RS-16-254, Supplemental Information Regarding Request for License Amendment to Revise Technical Specifications Section 5.5.12 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies

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Supplemental Information Regarding Request for License Amendment to Revise Technical Specifications Section 5.5.12 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies
ML16350A432
Person / Time
Site: Quad Cities  Constellation icon.png
Issue date: 12/15/2016
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF8387, CAC MF8388, QC-LAR-04, RS-16-254
Download: ML16350A432 (37)
v • d • e


Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-16-254 December 15, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Supplemental Information Regarding Request for License Amendment to Revise Technical Specifications Section 5.5.12 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies

References:

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Request for License Amendment to Revise Technical Specifications Section 5.5.12 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies," dated September 19, 2016
2. Letter from E. A. Brown (U.S. NRC) to B. C. Hanson (Exelon Generation Company, LLC), "Quad Cities Nuclear Power Station, Units 1 and 2 -

Acceptance Review Concerning the Permanent Extension of Type A and Type C Leak Rate Test Frequencies (CAC Nos. MF8387 and MF8388)

(PID L-LA-2016-29559) (RS-16-187)," dated December 2, 2016 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, respectively. The proposed change revises Technical Specifications (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," to allow for the permanent extension of the Type A Integrated Leak Rate Testing (ILRT) and Type C Leak Rate Testing frequencies.

In Reference 2, the NRC requested additional information that is needed to support the acceptance review of the license amendment request. In response to this request, EGC is providing the attached information.

EGC has reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in Attachment 1 of Reference 1. The additional information provided in this submittal does not affect the bases

December 15, 2016 U.S. Nuclear Regulatory Commission Page2 for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of December 2016.

Attachment:

QC-LAR-04, "Quad Cities ILRT Acceptance Review Supplemental Information" cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector- Quad Cities Nuclear Power Station Illinois Emergency Management Agency- Division of Nuclear Safety

ATTACHMENT QC-LAR-04, "Quad Cities ILRT Acceptance Review Supplemental Information"

RM DOCUMENTATION APPROVAL FORM RM DOCUMENTATION NO.. REV: 0 PAGENO. 1 STATION: Quad Cities Nuclear Power Station (QCNPS)

UNIT(*) AFFECTED: 1 and 2 TITLE: Quad Cities II.RT Acceptance Review Supplemental Information

SUMMARY

This document provides Exelon Generation's responses to NRC's request for supplemental information regarding Quad Cities ILRT License Amendment Request (LAR). The NRC,s request focuses on PRA quality.

This Is a Category I RM Document IAW ER-AA-600-1012. which requires independent review and approval.

[ l Review reQulred after periodic Update

[ X] Internal RM Documentation [ ] External RM Documentation Electronlc Calculation Data FHes: NIA Method of Review: [ X] Detalled [ ] Alternate [ ] Review of External Document Thi* RM documentation supersedes: NIA In Its entirety.

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Print s~n q\KF~rc Date QC-LAR-04 1

Response to Pre-RAIs Associated with ILRT Analysis for Quad Cities Nuclear Power Station

Background

Exelon Generation Company, LLC (EGC) submitted a License Amendment Request (LAR) for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 on September 19, 2016 to revise plant Technical Specifications to allow a permanent extension of the Type A Containment Integrated Leak Rate Test (ILRT) and Type C Leak Rate Test frequencies. Changes to the Type A ILRT frequency are supported by QCNPS probabilistic risk assessment (PRA) information and analysis.

The technical adequacy of the PRA used in developing QCNPS Units 1 and 2 LAR, QC-LAR-03, is consistent with the requirements of Regulatory Guide (RG) 1.200 Revision 2 (An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,) or where gaps exist, the gaps have been addressed, as is relevant to the ILRT interval extension. RG 1.200 endorses, with clarifications and qualifications, American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard ASME/ANS RA-Sa-2009. Section 1-5.4 of the PRA Standard discusses PRA maintenance and PRA upgrades and states:

  • Changes in PRA inputs or discovery of new information identified pursuant to 1-5.3 shall be evaluated to determine whether such information warrants PRA maintenance or PRA upgrade. (See Section 1-2 for the distinction between PRA maintenance and PRA upgrade.)
  • Upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the Peer Review Section of each respective Part of this Standard, but limited to aspects of the PRA that have been upgraded.

Section 1-2 of the PRA Standard provides the following definitions for PRA maintenance and PRA upgrade:

  • PRA maintenance: the update of the PRA models to reflect plant changes such as modifications, procedure changes, or plant performance (data).
  • PRA upgrade: the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progressions sequences. This could include items such as new human error analysis methodology, new data update methods, new approaches to quantification or truncation, or new treatment of common cause failure.

Appendix 1-A of the PRA Standard provides additional information and examples for distinguishing between PRA maintenance and PRA upgrade.

QC-LAR-04 2

NRC Questions Based on the information provided in the QCNPS LAR, the NRC staff, in a letter to EGC dated December 2, 2016, requests supplemental information to demonstrate the technical adequacy of the QCNPS PRA, as follows:

1. Table A-1 of Appendix A located in Attachment 3 to the LAR appears to provide a summary of gaps from a self-assessment performed against RG 1.200, Revision 0. To demonstrate technical adequacy of the internal events PRA against RG 1.200, Revision 2, at Capability Category I, as required for the ILRT LAR, provide the following:

(a) A list of Facts and Observations (F&Os) from all applicable peer reviews and self-assessments of the internal events PRA.

(b) For each F&O, include details of its disposition, and if open, provide an explanation of why not meeting the corresponding Capability Category I requirements has no impact on the application.

2. Appendix A located in Attachment 3 to the LAR states that the 2014A PRA model is the result of upgrading the PRA model in resolution of the self-assessment gaps, and the self-assessment F&O HR-D3 appears to indicate a possible upgrade to the pre-initiator HRA [human reliability analysis].

As defined in the PRA standard American Society of Mechanical Engineers and American Nuclear Society (ASME/ANS) - RA-Sa-2009, a PRA upgrade is the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Further, according to the PRA standard, new should be interpreted as new to the subject PRA even though the methodology in question has been applied in other PRAs (Section 1-A.1 of ASME/ANS RA-Sa-2009). For example, employing a different HRA approach to human error analysis constitutes a PRA upgrade as discussed in Example 24 in Section 1-A.3.24 of ASME/ANS RA-Sa-2009. Per the PRA standard, peer reviews are required for all PRA upgrades, therefore, the following information needs to be provided:

(a) Describe all changes, including any new analyses or incorporation of new methodology performed in the internal events PRA model since the latest full-scope peer review from 2000 and justify whether any of these changes fits the definition and criteria of ASME/ANS RA-Sa-2009 for a PRA upgrade.

(b) If any of these changes fit the definition and criteria of ASME/ANS RA-Sa-2009 for a PRA upgrade, perform a focused scope peer review on the affected supporting requirement, and provide the F&Os with a description of the impact on the ILRT application.

QC-LAR-04 3

Response to Question 1 As described in Appendix A of the LAR, QCNPS has conducted two independent peer reviews on its internal event PRA models.

