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NEI White Paper - Impact of Higher Source Term Fractions on EQ Doses
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Issue date:	06/13/2024
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WHITE PAPER

NEI White Paper: Impacts of Higher Source Term Release Fractions On Environmental Qualification

Prepared by the Nuclear Energy Institute June 2024

June 2024

Acknowledgements

This document was developed by the Nuclear Energy Institute. NEI acknowledges and appreciates the contributions of NEI members and other organizations in providing input, reviewing, and commenting on the document.

NEI Project Lead: Fran kie Pimentel

Technical Lead: Greg Broadbent, NEI Consultant - GB Atomics, LLC

Technical Advisors:

Jan Bostelman, Bostelman Engineering, LLC

Butch Bornt, Southern Company

Aladar Csontos, NEI

David Hindera, GE Vernova

Bill Kohlroser, Dominion

Samuel Lafountain, Southern Company

Alex Markivich, Dominion Energy

Rich McCarty, Winston & Strawn, LLP

Carlos Sisco, Winston & Strawn, LLP

Micheal Smith, NEI

Ron Wise, Nuclear Utility Group on Equipment Qualification

Notice

Neither NEI, nor any of its employees, members, supporting organizations, contractors, or consultants make any warranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights.

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Goal, Objectives, and Scope

The goal of this white paper is to provide a technical and licensing justification for the continued use of a TID-14844 based source term for environmental qualification of electric equipment subject to 10 CFR 50.49 and how this satisfies the applicable NRC requirements for those licensees who opt to modify their licensing basis associated with AST implementation.

The objective of this paper is to support the scope of environmental qualification analyses in association with Section 1.3.5 of Regulatory Guide 1.183 as it relates to determining whether licensee action is warranted. This paper supports the development of NRC regulatory positions in upcoming guidance with:

Regulatory positions developed from regulations and licensing precedents, and

Technical positions fro m quantitative evaluations of four source terms on the radiation environmental qualification analyses Key Insights and Conclusions

The key insights and conclusions from this white paper are:

The continued use of a TID-14844 based source term for environmental qualification of electric equipment is justified based on both regulatory and technical positions.

o The use of the TID -14844 source terms satisfies the requirements in 10 CFR 50.49 regarding the consideration of the most severe design basis accident. It also meets both 10 CFR 100.11 and 10 CFR 50.67 related to the design basis radiological accidents being assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

o The difference in accident source terms derived from SAND2011 -0128 and SAND2023-01313 and TID-14844 is primarily from improved understanding of the physics of core meltdown accidents and is not attributed to fuel burnup, use of MOX fuel, or elevated enrichment.

o For equipment inside containment, the airborne doses represent the highest contributor to the integrated accident dose used by EQ. As presented in Workshop #2 [16], the accident airborne dose based on TID-14844 continues to bound the airborne gamma and beta doses when changes in core inventories are ignored. When changes in core inventories are considered, the larger inventories of long-lived beta-emitting isotopes in fuel with higher exposures causes the integrated dose from the SAND2011 -0128 and SAND2023-01313 to exceed the TID dose in the long -term (i.e., after 100 days or longer). However, this impact is already considered in TID-based EQ analyses since any potential changes in core inventories are addressed through scaling factors, as needed.

o The accident pool dose based on TID-14844 is initially higher but is not bounding for the entire design basis survivability period. 1 The pool dose from SAND2023-01313 exceeds the

1 Also referred to as the post-accident operating time.

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pool dose derived from TID-14844 after approximately 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months for both the PWR and BWR cases. However, the substantial safety margin incorporated with the application of the TID source term more than accommodates the impact of these higher EQ doses.

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Table of Contents

Introduction..................................................................................................................................... 1 Key Terms Used in this Paper.......................................................................................................... 1 Environmental Qualification............................................................................................................ 2 Background...................................................................................................................................... 4 4.1 Release Fractions and Timing............................................................................................. 4 4.2 Airborne EQ Doses.............................................................................................................. 7 4.3 Pool\\Sump EQ..................................................................................................................... 9 Design Inputs................................................................................................................................. 11 5.1 Release Fractions and Durations...................................................................................... 11 5.2 Radionuclide Groups......................................................................................................... 12 5.3 Core Inventories................................................................................................................ 13 Methodology.................................................................................................................................. 17 6.1 Plant Dose Models............................................................................................................ 17 6.1.1 PWR Airborne Model........................................................................................... 18 6.1.2 BWR Airborne Model........................................................................................... 18 6.1.3 Pool Model........................................................................................................... 19 6.2 Beta Dose Model............................................................................................................... 19 6.3 Gamma Dose Model......................................................................................................... 20 6.3.1 Airborne Model.................................................................................................... 20 6.3.2 Pool Model........................................................................................................... 21 Results............................................................................................................................................ 22 Summary and Conclusions............................................................................................................. 29 References..................................................................................................................................... 30

Table of Figures

Figure 4-1 Total BWR Release Fractions....................................................................................................... 5 Figure 4-2 Total PWR Release Fractions....................................................................................................... 6 Figure 4-3 Release Durations........................................................................................................................ 6 Figure 4-4 Sandias Airborne Beta Dose for Surry........................................................................................ 7 Figure 4-5 Sandias Airborne Gamma Dose for Surry................................................................................... 8 Figure 4-6 Sandias Airborne Beta Dose for Grand Gulf............................................................................... 8 Figure 4-7 Sandias Airborne Gamma Dose for Grand Gulf.......................................................................... 9 Figure 4-8 Sandias Pool Dose Results for Grand Gulf................................................................................ 10

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Figure 4-9 Sandias Pool Dose Results for Surry......................................................................................... 11 Figure 6-1 Airborne Dose Models............................................................................................................... 19 Figure 6-2 Containment Air MCNP Model.................................................................................................. 21 Figure 6-3 Piping Dose Model..................................................................................................................... 22 Figure 7-1 BWR Airborne Beta Dose........................................................................................................... 23 Figure 7-2 BWR Airborne Gamma Dose...................................................................................................... 24 Figure 7-3 BWR Waterborne Gamma Dose................................................................................................ 25 Figure 7-4 PWR Airborne Beta Dose........................................................................................................... 26 Figure 7-5 PWR Airborne Gamma Dose...................................................................................................... 27 Figure 7-6 PWR Waterborne Gamma Dose................................................................................................ 28

Table of Tables

Table 5-1 BWR Release Fractions................................................................................................................ 11 Table 5-2 PWR Release Fractions................................................................................................................ 12 Table 5-3 Radionuclide Groups................................................................................................................... 12 Table 5-4 Fuel Assembly Characteristics..................................................................................................... 13 Table 5-5 Fuel Cycle Characteristics............................................................................................................ 13 Table 5-6 BWR Core Inventories................................................................................................................. 14 Table 5-7 PWR Core Inventories (Curies).................................................................................................... 16 Table A-1 Isotopes Modeled in Applicability Assessment........................................................................ A-2 Table A-2 Airborne Dose Contribution by Isotope.................................................................................... A-3 Table A-3 Waterborne Dose Contribution by Isotope.............................................................................. A-4

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INTRODUCTION

Existing radiation environmental qualification analyses are based on a plants core inventories and licensed source term. As a plants core inventories change, these analyses are addressed with simple scaling factors, if necessary. The current source term licensed for existing plants is Technical Information Document 14844 [1]; however, an updated source term involving significant increases in release fractions has been described in Regulatory Guide 1.183, Revision 1 and recent research in SAND2023-01313.

The purpose of this white paper is to provide a technical justification for the continued use of a TID -

14844 based source term for environmental qualification of electric equipment subject to 10 CFR 50.49 and how this satisfies the applicable NRC requirements for those licensees who opt to modify their licensing basis associated with AST implementation. This justification is intended to support the scope of environmental qualification analyses in association with Section 1. 3.5 of Regulatory Guide 1.183 as it relates to determining whether licensee action is warranted.

KEY TERMS USED IN THIS PAPER

Core Inventories: The mix of the radionuclides in the reactor fuel at the onset of an accident and are typically expressed in terms of Curies-per-Megawatt (Ci/MW). The core inventories are based primarily on core power, exposure, enrichment, and uranium mass.

