ML23019A219
| ML23019A219 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/01/2023 |
| From: | Tam Tran NRC/NMSS/DREFS/ELRB |
| To: | Peters K Comanche Peak Nuclear Power Co |
| References | |
| EPID L-2022-LNE-0004 | |
| Download: ML23019A219 (1) | |
Text
Ken J. Peters Senior Vice President and Chief Nuclear Officer Attention: Regulatory Affairs Comanche Peak Nuclear Power Plant Vistra Operations Company LLC 6322 N FM 56 P.O. Box 1002 Glen Rose, TX 76043
SUBJECT:
LICENSE RENEWAL SEVERE ACCIDENT MITIGATION ALTERNATIVES AUDIT PLAN REGARDING THE COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NUMBERS 1 AND 2, LICENSE RENEWAL APPLICATION (EPID NUMBER: L-2022-LNE-0004) (DOCKET NUMBERS: 50-445 and 50-446)
Dear Ken Peters:
The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing the Vistra Operations Company LLC license renewal application for Comanche Peak Nuclear Power Plant, Unit Numbers 1 and 2. The NRC staff plans to conduct an in-office audit related to the severe accident mitigation alternatives, beginning February 13, 2023, by teleconference and online reference portal. The audit activities will be conducted in accordance with the audit plan provided in Enclosure 1. The list of audit questions is provided in Enclosure 2.
If you have any questions, please contact me via email at Tam.Tran@nrc.gov.
Sincerely, Tam Tran, Project Manager Environmental Review License Renewal Branch Division of Rulemaking, Environmental, and Financial Support Office of Nuclear Material Safety and Safeguards Docket Nos. 50-445 and 50-446
Enclosures:
As stated cc w/encls: ListservFebruary 1, 2023 Signed by Tran, Tam on 02/01/23
ML23019A219 *via email OFFICE PM:REFS LA:REFS BC:REFS PM:REFS NAME TTran *AWalker-Smith TSmith TTran DATE 1/9/2023 1/24/2023 1/27/2023 1/31/2023
LICENSE RENEWAL AUDIT PLAN RELATED TO THE SEVERE ACCIDENT MITIGATION ALTERNATIVES - COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2
- 1. Background
By letter dated October 3, 2022 (Agencywide Documents Access and Management Systems (ADAMS) ML22276A082), Vistra Operations Company LLC (Vistra or the applicant), submitted to the U.S. Nuclear Regulatory Commission (NRC or staff) an application to renew the Comanche Peak Nuclear Power Plant, Units 1 and 2 (CPNPP), renewed facility operating licenses NPF-87 and NPF-89. The staff is reviewing the information contained in the environmental report (ER) of the license renewal application (LRA) per title 10 of the Code of Federal Regulations (10 CFR) part 51.
As part of the staffs review, a severe accident mitigation alternatives (SAMA) audit will be conducted for CPNPPs LRA. This audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of the licensing or regulatory decision. Specifically, the NRC staff will identify pertinent SAMA information and data and obtain clarifications regarding information provided in the ER.
Per NRC audit bases (see next paragraph, Audit Bases), the NRC staff prepares a regulatory audit plan that provides a clear overview of audit activities and scope, team assignments, and schedule.
- 2. Audit Bases
License renewal requirements are specified in 10 CFR 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants. Licensees are required by 10 CFR 54.23 to submit an ER that complies with the requirements in 10 CFR 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, as part of the LRA.
Review guidance for the staff is provided in NUREG-1555, Standard Review Plans for Environmental Reviews for Nuclear Power Plants: Supplement 1 - Operating License Renewal, as revised.
Section 51.53(c)(3)(ii)(L) of 10 CFR states that If the staff has not previously considered severe accident mitigation alternatives for the applicant's plant in an environmental impact statement or related supplement or in an environmental assessment, a consideration of alternatives to mitigate severe accidents must be provided. The NRC staff must consider severe accident mitigation alternatives for the renewal of CPNPP licenses.