The first independent PRA peer review was conducted under the auspices of the BWR Owners Group in 2000, following the Industry Peer Review process used at that time. This peer review included an assessment of the PRA model maintenance and update process. The Facts and Observations (F&Os) associated with the 2000 peer review are provided in Table 1.

Each F&O was assigned a level of importance, summarized as follows:

  • A - Extremely important and necessary to address to assure the technical adequacy of the PRA.
  • B - Important and necessary to address but may be deferred to the next PRA update.
  • C - Marginal importance but considered desirable to maintain maximum flexibility in the PRA applications.
  • D - Editorial or minor technical item.

NEI 05-04 Revision 2 (PRA Peer Review Process) indicates that an A or B F&O using the peer review process applicable at the 2000 peer review would equate to a Finding in the latest Peer Review process, and a C or D F&O using the previous peer review process would equate to a Suggestion in the latest Peer Review process. Accordingly, Table 1 addresses only the A and B F&Os from the 2000 Peer Review (i.e., those that would be categorized as Findings in the current process). It is also noted, that the 2000 Peer Review had no A F&Os and 26 B F&Os. All B F&Os were addressed as part of the 2005A PRA update. It is additionally noted that all C and D F&Os were addressed as part of the 2005A PRA update.

A second independent Focused Scope PRA Peer Review of the QCNPS Internal Flooding PRA model was performed in 2010 using the NEI 05-04 process, the ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009), and RG 1.200, Rev. 2. The F&Os from this focused peer review are provided in Table 2.

For each F&O in Tables 1 & 2, the current status related to the finding is noted (including details of its disposition), and its judged importance to the ILRT application. No F&Os were developed based on self-assessments. Identified gaps based on the self-assessment were provided previously in Table A-1 of the LAR.

All peer review F&Os have been addressed. Therefore, no explanation is required for not meeting Capability Category I requirements.

QC-LAR-04 4

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 1 Consider including loss of one division of DC as a IE-5 Addressed. Loss of one division of DC incorporated as No impact. The F&O is special initiator with its own event tree development. It part of the 2005A PRA update. Refer to documentation addressed.

may be beneficial to include the loss of one DC division in Initiating Event Notebook (QC PSA-001) and Event as a separate initiator since even though it may not trip Tree Notebook (QC PSA-002).

the plant, it would lead to a manual plant shutdown with (Ref. URE QC2000-006) degraded components if the DC division were not recovered.

The documentation states that this would be a low contributor to CDF based on the conditional core damage probability for the loss of feedwater events, but it would probably be best to verify this by including it as its own initiator to eliminate any uncertainty.

2 The on-line surveillance testing of interfacing system IE-14 Addressed. The methods used in NUREG/CR-5124 are No impact. The F&O is valves has not been included in the development of the applied and adequate for the assessment of ISLOCA. addressed.

ISLOCA initiating event frequency analysis. This could The ISLOCA assessment has been updated using Note ISLOCA releases be a dominant contributor to the frequency analysis. generic failure rates from NUREG/CR-6928. No further are EPRI Class 2 and evaluation of ISLOCA is required. as such, do not (Ref. URE QC2000-007 and URE QC2005-005) contribute to ILRT class 3b LERF.

3 The end states for the %DLOOP tree, DLOP-17, DLOP- AS-14 F&O addressed with 2005A PRA Update. The PRA No impact. The F&O is 20, DLOP-26 and DLOP-27 do not reflect the model has been revised. DLOOP end states reflect the addressed.

appropriate plant damage class (as indicated in the appropriate plant damage class.

ATWS presentation by the utility.)

(Ref. URE QC2000-011).

QC-LAR-04 5

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 4 There is conflicting information regarding the need for TH-10 F&O addressed with 2005A PRA Update. BSA-Q-96-01 No impact. The F&O is room cooling for RCIC. Calculation BSA-Q-97-04 implies that room cooling for RCIC would be required if addressed.

seems to indicate that it is not required, but BSA-Q CS and RCIC are both running. This is not a modeled (documentation only).

01 would indicate that it is required. (See QC-PSA-006, configuration due to the operators terminating CS if both Note 4 to Table 2-12) CS and RCIC are available. Documentation in Dependency Notebook states this.

(Ref. QC2000-013) 5 Revise the success criteria tables to match with the TH-12 F&O addressed with 2005A PRA Update. Table 3-1c No impact. The F&O is draft write-up on ATWS success criteria and check the ATWS Success Criteria provides the success criteria addressed.

PRA modeling is consistent with the success criteria. description and supporting text. Regarding FW, several footnotes provide success criteria associated with FW:

The success criteria notebook (QC-PSA-003)

Footnote 2 documents SLC initiation is a manual documents the ATWS criteria in Tables 3-1b. However, operation and also requires reactor water level control in the success criteria description or supporting text has the model. Footnote 11 documents FW as an not been provided. This makes it difficult to trace the acceptable injection system. Footnote 9 Loss of only bases for the ATWS success criteria.

FW does not prevent successful mitigation of ATWS.

Subsequent to this observation, written description of Remaining systems include HPCI and the low pressure ATWS success criteria (draft) was provided to the injection systems for coolant makeup.

Certification team. The write-up is and the success criteria tables are slightly inconsistent in terms of the Note 3 states, SLC initiation may be delayed for a need for FW Trip or runback. It appears that based on relatively long period of time (e.g., hours, characterized new information, the PRA team has decided that FW here as 40 min.) if sufficient turbine bypass capability trip is not required. The success criteria Tables still exists (i.e., >25%) and feedwater can be controlled.

indicate that FW runback is needed. It is not clear how [NEDE 24222] QC has approximately 34% TBV this issue was modeled in the PRA. capacity for the EPU configuration.

(Ref. URE QC2000-014)

QC-LAR-04 6

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 6 Various piping attached to the CCSTs (i.e. RCIC suction SY-6 There is now a locked fence and vehicle barriers No impact. The F&O is line) is not protected against inadvertent collision with protecting the CCST and area piping. The QC SRME addressed.

forklifts or other small vehicles. The presence of a performed a walkdown on 5/11/16 to confirm these vehicle and tire tracks indicate that this is a real barriers are still in place.

probability. This is a vulnerability that could cause a (Ref. URE QC2000-19) common cause failure of the CST and suction source for several systems.

Strongly suggest to station personnel that a vehicle barrier is important to prevent the potential issue from occurring. If this is not possible, include the vulnerability in the PSA model.

7 To maintain the quality of the model, make a procedural SY-26 F&O addressed on 10/18/2001. The Design Change No impact. The F&O is requirement that includes PSA Engineering in the Procedure requires a review to determine if the PRA is addressed.

review process. impacted. This procedure also provides criteria for screening and a requirement for risk management There is a procedure that requires PSA Engineering to personnel to create an Updating Requirements be interfaced with for changes made to the plant. There Document (URE) if the Design Change cannot be is not a procedural requirement to include PSA screened as no impact.

Engineering in the changes made to procedures, (Ref. URE 2000-025) surveillances, instructions, etc. that could affect the CDF.

