ML26036A178
| ML26036A178 | |
| Person / Time | |
|---|---|
| Issue date: | 02/12/2026 |
| From: | Andy Campbell NRC/RES/DSA |
| To: | Tony Nakanishi NRC/NRR/DRA |
| References | |
| SAND2026-15948 | |
| Download: ML26036A178 (0) | |
Text
MEMORANDUM TO:
Tony Nakanishi, Acting Director Division of Risk Assessment Office of Nuclear Reactor Regulation FROM:
Victor Hall, Acting Director Division of Systems Analysis Office and Nuclear Regulatory Research
SUBJECT:
PUBLICATION OF SAND2026-15948 FOR UPDATING REMOVAL RATES FOR PWR CONTAINMENTS IN RG 1.183 REV. 2 The purpose of this memorandum is to notify NRR of the recent publication of SAND2026-15948, Pressurized Water Reactor Containment Aerosol Settling Rate for Accident Source Terms (SNL, 2026) - ML26021A079. This report provides updated recommendations for containment aerosol settling rates for Pressurized Water Reactors (PWRs) that are consistent with the revised source term in SAND2023-01313 (SNL, 2023) for integration into DG1425 (RG 1.183 Rev. 2) (NRC, 2026).
Background
The NRC, with technical support from Sandia National Laboratories, recently developed an updated Alternative Source Term (AST) to facilitate the licensing of Accident Tolerant Fuels (ATFs), increased enrichments, and higher fuel burnups. These updated ASTs have been documented in SNL reports such as SAND2023-01313 (SNL, 2023), SAND2024-10670 (SNL, 2024), and SAND2024-10673 (SNL, 2024). These ASTs have been adopted by DG1425 (RG 1.183 Rev. 2) (NRC, 2026) to demonstrate compliance with 10 CFR 50.67, Accident Source Term. (64 FN 71990, Dec. 23, 1999).
Reduction in airborne radioactivity in the containment by natural deposition within the containment can be credited to demonstrate compliance with 10 CFR 50.67. Acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2, Containment Spray as a Fission Product Cleanup System, of the Standard Review Plan (NRC, 2007) and in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments. (NRC, 1996). These models are incorporated into the analysis code RADTRAD (NRC, 2016) which performs dose calculations for releases to the environment. In the past, staff have found acceptable the use of NUREG/CR-6189 at the 10th-percentile values. However, the NUREG/CR-6189 models are not recommended in DG1425. SAND2026-15948 provides PWR source term characteristics (removal rates and timings) based on best estimate MELCOR calculations that are applicable to the PWR source term in DG1425.
CONTACT: Shawn Campbell, RES/DSA/FSCB February 6, 2026 Signed by Hall, Victor on 02/06/26
T. Nakanishi Discussion SAND2026-15948 documents MELCOR-based, mechanistic calculations and uncertainty quantification to evaluate aerosol deposition in PWR containments during severe accidents. The analysis considers the same representative containment designs and accident sequences from SAND2023-01313, incorporating uncertainty parameters from the Surry and Sequoya State-of-the-Art-Consequence Analyses (SOARCA) (NRC, 2012) (NRC, 2019). New aerosol settling rate recommendations are developed for inclusion in section A-2.2 of a future revision of RG 1.183. The report also provides recommendations for the impact of containment sprays on the PWR containment removal rate.
Results from SAND2026-15948 show strong dependencies on both the sequence specifications and the plant design. The effect is so significant that the 10% and 90% results are dominated by one plant type and one or two similar sequences. Using the 10% and 90% values would neither meet the requirements and regulatory positions of an AST since these tail ends of the distribution do not represent the spectrum of results considered in the study. These requirements and regulatory positions are discussed in the Federal Register for 10 CFR 50.67, Accident Source Term, and regulatory positions in of RG 1.183 which specify a representative source term.
Key Findings MELCOR results show higher aerosol settling rates during both the in-vessel and long-term phases compared to NUREG/CR-6189. The increase is due to advanced, integrated modeling of thermal-hydraulics, aerosol physics, and sequence-specific containment conditions.
Sensitivity studies confirm that containment sprays at typical flow rates enhance aerosol removal and are consistent with NUREG/CR-5966 (D. Powers, 1993) median values.
Recommendations The updated removal rates should be applied to Section A-2.2 of DG-1425 (RG 1.183, Revision 2), if possible, for containment aerosol removal assumptions in AST evaluations for PWRs. Otherwise, the updated removal rates should be included in a future revision of RG 1.183. Table 1 provides the median coefficients recommended for dose modeling applications because they best represent the diverse accident sequences and plant designs considered in this study.
T. Nakanishi Table 1 Recommended effective aerosols settling coefficients for the high burnup source term from SAND2026-15948.
The thermal power scaling described in NUREG/CR-6189 is not necessary with these new recommended values since containment design and sequence diversity are adequately captured.
The median values from NUREG/CR-5966 should be used for the spray decontamination factor.
T. Nakanishi References D. Powers. (1993). NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays. Washington DC, (ML063480542): U.S. Nuclear Regulatory Commission.
NRC. (1996). NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments". Washington DC, (ML100130305): U.S. Nuclear Regulatory Commission.
NRC. (2007). NUREG 0800, Chapter 6.5.2, Containment Spray as a Fission Product Cleanup System. Washington, DC: US Nuclear Regulatory Commission.
NRC. (2012). NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA)
Report, Part 1 and Part 2". Washington DC, (ML12332A058): U.S. Nuclear Regulatory Commission.
NRC. (2016). NUREG/CR 6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation," April 1998 (ML15092A284)l NUREG/CR-7220, "SNAL/RADTRAD 4.0: Description of Models and Methods,". Washington, DC: U.S.
Nuclear Regulatory Commission.
NRC. (2019). NUREG/CR-7425, State-of-the-Art Reactor Consequence Analyses Project:
Sequoyah Integrated Deterministic and Uncertainty Analysis.. Washington DC, October 2019: U.S. Nuclear Regulatory Commission.
NRC. (2026). Regulatory Guide RG 1.183, Revision 2 (Draft Guide 1425), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC, October 2023 (ML24185A179). Washington DC: U.S.
Nuclear Regulatory Commission.
SNL. (2023). SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis.
Albuquerque, NM, (ML23097A087): Sandia National Laboratory.
SNL. (2024). SAND2024-10670, Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis". Albuquerque, NM, (ML24229A069): Sandia National Laboratories.
SNL. (2024). SAND2024-10673, "Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement". Albuquerque, NM, (ML24229A063): Sandia National Laboratories.
SNL. (2026). SAND2026-15948, "Pressurized Water Reactor Containment Aerosol Settling Rate for Accident Source Terms". Albuquerque, NM, (ML26021A079): Sandia National Laboratories.
ML26036A178 OFFICE RES/DSA/FSCB RES/DSA/FSCB: BC NRR/DRA/ARCB: BC RES/DSA NAME SCampbell HEsmaili DGarmon VHall DATE 2/6/2026 2/6/2026 2/6/2026 2/6/2026