theavailability
of SG PORVs -and, in fact, arguably
placed even greater emphasis
on thecontinued
availability
of those components
in their SGTR accident
analysis.
In particular,
July 1997 revisions
to the UFSAR for both units were revised to better track CNP EOPsidentifying
the SG PORVs (and not the steam generator
safety valves) as the initial means ofpreventing
overpressure
after loss of offsite power:In the event of a coincident
station blackout,
the steam dump valveswould automatically
close to protect the condenser.
The steam generator
pressure
would rapidly increase,
resulting
in steam discharge
to theatmosphere
through the steam generator
power operated
relief valves(and the steam generator
safety valves if their setpoint
had beenreached).
Reference
12 at Section 14.2.4 (emphasis
added). Later UFSAR revisions
to CNP's SGTRaccident
analysis
also incorporated
the original
FSAR's language
describing
the continued
availability
of SG PORVs despite a LOOP or station blackout
virtually
unchanged.
Reference
6at Section 14.2.4. Further,
I&M's review of pertinent
docketed
correspondence
with the NRCStaff has discovered
no evidence
of a departure
from CNP's original
assumption
of a unit-specific
LOOP coincident
with an SGTR accident.
b. The NRC Staff Has Reviewed
and Endorsed
CNP's Design Basis Assumptions
forSGTR Accidents
in Docketed
Correspondence
On October 24, 2000, I&M submitted
a license amendment
request (LAR, Reference
10) torevise the methodology
used in designing
CNP EOPs during a design basis SGTR accident.
The Westinghouse
Owners Group methodology
("SGTR AnalysisMethodology
to Determine
Margin to Steam Generator
Overfill"))
that I&M proposed
to adapt foruse within its SGTR accident
analysis
incorporated
lessons learned from operational
experience,
plant simulator
studies,
and advances
in computer
modeling
techniques
to bettercharacterize
fill conditions
during an SGTR accident.
Of particular
importance
to CNP was that the LOFTTR2 computer
program used in the WCAP-10698-P-A
methodology
simulated
the effects of operator
actions on margin to steam generator
overfill
during an SGTRaccident.
By incorporating
elements
of the WCAP-10698-P-A
methodology
for the simplified
calculations
of margin to steam generator
overfill
within its original
SGTR accident
analysisassumptions,
I&M could revise CNP EOPs to assure margins to steam generator
overfill
whileremaining
within the conservative
margins to radiological
consequences
described
in its originalSGTR accident
analysis.
Enclosure
2 to AEP-NRC-2013-53
Page 7Although
the NRC had previously
accepted
for use by licensees,
the NRCStaff had to evaluate
its application
analysis.
In a series ofdocketed
correspondence
with the NRC Staff detailing
how the WCAP-10698-P-A
would beused within CNP's SGTR accident
analysis,
I&M repeatedly
emphasized
that the newmethodology
would not disturb existing
license basis assumptions
in its SGTR accidentanalysis.
Specifically,
the safety analysis
for I&M's LAR noted that:The proposed
change ...does not affect any accident
initiators
orprecursors
.... The proposed
change also does not affect the ability ofoperators
to mitigate
the consequences
of an accident.
Reference
13, Attachment
1 at Page 4 (emphasis
added). I&M repeated
this claim in the LAR'sevaluation
of significant
hazards required
by 10 CFR 50.92(c):
[T]he new methodology
does not affect equipment
malfunction
probability
.... The proposed
change does not impact the design ofaffected
plant systems,
involve a physical
alteration
to the systems,
orchange the way in which systems are currently
operated,
such thatpreviously
unanalyzed
SGTRs would not occur. The change toincorporate
the WCAP-10698-P-A
methodology
does not introduce
anynew malfunctions
....Reference
13, Attachment
2 at Pages 2-3 (emphasis
added).Subsequent
docketed
correspondence
between I&M and the NRC Staff was even more explicitin describing
the retention
of existing
license basis assumptions
for SGTR accidents.
In aJune 29, 2001, response
(Reference
14) to a May 7, 2001, letter from the NRC Staff requesting
additional
information
(RAI) regarding
how I&M intended
to use the WCAP-10698-P-A
within itsSGTR accident
analysis,
I&M emphasized
that its use of the WCAP-10698-P-A
methodology
was "limited",
and that, by-and-large,
"CNP's present methodology
would be retained
forcalculating
the radiological
consequences
of the postulated
SGTR .... ." Reference
14,Attachment
1 at Page 1. In particular,
I&M noted that its analysis
retained
existing
licensing
basis assumptions
regarding
the availability
of certain systems,
components,
and instruments
(listed in a table within Reference
14) credited
for accident
mitigation
in an SGTR. Among theitems listed in that table were the "air-operated"
SG PORVs, which the notes accompanying
thetable stated were themselves
safety-grade
components
because they "form part of the mainsteam system pressure
boundary
upstream
of the SG stop valves,"
even though their "electrical
and control air appurtenances
[were] not safety-grade."
Reference
14, Attachment
1 at Pages3-4. Reference
14 also noted that I&M's limited use of the WCAP-10698-P-A
methodology
would not disturb CNP's existing
licensing
basis assumption
that an SGTR accident
would notinvolve a single failure.
Reference
14, Attachment
1 at Page 6.Reference
14 also communicated
I&M's intention
to retain CNP's existing
assumptions
regarding
the availability
of offsite power. Acknowledging
that the WCAP-10698-P-A
methodology
assumes that "the most challenging
SGTR scenario
with respect to SG fill includesa coincident
loss of offsite power", Reference
14 noted that the modified
SGTR analysis
wouldretain CNP's original
licensing
assumption
that SG PORVs would remain available
despite thefact that "offsite
power [was] not ...available."
Reference
14, Attachment
1 at Page 4.
Enclosure
2 to AEP-NRC-2013-53
Page 8Reference
14 contained
no suggestion
of a change in the scope of the LOOP assumed withinCNP's SGTR accident
analysis.
By letter dated October 24, 2001 (Reference
4), the NRC Staff approved
I&M's LAR in modifiedform to accommodate
CNP's existing
licensing
basis assumptions
for SGTR accidents.
