ML19302E819

Staff Technical Analysis in Support of the NuScale Design Certification Environmental Assessment

ML19302E819
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Issue date:	08/04/2020
From:	Donald Palmrose NRC/NMSS/DREFS/ENRB
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Donald Palmrose 415-3803
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Download: ML19302E819 (48)

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1 Staff Technical Analysis in Support of the NuScale Standard Design Certification Environmental Assessment Table of Contents 1.0 Introduction...................................................................................................................... 3 1.1 Description of the NuScale Design................................................................................. 3 1.2 Severe Accident Mitigation Alternatives and Severe Accident Mitigation Design Alternatives................................................................................................................................ 4 1.3 NuScale Probabilistic Risk Assessment......................................................................... 5 1.4 Review Guidance........................................................................................................... 6 1.5 Environmental Review Process...................................................................................... 6 2.0 Estimate of Risk for NuScale.......................................................................................... 7 2.1 Application of the NuScale PRA..................................................................................... 7 2.2 Offsite Consequences and Risk Analysis....................................................................... 8 2.2.1 Site Selection and Site Parameters......................................................................... 8 2.2.2 MACCS Input Quantification................................................................................... 9 2.2.3 Base Case and Sensitivity Cases........................................................................... 9 2.2.4 Staff Review.......................................................................................................... 10 3.0 Maximum Benefit Evaluation........................................................................................ 13 3.1 Introduction................................................................................................................... 13 3.2 Maximum Benefit.......................................................................................................... 13 3.2.1 Averted Public Exposure (APE)............................................................................ 14 3.2.2 Averted Offsite Property Damage Costs (AOC).................................................... 14 3.2.3 Averted Occupational Exposure (AOE)................................................................. 15 3.2.4 Averted Onsite Cost (AOSC)................................................................................. 16 3.2.5 Maximum Benefit for a Single NuScale Power Module......................................... 18 3.2.6 Maximum Benefit for Twelve NuScale Power Modules......................................... 19 3.3 Maximum Benefit Sensitivity Analyses......................................................................... 20 3.4 Staff Review................................................................................................................. 21 4.0 Potential Plant Improvements....................................................................................... 23 4.1 NuScale Process for Identifying Potential Plant Improvements................................... 23 4.2 Staff Review................................................................................................................. 24 5.0 Cost Benefit Analysis of Potential Plant Improvements............................................ 25 5.1 NuScale Cost Benefit Analysis..................................................................................... 25 5.1.1 Phase I Screening................................................................................................. 25

2 5.1.2 Phase II Cost-Benefit Analysis.............................................................................. 26 5.1.3 Screening Sensitivity............................................................................................. 26 5.2 Staff Review................................................................................................................. 27 6.0 Conclusions.................................................................................................................... 28 7.0 References...................................................................................................................... 30 APPENDIX A.............................................................................................................................. 33

3 1.0 Introduction By letter dated December31,2016, NuScale Power, LLC (hereinafter referred to as NuScale or the applicant), tendered its application with the U.S.Nuclear Regulatory Commission (NRC or Commission) for certification of an integrated pressurized water small modular reactor (SMR) standard nuclear reactor design (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17013A229). The applicant submitted this application in accordance with Title10 of the Code of Federal Regulations (10CFR)Part52, Licenses, Certifications, and Approvals for Nuclear Power Plants, SubpartB, Standard Design Certifications, and 10CFRPart 52, SubpartE, Standard Design Approvals. On March 15, 2017, the NRC staff accepted the Design Certification Application (DCA) for docketing and initiated its technical review under Docket Number 52-048.

In accordance with 10 CFR 52.47(b)(2) and 51.55(a), NuScale included in Part 3 of the DCA the following report: "Applicant's Environmental Report - Standard Design Certification," herein denoted as ER (NuScale 2016). NuScale submitted three subsequent revisions - Revision 1 through 3 (NuScale 2018a, 2018b, 2019a). This staff technical analysis report provides an evaluation of severe accident mitigation design alternatives (SAMDAs) for the NuScale design to ensure that plant design alternatives which may have the potential for improving severe accident performance are identified, evaluated, and providing the bases for not incorporating SAMDAs in the design. To the extent practicable, the evaluation addresses the potential costs and benefits of SAMDAs for the NuScale design.

The purpose of this technical analysis report is to document the staffs review and analysis of NuScales consideration of SAMDAs which included identifying a broad range of potential alternatives, then determining whether implementation of the alternative would be feasible and beneficial on a cost-risk reduction basis. Additionally, due to information provided in Revision 3 of the NuScale ER, the staff will describe unique circumstances where certain SAMDAs were not incorporated. This situation requires the staff to set certain SAMDA requirements that must be met at a combined license (COL) stage utilizing the NuScale Design Certification (DC) information, if the DC is approved.

1.1 Description of the NuScale Design The NuScale SMR is an integrated passive-designed small modular pressurized water reactor.

The NuScale SMR contains unique design features. This design encompasses an integral power module, called the NuScale Power Module (NPM), consisting of a reactor core, two steam generator tube bundles, and a pressurizer contained within a single reactor vessel, along with the compact steel containment vessel that immediately surrounds the reactor vessel. The NPM design eliminates the need for external piping to connect the steam generators (SGs) and pressurizer to the reactor pressure vessel where natural circulation provides reactor coolant system flow. The NPM is submerged in water in the reactor building safety-related pool, which is also the ultimate heat sink (UHS) for the reactor. The pool portion of the reactor building is located below grade and the reactor building is designed to uphold 12 SMRs. Each NPM is rated at 160MegaWatt-thermal (MWt), or 1,920MWt total, with approximately 50MegaWatt-electrical MWe per NPM for a total of 600MWe output. The NuScale SMR is designed to be scalable, such that from one to twelveNPMs can operate within a single reactor building.

Each NPM in the common reactor pool is in its own three-walled bay with the open wall facing the center of the pool. The bays are arranged into two rows with six bays per row along the

4 north and south walls of the reactor pool at the east end of the pool. A central channel between the bays allows for movement of the NPMs between the bays and the refueling pool. The NPM, reactor pool, and spent fuel pool are below grade. Also located below grade are most primary systems and some radioactive waste equipment. Hoisting and handling equipment is located above grade. Pipe fittings and electrical connections are provided above the reactor pool water level to permit manual connection and disconnection during NPM installation, refueling outages, and replacement or removal of NPMs.

The NuScale design features are:

No alternate current or direct current power required for safe shutdown and cooling; Compact helical coil SGs with reactor pressure primary coolant on the outside of the tubes; High-strength steel containment immersed in a pool of water; Sub-atmospheric containment pressure during normal operation; Small core in comparison to the current generation of nuclear reactor cores with a correspondingly small source term; and Comprehensive digital instrumentation and control for monitoring and control.

The design identifies these key features of a multi-module plant:

A scalable plant design, which allows for incremental plant capacity growth; A compact nuclear island; and The ability to operate in island mode.

Additionally, the NuScale design is intended to minimize human error through fail-safe design functionality, allows multi-modular control capability from a single control room with effective automation design, employs digital display design and soft control technology to enhance usability, and provides optimum workload management. The applicant further stated that the NuScale human factors engineering program leverages human performance and operating experience from nuclear and nonnuclear industries. Further details on the NuScale design can be found in the DCA Part 2, Tier 2, Chapter 1, Introduction and General Description of the Plant (NuScale 2019b) 1.2 Severe Accident Mitigation Alternatives and Severe Accident Mitigation Design Alternatives The term severe accident mitigation alternatives (SAMAs) refers to an additional feature or action which would prevent or mitigate the consequences of severe accidents. SAMAs would include the consideration of hardware modifications or design alternatives, procedure changes, and training program improvements. SAMDAs are a subset of SAMAs with just the consideration of hardware modifications or design alternatives.

The purpose of the evaluation of SAMAs is to determine whether there are SAMDAs, procedural modifications, or training activities that can be justified to further reduce the risks1 of severe accidents (i.e., the prevention of core damage or reducing the release to the surrounding environment of radioactive material resulting from core damage). For standard DCs, the assessment is only for SAMDAs because a DC applicant cannot assess changes in operating 1 Risk is defined as the probability of an accident multiplied by the magnitude of the consequences.

5 procedures and training programs. This is because these programs have not been developed at this stage of licensing. Rather, they are developed during the construction phase of a nuclear power plant as a licensee prepares for the beginning of operations for the selected reactor design. However, it is expected that risk insights would be considered by a licensee in the development of plant procedures and training.

Consistent with the objectives of standardization and early resolution of design issues, the Commission decided to evaluate SAMDAs as within the scope of a DC. There are several Commission policy statements, court actions, and a staff recommendation (SECY) that establish the basis to evaluate SAMDAs for a standard DC. In 1980, the Commission issued a policy statement on the consideration of severe accidents in Environmental Impact Statements (EISs) for new reactor applications submitted after July 1, 1980 (45 FR 40101, June 13, 1980). In a 1985 policy statement (50 FR 32138; August 8, 1985), the Commission defined the term severe accident as an event that is beyond the substantial coverage of design basis events, including events where there is substantial damage to the reactor core (whether or not there are serious offsite consequences). A 1989 court decision ruled that the consideration of SAMAs is required for plant operation (Limerick Ecology Action v. NRC, 869 F.2d 719 (3rd Cir. 1989)).

Subsequently, in the Staffs Requirements Memorandum to SECY-91-299 (ADAMS Accession Number ML003707922), the Commission approved:

Addressing SAMDAs for certified designs in a single rulemaking process that would consider both the 10 CFR 50.34(f) and the National Environmental Policy Act requirements in the 10 CFR Part 52 DC rulemaking; Consideration of the costs and benefits associated with the review of the SAMDAs for the standard DCs; and Staff direction to advise applicants for a DC that they will be required to assess SAMDAs and the applicable decision rationale as to why they will or will not benefit the safety of their designs.

Additionally, 10 CFR 52.47(a)(23) requires that applications to the NRC for a reactor DC include a description and analysis of design features for the prevention and mitigation of severe accidents. In addition, 10 CFR 52.47(a)(27) requires a description of the design-specific probabilistic risk assessment (PRA) and its results, and, under 10 CFR 52.47(b)(2), an application for a standard DC must contain an ER as required by 10 CFR 51.55. The ER, pursuant to 10 CFR 51.55(a), must address the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives in the design to be certified.

1.3 NuScale Probabilistic Risk Assessment The NuScale DCA Part 2, Tier 2, Sections 19.0, Probabilistic Risk Assessment and Severe Accident Evaluation, and 19.1, Probabilistic Risk Assessment, describe the PRA performed for the NuScale design and summarize the Level 1 and Level 2 PRA, which evaluates the risk associated with all modes of operation for both internal and external initiating events. The PRA was performed for a single module and used to develop insights for multiple modules. DCA Part 2, Tier 2, Section 19.1, includes major topics such as PRA quality, design features to minimize risk, methodology, data, uncertainties, sensitivities, insights, and results. Internal and external event PRAs for at-power, low-power shutdown, and other modes of operations are described, and the risk associated with multiple modules is also discussed. Top cutsets that could lead to radiological releases and candidate risk significant events from the NuScale PRA models,

6 described in NuScale DCA Part 2, Tier 2, Section 19.1, were used to assist in the development of NuScale Power Plant specific SAMDAs.

1.4 Review Guidance The staff reviewed the technical content of NuScales ER using guidance from NUREG-1555, Standard Review Plans for Environmental Reviews for Nuclear Power Plants: Environmental Standard Review Plan (NRC 2007). Specifically, NUREG-1555, Section 7.2, Severe Accidents, for the analysis of offsite consequences and NUREG-1555 Section 7.3, Severe Accident Mitigation Alternatives, for the identification and evaluation of design alternatives that reduce the radiological risk from a severe accident by preventing substantial core damage. The staff also applied the guidance in NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S.

Nuclear Regulatory Commission, Revision 4 (NRC 2004), which establishes a framework for evaluation of SAMAs including estimation of values and impacts for design alternatives, and NUREG/BR-0184, Regulatory Analysis Technical Handbook (NRC 1997) with respect to the cost-benefit methodology. Two environmental audits were conducted in accordance with Office of New Reactors Office Instruction NRO-REG-108, Regulatory Audits (NRC 2009).

In addition to the above guidance documents, NuScale also applied guidance from the Nuclear Energy Institute (NEI), namely NEI 05-01A, Severe Accident Mitigation Alternatives (SAMA)

Analysis - Guidance Document (NEI 2005). This guidance is an acceptable methodology to the NRC for the assessment of SAMAs for license renewals under 10 CFR 51.53(c)(3)(ii)(L) but has not been endorsed or accepted for the assessment of new reactor SAMDAs under 10 CFR 51.55(a). However, the staff recognizes that there is useful information and guidance contained in NEI 05-01A and it has been applied in other design certifications. For example, by providing guidance for the application of the cost formulas from NUREG/BR-0184 for assessing the maximum benefit, the identification of SAMDAs, and the process for screening and assessing whether a SAMDA is potentially cost-beneficial. Therefore, the staff accepts how NuScale applied the application of the NEI 05-01A guidance for this SAMDA assessment.

