ML19239A004

Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF

ML19239A004
Person / Time
Site:	Limerick
Issue date:	08/27/2019
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2018-LLA-0567
Download: ML19239A004 (23)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 August 27, 2019 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b."

References:

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b'," dated December 13, 2018 (ADAMS Accession No. ML18347B366).

2. Letter from D. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated February 14, 2019 (ADAMS Accession No. ML19045A011).

3. Electronic mail message from V. Sreenivas, U.S. Nuclear Regulatory Commission, to G. Stewart, Exelon Generation Company, LLC, "Limerick-Request for Additional Information: Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Completion Times -

RITSTF Initiative 4b (EPID L-2018-LLA-0567)," dated July 10, 2019 (ADAMS Accession No. ML19192A031).

4. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, Response to Request for Additional Information, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated August 12, 2019 (ADAMS Accession No. ML19224B705).

U.S. Nuclear Regulatory Commission Response to Request for Additional Information LAR to Adopt TSTF-505, Rev. 2 Docket Nos. 50-352 and 50-353 August 27, 2019 Page 2 By letter dated December 13, 2018 (Reference 1), as supplemented by letter dated February 14, 2019 (Reference 2), Exelon Generation Company, LLC (Exelon) requested an amendment to the Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (Limerick), Units 1 and 2, respectively.

The proposed amendment would modify Technical Specifications (TS) requirements to permit the use of risk-informed completion times (RICTs) in accordance with the Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18183A493).

During June 17, 2019 to June 21, 2019 the U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit at Exelons offices in Kennett Square, Pennsylvania to support development of its safety evaluation. Upon completion of the audit, the NRC staff determined that additional information is needed to complete its review of the LAR.

A draft request for additional information (RAI) was provided to G. Stewart (Exelon) by electronic email dated July 2, 2019. A conference call was subsequently held with the NRC on July 10, 2019 to provide clarification of the draft RAI questions. The formal RAI was issued by electronic email to G. Stewart (Exelon) on July 10, 2019 (Reference 3).

As noted in Reference 3, due to the need to perform sensitivity analyses and other additional studies, response to APLA RAI-02, RAI-03 and RAI-08 is required by August 30, 2019. The response to all other RAIs was submitted to the NRC by Reference 4.

The attachment to this letter provides a restatement of the NRC questions APLA RAI-02, RAI-03 and RAI-08 followed by our responses.

Exelon 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 the Reference 1 letter. Exelon has concluded that the information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92. In addition, Exelon has concluded that the information in this response 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.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania of this supplement to the application for license amendment by transmitting a copy of this letter and its attachment to the designated State Official.

U.S. Nuclear Regulatory Commission Response to Request for Additional Information LAR to Adopt TSTF-505, Rev. 2 Docket Nos. 50-352 and 50-353 August 27, 2019 Page 3 This letter contains no regulatory commitments.

If you should have any questions regarding this submittal, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 27th day of August 2019.

Respectfully, David P. Helker Sr. Manager - Licensing Exelon Generation Company, LLC

Attachment:

License Amendment Request - Response to Request for Additional Information cc: USNRC Region I, Regional Administrator USNRC Project Manager, Limerick USNRC Senior Resident Inspector, Limerick Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection

ATTACHMENT 1 License Amendment Request Limerick Generating Station, Units 1 and 2 NRC Docket Nos. 50-352 and 50-353 Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b."

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 1 of 19 Docket Nos. 50-352 and 50-353 By letter dated December 13, 2018 (Reference 1), as supplemented by letter dated February 14, 2019 (Reference 2), Exelon Generation Company, LLC (Exelon) requested an amendment to the Renewed Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (Limerick), Units 1 and 2, respectively.

The proposed amendment would modify Technical Specifications (TS) requirements to permit the use of risk-informed completion times (RICTs) in accordance with the Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18183A493).

During June 17, 2019 to June 21, 2019 the U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit at Exelons offices in Kennett Square, Pennsylvania to support development of its safety evaluation. Upon completion of the audit, the NRC staff determined that additional information is needed to complete its review of the LAR.

A draft request for additional information (RAI) was provided to G. Stewart (Exelon) by electronic email dated July 2, 2019. A conference call was subsequently held with the NRC on July 10, 2019 to provide clarification of the draft RAI questions. The formal RAI was issued by electronic email to G. Stewart (Exelon) on July 10, 2019 (Reference 3).

As noted in Reference 3, due to the need to perform sensitivity analyses and other additional studies, response to APLA RAI-02, RAI-03 and RAI-08 is required by August 30, 2019. The response to all other RAIs was submitted to the NRC by Reference 4.

A restatement of the NRC questions APLA RAI-02, RAI-03 and RAI-08 followed by our responses is provided below.

