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
INDIANA MICHIGAN POWER A unit of American Electric Power August 2, 2013 Docket Nos.: 50-315 50-316 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 Indiana Michigan Power.com AEP-NRC-2013-53
10 CFR 2.201 U.S. Nuclear Regulatory
Commission
Attn: Document Control Desk Washington, DC, 20555-0001
Donald C. Cook Nuclear Plant Units 1 and 2 Response 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. Nuclear Regulatory
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)," dated December 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, Units 1 and 2, Component
Design Bases Inspection
05000316/2012007," 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
05000316/2013010," dated July 8, 2013, (ADAMS Accession
No. ML13189A243).
This letter provides Indiana Michigan Power Company's (l&M's), Nuclear Plant (CNP) Units 1 and 2, response contesting
the documented
by Reference
4, Component
Design Bases 05000315/2013010;
licensee for Donald C. Cook Non-Cited
Violations (NCVs)Inspection (CDBI) Report In Reference
1, I&M identified
docketed correspondence
supporting
I&M's understanding
of CNP's licensing
basis to assume only a single-unit
loss of offsite power (LOOP) coincident
with a design basis 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 2 the results of its consultation
with the NRC Office of Nuclear Reactor Regulation
regarding
the NRC Staff'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 CNP between July 23, 2012, and December 31, 2012, would remain unresolved
items (URIs) pending the NRC Staffs resolution
of questions
regarding
the scope of a LOOP assumed within CNP's SGTR accident analysis.
In Reference
4, the NRC Staff resolved the URIs issued by Reference
3 and 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 operate SG PORVs as required by CNP's licensing
basis during an SGTR accident accompanied
by a LOOP 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 PORVs unavailable
after taking a control air compressor
out of service for maintenance.
Reference
4 characterized
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 the NCVs include the following:
- The NCVs are based on an erroneous
understanding
of CNP's licensing
basis. Contrary to the NCVs, CNP's licensing
basis assumptions
regarding
the initial conditions
for a SGTR accident have never considered
a coincident
LOOP involving
both units. Further, the NRC Staff'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 not represent
a fair reading of CNP's Updated Final Safety Analysis Report (UFSAR), and is inconsistent
with the NRC's current regulatory
position regarding
the loss of offsite power to non-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 fails to provide adequate protection
of public health and safety from either design basis events or beyond-design
basis external events. Further, the NRC Staff has not demonstrated
that its own position would provide a meaningful
improvement
in the protection
of public health and safety.* 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 scope of a LOOP assumed within CNP's design basis SGTR accident analysis, is inconsistent
with the NRC Staffs statements
in docketed correspondence, and is unrepresentative
of present licensee performance.
Enclosure
1 to this letter contains an affirmation
statement.
Enclosure
2 to this letter lays out in detail the regulatory
and factual support for I&M's response contesting
the NCVs.
U.S. Nuclear Regulatory
Commission
AEP-NRC-2013-53
Page 3 Regardless
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
to provide additional
defense-in-depth
and improved safety margins during an SGTR accident.
In March 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
during an 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 the operating
unit's control air compressor
when the opposite unit is shutdown and the shutdown unit's plant air compressor
is aligned to preferred
offsite power.This letter contains no new or revised commitments.
If you have any questions, please contact Mr. Michael K. Scarpello, Regulatory
Affairs Manager, at (269) 466-2649.Sincerely, Joel P. Gebbie Site Vice President DMB/kmh Enclosures:
1. Affirmation
2. Indiana Michigan Power Company's
Response to "Donald C. Cook Nuclear Power Plant, Units 1 and 2, Component
Design Bases Inspection
05000316/2013010," dated July 8,2013 c: C. A. Casto, NRC Region III J.T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosure E. Leeds, NRC NRR MDEQ-RMD/RPS
NRC Resident Inspector A. M. Stone, NRC Region III C. Tilton, NRC Region III T. J. Wengert, NRC Washington, DC R.P. Zimmerman, NRC Washington, DC
ENCLOSURE
I TO AEP-NRC-2013-53
AFFI RMATION I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President
of Indiana Michigan Power Company (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 herein pertaining
to I&M are true and correct to the best of my knowledge, information, and belief.Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED
BEFORE ME THIS____ DAY OF ,A)ws 2013 My Commission
Expires ( I 2 IO{
ENCLOSURE
2 TO AEP-NRC-2013-53
Indiana Michigan Power Company's
Response to "Donald C. Cook Nuclear Power Plant, Units 1 and 2, Component
Design Bases Inspection
05000316/2013010," dated July 8, 2013 1. Introduction
The Non-Cited
Violations (NCVs) within the Nuclear Regulatory
Commission (NRC) Staffs July 8, 2013, letter (Reference
1) to Indiana Michigan Power Company (I&M) are based on an erroneous
understanding
of the licensing
basis of Donald C. Cook Nuclear Plant (CNP). The NRC Staff's position that CNP's design basis Steam Generator
Tube Rupture (SGTR) accident assumes a coincident
loss of offsite power (LOOP) that can involve both units at CNP is inconsistent
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 Safety Analysis Report (UFSAR), and is likewise inconsistent
with relevant historical
and current regulatory
positions
of the NRC. Additionally, the NRC Staff has not demonstrated
that I&M's understanding
of CNP's licensing
basis fails to provide adequate protection
of public health and safety from either design basis events or beyond-design
basis external events. Lastly, the NRC Staff's determination
that the NCVs represent
a more-than-minor
performance
deficiency
with cross-cutting
aspects relies on an erroneous
understanding
of the scope of a LOOP assumed within CNP's design basis SGTR accident analysis, is inconsistent
with the NRC Staff's statements
in docketed correspondence, and is unrepresentative
of present licensee performance.