Environmental Qualification (EQ): In compliance with 10 CFR 50.49, the qualification of safety-related electrical equipment to the expected post-accident conditions in which their functions are credited.

These conditions include (i) temperature and pressure, (ii) humidity, (iii) chemical effects, and (iv) radiation. This paper addresses the applicable source terms for the radiation qualification.

Maximum Hypothetical Accident (MHA) Loss of Coolant Accident (LOCA): As described in Appendix A to Regulatory Guide 1.183 Rev. 1, a conservative surrogate accident that is intended to challenge aspects of the facility design. An MHA LOCA is typically assumed as the design basis case for evaluating the performance of release mitigation systems and the containment and for evaluating the proposed siting of a facility. This accident assumes a deterministic substantial core damage source term released into an intact containment.

Scaling Factors: Multipliers to the calculated EQ dose to accommodate changes in core inventories.

These scaling factors are developed by shielding calculations based on the radionuclides to which the component is exposed and the associated operating duration of the component.

Source Term: The magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release per 10 CFR 50.2. Four source terms are addressed in this white paper: (i) Technical Information Document 14844, (ii) Regulatory Guide 1.183 Revision 0, (iii) Regulatory Guide 1.183 Revision 1, and (iv) SAND2023-01313.

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ENVIRONMENTAL QUALIFICATION

The regulation containing the requirements for environmental qualification is 10 CFR 50.49. Regarding the radiation environment, 10 CFR 50.49(e)(4) requires:

The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.

This requirement has generally been met using the TID -14844 source term, originally developed to meet the requirements in 10 CFR 100.11. As noted in the regulation, this source term is

based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

The TID source term releases 100% of the core noble gases, 50% of the core iodine, and 1% of core solids (particulates) and is based on significant meltdown of the fuel. The largest activity release into containment from a design basis accident is typically associated with the Appendix K loss of coolant accident or PWR rod ejection in which a small fraction of the fuel rods is postulated to fail and release gap activity into the containment atmosphere. As such, the activity transient associated with the TID source term clearly bounds the most severe design basis accident applied in plant design basis, meeting the requirements of 10CFR50.49. This conclusion is no different than that developed previously by the NRC Staff.

As reflected in NUREG-0458 [17]:

The conservative positions taken in RG 1.89 assume an instantaneous release and uniform distribution in the containment atmosphere of 100% of the noble gases, 25% of the iodines, and 1% of the solids in the reactor core. An additional 25% of the iodines are assumed to plate out on the containment walls.

These estimates represent core degradation substantially beyond that which ECC systems are designed to prevent. With a minimum operation of the core cooling systems, as required by 10 CFR 50.46 and Appendix K, it is expected that the fission products released from such an event would be significantly lower than the assumed values (a factor of 10 to 100). In addition, other assumptions, such as uniform distribution and exposure calculations at the center of containment with no shielding, represent a significant conservatism. Therefore, a large safety margin exists in the level of environment qualification for radiation due to the source term assumed in staff licensing reviews.

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More recently, in response to an EQ source term proposed by NuScale, the Staff found, in SECY 0079

[18], that:

NuScale's proposal to use an accident source term without consideration of a core melt for environmental qualification, if technically justified, is consistent with 10 CFR 50.49 and presents an acceptable regulatory approach to addressing accident source term.

The SECY later acknowledges that NuScale's proposed approach to environmental qualification differs from past practice but is consistent with the language of 10 CFR 50.49 and compatible with past Commission direction. This SECY also notes that, although 10 CFR 50.49(e)(1) also references "the most severe design basis accident" for temperature and pressure effects, the staff has not used a core melt accident to define these parameters. The application of a source term without core melt as addressed in this SECY demonstrates that the TID source term maintains significant conservatism while still meeting the requirements in 10 CFR 50.49.

Consistent with the defense-in-depth philosophy, the continued use of the TID-14844 source term maintains a large safety margin between radiological conditions used to establish environmental qualification under 10 CFR 50.49 and the design basis for the ECC systems related to preventing core damage. Due to the significant safety margin that is inherent to the TID -14844 source term, there would be no discernible risk reduction associated with adopting an AST source term for EQ for licensees that choose to modify their licensing basis to higher burnup or enrichment than those specified in RG 1.183 Revision 0. This conclusion is consistent with the staffs resolution of GSI-187.

Overall, this approach continues to satisfy the requirement in 10 CFR 50.49(e)(4) for the radiation environment used for qualification of electrical equipment including that it must be associated with the most severe design basis accident during or following which the equipment is required to remain functional. The core inventories can be affected by many plant modifications such as power uprates, cycle extensions, or changes to fuel or core designs. In the event these plant changes increase the core inventory over that currently analyzed, the impacts on the plant environmental qualification program are assessed. These evaluations typically apply multipliers to the radiation profiles, called scaling factors, based on the difference between the updated core inventory and that applied in the EQ analysis of record. For licensees whose EQ programs are based on the TID source term, these evaluations would continue to apply the TID source term, as updated with the new core inventories. For those plants needing to perform AST-based EQ analyses, Appendix I to Reg Guide 1.183, Rev. 0 contains acceptable assumptions for evaluating radiation doses for environmental qualification.

The NRC has recently re-confirmed this position in association with a comment resolution on DG-1389

[11]. In the response to Comment 7 -1, the Staff indicates that:

The source term in Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, issued March 1962 (ML021750625), continues to be adequate for the environmental qualification (EQ) analysis for those plants that currently use TID-14844 for EQ in their current licensing basis and have not made significant plant modifications affecting source terms or the EQ analysis.

In addition to the licensing arguments above, this evaluation assesses, on a generic basis, the impact of a transition to the Revision 1 and SAND2023-01313 source term on TID-based environmental qualification profiles.

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B A C KG R O U N D

4.1 Release Fractions and Timing

The original source terms for most domestic nuclear plants are based on Technical Information Document (TID)-14844 [1] released in 1962. This document reports an accident source term that assumes the instantaneous release of 100% of the core noble gases, 50% of the iodines and 1% of the core solids, including cesium. The released iodine is primarily in the form of elemental iodine and half of the released iodine is assumed to remain airborne.

For the following 30 years, substantial additional information on fission product releases was developed based on significant severe accident research. In 1995, NRC issued NUREG-1465 [2] quantifying a new source term developed with the NRCs Source Term Code Package (STCP). This research concluded the releases would occur over some time (rather than instantaneous) and that the released iodine would be primarily in the aerosol species (rather than primarily elemental). The NUREG-1465 release fractions were adopted into Regulatory Guide 1.183 Revision 0 [3] for application by licensees. Since it was anticipated that these release fractions would be dependent on the fuel exposure, the applicability of the release fractions was limited to fuel burnups up to 62 G Wd/MTU (peak rod) in Footnote 10 of Reg Guide 1.183.

In support of industry initiatives to increase fuel burnup past 62 GWd/MTU, the NRC has re-examined these accident scenarios with the updated severe accident methodology in MELCOR. In October 2023, the NRC released Revision 1 [4] to Regulatory Guide 1.183. The release fractions and timings for the Maximum Hypothetical Accident (MHA) in this proposed revision were based on Sandia Report SAND2011- 0128 [5], which extended the maximum burnup from 62 to 68 GWd/MTU (peak rod -

average). PWR analyses also considered Mixed Oxide (MOX) fuel. This research reported significantly higher halogen releases in BWRs, as well as slightly lower alkali metal (i.e., cesium) releases. The changes in release fractions, however, were not attributed to the higher exposure but due to an improved understanding of accident progression based on the abstract to Reference 5.

Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term.

As part of ongoing research into proposed burnups as high as 80 GWd/MTU and enrichments up to 10 wt%, the NRC recently released SAND2023 -01313 [6] which reported even higher halogen releases than Reference 5. Consistent with the previous study, the abstract to SAND2023-01313 indicates:

Larger releases are observed, in part, because of increased sampling of low-pressure accident scenarios relative to NUREG-1465; low-pressure accident scenarios are known to cause larger releases to containment. Finally, this analysis demonstrates that in-containment source terms are essentially unchanged by increased burnup or elevated enrichment and that the most

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significant variation in source term continues to arise from differences between accident scenarios.