1.5 Environmental Review Process The staff conducted the environmental review of the NuScale DCA and the development of an environmental assessment in accordance with 10 CFR 51.30(d). Based on a review of the NuScale ER, the NRC staff conducted an audit from August 14, 2017, to February 28, 2018, of the documentation supporting the ER. The staff audited the methodology, models, assumptions, and calculation packages in support of the ER. The audit plan used to support these interactions can be found under NRC 2017 and the audit summary is under NRC 2018a. A second audit was conducted in accordance with the audit plan dated October 12, 2018 (NRC 2018b), and performed from October 16, 2018, through November 30, 2018. This second audit was necessary in resolving specific information needs to support the NRC staffs independent evaluation as described in an audit summary under NRC 2019a. Based on the information provided in Revision 3 of the ER, the results of the environmental audits, clarification meetings, and other independent sources of information available to the staff; the evaluation of the NuScale SAMDA analysis is presented below.

7 2.0 Estimate of Risk for NuScale In NuScale ER Appendix B Section B.2, Release Categories, the applicant describes the application of risk information from the NuScale Level 1 PRA and Level 2 PRA to be utilized in the assessment of the NuScale SAMDAs (NuScale 2019a). As previously discussed, NuScale DCA, Tier 2, Chapter 19 summarizes the information regarding the PRA and severe accident evaluations conducted by the applicant (NuScale 2019c). The NuScale PRA provides an evaluation of the risk of core damage and release of radioactive material associated with both internal and external events that can occur during plant operation at power or while shutdown.

The different components of the NuScale PRA are:

Internal events PRA; Low-power shutdown PRA; Internal flooding PRA; Internal fires PRA; External floods PRA; High winds PRA; and Multi-module PRA.

Based on the guidance in NURG-0800, Standard Review Plan (SRP) Chapter 19, Severe Accidents (NRC 2016), the staffs review of the NuScale DCA Tier 2, Chapter 19 with their regulatory findings are presented in the staffs Safety Evaluation Report (FSER) Chapter 19 (NRC 2019b).

2.1 Application of the NuScale PRA The PRA required for a DCA comprises two major areas of analysis: (1) identification of sequences of events that could lead to core damage and estimation of their frequencies of occurrence (the Level 1 PRA analysis); and (2) evaluation of the potential response of the containment to these sequences, with emphasis on the possible modes of containment failure and the corresponding radionuclide source terms (the Level 2 PRA analysis). NuScale analyzed the various combinations of events leading to radiological releases from the NPM, known as cutsets.

The products from the PRA cutsets are a range in the frequency of occurrence, or probability, for core damage along with variations in the associated radiological releases for these cutsets.

NuScale describes only two release categories (RCs) in the PRA described in NuScale DCA, Tier 2, Chapter 19 (NuScale 2019c). As noted in the NuScale ER, NuScale decided that a more precise treatment of radionuclide releases was desired for the purposes of a SAMDA assessment (See Section 5.2 and Section B.2 of NuScale 2019a). Due to number of possible cutsets to consider from the PRA to support SAMDAs, NuScale grouped many of the PRA cutset sequences into eight RCs based on initiating events and mitigation system availabilities.

The eight RCs applied by NuScale for the SAMDA assessment with their release frequencies per module-year are shown in Table 2-1.

8 Table 2-1: NuScale Release Categories for SAMDAs and Release Frequencies Release Category No.

Release Categorya Release Frequencyb (module-yr-1)

Fraction of Totalb 1

Chemical and volume control system (CVCS) loss-of-coolant accident inside of containment 2.15E-11 2.04E-05 2

Isolatable pipe break outside of containment 1.06E-12 1.01E-06 3

Unisolated CVCS pipe break outside of containment 1.79E-11 1.70E-05 4

Emergency core cooling system spurious actuation 2.44E-09 2.32E-03 5

Steam generator tube failure 1.39E-13 1.32E-07 6

General transient with reactor safety valve (RSV) stuck open 2.06E-10 1.96E-04 7

General transient with no RSVs 2.70E-10 2.56E-04 8

Dropped NPM during transport 1.05E-06 9.97E-01 Total 1.05E-06 1.00 a - Release categories based on information from Section 5.2 and B.2 of NuScale 2019a b - From Table B-25 of NuScale 2019a.

2.2 Offsite Consequences and Risk Analysis The scope of this review includes an analysis of the offsite population dose consequence from severe accidents, including socioeconomic impacts such as offsite property damage. For this analysis, the applicant must provide, and the staff must independently review, information that characterizes the risk profile of the plant. This includes the input contributors from the application of the NuScale PRA (i.e., core damage frequency (CDF) from dominant severe accident sequence as previously described and dose consequences from each release class, or category, for the associated source term. Along with the safety-related information, the applicant must select a Reference or surrogate, site with the characteristics, specifically population distribution and weather, to apply for assessing the offsite population dose and offsite property damage consequences. Outside of the release plume, its associated source term, and associated site parameters, the applicant must also select the other various MELCOR Accident Consequence Code System (MACCS) input parameter values to apply. The staff reviewed these analysis parameters and selected their own MACCS input parameter values for performing their confirmatory analysis.

This section will discuss the site selection, site parameters, and MACCS input parameter values established by NuScale. This will be followed by a summary of NuScales offsite consequence analysis with their results. A summary of the staffs independent confirmatory analysis and results will complete this section.

2.2.1 Site Selection and Site Parameters NuScale selected the Surry Power Station (Surry site) as the representative site for the purposes of assessing the offsite consequences for the NuScale SAMDA assessment. This site was analyzed as part of the severe accident assessment documented in NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, (NRC 1990) and the follow-on study to NUREG-1150 known as NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report (NRC 2012). Thus, this site has publicly available data sources and publicly available documents listing the MACCS input parameter values (see Appendix C of SNL 2013b). Based on the use in the NuScale ER of publicly available

9 information from NRC sources, the staff determined that the applicants selection of the Surry site is an acceptable representative site for assessing the offsite consequences for the NuScale design.

A description of the Surry site is provided in Appendix A of this report and by NuScale in Appendix B of the NuScale ER (NuScale 2019a). The NuScale ER sets the basis for the site parameters relied upon in the NuScale SAMDA assessment to include population distribution and economic factors out to 50 miles along with weather data for one year. The source of this Surry site information, as applied and updated by NuScale, came from Surry site SOARCA documentation as provided in Appendices C of NUREG/CR-7110, State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis (SNL 2013b). As a siting sensitivity case, NuScale also evaluated the offsite consequences for an alternative site, selecting the other SOARCA study site, the Peach Bottom Atomic Power Station (Peach Bottom site), applying the information from Appendix B from NUREG/CR-7110, State-of-the-Art Reactor Consequence Analyses Project Volume 1: Peach Bottom Integrated Analysis (SNL 2013a).

Descriptions of the Surry and Peach Bottom site can also be found in Appendix A of this report.

2.2.2 MACCS Input Quantification In addition to establishing the site parameter values, the applicant also should set the input parameter values for the other MACCS input parameters in the EARLY and CHRONC code modules (Chanin and Young 1998). The EARLY module assesses the time period immediately following a radioactive release. This module is where the NuScale design source term and plume segment values are provided to the MACCS code. Additionally, the EARLY module specifies the emergency response scenarios to be considered; including evacuation, sheltering, and dose-dependent responses. The CHRONC module evaluates the events following the emergency phase analyzed by the EARLY module. NuScale made the assumption in their base case model that there is not an evacuation scheme associated with a site-wide general emergency. The results from the CHRONC module includes total accumulated population dose, long term protective actions, individual health effects, and the economic costs related to the long-term protective actions. Current guidance for quantification of MACCS input parameters can be found in Sprung et al. 1990. The current version of MACCS does have input parameters that are not discussed in Sprung et al. 1990, and appropriate values would need to be assigned to these input parameters, such as from recent NRC studies (e.g., the SOARCA study documentation in NUREG/CR-7110 Volume 2 (SNL 2013)). NuScale applied values for these input parameters based on the SOARCA documents with adjustments of the economic input parameters based on the Consumers Price Index to bring them to 2017-dollar values.

2.2.3 Base Case and Sensitivity Cases NuScale Base Case Offsite Consequence Analyses NuScale performed base case offsite consequence calculations for each of the eight RCs with MACCS code package version 3.10.1.2. A summary of the eight RCs and their offsite calculations is provided in Section B.2. As previously described, the base case relied upon expected or recommended input parameter values from the SOARCA study. The offsite consequences for population doses and offsite property damage costs for each RC were multiplied by the RCs frequency (i.e., the summation of the CDF for each sequence in a RC) to obtain the appropriate risk value (person-rem per reactor year for population dose and dollars per reactor year for offsite property damage) for each RC. The overall results for offsite

10 population dose and offsite economic costs are shown in Tables B-27, 5-1, and 5-2 of the NuScale ER (NuScale 2019a). These base case RC results were applied as inputs to the averted public exposure and averted public offsite property damage costs as presented in Sections 5.2 and 5.3 of the NuScale ER (NuScale 2019a).

NuScale Sensitivity Offsite Consequence Analyses To assess how changes to important MACCS input parameter values would affect the maximum benefit results, NuScale performed eleven MACCS sensitivity cases, which are:

1. Standard Evacuation;

2. Updated Site Data File;

3. Release Site;

4. Height of Release;

5. Initial Core Inventory;

6. Start of Release;

7. Off-site Decontamination;

8. Other Economic Parameter;

9. Containment Deposition;

10. Aerosol Dry Deposition in the Environment; and

11. Plume Buoyancy.

The eleven cases are related to the MACCS calculations for offsite population dose and economic cost (i.e., offsite property damage) consequences as discussed in Section B.3 of the NuScale ER with results for the eight RCs being provided in Tables B-28 and B-29 (NuScale 2019a). The results of the eleven sensitivity cases were factored into NuScales maximum benefit evaluation, which will be discussed later in Section 6.3, Maximum Benefit Sensitivity Analysis, of this report.

2.2.4 Staff Review The staffs guidance for the review of the environmental impacts from severe accidents is found in Section 7.3, Severe Accidents, of NUREG-1555 (NRC 2007). As part of the staffs independent National Environmental Policy Act review of the NuScale SAMDA evaluation, confirmatory analyses of the offsite consequences and subsequent risk values were determined to be necessary. This required the building of a NuScale MACCS model by the staff based on publicly available information, such as the NuScale ER or prior NRC documents. Other items for the staffs MACCS model were supplemented by information in Part 2 of the DCA (NuScale 2019b), and by information obtained during the staffs two environmental audits (NRC 2017, NRC 2018a, NRC 2018b, and NRC 2019a). Because NuScale applied a number of the MACCS input parameter values from the SOARCA project (see Appendix B of NuScale 2019), prior staff-developed SOARCA MACCS models formed the starting point for the staff to independently build their own NuScale MACCS models (SNL 2013a and SNL 2013b).

For a small number of MACCS input parameters, the staff did not immediately have access to the value(s) needed since they were not specifically provided in the NuScale ER. The staff searched and found at least one input parameter value in the NuScale DCA Part 2 Teir 2 documents. For any others, the staff either applied their engineering judgement based on the experience from the SOARCA study or from other internal NRC MACCS analyses. Two examples of this last effort concerned the dose conversion factors (DCF) library and RC 8

11 application of organic and inorganic radioactive iodine (see Section A.4 of this report). The unverified information was later updated based on the two staff audits of the NuScale ER and supporting document (NRC 2018a and 2019a). Once this work was completed, the staff obtained good agreement to the results provided in the NuScale ER for all RCs. In most cases, the staffs results were in good agreement with the NuScale ER results (e.g., the same first two-digits for most MACCS early phase and long-term phase results)2.

After this verification of the performance of the staffs MACCS NuScale model, the staff adjusted several of the MACCS input parameter values to create the staffs MACCS confirmatory model.

As described in Section A.4 of Appendix A to this report, the staff updated some of the MACCS input parameter values based on recent internal NRC MACCS assessments. By applying revised input parameter values, the staff determined the impact of a more current state of input parameter valuations on the NuScale severe accident results. The staff applied the revised input parameter values to both the staff-developed Surry site base case and Peach Bottom site sensitivity case MACCS models.

The staff results for population dose and economic cost (i.e., offsite property damage) are compared to the NuScale results for both sites in Tables 2-2 through 2-5. Results are provided in absolute consequence units (person-rem per event or dollars per event) and in units of risk by multiplying the absolute consequences to the release frequency for an event (person-rem per reactor year (Ryr) and dollars per Ryr). The comparison of results shows some differences regarding the absolute consequence results between NuScales and the staffs MACCS analysis. However, the risk values between the two sets of results are not significantly different.