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, and December 31, 2012. As described
in Reference
2, the CDBI entailed a review of licensing
basis documentation
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
with CNP's licensing
basis requirements
for SGTR accidents.
As stated in Reference
2, the NRC Staff contended
during the CDBI that CNP was not in conformance
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 a SGTR accident included a coincident
LOOP affecting
both units at CNP, the NRC Staff reasoned that the only available
source of control air pressure during the most limiting SGTR accident would be the affected unit's dedicated
control air compressor (CAC) receiving
power from one of the two emergency
diesel generators (EDG). However, if the affected unit's CAC were unavailable
as a result of emergent or planned maintenance, then the NRC Staff reasoned that control air pressure would be unavailable
to operate the affected unit's SG PORVs. In reviewing
CNP operating
records, the NRC Staff identified
several occasions
in which CACs at
Enclosure
2 to AEP-NRC-2013-53
Page 2 CNP would have been unavailable
due to maintenance, but I&M had not declared the SG PORVs inoperable.
I&M disagreed
with the NRC Staff's characterization
of CNP's licensing
basis assumptions
for a SGTR event. Noting that the CNP licensing
basis for an SGTR event did not consider a coincident
multi-unit
LOOP, I&M contended
that the NRC Staffs finding was based on a beyond design basis accident scenario.
The NRC Staff requested
assistance
from the NRC Office of Nuclear 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 position and providing
docketed correspondence.
Reference
3 in particular
identified
a Safety Evaluation
Report (SER, Reference
4) dated October 24, 2001, explicitly
discussing
CNP's assumptions
for SGTR accident initial conditions, and revealing
the NRC Staff's evaluation
and endorsement
of I&M's understanding
of the CNP licensing
basis assumptions
for an SGTR accident.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 within CNP's UFSAR (Reference
6) in support of its position, interpreting
a handful of references
to the terms "LOOP" and "station" in descriptions
of CNP electrical
systems to mean that CNP's licensing
basis assumed a LOOP would affect both units 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 issue as 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's conclusions
regarding
CNP's licensing
basis assumptions
for SGTR accidents
without further explanation
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. On February 8, 2013, I&M provided Reference
7 to the NRC Staff, refuting Reference
5's interpretation
additional
detail regarding
the technical
and regulatory
bases supporting
I&M's understanding
of the CNP licensing
basis assumptions
for an SGTR accident.
During a May 20, 2013, technical
debrief of the CDBI findings, the NRC Staff repeated 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
of Technical
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
of a design basis SGTR accident)
and 3.7.4 (governing
the operability
of SG PORVs). Reference 1 states that I&M had violated Technical
Specification
5.4.1 because CNP EOPs could not ensure 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 3 failed on several occasions
to declare the SG PORVs unavailable
after taking a CAC out of service for maintenance.
Reference
1 characterized
the NCVs as representing
a more-than-minor, cross-cutting
performance
deficiency
involving
areas of human performance, the component
of decisionmaking, 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 accident accompanied
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 that CNP's licensing
basis did not assume a single failure of a non-safety-related
component (in particular, the unaffected
unit's PAC), during an SGTR event, Reference
1 contends that I&M had nevertheless
failed to demonstrate
that control air would reasonably
be available
during an SGTR event accompanied
by a multi-unit
LOOP. Similarly, Reference
1 asserts that even if the unaffected
unit's PAC would be available
during a design basis SGTR accident, I&M had failed to identify that assumption
within its SGTR accident analysis, and the NRC Staff had never explicitly
approved that assumption.
Further, Reference
1 endorsed Reference
5's interpretation
of the UFSAR's use of the term LOOP to refer to multi-unit
events, adding that the absence of CNP operating
procedures
preventing
alignment
of the same offsite power sources to 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 Valves In accordance
with Reference
6 (at Sections 10.2.2 and 14.2.4), the SG PORVs prevent overpressure
conditions
in the steam generators
by releasing
secondary
system steam to atmosphere
following
a loss of condenser
vacuum. The SG PORVs form part of the main steam system pressure boundary, and thus are safety-related
equipment
for main steam system pressure retention.
CNP operating
procedures
prescribe
operator actions in the event of a SGTR accident.
CNP operating
procedures
allow SG PORVs to be operated using motive force provided by control air supplied by either the compressed
air system shared between the two units, control air pressure supplied by a unit-specific
CAC, or installed
backup nitrogen tanks that can be aligned to the SG PORVs. In March 2013, I&M completed
installation
of a plant modification
and revised its operating
procedures
to ensure that the backup nitrogen tanks are immediately
and automatically
available
during an SGTR accident without the need for manual valve manipulation
outside the control room.b. CNP Compressed
Air System Section 9.8.2 of Reference
6 describes
the control air provided by CNP's compressed
air system 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 shared between both units at CNP. Each unit at CNP contains one CAC capable of providing
control
Enclosure
2 to AEP-NRC-2013-53
Page 4 air only within that unit, as well as a PAC capable of providing
control air to both units via a shared header. Both units share a single backup air compressor
capable of providing
control air to loads within either unit.During normal operations, control air pressure for operating
both units' SG PORVs is provided by one of the two PACs. Low pressure in the shared plant compressed
air header will result in the automatic
start and loading of the other unit's PAC. Low control air header pressure in one of 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 power transformers.