SAND2023- 01313 describes several model enhancements which tend to result in depressurized conditions, such as the high-temperature seizure of BWR safety relief valves and PWR hot leg creep rupture.

From the sources described above, the release fractions for the radionuclide groups that have significant releases are illustrated in Figure 4-1 and Figure 4-2 for BWRs and PWRs respectively. As shown in the figures, all models predict that effectively all noble gas inventory in the core would be released.

However, in Revision 1, although the BWR halogen releases significantly increased, the PWR halogen releases dropped slightly. Revision 1 proposes slightly lower cesium releases for both BWRs and PWRs while tellurium releases dramatically increased due to new knowledge of post-accident tellurium chemistry. In SAND2023-01313, the halogen and alkali metal radionuclide groups saw significant increases in their release fractions.

With the updated MELCOR methodology, improved heat transfer modeling resulted in significant extensions in the duration of the in-vessel release phase as observed in Figure 4-3. In addition, the duration of the gap release was also slightly extended in the recent research in Reference 6.

The impacts of these release fractions and durations can have significant impacts on plant environmental qualification (EQ) and are assessed in this evaluation.

B WR Total R e le ase F rac tions

1. 0

0. 9 TI D -1 48 44 R G 1. 18 3 R e v. 0 S e c t i on V.A Tab le 1

0. 8 R G 1. 18 3 R e v. 1 S AN D 20 23- 01 31 3

0. 7 Tab le 1 Ta b le 5-5

0. 6

0. 5

0. 4

0. 3

0. 2

0. 1

0. 0 No b le Gas es H a l og e n s Al k ali Metal s Tel lu riu m M et als Radionuclide Gr oup

Figure 4-1 Total BWR Release Fractions

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PWR Total R e le ase F rac tions

1. 0

0. 9 TI D -1 48 44 R G 1. 18 3 R e v. 0 S e c t i on V.A Tab le 2

0. 8 R G 1. 18 3 R e v. 1 S AN D 20 23- 01 31 3

0. 7 Tab le 2 Ta b le 5-6

0. 6

0. 5

0. 4

0. 3

0. 2

0. 1

0. 0 No b le Gas es H a l og e n s Al k ali Metal s Tel lu riu m M et als Radionuclide Gr oup

Figure 4-2 Total PWR Release Fractions

R e le ase Durations

9. 0

8. 0 TI D -1 48 44 R G 1. 18 3 R e v. 0 S e c t i on V.A

Tab le 4

7. 0 R G 1. 18 3 R e v. 1 S AN D 20 23- 01 31 3 Tab le 5 Tab le s 5-5 & 5-6

6. 0

Th e TI D -14844 r e l e a s e i s i n s t a n t a n eo u s.

5. 0 Durati on= 0 hours0 days 0 hours 0 weeks 0 months

4. 0

3. 0

2. 0

1. 0

0. 0 BWR G ap B W R I n -V e s se l P WR Gap P WR I n-Ves s el Release Phase

Figure 4-3 Release Durations

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4.2 Airborne EQ Doses

The impact of the AST on airborne EQ profiles was first quantified by Sandia as part of the Staffs re-baselining assessments [8]. The instantaneous release of 100% of the noble gases and 50% of the halogens was determined to bound the phased release of the AST release fractions of these radionuclide groups in both BWRs and PWRs. The small solids component of the TID source term was frequently ignored in many 2 airborne EQ calculations as being of negligible impact. With the smaller, phased releases, it was no surprise that the Sandia report concluded the AST-based EQ doses were less than those based on the TID source term. The integrated airborne doses calculated by Sandia are illustrated in Figure 4-4 through Figure 4-7.

Figure 4-4 Sandias Airborne Beta Dose for Surry

2 Sandia made this assumption in their airborne calculations based on Tables 5-6, 5-7, and 5-8 of the Sandia report.

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Figure 4-5 Sandias Airborne Gamma Dose for Surry

Figure 4-6 Sandias Airborne Beta Dose for Grand Gulf

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Figure 4-7 Sandias Airborne Gamma Dose for Grand Gulf

4.3 Pool\\Sump EQ

As the industry understanding of accident progression advanced by experience gained from the accident at Three Mile Island, it was recognized that the cesium releases may be greater than the guidance reported in TID-14844. As early as 1984, the NRC had recognized that considerable quantities may be appropriate for qualification doses as reported in Appendix D to Reg. Guide 1.89 [7]. As a particulate, cesium would eventually come to reside in the pool water or horizontal surfaces, increasing the dose rate and integrated doses in these areas.

New research into post-accident source term behavior quantified this cesium release in NUREG-1465 [3]

to be as high as 30% over the first ~2 hours of an accident. During the development of the alternative source terms (AST), it was recognized that this 30 -fold increase in cesium release could lead to higher integrated long-term doses, although the shorter -term doses may be lower due to the phased release of the new source term.

Although the TID-based airborne integrated EQ doses were shown to bound those associated with the AST, the Sandia assessment could not make a similar conclusion for the pool doses. As shown in Figure 4-8 and Figure 4-9 (extracted from Reference 8), the short-term integrated dose to the pool with the AST is less than that predicted by the TID source term due to the assumed instantaneous release of the TID source term. After some time, the AST predicts a higher integrated dose than TID -based doses due to the higher cesium releases. This cross-over point was quantified in Reference 8 to be at ~3500 and

~1000 hours for BWRs and PWRs, respectively.

As reported in Section 1.3.5 of Regulatory Guide 1.183, the NRC submitted the larger long-term water doses associated with the alternative source terms as candidate Generic Safety Issue (GSI) 187. An NRC panel subsequently concluded that this candidate generic issue should be dropped, as having no

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significant chance of meeting the incremental risk thresholds for backfit [9]. The panels rationale for this conclusion was based on the conservatisms in the sump dose calculation. Specifically, the panel noted that:

the design basis LOCA has no significant fuel damage,

the complete washdown of all source terms into the pool is not likely,

the suppression pool will not be a homogeneous mixture since some settling is expected,

equipment qualified for a given period may, in practice, remain available for a much longer period,

the lower decay heat load at later time periods will allow operators more opportunity for alternative strategies, and

the risk significant period is limited to the first days.

Based on these findings, plants transitioning to the AST from TID generally maintained their TID-based pool EQ bases. These findings are generally quantitative and not affected by the higher releases; however, the last conservatism could be challenged if the higher proposed source term release fractions cause the cross-over point to occur in the risk-significant first days. The impact on this point is quantified in Section 7 for sample BWRs and PWRs.

Figure 4-8 Sandias Pool Dose Results for Grand Gulf

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Figure 4-9 Sandias Pool Dose Results for Surry DESIGN INPUTS

5.1 Release Fractions and Durations

The release fractions and durations applied in this evaluation are those described in Section 4.1. These release fractions and durations are summarized below. In cases where SAND2023-01313 reports <1.0E -

6, this evaluation will conservatively apply a release fraction of 1.0E -6.