A. PROBABILISTIC RISK ANALYSIS (PRA) LICENSING BRANCH A (APLA)

APLA RAI Specific Key Assumptions and Sources of Uncertainty The NRC SE for NEI 06-09 states:

When key assumptions introduce a source of uncertainty to the risk calculations (identified in accordance with the requirements of the ASME standard), TR NEI 06-09, Revision 0, requires analysis of the assumptions and accounting for their impact to the RMTS [risk-managed technical specifications] calculated RICTs.

a. Regarding the uncertainty associated with continued injection from control rod drive (CRD) after containment failure, the disposition in LAR Table E9-1 states that RMAs will be implemented to address the uncertainty with this assumption for RICTs that are pertinent to loss of containment heat removal scenarios. Address the following:

i. Identify the Limiting Conditions for Operation (LCOs) proposed to be included in the RICT program for which loss of containment heat removal scenarios affect the RICT.

Response

The Limiting Conditions for Operation (LCOs) proposed to be included in the

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 2 of 19 Docket Nos. 50-352 and 50-353 Limerick RICT program for which the containment heat removal function is directly impacted are the following:

TS/LCO Condition 3.6.2.2.a One suppression pool spray loop inoperable 3.6.2.3.a One suppression pool cooling loop inoperable 3.7.1.1.a.2 One RHRSW pump in each subsystem inoperable 3.7.1.1.a.3 One RHRSW subsystem inoperable 3.7.1.1.a.3.a One RHRSW subsystem inoperable 3.7.1.1.a.3.b One RHRSW subsystem inoperable 3.7.1.2.a.3 One emergency service water loop inoperable Note that this list excludes LCOs 3.7.1.1.a.6 and 3.7.1.2.a.4 which are excluded from the RICT program scope in accordance with Reference 4.

ii. Provide the results of a sensitivity study of the impact on RICTs of this assumption to credit control rod drive injection after containment failure. Discuss the results of this sensitivity study in the context of the RICT estimates provided in Table E1-2 of Enclosure 1 of the LAR.

Response

Sensitivity cases were performed for the set of TS/LCOs listed above assuming that CRD fails after containment failure. This is a very bounding assessment because the most likely containment failure mode under overpressure situations is a leak in the upper drywell region, which would not render CRD inoperable after containment failure. The results of the bounding sensitivity cases with respect to the calculated RICT times reported in Table E1-2 of the Limerick TSTF-505 LAR are shown below.

The results of the bounding sensitivity cases generally show a nominal increase in CDF values, but the calculated values remain above the 30-day backstop. In these cases, the risk increase is almost entirely due to scenarios with operator action failures to vent containment given a loss of suppression pool cooling. Accounting for the RMA of briefing plant Operators on the importance of this action would greatly reduce the calculated risk increase from the bounding sensitivity case result. Also note that reasonable variations of this source of uncertainty (i.e., 2x or 5x the current value) would have much less of an impact on the CDF increase than does the bounding sensitivity case. As presented here, for the proposed LCOs, the sensitivity performed demonstrates that there is no impact on the RICT estimates shown in Table E1-2 of the LAR.

Original Sensitivity RICT Case RICT Estimate Estimate TS/LCO Condition (Days) (Days) 3.6.2.2.a One suppression pool spray loop inoperable 30.0 30.0

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 3 of 19 Docket Nos. 50-352 and 50-353 Original Sensitivity RICT Case RICT Estimate Estimate TS/LCO Condition (Days) (Days) 3.6.2.3.a One suppression pool cooling loop inoperable 30.0 30.0 3.7.1.1.a.2 One RHRSW pump in each subsystem inoperable 30.0 30.0 3.7.1.1.a.3 One RHRSW subsystem inoperable 30.0 30.0 3.7.1.1.a.3.a One RHRSW subsystem inoperable 30.0 30.0 3.7.1.1.a.3.b One RHRSW subsystem inoperable 30.0 30.0 3.7.1.2.a.3 One emergency service water loop inoperable 30.0 30.0 iii. Describe the RMAs (e.g., operator briefing on the significant human actions in the PRA that are pertinent to loss of containment heat removal scenarios) to be implemented for applicable RICTs and provide justification that these RMAs minimize the potential adverse impact on the RICT.

Response

Reliance on CRD after containment failure arises when all modes of containment heat removal fail. The significant human actions that would avoid that situation are the actions to align suppression pool cooling and the actions to vent containment at the Primary Containment Pressure Limit (PCPL) should suppression pool cooling fail.

RMAs associated with pre-job briefs for the importance of these actions will help reduce the likelihood of accident progression to containment failure and subsequently relying on CRD for continued injection. Beyond that, the importance of the potential need to refill the CST to maintain injection after containment failure or containment venting is the other operator action that can be included in the RMAs for these TS/LCOs. Note that, independent of the sensitivity study results presented here, these same actions would be identified as important from the base analysis for the configuration as well.

b. LAR Table E9-1 identifies that, given an uncontrolled flooding of the steam lines, a nominal failure probability of 1 x 10-3 is assigned to Safety Relief Valves (SRVs) being permanently disabled, which precludes the ability to depressurize the reactor pressure vessel through the SRVs. The disposition states that the 1 x 10-3 failure probability provides a slight conservative bias to the results such that the impact on RICT calculations is not unduly influenced. The LAR identifies that this uncertainty affects the RICTs for the LCOs associated with the High Pressure Injection Systems. The disposition further identifies that although the SRVs are designed to pass water, they are never tested in this fashion.