The CAC associated
with each unit at CNP can be powered by either offsite power source in normal operations, but can only receive power from its unit's CD EDG after offsite power has been lost to that unit. The CACs and PACs are both non-safety
related equipment
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, both PACs 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
of the operating
unit's CAC when the opposite unit is shutdown and the shutdown unit's PAC is aligned to preferred
offsite power.4. Regulatory
Basis for the Assumption
of Only a Single-Unit
LOOP within CNP's SGTR Accident Analysis a. CNP's Licensing
Basis Has from the Beginning
Assumed that an SGTR Accident Would Involve a Coincident, Single-Unit
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 Sections 14.2.4 and 14.2.7 of the FSAR for Units 1 and 2 dated February 2, 1971 (Reference
10), CNP's original licensing
basis evaluated
the radiological
consequences
of an SGTR accident by conservatively
estimating
the mass release of radioactivity
to the environment
over the 30-minute
time span between SGTR accident initiation
and subsequent
termination
of primary to 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 was considered
inherently
conservative
because it neglected
the reduction
in mass flow that would occur during this same time period.
Enclosure
2 to AEP-NRC-2013-53
Page 5 Inherent in that postulated
30-minute
mass release was an assumption
of the success of operator actions such as the operation
of SG PORVs to mitigate the event. Section 14.2.4 of Reference
10 in several places explicitly
credited the availability
of SG PORVs during a design basis SGTR regardless
of conditions.
Reference
10's evaluation
of SGTR accidents
omits any mention of the possibility
that compressed
air system components
could be unavailable
as a result of a single failure or maintenance, as it prefaced its elaboration
of the sequence of events initiated
by an SGTR event by stating that its analysis had "assum[ed]
normal operation
of the various plant control systems ....... Reference
10 at Section 14.2.4. Further, Reference
10 assumed that SG PORVs would remain available
regardless
of the status of offsite power, stating that when a unit was "without offsite power": Condenser
bypass valves will automatically
close and the steam generator
pressure will rapidly increase resulting
in steam discharge
to the atmosphere
through the steam generator
safety valves and/or the power 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 valves would automatically
close to protect the condenser.
The steam generator pressure would rapidly increase resulting
in steam discharge
to the atmosphere
through the steam generator
safety and/or power operated relief valves.Reference
10 at Section 14.2.4 (emphasis
added).I&M's assumption
that SG PORVs remained available
for mitigation
of an SGTR accident is consistent
with the description
of the compressed
air system elsewhere
within CNP's original FSAR. Among the design bases for CNP's compressed
air system within Reference
10 is a requirement
for continued
availability
of control air: The [compressed
air system] must provide a continuous
supply of compressed
air to vital systems under both normal and abnormal conditions.
Reference
10 at Section 9.8.2 (emphasis
added). With this in mind, each of CNP's PACs were designed to be "capable of supplying
the entire demand of both plant and control-instrument
air requirements
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 system components
within the opposite plant, the staggered
construction
and licensing
of CNP Units 1 and 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 the design 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 6 other unit, the fuel oil tank's capacity is based on the design operational
requirements
of a single EDG. Reference
6 at Section 8.4. Consequently, references
within Reference
10's SGTR accident analysis to a "loss of offsite power" or a "station blackout" referred to an event involving
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 in the 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
the availability
of SG PORVs -and, in fact, arguably placed even greater emphasis on the continued
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 EOPs identifying
the SG PORVs (and not the steam generator
safety valves) as the initial means of preventing
overpressure
after loss of offsite power: In the event of a coincident
station blackout, the steam dump valves would automatically
close to protect the condenser.
The steam generator pressure would rapidly increase, resulting
in steam discharge
to the atmosphere
through the steam generator
power operated relief valves (and the steam generator
safety valves if their setpoint had been reached).Reference
12 at Section 14.2.4 (emphasis
added). Later UFSAR revisions
to CNP's SGTR accident analysis also incorporated
the original FSAR's language describing
the continued availability
of SG PORVs despite a LOOP or station blackout virtually
unchanged.
Reference
6 at Section 14.2.4. Further, I&M's review of pertinent
docketed correspondence
with the NRC Staff 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
for SGTR Accidents
in Docketed Correspondence
On October 24, 2000, I&M submitted
a license amendment
request (LAR, Reference
10) to revise the methodology
used in designing
CNP EOPs during a design basis SGTR accident.The Westinghouse
Owners Group methodology (WCAP-10698-P-A
("SGTR Analysis Methodology
to Determine
Margin to Steam Generator
Overfill"))
that I&M proposed to adapt for use within its SGTR accident analysis incorporated
lessons learned from operational
experience, plant simulator
studies, and advances in computer modeling techniques
to better characterize
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 SGTR accident.
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 analysis assumptions, I&M could revise CNP EOPs to assure margins to steam generator
overfill while remaining
within the conservative
margins to radiological
consequences
described
in its original SGTR accident analysis.
Enclosure
2 to AEP-NRC-2013-53
Page 7 Although the NRC had previously
accepted WCAP-10698-P-A
for use by licensees, the NRC Staff had to evaluate its application
within CNP's SGTR accident analysis.