Table 5-1 BWR Release Fractions TID-14844 RG 1.183 Rev. 0 RG 1.183, Rev. 1 SAND2023-01313 Section V.A Tables 1 & 4 Tables 1 & 5 Table 5-5 Release Phase Total Gap In-Gap In-Vessel Gap In-Vessel Vessel Duration (hours) 0.00 0.50 1.50 0.16 7.81 0.70 6.70

Radionuclide Group Noble Gases 1.00 0.05 0.95 0.008 0.96 0.016 0.95 Halogens 0.50 0.05 0.25 0.003 0.54 0.005 0.71 Alkali Metals 0.01 0.05 0.2 0.003 0.14 0.005 0.32 Tellurium Group 0.01 0.00 0.05 0.003 0.39 0.003 0.56 Barium/Strontium 0.01 0.00 0.02 0.00 0.005 0.0006 0.005

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Noble Metals 0.01 0.00 0.0025 0.00 2.70E-03 <1.0E-06 0.006 Cerium Group 0.01 0.00 0.0005 0.00 1.60E-07 <1.0E-06 <1.0E-06 Lanthanides 0.01 0.00 0.0002 0.00 2.00E-07 <1.0E-06 <1.0E-06 Molybdenum Group 0.00 0.03 1.9E-05 0.12

Table 5-2 PWR Release Fractions TID-14844 RG 1.183 Rev. 0 RG 1.183, Rev. 1 SAND2023-01313 Section V.A Tables 2 & 4 Tables 2 & 5 Table 5-6 Release Phase Total Gap In-Gap In-Vessel Gap In-Vessel Vessel Duration (hours) 0.00 0.50 1.30 0.22 4.27 1.30 4.00

Radionuclide Group Noble Gases 1.00 0.05 0.95 0.022 0.94 0.026 0.93 Halogens 0.50 0.05 0.35 0.007 0.37 0.007 0.58 Alkali Metals 0.01 0.05 0.25 0.005 0.23 0.003 0.5 Tellurium Group 0.01 0.00 0.05 0.007 0.3 0.006 0.55 Barium/Strontium 0.01 0.00 0.02 1.40E-03 4.00E-03 0.001 0.002 Noble Metals 0.01 0.00 0.0025 0.00 6.00E-03 <1.0E-06 0.008 Cerium Group 0.01 0.00 0.0005 0.00 1.50E-07 <1.0E-06 <1.0E-06 Lanthanides 0.01 0.00 0.0002 0.00 1.50E-07 <1.0E-06 <1.0E-06 Molybdenum Group 0.00 0.10 2.0E-05 0.15

5.2 Radionuclide Groups

Reg Guide 1.183 Rev. 1 makes a change to the radionuclide grouping where a new group is added for Molybdenum, which includes the elements of Molybdenum (Mo), Technetium (Tc), and Niobium (Nb) based on Table 14 of SAND2011-0128. Zirconium is also relocated to the Cerium group in the later studies. For the different cases evaluated in this evaluation, the applied radionuclide groups are listed below with the impacted elements in red.

Table 5-3 Radionuclide Groups Radionuclide TID-14844 RG 1.183, Rev. 1 Group & &

Reg Guide 1.183 Rev. 0 SAND2023-01313 Noble Gases Xe, Kr Xe, Kr Halogens I, Br I, Br Alkali Metals Cs, Rb Cs, Rb Tellurium Group Te, Sb, Se Te, Sb, Se Barium/Strontium Ba, Sr Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Ru, Rh, Pd, Co Cerium Group Ce, Pu, Np Ce, Pu, Np, Zr

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Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, La, Nd, Eu, Pm, Pr, Sm, Sm, Y, Cm, Am Y, Cm, Am Molybdenum Group Mo, Tc, Nb

5.3 Core Inventories

Core inventories are highly dependent on burnup, initial enrichment, and uranium mass. The release fractions in Table 5-1 and Table 5-2 are based on different burnups, and, for SAND2023-01313, different enrichments. When plant changes result in adverse impacts to the core inventories, these changes are addressed in a licensees EQ program through multipliers, or scaling factors, on the TID-based dose.

Although the differences in core inventory between these source terms are expected to be small relative to the large changes in release fractions, the core inventories for each source term scenario are developed for this assessment.

Consistent with Reg Guide 1.183, Rev. 0, the ORIGEN methodology is applied to develop a specific set of core inventories for each source term. The fuel parameters are tabulated in the following tables.

Table 5-4 Fuel Assembly Characteristics PWR BWR

Plant Surry Grand Gulf Core Power 2587 4408 Number of Fuel Assemblies 157 800 Average Assembly Power (MW/Assy) 16.48 5.51 Uranium Mass (kg/Assy) 456 192

Table 5-5 Fuel Cycle Characteristics TID-14844 Reg Guide Reg Guide SAND 2023-1.183 Rev. 0 1.183 Rev. 1 01313 Bundle Average Enrichment (w/o) 5 5 5 6 Discharge Rod-Average Exposure 40 62 68 80 (GWd/MTU)

The BWR and PWR core inventories of the fission products are listed in Table 5-6 and Table 5-7, respectively. For the PWR inventories that are not fission products (Co-58, Co-60, Pu-238, Pu-239, Pu-240, Pu-241, Np-239, Cm-242, and Cm-244), the standard RADTRAD Curies-per-Megawatt multipliers are applied from Table.4.3.2-2 of NUREG/CR-6604 [15]. For BWRs, the activated corrosion product inventories are applied from Table.4.3.2-2 of NUREG/CR-6604.

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Table 5-6 BWR Core Inventories Reg Guide Reg Guide SAND 2023-Isotope TID-14844 1.183 Rev. 0 1.183 Rev. 1 01313 Am-241 1.753E+04 3.078E+04 3.382E+04 3.735E+04 Ba-137m 1.294E+07 1.969E+07 2.148E+07 2.500E+07 Ba-139 2.220E+08 2.168E+08 2.163E+08 2.158E+08 Ba-140 2.149E+08 2.093E+08 2.083E+08 2.086E+08 Ce-141 2.043E+08 1.987E+08 1.975E+08 1.978E+08 Ce-143 1.930E+08 1.841E+08 1.822E+08 1.825E+08 Ce-144 1.556E+08 1.608E+08 1.604E+08 1.636E+08 Cm-242 3.194E+06 7.687E+06 8.978E+06 9.940E+06 Cm-244 1.487E+05 1.118E+06 1.638E+06 2.026E+06 Cs-134 1.754E+07 3.797E+07 4.400E+07 5.238E+07 Cs-136 6.073E+06 1.043E+07 1.169E+07 1.372E+07 Cs-137 1.367E+07 2.079E+07 2.269E+07 2.641E+07 Cs-138 2.276E+08 2.212E+08 2.204E+08 2.201E+08 I-130 2.599E+06 5.076E+06 5.834E+06 6.444E+06 I-131 1.152E+08 1.182E+08 1.193E+08 1.184E+08 I-132 1.681E+08 1.711E+08 1.735E+08 1.711E+08 I-133 2.441E+08 2.427E+08 2.419E+08 2.419E+08 I-134 2.688E+08 2.655E+08 2.656E+08 2.647E+08 I-135 2.271E+08 2.265E+08 2.270E+08 2.261E+08 Kr-85 1.356E+06 1.864E+06 1.979E+06 2.306E+06 Kr-85m 3.579E+07 3.145E+07 3.049E+07 3.066E+07 Kr-87 6.933E+07 5.984E+07 5.774E+07 5.818E+07 Kr-88 9.808E+07 8.446E+07 8.142E+07 8.203E+07 La-140 2.259E+08 2.261E+08 2.177E+08 2.251E+08 La-141 2.033E+08 1.979E+08 1.970E+08 1.965E+08 La-142 1.979E+08 1.910E+08 1.899E+08 1.895E+08 Mo-99 2.230E+08 2.270E+08 2.286E+08 2.296E+08 Nb-95 2.120E+08 2.013E+08 1.983E+08 1.988E+08 Nb-97 2.060E+08 2.019E+08 2.013E+08 2.009E+08 Nd-147 8.067E+07 7.969E+07 7.961E+07 7.987E+07 Np-239 1.946E+09 2.327E+09 2.421E+09 2.289E+09 Pr-143 1.868E+08 1.777E+08 1.774E+08 1.766E+08 Pr-144 1.563E+08 1.616E+08 1.612E+08 1.644E+08 Pu-238 2.990E+05 9.753E+05 1.216E+06 1.641E+06 Pu-239 4.673E+04 5.497E+04 5.632E+04 5.682E+04 Pu-240 5.211E+04 8.559E+04 9.300E+04 9.834E+04 Pu-241 1.404E+07 2.182E+07 2.375E+07 2.486E+07 Rb-86 1.982E+05 3.555E+05 3.995E+05 4.663E+05 Rh-105 9.507E+07 1.174E+08 1.186E+08 1.212E+08 Rh-106 4.741E+07 7.189E+07 7.830E+07 7.888E+07