In both sets of analyses, the total risk is dominated by RC 8, the dropped NPM during transient scenario, and RC 3, the unisolated CVCS pipe break outside of containment scenario. The other RCs have three to seven or more orders of magnitude of lower severe accident risks.

These results are applied in the appropriate cost formulas of Section 6.2 of this report for their contributions in the maximum benefit assessment.

Table 2-2 Surry Site Base Case MACCS Offsite Population Dose Comparison Release Category Release Frequency (module-yr-1)

NuScalea Staff Person-rem/event Person-rem/Ryr Person-rem/event Person-rem/Ryr 1

2.15E-11 3.16E+01 6.80E-10 2.56E+01 5.50E-10 2

1.06E-12 2.29E+01 2.43E-11 1.79E+01 1.90E-11 3

1.79E-11 1.21E+06 2.16E-05 1.26E+06 2.26E-05 4

2.44E-09 2.06E+01 5.03E-08 1.61E+01 3.93E-08 5

1.39E-13 2.36E+05 3.27E-08 1.80E+05 2.50E-08 6

2.06E-10 1.50E+01 3.09E-09 1.18E+01 2.43E-09 7

2.70E-10 2.20E+01 5.94E-09 1.72E+01 4.64E-09 8

1.05E-06 7.57E+02 7.95E-04 7.78E+02 8.17E-04 Total 1.05E-06 8.17E-04 8.40E-04 a - From Tables 5-1 and B-27 of NuScale 2019a.

2 NuScale ER Revision 0 through Revision 2 only provided MACCS long-term phase results. The staff discussed this with NuScale during the second audit and NuScale provided a combination of the early phase and long-term phase results in NuScale ER Revision 3.

12 Table 2-3 Surry Site Base Case MACCS Offsite Property Damage Comparison Release Category Release Frequency (module-yr-1)

NuScalea Staff Dollars/event Dollars/yr Dollars/event Dollars/yr 1

2.15E-11 0.00E+00 0.00E+00 2.02E+02 4.34E-09 2

1.06E-12 0.00E+00 0.00E+00 1.53E+02 1.62E-10 3

1.79E-11 2.72E+09 4.87E-02 2.92E+09 5.23E-02 4

2.44E-09 0.00E+00 0.00E+00 1.23E+02 3.00E-07 5

1.39E-13 8.42E+07 1.17E-05 8.72E+07 1.21E-05 6

2.06E-10 0.00E+00 0.00E+00 5.07E+01 1.04E-08 7

2.70E-10 0.00E+00 0.00E+00 1.44E+02 3.89E-08 8

1.05E-06 4.77E+03 5.01E-03 8.31E+03 8.73E-03 Total 1.05E-06 5.37E-02 6.10E-02 a - From Tables 5-2 and B-27 of NuScale 2019a.

Table 2-4 Peach Bottom Site Sensitivity MACCS Offsite Population Dose Comparison Release Category Release Frequency (module-yr-1)

NuScale Staff Person-rem/eventa Person-rem/yr Person-rem/event Person-rem/yr 1

2.15E-11 4.52E+01 9.72E-10 4.30E+01 9.25E-10 2

1.06E-12 2.63E+01 2.79E-11 2.74E+01 2.90E-11 3

1.79E-11 1.05E+06 1.88E-05 1.42E+06 2.54E-05 4

2.44E-09 2.37E+01 5.78E-08 2.47E+01 6.03E-08 5

1.39E-13 2.23E+05 3.10E-08 2.38E+05 3.31E-08 6

2.06E-10 1.75E+01 3.61E-09 1.82E+01 3.75E-09 7

2.70E-10 2.52E+01 6.80E-09 2.63E+01 7.10E-09 8

1.05E-06 1.06E+03 1.11E-03 1.03E+03 1.08E-03 Total 1.05E-06 1.13E-03 1.11E-03 a - From Table B-28 for Case 3 of NuScale 2019a.

Table 2-5 Peach Bottom Site Sensitivity MACCS Offsite Property Damage Comparison Release Category Release Frequency (module-yr-1)

NuScale Staff Dollars/eventa Dollars/yr Dollars/event Dollars/yr 1

2.15E-11 2.68E+02 5.76E-09 9.37E+02 2.01E-08 2

1.06E-12 6.57E+00 6.96E-12 5.80E+02 6.15E-10 3

1.79E-11 8.19E+09 1.47E-01 1.87E+10 3.35E-01 4

2.44E-09 4.17E+00 1.02E-08 4.54E+02 1.11E-06 5

1.39E-13 2.85E+08 3.96E-05 8.01E+08 1.11E-04 6

2.06E-10 5.17E-01 1.07E-10 1.43E+02 2.95E-08 7

2.70E-10 5.40E+00 1.46E-09 5.32E+02 1.44E-07 8

1.05E-06 4.84E+04 5.08E-02 1.30E+05 1.37E-01 Total 1.05E-06 1.97E-01 4.71E-01 a - From Table B-29 for Case 3 of NuScale 2019a.

13 3.0 Maximum Benefit Evaluation 3.1 Introduction The methodology used by NuScale was based primarily on NRCs guidance for performing cost-benefit analysis, NUREG/BR-0184 (NRC 1997) and NEI 05-01A accepted by NRC for license renewal SAMAs (NEI 2005). However, as previously discussed in this report, the methodology presented in the industry guidance has been applied in new reactor licensing SAMDA assessments. As described in Section 5 of the ER (NuScale 2019a), the unmitigated risk monetary value, or the net present value (NPV) of current risk, is calculated using the following formula:

= (+ + + )

Where:

NPV = Net present value of current risk ($);

APE = Present value of averted public exposure ($);

AOC = Present value of averted offsite property damage costs ($);

AOE = Present value of averted occupational exposure ($);

AOSC = Present value of averted onsite costs ($); and COE = Cost of any enhancement implemented to reduce risk ($).

For the representation of the maximum benefit that could be provided, the maximum benefit is calculated to be the sum of the four averted cost categories. It is represented as:

= + + +

In accordance with NUREG/BR-0058 guidance (NRC 2004), present worth estimates should be developed using both the 7 percent and 3 percent discount rates, which is also discussed in Section 8.5 of NEI 05-01A (NEI 2005). NuScale conducted a maximum benefit base case analysis using the 7 percent discount rate and a sensitivity case for the Surry site using the 3 percent discount rate (Sensitivity Case No. 14). The details of the applicants maximum benefit analysis with the 7 percent discount rate for the Surry site is found in Section 5 of the ER (NuScale 2019a).

3.2 Maximum Benefit This section presents the results of the NuScale and the staffs analysis of the maximum benefit for the four cost categories presented in the prior equation in the order of averted public exposure, averted offsite property damage (i.e, offsite economic costs), averted occupational exposure, and averted onsite costs. This analysis costs the risks only for each of the eight individual RCs and does not assume that any additional damage occurs to the plant beyond the individual event. The maximum benefit presented in this section are provided first using the Surry site base case in the averted cost formulas for a single NPM, for the NuScale site (i.e., 12 NPMs), and for sensitivity cases. For the first two maximum benefit evaluations, NuScale results and the staffs independent confirmatory results are presented. For the sensitivity cases, NuScale results are presented along with comparisons to selected staff results for sensitivity cases were considered to provide key insights to the maximum benefit valuation.

14 3.2.1 Averted Public Exposure (APE)

The APE is associated with the present value for accident-related health effects as discussed in Section 5.7.1 of NUREG/BR-0184 (NRC 1997). The APE costs were calculated using the following formula:

= x x (1 )

Where:

APE = Present value of averted public exposure ($);

R = Monetary equivalent of unit dose ($5,100/person-rem);

DPA = Avoided accident offsite dose (person-rem per year) from ER Table 5-1 (NuScale 2019a) based on the summation over all RCs of the event frequency (events per year) times the public offsite dose (person-rem per event);

r = Real discount rate (7% per year); and tf = Years remaining until end of plant life (60 years).

Using the equation and parameters as defined above, NuScale calculated the APE for the eight RC events. The results of the NuScale and staff evaluations are provided in Table 3-1.

Table 3-1 Base Case APE Comparison for 7 Percent Discount Rate Release Category NuScalea Staff 1

$4.88E-05

$4.02E-05 2

$1.74E-06

$1.39E-06 3

$1.55E+00

$1.65E+00 4

$3.61E-03

$2.87E-03 5

$2.35E-03

$1.83E-03 6

$2.22E-04

$1.78E-04 7

$4.27E-04

$3.40E-04 8

$5.70E+01

$5.98E+01 Total APE

$5.86E+01

$6.14E+01 a - From Table 5-3 of NuScale 2019a.

3.2.2 Averted Offsite Property Damage Costs (AOC)

The AOC is associated with the present value for accident-related offsite property damage as discussed in Section 5.7.5 of NUREG/BR-0184 (NRC 1997). The AOC was calculated using the following formula:

= x (1 )

Where:

AOC = Present value of averted offsite property damage costs ($),

DDA = Avoided offsite property damage ($ cost per year) from Table 5-2 (NuScale 2019a) based on the summation over all RCs of the accident frequency (event per year) times property damage ( $ cost per event);

15 r = Real discount rate (7% per year),

tf = Years remaining until end of plant life (60 years).

Using the equation and parameters as defined above, the results of the NuScale and staff evaluations are provided in Table 3-2.

Table 3-2 Base Case AOC Comparison for 7 Percent Discount Rate Release Category NuScalea Staff 1

$0.00

$6.11E-08 2

$0.00

$2.28E-09 3

$6.85E-01

$7.35E-01 4

$0.00

$4.22E-06 5

$1.64E-04

$1.71E-04 6

$0.00

$1.47E-07 7

$0.00

$5.47E-07 8

$7.05E-02

$1.23E-01 Total AOC

$7.55E-01

$8.59E-01 a - From Table 5-3 of NuScale 2019a.

3.2.3 Averted Occupational Exposure (AOE)

The AOE is associated with the present value for accident-related occupational exposure as discussed in Section 5.7.3 of NUREG/BR-0184 (NRC 1997). The AOE values for each RC were calculated using the following formulas:

= +

= x x x (1 )

= x x x (1 )

x (1 )

Where:

AOE = Present value of averted occupational exposure ($),

WIO = Present value of averted immediate occupational exposure ($),

FA = Accident frequency (events per year from Table 5-1 (NuScale 2019a)),

DIOA = Avoided immediate occupational exposure (person-rem per year) based on 3,300 person-rem/event provided in Section 5.7.3.1 of NUREG/BR-0184 (NRC 1997),

R = Monetary equivalent of unit dose ($5,100/person-rem),

r = Real discount rate (7% per year),

tf = Years remaining until end of plant life (60 years).

WLTO = Present value of averted long-term occupational exposure ($),

DLTOA = Avoided long-term occupational exposure (person-rem per year) based on best-estimate of 20,000 person-rem/event provided in Section 5.7.3.1 of NUREG/BR-0184 (NRC 1997),

m = Years over which long-term doses accrue (10 years from NuScale 2019a).

Using the equation and parameters as defined above, the results of the NuScale and staff evaluations are provided in Table 3-3.

16 Table 3-3 Base Case AOE Comparison for 7 Percent Discount Rate Release Category NuScalea Staff 1

$2.73E-02

$2.78E-02 2

$1.35E-03

$1.37E-03 3

$2.27E-02

$2.32E-02 4

$3.10E+00

$3.16E+00 5

$1.76E-04

$1.80E-04 6

$2.62E-01

$2.67E-01 7

$3.43E-01

$3.49E-01 8

$1.33E+03

$1.36E+03 Total AOE

$1.34E+03

$1.36E+03 a - From Table 5-3 of NuScale 2019a.

3.2.4 Averted Onsite Cost (AOSC)

The AOSC is associated with the present value for onsite property damage from an accident:

cleanup and decontamination, long-term replacement power, and repair and refurbishment as discussed in Section 5.7.6 of NUREG/BR-0184 (NRC 1997). The total averted onsite property damage costs are the sum of these three types of costs. The AOSC were calculated using the following formula:

= + +

Where:

UCD = Present value of the cleanup and decontamination costs ($)

URP = Present value of the replacement power costs ($);and URR = Present value of the repair and refurbishment costs ($)

The assessments for UCD, URP, and URR are provided below.