In a series of docketed correspondence
with the NRC Staff detailing
how the WCAP-10698-P-A
would be used within CNP's SGTR accident analysis, I&M repeatedly
emphasized
that the new methodology
would not disturb existing license basis assumptions
in its SGTR accident analysis.
Specifically, the safety analysis for I&M's LAR noted that: The proposed change ...does not affect any accident initiators
or precursors
.... The proposed change also does not affect the ability of operators
to mitigate the consequences
of an accident.Reference
13, Attachment
1 at Page 4 (emphasis
added). I&M repeated this claim in the LAR's evaluation
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 of affected plant systems, involve a physical alteration
to the systems, or change the way in which systems are currently
operated, such that previously
unanalyzed
SGTRs would not occur. The change to incorporate
the WCAP-10698-P-A
methodology
does not introduce
any new malfunctions
....Reference
13, Attachment
2 at Pages 2-3 (emphasis
added).Subsequent
docketed correspondence
between I&M and the NRC Staff was even more explicit in describing
the retention
of existing license basis assumptions
for SGTR accidents.
In a June 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 its SGTR 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 for calculating
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 the items listed in that table were the "air-operated" SG PORVs, which the notes accompanying
the table stated were themselves
safety-grade
components
because they "form part of the main steam 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 Pages 3-4. 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 not involve 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 includes a coincident
loss of offsite power", Reference
14 noted that the modified SGTR analysis would retain CNP's original licensing
assumption
that SG PORVs would remain available
despite the fact that "offsite power [was] not ...available." Reference
14, Attachment
1 at Page 4.
Enclosure
2 to AEP-NRC-2013-53
Page 8 Reference
14 contained
no suggestion
of a change in the scope of the LOOP assumed within CNP's SGTR accident analysis.By letter dated October 24, 2001 (Reference
4), the NRC Staff approved I&M's LAR in modified form to accommodate
CNP's existing licensing
basis assumptions
for SGTR accidents.
In the SER submitted
with its approval of I&M's LAR, the NRC Staff acknowledged
that licensees
like I&M could not incorporate
the WCAP-10698-P-A
methodology
within their SGTR accident analysis in a uniform fashion because "variations
in plant designs prevent a single model from adequately
representing
all Westinghouse
Plants." Reference
4, SER at Page 2.Consequently, the NRC Staff devoted much of the SER to evaluating
the differences
between the generic WCAP-1 0698-P-A methodology
and I&M's proposed approach for incorporating
that methodology
within its licensing
basis.The NRC Staff noted that in the immediate
case, those differences
included I&M's intention
of retaining
CNP's existing assumptions
for SGTR accidents:
To implement
the WCAP, the licensee used the LOFTTR2 computer code and the plant-specific
current licensing
basis assumptions.
Reference
4, SER at Page 2 (emphasis
added). The NRC Staff explicitly
acknowledged
that CNP's licensing
basis assumptions
credited certain systems and components, including
the SG PORVs and their control air appurtenances, as remaining
available
for mitigation
of an SGTR accident: The licensee provided a list of systems, components, and instrumentation
that are used for SGTR accident mitigation.
They also specified
the safety classification
of the systems and power sources. However, the licensee listed several systems used for SGTR mitigation
that are not safety related and do not have safety related backups. The licensee justified
the use of the non-safety-related
equipment
by stating that these systems are credited in the current UFSAR Section 14.2.4 accident analysis.
Upon review of Section 14.2.4, the staff concludes
that the licensing
basis SGTR analysis does credit limited use of non-safety
grade equipment
for mitigating
the SGTR.Reference
4, SER at Page 3. Similarly, the NRC Staff acknowledged
that CNP's licensing
basis did 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
by the WCAP-10698-P-A
safety analysis, and did not provide it's [sic] effect on the margin to overfill.
The licensee based their decision not to assume the worst single failure on the fact that their current licensing
basis does not include a single failure.Reference
4, SER at Page 4. Further, the SER nowhere mentions that I&M intended to discard CNP's existing assumption
of a coincident
single-unit
LOOP during an SGTR accident, or that
Enclosure
2 to AEP-NRC-2013-53
Page 9 the LOOP assumed within the WCAP-10698-P-A
methodology
supplanted
CNP's existing licensing
basis assumptions
for SGTR accidents.
Although I&M's proposed retention
of CNP's existing licensing
basis assumptions
for SGTR accidents "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
the LOFTTR2 code into its licensing
bases for CNP and can use the LOFTTR2 code, with the current licensing
basis assumptions
as inputs for the overfill analysis of steam generator
tube rupture accidents.
This change to the licensing
basis does not affect accident initiators
or precursors.
This change also does not ...decrease the ability of the operators
to mitigate the consequences
of an accident.Reference
4, SER at Page 5 (emphasis
added). In justifying
its approval of a modified WCAP-10698-P-A
methodology
for use at CNP, the NRC Staff noted that I&M's adaptation
of the WCAP-10698-P-A
methodology
to CNP's existing licensing
basis assumptions
for SGTR accidents
did not affect conservative
estimates
of the radiological
consequences
of a design basis SGTR at CNP. Reference
4, SER at Page 3.I&M's subsequent
review of docketed correspondence
with the NRC Staff has identified
no further changes to CNP's licensing
basis assumptions
regarding
the availability
of SG PORVs in an SGTR accident, the absence of a single failure assumption
within CNP's SGTR accident analysis, 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 Reading of 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 SGTR Accidents
Does Not Address Pertinent
Docketed Correspondence
As noted earlier, the NCVs within Reference
1 are based on the NRC Staffs contention
that the coincident
LOOP assumed within CNP's licensing
basis SGTR accident analysis involves a loss of offsite power to both units at CNP. The NRC Staff's position is based on a single argument within Reference
5: that it follows from the use of the terms "LOOP" and "station" in a handful of CNP UFSAR sections, some of which are unrelated
to SGTR accident analysis, that a LOOP can 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.