- 3. Audit Scope
The scope of the SAMA audit is to review the SAMA analysis and results as documented in the CPNPP ER and supporting documents.
- 4. Information and Other Material Necessary for the Audit
As described in the list of audit questions (Enclosure 2).
Enclosure 1
- 5. Team Assignments
Name Responsibility J. Dozier Senior Reliability and Risk Analyst E. Dickson Physical Scientist T. Tran Project Manager
- 6. Logistics
The regulatory audit will begin as an in-office audit (at NRC Headquarter using Microsoft Teams and/or CPNPP online reference portal)) on February 13, 2023. Discussions between the NRC staff and Vistra staff will be held regarding topics and questions identified in Enclosure 2, list of audit questions.
- 7. Special Requests
The NRC staff requests the applicant to make available the license renewal application program basis documentation and other documents as requested for independent searches by the NRCs audit team via the electronic document portal.
- 8. Deliverables
An audit summary report is scheduled to be issued by NRC staff within 90 days from the end of the audit.
LIST OF AUDIT QUESTIONS COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION REVIEW IN-OFFICE POSTULATED ACCIDENT SEVERE ACCIDENT MITIGATION ALTERNATIVES
The Environmental Report (ER) states:
For CPNPP, the newer internal event information accounts for a decrease in CDF by a factor of seven. The conservatism in the upper bound estimates utilized in the 1996 GEIS account for other potential reductions in risk, including a factor of 5 for newer source term and population dose, an additional factor of 2 to account for conservatism built in the 1996 GEIS, and a factor of 3 to address areas of uncertainty. These factors are on the conservative end of the ranges provided. When these factors are applied, the net change in risk for CPNPP is reduction by a factor 12.3 (7 + 5 + 2 + 3 - 4.7 = 12.3).
Relative to this statement, the U.S. Nuclear Regulatory Commission (NRC) staff would like to discuss with the applicant regarding the items listed below.
For clarification, please confirm the following statements (staff understanding of the information from the ER):
- 1. The revision 3 Comanche Peak Nuclear Power Plant (CPNPP) Probabilistic Risk Assessment (PRA) had an internal events Core Damage Frequency (CDF) of approximately 9.30E-6/year. The revision 3 PRA internal events including internal flooding CDF is 9.37E-6/year. The noted CPNPP IPE CDF was 5.72E-5/year and was based on Unit 1 and determined to be applicable to both Unit 1 and Unit 2. The current model of record (revision 5) PRA has a CDF of approximately 1.1E-6/year for each unit.
The current internal events including internal flooding CDF is 1.22E-6 and 1.25E-6 per year for Unit 1 and Unit 2, respectively. These PRA model refinements represent an approximately 98 percent reduction in CDF from the Individual Plant Examination for External Event CDF (about a factor of 52) and an approximately 88 percent reduction in CDF from the revision 3 CDF (about a factor of 7) for each unit for the internal events (i.e., excluding internal flooding) PRA.
- 2. Vistra provided the base case CDF values used to evaluate severe accident mitigation alternatives (SAMAs) in the ER. The sum of the external events CDF (6.2E-5); fire, High Winds, and external Flooding CDFs (4.20E-05 per reactor-year, 3.7E-06 per reactor-year 1.59E-05 per reactor-year, respectively) is greater than the CPNPP internal event CDF(1.1E-6 per reactor-year), but is within the range of Pressurized Water Reactor (PWR) CDFs (4.4E-5 to 3.5E-4 per reactor-year) and only slightly above 5.9x10-5 per reactor-year which is the internal events mean value CDF for PWRs that the 2013 GEIS used to estimate probability-weighted, offsite consequences from airborne, surface water, and groundwater pathways, as well as the resulting economic impacts from such pathways.
- 3. The predicted early and latent fatalities and dose estimates per reactor-year (RY) for Comanche Peak are provided in table 5.6 of the 1996 Generic Environmental Impact Statement (GEIS). The very conservatively predicted latent total fatalities/RY (95 percent
Enclosure 2
upper-confidence bound) were determined to be 2.3E-03 in the 1996 GEIS. In the CPNPP ER, the total CDF (a surrogate for the individual latent cancer fatality risk) was calculated to be 4.30E-05 (over a factor of 500 improvement).