8 The data should be updated on a continuing basis. DA-6 F&O addressed with 2005A PRA Update. Note, data is No impact. The F&O is updated each PRA update. Scheduled updates occur at addressed.

The component probabilities data used in the evaluation approximately 4 year intervals.

are based on accumulated plant specific experience.

However, the last three years experience has not been included in the accumulated data. (See also URE 2000- (Ref. URE QC2000-028) 16)

QC-LAR-04 7

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 9 Prepare operator interviews or question sets to verify HR-14 The F&O was addressed with incorporation of a No impact. The F&O is that the QGA interpretation was done appropriately for structured interview process in the 2001 HRA Update. addressed.

the dominant HEPs in the model. Note, operator interviews are conducted for each scheduled update.

The Human Reliability Analysis relied on the analysts (Ref. URE QC2000-030) review and interpretation of the QGAs and other procedures. A major enhancement to the scrutability of the analysis would be to factor a structured interview or question process into the analysis and documentation.

This could also help to support the cases where execution time was estimated.

10 Flooding is an important issue that merits further DE-9 The F&O was addressed with information from a system No impact. The F&O is investigation by the Quad Cities PSA staff. At least two manager interview in 2002 and issuance of a flood addressed.

potential vulnerabilities exist at the Quad Cities Station model in 2005. It should be noted that the Focused in respect to internal flooding. 1) A rubber boot secured Internal Flooding peer review was performed in 2010.

by hose clamps on the RCIC suction line (torus room F&Os from that Peer Review are included in Table 2.

side). 2) Ventilation penetrations below the maximum Also noted, Eric Jebsen, Site Risk Management postulated torus room flood zone. Engineer, interviewed System Engineer Jim Wethington 2/14/02 who stated that the rubber boot was designed to

1) According to the system engineer, surveillance with a resist 11 ft. of water in the torus room. Its quality and cyclic frequency exists that cues him to perform a performance are monitored, and it is visually inspected walkdown of the torus room. PSA group might want to every 18 months.

validate that the rubber boot seal has some pedigree to (Ref. URE QC2000-033) ensure that it remains functional as long as it is installed.

2) The ventilation ductwork within the core spray room is stepped up such that the top is above the postulated flood level of the torus room. The PSA groups should consider ensuring a calculation has been performed validating the ductwork can support the weight of the water without collapsing and thereby flooding the pump room.

QC-LAR-04 8

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 11 Consider performing a detailed walkdown of the Quad DE-10 Issue addressed in 2001 based on the following: No impact. The F&O is Cities Generating Station of all areas modeled in the Walkdowns were done for IPE. Walkdowns have been addressed.

PSA. Recommend involving an individual with a strong performed for spatial-dependence issues, such as knowledge of the station to augment the two relatively revised fire study and revised flood study. Beyond this, new PSA Engineers in performance of their walkdown. Exelon analyst experience is sufficient to ensure quality of PRA. Walkdowns have also been performed for other PRA updates.

(Ref. URE QC2000-034) 12 The ATWS event tree and flag settings should be QU-8 Issue addressed in 2005A PRA Model Update. Cutset No impact. The F&O is reviewed for adequacy for this sequence. A proper #112 includes the appropriate Operator Action, addressed.

HEP should be applied to this sequence. 1ADOPINHIBIT-H-- Operator Fails to Inhibit ADS - (No FW Injection).

In the top 100 sequences, cutset 96 is an ATWS (Ref. URE QC2000-042) scenario with mechanical scram failure, following loss of feedwater event. Core damage occurs when operator fails to inhibit ADS due to low water level. The HEP credited is for failure to inhibit ADS w/feedwater injecting. This HEP appears to be misapplied for this sequence, since loss of feedwater is the initiator.

QC-LAR-04 9

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 13 The success criteria given one non-SBO diesel QU-15 F&O addressed with 2005A PRA Update. The following No impact. The F&O is available in the dual unit loop case should be examined is documented in the URE database: addressed.

to ensure it is appropriately captured in the model.

In accordance with QCOA 6100-03 the following limitations are identified: EDG: 2600Kw/3250KVA It is not clear that the EDGs (i.e. the non-SBO diesels)

(2860Kw/3575KVA short term); SBODG:

would have sufficient capacity to allow for RHR and 4350Kw/5437KVA (4785Kw/5981KVA short term);

RHRSW pumps to be running for both units in the dual 580amps on 13-1(14-1) to 13(14) and 13-1(14-1) to 23-loss of offsite power scenario.

1(24-1) breakers. The RHRSW pump utilizes 580-725Kw and the RHR pump uses 400-485KW. The EDG has sufficient capacity to power 2 RHR and 2 RHRSW, which is the design. The procedure identifies other loads including optional loads that could be loaded on the EDGs. The RHRSW is only needed to provide containment cooling (several hours into the event).

There is sufficient guidance (load on DGs and Xties) to permit powering the RHR and RHRSW pumps on both units. Several hours into the event would only require the operation of the RHR/RHRSW pumps on an intermittent basis, thus capability would be available to power other loads.

(Ref.URE QC2000-43) 14 An uncertainty analysis is an important part of QU-26 F&O addressed with 2005A PRA Update. An uncertainty No impact. The F&O is evaluating the model. The capability to perform such an analysis was performed to meet EPRI guidance. addressed.

analysis may be required for certain risk-informed (Reference URE QC2000-045) requests in the future. It may also be beneficial to include the need to perform an uncertainty analysis in the maintenance and update procedure.

QC-LAR-04 10

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 15 To establish reasonableness of the model development QU-26 F&O addressed with 2005A PRA Update. An uncertainty No impact. The F&O is and methodology employed, sensitivity and uncertainty analysis was performed to meet EPRI guidance. addressed.

assessments should be performed.

(Reference URE QC2000-045)

No special sensitivity or uncertainty cases have been performed. Typical sensitivity studies accompany the dominant sequences, initiators or other modeling feature (such as vessel rupture initiating event) which may dominate the uncertainty of the CDF point estimate of CDF. This information is needed to establish the acceptability of the final results.

16 Carry the support system dependency from L1 PRA to L2-7, 13 Issue addressed in 2002 PRA Update with new L2 No impact. The F&O is L2 PRA in the next revision. structure that carries over support systems to the L2 addressed.

PRA model. This structure has been preserved.

The transfer from L1 to L2 PRA is done by carrying the (Reference URE QC2000-046) plant damage state frequencies. The dependency of systems examined in the L2 PRA on the support systems that might have been disabled in the L1 PRA cannot be accounted for accurately.

17 Consider linking the Level 1 directly into the Level 2 L2-7, 20 F&O addressed with 2005A PRA Update. The Level 1 is No impact. The F&O is evaluation. now directly linked to the Level 2. addressed.

(Ref. URE QC2000-047)

The Level 2 evaluation is not directly linked with the Level 1 results. Instead, LERF multipliers are applied to each of the sequences from the Level 1. A more detailed linking approach may be desirable, especially with the currently high conditional LERF value of (~0.7 with Class 2 events included, ~0.4 without Class 2 events included). The conservatisms included by not directly linking the cutsets may be severely limiting for intended applications.