Averted Cleanup and Decontamination Costs (UCD)

The estimated clean-up and decontamination cost for severe accidents is defined in NUREG/BR-0814, Section 5.7.6.1 to be $1.5109 per accident, undiscounted and not adjusted for inflation (NRC 1997). The present value of the payments needed for cleanup and decontamination costs is calculated using the following formula with an additional factor to adjust for inflation:

=

x (1 )

Where:

PVCD = Present value of averted onsite cleanup costs exposure over cleanup period ($),

CCD = Total value of averted onsite cleanup costs ($1.5E+09/event),

CPI = Adjustment for inflation based on the ratio of CPI of the current year divided by the CPI in year 1993 dollars (see discussion in Section 5.7.6 of NUREG/BR-0184, NRC 1997) r = Real discount rate (7% per year),

m = Years over which long-term doses accrue (10 years),

17 PVCD = (($1.5109/event) / (10 years)) * ((1 - e-(0.07*10)) / 0.07),

PVCD = $1.08109.

The net present value of averted cleanup costs over the plant life is calculated using the following formula:

= x (1 )

Where:

UCD = Present value of averted onsite cleanup costs ($),

PVCD = Present value of averted onsite cleanup costs exposure over cleanup period ($),

tf = Years remaining until end of plant life (60 years).

Averted Replacement Power Cost (URP)

The estimated long-term replacement cost for severe accidents is defined in NUREG/BR-0814, Section 5.7.6.2 of NUREG/BR-0184, undiscounted and not adjusted for inflation (NRC 1997).

The averted replacement power costs as the net present value of replacement power over life of facility is calculated by the following formulas with an additional factor to adjust for inflation:

= $1.2 x 10 x x ( 910

)

x (1 )

=

x (1 )

Where:

URP = Net present value of replacement power over life of facility (dollars),

PVRP = Net present value of replacement power for a single event ($),

CPI = Adjustment for inflation based on the ratio of CPI of the current year divided by the CPI in year 1993 dollars (see discussion in Section 5.7.6 of NUREG/BR-0184, NRC 1997) r = Real discount rate (7% per year),

tf = Years remaining until end of plant life (60 years),

Rated Power = 50 MWe.

Averted Repair and Refurbishment Costs NuScale has stated that they assume the NPM undergoing core damage would not be repaired or refurbished and assigns a cost of zero for repair and refurbishment (NuScale 2019a).

Total AOSC As discussed at the start section 3.2.4, the equation for the AOSC is represented as:

= + + 0 The AOSC results from the NuScale and staff evaluations are provided in Table 3-4.

18 Table 3-4 Base Case On-site Cost AOSC Comparison for 7 Percent Discount Rate Release Category NuScalea Staff 1

$6.40E-01

$6.39E-01 2

$3.16E-02

$3.15E-02 3

$5.32E-01

$5.32E-01 4

$7.27E+01

$7.26E+01 5

$4.13E-03

$4.13E-03 6

$6.13E+00

$6.13E+00 7

$8.03E+00

$8.03E+00 8

$3.12E+04

$3.12E+04 Total AOSCRP

$3.13E+04

$3.13E+04 a - From Table 5-3 of NuScale 2019a.

3.2.5 Maximum Benefit for a Single NuScale Power Module Using the above equations and results, NuScale estimated the total unmitigated base case risk, or the maximum benefit, for a single NPM to be $32,700 at the 7 percent discount rate. The resulting Maximum Benefit represents the maximum risk benefit attainable if all core damage scenarios are eliminated over the 60-year plant life. As will be discussed in the next section, NuScale did not perform a sensitivity case for a single NPM using the 3 percent discount rate.

Table 3-5 is a summary of the total results generated for the Total Maximum Benefit.

Table 3-5 Base Case Maximum Benefit for Single NuScale Power Module Severe Accident Impact with 7 Percent Discount Rate Category NuScalea Staff APE

$5.86E+01

$6.14E+01 AOC

$7.55E-01

$8.59E-01 AOE

$1.34E+03

$1.36E+03 AOSC

$3.13E+04

$3.13E+04 Maximum Benefit

$3.27E+04

$3.27E+04 a - From Table 5-3 of NuScale 2019a.

Based on the results NuScale obtained and presents in ER Table 5-3, 99.7 percent of the Total Maximum Benefit is related to RC 8, namely the scenario of a single dropped NPM, with value of approximately $32,600. The combined Maximum Benefit for all other RCs, RC 1 through RC 7, is less than $100.

19 3.2.6 Maximum Benefit for Twelve NuScale Power Modules To account for seismic events and the overall plant design having twelve NPMs in a common Ultimate Heat Sink (UHS) pool within the same Reactor Building, NuScale assessed the maximum benefit for a scenario where multiple NPMs could be affected during the same event.

The NuScale PRA established that three NPMs can be damaged in a dropped NPM accident scenario (Section 19.1.7.4, Insights Regarding Low Power and Shutdown for Multi-Module Operation, of NuScale 2019b). In the discussion provided in Section 19.1.7.4, NuScale states if an operating module is struck near the top, damage may occur to the decay heat removal system components (piping or heat exchanger) or additional pipe breaks may occur if struck with enough force. However, NuScale contends that the dropped NPM would likely not have a strong enough impact for the CNV of an operating NPM to be breached due to the dropped NPM falling a short distance through a water medium providing resistance (NuScale 2019b).

Given this discussion in Chapter 19 of NuScale DCA Part 2, Tier 2, NuScale determined that the effect of seismic events and multiple NPMs co-located in the Reactor Building must be taken into account for determining the maximum benefit. Therefore, as discussed in NuScale DCA Part 2, Tier 2, Section 19.1.7.4 using a scenario based on RC 8, a dropped NPM tips over into an adjacent NPM bay hitting the upper half of an assumed operational NPM with such force as to knock it out of its bay. The bottom of the dropped NPM next slides along the bottom of the UHS pool directly across into the opposite NPM bay striking that NPM in its bottom half as to also knock that operational NPM out of its bay. The three NPMs next all end up horizontally on the bottom of the UHS pool in a manner that core cooling cannot be maintained, core damage ensues, and there is a breach of each containment vessel so that volatiles and other fission gases are released into the pool. However, due to the depth of the pool (approximately 50 feet),

most of the volatile gases and fission gases are scrubbed by the pool water. Only the noble gases, elemental iodine, and organic iodine are not scrubbed, being released into the Reactor Building, and subsequently released to the surrounding environment. Additionally, the resulting contamination in the Reactor Building is such that the remaining nine NPMs require replacement of their electricity production.

The other seven RCs remain the same but are now accounted for with all twelve modules and seismic external events. The result of this scenario regarding its effects on the various averted cost categories is as follows:

Three RC 8 releases into the surrounding site and offsite environment with contributions to the APE, AOC, and AOE cost categories; All twelve NPMs need replacement power; Cleanup and decontamination activities are conducted in the Reactor Building based on releases from the three NPMs; The overall probability of this event is the same as for RC 8 with no credit given for the likelihood that a dropped NPM would cause two other NPMs would be knocked out of their respective bays; An overall external events multiplier is applied where the seismic CDF is added to the total non-seismic CDF then divided by the total non-seismic CDF; and The summation of RC 1 through RC 7 maximum benefits is multiplied by 12 and by the seismic external events multiplier.

20 The resulting hypothetical maximum benefit becomes:

(3)

= (3 x + + + 12 x + 3 x x ) x

+ 12 x x

( )

Where:

Mexternal = Seismic external event multiplier of 1.03 for the Surry site (NuScale 2019a)

Applying the prior averted cost values results in the maximum benefits shown in Table 3-6 where, under NuScales analysis, RC 1 through RC 7 contribute $1,160 to the total maximum benefit of $136,000.

Table 3-6 Maximum Benefit for Twelve NuScale Power Modules with 7 Percent Discount Rate Category NuScalea Staff APE

$1.96E+02

$2.05E+02 AOC

$8.68E+00

$9.47E+00 AOE

$4.13E+03

$4.25E+03 AOSC

$1.31E+05

$1.31E+05 Total Maximum Benefit

$1.36E+05

$1.36E+05 a - From Table 5-4 of NuScale 2019a.

3.3 Maximum Benefit Sensitivity Analyses NuScale examined the maximum benefit sensitivity in terms of several MACCS input parameter values and modeling assumptions as well as three cost formula assumptions, which one of these cases applied the 3 percent discount rate sensitivity case. For Case 3, the Peach Bottom site sensitivity analysis, not only did NuScale apply Peach Bottom site-specific MACCS input parameter values as previously discussed, but also applied an external event multiplier factor of 2.91 based on the more seismically active Peach Bottom site as compared to the Surry site.

NuScale also performed three additional sensitivity cases to address cost formulas sensitivities related to the Surry site maximum benefit regarding the occupational doses (Case 12), the dollar-per-person-rem (Case 13), and the 3 percent discount rate (Case 14). The impact of all fourteen sensitivity cases on the maximum benefit value are shown in Table 5-5 of the NuScale ER (NuScale 2019a) and are reproduced in Table 3-8.

21 3.4 Staff Review As discussed and presented above, the staff performed a set of independent confirmatory maximum benefit analyses applying the staffs results from MACCS calculations for the Surry and Peach Bottom sites. The staffs maximum benefit results, including selected sensitivity cases, are similar to and consistent with the NuScale maximum benefit results. In Section 5.8 of the ER, NuScale presents their five insights from their sensitivity analyses (NuScale 2019a).

Table 3-8 Maximum Benefit Sensitivity Cases and Results Case Cases (Section, Title, Description)

NuScalea Selected Staff 5.7 Maximum Benefit: Surry site base case

$136,000

$135,700 1

B.3.1 Standard Evacuation: 10-mile Emergency Planning Zone

$152,000 2

B.3.2 Updated Site Data File: 2010 population and 2012 economic site data

$136,000 3

B.3.3 Release Site: Peach Bottom site

$383,000

$380,165 4

B.3.4 Height of Release: 24.689 m release height

$136,000 5

B.3.5 Initial Core Inventory: High burnup inventory

$136,000 6

B.3.6 Start of Release: Immediate Release

$136,000 7

B.3.7 Off-site Decontamination: Maximum TIMDEC and CDNFRM

$136,000 8

B.3.8 Other Economic Parameter: ER Table B-2, except DPRATE and DSRATE, multiplied by two

$136,000 9

B.3.9 Containment Deposition: Airborne and deposited release from containment

$136,000 10 B.3.10 Aerosol Dry Deposition in the Environment: 0.01m/s deposition velocity

$136,000 11 B.3.11 Plume Buoyancy: 100,000 W plume heat content

$136,000 12 5.8 High on-site dose estimate: 14,000 person-rem immediate and 30,000 person-rem long-term

$140,000 13 5.8 High Dollar-per-person-rem estimate: $7,500 per person-rem

$138,000 14 5.8 Three percent real discount rate

$341,000

$338,660 a - From Table 5-5 of NuScale 2019a.

In general, the staff agrees with the NuScale insights where certain aspects from the staffs independent evaluation are as follows:

1. By far, the dropped NPM accident scenario is the dominant scenario to be considered in the SAMDA cost-benefit evaluation that follows.

2. Offsite consequences/risks (dose and cost) and onsite occupational exposure results have a small contribution to the maximum benefit for the NuScale design based on the Surry site parameters.

3. The greatest effects on maximum benefit are 1) the site parameters as shown by the Surry versus Peach Bottom results, principally seismic; and 2) the value of the discount rate being applied) with approximately 2.5 to 3 times larger maximum benefits for the 3 percent discount rate over the 7 percent discount rate.

22 Additionally, the staff is applying NuScales assumed bounding maximum benefit scenario as it is currently provided in the NuScale ER. NuScale does state in NuScale DCA Part 2, Tier 2, Section 19.1.7.4 that the striking of an operational NPM by a dropped NPM could cause damage to the upright and operational NPM without it being dislodged from its bay. However, NuScale does not provide an analysis as to why this situation is less likely than a dropped NPM dislodging two operational NPMs from their bays. Nor does NuScale provide an analysis as to why a dropped NPM would have an impact great enough to dislodge two operational NPMs, with their seismic qualified structures. Given NuScale has a structural framework at the top of the NPM with a bio-shield structure covering the top of the NPM above the water level of the UHS, the force of impact from a dropped NPM may not be great enough to result in serious damage to an operational NPM above the decay heat removal system, let alone provide enough force to dislodge two NPMs from their respective bays.

Finally, the staff selected a subset of the NuScale sensitivity cases for the staffs independent confirmatory analysis. The staff recognizes, based on the above offsite consequence/risk and maximum benefit analyses that the averted public exposure (APE), averted offsite property damage costs (AOC), and averted occupational exposure (AOE) have a small contribution to the total maximum benefit. The NuScale sensitivity analysis clearly demonstrates this based on the results for Sensitivity Cases 1, 2, and 4 through 13. Additionally, the staffs confirmatory MACCS calculational results are very similar to NuScales results such that the staff sees no benefit from performing confirmatory analysis on the other sensitivity cases which varies MACCS input parameter valuations for the Surry site. Therefore, the staffs independent confirmatory analysis only focuses on the Peach Bottom site sensitivity case applying a 7 percent discount rate (Sensitivity Case 3) and the 3 percent discount rate for the Surry site sensitivity case (Sensitivity Case 14). The selected staffs maximum benefit sensitivity results are also presented in Table 3-8.