In theSER submitted
with its approval
of I&M's LAR, the NRC Staff acknowledged
that licensees
likeI&M could not incorporate
the WCAP-10698-P-A
methodology
within their SGTR accidentanalysis
in a uniform fashion because "variations
in plant designs prevent a single model fromadequately
representing
all Westinghouse
Plants."
Reference
4, SER at Page 2.Consequently,
the NRC Staff devoted much of the SER to evaluating
the differences
betweenthe generic WCAP-1 0698-P-A
methodology
and I&M's proposed
approach
for incorporating
thatmethodology
within its licensing
basis.The NRC Staff noted that in the immediate
case, those differences
included
I&M's intention
ofretaining
CNP's existing
assumptions
for SGTR accidents:
To implement
the WCAP, the licensee
used the LOFTTR2 computer
codeand the plant-specific
current licensing
basis assumptions.
Reference
4, SER at Page 2 (emphasis
added). The NRC Staff explicitly
acknowledged
thatCNP's licensing
basis assumptions
credited
certain systems and components,
including
the SGPORVs and their control air appurtenances,
as remaining
available
for mitigation
of an SGTRaccident:
The licensee
provided
a list of systems,
components,
and instrumentation
that are used for SGTR accident
mitigation.
They also specified
thesafety classification
of the systems and power sources.
However,
thelicensee
listed several systems used for SGTR mitigation
that are notsafety related and do not have safety related backups.
The licenseejustified
the use of the non-safety-related
equipment
by stating that thesesystems are credited
in the current UFSAR Section 14.2.4 accidentanalysis.
Upon review of Section 14.2.4, the staff concludes
that thelicensing
basis SGTR analysis
does credit limited use of non-safety
gradeequipment
for mitigating
the SGTR.Reference
4, SER at Page 3. Similarly,
the NRC Staff acknowledged
that CNP's licensing
basisdid not assume a worst single failure during an SGTR accident
as the WCAP-10698-P-A
methodology
did:[T]he licensee
did not assume the worst single failure as prescribed
bythe WCAP-10698-P-A
safety analysis,
and did not provide it's [sic] effecton the margin to overfill.
The licensee
based their decision
not to assumethe worst single failure on the fact that their current licensing
basis doesnot include a single failure.Reference
4, SER at Page 4. Further,
the SER nowhere mentions
that I&M intended
to discardCNP's existing
assumption
of a coincident
single-unit
or that
Enclosure
2 to AEP-NRC-2013-53
Page 9the LOOP assumed within the WCAP-10698-P-A
methodology
supplanted
CNP's existinglicensing
basis assumptions
for SGTR accidents.
Although
I&M's proposed
retention
of CNP's existing
licensing
basis assumptions
for SGTRaccidents
"varied significantly"
from the assumptions
underlying
the WCAP-10698-P-A
methodology,
the NRC Staff approved
I&M's use of some elements
of the WCAP-10698-P-A
methodology
identified
in the LAR and related correspondence:
[T]he NRC staff concludes
that the licensee
can incorporate
theLOFTTR2 code into its licensing
bases for CNP and can use theLOFTTR2 code, with the current licensing
basis assumptions
as inputs forthe overfill
analysis
tube rupture accidents.
Thischange to the licensing
basis does not affect accident
initiators
orprecursors.
This change also does not ...decrease
the ability of theoperators
to mitigate
the consequences
of an accident.
Reference
4, SER at Page 5 (emphasis
added). In justifying
its approval
of a modifiedWCAP-10698-P-A
methodology
for use at CNP, the NRC Staff noted that I&M's adaptation
ofthe WCAP-10698-P-A
methodology
to CNP's existing
licensing
basis assumptions
for SGTRaccidents
did not affect conservative
estimates
of the radiological
consequences
of a designbasis SGTR at CNP. Reference
4, SER at Page 3.I&M's subsequent
review of docketed
correspondence
with the NRC Staff has identified
nofurther changes to CNP's licensing
basis assumptions
regarding
the availability
of SG PORVs inan SGTR accident,
the absence of a single failure assumption
within CNP's SGTR accidentanalysis,
or the scope of a LOOP assumed in the SGTR analysis.
5. The NRC Staff's Understanding
of CNP's Licensing
Basis Assumptions
for SGTR Accidents
Does Not Address Pertinent
Docketed
Correspondence,
Is Unsupported
by a Fair Readingof the UFSAR, and is Inconsistent
with the NRC's Historical
and Current Regulatory
Positions
a. The NRC Staff's Reading of CNP's Licensing
Basis Assumptions
for SGTRAccidents
Does Not Address Pertinent
Docketed
Correspondence
As noted earlier,
the NCVs within Reference
1 are based on the NRC Staffs contention
that thecoincident
LOOP assumed within CNP's licensing
basis SGTR accident
analysis
involves
a lossof offsite power to both units at CNP. The NRC Staff's position
is based on a single argumentwithin Reference
5: that it follows from the use of the terms "LOOP" and "station"
in a handful ofCNP UFSAR sections,
some of which are unrelated
to SGTR accident
analysis,
that a LOOPcan refer to the denial of offsite power to one or both units at CNP.In support of this argument,
Reference
5 advances
only a handful of UFSAR passages.
Thefirst UFSAR passage referenced
in Reference
5 comes from Section 1.3.7 describing
theauxiliary
electrical
system for each of the two units at CNP:Donald C. Cook's UFSAR Section 1.3.7, "Electrical
System" states, "Themain generators
are 1800 rpm, Phase III, 60 cycle, hydrogen
and water
Enclosure
2 to AEP-NRC-2013-53
Page 10cooled units. The main transformers
deliver generator
power to the345kV and 765 kV switchyards.
The station auxiliary
power systemconsists
of auxiliary
transformers,
4160V and 600 V switchgear,
600Vmotor control centers,
120 V A-C vital instrument
buses and 250 V D-Cbuses."Reference
5 at Page 3 (emphasis
supplied
by NRC Staff). Based on the fact that UFSARSection 1.3.7 described
the identical
electrical
systems for both units, Reference
5 concluded
that the UFSAR passage's
reference
to "station"
must refer to both units at CNP, rather than toeach unit individually.