QC-LAR-04 11

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 18 Perform a more detailed L2 PSA. L2-24 F&O addressed with 2005A PRA Update. The Level 1 is No impact. The F&O is now directly linked to the Level 2 and the factors are no addressed.

The simplified approach of focusing only of LERF longer used.

sequences by using factors simplifies the approach (Ref. URE QC2000-049) significantly and possibly makes it conservative.

Because of this simplification, the approach may be too conservative for Risk-informed applications.

19 Develop working level procedures for each step in the MU-1, 3 Issue addressed in 2001 with updated Technical and No impact. The F&O is update process. Pay special attention to the steps Requirements (T&RM) guidance documents. addressed.

performed by plant personnel, especially those outside (Ref. URE QC2000-50) engineering.

Currently, these documents are now supplemented with The existing guidance in NEP-17-04 is at a Best Practice documents.

programmatic, conceptual level. There are very few working level documents in place to allow the update team to perform PSA updates without relying solely on the skill of the craft.

QC-LAR-04 12

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 20 Include the review elements listed below in the model MU-4 Issue addressed in 2002 with issuance of ER-AA-600- No impact. The F&O is update procedure and the implementing procedures for 1022, "FPIE PRA Model Capabilities. It requires that addressed.

NEP-17-04s high level guidance. Exelon PRA's meet Certification guidelines and the ASME PRA Standard.

Enhance the data collection phase of the update (Ref. URE QC2000-053) process to include the following elements that the Monitoring and Collecting New Information sub- Current data collection, monitoring of procedure element suggested was missing from the update changes (including emergency plan changes), industry procedure: studies, etc. are judged to conform with the current ASME/ANS Standard requirements.

Operating experience New maintenance policies Operator Training Program Emergency Plan changes Accident Management Programs Industry Studies Especially important in this process will be the true integration of the system engineers and operations personnel into the update process.

QC-LAR-04 13

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 21 Guidance needs to be provided which requires the MU-6 Issue addressed in 2001 with issuance of T&RM No impact. The F&O is designation of a pristine copy of the official model, to guidance documents. addressed.

be stored in a controlled manner. Controlled storage Guidance is currently found in ER-AA-600-1014 [URE options include offsite storage of the model on magnetic QC2000-053]

media, backup on computer network drives with special controls requiring special keys involving Information Services personnel, or non-re-writeable CD ROMs.

The update of PSA model files is mentioned in NEP 04, but the control of these models is not addressed.

The control of the FORTE and EOOS codes have been addressed in draft form to comply with NSP-CC-3021, but there is no guidance for the safeguard of the PSA model files.

PSA models and sensitivity studies are not stored in a controlled manner, and the official copies of the PSA models are limited to those available on the analysts personal computers hard drive and floppy disk box.

These two storage locations do not provide the safeguards needed to ensure model fidelity.

QC-LAR-04 14

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 22 Section 5.2.2 of NEP-17-04 describes the periodic MU-7, 11 Issue addressed in 2001 with issuance of updated No impact. The F&O is update process. The following findings apply to this guidance documents. T&RMs specify necessary addressed.

section of the process. actions.

(Reference URE QC2000-054)

Because this is a new procedure, the results of the use of this procedure cannot be reviewed. However, the elements of the process outlined can be compared to the certification criteria.

There are three aspects of the update process that are not adequately addressed. The first is the re-evaluation of Past PSA applications. The procedure briefly mentions that past PSA applications should be reviewed and updated as appropriate. This needs to be strengthened, especially if past PSA applications have served as the bases for risk informed Tech Spec or licensing basis submittals. It is vital to determine if PSA model changes will invalidate the bases for submittals made to the NRC.

The second aspect is uncertainty analysis. The update procedure is silent on the need to do uncertainty analyses for PSA model and results update. The reviewers have noted that in order to be successful with risk informed applications in the future, uncertainty needs to be addressed.

The third aspect is handling of the insights from the update, especially in terms of identification of vulnerabilities and enhancements. Detailed guidance is needed for the identification of vulnerabilities discovered during the update process, as well as enhancements for station procedures and other plant programs such as the emergency plan.

QC-LAR-04 15

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 23 Procedure NEP-17-04, as it is written, does not meet MU-9 Issue addressed in 2001 with issuance of updated No impact. The F&O is the intent of this sub-element to have knowledgeable guidance documents. T&RMs specify necessary addressed.

people review the results of the analysis. Particularly actions.

lacking is the input from Operations, Operator Training, ER-AA-600-1022, "FPIE PRA Model Capabilities," was and possibly from outside industry experts.

prepared subsequent to the QC PRA Peer Review/Certification. It requires that Exelon PRA's meet The update procedure could call for the formation of an Certification guidelines and the ASME PRA Standard expert panel, consisting of PSA, operations, operator training, systems engineering, maintenance rule, and (Reference URE QC2000-054).

possibly outside PSA industry expertise to review the results of the PSA update.

QC-LAR-04 16

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 24 This F&O addresses the deficiencies observed in the MU-12 Issue addressed in 2001 with issuance and No impact. The F&O is Maintenance and Update Process seen at the corporate implementation of updated guidance documents. addressed.

offices. T&RMs specify necessary actions.

Commonwealth Edison wrote NEP-17-04 to codify the (Ref. URE QC2000-056)

PSA Maintenance and Update Process for the entire Nuclear Operations Division. As of this Certification visit, this procedure has not been implemented for the Quad Cities PSA, and is only beginning to be implemented for the PSA Group.

The following are observations that pertain to activities performed solely at the corporate offices. They are listed by the steps in NEP-17-04 that apply.

5.1.1.3 The PSA supervisor/designee does not maintain a historical record of PSA updates 5.1.1.4 The supervisor/designee does not maintain an approved listing of the proper software/code for PSA application. The corporate offices do, however, abide by the requirements of NSP-CC-3021 for control of software codes, but this applies only to the EOOS/OSPRE codes that are run from the LAN/WAN.

5.1.1.5 The supervisor/designee does not maintain a record of the qualification of personnel assigned to the update tasks.

5.1.2.2 The PSA analysts are not performing the analysis of plant changes for the Quad Cities PSA model.

5.1.2.4 The PSA analysts are not reviewing the NFS calculation logs on a quarterly basis.

The corporate office needs to begin to implement the procedure as written (with enhancements suggested by this certification) and complement it with detailed implementing procedures for the elements of the update and maintenance process.

QC-LAR-04 17

TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 25 Commonwealth Edison wrote NEP-17-04 to codify the MU-12, 13 Issue addressed in 2001 with issuance and No impact. The F&O is PSA Maintenance and Update Process for the entire implementation of updated guidance documents. addressed.

Nuclear Operations Division. As of this Certification T&RMs specify necessary actions.

visit, this procedure has not been implemented for the (Ref. URE QC2000-057)

Quad Cities PSA, and is only beginning to be implemented for the PSA Group.

The following are observations that pertain to activities performed at the sites. They are listed by the steps in NEP-17-04 that apply.

