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Advanced Reactors GEIS Docs - PNNL - Advanced Nuclear Reactor Plant Parameter Envelope and Guidance

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PNNL - Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Attachments: McDowell and Goodman NRIC-PPE-Guidance-Feb-2021-Final.pdf

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance BK McDowell D Goodman NRIC-21-ENG-0001; PNNL-30992 l 2.18.2021

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Summary Pacific Northwest National Laboratory (PNNL) is supporting the National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) by developing advanced nuclear reactor plant parameter envelopes (PPEs) to facilitate environmental reviews of potential future advanced reactor demonstration projects at INL and elsewhere in the United States. Two PPEs are developed in this report for two size ranges: (1) microreactors, which are defined for this PPE as single units with outputs of 60 MWt or less, and (2) small- to medium-sized advanced reactors with outputs above 60 MWt up to 1,000 MWt.

This report describes the methodology for developing the PPEs, including reactor vendor responses to NRIC questionnaires, input from INL staff, independent assessments by subject matter experts, and a review of regulatory requirements a vendor would have to meet during construction and operation. This report presents the compiled PPEs for surrogate plants derived from these inputs, lists documentation supporting the PPE, and provides recommendations for its use when developing an environmental impact assessment.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 iii

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Acronyms and Abbreviations AC alternating current ALARA as low as reasonably achievable ANR advanced nuclear reactor BWR boiling-water reactor BWXT BWX Technologies CFR Code of Federal Regulations CITRC Critical Infrastructure Test Range Complex CT combustion turbine dBA A-weighted decibel DOE U.S. Department of Energy DoD U.S. Department of Defense EBR-II Experimental Breeder Reactor II EIS environmental impact statement ER Environmental Report ESP early site permit ESRP Environmental Standard Review Plan FLIBE fluorine-lithium-beryllium FHR fluoride salt-cooled high-temperature reactor FPR Fuel Processing Restoration FSAR Final Safety Analysis Report FTE full-time equivalent employee GAIN Gateway for Accelerated Innovation in Nuclear GDC General Design Criteria GE General Electric GEIS generic environmental impact statement GHG greenhouse gas HALEU high-assay low-enriched uranium HTGR high-temperature gas-cooled reactor HTTR high-temperature engineering test reactor INL Idaho National Laboratory LFR lead-cooled fast reactor LFTR liquid fluoride thorium reactor LMR liquid metal-cooled reactor LRWS liquid radioactive waste management system LWR light-water reactor MFC Materials and Fuels Complex MSR molten salt reactor MSRE Molten Salt Reactor Experiment MWe megawatt(s) electric MWt megawatt(s) thermal NAAQS National Ambient Air Quality Standards McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 v

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance NEI Nuclear Energy Institute NEPA National Environmental Policy Act NGCC natural gas combined cycle NRC U.S. Nuclear Regulatory Commission NRIC National Reactor Innovation Center ORIGEN Oak Ridge Isotope Generation code PBF Power Burst Facility PBR pebble bed reactor PCS power conversion system PNNL Pacific Northwest National Laboratory PPE plant parameter envelope PRISM Power Reactor Innovative Small Module PSEG Public Service Enterprise Group PWR pressurized water reactor RCCWS reactor component cooling-water system RCRA Resource Conservation and Recovery Act RD&D research, design, and development RIC Regulatory Information Conference ROW right-of-way RWDS radwaste drain system SCALE Standardized Computer Analyses for Licensing Evaluation SC-HTGR steam cycle high-temperature gas-cooled reactor SFR sodium fast reactor SME subject matter expert SMR small modular reactor SPE site parameter envelope TRISO tri-structural isotropic TRITON Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion (software)

TVA Tennessee Valley Authority UWS utility water system VTR Versatile Test Reactor ZPPR Zero Power Physics Reactor McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 vi

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Contents Summary...............................................................................................................................................iii Acronyms and Abbreviations ................................................................................................................ v 1.0 Introduction ................................................................................................................................ 1 1.1 NRIC Background............................................................................................................ 1 1.2 Advanced Reactor Types and Descriptions ..................................................................... 2 1.3 Report Content and Organization .................................................................................... 3 2.0 PPE/SPE Background and Adaptation ....................................................................................... 4 2.1 PPE Approach and NEPA Compliance ........................................................................... 4 2.2 Use of PPEs in Early Site Permit Applications................................................................. 4 2.3 Adaptation of the PPE/SPE Approach to NRIC at INL..................................................... 5 2.3.1 PPE Approach.................................................................................................... 5 2.3.2 SPE Approach.................................................................................................... 7 3.0 NRIC PPE Development ............................................................................................................ 9 3.1 Advanced Reactor Vendor Questionnaire Responses .................................................... 9 3.2 INL Input .......................................................................................................................... 9 3.3 Subject Matter Expert Input ........................................................................................... 10 3.4 NRC Advanced Reactor GEIS ....................................................................................... 10 3.5 Versatile Test Reactor Draft EIS.................................................................................... 11 3.6 Regulatory Limits ........................................................................................................... 11 4.0 Summary PPEs ........................................................................................................................ 12 4.1 Microreactor PPE .......................................................................................................... 12 4.2 Small- to Medium-Sized Advanced Reactor PPE .......................................................... 16 5.0 Guidance for Use of the PPE ................................................................................................... 20 5.1 Use of the PPE in the NEPA Process............................................................................ 20 5.2 Limitations ..................................................................................................................... 20 6.0 References ............................................................................................................................... 23 Appendix A - Vendor Questionnaire ..................................................................................................A.1 Appendix B - INL Supporting Information ..........................................................................................B.1 Appendix C - SME Reactor Plant Parameter Value Assessments ................................................... C.1 Appendix D - NRC Advanced Reactor GEIS Values ........................................................................ D.1 Appendix E - PPE Data Sources and Methodology ..........................................................................E.1 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 vii

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Figures Figure C.1. Utility Water System Diagram ...................................................................................... C.8 Tables Table 4.1. Microreactor PPE .......................................................................................................... 12 Table 4.2. Small- to Medium-Sized Advanced Reactor PPE ......................................................... 16 Table C.1. Clinch River ESP Projected Blowdown Constituents and Concentrations ................... C.9 Table C.2. Nuclear Power Plant Lifetime Greenhouse Gas Footprints(a) .................................... C.14 Table C.3. Minimum, Maximum, and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 0 (End of Operation) for Microreactors ............................. C.17 Table C.4. Minimum, Maximum, and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 1 ....................................................................................... C.19 Table C.5. Minimum, Maximum, and Average Radionuclide Activity at Time 0 .......................... C.20 Table C.6. Minimum, Maximum and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 0 (End of Operation) for Small- to Medium-Sized Advanced Reactors ................................................................................................... C.47 Table C.7. Minimum, Maximum, and Average Radionuclide Activity at Time 0 (End of Operation) for Small- to Medium-Sized Advanced Reactors ..................................... C.52 Table D.1. NRC Advanced Reactor GEIS Draft Plant Parameter Envelope ................................. D.1 Table D.2 NRC Advanced Reactor GEIS Draft Site Parameter Envelope ................................... D.4 Table E.1 Microreactor PPE Data Sources and Methodology ......................................................E.1 Table E.2 Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology ....E.11 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 viii

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 1.0 Introduction The National Reactor Innovation Center (NRIC) at Idaho National Laboratory (INL) in and near Idaho Falls, Idaho, was authorized under the Nuclear Energy Innovation Capabilities Act (Public Law 115-248) to provide innovators with resources and infrastructure for testing, demonstration, and performance assessment to accelerate demonstration and deployment of new advanced reactor technology concepts. NRIC plans to offer existing buildings and multiple undeveloped and previously developed sites at INL to advanced reactor developers for use in demonstrating a wide range of reactor technologies, designs, and sizes.

A National Environmental Policy Act (NEPA; 42 U.S.C. § 4321 et seq.) environmental review may be initiated to facilitate NRICs mission before any commitments by individual reactor vendors are made.

To evaluate the potential environmental impacts without knowing which particular designs may be deployed, NRIC has developed plant parameter envelopes (PPEs) based on potential advanced reactor demonstrations. These PPEs could be used by the U.S. Department of Energy, other federal agencies, and/or others to evaluate the largest or bounding environmental impacts of the deployment of any particular advanced reactor with design parameters falling within the envelopes. In this way, a NEPA review can be initiated prior to final design selection, thereby streamlining the regulatory review process for several buildings, sites, and reactor designs.

1.1 NRIC Background NRIC accelerates the deployment of advanced nuclear energy through its mission to inspire stakeholders and the public, empower innovators, and deliver successful outcomes. NRIC is a national program led by INL, allowing collaborators to harness the world-class capabilities of the U.S. National Laboratory System. NRIC is charged with and committed to demonstrating advanced reactors.

NRIC accelerates technology from proof of concept to proof of operation by allowing innovators to leverage the U.S. governments investment in nuclear energy research, development, demonstration, and deployment. By bridging world-leading laboratory infrastructure and expertise with the promise of visionaries working to commercialize new nuclear energy systems, NRIC is enabling a new era of clean, affordable, reliable energy.

NRIC intends to provide existing facilities and other undeveloped and previously developed sites at INL to advanced reactor vendors for prototype technology testing and deployment. The existing Materials and Fuels Complex (MFC)-767 (Experimental Breeder Reactor II [EBR-II]), Zero Power Physics Reactor (ZPPR), and Power Burst Facility Building 613 (PBF-613; Critical Infrastructure Test Range Complex [CITRC]) Communications Research Facility buildings will be modified for use as parts of a technology test bed (INL 2020a). Vendors will be able to install, test, and operate prototype reactor technology, then remove the prototype upon completion of testing. In addition, undeveloped and previously developed outdoor sites have been identified as potential locations for construction and operation of full-scale reactor prototypes (INL 2020a).

The national deployment of advanced reactors will require not only technical innovations, but innovations in the regulatory processes for siting and construction. To facilitate the U.S. Department of Energys (DOEs) NEPA review for reactor demonstration and deployment, NRIC has developed an approach adapted from the U.S. Nuclear Regulatory Commissions (NRCs) early site permit McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance (ESP) process under Title 10 of the Code of Federal Regulations Part 52 (10 CFR Part 52). The NRC developed the ESP process to give applicants the option to evaluate environmental impacts at development sites even before a specific reactor design is chosen, thereby allowing for early resolution of many environmental issues. After issuance of an ESP, a construction permit and operating license or a combined license (combined construction permit and operating license) would need to be completed before operation of any nuclear power plant could occur.

In 2003, to facilitate NRCs ESP process, the Nuclear Energy Institute (NEI) developed an approach to an ESP application based on a PPE and site parameter envelope (SPE) approach (NEI 2012). This report describes the adaptation of NEIs PPE/SPE approach to NRCs ESP process to possible deployment of advanced reactor prototypes at INL and potentially for reactor demonstrations and/or deployments elsewhere in the United States.

1.2 Advanced Reactor Types and Descriptions Many different non-light-water reactor (LWR) technologies are in development. In addition to these general reactor types, there are several design-specific variations in materials, coolants, and geometries within each type and/or hybrids within/across these technologies. Sodium-cooled designs are more mature technologies that involve operational experience (e.g., the Fast Flux Test Facility at the Hanford Site in southeastern Washington).

A brief description of potential types of advanced reactors is provided below for context in developing a PPE. This list of advanced reactor types is intended to provide familiarity with and an overview for reviewers of the potential types of reactors, but it is not intended to be entirely comprehensive nor to exclude any specific reactor designs that are adequately represented by the parameters established in the PPE.

x High-temperature gas-cooled reactors (HTGRs) refer to graphite-moderated, typically helium-cooled systems that use tri-structural isotropic fuel micro particles. The particles are packed into a graphite matrix to form either spherical or cylindrical fuel elements. The pebble bed version of the HTGR uses spherical billiard ball-sized fuel elements that flow continuously through the reactor.

The prismatic version of the HTGR uses the cylindrical fuel compacts in hexagonal blocks in a fixed geometry. HTGRs may be used for electricity production and/or process heat applications.

x Fluoride salt-cooled high-temperature reactors (FHRs) refer to a hybrid design that uses pebble fuel elements (like pebble bed HTGRs) and a fluoride salt coolant (like salt-cooled molten salt reactors). Some fixed-fuel FHR designs (like prismatic HTGRs) have been proposed, but none is currently under commercial consideration.

x Molten salt reactors (MSRs) come in several varieties. Some designs use molten fluoride salt, while others use chloride salts as the coolant. Some designs have stationary fuel rods or plates, while others have moving fuel pebbles or fissile material dissolved within the flowing coolant. In addition, some MSRs use a fast neutron spectrum, while others use a thermal spectrum.

x Liquid metal-cooled reactors (LMRs) are an advanced type of nuclear reactor in which the primary coolant is a liquid metal. LMRs are classified based on the liquid metal coolant used, such as sodium, lead-bismuth eutectic alloy, and lead-bismuth.

x Heat pipe reactors typically consist of a solid block core with the fuel in holes inside the solid block. Heat pipes are built into the block in a lattice configuration and remove the heat from the block as the liquid in the heat pipe is vaporized.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance x Integral pressurized water reactors are an advancement upon historical pressurized water reactor designs that use coolant and fuels similar to existing LWRs, but that have the primary coolant circuit components placed within the reactor pressure vessel, thereby eliminating the need for primary circuit pipework with the intention of enhancing safety and reliability.

1.3 Report Content and Organization This report describes the methodology for developing two PPEs for surrogate nuclear plants, one for microreactors, which are defined for this PPE as single units with outputs of 60 MWt or less, and one for small- to medium-sized advanced reactors 1 with outputs above 60 MWt up to 1,000 MWt.

Section 2.0 describes the background of the PPE and SPE, how the NRC has used the PPE and SPE approach, and how this approach was adapted for NRIC. Section 3.0 describes the various sources of information used to inform the development of the NRIC PPEs. Section 4.0 provides the summary PPEs, one for microreactors and one for small- to medium-sized advanced reactors.

Section 5.0 is intended to give the reader guidance for the use of the PPEs, including limitations and environmental impact assessment considerations. Sources and documentation cited in the narrative are listed in Section 6.0. The appendices contain supplemental detail, including a vendor questionnaire (Appendix A), INL supporting information (Appendix B), advanced reactor parameter value assessments (Appendix C), NRC advanced nuclear reactor (ANR) generic environmental impact statement (GEIS) values (Appendix D), and the PPE Data Sources and Methodology (Appendix E).

1 The IAEA defines small as anything below 300 MWe and medium as below 700 MWe. (IAEA 2001).

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 2.0 PPE/SPE Background and Adaptation The concept of the PPE and how it has been used in previous ESP applications to evaluate environmental impacts is described below. Also described is the SPE, which has been used to bound environmental impacts prior to selection of a site, and how it has been adapted to NRICs needs.

2.1 PPE Approach and NEPA Compliance NRCs Environmental Standard Review Plans (ESRPs) provide guidance for developing Environmental Impact Statements for several different licensing actions (NRC 2000). While the ESRPs were originally prepared to guide staff in the review of construction permits and operating licenses only, in 2000 the NRC broadened the ESRPs to cover additional options including ESPs.

The benefits of the ESP approach are that site safety, environmental protection, and emergency preparedness issues are resolve independent of a specific nuclear plant design. The development of the PPE associated with a surrogate reactor was originally developed and published in NEI 10-01

[Revision 1], Industry Guideline for Developing a Plant Parameter Envelope in Support of an Early Site Permit, which states that [the PPE] approach provides an equivalent level of finality to that achieved through an ESP based on a specific reactor design (NEI 2012). In 2002, NRC staff conducted an internal ESRP workshop to consider the implications of ESP reviews employing the PPE approach rather than a specific nuclear plant design. NRC concluded that the PPE can serve as the foundation for the environmental report, and that the PPE values would be provided as a surrogate for the design information identified in the ESRP. NRC confirmed this in a letter to NEI in 2003, stating that ESP applicants may use the PPE approach as a surrogate for actual facility information to support required safety and environmental reviews. (NRC 2003).

Use of the PPE approach to streamline NEPA compliance is consistent with the Council on Environmental Qualitys (CEQs) Final Guidance for Effective Use of Programmatic NEPA Reviews, which states that Programmatic NEPA reviews assess the environmental impacts of proposed policies, plans, programs, or projects for which subsequent actions will be implemented either based on the PEA or PEIS, or based on subsequent NEPA reviews tiered to the programmatic review (e.g.,

a site- or project-specific document) (CEQ 2014). Per CEQs guidance, in the absence of certainty regarding the environmental consequences of future proposed actions, agencies may be able to make broad program decisions and establish parameters for subsequent analyses based on a programmatic review that adequately examines the reasonably foreseeable consequences of a proposed program, policy, plan, or suite of projects. The CEQ Programmatic NEPA guidance provides an example of a programmatic EIS where the location, type, and timing of specific facilities was unknown. Therefore, the programmatic EIS appropriately focused on a bounded range of potential activities and their impacts. The development of PPE parameters facilitates the review of impacts associated with development of future plants by allowing analysis of broad impacts of the surrogate plant; at the time of the project-specific review, these impacts can be compared to the impacts of the surrogate plant and supplemented as necessary.

2.2 Use of PPEs in Early Site Permit Applications A PPE is a set of reactor and owner-engineered parameters that are expected to bound the characteristics of a reactor that might later be deployed at the ESP site. A PPE sets forth postulated values of parameters that provide sufficient details to support the NRC staff's review of an ESP application (NEI 2012). Using the PPE approach, the applicant for an ESP need not provide a McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 4

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance detailed designbut should provide sufficient bounding parameters and characteristicsof a reactor or reactors and the associated facilities so that an assessment of site suitability can be made.

Consequently, the ESP application may refer to a PPE as a surrogate for a nuclear power plant and its associated facilities.

The ESP application and review process uses the PPE to evaluate and resolve safety and environmental issues related to siting before a specific reactor design is chosen, allowing the applicant to bank the site and rely on this analysis for up to 20 years of future reactor siting. Analysis of environmental impacts based on the PPE for a surrogate plant permits an ESP applicant to defer the selection of a specific reactor design until the construction permit or combined construction permit and operating license stage.

The NRC has issued EISs for six early site permits (ESPs) to date, five of which (all but the Vogtle ESP) used the PPE approach to define a surrogate reactor (NRC 2020b).

Site Applicant Clinton ESP Site Exelon Generation Company, LLC Grand Gulf ESP Site System Energy Resources Inc.

North Anna ESP Site Dominion Nuclear North Anna, LLC Vogtle ESP Site Southern Nuclear Operating Company PSEG Site PSEG Power, LLC, and PSEG Nuclear, LLC (PSEG)

Clinch River Nuclear Site Tennessee Valley Authority (TVA)

It has used the PPE concept most recently in its review of the ESP application for the Clinch River Nuclear Site in Tennessee (NRC 2019c). The Tennessee Valley Authority (TVA) proposed an ESP at the Clinch River site for two or more small modular reactors (SMRs) using a bounding PPE developed by TVA based on four different light-water SMRsthe BWX Technologies (BWXT) mPower', Holtec SMR-160, NuScale, and Westinghouse SMRs with a total installed capacity of 800 MWe. The environmental impact statement (EIS) for the Clinch River ESP was completed in April 2019 (NRC 2019a).

The NRC is currently applying the ESP process and associated PPE approach in the development of its ANR GEIS (NRC 2020a). The NRC published a Notice of Intent initiating development of the ANR GEIS on April 30, 2020 (85 FR 24040).

2.3 Adaptation of the PPE/SPE Approach to NRIC at INL 2.3.1 PPE Approach Both the NRC ESP process and the preparation for deployment of advanced reactor prototypes in existing buildings and sites at INL are intended to streamline NEPA reviews. Similar to the NRC ESP process, NRIC anticipates DOE will prepare a NEPA review of potential deployment of advanced reactor prototypes at INL prior to NRIC receiving commitments from any specific advanced reactor vendors. Although the buildings and sites at INL are known, the key features of advanced reactor McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 5

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance prototypes (from an environmental review perspective) may vary considerably across reactor sizes and designs. As described below, NRIC has adapted the NRCs ESP process as one reasonable approach to an early NEPA review.

The NRCs use of a PPE in the ESP process has generally been for ranges of potential reactors that are of relatively similar size and designs. For instance, the four reactors used to develop the Clinch River ESP were all pressurized water reactors with multiple modules ranging from 60 MWe (NRC 2020c) to 225 MWe (NRC 2019b). Because the reactors were of similar design, choosing bounding parameters from the individual reactor parameters for the Clinch River project and other previous projects was relatively straightforward.

Choosing bounding parameters for potential reactor prototype deployments at INL, on the other hand, is more complicated because of the wide range of reactor designs. For example, the potential reactor prototype deployments at INL could range from less than 1 MWe to over 500 MWe. In addition, coolants could include a wide variety of types, including liquid metal, high-temperature gas, and molten salt. Some reactors may be constructed in a factory and delivered to the site in modules.

Some may produce electricity, and some only heat. Different types of fuel are envisioned, from TRISO to fuel incorporated directly into a molten salt.

Because of the potential for a wide range of reactor designs, NRIC developed a modified approach to identifying the PPE for a surrogate plant. As was true for the NRC approach, the advanced reactor PPEs developed for NRIC are intended to identify the characteristics of the range of anticipated reactor designs that in turn provide a set of parameters associated with the construction and operation of a surrogate plant. Under NRICs approach, surrogate plants are defined by PPEs that are generally based on reasonable values for a wide range of anticipated designs as opposed to easily identifiable bounding values for generally similar designs. In some cases, e.g., the plant footprint, these parameters could be the largest parameter values of the potential reactors that could be deployed. In other cases, there may not be a bounding value because of the wide range of potential designs, as would be the case for the nature of the fuel, coolant, or cooling technology. In those cases, the range of potential parameter values would be presented, and a reasonable value would be chosen. Furthermore, some prototype reactor demonstrations may not be designed to produce electricity; therefore, the size of these demonstrations in this NRIC PPE is described in terms of thermal output, or megawatts-thermal (MWt).

The NEI approach relies on obtaining plant design information by surveying vendors that could be chosen by an applicant for reactor deployment (NEI 2012). Because of uncertainties about demonstration projects and the early stage of the advanced reactor designs that could be sited at INL, this information may not be readily available for all reactor designs. As a result, NRIC has also adapted NEIs approach by supplementing vendor surveys with additional data sources, including INL site documents, subject matter expert assessments, PPE estimates in the NRC ANR GEIS, and any regulatory requirements a vendor would need to meet during construction and operation. In addition, the Versatile Test Reactor design provided an additional data source for advanced reactor plant parameters (DOE 2020c). These sources are described in Section 3.0. The combination of these sources informed the development of a robust PPE for a surrogate plant that can be analyzed in future NEPA processes.

NRIC has further adapted the PPE approach previously used by the NRC in ESP reviews based on reactor size. Two PPEs are developed in this report for two size ranges: (1) microreactors, which are defined for this PPE as single units with outputs of 60 MWt or less, and (2) small- to medium-sized McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 6

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance advanced reactors with outputs above 60 MWt up to 1,000 MWt. These two PPEs are described in Section 4.0.

The PPE values using NRICs adapted approach can be used as inputs to a NEPA review that would evaluate the environmental impacts of the surrogate plant or plants, assuming that these impacts would be reasonable estimates of the impacts of any particular design chosen. During the NEPA review for a specific design application, the impacts of the proposed reactor would be compared to the impacts of the surrogate reactor; assuming that the impacts are bounded by the analysis of the surrogate reactor, no additional analysis would be needed. For specific parameters exceeding the values identified in the PPE, additional reactor and/or site-specific documentation or analysis would be necessary. The benefit of two PPEs (microreactors and small- to medium-sized advanced reactors) is that the environmental impacts of two very different types and sizes of reactors can be assessed and disclosed as part of DOEs NEPA review.

Operation of advanced reactors will generate spent fuel. The ultimate disposition of spent fuel has not been identified by DOE and is not included in the PPE. Spent fuel management at INL pending disposition would include monitoring and storage in accordance with applicable DOE and other legal requirements. However, the PPE does not include parameters for construction, modification and operation of spent fuel storage facilities.

2.3.2 SPE Approach Site parameters are usually specified by a reactor vendor independent of the proposed site, and they represent postulated physical, environmental, and demographic features of an assumed site that is used as a basis for the design analysis. In NEIs approach, site parameters are provided as part of an applicants standard design certification and allow the NRC to evaluate the safety and environmental impacts of the specific reactor design on a postulated or typical site (NEI 2012). The SPE is used to identify the characteristics of suitable sites that would allow for deployment of the surrogate advanced reactor plant, minimizing and mitigating adverse environmental impacts to the extent possible. The SPE characteristics follow from the PPE characteristics; that is, the amount of resources necessary for siting the reactor is proportional to the resource impacts associated with the reactor design itself.

The SPE is a valuable tool when a site has not yet been selected; once the site is known, the environmental needs and characteristics of that site can be specifically considered, and the PPE can be applied.

The NRCs ANR GEIS is also being developed with both a PPE approach and an SPE approach to streamline potential advanced reactor deployments without needing to identify specific sites until an application is received (NRC 2020a; 85 FR 24040).

Adapting the SPE approach to evaluating the impacts of prototype reactors at INL, however, has two challenges. First, because many of the anticipated reactor designs are in the early stages of development, limited information is available regarding the owner-engineered parameters. Second, the SPE is not as relevant in cases where the buildings and sites are known, and specific information is either available or being developed to characterize the environmental setting. At INL, the Evaluation of Sites for Advanced Reactor Demonstrations at Idaho National Laboratory published in March 2020 (INL 2020a) identified a list of candidate site locations and areas within INL boundaries that would be suitable for onsite demonstration of advanced reactors. INL identified, ranked, and recommended suitable sites based on this evaluation. Based on this analysis, INL indicated that advanced reactor demonstrations would be most suitable at the EBR-II, ZPPR, and Fuel Processing McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 7

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Restoration (FPR) facilities, as well as at four undeveloped areas and two previously developed areas within the INL boundary.

To address these challenges, NRIC has included the site parameters that typically represent the postulated physical, environmental, and demographic features relevant to an environmental review and that may have been considered as site parameters in other reviews.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 3.0 NRIC PPE Development Two PPEs are developed in this report for two size ranges: (1) microreactors, which are defined for this PPE as single units with outputs of 60 MWt or less, and (2) small- to medium-sized advanced reactors with outputs above 60 MWt up to 1,000 MWt. The reactor sizes described in this PPE are expressed in terms of megawatts-thermal because not all demonstration reactors would produce electricity. Within the PPEs, certain parameters may have more than one value, depending on the anticipated water use associated with the design. These PPEs were developed using the following sources of information as described in the ensuing sections and appendices:

x Advanced Reactor Vendor Questionnaire (Appendix A) x INL Supporting Information (Appendix B) x Subject Matter Expertise (Appendix C x NRC Advanced Reactor GEIS PPE values (Appendix D) x parameters from the Versatile Test Reactor Draft EIS x limits on construction and operation activities imposed by regulations.

PPE values without identified sources were developed using professional judgment and the bases for these decisions are provided. References and sources for the remaining values are also provided in this report and within the individual appendices. Appendix E (PPE Data Sources and Methodology) summarizes the information evaluated and the values chosen for the PPEs.

3.1 Advanced Reactor Vendor Questionnaire Responses NRIC issued a questionnaire in June 2020 seeking vendor input on designs, features, and plant and site requirements of advanced reactor technologies to facilitate analysis of potential future deployment and implementation of these reactors. The questions were adapted from the NEI 10-01 PPE template (NEI 2012) and focused on the plant and site parameters that are most relevant to the potential analysis of environmental impacts.

The questionnaire was sent to vendors that may have interest in deploying advanced reactors at INL.

The list of recipients was developed using sources such as Third Way (2020), the DOE (2019b),

DOEs Office of Advanced Reactor Technologies (2020a), DOEs Gateway for Accelerated Innovation in Nuclear Program (GAIN 2020), NRICs industry engagement efforts, and others. As of October 2020, NRIC received a total of 11 responses: 6 responses were for designs in the microreactor size range, and 5 responses were for small- to medium-sized advanced reactor designs.

While multiple questions were included in the questionnaire, it was recognized that many vendors would not have answers to all of the questions, either because the design has not been finalized, the characteristics of a site location have not been determined, or the question is inapplicable to their design. In those cases, instead of providing a bounding plant parameter, staff used any input received from the vendors to inform and ground-truth the PPE values developed using other sources.

3.2 INL Input As part of the capabilities provided by NRIC, INL may provide support services to vendors during construction, operation, and decommissioning. For example, INL provided information about the site infrastructure available for reactor deployment, e.g., types of transmission lines at proposed sites (INL 2020a). Because INL might prepare DOE-owned high-assay, low-enriched uranium (HALEU) as feed McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 9

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance stock for reactor fuel vendors, INL provided a report, the Isotopic Characterization of HALEU from EBR-II Driver Fuel Processing (INL 2020b; TEV3537, Rev. 1). INL staff provided information about these support services to inform the PPE because the full scope of services was not available to reactor vendors at the time of this report (see Appendix B).

3.3 Subject Matter Expert Input Resource subject matter experts (SMEs) at the Pacific Northwest National Laboratory (PNNL) involved in the development of this report previously conducted many of the resource assessments in the EISs for the six NRC ESPs, including all of the assessments for the recent Clinch River EIS (NRC 2019a). Knowledge of how a PPE is applied to specific sites in past reviews was used to recommend appropriate and reasonably bounding PPE values for key environmental resource issues that should be considered in DOEs NEPA review, particularly for the deployment of reactor technologies in the existing buildings and surrogate plants at the sites identified in Evaluation of Sites for Advanced Reactor Demonstrations at Idaho National Laboratory (INL 2020a).

PNNL SMEs provided independent assessments of key parameters for a surrogate plant in the following technical areas:

x land use x water demand x transportation x workforce x waste x other air emissions x fission product inventory.

The SMEs reviewed literature provided by INL, including the 2020 site evaluation report (INL 2020a),

publicly available information regarding advanced reactor technologies and development, and previous NRC ESP NEPA reviews, including the Clinch River EIS (NRC 2019a), to develop and inform parameter values and assumptions.

The approaches for developing PPE values differed by technical area. The approaches and assumptions are listed in Appendix C.

3.4 NRC Advanced Reactor GEIS Preliminary NRC PPE and SPE values presented during the ANR GEIS public scoping meeting are included in Appendix D. The PPE and SPE in the ANR GEIS were, in general, downscaled from impacts associated with previous environmental reviews for LWRs, rather than incorporating plant parameter values associated with specific advanced reactor designs. These values are subject to change during the development of the draft and final ANR GEIS.

While the NRCs preliminary PPE and SPE values should be considered when developing the NRIC PPE and SPE, the two processes differ in multiple ways, including that the intent of the NRCs PPE and SPE is to identify thresholds for Category 1 issues (for which a generic analysis of environmental impacts is possible, provided that relevant assumptions in the PPE and SPE are met) and Category 2 issues (for which a meaningful generic analysis of environmental impacts is not possible without consideration of site-specific information). Therefore, for Category 1 issues, the NRC McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 10

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance has developed the PPE and SPE to identify the largest possible resource values to be able to conclude that impacts are small. The NRIC PPE and SPE, on the other hand, are simply intended to identify the characteristics associated with various advanced reactor technologies and resource needs without consideration of the relative magnitude of the effects of any given design. The impacts resulting from deployment of a surrogate reactor at INL would be assessed by DOE in a future NEPA review.

3.5 Versatile Test Reactor Draft EIS Relevant values from the Versatile Test Reactor (VTR) Draft EIS published in December 2020 (DOE 2020c) are included in the Small- to Medium-Sized Advanced Reactor PPE summarized in Section 4.2 and described in Appendix E, Table E.2. The intent of the VTR is to provide a capability for large-scale testing, accelerated testing, and qualification of advanced nuclear fuels, materials, instrumentation, and sensors. The Draft EIS evaluates the environmental impacts of alternatives for constructing and operating the VTR, which would be an approximately 300 MWt sodium-cooled, pool-type, metal-fueled reactor constructed at INL. While not all of the values associated with the VTR are relevant to a PPE that focuses on advanced reactor demonstrations, these parameters were added and considered as data sources in certain cases to help define and refine the bounding values in the Small- to Medium-Sized Advanced Reactor PPE.

3.6 Regulatory Limits Preliminary design information for advanced reactors may not be available for every plant parameter that may result in environmental impacts. In the absence of specific design information or other sources, regulatory limits, safety goals, or thresholds were evaluated as a set of parameters that would not be exceeded by the surrogate plant during construction or operation. Accordingly, environmental impacts resulting from releases have been evaluated and addressed, assuming regulatory limits as bounding values. For instance, the Clean Air Act (42 U.S.C. § 7401 et seq.)

establishes limits for criteria pollutants, including, for example, those that could be released during operation of emergency diesel generators. Regulatory limits used in the PPE are specified in Appendix E.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 11

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 4.0 Summary PPEs Two separate PPEs, one for microreactors and one for small- to medium-sized advanced reactors, are included in the sections below.

4.1 Microreactor PPE Table 4.1 lists plant parameter values for a surrogate microreactor plant, which is defined as a single unit with an output of less than 60 MWt, plus any associated support facilities. The parameters listed are those that have a nexus to the environment and that are important for determining the impacts of prototype construction and operation. Note that the values listed in the PPE are not meant to imply that an actual reactor with these parameters could be constructed and operated. See Appendix E for the data sources and methodology used to develop this table.

Table 4.1. Microreactor PPE Plant Design Parameter Definition PPE Value 1.0 Structure and Layout Structure height Vertical height from finished grade to the top of 28 ft the tallest power-block structure.

Stack height Vertical height from finished grade to the top of 50 ft the tallest exhaust stack.

Foundation Depth from finished grade to the bottom of the 20 ft embedment basemat or the most deeply embedded power-block structure (excavation depth is the same elevation as embedment depth).

Permanent disturbed Land area required to provide space for plant 8 ac acreage to support facilities, including any support facilities, plant operations switchyards, spent fuel management, and cooling towers.

2.0 Construction Temporary Land area required to provide space for 18 ac disturbance during construction support facilities.

construction Duration of Duration of construction activities onsite. 24 mo construction Construction Maximum number of people onsite during 150 workforce construction.

Construction noise Maximum expected sound level due to 101 dB at 50 ft (a) construction activities, measured at 50 ft from the noise source 3.0 Operations Design type Design of the plant, primarily the cooling Five different design types:

system. high-temperature gas, molten salt, liquid metal, heat pipe, and nuclear battery.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 12

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.1. Microreactor PPE (continued)

Plant Design Parameter Definition PPE Value Megawatts-thermal Thermal power generated by one plant module 60 MWt Megawatts-electric Best estimate of maximum megawatt electric 20 MWe (based on generator output estimated 33% thermal efficiency)

Plant operational life Operational life for which the plant is designed 30 yr or anticipated to be operated at INL.

Planned modules Number of modules that would be installed and One operated.

Stationary or mobile Planned design future use of the plant following Stationary demonstration Offsite power Power from utility systems essential to support Required per General safety class structures, systems, and Design Criterion 17 (10 CFR components (SSCs), such as electrical power Part 50, Appendix A, supply and water supply Criterion 17 - Electric Power Systems)

Normal plant heat sink Technology (or technologies) for the normal Mechanical draft cooling plant heat sink towers Support facilities Support facilities such as switchyards, spent Multiple support facilities, fuel management, and cooling towers. including: cooling-water system; switchyard/

transformers; chemical/gas/fuel storage, potable water supply; wastewater system, including retention basins and associated discharge equipment; liquid radwaste system; fire protection and emergency response buildings; Administration/Maintenance Building(s); Security Facility; Chemistry and Meteorology Facility; Radioactive Waste Storage Facility (Region/Country Dependent); spent fuel management facilities, various offsite facilities Operations staff Number of total permanent staff to support 50 operations of the plant.

Refueling/major Additional number of temporary staff required 100 maintenance staff to conduct refueling and major maintenance activities.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 13

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.1. Microreactor PPE (continued)

Plant Design Parameter Definition PPE Value Consumption of raw Average short-term consumptive use of water 450 gpm, for air-cooled water by the cooling-water systems (evaporation and reactors, 25 gpm drift losses) and service water systems, including potable and sanitary water use (if required).

Water discharge Average characteristics of plant water 400 gpm, for air-cooled discharges. reactors, 25 gpm Water source Source of any potable water or cooling water. Groundwater(b)

Water discharge Chemical and radionuclide constituents of the See Appendix C.

constituents plant discharges, and maximum and expected concentrations/activities in the discharge.

Waste streams Volume of radioactive and nonradioactive See Appendix C. It includes wastes generated during routine plant bounding values of operations. radioactive solid waste generation, hazardous waste, nonhazardous waste, and gaseous waste.

Air emissions Routine and periodic releases of criteria See Appendix C.

pollutants and greenhouse gases.

Stack exit velocity Exit velocity of the stack for dispersion 10 ft/s calculations.

Auxiliary systems Fuel source and size of auxiliary boilers, Two diesel emergency power systems and standby power 50-150 kW standby power systems generators Operation noise Maximum expected sound level produced by 65 dBA operation of cooling towers, measured at 1,000 ft from the noise source.

4.0 Fuel Fuel form Form of fuel associated with the plant design. Fuel types could include UO2, MOX, Metal (U, U alloys, Pu-containing alloys),

TRISO, molten salt, uranium nitride, uranium carbide, QUADRISO, cermet, accident-tolerant fuel.

Impacts associated with fuel form would differ depending on type; therefore fuel type is not a bounding PPE parameter.

Annual fuel Annual average fuel requirement (metric tons) 0.5 MT requirements per module. (5 MT initial fuel loading)

Fuel source Source location of the fuel or fuel feedstock. Offsite commercial source McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 14

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.1. Microreactor PPE (continued)

Plant Design Parameter Definition PPE Value Radionuclide Radionuclide inventory for irradiated fuel at See Appendix C inventory time of shipment (curies/metric ton of uranium

[Ci/MTU] by radionuclide).

Refueling Refueling frequency (yr) and MTU per 5 MTU (full core refueling),

refueling. online and continuous refueling Fission product Annual activity, by radionuclide, contained in See Appendix C inventory routine plant airborne effluent streams, excluding tritium.

5.0 Transportation Shipments of Total number of shipments and MTU for 10 shipments over the 30 yr unirradiated fuel unirradiated fuel shipped to reactor or site. life of the plant.

45 MTU total Shipments of Total number of shipments and volume of 49 shipments over the 30 yr radioactive waste radioactive waste shipments from reactor/site. life of the plant. Volume of each shipment is 2.34 m3.

Transport method Method of transporting reactor, fresh fuel, and Truck other large components to the site.

Spent fuel disposition Ultimate disposition of spent fuel. 89 irradiated fuel shipments over the 30 yr life of the plant.

Offsite storage or disposal.

Treatment, storage, and disposal in accordance with applicable legal requirements.

6.0 Decommissioning Prototype removal Vendor plans to remove prototype from the INL Yes site following demonstration.

Workforce Estimated number of temporary staff required 150 to conduct decommissioning activities.

Duration Duration of decommissioning activities onsite. 18 mo Waste generation Amount of waste generated during Bounded by the waste decommissioning activities streams evaluated in NUREG-0586 (a) Parameter value not included in vendor questionnaire. Construction noise value derived from Clinch River EIS (NRC 2019c).

(b) Site parameter value not included in vendor questionnaire. Groundwater source is assumed to bound obtaining water from INL plant services. Also assumed a new surface water intake and connection would not be required.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 15

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 4.2 Small- to Medium-Sized Advanced Reactor PPE Table 4.2 lists plant parameter values for a surrogate small- to medium-sized advanced reactor, which is defined as a single unit with an output of 1,000 MWt or less, plus any associated support facilities. The parameters listed are those that have a nexus to the environment and that are important for determining the impacts of prototype construction and operation. Note that the values listed in the PPE are not meant to imply that an actual reactor with these parameters could be constructed and operated. See Appendix E for the data sources and methodology used to develop this table.

Table 4.2. Small- to Medium-Sized Advanced Reactor PPE Plant Design Parameter Definition PPE Value 1.0 Structure and Layout Structure height Vertical height from finished grade to the top 75 ft of the tallest power-block structure.

Stack height Vertical height from finished grade to the top 87 ft of the tallest exhaust stack Foundation Depth from finished grade to the bottom of the 155 ft embedment basemat or the most deeply embedded power-block structure (excavation depth is the same elevation as embedment depth).

