JAFP-11-0119, Response to Request for Additional Information Regarding the Application for a Technical Specification Change to Relocate Specific Surveillance Frequency Requirements to a Licensee Controlled Program in Accordance with TSTF-425,...

Response to Request for Additional Information Regarding the Application for a Technical Specification Change to Relocate Specific Surveillance Frequency Requirements to a Licensee Controlled Program in Accordance with TSTF-425,...

ML112930085
Person / Time
Site:	FitzPatrick
Issue date:	10/19/2011
From:	Bronson K Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-11-0119, TAC ME6755, TSTF-425, Rev 3
Download: ML112930085 (21)

Text

JAFP-11-0119

October 19, 2011

United States Nuclear Regulatory Commission

Attn: Document Control Desk Washington, D.C. 20555

Subject:

Response to Request for Additional Information Regarding the Application for a Technical Specification Change to Relocate Specific Surveillance Frequency Requirements to a Licensee Controlled Program in Accordance with TSTF-425, Revision 3, (TAC No. ME6755).

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59

References:

1. Entergy Letter, from Kevin Bronson to the USNRC, Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program in Accordance with TSTF-425, Revision 3, JAFP 0088, dated July 22, 2011 (ML112060443). 2. NRC Request for Additional Information Regarding James A. FitzPatrick Nuclear Power Plant License Amendment to Relocate Specific Surveillance Frequency Requirements to a Licensee Controlled Program, by email dated September 15, 2011 (TAC No. ME6755). 3. Telephone Conference with the USNRC, Regarding Clarification of Request for Additional Information Questions, September 15, 2011.

Dear Sir or Madam:

By letter dated July 22, 2011 (Reference 1), Entergy Nuclear Operations, Inc. (Entergy) submitted for Nuclear Regulatory Commission (NRC) review and approval a change to the James A. FitzPatrick Nuclear Power Plant (JAF) Technical Specifications (TS). The proposed amendment would modify the JAF TS by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies." Kevin Bronson Site Vice President - JAF Entergy Nuclear NortheastEntergy Nuclear Operations, Inc.

James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-349-6024 Fax 315-349-6480 JAFP-11-0119 Page 2 of 2 Subsequent to the submittal of Reference 1, the NRC staff requestedadditionalinformation (RAI)necessary to perform their review (Reference 2).The RAI questions were discussed with the NRC and clarified on a teleconference on September 15, 2011 (Reference 3).The responses to those questions, as clarified, are included as an enclosure to this letter.There are no new commitments made in this letter.Questions concerning this response may be addressed to Mr.Joseph Pechacek, Licensing Manager, at (315)349-6766.I declare under penalty of perjury that the foregoing is true and correct.Executed on this 19th day of October 2011.Sincerely,[3('"Ih-Kevin Bronson Site Vice President KB/JP/jo

Enclosure:

Response to Request for Additional Information cc: Mr.William Dean Regional Administrator, Region I U.S.Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr.Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S.Nuclear Regulatory Commission Mail Stop 0-8-C2A Washington, D.C.20555-0001 Mr.Francis J.Murray Jr., President New York State Energy and Research Development Authority 17 Columbia Circle Albany, NY 12203-6399 Resident Inspector's Office U.S.Nuclear Regulatory Commission James A.FitzPatrick Nuclear Power Plant P.O.Box 136 Lycoming, NY 13093 Ms.Bridget Frymier New York State Department of Public Service 3 Empire State Plaza, 10 th Floor Albany, NY 12223-1350 Mr.Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10 th Floor Albany, NY 12223-1350 JAFP-11-0119 Enclosure Page 1 of 19 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION (RAI) RELATED TO AN AMENDMENT TO IMPLEMENT TSTF-425 REVISION 3 ENTERGY NUCLEAR NORTHEAST JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 RAI 1 In Section 2.2.3 of Attachment 2 of the LAR, the licensee identified that the September 2009 peer review identified 51 facts and observations (F&Os), 21 of which were considered to not meet at least capability category II of the applicable internal events probabilistic risk assessment (PRA) standard. Table 2-1 of Attachment 2 summarizes and assesses the open F&Os. There are 11 F&Os discussed in Table 2-1. a. How was the determination made for each F&O that capability category II of the standard was not met - was this internally determined by licensee staff, or by the peer review? If determined by licensee, then discuss this decision process.

Response: The peer review team members made the determination for each F&O that capability category II of the standard was not met based on their PRA experience and the supporting requirements of the ASME PRA Standard. b. The licensee is requested to confirm that the other 10 F&Os not summarized in Table 2-1 have been resolved and closed out, and that their disposition is reflected in the current PRA model proposed for use to support the Surveillance Frequency Control Program (Reference 3 of Attachment 2); if any of these 10 F&Os were closed out without making any changes to the PRA model, then the licensee is requested to provide summary information of the F&O, and justification as to why no change was required to resolve the issue. If the other 10 F&Os are not yet closed, the licensee should discuss and justify why these F&Os were omitted from Table 2-1.