Discuss the sensitivity of this assumption for the RICT application. If determined to be significant to the RICT application, discuss the RMAs that will be implemented to minimize the impact of this assumption.

Response

The PRA models an unlikely scenario in which steam lines may become flooded if all of the following occur:

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 4 of 19 Docket Nos. 50-352 and 50-353

• level is not maintained below Level 8 (the high-high level setpoint),

• the automatic HPCI, RCIC and FW system trip functions fail, and

• operators do not respond in time to take manual control of HPCI or RCIC after the Level 8 high-high level trip failure.

The water carried over into the steam lines is then postulated as having the potential to disable the SRVs from being able to subsequently perform their overpressure control function (i.e., fail to open to relieve pressure) even if the RPV water level drops later.

The above scenario, i.e., failure of the HPCI, RCIC or FW systems to trip on high reactor level, is included in the PRA model as leading to a steam line flooding scenario. Given the conditions occur that would allow initial flooding of the steam lines, a probability is assigned that this flooding permanently disables all of the SRVs. This is modeled with basic event APHSRXDXI(2), which is basically a common cause failure to open event for the SRVs.

Per the Limerick PRA Data Notebook:

The passing of water through the SRVs temporarily (i.e. should Level 8 trips fail and RPV water level rises to the main steam lines) should not render the SRVs totally unavailable. However, to account for this possibility, a failure probability of 1.0E-3 is assigned to this occurring (APHSRXDXI(2)) given that conditions arise to allow water ingress into the main steam lines.

Thus, the model assumes a failure probability of 1.0E-3 for the common cause failure to open of the SRV pressure relief function for the scenario described above, i.e., following flooding of the steam lines. Note that since this is a scenario in which relief through the SRVs is required, potential damage to the SRVs due to initial water relief would be a beneficial failure, i.e., a likely failure mechanism due to water relief through SRVs is damage to the valve seat resulting in leakage through the valve. Such leakage would be a beneficial failure that is not credited in the model.

For non-steamline flooding scenarios in the Limerick PRA, controlled reactor depressurization will be successful if at least two of 14 safety/relief valves open and remain open. For these scenarios, there is a Common Cause Failure (CCF) basic event (ARVALLCPI) modeled for 13 of 14 SRVs failing to open which has a probability value of 1.9E-6. This is the same common cause failure as APHSRXDXI(2) but in other scenarios in which there is no initial water relief challenge.

Thus, modeling all SRVs as unavailable for the scenario with water in the steam lines with a 1.0E-03 failure probability imposes an increase by more than a factor of 500 compared to the CCF failure to open basic event ARVALLCPI probability. This is considered to be a reasonable representation given the large number of SRVs relative to the number required, and likely represents a slight conservative bias.

In the base CDF model, the relative FV importance of event APHSRXDXI is 3.2E-4. Use of a probability of 1.0E-02 was examined as a sensitivity. With the 1.0E-02 common cause failure probability, i.e., a factor of 5000 greater than the nominal CCF, the relative

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 5 of 19 Docket Nos. 50-352 and 50-353 importance of event APHSRXDXI is 5.6E-3. This demonstrates that the model is not sensitive to the existing conservative value.

Additionally, a similar sensitivity case (i.e., increase APHSRXDXI to 1.0E-2) was performed for the specific configurations of HPCI and RCIC out of service. The results of these sensitivity cases with respect to the calculated RICT times reported in Table E1-2 are shown below. The results of the sensitivity cases generally show a nominal increase in CDF values, but the calculated RICT values remain above the 30-day backstop. In these cases, the risk increase is almost entirely due to scenarios with a loss of feedwater and condensate followed by a loss of HPCI and RCIC and operators fail to depressurize. Note that independent of the sensitivity study results presented here, the same RMAs would be identified as important from the base analysis for the configuration as well.

Original RICT Sensitivity Estimate Case RICT (Days) Estimate TS/LCO Condition (Days) 3.5.1.c.1 HPCI system inoperable 30.0 30.0 3.7.3.a RCIC system inoperable 30.0 30.0 APLA RAI Potential Credit for FLEX Equipment or Actions The NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269),

provides the NRCs staff assessment of the challenges of incorporating diverse and flexible (FLEX) coping strategies and equipment into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014). Though implementation of FLEX procedures is cited in the LAR as possible RMAs, the LAR and other docketed information do not indicate if Limerick has credited FLEX equipment or actions in the internal events, including internal flooding, or fire PRA models. As such, please address the following:

a. Discuss whether Exelon has credited FLEX equipment or mitigating actions into the Limerick internal events, including internal flooding, or fire PRA models. If not incorporated or their inclusion is not expected to impact the PRA results used in the RICT program, no additional response is requested.