The first UFSAR passage referenced
in Reference
5 comes from Section 1.3.7 describing
the auxiliary
electrical
system for each of the two units at CNP: Donald C. Cook's UFSAR Section 1.3.7, "Electrical
System" states, "The main generators
are 1800 rpm, Phase III, 60 cycle, hydrogen and water
Enclosure
2 to AEP-NRC-2013-53
Page 10 cooled units. The main transformers
deliver generator
power to the 345kV and 765 kV switchyards.
The station auxiliary
power system consists of auxiliary
transformers, 4160V and 600 V switchgear, 600V motor control centers, 120 V A-C vital instrument
buses and 250 V D-C buses." Reference
5 at Page 3 (emphasis
supplied by NRC Staff). Based on the fact that UFSAR Section 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 to each unit individually.
In the same vein, Reference
5 cites a passage from Section 1.3.8 of the UFSAR describing
the Safety Features associated
with each unit at CNP: Also, Section 1.3.8, "Safety Features," describes
the safety features incorporated
into the design of the plant, including
the fact that "even if external auxiliary
power to the station is lost concurrent
with an accident, power is available
for the engineered
safeguards
from on-site diesel generator
power to assure protection
of the public health and safety for any 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 the passage'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 (at Section 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 complete loss of the offsite grid accompanied
by a turbine trip at the station, or by a loss of the on-site AC distribution
system." Reference
5 at Page 4. The NRC Staff read this reference
to a "complete
loss of offsite grid accompanied
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
basis for every event evaluated
in UFSAR Section 14. Reference
5 at Page 4. Based on these examples, Reference
5 reports that NRR concurred
with NRC Staff that had performed
the CDBI that the LOOP assumed in CNP's SGTR analysis was a "station event, not a unit specific event." Reference
5 at Page 4.The NRC Staff's position and the UFSAR passages described
above represent
the only basis identified
by the NRC Staff for its position throughout
the multiple docketed communications
and meetings with I&M since the CDBI began in July 2012. The NRC Staff has identified
no regulatory
provisions
or policy guidance requiring
the assumption
of a LOOP affecting
both units for 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 within CNP's UFSAR supporting
its position.
Enclosure
2 to AEP-NRC-2013-53
Page 11 Further, the NRC Staff has yet to provide a meaningful
response to the analysis provided by I&M in References
3 and 7 in support of its understanding
of CNP's licensing
basis assumptions.
Reference
5 does not specifically
address the SGTR accident analysis assumptions
identified
within docketed correspondence
highlighted
within Reference
3: The scope of this TIA was limited to the licensing
basis as related to offsite power only. The staff did not evaluate other assertions
in the licensee's
white paper.Reference
5 at Page 4.1 Reference
2 merely repeated Reference
5's claims regarding
CNP's licensing
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 SGTR accident licensing
basis assumptions
advanced by I&M in References
3 and 7, a careful reading of the bases identified
in Reference
1 indicates
that the NRC Staff's reasoning
is circular in that it depends on, rather than proves the assumption
of a multi-unit
LOOP in CNP's SGTR accident analysis.
Specifically, in acknowledging
I&M's position that CNP's licensing
basis had never assumed a single failure of a non-safety-related
component (specifically
the unaffected
unit's PAC) during an SGTR event, Reference
1 contends that I&M had nevertheless
failed to demonstrate
that an unaffected
unit's PAC would reasonably
be available
during an SGTR accident affecting
one unit: The inspectors
agreed that certain older operating
plants are credited with the use of non-safety
related equipment
to mitigate events. 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 with I&M's observation
in Reference
7 that the original SER for Unit 1 did not consider that a CAC would be out of service for maintenance
pursuant to an assumed single failure, claiming that this demonstrates
that a CAC would have to be available
to supply control air pressure during a design basis SGTR accident, as its availability
would be a limiting condition
in CNP's SGTR accident analysis.However, the above arguments
do not prove the NRC's Staff understanding
of the scope of the LOOP 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 be reasonably
available
and bounded by a safety-related
path for main steam system pressure retention
purposes, regardless
of the status of the CAC on the affected unit. Similarly, the availability
of the affected unit's CAC is not a limiting condition
for CNP's SGTR accident analysis if the coincident
LOOP affects only the unit experiencing
the SGTR event such that the 1 The NRC Staff has not docketed correspondence
between Region III personnel
and NRR personnel
defining the scope of NRR personnel's
review of the competing
interpretations
of CNP's licensing
basis assumptions
for the LOOP assumed within CNP's SGTR design basis accident analysis.