Response:

The James A. FitzPatrick Nuclear Power Plant (JAF) PRA Baseline Model, documented in Reference 3 of Attachment 2 to the License Amendment Request (LAR), which includes a complete evaluation of the JAF risk profile for internal event challenges, was used for the peer review. Following the peer review, the 2010 PRA application model was developed to incorporate changes required from the peer review to assure that the PRA quality and expectations are met for all risk informed applications (e.g., Maintenance Rule, AOV / MOV Risk Rankings, Online Risk Model, and MSPI) according to Entergy's procedure EN-DC-151. The current model proposed for use to support the Surveillance Frequency Control Program is the 2010 PRA application model. The 10 F&Os not summarized in Table 2-1 of Attachment 2 to the LAR have been resolved and closed out. Although changes have been made to resolve each of those F&Os, only two of the 10 F&Os required changes to the actual model, and those changes are reflected in the 2010 PRA application model. The other eight changes are documentation changes only and will not have any impact on the implementation of this application. Table RAI-1 summarizes the status and potential impacts on this application for the 10 closed F&Os.

JAFP-11-0119 Enclosure Page 2 of 19 RAI 2 The summary of the F&O for supporting requirement (SR) AS-B7 is unclear, in that the table (Table 2-1 of Attachment 2) entry implies that the Control Rod Drive (CRD) system model logic and success criteria do not explicitly account for the fact that the CRD system capacity is insufficient for decay heat removal early in the sequence (i.e., immediately after a plant trip), but

that the CRD system is only credited on the success branch, which would seem to imply that the model logic is accurate. Further, the disposition of the importance of this F&O states that surveillance test interval changes would not be relevant to the modeling of long term use of CRD; this would appear to be inconsistent with the potential impact of improper modeling of CRD on the risk importance of CRD and other injection sources, which could impact the risk calculations made to address surveillance test interval changes for such systems. Finally, it is stated that random failures of early injection are truncated out during accident sequence quantification. While this may be accurate, it does not seem to be relevant to the issue of improper crediting of the CRD injection functi on. The U.S. Nuclear Regulatory Commission (NRC) staff is unable to reach a conclusion regarding this F&O and requests the following clarifications: a. Provide the full text of the F&O for staff review.

Response: The full text for F&O AS-B7 is as follows:

The use of CRD for makeup does not account for time dependency. Gate U3 is used after containment heat removal fails and venting is a success. The success criteria for CRD indicate that it is not a valid source of makeup immediately after a transient. It is implicit that the High Pressure Coolant Injection System (HPCI) or the Reactor Core Isolation Cooling System (RCIC) work until containment is vented.

Basis for Significance This is a finding since the time dependency of CRD is not adequately addressed.

Possible Resolution Explicit modeling of immediate high pressure makeup is required. b. Describe how the CRD injection function is modeled in the accident sequence logic, and specifically identify how this modeling will ultimately be revised to address the F&O (if applicable), or justify that the existing modeling is correct or conservative.

Response: The modeling of CRD does account for time dependency, albeit, implicitly and it recognizes the inability of CRD to provide early reactor vessel injection. The accident sequence logic does credit high pressure injection from t he CRD system, but only for those sequences where failure of containment decay heat removal (due to loss of the Torus Cooling and Drywell Sprays) occurs and plant operators successfully vent containment. Implicit in this development is the long term failure of HPCI/RCIC due to accident phenomena associated with loss of containment heat removal, and subsequent loss of the Low Pressure Coolant Injection (LPCI)/Core Spray Systems due to inadequate net positive suction head (NPSH) following containment venting via the torus vent path. At this point, following successful containment venting, the operators would attempt to align vessel injection using the CRD system.

JAFP-11-0119 Enclosure Page 3 of 19 To support crediting CRD for those sequences, a Modula Accident Analysis Program (MAAP) analysis was performed (MAAP case JA F-TW-1) that confirmed that 60 gpm from one CRD pump would provide sufficient in jection for those long-term sequences.

Although the current modeling does not inappropriately credit CRD for any sequences, the original LAR noted that core damage cutsets involving random failures of HPCI/RCIC/LPCI/Core Spray following failure of containment decay heat removal would expect to be truncated. This is based on the fact that random failures of HPCI/RCIC/LPCI/Core Spray (early or late) are addressed in the model for sequences involving success of containment decay heat removal. While corresponding sequences could be added representing random failure of HPCI/RCIC/LPCI/Core Spray following failure of containment decay heat removal (and the CRD system could then be questioned for that portion where containment venting is successful) all those sequences would be identical to existing sequences, except for the additional required failure of HPCI/RCIC/LPCI/Core Spray. The impact would therefore be negligible. As a result; the existing model will not impact the proposed application.

c. If the current modeling of the CRD injection function is non conservative, provide bounding sensitivity analyses which do not credit the CRD system.

