2CAN042403, Response to the Request for Additional Information Regarding ANO-2 License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended C
| ML24115A246 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/24/2024 |
| From: | Couture P Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 2CAN042403 | |
| Download: ML24115A246 (1) | |
Text
Entergy Operations, Inc. 1340 Echelon Parkway, Jackson, MS 39213 2CAN042403 10 CFR 50.90 April 24, 2024 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Response to the Request for Additional Information Regarding ANO-2 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" Arkansas Nuclear One, Unit 2 NRC Docket No. 50-368 Renewed Facility Operating License No. NPF-6 By letter dated April 5, 2023 (Reference 1), Entergy Operations, Inc. (Entergy) requested to change the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification (TS). The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times (RICT) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" (ML18253A085), dated November 21, 2018.
By letter dated January 11, 2024 (Reference 2), Entergy supplemented the original License Amendment Request (LAR) with information requested by the NRC staff during an online audit conducted between October 17 through October 18, 2023.
On March 18, 2024, the NRC stated that additional information was required for the staff to complete its review of the application. Entergy and the NRC staff determined that a clarification call to discuss the RAI was not necessary, and the RAI was sent to Entergy in final form by letter dated March 27, 2024 (Reference 3). The enclosure to this letter provides Entergy's response to Reference 3.
Entergy has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1. The supplemental information provided in this letter does not affect the bases for concluding that the proposed license amendments do not involve a significant hazards consideration.
This letter contains no new regulatory commitments.
Phil Couture Sr. Manager Fleet Regulatory Assurance - Licensing 601-368-5102
2CAN042403 Page 2 of 2 In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," Entergy is notifying the State of Arkansas of this amendment request by transmitting a copy of this letter and enclosures to the designated State Official.
If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 24th day of April 2024.
Sincerely, Phil Couture PC/mar
Enclosure:
Response to Request for Additional Information (RAI)
References:
- 1)
Letter from Entergy to NRC, "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 4' Arkansas Nuclear One - Unit 2" (ML23095A281) (2CAN042301) dated April 5, 2023
- 2)
Letter from Entergy to NRC, "Supplemental Information - Adopt Risk-Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b'"
(ML24011A293) (2CAN012403) dated January 11, 2024
- 3)
Email from T. Wengert (Senior Project Manager, NRC) to Riley Keele (Entergy), "FW: ANO TSTF-505 Post-Audit Final RAI (L-2023-LLA-0052)" (ML24088A008) (2CNA032401) dated March 27, 2024 cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official Philip Couture Digitally signed by Philip Couture Date: 2024.04.24 13:19:28 -04'00'
Enclosure 2CAN042403 Response to Request for Additional Information (RAI)
Enclosure 2CAN042403 Page 1 of 4 Response to Request for Additional Information (RAI)
By letter dated April 5, 2023 (ML23095A281), as supplemented by letter dated January 11, 2024 (ML24011A293), Entergy Operations, Inc. (the licensee) submitted a license amendment request (LAR) to modify the Arkansas Nuclear One - Unit 2 (ANO-2) technical specifications to permit the use of Risk-Informed Completion Times (RICT) in accordance with TSTF-505, Revision 2, "Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b."
The U.S. Nuclear Regulatory Commission (NRC) staff have reviewed the licensees submittals and concluded that additional information, as discussed below, is required to complete its review:
Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA) RAI 01 - Sources of PRA Model Uncertainty Regulatory Basis Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, states, in part, "NRC [U.S. Nuclear Regulatory Commission] reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application." The NRC staff evaluates the acceptability of the probabilistic risk assessment (PRA) for each new risk-informed application and as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision-making may vary with the relative weight given to the risk assessment element of the decision-making process. The NRC staff notes that the calculated results of the PRA are used directly to calculate a risk informed completion time (RICT), which subsequently determines how long structures, systems, and components (SSCs) -- both individual SSCs and multiple, unrelated SSCs -- controlled by technical specifications can remain inoperable. Therefore, the PRA results are given a very high weight in the staffs review of a TSTF-505 license amendment request (LAR). This is consistent with the guidance in Item 10 of Section 2.3.4, PRA Technical Adequacy of NEI 06-09, "Risk Informed Technical Specifications Initiative 4b; Risk Managed Technical Specifications (RMTS) Guidelines," Revision 0-A (ML122860402), which states, in part:
For LCOs [limiting conditions for operation] in which it is determined that identified uncertainties could significantly impact the calculated RICT, sensitivity studies should be performed for their potential impact on the RICT calculations.
(Reference EPRI-1009652 [ ] for one method to determine key uncertainties.) Insights obtained from these sensitivity studies should be used to develop appropriate compensatory risk management actions.
Enclosure 2CAN042403 Page 2 of 4 Issue In Audit Question 07.b the NRC staff noted that LAR Enclosure 9 Section 4, "Assessment of Level 2 Epistemic Uncertainty Impacts," states that no key sources of uncertainty for the RICT program were identified for the Level 2 PRA. However, in Table 8.4-3 of the ANO-2 key assumptions and sources of uncertainty analysis report provided on the audit portal, it is reported for the sensitivity study on the conditional steam generator tube rupture (SGTR) probability of burst (POB) that the RICT for LCO 3.7.1.5, which concerns main steam isolation valve operability, is very sensitive to this parameter (i.e., a 55 percent reduction in the RICT).
Based on this sensitivity study result it appears to the NRC staff that the conditional POB is a key source of uncertainty for the RMTS application.