Permanent disturbed Land area required to provide space for plant 50 ac acreage to support facilities, including any support facilities, plant operations switchyards, fuel management and cooling towers.

2.0 Construction Temporary Land area required to provide space for 100 ac disturbance during construction support facilities.

construction Duration of Duration of construction activities onsite. 54 mo construction McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 16

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.2. Small- to Medium-Sized Advanced Reactor PPE (continued)

Plant Design Parameter Definition PPE Value Construction Maximum number of people onsite during 1,400 workforce construction.

Construction noise Maximum expected sound level due to 101 dB at 50 ft(a) construction activities, measured at 50 ft from the noise source 3.0 Operations Design type Design of the plant, primarily the cooling High-temperature gas, molten system. salt, boiling water, liquid metal.

Megawatts-thermal Thermal power generated by one plant 1,000 MWt module Megawatts-electric Best estimate of maximum megawatt electric 333 MWe (based on generator output estimated 33% thermal efficiency)

Plant operational life Operational life for which the plant is designed 80 yr or anticipated to be operated at INL Planned modules Number of modules that would be installed One and operated Stationary or mobile Planned design future use of the plant Stationary following demonstration Offsite power Power from utility systems essential to support Two 230 kV transmission lines safety class structures, systems, and required. Offsite ROW 1,000 ft components (SSCs), such as electrical power x 100 ft (new) or within or supply and water supply adjacent to existing ROW Normal plant heat Technology (or technologies) for the normal Mechanical Draft Cooling sink plant heat sink Towers Support facilities Support facilities such as switchyards, fuel Multiple support facilities, management, and cooling towers including: cooling-water system; switchyard/

transformers; chemical/gas/fuel storage, potable water supply; wastewater system, including retention basins and associated discharge equipment; liquid radwaste system; fire protection and emergency response buildings; Administration/Maintenance Building(s); Security Facility; Chemistry and Meteorology Facility; Radioactive Waste Storage Facility (Region/Country Dependent);

various offsite facilities Operations staff Number of total permanent staff to support 207 operations of the plant.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 17

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.2. Small- to Medium-Sized Advanced Reactor PPE (continued)

Plant Design Parameter Definition PPE Value Refueling/major Additional number of temporary staff required 206 (413 total) (b) maintenance staff to conduct refueling and major maintenance activities.

Consumption of raw Average short-term consumptive use of water 5,850 gpm (water-cooled) water by the cooling-water systems (evaporation 415 gpm (air-cooled) and drift losses) and service water systems, including potable and sanitary water use (if required).

Water discharge Average characteristics of plant water 1,775 gpm (water-cooled) discharges. 415 gpm (air-cooled)

Water source Source of any potable water or cooling water. Groundwater (c)

Water discharge Chemical and radionuclide constituents of the See Appendix C.

constituents plant discharges, and maximum and expected concentrations/activities in the discharge.

Waste streams Volume of radioactive and nonradioactive See Appendix C. Appendix wastes generated during routine plant C.5 includes bounding values operations. of radioactive solid waste generation, hazardous waste, nonhazardous waste, and gaseous waste.

Air emissions Routine and periodic releases of criteria See Appendix C.

pollutants and greenhouse gases Stack exit velocity Exit velocity of the stack for dispersion 58 ft/s calculations Auxiliary systems Fuel source and size of auxiliary boilers, 50 MWt oil fired; 15 MWe emergency power systems, and standby Sentry turbine power systems Operation noise Maximum expected sound level produced by 65 dBA at site boundary operation of cooling towers, measured at 1,000 ft from the noise source 4.0 Fuel Fuel form Form of fuel associated with the plant design Molten salt, TRISO, uranium oxide, HALEU, U-Zr alloy.

Emission release mechanisms from molten salt are different from LWRs; expect that molten salt will have upper bounding impacts compared to other fuel technologies.

Annual fuel Annual average fuel requirement (metric tons) 8 MT requirements per module Fuel source Source location of the fuel or fuel feedstock Offsite commercial source McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 18

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table 4.2. Small- to Medium-Sized Advanced Reactor PPE (continued)

Plant Design Parameter Definition PPE Value Radionuclide Radionuclide inventory for irradiated fuel at See Appendix C inventory time of shipment (curies/metric ton of uranium

[Ci/MTU] by radionuclide)

Refueling Refueling frequency and MTU per refueling Daily refueling of 10.6 kg enriched U and 18 kg Th; annual requirement 3.9 MT enriched U, 6.6 MT Th.

Fission product Annual activity, by radionuclide, contained in See Appendix C inventory routine plant airborne effluent streams, excluding tritium 5.0 Transportation Shipments of Total number of shipments and MTU for 432 shipments over the 80 yr unirradiated fuel unirradiated fuel shipped to reactor or site life of the plant.

1,972 MTU total Shipments of Total number of shipments and volume of 2,160 shipments over the 80 radioactive waste radioactive waste shipments from reactor/site yr life of the plant. Volume of each shipment is 2.34 m3.

Transport method Method of transporting reactor, fresh fuel, and Truck other large components to the site Spent fuel disposition Ultimate disposition of spent fuel. 3,944 irradiated fuel shipments over 80 yr life of the plant.

Onsite storage, or offsite storage or disposal.

Treatment, storage, and disposal in accordance with applicable legal requirements.

6.0 Decommissioning Prototype removal Vendor plans to remove prototype from the Yes INL site following demonstration.

Workforce Estimated number of temporary staff required 450 total to conduct decommissioning activities Duration Duration of decommissioning activities onsite 10 yr Waste generation Amount of waste generated during Bounded by the waste decommissioning activities streams evaluated in NUREG-0586 (a) Parameter value not included in vendor questionnaire. Construction noise value derived from Clinch River EIS (NRC 2019c).

(b) Note that this parameter assumes that periodic refueling would occur, in order to better bound potential workforce impacts; however, the refueling parameter included below assumes continuous refueling.

(c) Site parameter value not included in vendor questionnaire. Groundwater source is assumed to bound obtaining water from INL plant services. Also assumed a new surface water intake and connection would not be required.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 19

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance 5.0 Guidance for Use of the PPE The use of the PPE to facilitate environmental impact assessments in future NEPA processes, and its limitations, are described below.

5.1 Use of the PPE in the NEPA Process The intent of the PPE is to describe a surrogate plant that would bound the upper limits of most potential demonstration reactor designs, or that would provide a representative resource value. As such, this surrogate plant can serve as the input for a NEPA review, allowing for analysis of the largest range of impacts anticipated by any given reactor prototype. The PPE does not define whether these impacts are small, moderate, or large, nor does it define a threshold of NEPA significance. The PPE provides DOE with a way to assess and disclose the reasonably foreseeable impacts from the deployment of micro and small- to medium-sized advanced reactor prototypes at INL.

Typically, NEPA reviews consist of evaluating a proposed actions impacts on the environment at a specific location. A future NEPA review for NRIC could assume the surrogate plant was sited in an existing building or at one of the specified sites at INL. The parameters developed are those that have the potential to impact the usual resource areas evaluated in a NEPA review (land use, hydrology, ecology, etc.). For instance, the plant footprint would be used to identify the maximum land that would be committed to the surrogate plant deployment and to assess the impact on changes in land use and the potential to disturb native habitat. By assessing the impact of the surrogate plant parameters on the site, the NEPA review informs the decision-maker and the public of the impacts that would be reasonably foreseeable.

During the NEPA review for deployment of a specific reactor design conducted after the initial NEPA review, the impacts of the proposed reactor would be compared to the impacts of the surrogate reactor. Assuming that the impacts are bounded by the analyses of the surrogate reactor, no additional analysis would be needed. In this situation, a Supplement Analysis may be the appropriate NEPA compliance mechanism for the project. For specific parameters exceeding the values identified in the PPE, additional design-specific or site-specific NEPA analysis would be necessary. This would likely be completed through the development of a supplemental environmental assessment or EIS, depending on the significance of the values exceeding the PPE.

To better inform stakeholders and the public about the potential impacts of the range of reactor designs and sizes expected to be deployed, it is recommended that the surrogate plants defined by the two PPEs presented in this report be assessed as separate deployments in the proposed action in any future NEPA review.

5.2 Limitations While the PPE contains useful and relevant information regarding advanced reactor technologies and deployment, the application of the PPE to a NEPA review has the following potential limitations:

1. Many parameters are not known by the vendors at their preliminary stages of design. As evidenced by the vendor responses to the PPE questionnaire, some vendors did not have information about the source of fuel, transportation of fresh fuel to and spent fuel from the site, decommissioning, and source term values. Because of the potential range of reactor sizes and McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 20

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance types and the early stage of plant designs for many vendors, it may not be possible to determine either a reasonable or bounding value for key plant parameters. For example, fission product inventories can be estimated, but routine releases and accident releases are design- and/or site-specific and must be considered as part of site-specific impact analysis. It was anticipated that many of these parameters would be unknown and therefore the information received from the vendors was supplemented with the other sources of information, as described in Section 3.0.

This information was used to create estimates of PPE values based on best professional judgment.

2. Some parameters are dependent on the type of reactor design, and therefore developing a single parameter value for one type of design would not bound values for different designs. For purposes of developing a value for the PPE, the waste streams associated with MSRs differ from LWRs and could result in additional operational waste streams to be analyzed. Using the anticipated releases from a MSRs as the PPE value when evaluating another reactor design (i.e., high-temperature gas, liquid metal, nuclear battery, and heat pipe) would not provide useful information for a NEPA review. However, MSRs are included in the Summary PPE in Section 4.1 because of the unique nature of the fuel and potential releases. This provides a basis for development of some of the other PPE values, including refueling, waste, and average/annual fuel requirements.

3. Bounding of solid, liquid, and gaseous waste would be dependent on compliance with applicable Federal regulations (such as 40 CFR Part 61 Subpart H; DOE O 458.1 [DOE 2020d], including DOE-STD-1196 [DOE 2011] and DOE-STD-1153 [DOE 2019a]; 10 CFR Part 20 Appendix B; Resource Conservation and Recovery Act of 1976 (RCRA); Clean Air Act; Clean Water Act),

State, local, and tribal regulations and permitting criteria. The best source of information for solid waste is developed from data about the transportation of shipments of radioactive waste as scaled from LWRs (Appendix C). However, the values for solid radioactive waste do not account for differences in the of advanced reactors and LWRs or any unique solid waste streams not found in a LWR. The best source of information for liquid radioactive effluent is developed from data about the constituents associated with the water discharges (Appendix C). Individual applications would have to provide additional information to address the generation and disposition of waste from the proposed action.

4. Some regulatory issues associated with advanced reactor deployment remain unresolved, including requirements associated with remote operation, emergency planning zones, and other issues differing from those associated with large LWRs. Resolution of these issues may affect the potential environmental impacts. For instance, the size of the emergency planning zone may affect the location of the nearest receptor.

5. The parameter values developed in the PPE are for deployment of a single reactor module. As part of vendor demonstrations, multiple modules may be deployed at the same time, or incrementally over time. The schedule for multiple module deployment will be needed to estimate the timing of impacts, and these schedules are not currently known. However, it is not necessarily the case that multiple deployments in a similar geographic area at the same or different times would result in additive impacts.

6. Any NEPA review based only on the PPE cannot fully cover deployment of advanced reactors at the INL site; supplemental NEPA analysis would be required to document that the impacts of any given proposed vendor demonstration would be within or similar to those discussed in the initial NEPA review. Some additional NEPA review could be necessary for parameters or actions that are not included in the PPEs or for deployment activities that are different than assumed. For example, if a demonstration project had design or operation features that did not fall within the McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 21

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance scope of the PPE considered in the initial NEPA review, then those features would be considered further and potentially evaluated as part of a supplemental NEPA analysis.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 22

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance INL (Idaho National Laboratory). 2017. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report. INL/EXT-16-40741, Revision 1, Idaho Falls, Idaho. Accessed July 24, 2020, at https://inldigitallibrary.inl.gov/sites/sti/sti/7365867.pdf.

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Accessed July 24, 2020, at https://www.nei.org/CorporateSite/media/filefolder/resources/reports-and-briefs/Report-Cost-Competitiveness-of-Micro-Reactors-for-Remote-Markets.pdf.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance NRC (U.S. Nuclear Regulatory Commission). 2002. Final Generic Environmental Impact Statement of Decommissioning of Nuclear Facilities: Regarding the Decommissioning of Nuclear Power Reactors.

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Lyons, Director, New Reactor Licensing Project Office, Office of Nuclear Reactor Regulation.

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NRC (U.S. Nuclear Regulatory Commission). 2019b. Evaluation of Emergency Preparedness Requirements for Small Modular Reactors: Clinch River Early Site Permit Application. Technical Session W14Innovation in New Light-Water-Reactor Reviews. Presented at the 31st Annual Regulatory Information Conference (RIC) Conference, March 12-14, 2019, Bethesda, Maryland.

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NRC (U.S. Nuclear Regulatory Commission). 2019c. Tennessee Valley Authority Clinch River Nuclear Site Early Site Permit ESP-006. Washington, D.C. Accessed July 24, 2020, at https://www.nrc.gov/docs/ML1935/ML19352D868.pdf.

NRC (U.S. Nuclear Regulatory Commission). 2020a. Advanced Nuclear Reactor Generic Environmental Impact Statement (GEIS). Washington, D.C. Accessed July 24, 2020, at https://www.nrc.gov/reactors/new-reactors/advanced.html#advRxGEIS.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Sowinski T. 2019. DOE-NE Micro-Reactor RD&D Program Mission and Objectives. Presented at the GAIN Micro-Reactor Workshop, June 18-19, 2019, Idaho National Laboratory, Idaho Falls, Idaho.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Appendix A - Vendor Questionnaire McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 A.1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 A.2

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 A.3

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 A.4

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 A.5

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Appendix B - INL Supporting Information Idaho National Laboratory (INL) staff provided the following documents in support of the plant parameter envelope development. This information was used to inform PPE values and SPE characteristics at the INL site.

x INL (Idaho National Laboratory). 2011. Idaho National Laboratory Comprehensive Land Use and Environmental Stewardship Report. INL/EXT-05-00726, Revision 1.

x INL (Idaho National Laboratory). 2013. Site Suitability and Hazard Assessment Guide for Small Modular Reactors. INL/EXT-13-29749.

x INL (Idaho National Laboratory). 2017. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report. INL/EXT-16-40741, Revision 1.

x INL (Idaho National Laboratory). 2019. Key Regulatory Issues in Nuclear Microreactor Transport and Siting. INL/EXT-19-55257.

x INL (Idaho National Laboratory). 2020a. Evaluation of Sites for Advanced Reactor Demonstrations at Idaho National Laboratory. INL/EXT-20-57821.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 B.1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Appendix C - SME Reactor Plant Parameter Value Assessments Appendix C describes the methodologies used by subject matter experts (SMEs) to estimate plant parameter envelope (PPE) values. These methodologies and assessments help define the PPE values associated with the surrogate plant.

Bounding plant parameters relevant to reactor design were determined via a review of publicly available documentation, including vendor websites and other literature, as well as vendor responses to the questionnaire presented in Appendix A. References reviewed beyond vendor websites include the following:

x Key Regulatory Issues in Nuclear Microreactor Transport and Siting (INL 2019) x Modernization of Technical Requirements for Licensing of Advanced Non-Light Water Reactors:

Westinghouse eVinci' Micro-Reactor Licensing Modernization Project Demonstration (Maioli et al. 2019) x Cost Competitiveness of Micro-Reactors for Remote Markets (NEI 2019a) x Micro-Reactor Regulatory Issues (NEI 2019b) x Roadmap for the Deployment of Micro-Reactors for U.S. Department of Defense Domestic Installations (NEI 2018) x Oklo Power Combined Operating License Application for the Aurora at INL (Oklo 2020c) x DOE-NE Micro-Reactor RD&D Program Mission and Objectives (Sowinski 2019) x Advanced Reactor Types Factsheet (DOE 2020b) x Project Pele Overview: Mobile Nuclear Power for Future DoD Need (Waksman 2020).

x Status Report - Steam Cycle High-Temperature Gas-Cooled Reactor (SC-HTGR) (Framatome 2019) x Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design (Cinotti et al. 2010) x Summary Description of the Fast Flux Test Facility (Cabell 1980).

Because many of the proposed reactor designs are still early in the development phase, many design details have not yet been determined. Based on the compiled information and comparison to existing plant designs for larger reactors (such as small modular reactors [SMRs], experimental reactors [Fast Flux Test Facility, Molten Salt Reactor Experiment, etc.], and LWRs), professional judgment was employed to determine bounding values. Where necessary, conservative assumptions were made.

C.1 Land Use Land use parameter values took into account input from the public scoping associated with the U.S.

Nuclear Regulatory Commission (NRC) Advanced Nuclear Reactor Generic Environmental Impact Statement (ANR GEIS), internal research and analysis of advanced reactor plant designs, and input received from the advanced reactor vendors in response to the questionnaire.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance For microreactors, among the vendor designs, the largest bounding value would be 8 ac of temporary disturbance and 7 ac of permanent disturbance. While the NRCs ANR GEIS is using a PPE value of 50 ac of temporary disturbance and 30 ac of permanent disturbance with a 50 ft maximum depth of excavation, the survey results and internal review found that most projects will use significantly less acreage with less excavation. Hence, the assumed land use requirements are slightly larger than the vendor information but smaller than the NRCs PPE values. Permanent disturbed acreage is assumed to be 8 ac or less, and temporary disturbed acreage is assumed to be 10 ac or less. Some reactors may require less acreage; for example, the total land use requirements associated with the Oklo Aurora reactor (Oklo 2020a) would be approximately 1 ac and would involve 0.3 ac of disturbed area.

For small- to medium-sized advanced reactors, the NRC ANR GEIS PPE values of 50 ac of temporary disturbance and 30 ac of permanent disturbance was instructive in developing the PPE values. However, the largest land use requirements provided in a vendor response were for 58 ac of temporary disturbance and 43 ac of permanent disturbance.

C.2 Water Demand Water demand for a power plant includes cooling-water use, non-cooling process water use, and potable/sanitary water use. For a wet-cooling system (e.g., a wet mechanical draft cooling tower),

total water demand will include the consumptive use for evaporative cooling and the nonconsumptive use for blowdown. For a dry-cooling system, water use for cooling will be minimal, but water will still be required for non-cooling process uses and for potable/sanitary uses.

For a power plant using a wet-cooling system, an estimate of the upper limit of consumptive water use for cooling can be made from the plants thermal power output and electrical power generated in power conversion, and the assumption that all waste heat is dissipated to the environment through the vaporization of water. The amount of waste heat is the difference between the plants thermal output and the work done (electricity generated). For example, a reactor that has an output of 60 MWt and a thermal efficiency of 37 percent would have an electrical power output of about 22 MWe. The cooling system in this case would need to dissipate 38 MW (60 MW - 22 MW).

The amount of heat required to evaporate a 1 kg mass of water is 2,256 kJ (i.e., the latent heat of vaporization of water is 2,256 kJ/kg). The mass rate of water evaporated by 1 MW (1,000 kJ/s) is therefore (1,000 kJ/s)/(2,256 kJ/kg) which equals 0.4433 kg/s. With continuous operation, this is about 38,300 L/d, which equates to 10,100 gal/d or 0.0157 ft3/s of water evaporated. For a plant that has an output of 60 MWt and 22 MWe, dissipating 38 MW using evaporative cooling would require the consumption of (38 x 38,300 L/d), or about 1.5 ML/d, which is approximately 0.38 Mgd or 270 gpm.

Cooling requirements of a particular plant would depend on the efficiency of the power conversion system, but the above analysis would provide a maximum estimate of consumptive cooling-water use for any plant using a wet-cooling system. The actual amount of cooling water consumed would depend on the cooling technology used and whether the heat from the plant was used in other industrial processes instead of being dissipated to the environment.

Blowdown for wet-cooling was estimated to be one-third of the consumptive use (evaporation), which is consistent with the assumption that the cooling system is operated at four cycles of concentration.

Non-cooling process water use was estimated as 7 percent of the total water demand, excluding the McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.2

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance potable/sanitary water use. This was based on the use of non-circulating water system water for the Clinch River plant. Potable and sanitary water use was estimated to be 100 gpd for each member of the operations workforce.

Appendix E presents bounding water use values using this approach for microreactors that have an output of 60 MWt or less and small- to medium-sized advanced reactors that have outputs of 1,000 MWt or less.

C.3 Transportation The number of shipments of unirradiated (fresh) fuel, irradiated (spent) fuel, and radioactive waste were estimated based on the shipment data presented in the Clinch River early site permit (ESP)

(NUREG 2226, NRC 2019a) for a surrogate SMR (see Tables 6-4, 6-10, and 6-14 in the Clinch ESP).

In the Clinch ESP, the surrogate SMR was developed by applying bounding parameters from four SMR designs (BWX Technologies mPower SMR, the Holtec SMR-160, the NuScale SMR, and the Westinghouse SMR). The bounding value for solid radioactive waste generation was 5,000 ft3/yr and a bounding activity of 57,200 Ci/yr (NUREG 2226; NRC 2019a). Overall, the generating output of the surrogate SMR was 800 MWe and the capacity factor was 90 percent. The shipment data for the four SMRs was normalized to a generating output of 1,100 MWe and a capacity factor of 80 percent for analysis in the Clinch River ESP.

The PPE for transportation was developed by further normalizing to the thermal output of the reactor, thermal to electrical energy conversion efficiency, capacity factor, and duration of operations for one microreactor and one small- to medium-sized advanced reactor:

x 60 MWt microreactor, 0.33 thermal to electrical energy conversion efficiency, 0.95 capacity factor, 30-year duration of operations x 1,000 MWt advanced non-LWR, 0.33 thermal to electrical energy conversion efficiency, 0.95 capacity factor, 80-year duration of operations.

Appendix E presents bounding values developed using this approach for the numbers of shipments associated with microreactors that have an output of 60 MWt or less and small- to medium-sized advanced reactors that have outputs of 1,000 MWt or less.

C.4 Workforce As indicated in Section 3.3, there is no recent nuclear power industry experience upon which to base PPE estimates for workforce parameters, including construction, operations, outages for refueling or maintenance activities, or any additional module installation into array-type facilities. Thus, to provide a basis for initial estimates of these workforce parameters, SMEs used recent electric power generation industry experience and adapted information relevant to nuclear-fueled facilities that might alter that experience. For example, construction cost and workforce estimates are generally based on electric power ratings (MWe) rather than the thermal ratings (MWt) of power plants. Using values scaled from the power capacity of a plant requires additional assumptions about the thermal efficiency of the plants being used as data sources. The values developed in this section are assumed to apply to microreactors in the range of 60 MWt and for small- to medium-sized advanced reactors bounded by the 1,000 MWt PPE value. For the purposes of developing the PPE values and to provide a conservative bounding estimate, thermal efficiency is estimated to be 33 percent.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.3

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Natural gas combustion turbine (CT) plant construction was judged to be most analogous to the expected early plant construction effort to install microreactors in the 60 MWt range. Estimates from PJM Interconnection, LLC (PJM 2018) for the costs of new CT plants were used to derive proxy values for the expected workforce parameters associated with a 20 MWe SMR. Smaller CT plants (under 100 MWe units) most closely approximate the expected construction effort needed to install a 20 MWe SMR and would likely provide bounding values.

Construction staffing PPE values are based on the PJM (2018) report and are derived using the cost per kilowatt reported for CT construction, assuming a greenfield site. The bounding estimate for the PJM service area for CT construction labor cost is $104/kW (PJM 2018). Scaling this value for 20 MWe yields an aggregate construction labor cost of $2.08 million. Average staffing during the CT construction period is approximately 65 workers. However, the effort is weighted toward the final 11 months (PJM 2018). The modular nature of the SMR technology and the relatively small expected facility footprint would greatly reduce the construction period required, compared to typical CT construction. By assuming a construction period of a maximum of 6 months and scaling PJM costs by the 20 MWe SMR costs above, the average construction workforce would be approximately 12 workers, but the effort would be expected to be weighted to the final 3 months of construction, with general site preparation work occurring over the initial 3 months. Thus, a reasonable approximation of the peak construction workforce would be 20 workers.

Operations staffing would be greatly minimized through deployment of automation and sensor technologies. Based on gas plant recent experience compiled by Black and Veatch (Wagman 2017),

a 565 MWe natural gas combined-cycle (NGCC) plant requires 27 full-time equivalent employees (FTEs) to operate on a typical 24-hour schedule. Increasing the plant size to 865 MWe requires only an additional six FTEs. Given the much larger capacity of the modern NGCC plant coupled with the use of nuclear-fueled technology in an SMR, it is assumed that 27 operations staff would be a bounding estimate for the 60 MWt first-of-a-kind implementation at Idaho National Laboratory (INL) of base-level operations. Presumably, the use of automation and sensor technologies would be maximized, but other safety- and security-related staffing would likely be required in addition to what might be expected at an NGCC plant.

To estimate the likely number of outage workers for the 60 MWt implementation at INL, SMEs reviewed recent experience at some plants in the existing fleet of large LWRs in geographically diverse areas of the United States. Specific cases include Byron 1 (Exelon 2020), Columbia Generating Station (Dobken and Markham 2017), Oconee 3 (Benson 2020), and Salem 1 (PSEG 2019). In these cases, the weighted average number of outage workers per megawatt-electrical unit of net capacity most recently equated to 1.04 workers/MWe. Scaling this experience to the 60 MWt size associated with the microreactor PPE would indicate that outage staffing would be approximately 21 workers. Given that the ANR technologies likely to be installed would have reduced maintenance requirements compared to large reactors, 21 workers could be considered a bounding estimate for outage staffing.

If an array-style design were implemented and capable of subsequently adding several 60 MWt modules, it is assumed that a workforce similar to the peak construction workforce would be needed for each module installation. Thus, 20 workers would be required to install a new module.

For the bounding case for a small- to medium-sized advanced reactor size of up to 1,000 MWt, workforce and related values are based on scaling down the values analyzed for the Clinch River ESP (NRC 2019a). Clinch River proposed a bounding value of 800 MWe (2,420 MWt) for a configuration of SMRs and represents the closest approximation available for NRIC consideration.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.4

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Assuming 33 percent thermal efficiency equates to approximately the 2,420 MWt used in the Clinch River case. Thus, workforce values reported for Clinch River were scaled using the factor:

1,000/2,420 = 0.413. This results in the maximum construction workforce onsite at one time of 909 workers at peak construction. Total construction staffing at this time would be 1,363 workers. The operations workforce would be 207 workers and the outage workforce would be 413 workers. It is assumed that the module replacement would occur concurrently with a typical refueling outage and use the same workforce of 413 workers.

Appendix E presents bounding workforce values for microreactors that have an output of 60 MWt or less and small- to medium-sized advanced reactors that have outputs of 1,000 MWt or less.

C.5 Waste C.5.1 Nonhazardous, Hazardous, and Radiological Waste Generation of nonhazardous, hazardous, and radiological waste materials would be expected at any nuclear reactor, regardless of the design. The Resource Conservation and Recovery Act (RCRA; 40 CFR Part 239-282) defines nonhazardous waste in Parts 239-259 and hazardous waste in Parts 260-273. DOE O 435.1 (DOE 2001) discusses radiological waste management.

Nonhazardous solid waste includes typical industrial waste such as metal, wood, and paper. A bounding estimate for nonhazardous waste could be 290 tons per month (NRC 2019a).

Examples of hazardous waste include lab packs, metals from shielding applications, and rags or wipes containing solvents. Universal waste is one class of hazardous waste that includes batteries, pesticides, equipment containing mercury, and lamps (bulbs) (40 CFR Part 273). RCRA also defines generator types for hazardous waste. This includes large quantity generators, small quantity generators, and very small quantity generators. Small quantity generators generate more than 100 kg, but less than 1,000 kg of hazardous waste per month. The ESP application for the Clinch River SMR expected the facility to qualify as a small quantity generator (TVA 2019). The ESP application for the Public Service Enterprise Group (PSEG) stated that PSEG maintains the program required of a small quantity generator (PSEG 2014). For nonradiological hazardous waste, it is assumed that hazardous waste amounts would remain below the criteria of a small quantity generator as defined by RCRA.

Examples of solid radioactive waste include low-level radioactive waste, such as radioactively contaminated protective clothing, tools, etc. The bounding value for solid radioactive waste generation was 5,000 ft3/yr and a bounding activity of 57,200 Ci/yr for a 1,100 MWe generating output (NRC 2019a). This value does not reflect differences in solid radioactive waste streams between LWR and advanced reactors designs and does not account for any waste streams unique to advanced reactors. Scaling of waste bounding values is discussed in Section C.3 for transportation and scaled values are presented in Appendix E.

C.5.2 Gaseous Waste Gaseous radioactive waste discharges will be controlled to the requirements of 10 CFR 20 and the ALARA principles of 10 CFR Part 50, Appendix I.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Typical gaseous radioactive wastes from large LWRs are released from vents on collection tanks and processing equipment and non-condensables in steam systems. The radioactive isotopes contained in these waste streams include fission products such as iodine, xenon and krypton, as well as activation products such as argon-41 and cobalt-60. The fission product inventories expected for advanced reactors are presented in Section C.7.3. Gaseous releases are typically collected and processed to decrease the radioactivity content to the point that they can be released to the environment through a controlled and monitored release point (plant vent or plant stack). The typical processing technique is one of holdup or delay to allow the short-lived activity to decay. Adsorption on activated charcoal or compression and storage are two methods used to create the necessary holdup time.

Specific systems for detecting minor leakage and ventilation systems for processing gases from plant systems will vary by prototype design. Microreactors would likely have no emissions of gaseous waste during routine operations. It is assumed that ventilation systems for small- to medium-sized advanced reactors will process gaseous releases by filtration, if needed, and direct the releases to a controlled and monitored release point. Part Gaseous radioactive waste discharges will be controlled to the requirements of Title 10 of the Code of Federal Regulations Part 20 (10 CFR Part 20) and the as low as reasonably achievable (ALARA) principles of 10 CFR Part 50, Appendix I. All emissions of gaseous radioactive effluents must comply with applicable regulatory limits established in 40 CFR Part 61 Subpart 61 Subpart H, and DOE O 458.1 (DOE 2020d), including DOE-STD-1196 (DOE 2011) and DOE-STD-1153 (DOE 2019a) (or 10 CFR Part 20 Appendix B depending on jurisdiction).

C.5.3 In-Vessel Solid Radioactive Waste Molten salt reactors are assumed in this work to generate the greatest quantity of in-vessel solid radioactive waste, consisting of both fuel and primary loop coolant. In several molten salt reactor designs, the fuel is intimately mixed with the coolant. It is conservative to assume that no waste treatment will occur and therefore the full inventory of fuel and coolant must be treated as in-vessel solid radioactive waste. Estimations of the fuel and coolant waste masses were made by scaling the Thorcon and Molten Salt Reactor Experiment (MSRE) reactors, respectively, from their nominal power ratings to the maximum microreactor power rating of 60 MWt and small- to medium-sized advanced reactor power rating of 1,000 MWt.

Each Thorcon molten salt reactor power module is designed to produce 250 MWe (ThorCon 2020).

According to data available on the company website, each module consumes 1,930 kg 19.7 percent enriched uranium and 3,290 kg thorium annually. This equates to 5.22 metric tons of combined uranium and thorium fuel waste per year; scaling this from 500 MWt to the maximum 60 MWt assumed for microreactors or 1,000 MWt for small- to medium-sized advanced reactors and assuming a 30- or 80-year demonstration period of operation, respectively, results in the conservative value of 19 and 836 MT fuel waste, respectively.

A similar scaling methodology was applied to calculate a bounding value for coolant waste mass.

MSRE, an 8 MWt molten salt reactor, had an initial coolant loading of 15,300 lb. (7 MT) (ORNL 2015).

Scaling this quantity to 60 MWt gives an initial coolant mass of 52.5 MT. Assuming the coolant would be replenished every 5 years and a 30-year demonstration period of operation, this results in 315 MT coolant waste. This waste could either take the form of high-level mixed waste or mixed transuranic waste depending upon whether there is any online fuel processing/waste management. A similar approach with an 80-year demonstration period was applied to obtain 13,920 MT of coolant waste for small- to medium-sized advanced reactors.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance C.5.4 Liquid Waste Radioactive waste system calculations are described in reactor final safety analysis reports for the design basis and realistic source terms for the primary and secondary coolants. The calculations are dependent on fuel design, steam generator design, reactor size, and other design factors. They are therefore specific to the reactor designs. All emissions of liquid radioactive effluents must comply with applicable regulatory limits established in DOE O 458.1 (DOE 2020d), including DOE-STD-1196 (DOE 2011) and DOE-STD-1153 (DOE 2019a) (or 10 CFR Part 20 Appendix B depending on the jurisdiction).

To develop reasonable estimates of waste volumes, SMEs reviewed the following documents:

x NuScale Design Control Document Final Safety Analysis Report (FSAR), Chapter 9, Auxiliary Systems (NuScale 2018a) x NuScale Design Control Document Final Safety Analysis Report (FSAR), Chapter 11, Radioactive Waste Management (NuScale 2018b) x Clinch River Nuclear Site Early Site Permit Application, Part 03Environmental Report (TVA 2019) x Final Safety Analysis Report (FSAR) - The Safety Case (Oklo 2020b) and Aurora Environmental ReportCombined License Stage (Oklo 2020a).

C.5.4.1 Liquid Waste Management System Description from NuScale (2018b)

The liquid radioactive waste management system (LRWS) is not safety related. NuScale inputs to LRWS include the following:

x primary coolant system letdown through chemical and volume control system and radwaste drain system (RWDS) equipment drains x RWDS floor drains, solid radioactive waste system decant water, and the reactor component cooling-water system (RCCWS) x detergent wastes from hand decontamination processes and personnel decontamination showers x chemical wastes collected by the RWDS.

Liquid radioactive waste streams are treated and monitored before being discharged to the utility water system (UWS) discharge basin. Annual activity discharged through liquid effluents is calculated and provided in FSAR Table 11.2-5 (Ci/yr by isotope) (NuScale 2018b).

The UWS receives nonradioactive wastewater from other sources (see the UWS description below) that dilutes the LRWS discharge in the UWS basin (Section 9.2.9 [NuScale 2018a]). A dilution factor of 5.34 cfs of the LRWS discharge was assumed in the calculation of discharge concentrations to meet the 10 CFR Part 20 Appendix B, Table 2 limits at the point of discharge. (Sum of discharge concentration fraction of limits was 0.243, 96 percent due to tritium.) An additional dilution factor of 270 cfs (e.g., a river) was used in LADTAP II to calculate unrestricted area doses to meet ALARA design objective dose limits in 10 CFR Part 50 Appendix I. Site-specific dilution flows will be needed for a combined license application or an operating license application.

The UWS (Figure C.1) provides the distribution of clarified water to the fire protection water tank, demineralized water system, potable water system, reactor building, control building, annex building, McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.7

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance radioactive waste building, turbine building, central utility building, and other plant users. The UWS also supplies raw water that has not been clarified to the two circulating water system cooling-tower basins and to the site cooling-water system cooling-tower basin for makeup water purposes. The water supplied by the UWS does not provide cooling functions. Raw water is the source of water for the UWS.

Figure C.1. Utility Water System Diagram C.5.4.2 Liquid Waste Management from Clinch River (TVA 2019)

The Clinch River Environmental Report (ER) Section 3.5.1 describes liquid radioactive waste system releases and provides a table of annual total (Table 3.5-1) and single-reactor (Table 3.5-2) releases (Ci/yr by isotope) (TVA 2019).

ER Section 3.6 describes liquid nonradioactive waste system releases, which may include cooling water, wastewaters from operating systems (e.g., demineralized water system), floor and equipment drain waters, stormwater runoff, and sanitary sewer effluents (TVA 2019). Wastewater may contain residual chemicals and biocides used to avoid scaling or fouling of plant systems. The specific chemicals/biocides used and their concentrations in the discharge would be plant specific. The Clinch River ER provided anticipated constituents and their concentrations in Table C.1 (TVA 2019). For a larger plant with evaporative water cooling, the blowdown from the cooling system is the dominant portion of the liquid nonradioactive waste stream. The concentration of constituents in blowdown will depend on how the cooling system is operated (e.g., cycles of concentration) and the quality of the cooling-water source. Sanitary effluents would be similar to those from the existing workforce McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.8

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance operations. For prototype reactors sited at INL, pretreatment of process and sanitary wastewaters would likely be required by existing discharge permits or would be similar to existing permit requirements if the discharges are authorized under a new permit.

Table C.1. Clinch River ESP Projected Blowdown Constituents and Concentrations Maximum Potential Constituent Concentration (ppm)(a)

Chlorine demand 1,000 Free available chlorine 0.5 Chromium -

Copper 6 Iron 3.5 Zinc 0.6 Phosphate 7.2 Sulfate 3,500 Oil and grease < 10 Total dissolved solids 17,000 Total suspended solids 150 Biological oxygen demand (BOD), 5-day <5 Calcium 260 Magnesium 85 Sodium 990 Manganese 0.1 Alkalinity as CaCO3 150 Nitrate (NO3) 52 Silicon dioxide (SiO2) 150 pH Range 7.5-8.5 (a) Assumed four cycles of concentration.

C.5.4.3 Liquid Waste Management from Oklo (2020a, 2020b)

Oklo FSAR Section 1.2.2.4.2.3 states that the Aurora does not use water for cooling, nor any other fluid that must be imported from offsite (Oklo 2020b). ER Section 2.4.2 describes the heat rejection from the power conversion system to the atmosphere via radiators. The heating, ventilation, and air conditioning system uses standard commercial heat exchangers without using any water (Oklo 2020a). ER Section 2.5 states that there are no permits required to ensure the Aurora INL site preparation and operation conforms with the Idaho Department of Environmental Quality regulations surrounding air and water quality. ER Section 3.1.1.3 states, the Aurora powerhouse does not use any water other than for incidental human use, which could be brought onsite and does not need be accessed from a local water line. According to the ER, there are no process water uses and no process water discharges. Potable and sanitary water use would be the only considerations in this case, which the ER states could be satisfied using temporary facilities (bottled water and portable toilets).

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Oklo FSAR Section 2.8.1 states that the power conversion system (PCS) is an off-the-shelf system, and that it can operate in turbine-bypass mode, in which heat is rejected to the environment (passively) (Oklo 2020b). Oklo FSAR Section 3.3.1.3 describes the PCS as using supercritical CO2 as the working fluid (Rankine cycle). Heat from the reactor is transferred to the PCS using heat pipes fully contained within the reactor module. Potassium is the heat pipe working fluid (solid until operation melts the potassium and eventually vaporizes it). (According to the FSAR Section 5.6.2.5, the ultimate heat sink [the system of structures and components and associated assured water supply and atmospheric condition(s) credited for functioning as a heat sink to absorb reactor residual heat and essential station heat loads after a normal reactor shutdown or a shutdown following an accident or transient including a loss-of-coolant accident (NRC 2015)] involves passive heat removal to air via natural convection in the reactor cavity surrounding the reactor module [and subsequently conduction and radiation from the reactor module and the building]).

C.5.4.4 Recommendations for NRIC PPE For a plant using evaporative cooling, liquid discharges will include cooling system blowdown. For most plants, discharge will include non-cooling process wastewater and potable/sanitary wastewater, with discharge rates assumed to be equivalent to the withdrawal rate (i.e., these water uses were assumed to be nonconsumptive).

Radionuclide releases in liquid wastes will be dependent on fuel design, steam generator design, reactor size, and other design factors. It is thus difficult to specify the constituents or inventory of radionuclides in the liquid radwaste system appropriate for use in the National Reactor Innovation Center (NRIC) PPE. However, it is safe to assume that any reactor that does have liquid radwaste will dilute that waste stream to meet the 10 CFR 20.1302 requirement to demonstrate that the annual average concentrations of radioactive material released in liquid effluents at the boundary of the unrestricted area (e.g., the point of discharge) do not exceed the values specified in 10 CFR Part 20 Appendix B, Table 2, Column 2. For a mixture of unidentified radionuclides, the lowest concentration value in Table 2 for any radionuclide not known to be absent from the effluent would be the applicable limit (Appendix B, Note 2). For a mixture of known radionuclides, the sum of the fractions (radionuclide effluent concentration divided by its limit) must be less than 1.0. For LWRs, the limiting radionuclide is typically tritium, but this may not be the case for non-LWR designs. For the NRIC PPE, an approach that would bound any liquid radwaste discharge would be to specify the PPE as the 10 CFR Part 20 Appendix B, Table 2, Column 2 effluent concentrations. The mixture limits would require that the actual discharge concentrations be less than these values.