In the same vein, Reference
5 cites a passage from Section 1.3.8 of theUFSAR describing
the Safety Features
associated
with each unit at CNP:Also, Section 1.3.8, "Safety Features,"
describes
the safety featuresincorporated
into the design of the plant, including
the fact that "even ifexternal
auxiliary
power to the station is lost concurrent
with an accident,
power is available
for the engineered
safeguards
from on-site dieselgenerator
power to assure protection
of the public health and safety forany loss of coolant accident."
Reference
5 at Page 3 (emphasis
supplied
by NRC Staff). Here, too, Reference
5 concludes
the fact that Section 1.3.8 describes
identical
safety features
at each unit means that thepassage's
reference
to "station"
must refer to both units at CNP, rather than only one unit.Lastly, Reference
5 points to language
within a passage from the accident
analysis
(atSection 14.1.12)
for "Loss of All AC Power to the Plant Auxiliaries"
at Unit 1:"A complete
loss of all (non-emergency)
AC Power (e.g., offsite power)may result in the loss of all power to the plant auxiliaries,
i.e., the RCPs,condensate
pumps, etc. The loss of power may be caused by a completeloss of the offsite grid accompanied
by a turbine trip at the station,
or by aloss of the on-site AC distribution
system."Reference
5 at Page 4. The NRC Staff read this reference
to a "complete
loss of offsite gridaccompanied
by a turbine trip at the station"
associated
with the design basis event postulated
within Section 14.1.12 to mean that a LOOP affecting
both units is within CNP's licensing
basisfor every event evaluated
in UFSAR Section 14. Reference
5 at Page 4. Based on theseexamples,
Reference
5 reports that NRR concurred
with NRC Staff that had performed
theCDBI that the LOOP assumed in CNP's SGTR analysis
was a "station
event, not a unit specificevent." Reference
5 at Page 4.The NRC Staff's position
and the UFSAR passages
described
above represent
the only basisidentified
by the NRC Staff for its position
throughout
the multiple
docketed
communications
andmeetings
with I&M since the CDBI began in July 2012. The NRC Staff has identified
noregulatory
provisions
or policy guidance
requiring
the assumption
of a LOOP affecting
both unitsfor a design basis SGTR accident.
The NRC Staff has advanced
no docketed
correspondence
in support of its understanding
of CNP's licensing
basis for SGTR accidents,
and has identified
no additional
passages
its position.
Enclosure
2 to AEP-NRC-2013-53
Page 11Further,
the NRC Staff has yet to provide a meaningful
response
to the analysis
provided
byI&M in References
3 and 7 in support of its understanding
of CNP's licensing
basisassumptions.
Reference
5 does not specifically
address the SGTR accident
analysisassumptions
identified
within docketed
correspondence
highlighted
within Reference
3:The scope of this TIA was limited to the licensing
basis as related tooffsite power only. The staff did not evaluate
other assertions
in thelicensee's
white paper.Reference
5 at Page 4.1 Reference
2 merely repeated
Reference
5's claims regarding
CNP'slicensing
basis, rather than address the detailed
licensing
basis interpretation
within Reference
7 provided
by I&M.Further,
although
Reference
1 suggests
that it addresses
the understanding
of CNP's SGTRaccident
licensing
basis assumptions
advanced
by I&M in References
3 and 7, a careful readingof the bases identified
in Reference
1 indicates
that the NRC Staff's reasoning
is circular
in thatit depends on, rather than proves the assumption
of a multi-unit
LOOP in CNP's SGTR accidentanalysis.
Specifically,
in acknowledging
I&M's position
that CNP's licensing
basis had neverassumed a single failure of a non-safety-related
component
(specifically
the unaffected
unit'sPAC) during an SGTR event, Reference
1 contends
that I&M had nevertheless
failed todemonstrate
that an unaffected
unit's PAC would reasonably
be available
during an SGTRaccident
affecting
one unit:The inspectors
agreed that certain older operating
plants arecredited
with the use of non-safety
related equipment
to mitigateevents. In these cases, the licensee
was required
to demonstrate
the non-safety-related
equipment
would reasonably
be available
and use of the equipment
was bound by a safety-related
path.Reference
1, Enclosure
at Pages 4 and 5. Similarly,
the NRC Staff in Reference
1 agrees withI&M's observation
in Reference
7 that the original
SER for Unit 1 did not consider
that a CACwould be out of service for maintenance
pursuant
to an assumed single failure,
claiming
thatthis demonstrates
that a CAC would have to be available
to supply control air pressure
during adesign basis SGTR accident,
as its availability
would be a limiting
condition
in CNP's SGTRaccident
analysis.
However,
the above arguments
do not prove the NRC's Staff understanding
of the scope of theLOOP assumed in CNP's SGTR accident
analysis.
Because the unaffected
unit's non-safety-
related PAC would remain available
during a single-unit
LOOP, control air pressure
would bereasonably
available
and bounded by a safety-related
path for main steam system pressureretention
purposes,
regardless
of the status of the CAC on the affected
unit. Similarly,
theavailability
of the affected
unit's CAC is not a limiting
condition
for CNP's SGTR accidentanalysis
if the coincident
LOOP affects only the unit experiencing
the SGTR event such that the1 The NRC Staff has not docketed
correspondence
between Region III personnel
and NRRpersonnel
defining
the scope of NRR personnel's
review of the competing
interpretations
ofCNP's licensing
basis assumptions
for the LOOP assumed within CNP's SGTR design basisaccident
analysis.
Enclosure
2 to AEP-NRC-2013-53
Page 12PAC on the unaffected
unit remains available
to provide control air pressure
to the affectedunit's SG PORVs. Lastly, the NRC Staff statement
quoted above is inconsistent
with the NRCStaff's statements
within Reference
4 endorsing
CNP licensing
basis assumptions
crediting
theavailability
air system components
during an SGTR accident.
b. The NRC Staff's Position
Is Unsupported
by a Fair Reading of the UFSARThe NRC Staff's categorical
statement
that every reference
to a LOOP within CNP's UFSARcan be understood
to refer to an event denying offsite power to one or both units at CNP isunsupported
by a careful reading of that document.