23 4.0 Potential Plant Improvements This section discusses the process NuScale applied for the identification of potential plant improvements for the NuScale design and the staffs evaluation of that process.

4.1 NuScale Process for Identifying Potential Plant Improvements NuScale presents their approach and methodology for identifying potential plant improvement, namely SAMDA candidates, in Section 3 of the ER and then presents in Section 4, SAMDA Candidate Identification, of the ER the steps they applied for the identification the SAMDA candidates (NuScale 2019a). As with past new reactor SAMDA evaluations, NuScale applies an industry-developed standard list of SAMA candidates based on the pressurized water reactor design, which are given in Table 14, STANDARD List of PWR SAMA Candidates, of NEI 05-01A (NEI 2005)3.

This list of SAMA candidates is supplemented by additional SAMDA candidates as determined from the risk significant events and insights from the NuScale PRA as presented in Section 19.1 of NuScale DCA Part 2, Tier 2. NuScale evaluated the basic events from the NuScale PRA for meeting specific risk significance criteria (see ER Table 4-1) or were in the top cutsets in the PRA for identifying SAMDA candidates specifically coupled to the NuScale design. These basic events and top cutsets are listed in ER Table 4-2. In Section 4.1, SAMDA Generation, NuScale describes thirteen NuScale components, systems, and events associated with the SAMDA candidates from the PRA evaluation of basic events and top cutsets. The thirteen NuScale components, systems, and events are:

1. Emergency core cooling system

2. Containment isolation

3. Control rod drive system

4. Reactor trip system

5. Decay heat removal system

6. Reactor coolant system

7. Backup power supply system

8. Containment flooding and drain system

9. CVCS

10. Module protection system

11. Reactor building crane (RBC) failures

12. Internal events

13. Fires, floods, high winds, external floods, seismic, and low power and shutdown events In general, for each item above, NuScale briefly describes the item or its purpose, references the corresponding Final Safety Analysis Report section of NuScale DCA Part 2, Tier 2 for further descriptions, presents general information about the PRA basic events related to the item, and specifies the SAMDAs by number being considered for their potential to reduce risk.

The incorporation of the standard list of pressurized water reactor SAMAs of NEI 05-01A Table 14 provides for 153 SAMA candidates. An additional 46 design-specific SAMDAs are derived from the evaluation of the NuScale PRA basic events and top cutsets. The results of NuScales 3 The term SAMA is applied here since the standard list of pressurized water reactor SAMAs from Table 14 of NEI 05-01A does contain SAMAs relating to procedures and training.

24 SAMDA identification process are presented in Appendix A, List of Candidate SAMDAs and Screening Results, of the ER (NuScale 2019a).

4.2 Staff Review The staff reviewed the process applied by NuScale for the selection of the initial set of SAMDAs and for the additional basic events as candidates for screening under the cost-benefit analysis.

The review included discussions with NuScale during the environmental audits (NRC 2017, NRC 2018a, NRC 2018b, and NRC 2019a) and coordinating with the staff reviewing the NuScale PRA risk insights provided in the NuScale DCA, Part 2, Tier 2, Chapter 19. The staff found the risk insights from the NuScale PRA adequately addressed the Commissions objectives regarding the prevention and mitigation of severe accidents. Details of the staffs review of the NuScale PRA are provided in Chapter 19, Probabilistic Risk Assessment and Severe Accidents, of the staffs Safety Evaluation Report (NRC 2019c).

Based on the above, the staff determined that the set of SAMDA candidates and basic events evaluated in the NuScale ER as supported by NuScale DCA, Part 2, Tier 2, Chapter 19, addresses the major contributors to core damage except where noted for the three SAMDA candidates (17, 85, and 195). For those SAMDA candidates that continue to SAMDA screening, NuScale applied a systematic and comprehensive process for identifying potential plant improvements for the NuScale design, and that the set of potential plant improvements identified by NuScale is reasonably comprehensive and, therefore, acceptable for further evaluation. NuScales SAMDA candidate search included reviewing insights from the design-specific PRA study as well as assessing SAMAs based on accepted industry guidance.

However, when the design of the RBC and associated cost information is available in a future COL application that references the NuScale SMR standard plant design, the COL applicant will need to address the additional SAMDA environmental analysis in the application.

25 5.0 Cost Benefit Analysis of Potential Plant Improvements The potential for plant improvements, or the assessment of SAMDA candidates, to be cost beneficial, as evaluated by NuScale and as reviewed by the staff is discussed in this section.

5.1 NuScale Cost Benefit Analysis NuScale applied a SAMDA screening process based on the process presented in Section 6 of NEI 05-01A and is presented in NuScale ER Section 6, Assessment of SAMDA Candidates (NuScale 2019a). In summary, NuScale performed three screening steps in assessing whether a SAMDA candidate should be considered for a cost-benefit analysis.

5.1.1 Phase I Screening SAMDA candidates are evaluated to determine if they are not applicable to the NuScale design based on the following considerations or categories:

1. Not Applicable: If a SAMDA candidate does not apply to the NuScale design, it is not retained.

2. Already Implemented: If a SAMDA candidate has already been implemented in the NuScale design, it is not retained.

3. Combined: If a SAMDA candidate is similar in nature and can be combined with another SAMDA candidate to develop a more comprehensive or NuScale plant-specific SAMDA candidate, only the combined SAMDA candidate is retained.

4. Excessive Implementation Cost: If a SAMA requires extensive changes that will obviously exceed the maximum benefit, even without an implementation cost estimate, it is not retained.

5. Very Low Benefit: If a SAMDA from an industry document (i.e., a generic SAMDA) is related to a non-risk significant system for which change in reliability is known to have negligible impact on the risk profile, it is not retained.

6. Not Required for Design Certification: If a SAMDA relates to the development of procedures and training to reduce and mitigate human errors, it is not retained.

Operator procedures and training processes are not developed as part of a DCA.

7. Considered for Further Evaluation: Any SAMDA candidate that was retained after the above considerations or categories would have a cost-benefit analysis performed on them.

NuScales Phase I screening has the following results from 199 identified SAMDA candidates:

Not applicable to the NuScale design: 45 SAMDA candidates.

Already implemented: 18 SAMDA candidates.

NuScale notes that some SAMDA candidates were an equivalent replacement that fulfilled the intent of the SAMDA.

Combined: 13 SAMDAs.

Excessive implementation cost: 1 SAMDA candidate Very low benefit: 34 SAMDA candidates.

Not required for design certification: 37 SAMDA candidates.

Considered for further evaluation: 51 SAMDA candidates.

26 Of the 51 SAMDA candidates considered for further evaluation under Phase II, 21 SAMDA candidates were from the 153 standard pressurized water reactor SAMAs of NEI 05-01A with the remaining 30 SAMDA candidates directly connected to the NuScale design.

NuScale has identified 37 SAMDAs candidates that were not applicable as part of their DCA.

Thirty-four of these SAMDAs regards operator training or operator procedure enhancements.

Since operator training and procedures are not established at the design certification state, there cannot be an assessment as to their potential to be cost beneficial. Three of the SAMDA candidates evaluated as Not Required for Design Certification (SAMDAs 17, 85, and 197) are not associated with operator training or operator procedure enhancements. Rather, two of the three SAMDA candidates (SAMDAs 17 and 85) are associated with potential enhancements associated with a multi-unit site. Namely, there are at least two complete plants each with 12 NPMs in their respective Reactor Buildings. The potential enhancement considered under SAMDA 17 is to create a cross-tie for diesel fuel oil at a multi-unit site. Under SAMDA 85, the potential enhancement is to provide cross-unit connection of uninterruptible compressed air supply. The third SAMDA candidate (SAMDA 197) is associated with hardware or control equipment design enhancements of the RBC, where NuScale must still finalize the design of the RBC and would do so at a later stage of the design process. With SAMDA 197, the potential enhancement could automate the NPM transport process to the extent permitted by the RBC control system design. Until the design of the RBC can be completed, any specific RBC SAMDAs cannot be determined to appropriately address reducing the risk from a dropped NPM along with an associated cost for any such specific RBC SAMDA.

5.1.2 Phase II Cost-Benefit Analysis The NuScale analysis of the SAMDAs considered for further evaluation is presented in ER Section 6.3, Phase II SAMDA Screening (NuScale 2019a). As NuScale states, their analysis entails subtracting the value of the severe accident risk associated with the design after the SAMDA has been incorporated in the design from the maximum benefit derived in Section 5.0 (which conservatively assumed that the implementation of any SAMDA would reduce the total plant risk to zero).

5.1.3 Screening Sensitivity Applying the prior maximum benefit sensitivity study, NuScale discussed in ER Section 6.4, Screening Sensitivity, that only the Peach Bottom site and the 3 percent discount rate sensitivity cases have the possibility to result in a SAMDA candidate in Phase II to be potentially cost-beneficial (NuScale 2019a. As was seen in the base case, none of the SAMDA candidates considered in Phase II are related to the RBC and the maximum benefits for RC 1 through RC 7 would again be a very small part of the $383,000 and $341,000 maximum benefits previously obtained for the two sensitivity cases. NuScale states $3,320 for Sensitivity Case 3 and $2,800 for Sensitivity Case 14. However, NuScale points out that the difference in seismic characteristics between Surry and Peach Bottom to show that the seismic risk is a significant contributor to the maximum benefit. Thus, NuScale examined the three SAMDA candidates that are related to seismic improvement (SAMDAs 140, 187, and 188). In all three of these cases, NuScale points out that the extent of the redesigns to increase seismic ruggedness of plant components for the site, base mat isolation, and of the twelve NPMs themselves would be extensive. Thus, the resulting implementation costs for such seismic improvements would be greater than $383,000 and potentially as high as $5,000,000.

27 5.2 Staff Review For the Phase I screening, the staff accepts that all SAMDA candidates not involving the RBC (SAMDA 1 through SAMDA 196, SAMDA 198, and SAMDA 199) fall under one of the screening criteria presented above in Section 5.1.1. The SAMDA candidates that were passed on for the Phase II cost-benefit analysis, in most cases, are candidates that could reduce the risks associated with RC 1 through RC 7. However, the maximum benefits for RC 1 through RC 7 for either a single NPM or all twelve NPMs are both very small. Thus, for the Phase II cost-benefit analysis, the staff finds it to be reasonable that the costs for these SAMDA candidates are all going to be much greater than RC 1 through RC 7s maximum benefits (less than $100 for the single NPM and $1,200 for the twelve NPMs). Therefore, none of the SAMDA candidates for reducing the risks from RC 1 through RC 7 under Phase II would be potentially cost beneficial.

As for the sensitivity regarding redesigns to increase seismic ruggedness, the staff notes that the NuScale design must satisfy all seismic safety criteria for the selected COL site. Thus, the COL applicant would assess the need to adjust the NuScale design such that SAMDAs 140, 187, and 188 would be incorporated as safety modifications to satisfy regulatory seismic requirements for ensuring adequate protection of the publics health and safety.

The staff notes that SAMDAs 17 and 85 would also be assessed at the COL stage, if applicable, due to their multi-unit aspects where a COL applicant should consider for addressing the site-specific external events and natural phenomena. SAMDA 197 regarding RBC risk reductions should also be considered at the COL stage since the RBC design has to be finalized at this stage for a license to be granted. Due to the sensitivity to site parameters, the staff also expects assessing SAMDA 197 cost-benefits during the COL stage would also be preferable. It is at the COL stage where the site parameters are fully developed with complete documentation in the COL application that would support a full and complete assessment of all possible SAMDA candidates for the severe accident scenario with the most significant risks.

28 6.0 Conclusions The staffs independent confirmatory MACCS results demonstrate good agreement with the population dose and offsite property damage risks results reported in Revision 3 of the NuScale ER. The staff could not confirm whether the changes made to the DCF file to mimic the DCF file used in the SOARCA study were correctly implemented for this analysis. The good agreement in results is likely due more to the low risk from the radiological releases as a result of the low event probabilities for seven of the eight RCs. For making an environmental finding on SAMDAs, the organ doses (besides whole-body dose) or cancer risk results, which would be sensitive to potential implementation errors, are not necessary. However, this situation of the DCF file and the modelling of releases of organic and inorganic iodine are items not seen before in past NRC severe accident studies or new reactor reviews. Therefore, they should be considered for further investigation by the NRC as items under the MACCS code maintenance program in the Office of Nuclear Regulatory Research.