Nonradioactive liquid waste would be dominated by the blowdown for a plant using evaporative cooling. For the NRIC PPE, it is assumed that nonradioactive liquid waste constituents and concentrations for a plant using an alternative cooling method (e.g., air-cooling) would be bounded by the blowdown from an evaporatively cooled plant. Blowdown concentrations will depend less on the size of the plant and more on the cycles of concentration at which the cooling system is operated; four cycles of concentration, which is not atypical for LWRs using surface water sources for cooling water, are assumed. Further, assuming that the additives to control scaling and biological growth used at Clinch River are typical, the NRIC PPE could use the Clinch River blowdown constituents and concentrations from the ER Table 3.6-1 (see above) (TVA 2019) for the nonradioactive liquid waste discharge. The specific mineralogical constituents of the blowdown at INL would vary from those listed in this table because of differences in the source waters. The Idaho Department of Environmental Quality (IDEQ 2018, p. 29) shows the presence of arsenic, barium, and chromium in INL groundwater; arsenic is present at a median concentration that would approach the drinking McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.10

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance water standard at four cycles of concentration and thus arsenic should be added to the PPE (at a concentration of 0.01 mg/L, which is the drinking water standard). The INL groundwater is likely to be higher in conductivity and alkalinity than the Clinch River water, which could require fewer cycles of concentration to be used at INL or the use of different anti-scaling chemicals. This would reduce discharge concentrations and could introduce other discharge constituents.

C.5.5 Decommissioning Waste At the end of the operating life of reactor prototypes, NRIC assumes the vendor and/or DOE would remove the facility from service, decommission the plant, and either remove or demolish plant structures and equipment. Decommissioning includes the reduction of residual radioactivity to a level that permits termination of an NRC license. It is assumed that a similar process would be required for termination of DOEs authorization of the prototype if an NRC license is not obtained. The regulations governing decommissioning of NRC-licensed power reactors are found in 10 CFR 50.75, 10 CFR 50.82, and 10 CFR 52.110. The radiological criteria for termination of the NRC license are in 10 CFR Part 20, Subpart E. Minimization of contamination and generation of radioactive waste requirements for facility design and procedures for operation are addressed in 10 CFR 20.1406. Additionally, 10 CFR 50.82 or 10 CFR 52.110, as applicable, provide that a NRC licensee shall not perform any decommissioning activities that result in significant environmental impacts not bounded by previously issued environmental review documents, such as NUREG-0586 Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (NRC 2002).

Radioactive gaseous, liquid and solid wastes, nonradioactive waste and mixed waste could be generated during advanced reactor plant decommissioning, depending on the scope of the activities and the size and condition of the plant upon decommissioning. Decommissioning activities associated with small- to medium-sized advanced reactors would be similar in scope to activities required to decommission a LWR. Decommissioning activities for a modular reactor constructed in a factory and delivered to a site would be significantly less. Some smaller reactor prototypes may be removed in total and returned to the manufacturer.

The NRC evaluated the environmental impacts of decommissioning in NUREG-0586 Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (NRC 2002). The anticipated volumes of wastes evaluated in NUREG-0586 were based on industry decommissioning experience as of 2002. Appendix G of NUREG-0586, Radiation Protection Considerations for Nuclear Power Facility Decommissioning summarizes effluent releases for operating facilities and decommissioning facilities. Appendix G includes occupational dose from reactors in the 60 MWe range for reference. Low-level waste volume estimates for decommissioning facilities are presented in Appendix K of NUREG-0586. The reactors considered in NUREG-0586 included non-LWR designs including fast breeder reactors such as FERMI Unit 1, and high-temperature gas-cooled reactors such as Peach Bottom Unit 1 and Fort St. Vrain.

The NRCs EIS for the Clinch River ESP (NRC 2019a) compared the impacts of decommissioning two or more SMRs with a combined capacity of 800 MWe to the impacts of decommissioning large LWRs described in NUREG-0586 (NRC 2002). The NRC found the following:

1. The quantities of Class C or greater than Class C wastes generated would be comparable to or less than the amounts of solid waste generated by reactors licensed before 2002.

2. The air-quality impacts of decommissioning are expected to be negligible at the end of the operating term.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance

3. Measures are readily available to avoid potential significant water-quality impacts from erosion or spills. The liquid radioactive waste system design includes features to limit release of radioactive material to the environment, such as pipe chases and tank collection basins. These features would minimize the amount of radioactive material in spills and leakage that would have to be addressed at decommissioning.

Microreactors will have significantly smaller reactor sizes, less infrastructure, and fewer balance-of-plant facilities than the LWRs described in NUREG-0586 (NRC 2002) and the SMRs described the Clinch River EIS (NRC 2019a). The plant facilities and infrastructure required for small- to medium-sized advanced reactors would be closer to, but smaller than those evaluated in the Clinch River EIS, based on the smaller bounding output of the proposed advanced reactors (1,000 MWt). Therefore, the amounts of solid waste and air emissions associated with decommissioning microreactors and small- to medium-sized advanced reactors would be bounded by the waste streams evaluated in NUREG-0586.

Demolition of plant structures and components that are not activated or otherwise contaminated would result in nonhazardous debris that could be disposed in commercial landfills.

C.6 Other Air Emissions Criteria pollutants (nitrogen oxides [NOx], carbon monoxide [CO], sulfur oxides [SOx], hydrocarbons in the form of volatile organic compounds, and particulate matter) may be emitted from a prototype plant during construction activities, as part of operations including periodic or routine operation of balance-of-plant equipment, such as emergency power, evaporator heating, plant space heating, and/or feed water purification, and also from vehicles. Emissions may occur during operation of auxiliary boilers for heating and startup and emergency power supply system diesel generators and/or gas turbines.

Because a specific ANR technology and supporting equipment have not been selected, detailed emission data are not available.

When a specific prototype is selected for deployment, modeling will be conducted, as required, to demonstrate that the project emissions will not result in exceedances of the National Ambient Air Quality Standards (NAAQSs). INL is not in a maintenance area, a nonattainment area, or a tribal nonattainment area administered by the U.S. Environmental Protection Agency (EPA 2020). Because INL is located in an attainment area for all NAAQS criteria pollutants, the proposed project is not subject to a Nonattainment New Source Review.

Regulatory limits were proposed by the NRC in the ANR GEIS public scoping effort as bounding parameters for air quality for advanced reactors, depending on the location. For reference, the NRC determined that emissions of criteria pollutants from one or more SMRs at the Clinch River site with a total installed capacity of up to 800 MWe would not have a noticeable impact on air quality (NRC 2019a). It is assumed that any new advanced reactor prototype constructed and operated at the INL site, up to and including a 1,000 MWt advanced reactor, would also comply with all regulatory requirements of the Clean Air Act, and therefore the de minimis levels of criteria pollutants would bound the potential emissions from any future reactor deployment. All emissions of gaseous radioactive effluents must comply with applicable regulatory limits established in 40 CFR Part 61 Subpart H and DOE O 458.1 (DOE 2020d), including DOE-STD-1196 (DOE 2011) and DOE-STD-1153 (DOE 2019a) (or 10 CFR Part 20 Appendix B, depending on the jurisdiction).

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance The NRC has evaluated greenhouse gas (GHG) emissions in previous National Environmental Policy Act (NEPA) reviews (for example, see NRC 2019a, Appendix K) for a 1,000 MWe reference plant.

The emissions assumed for the reference plant included emissions during all phases of construction, operation, and decommissioning of the plant and included emissions from the uranium fuel cycle such as from mining and milling uranium ore. It is assumed that the GHG emissions from a microreactor or a small- to medium-sized advanced reactor, including any fuel cycle emissions, would be no greater than the emissions from the reference reactor when downscaled based on reactor size. In particular, GHG emissions from a 60 MWt reactor would be no greater than 2 percent of the emissions from the 1,000 MWe reference reactor and GHG emissions from a small- to medium-sized advanced reactor with output 1,000 MWt or less would be no greater than 33 percent of the emissions of the 1,000 MWe reference reactor.

Based on NRCs approach, Table C.2 presents estimated GHG emissions for a 60 MWt prototype and a 1,000 MWt prototype to be constructed at INL, operated and decommissioned, including emissions from the fuel cycle. Emissions for a 60 MWt plant assume 30 years of operation. Emissions for a 1,000 MWt plant assume 80 years of operation.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table C.2. Nuclear Power Plant Lifetime Greenhouse Gas Footprints(a)

Total Emissions Advanced NRCs Total Emissions Reactors Activity Duration 1,000 MWe 60 MWt 1,000 MWt Source (year) Reference Plant(b) Plant(c) Plant(d)

Preconstruction/Construction 7 39,000 780 12,870 Equipment Preconstruction/Construction 7 43,000 860 14,190 Workforce Plant Operations 40 181,000 2,715 119,460 Operations Workforce 40 136,000 2,040 89,760 Uranium Fuel Cycle 40 10,100,000 151,500 6,666,000 Decommissioning Equipment 10 19,000 380 6,270 Decommissioning Workforce 10 8,000 160 2,640 SAFSTOR Workforce 40 10,000 200 3,300 TOTAL 10,536,000 158,635 6,914,490 (a) Derived from NRC 2019a, Appendix K.

(b) Emissions are rounded to the nearest 1,000 MT CO2e.

(c) Estimates for a 60 MWt plant are calculated as 2 percent of the reference plant estimates assuming 33 percent thermal efficiency and are based upon a 30-year plant operation cycle instead of NRCs assumed 40-year operation cycle.

(d) Emissions from a 1,000 MWt plant are calculated as 33 percent of the reference plant estimates assuming 33 percent thermal efficiency and are based upon a 80-year plant operation cycle instead of NRCs assumed 40-year operation cycle.

C.7 Fission Product Inventory Each company that proposes a small- to medium-sized advanced reactor concept does so with a distinct and usually proprietary design. Fission product inventories were calculated based on generalized designs, described in further detail below. Once the generalized designs were selected, depletion models were run using TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion) in Standardized Computer Analyses for Licensing Evaluation (SCALE) to create library files for later use. The model for each category of reactor was the simplest unit-cell representation possible. For example, the LeadCold reactor only used one fuel assembly in two-dimensional geometry. Using a full three-dimensional design would provide the most precise results for the specific reactor, but that level of precision is not required for calculation of bounding core inventories.

TRITON runs a defined number of depletion cases based on the specified power level (MW/MTHM).

Power levels and maximum burnup used for the depletion steps were representative of the reactors.

For example, LeadCold has an average power of 3.3 MW/MTHM with a maximum center assembly burnup of 58 GWD/MTHM. The number of depletion steps for all models was balanced to achieve both a fast runtime and enough resolution to be confident in the results. Each model (except the molten salt reactor designs) was run at several different uranium enrichments to provide more flexibility in subsequent analyses. The molten salt reactor models were run with and without thorium fuel.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Radionuclide inventories were then calculated using ORIGEN (Oak Ridge Isotope Generation Code) from the TRITON libraries. All solid-fuel reactors were run at 10 percent, 15 percent, and 19.95 percent enrichment, 25 library positions, and depletion to maximum burnup from the TRITON calculations. The input material was only UO2 with 1 MT of uranium. The input specific power was 20 MW/MTHM. The liquid-fuel reactors were run at a single library position that was closest to a k-effective of 1 from the TRITON runs. Input material was the respective fuel salt (1 MT heavy metal, split between uranium and thorium). If the run included thorium, a constant feed of 232Th was added into the system at a rate of 10 kg/yr. The molten salt reactors were depleted for 20 years at 20 MW/MTHM.

During actual operation, a molten salt reactor is likely to have continuous feed of fuel and removal of fission products and neutron poisons. Therefore, the molten salt reactors were depleted for a much longer period than any of the other reactors. However, the ORIGEN model only had a feed of 232Th in the thorium reactors, and no model incorporated the removal of fission products or neutron poisons; it is assumed that waste will remain onsite until decommissioning and is therefore available for release in an accident scenario.

After the depletion runs, ORIGEN performed decay-only calculations on the models out to 20 years, with small initial timesteps to account for short-lived isotopes. A decay time of 1 year was used for radionuclide inventory comparison.

C.7.1 Microreactors Each company that proposes a microreactor concept does so with a distinct and usually proprietary design, and thus, establishing core models for each design is inefficient. Therefore, each reactor that had some public information available was categorized by general reactor type. The analysis then uses these general cases, instead of specific designs, to calculate the radionuclide inventory. Each design, detailed below, was based on either an existing design or a simplified representation.

x Helium-cooled prismatic reactor, with tri-structural isotropic (TRISO) fuel. This design was based on the High-Temperature Engineering Test Reactor (HTTR) (IAEA 2004; Bess et al. 2020).

x Helium-cooled pebble-bed reactor, with TRISO fuel. This design was based on the HTR-10 (IAEA 2004).

x Lead-cooled fast reactor, with UO2 fuel. This design was based on the LeadCold reactor (Wallenius et al. 2018).

x Heat-pipe fast reactor, with UO2 fuel. This design was based on the Special Purpose Reactor (INL 2017; Hernandez et al. 2018).

x Molten salt thermal reactor, with liquid fuel. This design was based on the ThorCon reactor (Fei et al. 2019; EPRI 2015).

x Molten salt fast reactor, with liquid fuel. This was a simplified design that separated out the fuel salt and the cooling salt (Betzler et al. 2016).

Some of these designs, while not specifically microreactors, were taken as a scaled representation to derive neutronic properties similar to those of an anticipated microreactor design of the same category. The LeadCold reactor is the only microreactor design that is available to the public.

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Advanced Nuclear Reactor Plant Parameter Envelope and Guidance C.7.2 Small- to Medium-Sized Advanced Reactors Trying to establish models for each design can be difficult if not enough public information is available. However, reactors can be grouped based on input parameters so that the neutron spectra and self-shielding properties are similar. Therefore, one type of reactor, with the most publicly available information, can be used as an input library for ORIGEN. The different reactor types with public information used for this work are summarized below:

x Sodium fast reactor (SFR). This design is based on the General Electric (GE) Power Reactor Innovative Small Module (PRISM) (Triplett et al. 2010).

x Lead-cooled fast reactor (LFR). This design is based on the LeadCold reactor (Wallenius et al.

2018), but it was run at a higher specific power (MW/MTIHM) and out to a higher burnup.

x Boiling-water reactor (BWR). This design is based on the GE 10x10 BWR fuel distributed with SCALE.

x Pressurized water reactor (PWR). This design is based on the Westinghouse 17x17 PWR fuel distributed with SCALE (Wieselquist et al. 2020).

x Fluorine-lithium-beryllium (FLIBE) cooled pebble bed reactor (PBR). This design is based on the PB-FHR (Andreades et al. 2014).

x Helium-cooled PBR with TRISO fuel. This design is based on the HTR-10 (IAEA 2004).

x Helium-cooled prismatic reactor with TRISO fuel. This design is based on the HTTR (IAEA 2004).

x Molten salt thermal reactor with liquid fuel. This design is based on the Thorcon Reactor (Fei et al.

2019; EPRI 2015).

x Molten salt fast reactor with liquid fuel. This design is a simplified design that separated out the fuel salt and the cooling salt (Betzler et al. 2016).

With these groups, specific reactors can then be modeled with ORIGEN as long as they have enough publicly available information to provide accurate input parameters. However, some of the reactor designs above were also run as generic but possible configurations to account for possible small- to medium-sized advanced reactors using that specific technology.

C.7.3 Fission Product Inventory Comparison Comparison of reactor core inventories based on reactor design was limited to comparison of activity for each radionuclide in the inventory. Without understanding the mechanisms or chemical form of release it must be assumed that a release of a specific radionuclide will be equivalent to the release of that same radionuclide from another reactor design (i.e., same mode of release, chemical composition, particle size, etc.). Assuming that individual radionuclides are released in comparable ways across reactor designs, radionuclide inventories can then be compared side-by-side based on activity.

This analysis can only be used to compare designs for individual radionuclide activities. This analysis does not address which combination of radionuclides would lead to a higher dose consequence to people. Dose conversion factors for radionuclides vary from radionuclide to radionuclide and trying to determine the combination of radionuclides that will lead to a higher dose is a very involved, complicated problem. This question should only be addressed when the source term release quantity, mechanism, and chemical form are known.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.16

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Fission product core inventories were compared across the different reactor types described above.

The summary data set includes the reactor source term inventories organized alphabetically by radionuclide for all reactor designs included at Time 0 (end of operation) and Time 1 (1 year after end of operations). The molten salt reactors exhibited the highest activities for most radionuclides and had the highest total activities. From the summary data, a minimum, maximum, and average radionuclide activity can be deduced. However, not all radionuclides are mobile and available for release; Table C.3 and Table C.4 present a limited look at the radionuclides for microreactors that have potential mobility at Time 0 and Time 1. Calculations are based on observations of the Chernobyl and Fukushima accidents. The radionuclides compared here include noble gases as well as radionuclides known to be volatile or intermediately volatile as defined by Steinhauser et al. (2014). For reference, Table C.5 presents summary data for all radionuclides evaluated for microreactors. Similar information for small- to medium-sized advanced reactors is presented in Table C.6 (radionuclides with potential mobility at time 0) and Table C.7 (all radionuclides at time 0).

Table C.3. Minimum, Maximum, and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 0 (End of Operation) for Microreactors Minimum Average Percent Activity Maximum Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Kr79 3.44E+03 7.77E+04 2.51E+04 183.1 Kr81 2.40E+02 2.30E+06 3.28E+05 200.0 Kr81m 3.13E+01 9.52E+02 3.24E+02 187.3 Kr83m 9.55E+12 2.41E+13 1.45E+13 86.5 Kr85 1.95E+14 2.36E+15 8.07E+14 169.5 Kr85m 1.07E+14 2.93E+14 1.89E+14 93.3 Kr87 1.59E+10 4.45E+10 3.06E+10 94.6 Kr88 2.78E+13 7.85E+13 5.62E+13 95.3 Rn217 1.39E+06 2.20E+06 1.79E+06 44.9 Noble Gases Rn218 1.40E+03 1.27E+09 1.10E+08 200.0 Rn219 9.62E+03 2.60E+09 1.97E+08 200.0 Rn220 3.61E+06 8.73E+12 9.50E+11 200.0 Rn222 1.24E+03 4.28E+06 5.19E+05 199.9 Xe125 1.58E+03 6.63E+04 2.43E+04 190.7 Xe127 3.05E+06 2.53E+10 3.20E+09 200.0 Xe129m 5.86E+09 2.17E+13 3.13E+12 199.9 Xe131m 2.12E+14 3.83E+14 2.52E+14 57.7 Xe133 4.00E+16 5.72E+16 4.47E+16 35.4 Xe133m 1.08E+15 1.80E+15 1.28E+15 49.6 Xe135 1.20E+16 2.09E+16 1.47E+16 53.9 Xe135m 5.59E+14 7.74E+14 6.08E+14 32.3 Xe138 5.94E-15 7.90E-15 6.74E-15 28.4 Cs131 7.54E+05 7.18E+07 1.39E+07 195.8 Volatile Cs132 7.67E+09 8.57E+11 2.59E+11 196.5 Cs134 1.53E+14 4.04E+16 9.92E+15 198.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.17

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Average Percent Activity Maximum Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Cs134m 1.85E+11 1.70E+13 4.58E+12 195.7 Cs135 2.49E+10 3.11E+11 9.45E+10 170.4 Cs135m 1.90E+04 8.77E+05 1.76E+05 191.5 Cs136 2.90E+14 1.99E+16 3.37E+15 194.3 Cs137 1.74E+15 2.21E+16 9.11E+15 170.9 Cs138 7.07E+03 9.45E+03 7.98E+03 28.8 H3 7.23E+12 3.89E+16 3.49E+15 199.9 I123 1.01E+04 2.03E+05 6.42E+04 181.0 I125 3.37E+02 1.15E+06 2.15E+05 199.9 I126 2.26E+08 4.12E+10 1.20E+10 197.8 I128 1.06E-04 1.26E-02 2.24E-03 196.7 I129 3.62E+08 9.18E+09 2.57E+09 184.8 I130 1.29E+13 1.56E+15 2.77E+14 196.7 I131 1.82E+16 2.97E+16 2.12E+16 48.0 I132 2.39E+16 3.55E+16 2.69E+16 39.3 I132m 5.37E+08 5.13E+09 2.11E+09 162.1 I133 1.98E+16 2.72E+16 2.18E+16 31.3 I133m 3.18E+07 4.35E+07 3.65E+07 31.1 I134 1.09E+09 1.39E+09 1.22E+09 23.9 I135 3.25E+15 4.51E+15 3.54E+15 32.3 Te121 3.02E+05 8.02E+08 9.18E+07 199.8 Te121m 1.52E+05 3.96E+08 4.52E+07 199.8 Te123m 3.05E+09 1.02E+13 1.36E+12 199.9 Te125m 2.62E+13 2.12E+14 8.65E+13 156.0 Te127 1.12E+15 3.67E+15 1.80E+15 106.2 Te127m 1.43E+14 5.91E+14 2.46E+14 122.1 Te129 5.25E+14 1.54E+15 8.00E+14 98.2 Te129m 6.62E+14 1.95E+15 1.01E+15 98.5 Te131 4.40E+14 1.37E+15 6.93E+14 102.8 Te131m 1.68E+15 5.22E+15 2.64E+15 102.8 Te132 2.32E+16 3.45E+16 2.61E+16 39.3 Te133 6.77E+07 9.26E+07 7.77E+07 31.1 Te133m 3.18E+08 4.35E+08 3.65E+08 31.1 Te134 1.64E+06 2.04E+06 1.84E+06 21.8 Ru103 2.20E+16 5.94E+16 3.11E+16 91.9 Ru105 2.27E+14 1.19E+15 4.68E+14 136.0 Semi-Volatile Ru106 3.38E+15 4.07E+16 1.25E+16 169.3 Sr83 1.32E+02 1.16E+03 5.79E+02 159.1 Sr85 6.47E+06 8.00E+08 1.52E+08 196.8 Sr85m 1.40E+00 9.47E+01 1.89E+01 194.2 Sr87m 2.68E+07 5.13E+10 7.76E+09 199.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.18

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Average Percent Activity Maximum Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sr89 1.29E+16 3.42E+16 2.58E+16 90.4 Sr90 1.51E+15 1.72E+16 6.82E+15 167.8 Sr91 3.36E+15 6.92E+15 5.78E+15 69.3 Sr92 4.90E+13 8.75E+13 7.35E+13 56.4 Ba131 1.87E+03 4.81E+10 5.01E+09 200.0 Ba133 4.45E+06 1.78E+09 5.91E+08 199.0 Ba135m 1.67E+10 1.74E+14 2.72E+13 200.0 Ba136m 3.21E+13 2.20E+15 3.73E+14 194.3 Ba137m 1.65E+15 2.10E+16 8.62E+15 170.9 Ba139 2.60E+11 3.69E+11 2.95E+11 34.8 Ba140 3.48E+16 5.03E+16 3.96E+16 36.3 Ba141 6.67E-08 9.65E-08 7.53E-08 36.5 Total 9.05E+17 2.18E+18 1.30E+18 82.7 Table C.4. Minimum, Maximum, and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 1 (One Year After End of Operation) for Microreactors Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Kr81 2.40E+02 2.30E+06 3.28E+05 200.0 Kr83m 3.87E+05 1.08E+07 3.85E+06 186.1 Kr85 1.83E+14 2.21E+15 7.57E+14 169.5 Rn217 1.43E+05 4.26E+05 2.85E+05 99.3 Rn218 4.02E-02 5.00E+03 6.84E+02 200.0 Rn219 2.27E+04 3.34E+09 2.69E+08 200.0 Rn220 1.25E+07 1.24E+13 1.25E+12 200.0 Rn222 2.86E+03 4.63E+06 5.63E+05 199.8 Noble Gases Xe127 2.98E+03 2.47E+07 3.12E+06 200.0 Xe129m 2.68E-03 9.92E+00 1.43E+00 199.9 Xe131m 3.53E+05 6.02E+05 4.16E+05 52.2 Xe133 5.65E-05 8.06E-05 6.30E-05 35.1 Cs131 7.16E-06 4.37E-04 9.14E-05 193.6 Cs132 9.44E-08 1.06E-05 3.19E-06 196.5 Cs134 1.10E+14 2.89E+16 7.10E+15 198.5 Cs135 2.49E+10 3.11E+11 9.45E+10 170.4 Volatile Cs136 1.37E+06 9.38E+07 1.59E+07 194.3 Cs137 1.70E+15 2.16E+16 8.90E+15 170.9 H3 6.84E+12 3.68E+16 3.29E+15 199.9 I125 4.82E+00 1.64E+04 3.08E+03 199.9 I126 7.58E-01 1.38E+02 4.04E+01 197.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.19

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

I129 3.67E+08 9.19E+09 2.58E+09 184.7 I131 4.12E+02 6.82E+02 4.83E+02 49.4 Te121 3.29E+04 8.55E+07 9.76E+06 199.8 Te121m 3.28E+04 8.52E+07 9.73E+06 199.8 Te123m 3.68E+08 1.23E+12 1.63E+11 199.9 Te125m 2.42E+13 1.76E+14 7.28E+13 151.6 Te127 1.55E+13 5.91E+13 2.48E+13 116.8 Te127m 1.59E+13 6.04E+13 2.53E+13 116.8 Te129 2.29E+11 6.73E+11 3.48E+11 98.5 Te129m 3.63E+11 1.07E+12 5.52E+11 98.5 Ru103 3.55E+13 9.60E+13 5.02E+13 91.9 Ru106 1.71E+15 2.06E+16 6.34E+15 169.3 Sr85 1.32E+05 1.63E+07 3.11E+06 196.8 Semi-Volatile Sr89 8.75E+13 2.32E+14 1.75E+14 90.3 Sr90 1.47E+15 1.68E+16 6.65E+15 167.8 Ba131 5.54E-07 1.81E-05 7.14E-06 188.1 Ba133 4.17E+06 4.51E+10 4.90E+09 200.0 Ba136m 1.51E+05 1.04E+07 1.76E+06 194.3 Ba137m 1.61E+15 2.05E+16 8.43E+15 170.9 Ba140 8.91E+07 1.29E+08 1.01E+08 36.3 Total 4.21E+16 2.24E+17 1.01E+17 136.7 Table C.5. Minimum, Maximum, and Average Radionuclide Activity at Time 0 (End of Operation) for Microreactors Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ac225 2.07E+03 3.13E+10 3.19E+09 200.0 Ac226 1.59E+02 2.26E+09 3.10E+08 200.0 Ac227 1.06E+04 2.61E+09 1.92E+08 200.0 Ac228 8.59E+02 2.26E+10 2.38E+09 200.0 Ag106 1.42E+03 4.97E+04 1.34E+04 188.9 Ag106m 2.59E+03 8.70E+04 2.25E+04 188.4 Ag107m 2.60E+05 6.90E+06 2.45E+06 185.5 Ag108 1.12E+08 7.42E+10 1.38E+10 199.4 Ag108m 2.46E+05 1.46E+08 2.98E+07 199.3 Ag109m 7.53E+14 2.55E+16 6.82E+15 188.5 Ag110 1.28E+13 2.17E+16 4.34E+15 199.8 Ag110m 3.39E+11 6.47E+14 1.30E+14 199.8 Ag111 2.81E+14 4.29E+15 1.18E+15 175.4 Ag111m 2.83E+14 4.10E+15 1.15E+15 174.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.20

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ag112 1.62E+14 1.84E+15 5.58E+14 167.7 Ag113 9.09E+13 8.06E+14 2.62E+14 159.4 Ag113m 1.33E+14 1.18E+15 3.83E+14 159.4 Ag114 1.09E+14 7.26E+14 2.72E+14 147.8 Ag115 9.62E+13 4.40E+14 2.07E+14 128.3 Ag115m 5.81E+12 1.21E+13 7.25E+12 70.4 Ag116 1.04E+14 3.73E+14 1.94E+14 113.0 Ag116m 6.72E+12 2.54E+13 1.22E+13 116.3 Ag117 8.38E+13 2.70E+14 1.54E+14 105.1 Ag117m 1.47E+13 6.16E+13 2.95E+13 122.9 Ag118 5.89E+13 1.86E+14 1.17E+14 103.8 Ag118m 2.86E+13 8.65E+13 5.06E+13 100.5 Ag119 5.63E+13 1.50E+14 1.05E+14 90.8 Ag120 2.29E+13 7.77E+13 4.68E+13 108.8 Ag120m 4.73E+12 3.53E+13 2.07E+13 152.7 Ag121 1.79E+13 7.24E+13 3.82E+13 120.9 Ag122 2.34E+12 2.00E+13 8.49E+12 158.2 Ag122m 2.29E+12 1.77E+13 7.68E+12 154.3 Ag123 1.56E+12 1.84E+13 7.17E+12 168.8 Ag124 5.52E+11 8.71E+12 4.23E+12 176.2 Ag125 1.58E+10 1.83E+12 6.36E+11 196.6 Ag126 3.04E+09 3.59E+11 1.30E+11 196.6 Ag127 4.66E+08 5.50E+10 2.10E+10 196.6 Ag128 6.37E+07 7.52E+09 2.92E+09 196.6 Ag129 1.07E+07 2.90E+09 7.26E+08 198.5 Ag130 1.55E+08 3.95E+10 2.42E+10 198.4 Am239 3.61E+02 7.65E+07 5.29E+06 200.0 Am240 1.50E+06 5.63E+10 4.29E+09 200.0 Am241 1.08E+09 1.99E+14 1.88E+13 200.0 Am242 2.00E+10 2.61E+16 3.63E+15 200.0 Am242m 7.06E+06 3.39E+13 2.49E+12 200.0 Am243 3.12E+04 1.73E+13 3.23E+12 200.0 Am244 7.08E+05 4.31E+15 7.01E+14 200.0 Am244m 9.96E+06 6.45E+16 1.05E+16 200.0 Am245 5.87E+02 3.81E+12 5.64E+11 200.0 Am246 2.90E+04 3.83E+06 1.41E+06 197.0 Am246m 3.48E+05 3.54E+06 1.95E+06 164.3 As72 4.95E+02 2.77E+03 1.47E+03 139.3 As73 3.48E+04 1.40E+06 3.47E+05 190.3 As74 4.41E+06 2.08E+08 4.33E+07 191.7 As76 8.83E+10 1.77E+13 2.61E+12 198.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.21

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

As77 4.54E+13 1.35E+14 6.49E+13 99.1 As78 1.29E+14 2.98E+14 1.62E+14 79.5 As79 2.77E+14 7.44E+14 3.56E+14 91.4 As80 5.85E+14 1.29E+15 8.19E+14 75.0 As81 1.04E+15 2.01E+15 1.28E+15 63.3 As82 6.64E+14 1.60E+15 1.37E+15 82.6 As82m 1.97E+14 8.70E+14 3.32E+14 126.3 As83 1.01E+15 2.73E+15 1.90E+15 91.9 As84 5.96E+14 1.55E+15 1.22E+15 89.0 As85 2.33E+14 1.44E+15 1.04E+15 144.2 As86 9.41E+13 3.52E+15 2.27E+15 189.6 As87 1.91E+13 2.69E+14 1.90E+14 173.5 As88 3.34E+12 7.72E+14 4.77E+14 198.3 As89 3.63E+11 3.92E+12 1.75E+12 166.1 As90 1.99E+10 1.05E+11 4.99E+10 136.3 As91 1.01E+09 1.75E+10 6.70E+09 178.1 As92 1.67E+07 1.44E+09 5.01E+08 195.4 At217 2.07E+03 3.13E+10 3.19E+09 200.0 B12 1.25E+05 2.95E+09 5.16E+08 200.0 Ba131 1.99E+03 6.49E+04 2.56E+04 188.1 Ba133 4.46E+06 4.81E+10 5.24E+09 200.0 Ba135m 2.98E+10 3.11E+14 4.86E+13 200.0 Ba136m 3.39E+13 2.32E+15 3.94E+14 194.3 Ba137m 1.65E+15 2.10E+16 8.64E+15 170.8 Ba139 3.81E+16 5.45E+16 4.33E+16 35.4 Ba140 3.68E+16 5.31E+16 4.18E+16 36.3 Ba141 3.49E+16 5.05E+16 3.93E+16 36.7 Ba142 3.24E+16 4.79E+16 3.75E+16 38.6 Ba143 2.91E+16 3.63E+16 3.36E+16 22.3 Ba144 2.20E+16 2.76E+16 2.58E+16 22.6 Ba145 9.90E+15 1.35E+16 1.17E+16 31.1 Ba146 3.85E+15 6.77E+15 5.36E+15 55.1 Ba147 8.78E+14 1.94E+15 1.44E+15 75.2 Ba148 1.23E+14 3.41E+14 2.01E+14 93.9 Ba149 6.82E+12 4.80E+13 2.22E+13 150.2 Ba150 3.66E+11 5.13E+12 2.10E+12 173.4 Ba151 1.40E+10 8.48E+11 3.00E+11 193.5 Ba152 2.41E+08 1.35E+10 4.92E+09 193.0 Ba153 6.87E+06 6.48E+08 2.25E+08 195.8 Be10 1.09E+04 2.96E+11 4.30E+10 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.22

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Be11 1.13E+02 1.43E+12 1.62E+11 200.0 Be8 1.14E+04 9.78E+16 1.54E+16 200.0 Bi210 5.14E+02 6.53E+08 5.51E+07 200.0 Bi211 9.59E+03 2.60E+09 1.98E+08 200.0 Bi212 3.59E+06 8.74E+12 9.51E+11 200.0 Bi213 2.07E+03 3.12E+10 3.19E+09 200.0 Bi214 1.23E+03 4.28E+06 5.19E+05 199.9 Bk248 4.30E+05 3.47E+06 1.95E+06 155.9 Bk248m 2.60E+04 3.07E+07 8.15E+06 199.7 Bk249 8.25E+01 4.95E+12 6.55E+11 200.0 Bk250 5.88E+02 5.56E+13 7.39E+12 200.0 Bk251 5.45E+06 1.15E+10 3.10E+09 199.8 Br77 7.41E+03 1.85E+05 5.42E+04 184.6 Br77m 5.82E+03 1.43E+05 4.18E+04 184.3 Br78 9.00E+05 1.97E+07 7.24E+06 182.5 Br79m 4.88E+07 1.07E+09 4.07E+08 182.6 Br80 5.89E+09 3.55E+11 6.78E+10 193.5 Br80m 2.32E+09 1.11E+11 2.15E+10 191.8 Br82 1.34E+13 9.96E+14 1.75E+14 194.7 Br82m 1.18E+13 9.21E+14 1.61E+14 194.9 Br83 2.22E+15 5.57E+15 3.36E+15 86.1 Br84 3.33E+15 8.62E+15 5.82E+15 88.5 Br84m 1.11E+14 7.26E+14 2.30E+14 146.8 Br85 4.22E+15 1.14E+16 7.55E+15 91.9 Br86 5.30E+15 1.28E+16 1.00E+16 83.0 Br87 5.57E+15 1.28E+16 1.10E+16 78.5 Br88 4.40E+15 1.14E+16 9.22E+15 88.3 Br89 3.18E+15 7.79E+15 6.09E+15 84.1 Br90 1.97E+15 4.24E+15 3.28E+15 73.3 Br91 3.23E+14 1.74E+15 1.19E+15 137.5 Br92 6.44E+13 3.31E+14 1.89E+14 135.0 Br93 2.33E+13 1.06E+14 6.13E+13 127.9 Br94 2.66E+12 1.57E+13 7.85E+12 142.1 Br95 1.65E+10 1.30E+11 5.62E+10 154.9 Br96 5.96E+09 4.06E+10 1.93E+10 148.8 Br97 5.43E+07 7.78E+08 3.62E+08 173.9 C14 3.67E+06 5.53E+10 1.18E+10 200.0 C15 2.67E+07 2.50E+10 8.15E+09 199.6 Cd107 1.04E+03 1.15E+05 3.32E+04 196.4 Cd109 4.09E+06 3.54E+10 4.40E+09 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.23

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Cd111m 6.40E+08 2.16E+13 3.41E+12 200.0 Cd113m 1.25E+10 1.51E+12 3.23E+11 196.7 Cd115 9.70E+13 4.85E+14 2.20E+14 133.3 Cd115m 5.61E+12 3.61E+13 1.44E+13 146.2 Cd117 8.27E+13 2.78E+14 1.54E+14 108.2 Cd117m 2.09E+13 6.33E+13 3.63E+13 100.7 Cd118 8.83E+13 2.61E+14 1.66E+14 99.0 Cd119 5.91E+13 1.57E+14 1.06E+14 90.8 Cd119m 3.45E+13 9.03E+13 5.85E+13 89.4 Cd120 8.90E+13 2.28E+14 1.57E+14 87.6 Cd121 4.75E+13 1.26E+14 8.82E+13 90.5 Cd121m 2.78E+13 8.63E+13 5.18E+13 102.6 Cd122 8.73E+13 1.74E+14 1.33E+14 66.2 Cd123 5.13E+13 1.35E+14 8.45E+13 90.1 Cd124 1.52E+13 1.18E+14 7.19E+13 154.4 Cd125 4.05E+12 5.84E+13 3.18E+13 174.0 Cd126 1.60E+12 4.88E+13 3.46E+13 187.3 Cd127 6.27E+11 4.78E+13 3.12E+13 194.8 Cd128 1.97E+11 2.11E+13 1.35E+13 196.3 Cd129 3.71E+10 3.87E+12 6.40E+11 196.2 Cd130 1.30E+12 5.18E+14 3.16E+14 199.0 Cd131 2.66E+11 9.55E+13 5.83E+13 198.9 Cd132 2.04E+09 1.61E+11 5.78E+10 195.0 Ce137 2.03E+04 8.51E+05 2.16E+05 190.7 Ce139 7.63E+09 1.05E+12 2.99E+11 197.1 Ce139m 2.95E+09 4.41E+11 1.24E+11 197.3 Ce141 3.51E+16 5.21E+16 3.97E+16 39.0 Ce143 3.28E+16 4.76E+16 3.81E+16 36.7 Ce144 2.74E+16 3.91E+16 3.32E+16 35.4 Ce145 2.24E+16 2.95E+16 2.52E+16 27.2 Ce146 1.79E+16 2.32E+16 1.98E+16 25.6 Ce147 1.23E+16 1.70E+16 1.36E+16 32.2 Ce148 9.27E+15 1.14E+16 1.03E+16 20.4 Ce149 4.88E+15 6.88E+15 5.55E+15 34.0 Ce150 2.56E+15 3.68E+15 2.96E+15 36.1 Ce151 6.45E+14 1.19E+15 8.53E+14 59.7 Ce152 1.34E+14 3.27E+14 2.16E+14 83.6 Ce153 1.18E+13 6.04E+13 3.56E+13 134.6 Ce154 7.26E+11 7.75E+12 4.01E+12 165.7 Ce155 2.93E+10 7.76E+11 3.50E+11 185.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.24

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ce156 1.14E+09 4.95E+10 2.11E+10 191.0 Ce157 3.19E+07 2.00E+09 8.21E+08 193.7 Cf248 3.75E+04 2.76E+07 7.87E+06 199.5 Cf249 7.56E+07 2.48E+09 1.07E+09 188.2 Cf250 3.60E+03 4.05E+11 8.86E+10 200.0 Cf251 2.41E+07 3.02E+09 8.25E+08 196.8 Cf252 1.19E+03 7.19E+12 1.50E+12 200.0 Cf253 2.89E+08 9.58E+11 2.51E+11 199.9 Cf254 1.20E+05 1.94E+09 5.11E+08 200.0 Cf255 2.40E+05 3.22E+06 1.73E+06 172.3 Cm241 6.04E+03 4.85E+09 4.68E+08 200.0 Cm242 6.50E+09 2.02E+16 2.83E+15 200.0 Cm243 4.88E+05 3.71E+13 3.30E+12 200.0 Cm244 1.31E+05 1.02E+16 1.58E+15 200.0 Cm245 1.48E+04 2.01E+12 3.58E+11 200.0 Cm246 2.21E+04 3.43E+12 5.52E+11 200.0 Cm247 2.58E+06 4.74E+07 1.66E+07 179.3 Cm248 1.66E+07 1.67E+09 4.87E+08 196.0 Cm249 6.70E+01 6.20E+13 6.69E+12 200.0 Cm251 2.51E+05 4.43E+08 1.14E+08 199.8 Co65 8.18E+05 3.49E+07 1.35E+07 190.8 Co66 5.67E+08 2.32E+10 5.23E+09 190.4 Co67 1.98E+09 2.73E+10 1.21E+10 173.0 Co68 2.19E+09 3.20E+10 1.39E+10 174.4 Co69 2.55E+09 2.97E+10 1.35E+10 168.3 Co70 1.67E+09 2.02E+10 9.41E+09 169.5 Co71 1.06E+09 1.31E+10 5.39E+09 170.2 Co72 3.91E+08 4.69E+09 1.95E+09 169.2 Co73 1.02E+08 1.72E+09 7.49E+08 177.6 Co74 1.33E+07 3.51E+08 1.19E+08 185.4 Co75 1.52E+06 5.37E+07 1.71E+07 189.0 Cr66 3.65E+04 2.34E+06 9.32E+05 193.8 Cr67 6.93E+03 7.03E+05 2.98E+05 196.1 Cs131 8.10E+05 7.71E+07 1.49E+07 195.8 Cs132 8.54E+09 9.54E+11 2.88E+11 196.5 Cs134 1.53E+14 4.04E+16 9.92E+15 198.5 Cs134m 5.59E+13 5.14E+15 1.39E+15 195.7 Cs135 2.48E+10 3.11E+11 9.45E+10 170.4 Cs135m 2.87E+12 1.32E+14 2.66E+13 191.5 Cs136 3.05E+14 2.10E+16 3.55E+15 194.3 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.25

