View source for AEP-NRC-2015-07, Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, Containment Leakage Rate Testing Program, to Use NEI-94-01, Revision 3-A..

{{Adams
| number = ML15020A662
| issue date = 01/15/2015
| title = Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, Containment Leakage Rate Testing Program, to Use NEI-94-01, Revision 3-A..
| author name = Gebbie J
| author affiliation = Indiana Michigan Power Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000315, 05000316
| license number = 
| contact person = 
| case reference number = AEP-NRC-2015-07
| document type = Letter type:AEP
| page count = 12
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:z INDIANA                                                                              Indiana Michigan Power MICHIGAN                                                                             Cook Nuclear Plant
  *VPIER                                                                             One Cook Place Bridgmnan, MI 49106 A unitofAmerican Electric Power                                                       Indiana Michigan Power.com January 15, 2015                                                                AEP-NRC-2015-07 10 CFR 50.90 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0

==References:==
: 1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," AEP-NRC-2014-09, dated March 7, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14071A435.
: 2. Email from T. A. Beltz, NRC, to H. L. Etheridge, I&M, "Draft Requests for Additional Information Mechanical & Civil Engineering Branch of the Office of Nuclear Reactor Regulation Regarding a License Amendment Request for the Donald C. Cook Nuclear Plant, Units 1 and 2, to Revise Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" Indiana Michigan Power Company Docket Nos. 50-315 and 50-316 (TAC Nos. MF3568 and MF3569),"
dated August 25, 2014.
: 3. Email from T. A. Beltz, NRC, to H. L. Etheridge, I&M, "Draft Requests for Additional Information Mechanical & Civil Engineering Branch of the Office of Nuclear Reactor Regulation Regarding a License Amendment Request for the Donald C. Cook Nuclear Plant, Units 1 and 2, to Revise Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" Indiana Michigan Power Company Docket Nos. 50-315 and 50-316 (TAC Nos. MF3568 and MF3569),"
dated August 26, 2014.
: 4. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0,"
AEP-NRC-2014-73, dated September 30, 2014, ADAMS Accession No. ML14275A454.
vO\7

U. S. Nuclear Regulatory Commission                                             AEP-NRC-2015-07 Page 2
: 5. Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Supplemental Response to a Request for Additional Information Regarding the License Amendment Request to Revise Technical Specification 5.5.14, "Containment Leakage Rate Testing Program," to use NEI-94-01, Revision 3-A as Regulatory Guidance Versus the Current NEI 94-01, Revision 0," AEP-NRC-2014-100, dated December 16, 2014, ADAMS Accession No. ML14352A232.
: 6. Email from M. L. Chawla, NRC, to H. L. Etheridge, I&M, "Request for Additional Information -
Donald C. Cook Units 1 and 2 - LAR - To revise Technical Specification 5.5.14, "Containment Leakage Rate34 Testing Program" - MF3568 and MF3569," dated November 20, 2014.
This letter provides Indiana Michigan Power Company's (I&M), the licensee for Donald C. Cook Nuclear Plant Units 1 and 2, response to a Request for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding a license amendment request to change Technical Specification (TS) Section 5.5.14, "Containment Leakage Rate Testing Program."
By Reference 1, I&M submitted a request to amend the TS to Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to change TS 5.5.14, "Containment Leakage Rate Testing Program,"
to use Nuclear Energy Institute (NEI) 94-01, Revision 3-A as regulatory guidance versus the current NEI 94-01, Revision 0. By Reference 2, the NRC transmitted RAIs (RAI-EMCB-1,-2,-3, and -4) regarding the proposed amendment. By Reference 3, the NRC transmitted an additional RAI (RAI-SCVB-1) regarding the proposed amendment. References 4 and 5 provided I&M's response to References 2 and 3. By Reference 6, the NRC transmitted an additional RAI regarding the proposed amendment. This letter provides I&M's response to Reference 6. Enclosure 1 to this letter provides an affirmation statement. Enclosure 2 to this letter provides I&M's response to the NRC's RAI contained in Reference 6.
Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.              Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.