Response

Exelon has credited FLEX equipment and mitigating actions in the Limerick internal events, internal flooding, and fire PRA models.

b. If FLEX equipment or operator actions have been credited in the PRA, address the following, separately for the internal events, including internal flooding, and fire PRAs:

i. Summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 6 of 19 Docket Nos. 50-352 and 50-353 support this application. Include discussion of whether the credited FLEX equipment is portable or permanently installed equipment.

Response

Credit is taken for the following in each of the PRA models:

1) Deploying and aligning the portable FLEX 480V generators (limited to extended loss of AC power (ELAP) scenarios).

2) Deploying and aligning the portable FLEX pumps (limited to ELAP scenarios).

3) Prolonged RCIC operation via partial RPV depressurization and venting containment using the permanently installed Hardened Containment Vent System (HCVS).

ii. Discuss whether the credited equipment (regardless of whether it is portable or permanently-installed) are like other plant equipment (i.e. SSCs with sufficient plant-specific or generic industry data) and whether the credited operator actions are similar to other operator actions evaluated using approaches consistent with the endorsed ASME/ANS RA-Sa-2009 PRA standard.

Response

The portable FLEX pumps and generators are not like other plant equipment. The reliability data for these components is treated as noted in response to item iii below.

The HCVS includes permanently installed components similar to other permanently installed plant components.

The credited operator actions for each of the mitigating strategies listed above are shown below. These actions are similar to other operator actions included in the PRA models and are evaluated using approaches consistent with the endorsed ASME/ANS RA-Sa-2009 PRA standard as documented in the Limerick HRA notebook.

1) Success of the FLEX generators includes required operator actions for DC Load Shed (QHULSFDXI), deploy and start the FLEX generators (QHUFBXDXI), align the FLEX generators to the battery chargers (QHUFB1DXI or QHUFB2DXI), and refuel the FLEX generators (QHUREFDXI).

2) Success of the FLEX pumps includes required operator actions for aligning the FLEX pumps from the fire water system (QHUFFXDXI), aligning the FLEX pumps for RPV injection from the spray pond (QHULPADXI or QHULPBDXI), and refueling the FLEX pumps (QHUPRFDXI).

3) Success of prolonged RCIC operation includes required operator actions for performing partial RPV depressurization (RHUVT1DXI), opening of the hardened vent at the HCVS panel (QHUCVEDXI), aligning the FLEX pumps for suppression pool makeup from the spray pond (QHUSPADXI or QHUSPBDXI), and refueling the FLEX pumps (QHUPRFDXI).

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 7 of 19 Docket Nos. 50-352 and 50-353 iii. If any credited FLEX equipment is dissimilar to other plant equipment credited in the PRA (i.e. SSCs with sufficient plant-specific or generic industry data), discuss the data and failure probabilities used to support the modeling and provide the rationale for using the chosen data. Discuss whether the uncertainties associated with the parameter values are in accordance with the ASME/ANS PRA Standard as endorsed by RG 1.200, Revision 2.

Response

While industry data for FLEX components is being collected and is anticipated to be available prior to implementation of RICT at Limerick, an assumption is made that two-times the generic reliability values for similar equipment provides a reasonable approximation of the reliability of the FLEX equipment. The values used for the reliability data of the FLEX generators and FLEX pumps are shown below.

Failure Mode FLEX Generators FLEX Pumps Fail to Start 1.3E-2 6.3E-3 Fail to Run (24 Hours) 7.3E-2 9.5E-2 Total Failure Probability 8.6E-2 1.0E-1 The uncertainties associated with the data values are based on the uncertainty parameters from the generic data, and are in accordance with the ASME/ANS PRA Standard.

Use of these values should provide a reasonable approximation of the reliability of the FLEX equipment. This assumption will be revisited when the generic industry data for FLEX equipment becomes available.

iv. If any operator actions related to FLEX equipment are evaluated using approaches that are not consistent with the endorsed ASME/ANS RA-Sa-2009 PRA Standard (e.g., using surrogates), discuss the methodology used to assess human error probabilities for these operator actions. The discussion should include:

1. A summary of how the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of the ASME/ANS RA-Sa-2009 PRA Standard were evaluated.

Response

All the human error probabilities for FLEX components were evaluated with the same methodology used for all human error probabilities in the Limerick PRA models as documented in the Limerick HRA notebook. As such, the HEPs include an evaluation of the scenario specific performance shaping factors listed in (a)-(j) of supporting requirement HR-G3 of the ASME/ANS RA-Sa-2009 PRA Standard.