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Cs136m 3.59E+13 7.30E+14 1.92E+14 181.2 Cs137 1.74E+15 2.21E+16 9.11E+15 170.9 Cs138 4.00E+16 5.37E+16 4.50E+16 29.3 Cs138m 1.34E+15 4.90E+15 2.41E+15 114.1 Cs139 3.73E+16 5.10E+16 4.22E+16 31.2 Cs140 3.19E+16 4.06E+16 3.62E+16 24.0 Cs141 2.52E+16 3.31E+16 2.79E+16 27.1 Cs142 1.52E+16 1.85E+16 1.69E+16 19.5 Cs143 7.16E+15 9.97E+15 8.81E+15 32.8 Cs144 2.16E+15 3.75E+15 2.88E+15 53.9 Cs145 4.47E+14 9.29E+14 6.24E+14 70.1 Cs146 4.62E+13 1.49E+14 8.52E+13 105.2 Cs147 7.73E+12 2.39E+13 1.54E+13 102.2 Cs148 1.05E+11 2.26E+12 8.80E+11 182.4 Cs149 2.66E+09 4.16E+10 1.70E+10 176.0 Cs150 5.76E+07 4.31E+09 1.53E+09 194.7 Cs151 7.73E+06 7.55E+08 2.61E+08 195.9 Cu66 6.49E+08 3.27E+10 6.35E+09 192.2 Cu67 2.71E+09 5.12E+10 1.65E+10 179.9 Cu68 5.68E+09 8.75E+10 2.94E+10 175.6 Cu68m 1.22E+08 4.76E+09 9.07E+08 190.0 Cu69 1.36E+10 1.56E+11 5.62E+10 168.1 Cu70 3.02E+10 2.99E+11 1.09E+11 163.3 Cu70m 6.87E+09 9.29E+10 3.04E+10 172.5 Cu71 6.75E+10 8.79E+11 2.58E+11 171.5 Cu72 1.60E+11 1.37E+12 4.53E+11 158.1 Cu73 3.74E+11 2.21E+12 7.93E+11 142.2 Cu74 4.49E+11 1.88E+12 7.99E+11 123.0 Cu75 3.28E+11 2.11E+12 9.18E+11 146.2 Cu76 1.46E+11 1.04E+12 6.33E+11 150.6 Cu77 5.47E+10 5.28E+11 3.12E+11 162.4 Cu78 1.66E+10 1.53E+11 8.57E+10 160.9 Cu79 4.10E+08 3.36E+10 1.27E+10 195.2 Cu80 1.87E+08 3.44E+09 1.38E+09 179.4 Dy159 1.29E+05 3.09E+08 5.08E+07 199.8 Dy165 1.77E+11 4.46E+14 4.84E+13 199.8 Dy165m 1.71E+09 2.84E+14 3.06E+13 200.0 Dy166 4.28E+10 3.21E+12 5.43E+11 194.7 Dy167 1.10E+10 6.15E+11 1.49E+11 193.0 Dy168 3.41E+09 3.40E+11 7.37E+10 196.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.26

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Dy169 1.18E+09 1.33E+11 3.02E+10 196.5 Dy170 3.03E+08 4.76E+10 1.14E+10 197.5 Dy171 9.70E+07 1.46E+10 3.98E+09 197.4 Dy172 7.67E+07 7.13E+09 2.81E+09 195.7 Er165 5.14E+04 2.56E+09 1.86E+08 200.0 Er167m 2.25E+09 6.90E+13 6.35E+12 200.0 Er169 1.42E+09 6.27E+12 5.19E+11 199.9 Er171 1.91E+08 6.20E+10 1.09E+10 198.8 Er172 1.44E+08 2.44E+10 6.80E+09 197.6 Es253 2.41E+08 5.16E+11 1.35E+11 199.8 Es254 3.50E+05 1.22E+09 3.13E+08 199.9 Es254m 1.88E+07 1.24E+11 3.21E+10 199.9 Es255 5.24E+04 1.79E+09 6.22E+08 200.0 Eu149 8.40E+01 1.34E+05 3.32E+04 199.7 Eu152 3.13E+10 9.23E+12 1.48E+12 198.6 Eu152m 7.71E+11 5.70E+13 1.17E+13 194.7 Eu154 6.53E+12 3.10E+15 4.39E+14 199.2 Eu154m 1.56E+12 5.93E+14 1.12E+14 198.9 Eu155 7.47E+13 1.27E+15 2.90E+14 177.8 Eu156 1.78E+14 3.98E+16 6.81E+15 198.2 Eu157 8.62E+13 1.58E+15 3.71E+14 179.4 Eu158 4.56E+13 5.23E+14 1.55E+14 168.0 Eu159 1.95E+13 2.75E+14 7.76E+13 173.4 Eu160 7.64E+12 1.31E+14 3.50E+13 177.9 Eu161 2.68E+12 5.75E+13 1.48E+13 182.2 Eu162 6.39E+11 1.32E+13 3.56E+12 181.5 Eu163 1.43E+11 4.34E+12 1.05E+12 187.2 Eu164 1.77E+10 7.39E+11 1.86E+11 190.6 Eu165 2.44E+09 9.77E+10 3.13E+10 190.3 Eu166 2.62E+08 1.13E+10 5.08E+09 191.0 Eu167 2.83E+07 2.35E+09 8.76E+08 195.2 F20 5.93E+04 2.09E+17 1.59E+16 200.0 Fe65 8.18E+05 3.49E+07 1.35E+07 190.8 Fe66 2.91E+08 7.49E+09 2.26E+09 185.0 Fe67 4.49E+08 9.04E+09 3.58E+09 181.1 Fe68 2.69E+08 5.48E+09 2.16E+09 181.3 Fe69 9.39E+07 1.96E+09 7.88E+08 181.7 Fe70 2.54E+07 5.41E+08 2.29E+08 182.1 Fe71 3.91E+06 1.45E+08 4.48E+07 189.5 Fe72 4.90E+05 2.55E+07 6.68E+06 192.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.27

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Fr221 2.07E+03 3.13E+10 3.19E+09 200.0 Fr223 2.81E+02 3.60E+07 3.98E+06 200.0 Ga68 1.70E+03 4.94E+05 1.08E+05 198.6 Ga70 4.67E+07 1.81E+10 2.73E+09 199.0 Ga72 2.37E+11 2.65E+12 7.57E+11 167.1 Ga72m 9.01E+09 9.42E+10 2.84E+10 165.1 Ga73 7.73E+11 5.96E+12 1.83E+12 154.1 Ga74 2.35E+12 1.34E+13 4.41E+12 140.5 Ga74m 8.07E+10 7.48E+11 2.28E+11 161.1 Ga75 7.08E+12 3.67E+13 1.17E+13 135.3 Ga76 1.63E+13 6.05E+13 2.73E+13 115.2 Ga77 2.98E+13 9.47E+13 5.39E+13 104.4 Ga78 4.86E+13 1.15E+14 8.70E+13 81.0 Ga79 4.29E+13 1.38E+14 1.07E+14 105.3 Ga80 2.84E+13 8.95E+13 6.70E+13 103.6 Ga81 1.32E+13 5.90E+13 4.32E+13 126.9 Ga82 3.18E+12 3.91E+13 2.83E+13 169.9 Ga83 5.85E+11 4.75E+12 2.24E+12 156.2 Ga84 1.33E+11 6.76E+13 4.15E+13 199.2 Ga85 3.22E+09 9.18E+10 3.40E+10 186.4 Ga86 5.87E+08 1.94E+11 1.20E+11 198.8 Gd151 6.50E+05 1.84E+08 3.68E+07 198.6 Gd153 2.18E+09 7.97E+12 9.19E+11 199.9 Gd159 2.00E+13 1.43E+15 2.54E+14 194.5 Gd161 3.20E+12 7.60E+13 1.94E+13 183.9 Gd162 1.32E+12 2.70E+13 7.33E+12 181.4 Gd163 5.23E+11 1.43E+13 3.50E+12 185.9 Gd164 1.60E+11 5.24E+12 1.26E+12 188.2 Gd165 4.24E+10 1.46E+12 3.61E+11 188.7 Gd166 1.13E+10 3.30E+11 9.92E+10 186.8 Gd167 1.17E+09 5.54E+10 1.98E+10 191.7 Gd168 2.06E+08 1.12E+10 5.45E+09 192.7 Gd169 3.16E+07 2.42E+09 9.81E+08 194.8 Ge71 1.75E+05 7.88E+08 7.13E+07 199.9 Ge71m 4.25E+04 6.33E+05 1.91E+05 174.9 Ge73m 7.61E+11 5.92E+12 1.81E+12 154.4 Ge75 7.20E+12 3.78E+13 1.20E+13 136.0 Ge75m 3.23E+11 2.18E+12 6.33E+11 148.2 Ge77 4.45E+13 1.29E+14 6.51E+13 97.7 Ge77m 7.43E+11 5.51E+12 1.70E+12 152.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.28

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ge78 1.26E+14 2.90E+14 1.59E+14 78.7 Ge79 1.42E+14 4.04E+14 2.16E+14 96.3 Ge79m 8.31E+13 2.87E+14 1.14E+14 110.3 Ge80 3.87E+14 1.01E+15 6.82E+14 89.4 Ge81 3.93E+14 8.30E+14 7.26E+14 71.5 Ge81m 5.47E+12 2.45E+13 1.79E+13 127.0 Ge82 2.41E+14 8.63E+14 6.70E+14 112.8 Ge83 8.86E+13 2.95E+14 2.06E+14 107.5 Ge84 2.75E+13 2.17E+14 1.54E+14 154.9 Ge85 4.30E+12 3.77E+13 1.94E+13 159.1 Ge86 3.16E+12 3.58E+15 2.19E+15 199.6 Ge87 5.25E+10 1.30E+13 8.04E+12 198.4 Ge88 3.84E+09 3.08E+11 2.02E+11 195.1 Ge89 6.25E+07 3.87E+09 1.35E+09 193.6 H3 7.23E+12 3.89E+16 3.49E+15 199.9 He6 4.23E+03 3.73E+16 4.71E+15 200.0 Ho161 1.86E+03 1.57E+05 2.73E+04 195.3 Ho162 3.01E+04 2.45E+06 5.54E+05 195.1 Ho162m 3.00E+04 2.39E+06 5.36E+05 195.0 Ho163 3.26E+02 8.95E+02 5.66E+02 93.3 Ho163m 8.59E+04 5.41E+06 1.21E+06 193.7 Ho164 4.37E+06 2.71E+09 2.52E+08 199.4 Ho164m 3.00E+06 1.27E+09 1.27E+08 199.1 Ho166 5.63E+10 2.86E+14 2.85E+13 199.9 Ho166m 2.52E+05 7.42E+08 9.26E+07 199.9 Ho167 1.18E+10 1.65E+13 1.68E+12 199.7 Ho168 3.60E+09 3.68E+11 7.88E+10 196.1 Ho169 1.39E+09 1.69E+11 3.64E+10 196.7 Ho170 3.49E+08 6.36E+10 1.40E+10 197.8 Ho170m 4.60E+07 1.61E+10 2.62E+09 198.9 Ho171 1.72E+08 4.10E+10 8.30E+09 198.3 Ho172 1.29E+08 1.67E+10 5.61E+09 196.9 I123 3.56E+04 7.13E+05 2.26E+05 181.0 I125 3.41E+02 1.16E+06 2.17E+05 199.9 I126 2.38E+08 4.35E+10 1.27E+10 197.8 I128 2.34E+13 2.80E+15 4.96E+14 196.7 I129 3.62E+08 9.18E+09 2.57E+09 184.8 I130 4.94E+13 5.96E+15 1.06E+15 196.7 I130m 3.14E+13 3.85E+15 6.82E+14 196.8 I131 1.95E+16 3.16E+16 2.28E+16 47.3 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.29

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

I132 2.91E+16 4.43E+16 3.32E+16 41.5 I132m 8.69E+13 8.29E+14 3.42E+14 162.1 I133 4.28E+16 5.89E+16 4.71E+16 31.5 I133m 3.07E+15 6.03E+15 4.04E+15 65.0 I134 4.82E+16 6.51E+16 5.32E+16 29.9 I134m 2.50E+15 8.15E+15 4.37E+15 106.1 I135 4.09E+16 5.67E+16 4.45E+16 32.3 I136 1.60E+16 2.06E+16 1.80E+16 24.7 I136m 8.52E+15 1.44E+16 9.93E+15 51.6 I137 1.90E+16 2.51E+16 2.05E+16 27.8 I138 9.51E+15 1.33E+16 1.06E+16 33.3 I139 3.68E+15 5.65E+15 4.75E+15 42.1 I140 7.13E+14 1.54E+15 1.09E+15 73.3 I141 1.22E+14 4.09E+14 2.62E+14 108.1 I142 2.60E+13 6.41E+13 4.34E+13 84.5 I143 1.85E+11 6.63E+12 2.41E+12 189.1 I144 1.08E+10 1.02E+11 4.41E+10 161.7 In111 6.78E+01 7.11E+04 1.60E+04 199.6 In112 1.72E+03 1.08E+07 2.01E+06 199.9 In112m 1.39E+03 8.75E+06 1.62E+06 199.9 In113m 3.60E+01 8.36E+06 1.21E+06 200.0 In114 7.50E+08 2.95E+11 3.47E+10 199.0 In114m 4.37E+08 1.66E+11 1.99E+10 199.0 In115m 9.74E+13 4.85E+14 2.22E+14 133.1 In116 1.33E+12 1.80E+14 5.00E+13 197.1 In116m 2.18E+12 2.99E+14 8.27E+13 197.1 In117 6.34E+13 2.06E+14 1.16E+14 105.9 n117m 7.60E+13 2.55E+14 1.42E+14 108.2 In118 8.83E+13 2.62E+14 1.66E+14 99.1 In118m 1.74E+10 1.81E+11 6.05E+10 164.9 In119 4.57E+13 1.16E+14 7.64E+13 86.8 In119m 5.36E+13 1.43E+14 9.60E+13 91.0 In120 9.09E+13 2.34E+14 1.61E+14 88.2 In120m 3.43E+12 1.45E+13 6.98E+12 123.5 In121 6.38E+13 1.87E+14 1.15E+14 98.3 In121m 3.54E+13 9.47E+13 6.64E+13 91.2 In122 9.68E+13 2.20E+14 1.56E+14 77.6 In122m 1.81E+13 1.01E+14 4.56E+13 139.1 In123 5.57E+13 2.37E+14 1.12E+14 123.9 In123m 4.97E+13 1.04E+14 7.53E+13 70.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.30

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

In124 9.71E+13 1.74E+14 1.36E+14 56.9 In124m 2.10E+13 1.40E+14 6.39E+13 148.1 In125 6.04E+13 1.70E+14 1.01E+14 95.3 In125m 4.77E+13 1.68E+14 9.37E+13 111.6 In126 6.76E+13 1.65E+14 1.04E+14 83.5 In126m 1.95E+13 1.55E+14 6.97E+13 155.2 In127 2.02E+14 2.99E+14 2.69E+14 38.9 In127m 5.31E+13 2.32E+14 1.19E+14 125.6 In128 1.21E+14 2.36E+14 1.59E+14 64.6 In128m 9.96E+13 2.36E+14 1.45E+14 81.2 In129 6.52E+13 2.34E+14 1.67E+14 112.7 In129m 6.51E+13 2.28E+14 1.55E+14 111.2 In130 7.90E+13 5.91E+14 4.00E+14 152.8 In130m 5.15E+13 1.51E+14 1.07E+14 98.2 In131 5.62E+13 9.14E+13 7.57E+13 47.7 In131m 3.58E+13 6.25E+13 4.64E+13 54.3 In132 1.99E+13 8.40E+13 4.75E+13 123.3 In133 1.05E+12 7.40E+12 3.20E+12 150.2 In134 2.58E+10 4.51E+11 1.70E+11 178.3 In135 2.10E+08 1.59E+10 5.58E+09 194.8 Kr100 3.37E+07 6.81E+09 4.27E+09 198.0 Kr79 5.53E+03 1.25E+05 4.04E+04 183.1 Kr79m 2.78E+03 6.25E+04 2.03E+04 183.0 Kr81 2.40E+02 2.30E+06 3.28E+05 200.0 Kr81m 1.72E+07 4.78E+10 6.62E+09 199.9 Kr83m 2.30E+15 5.83E+15 3.40E+15 86.9 Kr85 1.95E+14 2.36E+15 8.07E+14 169.5 Kr85m 4.32E+15 1.19E+16 7.66E+15 93.3 Kr87 7.57E+15 2.12E+16 1.45E+16 94.7 Kr88 9.72E+15 2.74E+16 1.96E+16 95.3 Kr89 1.13E+16 2.93E+16 2.40E+16 88.8 Kr90 1.04E+16 2.87E+16 2.40E+16 93.8 Kr91 6.75E+15 2.09E+16 1.67E+16 102.5 Kr92 3.57E+15 1.10E+16 8.56E+15 101.8 Kr93 1.05E+15 3.60E+15 2.64E+15 109.6 Kr94 3.35E+14 1.08E+15 6.48E+14 105.1 Kr95 3.22E+13 1.60E+14 7.95E+13 132.8 Kr96 5.00E+13 2.34E+14 1.65E+14 129.6 Kr97 1.08E+12 9.03E+12 3.23E+12 157.4 Kr98 3.25E+10 9.96E+12 6.16E+12 198.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.31

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Kr99 2.06E+08 1.95E+10 6.76E+09 195.8 La135 9.69E+04 2.52E+06 8.95E+05 185.2 La137 8.12E+03 1.90E+06 4.90E+05 198.3 La140 3.71E+16 6.40E+16 4.43E+16 53.1 La141 3.51E+16 5.14E+16 3.96E+16 37.7 La142 3.34E+16 5.07E+16 3.87E+16 41.1 La143 3.25E+16 4.62E+16 3.77E+16 34.9 La144 2.90E+16 3.80E+16 3.36E+16 26.9 La145 2.09E+16 2.61E+16 2.38E+16 21.8 La146 9.18E+15 1.15E+16 1.02E+16 22.1 La146m 4.19E+15 5.64E+15 4.80E+15 29.6 La147 5.59E+15 8.00E+15 6.60E+15 35.6 La148 2.12E+15 3.13E+15 2.58E+15 38.2 La149 5.06E+14 1.09E+15 7.63E+14 72.9 La150 6.92E+13 2.56E+14 1.52E+14 114.9 La151 7.05E+12 4.62E+13 2.33E+13 147.0 La152 3.62E+11 5.50E+12 2.46E+12 175.3 La153 1.63E+10 5.87E+11 2.31E+11 189.2 La154 4.02E+08 2.54E+10 9.61E+09 193.8 La155 1.02E+07 7.18E+08 2.89E+08 194.4 Li8 1.26E+15 4.58E+17 1.36E+17 198.9 Mn66 7.52E+06 3.21E+08 1.24E+08 190.8 Mn67 6.04E+06 2.87E+08 1.04E+08 191.7 Mn68 9.87E+05 6.03E+07 2.32E+07 193.6 Mn69 1.26E+05 1.03E+07 3.44E+06 195.2 Mo101 3.45E+16 5.15E+16 3.83E+16 39.5 Mo102 2.93E+16 5.15E+16 3.44E+16 55.1 Mo103 2.18E+16 5.55E+16 3.00E+16 87.3 Mo104 1.42E+16 4.93E+16 2.28E+16 110.8 Mo105 8.02E+15 3.67E+16 1.52E+16 128.2 Mo106 3.68E+15 2.35E+16 8.56E+15 145.8 Mo107 1.17E+15 8.91E+15 3.17E+15 153.6 Mo108 3.35E+14 2.99E+15 1.06E+15 159.7 Mo109 1.04E+14 3.57E+14 2.00E+14 109.7 Mo110 7.24E+12 8.29E+13 3.89E+13 167.9 Mo111 6.99E+11 1.82E+13 6.88E+12 185.2 Mo112 6.28E+10 2.41E+12 8.61E+11 189.8 Mo113 3.27E+09 2.79E+11 9.74E+10 195.4 Mo114 4.57E+08 1.21E+10 5.01E+09 185.5 Mo115 1.18E+07 9.32E+08 3.76E+08 195.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.32

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Mo93 3.65E+03 3.65E+03 3.65E+03 0.0 Mo93m 1.30E+03 4.62E+04 1.63E+04 189.0 Mo99 3.94E+16 5.35E+16 4.30E+16 30.3 N16 1.13E+11 4.83E+16 4.50E+15 200.0 Na24 2.73E+06 2.73E+06 2.73E+06 0.0 Na24m 2.10E+06 2.10E+06 2.10E+06 0.0 Nb100 3.79E+16 5.03E+16 4.05E+16 28.1 Nb100m 1.97E+15 5.32E+15 3.05E+15 92.0 Nb101 3.33E+16 4.97E+16 3.66E+16 39.5 Nb102 1.86E+16 2.73E+16 2.07E+16 37.9 Nb102m 5.48E+15 1.32E+16 7.34E+15 83.0 Nb103 1.33E+16 2.97E+16 1.77E+16 76.1 Nb104 2.67E+15 8.37E+15 4.38E+15 103.3 Nb104m 2.11E+15 7.26E+15 3.47E+15 109.9 Nb105 1.86E+15 6.33E+15 3.23E+15 109.1 Nb106 1.34E+14 1.36E+15 5.52E+14 164.0 Nb107 2.35E+13 2.70E+14 1.17E+14 168.0 Nb108 1.53E+12 2.75E+13 1.26E+13 179.0 Nb109 1.26E+12 5.08E+12 3.09E+12 120.7 Nb110 1.72E+10 2.88E+11 1.39E+11 177.5 Nb111 3.14E+09 2.64E+11 9.19E+10 195.3 Nb112 4.04E+07 2.71E+09 9.43E+08 194.1 Nb113 4.41E+06 4.94E+08 1.70E+08 196.5 Nb91 1.73E+02 3.61E+02 2.75E+02 70.5 Nb92m 7.58E+03 3.00E+06 5.10E+05 199.0 Nb93m 1.33E+09 1.48E+11 4.45E+10 196.4 Nb94 1.46E+06 3.99E+07 1.28E+07 185.9 Nb94m 4.89E+09 5.85E+10 2.65E+10 169.2 Nb95 3.60E+16 4.87E+16 4.08E+16 30.0 Nb95m 3.89E+14 5.31E+14 4.42E+14 30.8 Nb96 1.44E+13 1.66E+14 6.42E+13 168.0 Nb97 3.63E+16 5.13E+16 4.04E+16 34.1 Nb97m 3.44E+16 4.83E+16 3.82E+16 33.6 Nb98 3.67E+16 4.80E+16 3.98E+16 26.6 Nb98m 1.94E+14 3.65E+14 2.50E+14 61.1 Nb99 2.32E+16 3.06E+16 2.51E+16 27.7 Nb99m 1.55E+16 2.21E+16 1.73E+16 35.4 Nd140 1.56E+02 1.53E+05 3.84E+04 199.6 Nd141 1.99E+07 7.99E+10 8.98E+09 199.9 Nd141m 4.04E+06 1.72E+10 1.91E+09 199.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.33

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Nd144 6.56E+01 6.76E+01 6.66E+01 3.1 Nd147 1.41E+16 1.93E+16 1.56E+16 31.2 Nd149 7.24E+15 1.28E+16 8.59E+15 55.5 Nd151 2.96E+15 7.19E+15 4.06E+15 83.3 Nd152 1.92E+15 5.03E+15 2.68E+15 89.7 Nd153 1.08E+15 3.15E+15 1.59E+15 98.0 Nd154 4.82E+14 1.85E+15 8.05E+14 117.1 Nd155 1.46E+14 7.66E+14 3.03E+14 136.1 Nd156 4.06E+13 2.91E+14 1.05E+14 151.1 Nd157 5.14E+12 7.99E+13 2.47E+13 175.8 Nd158 6.28E+11 1.45E+13 4.29E+12 183.4 Nd159 3.77E+10 1.41E+12 4.30E+11 189.6 Nd160 1.86E+09 7.95E+10 2.82E+10 190.9 Nd161 6.05E+07 2.91E+09 1.16E+09 191.8 Ne23 4.34E+05 7.57E+08 2.60E+08 199.8 Ni65 8.19E+05 3.50E+07 1.36E+07 190.8 Ni66 6.48E+08 3.25E+10 6.33E+09 192.2 Ni67 2.70E+09 4.98E+10 1.63E+10 179.4 Ni68 5.53E+09 8.15E+10 2.83E+10 174.6 Ni69 1.18E+10 1.23E+11 4.77E+10 164.9 Ni70 2.46E+10 2.24E+11 8.50E+10 160.4 Ni71 3.26E+10 3.48E+11 1.10E+11 165.7 Ni72 5.11E+10 3.48E+11 1.24E+11 148.8 Ni73 3.04E+10 2.12E+11 9.39E+10 149.7 Ni74 1.20E+10 9.40E+10 4.72E+10 154.6 Ni75 2.97E+09 3.11E+10 1.58E+10 165.1 Ni76 5.91E+08 8.70E+09 4.71E+09 174.5 Ni77 9.76E+07 1.32E+09 6.56E+08 172.4 Ni78 1.21E+07 2.05E+08 8.80E+07 177.6 Np235 1.21E+05 1.41E+08 2.75E+07 199.7 Np236 1.14E+03 1.12E+06 2.15E+05 199.6 Np236m 9.89E+08 3.73E+11 8.94E+10 198.9 Np237 1.97E+09 9.70E+10 2.08E+10 192.1 Np238 1.81E+14 4.44E+16 9.42E+15 198.4 Np239 1.37E+17 8.51E+17 4.23E+17 144.7 Np240 7.62E+12 5.91E+14 1.30E+14 194.9 Np240m 1.31E+13 1.03E+15 2.21E+14 195.0 Np241 5.21E+04 1.96E+08 2.12E+07 199.9 O19 7.39E+09 3.52E+15 3.27E+14 200.0 Pa229 3.11E+02 6.14E+06 3.54E+06 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.34

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pa230 1.61E+04 2.13E+09 2.94E+08 200.0 Pa231 5.54E+04 2.58E+10 2.47E+09 200.0 Pa232 8.71E+08 7.67E+14 6.18E+13 200.0 Pa233 2.33E+09 4.48E+17 5.43E+16 200.0 Pa234 2.60E+07 6.58E+15 6.17E+14 200.0 Pa234m 9.39E+09 6.92E+15 6.48E+14 200.0 Pa235 7.79E+04 3.61E+10 9.70E+09 200.0 Pb207m 3.15E+07 1.37E+08 8.44E+07 125.3 Pb209 2.14E+03 3.16E+10 3.22E+09 200.0 Pb210 5.25E+02 6.22E+08 5.16E+07 200.0 Pb211 9.59E+03 2.60E+09 1.98E+08 200.0 Pb212 3.58E+06 8.74E+12 9.51E+11 200.0 Pb214 1.23E+03 4.28E+06 5.19E+05 199.9 Pd101 2.54E+03 1.82E+05 6.57E+04 194.5 Pd103 3.34E+08 5.33E+11 5.98E+10 199.7 Pd107 4.70E+08 4.09E+10 8.93E+09 195.5 Pd107m 1.61E+11 6.10E+13 1.44E+13 198.9 Pd109 7.53E+14 2.55E+16 6.82E+15 188.5 Pd109m 1.75E+11 2.49E+14 5.54E+13 199.7 Pd111 2.85E+14 3.93E+15 1.12E+15 173.0 Pd111m 7.32E+10 5.77E+13 9.44E+12 199.5 Pd112 1.62E+14 1.82E+15 5.53E+14 167.4 Pd113 1.39E+14 1.21E+15 3.97E+14 159.0 Pd114 1.04E+14 7.22E+14 2.68E+14 149.9 Pd115 9.30E+13 4.31E+14 2.02E+14 129.1 Pd116 7.28E+13 3.46E+14 1.67E+14 130.4 Pd117 6.85E+13 2.04E+14 1.23E+14 99.5 Pd118 2.48E+13 1.04E+14 6.42E+13 122.7 Pd119 4.14E+12 4.63E+13 2.31E+13 167.2 Pd120 6.52E+12 3.29E+13 1.85E+13 133.8 Pd121 3.69E+11 7.97E+12 2.99E+12 182.3 Pd122 4.94E+10 2.28E+12 8.08E+11 191.5 Pd123 6.62E+09 4.90E+11 1.73E+11 194.7 Pd124 1.77E+09 8.08E+10 2.95E+10 191.4 Pm144 2.87E+02 2.87E+04 5.64E+03 196.0 Pm145 3.55E+04 3.32E+07 6.40E+06 199.6 Pm146 1.54E+08 2.33E+11 4.47E+10 199.7 Pm147 5.56E+15 1.22E+16 9.04E+15 74.4 Pm148 2.11E+14 6.37E+15 2.29E+15 187.2 Pm148m 1.65E+14 3.77E+15 8.99E+14 183.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.35

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pm149 7.35E+15 1.85E+16 9.67E+15 86.3 Pm150 1.58E+12 4.50E+14 5.71E+13 198.6 Pm151 2.97E+15 7.24E+15 4.08E+15 83.7 Pm152 1.93E+15 5.11E+15 2.71E+15 90.3 Pm152m 1.36E+13 1.42E+14 5.25E+13 165.0 Pm153 1.17E+15 3.60E+15 1.77E+15 101.8 Pm154 5.42E+14 2.18E+15 9.28E+14 120.3 Pm154m 4.31E+13 3.30E+14 1.23E+14 153.8 Pm155 2.77E+14 1.52E+15 5.82E+14 138.3 Pm156 1.22E+14 8.96E+14 3.09E+14 152.0 Pm157 4.15E+13 4.69E+14 1.42E+14 167.5 Pm158 1.01E+13 1.73E+14 4.67E+13 177.9 Pm159 1.83E+12 4.35E+13 1.15E+13 183.8 Pm160 1.98E+11 6.80E+12 1.77E+12 188.7 Pm161 1.78E+10 7.49E+11 2.02E+11 190.7 Pm162 4.12E+08 1.32E+10 6.18E+09 187.9 Pm163 2.15E+07 1.39E+09 5.46E+08 193.9 Po210 3.65E+02 6.19E+08 5.62E+07 200.0 Po211 3.39E+02 7.17E+06 1.17E+06 200.0 Po212 2.30E+06 5.60E+12 6.09E+11 200.0 Po213 2.03E+03 3.06E+10 3.12E+09 200.0 Po214 1.44E+04 2.07E+09 1.72E+08 200.0 Po215 9.59E+03 2.60E+09 1.97E+08 200.0 Po216 3.60E+06 8.74E+12 9.51E+11 200.0 Po218 1.23E+03 4.28E+06 5.19E+05 199.9 Pr139 8.42E+02 1.06E+07 1.32E+06 200.0 Pr140 4.86E+09 1.00E+12 2.18E+11 198.1 Pr142 1.26E+14 1.83E+16 3.19E+15 197.3 Pr142m 4.40E+13 6.37E+15 1.11E+15 197.3 Pr143 3.28E+16 4.73E+16 3.81E+16 36.0 Pr144 2.74E+16 3.93E+16 3.33E+16 35.7 Pr144m 2.67E+14 5.70E+14 3.65E+14 72.4 Pr145 2.24E+16 2.95E+16 2.52E+16 27.3 Pr146 1.80E+16 2.34E+16 1.99E+16 25.9 Pr147 1.39E+16 1.82E+16 1.52E+16 26.7 Pr148 1.01E+16 1.31E+16 1.10E+16 25.9 Pr148m 2.81E+14 1.82E+15 7.60E+14 146.6 Pr149 7.13E+15 1.06E+16 7.96E+15 39.3 Pr150 4.20E+15 7.73E+15 5.11E+15 59.1 Pr151 2.31E+15 4.68E+15 2.90E+15 67.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.36

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pr152 8.61E+14 2.37E+15 1.26E+15 93.4 Pr153 2.69E+14 9.49E+14 4.75E+14 111.6 Pr154 3.94E+13 2.63E+14 1.09E+14 147.8 Pr155 5.65E+12 5.32E+13 2.35E+13 161.6 Pr156 4.53E+11 8.10E+12 3.46E+12 178.8 Pr157 3.04E+10 8.87E+11 4.01E+11 186.8 Pr158 1.41E+09 5.57E+10 2.54E+10 190.2 Pr159 4.62E+07 2.08E+09 1.04E+09 191.3 Pu236 1.74E+08 7.91E+10 1.85E+10 199.1 Pu237 2.57E+07 7.17E+10 9.10E+09 199.9 Pu237m 1.04E+07 2.33E+10 3.66E+09 199.8 Pu238 1.20E+12 3.31E+15 4.61E+14 199.9 Pu239 4.35E+12 1.12E+14 3.45E+13 185.0 Pu240 5.36E+11 4.39E+13 1.53E+13 195.2 Pu241 1.39E+12 3.54E+16 5.10E+15 200.0 Pu242 2.05E+05 1.03E+12 1.71E+11 200.0 Pu243 8.60E+08 7.87E+16 1.36E+16 200.0 Pu244 8.65E+02 1.12E+06 6.09E+05 199.7 Pu245 1.87E+04 3.81E+12 6.08E+11 200.0 Ra222 1.52E+03 2.06E+09 1.72E+08 200.0 Ra223 9.59E+03 2.60E+09 1.97E+08 200.0 Ra224 3.60E+06 8.74E+12 9.51E+11 200.0 Ra225 2.18E+03 3.35E+10 3.35E+09 200.0 Ra226 1.25E+03 4.28E+06 5.19E+05 199.9 Ra227 3.97E+04 1.20E+09 9.68E+07 200.0 Ra228 1.01E+09 1.35E+09 1.18E+09 28.1 Rb100 3.14E+11 2.46E+14 1.50E+14 199.5 Rb101 6.00E+09 4.87E+10 2.11E+10 156.2 Rb102 5.88E+07 5.00E+09 1.73E+09 195.3 Rb81 1.25E+03 3.78E+04 1.08E+04 187.3 Rb83 9.80E+06 2.73E+08 9.75E+07 186.1 Rb84 2.42E+08 3.68E+10 9.24E+09 197.4 Rb86 8.59E+12 8.83E+14 1.84E+14 196.1 Rb86m 1.11E+12 1.08E+14 2.27E+13 195.9 Rb87 4.18E+05 7.12E+06 2.34E+06 177.8 Rb88 1.02E+16 2.94E+16 2.01E+16 96.8 Rb89 1.29E+16 3.42E+16 2.59E+16 90.3 Rb90 1.14E+16 2.96E+16 2.52E+16 89.0 Rb90m 4.09E+15 9.94E+15 5.60E+15 83.3 Rb91 1.65E+16 3.44E+16 3.01E+16 70.1 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.37

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Rb92 1.60E+16 2.98E+16 2.61E+16 60.1 Rb93 1.27E+16 2.32E+16 1.96E+16 58.9 Rb94 7.33E+15 1.19E+16 9.76E+15 47.2 Rb95 3.59E+15 5.69E+15 4.75E+15 45.3 Rb96 7.46E+14 2.12E+15 1.39E+15 96.0 Rb97 1.35E+14 2.94E+14 2.18E+14 74.1 Rb98 1.95E+13 1.07E+14 5.47E+13 138.5 Rb99 4.65E+11 1.60E+13 5.78E+12 188.7 Rh101 6.10E+04 1.44E+06 4.03E+05 183.8 Rh101m 2.30E+04 6.15E+06 1.09E+06 198.5 Rh102 1.82E+09 3.09E+11 6.78E+10 197.7 Rh102m 4.86E+08 8.43E+10 1.82E+10 197.7 Rh103m 2.22E+16 5.98E+16 3.13E+16 91.8 Rh104 4.40E+14 6.19E+16 1.51E+16 197.2 Rh104m 3.48E+13 4.70E+15 1.15E+15 197.1 Rh105 8.97E+15 4.24E+16 1.82E+16 130.2 Rh105m 2.64E+15 1.39E+16 5.46E+15 136.1 Rh106 3.38E+15 4.74E+16 1.32E+16 173.3 Rh106m 1.65E+11 7.05E+14 7.66E+13 199.9 Rh107 2.69E+15 3.17E+16 9.80E+15 168.7 Rh108 1.29E+15 2.23E+16 6.41E+15 178.2 Rh108m 5.20E+12 1.38E+14 4.26E+13 185.5 Rh109 7.47E+14 1.54E+16 4.32E+15 181.5 Rh110 4.58E+12 2.29E+14 5.63E+13 192.2 Rh110m 4.61E+14 7.53E+15 2.13E+15 176.9 Rh111 2.83E+14 3.82E+15 1.09E+15 172.3 Rh112 1.61E+14 1.64E+15 5.18E+14 164.2 Rh113 1.32E+14 1.02E+15 3.55E+14 154.3 Rh114 7.16E+13 4.87E+14 1.91E+14 148.7 Rh115 3.66E+13 2.02E+14 9.82E+13 138.8 Rh116 1.03E+13 7.50E+13 3.95E+13 151.9 Rh117 3.95E+12 2.67E+13 1.45E+13 148.5 Rh118 4.19E+11 1.21E+13 4.80E+12 186.5 Rh119 4.59E+10 3.09E+12 1.14E+12 194.1 Rh120 2.05E+10 6.92E+11 2.48E+11 188.5 Rh121 1.06E+09 1.13E+11 3.96E+10 196.3 Rh122 1.22E+08 1.42E+10 5.04E+09 196.6 Rn217 1.39E+06 2.19E+06 1.79E+06 44.6 Rn218 1.52E+03 2.06E+09 1.72E+08 200.0 Rn219 9.59E+03 2.60E+09 1.97E+08 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.38

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Rn220 3.60E+06 8.74E+12 9.51E+11 200.0 Rn222 1.23E+03 4.28E+06 5.19E+05 199.9 Ru103 2.24E+16 6.05E+16 3.16E+16 91.9 Ru105 9.31E+15 4.90E+16 1.93E+16 136.1 Ru106 3.38E+15 4.07E+16 1.25E+16 169.3 Ru107 2.67E+15 3.13E+16 9.66E+15 168.5 Ru108 1.28E+15 2.22E+16 6.36E+15 178.1 Ru109 7.33E+14 1.46E+16 4.13E+15 180.9 Ru110 4.56E+14 7.30E+15 2.07E+15 176.5 Ru111 2.62E+14 3.10E+15 9.14E+14 168.8 Ru112 1.25E+14 1.05E+15 3.62E+14 157.5 Ru113 6.29E+13 4.68E+14 1.72E+14 152.6 Ru114 2.10E+13 1.58E+14 6.36E+13 153.2 Ru115 3.05E+12 3.40E+13 1.51E+13 167.1 Ru116 3.47E+11 5.77E+12 2.90E+12 177.3 Ru117 3.15E+10 1.03E+12 4.16E+11 188.1 Ru118 2.64E+09 2.22E+11 7.91E+10 195.3 Ru119 2.02E+08 2.21E+10 7.85E+09 196.4 Ru120 2.02E+07 2.14E+09 7.56E+08 196.3 Sb118 7.83E+02 3.94E+04 8.89E+03 192.2 Sb118m 1.24E+03 6.42E+04 1.42E+04 192.5 Sb119 2.93E+05 1.71E+07 3.62E+06 193.3 Sb120 2.38E+07 4.04E+09 1.02E+09 197.7 Sb120m 2.14E+07 1.89E+09 5.20E+08 195.5 Sb122 2.39E+12 1.73E+14 3.74E+13 194.5 Sb122m 1.62E+11 1.11E+13 2.42E+12 194.2 Sb124 1.69E+12 2.15E+14 4.09E+13 196.9 Sb124m 7.95E+10 2.97E+12 7.44E+11 189.6 Sb125 1.26E+14 9.19E+14 3.80E+14 151.7 Sb126 5.78E+12 1.99E+13 9.54E+12 110.0 Sb126m 9.10E+12 1.55E+13 1.18E+13 51.8 Sb127 1.23E+15 4.11E+15 2.03E+15 107.7 Sb128 1.62E+14 7.04E+14 3.13E+14 125.1 Sb128m 2.42E+15 5.70E+15 3.29E+15 80.7 Sb129 4.02E+15 1.16E+16 6.16E+15 97.1 Sb129m 7.52E+13 1.39E+14 9.15E+13 59.3 Sb130 4.95E+15 7.78E+15 5.74E+15 44.5 Sb130m 6.75E+15 9.67E+15 7.47E+15 35.5 Sb131 1.65E+16 2.24E+16 1.81E+16 30.5 Sb132 1.20E+16 1.82E+16 1.33E+16 41.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.39