The UFSAR contains
no generic,controlling
definition
of the term LOOP requiring
it to be understood
as referring
to either asingle or multi-unit
event at every use within the UFSAR. Similarly,
the NRC Staff has identified
no regulatory
requirement,
policy guidance,
or docketed
correspondence
with I&M requiring
anyreference
to a LOOP to refer to either a single or multi-unit
event. Consequently,
whether aparticular
reference
to a LOOP within CNP's UFSAR refers to a LOOP affecting
one or bothunits at CNP must be determined
by reference
to a number of factors such as the textsurrounding
the UFSAR's reference
to the LOOP, the larger structure
of CNP's UFSAR, as wellas the relevant
historical
and regulatory
background.
i. The NRC Staff's Understanding
of the Scope of a LOOP Is Not Supported
bythe Surroundinq
TextA comparison
of the different
contexts
in which the term LOOP appears within CNP's SGTR andLoss of All AC Power to the Plant Auxiliaries
accident
analyses,
respectively,
does not supportthe NRC's generic interpretation
of the term. As noted earlier,
the NRC Staff's understanding
ofCNP's licensing
basis is based on the potentially
broad scope of the LOOP within UFSAR Unit 1Section 14.1.12,
"Loss of All AC Power to the Plant Auxiliaries."
The UFSAR's description
ofthe particular
LOOP at issue could involve:A complete
loss of all (non-emergency)
AC power (e.g., offsite power) ...result[ing]
in the loss of all power to the plant auxiliaries
.... The loss ofpower may be caused by a complete
loss of the offsite grid accompanied
by a turbine generator
trip at the station,
or by a loss of the on-site ACdistribution
system.Reference
5 at Page 4 (quoting
UFSAR Unit 1, Section 14.1.12.1)
(emphasis
added). Becausethe context of the UFSAR cited above passage is on its face ambiguous
regarding
the numberof units at CNP affected
by the LOOP, the NRC Staff contends
that it could, based only on agenerous
reading of the cited text alone, be read to refer to a LOOP to one or both units atCNP.The context surrounding
the use of the term LOOP within the SGTR accident
analysis
inUFSAR Units 1 and 2 Section 14.2.4 demands an entirely
different
conclusion
regarding
thenumber of units losing offsite power in a LOOP. Here, the UFSAR's use of the term LOOP isnot qualified
by the broad adjectives,
complete
loss, all power, the offsite grid, etc., used in theearlier accident
analyses
in a way that could arguably
suggest a LOOP denying power to bothunits; rather, CNP's SGTR accident
analysis
refers only to "offsite
power", or "a loss of offsitepower" or "a coincident
loss of offsite power." Reference
6 at Section 14.2.4.
Enclosure
2 to AEP-NRC-2013-53
Page 13ii. The NRC Staffs Understandinq
of the Meaninq of a LOOP Is Inconsistent
with the Structure
of CNP's UFSARThe structure
of the UFSAR also undercuts
the generic meaning attached
to the term LOOP bythe NRC Staff. According
to Reference
5, the potentially
broad scope of the LOOP described
inUFSAR Section 14.1.12 defines the meaning of the term throughout
the UFSAR. Reference
5at Page 4. However,
the NRC Staff provides
no justification
for why the particular
(broad)meaning it assigns to the term LOOP within UFSAR Section 14.1.12 is more appropriate
forgeneric application
throughout
the UFSAR than the more limited-scope
LOOP described
withinother sections
of the UFSAR such as Section 14.2.4.The NRC Staff's position
is also not supported
by the NRC and industry
guidance
regarding
theform and content of CNP's UFSAR. Consistent
with the scheme laid out in Regulatory
Guide1.70 (Reference
events and accidents
satisfying
aminimal threshold
for best-estimate
frequency
of occurrence,
which are then assigned
afrequency
grouping
based on criteria
established
by the American
Nuclear Society (ANS). Asstated in UFSAR Sections
14.0, ANS Condition
1 (normal operational
are omittedfrom CNP's UFSAR, while Condition
2 events (moderate
frequency)
appear mostly in UFSARSections
14.1, Condition
3 (infrequent)
events in UFSAR Section 14.2, and Condition
4 (unlikely
but limiting)
events mostly appear in UFSAR Section 14.3. Consistent
with Regulatory
Guide1.70, CNP's UFSAR analyzes
each of the events within the UFSAR individually
and for eachunit, to include a description
of the initial assumptions,
sequence
of events, and radiological
consequences
specific
to each event. Reference
15 at Pages 15-4 to 15-7.The NRC Staff's position
does not account for this structure.
ANS guidance
identifying
thethreshold
for consideration
of transient
events and accidents
within an FSAR requires
a minimalbest-estimate
frequency
of occurrence
of >l.OE-6/yr.
Reference
16 at 6. However,
when theNRC Staff used its Donald C. Cook Nuclear Plant Standardized
Plant Analysis
Risk (SPAR)Model to calculate
a best-estimate
frequency
of occurrence
for an SGTR with a coincident,
multi-unit
LOOP, it obtained
a value (2.12E-6/yr)
not much greater than the threshold
in ANSguidance;
further,
when accounting
for the risk that a CAC would be unavailable
formaintenance
for 30 days, the best-estimate
frequency
of occurrence
fell below (1.75E-7/yr)
theANS threshold.
Reference
1 at Enclosure
Page 7. Informal
calculations
by I&M incorporating
more recent industry
data on the frequency
of multi-unit
LOOPs provide more reason toconclude
that a multi-unit
LOOP is too remote an event to be considered
in CNP's design basisSGTR analysis.
According
to Reference
17, there was not one reactor trip coincident
with amulti-unit
LOOP reported
by the U.S. commercial
nuclear power industry
between 1986-2004.
Reference
17 at Page 51. Using this data, I&M's informal
calculation
of the probability
of anSGTR with a coincident,
multi-unit
LOOP yields a best-estimate
frequency
of occurrence
of6.33E-7/yr
-below the ANS threshold
for consideration
thebest-estimate
frequency
of occurrence
is even lower (1.91 E-8) when accounting
for the risk thata CAC would be unavailable
for any reason, including
maintenance.
