Text

Rev. 1 Public Comment Resolution (Pre-decisional) ACRS Version
	ML23243B014
Person / Time
Issue date:	08/31/2023
From:	Blumberg W; NRC/NRR/DRA/ARCB
To:	;
References
	DG-1389 RG-1.183, Rev 1
	Download: ML23243B014 (120)
	v • d • e

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Response to Public Comments on Draft Regulatory Guide (DG)-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors On April 21, 2022, the U.S. Nuclear Regulatory Commission (NRC) published a notice in the Federal Register (87 FR 23891) that Draft Regulatory Guide (DG)-1389, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (proposed Revision 1 of Regulatory Guide (RG) 1.183; Agencywide Documents Access and Management System (ADAMS) Accession No. ML21204A065), was available for public comment. The public comment period ended on June 21, 2022.

This document lists each public comment by correspondence (i.e., submittal number). Each comment or comment summary is followed by the NRC staffs response. Each comment is referred to by the submittal number listed below and each comment from the corresponding submittal (e.g., comments 10-1 through 10-18 were submitted by NuScale Power).

Comments on the subject DG are available electronically through the NRCs electronic reading room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can access ADAMS, which provides text and image files of the NRCs public documents. Comments were received from the following individuals or groups:

Submittal ADAMS Commenter Affiliation Commenter Name No. Accession No.

1 ML22152A081 Public/Anonymous Anonymous 2 ML22152A084 Public/Anonymous Anonymous 3 ML22152A086 Public/Anonymous Anonymous 4 ML22152A087 Public/Anonymous Anonymous 5 ML22165A056 Public Han-Chul Kim 6 ML22174A052 BWR Owners Group c/o GE Hitachi Denver Atwood Nuclear Energy 7 ML22174A053 Entergy Corp P. Coutur 8 ML22174A057 PWR Owners Group Michael Powell 9 ML22174A068 Public Brian Magnuson 10 ML22174A071 NuScale Power, LLC Carrie Fosaaen 11 ML22174A072 Nuclear Energy Institute (NEI) Frances A. Pimentel 12 ML22174A074 Public/Anonymous slafountai3 13 ML22174A081 Nuclear Utility Group on Equipment William A. Horin Qualification Winston & Strawn, LLP 14 ML22174A082 Public/Anonymous Anonymous 15 ML22174A083 Public/Anonymous Anonymous 16 ML22174A084 Public/Anonymous Anonymous 17 ML22174A085 Public/Anonymous Anonymous 18 ML22180A115 Dominion Energy Services, Inc. B.E. Standley 19 ML23060A242 Public Brian Magnuson 1

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Submission 1 ADAMS Accession No. ML22152A081 Name: Anonymous Email: cliffnukem@gmail.com 1-1 Anonymous Comment The defined composite worst-case approach introduces unnecessary conservatism. Applicants should be able to calculate burnup-dependent sets of radionuclide releases which more accurately reflect the physics. For example, an applicant should be allowed to postulate multiple FHA scenarios involving an incident with a 1st, 2nd, and 3rd burned fuel assembly. Each FHA scenario would employ a set of radionuclide release fractions representative of their burnup interval. This approach removes the non-physical, overly conservative aspects of assuming a single fuel assembly has a beginning-of-life Iodine-131 release fraction and end-of-life Krypton-85 release fraction.

Comment Response The NRC staff agrees with this comment. RGs describe one or more methods that the staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. Employing burnup-dependent release fractions may be acceptable for non-loss-of-coolant accident (non-LOCA) design-basis accidents (DBAs). There are several different ways that a licensee can employ burnup-dependent release fractions. If a licensee elects to employ such an approach, the method used will be considered on a case-by-case basis. In response to this comment, the NRC staff added the following sentences to RG 1.183, Revision 1, Appendix I, Analytical Technique for Calculating Fuel-Design or Plant-Specific Steady-State Fission Product Release Fractions for Non-Loss-of-Coolant Accident Events (previously labeled appendix J to DG-1389), to explain that alternative, more realistic methods that are not detailed in the RG, such as calculating and employing burnup-dependent release fractions, will be considered on a case-by-case basis:

The analytical technique outlined in this section is one acceptable means of calculating maximum steady state release fractions.

and One means of capturing more realism in the calculation of the steady state release fractions would be to calculate burnup-dependent release fractions for each radionuclide. The use of such means will be considered on a case-by-case basis.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 1-2 Anonymous Comment The analytical procedure appears to model an imaginary worst single rod which operated on a bounding rod power history. Licensees use high fidelity lattice physics and reload depletion models capable of predicting individual fuel rod power for the upcoming cycle. Together with the known power operating history from past reload cycles, the complete power history for each fuel rod is understood. a. Applicants should be able to calculate burnup-dependent release fractions for each fuel rod predicted to experience cladding failure during each DBA. b.

Applicants should be able to use the actual fuel rod power history and peaking factor to calculate the quantity of radionuclides (i.e., moles of gas) for each fuel rod predicted to experience cladding failure during each DBA. c. Applicants should be able to combine these individual releases to determine the RCS coolant activity (or release during a FHA outside reactor).

Comment Response The NRC staff partially agrees with this comment. The staff agrees that more explicit calculations when determining the quantity of radionuclides in each rod and the release of radionuclides to the RCS during non-LOCA DBAs may be acceptable. For example, the licensee may elect to use the burnup, power history, and peaking factor from the accident analyses (e.g., chapter 14 or 15) in the updated final safety analysis report (FSAR) for each rod predicted to fail.

The NRC staff disagrees that calculations based on actual power histories should be endorsed in RG 1.183, Revision 1, except on a case-by-case basis. Although licensees have the option to use cycle-specific calculations, doing so could be resource intensive for the NRC staff as well as the licensee. If actual power histories are used, this could result in the need to reanalyze DBA doses and NRC staff reviews of these changes on a cycle-specific or near cycle-specific basis.

Rather, traditionally, design-basis calculations bound cycle-to-cycle operation. Therefore, the NRC staff will consider such calculations on a case-by-case basis.

Regarding the calculation of burnup-dependent release fractions for non-LOCA DBAs, please see the NRC staffs response to Public Comment 1-1.

In response to this comment, the NRC staff added the following sentence to RG 1.183, Revision 1, Regulatory Position 3.1:

For non LOCA DBAs, the NRC staff will consider on a case by case basis the use of more explicit methods to calculate the fission product inventory and reactor coolant system activity, such as using the burnup, power history, and peaking factor from the accident analyses (e.g., chapter 14 or 15) in the updated FSAR, for each rod predicted to fail.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 1-3 Anonymous Comment Many of the conservative analytical techniques of combining worst-case parameters to achieve a composite worst-case result date back to a time when computers were limited and expensive.

Modern computation methods allow more rigorous engineering. The composite worst-case fuel rod is overly conservative. And this approach become even more non-physical when the DBA involves larger quantities of fuel rods where the assumption is that every failed fuel rod is the worst fuel rod operated at the TS/COLR peaking factor. Even for a FHA involving damage to a single fuel assembly, there is a distribution of power histories and thus differences in radionuclide quantities and release fractions. For DBAs involving larger populations of damaged fuel rods (e.g., PWR RCP locked rotor), this simplified approach is even more non-physical and introduces unnecessary conservatism. a. Given the capability to predict both burnup-dependent release fractions and quantities for each radionuclide for each failed fuel rod, applicants should be able to employ more explicit predictions (e.g., weighted average release) when determining RCS activity (or release during a FHA outside reactor).b. Knowledge of burnup-dependent fuel rod power distributions throughout the core are already part of the plants licensing basis and is used to predict the number of rods which experience boiling crisis (e.g., DNB) during DBAs.

Applicants should be able to use this same information to predict the releases from those same failed fuel rods.

Comment Response The NRC staff agrees with this comment. RGs describe one or more methods that the staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. The NRC staff revised RG 1.183, Revision 1, to state that more explicit methods for calculating RCS activity, such as using the burnup, power history, and peaking factor for each rod predicted to fail, will be considered on a case-by-case basis.

Specifically, the staff added the following sentence to Regulatory Position 3.1 of RG 1.183, Revision 1:

For non LOCA DBAs, the NRC staff will consider on a case by case basis the use of more explicit methods to calculate the fission product inventory and reactor coolant system activity, such as using the burnup, power history, and peaking factor from the accident analyses (e.g., chapter 14 or 15) in the updated FSAR, for each rod predicted to fail.

Please also see the response to Public Comment 1-2.

1-4 Anonymous Comment Section J-3 of the analytical procedure describes the application of uncertainties to achieve an upper tolerance value. a. Section J-3.2 states that the model uncertainties may be applied either deterministically or statistically when calculating long-lived radionuclide release fractions. I 4

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 agree that the applicant should have this option, especially when considering releases from multiple fuel rods. b. Section J-3.1 identifies that the 2011 ANS-5.4 standard recommends a deterministic multiplier of 5.0 on the best-estimate predictions for short-lived isotopes. It seems odd that NRC has defined such a detailed analytical procedure with all of these attributes, yet the results of these explicit calculations are multiplied by a factor of 5.0. The NRC should define an uncertainty distribution (e.g., mean, std. dev.) for this term so that it may also be sampled.

Comment Response Please note that the regulatory positions discussed in this public comment and noted as a Section in appendix J to DG-1389 are now located in appendix I to RG 1.183, Revision 1.

The NRC staff partially agrees with this comment. The staff agrees with the statement that the uncertainties can be applied either deterministically or statistically when calculating long-lived radionuclide release fractions.

The NRC staff disagrees with defining an uncertainty distribution for the uncertainty multiplier of 5.0. Notably, the 2011 American National Standards Institute/American Nuclear Society (ANSI/ANS) Standard 5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, recommends the uncertainty multiplier of 5.0 be applied to short-lived radionuclide release fractions. Deviation from this standard would require proper justification, and the staff is not aware of such technical justification. The 5.0 uncertainty factor is based on the standard deviation of the model prediction of the krypton-85m release fraction, as described in more detail in NUREG/CR-7003, Background and Derivation of ANS-5.4 Standard Fission Product Release Model, issued January 2010 (ML100130186). Appendix I to RG 1.183, Revision 1, provides one acceptable means of calculating steady-state release fractions. Any proposed deviations, such as a different procedure for calculating uncertainties for short-lived radionuclide release fractions, will be reviewed on a case-by-case basis. No changes were made in response to this comment.

Comment Submission 2 ADAMS Accession No. ML22152A084 Name: Anonymous Email: cliffnukem@gmail.com 2-1 Anonymous Comment Section 3.2 of DG-1389, Release Fractions, provides MHA LOCA core average release fractions (Tables 1 and 2) described as hybridized accident source terms from SAND-2011-0128. The hybridized source terms were derived from MELCOR simulations based on a defined set of accident sequences for a few typical NSSS designs. The range of applicability for these source terms needs to be more detailed to ensure appropriate future application. Are there changes in reactor power, power density, NSSS design, fuel design, ESFAS capabilities, etc.

which would challenge the applicability of these tables? Is the applicability limited to existing fleet of reactors? What about certified designs (e.g., AP1000, EPR, ESBWR)?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response Please note that the reference to Section 3.2 of DG-1389 in the comment is referring to Regulatory Position 3.2.

The NRC staff agrees with this comment and has added clarification to the guidance based on the comment. These source terms were derived from an examination of a set of accident sequences for current light-water reactor (LWR) designs and reflect the current understanding of severe accidents and fission product behavior since the publication of NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, issued February 1995 (ML041040063). Regulatory Position 3.2 of DG-1389 provided ranges of applicability for the use of these source terms. However, the NRC staff is adding clarification in section A and Regulatory Position 3.2 of RG 1.183, Revision 1, in response to this comment.

The staff added the following sentences to section A:

The updated source terms reflect the current understanding of severe accidents and fission product behavior since the publication of RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, issued July 2000 (Ref. 2), and include low burnup and high burnup LWR fuels. The use of these source terms is not endorsed for mixed oxide fuels.

The staff revised Regulatory Position 3.2, in part, to state the following:

The tables 1 and 2 source terms were derived by examining of a set of accident sequences for current LWR designs; they reflect the current understanding of severe accidents and fission product behavior since the publication of NUREG-1465.

2-2 Anonymous Comment The range of applicability includes chromium-coated cladding and chromia-doped fuel. I would imagine that differences in the accident progression would be introduced by increasing the chromium coating thickness from 8 microns to 100 microns. Similarly, fission product releases may be impacted, especially gap release phase, by increasing dopant concentrations well beyond solubility. Are ranges of applicability for these fuel design features appropriate?

Comment Response The NRC staff agrees that dopant concentration can impact fission gas release (FGR) (and thus non-LOCA and LOCA gap fractions). The impact of doping standard uranium dioxide is expected to be small compared to existing uncertainties in FGR behavior. It is also expected that, generally, doped fuel releases less fission gas under normal operating conditions and that gap fractions derived based on standard uranium dioxide are likely conservative for doped fuel.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 However, the 0.16 weight-percent doping is provided as design applicability based on the review performed in NUREG/CR-7282, Review of Accident Tolerant Fuel Concepts with Implications to Severe Accident Progression and Radiological Releases, issued July 2021 (ML21210A321). The NRC staff revised the applicability in Regulatory Position 3.2 to include a design applicability for chromia-doped uranium dioxide fuel of up to 0.16 weight-percent.

Reference 25, Applicability of Source Term for Accident Tolerant Fuel, High Burnup and Extended Enrichment, of RG 1.183, Revision 1, provides an upper limit on the chromium-coated cladding thickness evaluated:

However, the proposed coating layer thicknesses are less than 50 microns, which is a small fraction of the overall cladding thickness. Thus, the small increase in chromium that would be present in a molten corium pool is not expected to have a significant impact on the in-vessel release fractions given in Tables 10 and 11 of SAND2011-0128.

Based on the review performed in Reference 25, the NRC staff revised the applicability in RG 1.183, Revision 1, Regulatory Position 3.2, in response to this comment, to include a design applicability for chromium-coated cladding less than 50 microns.

2-3 Anonymous Comment Based on SAND-2011-0128, it does not appear that the hybridized source terms represent the composite worst-case releases (i.e., highest predicted release for each radionuclide from worst accident sequence in worst NSSS design). How then can it be determined that Table 1 is conservative for every BWR and Table 2 for every PWR?

Comment Response The NRC staff does not agree with this comment. The source terms from SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, issued January 2011 (ML20093F003), are based on the methods originally used in NUREG-1465, which provided the bases for RG 1.183, Revision 0, table 1 (PWR) and table 2 (BWR). They were independently peer-reviewed in ERI/NRC 11-211, Peer Review of Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup and Mixed Oxide Fuels, issued December 2011 (ML12005A043). In that peer review, the revised source terms used in tables 1 and 2 were found to be technically justified and appropriate, and there were ample justifications for revising the source term in NUREG-1465 for high-burnup fuel. Based on SAND2011-0128, the independent peer assessment results, and the use of previous methods to derive the table 1 and 2 values, the NRC staff determined that the source terms in tables 1 and 2 of DG-1389 represent a conservative source term for BWRs and PWRs for the purposes of the design-basis calculations. No changes were made in response to this comment.

2-4 Anonymous 7

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Based on SAND-2011-0128, differences between low- and high-burnup results do not appear significant. It appears that the main driver for differences between NUREG-1465 and SAND-2011-0128 is likely code-to-code differences in the MELCOR simulations. If this is true, then why is DG-1389 (future RG 1.183 Rev.01) being advertised as a high burnup source term? In light of the differences, maintaining RG 1.183 Rev.00 is confusing and needs to be justified. Will future applications, including the existing fleet, at a burnup limit of 62 GWd/MTU be able to choose between Rev.00 and Rev.01 source terms? If so, why?

Comment Response The NRC staff does not agree with this comment. The method proposed in DG-1389 is applicable to higher burnups than those in Revision 0 of RG 1.183. Both Revision 0 and Revision 1 of RG 1.183 will be available for use, because Revision 0 will not be withdrawn.

Each revision provides a method acceptable to the staff for compliance with the applicable regulations specified in the guidance. Use of combinations and permutations of regulatory positions from Revision 0 and Revision 1 will need appropriate technical justification and may require additional NRC review before approval. Notably, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

Comment Submission 3 ADAMS Accession No. ML22152A086 Name: Anonymous Email: cliffnukem@gmail.com 3-1 Anonymous Comment Table 7 of DG-1389 defines dose criteria for the various postulated DBAs. For next generation reactors, it appears that the NRC is considering a more rigorous probability-consequence correlation. In this guidance, the dose criteria, and presumably the underlying event probabilities, have not been updated since at least 2000 (RG 1.183 issued in 2000). The bases for Table 7, including event-specific probabilities, should be published and referenced in this guidance.

Comment Response The NRC staff does not agree with this comment. The dose criteria in table 7, Accident Dose Criteria for exclusion area boundary (EAB), low-population zone (LPZ), and control room locations were not revised as part of Revision 1 to RG 1.183. However, these criteria were originally provided in table 6 in RG 1.183, Revision 0, and now table 7 in RG 1.183, Revision 1, also provides the same criteria in two different units of dose (rem total effective dose equivalent (TEDE) and sievert TEDE). Accordingly, the basis for table 7 remains consistent with precedent 8

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 for considering the radiological consequences of these accidents (e.g., see table 6 of RG 1.183, Revision 0). Moreover, the NRC staff is not withdrawing RG 1.183, Revision 0, as it remains acceptable for use to meet the pertinent NRC regulations. No changes were made in response to this comment.

3-2 Anonymous Comment Please explain why the allowable consequences are identical for the MHA LOCA, BWR MSLB, PWR MSLB, and PWR SGTR. MHA LOCA involves very low probability initiating events along with delays/failures of multiple safety-related systems. I have difficulty believing that this beyond design basis event has the same probability of occurrence as a PWR SGTR. Note that the existing fleet has experienced multiple SGTR events.

Comment Response The dose criteria in table 7, Accident Dose Criteria for EAB, LPZ, and Control Room Locations, were not revised as part of Revision 1 to RG 1.183. However, these criteria were originally provided in table 6 in RG 1.183, Revision 0, and now table 7 of RG 1.183, Revision 1, also provides the same criteria in two different units of dose (TEDE and sievert TEDE). The EAB and LPZ accident dose criteria in table 7 for the MHA LOCA, and the fuel damage or pre-accident spike scenarios for the BWR MSLB, PWR MSLB, and PWR (SGTR) accidents, are based on what has been done historically to consider the expected frequency of occurrence for these events. No changes were made in response to this comment.

3-3 Anonymous Comment Please explain why the allowable consequences are identical for the BWR CRD, PWR CRE, and FHA. Note that the existing fleet routinely experiences fuel handling issues. CRD involves channel distortion and blade interference which is also quite common. And CRE involves instantaneous rupture of the reactor head (which would seem to be of similar probability as a piping break).

Comment Response The NRC staff agrees with the request to provide an explanation. The dose criteria in table 7, Accident Dose Criteria for EAB, LPZ, and Control Room Locations, were not revised as part of Revision 1 to RG 1.183. However, these criteria were originally provided in table 6 in RG 1.183, Revision 0, and now table 7 of RG 1.183, Revision 1, also provides the same criteria in two different units of dose (TEDE and sievert TEDE). The EAB and LPZ accident dose criteria in table 7 for the BWR control rod drop (CRD), PWR control rod ejection (CRE), and fuel handling 9

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 accident (FHA) are based on previous ways of considering the expected frequency of occurrence for these events. No changes were made in response to this comment.

3-4 Anonymous Comment The dose criteria for PWR Locker Rotor, 6.3 rem, seems reasonable given it involves failure of an active component.

Comment Response The NRC staff agrees that the dose criteria of 6.3 rem TEDE is reasonable. No changes were made in response to this comment.

Comment Submission 4 ADAMS Accession No. ML22152A087 Name: Anonymous Email: cliffnukem@gmail.com 4-1 Anonymous Comment Table 7 of DG-1389 includes dose criteria for all postulated DBAs, except design basis LOCA.

Why is there no design basis LOCA dose assessment and separate criteria? The MHA LOCA is a beyond design basis event and is a useful metric for reactor siting. However, given likely differences in event probabilities, the MHA LOCA should not be used to bound the design basis LOCA. DBAs are not allowed to digress to severe accidents involving core melt.

Comment Response The NRC staff disagrees with this comment. Table 7 provides the accident dose criteria for several DBAs. These criteria were not revised as part of Revision 1 to RG 1.183. However, in RG 1.183, Revision 0, these criteria were provided in table 6, and now table 7 of RG 1.183, Revision 1, also provides the same criteria in two different units of dose (TEDE and sievert TEDE). Traditionally, the dose acceptance criteria assigned for a double-ended rupture of the largest pipe in the RCS have been those in Title 10 of the Code of Federal Regulations (10 CFR) 50.67, Accident source term. However, since the requirements in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, ensure that the maximum fuel element cladding temperature is limited such that fuel melt would be precluded during a LOCA, the dose consequences from the design-basis MHA LOCA, which has substantial fuel melt, would exceed the dose consequences of what the comment defines as the design-basis LOCA event described in 10 CFR 50.46, as well as the guidance in Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping 10

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Breaks Within the Reactor Coolant Pressure Boundary, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.

The use of the term MHA LOCA in DG-1389 is intended to clarify an accident that is acceptable to meet the requirements in 10 CFR 50.67; that is, a hypothetical accident that is assumed to result in a substantial meltdown of the core with subsequent release of appreciable quantities of fission products into the containment (see, e.g., 10 CFR 50.34, Contents of applications; technical information, footnote 6; and 10 CFR 50.67, footnote 1, as well as similar footnotes in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants). There is no specific cause of the major accident hypothesized as described in footnote 1 of 10 CFR 50.67. The accident described in the regulation is a DBA. This accident description was used for siting in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance, then added to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR Part 52, specifying the radiological design basis for nuclear power plants. No changes were made in response to this comment.

4-2 Anonymous Comment The design basis LOCA is the design basis for many Technical Specification Limiting Conditions of Operation (LCO) and Limiting Safety System Settings (LSSS) parameters. How can changes to these TSs be assessed with no consideration of radiological consequences?

Comment Response The NRC staff disagrees with this comment. The design-basis LOCA and MHA LOCA accidents are intended for different purposes. The design-basis LOCA is intended, for example, to meet the requirements of 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. The MHA LOCA is a hypothetical accident that is acceptable to use to meet other regulations, such as 10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," and 10 CFR 50.67, "Accident Source Term. For RG 1.183, Revision 1, the MHA LOCA source terms were derived from an examination of a set of accident sequences for current LWR designs and reflect the current understanding of severe accidents and fission product behavior since the publication of NUREG-1465. See also the NRC staffs response to Public Comment 4-1. No changes were made in response to this comment.

4-3 Anonymous Comment The design basis LOCA is the design basis for many safety-related SSCs (e.g., reactor fuel, ECCS, containment, containment isolation, containment sprays). Many of these SSCs are designed to limit core damage during and following a LOCA. Yet, the degree of core damage 11

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 (i.e., number of fuel rod failures, releases from failed fuel rods) is not predicted nor judged against regulatory criteria. For all other DBAs, the performance of safety-related SSC is judged against core damage and unique radiological limits. How can changes to these SSCs be assessed with no consideration of radiological consequences?

Comment Response The NRC staff disagrees with this comment. The NRC staff does consider radiological consequences associated with changes to structures, systems, and components (SSCs). The regulation in 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, specifies the criteria that must be demonstrated to ensure that the emergency core cooling system (ECCS) is designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria in 10 CFR 50.46(b). The dose acceptance guidance for 10 CFR 50.46, is contained in Standard Review Plan, section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary.

Since the requirements in 10 CFR 50.46 ensure that the maximum fuel element cladding temperature is limited such that fuel melt would be precluded during a LOCA, the dose consequences from the design-basis MHA LOCA, which has substantial fuel melt, would exceed the dose consequences of what the comment defines as the design-basis LOCA event described in 10 CFR 50.46, as well as the guidance in NUREG-0800, section 15.6.5. No changes were made in response to this comment.

4-4 Anonymous Comment Other countries (e.g., Germany) have established dose criteria for design basis SBLOCA and LBLOCA, recognizing differences in event probability (break size) and differences from MHA LOCA. Why has the NRC not adopted this same regulatory position?

Comment Response This comment is out of scope of the proposed revisions to RG 1.183, Revision 1. Notably, RGs are not regulatory requirements. Instead, RGs describe one or more methods that the NRC staff considers acceptable for meeting the NRCs regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 4-5 Anonymous Comment Below are three examples illustrating how the lack of a design basis LOCA dose criteria has resulted in unintentional consequences.

a. Several of the advanced reactor designs (e.g., AP1000) rely upon passive systems to provide emergency coolant and reflood/refill the reactor following a major pipe break. For smaller pipe breaks, the plants are designed to quickly depressurize the RCS which allows the flow of emergency coolant into the core. However, rapid RCS depressurization increases the stress on the cladding due to differences between internal rod pressure and RCS pressure. This increased stress will likely result in a larger number of fuel rod failures (balloon/burst). Hence, the act of depressurizing the RCS increases the radiological source term and the dose to the public. Had a design basis LOCA limit been established, this problem would have been recognized.

b. Fuel designs and fuel utilization has evolved significantly over the past few decades. Several of these changes would promote a larger population fuel rod failures (balloon/burst) during a LOCA. For example, fuel rod cladding wall thickness has been reduced and cladding materials are weaker (e.g., SRA Zry-4 has a higher yield strength than RXA M5). Rod internal pressure is allowed to exceed system pressure (e.g., PWR operating system pressure of 2250 psi). Power uprates and higher discharge fuel burnup. Had a design basis LOCA limit been established, these changes would have been properly assessed against their impact to public dose. Since MHA LOCA dose assessment involves a substantial core melt, it is much less sensitive to fuel design.

c. As described in Section 3.2 of DG-1389, the potential impact of FFRD on the MHA LOCA source term was evaluations and found to be insignificant. However, this same conclusion would not be possible for a design basis LOCA. Research Information Letter (RIL) 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, describes significant quantities of transient fission gas release produces as a result of FFRD.

This additional source of radionuclides would likely increase public dose under design basis conditions. Furthermore, future regulatory decisions regarding FFRD would also likely be steered in a different direction in the presence of a design basis LOCA dose assessment.

Comment Response Please note that the reference to Section 3.2 of DG-1389 in the comment is referring to Regulatory Position 3.2.

The NRC staff disagrees with this comment. All three examples in this comment suggest that the lack of a design basis LOCA dose criterion has resulted in unacceptable consequences and that future regulatory decisions regarding fuel fragmentation, relocation, and dispersal (on FFRD) would likely be steered in a different direction in the presence of a design -basis LOCA dose assessment. This comment is out of scope of the revision to RG 1.183, Revision 1, because changes to 10 CFR 50.46 (which appear to be the focus of this comment as it discusses the design basis LOCA criterion) are not part of the revisions to RG 1.183, Revision

1. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 4-6 Anonymous Comment Design basis LOCA would appear to have the same probability of occurrence as a PWR CRE since both involve failure of robust, passive structures. Yet, PWR CRE is only allowed well within (i.e., 25%) of the upper dose limit. Please explain these differences.

Comment Response With regard to the differences between the accident dose criteria for the pressurized water reactor (PWR) control rod ejection (CRE) and the maximum hypothetical accident loss-of-coolant accident (MHA LOCA), the acceptance criteria in table 7 remain consistent with precedent for considering the radiological consequences for these accidents (e.g., see table 6 of RG 1.183, Revision 0). To note, these criteria were originally provided in table 6 of RG 1.183, Revision 0, and now table 7 of RG 1.183, Revision 1, also provides the same criteria in two different units of dose (rem TEDE and sievert TEDE). The exclusion area boundary (EAB) and low population zone (LPZ) accident dose criteria in table 7 for the PWR CRE are fractions (25 percent) of the criteria for the MHA LOCA dose reference values based on what has been done historically to consider the expected higher frequency of occurrence for the PWR CRE than for the MHA LOCA. No changes were made in response to this comment.

Comment Submission 5 ADAMS Accession No. ML22165A056 Name: Han-Chul Kim Email: k250khc@gmail.com 5-1 Kim Comment In section 3.3. Timing of Release Phases, there is a sentence: "Regardless of delays in the onset, the duration of the gap release phase is 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . In my understanding of the paragraph, "0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months " should be replaced by "0.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months ," which is the revised duration of the gap release phase.

Comment Response Please note that the reference to section 3.3 of DG-1389 in the comment is referring to Regulatory Position 3.3.

The NRC staff agrees with this comment. The NRC staff deleted the subject sentence in Regulatory Position 3.3 (Regardless of delays in the onset, the duration of the gap release phase is 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .) in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Submission 6 ADAMS Accession No. ML22174A052 Name: Denver Atwood, BWROG Chairman Email: dgatwood@southernco.com Address: c/o GE Hitachi Nuclear Energy, P.O. Box 780, 3901 Castle Hayne Road, M/C A-55, Wilmington, NC 28402 USA 6-1 BWR Owners Group Comment A comparison of the changes in maximum hypothetical accident (MHA) release fractions between BWRs (Table 1) and PWRs (Table 2) identified a significant increase in the BWR halogen release fractions with no indication in either SAND2011-0128 or DG-1389 as to the cause. This increase will adversely affect the ability of BWRs to comply with Reg Guide 1.183 Rev. 1. A comparison to the PWR analyses suggests that the accident sequences may be responsible for this impact.

The original alternative source term in NUREG-1465 was based on a range of accident sequences. As described in Section 3.1 of NUREG-1465, these sequences were based on the accident sequence data in NUREG-1150, which assessed the core damage risks for five representative plants using the PRA methods and severe accident modeling tools available in the late 1980s. The dominant sequences were evaluated with the Staffs Source Term Code Package (STCP) and the MELCOR code and are listed in Table 3.1 of NUREG-1465. An updated version of the core damage risk to the entire fleet is NUREG-1560, which was based on industry- prepared Individual Plant Evaluations (IPEs). As shown in Figure 3.6 of NUREG-1560 (reproduced below) for the majority of operating BWR plants (BWR/3 and BWR/4), there continues to be a broad diversity in the accident scenarios leading to core damage.