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Reg Guide Reg Guide SAND 2023-Isotope TID-14844 1.183 Rev. 0 1.183 Rev. 1 01313 Ru-103 1.637E+08 1.827E+08 1.875E+08 1.861E+08 Ru-105 9.981E+07 1.246E+08 1.310E+08 1.287E+08 Ru-106 4.720E+07 7.157E+07 7.797E+07 7.856E+07 Sb-124 1.018E+05 2.168E+05 2.533E+05 2.772E+05 Sb-125 1.503E+06 2.224E+06 2.407E+06 2.509E+06 Sb-127 1.106E+07 1.264E+07 1.298E+07 1.275E+07 Sb-129 3.496E+07 3.798E+07 3.871E+07 3.810E+07 Sr-89 1.331E+08 1.144E+08 1.097E+08 1.109E+08 Sr-90 1.083E+07 1.512E+07 1.611E+07 1.915E+07 Sr-91 1.634E+08 1.432E+08 1.387E+08 1.396E+08 Sr-92 1.720E+08 1.537E+08 1.497E+08 1.503E+08 Tc-99m 1.953E+08 1.989E+08 2.048E+08 2.011E+08 Te-125m 3.184E+05 4.866E+05 5.294E+05 5.558E+05 Te-127 1.095E+07 1.257E+07 1.335E+07 1.267E+07 Te-127m 1.449E+06 1.690E+06 1.747E+06 1.717E+06 Te-129 3.446E+07 3.746E+07 3.818E+07 3.760E+07 Te-129m 5.130E+06 5.566E+06 5.666E+06 5.597E+06 Te-131 1.024E+08 1.048E+08 1.056E+08 1.048E+08 Te-131m 1.624E+07 1.713E+07 1.727E+07 1.713E+07 Te-132 1.659E+08 1.685E+08 1.693E+08 1.684E+08 Te-133m 9.167E+07 8.678E+07 8.581E+07 8.585E+07 Xe -131m 1.289E+06 1.327E+06 1.335E+06 1.332E+06 Xe -133 2.349E+08 2.325E+08 2.392E+08 2.337E+08 Xe -133m 7.496E+06 7.586E+06 7.595E+06 7.569E+06 Xe -135 9.458E+07 8.492E+07 8.276E+07 8.531E+07 Xe -135m 4.512E+07 4.642E+07 4.679E+07 4.642E+07 Y -90 1.113E+07 1.561E+07 1.667E+07 1.984E+07 Y -91 1.677E+08 1.472E+08 1.421E+08 1.434E+08 Y -92 1.740E+08 1.556E+08 1.516E+08 1.522E+08 Y -93 1.963E+08 1.798E+08 1.762E+08 1.765E+08 Zr -95 2.114E+08 2.003E+08 1.974E+08 1.979E+08 Zr -97 2.042E+08 1.999E+08 1.993E+08 1.990E+08

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Table 5-7 PWR Core Inventories (Curies)

Reg Guide Reg Guide SAND 2023-Isotope TID-14844 1.183 Rev. 0 1.183 Rev. 1 01313 Ba-137m 8.788E+06 1.333E+07 1.453E+07 1.692E+07 Ba-139 1.266E+08 1.218E+08 1.208E+08 1.212E+08 Ba-140 1.227E+08 1.172E+08 1.160E+08 1.165E+08 Ce-141 1.159E+08 1.107E+08 1.095E+08 1.097E+08 Ce-143 1.075E+08 9.964E+07 9.796E+07 9.849E+07 Ce-144 9.626E+07 9.355E+07 9.199E+07 9.219E+07 Cs-134 1.238E+07 2.516E+07 2.888E+07 3.393E+07 Cs-136 4.393E+06 7.362E+06 8.244E+06 1.026E+07 Cs-137 9.259E+06 1.404E+07 1.531E+07 1.781E+07 Cs-138 1.327E+08 1.277E+08 1.267E+08 1.272E+08 I-130 1.212E+06 2.404E+06 2.789E+06 3.133E+06 I-131 6.941E+07 7.066E+07 7.089E+07 7.073E+07 I-132 1.014E+08 1.022E+08 1.024E+08 1.026E+08 I-133 1.428E+08 1.402E+08 1.397E+08 1.401E+08 I-134 1.603E+08 1.558E+08 1.549E+08 1.553E+08 I-135 1.360E+08 1.347E+08 1.345E+08 1.349E+08 Kr-85 9.105E+05 1.208E+06 1.270E+06 1.468E+06 Kr-85m 1.808E+07 1.472E+07 1.398E+07 1.411E+07 Kr-87 3.569E+07 2.846E+07 2.687E+07 2.709E+07 Kr-88 4.772E+07 3.755E+07 3.531E+07 3.560E+07 La-140 1.270E+08 1.236E+08 1.233E+08 1.255E+08 La-141 1.148E+08 1.094E+08 1.083E+08 1.086E+08 La-142 1.106E+08 1.042E+08 1.028E+08 1.032E+08 Mo-99 1.299E+08 1.278E+08 1.274E+08 1.281E+08 Nb-95 1.188E+08 1.081E+08 1.056E+08 1.055E+08 Nb-97 1.195E+08 1.143E+08 1.132E+08 1.138E+08 Nd-147 4.609E+07 4.508E+07 4.491E+07 4.526E+07 Pr-143 1.050E+08 9.788E+07 9.631E+07 9.597E+07 Rb-86 1.471E+05 2.508E+05 2.799E+05 3.341E+05 Rh-105 6.987E+07 8.519E+07 8.858E+07 8.901E+07 Ru-103 1.089E+08 1.228E+08 1.260E+08 1.261E+08 Ru-105 7.476E+07 9.248E+07 9.654E+07 9.666E+07 Ru-106 3.783E+07 5.715E+07 6.184E+07 6.346E+07 Sb-124 0.000E+00 0.000E+00 1.298E+05 1.490E+05 Sb-125 6.432E+05 9.369E+05 1.008E+06 1.079E+06 Sb-127 6.249E+06 6.965E+06 7.130E+06 7.124E+06 Sb-129 1.949E+07 2.140E+07 2.184E+07 2.183E+07 Sr-89 6.659E+07 5.244E+07 4.920E+07 4.938E+07

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Reg Guide Reg Guide SAND 2023-Isotope TID-14844 1.183 Rev. 0 1.183 Rev. 1 01313 Sr-90 7.110E+06 9.668E+06 1.024E+07 1.205E+07 Sr-91 8.412E+07 6.869E+07 6.530E+07 6.580E+07 Sr-92 9.000E+07 7.571E+07 7.256E+07 7.300E+07 Tc-99m 1.151E+08 1.139E+08 1.137E+08 1.146E+08 Te-125m 1.393E+05 2.127E+05 2.311E+05 2.510E+05 Te-127 6.157E+06 6.879E+06 7.043E+06 7.037E+06 Te-127m 1.043E+06 1.184E+06 1.216E+06 1.212E+06 Te-129 1.827E+07 2.010E+07 2.052E+07 2.052E+07 Te-129m 3.509E+06 3.875E+06 3.956E+06 3.949E+06 Te-131 5.897E+07 5.922E+07 5.930E+07 5.944E+07 Te-131m 1.330E+07 1.421E+07 1.442E+07 1.442E+07 Te-132 9.911E+07 9.937E+07 9.946E+07 9.971E+07 Te-133m 6.684E+07 6.434E+07 6.380E+07 6.402E+07 Xe-131m 9.376E+05 9.703E+05 1.006E+06 1.123E+06 Xe-133 1.431E+08 1.407E+08 1.403E+08 1.410E+08 Xe-133m 4.432E+06 4.447E+06 4.455E+06 4.476E+06 Xe-135 4.524E+07 3.789E+07 3.625E+07 3.768E+07 Xe-135m 2.949E+07 3.044E+07 3.068E+07 3.080E+07 Y-90 7.349E+06 1.009E+07 1.071E+07 1.260E+07 Y-91 8.723E+07 7.134E+07 6.770E+07 6.789E+07 Y-92 9.120E+07 7.667E+07 7.347E+07 7.392E+07 Y-93 1.029E+08 8.965E+07 8.673E+07 8.717E+07 Zr-95 1.179E+08 1.073E+08 1.048E+08 1.049E+08 Zr-97 1.187E+08 1.133E+08 1.122E+08 1.129E+08

METHODOLOGY

As described in Section 5 of the Sandia report, a computer code called EQDOSE was written specifically to assess the AST impacts on Environmental Qualification. This study will apply an update to the RAPTOR methodology in application at GGNS. This code is like the NRCs RADTRAD methodology [15] but can also develop integrated inventories. With these values, integrated energy releases can be calculated with the decay energies of each isotope. With gamma spectrum data and the MCNP shielding output, the integrated gamma doses at locations of interest can be assessed.