- 4. Although not a physical change to CPNPP or to the explicit PRA modeling, volume 2 of NUREG-7110 State-of-the-Art Reactor Consequence Analysis (SOARCA) was published in August 2013. This analysis updated the NRC's severe accident studies of the Surry Power Station (e.g., NUREG-1150), incorporating state-of-the-art analyses to evaluate offsite risk. The conclusions of the SOARCA analysis were that the calculated risks of public health consequences from severe accidents modeled in SOARCA are very small" and "The unmitigated versions of the scenarios analyzed in SOARCA have lower risk of early fatalities than calculated in the 1982 Siting Study SST 1 case in SOARCA. SOARCA's analyses show essentially zero risk of early fatalities. As stated in SOARCA, The actual risk of a prompt fatality (cf., table 7-13), using current best-estimate practices for calculating source terms, is about 5 orders of magnitude lower than using the SST1 source term would imply (cf., table 7-13 and table 7-18). Included in the SOARCA state-of-the-art analyses are evaluations of steam generator tube ruptures, demonstrating that their offsite consequences are less than previously modeled. The SOARCA analysis was not a complete analysis of all scenarios in the PRA, but it supports the conclusion that the offsite effects from a severe accident would be small. Comanche Peak is a very similar design to Surry, both are Westinghouse PWRs with large, dry containments, and the general conclusions of lower offsite consequences from the SOARCA apply to Comanche Peak as well.
- 5. More recent source term (timing and magnitude) at Comanche Peak has significantly smaller effects than had been quantified in the 1996 GEIS. For the Comanche Peak SAMA new and significant evaluation (described in ER section 4.15.3), SAMAs were evaluated for impact on CDF and source term category group frequencies if they were implemented. None of the SAMAs evaluated were found to reduce a significant source term category group frequency by at least 50 percent. Therefore, the offsite consequences of severe accidents initiated by the new source term during the subsequent license renewal term would not exceed the impacts predicted in the GEIS.
- 6. The 2020 population used in the Comanche Peak initial license renewal ER by Vistra was extrapolated to the year 2030 and found to be 6,987,542. In the ER, Vistra extrapolated the population within the 50-mile radius to the year 2054. Vistra projected the total population for the year 2054 to be 9,465,735. This is an increase of 35 percent (factor of 1.35) which is only slightly over the GEIS range of 5 to 30 percent change which the GEIS concludes does not generally result in significant increases in impacts.
The effect of the reduction in risk cited elsewhere far exceeds the effect of a population increase.
NEI 17-04 section 3.1, Data Collection, explains that the initial step of the assessment process is to identify the new information relevant to the SAMA analysis and to collect and develop those elements of information that will be used to support the assessment. The guidance document states that each applicant should collect, develop, and document the information
elements corresponding to the stage or stages of the SAMA analysis performed for the site.
Please be prepared to discuss this process.
Table E4.15-2 of the ER provides the Reduction in Maximum Benefit of Screened-In SAMAs.
Please be prepared to discuss a sample that resulted in higher thresholds (primarily small early release frequency (SERF).
Please be prepared to discuss the statement, Many individual STCs have a frequency that is insignificant, and while an insignificant STC could in theory be reduced by >50 percent, its impact on MB would be negligible. Additionally, many STCs have conditional offsite consequences that are negligible compared to the dominant STC groups (i.e., SERF, LLRF and LERF).
List of Acronyms:
ER Environmental Report CPNPP Comanche Peak Nuclear Power Plant CDF Core Damage Frequency GEIS Generic Environmental Impact Statement PRA Probabilistic Risk Assessment IPEEE Individual Plant Examination for External Event SAMA Severe Accident Mitigation Alternatives SOARCA State of the Art Reactor Consequence Analyses NEI Nuclear Energy Institute SERF Small early release frequency STC Source Term Category LLRF Large Late Release Frequency LERF Large Early Release Frequency