2. Whether maintenance procedures for the portable equipment were reviewed for possible pre-initiator human failures that renders the equipment unavailable during

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 8 of 19 Docket Nos. 50-352 and 50-353 an event, and if the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of the ASME/ANS RA-Sa-2009 PRA Standard.

Response

Maintenance procedures for the portable equipment were not reviewed for possible pre-initiator human failures. Basic events for maintenance unavailability of the pumps and generators are included in the model with screening values of 1E-02.

These maintenance terms represent failure of the entire system (i.e., all the FLEX generators for the unit or all the FLEX pumps for the unit) as credit for the N+1 FLEX components are not included in the model. With the screening values, the importance measures for these maintenance terms do not exceed the risk-significance threshold per the PRA standard. This low risk significance of the maintenance events, combined with the fact that the FLEX equipment failure probability values and combined required post-initiator HEPs used in the model far exceed this maintenance unavailability screening value, allows pre-initiator human failures for the FLEX components to be screened from inclusion in the model.

3. If the procedures governing the initiation or entry into mitigating strategies are ambiguous, vague, or not explicit, a discussion detailing the technical bases for probability of failure to initiate mitigating strategies.

Response

All the procedures for use of FLEX equipment or mitigating strategies at Limerick are clear, explicit, and built into the Emergency Operating Procedures (including the requirement for declaration of an ELAP). The declaration of an ELAP is clear and explicit in the Limerick E-1 procedure (LOSS OF ALL AC POWER (STATION BLACKOUT)).

c. The ASME/ANS RA-Sa-2009 PRA standard defines PRA upgrade as 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.

Section 1-5 of Part 1 of ASME/ANS RA-Sa-2009 PRA Standard states that 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.

Provide an evaluation of the model changes associated with incorporating mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

Response

As described in the other responses to this RAI, no new methodologies were implemented for the FLEX equipment and mitigating strategies that were added to the PRA model. The impacts of the modeling changes on the results were as expected (i.e., reductions in the

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 9 of 19 Docket Nos. 50-352 and 50-353 SBO and total loss of AC accident sequences when FLEX generators are credited, and reductions in the loss of containment heat removal scenarios when the HCVS system is credited). These changes do not represent a change in capability that impacts the significant accident sequences, but merely represent model updates to ensure that the models reflect the as-built, as-operated plant.

d. Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program. The NRC SE for NEI 06-09, Revision 0, states that this consideration is consistent with Section 2.3.5 of RG 1.177, Revision 1. NEI 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RICT calculation. NRC staff notes that the impact of model uncertainty could vary based on the proposed RICTs. NEI 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate compensatory RMAs including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in modeling FLEX equipment and actions related to assumptions regarding the failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application. In light of these observations:

i. Describe the sensitivity studies that will be used to identify the RICTs proposed in this application for which FLEX equipment and/or operator actions are key assumptions and sources of uncertainty (e.g., use of generic industry data for non-safety related equipment). Explain and justify the approach (e.g., any multipliers for failure probabilities) used to perform the sensitivity studies.

ii. Describe how the results of the sensitivity studies which identify FLEX equipment and/or operator actions as key assumptions and sources of uncertainty will be used to identify RMAs prior to the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A.

iii. Demonstrate the approaches described in items (i) and (ii) above using an example sensitivity study for the nominal configuration of a proposed RICT where the FLEX equipment and/or operator actions are identified as key assumptions and sources of uncertainty.

Response

Based on the types of scenarios for which credit for FLEX equipment is modeled as described above (i.e., reductions in the SBO and total loss of AC accident sequences when FLEX generators are credited), the following LCOs were identified as potentially being the most sensitive to assumptions related to the FLEX modeling.

TS/LCO Condition 3.7.1.2.a.3 One emergency service water loop inoperable

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 10 of 19 Docket Nos. 50-352 and 50-353 TS/LCO Condition 3.8.1.1.d One offsite circuit and one diesel generator inoperable 3.8.1.1.f One offsite circuit inoperable 3.8.1.1.g Two offsite circuits inoperable 3.8.2.1.a.3 Two battery chargers on one division inoperable 3.8.2.1.c Any battery(ies) on one division of required DC electrical power sources inoperable 3.8.3.1.a One required AC distribution system divisions not energized 3.8.3.1.b One required DC distribution system divisions not energized Note that this list excludes LCOs 3.8.1.1.b and 3.8.1.1.h which are excluded from the RICT program scope in accordance with Reference 4.

As described in the response to Item b.iii above, the current assumption is made that two-times the generic reliability values for similar equipment provides a reasonable approximation of the reliability of the FLEX equipment. This leads to probabilities of failure that in total are close to 0.1 or approximately 1-in-10. To further explore the sensitivity to this assumption, the TS/LCO conditions noted above were re-run for RICT using 5x the generic failure probabilities instead of 2x the generic failure probabilities. The increase in these values to a total system failure probability of approximately 1-in-4 represents a conservative bias to the analysis.