6.1 Plant Dose Models

The EQDOSE methodology includes some hard-coded models in addition to some plant-specific inputs.

The plant-specific inputs are listed in Tables 5-6, 5-7 and 5-9 of the Sandia report for the airborne gamma and beta dose models. Interestingly, the PWR gamma and beta doses are generated based on different plant-specific assumptions. The gamma dose does not credit containment spray while the beta

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dose credits substantial spray removal. For this evaluation, the airborne source term transient for this work will be based on a single model for each plant.

6.1.1 PWR Airborne Model

The PWR airborne EQ model is illustrated in Figure 6-1 and is nearly identical to that currently applied by the utility. The high degree of mixing between the sprayed and unsprayed volumes will lead to similar radionuclide concentrations in each compartment such that the beta dose will be similar in each volume. The larger open volume of the sprayed region will result in higher gamma doses in this compartment. Thus, both beta and gamma doses will be calculated in the sprayed region.

The PWR sprayed volume is 61 % of the total containment volume (1,819,000 ft3) per Section 14.5 of the Surry UFSAR. Containment sprays are credited to actuate at the onset of the event at 100 seconds and are assumed to run for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the release terminates. A mixing flow of 2 unsprayed volumes per hour is applied. A typical best-estimate spray removal rate of 10 hr-1 is assumed for aerosols terminated 2hours after the release ends. A typical spray removal rate of 2 0 hr-1 is assumed for elemental iodine until a DF of 200 is reached with a residual wall deposition rate of 0.25 hr-1 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . It should be noted that these removal mechanisms are applied to all four source term scenarios so that different values would shift all curves and have only a small effect on this relative evaluation.

6.1.2 BWR Airborne Model

The BWR airborne EQ model is illustrated in Figure 6-1 and, like the PWR model, consists of two containment compartments (sprayed and unsprayed regions) as well as a drywell. The BWR drywell volume is 2.7E5 ft 3 where radionuclide deposition is credited with the Powers 10% removal rates from Reference 15 and an elemental halogen plate-out rate of 0.866 hr-1 as calculated with Reference 10 for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the release ends. The BWR containment volume is 1.4E6 ft 3 with 60% of this volume in the sprayed compartment. In the containment, sprays are credited to actuate at 100 seconds and run for the same duration and same removal rate as the PWR scenario.

The TID scenario assumes source terms are released in all three volumes in proportion to their volume.

In all other cases, the source terms are released into the drywell.

The mixing flow between the drywell and unsprayed containment is modeled at 2000 cfm. For the Reg Guide 1.183, Rev. 0 case, these volumes are assumed to become well -mixed at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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S prayed S prayed CMT CMT

Un-Un-S prayed S prayed Drywell CMT CMT

PWR Dose Model BWR Dose Model

Figure 6-1 Airborne Dose Models

6.1.3 Pool Model

The pool model is based on the guidance in Section 5 of Appendix A to Reg Guide 1.183, Rev. 0. With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Table 5-1 and Table 5-2 of this white paper) are assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. The PWR sump volume is 55,986 ft 3 from the Surry UFSAR Section 14.5.5.2. The BWR suppression pool is based on GGNS with a volume of 4140 m3 (1.462E5 ft3) as applied in Table 5-10 of Reference 8.

6.2 Beta Dose Model

Beta doses are calculated based on the integrated inventory in the sprayed region using an infinite cloud model. The Sandia report calculates the beta doses from the following equation for each isotope, i, as:

= 2.13 6,

where:

= dose rate (rad/hr) from isotope i Ai = activity (Ci) of isotope i Vc = compartment volume (cm3)

, = average beta energy (MeV) per disintegration of isotope i (taken from Table A.1 of FGR-12

[14])

air = density of air (g/cm3)

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This equation can be derived as an infinite cloud model considering that:

1 Curie = 3.7E10 dis/s 1 Rad = 100 erg/g 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months = 3600 seconds 1 MeV = 1.6E -6 ergs.

The integrated volumetric energy release is calculated from the integrated activity for each isotope with the following equation

= 3.7 10 3600 /

= Integrated energy release (MeV/cc) from isotope i

= Integrated activity (Ci-hr) of isotope i

, = average beta energy (MeV/dis) of isotope i

The beta dose can be directly calculated from the volumetric beta energy release by assuming an infinite cloud via the following equation.

1.6E 6 ergs Rad

MeV 100 erg

where:

Di = integrated dose (Rad) from isotope i

= integrated energy release (MeV/cc) of isotope i r = air density (0.001293 g/cc)

6.3 Gamma Dose Model

Gamma doses are assessed for components exposed to the containment atmosphere and to recirculating fluids. These models are described below.

6.3.1 Airborne Model

Sandia applied a conservative spherical geometry to calculate the PWR airborne gamma results per Table 5-6 of the report, while the BWR doses applied a cylindrical model. Instead, in this evaluation, the gamma doses are calculated with a shielding model based on geometry that is representative of a large dome containment. A standardized model assuming a 62-foot diameter based on the GGNS geometry is applied to both the BWR and PWR plants. Multipliers are developed with MCNP5 and applied to the integrated gamma energy releases. The 18-group gamma energy spectrum of each isotope is taken from the ORIGEN 2.1 libraries [13].

The airborne location is taken to be the center of the containment dome with an energy deposition tally (MCNP Type 6). At this location, the open geometry leads to the largest gamma dose. As described in

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Section 5.2, the beta dose is calculated based on an infinite cloud assumption due to the short range of beta particles in air and the open geometry of this location. The density of air is assumed to be 0.001293 g/cc. This model is illustrated in Figure 6-2.

62 ft

dose point

2811

Figure 6-2 Containment Air MCNP Model

6.3.2 Pool Model

The Sandia report calculated the PWR pool doses based on an infinite pool with an activity concentration the same as the calculated pool concentration per Table 5-8 of Reference 8. For BWRs, the pool dose was calculated at the centerline top of a long pipe per Table 5-10 of Reference 8. Although pool activity can result in shine to some in-containment locations, this activity is a significant concern to sensitive components located near ECCS piping outside of containment. As such, a pool shielding model is developed to calculate representative doses from shine from typical BWR/PWR ECCS piping.

Based on GGNS drawings, it is assumed the ECCS piping class is 24-inch Schedule 30 carbon steel piping (r=7.82 g/cc) having a nominal wall thickness of 0.562 inches. The radionuclides are assumed to be homogeneously mixed within the pool water (r=1.0 g/cc) and doses are calculated at a point 1 meter from the pipe centerline of the middle of a 100-foot section of this piping. Only the gamma doses are calculated for this location due to the short range of beta particles in the water and the shielding effects of the piping.

The MCNP pipe model is a 100-foot cylindrical pipe centered at the origin with an ID and OD of 29.053 cm and 30.48 cm, respectively. Photon energies are assumed to be the average within each of the 18 -

energy bins. Doses are tallied with a ring detector at z=0 at a 1-meter radius from the pipe centerline. An energy-dependent multiplier is applied to the gamma flux using the multipliers in Table 11 of ANSI ICRP51 (1987) to generate doses in rad/hr. Backscatter off the walls of the ECCS room is ignored. This model is illustrated in Figure 6-3.

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Air Carbon Steel

100 ft Wa t e r OD ID

Figure 6-3 Piping Dose Model RESULTS

The results of this quantitative evaluation are generally consistent with the original Sandia results.