Response: Although, as discussed in the previous response, the current modeling of the CRD system in the accident sequence logic is reasonable, a bounding sensitivity analysis which does not credit the CRD system was performed.

Using the 2010 PRA application model the following values are presented:

Core Damage Frequency (CDF) with CRD credited is 2.3 x 10

-6 /ry CDF without CRD credited is 2.5 x 10

-6 /ry Large Early Release Frequency (LERF) with CRD credited is 2.6 x 10

-7 /ry LERF without CRD credited is 2.6 x 10

-7 /ry The increase in CDF is approximately 8.7 percent. There is no significant increase in LERF. As noted above, although this sensitivity has been performed to provide some perspective relative to the RAI concern, we believe that the current modeling of the CRD system reasonably reflects the actual plant and is appropriate for this application.

JAFP-11-0119 Enclosure Page 4 of 19 RAI 3 The summary of the F&O for SR HR-G7 identifies that in the modeling for human error dependencies, the order of the events is based on individual human error probabilities (HEPs) rather than the order in which the events will occur in a given accident sequence, which is the accepted method for analyzing the overall joint probability of multiple dependent human errors. The licensee discussion of the importance of this F&O states that lower probability human errors (i.e., those most likely to be successful) generally occur earlier in the sequence. The NRC staff is not aware of this fact as a generally accepted principle in PRA. The licensee also states that dependent probabilities are conservative because the dependency between execution portions of two actions are frequently low or zero, while the dependent probability applies to the overall HEPs. While this may be an accurate statement, the qualitative identification of conservatism in a calculation is not an adequate basis to judge the potential quantitative impact of an inaccurate calculation method.

The NRC staff is unable to reach a conclusion regarding this F&O and requests the following

clarifications:

a. Provide the basis for the assertion that lower probability human errors generally occur earlier in the sequence.

Response: The statement that the lower probability human error events generally occur earlier in the sequence was not meant to be an assertion about all PRAs but rather an observation of the combinations of human error probabilities which are important contributors specifically to the JAF PRA. Although the F&O states that the JAF approach does not conform to the "established calculation method", the actual supporting requirement in the ASME standard does not, in fact, provide or discuss an established calculation method for determining the appropriate joint probability. It simply requires the HRA to "calculate a joint human error probability that reflects the dependence". While the supporting requirement specifies that the influence of success or failure in preceding human actions be accounted for, it does not specify how the actual calculation should be performed. In fact, Note 1 in the ASME standard for supporting requirement HR-G7 recognizes that the state of the art with regard to assessing dependency is largely based on the analyst's judgment. Ultimately, we believe that the intent of the SR is to develop the best estimate of the joint probability and, therefore, it is not required or preferable to use a single approach that in some cases produces an unnecessarily conservative value. The HRA dependency analysis adopted for the JAF PRA was chosen to allow flexibility and generate more realistic estimates of the combined HEP (CHEP) in cases where the first action has a much higher HEP primarily due to a much smaller time window for performing the action. That is, if two operator actions involve similar diagnosis or cues, but the time available to perform the first sequential action is much smaller than the time available to perform the subsequent action, then the conditional HEP for the second action could be unrealistically high, resulting in a combined HEP which is higher than the independent HEP for the second action. However, it would stand to reason that the combined HEP should never be higher than any of the associated independent HEPs.

JAFP-11-0119 Enclosure Page 5 of 19 b. Discuss the scope of this F&O in terms of the number of dependency calculations which are impacted, the quantitative impact of individual calculations of dependency, and the potential cumulative impact on overall model results.

Response: Re-evaluation of the CHEPs to address the F&O for SR HR-G7 resulted in a cumulative CDF increase of only 1.73%, or 4.03E-08 per reactor-year. The increase in the calculated LERF is less than 0.02%. Of the 106 combinations of HEPs that were evaluated in the JAF PRA, only 40 are impacted by the F&O for SR HR-G7. In the re-evaluation of those 40 combinations, only 26 resulted in higher CHEP values, with only one CHEP contributing more than 1% to the total CDF. It should be noted that in many instances the re-evaluation of the CHEPs resulted in decreased CHEP values due to lower dependencies as a result of many hours being available between performance of the actions (e.g., different operating crews involved in the response) and credit for success of intervening actions.

Table RAI-2 shows the individual contribution of each CHEP on CDF, based on both the current CHEP in the JAF 2010 PRA application model as well as the sensitivity CHEP value to address the F&O for SR HR-G7.

JAFP-11-0119 Enclosure Page 6 of 19 RAI 4 The summary of the F&O for SR LE-E3 identifies that the definition of "early" is not consistent and that this may result in some large early release frequency (LERF) sequences being classified as non-LERF. The licensee discussion of the importance of this F&O states that changes to surveillance test interval would not impact the time available for protective actions prior to a release. The NRC staff believes that the potential impact of this F&O is that a change to a surveillance test interval may be calculated as having a lesser impact on LERF, since LERF sequences may not be properly identified. The fact that surveillance test intervals are unrelated to the time available for protective actions, while accurate, is not relevant in assessing the potential quantitative impact of this F&O. The NRC staff is unable to reach a conclusion regarding this F&O and requests the following clarifications:

a. Provide the full text of the F&O for staff review.