For the single NPM events, the staff found the development of the scenarios for RC 1 through RC 7, based on the PRA from Chapter 19 of Part 2, Tier 2 of the DCA, were reasonable. The staffs independent confirmatory analysis also derived similar results to NuScales analysis for these seven RCs. Based on the staffs review of NuScales SAMDA evaluation, the staff determined that NuScale adequately identified SAMDAs for release categories (RCs) 1 through 7 that could potentially reduce the risks and that these SAMDAs would not be cost-beneficial based on the given Surry site parameters.

NuScale has identified 37 SAMDAs candidates that were not applicable as part of their DCA.

Thirty-four of these SAMDAs regards operator training or operator procedure enhancements.

Since operator training and procedures are not established at the design certification stage, there cannot be an assessment as to their potential to be cost beneficial. Additionally, NuScale has two other SAMDA candidates, namely SAMDA 17 and SAMDA 85, that address multiple plant site risks. These two SAMDA candidates would need to be re-evaluated by a future applicant referencing the NuScale standard design that is proposing a site with multiple plants.

The third SAMDA candidate not evaluated (SAMDA 197) involves the RBC where an NPM is dropped damaging two other NPMs. SAMDA 197 cannot be evaluated as within the scope of this standard design certification since a final design of the RBC control system will be completed at a later design stage. As such, this SAMDA is not required for design certification.

As for the scenario that a dropped NPM damages two other operational NPMs, the staff cannot determine that damaging only one operational NPM by the dropped NPM is bounded by two operational NPMs being dislodged from their respective Reactor Building bays by the same dropped NPM. Section 19.1.7.4 of DCA Part 2, Tier 2, discusses a dropped NPM resulting in damage to the top of an NPM leading to a CVCS line break outside containment, which is RC 3.

NuScale contends that this type of damage is unlikely; however, NuScale does not provide further analysis in DCA Part 2, Tier 2, Section 19 to support this conclusion nor analysis to show dislodging two NPMs is more likely for the same strike from a dropped NPM.

For the sensitivity cases, the staff agrees with the general insights reached by NuScale in ER Section 5.8, for the given sensitivity case assumptions tiering from the three damaged NPMs scenario. These general insights should be applicable to a future COL site if that sites characteristics are similar to or bounded by the Surry site parameters.

Based on the staffs review of NuScales SAMDA evaluation, the staff determined that NuScale adequately identified SAMDAs for RC 1 through RC 7 that could potentially reduce the risks and

29 that these SAMDAs would not be cost-beneficial based on the given Surry site parameters.

Additionally, NuScale has two other SAMDA candidates, namely SAMDAs 17 and 85, that address multiple plant site risks. These two SAMDA candidates will need to be re-evaluated if a specific multi-unit site is proposed in a future licensing action referencing the NuScale certified design.

For the dropped NPM scenario of RC 8, NuScale will finalize the design of the RBC as part of a future licensing action but not within the design certification process currently before the NRC.

Thus, NuScale has not fully developed or evaluated the necessary potential SAMDA hardware modifications for the RBC and has not determined the related SAMDA implementation costs.

This situation for RBC SAMDA candidates applies for the single dropped NPM scenario or for the scenario with core damage to three NPMs due to a dropped NPM. The staff also determined the release frequency associated by NuScale to the core damage of three NPMs (1.05E-06 per year) is overly conservative. This overall probability could be reduced with additional analysis as to the probability that a dropped NPM would have enough force necessary to damage and/or knock two operational NPMs from their respective Reactor Building bays.

On May 11, 2020, NuScale and NRC discussed the potential need to submit a revision to the ER due to recent design changes that prevent postulated boron redistribution scenarios. The discussion was conducted as part of the NRC audit of these design changes. As a result of the meeting, NuScale submitted a letter (NuScale 2020b) to document its evaluation of the potential need to revise the ER to reflect those design changes. The boron redistribution design changes have been evaluated for their effect on the ER. NRC verified NuScales conclusion that the effect on the ER is limited to editorial changes associated with event sequence numbering and event descriptions for consistency with the Final Safety Analysis Report (FSAR) Chapter 19 changes that were included in DCA Revision 4.1 (NuScale 2020a). A COL applicant that references the NuScale design certification will need to provide a revised ER with the noted editorial changes.

The staff reviewed NuScales SAMDA analysis and determined the methods applied and the implementation of the methods are appropriate. Based on the staffs independent confirmatory evaluation as described in the previous sections, the staff finds the results of the NuScale risk and maximum benefit for the single NPM analysis with respect to the first seven RCs to be reasonable with no potentially cost-beneficial SAMDAs as assessed using the Surry site parameters. However, the staff cannot reach a finding regarding the maximum benefit and cost-benefits for potential enhancements regarding the RBC (i.e., RC 8). Therefore, a COL applicant that references the NuScale design certification rule will need to provide further SAMDA analyses once the design of the RBC is finalized and more information about procedures and training are available.

30 7.0 References Chanin, D. and M.L. Young. 1998. Code Manual for MACCS2: User's Guide, NUREG/CR-6613, Volume 1, May 1998, Sandia National Laboratories, Albuquerque, New Mexico. Accession No. ML17047A443.

NEI (Nuclear Energy Institute). 2005. Severe Accident Mitigation Alternatives (SAMA) Analysis, Guidance Document, NEI 05-01, Revision A. Washington, D.C. ADAMS Accession No. ML060530203.

NRC (U.S. Nuclear Regulatory Commission). 1990. Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants. NUREG-1150, December 1990, Washington, D.C. Accession No. ML040140729.

NRC (U.S. Nuclear Regulatory Commission). 1997. Regulatory Analysis Technical Evaluation Handbook. NUREG/BR-0184, January 1997, Washington, D.C. ADAMS Accession No. ML050190193.

NRC (U.S. Nuclear Regulatory Commission). 2007. Environmental Standard Review Plan Standard Review Plans for Environmental Reviews for Nuclear Power Plants. NUREG-1555, 2000 Main Report and 2007 Revisions, Washington, D.C. Available at http://www.nrc.gov/readingrm/doc-collections/nuregs/staff/sr1555/toc/.

NRC (U.S. Nuclear Regulatory Commission). 2004. NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission. September 2004. Washington, DC.

ADAMS Accession No. ML042820192.

NRC (U.S. Nuclear Regulatory Commission). 2009. Regulatory Audits, NRO-REG-108, April 2, 2009, Washington, D.C. ADAMS Accession No. ML081910260.

NRC (U.S. Nuclear Regulatory Commission). 2012. State-of-the-Art Reactor Consequence Analyses (SOARCA) Report. NUREG-1935. November 2012. ADAMS Accession No. ML12332A057.

NRC (U.S. Nuclear Regulatory Commission). 2016. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition - Severe Accidents. NUREG-0800, Chapter 19. Washington, D.C. Available at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/ch19/.

NRC (U.S. Nuclear Regulatory Commission). 2017. Audit Plan for the Regulatory Audit of Environmental Report for NuScale Power, LLC Regarding the Severe Accident Mitigation Design Alternative Analysis." August 22, 2017. Washington, D.C. ADAMS Accession No. ML17179A287.

NRC (U.S. Nuclear Regulatory Commission). 2018a. Summary Report for the Environmental Audit of Part 3 of the NuScale Design Certification Application. May 8, 2018. Washington, D.C.

ADAMS Accession No. ML18143B667.

NRC (U.S. Nuclear Regulatory Commission). 2018b. Audit Plan for the Second Regulatory Audit of Environmental Report for NuScale Power, LLC Regarding the Severe Accident

31 Mitigation Design Alternative Analysis. October 12, 2018. Washington, D.C. ADAMS Accession No. ML18284A259.

NRC (U.S. Nuclear Regulatory Commission). 2019a. Summary Report for the Environmental Audit of Part 3 of the NuScale Design Certification Application. March 15, 2019. Washington, D.C. ADAMS Accession No. ML19037A487.

NRC (U.S. Nuclear Regulatory Commission). 2019b. NuScale DCA - Safety Evaluation Report with Open Items - Chapter 19, "Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors". April 19, 2019. ADAMS Accession No. ML19073A071 NuScale (NuScale Power LLC). 2016. Applicants Environmental Report - Standard Design Certification, Part 3. Revision 0. December 2016. ADAMS Accession No. ML17013A296.

NuScale (NuScale Power LLC). 2018a. Applicants Environmental Report - Standard Design Certification, Part 3. Revision 1. December 2018. ML18086A196.

NuScale (NuScale Power LLC). 2018b. Applicants Environmental Report - Standard Design Certification, Part 3. Revision 2. October 2018. ADAMS Accession Number ML18310A346.

NuScale (NuScale Power LLC). 2019a. Applicants Environmental Report - Standard Design Certification, Part 3. Revision 3. August 2019. ADAMS Accession Number ML19241A432.

NuScale (NuScale Power LLC). 2019b. Part 02 - Final Safety Analysis Report (Rev. 3) - Part 02 - Tier 02 - Chapter 01 - Introduction and General Description of the Plant - Sections 01.01 -

01.10. Revision 3. August 22, 2019. ADAMS Accession No. ML19241A398.

NuScale (NuScale Power LLC). 2019c. Part 02 - Final Safety Analysis Report (Rev. 3) - Part 02

- Tier 02 - Chapter 19 - Probabilistic Risk Assessment and Severe Accident Evaluation -

Sections 19.00 - 19.05. Revision 3. August 22, 2019. ADAMS Accession No. ML19241A428.

NuScale (NuScale Power LLC). 2020a. Part 02 - Final Safety Analysis Report (Rev. 4.1) - Part 02 - Tier 02 - Chapter 19 - Probabilistic Risk Assessment and Severe Accident Evaluation -

Sections 19.00 - 19.05. Revision 4.1. June 19, 2020. ADAMS Accession No.ML20197A413.

NuScale (NuScale Power LLC). 2020b. Letter

Subject:

Submittal of Environmental Report:

Revision Status. July 10, 2020 ADAMS Accession No.ML20192A326 SNL (Sandia National Laboratories). 2013a. State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis. NUREG/CR-7110, Vol. 1, Rev. 1. May 2013, Sandia National Laboratories, Albuquerque, New Mexico. ADAMS Accession No. ML13150A053.

SNL (Sandia National Laboratories). 2013b. State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis, NUREG/CR-7110, Vol. 2, Rev. 1, August 2013, Sandia National Laboratories, Albuquerque, New Mexico. ADAMS Accession No. ML13240A242.

Webber (Webber, S.; Bixler, N.; McFadden, K.). 2019. SECPOP Version 4: Sector Population, Land Fraction, and Economic Estimation Program; Users Guide, Model Manual, and

32 Verification Report, NUREG/CR-6525, Rev. 2. June 2019. Sandia National Laboratories, Albuquerque, New Mexico. ADAMS Accession No. ML18057B293.

Sprung (Sprung, I., et al.). 1990. Evaluation of Severe Accident Risks: Quantification of Major Input Parameters, MACCS Input. NUREG/CR-4551, Vol. 2, Rev. 1, Part 7. December, 1990, Sandia National Laboratories, Albuquerque, New Mexico. Non-Publicly Available ADAMS Accession No. ML063540026.

33 APPENDIX A Development of NuScale MACCS Models in Support of the NuScale Design Certifications Environmental Assessment A.1 Introduction NuScale, the Applicant, submitted as one part of a design certification application (DCA) an Environmental Report (ER) to document the evaluation of severe accident mitigation design alternatives (SAMDAs) associated with their plant design (NuScale 2019). The NRC is required to include consideration of SAMDAs to conform with the National Environmental Policy Act (NEPA). The applicant performed MACCS calculations using eight severe accident source terms, or release categories, to determine the 0 to 50 mile whole-body population dose and 0-50 mile total economic cost. These offsite consequences were inputs to a SAMDA cost-benefit analysis to determine if there could be design alternatives that should be considered for incorporating into the NuScale design.

A.2 Objective The purpose of this appendix is to summarize the development of the staffs NuScale MACCS models and to document the staffs conclusions for the confirmatory assessment of the Applicants offsite consequence analysis as presented in their Environmental Report (NuScale 2019). The analysis performed by the Office of Nuclear Regulatory Research for the Office of New Reactors for MACCS technical support under User Need NRO-2018-01. This appendix also includes a summary of results from additional sensitivity calculations results. The staffs confirmatory calculations applied WinMACCS version 3.11.2 with MACCS version 3.11.0.1 (NRC 2019).

A.3.

Site Descriptions This analysis used MACCS models developed in support of the SOARCA Surry and Peach Bottom analyses (NRC 2013 and 2012). A brief description of those sites will be provided in Section A.3.1 and A.3.2. The initial model used as the starting point for the construction of the confirmatory model was the SOARCA Surry STSBO model from SNL 2013b. Model development for the Peach Bottom site applied the SOARCA Peach Bottom STSBO model from SNL 2013a.