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sb132m 5.84E+15 7.61E+15 6.57E+15 26.2 Sb133 1.19E+16 1.45E+16 1.31E+16 19.6 Sb134 2.06E+15 3.09E+15 2.52E+15 40.2 Sb134m 2.02E+15 3.00E+15 2.50E+15 38.9 Sb135 8.03E+14 1.71E+15 1.16E+15 72.4 Sb136 6.91E+13 2.71E+14 1.41E+14 118.8 Sb137 1.15E+13 4.31E+14 2.76E+14 189.6 Sb138 2.80E+11 4.01E+12 1.58E+12 173.9 Sb139 1.15E+10 2.87E+11 1.08E+11 184.5 Se75 4.31E+04 5.57E+07 7.49E+06 199.7 Se77m 1.55E+11 3.30E+12 4.70E+11 182.0 Se79 1.43E+09 3.42E+10 8.83E+09 183.9 Se79m 2.71E+14 7.41E+14 3.51E+14 92.8 Se81 1.15E+15 2.22E+15 1.38E+15 63.5 Se81m 8.60E+13 2.28E+14 1.23E+14 90.5 Se83 1.85E+15 4.82E+15 2.90E+15 89.1 Se83m 2.18E+14 5.28E+14 2.87E+14 83.1 Se84 3.15E+15 8.21E+15 5.65E+15 88.9 Se85 3.08E+15 7.19E+15 5.87E+15 80.1 Se86 2.53E+15 8.44E+15 6.76E+15 107.7 Se87 1.32E+15 4.92E+15 3.81E+15 115.2 Se88 4.64E+14 2.44E+15 1.74E+15 136.1 Se89 1.05E+14 5.84E+14 3.44E+14 138.9 Se90 3.90E+13 1.62E+14 9.43E+13 122.2 Se91 1.83E+12 2.03E+13 8.72E+12 166.9 Se92 1.44E+11 1.99E+12 8.04E+11 173.1 Se93 1.46E+10 1.02E+11 5.15E+10 149.7 Se94 2.96E+08 6.11E+09 2.49E+09 181.5 Sm145 8.34E+03 1.76E+07 3.66E+06 199.8 Sm146 1.25E+02 1.93E+04 4.92E+03 197.4 Sm147 3.70E+04 7.31E+05 2.75E+05 180.7 Sm151 1.17E+13 1.76E+14 6.30E+13 174.9 Sm153 1.28E+15 3.68E+16 8.89E+15 186.5 Sm155 3.03E+14 2.40E+15 7.60E+14 155.1 Sm156 1.67E+14 1.19E+15 4.17E+14 150.9 Sm157 8.27E+13 8.10E+14 2.54E+14 163.0 Sm158 4.30E+13 4.78E+14 1.42E+14 167.0 Sm159 1.44E+13 2.13E+14 5.91E+13 174.6 Sm160 4.28E+12 7.67E+13 2.03E+13 178.9 Sm161 7.92E+11 1.98E+13 4.93E+12 184.6 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.40

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sm162 8.10E+10 1.51E+12 4.72E+11 179.7 Sm163 6.23E+09 1.95E+11 6.14E+10 187.6 Sm164 4.23E+08 1.61E+10 7.28E+09 189.7 Sm165 2.74E+07 1.83E+09 7.12E+08 194.1 Sn113 1.07E+03 3.70E+05 1.24E+05 198.8 Sn113m 1.45E+03 2.90E+05 8.97E+04 198.0 Sn117m 2.22E+11 1.58E+12 5.64E+11 150.5 Sn119m 3.33E+12 8.72E+12 5.42E+12 89.6 Sn121 9.43E+13 2.66E+14 1.71E+14 95.2 Sn121m 3.32E+11 4.89E+12 1.75E+12 174.6 Sn123 1.24E+13 8.82E+13 3.38E+13 150.6 Sn123m 1.08E+14 3.62E+14 1.97E+14 108.0 Sn125 1.14E+14 4.44E+14 2.14E+14 118.1 Sn125m 1.56E+14 4.90E+14 2.66E+14 103.6 Sn126 2.83E+09 9.57E+10 2.63E+10 188.5 Sn127 7.54E+14 2.51E+15 1.23E+15 107.6 Sn127m 4.26E+14 1.44E+15 7.23E+14 108.8 Sn128 2.35E+15 5.24E+15 3.12E+15 76.2 Sn128m 1.12E+15 2.58E+15 1.51E+15 78.6 Sn129 2.12E+15 6.24E+15 3.38E+15 98.6 Sn129m 1.33E+15 2.44E+15 1.61E+15 59.3 Sn130 3.75E+15 4.59E+15 4.06E+15 20.2 Sn130m 3.71E+15 4.71E+15 4.08E+15 23.8 Sn131 2.81E+15 3.70E+15 3.24E+15 27.3 Sn131m 2.69E+15 3.54E+15 3.11E+15 27.2 Sn132 3.05E+15 4.49E+15 3.93E+15 38.4 Sn133 5.36E+14 1.13E+15 7.77E+14 71.5 Sn134 7.05E+13 2.45E+14 1.41E+14 110.8 Sn135 3.73E+12 2.04E+13 9.32E+12 138.1 Sn136 1.04E+11 1.33E+12 5.22E+11 171.1 Sn137 7.14E+09 1.32E+11 8.71E+10 179.4 Sr100 1.14E+14 4.54E+14 2.66E+14 119.8 Sr101 1.15E+13 6.33E+13 3.37E+13 138.7 Sr102 7.87E+11 6.44E+12 2.72E+12 156.4 Sr103 1.48E+10 1.56E+11 6.66E+10 165.3 Sr104 9.30E+08 1.91E+10 6.98E+09 181.4 Sr105 3.68E+07 3.61E+09 1.25E+09 196.0 Sr83 2.21E+02 1.94E+03 9.67E+02 159.1 Sr85 6.54E+06 8.09E+08 1.54E+08 196.8 Sr85m 3.60E+06 2.43E+08 4.84E+07 194.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.41

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sr87m 9.32E+09 1.89E+13 2.86E+12 199.8 Sr89 1.31E+16 3.46E+16 2.61E+16 90.4 Sr90 1.51E+15 1.72E+16 6.82E+15 167.8 Sr91 1.89E+16 3.89E+16 3.24E+16 69.3 Sr92 2.27E+16 4.06E+16 3.40E+16 56.4 Sr93 2.81E+16 4.40E+16 3.69E+16 44.2 Sr94 2.85E+16 4.02E+16 3.56E+16 33.9 Sr95 2.54E+16 3.37E+16 3.08E+16 28.0 Sr96 1.90E+16 2.59E+16 2.24E+16 31.0 Sr97 8.37E+15 1.26E+16 1.04E+16 40.1 Sr98 3.41E+15 6.35E+15 4.94E+15 60.2 Sr99 5.83E+14 1.63E+15 1.03E+15 94.7 Tb155 1.85E+03 4.01E+05 5.23E+04 198.2 Tb156 9.95E+04 5.92E+06 1.23E+06 193.4 Tb156m 9.53E+03 5.46E+05 1.11E+05 193.1 Tb157 2.66E+04 1.86E+06 4.01E+05 194.4 Tb158 7.57E+05 3.49E+07 8.60E+06 191.5 Tb158m 9.43E+06 3.80E+09 5.13E+08 199.0 Tb160 7.01E+11 5.53E+14 8.38E+13 199.5 Tb161 3.19E+12 2.52E+14 4.07E+13 195.0 Tb162 1.35E+12 2.89E+13 7.83E+12 182.2 Tb163 5.85E+11 1.61E+13 3.94E+12 186.0 Tb164 2.07E+11 7.05E+12 1.67E+12 188.6 Tb165 8.21E+10 2.94E+12 7.02E+11 189.1 Tb166 3.37E+10 1.07E+12 2.89E+11 187.8 Tb167 6.69E+09 3.74E+11 9.30E+10 193.0 Tb168 1.35E+09 1.28E+11 3.28E+10 195.8 Tb169 3.42E+08 2.88E+10 9.73E+09 195.3 Tb170 6.94E+07 5.10E+09 2.39E+09 194.6 Tb171 1.48E+07 1.37E+09 5.37E+08 195.7 Tc100 1.05E+15 4.20E+16 1.18E+16 190.2 Tc101 3.45E+16 5.16E+16 3.83E+16 39.6 Tc102 2.93E+16 5.17E+16 3.45E+16 55.2 Tc102m 5.26E+13 1.91E+14 9.71E+13 113.6 Tc103 2.24E+16 5.70E+16 3.08E+16 87.2 Tc104 1.50E+16 5.29E+16 2.43E+16 111.5 Tc105 9.26E+15 4.74E+16 1.87E+16 134.6 Tc106 4.92E+15 3.83E+16 1.30E+16 154.5 Tc107 2.46E+15 2.56E+16 7.93E+15 165.0 Tc108 9.17E+14 1.25E+16 3.54E+15 172.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.42

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Tc109 4.61E+14 5.70E+15 1.63E+15 170.1 Tc110 1.26E+14 9.80E+14 3.47E+14 154.5 Tc111 3.41E+13 1.62E+14 8.39E+13 130.5 Tc112 4.97E+12 4.07E+13 1.91E+13 156.4 Tc113 1.00E+12 1.18E+13 4.87E+12 168.8 Tc114 2.20E+11 2.70E+12 1.27E+12 169.9 Tc115 9.35E+09 4.35E+11 2.13E+11 191.6 Tc116 6.36E+08 4.47E+10 1.76E+10 194.4 Tc117 2.34E+07 1.88E+09 6.82E+08 195.1 Tc118 3.97E+06 4.55E+08 1.57E+08 196.5 Tc97m 1.03E+06 5.09E+07 1.35E+07 192.1 Tc98 1.04E+04 9.50E+05 2.12E+05 195.7 Tc99 2.51E+11 2.80E+12 1.11E+12 167.1 Tc99m 3.50E+16 4.70E+16 3.83E+16 29.2 Te121 3.08E+05 8.19E+08 9.37E+07 199.8 Te121m 1.53E+05 3.98E+08 4.54E+07 199.8 Te123m 3.07E+09 1.03E+13 1.37E+12 199.9 Te125m 2.62E+13 2.12E+14 8.64E+13 156.0 Te127 1.22E+15 3.95E+15 1.95E+15 105.3 Te127m 1.40E+14 5.90E+14 2.45E+14 123.2 Te129 3.84E+15 1.12E+16 5.90E+15 98.2 Te129m 6.72E+14 1.98E+15 1.02E+15 98.5 Te131 1.72E+16 2.53E+16 1.95E+16 37.8 Te131m 2.75E+15 8.60E+15 4.34E+15 103.0 Te132 2.87E+16 4.28E+16 3.24E+16 39.3 Te133 2.16E+16 2.93E+16 2.37E+16 30.3 Te133m 2.11E+16 2.89E+16 2.42E+16 31.2 Te134 3.84E+16 4.79E+16 4.31E+16 21.8 Te135 1.99E+16 2.44E+16 2.14E+16 20.1 Te136 7.42E+15 9.94E+15 8.51E+15 29.0 Te137 1.79E+15 3.30E+15 2.56E+15 59.6 Te138 3.35E+14 8.69E+14 5.36E+14 88.7 Te139 3.90E+13 1.41E+14 7.53E+13 113.5 Te140 4.91E+12 1.01E+14 6.82E+13 181.4 Te141 2.46E+11 4.98E+12 1.84E+12 181.1 Te142 1.29E+10 9.18E+10 3.94E+10 150.7 Th226 1.52E+03 2.06E+09 1.72E+08 200.0 Th227 9.94E+03 2.59E+09 1.96E+08 200.0 Th228 3.69E+06 8.71E+12 9.47E+11 200.0 Th229 1.60E+03 5.04E+09 4.25E+08 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.43

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Th230 2.47E+05 8.07E+08 8.79E+07 199.9 Th231 3.72E+09 7.88E+14 6.47E+13 200.0 Th232 4.23E+01 1.64E+09 1.09E+09 200.0 Th233 5.50E+08 4.62E+17 5.56E+16 200.0 Th234 5.50E+09 1.10E+10 9.58E+09 66.7 Tl207 9.56E+03 2.59E+09 1.97E+08 200.0 Tl208 1.29E+06 3.14E+12 3.42E+11 200.0 Tl209 6.65E+02 6.87E+08 1.25E+08 200.0 Tm167 5.64E+02 3.53E+04 9.59E+03 193.7 Tm168 1.40E+03 1.07E+07 1.20E+06 199.9 Tm170 1.95E+08 3.43E+12 2.66E+11 200.0 Tm171 1.36E+08 2.14E+11 1.92E+10 199.7 Tm172 1.53E+08 2.30E+11 2.30E+10 199.7 U230 1.22E+03 1.86E+08 2.46E+07 200.0 U231 4.62E+03 1.22E+10 9.13E+08 200.0 U232 2.19E+07 2.08E+13 1.95E+12 200.0 U233 1.45E+05 1.09E+13 8.55E+11 200.0 U234 6.10E+09 1.94E+12 3.59E+11 198.7 U235 3.92E+05 1.46E+10 6.26E+09 200.0 U235m 2.42E+14 3.17E+16 4.44E+15 197.0 U236 8.20E+09 4.84E+10 2.32E+10 142.0 U237 4.50E+15 5.28E+16 1.65E+16 168.6 U238 5.51E+09 1.10E+10 9.58E+09 66.6 U239 1.37E+17 8.52E+17 4.24E+17 144.6 U240 8.64E+02 1.12E+06 6.08E+05 199.7 Xe125 4.23E+03 1.77E+05 6.50E+04 190.7 Xe125m 1.12E+03 4.69E+04 1.72E+04 190.7 Xe127 3.11E+06 2.58E+10 3.26E+09 200.0 Xe127m 5.38E+05 4.34E+09 5.50E+08 200.0 Xe129m 6.34E+09 2.35E+13 3.39E+12 199.9 Xe131m 2.12E+14 3.86E+14 2.53E+14 58.2 Xe133 4.13E+16 5.94E+16 4.63E+16 36.0 Xe133m 1.16E+15 2.05E+15 1.41E+15 55.4 Xe134m 2.29E+14 1.22E+15 5.82E+14 136.9 Xe135 7.08E+15 5.95E+16 3.28E+16 157.5 Xe135m 7.92E+15 1.43E+16 9.78E+15 57.6 Xe137 3.83E+16 5.47E+16 4.30E+16 35.5 Xe138 3.62E+16 4.82E+16 4.11E+16 28.4 Xe139 2.69E+16 3.31E+16 3.07E+16 20.7 Xe140 1.77E+16 2.32E+16 2.08E+16 26.6 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.44

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Xe141 5.85E+15 8.91E+15 7.57E+15 41.5 Xe142 1.92E+15 3.63E+15 2.80E+15 61.6 Xe143 2.12E+14 5.16E+14 3.33E+14 83.6 Xe144 3.51E+13 1.35E+14 6.87E+13 117.1 Xe145 6.08E+11 1.25E+13 5.10E+12 181.4 Xe146 7.26E+10 9.01E+11 3.64E+11 170.2 Xe147 6.97E+09 2.49E+10 1.45E+10 112.4 Y100 3.69E+15 6.46E+15 4.90E+15 54.6 Y101 1.21E+15 2.77E+15 1.93E+15 78.6 Y102 2.81E+14 1.72E+15 1.27E+15 143.9 Y103 1.75E+13 1.35E+14 5.95E+13 154.1 Y104 2.70E+12 1.73E+13 7.76E+12 145.9 Y105 3.66E+10 3.18E+12 1.12E+12 195.5 Y106 1.25E+08 8.28E+09 3.27E+09 194.0 Y107 1.05E+07 1.19E+09 4.05E+08 196.5 Y108 5.56E+04 1.58E+07 2.83E+06 198.6 Y87 6.44E+05 1.60E+07 6.03E+06 184.6 Y87m 3.35E+03 3.61E+04 1.62E+04 166.0 Y88 5.94E+08 8.51E+10 2.20E+10 197.2 Y89m 1.30E+12 3.39E+12 2.54E+12 89.0 Y90 1.54E+15 1.77E+16 7.07E+15 168.0 Y90m 1.20E+11 2.94E+12 9.91E+11 184.3 Y91 1.90E+16 3.88E+16 3.26E+16 68.5 Y91m 1.11E+16 2.29E+16 1.91E+16 69.3 Y92 2.30E+16 4.13E+16 3.45E+16 57.1 Y93 2.91E+16 4.64E+16 3.79E+16 45.9 Y93m 1.01E+16 1.62E+16 1.31E+16 46.6 Y94 3.28E+16 4.76E+16 3.92E+16 36.9 Y95 3.50E+16 4.68E+16 3.98E+16 28.9 Y96 2.09E+16 2.74E+16 2.42E+16 26.8 Y96m 1.13E+16 1.88E+16 1.35E+16 50.2 Y97 1.80E+16 2.08E+16 1.94E+16 14.4 Y97m 1.09E+16 1.39E+16 1.20E+16 24.4 Y98 1.20E+16 1.46E+16 1.32E+16 20.0 Y98m 7.53E+15 1.06E+16 8.19E+15 33.9 Y99 1.23E+16 1.56E+16 1.40E+16 23.5 Yb169 3.54E+02 1.16E+07 1.12E+06 200.0 Yb169m 7.10E+04 7.10E+04 7.10E+04 0.0 Zn69 1.36E+10 1.61E+11 5.69E+10 168.9 Zn69m 4.97E+07 1.85E+09 3.23E+08 189.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.45

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Activity Maximum Activity Average Activity Percent Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Zn71 6.89E+10 9.09E+11 2.65E+11 171.8 Zn71m 5.92E+09 1.26E+11 3.04E+10 182.1 Zn72 2.32E+11 2.34E+12 7.07E+11 163.9 Zn73 7.52E+11 5.69E+12 1.75E+12 153.3 Zn74 2.20E+12 1.21E+13 4.01E+12 138.5 Zn75 4.73E+12 2.56E+13 8.58E+12 137.7 Zn76 6.15E+12 3.26E+13 1.61E+13 136.6 Zn77 4.97E+12 2.80E+13 1.93E+13 139.6 Zn78 4.01E+12 2.75E+13 1.89E+13 149.1 Zn79 1.31E+12 1.25E+13 8.49E+12 162.1 Zn80 3.46E+11 2.54E+12 1.59E+12 152.1 Zn81 6.13E+09 3.81E+11 1.54E+11 193.7 Zn82 9.21E+09 1.34E+11 7.25E+10 174.3 Zn83 3.25E+08 9.08E+09 3.37E+09 186.2 Zr100 3.50E+16 4.52E+16 3.75E+16 25.3 Zr101 1.91E+16 2.53E+16 2.10E+16 28.0 Zr102 1.12E+16 1.50E+16 1.33E+16 28.8 Zr103 3.07E+15 5.24E+15 3.95E+15 52.4 Zr104 5.49E+14 1.39E+15 9.11E+14 87.1 Zr105 1.49E+14 6.89E+14 5.14E+14 128.8 Zr106 2.68E+11 1.15E+13 3.27E+12 190.9 Zr107 1.54E+10 4.63E+11 2.20E+11 187.1 Zr108 6.82E+08 2.90E+10 1.32E+10 190.8 Zr109 1.12E+09 4.45E+09 2.77E+09 119.4 Zr110 9.76E+06 1.80E+08 8.90E+07 179.4 Zr88 8.06E+02 2.73E+04 7.94E+03 188.5 Zr89 1.62E+07 9.93E+09 1.47E+09 199.3 Zr89m 5.82E+05 9.38E+08 1.12E+08 199.8 Zr90m 5.84E+09 1.06E+12 2.33E+11 197.8 Zr93 3.36E+10 4.42E+11 1.78E+11 171.8 Zr95 3.60E+16 4.88E+16 4.08E+16 30.2 Zr97 3.61E+16 5.07E+16 4.02E+16 33.5 Zr98 3.61E+16 4.69E+16 3.91E+16 26.2 Zr99 3.57E+16 4.65E+16 3.94E+16 26.3 Total 4.13E+18 7.91E+18 5.25E+18 62.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.46

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table C.6. Minimum, Maximum and Average Radionuclide Activity for Radionuclides with Potential Mobility at Time 0 (End of Operation) for Small- to Medium-Sized Advanced Reactors Percent Minimum Activity Maximum Activity Average Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Kr100 6.71E+07 3.75E+10 8.22E+09 199.3 Kr79 1.87E+04 4.64E+06 9.02E+05 198.4 Kr79m 9.34E+03 2.35E+06 5.63E+05 198.4 Kr81 6.65E+03 3.57E+05 1.17E+05 192.7 Kr81m 1.54E+08 4.65E+10 1.22E+10 198.7 Kr83m 1.73E+15 3.58E+16 1.08E+16 181.6 Kr85 4.08E+14 2.18E+15 1.20E+15 137.0 Kr85m 3.87E+15 7.81E+16 2.29E+16 181.1 Kr87 7.48E+15 1.45E+17 4.22E+16 180.4 Kr88 1.01E+16 1.95E+17 5.59E+16 180.4 Kr89 1.26E+16 2.39E+17 6.76E+16 180.0 Kr90 1.24E+16 2.37E+17 6.64E+16 180.1 Kr91 8.06E+15 1.63E+17 4.64E+16 181.2 Kr92 4.23E+15 8.31E+16 2.46E+16 180.6 Kr93 1.22E+15 2.41E+16 7.88E+15 180.7 Kr94 3.42E+14 6.31E+15 2.33E+15 179.4 Kr95 2.92E+13 9.50E+14 3.07E+14 188.1 Kr96 6.12E+13 1.67E+15 4.35E+14 185.8 Kr97 1.10E+12 1.01E+14 2.59E+13 195.7 Kr98 7.81E+10 5.46E+13 1.16E+13 199.4 Kr99 1.86E+08 1.19E+11 3.02E+10 199.4 Rn217 1.10E+06 1.10E+06 1.10E+06 0.0 Rn218 1.13E+04 3.83E+08 3.50E+07 200.0 Rn219 1.73E+05 9.97E+09 1.66E+09 200.0 Rn220 2.47E+07 2.05E+13 1.71E+12 200.0 Rn222 2.80E+05 6.75E+05 4.78E+05 82.8 Xe125 7.80E+04 2.17E+07 5.57E+06 198.6 Xe125m 2.06E+04 5.75E+06 2.88E+06 198.6 Xe127 2.63E+07 3.80E+11 4.83E+10 200.0 Xe127m 4.41E+06 6.92E+10 9.03E+09 200.0 Xe129m 5.74E+10 4.49E+13 9.09E+12 199.5 Xe131m 1.36E+14 3.50E+15 1.13E+15 185.0 Xe133 2.50E+16 6.77E+17 2.07E+17 185.8 Xe133m 7.65E+14 2.14E+16 6.72E+15 186.2 Noble Gases Xe134m 2.95E+14 1.39E+16 3.74E+15 191.7 Xe135 7.01E+15 3.44E+17 8.74E+16 192.0 Xe135m 5.25E+15 1.58E+17 4.70E+16 187.1 Xe137 2.30E+16 6.09E+17 1.85E+17 185.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.47

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Percent Minimum Activity Maximum Activity Average Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Xe138 2.22E+16 5.68E+17 1.70E+17 185.0 Xe139 1.69E+16 4.05E+17 1.19E+17 183.9 Xe140 1.20E+16 2.59E+17 7.57E+16 182.2 Xe141 4.90E+15 9.33E+16 2.87E+16 180.0 Xe142 1.79E+15 3.23E+16 1.04E+16 179.0 Xe143 1.70E+14 3.56E+15 1.36E+15 181.7 Xe144 2.56E+13 7.85E+14 2.87E+14 187.4 Xe145 5.86E+11 7.53E+13 2.38E+13 196.9 Xe146 5.43E+10 5.36E+12 1.59E+12 196.0 Xe147 7.07E+09 1.40E+11 5.56E+10 180.7 Cs131 3.62E+06 1.27E+09 3.31E+08 198.9 Cs132 1.93E+10 2.46E+13 4.72E+12 199.7 Cs134 1.14E+15 8.37E+16 2.56E+16 194.6 Cs134m 1.61E+14 4.31E+16 9.74E+15 198.5 Cs135 3.27E+10 3.33E+11 1.13E+11 164.2 Cs135m 7.79E+12 9.82E+14 2.49E+14 196.8 Cs136 8.85E+14 5.47E+16 1.33E+16 193.6 Cs136m 7.44E+13 3.61E+15 1.09E+15 191.9 Cs137 4.48E+15 2.29E+16 1.38E+16 134.5 Cs138 2.43E+16 6.26E+17 1.89E+17 185.1 Cs138m 1.15E+15 4.47E+16 1.28E+16 190.0 Cs139 2.28E+16 5.80E+17 1.74E+17 184.9 Cs140 1.99E+16 4.77E+17 1.43E+17 184.0 Cs141 1.56E+16 3.76E+17 1.14E+17 184.1 Cs142 9.72E+15 2.24E+17 6.65E+16 183.4 Cs143 5.41E+15 1.13E+17 3.41E+16 181.8 Cs144 1.78E+15 3.46E+16 1.14E+16 180.4 Cs145 3.24E+14 6.80E+15 2.53E+15 181.8 Cs146 3.44E+13 8.80E+14 3.70E+14 185.0 Cs147 8.74E+12 1.55E+14 5.62E+13 178.6 Cs148 9.05E+10 1.35E+13 4.04E+12 197.3 Cs149 2.29E+09 2.48E+11 7.95E+10 196.3 Cs150 5.94E+07 2.62E+10 6.89E+09 199.1 Cs151 6.85E+06 4.59E+09 1.15E+09 199.4 H3 1.80E+13 1.85E+16 2.92E+15 199.6 I123 6.01E+04 7.56E+06 2.04E+06 196.8 I125 1.45E+04 4.81E+07 7.28E+06 199.9 I126 6.47E+08 1.39E+12 2.13E+11 199.8 I128 7.90E+13 1.15E+16 2.84E+15 197.3 Volatile I129 1.18E+09 6.47E+09 3.49E+09 138.3 I130 1.65E+14 1.73E+16 5.11E+15 196.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.48

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Percent Minimum Activity Maximum Activity Average Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

I130m 1.05E+14 1.11E+16 3.29E+15 196.3 I131 1.24E+16 3.45E+17 1.05E+17 186.1 I132 1.83E+16 5.00E+17 1.52E+17 185.9 I132m 1.63E+14 9.18E+15 2.37E+15 193.0 I133 2.59E+16 6.70E+17 2.05E+17 185.1 I133m 2.20E+15 6.72E+16 2.01E+16 187.3 I134 2.91E+16 7.55E+17 2.28E+17 185.1 I134m 2.27E+15 9.16E+16 2.53E+16 190.3 I135 2.46E+16 6.47E+17 1.97E+17 185.4 I136 1.07E+16 2.42E+17 7.46E+16 183.0 I136m 5.38E+15 1.65E+17 4.82E+16 187.4 I137 1.19E+16 2.92E+17 8.83E+16 184.3 I138 6.62E+15 1.52E+17 4.76E+16 183.3 I139 3.12E+15 5.86E+16 1.81E+16 179.8 I140 6.43E+14 1.23E+16 4.30E+15 180.1 I141 1.58E+14 2.44E+15 9.10E+14 175.7 I142 2.49E+13 4.92E+14 1.77E+14 180.7 I143 1.47E+11 4.03E+13 1.06E+13 198.5 I144 8.53E+09 6.16E+11 2.06E+11 194.5 Te121 1.35E+06 1.26E+10 1.67E+09 200.0 Te121m 6.86E+05 5.48E+09 7.02E+08 199.9 Te123m 3.75E+10 1.97E+13 3.24E+12 199.2 Te125m 5.13E+13 5.41E+14 2.26E+14 165.3 Te127 1.04E+15 3.29E+16 1.00E+16 187.8 Te127m 1.47E+14 2.85E+15 8.10E+14 180.4 Te129 3.17E+15 1.06E+17 3.09E+16 188.4 Te129m 5.53E+14 1.79E+16 5.23E+15 188.0 Te131 1.09E+16 2.88E+17 8.86E+16 185.4 Te131m 2.15E+15 7.63E+16 2.20E+16 189.0 Te132 1.80E+16 4.74E+17 1.46E+17 185.4 Te133 1.34E+16 3.37E+17 1.03E+17 184.7 Te133m 1.28E+16 3.35E+17 9.97E+16 185.3 Te134 2.40E+16 5.80E+17 1.72E+17 184.1 Te135 1.25E+16 2.91E+17 8.80E+16 183.5 Te136 5.51E+15 1.03E+17 3.20E+16 179.6 Te137 1.68E+15 3.00E+16 9.70E+15 178.9 Te138 2.70E+14 5.09E+15 2.04E+15 179.8 Te139 2.89E+13 8.28E+14 3.21E+14 186.5 Te140 6.18E+12 5.88E+14 1.47E+14 195.8 Te141 1.72E+11 2.98E+13 7.90E+12 197.7 Te142 9.34E+09 5.43E+11 1.69E+11 193.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.49

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Percent Minimum Activity Maximum Activity Average Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ba131 3.02E+04 6.14E+06 1.79E+06 198.0 Ba133 9.39E+07 4.06E+10 6.45E+09 199.1 Ba135m 1.15E+12 1.83E+14 5.75E+13 197.5 Ba136m 9.83E+13 6.06E+15 1.48E+15 193.6 Ba137m 4.24E+15 2.19E+16 1.32E+16 135.0 Ba139 2.32E+16 5.94E+17 1.80E+17 185.0 Ba140 2.24E+16 5.67E+17 1.71E+17 184.8 Ba141 2.12E+16 5.29E+17 1.60E+17 184.6 Ba142 1.99E+16 4.96E+17 1.49E+17 184.6 Ba143 1.82E+16 4.34E+17 1.28E+17 183.9 Ba144 1.40E+16 3.21E+17 9.36E+16 183.3 Ba145 6.97E+15 1.41E+17 4.22E+16 181.1 Ba146 3.49E+15 5.98E+16 1.82E+16 177.9 Ba147 9.56E+14 1.55E+16 4.92E+15 176.7 Ba148 9.25E+13 1.97E+15 8.07E+14 182.0 Ba149 5.27E+12 2.86E+14 1.02E+14 192.8 Ba150 3.05E+11 3.06E+13 9.77E+12 196.1 Ba151 1.26E+10 5.15E+12 1.34E+12 199.0 Ba152 2.71E+08 8.18E+10 2.28E+10 198.7 Ba153 6.32E+06 3.94E+09 9.94E+08 199.4 Ru103 1.84E+16 6.33E+17 1.81E+17 188.7 Ru105 1.00E+16 5.33E+17 1.39E+17 192.6 Ru106 5.85E+15 2.18E+17 6.48E+16 189.5 Ru107 3.27E+15 3.37E+17 8.10E+16 196.2 Ru108 1.91E+15 2.35E+17 5.46E+16 196.8 Ru109 1.19E+15 1.55E+17 3.56E+16 197.0 Ru110 6.99E+14 7.57E+16 1.73E+16 196.3 Ru111 3.42E+14 3.19E+16 7.36E+15 195.8 Ru112 1.63E+14 1.10E+16 2.74E+15 194.2 Ru113 7.76E+13 5.01E+15 1.27E+15 193.9 Ru114 2.71E+13 1.73E+15 4.70E+14 193.9 Ru115 4.63E+12 3.64E+14 1.05E+14 195.0 Ru116 5.37E+11 4.43E+13 1.72E+13 195.2 Ru117 3.90E+10 6.52E+12 2.09E+12 197.6 Ru118 2.84E+09 1.33E+12 3.66E+11 199.1 Ru119 1.95E+08 1.31E+11 3.56E+10 199.4 Ru120 1.78E+07 1.26E+10 3.41E+09 199.4 Semi-Volatile Ru97 3.36E+04 3.36E+04 3.36E+04 0.0 Sr100 1.45E+14 2.59E+15 9.24E+14 178.8 Sr101 1.35E+13 3.57E+14 1.15E+14 185.5 Sr102 6.99E+11 3.77E+13 1.09E+13 192.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.50

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Percent Minimum Activity Maximum Activity Average Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sr103 1.18E+10 9.50E+11 3.08E+11 195.1 Sr104 6.62E+08 1.14E+11 3.01E+10 197.7 Sr105 3.29E+07 2.19E+10 5.50E+09 199.4 Sr83 1.21E+03 4.43E+04 1.78E+04 189.4 Sr85 1.07E+07 3.07E+10 3.90E+09 199.9 Sr85m 5.66E+06 8.87E+09 1.20E+09 199.7 Sr87m 4.74E+10 1.63E+14 2.01E+13 199.9 Sr89 1.35E+16 2.84E+17 7.87E+16 181.8 Sr90 3.55E+15 1.64E+16 9.58E+15 128.6 Sr91 1.74E+16 3.45E+17 1.00E+17 180.7 Sr92 1.84E+16 3.79E+17 1.11E+17 181.5 Sr93 2.01E+16 4.32E+17 1.28E+17 182.3 Sr94 1.94E+16 4.30E+17 1.26E+17 182.8 Sr95 1.74E+16 3.77E+17 1.12E+17 182.3 Sr96 1.35E+16 2.74E+17 8.23E+16 181.2 Sr97 6.55E+15 1.24E+17 3.79E+16 180.0 Sr98 3.28E+15 5.40E+16 1.74E+16 177.1 Sr99 5.64E+14 9.89E+15 3.84E+15 178.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.51

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table C.7. Minimum, Maximum, and Average Radionuclide Activity at Time 0 (End of Operation) for Small- to Medium-Sized Advanced Reactors Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ac225 2.58E+04 1.58E+10 1.31E+09 200.0 Ac226 5.09E+03 1.57E+08 2.24E+07 200.0 Ac227 2.21E+04 1.04E+10 2.09E+09 200.0 Ac228 6.88E+03 8.78E+10 9.76E+09 200.0 Ag106 4.54E+03 6.30E+05 2.10E+05 197.1 Ag106m 9.58E+03 8.79E+05 2.76E+05 195.7 Ag107m 4.24E+05 1.79E+08 3.52E+07 199.1 Ag108 7.57E+08 1.93E+12 2.78E+11 199.8 Ag108m 2.57E+06 1.95E+09 2.55E+08 199.5 Ag109m 1.32E+15 2.67E+17 5.44E+16 198.0 Ag110 1.18E+14 1.99E+17 3.40E+16 199.8 Ag110m 4.98E+12 1.51E+15 3.96E+14 198.7 Ag111 3.90E+14 4.41E+16 9.58E+15 196.5 Ag111m 3.89E+14 4.16E+16 9.24E+15 196.3 Ag112 2.15E+14 1.79E+16 4.02E+15 195.2 Ag113 1.05E+14 7.48E+15 1.74E+15 194.5 Ag113m 1.53E+14 1.09E+16 2.54E+15 194.5 Ag114 1.14E+14 6.92E+15 1.73E+15 193.5 Ag115 8.91E+13 4.36E+15 1.23E+15 192.0 Ag115m 3.51E+12 7.92E+13 2.48E+13 183.0 Ag116 9.42E+13 4.00E+15 1.16E+15 190.8 Ag116m 6.37E+12 2.72E+14 7.59E+13 190.8 Ag117 7.35E+13 2.92E+15 8.92E+14 190.2 Ag117m 1.46E+13 6.75E+14 1.92E+14 191.5 Ag118 5.13E+13 2.03E+15 6.62E+14 190.1 Ag118m 2.49E+13 9.54E+14 2.94E+14 189.8 Ag119 4.63E+13 1.65E+15 5.63E+14 189.1 Ag120 1.79E+13 5.86E+14 2.34E+14 188.2 Ag120m 5.66E+12 3.64E+14 1.25E+14 193.9 Ag121 1.32E+13 4.57E+14 1.80E+14 188.8 Ag122 1.72E+12 1.21E+14 3.91E+13 194.4 Ag122m 1.68E+12 1.08E+14 3.54E+13 193.9 Ag123 1.20E+12 1.09E+14 3.11E+13 195.6 Ag124 6.64E+11 4.81E+13 1.40E+13 194.6 Ag125 1.41E+10 1.08E+13 2.89E+12 199.5 Ag126 2.73E+09 2.11E+12 5.78E+11 199.5 Ag127 4.19E+08 3.23E+11 9.06E+10 199.5 Ag128 5.72E+07 4.42E+10 1.25E+10 199.5 Ag129 1.08E+07 7.82E+09 2.30E+09 199.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.52

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ag130 3.23E+08 2.17E+11 4.52E+10 199.4 Am239 2.61E+05 1.12E+08 2.33E+07 199.1 Am240 2.80E+08 1.51E+11 2.27E+10 199.3 Am241 4.78E+11 4.70E+13 8.48E+12 196.0 Am242 9.36E+12 2.76E+16 6.51E+15 199.9 Am242m 1.46E+10 1.66E+12 4.67E+11 196.5 Am243 8.04E+07 1.44E+13 3.78E+12 200.0 Am244 1.91E+09 2.98E+16 4.31E+15 200.0 Am244m 2.68E+10 4.48E+17 6.47E+16 200.0 Am245 5.16E+04 1.38E+13 1.75E+12 200.0 Am246 2.16E+04 3.49E+06 1.02E+06 197.5 Am246m 2.00E+04 3.22E+06 9.39E+05 197.5 Am247 1.69E+05 1.69E+05 1.69E+05 0.0 As72 1.89E+03 1.03E+05 3.98E+04 192.8 As73 1.34E+05 7.88E+06 2.10E+06 193.3 As74 7.34E+06 1.89E+09 4.35E+08 198.5 As76 2.25E+11 2.25E+13 7.33E+12 196.0 As77 3.52E+13 6.57E+14 2.21E+14 179.7 As78 8.72E+13 1.81E+15 5.82E+14 181.6 As79 1.87E+14 3.84E+15 1.26E+15 181.4 As80 4.20E+14 9.05E+15 2.74E+15 182.3 As81 6.54E+14 1.51E+16 4.56E+15 183.3 As82 7.43E+14 1.39E+16 3.96E+15 179.7 As82m 1.57E+14 4.83E+15 1.46E+15 187.4 As83 1.06E+15 1.86E+16 5.66E+15 178.4 As84 7.04E+14 1.18E+16 3.71E+15 177.4 As85 2.73E+14 8.47E+15 2.64E+15 187.5 As86 1.23E+14 1.99E+16 4.53E+15 197.6 As87 2.29E+13 1.61E+15 4.23E+14 194.4 As88 6.95E+12 4.24E+15 9.01E+14 199.3 As89 4.22E+11 2.27E+13 6.24E+12 192.7 As90 1.43E+10 6.29E+11 2.05E+11 191.1 As91 9.93E+08 1.04E+11 2.77E+10 196.2 As92 1.43E+07 8.75E+09 2.20E+09 199.3 At217 2.58E+04 1.58E+10 1.31E+09 200.0 Au200 1.78E+05 1.78E+05 1.78E+05 0.0 B12 4.21E+05 1.39E+09 2.78E+08 199.9 Ba131 3.02E+04 6.14E+06 1.79E+06 198.0 Ba133 9.39E+07 4.06E+10 6.45E+09 199.1 Ba135m 1.15E+12 1.83E+14 5.75E+13 197.5 Ba136m 9.83E+13 6.06E+15 1.48E+15 193.6 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.53