Sincerely, Joel P. Gebbie Site Vice President JMT/amp

==Enclosures:==
: 1. Affirmation
: 2. Response to Request for Additional Information Regarding Technical             Specification Section 5.5.14, "Containment Leakage Rate Testing Program"

U. S. Nuclear Regulatory Commission                    AEP-NRC-2015-07 Page 3 c:     M. L. Chawla, NRC Washington, D.C.
J. T. King, MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson, AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2015-07 AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the U. S. 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 K         DAY OF            oO-4      , 2015                      DANIELLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 04,04-2018 My Com Notary Iac                                    Acting In the County. of My Commission Expires            -["           \

Enclosure 2 to AEP-NRC-2015-07 Response to Request for Additional Information Regarding Technical Specification Section 5.5.14, "Containment Leakage Rate Testing Program" By letter dated March 7, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14071A419), as supplemented by letter dated September 30, 2014 (ADAMS Accession No. ML14275A454), Indiana Michigan Power Company (I&M) requested amendments to Operating Licenses DPR-58 and DPR-74 in the form of changes to the Technical Specifications (TS) for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, respectively. The license amendment request (LAR) proposes a change to TS 5.5.14, "Containment Leakage Rate Testing Program," to allow a permanent extension of the Type A primary containment integrated leak rate test (ILRT) frequency from one in 10 years to one in 15 years and extension of the Type C containment isolation valve leakage test frequency to 75 months.        The U. S. Nuclear Regulatory Commission's (NRC) Probabilistic Risk Assessment (PRA) Licensing Branch has reviewed the risk-related information in the LAR and identified areas where additional information is needed to complete its review. By electronic mail dated November 20, 2014, the NRC transmitted a request for additional information (RAI) regarding the March 7, 2014 LAR. The text of the RAIs and I&M's responses are provided below.
NRC RAI 1, In the safety evaluation report for Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," the Nuclear Regulatory Commission (NRC) staff, in part, stated that for licensee requests for a permanent extension of the integrated leak rate testing (ILRT) surveillance interval to 15 years "[c]apability category I of ASME RA-Sa-2003 shall be applied as the standard,since approximate values of [core damage frequency (CDF)] and [large early release frequency (LERF)] and their distribution among release categoriesare sufficient for use in the EPRI methodology."
Enclosure 6 to the LAR states that peer reviews of the PRA model were performed in 2001 and 2009. Table 1.0 provided in Enclosure 6 to the LAR describes the status of the facts and observations (F&Os) from these peer reviews. Enclosure 6 to the LAR further states that a gap assessment was performed in 2012, which, per Section 4.6.2 of Enclosure 2 to the LAR, was performed against Regulatory Guide (RG) 1.200, Revision
: 2. Section 4.6.1 of Enclosure 2 to the LAR additionally states that a new Level 2 model was used to perform the plant-specific risk assessment and that this Level 2 model was peer reviewed in 2013 against ASME/ANS PRA Standard RA-Sa-2009 and RG 1.200, Revision 2.
: a. Enclosure 6 to the LAR states that the 2012 gap assessment identified 68 supporting requirements (SRs) with potential gaps. Provide a list of any SRs considered not met to AEP-NRC-2015-07                                                             Page 2 at Capability Category I. Justify why not meeting each Capability Category I requirement will have no impact on the ILRT extension application.
: b. Section 4.6.2 of Enclosure 2 to the LAR states that the 2013 focused-scope peer review of the Level 2 model identified four SRs that were 'Not Met'. Provide the F&Os associated with these SRs and justify why not meeting each Capability Category I requirement will have no impact on the ILRT extension application I&M Response The RAI states that the 2012 gap assessment identified 68 Supporting Requirements (SRs) with potential gaps.      These were gaps to meeting applicable Capability Category (CC) II requirements in the 2009 American Society of Mechanical Engineers (ASME) PRA Standard. In addition, the RAI states that the 2013 focused-scope peer review of the Level 2 model identified 4 SRs that did not meet their applicable CC II requirements in the 2009 ASME PRA Standard.
However, these model reviews did not make any additional statements with regard to whether the SRs of interest met any specific CC requirements of any issue of the ASME PRA Standard.
Since the CC I requirements of the 2003 ASME PRA Standard are applicable to support ILRT interval extension requests, this response provides assurance that either the SRs of interest meet these requirements, or provides justification why not meeting these requirements will have no impact on the ILRT extension application.
The SRs of interest were reviewed, as were the nature of the potential gaps and/or basis for determining NOT MET status compared to the 2009 ASME Standard CC II requirements, including clarifications provided by the NRC (Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2). In addition, the historical record of independent review of the PRA model against the evolving ASME PRA Standard requirements for many of the SRs was also consulted.
Based on these review efforts, 11 SRs were determined to not be met at the CC I level for the 2003 ASME PRA Standard in whole or in part by the PRA model. Each of these SRs is discussed below in more detail, including a justification of why not meeting the CC I requirements of the 2003 ASME PRA Standard will have no impact on the ILRT extension application. Of these 11 SRs discussed below, the first seven are from the set of 68 gaps identified by the 2012 gap assessment. The discussion of these seven SRs are in response to RAI Section a. The last four SRs are from the 2013 Level 2 PRA focused-scope peer review.