Response: The full text for F&O LE-E3 is as follows:

The definition of 'Early' is inconsistent within the documents and may classify some LERF sequences as non-LERF.

(This F&O originated from SR LE-E3)

Basis for Significance The definition of 'Early' is critical to the identification of LERF or non-LERF sequences.

Possible Resolution Ensure a consistent definition of 'Early' that is based on the time of General Emergency declaration and link each accident progression sequence to a declaration time for comparison to the 'Early' criterion. b. Provide the basis for defining "early" in the actual PRA model of LERF sequences.

Response:

Following is the definition of "early" from the PRA model update report:

The rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions. This involves CET endstates in which containment failure occurs within 0 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the start of the initiating event. (0-6 hours is conservatively assumed to include cases in which minimal offsite protective measures have been observed to be performed in non-nuclear accidents).

The above definition is consistent with ASME/ANS standard definition of a large early

release.

JAFP-11-0119 Enclosure Page 7 of 19

c. Discuss where the model may be non conservative (i.e., treating potential LERF sequences as non-LERF) and include a quantitative assessment of the non conservatisms.

Response:

Based on the definition provided in response to 4b of this RAI, the LERF sequences are binned accordingly. Specifically the LERF sequences are identified by their containment

failure (or bypass) mode and confirmed by accident progression calculations documented in Section J1.5.5 and Tables J1.5-6 and J1.5-8 of Appendix J1 of the JAF PRA Model Update

Report. In regard to the Peer Review F&O, while we do not object to linking the definition of an early release to the amount of time after a General Emergency is declared, given the uncertainty related to implementation of protective actions, and the likelihood of proactive implementation, we do not consider the current JAF approach unreasonable. We consider the attempt to link scenario timing to specific plant EALs and anticipated actions on the part of the plant and offsite authorities to be an enhancement to the definition of large early release contained in the combined ASME/ANS PRA standard.

Although we believe LERF sequences to be accurately modeled, sensitivity was performed to address the concern raised in the SR LE-E3 F&O and provide the quantitative assessment requested in the RAI. This was done by examining the fifteen core damage accident classes used in the PRA model. Based on this examination, accident sequences involving the loss of containment decay heat removal (Accident Class II) are the only class expected to have the potential for reclassification of non-LERF sequences into LERF sequences if use of this alternate approach resulted in a delay in the assumed time to take protective actions. This Class considers all releases to be late releases based on the extended time for core damage to occur following a plant initiator.

According to a new EAL scheme based on NEI 99-01 that is being implemented by October, 20, 2011, the time to declare a General Emergency is normally triggered by the loss of any two barriers and the loss or potential loss of a third barrier. It can also be declared at the discretion of the Emergency Director. To evaluate the impact of re-binning these non-LERF Accident Class II sequences, the time at which a General Emergency would be declared was estimated based on the core damage accident class definition, conditions for General Emergency declaration, and MAAP results. The salient information for this re-assessment of the loss of containment decay heat removal accident sequences is provided in Table RAI-3.

As shown in that table, the additional time to core damage beyond the expected time at which the General Emergency would be declared is still well in excess of the anticipated time required for evacuation. Based on this, there is no expected change in LERF by applying the EAL specific approach. Since it cannot be guaranteed, however, that the plant will perform as expected, Table RAI-3 proposes a sensitivity in which some probability of failing to properly implement the EALs is applied.

JAFP-11-0119 Enclosure Page 8 of 19 Re-binning the non-LERF sequences as LERF sequences based on the sensitivity described in Table RAI-3 and using the 2010 PRA application model results in an increase in the calculated baseline LERF of 3 x 10

-8/ry.

Furthermore, since only Class II accident sequences are impacted and not all of Class II accident sequences result in high releases, any increase in the LERF for these accident sequences due to any proposed change in a component Surveillance Test Interval would be bounded by the change in CDF.

As a result, the existing model will not impact the proposed application.

JAFP-11-0119 Enclosure Page 9 of 19 Table RAI-1 Summary of Closed F&Os Identified as Not Meeting Capability Category II of the ASME/ANS PRA Standard DESCRIPTION OF F&O APPLICABLE SRs CURRENT STATUS I COMMENT IMPACT ON APPLICATION The time dependency of the DC to support HPCI (U1) and RCIC (U2) is inadequate. The battery and chargers are both in an AND gate. This is incorrect for mission times greater than four hours. The battery, with load shed, will last four hours and needs the charger after that; therefore, the sole dependency should be on the battery charger. AS-B7 Closed. Modified dependency logic for HPCI and RCIC fault trees (gates GHCI-FLC108 and GRCI-FLC-1391, respectively) to reflect that continued operation requires availability of both the charger and battery. None. This F&O has been addressed and incorporated

into the model. Appendices E1-E36 show many systems

where support systems ar e explicitly modeled or assumed to be necessary in the absence of engineering analyses to determine whether they are needed. Example: battery ventilation system. SY-B6 Closed. Plant engineering analyses, where available, were used in modeling support systems. In the specific case of the battery ventilation system, the calculation did not provide sufficient basis for concluding that

ventilation was unnecessary. Therefore, the Battery Room Ventilation System was included as a support system to the battery chargers.