SAND2011-0128 updates NUREG-1465 with higher core exposures utilizing the latest NRCs MELCOR methodology. The accident sequences that were analyzed to develop the PWR release fractions are listed in Table 5 of the Sandia report and include a variety of accident types. However, for the BWR release fractions, Table 3 of the Sandia report indicates that nearly all of the evaluations were based on station blackout (SBO) sequences. As illustrated in Figure 3.6 of NUREG-1560, the core damage risk from SBO is similar to that from other accident sequences. Considering the recent modifications to address Fukushima SBO risk, the core damage risk from SBO is considerably lower than the NUREG-1560 data taken from the late 1990s.

Section A-2.1 of DG-1389 characterizes these release fractions as airborne releases into the BWR drywell. It is also unclear from the Sandia report whether suppression pool scrubbing was credited in determining the release fractions from a BWR SBO. Credit for pool scrubbing can significantly decrease the airborne activity since the SBO-related releases would be released via spargers underwater in the suppression pool.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Consistent with the PWR analysis in SAND2011-0128, the BWR release fractions should be re-evaluated based on accident sequences that are a more representative of BWR core damage risks. Any sequences that involve releases through the pool spargers should take credit for suppression pool scrubbing.

Comment Response Please note that the reference to Section A-2.1 of DG-1389 in the comment is referring to Regulatory Position A-2.1.

The NRC staff does not agree with the comment. Section 4.1.2, The In-Vessel Release Phase, of SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, issued January 2011 (ML20093F003), discusses some of the reasons for an increase in halogen release fractions. In addition, a revision to SAND2011-0128 is not within the scope of this revision of RG 1.183; therefore, the staff made no revisions to source terms in tables 1 and 2 in response to this public comment. Regarding credit for suppression pool scrubbing, RG 1.183, Revision 1, Regulatory Position A-2.5, covers the staffs position on suppression pool scrubbing, which is further discussed in the staffs response to Public Comment 11-28.

6-2 BWR Owners Group Comment The Clifford memo entitled TECHNICAL BASIS FOR DRAFT RG 1.183 REVISION 1 (2021)

NON- LOCA FISSION PRODUCT RELEASE FRACTIONS (ADAMS Accession No. ML21209A524) states that the BWR gap fractions are based on a BWR 10x10 fuel design; however, the design specifics or model name is not provided. Since there are at least 6 different BWR 10x10 designs, clarification is needed to ensure which designs are addressed by the listed DG-1389 non-LOCA release fractions. The listed release fractions may also be applicable to more advanced fuel designs such as 11x11 designs.

Commenters Proposed Resolution State the BWR 10x10 fuel design applied to develop the BWR release fractions in Table 3.

Indicate whether the results are applicable to BWR 11x11 designs that meet the power history inputs in Figure 1.

Comment Response The NRC staff agrees with this comment. A representative General Electric 10x10 design was chosen that is more limiting than the remaining 10x10 and 11x11 fuel designs that are currently approved. Therefore, table 3 will apply to all currently approved BWR 10x10 and 11x11 fuel designs, upon issuance of the final RG 1.183, Revision 1.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 In response to this comment, the NRC staff incorporated changes into Regulatory Position 3.2 of RG 1.183, Revision 1, to clarify the applicability by adding the words (as of the issuance of this RG) to clarify the words is limited to currently approved designs.

6-3 BWR Owners Group Comment The BWR part-length rods tend to experience somewhat more aggressive power profiles than the full-length rods. Did the NRC analysis that developed the release fractions Table 3 apply the reported BWR power profile to the PLRs or was a more aggressive history assumed?

Commenters Proposed Resolution Confirm applicability of Figure 1 power history to BWR PLRs or provide applied power history Comment Response The NRC staff agrees that BWR part-length rods typically have a more aggressive power history than full-length rods due to being located at the bottom of the core, where there is generally a higher neutron flux. Therefore, BWR part-length rods may have a higher rod average power than adjacent full-length rods of the same enrichment. As a result, BWR part-length rods may not be able to meet the rod average power envelope in figure 1, Maximum Allowable Power Operating Envelope for Steady State Release Fractions. In response to this comment, the NRC staff revised Regulatory Position 3.2 in RG 1.183, Revision 1, to now state the following:

If BWR part-length fuel rods are treated as full-length fuel rods with respect to overall quantity of fission products and are operated below the burnup-dependent BWR peak power envelope in figure 1, then table 3 steady-state fission product release fractions apply to these part-length fuel rod designs.

Also, in response to this comment, the NRC staff revised figure 1 to include peak linear heat generation rate (LHGR) envelopes.

In other words, the table 3 release fractions can be applied to BWR part-length rods if the peak LHGR of the part-length rod adheres to the peak power envelope curve of figure 1 and the part-length rods are treated as full-length rods with respect to the overall quantity of fission products.

6-4 BWR Owners Group Comment BWR fuel designs are increasingly utilizing part-length fuel rods (PLRs) to optimize fuel exposure and shutdown margin. As the number of PLRs increases, the DG-1389 approach of 17

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 treating PLRs as full-length rods tends to significantly over-estimate the bundle source term.

The DG-1389 approach to the PLR release fraction is just one approach to considering the PLR gap release. Another approach is to apply the guidance in DG-1389 Appendix J to calculate the actual release fraction in the PLRs.

Commenters Proposed Resolution Acknowledge that the DG-1389 approach to BWR part-length rods being treated as full-length rods with respect to overall quantity of fission products is one approach to the PLR issue and that other approaches are acceptable.

Comment Response The NRC staff agrees with this comment. RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. Treating BWR part-length rods as full-length rods with respect to the overall quantity of fission products, as outlined in Regulatory Position 3.2 of RG 1.183, Revision 1, is one acceptable approach to modeling part-length rods. Another acceptable approach is to use the procedure in RG 1.183, Revision 1, appendix I, Analytical Technique for Calculating Fuel-Design or Plant-Specific Steady-State Fission Product Release Fractions for Non-Loss-of-Coolant Accident Events (previously labelled appendix J in DG-1389), to calculate release fractions specific to part-length rods. No changes were made in response to this comment because the discussion in Regulatory Position 3.2, Release Fractions, in RG 1.183, Revision 1 states that appendix I is an alternative approach to the guidance in Regulatory Position 3.2 for calculating non-LOCA release fractions. Specifically, Regulatory Position 3.2 of RG 1.183, Revision 1 states, Appendix I provides an acceptable analytical technique for calculating plant-specific or fuel rod design-specific fission product release fractions. The NRC staff notes that in RG 1.183, Revision 1, the statement in the previous sentence now refers to appendix I rather than appendix J because of the deletion of DG-1389, Appendix I, Analysis Decision Flowchart, in response to Public Comment 11-56.

6-5 BWR Owners Group Comment Section 3.7 of Appendix A to Reg Guide 1.183 Rev. 0 contained a paragraph providing specific guidance regarding mixing in Mark III containments, including uniform mixing between the drywell and containment after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . All BWR/6 Mark-III units currently apply this guidance.

This paragraph has been deleted in DG-1389.In the absence of any guidance, what is an acceptable approach to mixing in Mark-III containments? Are the previous Mark-III mixing assumptions still acceptable for amendments prepared under the proposed revision? Is the 2-hour timing for uniform mixing from Rev. 0 still applicable considering the new release timing in Table 5? As reported in Item 2 in RIS 2006-04, are mixing approached based on thermal-hydraulic conditions still acceptable? If so, what would be the appropriate accident to apply for the MHA analysis? Should mixing in the drywell be considered substantial enough in the MHA such that the reactor vessel volume can be included in the drywell volume?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution NRC should confirm the continued applicability of the previous guidance or provide alternate guidance to BWR mixing between the drywell and wetwell/containment.

Comment Response Please note that the reference to Section 3.7 of appendix A to RG 1.183, Revision 0, is referring to Regulatory Position 3.7 of appendix A.

The NRC staff disagrees with this comment. Regulatory Position 3.7 of appendix A to RG 1.183, Revision 0, was challenged in Differing Professional Opinion (DPO)-2020-002 (ML21067A645).

As a result of the disposition of DPO-2020-002, DG-1389 provides guidance that safety-related systems may be credited to provide mixing of the MHA LOCA substantial core melt source term at the response times associated with these systems (see DG-1389, footnote 2). The previous 2-hour delay in containment mixing has been eliminated in DG-1389. Mixing in the drywell can be considered using the guidance in item 2 in Regulatory Issue Summary (RIS) 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ML053460347). Plant response to the MHA LOCA should be independent of the source term timing assumptions. The release of appreciable quantities of fission products into the containment would initiate plant response based on high containment radiation levels.

Containment leakage assumptions should continue to be based on the current design-basis pressure profile. The NRC staffs response to Public Comment 2-4 discusses the availability of RG 1.183, Revision 0, and both Revision 0 and Revision 1 being acceptable for use. No changes were made in response to this comment.

6-6 BWR Owners Group Comment This section and the removal coefficients in Table A-1 do not permit credit for the inboard MSL piping. The basis for this lack of credit seems to originate from Section 6.3 of SAND2008-6601 Analysis of Main Steam Isolation Valve Leakage in Design Basis Accidents Using MELCOR 1.8.6 and RADTRAD dated October, 2008 (ADAMS Accession No. ML083180196) which states that at times in the simulation the temperature of portions of the in-board MSL piping are predicted to be high enough to vaporize fission products that had been previously deposited. However, many plants already credit this volume and RIS 2006-04 specifically permits this credit. Section 2 of RIS 2006-04 states modeling of MSL piping may include volumes between the reactor pressure vessel and the inboard MSIV (inboard volume), between the inboard and outboard valves (in-between volume), and outside of the outboard valve (outboard volume). These inboard steam lines represent significant lengths of safety-related horizontal surfaces that are amenable for source term deposition. It is recognized that these volumes would need to be subtracted from the overall drywell volume in performing the dose analysis. DG-1389 should allow licensees to take credit for this piping if analyses show temperatures that will not vaporize the deposited source term. Alternatively, re-vaporization of the deposited source terms can be modeled.

Commenters Proposed Resolution 19

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Add statement allowing licensees to take credit for the inboard MSL piping with appropriate supporting analysis.

Comment Response The NRC staff does not agree with the proposed revision that pertains to crediting removal in the inboard main steamline isolation valves (MSIVs). Although the staff has previously allowed credit for inboard deposition for some licensees (RIS 2006-04 discusses the considerations involved in such credit), DG-1389 provides a revised methodology that differs from RIS 2006-04 for modeling the MSIV leakage pathway. Although not providing credit for removal in the inboard main steamline volumes in RG 1.183, Revision 1, the regulatory position is supported by SAND2008-6601, but there are other reasons for not crediting these inboard volumes.

SAND2008-6601 showed that these volumes, which are connected directly to the steam dome, have higher concentrations of radioactivity than the containment during the beginning of the accident. However, the MSIV leakage model proposed in RG 1.183, Revision 1, removes the mechanistic modeling of the source term in the volume upstream of the inboard MSIV that was originally proposed in DG-1199, Draft Regulatory Guide DG-1199, (Proposed Revision 1 of Regulatory Guide 1.183, dated July 2000), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (proposed Revision 1 of Regulatory Guide 1.183), issued October 2009 (ML090960464). The change to model the accident as a maximum hypothetical accident, in part, removed the mechanistic modeling of the source term in volume, upstream of the inboard MSIV, originally proposed in DG-1199. Since the MSIVs serve as containment boundary valves, the steamline volume upstream of the inboard MSIV is treated deterministically as having the containment concentration (see DG-1389, Regulatory Position A-5.1). As such, the concentrations inside the steamline upstream of the inboard MSIV, inside the steam dome, and outside the RCS are assumed to be the same. In addition, the deposition methods in DG-1389 are based on the assumption that the volume at which the deposition is credited is inactive and the only two forces acting on the aerosol are buoyancy and gravity. This assumption is applicable only for volumes that are closed or near closed to outside forces (such as the volume between two closed valves). The inboard volume is not such a volume, as it is open to the reactor vessel, which contains enough energy to cause convection currents throughout this piping (note that it does not take much force to move these aerosols).

No changes were made in response to this comment.

6-7 BWR Owners Group Comment Multiple BWRs currently have credit for aerosol removal from drywell sprays as well as aerosol deposition within in the main steam lines in their current licensing basis. Section A-5 presents three acceptable methods for calculating aerosol deposition within the main steam lines, but caveats the methodology with, however, these methods are not valid if credit has been taken for aerosol removal from drywell sprays.

Given the prevalence of credit for both sprays and steam line deposition, there should be a model presented that the Staff finds acceptable for crediting both, or modifications to the presented models if the licensee wants to credit spray removal (e.g., different aerosol size distribution)?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Provide additional guidance on MSL deposition for BWRs that credit drywell sprays.

Comment Response Please note that the reference to Section A-5 in the comment is referring to Regulatory Position A-5.

The NRC staff does not agree with this comment. While some BWRs currently have credit for drywell sprays, the majority of BWRs do not credit drywell sprays, and proprietary or more complex methods are typically used to model the MSIV leakage pathway for those that do. This RG provides one or more methods that the NRC staff considers acceptable for meeting the NRCs regulatory requirements, and licensees may use alternatives to this guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. Any integrated removal model (such as crediting containment sprays and removal via suppression pool) should account for all interactions and would be evaluated on a case-by-case basis. No changes were made in response to this comment.

6-8 BWR Owners Group Comment The time periods in DG-1389 Table A-1 do not correspond to those in Table 6-1 of SAND2008-6601. Specifically, SAND2008-6601 has a 0-2 hour time interval where higher removal coefficients can be applied. Then, the next time step is 2-12 hours with a third time period of 12+ hours. DG-1389 only contains a 0-10 hour period and one for 10+ hours.

Commenters Proposed Resolution Correct Table A-1 with the time periods and removal coefficients from the underlying reference.

Comment Response The NRC staff partially agrees with this comment. The staff agrees that time periods in DG-1389, table A-1, do not correspond to those in table 6 of SAND2008-6601. However, this offset of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is based on differences in the timing of the credit for reflooding the reactor between what was assumed in SAND2008-6601 and what is assumed in the proposed revised RG 1.183. In the proposed RG 1.183, Revision 1, guidance, the MHA LOCA modeling assumptions credit reflood soon after the start of the accident rather than at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , as assumed in SAND2008-6601. Table A-1 in DG-1389 accounts for these differences by using the values in SAND2008-6601, table 6.1, for 2-12 hours and applying them to the 0-10-hour period in table A-1. Likewise, the values in SAND2008-6601, table 6-1, for 12+ hours are applied to the 10+ hour period in table A-1. This aligns the deposition values in SAND2008-6601 after reflood to those after the time of assumed reflood in proposed RG 1.183, Revision 1. The NRC staff notes that this comment is also related to Public Comment 11-38. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 6-9 BWR Owners Group Comment Description of concern:

A pre-accident iodine spike occurs when a plant enters a time-limited action statement related to the limiting condition for operation (LCO) for specific RCS activity. The inclusion of this abnormal operation condition in the licensing basis analyses is unique. Typically, abnormal operation conditions, which require a time-limited action statement to restore the system to operable status, are not included in the licensing basis analyses.

A pre-accident iodine spike occurs independent of and prior to the main steam line break (MSLB). Thus, the MSLB with a pre-accident iodine spike constitutes two independent overlapping design basis conditions, the combination of which is typically considered beyond design basis.

As supporting evidence, the probability of a MSLB occurring during the LCO period of a DEI excursion can be shown to be beyond the scope of a plants licensing basis. Based on the industry average parameter estimates in NUREG/CR-6928, the mean frequency of a BWR steam line break outside containment is 2.20E-3/year. This frequency was developed by Idaho National Laboratory and is based on data as recent as 2020. This value is used in the NRC SPAR models and industry PRA models. NRC: Industry Average Parameter Estimates (inl.gov)

The likelihood of this event occurring coincident with the plant being in a DEI Limiting Condition is extraordinarily low. Standard Technical Specifications 3.4.7 and 3.4.8 in NUREGs-1433 and 1434 respectively report this LCO period is only 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or 0.55% of the year for all BWR types. Fuel performance indicators would typically direct a plant to shut down well before it even enters this LCO.

Considering that the MSLB accident is independent of the chemistry excursion that leads to the DEI exceeding the maximum steady-state value (typically 0.2 µCi/g), the probability of a MSLB occurring during this LCO period is then only 1.21E-5. Including a subsequent LOOP (mean frequency on the order of E-02) in this scenario leads to an even lower frequency (on the order of E-07).

This accident frequency is well below the Design Basis AccidentLimiting Fault range of 1E-4/year to 1E-6 described in Annex II of IAEA Publication SSG-2, Rev. 1 Deterministic Safety Analysis for Nuclear Plants 2019. As such, the MSLB case assuming a pre-existing spike should not be included in the plant design bases.

Commenters Proposed Resolution Eliminate the pre-accident iodine spike as a requirement for licensing basis analyses. The inclusion of an abnormal operation condition in the licensing basis analyses is unique to the pre-accident iodine spike scenarios. These unique requirements should be either removed or justified.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff does not agree with this comment. As stated in 10 CFR 50.67, footnote 2 The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

The supporting information provided in the comment to remove traditional deterministic analyses uses probabilities to justify not considering the pre-accident spike. This approach conflicts with the NRCs policy statement on the use of probabilistic risk assessment (PRA) methods (60 FR 42622, August 16, 1995), which calls for the use of PRA technology in all regulatory matters in a manner that complements the deterministic approach and supports the defense-in-depth philosophy. In this case, probability is being used to propose the removal of deterministic methods rather than complement them.

Currently, the safety analysis guidance in RG 1.183, Revision 0, considers two cases of reactor coolant iodine-specific activity. One case assumes specific activity at typically 1.0 microcurie per gram dose equivalent iodine (I)-131 (DEI-131), with a concurrent iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a steamline break (by a factor of 500) or SGTR (by a factor of 335),

respectively. The second case assumes the initial reactor coolant iodine activity at typically 60.0 microcuries per gram DEI-131 due to an iodine spike caused by a reactor or an RCS transient before the accident. Operation with iodine-specific activity levels greater than typically 1 microcurie per gram DEI is permissible if the activity levels do not exceed the analyzed preexisting spike level condition of typically 60 microcuries per gram for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

These limits are also used for establishing standardization in radiation shielding and plant personnel radiation protective practices and usually satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The allowance to operate up to the preexisting spike level of typically 60 microcuries per gram DEI provides an upper bound on allowed reactor coolant activity consistent with the accident analysis. The technical specifications are derived from this safety analysis. If this analyzed condition did not exist, then operation above 1 microcurie per gram DEI would require a shutdown (e.g., NUREG-1431, Revision 5, Standard Technical SpecificationsWestinghouse, issued September 2021 (ML21259A155 and ML21259A159), Technical Specification 3.4.16, Condition C, states that if the DEI is greater than the analyzed condition of typically 60 microcuries per gram, then the reactor must be brought to Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ) because only the coexisting spike analysis would exist. As noted in the Statements of Consideration for 10 CFR 50.67 (64 FR 71990; December 23, 1999), These accident analyses are intentionally conservative in order to address uncertainties in accident progression, fission product transport, and atmospheric dispersion. No changes were made in response to this comment. The NRC staff also notes that this comment is similar to Public Comment 8-6.

Comment Submission 7 23

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 ADAMS Accession No. ML22174A053 Email: pcoutur@entergy.com Organization: Entergy 7-1 Entergy Comment In the original implementation of the alternative source term, the EQ impacts of a simple conversion to the AST with no additional changes or relaxations was evaluated as part of the re-baselining work. The Sandia evaluation in ADAMS Accession No. ML20154G363 determined the AST airborne EQ doses were lower than those based on TID-14844 while the pool doses were addressed and dispositioned under Generic Issue 187. Section 1.3.5 indicates that the EQ analyses may only be affected by proposed plant modifications; however, no additional benchmarking on plant impacts appears to have been performed to support this DG.

Considering the significant changes to the timing and release fractions, are the EQ-related conclusions from the original AST effort still applicable?

Commenters Proposed Resolution Confirm applicability of original AST conclusion on EQ.

Comment Response Please note that the reference to Section 1.3.5 in the comment is referring to Regulatory Position 1.3.5.

The NRC staff partially agrees with this comment. 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. Plants are not required to update their EQ analysis based on the update to RG 1.183, Revision 1. However, the NRC staff has not assessed the acceptability of continuing to use TID-14844 source term methodology for EQ for facilities that have modified their licensing basis to higher burnup or enrichment. Therefore, when these types of significant plant modifications occur, licensees must appropriately evaluate the impact on EQ, consistent with the modifications made, and make updates as appropriate (e.g., the licensee would need to determine whether changes to the source term methodology for EQ, or other related changes to EQ, are necessary and evaluate the radiological impacts to equipment). No changes were made in response to this comment.

7-2 Entergy 24

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment The halogen component of the BWR core release fractions have nearly doubled from those the

[sic] Revision 0. The SAND2011-0128 report provides no details on why the BWR halogen releases increased so dramatically while the PWR releases decreased. The small burnup extension considered in these updated analyses would not be expected to lead to such a large difference for just one plant type.

As the operator of two large BWRs, these higher release fractions will make it difficult to apply this new Reg Guide considering the current margins to the regulatory acceptance criteria.

Commenters Proposed Resolution Provide the basis for the significant increase in the BWR halogen release fractions in Table 1.

Comment Response The NRC staff disagrees with the comment. The SAND2011-0128 report abstract and Section 1.2, Insights Since Publication of the Alternative Source Term, provide high-level details regarding differences between the recommendations in NUREG-1465 and SAND2011-0128. Specific details on why the BWR halogen releases increased from those in NUREG-1465 can be found in SAND2011-0128, section 4.0, and are discussed for each release phase. The updated source terms reflect the current understanding of severe accidents and fission product behavior since the publication of RG 1.183, Revision 0, which is based on NUREG-1465.

Since the development of the NUREG-1465 source term, modeling of core degradation has improved greatly, largely by identifying and modeling efficient mechanisms for distributing heat from the degrading reactor fuel to the RCS, especially by natural convection processes.

Consequently, degrading core material is not predicted to become as hot, as rapidly, as it was in calculations of reactor accidents using the source term code package that was the basis of the NUREG-1465 source term.

The degrading core materials are hot enough to sustain the continued release of volatile radionuclides. Because the period between the onset of core degradation and the penetration of the reactor vessel by core debris is longer, the prolonged core degradation allows more of the volatile radionuclides to be released during the in-vessel release phase of an accident. As a result, the release fraction for volatile radionuclides, such as iodine, is larger than that of the NUREG-1465 source term for BWRs. Specifically, the total halogens release fraction increased from 0.3 in RG 1.183, Revision 0, to 0.543 in DG-1389.

The same modeling improvements affect BWR and PWR technologies differently. This may be due to several factors including variability in accident progression, vessel internal structure temperature differences between BWRs and PWRs, and the potential scrubbing of fission products in PWRs by the pressurizer relief tank, thereby dampening the impact of larger releases than in NUREG-1465 on releases to containment. Overall, the combined effects of modeling changes resulted in halogens increasing in BWRs but slightly decreasing from 0.4 in RG 1.183, Revision 0, to 0.377 in DG-1389 for PWRs. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 7-3 Entergy Comment Tables 1 and 2 report release fractions from a new Molybdenum group; however, the elements in this new group are not listed in Table 6. Which release group is Zirconium (Zr) considered to reside? DG-1389 is consistent with Rev. 0 and indicates it is part of the Lanthanides while Table 14 in the underlying Sandia report (SAND 2011-0128) reports Zirconium as part of the Cerium group.

Commenters Proposed Resolution It is expected that the final grouping is consistent with SAND-2011-0128 since that is the basis for the release fractions in Tables 1 & 2. Update the Table 6 for the new proposed Molybdenum group. Change Zr grouping to Cerium group.

Comment Response The NRC staff agrees with this comment. The staff revised table 6 to be consistent with the groupings in SAND2011-0128, table 14, and ERI/NRC 11-211, section 4.3, in response to Public Comment 7-3 and similar comments in Public Comments 11-15, 11-20, and 15-5.

7-4 Entergy Comment The reported onset time of the early in-vessel release phase for BWRs is listed as 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

This value does not appear to include the delay from the BWR coolant release phase of 2 minutes. If the gap release begins at 2 minutes and lasts 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in duration, the onset time of the early in-vessel release phase would be 0.193 hours0.00223 days 0.0536 hours 3.191138e-4 weeks 7.34365e-5 months if it occurred consecutively after the gap release phase as the text indicates. However, Table 5 reports 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months as the onset time implying the early in-vessel phase begins before the gap release phase ends.

Commenters Proposed Resolution To accurately consider the 2-minute coolant release phase, the reported onset time of the early in-vessel release phase for BWRs should be updated to 0.193 hours0.00223 days 0.0536 hours 3.191138e-4 weeks 7.34365e-5 months .

Comment Response The NRC staff agrees with this comment. The staff revised table 5 to be consistent with the intent of this comment for BWRs. In addition, for consistency, it also updated the table for PWRs. The value for BWRs was changed to 0.19 hour2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months rather than 0.193 hour0.00223 days 0.0536 hours 3.191138e-4 weeks 7.34365e-5 months .

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 7-5 Entergy Comment The inputs to the design basis analysis are numerous and although each of these inputs would have a distribution of applicable values, this DG typically requires values that represent the worst-case (i.e., generally 5th or 10th percentile of the distribution). This conservatism in the inputs is compounded in the evaluation process and leads to an extremely conservative result.

As currently performed for 50.46 LOCA analyses, the required reasonable assurance in 10 CFR 50.67 may be provided with a statistical analysis based on probability distributions that have been rigorously developed from the underlying data. This change in methodology would likely require additional NRC review.

Commenters Proposed Resolution Confirm that the caution proposed for deviating from the standard methodology does not preclude statistical approaches to assessing the dose consequences.

Comment Response The NRC staff agrees with this comment. RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

7-6 Entergy Comment DG-1389 states that the control room methodology in RG 1.194 may be used to estimate the offsite X/Q out to distances of 1,200 m. This is a new position and the basis for this appears to be DG-4030; however, the associated Reg Guide 1.249 has not been released.

Commenters Proposed Resolution Reference the basis for this new position Comment Response The NRC staff agrees with this comment. The issuance of the revision to RG 1.249 (DG-4030)

Use of ARCON Methodology For Calculation Of Accident-Related Offsite Atmospheric Dispersion Factors has been delayed. Accordingly, the NRC staff revised Regulatory Position 5.3 to remove any references to the methods for estimating the offsite atmospheric dispersion factors using the control room methodology in RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants. The NRC staff revised RG 1.183, Revision 1, in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 7-7 Entergy Comment Aerosol deposition using the NUREG/CR-6189 methods are credited at some Entergy plants.

Section 6.5.2 of DG-1389 indicates that NUREG/CR-6189 continues to be acceptable for removal of iodine and aerosols. However, based on Section 1 of NUREG/CR-6189, the models are based on the release fractions and timing in NUREG-1465. Considering the significant changes to the release fractions and timings in DG-1389, there may be significant impacts to the deposition rates in this NUREG.

Commenters Proposed Resolution Confirm the continued applicability of the NUREG/CR-6189 aerosol removal rates.

Comment Response Please note that the reference to Section 6.5.2 of DG-1389 in the comment is referring to Regulatory Position A-2.2.

The NRC staff partially agrees with this comment. NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Contaminants, issued July 1996 (ML100130305), is based upon the NUREG-1465 source term, which is no longer the basis for the source term in RG 1.183, Revision 1. However, the methods in NUREG/CR-6189 may be credited on a case-by-case basis if they are adjusted to incorporate the revised MHA source term in Revision 1 of RG 1.183. The NRC staff revised Regulatory Position A-2.2 in response to this comment. The revision states that the NRC staff no longer accepts the reductions calculated in NUREG/CR-6189, but that the methods in this NUREG could be credited on a case-by-case basis if they are adjusted to incorporate the revised MHA source term in Revision 1 of RG 1.183.

7-8 Entergy Comment Section 3.7 of Appendix A to Reg Guide 1.183 Rev. 0 contained a paragraph providing specific guidance regarding mixing in Mark III containments, including uniform mixing between the drywell and containment after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This paragraph has been deleted in the proposed revision. Entergy operates 2 BWR/6 plants with Mark-III containments that credit the mixing guidance from Revision 0. In the absence of any guidance, what is an acceptable approach to mixing in Mark-III containments? Are the previous Mark III mixing assumptions still acceptable for amendments prepared under the proposed revision? Is the 2-hour timing for uniform mixing from Rev. 0 still applicable considering the new release timing in Table 5? With the new MHA-LOCA scenario, should drywell-wetwell/containment mixing even need to be considered?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution NRC should confirm the continued applicability of the previous guidance or provide alternate guidance.

Comment Response Please note that the reference to Section 3.7 of appendix A (RG 1.183, Revision 0) in the comment is referring to Regulatory Position 3.7 (now Regulatory Position A-2.7 in RG 1.183, Revision 1).

Please see the NRC staffs response to Public Comment 6-5, as that comment is essentially the same. No changes were made in response to this comment.

7-9 Entergy Comment The DG indicates that aerosol deposition may be considered in bypass pathways. However, plate-out of elemental iodine may also be a significant removal mechanism in gas-filled secondary bypass leakage pathways. In addition to aerosol deposition, the plate-out of elemental iodine should not be excluded.

Commenters Proposed Resolution Add statement allowing iodine plate-out.

Comment Response The NRC staff agrees with this comment. The staff revised Regulatory Position A-3.5 of RG 1.183, Revision 1, to state, Similarly, deposition of aerosols and elemental halogens in gas-filled lines, [emphasis added] so that the case-by-case consideration is not restricted to just aerosol deposition. Note that halogens rather than iodine was used in Regulatory Position A-3.5, because iodine is a halogen and because of the NRC staff response to a similar comment provided in Public Comment 11-29.

7-10 Entergy Comment The DG states that the chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to be 95-percent cesium iodide (CsI), 4.85- percent elemental iodine, and 0.15-percent organic iodide. While DG-1389 only- reports the iodine species distribution released into the pool, Reg Guide 1.183 Rev. 0 reported that the iodine chemical species above the water is 57% elemental and 43% organic. For Rev. 1 applications, what iodine species 29

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 distribution should be applied for the early 2-hour airborne release considering the new fuel handling accident model? Based on the calculated overall pool DF from Equations 3-1, 3-2, and 3-3, is a species-dependent DF approach acceptable with organic iodine having a DF of 1? Is a purely elemental iodine species appropriate for the long-term release since this is comprised of re-evolved elemental iodine?