The discussion of these four SRs are in response to RAI Section b.
Justification for No Impact - 2003 Standard SR IE-Cl0 SR IE-C10 of the 2003 ASME Standard pertains to comparing and resolving the results of the initiating event analysis with generic data sources to provide a reasonableness check of the results. This SR was determined to not meet the 2003 Standard CC I requirements because the Peer Review teams found no explicit documentation demonstrating performance of such a reasonableness check for initiating event analysis. However, the various Peer Reviews, Gap Analysis and Gap Assessment performed over the years by various vendors and Peer to AEP-NRC-2015-07                                                              Page 3 reviewers scrutinized the model results and did not identify any initiating event analysis category significantly differing from generic industry information. For that reason, this is considered solely a documentation issue and CNP model results are believed to be consistent with general industry results and are believed to be adequate for the application to extend the CNP ILRT frequency Justification for No Impact - 2003 Standard SR AS-C3 SR AS-C3 of the 2003 ASME Standard had a shortcoming identified by the 2001 Peer Review that a search for unique or unusual sources of uncertainty not present in the typical or generic plant analysis had not been explicitly documented. The Peer Review comments stated that event tree documentation discussed potential conservatisms, and that overall it was considered that the effect of any uncertainties would not significantly alter model results or insights. As a result, this is considered a documentation issue only, with no significant impact on model results or insights.
Justification for No Impact - 2003 Standard SR SC-C2 & SC-C3 SR SC-C2 & SC-C3 of the 2003 ASME Standard did not meet the CC I requirements for the 2003 version of this SR due to documentation shortcomings. These shortcomings were that sufficient documentation was not included in the notebooks discussing: thermal-hydraulic Modular Accident Analysis Program (MAAP) code limitations; processes used to define initiating event success criteria; and a search for unique sources of uncertainty. However, based on Peer Review comments, and associated later Gap Assessment feedback, the documentation inadequacies are not expected to affect the specific success criteria developed, nor the numerical estimates obtained by the PRA model. Peer Review comments associated with related SRs generally concluded the model was overly conservative but satisfactory for all applications and determining sensitivities Justification for No Impact - 2003 Standard SR QU-E2 SR QU-E2 of the 2003 ASME Standard pertains to identifying key sources of model uncertainty, and the assumptions made or models adopted in response to those uncertainties. Peer Review comments associated with this SR were made regarding a lack of sufficient documentation of a formal search for unique or unusual uncertainty sources. The overall QU element assessment was graded a 3, and stated only that, "Consideration should be given to expanding this evaluation to identify sources of uncertainty in the PRA, to aid in risk-informed applications."
However, 2001 Westinghouse Owner's Group Peer Review comments and later model assessments, such as the 2004 Gap analysis, indicate this is a documentation issue that should not significantly affect model results or insights.
Justification for No Impact - 2003 Standard SR LE-A4 SR LE-A4 of the 2003 ASME Standard pertains to the transfer of Level 1 information to the Large Early Release Frequency (LERF)/Level 2 model. The 2012 Gap Assessment identified a gap in some Level 1 sequences where Auxiliary Feedwater (AFW) is not questioned, and AFW is assumed to succeed; this gap represents a failure to meet the CC I requirements. The to AEP-NRC-2015-07                                                           Page 4 potential impact of this gap is considered numerically small since AFW is very reliable at CNP, given each unit has two motor-driven AFW pumps and one turbine-driven AFW pump, and each motor-driven AFW pump can be cross-tied to supply AFW to the other unit. This gap was explicitly addressed as part of the ILRT extension effort. The 2013 Level 2 Focused-Scope Peer Review identified this deficiency as corrected and stated that the corresponding 2009 Standard (SR LE-A4) met the CC I-Ill requirements.
Justification for No Impact - 2003 Standard SR LE-D5 SR LE-D5 of the 2003 ASME Standard pertains to the treatment of Thermally-Induced Steam Generator Tube Rupture (ISGTR) in the LERF/Level 2 model. The 2012 Gap Assessment identified a non-conservatism in the modeling approach. This gap was explicitly addressed as part of the ILRT extension effort. The 2013 Level 2 Focused-Scope Peer Review identified this deficiency as corrected and stated that the corresponding 2009 Standard (SR LE-D6) met the CC I requirements due to the conservative treatment.