5.1.3.1 The QC site PSA analyst is not actively involved with the PSA update as part of the update team.

5.1.3.2 The QC site PSA analyst does not perform independent reviews of PSA updates performed for other stations.

5.1.3.4/5.2.3.3 The QC site PSA analyst does not review the site calculation logs on a quarterly basis for impacts to the PSA model.

5.1.5.1 The system engineers do not receive training on the PSA, and therefore cannot be held responsible for being knowledgeable for information contained in the PSA system NBs 5.1.5.2 The system engineers (or the designees) do not receive training on the PSA, and therefore cannot be held responsible for considering the impact of the PSA model when the system engineer initiates or is made aware of plant changes.

5.1.6.1 The procedure writers, operating procedures/policy reviewers, and EOP writers do not receive training on the PSA, and therefore cannot be held responsible for considering the impact of procedure or policy changes on the PSA model.

The lack of quality input from the plant prevents the kind of independent review expected in element MU-13.

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TABLE 1 STATUS OF F&OS SIGNIFICANCE A OR B FROM THE 2000 PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE TO NO. SUBELEMENT APPLICATION 26 The PRA engineer is automatically not in the loop for MU-12 Issue addressed in 2001 with issuance and No impact. The F&O is changes to plant procedures and Technical implementation of updated guidance documents as well addressed.

specifications that could impact the PRA results. It as Design Procedures. Procedures CC-AA-10 and CC-appears that PRA engineer is consulted on plant AA-102 require PRA consideration during plant modifications, though it is not clear that there is a modifications.

formalized procedure for this.

Procedure ER-AA-600 T&RMs require procedural changes be reviewed.

(Ref. URE QC2000-058)

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TABLE 2 STATUS OF FINDING F&OS FROM THE 2010 INTERNAL FLOODING FOCUSED SCOPE PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE NO. SUBELEMENT TO APPLICATION 1-2 The quantitative screening criteria described in Section 2.3.6 IFQU-A3 Addressed. The Quad Cities Internal Flood No impact. The and Section B.3 is less conservative than the screening criteria IFSO-A3 PRA evaluation is thorough, comprehensive, F&O is addressed.

in the ASME/ANS PRA Standard. IFQU-A3 states that a flood and very detailed. The screening process area can be screened if the product of the sum of frequencies follows the qualitative process used cited in of flood scenarios for a flood area calculated based upon the the ASME/ANS PRA Standard in IFSO A1 &

bounding CCDP is less that 1e-9. Section 2.3.6 does not use IFSO A2.

the bounding CCDP but rather sums the CDFs for the floods.

This would result in screening more areas out than would be In addition, very selective use of the screened using the criteria of this SR. The screening criterion quantitative screening criteria is performed.

of this SR is not used.

The documentation has been rewritten to more clearly state what is done and how it meets the Standard.

(Ref. URE QC2011-006) 1-6 Potential sources of flooding have been identified for each flood IFSO-A1 Addressed. Evaluated and added text to No impact. The area, and include consideration of equipment such as pumps, IFSN-A1 Appendix D to identify that flood scenarios F&O is addressed.

tanks, piping, etc. Flooding from external sources and in- IFSN-A8 via inadvertently open corner room drain leakage from other flood areas was also considered. This is lines would not be a significant contributor to documented in Sections 2.3.3, Table B.3-3 of the IF Notebook. (Note: All three SRs risk. Failure of the ball drain lines is not were determined MET explicitly modeled.

Check valve backflow failure rate is 1e-3/demand. QC does not at Capability Category currently rely on check valve modeling as RHRSW has 3 valves II) in series, and the corner rooms have a check valve and In addition, deleted references to backflow normally closed ball valve in series. check valves in drain lines because these do not exist in QC corner room drain lines.

However, the drain ball valves are routinely operated by (Ref. URE QC2011-007) operators to remove water from the corner room drains and could potentially be left open. Leak testing occurs once each 4 years. (see procedures QCOS 0020-04 and QCOS 0010-11).

The potential for human error in leaving the ball valves open is not addressed in the model or documentation.

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TABLE 2 STATUS OF FINDING F&OS FROM THE 2010 INTERNAL FLOODING FOCUSED SCOPE PEER REVIEW F&O DESCRIPTION OF FINDING APPLICABLE CURRENT STATUS / COMMENT IMPORTANCE NO. SUBELEMENT TO APPLICATION 1-7 The documentation of EQ equating to operability during a spray IFSN-A7 Addressed. The description in 2.2.13.3 of the No impact. The event does not meet the standard for expert judgment. QC Internal Flooding Notebook (QC-PSA- F&O is addressed.

Additionally, there is no data or analysis to justify this position. 012) is clarified to indicate that no credit is This was the only case identified in which this SR was used. included for survivability of instruments that are sprayed.

(Ref. URE QC2011-013)

QC-LAR-04 21

Response to Question 2 There have been five updates to the internal events PRA model since the performance of the 2000 peer review of the 1998-1999 (i.e., QC99A) PRA model. The CDF and LERF values from the peer reviewed model and model of record updates since the peer review are provided below.

Models U1 CDF U1 LERF U2 CDF U2 LERF Q199A and Q299A 4.6E-06 3.3E-06 4.6E-06 3.3E-06 (2000 Peer Review Model)

Q102C and Q202C(1) 2.2E-06 2.6E-07 2.2E-06 2.6E-07 Q105A and Q205A 7.36E-06 6.64E-07 7.36E-06 6.64E-07 Q105B and Q205B (2010 Flood Peer Review 5.56E-06 6.47E-07 5.56E-06 6.47E-07 Model)

Q110A and Q210A 2.93E-06 4.02E-07 2.93E-06 4.02E-07 Q114A and Q214A 2.91E-06 1.97E-07 2.91E-06 1.97E-07 A summary of changes made to the internal events PRA models of record are documented in the introduction sections of the respective FPIE PRA Update Quantification Summary Notebooks (QC-PRA-013). Additionally, changes in the FPIE PRA models are tracked in the Quad Cities Updating Requirements Evaluation (URE) database which tracks PRA observations and open items identified in between scheduled FPIE PRA Update periods. URE items that involve model changes that were addressed in the updated models are also listed and described in the Quantification Summary Notebooks. The summary of changes and table of UREs involving model changes were used to identify the changes made since the Q199A and Q299A models were peer reviewed.

A summary of the identified changes for each of the PRA model updates since the 2000 peer review are discussed below. For each change, a discussion is also provided to identify whether the change item is PRA Maintenance or PRA Upgrade. A reference to the specific PRA Standard Appendix 1-A Example in which the change item relates is provided when possible.

These changes, including any new analyses or incorporation of new methodologies performed in the internal events PRA model since the last full-scope peer review from 2000, were reviewed for this RAI response. The following changes meet the definition and criteria of ASME/ANS RA-Sa-2009 for a PRA upgrade:

1. Level 2 Linking to Level 1(Item 2) - The Level 2 modeling was modified as part of the 2002 update to include direct linking of Level 1 and Level 2 modeling, along with support system dependencies.

(1)

Includes changes associated with earlier 2002 model changes for models Q102A, Q102B, Q202A, and Q202B.