Commenters Proposed Resolution Provide additional guidance on acceptable airborne iodine species assumptions.

Comment Response The NRC staff agrees with these comments. Regarding the comments question about which iodine chemical species should be applied for the early 2-hour airborne release, the staff revised Regulatory Positions B-1.3 and B-2 to clarify the methods that can be used to determine the iodine chemical species above the pool for the early 2-hour airborne release.

The comment also asked about whether a species-dependent decontamination factor (DF) approach is acceptable. A species-dependent DF approach is acceptable to determine the iodine chemical species above the pool during the early 2-hour airborne release. The NRC staff revised Regulatory Positions B-1.3 and B-2 to clarify how this approach could be used by stating that elemental iodine is decontaminated by the overlying pool water; Regulatory Position B-2 was also revised to provide the species-dependent DFs for elemental and organic iodine.

The comment further asked whether a purely elemental iodine species is appropriate for long-term release. It is appropriate, and the NRC staff revised Regulatory Position B-1.3 to clarify the solid cesium iodide and elemental iodine scrubbed by the pool in Phase 1 re-evolves as elemental iodine for the long-term phase. The NRC staff incorporated the revisions discussed in response to these comments into RG 1.183, Revision 1.

7-11 Entergy Comment For the fuel handling accident, DG-1389 indicates that the fission product release from the breached fuel is based on the non-LOCA gap fractions in Regulatory Position 3.2; however, the NRCs assessment in the Staffs example in ADAMS Accession No. ML21190A040 appears to apply a much larger value of 23% for Iodine-129 which is significantly higher than the 4% from Table 4 for other halogens.

Commenters Proposed Resolution Confirm the applicability of Tables 3 and 4 for the other halogens for application in the fuel handling accident.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff agrees with this comment. Table 3, BWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, and Table 4, PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, are applicable for the release fractions for other halogens in the FHA, as described in Regulatory Position B-1 and Regulatory Position 3 of RG 1.183, Revision 1. Specifically, Regulatory Position B-1 states that Regulatory Position 3 of this guide provides acceptable assumptions regarding core inventory and the release of radionuclides from the fuel. Regulatory Position 3 references, in part, tables 3 and 4. The values provided by the NRC staff in a memorandum dated August 30, 2021 (ML21190A040) are intended to be examples. Regulatory Position 3.2, tables 3 and 4, gives guidance to licensees for the non-LOCA gap fractions values to use. No changes were made in response to this comment.

7-12 Entergy Comment Section B-2 states that the decontamination factor (DF) can be calculated from Equations in B-1, B-2, and B-3, if the water depth is between 19 and 23 feet. There is no guidance in the event the water depth is greater than 23 feet. Some Entergy plants consider drops over the core where there is significantly greater than 23 feet of water coverage. The proposed DF model would be expected to be valid for these larger water depths.

Commenters Proposed Resolution The Reg Guide should allow the application of the DF equations in Appendix B-2 for water depths greater than 23 feet. Since these equations are not directly dependent on depth, it would be expected that they would yield a conservatively low DF than an actual value for cases that credit the additional scrubbing depth.

Comment Response Please note that the reference to Section B-2 in the comment is referring to Regulatory Position B-2.

The NRC staff disagrees with this comment. The model described in the NRC staff technical paper Re-evaluation of the Fission Product Release and Transport for the Design-Basis Accident Fuel Handling Accident (ML19248C647), is based on the boundary conditions of experiments with a pool depth of 23 feet. Regulatory Position B-2 equations are, in part, a function of bubble rise time through the experimental pool depth of 23 feet of water. No changes were made in response to this comment.

Comment Submission 8 ADAMS Accession No. ML22174A057 31

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Name: Michael Powel, Chairman and COO Email: holdercm@westinghouse.com Organization: PWR Owners Group Address: Program Management Office, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066 8-1 PWR Owners Group Comment It is not clear if licensees can transition one (or a few) analyses to revision 1, or if full implementation for all design basis accidents is required. In revision 0 of RG 1.183, partial implementation was allowed.

Full implementation of DG-1389 could be cost prohibitive. Additionally, vague guidance could result in submittals that do not meet the intent of the Reg Guide.

Commenters Proposed Resolution Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations.

Selective implementation is acceptable, provided that each accident analysis uses either revision 0 or revision1. A combination of the methods contained in revision 0 or revision 1 of RG 1.183 in a single analysis would need additional justification. Basis for Proposal: It is assumed that the NRC intends to allow partial implementation of DG-1389 but did not intend the guidance in the DG to be mixed in a single analysis.

Comment Response The NRC staff partially agrees with this comment but not the proposed resolution. Both Revision 0 and Revision 1 of RG 1.183 will remain available for use, because Revision 0 will not be withdrawn. Each revision provides a method acceptable to the NRC staff for compliance with the applicable regulations specified in the guidance. Use of combinations and permutations of regulatory positions from Revision 0 and Revision 1 will need appropriate technical justification and may require additional NRC review before approval. Many considerations may factor into a licensees or applicants voluntary implementation, including, for example, the current licensing and design basis. Regulatory Position 1.3.2 and Regulatory Position 5.1.4 provide guidance for these considerations and on selective and full implementation of an AST. However, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. As DG-1389 already contained reanalysis guidance and guidance on the applicability of the prior licensing basis, no changes were made in response to this comment. However, please see the NRC staffs response to Public Comment 11-2 for a related request to revise the same paragraph proposed to be revised in this comment.

8-2 PWR Owners Group 32

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment The statement implies that so called transient fission gases releases (TFGR) due to high burnup pellet fragmentation must be considered for non-LOCA DBAs that predict fuel failure but does not provide details how this requirement could be met. Additional fragmentation-based accident transient fission gas releases are tied to certain mechanisms, such as increases in fuel temperature, that occur in specific event scenarios. For Non-LOCA, TFGR guidance is currently only available for reactivity insertion accidents (RG 1.236). There is no guidance available to address non-LOCA accidents aside from reactivity insertion accidents and little experimental data to develop a model or support the concern. The mechanism for transient fission gas release and test data supports a fuel temperature dependent relationship. The first phase of release is in the range of 500-800°C.

Commenters Proposed Resolution Appropriate additional high burnup-related TFGR guidance for non-LOCA DBAs should be provided. Non-LOCA DBA without fuel overheating are not impacted. For non-LOCA DBA with fuel overheating, guidance should be specific. Identify the specific non-LOCA DBAs that are impacted and include applicable TFGR guidance in the relevant Appendices. Existing data indicates significant transient gases require the accidental conditions associated with RIA

[reactivity initiated accidents] or the higher temperature thresholds associated with LOCA-type events. The Draft Guide should not extrapolate TFGR to other non-LOCA accidents without adequate references.

Comment Response The NRC staff partially agrees with this comment. The staff agrees that the guidance related to TFGR for non-LOCA and non-reactivity-initiated events should be clarified. The staff disagrees that guidance should be strictly limited to only high-temperature DBAs. There is neither a consensus on the mechanism of TFGR nor a significant body of experimental data for the computation of TFGR for non-LOCA and non-reactivity-initiated DBAs to provide specific guidance at this time. As a result, future applications should address this using engineering judgment or experimental data. Though not directly applicable, NRC RIL 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, issued December 2021 (ML21313A145), provides TFGR data for design-basis LOCAs that may be used to provide a bounding estimate of TFGR for other high-temperature DBAs.

Concerning lower temperature DBAs, such as FHAs, despite a lack of data, the fracturing of the fuel could potentially lead to the release of additional fission gas from the fuel even at low temperatures, especially because microcracking of the uranium dioxide fuel, leading to the burst release of fission gas from the grain boundaries, is one postulated mechanism of TFGR. The NRC staff may decide to issue more specific guidance for calculating TFGR for non-LOCA and non-reactivity-initiated DBAs in the future if more data become available. To note, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. The staff revised RG 1.183, Revision 1, Regulatory Position 3.2, in response to this comment to clarify the guidance related to TFGR for non-LOCA and non-reactivity-initiated events.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 8-3 PWR Owners Group Comment Akin to the limitation allowance for operator action in the current draft text, the guidance should not exclude all potential use of non-safety related equipment. Consistent with the historic practice in safety analysis, justifiable credit for non-safety grade equipment has been acceptable.

Commenters Proposed Resolution The DG should provide additional guidance for crediting non-safety grade equipment during an accident. There is currently no guidance for providing acceptable justifications for crediting non-safety grade equipment to mitigate an accident within the draft guide. Add text to the RG discussing the elements necessary to justify crediting the use of non-safety equipment.

Examples of crediting the use of non-safety equipment include the crediting the Hatch 1 & 2 turbine building ventilation system described in ADAMS Accession No. ML072910399 and approved via ADAMS Accession No. ML081770075 and the Oconee QA Condition 5 program described in Attachments 4, 4a and 4b of ADAMS Accession No. ML15238A066.pdf and accepted in ADAMS Accession No. ML16141A933. The NRC should provide additional guidance to credit non-safety related equipment following an accident to ensure that these justifications are addressed consistently across the LWR fleet.

Comment Response The NRC staff disagrees with the proposed change to add further guidance. The staff notes that Regulatory Position 1.4 of RG 1.183, Revision 1, states, in part, that [t]he licensee may elect to use risk insights in support of proposed changes to the design basis that are not addressed in currently approved NRC staff positions. Additionally, guidance can be found in RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, issued January 2018 (ML17317A256). No changes were made in response to this comment.

8-4 PWR Owners Group Comment Control rooms are designed to maintain habitable and safe conditions under accident conditions, including loss-of-coolant accidents. Control room dose calculations are performed to demonstrate the adequacy of the control room design with respect to radiological protection.

Control rooms are not designed to maintain habitable and safe conditions for those external to the boundary of the control room.

10 CFR 50.67(b)(2)(iii) states Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident. This echoes the language of 10 CFR 50, Appendix A, General Design Criterion 19. With respect to radiological consequences, 10 CFR 50, Appendix A, General 34

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Design Criterion 19 states Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

In accordance with General Design Criteria 19, control rooms are designed to protect individuals occupying an accessible control room. With respect to radiological consequences, a key focus of the design is on HVAC considerations, including a safety-related emergency mode with HEPA filters and charcoal beds, with provisions made for limiting unfiltered inleakage. The physical structure, i.e. walls, ceiling, and floor, of the control room is designed to provide shielding to occupants from a postaccident radiological release. Post-accident access to the control room for the operators is not limited by the control room design or control room habitability system design, except to the extent where the control room must remain accessible.

i.e., the operators can enter and exit the control room freely. The existing analysis assumption of an operator remaining in the control room for 30 days (with allowances for occupancy factors and reasonable in-leakage which covers ingress and egress through the control room envelope boundary) is intended to be an overall conservative approach with respect to the expected rotation of personnel. Demonstrating doses less than the design limit confirms the adequacy of the design of the control room (along with the other engineered safety features of the plant design). This is the dose reported in Chapter 15.

Doses accrued by operators when not in the control room are not explicitly accounted for in the design of the control room as the occupancy factor assumptions are intended to provide an overall conservative approach. In addition, 10CFR50.47(a)(9), (10), and (11) provide sufficient requirements for the Emergency Response organizations to measure and limit doses to emergency response personnel including control room operators.

It is understood that a small subset of licensees include transit dose in their respective Chapter 15 control room doses. This is not a widespread practice. As such, there is no definitive guidance for calculating transit doses.

Commenters Proposed Resolution Please remove the discussion of transit dose. Control room doses should be limited to doses accrued while occupying the control room provided.

Comment Response In consideration of the comments, the NRC staff has removed the language associated with transit dose in Regulatory Position 4.2 and restored the original language from RG 1.183, Revision 0. The staff is evaluating this issue and will determine whether to address it in a future revision to this RG.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 8-5 PWR Owners Group Comment The holistic impact of transitioning from Revision 0 to Revision 1 is not clear. Evaluations of a sample plant(s) would illustrate the impacts of transitioning from Revision 0 to Revision 1 on plant analyses and plant maintenance. DG-1389 represents a significant change in methodology compared to RG 1.183. The difference is comparable to the transition from TID to AST circa 1999. Major changes have been made to the release fraction, timing, and accident definition. It is not clear how the proposed changes will impact existing PWR dose analyses or plant maintenance requirements based on those analyses. For the transition from AST to TID, sample plant evaluations were developed to illustrate the impact. For example, the MHA release duration in DG-1389 (4.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) is extended compared to RG 1.183 Revision 0 LOCA (1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ). This increase in duration could result in additional duration-of-operation requirements on containment sprays.

Commenters Proposed Resolution Please provide evaluation(s) of sample plant(s) to illustrate the impacts of transitioning from Revision 0 to Revision 1. The impact of transitioning from Revision 0 to Revision 1 should be illustrated by example, similar to the transition from TID to AST (SECY-98-154).

Comment Response The NRC staff disagrees with this comment. RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. Individual plants may consider the impact of transitioning from Revision 0 to Revision 1 based on their design and licensing basis prior to using Revision 1, because such evaluations would be best accomplished on a site-specific basis by licensees, since generic examples may yield significantly different results than those performed on a site-specific basis. In addition, Revision 0 is not being withdrawn and will remain available for use. Accordingly, the option to continue using Revision 0 still exists. The NRC staff also notes that the analysis documented in SECY 98-154, Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors, dated June 30, 1998 (ML992880064) was performed by the NRC staff to evaluate and provide the technical basis for rulemaking (which became 10 CFR 50.67). Revision 1 to RG 1.183 is not associated with a change to the NRCs regulations.

No changes were made in response to this comment.

8-6 PWR Owners Group Comment A pre-accident (also called pre-incident) iodine spike occurs when a plant enters a time-limited action statement related to the limiting condition for operation (LCO) for specific RCS activity.

The abnormal operation condition occurs independent of and prior to the design basis event 36

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 (e.g., MSLB or SGTR). Thus, a MSLB or SGTR with a pre-accident iodine spike constitutes two independent overlapping design basis conditions, the combination of which is typically considered beyond design basis. This is unique to the dose analyses. While EAB and LPZ dose limits are higher for this scenario, the Control Room dose limit is unchanged. It is recommended that the pre-accident iodine spike be removed as a requirement for licensing basis dose analyses.(see original incoming for further information)

Commenters Proposed Resolution Eliminate the pre-accident iodine spike as a requirement for licensing basis analyses. The inclusion of an abnormal operation condition in the licensing basis analyses is unique to the pre-accident iodine spike scenarios. These unique requirements should be either removed or justified.

Comment Response Please see the response to Public Comment 6-9.

8-7 PWR Owners Group Comment This is a new footnote that was not included in RG 1.183 Rev. 0. The intent of this new footnote is unclear. There are multiple interpretations for this new footnote. Within context, it is understood to mean that the same set of nuclides should be used in the dose calculations both internal and external to the control room, with recognition that the relative dose significance of a given nuclide in each calculation may vary. Clarification by the NRC will ensure proper application by licensees.

Commenters Proposed Resolution Please clarify the footnote. Suggested text is provided in italics:15. The nuclides used for modeling dose from airborne radioactivity inside the control room may not be conservative for determining the dose from radioactivity outside the control room. Recognizing that the relative significance of a given nuclide may vary, the same set of nuclides should be used in the dose calculations both internal and external to the control room.

Comment Response The NRC staff agrees with this comment but disagrees with the proposed resolution. The staff removed proposed footnote 15 in DG-1389 in response to this comment. The footnote removed read, The nuclides used for modeling dose from airborne radioactivity inside the control room may not be conservative for determining the dose from radioactivity outside the control room.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 8-8 PWR Owners Group Comment It is not clear in the wording of Table 5 and surrounding text whether licensees should treat the times listed at the end of each phase as duration or end times. If they are treated as duration as described, there would be overlap between phases. However, the text states the early in-vessel release phase begins immediately following the gap release phase.

Commenters Proposed Resolution The DG should be clear in the intent of the modeling of release phases, so all licensees model the accident consistently and correctly. Please revise the word duration in Table 5 and surrounding text to state end time or similar wording, to clarify there is no overlap of the phases, and the final times listed are the end of any releases (4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months or 8.0 hours0 days 0 hours 0 weeks 0 months ).

Comment Response The NRC staff agrees with this comment. The staff revised table 5 in response to this comment, to add clarity by changing the onset times for the early in-vessel release phase to include the onset time for the gap release phase, changing duration to end time, and adjusting the values accordingly. The NRC staff also removed the sentence regarding the duration of the gap release being 0.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months in the surrounding text to address an inconsistency. Please also see the NRC staffs response to Public Comment 7-4 for a similar change for BWRs.

8-9 PWR Owners Group Comment It is overly conservative to model the maximum Technical Specification leak rate for the duration of the event, especially in the later stages of the accident where the pressure differential across the steam generator tubes may be significantly reduced. This results in unrealistic activity release to the secondary system and the environment in events that model these releases (MSLB, SGTR, locked rotor, rod ejection). In later stages of the event(s), the pressure differential across the steam generator tubes is significantly reduced from event initiation. For instance, the TS leak rate is intended to bound the pressure differential during instances of a MSLB when the faulted steam generator blows down to atmospheric (or near atmospheric) conditions while the primary system remains at higher pressures (NOP down to ~1000 psia).

Later in the transient, residual heat removal cut-in conditions are reached, with the primary system pressure reduced to ~375 psig. MSLB conditions (and other secondary release events) are not anticipated to change the tube defect level in the steam generators; therefore, the maximum primary-to-secondary leakage should decrease significantly from initial conditions.

Commenters Proposed Resolution The NRC should add the following statement to Position E-6.1 (and related statements in other appendices) to allow for a licensee to credibly justify lower leakage rates during later stages of 38

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 the events. Licensees may provide adequate technical justification to consider lower primary-to-secondary leakage rates at later stages of the transient when pressure differentials across the steam generator tubes are reduced from those representing maximum leakages. This statement should also be added to the other appendices modeling primary to secondary leakage (Appendix F for MSLB, Appendix G for Locked Rotor, and Appendix H for Rod Ejection). It is overly conservative to model the maximum Technical Specification leak rate for the duration of the event, especially in the later stages of the accident where the pressure differential across the steam generator tubes may be significantly reduced.

Comment Response The NRC staff agrees with this comment. It is typical for licensees to model the fraction of the ruptured tube break flow that flashes to vapor in the steam generator to reflect ruptured steam generator tube flow data generated by their thermal-hydraulic analysis code. The staff has found this acceptable in previous license amendment requests. The NRC staff revised the guidance in RG 1.183, Revision 1, appendix E, Regulatory Position E-6.1, to reflect the recommended text as follows: The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses. Additionally, the text proposed in response to this comment has also been added to the guidance in other appendices describing modeling of primary-to-secondary leakage in PWR accidents, including appendix F, Regulatory Position F-6.4, for the PWR MSLB accident; appendix G, Regulatory Position G-5.1, for the PWR locked rotor accident; and appendix H, Regulatory Position H-7.1, for the PWR CRE accident. In addition, guidance on the modeling of primary-to-secondary leakage for facilities licensed with or applying for alternative repair criteria was being developed in DG-1074, Steam Generator Tube Integrity. However, the development of the guidance in DG-1074 has been cancelled, so the reference to this guidance was removed from appendices E, F, G, and H.

8-10 PWR Owners Group Comment The guidance in E-2.2 and F-2.2 appear to be inconsistent with table 7.

Commenters Proposed Resolution Guidance should be self-consistent throughout. Additionally comment 8-6 requests that the pre-accident spike be removed. Remove Pre-Accident spike from Table 7 and delete E-2.1 and F-2.1. This preserves intent of E-2.2 and F-2.2 and is consistent with the rest of the comment package.

Comment Response The NRC staff partially agrees. In response to this comment, the term pre-accident spike is now used throughout the guidance; coincident iodine spike in table 7 was also changed to concurrent iodine spike, consistent with Regulatory Positions E-2.2 and F-2.2. However, the staff does not agree with the removal of pre-accident spike, for the reasons described in the staffs response to Public Comment 6-9.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 8-11 PWR Owners Group Comment This equation, from the Murphy-Campe paper, is for Noble Gases. Use of AST will have many more isotopes in the control room than the Noble Gases (with the associated photon differences). Other equations could apply to other isotopes in the control room. The requirement to apply the equation provided in the RG is overly restrictive and may prevent a licensee from pursuing other justifiable conversion factors.

Commenters Proposed Resolution The RG should include an allowance for a licensee to proposal alternate conversion factors with appropriate justification, Licensees may choose to create their own finite volume correction factor based upon the isotopic mix (versus time) developed in their dose analysis. The Licensee needs to provide adequate justification supporting the alternate conversion factors. To assure that appropriate credit for shape factor is considered/used while calculating control room dose.

Comment Response The NRC staff partially agrees with this comment. The staff agrees that licensees may use a revised method other than that in RG 1.183, Revision 1, if the licensees demonstrate that the revised method satisfies the applicable NRC requirements. The NRC staff disagrees that the RG needs to be revised to explicitly allow for licensees to propose alternate conversion factors.

Specifically, this RG is intended to provide an acceptable method to demonstrate compliance with the applicable NRC regulations. Accordingly, licensees are not precluded from submitting other methods that may be acceptable with appropriate technical justification. However, endorsement of all the permutations of alternative methods that could be proposed is not necessary and would make the RG too voluminous. The NRC staff also notes that the NRC code RADTRAD (RADionuclide, Transport, Removal, and Dose Estimation) contains, and automatically applies, the control room geometry correction factor for all radionuclide dose conversion factors. No change was made in response to this comment.

8-12 PWR Owners Group Comment In PWRs, if forced circulation is not available (i.e. loss of off-site power), once the shutdown cooling system is placed into service, flow in the faulted loop may stall with the steam generator isolated. This means that although the active portions of the RCS will cool down via the shutdown cooling system and by dumping steam from the active SGs, the local fluid conditions in the faulted loop outside of the RHR cooling circuit could remain above 212F for an extended period of time. There may be inconsistent interpretations of how to determine when the radiological release from the faulted steam generator has terminated for the purpose of radiological analysis.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Update item F-6.4 to state that the release of radioactivity from faulted and unaffected steam generators may be analytically assumed to terminate when the bulk of the primary system is less than 212°F. Recommended F-6.4 text is below, with the purposed change in italics. The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature in the bulk of the primary system is less than 100 degrees Celsius (212 degrees Fahrenheit). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated. Note that similar guidance in E-6.3 for the unaffected SGs should be updated similarly. Once shutdown cooling is placed in service, RCS pressures and temperatures are significantly reduced. In addition, the local masses of fluid that would not be in communication with the active portions of the RCS are finite. The radiological release rate and consequence during this condition is very low. Explicit accounting for additional radiological release during this condition would add analysis complexity and burden without a significant change to the overall radiological consequence result.

Comment Response The NRC staff agrees with this comment. The staff made the proposed changes to RG 1.183, Revision 1, Regulatory Positions E-6.3, F-6.4, and G-5.3, in response to this comment.

Comment Submission 9 ADAMS Accession No. ML22174A068 Name: Brian Magnuson Email: magnuson28@msn.com Address: Aurora, IL 60504 41

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 9-1 Magnuson Comment Page 422 of 448 of Comment Submission 9 states: I am opposed to DG-1389. My prior public comments referenced SAND2008-6601 which determined the BWR MSIV source term methodologies provided in RG 1.183 (Revision 0) are non-conservative and conceptually inaccurate in 2008. Additionally, my prior comments expounded on SAND2008-6601 and identified other examples in which RG 1.183 methodologies violate the laws of physics.

RG 1.183 allows nuclear power plants (NPPs) to ignore the laws of physics in accident dose calculations that are used to demonstrate compliance with nuclear safety regulations, including General Design Criterion-19 (Appendix A to 10 CFR Part 50). In other words, the errors in RG 1.183 financially benefit nuclear power plants at the expense of public safety. It appears DG-1389 may correct a few of the technical errors in Revision 0 of RG 1.183; however, any corrections would be negated because it states: Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations. RG 1.183 Revision 0 has a broad range of safety ramifications. Until the NRC has reconciled the errors that SAND2008-6601 identified and I reported in prior public comments, it seems imprudent of the NRC to claim it is an acceptable method for demonstrating compliance with regulations. In effect, the errors identified in RG 1.183 Revision 0 provide a means for nuclear power plants to ignore the laws of physics in accident dose calculations in order to feign compliance with federal nuclear safety regulations.

[A similar comment is also found on pages 1 and 2 of 448 of Comment Submission 9.]

Comment Response The NRC staff disagrees with this comment. Regarding the NRC staff allowing licensees and applicants to use previous versions of RGs, please see the staffs response to Public Comment 15-6. The methods, techniques, or data described in an RG provide a method deemed acceptable to the NRC staff for compliance with the applicable regulations. However, applicants or licensees may choose to use alternative approaches as long as the applicant or licensee provides sufficient information that demonstrates that the requirements in the NRCs regulations are satisfied.

If a licensee or applicant chooses to select specific positions from different versions of an RG, then it would need to provide the appropriate technical justification.

Regarding the comment that the NRC needs to reconcil[e] the errors that SAND2008-6601 identified and I reported in prior public comments, please see the NRC staffs responses to Public Comments 9-12, 9-14, and 15-6.

No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 9-2 Magnuson Comment Page 422 of 448 of Comment Submission 9 states: I am opposed to using the DG-1389 term maximum hypothetical accident (MHA) loss-of- coolant accident (LOCA). An NRC Regulatory/Draft guide cannot legally be used to redefine the accident described in the applicable regulations. For example, the applicability of Appendix A to Part 50, General Design Criterion19 cannot be limited. Nevertheless, the apparent attempt drew attention to the most egregious contravention of RG 1.183 (and DG-1389).

[A similar comment is also found on page 2 of 448 of Comment Submission 9.]

Comment Response The NRC staff disagrees with this comment. The use of the term MHA LOCA in DG-1389 is intended to clarify the accident that the staff finds acceptable to use to meet the NRCs regulations; that is, a hypothetical accident that is assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products into the containment (see, for example, 10 CFR 50.34, footnote 6; 10 CFR 50.67, footnote 1; as well as similar footnotes in 10 CFR Part 52). There is no specific cause of the accident described in the regulation. The accident described in the regulation is a design basis accident. This accident description was used for siting in section 10 CFR 100.11, then added to 10 CFR Part 50 and 10 CFR Part 52 specifying the radiological design basis for nuclear power plants. No change was made in response to this comment.

9-3 Magnuson Comment Page 429 of 448 of Comment Submission 9 states: The term maximum hypothetical accident (MHA) appears misleading. Because Regulatory Guides are not regulations; it seems inappropriate to use an NRC Regulatory Guide to bound applicable regulations to the DG-1389 definition of the MHA. Otherwise stated, the NRC does not have the authority to redefine the applicability of federal nuclear safety regulations (or the current licensing basis (CLB) of nuclear power plants) by using Regulatory Guide 1.183. For example, the requirements of 10 CFR 50, Appendix A, General Design Criterion19 are not limited to a LOCA or the MHA described in DG-1389 Comment Response The NRC staff agrees that RGs are not regulations and rather just describe one or more methods that the staff considers acceptable for meeting the agencys regulatory requirements.

Licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. The NRC staff does not agree that the term maximum hypothetical accident is misleading because RG 1.183, Revision 1, is clarifying an accident that the staff finds would be acceptable to use to meet the description in the 43

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 applicable regulations (see, for example, 10 CFR 50.34, footnote 6; 10 CFR 50.67, footnote 1; as well as similar footnotes in 10 CFR Part 52).

Additionally, please see the NRC staff response to Public Comment 9-2. No changes were made in response to this comment.

9-4 Magnuson Comment Page 425 of 448 of Comment Submission 9 states: In 2019, the NRC further acknowledges that the control room accident dose criterion has proven to be challenging to demonstrate with most plants having very little margin to the regulation [Appendix B to Part 50, GDC-19]. Does very little margin to GDC-19 provide sufficient safety margins with adequate defense in depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs as is apparently required by RG 1.183?

Comment Response The NRC staff agrees with the request to answer the question provided. For clarification purposes, the language quoted by the comment (i.e., The control room accident dose criterion has proven to be challenging to demonstrate with most plants having very little margin to the regulation) is contained in Petition for Rulemaking (PRM) 50-121, Voluntary Adoption of Revised Design Basis Accident Dose Criteria, dated November 23, 2019 (ML20050M894).

When the NRC staff responded to this comment in June 2023, PRM-50-121 continued to be under NRC review, and so the NRC Commission had not made a conclusion on these statements. This language was provided by the petitioner of the PRM and is not a statement or acknowledgement made by the NRC staff. Notably, this statement is not contained in DG-1389.

Additionally, to clarify, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. As a result, RG 1.183 is not a regulatory requirement but rather is guidance that provides a method(s) acceptable to the NRC staff for meeting specific regulations. For context, the comment is quoting from the latter part of the following sentence from page 8 of DG-1389:

These license amendment requests should demonstrate that the facility, as modified, will continue to provide sufficient safety margins with adequate defense in depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs.

The NRC staff conducts reviews and makes findings consistent with agency regulations and guidance, relevant to the application provided to the staff. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 9-5 Magnuson Comment Page 425 of 448 of Comment Submission 9 states: As documented in my previous comments, the NRC clearly knows that the Protection by Multiple Fission Product Barriers are grossly inadequate. They cannot protect people (and the environment) from severe nuclear accidents as required by 10 CFR 100.11 and 10 CFR 50.67. In fact, the inferior design of these barriers will cause them to overheat, create explosive gases, and catastrophically self-destruct during credible accidents. Because many of us watched the containment barriers at Fukushima explode, the NRC was compelled to require similar nuclear power plants to install Hardened Containment Vents. In recognition that containment barriers will fail, the NRC now requires nuclear power plant operators to use the Hardened Containment Vents, during credible accidents, to release large quantities of highly radioactive material directly to the environment to prevent their containment barriers from self-destructing and releasing much more radioactive material. After studying severe accidents for decadesadmitting that containment barriers will fail in credible nuclear accidents and watching the containment barriers at Fukushima catastrophically fail, the NRC wrongly allows nuclear power plants to assume that containments will not fail in accident dose calculations using RG 1.183. Footnote 1 of DG-1389 states: These evaluations assume containment integrity with offsite hazards evaluated based on design basis containment leakage. Footnote 2 of DG-1389 states: The purpose of this approach would be to test the adequacy of the containment and other safety-related systems. These footnotes give rise to a circular position. DG-1389 proffers that containment barriers can be adequately tested by using design-basis evaluations that assume they will not fail. This fallacy epitomizes the NRCs design-basis contravention.

