AEP-NRC-2013-53, Response to the Non-Cited Violations Resulting from Component Design Bases Inspection 05000315/2013010; 05000316/2013010

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Response to the Non-Cited Violations Resulting from Component Design Bases Inspection 05000315/2013010; 05000316/2013010
ML13224A246
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 08/02/2013
From: Gebbie J
Indiana Michigan Power Co
To:
Document Control Desk, NRC/RGN-III
References
AEP-NRC-2013-53 IR-13-010
Download: ML13224A246 (25)
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Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, MI 49106 A unitof American Electric Power Indiana Michigan Power.com August 2, 2013 AEP-NRC-2013-53 10 CFR 2.201 Docket Nos.: 50-315 50-316 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), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response contesting the Non-Cited Violations (NCVs) documented by Reference 4, Component Design Bases Inspection (CDBI) Report 05000315/2013010; 05000316/2013010.

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. 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 ( IIO{ 2

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 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 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 GeneratorPower OperatedRelief 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 CompressedAir 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 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.

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 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.

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 operatorsto 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, 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.

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 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 PertinentDocketed 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 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.

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, ratherthan 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.

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 FairReading 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 Understandingof 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 distributionsystem.

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.

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 to AEP-NRC-2013-53 Page 14 (Loss of All AC Power to the Plant Auxiliaries) automaticallycarry 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:

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-SituatedFacilities 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) 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 regulatoryrequirements 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.

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 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.

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.

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
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