As presented in Workshop #2 [16], the airborne integrated dose based on TID -14844 continues to bound the airborne gamma and beta doses when changes in core inventories are ignored. As shown in Tables 5-6 and 5-7, the inventories of the long-lived beta-producing isotope of Kr-85 are significantly larger at the higher exposures and consequently increase the integrated dose at the longer durations for the higher exposure cases. When changes in core inventories are considered, the larger inventories of long-lived beta-emitting isotopes in fuel with higher exposures causes the integrated dose from Reg Guide 1.183 and SAND2023 -01313 to exceed the TID dose in the long-term (i.e., after 50 days or longer). However, this impact is already considered in TID-based EQ analyses since any potential changes in core inventories are addressed using scaling factors, as necessary.

The integrated doses from waterborne radiation are not bounded by the TID source term. The larger release fractions of cesium move the cross-over point earlier. For both PWRs and BWRs, doses based on the TID source term were not bounding as early as ~10 hours. This result is attributed to the larger release fractions of alkali metals and tellurium in Reg Guide 1.183 Rev. 1 and SAND2023-01313 since it was also observed when the core inventories are held constant as presented in Workshop #2. As described in the supporting Sandia analyses, these higher release fractions are not due to increased fuel exposure or enrichment but result from severe accident modeling that reflects an improved understanding of accident progression.

The results of the quantitative evaluation are illustrated in the following figures.

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Figure 7-1 BWR Airborne Beta Dose

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Figure 7-2 BWR Airborne Gamma Dose

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Figure 7-3 BWR Waterborne Gamma Dose

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PWR A irborne I nte grate d B e ta Dose 1.E+09

1.E+08

1.E+07

1.E+06

1.E+05 TID-14844 R eg Gui de 1.183 R ev. 0 1.E+04 R eg Gui de 1.183 R ev. 1 SA ND2023-0131 3

1.E+03

1.E+02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 Hour s

Figure 7-4 PWR Airborne Beta Dose

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PWR A irborne I nte grate d Gamma Dose 1.E+07

1.E+06

1.E+05

1.E+04 TID-14844 R eg Gui de 1.183 R ev. 0 R eg Gui de 1.183 R ev. 1 SA ND2023-0131 3 1.E+03

1.E+02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 Hour s

Figure 7-5 PWR Airborne Gamma Dose

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PWR Pool I nte grate d Gamma Dose 1.E+07

1.E+06

1.E+05

1.E+04

1.E+03 TID-14844 R eg Gui de 1.183 R ev. 0 1.E+02 R eg Gui de 1.183 R ev. 1 SA ND2023-0131 3

1.E+01

1.E+00 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 Hour s

Figure 7-6 PWR Waterborne Gamma Dose

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SUMMARY

AND CONCLUSIONS

The continued use of the TID-14844 source term for environmental qualification of electric equipment subject to 10 CFR 50.49 is justified. As demonstrated in this white paper, the airborne EQ radiation profiles developed with TID -14844 source terms continue to bound those developed with the existing revisions of Reg Guide 1.183 and the source term reported in SAND 2023-01313. Although TID-14844 is not bounding for the later periods of the accident for equipment exposed to the sump or suppression pool water, this increase in pool dose is not considered significant from a risk perspective since it is offset by the substantial safety margin (e.g., several orders of magnitude) that exists between the degree of fuel failure assumed by TID-14844 and the design basis requirements for the ECCS stipulated in 10 CFR 50.46 and Appendix K, including an assumed single failure. Even when accounting for the increase in pool dose, the magnitude of this safety margin accounts is sufficient to account for uncertainties associated with accident progression. Crediting the inherent safety margin in the TID -

14844 source term to offset an increase in pool dose is analogous to the approach used to eliminate the hydrogen release from the design basis LOCA under the risk informed revision to 10 CFR 50.44 [68 FR 54123, Sept 16, 2003]. Consistent with the defense-in-depth philosophy, the continued use of the TID-14844 source term maintains a large safety margin between radiological conditions used to establish environmental qualification under 10 CFR 50.49 and the design basis for the ECC systems related to preventing core damage. Overall, this approach would continue to satisfy the requirement in 10 CFR 50.49(e)(4) for the radiation environment used for qualification of electrical equipment must be associated with the most severe design basis accident during or following which the equipment is required to remain functional.

If plant EQ programs were to be modified to another source term, significant updates would be required to the EQ design basis calculations in addition to likely plant modifications. Although the EQ analyses may have been previously updated through the application of simple scaling factors to address changes in core inventories, the adoption of a new source term would require more substantial analytical changes, including applying new computational methods and adding many additional radionuclides. The large increase in doses to components exposed to the pool/sump water would require significant plant modifications in these compartments to either (i) shield these components, (ii) relocate them to lower dose areas, (iii) qualify them to the new doses, or (iv) replace them with qualified components. These increased waterborne doses could also lead to higher doses in the surrounding areas and the potential for these areas to exceed the harsh/mild radiation threshold 3, requiring any credited electrical equipment be added to the EQ program with the potential for similar necessary modifications. These potential adverse impacts are not due to the use of MOX or higher fuel burnups or enrichments but are solely driven by the higher release fractions resulting from changes to the accident models.

The following justification is specific to licensees that choose to modify their licensing basis to higher burnup or enrichment than those specified in RG 1.183 Revision 0 (above the 62GWD/MTU or 5 weight-percent U-235 enrichment).

1. The use of the TID -14844 source terms satisfies the requirements in both 10 CFR 100.11 and 10 CFR 50.67 related to the design basis radiological accidents being assumed to result in

3 This radiation threshold is typically very low as reflected in Section 3.11 of NUREG-0800 (e.g., 1E+03 rad for electronics and 1E+04 rad for electrical equipment).

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substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

2. The difference in accident source terms derived from SAND2011 -0128 and SAND2023-01313 and TID-14844 is primarily from improved understanding of the physics of core meltdown accidents and is not attributed to fuel burnup, use of MOX fuel, or elevated enrichment.

3. As presented in Workshop #2 [16], the accident airborne dose based on TID-14844 continues to bound the airborne gamma and beta doses when changes in core source terms are ignored.

When changes in core inventories are considered, the larger inventories of long-lived beta-emitting isotopes in fuel with higher exposures causes the integrated dose from the SAND2011 -

0128 and SAND2023- 01313 to exceed the TID dose in the long -term (i.e., after 50 days or longer). However, this impact is already considered in TID-based EQ analyses since any potential changes in source term are addressed using scaling factors, as necessary. For equipment inside containment the airborne doses represent the highest contributor to the integrated accident dose used by EQ.

4. The accident pool dose based on TID-14844 is initially higher but is not bounding for the entire design basis survivability period. 4 The pool dose from SAND2023-01313 exceeds the pool dose derived from TID-14844 after approximately 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months for both the PWR and BWR cases. The SAND2023- 01313 source term results in a 30-day integrated accident pool dose that is approximately 5 times higher for both BWRs and PWRs compared to TID -14844. At later times, these differences will increase. This increase in pool dose is offset by the substantial safety margin (e.g., 1 to 2 orders of magnitude) that exists between the degree of fuel failure assumed by TID-14844 and the design basis requirements for the ECCS stipulated in 10 CFR 50.46 and Appendix K, including an assumed single failure. The magnitude of this safety margin accounts for uncertainties associated with accident progression. Accident management and plant recovery actions can also help mitigate the impact of higher pool doses in the longer term.

REFERENCES

1. Technical Information Document 14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 23, 1962.

2. NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, dated February 1995.

3. USNRC Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants, July 2000.

4. USNRC Regulatory Guide 1.183, Revision 1 to Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants, October 2023.

5. SAND2011- 0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, dated January 2011.

6. SAND2023- 01313, High Burnup Fuel Source Term Accident Sequence Analysis, dated April 2023.

7. Regulatory Guide 1.89, Rev. 1, Environmental Qualification of Certain Electric Equipment Important To Safety for Nuclear Power Plants, June 1984.

4 Also referred to as the post-accident operating time.

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8. Letter Report, Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using the Revised Source Term, dated September 28, 1998.

9. NRC Memorandum, Initial Screening of Candidate Generic Issue 187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump, dated April 30, 2001.