Further,
although
Regulatory
Guide 1.70 states that the input parameters
and initial conditions
for each accident
should be "clearly
identified"
within its analysis,
the NRC Staff's contention
assumes that the assumptions
regarding
the potential
scope of one UFSAR Section 14 analysis
Enclosure
2 to AEP-NRC-2013-53
Page 14(Loss of All AC Power to the Plant Auxiliaries)
automatically
carry over wholesale
to subsequent
accident
analyses
(SGTR). Reference
15 at Page 15-5.Additionally,
the NRC Staff's contention
that its reading of the scope of the LOOP within UFSARSection 14.1.12 should apply to the LOOP assumed in CNP's Section 14.2.4 SGTR analysis.
compares
accidents
with very different
frequencies.
The Loss of All AC Power to the PlantAuxiliaries
is an ANS Condition
II event, while the SGTR accident
is a Condition
III event.Reference
6 at Section 14.0. Further,
because a dual-unit
LOOP can be expected
to occurmuch less frequently
than a single-unit
LOOP, application
of the NRC Staff's reading of thescope of the term LOOP within CNP's SGTR analysis
represents
a significant
change in theinitial assumptions
and anticipated
frequency
for that particular
accident.
That revisedfrequency
of CNP's design basis SGTR accident
could conceivably
require the assignment
ofnew ANS Conditions
to either the UFSAR Loss of All AC Power to the Plant Auxiliaries
analysis(Reference
6 at Section 14.1.12),
or its SGTR accident
analysis
(Reference
6 at Section14.2.4),
which in turn would require the re-organization
theNRC Staff's position
does not account for the significance
attached
by NRC guidance
to thedistinction
between different
ANS Conditions
and (by extension)
types of design basis events oraccidents.
The NRC Staff's references
to the use of the word "station"
within the UFSAR's description
ofCNP systems is similarly
not helpful for determining
the scope of the LOOP assumed in CNP'sSGTR accident
analysis.
In support of its contention
that every use of the term LOOP refers toeither a single or multi-unit
event, Reference
5 points to a handful of examples
of the UFSAR'suse of the word "station"
in descriptions
of CNP Electrical
System (at Section 1.3.7) and SafetyFeatures
(at Section 1.3.8) that the NRC Staff understands
to refer to both units at CNP.However,
the NRC Staff nowhere explains
why a handful of references
to the word "station"
within the system descriptions
in Sections
1.3.7 and 1.3.8 define the use of that and otherterms (e.g., LOOP) throughout
the UFSAR. Regulatory
Guide 1.70 understood
the systemdescriptions
within the first section of a licensee's
UFSAR to be distinct
from the accidentanalyses
described
in a later section of the UFSAR:The first chapter of the SAR should present an introduction
to the reportand a general description
of the plant. This chapter should enable thereader to obtain a basic understanding
of the overall facility
withouthaving to refer to the subsequent
chapters.
Reference
15 at Page 1-1 (emphasis
added). In contrast,
the NRC Staff's position
determines
the meaning of ambiguous
terms ("station",
"LOOP") in the UFSAR's SGTR accident
analysisassumptions
not by reference
to surrounding
text, but by reference
to language
in an entirelydifferent
UFSAR section.
The NRC Staff's more fluid distinction
between UFSAR sections
isdifficult
to reconcile
with the approach
endorsed
within Regulatory
Guide 1.70.Although
the NRC Staff in Reference
1 states that the difference
between UFSAR sectionsidentified
above supports
its understanding
of CNP's licensing
basis, the NRC Staffs position
iserroneous.
Conceding
that high-level
system descriptions
within Section 1 of CNP's UFSAR donot prescribe
accident
analyses
assumptions
within subsequent
UFSAR sections,
the NRC Staffincorrectly
asserts that:
Enclosure
2 to AEP-NRC-2013-53
Page 15This argument
supports
the inspectors'
position
that the licenseecannot take credit for the unaffected
unit's non-safety-related
PACunless explicitly
approved
by the NRC and described
in the SGTRanalysis.
Reference
1, Enclosure
at Page 5 (emphasis
added). Notwithstanding
the fact the languagewithin Section 1 of CNP's UFSAR is unhelpful
for interpreting
language
describing
UFSARaccident
analysis
assumptions,
it does not follow that Section l's high-level
description
of thecomponents
comprising
CNP systems would not control throughout
the UFSAR. Regulatory
Guide 1.70 states that Section 1 of CNP's UFSAR exists precisely
so that I&M would not haveto describe
CNP systems and components
multiple
times. Reference
15 at Page 1-1. BecauseSection 1.3.9.h of CNP's UFSAR describes
CNP's compressed
air system as a shared systemof which both units' PACs and CACs are components,
the NRC Staffs explicit
endorsement
within the SER in Reference
4 of the continued
availability
of motive force to the SG PORVsfrom CNP's control air appurtenances
and equipment
permits I&M to take credit for theunaffected
unit's PAC in CNP's SGTR accident
analysis.
Further,
by the NRC Staff's logic, I&Mwould not be able to take credit for the operation
of any CAC or PAC within CNP's SGTRaccident
analysis,
as neither of those components
is explicitly
mentioned
in the UFSAR's SGTRaccident
analysis.
Additionally,
even if the NRC Staff's approach
were appropriate,
the cited examples
of the term"station"
within Section 1 of the UFSAR do not support its position.
Reference
6 Section 1.3.7states:"The station auxiliary
power system consists
of auxiliary
transformers,
4160 v and 600 v switchgear,
600 v motor control centers,
120 v-a-c vitalinstrument
buses and 250 v d-c buses."However,
the NRC Staffs suggestion
that the term "station"
in this context necessarily
refers toboth units at CNP is incorrect.