None. Support system modeling was included where needed or best estimate analyses were not available or sufficient.

The majority of support systems are based on conservative success criteria and timing, but the criteria are not justified based on the impact on risk significant contributors.

SY-B7 Closed. A further review of the existing support system success criteria documentation determined that use of conservative success criteria and timing was limited and where used, it had little impact

on results. None. This was a documentation issue only and has been addressed.

JAFP-11-0119 Enclosure Page 10 of 19 Table RAI-1 Summary of Closed F&Os Identified as Not Meeting Capability Category II of the ASME/ANS PRA Standard DESCRIPTION OF F&O APPLICABLE SRs CURRENT STATUS I COMMENT IMPACT ON APPLICATION RCIC/HPCI flow is delivered to the reactor vessel via feedwater line A/B. The HPCI fault tree includes the failure of the downstream

feedwater check valve and manual valve but the RCIC fault tree does not.

SY-B13 Closed. The RCIC system fault tree has been revised to include the "A" feedwater

check valve and manual valve. None. This F&O has been addressed and incorporated

into the model. The documentation contained no evidence of checking the reasonableness of the posterior distribution.

DA-D4 Closed. Detailed discussions that describe the process of confirming the reasonableness of the posterior distribution mean value have been added to Appendix D. None. This was a documentation issue only and has been addressed. Significant contributors to CDF by initiating events, accident sequence, systems and operator errors were identified and discussed in the Summary Report, Sections 3.2 through 3.6. A listing of basic event importance is provided in Appendix I.3. It does not appear the SSCs and HFEs associated with initiating events modeled by fault tree were included. QU-D6 Closed. Significant contributors to CDF, such as initiating events, accident sequences, equipment failures, common cause failures, and operator errors were identified. The importance of SSCs and HFEs that contribute to initiating event frequencies and event mitigation are identified in the results to the extent possible. In general, it was found that basic events associated with initiating events which are important contributors to CDF will also be important for accident mitigation and their importance will be reasonably accounted for using the current method for

importance ranking.

None. The importance of systems components and operator actions has been reasonably accounted for in the PRA model. In any case, baseline importance ranking is not relevant to this application.

JAFP-11-0119 Enclosure Page 11 of 19 Table RAI-1 Summary of Closed F&Os Identified as Not Meeting Capability Category II of the ASME/ANS PRA Standard DESCRIPTION OF F&O APPLICABLE SRs CURRENT STATUS I COMMENT IMPACT ON APPLICATION While descriptions and some discussion of the top ten accident sequences are provided, additional information with regard to the contributors to the sequence frequency should be included.

QU-F3 Closed. Each of the top 95% accident sequences was provided with a description which summarizes the failures and/or successes for that sequence. In addition, detailed descriptions were provided for all sequences in Appendix F. None. This was a documentation issue only and has been addressed. Accident sequences that credited reactor building plate-out (reducing the release magnitude one classification) have not been

justified.

LE-C1 Closed. The appendix addressing radionuclide release impacts was modified to document that for sequences where the Reactor Building (RB) node is effective, MAAP calculations indicate a substantial reduction in the release. None. This was considered a documentation issue only and has been addressed. LERF sequences are identified by their containment failure (or bypass) mode and confirmed by accident progression calculations documented in Section J1.5.5 and Tables J1.5-6 and J1.5-8. Some MAAP runs in Table J1.5-6 labeled as Large-Early

are not included in LERF because the MAAP run may only be implemented as a conservative run for a late scenario.

LE-E3 Closed. Added the following sentence to the last paragraph of Section J1.5.5:

"Note that not all the MAAP analyses represented in Table J1.5-6 and Table J1.5-8 were used. The MAAP analyses selected reflect the most appropriate cases in terms of predicting the accident progression for

each of the large early release endstates."

None. This was a documentation issue only and has been addressed. Limitations discussed in Appendix J are focused on model conservatisms and other LERF issues, but these limitations are not addressed in terms of their impact on applications.

LE-G5 Closed. A discussion was added in Appendix J to document that no limitations of the LERF quantification process were identified that would impact applications. None. This was a documentation issue only and has been addressed.