Comment Response The NRC staff disagrees with this comment. RG 1.183 provides guidance for the evaluation of DBAs. Severe accident evaluations that assume containment failure or require the use of hardened containment venting are not covered in RG 1.183. To note, the NRC staff has evaluated severe accident consequences, and documented the results in NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, issued November 2012 (ML12332A053 [package]).

No changes were made in response to this comment.

9-6 Magnuson Comment Page 427 of 448 of Comment Submission 9 states: If 10 CFR 50.67 is to be effective, it seems that any revision to RG 1.183 must account for the source terms of credible nuclear accidents that result in containment failures or require the use of Hardened Containment Vents.

Comment Response The NRC staff disagrees with this comment. RG 1.183, Revision 1, provides guidance for the evaluation of DBAs. Severe accident evaluations that assume containment failure or require the 45

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 use of hardened containment venting are not covered in RG 1.183, Revision 1. To note, the NRC staff has evaluated severe accident consequences and documented the results in NUREG-1935, "State-of-the-Art Reactor Consequence Analyses (SOARCA) Report" (ADAMS Package No. ML12332A053). No changes were made in response to this comment.

9-7 Magnuson Comment Page 428 of 448 of Comment Submission 9 states: Ironically, the design-basis contravention must deviate from design-basis, because even using the contravention, nuclear power plants cannot comply with GDC-19. This is why RG 1.183 and DG-1389 lowers design-basis standards. They credit the use of non-safety related and non-seismically qualified systems and componentsthat are not credited in design-basis analyses. Proffered inspections of non-safety related equipment (e.g., piping, condensers) do not satisfy legitimate design-basis analyses that can only credit safety-related equipment. RG-1389s seismically rugged is simply artifice; legitimate design-basis seismic analyses refer to these components as seismically unqualified.

As such, DG-1389 is contrary to NRC Regulatory Issue Summary 2001-19: Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests. It appears the NRCs RG 1.183 efforts are narrowly focused on obscuring known design deficienciesproviding nuclear power plants with questionable and unscientific methods to perform accident dose calculations so they can feign compliance with regulations; circumvent deterministic regulations; bring them into compliance or; otherwise increase their profit margins. I am opposed to RG 1.183 and DG-1389 because they provide the means to falsify accident dose calculations and feign compliance with nuclear safety regulations.

Comment Response The NRC staff disagrees with this comment. The comment stated that the Proffered inspections of non-safety related equipment (e.g., piping, condensers) do not satisfy legitimate design-basis analyses that can only credit safety-related equipment. The staff notes that RG 1.183, Revision 0, already provides an acceptable approach for crediting the condenser and pathways to the condenser in a licensees design basis dose calculations and that the NRC staff has reviewed and approved this credit for several licensees. The staff performs its review of such credit consistent with Regulatory Position 1.5, Submittal Requirements and Information, which states, Per 10 CFR 50.67, [t]he NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that the identified dose criteria are met. As discussed in the technical basis information added to Regulatory Position A-5.5 of RG 1.183, which was previously described in the now deleted Reference A-14, Draft Technical Assessment of Hold-up and Retention of Main Steam Isolation Valve Leakage Within the Main Steam Lines and Main Condenser, of DG-1389, the staff determined that the approach in RG 1.183, Revision 0, for demonstrating the seismic robustness of the condenser and pathways to the condenser included unnecessary conservatism because operating experience and past analyses demonstrated substantial seismic margin.

RIS 2001-019, Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests, dated October 18, 2001 46

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 (ML011860407), notes in the Summary of Issue section that the NRC staff generally does not accept analyses that credit plant features that are not safety-related unless they were previously accepted by the NRC staff in a site-specific licensing action. The provisions of Regulatory Position A-5.4 describe a means of establishing a pathway for hold-up and deposition of MSIV leakage within the main steam piping and main condenser, including either qualifying the pathway as seismic Category I and safety-related in accordance with RG 1.29, Seismic Design Classification for Nuclear Power Plants, or, for BWR operating license holders, demonstrating that the pathway is seismically robust. The provisions of Regulatory Position A-5.5, which provide an alternative risk-informed method for demonstrating seismic robustness of the condenser and pathways to the condenser, would be subject to NRC staff review and approval during a site-specific licensing action. Therefore, the staff finds Regulatory Positions A-5.4 and A-5.5 to be consistent with RIS 2001-019. The NRC staff revised RG 1.183, Revision 1, to include the technical basis supporting the methods for demonstrating seismic robustness of non-safety-related equipment and to indicate that these methods are applicable only to BWR operating license holders.

9-8 Magnuson Comment Page 435 of 448 of Comment Submission 9: The ICRP-103 evaluation confirmed the NRCs overriding concern that the implementation of the new ICRP 103 values could cause these plants to exceed the regulatory dose limits. Therefore, the NRC concluded: The NRC staffs decision to discontinuing the rulemaking activities associated with potential changes to the radiation protection and reactor effluents regulations was based on the knowledge that the current NRC regulatory framework continues to provide adequate protection of the health and safety of workers, the public, and the environment. The NRCs reason for rejecting ICRP-103 is my overring concern with RG 1.183 (Revision 0) and DG-1389.

Page 438 of 448 of Comment Submission 9 also discusses the use of ICRP-103 and its potential impact on control room habitability. It states:

The NRCs Office of Nuclear Regulatory Research completed a study entitled, Control Room Dose Evaluation Using ICRP 103 Dose Conversion Factors, letter report (ADAMS Accession No. ML17156A603), which concludes that:

Application of the ICRP 103 DCFs will result in an increase in the range of 23 to 25% in the TEDE doses for the control room. The degree of impact will depend on the amount of credit taken for various iodine removal mechanisms both natural and engineered. However, if the ICRP recommendations are ever incorporated into NRCs regulations and guidance, the incorporation of a thyroid weighting factor of 0.04 will decrease the already small margin many licensees have in their control room dose consequence analysis.

Comment Response This comment is out of scope, as the evaluation of ICRP-103, The 2007 Recommendations of the International Commission on Radiological Protection, issued 2007, is not related to the effort to prepare RG 1.183, Revision 1. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 9-9 Magnuson Comment Within the References section starting on page 429, pages 438 of 448 of Comment Submission 9, states: GDC-19 requires that, Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. The NRC has not emphasized the issue of control room access in any of the regulatory guides dealing with control room habitability. As such most licensees do not include an evaluation of access dose in their control room dose consequence analysis. Including access dose in the calculation of the total control room would decrease the already small margin most licensees have in their control room dose consequence analysis.

Comment Response In consideration of the comments, the NRC staff has removed the language associated with transit dose in Regulatory Position 4.2 and restored the original language from RG 1.183, Revision 0. The staff is evaluating this issue and will determine whether to address it in a future revision to this RG.

9-10 Magnuson Comment Page 158-159 of 448 of Comment Submission 9 states: NUREG-1465 (1995) Accident Source Terms for Light-Water Nuclear Power Plants: Recent information has indicated that high burnup fuel, that is, fuel irradiated at levels in excess of about 40 GWD/MTU, may be more prone to failure during design basis reactivity insertion accidents (RIA) than previously thought.

Preliminary indications are that high burnup fuel also may be in a highly fragmented or powdered form, so that failure of the cladding could result in a significant fraction of the fuel itself being released. The underlying concern identified here, is a cladding failure source term release could exceed that of a fuel melt source term release. What should be considered in RG 1.183-Revision 1, is the radiological consequences of a lessor and more likely accident may be worse than the maximum credible accident assumed in licensees current licensing bases.

Reports and studies (e.g., Resolution of Generic Safety Issues: Issue 170: Fuel Damage Criteria for High Burnup Fuel (Rev. 2)) have evaluated high burnup fuel and approved higher burn-up levels, but they have neither disputed the fuel disintegration caused by high-burnup nor evaluated the consequences of a powdered fuel source term. Until this NUREG-1465 concern has been eliminated, any revision to RG 1.183 should include a powdered fuel source term.

[A similar comment is also found on Page 167 of 448 of Comment Submission 9.]

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff disagrees with this comment but, in response, added clarifying guidance regarding fuel dispersal. The staff examined the effect of the high-burnup phenomena of fuel fragmentation, relocation, and dispersal (FFRD) on the DG-1389 MHA source terms and concluded that the effect would be minimal. The DG-1389 containment MHA source term was found to be bounding compared to equivalent analyses that explicitly consider the effects of FFRD. Additional details of this study can be found in the enclosure FFRD Impact on the Containment Source Term, dated June 15, 2021 (ML21197A069), to the Letter Report on Evaluation of the Impact of Fuel Fragmentation, Relocation, and Dispersal for the Radiological Design Basis Accidents in Regulatory Guide 1.183, dated July 20, 2021 (ML21197A067).

Furthermore, for non-LOCA events, DG-1389 states that fragmentation-induced transient fission gas release should be addressed, but guidance is not provided for non-LOCA events that have fuel dispersal. Additionally, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. The NRC staff added the following sentence to RG 1.183, Revision 1, Regulatory Position 3.2, to clarify the non-LOCA guidance in response to this comment: This RG does not provide guidance related to an acceptable treatment of fuel dispersal during non-LOCA DBAs.

9-11 Magnuson Comment Page 159 of 448 of Comment Submission 9 states: Notably, DG-1199 significantly increased Non-LOCA nobel gas release fractions (above RG 1.183 Revision 0) and returned them to NUREG-1465 levels. Excessive MISV leakage rates and realizations from the TMI accident prompted control room habitability studies and modifications to install Control Room Emergency Ventilation/Filter Systems. Subsequently, RG 1.183- Revision 0 required Control Room Operator doses to be evaluated for specific accidents, including the Non-LOCA fuel handling accident (FHA); however, missing from RG 1.183-Revision 0 is a requirement to evaluate doses to those workers/fuel handlers that would be in close proximity to this accident. Given the concerns identified in NRC Information Notice No. 90-08: KR-85 Hazards From Decayed Fuel and estimations based on FHA doses to control room operators, workers near spent fuel pools during would undoubtedly be overexposed (> 5 Rem TEDE).Because no amount of water in spent fuel pools will not prevent the release of noble gas (Kr [krypton]-85, a pure beta emitter) in a FHA, revisions to RG 1.183 should require the calculation of spent fuel pool doses to ensure workers are aware of the hazards. This calculation could also be used to ensure the viability of FLEX actions to intended to mitigate an extended loss of spent fuel pool cooling. [A similar comment is also found on Page 167 of 448 of Comment Submission 9.]

Pages 161-162 of 448 of Comment Submission 9 states: Prior to the accident at Three Mile Island (1979) and years afterward, control room operators were not protected by emergency air filtration systems. Operator doses from a DBA FHA (and other DBAs) were not publicly communicated because they exceeded General Design Criterion 19 limits (< 5 Rem whole body). After RG 1.183 was approved, the NRC required control room emergency filtration systems to be installed, and when their dose reduction factors were applied, operator doses 49

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 were restored to within the new limits of 10 CFR 50.67 (< 5 Rem TEDE). Even still, today control room operator doses are often the most limiting regulatory dose. While there may be margin regarding the iodine doses to control room operators, there is no margin regarding the Kr-85 doses in a DBA FHA. No amount of water in spent fuel pools will mitigate or prevent the release of Kr-85 in a FHA, and noble gasses cannot be filtered. Consideration of KR- 85 Hazards From Decayed Fuel (Information Notice No. 90-08) is conspicuously missing from RG 1.183-Revision 0. Any revision RG 1.183 should address IN 90-08 concerns and require that doses to fuel handlers/workers in the area of a FHA be calculated. [A similar comment is also found on Page 169 of 448 of Comment Submission 9.]

Comment Response The NRC staff disagrees with the comment. RG 1.183 provides a method acceptable to the NRC staff for demonstrating compliance with, for example, 10 CFR 50.67, but this method is not for evaluating worker doses near the spent fuel pool. Regulatory Position 4.2.1 lists sources of radiation that should be considered in the determination of exposure to control room personnel.

This list includes shine from radioactive materials external to the control room. Regulatory Position 3.4 states that table 6 includes a list of elements that should be considered in these design-basis analyses. Table 6 includes krypton. Therefore, krypton would be included in the fuel handling accident dose calculations. No changes were made in response to this comment.

9-12 Magnuson Comment Page 160 of 448 of Comment Submission 9 states: Does the NRC mean say LARs from last year (2019) cause a 11-year delay? DG-1199 (RG 1.183 Revision 1 Draft) was published by the NRC in 2009. In consideration of The NRC Approach to Open Government, please explain the 11-year delay. SAND2008-6601 clearly explains/illustrates that RG 1.183 MSIV Leakage source terms and metrologies are non-conservative and conceptually in error. Given this, why did the NRC approve the use of non-conservative and conceptually inaccurate guidance to increase MSIV leakage?

Pages 168-169 of 448 of Comment Submission 9 states: Does the NRC mean say LARs from last year (2019) cause a 10-year delay? DG-1199 was approved (but not issued) by the NRC in 2010. In consideration of The NRC Approach to Open Government, please explain the 10-year delay. Because SAND2008-6601 clearly explains/illustrates that RG 1.183 MSIV Leakage source terms and metrologies are non-conservative and conceptually in error, it does not seem that LARs to increase MSIV leakage are in the best interest of public health and safety.

Page 160 and 168 of 448 of Comment Submission 9 states: SAND2008-6601 determined RG 1.183 BWR MSIV leakage source terms and methodologies are non-conservative and conceptually in error. These conceptual errors (and others) should be corrected in any revision to RG 1.183.

Page 158 of 448 of Comment Submission 9 states: DG-1199 (Draft RG 1.183 Revision 1) was the first effort to revise RG 1.183. It was prompted by SAND2008-6601 and published by the 50

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 NRC in 2009; however, it was never implemented. After eleven years, what prompted this effort?

Page 167 of 448 of Comment Submission 9 states: DG-1199 (Draft RG 1.183 Revision 1) was approved (but not issued) by the NRC in 2010. After ten years, what prompted this effort?

Comment Response The NRC staff disagrees with the comment. To note, regarding the time period between DG-1199 and DG-1389, the license amendment requests listed in this comment did not exist 11 years ago, and therefore the staff did not intend to imply that these license amendment requests caused an 11-year gap in providing the draft of Revision 1 to RG 1.183, published in 2022 (DG-1389).

Regarding the request for an explanation of the 11-year gap, the following is provided.

After the issuance of DG-1199, the staff received a number of public comments and spent significant effort in addressing those comments, including resolving different NRC staff views on the approach to responding to certain public comments. The efforts included soliciting an independent review by Sandia National Laboratories (SNL) of certain aspects of DG-1199. In 2017, the NRC staff received final responses from SNL associated with its independent review (ML19094A465). The NRC staff resumed work on RG 1.183, Revision 1, in late 2020 after considering the significant insights gained since the initial issuance of DG-1199, including the following:

2017 SNL responses

the 2019 Commissions approval of a revision to Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, dated September 20, 2019 (ML18093B087), as stated in SRM-SECY-18-0049, Staff RequirementsSECY-18-0049Management Directive and Handbook 8.4, Management of Backfitting, Issue Finality, and Information Collection, dated May 29, 2019 (ML19149A294 (package))

lessons learned from the four main steam isolation valve safety evaluations completed in 2020 (see ML20140A070, ML20150A328, ML20241A190, and ML20265A240)

research pertaining to state-of-the-art source term knowledge, such as the fuel fragmentation, relocation and dispersal.

Regarding the assertion in SAND2008-6601, which determined the BWR MSIV containment aerosol concentrations provided in RG 1.183, Revision 0, are non-conservative and conceptually inaccurate, please see DPO-2020-002 (ML21067A645), which disputed that the methodology was non-conservative and conceptually inaccurate. The assumptions used in SAND2008-6601 did not credit the mixing of the deterministic fuel melt source term for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the accident evaluation period. After 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the evaluation credited safety-related systems to mix the source term throughout the containment. DPO-2020-002 disputed this delayed period for the crediting of safety-related systems. The case file for DPO-2020-002 includes the memorandum from the director of the Office of Nuclear Regulatory Research, which recommended the following:

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023

[The Office of Nuclear Reactor Regulation] should update guidance to allow credit for safety-related systems to distribute the deterministic fuel melt source into the containment atmosphere when assessing the initial dose consequence from main steam isolation valve leakage.

The staff incorporated this recommendation (that these safety-related systems should be credited) into DG-1389 (see DG-1389, footnote 2).

The comments question as to why did the NRC approve the use of nonconservative inaccurate guidance to increase MSIV leakage is unclear but appears to refer to guidance used for the approval of the cited amendments. Addressing the basis for the NRC staffs approval for those license amendments pertains to issues related to those amendments and not to any specific issue associated with the regulatory guidance update. Information on the basis for the NRC staffs approval is contained in the agencys safety evaluation for those license amendments. As a result, no changes were made in response to these comments and questions.

9-13 Magnuson Comment Page 161 of 448 of Comment Submission 9 states: DG-1199 was prompted by SAND2008-6601, which determined RG 1.183-Revision 0 source terms and methodologies are conceptually inaccurate. The intent of DG-1199 was to correct the fundamental errors in RG 1.183-Revision 0. Is this still the intent of RG 1.183-Revision 1?

Comment Response The NRC staff agrees with the request to answer the question provided. The intent of the proposed RG 1.183, Revision 1, is to incorporate over 20 years of experience and lessons learned since the release of RG 1.183, Revision 0. Notably, the Reason for Revision in section B of the RG provides examples of the updates and revisions made as part of this revision. No changes were made in response to this question.

9-14 Magnuson Comment Page 161 of 448 of Comment Submission 9 states: RG 1.183 states: The design basis accident source term is a fundamental assumption upon which a significant portion of the facility design is based. Considering the significance of the accident source term, why would the NRC continue to allow licensees to use RG 1.183-Revision 0? Is not negligent to allow licensees to base nuclear power safety (systems) on conceptually inaccurate and non-conservative accident source terms?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Page 164 of 448 of Comment Submission 9 states: The design basis accident source term is a fundamental assumption upon which a significant portion of every nuclear power plant design is based; therefore, RG 1.183-Revision 0 is, essentially, a generic safety issue. The NRCs failure to act on this fundamental safety issue prompted PRM-50-12210 CFR Part 2.802 request for rulemaking.

Page 169 of 448 of Comment Submission 9 states: According to RG 1.183, The design basis accident source term is a fundamental assumption upon which a significant portion of the facility design is based. Given this and SAND2008-6601, how does the existence (coexistence) and continued use of the non-conservative and conceptual errors in RG 1.183 benefit the health and safety of the public?

Page 162 of 448 of Comment Submission 9 states: SAND2008-6601 is a scientific study performed by Sandia National Laboratories on behalf of the NRC that clearly explains/illustrates that RG 1.183 BWR MSIV source terms and metrologies are non-conservative and conceptually in error. It is the technical basis for the proposed DG-1199 MSIV modeling changes. Nuclear power plant owners (licensees) have not adopted SAND2008-6601 (and have resisted DG-1199) because it is unlikely that they can comply with 10 CFR 50.67 if accurate MSIV leakage models and source terms are used. Please refer to the following January 2010 letters. [See pages 162-163 of Comment Submission 9 for the letters identified and quoted in the comment.]

Page 163 of 448 of Comment Submission 9 states: Has the NRC disavowed SAND2008-6601?

If not, why has the NRC allowed licensees to use non-conservative and conceptually inaccurate MSIV leakage models and source terms for the past ten years? If not, why would the NRC allow RG 1.183-Revision 0 to co-exist with RG 1.183-Revision 1?

Page 164 of 448 of Comment Submission 9 states: During the RG 1.183 public meeting on November 19, 2020, an industry member commented that the incorrect methods, described in RG 1.183, to calculate the radiological dose consequences, were used to assess Operability of structures, systems and components required by plant Technical Specifications. Again, why would the NRC allow RG 1.183-Revision 0 to co-exist with RG1.183-Revision 1?

Page 160 of 448 of Comment Submission 9 states: With known, fundamental errors in RG 1.183-Revision 0, why would the NRC allow it to co-exist?

Pages 169-170 of 448 of Comment Submission 9 states: SAND2008-6601 clearly explains/illustrates that RG 1.183 BWR MSIV source terms and metrologies are non-conservative and conceptually in error. It identifies a safety concern (with a complex array of regulatory implications); however, this concern was not enough to motivate nuclear power plant owners/operators to adopt SAND2008-6601 or otherwise correct the nonconservative errors in RG 1.183that adversely affect the health and safety of the public. This is the crux of the matter and the reason for PRM-50-122.

Comment Response Please see the NRC staffs responses to Public Comments 9-12 and 15-6. No changes were made in response to this comment. For the NRC staffs response to the comments overall question on the use of SAND2008-6601 results and the assertion that the RG 1.183 source term and methodology are nonconservative and conceptually inaccurate, please see the 53

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 response to Public Comment 9-12. Regarding whether SAND2008-6601 was disavowed, SAND2008-6601 was used to add acceptable main steamline removal coefficients to RG 1.183, Revision 1; however, the source term insights from SAND2008-6601 were not used because of issues discussed in the response to Public Comment 9-12 that are related to DPO-2020-002.

Regarding the referenced PRM-50-122, the NRC denied that petition for rulemaking in July 2022 (see ML22196A047 (package) and 87 FR 44281; July 26, 2022).

9-15 Magnuson Comment Page 167 of 448 of Comment Submission 9 states: Insights from Fukushima were previously incorporated into RASCAL (NUREG-1430) source terms and methodologies. Will these same insights be incorporated into RG 1.183 Revision 1? Why is the revision to RG 1.183 lagging behind revisions to RASCAL? Also, please explain why RASCAL does not use RG 1.183 source terms and methodologies. [A similar comment is also found on page 158 of Comment Submission 9.]

Comment Response The NRC staff disagrees with this comment as the computer code RASCAL (Radiological Assessment System for Consequence Analysis for radiological emergencies; see NUREG-1940, RASCAL 4: Description of Models and Methods, issued December 2012 (ML13031A448)), and RG 1.183 are used for different regulatory purposes. Both, however, use the 10 CFR 100.11 description of the source term magnitude. RG 1.183 is used to assess the acceptability of safety-related SSCs designed to mitigate consequences. The assessments which utilize RG 1.183 are classic engineering analyses. RASCAL is used to assess emergency response to a release of radioactive material to the environment (not the design of safety-related SSCs to mitigate the source term). Because the purposes of RG 1.183 and RASCAL are different, changes made in RG 1.183 may not be appropriate for incorporation into RASCAL.

Historically, the development of RASCAL and supporting analysis follow the source term adoptions of RG 1.183 and its previous iterations (RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, issued May 2003 (ML031490640); and the former RG 1.3, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, and RG 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, both withdrawn December 2016).

There are several versions of this particular regulatory source term. All modern versions use the NRC/SNL code MELCOR to derive it (NUREG-1465, NUREG-1935 (SOARCA),

SAND2011-0128, post-Fukushima work).

No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Submission 10 ADAMS Accession No. ML22174A071 Name: Carrie Fosaaen, Director, Regulatory Affairs Organization: NuScale Power, LLC Email: areeve@nuscalepower.com Address: 1100 NE Circle Blvd., Suite 200, Corvallis, Oregon 97330 10-1 NuScale Comment The document purpose states that this guide establishes an AST based in part on SAND-2011-0128, Accident Source Terms for Light Water Nuclear Power Plants Using High Burnup of [sic] MOX Fuel. Section 1.1 states as used in this guide, the AST is an accident source term that is derived principally from SAND-2011-0128 and differs from the TID-14844 and NUREG-1465 source terms used in the original and revised design and licensing of operating reactor facilities. Section B, Background, states Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations. Taken together these comments imply that Revision 1 applies to high-burnup fuel (only) while Revision 0 continues to apply to low-burnup applications. The RG should be clarified to discuss the scope of SAND-2011-0128 and its use within RG 1.183 (covering both LBU and HBU applications) and the relationship to and continued availability of NUREG- 465 as implemented by Revision 0.

Commenters Proposed Resolution The Purpose, Reason for Revision, and Background should be clarified to discuss the scope of SAND-2011-0128 and its use within RG 1.183 (covering both LBU and HBU applications) and the relationship to and continued availability of NUREG-1465 as implemented by Revision 0.

The statement that Revision 0 remains available should clarify the rationale for such, and that continued availability should be elevated beyond the background portion of the RG. Section A, Applicability, should explicitly define applicability to both LBU and HBU applications, and note continued availability of Revision 0.

Comment Response The NRC staff agrees, in part, with this comment. The staff revised Section A, Purpose, to clarify that RG 1.183, Revision 1, source terms are applicable to low-burnup and high-burnup LWR fuels. Information on the scope and use of the SAND2011-0128 report as a basis for the source term is provided in the relevant staff regulatory guidance in section C of the DG; therefore, the staff did not find it necessary to add the requested additional information in the Background discussion in response to this comment.

Additionally, the NRC staffs response to Public Comment 11-2 discusses the continued availability of RG 1.183, Revision 0, and both Revision 0 and Revision 1 being acceptable for use, which resulted in a change to Regulatory Position C-1.1.5, Applicability to New Light-Water Reactor Applications, Including Advanced Evolutionary and Passive Designs, of RG 1.183, Revision 1. The NRC staff added information to the Background section of RG 1.183, Revision 1, on the applicability of Revision 0 for an increase in burnup above 55

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 62,000 megawatt-days per metric ton of uranium (MWd/MTU) and 5 weight-percent uranium-235 enrichment in response to Public Comments 11-2 and 13-2. Both Revision 0 and Revision 1 of RG 1.183 will be available for use, because Revision 0 will not be withdrawn.

Each revision provides a method acceptable to the NRC staff for compliance with the applicable regulations specified in the guidance. Use of combinations and permutations of regulatory positions from Revision 0 and Revision 1 will need appropriate technical justification and may require additional NRC review before approval.

10-2 NuScale Comment The document purpose states that this guide establishes an AST based in part on SAND-2011-0128, Accident Source Terms for Light Water Nuclear Power Plants Using High Burnup of [sic] MOX Fuel. Section 3.2 states that the release fractions are not endorsed for MOX fuel. The applicability of this RG for MOX fuel should be addressed up front in the Purpose and Applicability discussions of Section A.

Commenters Proposed Resolution Address RG applicability for MOX fuel in the Purpose and Applicability of Section A.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position 3.2.

The NRC staff agrees with this comment. The staff revised RG 1.183, Revision 1, section A, to include the following language:

The updated source terms reflect the current understanding of severe accidents and fission product behavior since the publication of RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, issued July 2000, (Ref 2), and include low-burnup and high-burnup LWR fuels. The use of these source terms is not endorsed for mixed oxide fuels.

10-3 NuScale Comment The first sentence of the second bullet identifies the five active subparts of 10 CFR 52: early site permits (ESP), standard design certifications (DC), combined licenses (COL), standard design approvals (SDA), and manufacturing licenses. The second sentence discussing the requirement to evaluate offsite radiological consequences only discusses two of the five subparts: DC 56

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 (52.47) and COL (52.79). It is not clear why only these two subparts are addressed and the other three are omitted. ESP (52.17), SDA (52.137), and ML (52.157) all contain equivalent requirements.

Commenters Proposed Resolution Include references to 10 CFR 52.17, 10 CFR 52.137, and 52.157 in the second sentence of the second bullet.

Comment Response The NRC staff agrees with this comment. To provide clarity, the staff revised the bullet points under the item for 10 CFR Part 52 in Applicable Regulations within section A of RG 1.183, Revision 1, to include citations of all applicable requirements in 10 CFR Part 52.

10-4 NuScale Comment The reason provided for revision of the guide is to address new issues since the original guide was issued. The issues are then listed. Although the list is helpful to understand what changed, in the case of technical changes to the previous guidance, no information is provided as to the need for the specific changese.g., it is not clear if the changes were made to address non-conservatisms in the old guidance or to eliminate over-conservatisms in the old guidance.

Although the Revision 0 of RG 1.183 will remain available for use, a user needs to be apprised of the purpose of the changes in order to determine which version is appropriate for their use.

As an example, the new guidance makes small changes to some release fractions and large changes to others. Some of the changes are increases and some are decreases. The intent of the overall net impact of these release fractions changes is not clear. It would also be helpful to know whether the cumulative impact of all the changes in the new guidance would be expected to calculate an increase or decrease in dose relative to the old guidance.

Commenters Proposed Resolution For each of the listed technical changes, the reason for revision should be clearly stated. The reasoning should indicate whether the new guidance is more or less conservative than the prior guidance. If the new guidance is simply more accurate than the old guidance and both are conservative, then that should be stated.

Comment Response The NRC staff does not agree with this comment. While the staff appreciates that additional information explaining differences between this version of the draft guidance and the previous RG may be beneficial to certain users of the RG, the purpose of this RG is to provide one acceptable method for complying with the regulations. Because the staff is not withdrawing the previous version of the RG, applicants and licensees would be free to choose either, if applicable. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 10-5 NuScale Comment Footnote 1 provides a definition of a maximum hypothetical accident (MHA). In the definition, the MHA is also referred to parenthetically as a maximum credible accident (MCA). The MHA has historically been referred to as the MCA in some contexts (e.g. TID-14844), however the context of the statement might lead users to believe that the MHA LOCA is itself credible.

Further, the purpose of the MHA in relation to other regulatory requirements should be clarified.