A.3.1 Surry Site Description The Surry site has been selected by NuScale as the representative site for evaluating the postulated offsite severe accident risks necessary for performing a SAMDA assessment. A full description of the region surrounding the Surry Power Station can be found in the Dominion Energy Virginia Environmental Impact Statement (Dominion 2018). The Surry site is operated by Dominion Energy Virginia and is located in the northeastern corner of Surry County, Virginia.

Geographically, the plant is located in southeastern Virginia on Gravel Neck Peninsula, on the southern side of the James River, approximately 24 miles upstream from where the James River enters the Chesapeake Bay. The largest population center in the region is Virginia Beach, which is approximately 24 miles east-southeast of the plant. Figure A-1, taken from Dominion

34 2018, presents the geography 50 miles around the plant. The plant is surrounded by the James River on three sides, where the river can be up to three miles wide.

Table A-1 presents a breakdown of the population around the site, which is extracted from the site file for use in this analysis. In summary, the population distribution used in the original SOARCA Surry analysis (NRC 2013) is extrapolated to the year 2060 using a scale factor of 1.406 provided in the Applicants analysis (NuScale 2019). The original Surry site file was based on 2000 Census data which was escalated to the year 2005 (USCB 2005). The Applicant calculated the population scaling factor as the ratio of the estimated 2060 population taken from Census 2015 (416,795,000) and the July 1, 2005 population taken from Census 2005 (296,410,404) (USCB 2015; USCB 2005).

The applicant used the weather file which was used for the original SOARCA Surry analysis for their calculations, which can be found in NRC 2015a. This weather was for the year 2004 and uses weather readings taken from the plant. The total rainfall for that year was 51.22 inches, where the monthly rainfall is presented in Figure A-2.

Table A-2 summarizes the 2004 weather year at the Surry site used in this analysis. The site had stable or neutral weather conditions for a large fraction of the year. Measurable precipitation was recorded in 521 hours0.00603 days <br />0.145 hours <br />8.614418e-4 weeks <br />1.982405e-4 months <br /> in the given year, which is approximately 6 percent of the year.

The average wind speed at the site in 2004 was 2.27 m/s. Figure A-3 presents 64-sector windrose data. That is, it presents the probability that the wind is blowing in a particular direction for a given year. The fraction of the 50-mile population in a given direction is presented in Figure A-4 for comparison purposes.

A.3.2 Peach Bottom Site Description A full description of the region surrounding the Peach Bottom Atomic Power Station can be found in the Exelon Generation Company Environmental Impact Statement (Exelon 2018). The Peach Bottom site is jointly owned by Exelon Generation Company and PSEG Nuclear, and is located in the Peach Bottom Township, York County, Pennsylvania. Geographically, the plant is located on the west side of Conowingo Pond on the Susquehanna River, approximately 18 miles upstream of the location where the Susquehanna River enters the Chesapeake Bay and 8 miles upstream from the Conowingo Dam. Exelon 2018 states no major metropolitan centers are within 6 miles of the site. The site is approximately 19 miles southwest of Lancaster, Pa, 30 miles southeast of York, PA and 38 miles north of Baltimore, MD. The areas within 6 miles of the site are predominantly rural, characterized by farmlands and woods. Figure A-5, taken from Exelon 2018, presents the geography 50 miles around the plant.

Table A-3 presents a breakdown of the population around the site, which is extracted from the site file used for the Peach Bottom analysis. In summary, the population distribution used in the original SOARCA Peach Bottom analysis of NRC 2013 is extrapolated to the year 2060 using a scale factor of 1.406 provided in the Applicants analysis, as discussed in Section B.1.4 (NuScale 2019). The original Peach Bottom site file was based on 2000 Census data which was escalated to the year 2005.

35 Figure A-1.

Map of 50-mile radius surrounding the Surry Power Station. (Dominion 2018)]

Table A-1.

Population distribution surrounding the Surry Plant used in this analysis.

Distance Interval (mi)

Radial Interval Population (individuals)

Cumulative Population (individuals)

Interval Population Density (individuals/mi2) 0-1 0

0 0.0 1-2 52 52 5.5 2-5 5825 5877 88.2 5-10 163584 169461 694.9 10-20 444095 613556 471.0 20-30 504088 1117644 321.0 30-40 1052173 2169817 478.6 40-50 1018303 3188120 360.0

36 Figure A-2.

2004 monthly rainfall totals at the Surry site.

Table A-2.

Summary of 2004 meteorology at the Surry site.

Category Metric Frequency Stability Class Unstable (%)

4%

Neutral (%)

29%

Stable (%)

67%

Precipitation Total (in) 51.22 Nonzero hours 521 Percentage of year (%)

6%

Wind Speed Average (m/s) 2.27

37 Figure A-3.

2004 windrose data at the Surry site in 64 sectors.

Figure A-4.

Fraction of the scaled 50-mile population surrounding the Surry site in a given direction.

38 Figure A-5.

Map of 50-mile radius surrounding the Peach Bottom Power Station.

(Exelon 2018)

Table A-3.

Population distribution surrounding the Peach Bottom Power Station used in this analysis.

Distance Interval (mi)

Radial Interval Population (individuals)

Cumulative Population (individuals)

Interval Population Density (individuals/mi2) 0-1 186 186 59.2 1-2 480 666 50.9 2-5 11170 11836 169.2 5-10 49231 61067 209.1 10-20 581105 642172 616.3 20-30 1293727 1935899 823.8 30-40 2889208 4825107 1314.1 40-50 2773916 7599023 980.7

39 The applicant used the weather file which was used for the original SOARCA Peach Bottom analysis for their calculations, which can be found in NRC 2015b. This weather was for the year 2006 and uses weather readings taken from the plant. The total rainfall for that year was 44.42 inches, where the monthly rainfall is presented in Figure A-6.

Table A-4 summarizes the 2006 weather year at the Peach Bottom Power Station used in this analysis. The site had stable weather or neutral weather conditions for a large fraction of the year. Measurable precipitation was recorded in 602 hours0.00697 days <br />0.167 hours <br />9.953704e-4 weeks <br />2.29061e-4 months <br /> in the given year, which is approximately 7 percent of the year.

The average wind speed at the Peach Bottom site in 2006 was 2.12 m/s. Figure A-7 presents 64-sector windrose data. That is, it presents the probability that the wind is blowing in a particular direction for a given year. The fraction of the 50-mile population in a given direction is presented in Figure A-8 for comparison purposes.

Figure A-6. 2004 monthly rainfall totals at the Peach Bottom site.

40 Table A-4. Summary of 2004 meteorology at the Peach Bottom site.

Category Metric Frequency Stability Class Unstable (%)

18%

Neutral (%)

25%

Stable (%)

58%

Precipitation Total (in) 51.22 Nonzero hours 602 Percentage of year (%)

7%

Wind Speed Average (m/s) 2.12 Figure A-7. 2006 windrose data at the Peach Bottom site in 64 sectors.

41 Figure A-8. Fraction of the scaled 50-mile population surrounding the Peach Bottom site in a given direction.

A.4 NRC Development of a NuScale MACCS Model As part of the staffs NEPA review of the NuScale ER, independent confirmatory analysis of the offsite consequences and subsequent risk values were determined to be necessary. This required the building of a NuScale MACCS model by the staff based on publicly available information, such as the NuScale ER or prior NRC documents, as supplemented by information in Part 2 of the DC application, and by information obtained during the staffs two environmental audits. Because NuScale applied a number of the MACCS input parameter values from the State-of-the-Art Reactor Consequences Analyses (SOARCA) project (see Appendix B of NuScale 2019), prior staff-developed SOARCA MACCS models formed the starting point for the staff to independently build their own NuScale MACCS models (SNL 2013a, SNL 2013b).

The details regarding the staffs development of a Surry baseline model and a Peach Bottom sensitivity model are provided in the MACCS Model Development in Support of the NuScale DCA Environmental Assessment: Summary Report under Section 4, Methodology (NRC 2019).

The staff first applied the input parameter values as presented and described in the NuScale ER to verify NuScales results. With this verification, the staff adjusted several of the MACCS input parameter values to create the staffs MACCS confirmatory model. During the review of the NuScale MACCS model during the two audits, the staff found that there were two noticeable differences in what NuScale did in their MACCS model versus current MACCS modeling guidance. The first is related to the dose conversion factors (DCF) data library and the second concerns the modeling of the release of iodine for the dropped NPM release category. The remainder of this section discusses these two modeling issues.

42 The ingestion model and economic costs were not considered in the original SOARCA studies, except for use in long-term protective action decision-making (i.e. the decision to condemn property rather than decontaminate it). As previously noted, NuScales input models were examined to determine if the staff should apply new input parameters values. For the staffs independent confirmatory analysis, MACCS input parameter values were set based on recent internal NRC MACCS assessments and are discussed in NRC 2019c. Based on this review by the staff, new site files were created based on the 2010 census and 2012 economic county level economic data, where the population was scaled to the year 2060 and the economic values in the site file were escalated to 2018. The staffs MACCS confirmatory models were updated with new parameters. Outside of the MACCS modelling variations discussed in this section, NuScale principally applied the input parameter values from the Surry and Peach Bottom SOARCA analysis, which are documented in NUREG/CR-7110 (SNL 2013a; SNL 2013b).

A.4.1 Dose Conversion Factors Data Library The NuScale ER states the dose conversion factors (DCF) data library in MACCS, the Fgr13dcf.inp file, was updated by replacing the chronic inhalation dose factors for the 69 SOARCA nuclides updated with the mean values from NUREG/CR-7155 SOARCA Uncertainty Analysis Table 4.2-8. While unstated in the NuScale ER, examination of supporting documentation during a 2018 audit (NRC 2019a) indicate additional measures were taken to mimic the DCF file used in the SOARCA analysis. The intent was to allow NuScale to use the same cancer risk coefficients used in the SOARCA analysis. The staff could not verify from the information provided in the NuScale ER if the updated DCF values were implemented correctly.

Rather than attempt to replicate the steps NuScale described to create a modified dose conversion factor file starting with the distributed Fgr13dcf.inp file, the staffs analysis uses the methodology the applicant is trying to replicate. That is to say, this staffs analysis will use the original dose conversion factor file created for the SOARCA program, which is consistent with the cancer risk coefficients defined in the SOARCA models.

The food model was required to be turned on for this analysis to allow the calculation of economic costs. The NuScale ER does not discuss the implementation of the food ingestion model. WinMACCS 3.10 is distributed with a file which defines the dose conversion factors due to the ingestion of food, fgr13samp_a.bin, which is generated by the COMIDA2 food ingestion model. The staffs confirmatory analysis applied the updated COMIDA2 file developed for use in WinMACCS version 3.11.2, which is consistent with the cancer risk coefficients defined in the original SOARCA analysis, comida2_FGR13GyEquivDCF.bin. As in the case for the dose conversion file FGR13GyEquivDCF.INP, this file is currently not available for public use.

A.4.2 Iodine Release for Dropped NPM Release Category The NuScale ER does not provide extensive information regarding the creation of the MACCS model required to perform the calculation for the dropped NPM scenario, or Release Category 8 as designated in Appendix B of the NuScale ER. Because the dropped NPM is assumed to end up lying horizontally at the bottom of the Ultimate Heat Sink pool, NuScale states that most of released isotopes are scrubbed by traversing the pool, where the only releases occur from the noble gases, elemental iodine, and organic iodine. The organic iodine was taken as 0.15% of the initial iodine core inventory. NuScale must somehow account for two different chemical types of iodine being released from the dropped NPM. However, MACCS does not allow a user to use the same isotope within multiple classes, e.g. I-131 cannot appear in class 2 and class 4.

43 There is no elaboration in the NuScale ER on how NuScale created a MACCS model which can handle the release of iodine in both forms.

For the staffs confirmatory analysis, new isotopes were defined for use in WinMACCS with the appropriate properties assigned, such as half-lives and dose conversion factors. Doing this requires multiple changes to the initialization files used by WinMACCS, in addition to changes in the model input. Rather than create a special model that only includes the three isotope groups of interest, a tenth isotope group is added to the model to minimize the number of required changes. Also, this insures all the original isotope classes are available if iodine decays to one of the other chemical groups. The isotopes of interest for the new organic iodine isotopes to be defined include I-131, I-132, I-133, I-134, and I-135. A new chemical class was created which represented the organic iodine and was called CH3I for the confirmatory calculations.

The particle size distribution for this release category is taken from Table B-23 of the NuScale ER (NuScale 2019a). In Table B-23, organic iodine is defined with a particle size distribution of 0.1 for each of the 10 particle size bins. This is the same as the treatment for noble gases, therefore the organic iodine is defined to be non-depositing for both dry and wet deposition, similar to the noble gas. The release fractions, along with the remaining definition of the plume segments, are defined in Table B-24 of the NuScale ER (NuScale 2019a).

The isotopic inventory of the nuclides is assumed to come from Table B-5 of the NuScale ER.