JAFP-11-0119 Enclosure Page 12 of 19 Table RAI-2 Summary of Impact on Combined (Dependent) Human Error Probabilities (CHEPs) JAF 2010 PRA Model Sensitivity for HR-G7 COMBINED HUMAN ERROR PROBABILITY EVENT APPLICABLE COMBINATIONS OF HUMAN FAILURE EVENTS CHEP Value %CDF CHEP Value %CDF CHEP-3AALT-LVENT LIP-XHE-FO-3AALT

• NVP-XHE-FO-LVENT 7.69E-04 [1] 0.019% 5.30E-04 0.013% CHEP-3AALT-RVENT LIP-XHE-FO-3AALT

• NVP-XHE-FO-RVENT 3.48E-04 0.027% 2.40E-04 0.018% CHEP-3BALT-LVENT LIP-XHE-FO-3BALT

• NVP-XHE-FO-LVENT 7.69E-04 0.018% 5.30E-04 0.012% CHEP-3BALT-RVENT LIP-XHE-FO-3BALT

• NVP-XHE-FO-RVENT 3.48E-04 0.006% 2.40E-04 0.004% CHEP-CTSLT-RB272 CTS-XHE-FO-AILT

• FLD-XHE-FO-RB272 1.42E-02 0.035% 9.80E-03 0.024% CHEP-FXTHX-W2 DWS-XHE-FO-W2

• FXT-XHE-FO-RHRSW 2.35E-05 0.003% 1.00E-06 0.000% CHEP-FXTHX-X1LT ADS-XHE-FO-X1LT

• FXT-XHE-FO-RHRSW 5.10E-06 0.007% 1.00E-06 0.001% CHEP-LVENT-RB272 FLD-XHE-FO-RB272

• NVP-XHE-FO-LVENT 7.69E-04 0.025% 5.30E-04 0.016% CHEP-NT4F-43ESWA FLD-XHE-FO-NT4F

• FXT-XHE-FO-ESWA

• NR-FLD-AOP43 1.27E-04 0.000% 1.47E-04 0.000% CHEP-NT4F-43ESWB FLD-XHE-FO-NT4F

• FXT-XHE-FO-ESWB

• NR-FLD-AOP43 1.17E-04 0.000% 9.79E-05 0.000% CHEP-NT4F-AOP43 FLD-XHE-FO-NT4F

• NR-FLD-AOP43 6.89E-04 0.036% 2.00E-04 0.010% CHEP-NT4F-ESWALV FLD-XHE-FO-NT4F

• FXT-XHE-FO-ESWA

• NVP-XHE-FO-LVENT 1.39E-05 0.000% 5.25E-05 0.000% CHEP-NT4F-ESWBLV FLD-XHE-FO-NT4F

• FXT-XHE-FO-ESWB

• NVP-XHE-FO-LVENT 9.29E-06 0.000% 3.50E-05 0.000% CHEP-NT4MF-43ESWA FLD-XHE-FO-NT4MF

• FXT-XHE-FO-ESWA

• NR-FLD-AOP43 3.19E-04 0.000% 8.23E-04 0.000% CHEP-NT4MF-43ESWB FLD-XHE-FO-NT4MF

• FXT-XHE-FO-ESWB

• NR-FLD-AOP43 2.92E-04 0.000% 5.49E-04 0.000% CHEP-NT4MF-AOP43 FLD-XHE-FO-NT4MF

• NR-FLD-AOP43 3.86E-03 0.215% 1.30E-02 0.713% CHEP-NT4MF-ESWALV FLD-XHE-FO-NT4MF

• FXT-XHE-FO-ESWA

• NVP-XHE-FO-LVENT 3.49E-05 0.000% 2.94E-04 0.000% CHEP-NT4MF-ESWBLV FLD-XHE-FO-NT4MF

• FXT-XHE-FO-ESWB

• 2.33E-05 0.000% 1.96E-04 0.000%

JAFP-11-0119 Enclosure Page 13 of 19 Table RAI-2 Summary of Impact on Combined (Dependent) Human Error Probabilities (CHEPs) JAF 2010 PRA Model Sensitivity for HR-G7 COMBINED HUMAN ERROR PROBABILITY EVENT APPLICABLE COMBINATIONS OF HUMAN FAILURE EVENTS CHEP Value %CDF CHEP Value %CDF NVP-XHE-FO-LVENT CHEP-RR3F-43ESWA FLD-XHE-FO-RR3F