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ba137m 4.24E+15 2.19E+16 1.32E+16 135.0 Ba139 2.32E+16 5.94E+17 1.80E+17 185.0 Ba140 2.24E+16 5.67E+17 1.71E+17 184.8 Ba141 2.12E+16 5.29E+17 1.60E+17 184.6 Ba142 1.99E+16 4.96E+17 1.49E+17 184.6 Ba143 1.82E+16 4.34E+17 1.28E+17 183.9 Ba144 1.40E+16 3.21E+17 9.36E+16 183.3 Ba145 6.97E+15 1.41E+17 4.22E+16 181.1 Ba146 3.49E+15 5.98E+16 1.82E+16 177.9 Ba147 9.56E+14 1.55E+16 4.92E+15 176.7 Ba148 9.25E+13 1.97E+15 8.07E+14 182.0 Ba149 5.27E+12 2.86E+14 1.02E+14 192.8 Ba150 3.05E+11 3.06E+13 9.77E+12 196.1 Ba151 1.26E+10 5.15E+12 1.34E+12 199.0 Ba152 2.71E+08 8.18E+10 2.28E+10 198.7 Ba153 6.32E+06 3.94E+09 9.94E+08 199.4 Be10 2.29E+05 1.69E+11 4.51E+10 200.0 Be11 6.65E+05 1.41E+12 3.18E+11 200.0 Be8 4.57E+05 1.47E+17 3.60E+16 200.0 Bi206 9.33E+06 9.33E+06 9.33E+06 0.0 Bi207 6.24E+09 6.24E+09 6.24E+09 0.0 Bi208 5.71E+09 5.71E+09 5.71E+09 0.0 Bi210 6.46E+03 4.44E+15 8.88E+14 200.0 Bi210m 2.58E+09 2.58E+09 2.58E+09 0.0 Bi211 1.73E+05 9.56E+11 1.61E+11 200.0 Bi212 2.47E+07 2.05E+13 1.71E+12 200.0 Bi213 2.58E+04 1.58E+10 1.31E+09 200.0 Bi214 2.80E+05 6.75E+05 4.78E+05 82.8 Bk248 1.24E+05 1.90E+05 1.57E+05 42.3 Bk248m 6.86E+05 9.77E+06 3.42E+06 173.8 Bk249 6.86E+05 4.16E+11 6.57E+10 200.0 Bk250 2.84E+05 6.39E+12 9.06E+11 200.0 Bk251 9.86E+06 3.78E+09 6.70E+08 199.0 Br77 3.11E+04 2.05E+06 5.23E+05 194.0 Br77m 2.40E+04 1.59E+06 4.63E+05 194.0 Br78 3.28E+06 2.20E+08 5.80E+07 194.1 Br79m 1.85E+08 1.20E+10 3.05E+09 193.9 Br80 1.30E+10 9.44E+11 3.22E+11 194.6 Br80m 4.11E+09 3.26E+11 1.13E+11 195.0 Br82 3.21E+13 3.00E+15 8.91E+14 195.8 Br82m 2.89E+13 2.75E+15 8.16E+14 195.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.54

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Br83 1.73E+15 3.55E+16 1.07E+16 181.5 Br84 3.08E+15 5.93E+16 1.76E+16 180.2 Br84m 8.96E+13 3.55E+15 1.03E+15 190.1 Br85 3.84E+15 7.69E+16 2.24E+16 181.0 Br86 5.32E+15 1.02E+17 2.96E+16 180.3 Br87 6.04E+15 1.12E+17 3.23E+16 179.5 Br88 5.25E+15 9.28E+16 2.66E+16 178.6 Br89 3.78E+15 6.11E+16 1.88E+16 176.7 Br90 2.20E+15 3.39E+16 1.09E+16 175.7 Br91 3.79E+14 9.81E+15 3.36E+15 185.1 Br92 7.43E+13 1.88E+15 6.20E+14 184.8 Br93 1.86E+13 6.57E+14 2.49E+14 189.0 Br94 2.95E+12 1.76E+14 5.33E+13 193.4 Br95 1.21E+10 7.95E+11 2.46E+11 194.0 Br96 7.55E+09 2.37E+11 7.22E+10 187.6 Br97 7.98E+07 5.88E+09 2.16E+09 194.6 C14 2.55E+06 2.00E+11 2.78E+10 200.0 C15 5.75E+07 2.16E+11 3.95E+10 199.9 Cd107 4.15E+04 2.49E+07 4.73E+06 199.3 Cd109 4.04E+07 2.88E+11 3.87E+10 199.9 Cd111m 1.45E+10 7.09E+13 1.18E+13 199.9 Cd113m 7.09E+09 1.53E+13 1.88E+12 199.8 Cd115 8.94E+13 4.68E+15 1.30E+15 192.5 Cd115m 5.34E+12 2.76E+14 8.00E+13 192.4 Cd117 7.39E+13 3.02E+15 9.20E+14 190.4 Cd117m 1.82E+13 6.96E+14 2.15E+14 189.8 Cd118 7.58E+13 2.88E+15 9.33E+14 189.7 Cd119 4.93E+13 1.75E+15 5.75E+14 189.1 Cd119m 2.89E+13 1.01E+15 3.22E+14 188.9 Cd120 7.35E+13 2.55E+15 8.42E+14 188.8 Cd121 3.89E+13 1.34E+15 4.66E+14 188.7 Cd121m 2.47E+13 9.68E+14 2.98E+14 190.1 Cd122 6.53E+13 1.74E+15 6.29E+14 185.5 Cd123 4.37E+13 8.40E+14 3.41E+14 180.2 Cd124 1.82E+13 6.38E+14 2.09E+14 188.9 Cd125 4.89E+12 3.16E+14 9.56E+13 193.9 Cd126 2.08E+12 2.74E+14 7.91E+13 197.0 Cd127 9.33E+11 2.66E+14 6.35E+13 198.6 Cd128 3.19E+11 1.17E+14 2.70E+13 198.9 Cd129 2.64E+10 4.00E+12 1.25E+12 197.4 Cd130 3.59E+12 2.84E+15 5.89E+14 199.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.55

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Cd131 6.98E+11 5.23E+14 1.09E+14 199.5 Cd132 2.40E+09 9.94E+11 2.64E+11 199.0 Ce137 9.36E+04 7.66E+06 1.89E+06 195.2 Ce139 2.07E+10 2.56E+13 4.47E+12 199.7 Ce139m 7.65E+09 1.05E+13 2.08E+12 199.7 Ce141 2.13E+16 5.22E+17 1.60E+17 184.3 Ce143 2.03E+16 4.92E+17 1.48E+17 184.2 Ce144 1.86E+16 3.61E+17 1.18E+17 180.4 Ce145 1.38E+16 3.34E+17 1.00E+17 184.1 Ce146 1.10E+16 2.66E+17 8.08E+16 184.1 Ce147 7.68E+15 1.93E+17 5.96E+16 184.7 Ce148 5.82E+15 1.36E+17 4.10E+16 183.6 Ce149 3.29E+15 7.84E+16 2.43E+16 183.9 Ce150 1.83E+15 4.18E+16 1.30E+16 183.2 Ce151 4.81E+14 1.33E+16 4.03E+15 186.0 Ce152 1.04E+14 3.55E+15 1.08E+15 188.6 Ce153 1.10E+13 5.86E+14 1.94E+14 192.6 Ce154 8.67E+11 6.37E+13 2.26E+13 194.6 Ce155 4.81E+10 4.89E+12 1.92E+12 196.1 Ce156 2.31E+09 3.10E+11 1.14E+11 197.0 Ce157 7.42E+07 1.25E+10 4.32E+09 197.6 Cf248 5.71E+04 8.61E+05 4.95E+05 175.1 Cf249 1.60E+03 5.40E+07 1.27E+07 200.0 Cf250 1.69E+04 1.30E+10 2.43E+09 200.0 Cf251 9.10E+05 9.91E+07 2.13E+07 196.4 Cf252 8.75E+03 4.45E+10 8.49E+09 200.0 Cf253 1.63E+08 1.26E+10 2.48E+09 194.9 Cf254 3.26E+05 2.77E+07 5.43E+06 195.3 Cf255 0.00E+00 0.00E+00 0.00E+00 0 Cm240 9.93E+03 7.36E+05 3.81E+05 194.7 Cm241 6.81E+06 1.33E+10 4.72E+09 199.8 Cm242 5.95E+12 1.16E+16 3.54E+15 199.8 Cm243 2.22E+09 9.45E+12 2.51E+12 199.9 Cm244 1.61E+09 6.04E+15 1.25E+15 200.0 Cm245 4.99E+04 1.57E+12 2.41E+11 200.0 Cm246 6.44E+02 7.49E+11 1.24E+11 200.0 Cm247 2.74E+05 6.41E+06 1.62E+06 183.6 Cm248 2.39E+06 4.32E+07 1.09E+07 179.1 Cm249 2.26E+04 9.71E+12 1.24E+12 200.0 Cm251 3.29E+04 2.03E+08 3.58E+07 199.9 Co65 6.28E+05 2.13E+08 5.66E+07 198.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.56

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Co66 4.98E+08 5.87E+10 1.79E+10 196.6 Co67 1.44E+09 1.69E+11 4.74E+10 196.6 Co68 1.67E+09 2.04E+11 5.86E+10 196.8 Co69 1.96E+09 2.00E+11 5.84E+10 196.1 Co70 1.33E+09 1.40E+11 4.17E+10 196.2 Co71 7.77E+08 7.01E+10 2.17E+10 195.6 Co72 2.76E+08 2.59E+10 7.51E+09 195.8 Co73 9.68E+07 1.00E+10 2.79E+09 196.2 Co74 1.32E+07 1.55E+09 4.44E+08 196.6 Co75 1.67E+06 2.17E+08 6.25E+07 197.0 Cr66 2.83E+04 1.42E+07 3.83E+06 199.2 Cr67 6.51E+04 4.26E+06 1.29E+06 194.0 Cs131 3.62E+06 1.27E+09 3.31E+08 198.9 Cs132 1.93E+10 2.46E+13 4.72E+12 199.7 Cs134 1.14E+15 8.37E+16 2.56E+16 194.6 Cs134m 1.61E+14 4.31E+16 9.74E+15 198.5 Cs135 3.27E+10 3.33E+11 1.13E+11 164.2 Cs135m 7.79E+12 9.82E+14 2.49E+14 196.8 Cs136 8.85E+14 5.47E+16 1.33E+16 193.6 Cs136m 7.44E+13 3.61E+15 1.09E+15 191.9 Cs137 4.48E+15 2.29E+16 1.38E+16 134.5 Cs138 2.43E+16 6.26E+17 1.89E+17 185.1 Cs138m 1.15E+15 4.47E+16 1.28E+16 190.0 Cs139 2.28E+16 5.80E+17 1.74E+17 184.9 Cs140 1.99E+16 4.77E+17 1.43E+17 184.0 Cs141 1.56E+16 3.76E+17 1.14E+17 184.1 Cs142 9.72E+15 2.24E+17 6.65E+16 183.4 Cs143 5.41E+15 1.13E+17 3.41E+16 181.8 Cs144 1.78E+15 3.46E+16 1.14E+16 180.4 Cs145 3.24E+14 6.80E+15 2.53E+15 181.8 Cs146 3.44E+13 8.80E+14 3.70E+14 185.0 Cs147 8.74E+12 1.55E+14 5.62E+13 178.6 Cs148 9.05E+10 1.35E+13 4.04E+12 197.3 Cs149 2.29E+09 2.48E+11 7.95E+10 196.3 Cs150 5.94E+07 2.62E+10 6.89E+09 199.1 Cs151 6.85E+06 4.59E+09 1.15E+09 199.4 Cu66 6.02E+08 6.52E+10 2.05E+10 196.3 Cu67 2.10E+09 2.06E+11 6.14E+10 196.0 Cu68 4.78E+09 3.77E+11 1.20E+11 195.0 Cu68m 1.63E+08 1.03E+10 3.89E+09 193.7 Cu69 1.25E+10 7.67E+11 2.62E+11 193.6 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.57

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Cu70 2.92E+10 1.45E+12 5.46E+11 192.1 Cu70m 8.96E+09 5.55E+11 1.82E+11 193.6 Cu71 6.62E+10 3.21E+12 1.20E+12 191.9 Cu72 1.35E+11 5.21E+12 1.98E+12 189.9 Cu73 2.70E+11 6.86E+12 2.85E+12 184.8 Cu74 3.23E+11 7.10E+12 2.77E+12 182.6 Cu75 3.80E+11 7.18E+12 2.60E+12 179.9 Cu76 1.66E+11 5.54E+12 1.65E+12 188.3 Cu77 6.17E+10 2.85E+12 8.16E+11 191.5 Cu78 1.91E+10 8.89E+11 2.56E+11 191.6 Cu79 4.16E+08 2.06E+11 5.31E+10 199.2 Cu80 2.19E+08 2.06E+10 5.46E+09 195.8 Dy157 7.06E+03 1.24E+05 7.25E+04 178.5 Dy159 7.08E+05 2.73E+10 2.72E+09 200.0 Dy165 4.45E+11 8.18E+14 1.21E+14 199.8 Dy165m 1.42E+10 5.17E+14 7.51E+13 200.0 Dy166 4.22E+10 2.05E+13 3.69E+12 199.2 Dy167 1.28E+10 3.48E+12 8.94E+11 198.5 Dy168 5.13E+09 1.89E+12 4.04E+11 198.9 Dy169 2.23E+09 7.11E+11 1.61E+11 198.8 Dy170 6.87E+08 2.42E+11 5.55E+10 198.9 Dy171 1.53E+08 6.64E+10 1.80E+10 199.1 Dy172 8.00E+07 4.48E+10 1.25E+10 199.3 Er163 2.01E+04 2.01E+04 2.01E+04 0.0 Er165 3.14E+05 1.88E+09 4.39E+08 199.9 Er167m 5.59E+09 2.95E+13 4.60E+12 199.9 Er169 2.84E+09 1.87E+12 3.72E+11 199.4 Er171 3.71E+08 2.32E+11 4.64E+10 199.4 Er172 2.03E+08 1.01E+11 2.94E+10 199.2 Es253 7.05E+07 3.40E+09 7.36E+08 191.9 Es254 9.19E+04 6.81E+06 1.41E+06 194.7 Es254m 1.42E+07 2.18E+09 4.08E+08 197.4 Es255 3.31E+05 3.58E+07 6.77E+06 196.3 Eu149 1.48E+04 1.88E+06 4.83E+05 196.9 Eu152 1.71E+10 6.12E+12 1.42E+12 198.9 Eu152m 3.33E+12 1.03E+14 2.32E+13 187.5 Eu154 7.65E+13 2.41E+15 7.78E+14 187.7 Eu154m 5.16E+12 4.16E+15 8.29E+14 199.5 Eu155 1.36E+14 3.57E+15 7.97E+14 185.3 Eu156 2.53E+14 2.31E+17 4.55E+16 199.6 Eu157 1.13E+14 4.08E+16 6.45E+15 198.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.58

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Eu158 5.43E+13 5.59E+15 1.27E+15 196.1 Eu159 2.26E+13 2.75E+15 6.31E+14 196.7 Eu160 8.59E+12 1.25E+15 2.77E+14 197.3 Eu161 3.06E+12 5.34E+14 1.15E+14 197.7 Eu162 5.81E+11 1.06E+14 2.40E+13 197.8 Eu163 1.38E+11 3.02E+13 6.39E+12 198.2 Eu164 2.08E+10 4.58E+12 1.05E+12 198.2 Eu165 3.50E+09 5.79E+11 1.67E+11 197.6 Eu166 3.42E+08 8.27E+10 2.52E+10 198.4 Eu167 2.83E+07 1.49E+10 4.00E+09 199.2 F20 2.38E+05 3.66E+17 6.24E+16 200.0 Fe65 6.28E+05 2.13E+08 5.66E+07 198.8 Fe66 2.17E+08 2.87E+10 8.10E+09 197.0 Fe67 3.43E+08 5.51E+10 1.45E+10 197.5 Fe68 1.99E+08 3.35E+10 8.92E+09 197.6 Fe69 6.57E+07 1.21E+10 3.28E+09 197.8 Fe70 1.78E+07 3.37E+09 9.35E+08 197.9 Fe71 2.74E+06 5.95E+08 1.72E+08 198.2 Fe72 3.79E+05 8.33E+07 2.46E+07 198.2 Fr221 2.58E+04 1.58E+10 1.31E+09 200.0 Fr222 9.39E+03 9.39E+03 9.39E+03 0.0 Fr223 2.14E+03 1.44E+08 4.81E+07 200.0 Ga68 3.35E+03 1.59E+07 2.54E+06 199.9 Ga70 1.38E+08 8.34E+10 1.55E+10 199.3 Ga72 2.18E+11 9.15E+12 3.43E+12 190.7 Ga72m 8.78E+09 3.83E+11 1.38E+11 191.0 Ga73 6.25E+11 1.93E+13 7.29E+12 187.5 Ga74 1.77E+12 4.47E+13 1.69E+13 184.8 Ga74m 8.62E+10 4.15E+12 1.24E+12 191.9 Ga75 5.04E+12 1.05E+14 3.86E+13 181.7 Ga76 1.29E+13 2.36E+14 8.51E+13 179.2 Ga77 3.00E+13 4.90E+14 1.66E+14 176.9 Ga78 5.09E+13 8.53E+14 2.66E+14 177.5 Ga79 5.08E+13 9.67E+14 3.06E+14 180.1 Ga80 3.36E+13 6.25E+14 1.97E+14 179.6 Ga81 1.56E+13 3.75E+14 1.19E+14 184.0 Ga82 3.82E+12 2.29E+14 6.63E+13 193.4 Ga83 6.83E+11 2.79E+13 7.98E+12 190.4 Ga84 4.30E+11 3.70E+14 7.77E+13 199.5 Ga85 2.91E+09 5.47E+11 1.42E+11 197.9 Ga86 1.47E+09 1.07E+12 2.28E+11 199.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.59

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Gd151 4.12E+05 3.04E+09 3.83E+08 199.9 Gd153 4.01E+10 9.86E+12 1.41E+12 198.4 Gd155m 3.67E+10 7.10E+10 5.38E+10 63.7 Gd159 2.55E+13 8.43E+15 1.44E+15 198.8 Gd161 3.93E+12 6.77E+14 1.51E+14 197.7 Gd162 1.24E+12 2.22E+14 5.11E+13 197.8 Gd163 4.76E+11 1.05E+14 2.23E+13 198.2 Gd164 1.50E+11 3.59E+13 7.89E+12 198.3 Gd165 4.24E+10 9.65E+12 2.21E+12 198.3 Gd166 1.27E+10 2.27E+12 6.21E+11 197.8 Gd167 1.50E+09 3.24E+11 1.03E+11 198.2 Gd168 4.83E+08 7.99E+10 2.90E+10 197.6 Gd169 6.09E+07 1.55E+10 4.88E+09 198.4 Ge69 6.53E+02 2.17E+05 9.17E+04 198.8 Ge71 7.89E+05 1.84E+09 2.44E+08 199.8 Ge71m 9.99E+04 6.93E+06 1.87E+06 194.3 Ge73m 6.16E+11 1.91E+13 7.21E+12 187.5 Ge75 5.12E+12 1.08E+14 3.98E+13 181.9 Ge75m 2.54E+11 7.17E+12 2.44E+12 186.3 Ge77 3.45E+13 6.34E+14 2.17E+14 179.3 Ge77m 6.64E+11 2.50E+13 7.72E+12 189.6 Ge78 8.58E+13 1.75E+15 5.63E+14 181.3 Ge79 1.19E+14 2.16E+15 7.05E+14 179.2 Ge79m 5.53E+13 1.23E+15 4.11E+14 182.8 Ge80 3.54E+14 6.96E+15 2.07E+15 180.6 Ge81 3.93E+14 7.56E+15 2.19E+15 180.2 Ge81m 6.47E+12 1.56E+14 4.92E+13 184.0 Ge82 2.87E+14 6.25E+15 1.81E+15 182.5 Ge83 1.04E+14 1.73E+15 6.35E+14 177.3 Ge84 3.25E+13 1.16E+15 3.88E+14 189.1 Ge85 4.94E+12 2.17E+14 6.28E+13 191.1 Ge86 1.67E+13 1.96E+16 4.06E+15 199.7 Ge87 1.12E+11 7.12E+13 1.53E+13 199.4 Ge88 5.63E+09 1.71E+12 4.09E+11 198.7 Ge89 4.85E+07 2.33E+10 5.93E+09 199.2 H3 1.80E+13 1.85E+16 2.92E+15 199.6 He6 8.63E+05 5.58E+16 1.59E+16 200.0 Hg203 1.37E+10 1.37E+10 1.37E+10 0.0 Hg205 5.43E+10 5.43E+10 5.43E+10 0.0 Hg206 3.07E+08 3.07E+08 3.07E+08 0.0 Ho161 8.27E+03 1.85E+06 4.60E+05 198.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.60

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ho161m 0.00E+00 0.00E+00 #DIV/0! #DIV/0!

Ho162 7.90E+04 1.67E+07 3.38E+06 198.1 Ho162m 7.64E+04 1.62E+07 3.30E+06 198.1 Ho163 1.71E+03 4.66E+03 2.83E+03 92.6 Ho163m 1.66E+05 3.55E+07 7.08E+06 198.1 Ho164 8.83E+06 4.10E+09 1.07E+09 199.1 Ho164m 5.81E+06 2.24E+09 5.68E+08 199.0 Ho166 1.10E+11 3.24E+14 4.74E+13 199.9 Ho166m 1.15E+06 2.74E+08 8.04E+07 198.3 Ho167 1.62E+10 2.13E+13 3.35E+12 199.7 Ho168 5.44E+09 2.03E+12 4.31E+11 198.9 Ho169 2.59E+09 8.86E+11 1.94E+11 198.8 Ho170 8.06E+08 3.15E+11 6.74E+10 199.0 Ho170m 1.11E+08 7.36E+10 1.19E+10 199.4 Ho171 3.18E+08 1.85E+11 3.69E+10 199.3 Ho172 1.66E+08 7.15E+10 2.45E+10 199.1 I123 6.01E+04 7.56E+06 2.04E+06 196.8 I125 1.45E+04 4.81E+07 7.28E+06 199.9 I126 6.47E+08 1.39E+12 2.13E+11 199.8 I128 7.90E+13 1.15E+16 2.84E+15 197.3 I129 1.18E+09 6.47E+09 3.49E+09 138.3 I130 1.65E+14 1.73E+16 5.11E+15 196.2 I130m 1.05E+14 1.11E+16 3.29E+15 196.3 I131 1.24E+16 3.45E+17 1.05E+17 186.1 I132 1.83E+16 5.00E+17 1.52E+17 185.9 I132m 1.63E+14 9.18E+15 2.37E+15 193.0 I133 2.59E+16 6.70E+17 2.05E+17 185.1 I133m 2.20E+15 6.72E+16 2.01E+16 187.3 I134 2.91E+16 7.55E+17 2.28E+17 185.1 I134m 2.27E+15 9.16E+16 2.53E+16 190.3 I135 2.46E+16 6.47E+17 1.97E+17 185.4 I136 1.07E+16 2.42E+17 7.46E+16 183.0 I136m 5.38E+15 1.65E+17 4.82E+16 187.4 I137 1.19E+16 2.92E+17 8.83E+16 184.3 I138 6.62E+15 1.52E+17 4.76E+16 183.3 I139 3.12E+15 5.86E+16 1.81E+16 179.8 I140 6.43E+14 1.23E+16 4.30E+15 180.1 I141 1.58E+14 2.44E+15 9.10E+14 175.7 I142 2.49E+13 4.92E+14 1.77E+14 180.7 I143 1.47E+11 4.03E+13 1.06E+13 198.5 I144 8.53E+09 6.16E+11 2.06E+11 194.5 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.61

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

In111 2.03E+03 6.84E+05 1.90E+05 198.8 In112 8.42E+03 3.31E+08 4.19E+07 200.0 In112m 6.80E+03 2.67E+08 3.38E+07 200.0 In113m 2.14E+03 3.65E+07 7.68E+06 200.0 In114 1.14E+09 2.41E+12 2.64E+11 199.8 In114m 7.12E+08 1.31E+12 1.46E+11 199.8 In115m 8.94E+13 4.89E+15 1.33E+15 192.8 In116 4.83E+12 1.72E+15 3.72E+14 198.9 In116m 7.94E+12 2.86E+15 6.16E+14 198.9 In117 5.62E+13 2.25E+15 6.88E+14 190.2 In117m 6.79E+13 2.77E+15 8.45E+14 190.4 In118 7.59E+13 2.88E+15 9.34E+14 189.7 In118m 2.46E+10 1.30E+12 3.95E+11 192.6 In119 3.79E+13 1.30E+15 4.17E+14 188.7 In119m 4.48E+13 1.59E+15 5.23E+14 189.1 In120 7.53E+13 2.63E+15 8.63E+14 188.9 In120m 3.58E+12 1.49E+14 4.35E+13 190.6 In121 5.57E+13 2.11E+15 6.56E+14 189.7 In121m 2.97E+13 1.06E+15 3.59E+14 189.1 In122 7.57E+13 2.28E+15 7.80E+14 187.2 In122m 2.03E+13 1.07E+15 3.00E+14 192.6 In123 4.82E+13 1.82E+15 5.81E+14 189.7 In123m 3.60E+13 8.67E+14 3.38E+14 184.1 In124 7.02E+13 1.62E+15 6.10E+14 183.4 In124m 2.44E+13 1.37E+15 4.01E+14 193.0 In125 4.78E+13 1.46E+15 5.00E+14 187.3 In125m 4.00E+13 1.39E+15 4.81E+14 188.8 In126 5.18E+13 1.48E+15 5.02E+14 186.5 In126m 2.35E+13 1.41E+15 4.23E+14 193.4 In127 1.58E+14 3.15E+15 9.43E+14 181.0 In127m 5.01E+13 2.22E+15 6.59E+14 191.2 In128 9.45E+13 2.71E+15 8.24E+14 186.5 In128m 8.20E+13 2.68E+15 7.95E+14 188.1 In129 7.56E+13 1.44E+15 4.66E+14 180.0 In129m 7.54E+13 1.33E+15 4.44E+14 178.5 In130 9.39E+13 3.83E+15 9.35E+14 190.4 In130m 4.49E+13 1.68E+15 5.64E+14 189.6 In131 4.47E+13 9.70E+14 2.82E+14 182.4 In131m 2.66E+13 7.06E+14 2.14E+14 185.5 In132 2.45E+13 4.86E+14 1.74E+14 180.8 In133 7.60E+11 4.42E+13 1.36E+13 193.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.62

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

In134 1.87E+10 2.70E+12 7.46E+11 197.2 In135 1.80E+08 9.58E+10 2.43E+10 199.3 Kr100 6.71E+07 3.75E+10 8.22E+09 199.3 Kr79 1.87E+04 4.64E+06 9.02E+05 198.4 Kr79m 9.34E+03 2.35E+06 5.63E+05 198.4 Kr81 6.65E+03 3.57E+05 1.17E+05 192.7 Kr81m 1.54E+08 4.65E+10 1.22E+10 198.7 Kr83m 1.73E+15 3.58E+16 1.08E+16 181.6 Kr85 4.08E+14 2.18E+15 1.20E+15 137.0 Kr85m 3.87E+15 7.81E+16 2.29E+16 181.1 Kr87 7.48E+15 1.45E+17 4.22E+16 180.4 Kr88 1.01E+16 1.95E+17 5.59E+16 180.4 Kr89 1.26E+16 2.39E+17 6.76E+16 180.0 Kr90 1.24E+16 2.37E+17 6.64E+16 180.1 Kr91 8.06E+15 1.63E+17 4.64E+16 181.2 Kr92 4.23E+15 8.31E+16 2.46E+16 180.6 Kr93 1.22E+15 2.41E+16 7.88E+15 180.7 Kr94 3.42E+14 6.31E+15 2.33E+15 179.4 Kr95 2.92E+13 9.50E+14 3.07E+14 188.1 Kr96 6.12E+13 1.67E+15 4.35E+14 185.8 Kr97 1.10E+12 1.01E+14 2.59E+13 195.7 Kr98 7.81E+10 5.46E+13 1.16E+13 199.4 Kr99 1.86E+08 1.19E+11 3.02E+10 199.4 La135 3.83E+05 2.84E+07 7.18E+06 194.7 La137 1.67E+05 2.17E+06 8.73E+05 171.5 La140 2.37E+16 5.95E+17 1.82E+17 184.7 La141 2.13E+16 5.33E+17 1.62E+17 184.6 La142 2.04E+16 5.14E+17 1.54E+17 184.7 La143 2.01E+16 4.89E+17 1.46E+17 184.2 La144 1.81E+16 4.30E+17 1.27E+17 183.8 La145 1.31E+16 3.06E+17 9.11E+16 183.6 La146 6.06E+15 1.26E+17 3.80E+16 181.7 La146m 2.57E+15 6.65E+16 1.98E+16 185.1 La147 4.07E+15 9.12E+16 2.86E+16 182.9 La148 1.58E+15 3.31E+16 1.05E+16 181.8 La149 3.77E+14 1.01E+16 3.45E+15 185.5 La150 5.51E+13 2.10E+15 7.51E+14 189.8 La151 6.06E+12 2.85E+14 1.17E+14 191.7 La152 3.95E+11 3.37E+13 1.27E+13 195.4 La153 2.06E+10 3.60E+12 1.13E+12 197.7 La154 6.22E+08 1.56E+11 4.66E+10 198.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.63

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

La155 2.43E+07 4.44E+09 1.51E+09 197.8 Li8 1.38E+09 8.08E+17 4.50E+17 200.0 Mg27 1.10E+05 1.10E+05 1.10E+05 0.0 Mn66 5.77E+06 1.96E+09 5.20E+08 198.8 Mn67 4.48E+06 1.73E+09 4.48E+08 199.0 Mn68 7.70E+05 3.66E+08 9.76E+07 199.2 Mn69 9.92E+04 4.88E+07 1.37E+07 199.2 Mo101 2.18E+16 5.90E+17 1.83E+17 185.8 Mo102 1.98E+16 5.89E+17 1.73E+17 187.0 Mo103 1.75E+16 6.23E+17 1.73E+17 189.1 Mo104 1.29E+16 5.45E+17 1.45E+17 190.7 Mo105 8.13E+15 4.02E+17 1.04E+17 192.1 Mo106 4.29E+15 2.54E+17 6.31E+16 193.4 Mo107 1.45E+15 9.52E+16 2.32E+16 194.0 Mo108 4.55E+14 3.20E+16 7.77E+15 194.4 Mo109 8.79E+13 3.81E+15 1.07E+15 191.0 Mo110 6.86E+12 4.82E+14 1.29E+14 194.4 Mo111 9.49E+11 1.08E+14 2.79E+13 196.5 Mo112 5.45E+10 1.44E+13 3.72E+12 198.5 Mo113 2.72E+09 1.66E+12 4.36E+11 199.3 Mo114 6.24E+08 9.10E+10 2.70E+10 197.3 Mo115 3.03E+07 5.83E+09 2.02E+09 197.9 Mo91 4.83E+11 4.83E+11 4.83E+11 0.0 Mo93 1.92E+03 1.32E+13 4.40E+12 200.0 Mo93m 6.80E+03 7.72E+13 6.43E+12 200.0 Mo99 2.38E+16 6.17E+17 1.97E+17 185.2 N16 1.12E+11 7.26E+16 1.40E+16 200.0 Na22 2.03E+10 2.03E+10 2.03E+10 0.0 Na24 1.70E+05 4.04E+17 1.35E+17 200.0 Na24m 1.31E+05 3.10E+17 1.03E+17 200.0 Na25 7.70E+09 7.70E+09 7.70E+09 0.0 Nb100 2.29E+16 5.81E+17 1.77E+17 184.9 Nb100m 1.56E+15 6.04E+16 1.69E+16 189.9 Nb101 2.08E+16 5.68E+17 1.72E+17 185.8 Nb102 1.25E+16 3.12E+17 9.56E+16 184.6 Nb102m 4.00E+15 1.49E+17 4.11E+16 189.5 Nb103 1.05E+16 3.33E+17 9.53E+16 187.7 Nb104 2.29E+15 9.18E+16 2.56E+16 190.2 Nb104m 1.88E+15 7.95E+16 2.12E+16 190.8 Nb105 1.65E+15 6.93E+16 1.92E+16 190.7 Nb106 1.84E+14 1.43E+16 3.76E+15 194.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.64

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Nb107 3.52E+13 2.84E+15 7.85E+14 195.1 Nb108 3.05E+12 2.88E+14 8.34E+13 195.8 Nb109 1.48E+12 2.63E+13 9.17E+12 178.7 Nb110 1.87E+10 1.70E+12 4.49E+11 195.6 Nb111 2.63E+09 1.60E+12 4.02E+11 199.3 Nb112 3.16E+07 1.62E+10 4.20E+09 199.2 Nb113 3.94E+06 2.91E+09 7.72E+08 199.5 Nb90 7.28E+07 7.28E+07 7.28E+07 0.0 Nb90m 5.03E+07 5.03E+07 5.03E+07 0.0 Nb91 7.07E+10 7.07E+10 7.07E+10 0.0 Nb91m 5.43E+12 5.43E+12 5.43E+12 0.0 Nb92 5.82E+06 5.82E+06 5.82E+06 0.0 Nb92m 6.02E+04 7.18E+14 5.99E+13 200.0 Nb93m 1.17E+10 9.17E+11 1.04E+11 195.0 Nb94 5.97E+06 1.37E+09 1.48E+08 198.3 Nb94m 7.68E+09 1.17E+13 1.25E+12 199.7 Nb95 2.42E+16 5.39E+17 1.59E+17 182.8 Nb95m 2.61E+14 5.78E+15 1.72E+15 182.7 Nb96 1.46E+13 7.20E+15 1.67E+15 199.2 Nb97 2.22E+16 5.52E+17 1.67E+17 184.5 Nb97m 2.10E+16 5.21E+17 1.58E+17 184.5 Nb98 2.21E+16 5.56E+17 1.70E+17 184.7 Nb98m 1.18E+14 3.50E+15 1.04E+15 187.0 Nb99 1.41E+16 3.56E+17 1.08E+17 184.8 Nb99m 9.40E+15 2.52E+17 7.71E+16 185.6 Nd140 1.96E+03 6.13E+06 1.26E+06 199.9 Nd141 1.84E+08 8.51E+11 1.20E+11 199.9 Nd141m 3.79E+07 1.81E+11 2.55E+10 199.9 Nd147 8.64E+15 2.05E+17 6.44E+16 183.8 Nd149 4.74E+15 1.36E+17 4.09E+16 186.5 Nd151 2.26E+15 7.84E+16 2.23E+16 188.8 Nd152 1.50E+15 5.56E+16 1.52E+16 189.5 Nd153 9.01E+14 3.41E+16 9.16E+15 189.7 Nd154 4.38E+14 1.98E+16 5.07E+15 191.3 Nd155 1.49E+14 8.10E+15 2.02E+15 192.8 Nd156 4.77E+13 3.08E+15 7.46E+14 193.9 Nd157 7.09E+12 8.28E+14 1.89E+14 196.6 Nd158 9.12E+11 1.50E+14 3.36E+13 197.6 Nd159 6.80E+10 1.45E+13 3.33E+12 198.1 Nd160 4.04E+09 8.17E+11 2.04E+11 198.0 Nd161 1.57E+08 2.98E+10 8.01E+09 197.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.65

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ne23 9.83E+07 5.13E+14 1.71E+14 200.0 Ni65 6.28E+05 2.16E+08 5.72E+07 198.8 Ni66 6.00E+08 6.51E+10 2.05E+10 196.3 Ni67 2.08E+09 2.05E+11 6.08E+10 196.0 Ni68 4.57E+09 3.65E+11 1.15E+11 195.1 Ni69 1.03E+10 6.53E+11 2.17E+11 193.8 Ni70 2.20E+10 1.12E+12 4.03E+11 192.3 Ni71 2.67E+10 1.19E+12 4.52E+11 191.2 Ni72 3.62E+10 1.22E+12 4.38E+11 188.5 Ni73 2.94E+10 9.41E+11 2.96E+11 187.9 Ni74 1.36E+10 4.64E+11 1.41E+11 188.6 Ni75 3.35E+09 1.55E+11 4.49E+10 191.5 Ni76 6.60E+08 4.85E+10 1.25E+10 194.6 Ni77 1.10E+08 7.88E+09 2.06E+09 194.5 Ni78 1.83E+07 1.21E+09 3.20E+08 194.0 Np234 6.45E+02 2.60E+05 1.33E+05 199.0 Np235 1.62E+06 3.27E+09 5.00E+08 199.8 Np236 8.82E+03 1.10E+07 2.29E+06 199.7 Np236m 5.26E+09 1.10E+13 2.16E+12 199.8 Np237 8.56E+09 1.25E+11 3.99E+10 174.4 Np238 1.20E+15 4.70E+17 9.53E+16 199.0 Np239 8.00E+16 9.23E+18 2.45E+18 196.6 Np240 8.34E+12 3.25E+16 5.63E+15 199.9 Np240m 1.44E+13 5.48E+16 9.48E+15 199.9 Np241 2.19E+06 3.27E+09 7.36E+08 199.7 O19 9.27E+09 5.32E+15 1.03E+15 200.0 Pa229 6.70E+02 7.52E+06 3.76E+06 200.0 Pa230 1.33E+05 3.21E+09 3.22E+08 200.0 Pa231 3.75E+05 1.28E+11 1.28E+10 200.0 Pa232 2.45E+08 9.77E+14 8.15E+13 200.0 Pa233 1.38E+10 8.78E+17 7.32E+16 200.0 Pa234 4.52E+07 1.67E+15 1.39E+14 200.0 Pa234m 9.60E+09 1.75E+15 1.46E+14 200.0 Pa235 4.14E+05 1.49E+11 1.93E+10 200.0 Pb203 2.16E+12 2.16E+12 2.16E+12 0.0 Pb205 2.38E+08 2.38E+08 2.38E+08 0.0 Pb207m 4.09E+03 1.75E+14 5.83E+13 200.0 Pb209 4.07E+04 4.78E+14 3.98E+13 200.0 Pb210 6.47E+03 4.24E+07 1.03E+07 199.9 Pb211 1.73E+05 9.97E+09 1.66E+09 200.0 Pb212 2.47E+07 2.05E+13 1.71E+12 200.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.66

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pb214 2.80E+05 6.75E+05 4.78E+05 82.8 Pd101 8.54E+04 1.52E+07 3.87E+06 197.8 Pd103 2.70E+09 4.91E+12 9.27E+11 199.8 Pd107 2.29E+09 3.39E+10 1.44E+10 174.7 Pd107m 6.01E+11 9.66E+14 1.47E+14 199.8 Pd109 1.33E+15 2.67E+17 5.45E+16 198.0 Pd109m 1.65E+12 2.33E+15 3.88E+14 199.7 Pd111 3.92E+14 3.84E+16 8.79E+15 196.0 Pd111m 2.32E+11 2.71E+14 5.98E+13 199.7 Pd112 2.15E+14 1.70E+16 3.89E+15 195.0 Pd113 1.60E+14 1.13E+16 2.64E+15 194.4 Pd114 1.10E+14 6.88E+15 1.72E+15 193.7 Pd115 8.67E+13 4.28E+15 1.20E+15 192.1 Pd116 7.36E+13 3.68E+15 1.06E+15 192.2 Pd117 5.82E+13 2.21E+15 6.89E+14 189.7 Pd118 2.23E+13 9.88E+14 3.53E+14 191.2 Pd119 4.49E+12 3.26E+14 1.25E+14 194.6 Pd120 7.40E+12 1.84E+14 6.25E+13 184.5 Pd121 2.84E+11 4.74E+13 1.37E+13 197.6 Pd122 3.85E+10 1.35E+13 3.67E+12 198.9 Pd123 5.33E+09 2.88E+12 7.80E+11 199.3 Pd124 1.56E+09 4.73E+11 1.29E+11 198.7 Pm144 2.23E+03 9.62E+05 2.07E+05 199.1 Pm145 4.79E+05 7.65E+08 1.20E+08 199.7 Pm146 1.43E+08 3.01E+12 5.13E+11 200.0 Pm147 7.17E+15 4.54E+16 1.70E+16 145.4 Pm148 3.03E+14 9.72E+16 2.21E+16 198.8 Pm148m 2.52E+14 1.77E+16 4.30E+15 194.4 Pm149 4.72E+15 2.23E+17 5.84E+16 191.7 Pm150 2.34E+12 1.26E+16 1.89E+15 199.9 Pm151 2.27E+15 7.82E+16 2.23E+16 188.7 Pm152 1.51E+15 5.67E+16 1.55E+16 189.6 Pm152m 2.67E+13 1.82E+15 4.43E+14 194.2 Pm153 1.00E+15 3.91E+16 1.05E+16 190.0 Pm154 5.04E+14 2.34E+16 5.97E+15 191.6 Pm154m 6.54E+13 3.55E+15 9.08E+14 192.8 Pm155 2.95E+14 1.61E+16 4.06E+15 192.8 Pm156 1.49E+14 9.48E+15 2.31E+15 193.8 Pm157 5.25E+13 4.85E+15 1.11E+15 195.7 Pm158 1.23E+13 1.77E+15 3.83E+14 197.2 Pm159 2.42E+12 4.43E+14 9.49E+13 197.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.67