NuScale states that the initial core inventory of organic iodine, designated as Io for iodine-organic, is taken as 0.15% of the total core inventory. Therefore, five new isotopes (Io-131, Io-132, Io-133, Io-134, and Io-135) are included in the list of isotopes, where the assigned core inventory, in Bq, is taken as 0.15% of the corresponding iodine inventory defined in NuScale ER Table B-5.

All of the changes indicated above define an inventory and WinMACCS model which includes a new and unique chemical class. Additional changes to the WinMACCS initialization files, and DCF files, are needed for this calculation to work properly. The INDEXR.DAT file included with the WinMACCS installation provides decay information for the calculation. The appropriate entries are copied and placed at the end of the file, and the duplicate isotopes are provided a new name for reference. The decay products of these new isotopes are directed to the normal decay products. A special WinMACCS executable directory was created to execute these problems, called WinMACCS 3.11.2_NuScale+Io, which contains the modified INDEXR.DAT file. Using the executable providing in this directory will initialize WinMACCS with the additional decay information for the newly created organic iodine.

Additionally, the newly defined organic iodine isotopes do not have DCF factors available. To overcome this limitation, a new DCF file is created. Starting with the DCF filed used for the other seven release categories (FGR13GyEquivDCF.INP), the number of nuclides defined on the file is increased to accommodate the new isotopes. Next, the biokinetic information for the appropriate isotopes were copied and pasted at the bottom of the list, where the name of the isotope is updated to a new reference name. Finally, the DCF values for the new isotopes are copied and placed at the bottom of the DCF file, and the names are updated.

While the method employed in the confirmatory analysis would allow the calculation of the release of iodine in multiple chemical forms, a number of assumptions are made. Creating a DCF file as described above will allow MACCS to calculate dose contributions due to the organic iodine as defined, but this assumes that organic and elemental iodine behave identically upon ingestion, that is the same underlying biokinetic models are applicable for both elemental and organic iodine. Additionally, while these changes can be performed for the DCF files, the

44 comida files used for the long-term ingestion pathways model are binary and cannot be easily modified. Without changes to this file, the organic iodine does not contribute to the long-term phase dose. The organic iodine radionuclides are also assumed to be non-depositing.

None of the above assumptions are addressed in NuScales ER documentation. To examine the influence of these assumptions, a sensitivity was performed to combine the elemental and organic releases into a single elemental class, therefore all the iodine released during the event will be treated as elemental iodine, which will allow the iodine original released as organic iodine to deposit into the environment and use the same DCF values as the elemental iodine. This sensitivity will be discussed further in Section A.5.4.

A.5 Staffs Confirmatory Analysis Results To perform its independent confirmatory analysis, the staff selected specific MACCS calculations performed by NuScale based on the staffs assessment of their overall impact toward the subsequent cost-benefit analysis. The staff identified the Surry site baseline case, the Peach Bottom site sensitivity case, and the Surry site sensitivity case with the high burn-up core inventory as the NuScale calculations with the potential to have the most significant impact on the SAMDA cost-benefit assessment. Additionally, as previously described, the staff evaluated an alternative calculation of the dropped NPM scenario applying only one radioactive iodine chemical group for the Surry site baseline case. The staff applied WinMACCS Version 3.11.2 while NuScale applied WinMACCS Version 3.10.1.2. The results of the staffs analysis for these confirmatory analyses are presented below.

A.5.1 Surry Site Baseline Results The results of the staffs confirmatory analysis are presented in Table A.5-1 for the offsite population exposure and for offsite property damage (i.e., total cost).

A.5.2 Peach Bottom Site Sensitivity Results The results of the Peach Bottom site sensitivity analysis for the staff and NuScale are also presented in Table A.5-1 for the offsite population exposure and for offsite property damage (i.e.

total cost). Note that the Peach Bottom site results are higher than for the Surry site baseline results due to the greater numbers for the population distribution and greater urban areas. Thus, the site parameters can have a noticeable effect on the value of the risks.

A.5.3 Surry Site Sensitivity Results Applying the high burn-up core inventory assumption, the results of this sensitivity analysis for the Surry site are presented in Table A.5-1 for the offsite population exposure and for offsite property damage (i.e., total cost). While the sensitivity case consequences are higher, as one would expect.

45 Table A.5-1. Overall Population Dose and Total Cost for the Sensitivity Cases With Updated Input Parameters.

Release Category Surry Site Peach Bottom Site Surry Site with High Burnup Core Inventory Overall Population Dose (person-rem/event)

Total Cost

($-2018)

Overall Population Dose (person-rem/event)

Total Cost

($-2018)

Overall Population Dose (person-rem/event)

Total Cost

($-2018)

RC01 2.56E+01 2.02E+02 4.30E+01 9.37E+02 5.31E+01 6.83E+02 RC02 1.79E+01 1.53E+02 2.74E+01 5.80E+02 4.69E+01 6.14E+02 RC03 1.26E+06 2.92E+09 1.42E+06 1.87E+010 2.16E+06 7.99E+09 RC04 1.61E+01 1.23E+02 2.47E+01 4.54E+02 4.20E+01 5.40E+02 RC05 1.80E+05 8.72E+07 2.38E+05 8.01E+08 4.04E+05 4.81E+08 RC06 1.18E+01 5.07E+01 1.82E+01 1.43E+02 3.03E+01 3.55E+02 RC07 1.72E+01 1.44E+02 2.63E+01 5.32E+02 4.51E+01 5.88E+02 RC08 7.78E+02 8.31E+03 1.03E+03 1.30E+05 8.38E+02 9.70E+03 A.5.4 Dropped NPM Iodine Release Sensitivity Results NuScale modeled Release Category 8, the dropped NPM scenario, with two separate chemical groups to represent iodine, in both elemental and organic forms, but the underlying assumptions of this method were not discussed in the NuScale ER The assumptions applied by the staff are:

1) organic iodine was treated as non-depositing, 2) elemental and organic iodine behave identically upon intake (e.g. inhalation), and 3) organic iodine does not contributed to the food ingestion dose in the long-term phase. Assumption 1 arises from how the model is set up, but there is evidence that while organic iodine has a low deposition velocity, it is not non-depositing.

This also ignores the potential for organic iodine to decompose into elemental iodine upon exposure to sunlight in the environmental. Assumption 2 arises from the fact that the DCF values present were copied from the DCF file and assigned to the newly defined isotopes. The DCF values are a function of not just dose but biokinetics models, therefore this assumption means both the elemental and organic form behave the same biokinetically upon intake.

Assumption 3 arises from the fact that the file providing DCF factors for the food ingestion model are within a binary file which cannot be easily modified. Therefore, by applying Assumption 3, the potential dose contribution of the organic iodine would be ignored.

This sensitivity calculation was performed to examine these assumptions. The goal of this sensitivity was not to provide a method for the treatment of organic iodine releases, but rather to provide a bounding calculation to remove the potential influence of the assumptions. This sensitivity was set up to combine the elemental and organic releases by activity weighting the respective release fractions. The result would yield a single release fraction which is treated like elemental iodine, therefore it would be able to: 1) deposit into the environment, 2) it would be modeled explicitly as elemental iodine so using the provided DCF values would be appropriate, and 3) it would be able to contribute to food ingestion.

46 Table A.5-2. Overall Whole-Body Population Dose and Total Economic Costs for Release Category 8 Sensitivity Case with Combine Elemental and Organic Iodine Release Fractions.

Release Category Overall Whole-Body Population Dose (person-rem)

Total Economic Costs

($ - 2018)

RC 8 (Base Case) 7.78E+02 8.31E+03 RC 8 (Combined Iodine) 9.09E+02 1.40E+04 A.5.5 Staff Observations The ingestion model and economic costs were not considered in the original SOARCA studies, except for use in long-term protective action decision-making (i.e. the decision to condemn property rather than decontaminate it). As previously noted, NuScales input models were examined to determine if the staff should apply new input parameters values. Based on this review by the staff, new site files were created based on the 2010 census and 2012 economic county level economic data, where the population was scaled to the year 2060 and the economic values in the site file were escalated to 2018. The Surry, Peach Bottom, and Surry with high burnup inventory confirmatory models were updated with new parameters. Because the modelling of organic iodine releases for its potential dose contribution cannot currently be accounted for in the MACCS code, this radionuclides dose pathway should be considered for further investigation as items under the MACCS code maintenance program in the Office of Nuclear Regulatory Research.

A.6 Conclusions WinMACCS models, using WinMACCS version 3.11.2, were created by the staff which are capable of replicating the results presented by NuScale for three scenarios: 1) assuming a plant is sited at the Surry site, 2) assuming the plant was sited at the Peach Bottom site, and 3) assuming a high burnup core inventory for the Surry site (NuScale 2019).

While the confirmatory models developed for this analysis confirm NuScales results, the DCF modification process followed by NuScale was not confirmed explicitly. NuScale intended to replicate the FGR13GyEquivDCF.inp file developed and used for the original SOARCA studies, which is not currently available to the public, so the same cancer risk factors could be used. The confirmatory models used the original FGR13GyEquivDCF.inp file. NuScale did not report results where discrepancies could occur, such as organ doses or cancer risks. Therefore, the staff cannot determine if the changes made by NuScale were correctly implemented. However, the staffs confirmatory calculations show the modifications will likely have no significant impact on the reported results.

Additionally, NuScale did not provide a thorough description of the MACCS model used in the calculation for the dropped NPM scenario, namely Release Category 8. This source term separated the iodine release into an elemental and organic chemical class. MACCS cannot easily assign the same isotope into multiple chemical groups. The staff developed a confirmatory model to treat the elemental and organic iodine species as separate chemical classes, which was able to replicate the NuScales results. This model required changes to the DCF file and an MACCS initialization file to provide dose conversion factors for the new chemical group isotopes and the INDEXR.DAT file to provide decay information. NuScale did

47 not address any of the assumptions that this method contains, which led to the need for the performance of a sensitivity calculation by the staff.

Given the above factors, principally due to the staffs confirmation of the NuScale results, the staff will apply the NuScale offsite population exposure and offsite property damage results in the staffs cost-benefit confirmatory analysis.

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Exelon (Exelon Generation Company, LLC). 2018. Applicants Environmental Report -

Operation License Renewal Stage - Subsequent License Renewal: The Second License Renewal - Peach Bottom Atomic Power Station. July 2018. ADAMS Accession Number ML18201A219.

NuScale (NuScale Power LLC). 2019a. Applicants Environmental Report - Standard Design Certification, Part 3. Revision 3. August 2019. ADAMS Accession Number ML19241A432 NRC (U.S. Nuclear Regulatory Commission). 2012. State-of-the-Art Reactor Consequences Analyses Project, Volume 1: Peach Bottom Integrated Analysis, NUREG/CR-7110, Revision 1, January 2012.

NRC (U.S. Nuclear Regulatory Commission). 2013. State-of-the-Art Reactor Consequences Analyses Project, Volume 2: Surry Integrated Analysis, NUREG/CR-7110, Revision 1. August 2013.

NRC (U.S. Nuclear Regulatory Commission). 2015a. SOARCA Surry 2004 Meteorological File, April 2015. ADAMS Accession Number ML15097A117.

NRC (U.S. Nuclear Regulatory Commission). 2015b. SOARCA Peach Bottom 2006 Meteorological File, April 2015, ADAMS Accession Number ML15097A114.

NRC (U.S. Nuclear Regulatory Commission). 2019a. Summary Report for the Environmental Audit of Part 3 of the NuScale Design Certification Application. March 2019. ADAMS Accession Number ML19037A487.

NRC (U.S. Nuclear Regulatory Commission). 2019b. MACCS Model Development in Support of the NuScale DCA Environmental Assessment: Summary Report. October 2019. ADAMS Accession Number ML19284B128, Non-Publicly Available.

SNL (Sandia National Laboratories). 2013a. State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis. NUREG/CR-7110, Vol. 1, Rev. 1. May 2013.

Sandia National Laboratories, Albuquerque, New Mexico. ADAMS Accession No. ML13150A053.

48 SNL (Sandia National Laboratories). 2013b. State-of-the-Art Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis, NUREG/CR-7110, Vol. 2, Rev. 1. August 2013.

Sandia National Laboratories, Albuquerque, New Mexico. ADAMS Accession No. ML13240A242.

USCB (U.S. Census Bureau). 2005. U.S. Department of Commerce. Estimates of the Population for the United States and States, and for Puerto Rico: April 1, 2000 to July 1, 2005 (NST-EST2005-01). https://www2.census.gov/programs-surveys/popest/tables/2000-2005/state/totals/nst-est2005-01.xls USCB (U.S. Census Bureau). 2015. U.S. Department of Commerce. Economics and Statistics Administration. U.S. Census Bureau. Projections of the Size and Composition of the U.S.

Population: 2014 to 2060: Population Estimates and Projections - Current Population Reports.

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