QC-LAR-04 22

  • Impact to ILRT - The model changes to link the Level 2 model logic to the Level 1 logic (as compared to only modeling Level 2 point estimates within the Level 1 logic) would have negligible impact on the Level 2 results. The Level 2 logic was, in general, of equal level of detail prior to and after linking. It is also noted that in using the simplified EPRI methodology, the risk assessment results are more dependent upon CDF leading to a containment intact condition (where containment leakage is of material interest) than differentiating between different containment failure states (where potential containment leakage is irrelevant for a failed containment). Therefore, potential impacts to the ILRT risk assessment associated with the model change to link Level 2 logic to Level 1 logic are judged to be negligible.
2. Support System Initiators (Item 81)- Fault trees for Support System Initiating Events (SSIE) for TBCCW, RBCCW, and SW were incorporated directly into the main model for the 2014A update. Prior to this, these SSIE Fault Trees were separate fault tree files, which were quantified individually to provide a point estimate value for the initiator in the main model.
  • Impact to ILRT - Similar to the Level 2 logic model linking changes, the incorporation of the support system initiating event (SSIE) logic within the main model logic (as compared to only including the SSIE point estimate results in the Level 1 model) would have negligible impact on the risk results. The SSIE logic was, in general, of equal level of detail prior to and after linking. Therefore, potential impacts to the ILRT risk assessment associated with the SSIE model change are judged to be negligible.
3. Internal Flood Model Logic Integration (Item 45) - Internal Flood accident sequences were integrated into the full power internal events model as part of the 2005A update.
  • Impact to ILRT - Similar to the Support System Initiator logic model linking changes, the incorporation of the internal flood logic within the main model logic (as compared to only including the separately quantified flood initiator point estimate results in the Level 1 model) would have negligible changes to the risk results. Additionally, the internal flooding focused scope peer review occurred after this logic change and was therefore independently reviewed. The F&Os from that peer review are provided and assessed in the response to RAI #1.

Therefore, this upgrade has been adequately addressed for the ILRT application.

Appendix A in Attachment 3 to the LAR states for HR-D3 that, a possible upgrade to the pre-initiator HRA to include specific quantifications for each pre-initiator HEP would be strict compliance with the standard. This wording dates back to the 2004 Self-Assessment for Quad Cities. The use of the word upgrade in this gap assessment is inadvertently misleading. It was meant to signify an improvement to the pre-initiator HRA, not an upgrade as defined in the PRA Standard which would require a methods change and thus a Peer Review. The pre-initiator calculations have been expanded to include more components, however the same methodology has been used since the 2000 Peer Review.

As noted previously, a PRA upgrade is defined as the incorporation into the PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Appendix 1-A of the PRA Standard also notes that consideration should be given to the scope or number of PRA maintenances performed, and that the integrated nature of several changes may make a peer QC-LAR-04 23

review desirable. Although the changes to the QCNPS PRA models since the 2000 Peer Review have not involved new methods, multiple model revisions have occurred over the intervening years. These revisions have resulted in changes to the risk metrics of CDF and LERF, as presented above. For this reason, QCNPS plans to perform a full-scope Peer Review to the PRA Standard in early 2017, primarily to support more involved risk informed applications such as 50.69 and Initiative 4b which generally require that the PRA Standard supporting requirements meet Capability Category II. In comparison to these applications, the ILRT application uses a simplified risk assessment approach (i.e., NRC endorsed EPRI methodology) with conservatisms incorporated into the approach (e.g., all EPRI Class 3b frequency is LERF).

An ILRT application is only required to have the PRA Standard supporting requirements meet Capability Category I. As indicated in Table 1-1.3-2 of the PRA Standard, Capability Category I requirements (as compared to Capability Category II requirements) reflect a lower required scope and level of detail (e.g., risk contributors at the system and train level are acceptable),

less plant-specificity (e.g., use of generic data / models is acceptable), and less realism (e.g.,

departures from realism are allowed to have a moderate impact on conclusions). Using the simplified EPRI methodology, an ILRT risk evaluation has less sensitivity to PRA model details.

Since none of the model upgrades identified above are judged to substantially impact the ILRT risk assessment, the existing PRA technical adequacy is judged to adequately support the ILRT application.

QC-LAR-04 24

Q102C and Q202C Model Changes The Q102C and Q202C models were developed as a result of a regular scheduled update.

Major changes incorporated into the model include the following data, plant, procedure, and analysis changes:

1. Extended Power Uprate (EPU) plant configuration and MAAP 4.0.4 analysis o PRA Maintenance (Examples 23 & 17). HEPs are updated using the same HRA methodology. No new methods are employed.
2. Level 2 modeling modified to include direct linking of Level 1 & Level 2 modeling along with support system dependencies.

o PRA Upgrade (Example 12) This change is judged to reflect significant model manipulations to link the Level 2 logic to the Level 1 logic. This change was in response to Peer Review F&Os. Although the impact on the ILRT would be negligible since the level of detail in the model generally remained the same before and after linking, it could constitute a PRA Upgrade.

3. Revised human reliability analysis (HRA) based on the most recent operator interviews and comments from SRME o PRA Maintenance (Example 20 & 23). HEP modeling updated, Internal Flooding operator actions added to model in accordance with existing HRA methodology, no new methodology employed.
4. Completed URE, OPEX, and Nuclear Operations Notification (NON) review efforts o PRA Maintenance. Addressing plant and procedure changes (Example 22) as well as correcting identified errors (Example 6) and incorporating minor enhancements (Example 7) represent normal update activities. No new methodologies were employed.
5. Maintenance unavailability data based on the most recent plant operating experience o PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.
6. Bayesian updated initiating event frequencies utilizing Quad Cities most recent operating experience o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed for such data.
7. Individual component random failure probabilities Bayesian updated (as applicable) based upon the most recent plant specific data and the most current generic sources o PRA Maintenance (Examples 2 and 3). Although this was the first Bayesian Update for random failure probabilities used in Component Data, the same Bayesian Update methodology that was used for Initiating Event frequencies was employed. Therefore, this would not be a new methodology just an expansion of current methodology to a different data set.

QC-LAR-04 25

8. Common cause failure (CCF) calculations revised to incorporate the updated individual random basic event probabilities and the most up to date Multiple Greek Letter (MGL) parameters from NUREG/CR-5497 and NUREG/CR-5485 o PRA Maintenance (Example 3 and 26). Using new data, no new methodology employed.
9. Revised LOOP/DLOOP analysis for initiating event frequencies and non-recovery probabilities based upon a Midwest regional data filtering approach o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.
10. Revised DC distribution system CCF modeling (CCF events set to zero) to prevent double counting o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.
11. Revised mechanical and electrical ATWS probabilities, based on information in NUREG/CR-5500 o PRA Maintenance (Example 2). Using new generic data, no new methodology employed.
12. Credit for repair/recovery of RHR for long term loss of DHR events o PRA Maintenance (Examples 7 & 9). Logic model enhancement to address an omission. No new methodology employed compared to the prior model.
13. Revision of medium water LOCA and medium steam LOCA event trees to reference appropriate RHR repair/recovery logic o PRA Maintenance (Examples 6 & 7). Logic model enhancement, no new methodology employed compared to the prior model.
14. Change of HEP for Diesel Fire Pump pre-initiator (basic event BFPHU4101ABH--)

from 5.5E-2 to 1E-4 to be consistent with the QC HRA Notebook o PRA Maintenance (Examples 6 & 20). HEP modeling enhancement, no new methodology employed.