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pm160 2.92E+11 6.84E+13 1.46E+13 198.3 Pm161 2.94E+10 7.51E+12 1.62E+12 198.4 Pm162 5.54E+08 9.74E+10 3.17E+10 197.7 Pm163 2.49E+07 8.89E+09 2.56E+09 198.9 Po208 5.06E+06 5.06E+06 5.06E+06 0.0 Po209 1.78E+09 1.78E+09 1.78E+09 0.0 Po210 5.78E+03 4.36E+15 1.09E+15 200.0 Po211 4.79E+02 5.83E+09 1.46E+09 200.0 Po211m 4.76E+07 4.76E+07 4.76E+07 0.0 Po212 1.58E+07 1.31E+13 1.09E+12 200.0 Po213 2.53E+04 1.54E+10 1.29E+09 200.0 Po214 1.13E+04 3.84E+08 3.51E+07 200.0 Po215 1.73E+05 9.97E+09 1.66E+09 200.0 Po216 2.47E+07 2.05E+13 1.71E+12 200.0 Po218 2.80E+05 6.75E+05 4.78E+05 82.8 Pr139 1.39E+05 1.32E+08 3.34E+07 199.6 Pr140 1.23E+10 1.52E+13 3.00E+12 199.7 Pr142 3.76E+14 4.86E+16 1.39E+16 196.9 Pr142m 1.31E+14 1.70E+16 4.86E+15 196.9 Pr143 1.95E+16 4.89E+17 1.46E+17 184.6 Pr144 1.86E+16 3.69E+17 1.20E+17 180.9 Pr144m 1.79E+14 1.12E+16 2.45E+15 193.7 Pr145 1.38E+16 3.34E+17 1.00E+17 184.1 Pr146 1.11E+16 2.67E+17 8.12E+16 184.1 Pr147 8.46E+15 2.12E+17 6.44E+16 184.6 Pr148 6.21E+15 1.52E+17 4.64E+16 184.3 Pr148m 3.91E+14 2.04E+16 5.33E+15 192.5 Pr149 4.53E+15 1.21E+17 3.67E+16 185.6 Pr150 2.97E+15 8.69E+16 2.53E+16 186.8 Pr151 1.68E+15 5.23E+16 1.49E+16 187.6 Pr152 7.06E+14 2.58E+16 6.97E+15 189.3 Pr153 2.34E+14 1.01E+16 2.73E+15 190.9 Pr154 4.35E+13 2.75E+15 7.06E+14 193.8 Pr155 7.33E+12 5.53E+14 1.52E+14 194.8 Pr156 7.57E+11 8.43E+13 2.32E+13 196.4 Pr157 6.56E+10 9.18E+12 2.64E+12 197.2 Pr158 3.64E+09 5.77E+11 1.67E+11 197.5 Pr159 1.44E+08 1.88E+10 6.34E+09 197.0 Pt200 7.33E+04 7.33E+04 7.33E+04 0.0 Pu236 1.58E+09 1.17E+12 1.95E+11 199.5 Pu237 4.38E+08 1.21E+12 1.75E+11 199.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.68

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Pu237m 1.51E+08 4.50E+11 9.07E+10 199.9 Pu238 3.30E+13 1.77E+15 5.23E+14 192.7 Pu239 4.42E+12 1.85E+14 5.26E+13 190.7 Pu240 1.16E+13 1.65E+14 3.93E+13 173.8 Pu241 1.33E+14 3.48E+16 8.41E+15 198.5 Pu242 1.05E+08 8.06E+11 2.76E+11 199.9 Pu243 4.57E+11 6.40E+17 1.02E+17 200.0 Pu244 2.45E+03 5.90E+05 2.25E+05 198.3 Pu245 5.16E+04 1.37E+13 1.74E+12 200.0 Ra222 1.13E+04 3.83E+08 3.50E+07 200.0 Ra223 1.73E+05 9.97E+09 1.66E+09 200.0 Ra224 2.47E+07 2.05E+13 1.71E+12 200.0 Ra225 2.63E+04 1.60E+10 1.34E+09 200.0 Ra226 2.80E+05 6.75E+05 4.78E+05 82.7 Ra227 3.59E+03 3.84E+07 4.68E+06 200.0 Ra228 1.40E+10 1.40E+10 1.40E+10 0.0 Rb100 1.31E+12 1.35E+15 2.80E+14 199.6 Rb101 6.18E+09 2.83E+11 8.25E+10 191.5 Rb102 5.03E+07 3.02E+10 7.64E+09 199.3 Rb81 5.93E+03 4.25E+05 1.42E+05 194.5 Rb83 4.57E+07 2.34E+09 6.43E+08 192.3 Rb84 6.10E+08 5.60E+11 1.16E+11 199.6 Rb86 2.31E+13 5.45E+15 1.15E+15 198.3 Rb86m 2.88E+12 6.79E+14 1.46E+14 198.3 Rb87 1.08E+06 4.48E+06 2.84E+06 122.2 Rb88 1.02E+16 2.00E+17 5.76E+16 180.6 Rb89 1.34E+16 2.60E+17 7.42E+16 180.4 Rb90 1.36E+16 2.50E+17 7.04E+16 179.4 Rb90m 2.59E+15 6.27E+16 1.84E+16 184.1 Rb91 1.63E+16 3.17E+17 9.10E+16 180.4 Rb92 1.46E+16 2.87E+17 8.30E+16 180.7 Rb93 1.16E+16 2.18E+17 6.43E+16 179.8 Rb94 6.31E+15 1.12E+17 3.47E+16 178.7 Rb95 3.15E+15 5.69E+16 1.78E+16 179.0 Rb96 8.10E+14 1.32E+16 4.91E+15 176.8 Rb97 1.47E+14 2.29E+15 7.20E+14 175.8 Rb98 1.60E+13 5.67E+14 1.85E+14 189.0 Rb99 3.66E+11 9.83E+13 2.59E+13 198.5 Rh101 1.52E+05 6.27E+07 9.29E+06 199.0 Rh101m 6.97E+04 7.79E+08 1.12E+08 200.0 Rh102 4.47E+09 7.01E+12 1.10E+12 199.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.69

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Rh102m 1.88E+09 1.19E+12 1.85E+11 199.4 Rh103m 1.82E+16 6.27E+17 1.79E+17 188.7 Rh104 1.57E+15 5.48E+17 1.21E+17 198.9 Rh104m 1.24E+14 4.17E+16 9.20E+15 198.8 Rh105 9.90E+15 3.63E+17 1.13E+17 189.4 Rh105m 2.84E+15 1.51E+17 3.95E+16 192.6 Rh106 5.98E+15 3.45E+17 8.38E+16 193.2 Rh106m 5.43E+11 1.39E+16 2.06E+15 200.0 Rh107 3.30E+15 3.41E+17 8.22E+16 196.2 Rh108 1.92E+15 2.37E+17 5.49E+16 196.8 Rh108m 9.86E+12 1.46E+15 3.67E+14 197.3 Rh109 1.23E+15 1.62E+17 3.71E+16 197.0 Rh110 1.18E+13 1.89E+15 3.80E+14 197.5 Rh110m 7.11E+14 7.76E+16 1.77E+16 196.4 Rh111 3.90E+14 3.77E+16 8.54E+15 195.9 Rh112 2.14E+14 1.59E+16 3.76E+15 194.7 Rh113 1.55E+14 1.02E+16 2.49E+15 194.0 Rh114 8.42E+13 5.14E+15 1.35E+15 193.5 Rh115 3.94E+13 2.18E+15 6.43E+14 192.9 Rh116 1.27E+13 8.14E+14 2.57E+14 193.8 Rh117 3.91E+12 2.05E+14 8.11E+13 192.5 Rh118 4.81E+11 7.63E+13 2.41E+13 197.5 Rh119 5.45E+10 1.87E+13 5.41E+12 198.8 Rh120 1.49E+10 4.08E+12 1.12E+12 198.5 Rh121 9.59E+08 6.63E+11 1.79E+11 199.4 Rh122 1.10E+08 8.31E+10 2.26E+10 199.5 Rn217 1.10E+06 1.10E+06 1.10E+06 0.0 Rn218 1.13E+04 3.83E+08 3.50E+07 200.0 Rn219 1.73E+05 9.97E+09 1.66E+09 200.0 Rn220 2.47E+07 2.05E+13 1.71E+12 200.0 Rn222 2.80E+05 6.75E+05 4.78E+05 82.8 Ru103 1.84E+16 6.33E+17 1.81E+17 188.7 Ru105 1.00E+16 5.33E+17 1.39E+17 192.6 Ru106 5.85E+15 2.18E+17 6.48E+16 189.5 Ru107 3.27E+15 3.37E+17 8.10E+16 196.2 Ru108 1.91E+15 2.35E+17 5.46E+16 196.8 Ru109 1.19E+15 1.55E+17 3.56E+16 197.0 Ru110 6.99E+14 7.57E+16 1.73E+16 196.3 Ru111 3.42E+14 3.19E+16 7.36E+15 195.8 Ru112 1.63E+14 1.10E+16 2.74E+15 194.2 Ru113 7.76E+13 5.01E+15 1.27E+15 193.9 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.70

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Ru114 2.71E+13 1.73E+15 4.70E+14 193.9 Ru115 4.63E+12 3.64E+14 1.05E+14 195.0 Ru116 5.37E+11 4.43E+13 1.72E+13 195.2 Ru117 3.90E+10 6.52E+12 2.09E+12 197.6 Ru118 2.84E+09 1.33E+12 3.66E+11 199.1 Ru119 1.95E+08 1.31E+11 3.56E+10 199.4 Ru120 1.78E+07 1.26E+10 3.41E+09 199.4 Ru97 3.36E+04 3.36E+04 3.36E+04 0.0 Sb118 3.32E+03 6.30E+05 1.91E+05 197.9 Sb118m 5.13E+03 1.06E+06 3.14E+05 198.1 Sb119 9.96E+05 9.72E+07 2.89E+07 195.9 Sb120 5.34E+07 1.24E+11 2.15E+10 199.8 Sb120m 4.22E+07 5.69E+10 1.00E+10 199.7 Sb122 7.11E+12 1.06E+15 2.64E+14 197.3 Sb122m 4.67E+11 6.96E+13 1.74E+13 197.3 Sb124 5.66E+12 6.78E+14 1.73E+14 196.7 Sb124m 1.64E+11 1.79E+13 4.55E+12 196.4 Sb125 2.23E+14 2.53E+15 1.08E+15 167.6 Sb126 4.88E+12 2.84E+14 8.02E+13 193.2 Sb126m 6.10E+12 1.87E+14 6.08E+13 187.4 Sb127 1.07E+15 3.87E+16 1.10E+16 189.2 Sb128 1.54E+14 6.15E+15 1.73E+15 190.2 Sb128m 1.80E+15 5.47E+16 1.64E+16 187.3 Sb129 3.33E+15 1.11E+17 3.23E+16 188.4 Sb129m 4.94E+13 1.42E+15 4.28E+14 186.5 Sb130 3.20E+15 8.94E+16 2.70E+16 186.2 Sb130m 4.12E+15 1.07E+17 3.32E+16 185.2 Sb131 1.04E+16 2.57E+17 7.95E+16 184.5 Sb132 7.77E+15 2.09E+17 6.30E+16 185.7 Sb132m 4.13E+15 8.34E+16 2.59E+16 181.1 Sb133 7.87E+15 1.75E+17 5.29E+16 182.8 Sb134 1.59E+15 3.24E+16 1.05E+16 181.3 Sb134m 1.62E+15 3.23E+16 1.03E+16 180.8 Sb135 6.24E+14 1.26E+16 4.68E+15 181.1 Sb136 5.08E+13 1.62E+15 6.08E+14 187.9 Sb137 1.52E+13 2.44E+15 5.57E+14 197.5 Sb138 2.11E+11 2.40E+13 7.04E+12 196.5 Sb139 9.43E+09 1.72E+12 4.84E+11 197.8 Se75 5.30E+05 1.41E+08 3.09E+07 198.5 Se77m 1.24E+11 3.09E+12 1.01E+12 184.6 Se79 3.99E+09 1.73E+10 1.04E+10 125.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.71

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Se79m 1.83E+14 3.77E+15 1.24E+15 181.5 Se81 6.97E+14 1.66E+16 5.06E+15 183.8 Se81m 5.88E+13 1.90E+15 5.70E+14 188.0 Se83 1.52E+15 3.00E+16 9.15E+15 180.8 Se83m 1.30E+14 3.49E+15 1.04E+15 185.6 Se84 3.02E+15 5.72E+16 1.70E+16 180.0 Se85 3.15E+15 5.99E+16 1.73E+16 180.0 Se86 3.03E+15 6.51E+16 1.82E+16 182.2 Se87 1.58E+15 3.61E+16 1.02E+16 183.3 Se88 5.50E+14 1.52E+16 4.51E+15 186.0 Se89 1.22E+14 3.34E+15 1.06E+15 185.9 Se90 4.57E+13 9.19E+14 3.12E+14 181.0 Se91 2.39E+12 1.18E+14 3.25E+13 192.1 Se92 1.67E+11 1.18E+13 3.18E+12 194.4 Se93 1.06E+10 5.86E+11 1.61E+11 192.9 Se94 3.28E+08 4.13E+10 1.25E+10 196.9 Sm145 4.16E+05 4.70E+08 6.73E+07 199.6 Sm146 1.54E+03 1.00E+05 4.40E+04 193.9 Sm147 9.67E+04 9.93E+05 3.59E+05 164.5 Sm151 1.10E+13 4.10E+14 1.12E+14 189.5 Sm153 1.12E+15 2.95E+17 6.17E+16 198.5 Sm155 3.39E+14 2.16E+16 5.30E+15 193.8 Sm156 2.05E+14 1.27E+16 3.14E+15 193.6 Sm157 1.06E+14 8.40E+15 2.00E+15 195.0 Sm158 5.00E+13 4.91E+15 1.13E+15 196.0 Sm159 1.71E+13 2.15E+15 4.82E+14 196.8 Sm160 5.03E+12 7.52E+14 1.64E+14 197.3 Sm161 9.58E+11 1.91E+14 4.01E+13 198.0 Sm162 8.32E+10 1.24E+13 3.06E+12 197.3 Sm163 8.51E+09 1.27E+12 3.40E+11 197.3 Sm164 5.59E+08 1.11E+11 3.64E+10 198.0 Sm165 3.07E+07 1.19E+10 3.33E+09 199.0 Sn113 1.30E+05 2.15E+07 6.44E+06 197.6 Sn113m 1.11E+05 2.50E+07 7.57E+06 198.2 Sn117m 2.05E+11 1.19E+13 3.23E+12 193.2 Sn119m 2.79E+12 6.38E+13 2.43E+13 183.2 Sn121 8.16E+13 2.95E+15 9.55E+14 189.2 Sn121m 1.15E+12 7.32E+12 3.04E+12 145.8 Sn123 1.43E+13 6.51E+14 1.94E+14 191.4 Sn123m 8.79E+13 2.93E+15 9.89E+14 188.3 Sn125 1.07E+14 4.42E+15 1.26E+15 190.6 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.72

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Sn125m 1.28E+14 4.25E+15 1.37E+15 188.3 Sn126 1.03E+10 7.14E+10 3.38E+10 149.7 Sn127 6.37E+14 2.36E+16 6.69E+15 189.5 Sn127m 3.70E+14 1.37E+16 3.92E+15 189.5 Sn128 1.72E+15 5.07E+16 1.53E+16 186.9 Sn128m 8.24E+14 2.47E+16 7.44E+15 187.1 Sn129 1.76E+15 6.01E+16 1.77E+16 188.6 Sn129m 8.71E+14 2.50E+16 7.54E+15 186.5 Sn130 2.38E+15 5.47E+16 1.67E+16 183.3 Sn130m 2.41E+15 5.55E+16 1.72E+16 183.3 Sn131 2.05E+15 4.15E+16 1.32E+16 181.1 Sn131m 1.97E+15 3.99E+16 1.27E+16 181.2 Sn132 2.62E+15 5.26E+16 1.67E+16 181.0 Sn133 4.34E+14 8.40E+15 3.05E+15 180.4 Sn134 7.01E+13 1.43E+15 5.26E+14 181.3 Sn135 2.67E+12 1.21E+14 3.79E+13 191.4 Sn136 7.50E+10 7.98E+12 2.24E+12 196.3 Sn137 8.81E+09 6.70E+11 2.15E+11 194.8 Sr100 1.45E+14 2.59E+15 9.24E+14 178.8 Sr101 1.35E+13 3.57E+14 1.15E+14 185.5 Sr102 6.99E+11 3.77E+13 1.09E+13 192.7 Sr103 1.18E+10 9.50E+11 3.08E+11 195.1 Sr104 6.62E+08 1.14E+11 3.01E+10 197.7 Sr105 3.29E+07 2.19E+10 5.50E+09 199.4 Sr83 1.21E+03 4.43E+04 1.78E+04 189.4 Sr85 1.07E+07 3.07E+10 3.90E+09 199.9 Sr85m 5.66E+06 8.87E+09 1.20E+09 199.7 Sr87m 4.74E+10 1.63E+14 2.01E+13 199.9 Sr89 1.35E+16 2.84E+17 7.87E+16 181.8 Sr90 3.55E+15 1.64E+16 9.58E+15 128.6 Sr91 1.74E+16 3.45E+17 1.00E+17 180.7 Sr92 1.84E+16 3.79E+17 1.11E+17 181.5 Sr93 2.01E+16 4.32E+17 1.28E+17 182.3 Sr94 1.94E+16 4.30E+17 1.26E+17 182.8 Sr95 1.74E+16 3.77E+17 1.12E+17 182.3 Sr96 1.35E+16 2.74E+17 8.23E+16 181.2 Sr97 6.55E+15 1.24E+17 3.79E+16 180.0 Sr98 3.28E+15 5.40E+16 1.74E+16 177.1 Sr99 5.64E+14 9.89E+15 3.84E+15 178.4 Tb155 8.67E+03 1.79E+06 4.58E+05 198.1 Tb156 2.77E+05 3.09E+07 8.47E+06 196.4 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.73

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Tb156m 2.45E+04 3.96E+06 9.56E+05 197.5 Tb157 1.07E+05 5.14E+06 1.03E+06 191.9 Tb158 8.11E+05 4.70E+08 6.71E+07 199.3 Tb158m 2.29E+07 4.47E+10 7.90E+09 199.8 Tb160 2.63E+12 1.08E+15 2.42E+14 199.0 Tb161 4.07E+12 1.83E+15 3.06E+14 199.1 Tb162 1.28E+12 3.28E+14 6.46E+13 198.5 Tb163 5.30E+11 1.18E+14 2.52E+13 198.2 Tb164 1.92E+11 4.79E+13 1.05E+13 198.4 Tb165 7.78E+10 1.95E+13 4.37E+12 198.4 Tb166 3.33E+10 7.37E+12 1.88E+12 198.2 Tb167 8.38E+09 2.08E+12 5.35E+11 198.4 Tb168 2.73E+09 7.54E+11 1.77E+11 198.6 Tb169 7.65E+08 1.70E+11 5.09E+10 198.2 Tb170 1.42E+08 3.34E+10 1.17E+10 198.3 Tb171 1.76E+07 8.51E+09 2.43E+09 199.2 Tc100 2.82E+15 3.57E+17 9.14E+16 196.9 Tc101 2.18E+16 5.91E+17 1.83E+17 185.8 Tc102 1.98E+16 5.91E+17 1.73E+17 187.0 Tc102m 5.01E+13 2.19E+15 5.96E+14 191.0 Tc103 1.79E+16 6.41E+17 1.78E+17 189.1 Tc104 1.38E+16 5.85E+17 1.56E+17 190.8 Tc105 9.93E+15 5.19E+17 1.33E+17 192.5 Tc106 5.81E+15 4.15E+17 1.02E+17 194.5 Tc107 2.85E+15 2.74E+17 6.52E+16 195.9 Tc108 1.25E+15 1.32E+17 2.97E+16 196.2 Tc109 5.87E+14 6.03E+16 1.35E+16 196.1 Tc110 1.48E+14 1.01E+16 2.39E+15 194.2 Tc111 3.17E+13 1.67E+15 4.83E+14 192.6 Tc112 4.17E+12 2.52E+14 9.44E+13 193.5 Tc113 7.68E+11 7.15E+13 2.23E+13 195.7 Tc114 2.99E+11 2.29E+13 7.95E+12 194.8 Tc115 3.07E+10 3.62E+12 1.31E+12 196.6 Tc116 1.20E+09 2.77E+11 8.84E+10 198.3 Tc117 2.68E+07 1.13E+10 3.20E+09 199.1 Tc118 3.57E+06 2.67E+09 7.15E+08 199.5 Tc96 8.68E+04 8.68E+04 8.68E+04 0.0 Tc97m 4.30E+06 1.06E+09 2.43E+08 198.4 Tc98 4.46E+04 5.75E+06 1.14E+06 196.9 Tc99 6.66E+11 2.79E+12 1.72E+12 122.9 Tc99m 2.09E+16 5.42E+17 1.75E+17 185.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.74

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Te121 1.35E+06 1.26E+10 1.67E+09 200.0 Te121m 6.86E+05 5.48E+09 7.02E+08 199.9 Te123m 3.75E+10 1.97E+13 3.24E+12 199.2 Te125m 5.13E+13 5.41E+14 2.26E+14 165.3 Te127 1.04E+15 3.29E+16 1.00E+16 187.8 Te127m 1.47E+14 2.85E+15 8.10E+14 180.4 Te129 3.17E+15 1.06E+17 3.09E+16 188.4 Te129m 5.53E+14 1.79E+16 5.23E+15 188.0 Te131 1.09E+16 2.88E+17 8.86E+16 185.4 Te131m 2.15E+15 7.63E+16 2.20E+16 189.0 Te132 1.80E+16 4.74E+17 1.46E+17 185.4 Te133 1.34E+16 3.37E+17 1.03E+17 184.7 Te133m 1.28E+16 3.35E+17 9.97E+16 185.3 Te134 2.40E+16 5.80E+17 1.72E+17 184.1 Te135 1.25E+16 2.91E+17 8.80E+16 183.5 Te136 5.51E+15 1.03E+17 3.20E+16 179.6 Te137 1.68E+15 3.00E+16 9.70E+15 178.9 Te138 2.70E+14 5.09E+15 2.04E+15 179.8 Te139 2.89E+13 8.28E+14 3.21E+14 186.5 Te140 6.18E+12 5.88E+14 1.47E+14 195.8 Te141 1.72E+11 2.98E+13 7.90E+12 197.7 Te142 9.34E+09 5.43E+11 1.69E+11 193.2 Th226 1.45E+04 3.83E+08 3.85E+07 200.0 Th227 1.73E+05 9.92E+09 1.65E+09 200.0 Th228 2.45E+07 2.04E+13 1.70E+12 200.0 Th229 2.10E+03 7.25E+09 1.45E+09 200.0 Th230 1.90E+04 4.39E+08 4.27E+07 200.0 Th231 3.09E+08 1.38E+15 1.15E+14 200.0 Th232 2.31E+10 2.31E+10 2.31E+10 0.0 Th233 1.63E+09 8.82E+17 7.35E+16 200.0 Th234 8.32E+09 1.12E+10 9.82E+09 29.6 Tl202 1.40E+08 1.40E+08 1.40E+08 0.0 Tl204 3.55E+10 3.55E+10 3.55E+10 0.0 Tl206 1.75E+12 1.75E+12 1.75E+12 0.0 Tl207 1.73E+05 9.75E+11 1.64E+11 200.0 Tl208 8.86E+06 7.36E+12 6.14E+11 200.0 Tl209 5.68E+02 3.47E+08 1.16E+08 200.0 Tm167 2.50E+03 2.27E+05 6.81E+04 195.6 Tm168 5.93E+03 9.37E+07 1.23E+07 200.0 Tm170 9.29E+08 3.13E+11 8.17E+10 198.8 Tm171 2.99E+08 5.86E+10 1.59E+10 198.0 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.75

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Tm172 2.29E+08 2.17E+11 4.93E+10 199.6 U230 1.03E+04 2.53E+08 3.18E+07 200.0 U231 6.51E+03 2.88E+09 2.62E+08 200.0 U232 5.37E+07 3.86E+13 3.22E+12 200.0 U233 8.52E+04 2.04E+13 1.70E+12 200.0 U234 1.21E+09 6.24E+11 1.19E+11 199.2 U235 1.41E+06 8.01E+09 2.95E+09 199.9 U235m 7.79E+12 1.81E+16 2.95E+15 199.8 U236 1.34E+10 6.77E+10 3.40E+10 133.9 U237 7.09E+15 7.55E+17 1.46E+17 196.3 U238 8.32E+09 1.12E+10 9.82E+09 29.9 U239 8.00E+16 9.34E+18 2.47E+18 196.6 U240 2.45E+03 5.86E+05 2.24E+05 198.3 Xe125 7.80E+04 2.17E+07 5.57E+06 198.6 Xe125m 2.06E+04 5.75E+06 2.88E+06 198.6 Xe127 2.63E+07 3.80E+11 4.83E+10 200.0 Xe127m 4.41E+06 6.92E+10 9.03E+09 200.0 Xe129m 5.74E+10 4.49E+13 9.09E+12 199.5 Xe131m 1.36E+14 3.50E+15 1.13E+15 185.0 Xe133 2.50E+16 6.77E+17 2.07E+17 185.8 Xe133m 7.65E+14 2.14E+16 6.72E+15 186.2 Xe134m 2.95E+14 1.39E+16 3.74E+15 191.7 Xe135 7.01E+15 3.44E+17 8.74E+16 192.0 Xe135m 5.25E+15 1.58E+17 4.70E+16 187.1 Xe137 2.30E+16 6.09E+17 1.85E+17 185.4 Xe138 2.22E+16 5.68E+17 1.70E+17 185.0 Xe139 1.69E+16 4.05E+17 1.19E+17 183.9 Xe140 1.20E+16 2.59E+17 7.57E+16 182.2 Xe141 4.90E+15 9.33E+16 2.87E+16 180.0 Xe142 1.79E+15 3.23E+16 1.04E+16 179.0 Xe143 1.70E+14 3.56E+15 1.36E+15 181.7 Xe144 2.56E+13 7.85E+14 2.87E+14 187.4 Xe145 5.86E+11 7.53E+13 2.38E+13 196.9 Xe146 5.43E+10 5.36E+12 1.59E+12 196.0 Xe147 7.07E+09 1.40E+11 5.56E+10 180.7 Y100 2.88E+15 6.06E+16 2.07E+16 181.9 Y101 1.15E+15 2.05E+16 7.28E+15 178.8 Y102 3.34E+14 1.14E+16 3.19E+15 188.6 Y103 1.32E+13 8.11E+14 2.72E+14 193.6 Y104 2.35E+12 1.01E+14 3.18E+13 190.9 Y105 3.91E+10 1.95E+13 5.04E+12 199.2 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.76

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Y106 5.53E+08 6.81E+10 2.09E+10 196.8 Y107 1.00E+07 7.02E+09 1.89E+09 199.4 Y108 8.51E+04 2.85E+07 1.05E+07 198.8 Y87 2.76E+06 1.79E+08 4.67E+07 193.9 Y87m 1.45E+04 6.73E+06 2.32E+06 199.1 Y88 1.50E+09 1.38E+12 2.85E+11 199.6 Y89m 1.57E+12 2.77E+13 8.76E+12 178.5 Y90 3.60E+15 1.78E+16 1.06E+16 132.6 Y90m 2.19E+11 4.01E+13 9.10E+12 197.8 Y91 1.75E+16 3.79E+17 1.06E+17 182.4 Y91m 1.03E+16 2.03E+17 5.89E+16 180.8 Y92 1.85E+16 3.84E+17 1.12E+17 181.6 Y93 2.05E+16 4.43E+17 1.32E+17 182.3 Y93m 7.05E+15 1.53E+17 4.55E+16 182.4 Y94 2.11E+16 4.78E+17 1.42E+17 183.1 Y95 2.17E+16 5.04E+17 1.51E+17 183.5 Y96 1.43E+16 2.98E+17 8.94E+16 181.7 Y96m 6.82E+15 1.87E+17 5.65E+16 185.9 Y97 1.13E+16 2.53E+17 7.65E+16 182.9 Y97m 6.65E+15 1.65E+17 4.98E+16 184.5 Y98 8.22E+15 1.75E+17 5.44E+16 182.0 Y98m 4.71E+15 1.21E+17 3.70E+16 185.0 Y99 8.93E+15 1.86E+17 5.80E+16 181.7 Yb169 1.70E+04 1.55E+07 2.20E+06 199.6 Yb169m 6.70E+03 1.71E+06 4.98E+05 198.4 Zn69 1.26E+10 7.78E+11 2.65E+11 193.6 Zn69m 7.73E+07 5.23E+09 1.63E+09 194.2 Zn71 6.82E+10 3.34E+12 1.24E+12 192.0 Zn71m 8.46E+09 5.69E+11 1.68E+11 194.1 Zn72 2.14E+11 8.63E+12 3.27E+12 190.3 Zn73 5.99E+11 1.79E+13 6.86E+12 187.0 Zn74 1.62E+12 3.74E+13 1.48E+13 183.4 Zn75 3.78E+12 6.87E+13 2.55E+13 179.1 Zn76 7.20E+12 1.20E+14 4.30E+13 177.3 Zn77 5.80E+12 1.48E+14 4.78E+13 184.9 Zn78 4.75E+12 1.50E+14 4.54E+13 187.8 Zn79 1.55E+12 6.44E+13 2.10E+13 190.6 Zn80 4.03E+11 1.45E+13 4.47E+12 189.2 Zn81 7.45E+09 2.39E+12 6.51E+11 198.8 Zn82 1.09E+10 7.41E+11 2.15E+11 194.2 Zn83 3.06E+08 5.41E+10 1.41E+10 197.8 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.77

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Minimum Maximum Average Percent Activity Activity Activity Difference Radionuclide (Bq/MTHM) (Bq/MTHM) (Bq/MTHM) (Min/Max)

Zr100 2.14E+16 5.28E+17 1.60E+17 184.5 Zr101 1.27E+16 2.94E+17 9.07E+16 183.4 Zr102 8.49E+15 1.76E+17 5.45E+16 181.6 Zr103 2.29E+15 4.89E+16 1.67E+16 182.1 Zr104 4.13E+14 1.24E+16 4.40E+15 187.1 Zr105 1.79E+14 4.88E+15 1.36E+15 185.9 Zr106 1.04E+12 1.19E+14 2.71E+13 196.6 Zr107 3.25E+10 3.23E+12 1.31E+12 196.0 Zr108 1.97E+09 2.19E+11 7.89E+10 196.4 Zr109 1.34E+09 2.37E+10 8.43E+09 178.6 Zr110 1.11E+07 1.10E+09 2.92E+08 196.0 Zr88 4.55E+03 1.37E+09 1.52E+08 200.0 Zr89 1.44E+08 1.12E+13 1.05E+12 200.0 Zr89m 8.43E+06 3.09E+12 2.63E+11 200.0 Zr90m 1.42E+10 1.41E+13 3.22E+12 199.6 Zr93 1.01E+11 3.75E+11 2.40E+11 115.0 Zr95 2.42E+16 5.34E+17 1.58E+17 182.7 Zr97 2.21E+16 5.48E+17 1.66E+17 184.5 Zr98 2.18E+16 5.46E+17 1.67E+17 184.7 Zr99 2.15E+16 5.41E+17 1.64E+17 184.7 McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 C.78

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Appendix D - NRC Advanced Reactor GEIS Values The following draft plant parameter envelope (PPE) and site parameter envelope (SPE) tables were presented to the public in May 2020 during the U.S. Nuclear Regulatory Commissions (NRCs) scoping process associated with the Advanced Nuclear Reactor (ANR)

Generic Environmental Impact Statement (GEIS). These tables are subject to change based upon additional NRC analysis.

Table D.1. NRC Advanced Reactor GEIS Draft Plant Parameter Envelope Parameter Value/Description Assumptions Site size 100 ac 1. Meets NRC Siting Regulations

2. Stand-alone site or designated portion of larger site (e.g.,

Government reservation, military base, or existing power plant site)

3. Complies with applicable zoning

4. Not inconsistent with any comprehensive plans or other land use plans Permanent footprint of disturbance 30 ac 1. No prime farmland, or not adjacent to actively used farmland

2. No wetlands, floodplains, surface water features, riparian habitat, climax or old-growth vegetation, or dedicated conservation land Temporary footprint of disturbance Additional 20 ac 1. Restored to original grade and seeded or planted with indigenous vegetation once construction is complete.

2. Meets assumptions for permanent footprint Offsite right-of-way 1,000 ft x 100 ft (new right-of-way) 1. Meets assumptions for site size or unlimited length (within or 2. Does not cross or pass adjacent to parks, wildlife refuges, or adjacent to existing right-of-way) conservation lands

3. Does not cross Wild and Scenic River or National Heritage River, or river of similar state designation Cooling and service water intake 1000 Gallons per minute (gpm) 1. If water-cooled, maximum amount of water removed from surface water bodies for cooling-water makeup Consumptive water use 400 gpm 1. Consumption through evaporative loss during the cooling process McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 D.1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table D.1. NRC Advanced Reactor GEIS Draft Plant Parameter Envelope (continued)

Parameter Value/Description Assumptions Plant water discharge 600 gpm 1. Amount of water discharged to waterbody after use for plant purposes including cooling (i.e., blowdown), and service water system.

2. Also includes other discharges from potable and sanitary systems (if applicable).

Blowdown temperature and Within applicable Clean Water Act 1. Discharge results mainly from plant blowdown.

constituent concentrations limits 2. Discharges are regulated under a clean water act permit and meets established discharge limits for temperature and for quantity of waste and concentration of each constituent.

Potable and sanitary water use 5 gpm 1. If groundwater is used, pumping rates fall within permittable and discharge limits.

2. If municipal water and sewage is used, usage amount is available and within capacity of the system.

Emissions from construction Criteria pollutants are less than 1. Clean Air Act requires a conformity determination for equipment and standby power Clean Air Act de minimis levels. maintenance or nonattainment areas that exceed de minimis equipment during operations values. Not applicable to attainment areas.

Megawatts thermal (MWt) 60 MWt 1. Total thermal power generated by all units on site; can be more than one unit, however total thermal power is 60 MWt.

The operational life for which the 80 yr 1. Bounding value. Assumes 40-year license with two 20-year plant is designed license renewals for operational life Building height 50 ft 1. Tallest structure, other than meteorology tower Foundation embedment 50 ft 1. The depth from finished grade to the bottom of the basemat for the most deeply embedded power-block structure Maximum number of construction 150 people 1. Maximum number of construction workforce, half of whom in-workers migrate to the host county is less than 5% of total host county populations The number of total permanent 50 people 1. Maximum operations workforce all of whom in- migrate to the staff to support operations host county.

The additional number of 100 people temporary staff for refueling outage 1. No refueling workers in-migrate to the host county.

Noise generation 65 dBA 1. At site boundary Station capacity factor 95% or greater 1. The percentage of time that a plant is capable of providing power to the grid McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 D.2

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table D.1. NRC Advanced Reactor GEIS Draft Plant Parameter Envelope (continued)

Parameter Value/Description Assumptions The normal plant operating cycle 2 to 20 yr length

1. Different designs have different operating cycle lengths Electrical output in megawatts- 20 MWe 1. Most nuclear steam supply system designs are approximately electric (MWe) 33 to 37 percent efficient applying a Rankine cycle without superheated steam.

2. It is acceptable that if the efficiency is higher than the bounding value can be slightly higher than 20 MWe.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 D.3

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table D.2 NRC Advanced Reactor GEIS Draft Site Parameter Envelope Parameter Value/Description Assumptions Water for sanitary and potable water Up to 5 gpm supply provided by 1. If groundwater is used, pumping rates fall within permittable uses municipal systems or groundwater limits and the aquifer supplying water must support the resources. required amount at a rate that is sustainable and does not impact offsite uses or users.

2. If municipal water supply is used, usage amount is available and within capacity of the system.

3. Sanitary discharge to sewage treatment plant is within available capacity and permittable. The plant is allowed to hook up to municipal water and sewage system with sufficient capacity.

Surface water availability If plant uses surface water, monthly 1. Not applicable if plant is air-cooled.

minimum flow is 75 cubic feet per 2. Maximum average plant water withdrawals are less than 3%

second (cfs). percent of minimum monthly flow of water body.

3. Water availability is demonstrated by state-issued withdrawal permit.

4. Withdrawals do not prevent the maintenance of applicable instream flow requirements.

5. Water rights are obtainable, if needed, and amount is available without impact to other uses and users.

6. Large water bodies such as the oceans, Great Lakes are presumed to have sufficient water availability.

7. Coastal Zone Management Act consistency determination obtained.

Surface water discharge If plant is discharging to surface 1. Not applicable if plant is air-cooled.

water, monthly minimum flow is 75 2. Maximum average plant discharge is small in comparison to cfs monthly minimum flow of water body (<3 percent) and thermal and chemical components within the discharge would be diluted quickly.

3. Discharge is in accordance with state/local permits.

4. Altered current patterns and salinity gradients would be localized.

5. Large water bodies such as the oceans, Great Lakes are presumed to have sufficient water capacity for dilution as long as restrictions on localized impacts are met.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 D.4

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table D.2 NRC Advanced Reactor GEIS Draft Site Parameter Envelope (continued)

Parameter Value/Description Assumptions Groundwater- availability and quality Pumping rate of < 100 gpm 1. Pumping rate is sustainable, is a small percent of flow within (regardless of proposed purpose) the aquifer and does not impact availability to offsite uses and users

2. Withdrawal rates are within limits which are permittable by applicable state or local agencies

3. Groundwater usage does not impact quality within the aquifer Air quality Attainment, maintenance area, or 1. Emission of criteria pollutants are less than de minimis levels nonattainment Economics Annual property tax for the 1. Overnight construction cost of the proposed project is no proposed project is less than ten more than $500 million USD percent of the total property tax revenue of the host county McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 D.5

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Appendix E - PPE Data Sources and Methodology These tables describe the sources of information that were used to develop the bounding value, as well as the source or rationale for the identified value. The plant parameter envelope (PPE) table references certain values and tables that are included in Appendix C; these locations are noted in the table as applicable. The bounding value in both Table E.1 and Table E.2 are denoted using bold font.

Table E.1 Microreactor PPE Data Sources and Methodology(a)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale What is your design None HTGR, MSR, HTGR, MSR, Not applicable Not Applicable for Plant type itself is not relevant to the environmental type? provided LMR, heat LMR, and bounding analysis; parameters therein that have an pipe, and nuclear battery parameters environmental interface are considered.

nuclear battery How many units do you None Not evaluated 1 Not applicable 1 While the PPE considers installation of one unit, plan to install? provided multiple units may be proposed, for instance to demonstrate the capability of following increases in electricity demand over time. This would have potential impacts on the extent and timing of resource analyses, including cumulative impacts.

What is the output of 60 13 MWe 50 MWt Not applicable 60 MWt Value is bounded by a larger microreactor that your design (per unit)? MWt/20MWe 17 MWe maximizes the difference between thermal and electrical output and will generally lead to greater resource needs, such as cooling water. Therefore, NRC's proposed microreactor limit was selected as the bounding value.

Is your reactor designed None Not evaluated No Not applicable No While the PPE representative value indicates that the to be mobile? provided reactor is not mobile, some designs may include Plant Design mobile reactors. If so, there would be additional transportation and workforce related issues that would have to be considered.

If the reactor is None Not evaluated 15 shipments, Not applicable 30 shipments, The largest value among the microreactor vendor designed to be provided including the including the responses, scaled to a 60 MWt reactor.

transportable, what are reactor, fuel, reactor, fuel, and the total number of and core core assembly shipments and weight of assembly reactor, fuel, and its packaging?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.1

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale Describe your power None Not evaluated Varied; N/A Power conversion does not itself have an conversion system. provided including environmental nexus.