In its response to this audit question provided in the LAR supplement dated January 11, 2024, the licensee explains that the conditional POB used in the ANO-2 PRA model is based on plant-specific steam generator tube inspection data obtained under the Steam Generator Integrity Program and is calculated utilizing methods developed by the Electric Power Research Institute (EPRI). It is further explained that the conditional POB value used in the sensitivity study is not a plant-specific value but rather a generic value developed by the NRC staff in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5, Regarding Steam Generator Tube Integrity" (ML082400710). However, while the response acknowledges that there is uncertainty in the conditional POB, no information is provided to show that the conditional POB has an inconsequential impact on the RICT calculations.
Request for Additional Information Justify that the conditional POB has an inconsequential impact on the RICT calculations.
If, in response to part a), the licensee cannot justify that the conditional POB has an inconsequential impact on the estimated RICTs, then identify potential additional contingencies or actions that can compensate for this uncertainty.
Entergy Response The ANO-2 full-power internal events and internal flooding model (FPIE/IF) modeled a conditional SGTR POB using a generic value of 0.027. This value was subsequently updated using a site-specific POB of 1.4E-03 for the FPIE/IF model after the source of uncertainty calculation was completed for the RICT program. The Fire PRA was already using the site-specific POB. As a result of this change, both the internal events and the Fire PRA are using the same value. Reviewing the sensitivity results, the driver for LCO 3.7.1.5, main steam isolation valve operability, was the result of the conservative values assumed in the sensitivity for the FPIE/IF model.
The sensitivity analysis cited in the RAI multiplied the POB by a factor of three (3) for all models, including the generic value used in the internal events. The sensitivity results were being skewed based on conservative estimates used in the FPIE/IF model and have been replaced with a follow up sensitivity study using realistic estimates that better reflect the final PRA configuration for the RICT program.
Enclosure 2CAN042403 Page 3 of 4 To better understand the impact the POB has on the RICT program, the sensitivity analysis has been reanalyzed for LCO 3.7.1.5, which concerns main steam isolation valve operability, using the site-specific probability in an updated FPIE/IF model.
Both the baseline and the sensitivity were evaluated for completeness.
When the FPIE/IF model uses a POB of 1.4E-03, the base RICT time for LCO 3.7.1.5 is increased to 30 days. When this value is adjusted by a factor of three (3) (4.2E-03) in all ANO-2 PRA models, the RICT time for LCO 3.7.1.5 is 25.05 days.
The site-specific POB will be used in all the RICT models and is not a key source of uncertainty for the RICT program as the difference in time between 25.05 days and 30 days is relatively inconsequential, especially when considering the level of confidence of the source data. It is known that the value used for the POB can have a significant impact on the Large Early Release Fraction (LERF) estimates. The generic data was based off data in the 1980s and is not indicative of the US nuclear fleet today. The site-specific data is based on actual wear test data and was recently confirmed using the latest site data. The 3X (times three) multiplier was a generic adjustment to bound the 95th percentile of an unknown statistical parameter. The latest set of data (not currently in the model) provided a 95th confidence interval of approximately 1.5X, which is significantly less than the 3X multiplier and would further reduce the gap between 25.05 and 30 days. Finally, the statistical POB of 1.4E-03, is approximately the same value as the yearly mean for the industry initiating event frequency of SGTRs, 1.78E-03 (Reference 1). Given that the initiating event data supports tube ruptures having a low frequency of occurrence, and the confidence of the data supports use of a lower multiplier, the sensitivity results from 25.05 to 30 days is not a source of uncertainty when all models are using a site-specific parameter for the POB. Therefore, the POB is not a source of uncertainty in the RICT program.
References:
1.
Idaho National Laboratory (INL), "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants: 2020 Update,"
(INL/EXT-21-65055), dated November 2021
Enclosure 2CAN042403 Page 4 of 4 APLA RAI 02 - FLEX Fire PRA Update of Failure Probabilities Regulatory Basis Item 10 of Section 2.3.4 of NEI 06-09 states, in part: "PRA modeling (i.e., epistemic) uncertainties shall be considered in application of the PRA base model results to the RMTS program." The NRC safety evaluation for NEI 06-09 states that this consideration is consistent with Section 2.3.5 of RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications," Revision 1, (ML100910008).
NRC staff memorandum dated May 6, 2022, "Updated Assessment of the Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments" (ML22014A084), provides the staffs assessment of challenges in incorporating Diverse and Flexible Mitigation Strategies (FLEX) equipment and strategies into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.200.
Conclusion 6, concerning FLEX equipment failure data, in the May 6, 2022, NRC memorandum, instructs that licensees should not use failure rates of permanently installed equipment to represent portable equipment, even if sensitivity analyses are performed.
Issue In response to APLA Question 08 in the LAR supplement dated January 11, 2024, the licensee stated that the Fire PRA is being updated using PWROG-18042-N failure rates and is to be completed to support the RICT program implementation. The NRC staff notes that this planned future action by the licensee appears to be an implementation item for this application, however no implementation item was provided in the LAR supplement.
Request for Additional Information Propose a mechanism, such as an implementation item, that provides reasonable assurance to the NRC staff that the Fire PRA model will be updated using the FLEX equipment failure rates in PWROG-18042-N prior to implementing the RICT program.
Entergy Response Entergy agrees and proposes to add the following implementation item to provide reasonable assurance that the Fire PRA model shall be updated using the FLEX equipment failure rates in PWROG-18042-N prior to implementing the RICT program for ANO-2:
Proposed Implementation Item Implementation Item Completion Date
- 1. FLEX Equipment Failure Rates - The Fire PRA model shall be updated using the FLEX equipment failure rates in PWROG-18042-N.
Prior to implementing the RICT program for ANO-2 and within the Technical Specification amendment implementation grace period specified in the Safety Evaluation Report (SER)