ML13224A246

Donald C. Cook, Units 1 and 2 - Response to the Non-Cited Violations Resulting from Component Design Bases Inspection 05000315/2013010; 05000316/2013010

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

05000315/2013010;

05000316/2013010

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

05000315/2012007;

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

05000315/2013010;

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;

05000316/2013010.

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

of steam generator

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

of SG PORVs). Reference

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

05000315/2013010;

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

05000315/2013010;

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

of CNP's UFSAR and providing

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

NCVs of CNP Technical

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

a. CNP Steam Generator

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

LOOP CNP'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 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

steam generator

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

steam generator

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

of SG PORVs and compressed

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 CNP's UFSAR 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

15), CNP's UFSAR evaluates

transient

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

transients)

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

of SG PORVs and supporting

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

at CNP since the 2012 CDBI.

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

05000315/2013010;

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

05000315/2012007;

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