15. Revision of loss of single 125 VDC bus initiating event frequencies to be consistent with the QC Initiating Events Notebook o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.
16. Inclusion of minor revisions to basic event text descriptions to address Exelon and ERIN comments.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

17. New probabilities for the LERF events based on the 2002 Level 2 notebook were added.

o PRA Maintenance (Example 2). Similar to using new plant-specific data, no new methodology employed.

QC-LAR-04 26

Q105A and Q205A Model Changes The 2005A model was the result of a regularly scheduled update per Exelon Risk Management T&RMs. Major changes incorporated into the model include the following data, plant, procedure, and analysis changes:

18. Bayesian updated initiating event frequencies utilizing the most recent Quad Cities operating experience.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

19. Allocation of LOCA frequencies on a location and size specific basis. (i.e., the LOCA locations have been subdivided for more accurate assessments of their consequences.)

o PRA Maintenance (Example 1). Medium and Large LOCA Initiators were broken up into various initiators depending on location. This is considered a refinement of the data without employing any new methodology.

20. Revised LOOP/DLOOP analysis for initiating event frequencies and non-recovery probabilities including the 2003 Northeast Blackout using the latest INEEL analysis.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

21. A separate Initiating Event fault tree has been developed for the support system induced initiator for loss of RBCCW.

o PRA Maintenance (Example 1) Initiating Events fault trees were already used for Loss of TBCCW and Loss of SW. The addition of an Initiating Event fault tree for loss of RBCCW follows the same methodology previously used.

Loss of RBCCW was previously subsumed in the Turbine Trip initiator.

22. Revised component failure data including extensive use of plant-specific component failure data gathered from the Quad Cities Maintenance Rule program.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

23. Individual component random failure probabilities Bayesian updated (as applicable) based upon the most recent plant specific data and the most current generic sources.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

24. Common cause failure (CCF) calculations revised to incorporate the updated individual random basic event probabilities and the most up to date Multiple Greek Letter (MGL) parameters from NUREG/CR-5497 and NUREG/CR-5485.

o PRA Maintenance (Example 3 and 26). Using new data, no new methodology employed.

QC-LAR-04 27

25. Maintenance unavailability data based on the most recent Quad Cities operating experience.

o PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.

26. Extensive HRA re-assessment based on operating crew interviews using the latest EOPs and support procedures.

o PRA Maintenance (Examples 20 & 22). HEP modeling enhancement, no new methodology employed.

27. Use of MAAP 4.0.5 deterministic calculations to support the success criteria and HRA calculations (i.e., cues and time available for actions) for the Extended Power Uprate (EPU) configuration.

o PRA Maintenance (Example 17). Changed from MAAP 4.0.4 to 4.0.5 which is only a minor code revision rather than implementing a new code or method (e.g., not a change from MELCOR to MAAP).

28. Addition of recirculation pump seal leakage scenarios.

o PRA Maintenance (Example 9). Logic model enhancement (corrects an omission), no new methodology employed compared to the prior model.

29. Addition of about 10 new HRA pre-initiating events consistent with existing events already included in the model.

o PRA Maintenance (Example 20). HEP modeling updated consistent with existing model methods, no new methodology employed.

30. Added alternate configuration logic for systems with alternate / standby trains.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

31. The water hammer potential due to alternate LPCI or CS alignment is explicitly included.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

32. The conditional probability of a DLOOP given a transient or LOCA signal event is incorporated into the PRA modeling.

o PRA Maintenance (Example 14). Logic model enhancement, no new methodology employed compared to the prior model. Conditional probability of DLOOP is small compared to DLOOP frequency from offsite sources and results in a small change in risk insights.

33. Recognition of the potential for SBCS injection of dirty water into the FW reg valve stacked disk design resulting in clogging and ineffective RPV injection.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

QC-LAR-04 28

34. The available FPS calculation for RPV injection was not performed for EPU or with containment backpressure and elevation head accounted for. The analysis of FPS does not support early RPV makeup success and has been deleted from the model.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

35. CCST support of condensate is adequate when the makeup volume and flow rate requirements are small (single unit challenges).

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

36. The potential for reactor scram and failure of automatic initiation signals due to reference leg leak down was added.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

37. Addition of failure mode for HPCI if torus temperature is above 140°F when suction is from the torus.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

38. Shorter HPCI allowed mission time of 8 hrs.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

39. Addition of extreme weather impacts on SBO DG building.

o PRA Maintenance (Example 7). Dependency added to the SBO DG modeling in the form of a basic event representing the conditional probability of a failure given an extreme weather LOOP.

40. Separation of LOOP/DLOOP frequency contributors and AC recovery o PRA Maintenance (Example 1). LOOP/DLOOP Initiating Event frequencies broken out into plant centered, switchyard centered, grid related, and weather related frequencies and associated non-recovery probabilities.
41. Longer DC battery life of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for cases with DC load shed.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

42. Shorter DC battery life of 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for cases without successful DC load shed.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

43. SSMP standby condition with the room cooler defeated due to open bypass around the room cooler (requires local manual action to place SSMP in successful condition for accident mitigation).

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

QC-LAR-04 29

44. RCIC trip under small LOCA conditions as calculated by MAAP indicates RCIC does not provide success for small water or steam LOCA events.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

45. Internal Flood accident sequences have been integrated into the full power internal events model.

o PRA Upgrade (Example 5). This model change is judged to reflect significant model manipulations and could be considered an upgrade.

However, the internal flood modeling was subsequently the subject of a focused scope peer review in 2010. Therefore, this potential upgrade would have been evaluated during that review.

Q105B and Q205B Model Changes The 2005B PRA model revision was a limited interim PRA model to remove conservatism for station Emergency Diesel Generators. The 2005B PRA model update incorporated the following modification:

46. Probability of basic event BSBPH-XWTHR--F-- Failure of the SBO Diesel Generator Building Given Extreme Weather Induced LOOP changed from 1.0 to 0.2 in the basic event database based on a technical evaluation.

o PRA Maintenance (Example 7). Logic model enhancement, no new methodology employed compared to the prior model.

Q110A and Q210A Model Changes The 2010A model was the result of a regularly scheduled update per Exelon Risk Management T&RMs. Major changes incorporated into the model include the following data, plant, procedure, and analysis changes:

47. Bayesian updated initiating event frequencies utilizing most recent Quad Cities operating experience.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

48. Revised component failure data including extensive use of plant-specific component failure data gathered from the Quad Cities MSPI and Maintenance Rule programs.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

49. Individual component random failure probabilities Bayesian updated (as applicable) based upon the most recent plant specific data and the most current generic data sources.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

QC-LAR-04 30

50. Common cause failure (CCF) calculations revised to incorporate the updated individual random basic event probabilities and the most up to date Alpha Factor parameters from NUREG/CR-5497 and NUREG/CR-5485 on the NRC website.

o PRA Maintenance (Example 3 and 26). The change from MGL to Alpha factor method (NUREG/CR-5497) reflects Example 26 because MGL and Alpha methods both correlate to the same result. This change is less extensive than a method change from the Beta Factor approach of Example 27.