Enclosure
2 to AEP-NRC-2013-53
Page 12 PAC on the unaffected
unit remains available
to provide control air pressure to the affected unit's SG PORVs. Lastly, the NRC Staff statement
quoted above is inconsistent
with the NRC Staff's statements
within Reference
4 endorsing
CNP licensing
basis assumptions
crediting
the availability
air system components
during an SGTR accident.b. The NRC Staff's Position Is Unsupported
by a Fair Reading of the UFSAR The NRC Staff's categorical
statement
that every reference
to a LOOP within CNP's UFSAR can be understood
to refer to an event denying offsite power to one or both units at CNP is unsupported
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 a single 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
any reference
to a LOOP to refer to either a single or multi-unit
event. Consequently, whether a particular
reference
to a LOOP within CNP's UFSAR refers to a LOOP affecting
one or both units at CNP must be determined
by reference
to a number of factors such as the text surrounding
the UFSAR's reference
to the LOOP, the larger structure
of CNP's UFSAR, as well as the relevant historical
and regulatory
background.
i. The NRC Staff's Understanding
of the Scope of a LOOP Is Not Supported
by the Surroundinq
Text A comparison
of the different
contexts in which the term LOOP appears within CNP's SGTR and Loss of All AC Power to the Plant Auxiliaries
accident analyses, respectively, does not support the NRC's generic interpretation
of the term. As noted earlier, the NRC Staff's understanding
of CNP's licensing
basis is based on the potentially
broad scope of the LOOP within UFSAR Unit 1 Section 14.1.12, "Loss of All AC Power to the Plant Auxiliaries." The UFSAR's description
of the 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 of power 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 AC distribution
system.Reference
5 at Page 4 (quoting UFSAR Unit 1, Section 14.1.12.1) (emphasis
added). Because the context of the UFSAR cited above passage is on its face ambiguous
regarding
the number of units at CNP affected by the LOOP, the NRC Staff contends that it could, based only on a generous reading of the cited text alone, be read to refer to a LOOP to one or both units at CNP.The context surrounding
the use of the term LOOP within the SGTR accident analysis in UFSAR Units 1 and 2 Section 14.2.4 demands an entirely different
conclusion
regarding
the number of units losing offsite power in a LOOP. Here, the UFSAR's use of the term LOOP is not qualified
by the broad adjectives, complete loss, all power, the offsite grid, etc., used in the earlier accident analyses in a way that could arguably suggest a LOOP denying power to both units; rather, CNP's SGTR accident analysis refers only to "offsite power", or "a loss of offsite power" or "a coincident
loss of offsite power." Reference
6 at Section 14.2.4.
Enclosure
2 to AEP-NRC-2013-53
Page 13 ii. The NRC Staffs Understandinq
of the Meaninq of a LOOP Is Inconsistent
with the Structure
of the UFSAR also undercuts
the generic meaning attached to the term LOOP by the NRC Staff. According
to Reference
5, the potentially
broad scope of the LOOP described
in UFSAR Section 14.1.12 defines the meaning of the term throughout
the UFSAR. Reference
5 at 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
for generic application
throughout
the UFSAR than the more limited-scope
LOOP described
within other 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
the form and content of CNP's UFSAR. Consistent
with the scheme laid out in Regulatory
Guide 1.70 (Reference
events and accidents
satisfying
a minimal threshold
for best-estimate
frequency
of occurrence, which are then assigned a frequency
grouping based on criteria established
by the American Nuclear Society (ANS). As stated in UFSAR Sections 14.0, ANS Condition
1 (normal operational
are omitted from CNP's UFSAR, while Condition
2 events (moderate
frequency)
appear mostly in UFSAR Sections 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
Guide 1.70, CNP's UFSAR analyzes each of the events within the UFSAR individually
and for each unit, 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
the threshold
for consideration
of transient
events and accidents
within an FSAR requires a minimal best-estimate
frequency
of occurrence
of >l.OE-6/yr.
Reference
16 at 6. However, when the NRC 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 ANS guidance;
further, when accounting
for the risk that a CAC would be unavailable
for maintenance
for 30 days, the best-estimate
frequency
of occurrence
fell below (1.75E-7/yr)
the ANS 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 to conclude that a multi-unit
LOOP is too remote an event to be considered
in CNP's design basis SGTR analysis.
According
to Reference
17, there was not one reactor trip coincident
with a multi-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 an SGTR with a coincident, multi-unit
LOOP yields a best-estimate
frequency
of occurrence
of 6.33E-7/yr
-below the ANS threshold
for consideration
within CNP's UFSAR. Further, the best-estimate
frequency
of occurrence
is even lower (1.91 E-8) when accounting
for the risk that a 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 UFSAR Section 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 Plant Auxiliaries
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 occur much less frequently
than a single-unit
LOOP, application
of the NRC Staff's reading of the scope of the term LOOP within CNP's SGTR analysis represents
a significant
change in the initial assumptions
and anticipated
frequency
for that particular
accident.
That revised frequency
of CNP's design basis SGTR accident could conceivably
require the assignment
of new 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 Section 14.2.4), which in turn would require the re-organization
of CNP's UFSAR. Consequently, the NRC Staff's position does not account for the significance
attached by NRC guidance to the distinction
between different
ANS Conditions
and (by extension)
types of design basis events or accidents.
The NRC Staff's references
to the use of the word "station" within the UFSAR's description
of CNP systems is similarly
not helpful for determining
the scope of the LOOP assumed in CNP's SGTR accident analysis.