The results of the bounding sensitivity cases with respect to the calculated RICT times reported in Table E1-2 are shown below. The results of the sensitivity cases generally show a nominal increase in CDF and LERF values which corresponds to only a small change in the calculated RICT estimates.

Sensitivity Original RICT Case RICT Estimate Estimate TS/LCO Condition (Days) (Days) 3.7.1.2.a.3 One emergency service water loop inoperable 30.0 30.0 3.8.1.1.d One offsite circuit and one diesel generator inoperable 30.0 30.0 3.8.1.1.f One offsite circuit inoperable 30.0 30.0 3.8.1.1.g Two offsite circuits inoperable 30.0 28.9 3.8.2.1.a.3 Two battery chargers on one division inoperable 30.0 30.0 3.8.2.1.c Any battery(ies) on one division of required DC electrical 16.2 15.5 power sources inoperable 3.8.3.1.a One required AC distribution system divisions not 7.8 7.5 energized 3.8.3.1.b One required DC distribution system divisions not 16.2 15.5 energized In the sensitivity cases performed above, the changes to the FLEX reliability values led to shifts in the cutsets which included those reliability values, but this did not result in the identification of any new important operator actions nor did it identify any additional RMAs.

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 11 of 19 Docket Nos. 50-352 and 50-353 APLA RAI PRA Modeling of Instrumentation and Controls The proposed TS limiting conditions for operations (LCOs) include those related to instrumentation and controls (I&C).

PRA technical acceptability attributes are provided in Section 2.3.4 of NEI 06-09, Revision 0-A, and in RG 1.200, Revision 2. The licensee has previously received approval for changes to its I&C completion times, bypass test times, and surveillance intervals consistent with guidance in Technical Specifications Task Force (TSTF) traveler TSTF-411 and TSTF-418. However, the licensee does not address whether the I&C is modeled in sufficient detail to support implementation of TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493). The following additional information is requested:

a. Explain how instrumentation is modelled in the PRA. This should include, but not be limited to, the scope of the I&C equipment (e.g., channels, relays, logic) and associated TS functions for which a RICT would be applied, and PRA modeling of the I&C and functions including how these are modeled in sufficient detail and based on plant-specific data, etc.

Response

Instrumentation is explicitly modeled in the PRA as required to support the modeled system.

The failure data is handled in the same fashion as other components with Type Codes assigned for each component type. Each of the instrumentation Technical Specifications in the scope of the RICT program is listed below with the associated scope of PRA modeling (examples of modeled components and component types). For those functions that are not explicitly modeled, a conservative surrogate event is used to calculate the RICT as noted in the LAR. Additional instrumentation not in the table is similarly modeled to support other systems and operator actions.

RPS INSTRUMENTATION MODELED TS TABLE 3.3.1-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

1. Intermediate Range Not modeled - channel failure N/A Monitors used as surrogate

2. APRM upscale APRM relays (C71A-K12A) Relay APRMs APRM VOTERs Voter Circuit

3. RPV high pressure RPV high pressure relay C71A- Relay K5A PT-042-1N078A Pressure Transmitter PIS-041-1N678A Trip unit

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 12 of 19 Docket Nos. 50-352 and 50-353 RPS INSTRUMENTATION MODELED TS TABLE 3.3.1-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

4. Reactor vessel water RPV low level relay (C71A-K6A) Relay level -low, level 3 LT-042-1N080A Level Transmitter LIS-042-1N680A Trip unit

5. Main Steam Isolation MSIV Closure relays (C71A- Relay Valve (MSIV) closure K3A)

MSIV Closure bypass relays Relay (C71A-K11A)

MSIV inboard and outboard limit Limit Switch switches (ZS-041-122A)

6. Deleted from TS

7. Drywell pressure -high High Drywell Pressure relay Relay (C71A-K4A)

PT-042-1N050A Pressure Transmitter PIS-042-1N650A Trip unit

8. Scram Discharge Not modeled - channel failure N/A Volume level- High used as surrogate

9. Turbine Stop Valve TSV closure relay (C71A-K10A) Relay Closure TSV position switches Pressure Switch ZS-001-104A

10. Turbine Control TCV closure relay (C71A-K08A) Relay Valve Fast Closure 30%power bypass relay(C71A- Relay K09A)

PS-001-102A Pressure Switch

11. Reactor Mode Switch Mode switch Switch Shutdown position

12. Manual Scram Not modeled - channel failure N/A used as surrogate N/A Scram relays (C71A-K14A) Relay

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 13 of 19 Docket Nos. 50-352 and 50-353 ISOLATION ACTUATION INSTRUMENTATION TS TABLE 3.3.2-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

1. Main Steam Line Isolation

a. Reactor Water level LS-042-1N684A Trip Unit LT-042-1N081A Transmitter b.-h. Input to the high pressure break N/A outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