Indeed, each unit at CNP has the components
(redundant
auxiliary
transformers,
multiple
600 v switchgear,
independent
120 v-a-c vital instrument
busesand 250 v-d-c buses, and 4160 v and 600 v switchgear)
the NRC Staff suggests
represents
ashared system between CNP units. Similarly,
both units have the EDGs and turbinesmentioned
in the cited passage from UFSAR Section 1.3.8. Further,
the NRC Staff's claim thatthe use of the term "station"
within Section 1.3.8's description
of CNP Safety Features
provesthat there is only one, shared auxiliary
power system at CNP is at odds with surrounding
text notexamined
by the NRC Staff. Specifically,
UFSAR Section 1.3.9, "Shared Facilities
andEquipment,"
begins by noting that:Separate
and similar systems and equipment
are provided
for each unit,except as noted below.Reference
6 at Section 1.3.9 (emphasis
added). The auxiliary
power system is absent fromSection 1.3.9's list of shared systems and equipment.
iii. The NRC Staff's Understanding
of the Term LOOP Is at Odds with theReaulatorv
History of CNP and Similarlv-Situated
Facilities
Enclosure
2 to AEP-NRC-2013-53
Page 16The NRC Staff's understanding
of the term LOOP also does not account for docketedcorrespondence
acknowledging
the retention
of the assumptions
within CNP's original
SGTRaccident
analysis.
As explained
at length earlier,
the NRC Staff in 2001 reviewed
and explicitly
approved
I&M's retention
of CNP's original
licensing
basis assumptions
for SGTR accidents,
including
the assumption
of a single-unit
LOOP only. Consequently,
the NRC Staff'sunderstanding
of the scope of the term LOOP assumed within CNP's SGTR accident
analysisnot only re-writes
CNP's UFSAR, but also re-writes
nearly forty years' worth of pertinent
docketed
correspondence.
Further,
as explained
earlier,
the NRC Staffs reading of the term LOOP within CNP's SGTRaccident
analysis
is also inconsistent
with the regulatory
history of CNP and other multi-unit
facilities
of similar vintage.
The two units at CNP were licensed
and constructed
on a staggered
schedule,
with construction
on Unit 1 beginning
before Unit 2 such that Unit 1 received
itsoperating
license several years before Unit 2 (1974 as opposed to 1977). Consequently,
theSGTR accident
analysis
within CNP's original
licensing
basis did not, as a practical
matter,assume a multi-unit
LOOP.Further,
the CNP is not the only licensee
that assumes only a single-unit
LOOP within thedesign basis accident
analyses
for the units at its facility.
I&M's informal
polling of other multi-unit facilities
licensed
in approximately
the same timeframe
as CNP reveals that many of thoselicensees
understand
the licensing
basis assumptions
for units at their facility
to assume only asingle-unit
LOOP during SGTRs and other accidents.
Further,
among those licensees
whoselicensing
basis currently
assumes multi-unit
LOOPs were some who acknowledged
that theircurrent licensing
basis assumptions
are a departure
from original
licensing
basis assumptions
that understood
LOOPs to affect only a single unit at their facility.
Lastly, the Commission's
current regulations
and guidance
governing
the availability
of offsitepower reflect the unit-specific
approach
to electric
system design within licensing
basis accidentassumptions
at CNP and other similarly-situated
facilities.
Most prominently,
the current StationBlackout
Rule at 10 CFR 50.63 (Reference
8) is unit-specific
in its approach
to the availability
ofAC power, including
offsite power. Although
the NRC has recently
published
a Federal Registernotice (Reference
18 at 16179) indicating
a desire to revise its Station Blackout
Rule and otherregulations
and guidance
to adopt a facility-wide
perspective
on continuity
of electrical
power,interpreting
the language
within CNP's licensing
basis against that proposed
approach
would bepremature,
regardless
of whether the NRC Staff can (as Reference
1 asserts)
conceive
ofscenarios
in which plant configuration
would make a multi-unit
LOOP a credible
event at CNP.6. The NRC Staffs Position
Is Unnecessary
for Assuring
Adequate
Protection
Against EitherDesign Basis Events or Beyond-Design
Basis External
EventsNRC Orders issued following
the earthquake
and tsunami at the Fukushima
Dai-ichi
nuclearpower plant in March 2011 acknowledge
that existing
defense-in-depth
approaches
at licensedfacilities
provide adequate
protection
of public health and safety against design basis accidents.
Specifically,
states:To protect public health and safety...
the NRC's defense-in-depth
strategy
includes
multiple
layers of protection:
(1) prevention
of accidents
by virtue of the design, construction,
and operation
of the plant; (2)
Enclosure
2 to AEP-NRC-2013-53
Page 17mitigation
features
to prevent radioactive
releases
should an accidentoccur; and (3) emergency
preparedness
programs
that include measuressuch as sheltering
and evacuation
.... These defense-in-depth
featuresare embodied
in the existing
regulatory
requirements
and thereby provideadequate
protection
of the public health and safety.Reference
19 at Page 5 (emphasis
added). Compliance
with those NRC requirements,
theNRC concluded,
"presumptively
assures adequate
protection"
of public health and safety frominadvertent
release of radioactive
materials
during a design basis accident.
Reference
19 atPages 4-5.As explained
at length earlier,
the NRC Staff's contention
within Reference
1 that CNP is not incompliance
with licensing
basis requirements
for a design basis SGTR accident
is incorrect.
CNP's licensing
basis has never assumed that the LOOP coincident
with a design basis SGTRaccident
involves
both units at CNP, and the NRC Staff has presented
no meaningful
evidencein support of a contrary
position.
Further,
as recently
as 2001, the NRC Staff endorsed
themeasures
(including
the crediting
of the continued
availability
compressed
air system components)
I&M employs for mitigating
the risk of inadvertent
releaseof radioactive
materials
during a design basis SGTR accident
at CNP. Reference
4 concludes
that I&M's approach
to mitigating
the consequences
of a design basis SGTR provides"reasonable
assurance"
of protection
of public health and safety, and "will be conducted
incompliance
with the Commission's
regulations.
... "Further,
as noted earlier,
I&M has supplemented
the mitigation
measures
for SGTR accidents
evaluated
within Reference
4 to provide additional
defense-in-depth
from design basis SGTRaccidents.