Justification for No Impact - 2009 SR LE-F3 (2003 SR LE-F2)
The 2013 Focused-Scope Peer Review of the Level 2 model concluded that 2009 Standard SR LE-F3 CC I-Ill was NOT MET, and assigned Finding LE-F3-01 to the SR. This Finding and Observation (F&O) states the following:
"This SR points to the requirements of the Level 1 QU SRs. Examples of these SRs (i.e.,
QU-D1 to QU-D7 and QU-E1 to QU-E4) require that sources of model uncertainty and assumptions are identified, the uncertainty interval should be estimated, and the impact on the PRA model should be identified. This was not performed and results in the SR Not Met.
SR LE-G4 requires the sources of model uncertainty and related assumptions be documented, including results and important insights from sensitivity studies. Although some general assumptions are scattered throughout PRA-L2-Model, Rev. 0, there potentially are many more assumptions that are not provided in the notebook. Results in SR LE-G4 NOT MET."
The 2003 Standard SR corresponding to the 2009 Standard SR LE-F3 is LE-F2 and the CC I requirements are as follows:
"PROVIDE a qualitative assessment of the key sources of uncertainty.
Examples:
(a) Identify bounding assumptions.
(b) Identify conservative treatment of phenomena."
Although the requirements to meet CC I for the corresponding 2003 Standard SR are less than for the 2009 Standard SR, SR LE-F2 is also judged to have NOT MET the CC I requirements for the 2003 Standard.
However, it is important to note that the Level 2 Focused-Scope Peer Review Finding which affected the most SRs was LE-Bl-01 (although it did not result in any of these SRs being assigned a status of NOT MET). This F&O states the following:
to AEP-NRC-2015-07                                                              Page 5 "The use of NUREG/CR-6595 generic analyses, without significant plant-specific supporting calculations, produces a CC I model.
NUREG/CR-6595 generic analyses were used to characterize containment challenges for other initiators. Some of these challenges are also significant. (LE-B2, LE-D1, LE-E6)
The only explicit operator action included in the analysis is for the initiation of the Distributed Ignition System (DIS) prior to hydrogen build-up in containment. Implicit operator actions are captured via, for example, the use of NUREG-1570 for ISGTR and in the refilling of SGs following a main steam line break (MSLB) with AFW failure. (LE-C2)
No credit is taken for equipment or human actions that containment failure could impact.
(LE-C11, LE-C12)
Values from NUREG-1570 are used in the ISGTR analysis, and justified, which is considered conservative and yields a CC I. (LE-D6)"
As this F&O makes clear, a principle criticism of the Level 2 modeling is that it is overly conservative, in the sense that it is over-predicting the LERF value. As a result, the expected outcome of addressing this F&O is a list of risk-ranked, model enhancements that, as implemented, would reduce the overall LERF and likely improve the metrics provided in the ILRT extension submittal. As a result, not meeting the CC I requirements for 2009 SR LE-F3 will have no impact on the ILRT extension application.
Justification for No Impact - 2009 SR LE-G4 (2003 SR LE-G7)
The 2013 Focused-Scope Peer Review of the Level 2 model concluded that 2009 Standard SR LE-G4 CC I-Ill was NOT MET, and assigned Finding LE-F3-01 to the SR. This F&O states the following:
"This SR points to the requirements of the Level 1 QU SRs. Examples of these SRs (i.e.,
QU-D1 to QU-D7 and QU-E1 to QU-E4) require that sources of model uncertainty and assumptions are identified, the uncertainty interval should be estimated, and the impact on the PRA model should be identified. This was not performed and results in the SR NOT MET.
SR LE-G4 requires the sources of model uncertainty and related assumptions be documented, including results and important insights from sensitivity studies. Although some general assumptions are scattered throughout PRA-L2-Model, Rev. 0, there potentially are many more assumptions that are not provided in the notebook. Results in SR LE-G4 NOT MET."
The 2003 Standard SR corresponding to the 2009 Standard SR LE-G4 is LE-G7 and the CC I-Ill requirements are as follows:
"DOCUMENT sources of uncertainty."
SR LE-G7 is judged to have NOT MET the CC I requirements for the 2003 Standard.
to AEP-NRC-2015-07                                                              Page 6 However, it is important to note that the Level 2 Focused-Scope Peer Review Finding which affected the most SRs was LE-Bl-01 (although it did not result in any of these SRs being assigned a status of NOT MET). This F&O states the following:
"The use of NUREG/CR-6595 generic analyses, without significant plant-specific supporting calculations, produces a CC I model.
NUREG/CR-6595 generic analyses were used to characterize containment challenges for other initiators. Some of these challenges are also significant. (LE-B2, LE-D1, LE-E6)
The only explicit operator action included in the analysis is for the initiation of the Distributed Ignition System (DIS) prior to hydrogen build-up in containment. Implicit operator actions are captured via, for example, the use of NUREG-1570 for ISGTR and in the refilling of SGs following an MSLB with AFW failure. (LE-C2)
No credit is taken for equipment or human actions that containment failure could impact.