2. RHR Shutdown Cooling Isolation

a. Reactor Water Input to the high pressure break N/A level outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

b. Reactor low See above pressure permissive

c. Manual initiation See above

3. RWCU

a. High delta flow Input to the high pressure break N/A outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

b. High area temp. See above

c. High delta temp. See above

d. SLCS initiation C41A-K4A Relay

e. Reactor low level Input to the high pressure break N/A outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

f. Manual initiation See above

4. HPCI Isolation

a. High steam line Input to the high pressure break N/A delta pressure outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 14 of 19 Docket Nos. 50-352 and 50-353 ISOLATION ACTUATION INSTRUMENTATION TS TABLE 3.3.2-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

b. Low steam supply See above pressure

c. High turbine See above exhaust pressure

d. High room temp. See above

e. High room See above differential temp.

f. High pipe tunnel See above temp.

g. Manual initiation See above

h. HPCI steam line See above delta pressure timer

5. RCIC

a. High steam line Input to the high pressure break N/A delta pressure outside containment initiator; signal contribution will be treated as failed for RICT calculation when out of service

b. Low steam supply See above pressure

c. High turbine exhaust See above pressure

d. High room temp. See above

e. High room See above differential temp.

f. High pipe tunnel See above temp.

g. Manual initiation See above

6. Primary Containment

a. Reactor water level LT-042-1N081A Transmitter

b. Drywell pressure PT-042-1N078A Transmitter c-g. (Rad monitors or Not modeled - generic isolation N/A deleted) failure event used as surrogate

h. High drywell See above pressure/low reactor pressure

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 15 of 19 Docket Nos. 50-352 and 50-353 ISOLATION ACTUATION INSTRUMENTATION TS TABLE 3.3.2-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

i. PCIG to drywell delta See above pressure low

j. Manual initiation See above

7. Secondary Out of Scope Containment ECCS INSTRUMENTATION MODELED TS TABLE 3.3.3-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS* TYPE N/A Core Spray initiation logic Relay Logic N/A RHR LPCI initiation logc Relay Logic N/A HPCI intitiation logic Relay Logic N/A ADS initiation logic Relay Logic Reactor low level(s) LT-042-1N091A Transmitter LIS-042-1N691A Trip Unit LT-042-1N091B Transmitter LS-042-1N692B Trip Unit High drywell pressure PT-042-1N094A Transmitter PIS-042-1N694A Trip Unit RPV pressure low E21A-K25A/B Relay permissive E21A-K27A/B Relay PT-042-1N090A Transmitter PIS-042-1N690A Trip Unit Injection valve differential PDT-051-1N058A Transmitter pressure low PDISL-051-1N658A Trip Unit Manual Not modeled - associated N/A operator action used as surrogate LPCI, RHR Pump PT-052-1N055A Transmitter discharge pressure (ADS permissive)

PIS-052-1N655A Trip Unit PT-051-1N055A-D Transmitter PIS-051-1N655A Trip Unit PT-051-1N056A Transmitter PIS-051-1N656A Trip Unit CST low level LT-055-1N061B Transmitter LIS-055-1N661B Trip Unit LT-049-1N035A Transmitter Suppression pool level Not modeled - channel N/A high failure used as surrogate Reactor high level LS-042-1N693B Trip Unit

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 16 of 19 Docket Nos. 50-352 and 50-353 ECCS INSTRUMENTATION MODELED TS TABLE 3.3.3-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS* TYPE ADS timer B21C-K8A Timer 4KV bus under voltage- Not modeled - channel N/A degraded voltage failure used as surrogate 4KV bus under voltage E21A-K18A Relay loss of voltage 127-115 Relay 162-11502 (D11) Time Delay Relay 162-11509 (D11) Time Delay Relay 183-11509 (D11) Relay

• Note: Parameters and components are shared among ADS, Core Spray, RHR-LPCI and PCI and are therefore listed once by parameter.

RRCS/SLCS INSTRUMENTATION (ATWS RPT) MODELED TS TABLE 3.3.4.1-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS TYPE

1. Reactor level LT-042-1N402A Level Transmitter ATM 2A Trip Module

2. Reactor pressure PT-042-1N403A Pressure Transmitter N/A ATM 3A Trip Module N/A Circuit card A1 Logic Device EOC-RPT INTRUMENTATION MODELED TS TABLE 3.3.4.2-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENT TYPE N/A C71A-K50A/B Relay

1. TSV closure ZS-001-104A Limit switch

2. TCV closure C71A-K10A Relay PS-001-102A Pressure switch C71A-K8A Relay

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 17 of 19 Docket Nos. 50-352 and 50-353 RCIC INSTRUMENTATION MODELED TS TABLE 3.3.5-1 EXAMPLE MODELED COMPONENT FUNCTION COMPONENTS* TYPE N/A RCIC LOGIC Logic

a. Reactor low level LT-042-1N097A/E Transmitter LIS-042-1N697A/E Trip Unit LIS-042-1N692A/E Trip Unit LIS-042-1N091A/E Transmitter

b. Reactor high level LIS-042-1N693A/E Trip Unit LIS-042-1N698A/E Trip Unit LT-042-N097A/E Transmitter

c. CST level LIS-049-1N635A/E Trip Unit LT-049-1N035A Transmitter

d. Manual initiation Not modeled - associated N/A operator action used as surrogate FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION MODELED TS TABLE 3.3.9-1 EXAMPLE MODELED TYPE FUNCTION COMPONENT