Specifically,
I&M in March 2013, completed
installation
of a plant modification
andrevised CNP operating
procedures
to ensure that backup nitrogen
tanks are immediately
andautomatically
available
during an SGTR for operation
of SG PORVs without the need for manualvalve manipulation
outside the control room. I&M has also revised CNP Work Controlprocesses
to provide additional
defense-in-depth
from a loss of control air pressure
byrestricting
removal for maintenance
of the operating
unit's CAC when the opposite
unit isshutdown
and the shutdown
unit's PAC is aligned to preferred
offsite power.In contrast,
the NRC Staff has not demonstrated
that its position
would result in any meaningful
contribution
to adequate
protection
of public health and safety from design basis SGTRaccidents
at CNP. As noted earlier,
the most recent published
industry
data on the frequency
ofLOOPs within Reference
17 indicates
that the best-estimate
frequency
of occurrence
for a multi-unit LOOP coincident
with an SGTR would fall well below the minimal threshold
within ANSguidance
(Reference
16) for consideration
within CNP's design basis. Moreover,
the difference
in core damage frequency
from adopting
the NRC Staff's position
regarding
the scope of theLOOP accompanying
a design basis SGTR accident
is so small (2.4E-8/yr)
as to provide nomeaningful
advantage
over I&M's understanding
of CNP's licensing
basis for assuring
adequateprotection
of public health and safety. Reference
1, Enclosure
at Page 1. Further,
even thismarginal
difference
in core damage frequency
between I&M's and the NRC Staff's positions
islikely overstated,
as the core damage frequency
calculation
within Reference
1 (Enclosure
atPages 6-7) does not account for the additional
defense-in-depth
measures
implemented
Enclosure
2 to AEP-NRC-2013-53
Page 18Lastly, the NRC Staff has provided
no basis to conclude
that I&M has failed to provide adequateprotection
against beyond-design
basis scenarios
involving
an SGTR accompanied
by acoincident,
multi-unit
LOOP. As explained
in Order EA-12-049,
the events at Fukushima
Dai-ichi
demonstrated
the need for licensees
to adopt additional
defense-in-depth
measures
tomitigate
the consequences
of beyond-design
basis external
events, such as those resulting
inthe extended
loss of electrical
power at multiple
units at a facility.
Reference
19 at Pages 4-6.Subsequent
NRC guidance
(Reference
20 at Page 4) endorsed
licensees'
use of the NuclearEnergy Institute's
(NEI's) Diverse and Flexible
Mitigation
Capability
(FLEX) strategy
(Reference
21) to satisfy Order EA-12-049's
requirements
for assuring
adequate
protection
against beyond-design basis external
events resulting
in extended
loss of electrical
power (including
offsitepower) at both units at a multi-unit
facility.
As required
by Order EA-1 2-049, I&M has submitted
an Overall Integrated
Plan (Reference
22) for mitigation
of beyond-design
basis external
eventsat CNP. I&M's Overall Integrated
Plan incorporates
the FLEX strategy
endorsed
by the NRCStaff in Reference
20 for use by licensees
in satisfying
the requirements
within Order EA-12-049
for mitigation
measures
providing
adequate
protection
from beyond-design
basis events suchas a multi-unit
LOOP accompanying
an SGTR.7. The NRC Staff's Determination
that the NCVs Represent
a More-than-Minor
Performance
Deficiency
Involving
Cross-Cutting
Aspects Lacks MeritIn Reference
1, the NRC Staff contends
that the NCVs represent
a more-than-minor
performance
deficiency
involving
cross-cutting
areas of human performance,
the component
ofdecision
making, and the aspect of conservative
assumptions.
Reference
1 Enclosure,
atPages 1 and 2. The NRC Staff stated that the NCVs involved
cross-cutting
aspects becauseI&M's plant procedures
assumed that the unaffected
unit's compressed
air system equipment
would be available
during an SGTR accident,
despite the fact that the NRC Staff nowunderstands
CNP's licensing
basis to assume that an SGTR accident
would be accompanied
bya multi-unit
LOOP. Reference
1 Enclosure,
at Pages 1 and 2.The NRC Staff's conclusion
that the NCVs involve cross-cutting
aspects,
however,
incorrectly
assumes the validity
of NCVs identified
within Reference
1. As explained
at length above, thoseNCVs are based on an erroneous
understanding
of the scope of the coincident
LOOP withinCNP's design basis SGTR accident
analysis:
contrary
to the NRC Staffs current position,
CNP's licensing
basis has only ever assumed a single-unit
LOOP as an initial condition
in anSGTR event. Consequently,
the unaffected
unit's PAC will remain available
to provide controlair pressure
to operate SG PORVs in the affected
unit in the event of an SGTR event,regardless
of the status of the CAC of the affected
unit. Further,
the NRC Staff in the 2001 SERwithin Reference
4 endorsed
I&M's claims regarding
the continued
availability
of control air tooperate an affected
unit's SG PORVs during an SGTR accident,
notwithstanding
a coincident
LOOP. Because the NCVs within Reference
1 are incorrect,
the NRC Staff's conclusion
thatthose NCVs involve cross-cutting
aspects is similarly
incorrect.
Additionally,
even if the NRC Staff's current understanding
of CNP's licensing
basis werecorrect,
the NCVs identified
within Reference
1 would not involve cross-cutting
aspects.Although
Reference
1 (Enclosure,
Page 7) criticizes
I&M for not having adopted requirements,
EOPs, and work control procedures
positively
demonstrating
safety, the NRC Staff nowhereexplains
how I&M's requirements
were inconsistent
with reactor safety and public health. Asnoted earlier,
the NRC Staff concluded
in the SER (Pages 3 to 5) within Reference
4 that the
Enclosure
2 to AEP-NRC-2013-53
Page 19changes to CNP's licensing
basis proposed
by I&M in its 2000 LAR would not increase
the riskor consequences
of an SGTR accident
beyond the conservative
estimates
within CNP's originallicensing
basis. In arriving
at this conclusion,
the NRC Staff explicitly
noted that I&M hadrevised its EOPs for SGTR accidents
to improve margin to steam generator
overfill.