(LE-C11, LE-C12)
Values from NUREG-1570 are used in the ISGTR analysis, and justified, which is considered conservative and yields a CC I. (LE-D6)"
As this F&O makes clear, a principle criticism of the Level 2 modeling is that it is overly conservative, in the sense that it is over-predicting the LERF value. As discussed above, the expected outcome of addressing this F&O is a list of model enhancements that, if implemented, would reduce the overall LERF and likely improve the metrics provided in the ILRT extension submittal. Since addressing this SR relates to the part of addressing the F&O that creates the list of model enhancements, as opposed to implementing the potential model enhancements, not meeting the CC I requirements for 2009 SR LE-G4 will have no impact on the ILRT extension application.
Justification for No Impact - 2009 SR LE-G5 (2003 SR LE-G8)
The 2013 Focused-Scope Peer Review of the Level 2 model concluded that 2009 Standard SR LE-G5 CC I-Ill was NOT MET, and assigned Finding LE-G5-01 to the SR. This F&O states the following:
"This SR requires that limitations in the LERF analysis that would impact applications be documented. No limitations are discussed in PRA-L2-Model, Rev. 0 resulting in the SR NOT MET."
The 2003 Standard SR corresponding to the 2009 Standard SR LE-G5 is LE-G8 and the CC I-Ill requirements are as follows:
"IDENTIFY limitations that would impact applications."
SR LE-G8 is judged to have NOT MET the CC I requirements for the 2003 Standard.
to AEP-NRC-2015-07                                                              Page 7 However, it is important to note that the Level 2 Focused-Scope Peer Review included the following suggestion related to this topic:
"The AFW system model was used to develop point estimate failure probabilities for use in the Level 2 model. While this is acceptable, as dependencies appear to have been correctly identified, the use of the model for applications is hampered by the use of point estimate failure probabilities.
Both the containment isolation and AFW system models were used to develop point estimate failure probabilities for use in the Level 2 model. While this is acceptable, as dependencies appear to have been correctly identified, the use of the model for applications is hampered by the use of point estimate failure probabilities. (LE-D7, LE-G5)"
Although addressing the limitation identified would improve the general applicability of the Level 2 model, the Peer Review did not consider that the limitation affected the particular application (i.e., the ILRT extension) that was the genesis for both the Level 2 model and the Focused-Scope Peer Review. In fact, the suggestion explicitly states that "dependencies appear to have been correctly identified." As a result, for this application, not meeting the CC I requirements for LE-G5 is a documentation deficiency that will have no impact on the ILRT extension application.
Justification for No Impact - 2009 SR LE-G6 (No Equivalent 2003 SR)
The 2013 Focused-Scope Peer Review of the Level 2 model concluded that 2009 Standard SR LE-G6 CC I-Ill was NOT MET, and assigned Finding LE-C10-01 to the SR. This F&O states:
"Section 7.2 reviews the significant accident progression sequences. However, there is no discussion in PRA-L2-Model, Rev. 0 regarding a review of the significant accident progression sequences and continued operation that could reduce LERF.
SR LE-G6 requires 'DOCUMENT the quantitative definition used for significance accident progression sequence.        If other than the definition is used in Section 2, JUSTIFY the alternative.' No definitions for significant accident progression sequence were provided in PRA-L2-Model, Rev. 0, resulting in a NOT MET."
The 2003 Standard has no SR corresponding to the 2009 Standard SR LE-G6.
Further, the possible resolution of this F&O identified in the 2013 Level 2 Focused-Scope Peer review states:
"SR LE-C10 CC I requires no review. To obtain CC II, a review of the significant accident progression sequences and a discussion of possible continued equipment/human operation that could reduce LERF should be included."
Since the 2003 Standard has no specifically comparable SR, this gap in the 2009 Standard SR does not reflect any shortcoming of the model in meeting the CC I requirements for the 2003 Standard in this aspect. Further, the identified documentation shortcoming is not required to to AEP-NRC-2015-07                                                        Page 8 meet the 2009 Standard SR CC I requirements. As a result, not meeting the CC II requirements for SR LE-G6 will have no impact on the ILRT extension application.
Conclusion Based on the above discussion, I&M concludes that PRA model used in support of the ILRT extension application generally meets the 2003 ASME Standard CC I requirements with few exceptions. Further, these few exceptions have been justified as not materially affecting the ILRT extension application.}}