10. Section 6.5.2 of NUREG-0800, Revision 4, Standard Review Plan, Containment Spray as a Fission Product Cleanup System, March 2007.

11. Response to Public Comments on Draft Regulatory Guide (DG)-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, (ML23243B014).

12. NUREG/CR-4467, Relative Importance of Individual Elements to Reactor Accident Consequences Assuming Equal Release Fractions, March 1986.

13. ORNL/TM-7175, A Users Manual For the ORIGEN2 Computer Code, July 1980.

14. Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil, dated September 1993.

15. NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, dated December 1997.

16. ML24043A116, Presentation at RG 1.183 Public Workshop (2 of 3), Impact of RG 1.183 R1 and SAND2023- 01313 on Environmental Qualification.

17. NUREG-0458, Short Term Safety Assessment on the Environmental Qualification of Safety -

Related Electrical Equipment on SEP Operating Reactors, May 1978.

18. SECY-19-0079, Staff A pproach to Evaluate Accident Source Terms f or the NuScale Power Design Certification Application dated August 16, 2019.

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APPENDIX A. SELECTION OF ISOTOPES

The important isotopes to the radiological analyses have been developed in NUREG/CR-4467 [22].

Although the scenario in this NUREG is somewhat different than that applied in the alternative source term, the sixty isotopes identified in this report have been implemented in radiological codes like RADTRAD. However, there is no standard guidance on the isotopes important to plant EQ analyses. This report develops a set of isotopes applicable to the GGNS EQ analyses.

The EQ impact of an isotope is primarily dependent on three characteristics:

1. the amount of the isotope that is released from the reactor,

2. the isotopes half-life, and

3. the shielding significance of the isotope and its daughters.

This evaluation applies the PWR core inventories and release fractions developed for the SAND2023 -

01313 case since they release the largest radionuclide concentrations. The ORIGEN evaluation reports inventories for 457 fission product isotopes for this case. The isotopes in the chemical groups with larger release fractions per Table 5-2 (noble gases, halogens, alkali metals, telluriums, and molybdenums) were then selected based on those isotopes with half -lives longer than 60 seconds. This selection process led to a final list of 111 isotopes. Table A-1 lists all the isotopes considered in this evaluation.

The integrated activity for each isotope for the airborne PWR scenario described in Section 5.1.1 and the waterborne scenario in Section 5.1.3 were developed including consideration of progeny production. A typical EQ duration of 180 days was applied in generating the integrated activities. The integrated volumetric total energy release from each isotope was then calculated based on the integrated activities.

The calculated energy releases of the isotopes producing >0.0 5% of the total energy release to the airborne and waterborne scenarios are listed in Tables A-2 and A-3 respectively. The following conclusions are drawn from these tables.

Xe-131m, Xe-133m and Xe-135m are important contributors to the airborne dose

The larger halogen and cesium release fractions increase the importance of I-130 and Cs-138

Ba-137m is a significant beta and gamma contributor as the progeny of Cs-137

The larger release fraction of the tellurium group leads to the inclusion of Te -125m, Te-131, Te-133m, Sb-124, and Sb-125

The large release fraction of the new molybdenum group increases the contribution from Nb-97

Consequently, these 12 isotopes are added to the standard 60 isotopes listed in NUREG/CR-4467 for the EQ evaluations developed in this white paper.

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Table A-1 Isotopes Modeled in Applicability Assessment

Co-58 Sr-91 Tc-102m Te-131m Cs-138 Co-60 Sr-92 Tc-103 Te-132 Cs-138m Se-79m Sr-93 Tc-104 Te-133 Cs-139 Se-81 Sr-94 Tc-105 Te-133m Cs-140 Se-81m Y-90 Rh-105 Te-134 Ba-135m Se-83 Y-91 Ru-103 I-128 Ba-137m Se-83m Y-92 Ru-105 I-129 Ba-139 Se-84 Y-93 Ru-106 I-130 Ba-140 Br-82 Nb-95 Sb-122 I-130m Ba-141 Br-82m Nb-95m Sb-124 I-131 Ba-142 Br-83 Nb-96 Sb-125 I-132 La-140 Br-84 Nb-97 Sb-127 I-132m La-141 Br-84m Nb-97m Sb-128a I-133 La-142 Br-85 Nb-98 Sb-128b I-134 Ce-141 Kr-83m Nb-98m Sb-129 I-135 Ce-143 Kr-85 Nb-99m Sb-130 Xe-131m Ce-144 Kr-85m Zr-95 Sb-130m Xe-133 Pr-143 Kr-87 Zr-97 Sb-131 Xe-133m Nd-147 Kr-88 Mo-99 Sb-132 Xe-135 Pu-238 Kr-89 Mo-101 Sb-132m Xe-135m Pu-239 Rb-86 Mo-102 Sb-133 Xe-137 Pu-240 Rb-88 Mo-103 Te-125m Xe-138 Pu-241 Rb-89 Mo-104 Te-127 Cs-134 Np-239 Rb-90 Tc-99m Te-127m Cs-134m Am-241 Rb-90m Tc-100 Te-129 Cs-135m Cm-242 Sr-89 Tc-101 Te-129m Cs-136 Cm-244 Sr-90 Tc-102 Te-131 Cs-137

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Table A-2 Airborne Dose Contribution by Isotope

Fraction of Fraction of Isotope Total Beta Isotope Total Gamma Energy Release Energy Release Xe-133 59.33% Cs-134 32.35%

Kr -85 26.27% Xe -133 32.28%

Xe -135 3.99% Ba-137m 6.64%

Cs-134 2.12% I-132 5.68%

Cs-137 1.37% Xe -135 5.04%

Xe -131m 1.12% Nb-95 4.51%

Xe -133m 1.09% Kr -88 3.66%

I-132 0.77% I-131 1.83%

I-131 0.57% Cs-136 1.64%

Ba-137m 0.45% I-135 1.16%

Kr -88 0.42% I-133 0.96%

I-133 0.40% Te-132 0.56%

Kr -85m 0.24% I-134 0.47%

Kr -87 0.24% Xe -133m 0.38%

Te-129 0.20% Kr -85 0.34%

Mo-99 0.18% Sb-125 0.32%

I-135 0.17% Xe -131m 0.25%

Nb-95 0.16% Te-131m 0.25%

Te-132 0.15% Kr -85m 0.24%

Te-129m 0.13% Kr -87 0.23%

Te-127 0.10% Sb-127 0.13%

I-134 0.07% Xe -135m 0.13%

Cs-136 0.07% Sb-129 0.12%

Mo-99 0.11%

Nb-97 0.09%

Tc-99m 0.09%

Sb-124 0.08%

Te-133m 0.08%

Cs-138 0.07%

I-130 0.05%

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Table A-3 Waterborne Dose Contribution by Isotope

Fraction of Fraction of Isotope Total Beta Isotope Total Gamma Energy Release Energy Release Cs-134 33.83% Cs-134 61.24%

Cs-137 21.85% Ba-137m 12.60%

I-132 9.44% Nb-95 8.32%

Ba-137m 7.18% I-132 8.30%

I-131 6.89% Cs-136 2.92%

I-133 2.84% I-131 2.62%

Te-129 2.77% Te-132 0.83%

Nb-95 2.50% I-133 0.80%

Mo-99 2.11% Sb-125 0.61%

Te-129m 1.98% I-135 0.49%

Te-132 1.88% Te-131m 0.27%

Te-127 1.51% Sb-127 0.21%

Cs-136 0.98% Sb-124 0.16%

Sb-125 0.74% Mo-99 0.15%

I-135 0.60% Tc-99m 0.12%

Sb-127 0.49% Te-129 0.06%

Te-127m 0.45% Te-129m 0.06%

Sr-89 0.42%

Rb-86 0.22%

Te-131m 0.20%

Te-125m 0.19%

Sb-124 0.18%

Te-131 0.16%

Ba-140 0.15%

Sr-90 0.09%

Tc-99m 0.08%

I-134 0.06%

Sb-129 0.06%

ML24165A085

Text

MeV 100 erg

SUMMARY

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