• FXT-XHE-FO-ESWA

• NR-FLD-AOP43 4.24E-05 0.000% 4.44E-05 0.000% CHEP-RR3F-43ESWB FLD-XHE-FO-RR3F

• FXT-XHE-FO-ESWB

• NR-FLD-AOP43 4.11E-05 0.000% 4.07E-05 0.000% CHEP-RR3F-ESWALV FLD-XHE-FO-RR3F

• FXT-XHE-FO-ESWA

• NVP-XHE-FO-LVENT 3.60E-06 0.000% 2.28E-05 0.000% CHEP-RR3F-ESWBLV FLD-XHE-FO-RR3F

• FXT-XHE-FO-ESWB

• NVP-XHE-FO-LVENT 3.49E-06 0.000% 2.28E-05 0.000% CHEP-RR4F-43ESWA FLD-XHE-FO-RR4F

• FXT-XHE-FO-ESWA

• NR-FLD-AOP43 2.00E-04 0.000% 4.02E-04 0.000% CHEP-RR4F-43ESWB FLD-XHE-FO-RR4F

• FXT-XHE-FO-ESWB

• NR-FLD-AOP43 1.83E-04 0.000% 2.68E-04 0.000% CHEP-RR4F-AOP43 FLD-XHE-FO-RR4F

• NR-FLD-AOP43 1.89E-03 0.043% 6.24E-03 0.140% CHEP-RR4F-ESWALV FLD-XHE-FO-RR4F

• FXT-XHE-FO-ESWA

• NVP-XHE-FO-LVENT 2.18E-05 0.000% 1.44E-04 0.000% CHEP-RR4F-ESWBLV FLD-XHE-FO-RR4F

• FXT-XHE-FO-ESWB

• NVP-XHE-FO-LVENT 1.46E-05 0.000% 9.57E-05 0.000% CHEP-RVENT-RB272 FLD-XHE-FO-RB272

• NVP-XHE-FO-RVENT3.48E-04 0.010% 2.40E-04 0.007% CHEP-ST3F-43ESWA FLD-XHE-FO-ST3F

• FXT-XHE-FO-ESWA

• NR-FLD-AOP43 1.65E-04 0.000% 2.80E-04 0.000% CHEP-ST3F-43ESWB FLD-XHE-FO-ST3F

• FXT-XHE-FO-ESWB

• NR-FLD-AOP43 1.51E-04 0.000% 1.87E-04 0.000% CHEP-ST3F-AOP43 FLD-XHE-FO-ST3F

• NR-FLD-AOP43 1.31E-03 0.018% 3.37E-03 0.046% CHEP-ST3F-ESWALV FLD-XHE-FO-ST3F

• FXT-XHE-FO-ESWA

• NVP-XHE-FO-LVENT 1.80E-05 0.000% 1.00E-04 0.000% CHEP-ST3F-ESWBLV FLD-XHE-FO-ST3F

• FXT-XHE-FO-ESWB

• NVP-XHE-FO-LVENT 1.20E-05 0.000% 6.67E-05 0.000% CHEP-X1LT-V1LT ADS-XHE-FO-X1LT

• CDS-XHE-FO-V1LT 5.03E-06 0.000% 1.60E-05 0.000%

JAFP-11-0119 Enclosure Page 14 of 19 Table RAI-2 Summary of Impact on Combined (Dependent) Human Error Probabilities (CHEPs) JAF 2010 PRA Model Sensitivity for HR-G7 COMBINED HUMAN ERROR PROBABILITY EVENT APPLICABLE COMBINATIONS OF HUMAN FAILURE EVENTS CHEP Value %CDF CHEP Value %CDF CHEP-X1LT-V1T2 ADS-XHE-FO-X1LT

• CDS-XHE-FO-V1T2 5.10E-06 0.000% 1.00E-06 0.000% CHEP-X1T2-F1HHCI ADS-XHE-FO-X1T2

• FLD-XHE-FO-RR4F

• NR-HPCI (transients) 8.49E-06 0.000% 7.26E-04 0.000% CHEP-X1T2-F1HRCI ADS-XHE-FO-X1T2

• FLD-XHE-FO-RR4F

• NR-RCIC (transients) 8.68E-06 0.000% 7.42E-04 0.000% CHEP-X1T2-HCI ADS-XHE-FO-X1T2

• HCI-XHE-FO-RESET 1.06E-04 0.028% 7.30E-04 0.401% CHEP-X1T2-HPI ADS-XHE-FO-X1T2

• NR-HPCI

• NR-RCIC (transients) 1.86E-05 0.000% 1.85E-05 0.000% CHEP-X1T2-V1T2 ADS-XHE-FO-X1T2

• CDS-XHE-FO-V1T2 3.72E-05 1.523% 5.58E-05 2.280%

[1] 7.69E-04 = 7.69 x 10

-4 JAFP-11-0119 Enclosure Page 15 of 19 Table RAI-3 Basis for General Emergency Declaration for Accident Class II Sequences Accident Class Designator Definition Conditions for General Emergency Declaration Timing of General Emergency Declaration Class IIA Transient sequence with loss of all containment heat removal. Core

damage induced post

containment failure.

9.1.8 General Emergency Any event, as determined by the Shift Manager or Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third, refer to "Fission Product Barrier Loss/Potential Loss Matrix" This Emergency Action Level (EAL) allows for classification of emergencies that do not fit into any other General Emergency (GE) criteria.

The EALs of this category provide the Emergency Director (ED) or Shift Manager (SM) the latitude to classify emergency conditions consistent with the established classification criteria, based upon judgment.

Core damage accident subclass IIA involves a plant transient with loss of all means to remove decay heat from containment using the RHR system and containment vent. For these sequences, core damage is postulated

to occur post containment failure.