51. Maintenance unavailability data based on the most recent Quad Cities operating experience.

o PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.

52. HRA re-assessment based on operating crew interviews using the latest EOPs and support procedures.

o PRA Maintenance (Examples 20 & 22). HEP modeling updated consistent with existing model methods, no new methodology employed.

53. HRA re-assessment based on enhanced thermal hydraulic calculations (e.g., time available to emergency depressurize the RPV).

o PRA Maintenance (Examples 17 & 20). HEP modeling enhancement, no new methodology employed.

54. Performed additional MAAP 4.0.5 deterministic calculations to support the success criteria and HRA calculations (i.e., cues and time available for actions).

o PRA Maintenance (Examples 8 & 10). Refinement for early and late credit for injection source based on thermal hydraulic calculations.

55. Addition of ASD logic to replace recirculation pump MG sets.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

56. Update of the Quad Cities evacuation study for use in evaluating offsite consequence categories (i.e., LERF).

o PRA Maintenance (Example 6). Logic model enhancement, updating already existing modeling, no new methodology employed compared to the prior model.

57. The potential for ECCS pipe rupture due to water hammer events is re-examined.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

58. Revised Small LOCA Water sequences to remove credit for Condensate by itself and add new logic to credit SBCS supply for long term hotwell makeup to support Condensate injection.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

QC-LAR-04 31

59. Internal Flood initiators are recalculated using updated pipe-break frequencies from the 2006 EPRI report.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

60. Added the diesel fuel oil pumps for the Station Emergency diesels into the models as separate components o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.
61. Revised the containment vent valve power supplies and their backups.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

62. Revised mitigation capability during total Loss of 125 VDC events.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

63. Revised SBLC success criteria to require only 1 of 2 SBLC pumps to reflect change in procedures for use of enriched boron.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

64. Added room cooling dependencies for Diesel Generator Cooling Water (DGCW) pumps o PRA Maintenance (Example 7). Model enhancement to correct error or omission.
65. Revised credit for operator action to control containment vent evolution (i.e., control containment pressure within narrow band).

o PRA Maintenance (Example 20). HEP modeling enhancement, no new methodology employed.

Q114A and Q214A Model Changes The 2014A model was the result of a regularly scheduled update per Exelon Risk Management T&RMs. Major changes incorporated into the model include the following data, plant, procedure, and analysis changes:

66. Bayesian updated initiating event frequencies utilizing most recent Quad Cities operating experience. This was not the first time Bayesian updating was performed.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

67. Revised component failure data including extensive use of plant-specific component failure data gathered from the Quad Cities Maintenance Rule program.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed.

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68. Individual component random failure probabilities Bayesian updated (as applicable) based upon the most recent plant specific data and the most current generic data sources.

o PRA Maintenance (Examples 2 and 3). Using new plant-specific and new generic data, no new methodology employed. This was not the first time Bayesian updating was performed.

69. Common cause failure (CCF) calculations revised to incorporate the updated individual random basic event probabilities and Alpha Factor parameters from NUREG/CR-5497 and NUREG/CR-5485 on the NRC website from 2010.

o PRA Maintenance (Example 3 and 26). Using new data, no new methodology employed.

70. Maintenance unavailability data based on the most recent Quad Cities operating experience.

o PRA Maintenance (Examples 2 and 19). Using new plant-specific data, no new methodology employed.

71. HRA updated to use the latest EOPs and SAGs; however, there is no change in HRA methods.

o PRA Maintenance (Example 20). HEP modeling enhancement, no new methodology employed.

72. The use of the EPRI HRA Calculator is implemented for the 2014A update to provide a consistent format for reporting the HRA results. There is no change in the HRA methods used.

o PRA Maintenance (Example 20). HEP modeling enhancement, no new methodology employed.

73. The previous model had approximately 70 pre-initiator Human Error Probabilities (HEPs) and about 20 more have been added to the PRA model. This process included a systematic identification of new pre-initiators and the quantification of these pre-initiators.

o PRA Maintenance (Example 20). Pre-initiator HEPs were already included in the model (e.g., mis-calibration of the low pressure permissives). Additional pre-initiators were added to the model as a modeling enhancement, no new methodology employed.

74. The evaluation of containment heat removal and RPV injection success following offsite AC power recovery in the LOOP and DLOOP event trees is added to the model to more explicitly model the recovery/restoration capabilities of the plant.

o PRA Maintenance (Example 10). Similar to Example 10, this modeling change was performed for completeness purposes in response to industry peer review comments.

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75. HRA dependency modeling is refined to incorporate additional HEP dependencies directly into the fault tree models and reduce reliance on the use of QRecover.

o PRA Maintenance (Example 20). HEP modeling update, new dependency groups were added directly to the model consistent with previous methods and other dependent groups treated as such (e.g.,

dependent HEP of Suppression Pool Cooling (SPC) and Containment Venting), no new methodology employed.

76. Update of the Quad Cities evacuation study for use in evaluating offsite consequence categories (i.e., LERF).

o PRA Maintenance (Example 22). Updates based on latest procedures, no new methodology employed.

77. Update the Quad Cities Emergency Action Levels (EALs) for use in evaluating offsite consequence categories (i.e., LERF).

o PRA Maintenance (Example 22). Updates based on latest procedures, no new methodology employed.

78. Internal floods that do not cause an automatic scram have been modeled as Manual Shutdowns in the 2014A model update. These were previously treated as turbine trip.

o PRA Maintenance (Example 6). This modeling change was simple in structure to more accurately model flooding impacts for as-built, as-operated plant.

79. Insights from the Fukushima Daiichi accidents that occurred in March 2011 are incorporated in the Level 2 PRA. Modeling probabilities and logic were updated in view of the Fukushima event to better model plant response based on operating experience.

o PRA Maintenance (Example 6). Logic model enhancement, no new methodology employed compared to the prior model.

80. Additional Level 2 CET structural changes were made to allow sensitivity calculations to be performed representing Fukushima observed events, e.g., no shell failure at RPV breach.

o PRA Maintenance (Example 6). These changes were performed for sensitivity case analysis only and do not affect the base model results.

The established success criteria are maintained. As such, they are considered modeling changes to address omissions and are considered maintenance.

81. Initiating Event Fault Trees for RBCCW, TBCCW, and SW were incorporated directly into the model. Prior to this, these SSIE Fault Trees were separate fault tree files which were quantified individually to provide a point estimate value for the initiator in the model.

o PRA Upgrade (Example 5). Although the fault tree files generally remained the same, their incorporation into the model is conservatively judged to be a PRA Upgrade event. The impact on the ILRT would be negligible since the level of detail in the model generally remained the same before and after incorporation.

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