Commenters Proposed Resolution Revise footnote 1 to clarify. Suggested revision, in marked-up form, is as follows:

The maximum hypothetical accident (MHA) (also referred to as the maximum credible accident) is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any accident whose occurrence during the lifetime of the facility would appear to be credible (note: historical references such as TID-14844 may refer to the MHA as the maximum credible accident, although this terminology does not indicate the MHA is itself credible). As used in this guide, the term LOCA refers to any accident that causes a loss of core cooling. The MHA LOCA refers to a loss of core cooling resulting in substantial meltdown of the core with subsequent release into containment of appreciable quantities of fission products. Although the design basis of a facilitys ECCS is required to prevent a LOCA from resulting in substantial meltdown of the core, the MHA evaluation is conducted to assess the performance of containment, fission product control systems, and site characteristics in providing defense-in-depth against a more severe accident than allowable within the plants design basis. These evaluations assume containment integrity with offsite hazards evaluated based on design basis containment leakage.

Comment Response The NRC staff disagrees with this comment. Please refer to the staffs response to Public Comment 4-1 for more information on the MHA LOCA. The NRC staff considers the MHA LOCA to be a credible DBA. Although the design basis of a facilitys ECCS is required to prevent a LOCA from resulting in substantial meltdown of the core, the MHA evaluation is, in part, conducted to assess the performance of containment, fission product control systems, and site characteristics in providing defense in depth against a more severe accident than allowable within the plants ECCS design basis.

The MHA may be used to show compliance with 10 CFR 50.34, footnote 6; 10 CFR 50.67, footnote 1; and 10 CFR 100.11, footnote 1, as well as similar footnotes in 10 CFR Part 52.

Although the regulations include consideration of plant systems, structures, and components and site characteristics, they do not provide prescriptive requirements on how the transport of fission products from the core to the environment should be modeled. Instead, guidance in RG 1.183, Revision 1, provides an acceptable source term, methods, and assumptions for modeling the plant response and transport of radionuclides to the environment and calculation 58

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 of radiological consequences for demonstration of compliance with the radiological performance criteria. No changes were made in response to this comment.

10-6 NuScale Comment The third paragraph contrasts the whole body and thyroid dose criteria of 10 CFR 100.11 with the total effective dose equivalent criteria in 10 CFR 50.34, 10 CFR 52, and 10 CFR 50.67. In addition to those regulations, 10 CFR 100.21 provides a connection from Part 100 to 10 CFR 50.34 for newer plants. It may be beneficial to identify 10 CFR 100.21 in the list of applicable regulations.

Commenters Proposed Resolution Add a reference to 10 CFR 100.21 where applicable.

Comment Response For context, this public comment refers to the third paragraph on page 6 of DG-1389. The NRC staff agrees with the comment that it is beneficial to identify the nonseismic siting requirement in 10 CFR 100.21, Non-seismic siting criteria, where relevant. However, the staff disagrees that the specific location referenced in the comment (the Background section of section B) requires clarifying changes because the TEDE criteria are not included directly in 10 CFR 100.21.

Section B of RG 1.183, Revision 1, discusses the basis for guidance documents that were developed before the issuance of 10 CFR 100.21 and the inconsistency of that guidance with the TEDE criterion. The regulation in 10 CFR 100.21 refers to the criteria in 10 CFR 50.34(a)(1),

which are already included in the list of applicable regulations in section A of RG 1.183, Revision 1. In response to Public Comment 10-8, the NRC staff clarified the siting requirements in Regulatory Position 1.3.1 of RG 1.183, Revision 1, including the addition of 10 CFR 100.21.

The NRC staff revised Regulatory Position 1.3.1, item f, to include a reference to 10 CFR 100.21 in response to this comment.

10-7 NuScale Comment Section 1.1 states ASTs may be used for advanced LWRs under 10 CFR Part 50 and 10 CFR Part 52 and for operating reactors under 10 CFR 50.34 and 10 CFR 50.67. Advanced in this context seems to refer to new reactors; advanced implies a more limited applicability of ASTs.

Commenters Proposed Resolution Replace advanced with new.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response Please note that the reference to Section 1.1 in the comment is referring to Regulatory Position 1.1.

The NRC staff agrees with the comment, in that the general applicability of the guidance is to any new power plant applications for an LWR design. The staff also notes that new LWR applications may include advanced evolutionary and passive LWR designs. Evolutionary and passive LWR designs have been used previously in the context of referring to advanced LWRs (see, for example, the discussion in SECY-94-302, Source Term-Related Technical and Licensing Issues Pertaining to Evolutionary and Passive Light-Water-Reactor Designs, dated December 19, 1994 (ML003708141)). To address this comment, the NRC staff incorporated changes into RG 1.183, Revision 1, in the Related Guidance part of section A; in the Reason for Revision and Background parts of section B; and in section C, Regulatory Positions 1.1 and 1.1.5, to clarify that the guidance can be used for new LWR applications, including those for advanced evolutionary or passive LWR designs.

10-8 NuScale Comment Subsection 1.3.1 provides a list of regulatory requirements. The text indicates that the list is not all inclusive. However, the list neglects to include significant references, such as 10 CFR 50.34(a)(1) and 10 CFR 100.21.

Commenters Proposed Resolution Add 10 CFR 50.34(a)(1) and 10 CFR 100.21 to the list of regulatory requirements.

Comment Response Please note that the reference to Subsection 1.3.1 in the comment is referring to Regulatory Position 1.3.1.

The NRC staff agrees with this comment. The staff revised RG 1.183, Revision 1, section C, Regulatory Position 1.3.1, items f and g, to include citations to applicable requirements for power reactor license applications to provide clarity and completeness. Additionally, the NRC staff clarified that the guidance in RG 1.183, Revision 1, is relevant to power reactor siting (item f) and new power reactor applications under 10 CFR Part 50 (item g).

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 10-9 NuScale Comment Item g of the list in Subsection 1.3.1 is related to 10 CFR 52. However, the item only includes three of the five active subparts of 10 CFR 52 (ESP, DC, COL). The other two active subparts of 10 CFR 52 (SDA and ML) should be added to Item g.

Commenters Proposed Resolution Add SDA and ML to Item g.

Comment Response Please note that the reference to Subsection 1.3.1 in the comment is referring to Regulatory Position 1.3.1.

The NRC staff agrees with this comment. The staff revised section C, Regulatory Position 1.3.1, item g, to include standard design approvals and manufacturing licenses to provide clarity and completeness, because they are included in 10 CFR Part 52 as nuclear power facility licenses.

10-10 NuScale Comment Subsection 3.1 states that the core inventory should be determined using an appropriate isotope generation and depletion computer code. The previous revision of this RG provided examples of computer codes that met this guidance. It is not clear why the names of the computer codes have been removed. Does the removal of the examples imply that the computer codes in the previous revision of the RG are no longer adequate? Or is the intent of eliminating the examples to provide flexibility to use other computer codes?

Commenters Proposed Resolution Clarify the guidance regarding what is an appropriate computer code for depletion.

Comment Response Please note that the reference to Subsection 3.1 in the comment is referring to Regulatory Position 3.1.

The NRC staff disagrees with this comment. The staff did not include specific examples of isotope generation and depletion codes in RG 1.183, Revision 1, to allow the licensee or applicant flexibility in the choice of specific codes used and also to avoid the case in which the computer code referenced as an example becomes unsupported by the developer or out of date during the time that the RG is in use by the licensee or applicant, as has occurred previously.

No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 10-11 NuScale Comment Subsection 3.2 provides tables of release fractions. These tables are easy to use and relatively similar to the previous revision. However, the text indicates that the actual release should be based on the tables plus any TFGR prompted by the accident conditions. Correlations are provided to determine TFGR as a function of burnup and increase in radial average fuel enthalpy. The guidance indicates that non-LOCA DBAs such as fuel handling accidents should consider TFGR. Transient analyses are typically not performed for fuel handling accidents and increase in enthalpy is not applicable. There is no guidance provided for determining TFGR for fuel handling accidents. The guidance for TFGR appears to be incomplete.

Commenters Proposed Resolution Clarify the applicability of TFGR to the various design basis events which consider dose consequences. Provide guidance for determining TFGR for fuel handling accidents.

Comment Response Please note that the reference to Subsection 3.2 in the comment is referring to Regulatory Position 3.2.

The NRC staff agrees in part, and disagrees in part, with this comment. As stated in RG 1.183, Revision 1, Regulatory Position 3.2, TFGR should be addressed [f]or the remaining non-LOCA DBAs that predict fuel rod cladding failure. Please see the staffs response to Public Comment 8-2 for more details and changes to RG 1.183, Revision 1, concerning additional TFGR guidance. In response to this comment, the NRC staff incorporated changes into Regulatory Position 3.2 of RG 1.183, Revision 1, to clarify the guidance on TFGR in response to this comment.

10-12 NuScale Comment Subsection 3.3 provides a table with the onset time and duration of release phases. Table 5 indicates that the duration of the gap release phase is 0.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months for PWRs and 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months for BWRs. The paragraph of text below Table 5 concludes with a statement that the duration of the gap release phase is 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . This appears to be an inconsistency that causes confusion about what gap release duration should be used.

Commenters Proposed Resolution Clarify the relationship between the durations in Table 5 and the durations in the text. Clearly identify the applicable duration.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response Please note that the reference to Subsection 3.3 in the comment is referring to Regulatory Position 3.3.

The NRC staff agrees with this comment. The staff revised table 5 to add clarity by changing the onset times for the early in-vessel release phase to include the onset time for the gap release phase. The NRC staff also deleted the sentence in the paragraph below table 5 regarding the duration of the release.

10-13 NuScale Comment Subsection 4.2 introduces the concept of transit dose. The guidance indicates that the licensing basis for some licensees may include transit dose and that those licensees should include the transit dose when evaluating the 5 rem TEDE control room dose limit associated with 10 CFR 50.67. The concept of transit dose is not found in the supporting regulations. The regulatory basis for this concept is not provided. It is also not clear why it would be applicable to some licensees and not to others. In addition, the only guidance provided for evaluation of this transit dose is on a case-by-case basis. This RG is intended to also provide useful information for new reactor applicants and this paragraph addressing transit dose ends with a statement that new reactor licensees may use this guidance in demonstrating CR habitability. Therefore, it is important to clarify whether evaluation of transit dose is necessary, with appropriate regulatory basis, and if so to provide prescriptive guidance on including transit dose. Further, clarify whether the statement that new reactor licensees may use this guidance refers to the preceding discussion of transit dose or to the remaining discussion of Subsection 4.2.

Commenters Proposed Resolution The regulatory basis, if any, for the concept of transit dose should be provided. If there is none, then the concept should be clearly confined to those existing licensees for which transit dose is part of their licensing basis. If there is a basis to apply the guidance to new reactor applicants, then the guidance should provide prescriptive instructions for evaluating transit dose. In addition, the end of the paragraph should clarify the meaning of this guidance with respect to new reactor licensees.

Comment Response Please note that the reference to Subsection 4.2 in the comment is referring to Regulatory Position 4.2.

In consideration of the comments, the NRC staff has removed the language associated with transit dose in Regulatory Position 4.2 and restored the original language from RG 1.183, Revision 0. The staff is evaluating this issue and will determine whether to address it in a future revision to this RG.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 10-14 NuScale Comment Footnote 15 states that the nuclides used for modeling dose from airborne radioactivity inside the control room may not be conservative for determining the dose from radioactivity outside the control room. No guidance is provided for how it is determined whether or not the nuclide set is conservative. Similarly, no guidance is provided for determining an alternative set of nuclides that would be conservative. If it is not acceptable to use the set of nuclides from inside the control room, then guidance should be provided for what is acceptable.

Commenters Proposed Resolution Footnote 15 should be removed or else revised to provide guidance as to what is an acceptable set of nuclides for determining dose outside the control room.

Comment Response The NRC staff agrees with the comment and removed footnote 15, which was included in DG-1389.

10-15 NuScale Comment Subsection 5.3 states that RG 1.145 and another reference document methodologies used in the past for determining atmospheric dispersion (X/Q) values. The previous revision of this RG provided examples of computer codes that met this guidance. It is not clear why the names of the computer codes have been removed. In addition, it is not clear what is meant by methodologies used in the past. It seems to imply that the methodologies are out-of-date and possibly should not be used in future. If this is not the intent, the guidance should be clarified to indicate that these methodologies used in the past are still acceptable for current and future use.

Commenters Proposed Resolution Clarify guidance regarding what is an appropriate computer code for dispersion. Clarify whether previous guidance in RG 1.145 is still acceptable.

Comment Response Please note that the reference to Subsection 5.3 in the comment is referring to Regulatory Position 5.3.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff agrees with the comment but disagrees with the proposed resolution. In response to this comment, the staff did clarify the regulatory guidance in RG 1.183, Revision 1, by deleting in the past from the first paragraph of Regulatory Position 5.3 to avoid the perception that the methodologies are out of date and should not be used in the future.

Regarding the comments request to clarify whether the previous guidance in RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, is still acceptable, the second paragraph of Regulatory Position 5.3 states, RG 1.145 and RG 1.194 should be used if the FSAR /Q values are to be revised. This statement is intended to confirm that RG 1.145 remains an acceptable method.

Additionally, dispersion codes are not named in DG-1389, because RG 1.145 and RG 1.194 list methodologies acceptable to meet the regulations. In response to this comment, the NRC staff revised RG 1.183, Revision 1, as noted.

10-16 NuScale Comment Subsection B-2 provides very detailed guidance for determining the decontamination factor (DF) for the fuel handling accident. The very detailed guidance has a very narrow applicability of water depths (19 to 23 ft). No guidance is given for depths less than 19 ft or greater than 23 ft.

The previous guidance gave a bounding DF which could be used for all depths greater than or equal to 23 ft. As a result, many licensees and new reactor applicants rely on design features to ensure that the depth is at least 23 ft. Typically, the depth is larger than 23 ft, as more water is thought to be better for shielding purposes, etc. Under the new guidance, licensees and new reactor applicants with a depth of greater than 23 ft would have to justify the selection of DF on a case-by-case basis. This new guidance appears to encourage licensees and new reactor applicants to decrease the water depth to less than 23 ft in order to fall in the applicable range of the equations. This seems nonconservative. The old guidance, which provides a bounding DF for depths greater than 23 feet, is much more straightforward and easy to apply. In addition, the equation provided for depths between 19 and 23 feet is a complicated function of pin pressure that is not easily comparable to the old guidance. The equation has a discontinuity around 5000 psig. For pin pressures less than 5000 psig, the minimum DF calculated by the equation appears to be approximately 250. If the new equation always yields values greater than 250 (for pin pressures less than 5000 psig), then the guidance should indicate that a value of 250 may be used in lieu of calculating pin pressures and using the equation. This would also confirm that the prior guidance was conservative.

Commenters Proposed Resolution The guidance should provide a bounding DF that can be used as an alternative to calculating DF based on pin pressure. The guidance should also be expanded to include depths greater than 23 ft, as this is the most common depth designed for by licensees and new reactor applicants.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response Please note that the reference to Subsection B-2 in the comment is referring to Regulatory Position B-2.

The NRC staff disagrees with the comment that the Regulatory Position B-2 guidance needs to address depths greater than 23 feet and less than 19 feet, and that a bounding decontamination factor (DF) needs to be provided. RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. To note, the model described in the NRC staff technical paper Re-evaluation of the Fission Product Release and Transport for the Design-Basis Accident Fuel Handling Accident, issued December 2019 (ML19248C647), is based on the boundary conditions of experiments with a pool depth of 23 feet. Depths greater than 23 feet are, therefore, not addressed in the guidance. However, for depths greater than 23 feet, a conservative approach exists to use the DFs at 23 feet or alternatively justify DFs based on applicable experiments. While depths less than 19 feet are outside of the range of the model applicability and that of the model developed by the NRC staff, depths less than 19 feet are very uncommon. Additionally, the NRC staff notes that a bounding DF can be calculated using the plant-specific design parameters in RG 1.183, Revision 1. No changes were made in response to this comment.

10-17 NuScale Comment Appendices E and F are provided for the PWR SGTR and PWR MSLB events, respectively. In the previous revision of the guidance, these two Appendices were exactly reversed, with Appendix E being for the PWR MSLB event and Appendix F being for the PWR SGTR event.

The Appendices for the event-specific guidance for all other events is maintained consistent between the two revisions. It is not clear the benefit of switching the order of Appendices E and F. In addition, it may cause confusion to licensees and applicants who are accustomed to the Appendices being aligned to specific events.

Commenters Proposed Resolution Reverse the order of Appendices E and F so that consistency with the order of the previous revision is maintained.

Comment Response The NRC staff disagrees with reversing the order of appendices E and F to maintain consistency with RG 1.183, Revision 0. The staff revised the order of appendices E and F in RG 1.183, Revision 1, to be consistent with similar guidance in RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, issued May 2003 (ML031490640). To note, RG 1.195 was also used for modeling the same accidents as those in RG 1.183, Revision 1. For awareness, the titles of these appendices do identify the accident to which the regulatory positions pertain for 66

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 modeling (e.g., Appendix E, Assumptions for Evaluating the Radiological Consequences of a Pressurized-Water Reactor Steam Generator Tube Rupture Accident). No changes were made in response to this comment.

10-18 NuScale Comment In the prior guidance, the MSLB transport model was based on a figure included with the MSLB appendix. The figure was similar to the figure provided as Figure E-1 in the SGTR appendix of the new guidance. The new guidance is not clear on whether the new Figure E-1 applies to the MSLB transport model or not. If it does apply, the cross-reference should be included. If it does not apply, no explanation is provided for why it no longer applies.

Commenters Proposed Resolution Clarify whether the new Figure E-1 applies to the MSLB transport model or not and provide either the applicable cross-reference or the explanation for why it no longer applies.

Comment Response The NRC staff does not agree with the comment. The order of appendices E and F, provided in RG 1.183, Revision 0, were revised in DG-1389 to be consistent with similar guidance in RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors, issued May 2003 (ML031490640),

which does not contain or reference a figure in appendix F for the MSLB accident. For consistency, RG 1.183, Revision 1, does not contain or reference a figure in appendix F for the MSLB accident. Notably, in RG 1.183, Revision 1, Regulatory Positions F-6 through F-6.6.3 contain the MSLB accident transport model guidance, so cross-referencing would not be necessary. No changes were made in response to this comment.

Comment Submission 11 ADAMS Accession No. ML22174A072 Name: Frances Pimentel, Senior Project Manager Organization: Nuclear Energy Institute (NEI)

Email: atb@nei.org Address: 1201 F Street, NW, Suite 1100, Washington, DC 20004 11-1 Nuclear Energy Institute Comment The NRC has identified that licensees may elect to use risk insights in support of proposed changes to the design basis that are not addressed in currently approved NRC staff positions.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC should provide some more explanation on the parameters/limitations of using risk insights.

Commenters Proposed Resolution NRC should provide some additional explanation on the acceptable uses of risk insight in conjunction with the radiological analyses performed with this RG.

Comment Response The NRC staff disagrees with the comment. The methods, techniques, or data described in an RG provide a method deemed acceptable to the NRC staff for compliance with the applicable regulations. However, applicants or licensees may choose to use alternative approaches if the applicants or licensees provide sufficient information that demonstrates the requirements in the NRCs regulations are satisfied. To note, Regulatory Position 1.4 of RG 1.183, Revision 1, does include references to both RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, issued January 2018, and LIC-206, Integrated Risk-Informed Decision Making for Licensing Reviews, dated June 26, 2021 (ML21167A142) (regarding how the staff performs integrated risk-informed decision-making for licensing reviews), for additional guidance related to risk implications.

No changes were made in response to this comment.

11-2 Nuclear Energy Institute Comment The last paragraph of the section states: Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations. A combination of the methods contained in Revision 0 or Revision 1 of RG 1.183 would need additional justification. The NRC should provide additional clarifications in this paragraph (or Section 1.2) to inform the licensee as to acceptable instances for combination of Revision 0 and 1 methods.

Commenters Proposed Resolution NRC should include a statement like the following: For example, across a licensing basis, different revision methods may be adopted provided the individual analyses fully adopt a single revision (i.e., individual analysis inputs, assumptions, or modeling aspects may not adopt different revision methods within the same analysis). Additional details should also be included in Section 1.2Scope of Implementation.

Comment Response Please note that the reference to Section 1.2 in the comment is referring to Regulatory Position 1.2.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff agrees with the comments request that an example be added to RG 1.183, Revision 1, to provide guidance on the use of Revision 0 and Revision 1, but not with the proposed wording. The staff has added examples to the Background section and Regulatory Position 1.1.5 of RG 1.183, Revision 1, in response to this comment. Please also see the NRC staffs response to Public Comment 8-1, which is a related comment that asks the staff to revise the language to the paragraph discussed in this comment.

11-3 Nuclear Energy Institute Comment The last sentence states: Once the staff has approved the initial AST implementation and it has become part of the facility design basis This sentence should be clarified to identify the initial AST implementation associated with Revision 1 of RG 1.183.

Commenters Proposed Resolution The NRC should clarify the sentence as: Once the staff has approved the initial AST implementation in accordance with Revision 1 and it has become part of the facility design basis Comment Response The NRC staff agrees with this comment but changed the proposed language to clarify the sentence. Specifically, the staff revised Regulatory Position 1.1.1 in response to this comment to state the following:

If the initial AST implementation, consistent with the guidance in RG 1.183, Revision 1, is approved by the staff and becomes part of the facility design basis, licensees may use 10 CFR 50.59, Changes, tests and experiments, and its supporting guidance to assess facility modifications and changes to procedures that are described in the updated FSAR.

11-4 Nuclear Energy Institute Comment SECY-98-154, Reference 19, performed a re-baselining of sample radiological consequences analyses when transitioning from TID to AST source terms. Has the NRC performed similar re-baselining studies for the major method changes being made in DG-1389 compared to RG 1.183, Revision 0? Notably, the NRC should investigate sample PWR and BWR analyses for the LOCA release fraction and timings, non-LOCA release fractions (including TFGR components), updated FHA modeling, etc. If this has already been performed, the wording for this section should be supplemented to include reference to these studies.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Supplement Section 1.3.2 by incorporating results from supporting analyses performed by staff when developing the proposed revisions contained in DG-1389.

Comment Response Please note that the reference to Section 1.3.2 in the comment is referring to Regulatory Position 1.3.2.

The NRC staff disagrees with this comment. The staff previously performed rebaselining to evaluate and provide the technical basis for rulemaking (which became 10 CFR 50.67).

However, this revision of RG 1.183 does not involve a change in regulations but is only a change in the regulatory guidance. Therefore, the NRC staff did not perform a rebaselining similar to that performed for SECY-98-154.

With regard to the proposed resolution to supplement Regulatory Position 1.3.2, the NRC staff does not agree with the proposed change. Specifically, the information requested does not provide guidance or regulatory positions on methods that could be acceptable to meet the requirements in 10 CFR 50.67. Therefore, no changes were made in response to this comment.

11-5 Nuclear Energy Institute Comment RG 1.183, Revision 0 identified specific computer codes acceptable for use in core inventory calculations (ORIGEN2 and ORIGEN-ARP). These code examples were deleted for DG-1389.

The NRC should continue to include examples of core inventory codes which are acceptable for use.

Commenters Proposed Resolution The NRC should continue to include examples of core inventory codes which are acceptable for use in core inventory calculations.

Comment Response The NRC staff disagrees with this comment. Please see the staffs response to Public Comment 10-10. No changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 11-6 Nuclear Energy Institute Comment Based on the directive to maximize fission product inventory it is implied that bounding core parameters should be used, however downstream discussion (e.g., Section 3.2) states that the inventory should be for an equilibrium core. The RG should be consistent in prescribing if a bounding or an equilibrium core should be applied to dose calculations.

Commenters Proposed Resolution The RG should be consistent in prescribing if a bounding or an equilibrium core should be applied to dose calculations.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position 3.2.

The NRC staff agrees with this comment. Regulatory Position 3.1 already discusses this topic and stated, as proposed in DG-1389, The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. Additionally, footnote 10 of RG 1.183, Revision 1, discusses how the maximum inventory at end of life should be used for certain nuclides. In response to this comment, the NRC staff revised the above statement in Regulatory Position 3.1 in RG 1.183, Revision 1, from that proposed in DG-1389 by deleting equilibrium or to reach and changed Regulatory Position 3.2 by replacing equilibrium with maximum. A conforming change was also made to footnote 10 to replace equilibrium with maximum values.

11-7 Nuclear Energy Institute Comment Its not clear that this directive can be applied. Many BWR units do not have radial peaking factors in the COLR nor Technical Specifications.

Commenters Proposed Resolution Remove discussion of COLR/TS from this section.

Comment Response The NRC staff agrees with this comment. In response to this comment, the staff revised Regulatory Position 3.1 to state the following:

To account for differences in power level across the core, the analysis should apply the radial peaking factors (for PWRs, these are contained in the facilitys 71

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 core operating limits report or technical specifications) in determining the inventory of the damaged rods.

11-8 Nuclear Energy Institute Comment Footnote 10 identifies that the data in this section does not apply to cores containing mixed oxide fuel (MOX). The section also identifies that Accident Tolerant Fuel (ATF) concepts, excluding near-term ATF concepts, are also not applicable to the data in this section.

Footnote 10 should also include these ATF concepts as not being applicable for the data in Section 3.2.

Commenters Proposed Resolution Footnote 10 should also include these ATF concepts as not being applicable for the data in Section 3.2.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position 3.2.

The NRC staff agrees with the comment. The staff revised and expanded footnote 10 in DG-1389 to read, The data in this Regulatory Position do not apply to cores containing mixed oxide fuel or to near-term ATF FeCrAl and long-term ATF concepts. Please note that footnote 10 is now footnote 9 in RG 1.183, Revision 1, to reflect that it is applicable to the accident source term rather than to only the release fractions in Regulatory Position 3.2.

11-9 Nuclear Energy Institute Comment It states the steady-state fission product release fractions in Table 3 can only be used if BWR part-length rods are treated as full-length rods with respect to overall quantity of fission products. BWR fuel bundles can have up to 20 part-length rods. Assuming all the part-length rods have the same fission products of a full-length rod penalizes the source term by a large amount. The consequence of this requirement would be the source term has many pins more worth of inventory than it actually has.

Commenters Proposed Resolution Please remove the requirement that BWR part-length rods are treated as full-length rods with respect to overall quantity of fission products.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff disagrees with this comment. Table 3 values were derived using calculations with full-length rods. BWR part-length rods are often placed at the bottom of the core, which generally has a higher neutron flux than the top of the BWR core, so the part-length rods will typically have a more aggressive power profile than full-length rods. Since gap release fractions are dependent on fuel design and power history, using the table 3 release fractions for part-length rods could lead to a nonconservative calculation for the quantity or fission products released, unless the part-length rods are conservatively treated as full-length rods, as identified in RG 1.183, Revision 1. Alternatively, the analytical procedure outlined in RG 1.183, Revision 1, appendix I, Analytical Technique for Calculating Fuel-Design or Plant-Specific Steady-State Fission Product Release Fractions for Non-Loss-of-Coolant Accident Events, can be used to calculate release fractions specific to part-length rods. No changes were made in response to this comment. However, please note that the staff corrected the units for the half-life of xenon (Xe)-135m, Xe-137, Xe-138, and Kr-89 in table J-1 of DG-1389 (now table I-1 in RG 1.183, Revision 1) from months to minutes to be consistent with table 1 of NUREG/CR-7003 and added the units for the decay constants (1/sec).

11-10 Nuclear Energy Institute Comment Regulatory Position 3.2 includes the following statement: If it can be demonstrated that local power level, rate of fission gas release, and cumulative fission gas release remain less than the limiting co-resident UO2 [uranium dioxide] fuel rod, then Table 3 and 4 steady-state fission product release fractions apply to fuel rod designs containing integral burnable absorbers (e.g., Gadolinia).

Commenters Proposed Resolution The NRC should clarify the level of justification needed to confirm fission gas release rates and cumulative fission gas release for fuel designs containing integral burnable absorbers.

Comment Response The NRC staff agrees with this comment. One acceptable means to demonstrate that the rate of FGR and the cumulative FGR for integral burnable absorber fuel is bounded by co-resident uranium dioxide fuel is through the use of an NRC-approved fuel performance code, if that code has models for predicting FGR that are applicable to the integral burnable absorber fuel design.

If the licensees integral burnable absorber design lies outside the range of applicability of its fuel performance code (e.g., using gadolinia concentrations higher than those approved for the code or using alternate burnable absorber compounds for which the NRC staff did not review FGR models), then experimental data that justify the expansion of the applicability of the licensees current FGR models to the integral burnable absorber designs could be provided.

The NRC staff revised Regulatory Position 3.2 to include the following sentence: One acceptable means of demonstrating this is by using an NRC-approved fuel performance code that has fission gas release models that are applicable to the integral burnable absorber fuel designs. To note, RGs describe one or more methods that the NRC staff considers acceptable 73

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements.

11-11 Nuclear Energy Institute Comment A factor is presented to adjust the transient fission product release correlations for short-lived isotopes. This factor is not prominently presented with the correlations and may be missed by applicants or licensees.

Commenters Proposed Resolution The NRC should present the factor for short-lived isotopes more prominently with the correlations presented in the section.

Comment Response The NRC staff agrees with this comment. The staff revised Regulatory Position 3.2 to include separate equations for the long-lived and short-lived isotopes to make the factor for short-lived isotopes clearer.

11-12 Nuclear Energy Institute Comment Section 3.2 contains the following guidance: For the remaining non-LOCA DBAs which predict fuel rod cladding failure, such as PWR reactor coolant pump locked rotor and fuel handling accident, additional fission product releases may occur as a result of fuel pellet fragmentation (e.g., fracturing of high burnup rim region) due to loss of pellet-to-cladding mechanical constraint or impact loads. TFGR has been experimentally observed under a variety of accident conditions and should be addressed in future applications. However, the regulatory guide does not provide guidance regarding an acceptable treatment for any additional fission product releases for these accidents. Therefore, the current guidance within the draft regulatory guide is incomplete.

Commenters Proposed Resolution Recommend that the current draft regulatory guide be updated to provide complete guidance for these remaining accidents.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position 3.2.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff partially agrees with this statement. The staff revised RG 1.183, Revision 1, in response to this comment to include additional general guidance, as described in the response to Public Comment 8-2. However, if more data become available, the NRC staff may decide to issue more specific guidance for calculating TFGR for non-LOCA and non-reactivity-initiated DBAs in the future.