The ED would have the following conditions to declare a GE:

1. Primary containment pressure greater than 56 psig and rising.

Note that for core damage accident subclass IIA events, the time required to exceed 56 psig is 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; this precedes

containment overpressure failure by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and core damage by more than 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> (based on plant transient MAAP case JAF-TW-1).

JAFP-11-0119 Enclosure Page 16 of 19 Table RAI-3 Basis for General Emergency Declaration for Accident Class II Sequences Accident Class Designator Definition Conditions for General Emergency Declaration Timing of General Emergency Declaration 2. Potential loss in maintaining RPV water level above top-of-active fuel due to inadequate RPV injection (due to NPSH conditions inside the torus).

Based on these conditions, many hours are available for evacuation after reaching the plant conditions requiring declaration of a General Emergency such that effective

evacuation would be expected. Despite this, we have performed a sensitivity analysis assuming that 95 percent of the time a GE would be declared in sufficient time for effective evacuation (assumed to be 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before release) and 5 percent of the time a GE would be declared without at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

warning for evacuation.

Class IIL Loss of containment heat removal with RPV breach but no initial core damage;

core damage after

containment failure.

9.1.8 General Emergency Any event, as determined by the Shift Manager or Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third, refer to "Fission Product Barrier Loss/Potential Loss

Matrix" This EAL allows for classification of emergencies that do not fit into any other GE criteria. The EALs of this category provide the ED or SM the latitude to classify emergency conditions consistent with the established classification criteria, based upon judgment.

Core damage accident subclass IIL involves a LOCA with loss of all means to remove decay

heat from containment using the RHR system and containment vent. For these sequences core damage is postulated to occur post

containment failure.

JAFP-11-0119 Enclosure Page 17 of 19 Table RAI-3 Basis for General Emergency Declaration for Accident Class II Sequences Accident Class Designator Definition Conditions for General Emergency Declaration Timing of General Emergency Declaration The ED would have the following conditions to declare a GE:

1. RCS leakage greater than 50 gpm inside the drywell.

2. Primary containment pressure greater than 56 psig and rising.

Note that for core damage accident subclass IIL events, the time required for primary containment pressure to exceed 56 psig) is 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, this precedes containment overpressure failure by 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, and core damage by more than 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> (based on large LOCA MAAP case

JAF-AW-1).

3. Potential loss in maintaining RPV water level above top-of-active fuel (due to inadequate RPV injection due to NPSH conditions inside the torus).

Although the time available for this subclass is shorter than for Class IIA, based on these conditions, there are still many hours available

for evacuation after reaching the plant conditions requiring declaring a General Emergency such that effective evacuation would be expected.

Therefore, given the somewhat shorter time JAFP-11-0119 Enclosure Page 18 of 19 Table RAI-3 Basis for General Emergency Declaration for Accident Class II Sequences Accident Class Designator Definition Conditions for General Emergency Declaration Timing of General Emergency Declaration available, we have performed a sensitivity analysis assuming that 80 percent of the time a GE would be declared in sufficient time for evacuation (assumed to be 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before release) and 20 percent of the time a GE would be declared without at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

warning for evacuation.

Class IIV Main condenser and RHR fail and the torus vent opens for containment

pressure relief. Core damage occurs some time following torus vent

initiation.

9.1.8 General Emergency Any event, as determined by the Shift Manager or Emergency Director, that could lead or has led to a loss of any two fission product barriers and loss or potential loss of the third, refer to "Fission Product Barrier Loss/Potential Loss

Matrix" This EAL allows for classification of emergencies that do not fit into any other GE criteria. The EALs of this category provide the ED or SM the latitude to classify emergency conditions consistent with the established classification criteria, based upon judgment.

Core damage accident subclass IIV involves a LOCA or plant transient with subsequent loss of containment decay heat using the RHR system followed by su ccessful containment venting. For these sequences core damage is postulated to occur post containment venting due to other mitigating system failures.

The ED would have the following conditions to declare a GE:

1. Primary containment pressure greater than 56 psig and rising.

JAFP-11-0119 Enclosure Page 19 of 19 Table RAI-3 Basis for General Emergency Declaration for Accident Class II Sequences Accident Class Designator Definition Conditions for General Emergency Declaration Timing of General Emergency Declaration Note that for core damage accident subclass IIV events, the time required for primary containment pressure to exceed 56 psig is 21.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; this precedes core damage by 4.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (based on small LOCA MAAP case JAF-SW2-2).

2. Potential loss in maintaining RPV water level above top-of-active fuel due to inadequate RPV injection (due to NPSH conditions inside the torus).

3. RCS leakage greater than 50 gpm inside the drywell (if initiator is an LOCA).

Based on these conditions, it's likely that a GE would be declared in sufficient time for

evacuation.

However, the total Cesium Iodine (CsI)release for core damage accident Class IIV is less than (10% of the total CsI ) of the criteria used to define a large release (for the FitzPatrick PRA), therefore, core damage accident Class IIV sequences would continue to be binned as non-LERF accident sequences, regardless of

timing.