1. Reactor high level Logic reactor high level Logic LT-042-1N004A Level transmitter

b. Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties be considered in application of the PRA base model results to the RICT program. The NRC SE for NEI 06-09, Revision 0, states that this consideration is consistent with Section 2.3.5 of RG 1.177, Revision 1. NEI 06-09, Revision 0-A, further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties which could potentially impact the results of a RICT calculation and that sensitivity studies should be used to develop appropriate compensatory RMAs.

Regarding digital I&C, NRC staff notes the lack of consensus industry guidance for modeling these systems for plant PRAs to be used in risk-informed applications. In addition, known modeling challenges exist due to the lack of industry data for digital I&C systems and components and the complexities associated with modeling software failures including common cause software failures. Given these needs and challenges, if the modeling of digital I&C systems or components is included in the RTR model, then address the following:

i. Provide the results of a sensitivity study demonstrating that the uncertainty associated with modeling the digital I&C systems or components has an inconsequential impact on the LCOs included in the RICT program, or

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 18 of 19 Docket Nos. 50-352 and 50-353 ii. If the modeling of digital I&C systems or components is determined to be a key source of uncertainty for the application, identify impacted LCOs and describe how sensitivity studies are used to identify RMAs to minimize the potential adverse impacts of this source of uncertainty, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A

Response

Limerick has a digital feedwater control system, installed in 2004 (Unit 1) and 2005 (Unit 2).

In the feedwater level control logic, a false signal from the redundant reactivity control system could lead to inadvertent termination of automatic feedwater level control. This event is modeled in the PRA by a basic event (FPHRRCDXI) included in the feedwater system level control logic. In the feedwater level control logic, a high reactor water level (level 8) signal trips the feedwater turbines. An event for a false level 8 signal is included in the feedwater system fault tree.

The Limiting Conditions for Operation (LCOs) proposed to be included in the Limerick RICT program for which loss of digital feedwater most affect the RICT are the following:

TS/LCO Condition 3.5.1.c.1 HPCI system inoperable 3.7.3.a RCIC system inoperable Sensitivity cases were performed for the set of TS/LCOs above assuming that the failure rates of each of the basic events associated with digital feedwater were 100 times greater than the base values. For these sensitivities, the probabilities for the following basic events associated with digital feedwater were each increased by a factor of 100.

Basic Event Description FAF450HWI AF450 FAILS TO AUTO CONTROL LEVEL FAF100HWI FIELD BUS AF100A AND AF100B FAIL FPHRRCDXI FALSE SIGNAL FROM REDUNDANT REACTIV. CONTROL SYSTEM FPHL8HDXI FALSE LEVEL 8 SIGNAL The results of the sensitivity cases with respect to the calculated RICT times reported in Table E1-2 of the Limerick TSTF-505 LAR are shown below. The results of the sensitivity cases generally show a very nominal increase in CDF values but the calculated RICT values decreased by less than 1% and remained above the 30-day backstop in each case. In these cases, the risk increase is almost entirely due to scenarios with operator actions involving failure to vent containment and failure to swap feedwater to manual control given a loss of automatic control. The RMA of briefing Operators on the importance of these actions would greatly reduce the calculated risk increase from the sensitivity case result. As presented here, for the proposed LCOs, the sensitivity performed demonstrates that there is no impact on the RICT estimates shown in Table E1-2 of the LAR.

Response to Request for Additional Information Attachment 1 LAR to Adopt TSTF-505, Rev. 2 Page 19 of 19 Docket Nos. 50-352 and 50-353 Sensitivity Original RICT Case RICT Estimate Estimate TS/LCO Condition (Days) (Days) 3.5.1.c.1 HPCI system inoperable 30.0 30.0 3.7.3.a RCIC system inoperable 30.0

30.0 REFERENCES

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b'," dated December 13, 2018 (ADAMS Accession No. ML18347B366).

2. Letter from D. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated February 14, 2019 (ADAMS Accession No. ML19045A011).

3. Electronic mail message from V. Sreenivas, U.S. Nuclear Regulatory Commission, to G.

Stewart, Exelon Generation Company, LLC, "Limerick-Request for Additional Information:

Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Completion Times -RITSTF Initiative 4b (EPID L-2018-LLA-0567)," dated July 10, 2019 (ADAMS Accession No. ML19192A031).

4. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information, License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," dated August 12, 2019 (ADAMS Accession No. ML19224B705).