11-13 Nuclear Energy Institute Comment The paragraph starting with For the remaining non-LOCA DBs which predict fuel rod cladding failure is not clear on the background of the issue or the success pathways afforded to the licensees to address the NRCs concerns on the topic in question. Ideally, background documents for further description of the issue and/or documents identifying acceptable methods of evaluation are cited. The NRC should include these documents in order for the licensees to fully understand and properly address the issue in question. If no documents exist, the NRC should develop these positions.

Commenters Proposed Resolution Include background documents for further description of the issue and/or documents identifying acceptable methods of evaluation so that licensees can fully understand and properly address the issue in question. If no documents exist, the NRC should develop these positions.

Comment Response The NRC staff partially agrees with this comment. Changes were incorporated into RG 1.183, Revision 1, in response to this comment to include additional general guidance, as described in the NRC staffs response to Public Comment 8-2. The staffs response to Public Comment 8-2 also discusses that there is neither a consensus on the mechanism of TFGR nor a significant body of experimental data for the computation of TFGR for non-LOCA and non-reactivity-initiated DBAs to provide specific guidance at this time. As a result, future applications should address this using engineering judgment or experimental data.

11-14 Nuclear Energy Institute Comment Tables 1 and 2 present core inventory fraction releases into containment for BWRs and PWRs.

The release groups are updated from Revision 0 of RG 1.183. One significant difference is that the tellurium group (both PWRs and BWRs) and barium, strontium group (PWRs only) identify a release fraction during the gap release phase.

From page 34 of the SAND-2011-0128 report: At the same time, the Gap Release Phase was calculated to be long enough that some core degradation characteristic of the In-vessel Phase 75

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 of release as prescribed in the NUREG-1465 Source Term did take place. This is indicated by small amounts of tellurium release during the Gap Release Phase and in the case of PWR accidents small amounts of alkaline earth release. Ordinarily tellurium and alkaline earths are not thought to be contributors to the gap release. They contribute here because some portions of the core had entered into what would be categorized phenomenologically as in-vessel release before the criterion to terminate gap release had been reached.

This confirms that, while the timing of the Te, Ba, and Sr groups does release during the gap fraction phase, these releases are more accurately associated with the early in-vessel phase.

The continued identification of these groups release as gap phase release creates confusion with subsequent gap fraction tables (Tables 3 and 4) which do not identify these nuclide groups as being released. [Note that it is not requested to add Te, Ba, and Sr to the gap releases.]

Commenters Proposed Resolution The NRC should clarify that the Te, Ba, and Sr groups do contribute releases during the gap release phase for LOCA releases in Tables 1 and 2; however, these releases are more accurately classified as early in-vessel releases and these groups do not need to be considered in gap releases for other events.

Comment Response The NRC staff does not agree with the comment that the inclusion of tellurium (Te), barium (Ba),

and strontium (Sr) groups in the gap-release phases for tables 1 and 2 would create confusion with subsequent gap fraction tables (tables 3 and 4). These tables were derived from different technical research and pertain to different accidents. To note, because of a discretized core in the SAND2011-0128 analyses rather than the single-node core in prior analyses (NUREG-1465), the SAND2011-0128 source term analyses did not experience a distinct gap release phase that was previously observed (sections of the core experienced in-vessel releases before other sections had not released their gap inventories). This necessitated creating a technical criterion to designate the end of the gap release phase that was consistent with prior analyses and resulted in the assignment of some in-vessel releases to the gap release. The small amount of these releases assigned to the gap release phase is not significant to overall dose or relative to the early in-vessel releases of the same chemical groups. No changes were made in response to this comment.

11-15 Nuclear Energy Institute Comment Tables 1 and 2 report release fractions from a new Molybdenum group; however, the elements in this new group are not listed in Table 6. Which release group is Zirconium (Zr) considered to reside? DG-1389 is consistent with Revision 0 and indicates it is part of the Lanthanides while Table 14 in the underlying Sandia report (SAND 2011-0128) reports Zirconium as part of the Cerium group.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution It is expected that the final grouping is consistent with SAND 2011-0128 since that is the basis for the release fractions in Tables 1 & 2. Update the Table 6 for the new proposed Molybdenum group. Change Zr grouping to Cerium group.

Comment Response The NRC staff agrees with this comment. The staff revised table 6 to be consistent with the groupings in SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, (ML20093F003), Tabletable 14, and ERI/NRC 11-211, Peer Review of Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup and Mixed Oxide Fuel, (ML12005A043), section 4.3, in response to this comment as well as similar comments in Public Comments 7-3, 11-20, and 15-5.

11-16 Nuclear Energy Institute Comment Figure 1 presents the maximum allowable power operating envelope for steady-state release fractions. The envelope is identified by rod average power (kW/ft) and rod average burnup (GWD/MTU). The figure further identifies a Peak LHGR (Linear Heat Generation Rate) of 15.0 kW/ft for BWRs and 14.0 kW/ft for PWRs. From the Ref. 24 Technical Basis for Non-LOCA Fission Product Release Fractions, the Peak LHGR is a combination of rod average power and the axial power profile maximum factors. For example, using a rod average power of 12.2 kW/ft from Figure 1 and an axial power profile maximum factor of approximately 1.15 for PWRs from Figure 2 of Ref. 24, the resulting Peak LHGR would be approximately 14.0 kW/ft. The maximum linear heat rate is a typical term for licensees (generally defined in safety analyses and/or Technical Specifications) and is the product of the rod average power (or linear heat generation rate determined by the core rated thermal power and the linear component of all power producing rods in the core) and the hot channel factor (FQ). These maximum linear heat rates can exceed the definition of Peak LHGR associated with Figure 1 of DG-1389. Without further explanation of the definition of Peak LHGR associated with the figure, misinterpretation and confusion of the Figure 1 envelope may result between the licensee and regulator.

Commenters Proposed Resolution Include the following statement in a footnote for Figure 1 or incorporated into the discussion of Section 3: From Ref. 24, the Peak LHGR is defined as the product of the peak fuel rod average power and the peak fuel rod axial power distribution. This Peak LHGR may differ from the definition used in licensees Technical Specification or Core Operating Limits Report. A Peak LHGR derived consistent with the definition of Ref. 24 should be used for comparison of applicability to Figure 1.

Comment Response Please note that the reference to Section 3 in the comment is referring to Regulatory Position 3.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff disagrees with this comment. The peak LHGR depicted in figure 1 should bound the peak LHGR in the core, as defined in a licensees technical specifications or core operating limits report, to confirm the applicability of tables 3 and 4. Licensees can use appendix I to RG 1.183, Revision 1, if they cannot meet the peak LHGR of figure 1. However, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

11-17 Nuclear Energy Institute Comment The Figure 1 rod average power envelopes were derived in Ref. 24. Ref. 24 does not indicate if uncertainties were applied to the rod average powers or normalized axial power distributions in Figure 2 of Ref. 24. It is assumed that uncertainties are not applied; however, the NRC should confirm that uncertainties do not need to be considered when comparing against the bounding power profile. Note that for typical applications for determining release inventories, uncertainties may be accounted for in peaking factors such as the radial peaking factor (applied uncertainty),

which is applied separately to the core inventory from the gap fractions in the radiological analyses. As such, applying uncertainties in determining the gap fractions may result in double-accounting and should not be advised.

Commenters Proposed Resolution The NRC should confirm that uncertainties do not need to be applied to the power inputs used in comparing to the bounding power profile and include this clarifying information in the discussion in Section 3.2.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position 3.2.

The NRC staff disagrees with this comment. Licensees should make adjustments to power histories to account for power uncertainties and plant maneuvering before comparing to the power envelope in figure 1 of RG 1.183, Revision 1. The NRC staff used the analytical procedure outlined in RG 1.183, Revision 1, appendix I, to calculate the release fractions in tables 3 and 4 in section C of the RG. Attachment 3, Sample Calculation of Steady-State Release Fractions, of Reference 24 (U.S. NRC Internal Memorandum from Paul Clifford to Robert Lukes, Technical Basis for Draft RG 1.183 Revision 1 (2021) Non-LOCA Fission Product Release Fractions, dated July 28, 2021 (ML21209A524)), in RG 1.183, Revision 1, walks through a sample calculation using the analytical procedure in DG-1389, appendix J (now RG 1.183, Revision 1, appendix I) to perform cycle-specific calculations to verify the continued applicability of plant-specific, steady-state release fractions docketed in the updated FSAR (e.g., chapter 15) radiological consequences. Attachment 3 of Reference 24 states the following:

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 For this cycle, the licensee surveys the reload depletion and identifies the limiting fuel rod power histories for long-lived stable isotopes and short-lived volatile isotopes. Examples are shown in Figure A3-1. The licensee then makes adjustments to account for power uncertainties and plant maneuvering.

As shown in the excerpt above, adjustments are made in the licensees rod power histories in the sample calculation in attachment 3 of Reference 24 to account for power uncertainties and plant maneuvering. Therefore, the same should be done when the licensee is comparing power histories to the bounding power profile of figure 1 of RG 1.183, Revision 1. In response to this comment, the NRC staff added the following sentence to Regulatory Position 3.2 of RG 1.183, Revision 1: Licensees should make adjustments to account for power uncertainties and plant maneuvering when comparing operating power histories to figure 1.

11-18 Nuclear Energy Institute Comment Table 5 in Section 3.3 lists the duration of the gap release phase is 0.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months for PWRs and 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months for BWRs. However, the last sentence of the following paragraph says: Regardless of delays in the onset, the duration of the gap release phase is 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The values in Table 5 are not consistent with this statement.

Commenters Proposed Resolution The NRC should delete or correct the sentence in question as it contradicts gap phase durations in Table 5.

Comment Response Please note that the reference to Section 3.3 in the comment is referring to Regulatory Position 3.3.

The NRC staff agrees with this comment. The staff deleted the sentence mentioning the duration of the release in the paragraph below table 5 in DG-1389.

11-19 Nuclear Energy Institute Comment DG-1389 states: The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase. RADTRAD models core release as a constant fraction of the core inventory over the release duration. Including decay of the core inventory will make this release slightly non-linear, decreasing over time as the core inventory decreases due to decay.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Clarify that the statement in the guidance excludes the effects of decay.

Comment Response The NRC staff agrees with this comment. The staff revised footnote 11 in Regulatory Position 3.3 to clarify that the regulatory position quoted in the comment excludes the effects of radioactive decay in the core inventory.

11-20 Nuclear Energy Institute Comment Table 6 presents the radionuclide groups which should be considered in design basis analyses.

The table presents molybdenum as included in the noble metals group. This is inconsistent with Tables 1 and 2 which present molybdenum as a separate group with different release fractions than the noble metals group.

Commenters Proposed Resolution The NRC should present molybdenum as a separate group in Table 6 in order to maintain consistency with Tables 1 and 2.

Comment Response The NRC staff agrees with this comment. The staff revised table 6 to be consistent with the groupings in SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, (ML20093F003), table 14, and ERI/NRC 11-211, Peer Review of Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup and Mixed Oxide Fuel, (ML12005A043), section 4.3, in response to this comment as well as similar comments in Public Comments 7-3, 11-15, and 15-5.

11-21 Nuclear Energy Institute Comment DG-1389 includes a statement that the transit dose to personnel traveling to and from the control room should be considered for licensees whose licensing basis includes transit dose.

This statement should be clarified or removed based on the following points:

For licensees who do not currently include transit dose in their licensing basis, will the NRC force the adoption of transit dose if this regulatory guide is adopted?

For licensees who do currently include transit dose in their licensing basis, will the NRC 80

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 allow for removal of the transit dose based on the positions summarized below:

GDC 19 identifies that Adequate radiation protection shall be provided to permit access and occupancy of the control room. GDC 19 does not identify that access includes the need for the control room to provide protections outside of the control room envelope. Additionally, radiological consequences analyses are intended to provide an evaluation of the design and performance of structures, systems, and components of the facility. This intent is to confirm the design, construction, and siting of the facility and applicable safety features are adequate to limit dose exposure; thus, operator actions outside of the CR envelope are not historically considered in the design basis analyses.

During emergency situations, personnel dose is governed by ALARA principles such that Emergency Planning will limit the dose received by operators during transit to and from the control room. 10 CFR 50.47(a)(9), (10), and (11) provide sufficient requirements for the Emergency Response organizations to measure and limit doses to emergency response personnel including control room operators.

Commenters Proposed Resolution The NRC should remove these statements from the RG on the basis that transit doses would be addressed by ALARA principles through Emergency Planning. These measures may include alternate travel pathways to and from the control room such that operators are not traversing the radioactive plume and personal protective equipment (e.g., respirators).

Comment Response In consideration of the comments, the NRC staff has removed the language associated with transit dose in Regulatory Position 4.2 and restored the original language from RG 1.183, Revision 0. The staff is evaluating this issue and will determine whether to address it in a future revision to this RG.

11-22 Nuclear Energy Institute Comment The occupancy factors in RG 1.183 R0 are based on RG 1.183 R0 Reference 22 (Reference 35 in Rev. 1). Reference 22 is the paper by K. G. Murphy and K. W. Campe, Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19, published August 1974. In that paper, the occupancy factors are described as follows:

an allowance may be considered for the time the operator leaves the plant vicinity. This is described as the occupancy factor.

Table 1 in the paper by K. G. Murphy and K. W. Campe provides the occupancy factors included in the determination of the X/Q values using the K. G. Murphy and K. W. Campe methodology. Therefore, it is reasonable to conclude the occupancy factors (1, 0.6 and 0.4) assumed for the control room analysis already implicitly account for the time for the operator travel from the parking lot to the control room. The occupancy factor discussion is in the X/Q 81

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 section (page 13) which is consistent with this argument. No additional discussion is required with those occupancy factors. Further the Murphy and Campe paper interprets the GDC in this way: Whole body gamma radiation from direct shine radiation sources external to the control room and from the airborne activity within the control room should not exceed a total of 5 rem.

Direct shine here is from outside radioactivity directly to operators within the control room. This would further exclude a transit dose for the operators. As well, in NUREG-0800, Section 6.4, referring to individuals within the control room says: In accordance with GDC 19, these doses to an individual should not be excluded for any postulated accident. The whole body gamma doses consists of the contributions from airborne radioactivity inside and outside the control room, as well as direct shine from all radiation sources. Industry has general evaluated the other sources in terms of containment shine and filter shine.

Commenters Proposed Resolution Remove the following discussion as it is unnecessary: The following guidance should be used in determining the TEDE dose for demonstrating compliance with 10 CFR 50.67(b)(2)(iii). For the purpose of this RG, a transit dose is considered to be the dose that is accumulated as personnel travel to and from the control room for the duration of an accident once onsite (e.g., dose from site boundary to the control room). Licensees whose licensing basis includes transit dose should include the transit dose to demonstrate compliance with 10 CFR 50.67(b)(2)(iii). The licensees results for the evaluation of transit dose will be evaluated on a case-by-case basis. New reactor licensees that are required to show compliance with GDC 19 or similar control room radiological habitability principal design criteria may use this guidance.

Comment Response In consideration of the comments, the NRC staff has removed the language associated with transit dose in Regulatory Position 4.2 and restored the original language from RG 1.183, Revision 0. The staff is evaluating this issue and will determine whether to address it in a future revision to this RG.

11-23 Nuclear Energy Institute Comment Footnote 15 states that nuclides used for modeling dose from airborne radioactivity inside the control room may not be conservative for determining the dose from radioactivity outside of the control room. This statement is vague and does not provide the licensee with details as to which nuclides the NRC is concerned about for the different control room dose aspects (inhalation/immersion versus plume shine dose). Without further explanation, it is assumed that the NRC is in agreement that the sets of nuclides used by licensees acceptably addresses the footnote.

Commenters Proposed Resolution The NRC should provide clarification for Footnote 15 in terms of which nuclides are of concern for the different aspects of the control room dose. Lack of explanation infers that the NRC 82

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 accepts the licensees currently analyzed source terms, and no updates are needed.

Comment Response The NRC staff agrees with this comment. The staff removed footnote 15 in DG-1389. The footnote removed read, The nuclides used for modeling dose from airborne radioactivity inside the control room may not be conservative for determining the dose from radioactivity outside the control room.

11-24 Nuclear Energy Institute Comment This section provides the expression to correct the semi-infinite cloud dose to a finite cloud dose for external exposure. This expression is generally incorporated into radiological consequences codes (e.g., RADTRAD) and does not specifically need to be considered as an additional factor applied to the control room dose results.

Commenters Proposed Resolution The NRC should note that the expression may be incorporated into radiological consequences codes (e.g., RADTRAD) and does not need to be specifically applied to the control room dose results.

Comment Response Please note that the reference to This section in the comment is referring to Regulatory Position 4.2.7.

The NRC staff agrees with this comment. The staff added the following sentence to Regulatory Position 4.2.7: The expression may be already incorporated into certain radiological assessment codes and would not need to be separately added to the dose results of finite volumes such as control room doses when using those codes.

11-25 Nuclear Energy Institute Comment Table 7 provides the analysis release duration for the various accidents presented in the regulatory guide. The following items should be addressed by the NRC:

The PWR Steam Generator Tube Rupture, PWR Main Steamline Break, and PWR Locked Rotor identify that the analysis release duration is Until cold shutdown is established. For consistency with the event-specific guidance presented in Appendices 83

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 E, F, and G, this wording should be changed to Until shutdown cooling is in operation and releases from the SG(s) have been terminated.

For Fuel Handling Accident, the analysis release duration should be updated for consistency with the updated FHA model presented in Appendix B. The updated model presents releases in two phases which potentially span the entire standard 30 days of event duration.

Commenters Proposed Resolution For consistency with the event-specific guidance presented in Appendices E, F, and G, change wording from the analysis release duration is Until cold shutdown is established to Until shutdown cooling is in operation and releases from the SG(s) have been terminated. In the Fuel Handling Accident, update the analysis release duration with the updated FHA model presented in Appendix B.

Comment Response The NRC staff agrees with this comment. The staff revised the analysis release duration information in table 7, Accident Dose Criteria for EAB, LPZ, and Control Room Locations, based on this comment. The table 7 analysis release durations for accidents in appendices B, E, F, and G were made consistent with the language in these appendices. In addition, the table 7 analysis release duration for the PWR rod ejection accident in appendix H was also made consistent with the language used in appendices E, F, and G.

11-26 Nuclear Energy Institute Comment Guidance for the modeling of the limiting X/Qs for the LPZ and control room is presented and directs the licensee to model the period of most unfavorable atmospheric dispersion factors coincident with the time period of most adverse environmental release. This guidance has been included previously for the control room location in RG 1.194 (and identified in the 2015 periodic review of RG 1.194 as an item for alignment). However, the guidance has not been previously incorporated in or identified as needed for RG 1.183. [Also, no identification of this guidance has been made for RG 1.145 for the LPZ location.]

The current widely used practice for modeling X/Q values for the LPZ and control room is to align the values in correct time period order, e.g., 0-2 hrs, 2-8 hrs, etc. (similar to the first portion of Figure 2 in DG-1389). However, it is recognized that this may not align with the period of most adverse environmental release in instances of delayed releases from damaged/melted fuel (in cases of an MHA) or prolonged buildup and transit from the primary to secondary systems to the environment (in cases of a locked RCP rotor).

The interpretation of the guidance presented in DG-1389 is that the period of most adverse release aligns with the period of highest dose. However, the NRC has not provided sufficient guidance for consideration of situations in which the period of most adverse environmental release does not align with the period of highest dose. This may be the case for instances in the 84

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 control room depending on specific modeling assumptions.

For instance, for a PWR MHA, a staged release is modeled in accordance with Table 2 of RG 1.183. This results in the limiting environmental release typically occurring after the initial 0-2 hour time period. The control room is typically isolated early in the event, accounting for some delay, such that only gap activity enters the control room envelope during unfiltered, normal ventilation mode. Therefore, during the period of greatest environmental release, the activity entering the control room would be subject to emergency mode flows (possibly lower than normal mode), filtration (~90-95% iodine and particulates), and lower X/Q values associated with emergency mode intakes (depending on the receptor positioning). As a result, applying the proposed guidance would decrease the control room dose consequences relative to the current practices.

Additionally, no guidance has been provided for cases in which multiple release pathways are modeled with varying periods of limiting environmental releases and/or dose.

For simplicity, and to avoid potential non-conservative applications of the guidance, it is recommended that the NRC remove this position from DG-1389 and maintain the current practice of applying LPZ and CR values from highest X/Q at event initiation (T=0) for all events, regardless of release magnitude. It is also recommended that this position be removed from RG 1.194.

Commenters Proposed Resolution The NRC should re-assess the need to include this guidance for atmospheric dispersion factor modeling. The guidance, as presented, may result in a misappropriation of the limiting two-hour X/Q values for the time period of greatest release as this may not align with the time period of greatest dose contribution for the control room.

It is recommended that the NRC remove this position from DG-1389 and maintain the current practice of applying LPZ and CR values from highest X/Q at event initiation (T=0) for all events, regardless of release magnitude. It is also recommended that this position be removed from RG 1.194.

Comment Response The NRC staff partially agrees with this comment. In response to this comment, the NRC staff has clarified the guidance in Regulatory Position 5.3 to address the comments concern that the guidance in RG 1.183, Revision 1, could decrease the control room dose consequences (as compared to models that align the atmospheric dispersion values, as shown in RG 1.183, Revision 1, figure 2a). To prevent underpredicting control room and LPZ doses, the NRC staff added the underlined text to the following sentence to clarify the meaning of adverse release:

To ensure a conservative dose analysis, the period of the most adverse release of radioactive materials to the environment, with respect to doses, should be assumed to occur coincident with the period of most unfavorable atmospheric dispersion.

RG 1.183, Revision 1, provides one acceptable methodology to align the atmospheric dispersion factors with the most adverse release of radioactivity. Existing computer codes, such as RADTRAD/SNAP (Symbolic Nuclear Analysis Package), do not perform this calculation automatically. To determine the periods of release that are most limiting to dose values, a 85

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 constant atmospheric release value can be assumed in the RADTRAD/SNAP code models, or a spreadsheet can be used. The time-dependent atmospheric release factors would then be applied as described in Regulatory Position 5.3 and figure 2b in that position. The recommendation that a regulatory position be removed from RG 1.194 is out of scope for the revision to RG 1.183, Revision 1, and no change was made in response to that recommendation.

11-27 Nuclear Energy Institute Comment This section indicates that the aerosol deposition models in NUREG/CR-6189 are still applicable; however, based on Section 1 of NUREG/CR-6189, the models are based on the release fractions and timing in NUREG-1465. Considering the significant changes to the release fractions and timings in DG-1389, there may be significant impacts to the deposition rates in this NUREG.

Commenters Proposed Resolution Confirm the continued applicability of the NUREG/CR-6189 aerosol removal rates.

Comment Response Please note that the reference to section in the comment is referring to Regulatory Position A-2.2.

Please see the NRC staffs response to Public Comment 7-7.

11-28 Nuclear Energy Institute Comment Revise A-2.5 to allow credit for suppression pool scrubbing based on:

NUREG/CR-6153provides models for accident dose calculations using the AST for the purposes of crediting iodine decontamination provided by suppression pools

Accounts for the changing aerosol distribution following passage through the pool State-of-the-Art Reactor Consequence Analyses (SOARCA) project results (ADAMS Accession No. ML20304A339NRC Brochure) include the suppression pool in their models and indicate that all modeled accident scenarios, progress more slowly and release smaller amounts of radioactive material than calculated in earlier studies.

Commenters Proposed Resolution 86

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Change A-2.5 to reference NUREG/CR-6153 and allow credit for reduction in airborne radioactivity in the containment by suppression pool scrubbing in BWRs.

Comment Response While the NRC staff agrees that the need to address changing aerosol sizes due to pool scrubbing would be important to any credit for downstream removal, such as in the main steamlines, the main steamline models in RG 1.183, Revision 1, do not account for suppression pool scrubbing.

It should also be noted that the use of NUREG-1935 (SOARCA) to inform the design-basis analysis is complex. SOARCA is a realistic study not created for design-basis licensing analysis.

As such, there are significant differences between the SOARCA models and those used for design-basis licensing.

The NRC staff does not agree with the proposed resolution to change Regulatory Position A-2.5 to reference NUREG/CR-6153, A Simplified Model of Decontamination by BWR Steam Suppression Pools, issued May 1997. The NRC staff has not reviewed NUREG/CR-6153 for endorsement in design-basis analyses. Likewise, the NRC staff has not reviewed NUREG/CR-6153, nor has the NRC staff developed a revised steamline model to account for changing aerosol sizes, for endorsement in Revision 1 of RG 1.183.

Regulatory Position A-2.5 of DG-1389 already allows for consideration of suppression pool scrubbing on an individual case basis and provides guidance on what an evaluation should consider to credit suppression pool scrubbing. However, the NRC staff revised Regulatory Position A-2.5 in response to this comment to acknowledge that, although suppression pool scrubbing has not been credited historically for operating BWR reactors, the staff may consider such reductions on an individual case basis. Additionally, the NRC staff added footnote 1, which provides an example of modeling radionuclide transport in containment with scrubbing credit.

11-29 Nuclear Energy Institute Comment For gas-filled secondary containment bypass leakage paths, this section states deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis. Plate-out of elemental iodine may also be a significant removal mechanism in gas-filled secondary bypass leakage pathways. In addition to aerosol deposition, the plate-out of elemental iodine should not be excluded.

Commenters Proposed Resolution Add deposition of aerosol radioactivity and plate-out of elemental halogens Comment Response 87

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Please note that the reference to section in the comment is referring to Regulatory Position A-3.5.

The NRC staff agrees with this comment. Please see the NRCs staff response to Public Comment 7-9, which provided a similar comment.

11-30 Nuclear Energy Institute Comment Multiple BWRs currently have credit for aerosol removal from drywell sprays as well as aerosol deposition within in the main steam lines (some also have Condenser removal) in their current licensing basis. Section A-5 presents three acceptable methods for calculating aerosol deposition within the main steam lines, but this section states, however, these methods are not valid if credit has been taken for aerosol removal from drywell sprays. Given the prevalence of credit for both sprays and steam line deposition, why is there not a model presented that the Staff finds acceptable for crediting both, or modifications to the presented models if the licensee wants to credit spray removal (e.g., different aerosol size distribution)?

Commenters Proposed Resolution Provide a model where it is acceptable to credit aerosol removal from drywell sprays as well as aerosol deposition within in the main steam lines.

Comment Response Please note that the reference to Section A-5 in the comment is referring to Regulatory Position A-5.

The NRC staff disagrees with this comment. Notably, only a few BWRs credit drywell sprays for aerosol removal, and those that do typically have used proprietary or more complex methods for modeling the MSIV leakage pathway. Because regulatory guidance is developed for most (but not all) designs, the NRC staff did not develop or add the requested regulatory guidance.

However, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

11-31 Nuclear Energy Institute Comment What is the technical justification for drywell sprays aerosol removal not being valid if used alongside main steam isolation valve leakage? The removal mechanisms associated with spray described in NUREG/CR-5966 are largely different than the removal mechanisms associated 88

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 with deposition described in SAND 2008-6601 and AEB 98-03. Section 6.2 of SAND 2008-6601 states that sprays should not be used in conjunction with main steam line deposition, but this discussion is based on the steam dome being the source of radioactivity rather than the accepted position of the drywell being the source of radioactivity.

Commenters Proposed Resolution Revise words to state that the impact of sprays on aerosol removal should be determined in the submittal, and do not state that removal by both sprays and deposition should not be used.

Comment Response The NRC staff disagrees with this comment. The three steamline deposition models proposed in the regulatory guidance make an assumption regarding the containment source size distribution that does not consider the impact of sprays. This assumption and modeling are not consistent with the source size distribution that would be present if sprays are credited. Sprays impact the aerosol size distribution in containment. NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, issued June 1993 (ML063480542), provides details on how sprays impact aerosols. NUREG/CR-5966 indicates that the sprays shift the sizes of aerosols in the containment towards those that are removed most slowly (the mean aerosol size decreases as the sprays operate). A different aerosol distribution due to crediting sprays would make these models invalid for use when both sprays and steamline deposition are credited. No changes were made in response to this comment.

11-32 Nuclear Energy Institute Comment DG-1389 states that the reported 3 aerosol methods are not valid if credit has been taken for aerosol removal by drywell sprays. Are other aerosol removal mechanisms similarly affected?

Can a licensee credit the natural removal mechanisms in NUREG/CR-6189 consistent with Section A-2.2 and apply the reported MSL models? Although the deposition models in NUREG/CR-6189 may affect the aerosol size distribution, this impact would be expected to be similar to the impact of the MSL models themselves such that the aerosol distribution entering the MSLs are not significantly different from that assumed in the models.

Commenters Proposed Resolution Confirm acceptability of applying NUREG/CR-6189 aerosol deposition in addition to reported MSL deposition models.

Comment Response Please note that the reference to Section A-2.2 in the comment is referring to Regulatory Position A-2.2.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff agrees with this comment. The three aerosol removal methods did not consider the impact of containment removal due to also crediting NUREG/CR-6189 and other containment removal mechanisms. Based upon this comment, the NRC staff modified Regulatory Position A-5 to clarify that the use of these three steamline deposition methods is not valid when also crediting other containment aerosol removal mechanisms without accounting for the change in particle size distributions due to these containment removal processes. See also the NRC staffs response to Public Comment 7-7 for other related items regarding NUREG/CR-6189, as well Public Comment 11-27.

11-33 Nuclear Energy Institute Comment DG-1389 reports the first MSL aerosol deposition model as: Direct adoption of the SAND 2008-6601 (Ref. A-12) recommendations without scaling R*-factors; This approach needs additional explanation. As defined in Section 1.1 of SAND2008-6601, the R* factor is defined as the ratio of NUREG-1465 containment airborne concentrations to MELCOR containment airborne concentrations. A separate factor, RM, models the ratio of the steam dome concentration to the drywell concentration determined by the MELCOR full plant analyses. Backing the R* factor out of the process does not appear to be possible since, per Section 5.2 of SAND2008-6601, the RM and R* factors are combined and their product is applied to develop the results in Table 5-3. This section should adopt the MHA approach where the drywell and steam dome are well-mixed. As such, there should be no scaling factor applied to model the increased concentration in the steam dome.

Commenters Proposed Resolution Revise first aerosol deposition model to without scaling RM and R* factors.