In support of its contention
that every use of the term LOOP refers to either a single or multi-unit
event, Reference
5 points to a handful of examples of the UFSAR's use of the word "station" in descriptions
of CNP Electrical
System (at Section 1.3.7) and Safety Features (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 other terms (e.g., LOOP) throughout
the UFSAR. Regulatory
Guide 1.70 understood
the system descriptions
within the first section of a licensee's
UFSAR to be distinct from the accident analyses described
in a later section of the UFSAR: The first chapter of the SAR should present an introduction
to the report and a general description
of the plant. This chapter should enable the reader to obtain a basic understanding
of the overall facility without having 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 analysis assumptions
not by reference
to surrounding
text, but by reference
to language in an entirely different
UFSAR section. The NRC Staff's more fluid distinction
between UFSAR sections is difficult
to reconcile
with the approach endorsed within Regulatory
Guide 1.70.Although the NRC Staff in Reference
1 states that the difference
between UFSAR sections identified
above supports its understanding
of CNP's licensing
basis, the NRC Staffs position is erroneous.
Conceding
that high-level
system descriptions
within Section 1 of CNP's UFSAR do not prescribe
accident analyses assumptions
within subsequent
UFSAR sections, the NRC Staff incorrectly
asserts that:
Enclosure
2 to AEP-NRC-2013-53
Page 15 This argument supports the inspectors'
position that the licensee cannot take credit for the unaffected
unit's non-safety-related
PAC unless explicitly
approved by the NRC and described
in the SGTR analysis.Reference
1, Enclosure
at Page 5 (emphasis
added). Notwithstanding
the fact the language within Section 1 of CNP's UFSAR is unhelpful
for interpreting
language describing
UFSAR accident analysis assumptions, it does not follow that Section l's high-level
description
of the components
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 have to describe CNP systems and components
multiple times. Reference
15 at Page 1-1. Because Section 1.3.9.h of CNP's UFSAR describes
CNP's compressed
air system as a shared system of 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 PORVs from CNP's control air appurtenances
and equipment
permits I&M to take credit for the unaffected
unit's PAC in CNP's SGTR accident analysis.
Further, by the NRC Staff's logic, I&M would not be able to take credit for the operation
of any CAC or PAC within CNP's SGTR accident analysis, as neither of those components
is explicitly
mentioned
in the UFSAR's SGTR accident 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 1.3.7 states: "The station auxiliary
power system consists of auxiliary
transformers, 4160 v and 600 v switchgear, 600 v motor control centers, 120 v-a-c vital instrument
buses and 250 v d-c buses." However, the NRC Staffs suggestion
that the term "station" in this context necessarily
refers to both 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
buses and 250 v-d-c buses, and 4160 v and 600 v switchgear)
the NRC Staff suggests represents
a shared system between CNP units. Similarly, both units have the EDGs and turbines mentioned
in the cited passage from UFSAR Section 1.3.8. Further, the NRC Staff's claim that the use of the term "station" within Section 1.3.8's description
of CNP Safety Features proves that there is only one, shared auxiliary
power system at CNP is at odds with surrounding
text not examined by the NRC Staff. Specifically, UFSAR Section 1.3.9, "Shared Facilities
and Equipment," 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 from Section 1.3.9's list of shared systems and equipment.
iii. The NRC Staff's Understanding
of the Term LOOP Is at Odds with the Reaulatorv
History of CNP and Similarlv-Situated
Facilities
Enclosure
2 to AEP-NRC-2013-53
Page 16 The NRC Staff's understanding
of the term LOOP also does not account for docketed correspondence
acknowledging
the retention
of the assumptions
within CNP's original SGTR accident 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's understanding
of the scope of the term LOOP assumed within CNP's SGTR accident analysis not 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 SGTR accident 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 its operating
license several years before Unit 2 (1974 as opposed to 1977). Consequently, the SGTR 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 the design 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 those licensees
understand
the licensing
basis assumptions
for units at their facility to assume only a single-unit
LOOP during SGTRs and other accidents.
Further, among those licensees
whose licensing
basis currently
assumes multi-unit
LOOPs were some who acknowledged
that their current 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 offsite power reflect the unit-specific
approach to electric system design within licensing
basis accident assumptions
at CNP and other similarly-situated
facilities.
Most prominently, the current Station Blackout Rule at 10 CFR 50.63 (Reference
8) is unit-specific
in its approach to the availability
of AC power, including
offsite power. Although the NRC has recently published
a Federal Register notice (Reference
18 at 16179) indicating
a desire to revise its Station Blackout Rule and other regulations
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 be premature, regardless
of whether the NRC Staff can (as Reference
1 asserts) conceive of scenarios
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 Either Design Basis Events or Beyond-Design
Basis External Events NRC Orders issued following
the earthquake
and tsunami at the Fukushima
Dai-ichi nuclear power plant in March 2011 acknowledge
that existing defense-in-depth
approaches
at licensed facilities
provide adequate protection
of public health and safety against design basis accidents.
Specifically, EA-12-049
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 17 mitigation
features to prevent radioactive
releases should an accident occur; and (3) emergency
preparedness
programs that include measures such as sheltering
and evacuation
.... These defense-in-depth
features are embodied in the existing regulatory
requirements
and thereby provide adequate protection
of the public health and safety.Reference
19 at Page 5 (emphasis
added). Compliance
with those NRC requirements, the NRC concluded, "presumptively
assures adequate protection" of public health and safety from inadvertent
release of radioactive
materials
during a design basis accident.
Reference
19 at Pages 4-5.As explained
at length earlier, the NRC Staff's contention
within Reference
1 that CNP is not in compliance
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 SGTR accident involves both units at CNP, and the NRC Staff has presented
no meaningful
evidence in support of a contrary position.