Reference
4, SER at 4. Further,
the core damage frequency
data provided
by the NRC Staff inReference
1 (Enclosure
at Page 1) is consistent
with the NRC Staffs conclusions
withinReference
4, as the difference
in core damage frequency
from assuming
a dual-unit
LOOP isonly marginally
different
(2.4E-8/yr)
from scenarios
involving
a single-unit
LOOP.Further,
the NRC Inspection
Manual states that for an NCV to have cross-cutting
aspects,
theperformance
deficiency
at issue must be "recent (i.e., nominally
within the last three years)."Reference
23, at Page 3. However,
as explained
at length above, the NCVs in Reference
1 arebased on an understanding
of CNP's licensing
basis that has been in place since the originallicensing
of Unit 1 at CNP around forty years ago, and which was endorsed
by the NRC Staff asrecently
as 2001. Consequently,
the NCVs within Reference
1 do not satisfy NRC Inspection
Manual standards
for determining
whether NCVs have cross-cutting
aspects.Nor can the NRC Staff claim that I&M's failure to correct the longstanding
performance
deficiency
until recently
is indicative
of present performance.
Although
the NRC Inspection
Manual allows for a cross-cutting
determination
if "the performance
deficiency
occurred
morethan three years ago, but the performance
characteristic
has not been corrected
or eliminated",
it severely
limits the application
of this exception
to "some rare or unusual cases". Reference
23at Page 3. Reference
1 provides
no justification
for why the NCVs represent
a "rare or unusualcase" warranting
application
of this exception.
Further,
as explained
above, I&M'sunderstanding
of its licensing
basis is not rare or unusual;
in fact, multiple
plants of similarvintage and configuration
have the same licensing
basis assumptions
regarding
the scope of aLOOP during an SGTR or other accident.
8. Conclusion
For the reasons identified
above, both the NCVs identified
within Reference
1 and the NRCStaff's determination
that those NCVs involve cross-cutting
aspects are incorrect.
Enclosure
2 to AEP-NRC-2013-53
Page 20REFERENCES:
1. Letter from G. Shear, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Power Plant,Units 1 and 2, Component
Design Basis Inspection
dated July 8, 2013.2. Letter from A. M. Stone, NRC, to L. J. Weber, I&M, "D. C. Cook Nuclear Power Plant,Units 1 and 2, Component
Design Bases Inspection
dated January 11, 2013.3. Letter from W. Hodge, I&M, to C. Tilton, NRC, "D. C. Cook CDBI Response
to Question2012-CDBI-298,"
dated November
15, 2012.4. Letter from J. F. Stang, NRC, to R. P. Powers, I&M, "Donald C. Cook Nuclear Plant,Units 1 and 2 -Issuance
of Amendments
(TAC Nos. MB0739 and MB0740),"
datedOctober 24, 2001.5. Letter from K. O'Brien,
NRC, to S. Bahadur,
NRC, "Task Interface
Agreement
-Licensing
Basis for Donald C. Cook Nuclear Power Plant, Units 1 and 2, During a SteamGenerator
Tube Rupture Event Coincident
with a Loss of Offsite Power (TIA 2012-11),"
dated December
7, 2012.6. Donald C. Cook Nuclear Plant Updated Final Safety Analysis
Report Rev. 24, datedMarch 17, 2012.7. Letter from I&M to Ann Marie Stone and Caroline
Tilton, NRC, "Response
to NRCInspection
Report Issued January 11, 2013 Containing
the Results of the Component
Design Basis Inspection
Conducted
Between July 23, 2012 and December
3, 2012,"dated February
8, 2013.8. 10 CFR 50.63, "Loss of All Alternating
Current Power."9. Donald C. Cook Nuclear Plant Preliminary
Safety Analysis
Report for Units 1 and 2,dated December
18, 1967.10. Donald C. Cook Nuclear Plant Final Safety Analysis
Report for Units 1 and 2, datedFebruary
2, 1971.11. Amendments
to Donald C. Cook Nuclear Plant Final Safety Analysis
Report for Units 1and 2, dated November
11, 1977.12. Amendments
to the Donald C. Cook Nuclear Plant Final Safety Analysis
Report for Units1 and 2, dated July 1997.13. Letter from R.P. Powers, I&M, to the NRC Document
Control Desk, "Letter C1000-11,
Donald C. Cook Nuclear Plant Units 1 and 2 License Amendment
Request for Changesin Steam Generator
Tube Rupture Analysis
Methodology,"
dated October 24, 2000.
Enclosure
2 to AEP-NRC-2013-53
Page 2114. Letter from M. W. Rencheck,
I&M, to the NRC Document
Control Desk, "Letter C0601-21, Donald C. Cook Nuclear Plant Units 1 and 2 Response
to Request for Additional
Information
Regarding
License Amendment
for 'Changes
TubeRupture Analysis
Methodology
(TAC Nos. MB0739 and MB0740),"
dated June 29, 2001.15. NRC Regulatory
Guide 1.70, "Standard
Format and Content of Safety Analysis
Reportsfor Nuclear Power Plants, Rev. 3, " dated November
1978.16. American
Nuclear Society,
"Nuclear
Safety Criteria
for the Designof Stationary
Pressurized
Water Reactor Plants,"
dated 1983.17. NUREG/CR-6890,
"Reevaluation
of Station Blackout
Risk and Nuclear Power Plants:Analysis
of Loss of Offsite Power Events 1986-2004,"
dated December
2005.18. 77 Federal Register
16175, "NRC Advanced
Notice of Proposed
Rulemaking:
StationBlackout,"
dated March 19, 2012.19. NRC Order Number EA-12-049,
"Order Modifying
Licenses
with Regard toRequirements
for Mitigation
Strategies
for Beyond-Design-Basis
External
Events,"
datedMarch 12, 2012.20. NRC Interim Staff Guidance
"Compliance
with Order EA-12-049,
Order Modifying
Licenses
with Regard to Requirements
for Mitigation
Strategies
forBeyond-Design-Basis
External
Events, Rev. 0," dated August 29, 2012.21. NEI 12-06, "Diverse
and Flexible
Coping Strategies
(FLEX) Implementation
Guide, Rev.0," dated August 2012.22. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2Overall Integrated
Plan In Response
to March 12, 2012 Commission
Order Modifying
Licenses
with Regard to Requirements
for Mitigation
Strategies
for Beyond-Design-
Basis External
Events (Order Number EA-12-049),"
dated February
27, 2013.23. NRC Inspection
Manual Chapter 0612, "Power Reactor Inspection
Reports,"
datedJanuary 24, 2013