Rankine Cycle, Brayton Cycle, and Air cooled DOWTHERM

' heat transfer fluid Will offsite power Required. Required, None provided General Required. Offsite The requirement for access to the existing onsite INL sources be required to Offsite ROW assuming Design ROW 1,000 ft x 100 transmission system would bound all designs. Both maintain functioning of 1,000 ft x compliance Criteria 17 ft (new) or within or substation and transmission interconnections are structures, systems, and 100 ft (new) with General adjacent to existing assumed to be required. Length and breadth of components important or within or Design Criteria ROW transmission line right-of-way and size of the to safety following loss adjacent to 17 switchyard will depend on final site location.

of onsite AC power? If existing so, what transmission ROW voltage would be required from offsite power sources?

What support facilities None Fuel storage Varied Not applicable Fuel storage and Representative support facilities as informed by SME (fuel storage and provided and handling; handling; waste analysis and review of publicly available information handling, waste waste treatment; reactor on microreactors, small- to medium-sized advanced Plant Design (continued) treatment, etc.) are treatment; pre-heating and reactors, and vendor responses. The existence of necessary for your plant reactor pre- metal melting; these structures themselves does not have an design? heating and control building; environmental nexus; however, the land use metal melting; power conversion requirements and resource needs associated with control these facilities will have an environmental nexus and building; power should be considered.

conversion What is the tallest 50 ft 28 ft structure 28 ft structure Not applicable 28 ft structure Selected largest values from vendor responses and Plant structure and what is (structure) NRC ANR GEIS PPE to better bound potential visual, the maximum structure 45 ft stack 45 ft stack 50 ft stack height scenic, and land use impacts Structure height (structure, ft)? None height What is the stack provided and Footprint height? (stack)

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.2

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale What is the maximum 50 ft Not Evaluated 20 ft Not applicable 20 ft Selected largest value consistent with vendor depth of excavation? responses. The NRC ANR GEIS value appears larger than necessary for the planned microreactor deployments.

What is the temporary 50 ac 10 ac 8 ac Not applicable 18 ac Selected values that bound vendor responses, with disturbed acreage slight rounding up to account for potential larger during construction, projects. The NRC ANR GEIS value appears larger including parking and than necessary for the planned microreactor laydown? deployments.

What is the permanent 30 ac 8 ac 7 ac Not applicable 8 ac Selected values that bound vendor responses. The disturbed acreage, NRC ANR GEIS value appears larger than necessary including parking lots, for the planned microreactor deployments.

ponds, substations, and other plant support facilities?

What is the maximum None 101 dB at 50 ft Question not Not applicable 101 dB at 50 ft Questionnaire did not include this parameter. SME expected sound level provided asked estimate is from the Clinch River EIS PPE (NRC due to construction 2019c).

activities, measured at 50 ft from the noise source?

Are there large None Not evaluated 160 tons/15 m3 Not applicable 160 tons/15 m3 Responses did not pose any particular environmental quantities of any unique provided lead; borated lead; borated poly; challenges. Any particular unique materials are materials (perhaps poly; graphite; graphite; sodium; necessarily specific to a given design proposed for items not normally used sodium 52.5 MT molten salt deployment. For purposes of impact analysis, the in general office or SME estimate of unique materials should bound industrial buildings) that applicable resource impacts.

will be used in plant construction (e.g.,

Plant Structure and Footprint (continued) graphite)? If so, what are these anticipated volumes?

What is the operational 80 yr Not evaluated 30 yr Not applicable 30 yr Selected longest vendor response value. Prototype life for which the plant is deployment at INL would likely be shorter than the 80 Operational designed? How long do yr operational period chosen by NRC for a 2 to 20 yr 10 yr 10 yr you intend to operate operating commercial reactor.

Parameters the reactor prototype? cycle length McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.3

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale Do you anticipate None Not evaluated No Not applicable No While the PPE assumes that additional modules installing additional provided would not be added over time, particularly for modules incrementally demonstration projects, it is possible that multiple over time? modules could be proposed for certain microreactor applications. This would have additive implications as well as potential cumulative impacts.

What is the reactor heat Water 52.5 MT Liquid lead Not applicable 52.5 MT molten salt Molten salt value obtained by scaling Molten Salt transfer material molten salt 73 tons Initially initial loading Reactor Experiment (MSRE) (8 MWt) coolant quantity (coolant)? How much is initial loading 0 tons Annually to 60 MWt (ORNL 2015) required 150 MT lead initial initially/annually? 150 MT lead loading Lead value obtained by scaling and rounding vendor initial loading response (30 MWt) to 60 MWt What is the anticipated None Mechanical Varied Not applicable Mechanical draft It is anticipated that mechanical draft cooling towers, technology (or provided draft cooling cooling tower in general, will have the most resource-intensive type technologies) for the tower of plant heat sink.

normal plant heat sink?

What are the maximum 1,000 gpm 335 gpm 450 gpm Not applicable 450 gpm (average) The bounding vendor value was chosen as the PPE and average daily water (average) (average) value because it exceeds the SME calculated value.

use requirements for For air-cooled plant cooling and For air-cooled reactors, 25 gpm For air-cooled reactors, the PPE water use includes service water systems, reactors, 25 non-cooling uses, which were based upon scaling including potable and gpm non-cooling-water use from the Clinch River EIS, and sanitary water use (if potable/sanitary use assumed to be 100 gpd per required)? member of the vendor-provided or estimated operations work force.

What are the expected 600 gpm 102 gpm 400 gpm Not applicable 400 gpm The vendor value was chosen as the PPE value for characteristics of plant consistency with the water demand estimate.

water discharges (if Operational Parameters (continued)

For air-cooled For air-cooled any)? reactors, 25 reactors, 25 gpm For air-cooled reactors, the PPE discharge includes gpm non-cooling system wastewater and potable/sanitary wastewater, assumed to be equivalent to the water use rate for these purposes.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.4

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale Blowdown temperature Within Not evaluated Not EvaluatedWithin Within applicable Questionnaire did not include this parameter.

and constituent applicable applicable Clean Water Act Discharges mainly from plant blowdown are regulated concentrations Clean Water Clean Water limits under a Clean Water Act permit.

Act limits Act limits What are the chemical Within See Clinch None Provided DOE O 458.1 See Clinch River Clinch River ER provided anticipated constituents and and radionuclide applicable River Table - (DOE-STD- Table - Projected concentrations associated with blowdown, which constituents of the plant Clean Water Projected 1196 and Blowdown would be assumed to be the dominant portion of discharges, and Act limits Blowdown DOE-STD- Constituents and liquid nonradioactive waste. Not all of these maximum and expected Constituents 1153) (or 10 Concentrations, constituents would be relevant to each microreactor concentrations/activities and CFR Part 20 Table C.2. design, but these values represent a reasonable in the discharge (if Concentrations Appendix B) estimate for values that could be included in a available)? (Table C.2). for both liquid surrogate plant.

and gaseous effluents and 40 CFR Part 61 Subpart H for gaseous effluents What is the fuel source None Not evaluated Diesel Not applicable Two diesel The largest value from vendor responses was and size of auxiliary provided 50-150 kW 50-150 kW standby selected. Two generators are assumed for boilers, emergency Standby Power power generators redundancy to power plant safety systems in the power systems and event of loss of offsite power.

standby power systems (if applicable) (fuel source, MW)?

Emissions from Criteria Not evaluated Not evaluated Criteria Criteria pollutants Questionnaire did not include this parameter. Clean construction equipment pollutants pollutants are are less than Clean Air Act requires a conformity determination for Operational Parameters (continued) and standby power are less than less than Air Act de minimus maintenance or nonattainment areas that exceed de equipment during Clean Air Act Clean Air Act levels minimus values. Not applicable to attainment areas, operations de minimus de minimus so this would be bounding for INL.

levels levels McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.5

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale How much hazardous, None 19 MT None provided Small quantity 19 MT radioactive Reference molten salt reactor consumes 1,930 kg of radioactive, and mixed provided radioactive generators waste (fuel) 19.7 percent enriched U and 3,290 kg of Th annually; waste would be waste (fuel) produce more scaled from reference reactors power (500 MWt) to generated during than 100 315 MT molten salt microreactor power (60 MWt). Assumes 30 yr operations, and where 315 MT molten kilograms, but (mixed) demonstration.

would it be salt (mixed) less than dispositioned? 1,000 Initial loading of molten salt value taken by scaling Hazardous kilograms of MSRE (8 MWt) coolant quantity to 60 MWt (ORNL waste hazardous 2015). Assuming MSRE initial loading of 52.5 MT of generation waste a molten salt would be replenished every 5 yr, two amount would month. loadings would be needed for the assumed 80 yr be within the demonstration. The molten salt spent fuel and coolant criteria of a would be classified as either high-level mixed waste small quantity or mixed transuranic waste (depending on the spent generator fuel processing).

These waste volumes reflect waste that would be generated from within the reactor vessel. For estimates of total radioactive waste generation (excluding spent fuel) see estimates of the total number of shipments and volume of radioactive waste.

RCRA requires waste management for hazardous waste and sets a volume amount of generations of no more than 1,000 kg a month. This volume could be used as a bounding value.

What is the stack exit None The largest value from vendor responses was Operational Parameters (continued)

Not evaluated 10 ft/s Not applicable 10 ft/s velocity? Provided selected.

What amount of noise 65 dBA at Not evaluated None provided Not applicable 65 dBA at site The value from the NRC estimate in ANR GEIS was would be generated 50 site boundary selected and it is consistent with NRC Environmental ft from the source and at boundary Standard Review Plans (NRC 2013).

the site boundary?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.6

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale What is the form of the None Molten salt Molten salt Not applicable Molten salt Fuel types could include UO2, MOX, Metal (U, U fuel associated with provided alloys, Pu-containing alloys), TRISO, molten salt, your design? uranium nitride, uranium carbide, QUADRISO, cermet, accident-tolerant fuel. Emission release mechanisms from molten salt are different from LWRs; expect that molten salt will have upper bounding impacts compared to other fuel technologies.

What is the annual None 0.5 MT (5 MT 0.5 MT Not applicable 0.5 MT (5 MT initial The largest value from vendor responses was average fuel provided initial fuel fuel loading) selected.

requirement (metric loading) tons) per module?

Where would fuel be None Not evaluated Existing DOE Not applicable Offsite commercial Multiple vendors assumed that the fuel would come obtained? provided supply source from an existing DOE supply at INL, while other vendors would source the fuel from offsite commercial sources. For purposes of developing a surrogate reactor, the PPE assumes that fuel would Fuel be obtained from offsite sources.

What is the total number None 10 shipments 1 shipment, Not applicable 10 shipments over Unirradiated fuel shipments scaled to 60 MWt from of shipments and MTU provided over the 30 yr 3 MTU the 30 yr life of the surrogate SMR from Clinch River ESP (NUREG-for unirradiated fuel life of the plant. plant. 2226, NRC 2019a, Table 6-4).

shipped to reactor or site? 45 MTU total 45 MTU total MTU scaled to 1,000 MWt from Clinch River ESP (NUREG-2226, NRC 2019a, Table 6-10).

Total number of None 49 shipments 1 shipment, 14 Not applicable 49 shipments over Radioactive waste shipments and volume scaled to shipments and volume provided over the 30 yr m3 the 30 yr life of the 60 MWt from surrogate SMR from Clinch River ESP of radioactive waste life of the plant. plant. Volume of (NUREG-2226, NRC 2019a, Table 6-14).

shipments from Volume of each shipment is reactor/site? each shipment 2.34 m3. Total These values are used as a bounding measure of is 2.34 m3. volume = 113 m3 radioactive waste generation (excluding spent fuel) but do not account for differences in design or unique waste streams from advanced reactors.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.7

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D and C) Appendix A) Limit Value Bounding Value Source/Rationale What is the radionuclide None See Fission None provided Not applicable Fission Product See Appendix C.7 inventory for irradiated provided Product Inventory fuel at time of shipment Inventory (Appendix C.7)

(Ci/MTU by (Appendix C.7) radionuclide)?

How will the reactor, None Truck Truck or rail Not applicable Truck Truck transportation is assumed based upon internal fresh fuel and other provided research value.

large components be transported to the site?

Is the reactor designed None 5 MTU [full Yes, online Not applicable Yes, 5 MTU (full Assumed that the reactor would be refueled in order to be refueled? If so, at provided core refueling] and continuous core refueling), to develop a more robust bounding impact. Online what frequency (year)? refueling online and and continuous refueling was assumed, which may What MTU per continuous increase impacts associated with radioactive and refueling? refueling nonradioactive emissions.

What are the source None See Fission None provided Not applicable Fission Product See Appendix C.7. The analysis uses general cases, terms for routine provided Product Inventory instead of specific designs, to calculate the releases (if any) per Inventory (Appendix C.7) radionuclide inventory.

module and design- (Appendix C.7) basis accidents?

Are there any unique None Not Evaluated None Not applicable None No unique fuel storage or cooling requirements fuel storage or cooling provided identified by microreactor vendors.

requirements Fuel (continued) associated with the fuel?

How and where would None 89 shipments Onsite storage Not applicable 89 irradiated fuel HALEU and all spent fuel used for the Oklo spent fuel be provided of irradiated shipments over 30 application would stay at the INL site post-dispositioned? fuel over the yr life of the plant. demonstration. (Oklo 2020c) 30 yr life of the plant. Offsite storage or Irradiated fuel shipments scaled to 60 MWt from disposal. surrogate SMR from Clinch River ESP (NUREG-Onsite storage, Treatment, storage, 2226, NRC 2019a, Table 6-10). The Clinch River or offsite and disposal in ESP assumed that fuel would be dispositioned to storage or accordance with Yucca Mountain. This assumption is not carried disposal applicable legal forward into this PPE, but the number of shipments is requirements. scaled as a bounding value.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.8

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D and C) Appendix A) Limit Value Bounding Value Source/Rationale How many workers will 150 150 None provided Not applicable 150 The largest number of workers from NRC ANR GEIS be onsite for was selected to bound impacts.

construction?

What is the anticipated None 6 months 24 months Not applicable 24 months The largest value from the vendor responses was construction period? provided selected and is consistent with the SME estimate.

What is the number of 50 27 None provided Not applicable 50 The largest value from NRC ANR GEIS was selected total permanent staff to to bound impacts.

support operations?

What is the number of 100 21 None provided Not applicable 100 The largest value from the NRC ANR GEIS was temporary staff during selected to bound impacts.

refueling (if planned)?

What is the number of None 20 None provided Not applicable N/A Assumed a single module for purposes of this PPE, temporary staff during Workforce provided thus no temporary staff are needed.

additional module installation (if planned)?

What are the distances None 500 ft 500 ft Not applicable 500 ft Internal research estimate is consistent with vendor from radiation sources provided response.

to the nearest involved worker?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.9

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.1. Microreactor PPE Data Sources and Methodology(a) (continued)

Information Sources Internal NRC ANR Research Vendor GEIS Value Value (from Bounding PPE (from Appendices B Value (from Regulatory Microreactor Section Parameter Appendix D) and C) Appendix A) Limit Value Bounding Value Source/Rationale Do you plan to None Not evaluated Yes Not applicable Yes It is assumed that the prototype would be decommission and provided decommissioned to bound impacts associated with remove the prototype land use, fuel, transportation, and workforce.

from the INL site?

What is the number of None 150 None provided Not applicable 150 It is assumed that the number of staff needed during temporary staff during provided decommissioning would be similar to those needed decommissioning (if during construction.

planned)?

What is the number of None Not evaluated 18 months Not applicable 18 months Selected the largest value from vendor responses.

months from start of provided decommissioning to completion (if planned)?

How much waste would None Bounded by None provided Not applicable Bounded by the The anticipated volumes of wastes evaluated in be generated during provided the waste waste streams NUREG-0586 were based on industry decommissioning (if streams evaluated in decommissioning experience as of 2002. Appendix G planned)? evaluated in NUREG-0586 of NUREG-0586, Radiation Protection Decommissioning NUREG-0586 Considerations for Nuclear Power Facility Decommissioning summarizes effluent releases for operating facilities and decommissioning facilities.

Low-level waste volume estimates for decommissioning facilities are presented in Appendix K of NUREG-0586.

AC = alternating current; ANR = advanced nuclear reactor; CFR = Code of Federal Regulations; DOE = U.S. Department of Energy; ; EIS = environmental impact statement; ER

Environmental Report; ESP = early site permit; GDC = General Design Criteria; GEIS = generic environmental impact statement; HALEU = high-assay low-enriched uranium; HTGR = high-temperature gas-cooled reactor; INL = Idaho National Laboratory; LMR = liquid metal reactor; LWR = light-water reactor; MSR = molten salt reactor; MSRE

Molten Salt Reactor Experiment; NRC = U.S. Nuclear Regulatory Commission; PPE = plant parameter envelope; RCRA = Resource Conservation and Recovery Act; ROW =

right-of-way; SME = subject matter expert; SMR = small modular reactor.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.10

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2 Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology Information Sources Internal NRC ANR Research Vendor Versatile Test Small- to Medium-GEIS Value(a) Value (From Bounding Reactor Draft Sized Advanced PPE (From Appendices B Value (From EIS Value Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) Limit Value Value Source/Rationale What is your design None HTGR, BWR, GE Hitachi Not applicable Not Applicable for Plant type itself is not relevant to the type? provided LMR PRISM bounding environmental analysis; parameters therein design parameters that have an environmental interface are considered.

How many units do None Not evaluated 1-4 units 1 Not applicable 1 While the PPE considers installation of one you plan to install? provided unit, multiple units may be proposed, for instance to demonstrate the capability of following increases in electricity demand over time. This would have potential impacts on the extent and timing of resource analyses, including cumulative impacts.

What is the output of 60 Not evaluated 950 MWt 300 MWt Not applicable 1,000 MWt Based upon the largest vendor response, your design (per unit)? MWt/20MWe 1,000 MWt was selected as a bounding value to account for potentially larger plants.

Is your reactor None Not evaluated No No Not applicable No Reactor is not assumed to be mobile.

designed to be mobile? provided If the reactor is None Not evaluated N/A N/A Not applicable N/A Reactor is not assumed to be transportable.

designed to be provided transportable, what are the total number of shipments and weight of reactor, fuel, and its Plant Design packaging?

Describe your power None Not evaluated Rankine Cycle N/A N/A Power conversion does not itself have an conversion system. provided environmental nexus.

Will offsite power Required. Required, Not required Yes, two 230 General Two 230 kV GDC 17 requires two offsite sources of sources be required to Offsite ROW assuming kV Design transmission lines power. The requirement for access to the maintain functioning of 1,000 ft x 100 compliance transmission Criteria 17 required. Offsite existing onsite INL transmission system structures, systems, ft (new) or with General lines available ROW 1,000 ft x 100 would bound all designs. Both substation and components within or Design Criteria ft (new) or within or and transmission interconnections are important to safety adjacent to 17 adjacent to existing assumed to be required. Length and breadth following loss of onsite existing ROW ROW of transmission line right-of-way and size of AC power? If so, what the switchyard will depend on final site transmission voltage location.

would be required from offsite power sources?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.11

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What support facilities None Not evaluated Cooling-water Feedstock Not applicable Multiple support Vendor responses included representative (fuel storage and provided system; preparation facilities facilities that may be required associated handling, waste switchyard/ facility, fuel with any given design.

treatment, etc.) are transformers; fabrication necessary for your chemical/gas/ facility, plant design? fuel storage, experiment potable water support supply; areas, post-wastewater irradiation system, examination including facility, spent retention fuel treatment basins and facility, onsite associated spent fuel pad discharge equipment; liquid radwaste system; fire protection and emergency response buildings; Administration/

Maintenance Building(s);

Security Facility; Plant Design (continued)

Chemistry and Meteorology Facility; Radioactive Waste Storage Facility (Region/

Country Dependent);

various offsite facilities McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.12

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What is the tallest 50 ft Not evaluated 75 ft structure 90 ft Not applicable 75 ft structure Selected largest values from vendor structure and what is (structure) (experiment responses to better bound potential visual, the maximum structure 87 ft stack support area) 87 ft stack height scenic, and land use impacts.

height (structure, ft)? None height 190 ft (cooling What is the stack Provided chimneys) Although the VTR is larger, aboveground height? (stack) PPE estimates are generally consistent; the height of the aboveground portion of the VTR stack is generally consistent with the vendor-provided estimates.

What is the maximum 50 ft Not evaluated 155 ft 93 ft Not applicable 155 ft Selected largest value consistent with depth of excavation? vendor responses.

What is the temporary 50 ac 60 ac 58 ac 100 ac Not applicable 100 ac Selected largest value consistent with disturbed acreage vendor responses and the VTR estimated during construction, acreage.

including parking and laydown?

What is the permanent 30 ac 50 ac 43 ac 25 ac Not applicable 50 ac Selected largest value consistent with disturbed acreage, vendor responses, rounded up to consider including parking lots, potential additional acreage needed for air-ponds, substations, cooling.

and other plant support facilities?

What is the maximum None 101 dB at 50 ft Question not Imperceptible Not applicable 101 dB at 50 ft Questionnaire did not include this Plant Structure and Footprint expected sound level provided asked at the INL site parameter. SME estimate is from the Clinch due to construction boundary and River EIS PPE (NRC 2019c).

activities, measured at the closest 50 ft from the noise receptor source?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.13

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale Are there large None 230 MT Graphite, 280 No Not applicable 280 m3 lead Responses did not pose any particular quantities of any provided Graphite, 65 m3 lead information environmental challenges. Any particular unique materials 3 m lead, 2,020 unique materials are necessarily specific to provided (perhaps items not m3 sodium a given design proposed for deployment. For Plant Structure and normally utilized in purposes of impact analysis, the SME general office or estimate of unique materials should bound industrial buildings) applicable resource impacts.

that will be utilized in plant construction Footprint (continued) (e.g., graphite)? If so, what are these anticipated volumes?

What is the operational 80 yr Not evaluated 80 yr 60 yr Not applicable 80 yr Selected longest vendor response value.

life for which the plant Prototype deployment at INL would likely be is designed? How long 2 to 20 yr 80 yr 80 yr shorter than the 80 yr operational period do you intend to operating chosen by NRC for a commercial reactor.

operate the reactor cycle length prototype?

Do you anticipate None Not evaluated Yes No Not applicable Yes Vendor responses stated that multiple installing additional provided modules may be installed. This would have modules incrementally additive implications as well as potential over time? cumulative impacts.

What is the reactor Water 870 MT molten Liquid metal Sodium Not applicable Various Molten salt value taken by scaling reference heat transfer material salt, 65 m3 (e.g., sodium, MSR (8 MWt) coolant quantity to 1,000 MWt (coolant)? How much lead, 2,020 m3 lead, lead- (ORNL 2015) is required sodium bismuth); gas initially/annually? (e.g., helium); Lead value taken by scaling reference LFR water (280 MWt) coolant quantity to 1,000 MWt (Cinotti et al. 2010)

Operational Parameters Sodium value taken by scaling reference SFR (400 MWt) coolant quantity to 1,000 MWt (Cabell 1980)

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.14

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What is the anticipated None Mechanical Mechanical Air-cooled Not applicable Mechanical draft It is anticipated that mechanical draft cooling technology (or provided draft cooling draft cooling heat cooling tower towers, in general, will have the most technologies) for the tower tower or air- exchangers resource-intensive type of plant heat sink.

normal plant heat sink? cooled condenser What are the maximum 1,000 gpm 5,850 gpm 7,500 gpm 6.8 million Not applicable For water-cooled Cooling-water use was estimated assuming and average daily maximum; gallons/yr/365 reactors, 5,850 gpm the bounding reactor operates at 33 percent water use For air-cooled 4,200 gpm = about (maximum) thermal efficiency, which is considered a requirements for plant reactors, 415 average 18,000 5,850 gpm (average) lower bound on the efficiency. The 7,500 cooling and service gpm gallons per gpm vendor value seems excessively high Operational Parameters water systems, day For air-cooled given the power output and efficiency of the including potable and reactors, 415 gpm reactor. This water use estimate would also sanitary water use (if bound those reactor designs that are only required)? using process heat rather than electricity.

For air-cooled reactors, the PPE water use includes non-cooling uses, which were based upon scaling non-cooling-water use from Clinch River EIS, and potable/sanitary (continued) use assumed to be 100 gpd per member of the vendor-provided or estimated operations work force.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.15

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What are the expected 600 gpm 1,775 gpm Various, 4.4 million Not applicable 1,775 gpm Discharge includes blowdown from the characteristics of plant including no gallons cooling towers with contributions from non-water discharges (if For air-cooled discharges annually For air-cooled cooling systems and potable/sanity uses.

any)? reactors, 415 anticipated (about 8.4 reactors, 415 gpm The blowdown rate depends on the cycles of gpm gpm if concentration during tower operation, which continuous), was assumed to be four. Two cycles of including the concentration would result in a larger volume discharge rate, but four cycles of required for concentration was selected for the PPE personnel use because this maximizes nonradioactive and sanitation, concentrations in the discharge. Minimizing fire protection the liquid discharge rate is likely to be water, and desirable at INL.

demineralized For air-cooled reactors, the PPE discharge water. includes non-cooling system wastewater and No water potable/sanitary wastewater, assumed to be required for equivalent to the water use rate for these reactor purposes.

operation Blowdown Within Not evaluated Question not No information Within Within applicable Questionnaire did not include this Temperature and applicable asked provided applicable Clean Water Act parameter. Discharges mainly from plant Constituent Clean Water Clean Water limits blowdown are regulated under a Clean Concentrations Act limits Act limits Water Act permit.

What are the chemical Within See Clinch TBD No information DOE O 458.1 See Clinch River Radionuclides in liquid discharge will be and radionuclide applicable River Table - provided (DOE-STD- Table - Projected dependent on the specific reactor. Discharge constituents of the Clean Water Projected 1196 and Blowdown can be assumed to be diluted to meet the 10 plant discharges, and Act limits Blowdown DOE-STD- Constituents and CFR Part 20 Appendix B, Table 2 limits at maximum and Constituents 1153) (or 10 Concentrations, the point of discharge. Nonradioactive expected and CFR Part 20 Table C.2. constituents of the liquid discharge will be concentrations/activitie Concentrations Appendix B) determined by the source water used for s in the discharge (if (Table C.2) for both liquid cooling, by the cycles of concentration used Operational Parameters (continued) available)? and gaseous in cooling-tower operation, and by additives effluents and used in plant processes. The Clinch River 40 CFR Part discharge was assumed to be representative 61 Subpart H with some consideration of the typical for gaseous source water at INL. The microreactor PPE effluents assumed four cycles of concentration, which would also be a reasonable bounding McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.16

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance assumption for the small- to medium-sized advanced reactors.

Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What is the fuel source None 50 MWt oil 4 MWe natural 4.7 million ft3 Not applicable 50 MWt oil fired; 15 Selected based upon a review of publicly and size of auxiliary provided fired; 15 MWe gas or diesel of propane MWe Sentry turbine available documentation, including vendor boilers, emergency Sentry turbine auxiliary boiler; per year websites and other literature.

power systems and 1 MWe standby power standby power systems (if applicable) (gas, diesel, (fuel source, MW)?

Operational Parameters battery)

Emissions from Criteria Not Evaluated Not asked Well below Criteria Criteria pollutants Questionnaire did not include this construction equipment pollutants are EPA PSD pollutants are are less than Clean parameter. Clean Air Act requires a and standby power less than permitting less than Air Act de minimus conformity determination for maintenance or equipment during Clean Air Act threshold of Clean Air Act levels nonattainment areas that exceed de operations de minimus 250 tons per de minimus minimus values. Not applicable to attainment levels year for a levels areas, so this would be bounding for INL.

(continued) criteria pollutant McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.17

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale How much hazardous, None 836 MT Various 45 spent Not applicable 836 MT radioactive Reference molten salt reactor consumes radioactive, and mixed Provided radioactive driver fuel waste (fuel) 1,930 kg of 19.7 percent enriched U and waste would be waste (fuel) assemblies 3,290 kg of Th annually; scaled from generated during per year, 13,920 MT molten reference reactors nominal power (500 operations, and where 13,920 MT initially stored salt (mixed) MWt) to 1,000 MWt. Assumes 30 yr would it be molten salt onsite demonstration.

dispositioned? (mixed) 18,460 total Initial loading of molten salt value taken by shipments scaling reference MSR (8 MWt) coolant quantity to 1,000 MWt (ORNL 2015).

Assuming initial loading of 870 MT of molten salt would be replenished every 5 yr, 16 loadings would be needed for assumed 80 yr demonstration. The molten salt spent fuel and coolant would be classified as either high-level mixed waste or mixed transuranic waste (depending on the spent fuel processing)

Vendor responses indicate that fuel would come from existing INL feedstock. As a result, the PPE assumes that disposition would remain at INL.

These waste volumes reflect waste that would be generated from within the reactor vessel. For estimates of total radioactive waste generation (excluding spent fuel), see estimates of the total number of shipments and volume of radioactive waste.

What is the stack exit None Not evaluated 58 ft/s No Not applicable 58 ft/s The largest value from vendor responses velocity? provided information was selected.

Operational Parameters (continued) provided What amount of noise 65 dBA at site Not evaluated 70 dBA at 50 ft Imperceptible Not applicable 65 dBA at site The value from the NRC estimate in ANR would be generated 50 boundary from cooling at the INL site boundary GEIS was selected and it is consistent with ft from the source and tower; <55 boundary and NRC Environmental Standard Review Plans at the site boundary? dBA at site the closest (NRC 2013).

boundary receptor McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.18

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What is the form of the None Molten salt Various, Uranium- Not applicable Molten Salt Molten salt, TRISO, Uranium Oxide, HALEU, fuel associated with provided including metal plutonium- U-Zr alloy. Emission release mechanisms your design? fuel, TRISO in zirconium from molten salt are different from LWRs; graphite blocks alloy fuel (U- expect that molten salt will have upper 20Pu-10Zr) or pebbles, bounding impacts compared to other fuel uranium oxide. technologies.

What is the annual None 8 MT 4.9 MT 1.8 MT 8 MT The largest value from vendor responses average fuel provided annually was selected, scaled to 1000 MWt.

requirement (metric (~1.3-1.4 MT tons) per module? uranium,

~0.4-0.54 MT plutonium, plus 10% Zr)

Where would fuel be None Not evaluated Various U from within Not applicable Offsite commercial Multiple vendors assumed that the fuel obtained? provided DOE complex source would come from an existing DOE supply at and from INL, while other vendors would source the commercial fuel from offsite commercial sources. For vendors, Pu from within purposes of developing a surrogate reactor, the DOE the PPE assumes that fuel would be complex. Fuel obtained from offsite sources.

Fuel manufactured at either INL or SRS What is the total None 432 shipments 255 fuel blocks 460-550 kg Pu Not applicable 432 shipments over Unirradiated fuel shipments scaled to 1,000 number of shipments provided over the 80 yr per and 1.61-1.92 the 80 yr life of the MWt from surrogate SMR from Clinch River and MTU for life of the plant. module/year; MTU per year plant. ESP (NUREG-2226, NRC 2019a, Table 6-unirradiated fuel 1,972 MTU 10 shipments needed for 1,972 MTU total. 4).

feedstock. 3 shipped to reactor or total. per assemblies per MTU scaled to 1,000 MWt from Clinch River site? module/year shipment. 22 ESP (NUREG-2226, NRC 2019a, Table 6-shipments for 10).

initial loading, VTR refueling will occur more frequently then 15 shipments than for demonstration reactors considered annually = 922 in this PPE to support its research mission.

total While the VTR value would bound impacts, unirradiated this value is not representative of the fuel shipments anticipated fueling cycle for anticipated for 60 yr advanced reactor designs.

operational life McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.19

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale Total number of None 2,160 ~4 Total Up to 423 Not applicable 2,160 shipments Radioactive waste shipments and volume shipments and volume provided shipments over Shipments, shipments over the 80 yr life of scaled to 1,000 MWt from surrogate SMR of radioactive waste the 80 yr life of ~24 M3 annually = the plant. Volume of from Clinch River ESP (NUREG-2226, NRC shipments from the plant. 25,380 each shipment is 2019a, Table 6-14).

reactor/site? Volume of shipments for 2.34 m3.

each shipment 60 yr Total volume= 4,981 These values are used a bounding measure is 2.34 m3. operational m3. of radioactive waste generation (excluding Total volume = life spent fuel) but do not account for differences 4,981 m3. in design or unique waste streams from advanced reactors Similar to the shipments of unirradiated fuel, the VTR parameter for radioactive waste shipments exceeds the PPE bounding parameter related to NRIC prototypes, based upon a different research and mission goal. While the VTR value would bound impacts, this value is not representative of the anticipated radioactive waste shipments Fuel (continued) for anticipated advanced reactor designs.

What is the None See Fission No information No Not applicable Fission Product See Appendix C.7 radionuclide inventory provided Product provided information Inventory (Appendix for irradiated fuel at Inventory provided C.7) time of shipment (Appendix C.7)

(Ci/MTU by radionuclide)?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.20

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale How will the reactor, None Truck or rail Unirradiated Not applicable Truck Truck transportation is assumed for both fresh fuel and other provided fuel to be microreactors and small- to medium-sized large components be shipped via advanced reactors transported to the site? Secure Transporta-tion Asset (STA; truck transport only)

Is the reactor designed None Daily refueling Various, Typically <17 Not applicable Daily refueling of Assumed that the reactor would be refueled to be refueled? If so, at Provided of 10.6 kg depending on driver fuel 10.6 kg enriched U in order to develop a more robust bounding what frequency (year)? enriched U and whether assemblies and 18 kg Th; impact. Online and continuous refueling What MTU per 18 kg Th; counting total (~1/4 of core) annual requirement assumed, which may increase impacts refueling? annual amount of fuel replaced per 3.9 MT enriched U, associated with radioactive and requirement or annual fuel refueling/ 6.6 MT Th. nonradioactive emissions.

3.9 MT consumption three enriched U, 6.6 ~100-day Continuous refueling quantity scaled from MT Th. operating reference MSR (500 MWt) to 1,000 MWt.

cycles per year/~0.6 MT Th per Fuel (continued) refueling Up to 45 driver fuel assemblies per year/

~1.8 MT Th What are the source None See Fission TBD No Not applicable Fission Product See Appendix C.7. The analysis uses terms for routine provided Product information Inventory (Appendix general cases, instead of specific designs, to releases (if any) per Inventory provided C.7) calculate the radionuclide inventory.

module and design- (Appendix C.7) basis accidents?

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.21

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale Are there any unique None Sodium or lead TBD Cool in-vessel Not applicable Sodium or lead pool No unique fuel storage or cooling fuel storage or cooling provided pool depending for ~1 year, depending on requirements identified by small- to medium-requirements on coolant then wash off coolant type; sized advanced reactor vendors.

associated with the type; separate external separate storage sodium and fuel? storage area transfer to area required with Fuel from liquid metal reactors will require required for cask on spent liquid metal reactors compatible pool/storage vessel.

liquid metal fuel pad to reactors cool further.

Fuel then chopped, consolidated, sodium removed, and diluted (likely with scrap metal from driver fuel assembly).

Mixture packaged in containers, placed in storage casks, and stored on spent fuel pad Fuel (continued) until shipped offsite How and where would None 3,944 Various; cask Ultimately an Not applicable 3,944 shipments of HALEU and all spent fuel used for the Oklo spent fuel be provided shipments of stored for offsite storage irradiated fuel over application would stay at the INL site post-dispositioned? irradiated fuel future or disposal the 80 yr life of the demonstration. (Oklo 2020c) over the 80 yr disposition, facility plant.

life of the plant. either onsite, Onsite storage, or Irradiated fuel shipments scaled to 1,000 Onsite storage, intermediate offsite storage or MWt from surrogate SMR from Clinch River or offsite offsite storage, disposal ESP (NUREG-2226, NRC 2019a, Table 6-storage or bore-hole, or 10).

disposal permanent repository.

Recycling is possible.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.22

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale How many workers will 150 909 maximum 900 1,400 Not applicable 909 maximum onsite Scaled down the values analyzed for the be onsite for onsite at one at one time Clinch River ESP to a 1,000 MWt reactor.

construction? time construction construction workforce; workforce; maximum total maximum total construction construction workforce of 1,400 workforce of at peak 1,363 at peak What is the anticipated None 45 months 54 months 51 months Not applicable 54 months Consistent with vendor response construction period? provided What is the number of 50 207 420 600 Not applicable 207 Scaled down the values analyzed for the total permanent staff to Clinch River ESP to a 1,000 MWt reactor.

support operations?

The VTR is a research reactor that will involve a larger staff to support research operations; this is a higher value than would Workforce be expected for demonstration prototypes.

The construction workforce estimate is consistent with the VTR data (DOE 2020c).

and bounds the VTR value peak. The VTR value of 600 also includes the workforce associated with activities outside of reactor operations (e.g., fuel fabrication, post-irradiation examination of experiments, experiment design and support staff) that are not relevant to the demonstration reactors considered in this PPE.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.23

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale What is the number of 100 413 <125 No Not applicable 413 Scaled down the values analyzed for the temporary staff during information Clinch River ESP to a 1,000 MWt reactor.

refueling (if planned)? provided What is the number of None 413 900 N/A Not applicable 413 Scaled down the values analyzed for the Workforce temporary staff during provided Clinch River ESP to a 1,000 MWt reactor.

additional module installation (if planned)?

(continued) What are the distances None 500 ft ~20 m Nearest Not applicable ~20 m Consistent with vendor response from radiation sources provided uninvolved to the nearest involved worker: 330 ft worker?

Do you plan to None Not evaluated TBD No Not applicable Yes It is assumed that the prototype would be decommission and provided information decommissioned to bound impacts remove the prototype provided associated with land use, fuel, from the INL site? transportation, and workforce.

What is the number of None Not evaluated 450 No Not applicable 450 Consistent with vendor response temporary staff during provided information decommissioning (if provided planned)?

What is the number of None Not evaluated 10 yr No Not applicable 10 yr Selected the largest value from vendor months from start of provided information responses.

decommissioning to provided completion (if planned)?

How much waste None Bounded by No Not applicable Bounded by the The anticipated volumes of wastes would be generated provided the waste information waste streams evaluated in NUREG-0586 were based on during streams provided evaluated in NUREG- industry decommissioning experience as of decommissioning (if evaluated in 0586 2002. Appendix G of NUREG-0586, Decommissioning planned)? NUREG-0586 Radiation Protection Considerations for Nuclear Power Facility Decommissioning summarizes effluent releases for operating facilities and decommissioning facilities.

Low-level waste volume estimates for decommissioning facilities are presented in Appendix K of NUREG-0586.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.24

Advanced Nuclear Reactor Plant Parameter Envelope and Guidance Table E.2. Small- to Medium-Sized Advanced Reactor PPE Data Sources and Methodology (continued)

Information Sources Internal NRC ANR Research Vendor Small- to Medium-GEIS Value(a) Value (From Bounding Versatile Test Sized Advanced PPE (From Appendices B Value (From Reactor Draft Regulatory Reactor Bounding Section Parameter Appendix D) and C) Appendix A) EIS Value Limit Value Value Source/Rationale (a) Note that because NRCs ANR GEIS PPE/SPE generally focuses on microreactors, many of the parameters are not applicable to these small- to medium-sized advanced reactors.

However, some of the parameters would provide appropriate bounding values regardless of the size of the reactor and are therefore included in this table.

AC = alternating current; ANR = advanced nuclear reactor; BWR = boiling-water reactor; CFR = Code of Federal Regulations; DOE = U.S. Department of Energy; EIS = environmental impact statement; ER = Environmental Report; ESP = early site permit; GDC = General Design Criteria; GEIS = generic environmental impact statement; HALEU = high-assay low-enriched uranium; HTGR = High-Temperature Gas-Cooled Reactor; INL = Idaho National Laboratory; LFR = Lead-Cooled Fast Reactor; LMR = Liquid Metal Reactor; LWR = light-water reactor; MSR = Molten Salt Reactor; MSRE = Molten Salt Reactor Experiment; NRC = U.S. Nuclear Regulatory Commission; PPE = plant parameter envelope; RCRA = Resource Conservation and Recovery Act; ROW = right-of-way; SME = subject matter expert; SMR = small modular reactor; TBD = to be developed; Th = Thorium; TRISO = tri-structural isotropic.

McDowell, Goodman NRIC-21-ENG-0001; PNNL-30992 2/18/2021 E.25