Further, as recently as 2001, the NRC Staff endorsed the measures (including
the crediting
of the continued
availability
compressed
air system components)
I&M employs for mitigating
the risk of inadvertent
release of 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
in compliance
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 SGTR accidents.
Specifically, I&M in March 2013, completed
installation
of a plant modification
and revised CNP operating
procedures
to ensure that backup nitrogen tanks are immediately
and automatically
available
during an SGTR 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 the operating
unit's CAC when the opposite unit is shutdown 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 SGTR accidents
at CNP. As noted earlier, the most recent published
industry data on the frequency
of LOOPs 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 ANS guidance (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 the LOOP accompanying
a design basis SGTR accident is so small (2.4E-8/yr)
as to provide no meaningful
advantage
over I&M's understanding
of CNP's licensing
basis for assuring adequate protection
of public health and safety. Reference
1, Enclosure
at Page 1. Further, even this marginal difference
in core damage frequency
between I&M's and the NRC Staff's positions
is likely overstated, as the core damage frequency
calculation
within Reference
1 (Enclosure
at Pages 6-7) does not account for the additional
defense-in-depth
measures implemented
Enclosure
2 to AEP-NRC-2013-53
Page 18 Lastly, the NRC Staff has provided no basis to conclude that I&M has failed to provide adequate protection
against beyond-design
basis scenarios
involving
an SGTR accompanied
by a coincident, 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 to mitigate the consequences
of beyond-design
basis external events, such as those resulting
in the 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 Nuclear Energy 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
offsite power) 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 events at CNP. I&M's Overall Integrated
Plan incorporates
the FLEX strategy endorsed by the NRC Staff 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 such as 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 Merit In 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
of decision making, and the aspect of conservative
assumptions.
Reference
1 Enclosure, at Pages 1 and 2. The NRC Staff stated that the NCVs involved cross-cutting
aspects because I&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 now understands
CNP's licensing
basis to assume that an SGTR accident would be accompanied
by a 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, those NCVs are based on an erroneous
understanding
of the scope of the coincident
LOOP within CNP'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 an SGTR event. Consequently, the unaffected
unit's PAC will remain available
to provide control air 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 SER within Reference
4 endorsed I&M's claims regarding
the continued
availability
of control air to operate 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
that those NCVs involve cross-cutting
aspects is similarly
incorrect.
Additionally, even if the NRC Staff's current understanding
of CNP's licensing
basis were correct, 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 nowhere explains how I&M's requirements
were inconsistent
with reactor safety and public health. As noted 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 19 changes to CNP's licensing
basis proposed by I&M in its 2000 LAR would not increase the risk or consequences
of an SGTR accident beyond the conservative
estimates
within CNP's original licensing
basis. In arriving at this conclusion, the NRC Staff explicitly
noted that I&M had revised 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 in Reference
1 (Enclosure
at Page 1) is consistent
with the NRC Staffs conclusions
within Reference
4, as the difference
in core damage frequency
from assuming a dual-unit
LOOP is only 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, the performance
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 are based on an understanding
of CNP's licensing
basis that has been in place since the original licensing
of Unit 1 at CNP around forty years ago, and which was endorsed by the NRC Staff as recently 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 more than 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
23 at Page 3. Reference
1 provides no justification
for why the NCVs represent
a "rare or unusual case" warranting
application
of this exception.
Further, as explained
above, I&M's understanding
of its licensing
basis is not rare or unusual; in fact, multiple plants of similar vintage and configuration
have the same licensing
basis assumptions
regarding
the scope of a LOOP during an SGTR or other accident.8. Conclusion
For the reasons identified
above, both the NCVs identified
within Reference
1 and the NRC Staff's determination
that those NCVs involve cross-cutting
aspects are incorrect.
Enclosure
2 to AEP-NRC-2013-53
Page 20 REFERENCES:
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
05000316/2013030," 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
05000316/2012007," dated January 11, 2013.3. Letter from W. Hodge, I&M, to C. Tilton, NRC, "D. C. Cook CDBI Response to Question 2012-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)," dated October 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 Steam Generator
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, dated March 17, 2012.7. Letter from I&M to Ann Marie Stone and Caroline Tilton, NRC, "Response
to NRC Inspection
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, dated February 2, 1971.11. Amendments
to Donald C. Cook Nuclear Plant Final Safety Analysis Report for Units 1 and 2, dated November 11, 1977.12. Amendments
to the Donald C. Cook Nuclear Plant Final Safety Analysis Report for Units 1 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 Changes in Steam Generator
Tube Rupture Analysis Methodology," dated October 24, 2000.
Enclosure
2 to AEP-NRC-2013-53
Page 21 14. 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 in Steam Generator
Tube Rupture Analysis Methodology (TAC Nos. MB0739 and MB0740)," dated June 29, 2001.15. NRC Regulatory
Guide 1.70, "Standard
Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, " dated November 1978.16. American Nuclear Society, ANSI/ANS-51.1-1983, "Nuclear Safety Criteria for the Design of 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:
Station Blackout," dated March 19, 2012.19. NRC Order Number EA-12-049, "Order Modifying
Licenses with Regard to Requirements
for Mitigation
Strategies
for Beyond-Design-Basis
External Events," dated March 12, 2012.20. NRC Interim Staff Guidance JLD-ISG-2012-01, "Compliance
with Order EA-12-049, Order Modifying
Licenses with Regard to Requirements
for Mitigation
Strategies
for Beyond-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 2 Overall 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," dated January 24, 2013