Comment Response The NRC staff agrees with this comment. The NRC staff modified Regulatory Position A-5 in DG-1389 to state without scaling R*-or RM factors; in response to this comment.

11-34 Nuclear Energy Institute Comment It is understood that there are several deposition and removal mechanisms (see Table 4.1-3 of the State-of-the-Art Report for examples) present inside of steam lines and it is challenging to provide a comprehensive model due to lack of experimental data and disassociation between the reality of core cooling being maintained during a design basis accident and fuel damage being assumed per DG-1389. The State-of-the-Art Report attempts to capture all of these uncertainties but cannot and does not form a conclusion based on all of the available literature.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Based on previous approval of AST applications and the lack of concrete knowledge surrounding removal in steam lines, it is recommended to allow licensees to continue to use their existing main steam line removal models if pursuing licensing actions unrelated to the steam lines themselves.

Commenters Proposed Resolution Change footnote 4 to state that previously approved methods are not superseded by the RG 1.183 Revision 1 methods if unaltered during submittals associated with licensing actions unrelated to steam line removal.

Comment Response The NRC staff disagrees with this comment. RG 1.183, Revision 0, does not include regulatory positions endorsing aerosol removal mechanisms in the main steamlines. RG 1.183, Revision 1, contains three aerosol removal models based upon fundamental aerosol physics that are well known. Revision 1 provides a method acceptable to the NRC staff for demonstrating compliance with 10 CFR 50.67. Licensees may use alternatives to this guidance if the licensees demonstrate that the alternative satisfies the applicable NRC requirements. Additionally, Regulatory Position 5.1.4 provides the guidance suggested regarding prior design bases that are unrelated to or unaffected by an AST. No changes were made in response to this comment.

11-35 Nuclear Energy Institute Comment Aerosol deposition in vertical volumes is non-zero. SAND 2008-6601 Section 3.3 states that some deposition occurs in vertical surfaces. This is confirmed by Figure 2-16 and Figure 2-17 of the State-of-the-Art Report which show strong correlations between temperature driven deposition by thermophoresis and condensing vapor deposition by diffusiophoresis. These removal mechanisms are largely independent of gravitational settling in horizontal segments.

Commenters Proposed Resolution Remove in horizontal volumes or provide different guidance for horizontal vs. vertical volumes when applying Method 1: Direct adoption of the SAND 2008-6601 recommendations without scaling RM and R* factors.

Comment Response The NRC staff agrees that the aerosol deposition is nonzero, but in the model developed in SAND2008-6601, horizontal lengths were deemed to be more important than vertical main steamline pipe because aerosol deposition primarily occurs on horizontal rather than vertical surfaces. Because the recommended model developed in SAND2008-6601 does not include deposition on these vertical surfaces, no changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 11-36 Nuclear Energy Institute Comment Section A-5.6 implies that only plants other than Mark I, II, or III will be considered on a case-by-case basis. All models should be considered on a case-by-case basis considering the models described by positions A-5.6.1, A-5.6.2, and A- 5.6.3 do not consider all removal mechanisms present in the steam lines.

Commenters Proposed Resolution Revise statement to allow case by case review for all technologies.

Comment Response Please note that the reference to Section A-5.6 in the comment is referring to Regulatory Position A-5.6.

The NRC staff partly agrees with the comment but does not agree with the proposed resolution.

The NRC staff agrees that Regulatory Positions A-5.6.2 and A-5.6.3 do not consider all removal mechanisms, as they consider only removal due to gravitational settling. However, Regulatory Position A-5.6.1 provides a model that includes all significant removal mechanisms.

Regarding the proposed resolution, the NRC staff does not agree that Regulatory Position A-5.6 needs to be revised to allow case-by-case reviews for all technologies. The SAND2008-6601 model is specific to large, light-water BWR-type designs but may not be appropriate for other technologies. While the multi-group and numerical integration methods are technology neutral, the assumptions in the regulatory positions for these methods, such as the aerosol sizes, may not be technology neutral or appropriate for technologies other than BWR designs with Mark I, II, or III containment designs. In response to this comment, the staff revised Regulatory Position A-5.6 to state the following:

For BWRs with Mark I, II, or III containment designs, aerosol deposition in horizontal volumes that meet Regulatory Position A-5.4 or A-5.5 may be credited as described below.5 The NRC staff will consider aerosol deposition models for BWR designs other than those with Mark I, II, or III containment designs on a case-by-case basis.

11-37 Nuclear Energy Institute Comment Section A-5.6 indicates that the multi-group method will be evaluated on an individual case-by-case basis. This statement implies that the application of this method may not be fully approved, thereby increasing the potential for regulatory uncertainty in future submittals. Why does the approach described in Section A-5.6.2, including an AMMD of 2.0 µm 2,000 groups, and 10,000 trials, require additional regulatory review?

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Commenters Proposed Resolution Confirm that the approaches in Sections A-5.6.2 and A-5.6.3 are acceptable models for aerosol deposition in the MSLs.

Comment Response Please note that the reference to Section A-5.6.2 and Section A-5.6.3 in the comment are referring to Regulatory Position A-5.6.2 and Regulatory Position A-5.6.3.

The NRC staff agrees with this comment. The staff has deleted the phrase which will be evaluated on an individual case-by-case basis. Furthermore, Reference A-20 (Technical Basis for Draft RG 1.183 Revision 1 (2021) Re-Evaluated AEB-98-03 Settling Velocity Method, the Multi-Group Method, and the Numerical Integration Method, dated July 29, 2021 (ML21141A006)) describes the technical basis for the selected parameters described in Regulatory Position A-5.6.2. The methods and positions provided in Regulatory Positions A-5.6.2 and A-5.6.3, when used with the other regulatory positions within RG 1.183, Revision 1, provide an acceptable approach for modeling aerosol deposition within the main steamlines.

11-38 Nuclear Energy Institute Comment The basis for not being able to credit the piping upstream of the inboard MSIV seems to come from Section 6.3 of SAND2008-6601 which states that at times in the simulation the temperature of portions of the in-board MSL piping are predicted to be high enough to vaporize fission products that had been previously deposited. The MELCOR simulation in SAND2008-6601 is not representative of plant specific thermo-hydraulic conditions.

(e.g., Figure 2-19 of SAND2008-6601 shows a long-term temperature of approximately 800F in the steam dome while most BWR analyses would show long term temperatures of less than 300F in the core with ECCS operational). Credit should be able to be taken for deposition in the inboard lines if plant specific analysis shows low temperatures considering RIS 2006-04 states that deposition of particles in the inboard volume occurs. In addition, the time periods in DG-1389 are inconsistent with SAND2008-6601.

Commenters Proposed Resolution Revise table to be consistent with SAND 2008-6601 and add in calculated removal for in-board lines.

Comment Response The NRC staff disagrees that credit should be taken for deposition in the inboard lines. Please see the NRC staffs response to Public Comment 6-6 for additional information.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff agrees that time periods in table A-1 of DG-1389 are different than those in SAND2008-6601. Please see the NRC staffs response to Public Comment 6-8 for the reason for these differences.

11-39 Nuclear Energy Institute Comment Calculations typically consider a single failure which is either assumed to be a break in a steam line leading to not modeling holdup/deposition in a steam line upstream of the inboard MSIV or assumed to be a stuck open MSIV which leads to combining the inboard volume with the between MSIV volume in a single steam line. The discussion states that a pipe break is not assumed, so should all applicants consider the limiting single failure to be a stuck open MSIV (with respect to deposition in the steam lines)?

Commenters Proposed Resolution Clarify that there is no in-board pipe break in the MHA scenario.

Comment Response The NRC staff agrees with this comment. No inboard pipe break is assumed in the MHA scenario. No changes were made in response this comment.

11-40 Nuclear Energy Institute Comment The State-of-the-Art report recommends a AMMD of 3 µm for containment and Section A-5.6.2 describes the approved approach as applying an aerodynamic mass mean diameter (AMMD) of 2.0 µm. This appears to be an average value of the RCS and containment AMMD values of 1.0 and 3.0 µm respectively. Considering the MHA assumption of a well-mixed drywell, the containment AMMD of 3.0 µm would be the applicable parameter on which to base the aerosol size distribution entering the main steam lines.

Commenters Proposed Resolution Revise the text to apply a 3.0 µm AMMD or explain the basis for the suggested value of 2.0 µm.

Comment Response Please note that the reference to Section A-5.6.2 in the comment is referring to Regulatory Position A-5.6.2.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 The NRC staff disagrees with the comment and suggestion to revise the assumed aerosol sizes to 3 microns AMMD. The staff reviewed the open literature and considered the potential sources of aerosols for this leakage pathway. Based upon this review and the potential range of sizes for sources of aerosols for this pathway, the NRC staff selected an AMMD value of 2 to be reasonable for DBA radiological consequence analyses. No changes were made in response to this comment. The staff addresses a similar comment regarding aerosol sizes, but requesting an aerosol distribution of 1.0 micron AMMD, in response to Public Comment 17-1.

11-41 Nuclear Energy Institute Comment These methods are based on the AEB 98-03 methodology which does not account for thermophoresis, impaction, diffusiophoresis, flow irregularities, and hygroscopicity. The removal calculated is largely correlated to particle size but because it does not include these other mechanisms it is overly conservative. For example, Section 2.6.3 of the State- of-the-Art Report states that thermophoretic deposition velocity is not an especially strong function of particle size.

Commenters Proposed Resolution Clarify that other associated removal mechanisms may be calculated separately and included in the TEARE.

Comment Response The NRC staff agrees with this comment. While other removal processes are not included in Models 2 and 3 of Regulatory Position A-5, the staff agrees that credit for other valid removal mechanisms could potentially be added to Models 2 and 3 (two models that calculate a total effective aerosol removal efficiency (TEARE)). Notably, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. Although the NRC staff does not currently have a model developed for applying other removal mechanisms to Models 2 and 3, alternatives may be used if a licensee demonstrates that the alternatives satisfy the applicable NRC requirements. Additionally, other removal processes are included in Model 1 derived from SAND2008-6601, and, like Model 1, they can be included in the removal coefficients or TEARE.

No changes were made in response to this comment.

11-42 Nuclear Energy Institute Comment Section A-5.6.4 indicates that aerosol deposition in the condenser using a multi-group or numerical integration approach needs to be evaluated on an individual case basis. Why would the multi-group or numerical integration approaches not be applicable in the condenser? These 95

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 models consider the impact of easier-to-remove particles being removed in the upstream compartments, leading to less deposition in the condenser. Can the condenser deposition coefficients in Table A-1 be applied instead of a value developed with the multi-group or numerical integration approaches?

Commenters Proposed Resolution Clarify use of Table A-1 condenser deposition coefficients or clarify why the multi-group method is not applicable to the condenser.

Comment Response Please note that the reference to Section A-5.6.4 in the comment is referring to Regulatory Position A-5.6.4.

The NRC staff agrees with this comment. Regulatory Position A-5.6.4 provides guidance on calculating TEAREs using the multi-group and numerical integration methods described in the guidance. Table A-1 removal coefficients are derived from other methods described in SAND2008-6601. The NRC staff has not evaluated a method that combines using table A-1 removal coefficients with TEAREs. Therefore, RG 1.183, Revision 1, does not provide a method acceptable for demonstrating compliance with 10 CFR 50.67 when the results of these methods are combined. However, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements.

In response to the comment to clarify the applicability of the multi-group method to the condenser, the NRC staff revised Regulatory Position A-5.6.4 to read as follows: Aerosol deposition removal coefficients for the condenser using a multi-group method and numerical integration are acceptable and will be evaluated on an individual case basis.

11-43 Nuclear Energy Institute Comment Footnote 1 in Appendix B presents that if a postulated event (heavy load drop) occurs for an FHA, the activity release may be based on core average gap fractions presented for the LOCA in Tables 1 and 2. The following issues should be addressed by the NRC:

Tables 1 and 2 identify gap fraction releases for the Te, Ba, and Sr groups. As identified in a previous comment, the SAND-2011-0128 report, in which Tables 1 and 2 are based, clarified that these releases listed in the gap phase were more accurately classified as in-vessel releases that occurred during the gap release phase. As a heavy load drop event is a shutdown event and fuel damage should only result from a load impact, there would not be a driver for further fuel failure associated with in-vessel releases (which results from fuel melt). Therefore, these nuclide groups should be excluded from consideration in the fuel handling accidents involving the entire core.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023

The Tables 1 and 2 release fractions are associated with the timings presented in Table 5. Table 5 presents gap phase durations of 0.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months (PWRs) and 0.16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months (BWRs). The timings in Table 5 are based on simulations of various plant transients which do not include fuel handling accidents / heavy load drop events. Therefore, it should be specified that the use of Tables 1 and 2 do not need to explicitly follow the gap phase durations from Table 5 and can instead use the durations presented in this appendix (Appendix B).

Commenters Proposed Resolution Replace the footnote wording with the following to clarify the gap fractions and timings to apply for the heavy load drop accident: These assumptions may also be used in assessing the radiological consequences of a heavy load drop over fuel accident. If the event is postulated to damage all of the rods in the core, the release activity may be based on the core-average gap fractions of Tables 1 and 2 in the main text of the guide, and the radial peaking factor may be omitted. Gap fractions may be limited to the noble gas, halogen, and alkali metal groups, consistent with Tables 3 and 4. Additionally, the release timings and durations associated with this appendix may be used in lieu of release timings and durations associated with Regulatory Position 3.2 for Tables 1 and 2.

Comment Response The NRC staff agrees that the heavy load drop event would not include radionuclides from a source term generated from a fuel melt event. However, the NRC staff disagrees with the recommended language in the proposed resolution. Instead, the staff deleted the footnote on page B-1 of DG-1389 in response to this comment, and licensees would follow source term guidance already provided in Regulatory Position B-1. Regulatory Position 3.2 of RG 1.183, Revision 1, states, For non-LOCA DBAs, table 3 (for BWRs) and table 4 (for PWRs) list the maximum steady-state fission product release fractions residing in the fuel rod void volume (plenum and pellet-to-cladding gap), by radionuclide groups, available for release upon cladding breach. Additionally, the staff removed the statement (e.g., heavy load drop accident) in the third paragraph of Regulatory Position 3.2 in DG-1389.

11-44 Nuclear Energy Institute Comment In the current licensing basis fuel handling accidents, the impacts of stable and long-lived iodine isotopes (e.g., I-127 and I-129) are not typically considered due to their negligible dose consequences. For the revised fuel handling accident approach, these isotopes are important to develop the pool iodine concentration as they represent most of the iodine inventory in the fuel rods. The NRCs assessment in the staffs example in ADAMS Accession No. ML21190A040 appears to apply a much larger value of 23% for Iodine-129 which is significantly higher than the 4% from Table 4 for other halogens.

Commenters Proposed Resolution Confirm the applicability of Tables 3 and 4 for I-127 and I-129 as other halogens.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff agrees that other halogens in table 3, BWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, and table 4, PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap include I-127 and I-129. These values in tables 3 and 4 are used to estimate the total iodine in the fuel rod plenum and gap that can be released to the pool during a design-basis FHA. Accounting for both the radioactive and nonradioactive iodine in the pool is important in determining how much radioactive iodine evolves from the pool and, therefore, has an impact on the consequences calculated pursuant to 10 CFR 50.67. The NRC staff revised Regulatory Position B-3, Phase 2 ReleaseRe-Evolution Release, to clarify that the total iodine considered for the calculation of iodine release from the pool for the Phase 2 calculation should include all radioactive and stable iodine (such as I-127), because the regulatory position previously only specified I-131 and I-129 in its equations. The appendix B model includes releases of radioactive iodine evolving from the pool for the purposes of computing radiological consequence.

11-45 Nuclear Energy Institute Comment DG-1389 only reports the iodine species distribution released into the pool water. Reg Guide 1.183 Rev. 0 reported that the iodine chemical species above the water is 57% elemental and 43% organic. For Rev. 1 applications, what iodine species distribution should be applied for the early 2-hour airborne release considering the new fuel handling accident model? Based on the calculated overall pool DF from Equations 3-1, 3-2, and 3-3, is a species-dependent DF approach acceptable with organic iodine having a DF of 1? Is a purely elemental iodine species appropriate for the long-term release since this is comprised of re-evolved elemental iodine?

Commenters Proposed Resolution Provide additional guidance on acceptable airborne iodine species assumptions.

Comment Response Please see the NRC staffs response to Public Comment 7-10.

11-46 Nuclear Energy Institute Comment There can be fuel handling accident scenarios where the water depth may be greater than 23 feet such as drops over the core. However, Section B-2 states that the DF can be calculated from Equations in B-1, B-2, and B-3, if the water depth is between 19 and 23 feet. There is no guidance in the event the water depth is greater than 23 feet. Since the DG-1389 equations are 98

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 not dependent on depth, it would be expected that they would yield a conservatively low DF than an actual value for cases that credit the additional scrubbing depth.

Commenters Proposed Resolution Revise statement as follows: For water depths greater than or equal to 19 feet, an overall iodine DF based on pin pressure is computed as follows:

Comment Response Please note that the reference to Section B-2 in the comment is referring to Regulatory Position B-2.

The NRC staff disagrees with this comment. Please see the NRC staffs response to Public Comment 10-16 regarding the comment that there is no guidance for depths greater than 23 feet and the request to revise Regulatory Position B-2 used for calculating DFs.

11-47 Nuclear Energy Institute Comment The mass transfer coefficient (KL) is presented as a constant value (3.66x10-6 m/s). From of Reference B-1, this is a combined liquid-gas phase coefficient that does not consider recirculation in the pool and is based on the assumption of a high flow rate to clear the building air.

Commenters Proposed Resolution The NRC should clarify that the mass transfer coefficient remains applicable for all conditions (i.e., recirculation in the pool and low flow rate to clear the building air) and no adjustments need to be made for changing conditions.

Comment Response The NRC staff disagrees with this comment. This mass transfer coefficient was intended to be bounding for pool-to-atmosphere mass transfer in the FHA scenario by scaling experimental mass transfer based on computational fluid dynamics recirculation analyses due to natural recirculation in the pool. Lower recirculation leads to a lower liquid-phase mass transfer coefficient. The liquid-phase mass transfer was the rate-limiting process, so higher gas flow rates and gas-phase mass transfer rate do increase the overall mass transfer coefficient. Details of the evaluation of the pool mass transfer coefficient can be found in RG 1.183, Revision 1, Reference B-1, Re-evaluation of the Fission Product Release and Transport for the Design-Basis Accident Fuel Handling Accident (ML19248C647), section 4.3.3, which was based on experiments intended to bound possible iodine evolution scenarios by considering both stagnant and recirculating conditions. However, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and 99

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

11-48 Nuclear Energy Institute Comment Step 2 of the FHA model instructs the analyst to calculate the amount of iodine in the fuel pin gap. The calculations consider I-131 for the amount of radioactive iodine (with subsequent clarification to include other radioactive iodine isotopes to the total) and consider I-129 for the amount of non-radioactive iodine (with no subsequent clarification to include other non-radioactive iodine isotopes). The radioactive and non-radioactive iodines are then combined for a total iodine which is used in the downstream calculations. [Note that the individual components are not used in any downstream calculations within the model and calculations involving only the radioactivity iodine can be removed from the model development.] It was observed in the calculations that a higher total iodine amount is conservative for calculating the release rate to be used in the analysis. As such, additional non-radioactive iodine isotopes (I-127) should conservatively be considered.

Commenters Proposed Resolution The NRC should specify that additional non-radioactive iodine isotopes (especially I-127) should be considered to maximize the amount of iodine considered in Step 2 of the FHA model calculation.

Comment Response The NRC staff agrees that the guidance in RG 1.183, Revision 1, should specify that nonradioactive iodine is considered as part of the FHA model calculation. This information is identified in Regulation Position B-3. Specifically, Regulatory Position B-3, Step 1, Calculate amount of iodine in the fuel pin gap , states that [b]oth the radioactive and the nonradioactive iodine (e.g., I-131 and I-127) in the pool affect the radioactive iodine evolution.

The staff revised Regulatory Position B-3 to clarify that the total iodine considered for the calculation of iodine release from the pool for the Phase 2 calculation should include all radioactive and stable iodine (such as I-127), because the regulatory position previously only specified I-131 and I-129 in its equations. Please also see the NRC staffs response to Public Comment 11-44 for a similar comment regarding I-127.

11-49 Nuclear Energy Institute Comment Equation B-10 in Step 2 calculates the [I2]/[I-]2 concentration ratio using the following equation:

Ri = [I2]/I-]2 = Ch2 / 6.05E-14 + 1.47E-09 Ch) 100

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Constants 6.0603E-14 and 1.4708E-09 originate from Ref. B-2 (NUREG/CR-5950); however, the constants are not presented in the reference to that degree of accuracy. The constants presented in NUREG/CR-5950 are 6.05E-14 and 1.47E-09 [NUREG/CR-5950 page 13 and Appendix C, page C.3]. It is unclear where the discrepancy is occurring.

Commenters Proposed Resolution The NRC should provide clarification as to the discrepancy in constants between NUREG/CR-5950 and DG-1389 or update Equation B-10 to present the constants to the degree of precision from NUREG/CR-5950.

Comment Response The NRC staff agrees with this comment. The staff adjusted the constants in Equation B-9 (previously Equation B-10 in DG-1389) to reflect the significant figures presented in NUREG/CR-5950, Iodine Evolution and pH Control, issued December 1992 (ML063460464).

11-50 Nuclear Energy Institute Comment Position B-5.1 provides a release duration time for the first phase releases (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ). The position does not provide any specifics for onset and duration timing for the second phase releases (re-evolution releases). From the Ref. B-1 model background, this release phase should begin immediately after the first phase (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) and be modeled until the typical end of the radiological considerations (30 days). It is assumed that there is no overlap of release phases. Note that this comment is also applicable to Position B-6.3.

Commenters Proposed Resolution The NRC should clarify the starting time and duration of the re-evolution phase.

Comment Response The NRC staff agrees with the comment to clarify the starting time and duration of the re-evolution phase. Rather than providing these clarifications in both Regulatory Positions B-5.1 and B-6.3, the NRC staff is giving them in a different regulatory position to avoid their unnecessary duplication. The NRC staff revised Regulatory Position B-1.3 to clarify the onset and end time for Phase 2 (re-evolution phase). To be consistent with the changes made to the Phase 2 description, the NRC staff also revised the description of Regulatory Position B-1.3, Phase 1, to also provide the onset and end time for Phase 1.

11-51 Nuclear Energy Institute Comment 101

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Footnote 4 breaks across pages B-5 and B-6. This footnote should be contained on a single page. Note that this footnote is referenced in Positions B-6.1 and B-6.2 which are currently on separate pages, so Footnote 4 may have to be repeated on separate pages. Likewise, Footnote 3 for Position B-6.4 is not located on the same page as the position.

Commenters Proposed Resolution See comment for the items to address.

Comment Response The NRC staff agrees with this comment. The staff properly aligned the footnotes in RG 1.183, Revision 1, with the sections.

11-52 Nuclear Energy Institute Comment Revision 0 of RG 1.183 presented the PWR Main Steamline Break event in Appendix E. There is no indication as to the motive behind the switch in appendices for DG-1389. This may cause confusion to licensees which need to compare/contrast the RG revisions or those that elect to implement and maintain different RG revision methods for these events.

Commenters Proposed Resolution The NRC should consider maintaining consistency in the event order presented in the regulatory guide appendices.

Comment Response The NRC staff disagrees with this comment. Please see the NRC staffs response to Public Comment 10-17 for more information regarding the order of the appendices in RG 1.183, Revision 1.

11-53 Nuclear Energy Institute Comment The page numbering for Appendix E is not correct as the first page of the appendix is identified as page E-3.

Commenters Proposed Resolution The NRC should correct the page number in Appendix E and also review the entirety of the regulatory guide to correct any further editorial errors.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response The NRC staff agrees with this comment and has revised RG 1.183, Revision 1, in response to this public comment.

11-54 Nuclear Energy Institute Comment Position E-6.1 states that the primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate for the limiting condition for operation specified in the TSs. This leak rate is associated with significant pressure differentials across the steam generator tubes which typically occur early in the transient. In later periods of the transient, and especially during longer-term steaming for cooldown, the pressure differential is significantly decreased.

Licensees should be able to credit this significant reduction in leakage if it is shown to be credible. This comment is also applicable for other steam release events which model primary-to-secondary leakage. [MSLB in Appendix F, Locked Rotor in Appendix G, and Rod Ejection in Appendix H]

Commenters Proposed Resolution The NRC should add a statement that licensees may credit a reduction in the TS leakage rates for later periods of the transient when pressure differentials across the steam generator tubes are significantly reduced based on adequate technical justifications.

Comment Response The NRC staff agrees with this comment and has made changes to RG 1.183, Revision 1, based upon this and another similar public comment. Please see the NRC staffs response to Public Comment 8-9 for more details.

11-55 Nuclear Energy Institute Comment Position E-6.6 identifies that the potential impact of tube uncovery on the transport model parameters needs to be considered. The issue of tube uncovery was addressed by the Westinghouse Owners Group (WOG) in WCAP-13247, Report on the Methodology for the Resolution of the Steam Generator Tube Uncovery Issue, March 1992. The WOG program concluded that the effect of tube uncovery would be essentially negligible and the issue could be closed without any further investigation or generic restrictions. This position was accepted by the NRC in a letter dated March 10, 1993, from Robert C. Jones, Chief of the Reactor System Branch, to Lawrence A. Walsh, Chairman of the WOG. The letter states ... the Westinghouse analyses demonstrate that the effects of partial steam generator tube uncovery on the iodine release for SGTR and non-SGTR events is negligible. Therefore, we agree with your position on 103

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 this matter and consider this issue to be resolved. Consistent with this position, tube uncovery should not need to be addressed further for U-tube style steam generators. This comment also applies to Positions F-6.6.3 for MSLB, G-5.5 for Locked Rotor, and H-7.4 for Rod Ejection.

Commenters Proposed Resolution The NRC should update the wording of this position to acknowledge for U-tube steam generators that the issue of short-term tube uncovery has previously been resolved and does not need to be considered in future applications.

Comment Response The NRC staff does not agree with the recommendation to state within the referenced regulatory positions that short-term tube uncovery does not need to be considered in future applications. The referenced regulatory positions are based upon Information Notice No. 88-31, Steam Generator Tube Rupture Analysis Deficiency, dated May 25, 1988 (ML031150151).

Information Notice No. 88-31 applies to all Westinghouse and Combustion Engineering designs.

However, the WCAP-13247 study was performed for Westinghouse design steam generators considered most susceptible at that time (e.g., Models 44 and 51). However, the applicability of WCAP-13247 and the NRC staffs March 10, 1993, letter referenced in this comment to not consider the impact of uncovery may be acceptable if the applicant can demonstrate that it is applicable to current steam generators. No changes were made in response to this comment.

11-56 Nuclear Energy Institute Comment Appendix I is not referenced within DG-1389 and the purpose and importance of this appendix is not understood. Given the unspecified use and importance of this appendix, it is recommended the NRC remove the appendix and flowchart.

Commenters Proposed Resolution NRC should delete this appendix or provide reference in document as to why this is needed.

Comment Response The NRC staff agrees with this comment. Appendix I, Analysis Decision Flowchart, was deleted and will not be included in RG 1.183, Revision 1.

11-57 Nuclear Energy Institute Comment The first paragraph of this appendix cross-references Section C of the main body of this guide 104

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 for the release fractions in Tables 3 and 4. This is an editorial error as Tables 3 and 4 are located in Section 3.2.

Commenters Proposed Resolution The NRC should correct this cross-reference and review DG-1389 for any other potential cross-referencing errors. Section C position 3.2.

Comment Response Please note that the reference to Section 3.2 in the comment is referring to Regulatory Position A-3.2.

The NRC staff agrees with this comment. The NRC staff revised RG 1.183, Revision 1, to fix the cross-referencing error.

11-58 Nuclear Energy Institute Comment Of particular concern is that the DG does not include the results of the staffs PWR and BWR analyses to enable an assessment of the impact to analyzed doses resulting from the updated guidance, such as impact from the new release fractions and timing. Industry views the proposed changes in this DG just as significant as when the guidance transitioned from the TID Source Term to the Alternative Source Term. As proposed, the conservatisms and changes incorporated in this revision precludes its use by many plants that are interested in implementing near-term ATF design concepts, fuel burnup extension to 68 GWd/MTU (peak rod average), and 235U enrichments up to 8.0wt %.

Comment Response This comment is taken from the Nuclear Energy Institute (NEI) letter dated June 21, 2022, that transmitted the NEI comments on DG-1389. The above comment reiterates the information in NEI Comment 11-4 regarding a request that the NRC use the updated guidance in DG-1389 to determine the impact of the DG on a sample of PWRs and BWRs (like the rebaselining analyses documented in SECY-98-154, Results of the Revised (NUREG 1465) Source Term Rebaselining for Operating Reactors, dated June 30, 1998, (ML992880064). Public Comment 11-4 also requested that the NRC include the results of these analyses in DG-1389.

The NRC staff disagrees with the comment. As described in the response to Public Comment 11-4, the NRC staff previously performed rebaselining to evaluate and provide the technical basis for rulemaking when transitioning to what became 10 CFR 50.67. However, the revision of RG 1.183 from Revision 0 to Revision 1 does not involve a change in regulations but is only a change in the regulatory guidance. Therefore, the NRC staff did not perform a rebaselining similar to that performed for SECY-98-154. Also, the information requested does not provide guidance or regulatory positions on methods that could be acceptable to meet the requirements in 10 CFR 50.67. Therefore, no changes were made in response to this comment.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 11-59 Nuclear Energy Institute Comment A comparison of the changes in MHA release fractions between BWRs (Table 1) and PWRs (Table 2) identified a significant increase in the BWR halogen release fractions with no indication in either SAND2011-0128 or DG-1389 as to the cause. This increase may adversely affect the ability of BWRs to comply with RG 1.183 Rev. 1. A comparison to the PWR analyses suggests that the accident sequences may be responsible for this impact. SAND2011-0128 updates NUREG-1465 with higher core exposures utilizing the latest NRCs MELCOR methodology. The accident sequences that were analyzed to develop the PWR release fractions are listed in Table 5 of the Sandia report and include a variety of accident types.

However, for the BWR release fractions, Table 3 of the Sandia report indicates that nearly all the evaluations were based on station blackout (SBO) sequences. However, the risk from SBO events has been substantially reduced by the industrys implementation of FLEX. Consistent with the PWR analysis in SAND2011-0128, the BWR release fractions should be re-evaluated to ensure that they are based on an appropriate set of accident sequences that more accurately reflect BWR risk profiles.

Additionally, it is unclear from the SAND2011-0128 report whether suppression pool scrubbing was credited in determining the release fractions from a BWR SBO. Credit for suppression pool scrubbing can significantly decrease the airborne activity since the SBO-related releases would be released via spargers submerged in the suppression pool. Therefore, any sequences that involve releases through the pool spargers should take credit for suppression pool scrubbing Comment Response This comment is taken from the NEI letter dated June 21, 2022, that transmitted the NEI comments on DG-1389. The NRC staff disagrees with the comment.

Regarding the comment that neither SAND2011-0128 nor DG-1389 provided the cause for the increase in halogens, please see the NRC staff response to Public Comment 7-2.

The method for modeling suppression pool scrubbing is consistent with NUREG-1465 that provided the basis for the source term in RG 1.183, Revision 0. SAND2011-0128 presents an in-containment source term consistent with past regulatory practices used to demonstrate compliance with applicable regulatory requirements. For BWRs, the containment is composed of both a wet well, which contains a suppression pool, and a drywell. The modeling simulations used to derive the in-containment source term include the effects of suppression pool scrubbing.

Therefore, the BWR in-containment source term includes fission products retained within both volumes of the containment. Please also see NRC staff responses to Public Comments 6-1 and 11-28 for related comments on suppression pool scrubbing.

No changes were made in response to this comment.

11-60 Nuclear Energy Institute 106

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Also, although new guidance was added for crediting holdup and retention of MSIV leakage within the main steamlines and condenser for BWRs, it contains more conservative assumptions than previous models that limits its effectiveness. For example, Section A-5 of the DG presents three acceptable methods for calculating aerosol deposition within the main steam lines, but also states, these methods are not valid if credit has been taken for aerosol removal from drywell sprays. No technical justification is provided for why aerosol removal from drywell sprays is not valid if used with credit for main steamline deposition. As a result, the effectiveness of the application of these models is significantly reduced. Considering the number of BWRs currently modeling both removal mechanisms, the DG should provide guidance for crediting both of these important mitigative features.

Comment Response Please note that the reference to Section A-5 in the comment is referring to Regulatory Position A-5.

This comment is taken from the NEI letter dated June 21, 2022, that transmitted the NEI comments on DG-1389. The NRC staff disagrees with the comment.

Regarding the technical justification for why aerosol removal from drywell sprays is not valid if used with credit for main steamline deposition, please see the NRC staff response to Public Comment 11-31, which asked for this technical justification.

Regarding the comment that the RG should provide guidance for crediting both sprays and steamline deposition, please see the NRC staff response to Public Comment 6-7, which provided the same recommendation.

No changes were made in response to this comment.

11-61 Nuclear Energy Institute Comment Further, the DG states, Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations. A combination of the methods contained in revision 0 or revision 1 of RG 1.183 would need additional justification. However, it is not clear in the proposed document if licensees can transition one (or a few) analyses to revision 1, or if full implementation for all design basis accidents is required. Clarification should be added to discuss that selective implementation is acceptable, provided that each accident analysis uses either revision 0 or revision 1 and to specify that a combination of the methods contained in revision 0 or revision 1 of RG 1.183, in a single analysis, would need additional justification.

Comment Response 107

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 This comment is taken from the NEI letter dated June 21, 2022, that transmitted the NEI comments on DG-1389. The NRC staff agrees with the need to add clarification to the guidance.

In response to this comment and Public Comment 11-2, which also requested related clarifications, the NRC staff added guidance on the use of Revision 0 and Revision 1 to RG 1.183, Revision 1. The NRC staff added examples to the Background section and Regulatory Position 1.1.5 of RG 1.183, Revision 1, in response to this comment. Please also see the NRC staff response to Public Comment 8-1, which is a related comment that requests the staff to revise the language to the paragraph discussed in Public Comment 11-2.

Comment Submission 12 ADAMS Accession No. ML22174A074 Name: Anonymous Email: slafountai3@gmail.com 12-1 Anonymous (slafountai3)

Comment Regulatory Guide 1.183 Revision 0, Appendix C (BWR rod drop accident) and Appendix H (PWR rod ejection accident), both contain explicit guidance on the release fractions to assume associated with the fraction of fuel that reaches or exceeds the initiation temperature for fuel melting (e.g., Appendix C says that 100% of the noble gases and 50% of the iodines contained in the fraction of fuel that melts are released to the reactor coolant). The equivalent appendices in DG-1389 do not contain guidance on event specific release fractions to assume for fuel that reaches the melting temperature. It is not clear if the earlier reference in these appendices to Regulatory Position 3.2 for the basis of the fission product release from breached fuel is intended to apply to just the gap release, or if this is telling the analyst to use in-vessel release fractions from Tables 1 or 2 as applicable, or something else entirely.

Commenters Proposed Resolution Recommend revision of Positions C-1 and H-1 to more clearly provide guidance on how to treat releases from any fraction of fuel postulated to melt as a result of these reactivity insertion events.

Comment Response The NRC staff agrees with this comment. The NRC staff revised Regulatory Positions C-1 and H-1 in RG 1.183, Revision 1, to include guidance from RG 1.183, Revision 0, which was removed in DG-1389, regarding more specific guidelines for the treatment of limited fuel melting in PWR rod ejection and BWR rod drop accident source terms.

Comment Submission 13 ADAMS Accession No. ML22174A081 108

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Name: William H. Horin Organization: Nuclear Utility Group on Equipment Qualification (NUGEQ)

Email: whorin@winston.com Address: Winston & Strawn LLP, 1901 L Street, NW, Washington, DC 20036-0081 13-1 Nuclear Utility Group on Equipment Qualification Comment General Observation: Some of the drivers for revising RG 1.183 are also generically applicable to radiological source terms based on TID-14844. Specifically, Section B. DiscussionReason for Revision (6), to add guidance for Accident Tolerant fuel (ATF), Mixed Oxide (MOX) fuel, higher burnup or higher enrichment source term analysis could also be relevant to source terms based on TID-14844 in a similar manner to Alternative Source Terms (AST). As a result, a corresponding revision to RG 1.195 appears warranted.

Commenters Proposed Resolution NUGEQ is not aware of any ongoing actions by the staff to update/revise RG 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design-Basis Accidents at Light-Water Nuclear Power Reactors. This regulatory guide was last reviewed by the NRC in 2016. Any update to RG 1.195 should provide guidance on using TID-14844 timing and distribution assumptions with high burnup or high enrichment fuel to address the regulatory position in C.3.1 of RG 1.195, Core inventory factors (curies per megawatt thermal) provided in TID-14844 and used in some analysis computer codes were derived for low-burnup, low-enrichment fuel and should not be used with higher burnup and higher enrichment fuels.

Comment Response This comment is outside the scope because revisions to RG 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Plants are not part of the revisions to RG 1.183, Revision 1. No changes were made in response to this comment.

13-2 Nuclear Utility Group on Equipment Qualification Comment The proposed wording in the background section should be clarified to clearly reflect that the guidance in Appendix I to Revision 0 of RG 1.183 is specific to operating reactors that have amended their licensing basis to use AST for EQ.

Commenters Proposed Resolution The intent of the comment is to reflect closure of the staffs interim position in Section C.6, Assumptions for Evaluating the Radiation Doses for Equipment Qualification from RG 1.183 R0. Also See Comment #3.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Response Please note that the reference to Section C.6 in the comment is referring to Regulatory Position 6 in RG 1.183, Revision 0.

The NRC staff disagrees with this comment. The EQ guidance in RG 1.183, Revision 0, appendix I, is not only used by operating reactors that have amended their licensing bases for AST but also by 10 CFR Part 52 licensees and design certification holders. The staff has removed the reference to EQ guidance provided in appendix I to RG 1.183, Revision 0, from RG 1.183, Revision 1, in light of the issuance of RG 1.89, Revision 2, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, issued April 2023. Updated radiological EQ guidance, similar to that provided in appendix I to RG 1.183, Revision 0, has been incorporated into RG 1.89, Revision 2.

As indicated in the Background section of DG-1389, Revision 0 of RG 1.183 will continue to be available for use by licensees and applicants as a method acceptable to the NRC staff for demonstrating compliance with the regulations when applied within the applicability of that method. However, in using RG 1.183, Revision 0, if the proposed site-specific implementations of the AST conflict with the licensing basis or guidance (e.g., increase in burnup above the 62,000 MWd/MTU of 5 weight-percent uranium-235 enrichment assumed in the Revision 0 methods), licensees need to consider and address any new or unreviewed issues created and ensure that the proposed implementation of the guidance in RG 1.183, Revision 0, is technically justified. In addition, licensees following the guidance in RG 1.183, Revision 0, for EQ are not required to update to RG 1.89, Revision 2. In order to clarify the staff position regarding the use of EQ guidance in RG 1.183, Revision 0, for licensees and applicants, the NRC staff incorporated changes into the Background section of RG 1.183, Revision 1, consistent with this response.

13-3 Nuclear Utility Group on Equipment Qualification Comment This comment is intended to cover those Part 50 licensees who have adopted full or selective implementation of § 50.67 as described in Sections 1.1.3 and 1.2 of DG-1389, but retained TID-14844 as the source term for environmental qualification of equipment. DG-1389 should clarify or otherwise specifically address the ability of a licensee to continue to use source terms based on TID-14844 for Environmental Qualification consistent with the licensing basis of the plant. Specifically, the wording in Section C.1.3.5 of DG-1389 should be reworded to reflect, that consistent with a plants licensing basis and the resolution of GSI-187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump licensees may continue to utilize the TID-14844 radiological source term to establish environmental qualification of equipment subject to 10 CFR 50.49.

Commenters Proposed Resolution To clarify that the applicability of AST to EQ is specific to licensees who have amended their licensing basis to apply AST for environmental qualification of equipment under § 50.49. This clarification would result in consistency with the resolution of GSI-187 as well as the Statement 110

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 of Considerations for 10 CFR 50.67, which states The NRC considered the applicability of the revised source terms to operating reactors and determined that the current analytical approach based on the TID-14844 source term would continue to be adequate to protect public health and safety, and that operating reactors licensed under this approach would not be required to reanalyze accidents using the revised source terms. See 64 FR 71992.See References 1 and 2 below for the basis for closure of GSI-187. The resolution of GSI-187 occurred after the issuance of RG 1.183, R0. As noted in Reference 2 [ML011210348], The panel has decided that the candidate generic issue should be dropped, as having no significant chance of meeting the incremental risk thresholds for backfit as described in the MD 6.4 Handbook.

REFERENCES:

1) NUREG-0933, Main Report with Supplements 1-35, Section 3. New Generic Issues Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification. [ML21251A113]

2) Memorandum for A. Thadani from J. Rosenthal, Initial Screening of Candidate Generic Issue 187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump April 30, 2001. [ML011210348]

Comment Response Please note that the reference to Sections 1.1.3 and 1.2 in the comment is referring to Regulatory Position 1.1.3 and 1.2.

The NRC staff partially agrees with this comment. The NRC staff agrees that the TID-14844 source term continues to be adequate for those plants that currently use TID-14844 for the EQ analysis in their current licensing basis and have not made significant plant modifications affecting source terms or the EQ analysis. The NRC staff notes that these plants are not required to update their EQ analyses based on the update to RG 1.183.

However, the staff disagrees that the findings in Generic Safety Issue (GSI)-187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump, dated April 30, 2011, are directly applicable to RG 1.183, Revision 1, because GSI-187 reviewed the differences to EQ dose comparing TID-14844 to Revision 0 of RG 1.183. Since RG 1.183, Revision 1, provides a different source term than RG 1.183, Revision 0, the conclusions made with the resolution of GSI-187 are not necessarily applicable to RG 1.183, Revision 1. Therefore, the acceptability of continuing to use TID-14844 source term methodology for EQ for licensees that have modified their licensing bases to higher burnup or enrichment than those specified in RG 1.183, Revision 0 (above the 62,000 MWd/MTU of 5 weight-percent uranium-235), has not been assessed. Therefore, when these types of significant plant modifications occur, licensees need to evaluate the impacts on EQ consistent with the modifications made and make updates as appropriate. If the licensee wishes to continue to use TID-14844 for EQ after increasing above the parameters specified in RG 1.183, Revision 0, or when making other plant modifications that affect the applicability of previously approved methods or assumptions, the licensee must demonstrate that how it is using the guidance will satisfy applicable NRC requirements. No changes were made as a result of this comment because the guidance provided in Regulatory Position 1.3.5 is adequate.

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Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Comment Submission 14 ADAMS Accession No. ML22174A082 Name: Anonymous Email: donnie766@gmail.com 14-1 Anonymous Comment Reference A-14 concludes that there is high confidence that the main steam lines and structures, systems and components in the alternative pathway will be available for fission product dilution, hold-up, and retention. We disagree with this conclusion, that is based almost entirely on the inherent seismic robustness of the condenser and its associated piping and does not address the functionality of the components. While Regulatory Position A-5.4 touches on these issues it is incomplete. The reliance for credit for the accident analysis needs to be supported by past alternative source term application seismic walkdowns. These walkdowns identified the need for plant modifications to address a variety of seismic related issues, such as block wall impacts on instrument lines that could cause a divergent release path away from the condenser, as well as other types of interactions and anchorage issues. It should address what actions need to be taken to make the condenser available including the need to isolate potentially divergent paths, and what program, procedures, and plant modifications (such as establishing emergency power to the pathway valves and boundary valves that must isolate) must be established to ensure this pathway will be maintained as highly reliable mitigation. Only after these actions are addressed and taken by the licensee should the staff conclude there is high confidence that the pathway to the condenser to warrant credit in the accident analysis.

Comment Response The NRC staff partially agrees with this comment. The comment identifies the need for information on actions and procedures or plant modifications for the alternative pathway. The staff agrees that the functionality of the alternative pathway needs to be supported by procedures and, potentially, by plant modifications to establish the defined pathway. The staff included information in Regulatory Position A-5.4 in RG 1.183, Revision 1, to address the functionality of the components in the alternative pathway identified in the comment. Regulatory Position A-5.4 includes information that should be submitted regarding the basis for the alternative pathways functional reliability (commensurate with its intended safety-related function) and emergency operating procedures that may be required if a highly reliable power source is available or identification of necessary operator actions if a highly reliable power source is unavailable. The NRC staff concludes that the information identified in Regulatory Position A-5.4 includes items in addition to those given in Regulatory Position A-5.5 and would provide support for a reasonable assurance finding regarding the reliability of the pathway to the main condenser, including the seismic capacity of the SSCs in the pathway.

The comment states that condenser credit in the accident analysis is warranted only after the licensee takes any actions related to the alternative pathway. The comment notes that credit for the pathway to the main condenser needs to be supported by past AST application seismic walkdowns, in addition to the information in Reference A-14 in DG-1389. The staff disagrees with these comments. The reasons discussed in the comment are addressed by the staffs ability to ensure that the design basis, as established by the amendment, is satisfied. Past 112

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 applications, such as for an AST, and the corresponding safety evaluation will be available to the staff to identify any information relevant to its review. The NRC staff revised RG 1.183, Revision 1, to include the technical basis supporting the methods for demonstrating seismic robustness of non-safety-related equipment and removed the citation for Reference A-14 that was included in DG-1389 because the regulatory position includes the relevant basis information.

14-2 Anonymous Comment Do you have any idea how much changing all the documents from Loss of coolant accident to MH-LOCA costs a licensee in time, just for the sake of changing back to a term that was removed over 20 years ago? When this terminology is used the source term should not be used incorrectly. It should represent the maximum hypothetical value, not a 50th percentile value for the release fractions and duration.

Comment Response The NRC staff disagrees with this comment. There is no expectation that licensees need to change their current licensing basis documents to accommodate the term MHA LOCA. RGs generally provide an acceptable method for complying with the applicable regulations.

Licensees and applicants are not required to follow the processes and procedures discussed in regulatory guides and may choose to comply with the regulations using alternative approaches.

Thus, the staff does not anticipate that existing licensees would choose to adopt this RG merely to adopt its terminology. However, a recent NRC staff review of current FSARs indicates that several licensees referenced the term MHA for the accident analyzed to show compliance with the regulatory dose acceptance criteria.

With regard to the question about a 50th percentile value for the release, the NRC staff disagrees with the assertion that the MHA LOCA terminology is being used incorrectly. The use of the term MHA LOCA in DG-1389 is intended to clarify the accident that the staff finds acceptable to use to meet the description provided in 10 CFR 50.34 and 10 CFR 50.67, as well as various sections of 10 CFR Part 52; that is, an accident assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products into the containment. The regulations (10 CFR 100.11, 10 CFR 50.34, and 10 CFR 50.67, as well as various sections of 10 CFR Part 52) refer to a substantial meltdown of the core, not complete core melt. The phrase substantial meltdown of the core was detailed in TID-14844, which was the technical basis for 10 CFR 100.11. Later, the phrase substantial meltdown of the core was examined in more detail in NUREG-1465. Insights from NUREG-1465 were incorporated into RG 1.183, Revision 0, providing an alternative set of assumptions that could be used to show compliance with revised dose acceptance criteria expressed in TEDE. No changes were made in response to this comment.

Comment Submission 15 ADAMS Accession No. ML22174A083 113

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Name: Anonymous Email: cliffnukem@gmail.com 15-1 Anonymous Comment Since the source term is being applied to all LWR reactors the maximum site-specific releases, like those for the non-LOCA DBA, for all reactors (not just 4 reactors) needs to be calculated and put in Tables 1 and 2.

Comment Response The NRC staff disagrees with this comment to revise tables 1 and 2 to include calculated releases for all LWR reactors. Tables 1 and 2 provide representative source terms consistent with past practices in NUREG-1465 and RG 1.183, Revision 0. The referenced MHA LOCA updated source terms in tables 1 and 2 reflect the current understanding of severe accidents and fission product behavior since the publication of RG 1.183, Revision 0. For additional information regarding using the representative NUREG-1465 source terms as they relate to the 10 CFR 50.67 rule, see the Statements of Consideration for 10 CFR 50.67 (64 FR 71990, Use of Alternative Source Terms at Operating Reactors, dated; December 23, 1999). To note, RGs describe one or more methods that the NRC staff considers acceptable for meeting the agencys regulatory requirements, and licensees may use alternatives to the guidance if the licensees demonstrate that the alternatives satisfy the applicable NRC requirements. No changes were made in response to this comment.

15-2 Anonymous Comment The source term in DG-1389 should be restricted to use by light water reactor fuel (Uranium dioxide or mixed oxide) with zirconium clad and a maximum burnup of 68,000 MWd/MTU.

Comment Response The NRC staff disagrees with this comment but agrees with including restrictions to clarify the applicability of the source term. The maximum hypothetical accident loss-of-coolant accident (LOCA) source term provided in Regulatory Position 3.2 is applicable to applicants and licensees using zirconium-alloy cladded uranium dioxide (UO2) fuel rod designs with reactor core burnups up to a maximum rod-average of 68 GWd/MTU (and fuel enrichments up to 8 weight-percent uranium-235), including chromium-coated cladding (thicknesses less than 50 microns) and chromia-doped (up to 0.16 weight-percent) fuel. The non-LOCA source term models apply to applicants and licensees with reactor core burnups up to a maximum rod-average burnup of 68 GWd/MTU (and fuel enrichments up to 8 weight-percent uranium-235) for currently approved (as of the issuance of this RG) zirconium-alloy cladded UO2 fuel rod designs at power levels below the burnup-dependent power envelopes depicted in figure 1 of RG 1.183, Revision 1. This applicability differs from the applicability proposed by the comment. To note, 114

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 the staff revised this applicability in Regulatory Position 3.2 from that proposed in DG-1389 in response to Public Comment 2-2 and to reflect the staffs response to this comment. For clarity, the restrictions provided in this response were also added to the Staff Regulatory Guidance in section C of RG 1.183, Revision 1 (before Regulatory Position 1). The basis for this applicability is provided within Regulatory Position 3.2. However, the NRC staff is not aware of technical support or basis for restricting the source term as proposed in the comment.

15-3 Anonymous Comment Restore the terminology loss-of-coolant accident, rather than using a new terminology and new philosophical scenario of the accident that must be considered. It took decades of research to get away from the maximum hypothetical accident and move to a revised source term and more reasonable realistic methodologies that use the scenarios that dont require modeling that conflicts with structures, systems, and components and the phenomenology in codes like MELCOR. In addition, the terminology maximum hypothetical LOCA conflicts with or does not exist in the regulations such as General Design Criteria 19 and 10 CFR 50.67.

Comment Response See responses to Public Comments 4-1, 10-5, and 14-2. No changes were made in response to this comment.

15-4 Anonymous Comment The new guidance specifies an MHA source term in Section 3 that is being applied to a wide range of facility designs. However, the source term is only based upon accident modeling for four facilities that include Surry, Sequoyah, Peach Bottom and Grand Gulf. It our understanding that NUREG-1465 was based upon a source term that was approximately the 70th percentile phase durations and release factions and SAND-2011 contains source terms based upon the 50th percentile phase durations and release factions. These are clearly not maximum hypothetical release source terms but are average hypothetical release fractions and durations.

Therefore, Tables in Section 3 of the revision (for example Tables 1 and 2 values) in RG-1.183 need to be revised to include values that represent maximum hypothetical releases consistent with the 95th percentile if they are to be truly considered maximum hypothetical releases Comment Response Please note that the reference to Section 3 in the comment is referring to Regulatory Positions 3.1, 3.2, and 3.3.

See the NRC staffs responses to Public Comments 4-1, 10-5, and 14-2. No changes were 115

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 made in response to this comment.

15-5 Anonymous Comment Regulatory Position 3.4, Table 6, should be updated to include the molybdenum group consistent with page 18, Section 4.3 of ERI/NRC 11-211, Peer Review of Accident Source Terms for Light-water Nuclear Power Plants using High-Burnup and Mixed Oxide Fuels. In addition, Zr in this table should be moved from the Lanthanide group to the Cerium group consistent with Table 14 of SAND 2011-0128.

Comment Response The NRC staff agrees with this comment. The staff revised table 6 to be consistent with the groupings in SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, (ML20093F003), table 14, and ERI/NRC 11-211, section 4.3, in response to this and similar comments in Public Comments 7-3, 11-15, and 11-20.

15-6 Anonymous Comment Given the history of regulatory guide updates it is a rare exception that the previous version of the regulatory guidance would continue to be acceptable, especially when obvious errors are corrected in the revised version of that guidance. If necessary, a backfit analysis needs to be performed to supersede the known to be erroneous Revision 0 of regulatory guide 1.183 gap fractions and other items fixed in DG- 1389. Revision 0 should not be allowed to continue to exist.

Comment Response The NRC staff does not agree with this comment. Previous versions of RGs (i.e., RG 1.183, Revision 0) will continue to be available for use until the NRC staff determines they need to be withdrawn. NRC Management Directive 6.6, Regulatory Guides, dated July 19, 2022 (ML22010A233), directs the staff to periodically review RGs to determine whether the guidance has become outdated or is no longer needed. A withdrawal is taken to signal that the document no longer reflects the preferred method of meeting the regulatory need. Once a decision is made to withdraw an RG, the staff will prepare a justification and incorporate it into a Federal Register notice announcing the action. However, the withdrawal of an RG does not alter any prior or existing NRC licensing approval or the acceptability of licensee commitments to the RG, and current licensees may continue to use it as documented in the licensees licensing basis.

The NRC staff is not withdrawing RG 1.183, Revision 0, because, despite long-lived, stable radionuclides (e.g., krypton (Kr)-85, cesium-134) release fractions being underpredicted (see 116

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 RG 1.183, Revision 0, table 3), short-lived, volatile radionuclides (e.g., I-131) are overpredicted.

Given the significance of I-131 on predicted radiological consequences, the staff believes that individual plant dose calculations remain conservative.

Because the NRC is not withdrawing RG 1.183, Revision 0, a backfit assessment is not necessary. Since the staff positions in RGs do not constitute requirements, a backfit is not imposed simply by issuing an updated RG. If the commenter wishes to petition the NRC to pursue imposing new requirements on plants licensed to RG 1.183, Revision 0, the commenter may use another NRC process (for example, the process in 10 CFR 2.206, Requests for action under this subpart, or the rulemaking process).

No changes were made in response to this comment.

Comment Submission 16 ADAMS Accession No. ML22174A084 Name: Anonymous Email: jimm@gmail.com 16-1 Anonymous Comment We agree that the containment sprays should not be credited in Regulatory Position A-4, especially when the steam line deposition models proposed in DG-1389 regulatory position A-5 are credited.

Comment Response The NRC staff agrees with this comment. No changes were made in response to this comment.

Comment Submission 17 ADAMS Accession No. ML22174A085 Name: Anonymous Email: nuclearrandy@gmail.com 17-1 Anonymous Comment The change from LOCA to MHA LOCA does not acknowledge progress made beginning with the original Perry amendment for 10 CFR 50.67. The DPO 2020-002 idea that the current RG 1.183, Rev. 0 methods leads to an inappropriate need to impose non-physical assumptions and ignores the physical processes that would lead to such a condition. Therefore, it too, also creates the need to impose nonphysical assumptions. Lines must break for the radioactivity to 117

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 get to the containment and melt the fuel. Now these facts are being ignored with something that is from the early ages of nuclear when we didnt have codes like MELCOR to model the physical phenomena involved. It also ignores the fact that the steamline is connected to the reactor and the source term aerosol sizes will be smaller than those outside the reactor vessel.

Commenters Proposed Resolution Given the distribution in the reactor coolant system is 1 to 2 microns AMMD, using a 2-micron value will lead to non-conservative doses. Keep the methods as they exist in all the license amendments cited in Appendix A of DPO-2020-002 where the cessation of the core melt accident is consistent with the reflood necessary to stop the accident. Given the uncertainty of the accident progression and the location of the reactor vessel as being the closest to the MSIVs, change the aerosol source size distribution to a value that is appropriate for the aerosol distribution coming from the core (1 AMMD) rather than using the value at the top end of the distribution for reactor coolant source term aerosols.

Comment Response The NRC staff disagrees with the comment and suggestion to revise the assumed aerosol sizes to 1 microns AMMD. The staff reviewed the open literature and considered the potential sources of aerosols for this leakage pathway. Based upon this review and the potential range of sizes for sources of aerosols for this pathway, the NRC staff selected an AMMD value of 2 to be reasonable for DBA radiological consequence analyses. No changes were made in response to this comment. The staff addressed a similar comment regarding aerosol sizes but requesting an aerosol distribution of 3.0 micron AMMD in its response to Public Comment 11-40.

17-2 Anonymous Comment We believe that the regulatory position that does not credit deposition up to the main steam lines is appropriate. These volumes are part of the source term volumes. An assumption of instantaneous and homogeneous mixing in these volumes would conflict with the assumption used in the deposition models that only consider gravity and buoyancy are acting on the radioactive particles in these volumes.

Comment Response The NRC staff agrees with this comment. No changes were made in response to this comment because the proposed regulatory guidance already excludes credit for deposition inboard of the main steam isolation valves.

Comment Submission 18 ADAMS Accession No. ML22180A115 Name: B.E. Standley, Director, Nuclear Regulatory Affairs 118

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Organization: Dominion Energy Services, Inc.

Address: Dominion Energy Services, Inc., 5000 Dominion Boulevard, Glen Allen, VA 23060 18-1 Dominion Comment Figure 1: Maximum Allowable Power Operating Envelope for Steady-State Release Fractions provides a bounding power envelope for the applicability of steady-state fission pro duct release fractions residing in the fuel rod plenum and gap. This figure has a title that identifies a peak linear heat generation rate (LHGR) on which the power envelope is based. This title presents:

Peak LHGR = 15.0 KW/ft BWR, 14.0 KW/ft PWR.

Reference 24 of DG-1389 provides the technical basis for the non-LOCA fission product release fractions and defines the Peak LHGR as a combination of rod average power and the axial power profile maximum factors. For example, using a rod average power of 12.2 kW/ft from Figure 1 and an axial power profile maximum factor of approximately 1.15 for PWRs from Figure 2 of Reference 24, the resulting Peak LHGR would be approximately 14.0 kW/ft (consistent with Figure 1).

The maximum linear heat rate is a typical term for licensees (generally defined in safety analyses and/or Technical Specifications) and is the product of the rod average power (or linear heat generation rate determined by the core rated thermal power and the linear component of all power producing rods in the core) and the hot channel factor (Fa). These maximum linear heat rates can exceed the definition of 1 Peak LHGR associated with Figure 1 of DG-1389, which will result in confusion when determining the applicability of the Figure 1 envelope.

Commenters Proposed Resolution The NRC should either:

Delete the "Peak LHGR1 parenthetical from the title of Figure 1 of DG-1389, OR

Include a statement in association with Figure 1 to provide a definition of "Peak LHGR" being used in defining the envelopes.

Suggested wording for the statement is: From Ref. 24, the Peak LHGR is defined as the product of the peak fuel rod average power and the peak fuel rod axial power distribution. This Peak LHGR may differ from the definition used in licensees' Technical Specifications or Core Operating Limits Report. A Peak LHGR derived consistent with the definition of Ref. 24 should be used for comparison of applicability to Figure 1.

Comment Response Please see the NRC staffs response to Public Comment 11-16.

Comment Submission 19 ADAMS Accession No. ML22174A068 119

Pre-decisional copy for the Advisory Committee on Reactor Safeguards to support the public meeting on September 6, 2023 Name: Brian Magnuson Email: magnuson28@msn.com Address: Aurora, IL 60504 19-1 Magnuson These comments, designated as Comment Submission 19, were provided on June 20, 2022 (ML23060A242). These comments were also provided within a second set of comments, designated as Comment Submission 9 (beginning at page 422 of Comment Submission 9 (ML22174A068)). Therefore, the NRC staff provided responses only to Comment Submission 9 (e.g., Public Comment numbers 9-1, 9-2), because the comments in Comment Submission 19 were identical.

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