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Effluent Release (GALE Replacement) Methodology and Results

TR-123242-NP Revision 1 Licensing Technical Report

Effluent Release (GALE Replacement) Methodology and Results

August 2023 Revision 1 Docket: 52-050

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Effluent Release (GALE Replacement) Methodology and Results

TR-123242-NP Revision 1 Licensing Technical Report

COPYRIGHT NOTICE

This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copi es of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a licens e, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations.

Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.

TR-123242-NP Revision 1 Licensing Technical Report

Department of Energy Acknowledgement and Disclaimer

This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights.

Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not ne cessarily state or reflect those of the United States Government or any agency thereof.

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Table of Contents

Abstract................................................................... 1 Executive Summary.......................................................... 2 1.0 Introduction.......................................................... 3 1.1 Purpose.............................................................. 3 1.2 Scope................................................................ 3 1.3 Abbreviations.......................................................... 4 2.0 Background.......................................................... 5 2.1 GALE Code Applicability................................................. 5 2.2 Theory............................................................... 7 2.3 Regulatory Requirements................................................ 8 3.0 Source Term Production................................................ 9 3.1 Water Activation Products................................................ 9 3.1.1 Tritium........................................................ 10 3.1.2 Carbon-14..................................................... 13 3.1.3 Nitrogen-16..................................................... 14 3.1.4 Argon-41....................................................... 14 3.2 Corrosion and Wear Activation Products.................................... 15 3.2.1 Mechanism Overview............................................. 15 3.2.2 Modeling Corrosion and Wear Activation Products...................... 16 3.3 Fission Products....................................................... 17 3.3.1 Software Use and Qualification..................................... 17 3.3.2 TRITON Code Sequence.......................................... 18 3.3.3 ORIGEN (ORIGEN-ARP and ORIGEN-S) Code Sequences............... 19 4.0 Radionuclide Transport, Removal Mechanisms, and Release................ 20 4.1 Primary Coolant Water System........................................... 20 4.1.1 Water Activation Products......................................... 20 4.1.2 CRUD......................................................... 22 4.1.3 Fission Products................................................. 22 4.1.4 Primary Coolant Activity Concentrations.............................. 23 4.2 Secondary Coolant Water System......................................... 24 4.3 Chemical and Volume Control System..................................... 27 4.4 Reactor Pool and Spent Fuel Pool......................................... 28

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Table of Contents

4.5 Airborne Activity....................................................... 30 4.5.1 Waste Gas Processing System..................................... 30 4.5.2 Steam Generator Blowdown System................................. 30 4.5.3 Condenser Air Ejector Exhaust..................................... 30 4.5.4 Containment Purge Exhaust....................................... 31 4.5.5 Ventilation Exhaust Air from the Radioactive Waste Building and the Reactor Building................................................. 31 4.5.6 Steam Leakage from Secondary System.............................. 31 4.5.7 Reactor Pool Evaporation......................................... 31 4.5.8 Inadvertent Emergency Core Cooling System Actuation Anticipated Operational Occurrence........................................... 32 4.6 Gaseous Radioactive Waste System....................................... 32 4.6.1 Activity Input to the Guard Bed...................................... 32 4.6.2 Activity Input to the Decay Beds..................................... 32 4.7 Liquid Radioactive Waste System......................................... 33 4.7.1 Overall Liquid Radioactive Waste System Flow and Parameters........... 33 4.7.2 Activity Input to Liquid Radioactive Waste Collection Tanks............... 35 4.7.3 Activity Input to the Oil Separators................................... 36 4.7.4 Low-Conductivity Waste Sample Tanks............................... 36 4.7.5 High-Conductivity Waste Sample Tanks.............................. 36 4.8 Plant Effluent Release.................................................. 3 6 4.8.1 Gaseous Effluent Release......................................... 36 4.8.2 Liquid Effluent Release........................................... 38 5.0 Fuel Failure Fraction.................................................. 39 5.1 US Pressurized Water Reactor Fuel Failure History........................... 39 5.2 Fuel Failure Fraction Conclusions......................................... 40 6.0 Summary and Conclusion.............................................. 41 7.0 References.......................................................... 42 7.1 Source Documents..................................................... 42 7.2 Referenced Documents................................................. 42 Appendix A Summary Tables................................................A-1

Effluent Release (GALE Replacement) Methodology and Results

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List of Tables

Table 1-1 Abbreviations.................................................... 4 Table 2-1 (GALE) Applicability Range......................................... 6 Table 3-1 CRUD Isotopic Primary Concentrations............................... 16 Table 4-1 Fuel Isotopic Escape Coefficients................................... 22 Table 4-2 NUREG-0017 and Corresponding NuScale Parameters.................. 25 Table 4-3 Charcoal Decay Bed Information.................................... 33 Table 4-4 Processing Paths for Liquid Radioactive Waste......................... 34 Table 4-5 Decontamination Factors Used in Liquid Radioactive Waste System Processing for Effluent Release..................................... 34 Table 4-6 Expected Liquid Waste Inputs...................................... 34 Table 5-1 Fuel Failure Values.............................................. 39 Table 6-1 Primary Contributors and Methodology Employed for Effluents............. 41 Table A-1 Maximum Fuel Inventory per Assembly (Ci)........................... A-1 Table A-2 Primary and Secondary Coolant Radionuclide Activity Concentrations (Mode 1)....................................................... A-3 Table A-3 Gaseous and Liquid Yearly Efflue nt Release Values for a NuScale Power Plant (with Six Operating Modules).................................. A-5 Table A-4 Fuel Failure Data for U.S. Pressurized Water Reactors with Zirconium-Alloy Cladding....................................................... A-8

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List of Figures

Figure 2-1 NuScale US460 Plant Layout........................................ 8 Figure 3-1 Time Dependent NuScale Isotopic Tritium Production Breakdown in Primary Coolant........................................................ 11 Figure 3-2 Total NuScale Isotopic Tritium Production Breakdown in Primary Coolant.... 12 Figure 3-3 Comparison of GALE, Electric Power Research Institute, and NuScale Yearly Tritium Production.......................................... 13 Figure 4-1 Water Injection and Bleed in the Primary Coolant....................... 21 Figure 4-2 Tritium Reactor Coolant System Balance............................. 21

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Abstract

This technical report describes the methodology used to calculate normal operation, including anticipated operational occurrences (AOOs), a nnual radioactive gaseous and liquid effluents to the environment from an operating NuScale US 460 Power Plant. The application of this methodology demonstrates compliance with regulatory requirements for normal radioactive effluents. There are no exemptions from existing regulations related to radioactive effluents.

Regulatory requirements for effluents consist of a combination of annual release quantities, site boundary concentrations, and doses to members of the public. The methodology presented in this report uses first principles-based calculations where appropriate, combined with recent nuclear industry experience where applicable, and lessons learned where available, to determine NuScale-appropriate primary and secondary coolant concentrations of fission products, along with activated corrosion and wear products and coolant water activation products. These in-plant source terms form the basis for the evaluation of effluents.

The development of an alternate effluent release methodology is necessary because the existing PWRGALE code was developed in the 1980s for eval uation of the traditional large pressurized water reactors (PWRs) of that time and does not appropriately address unique characteristics of the NuScale Power Plant (NPP). The small modular reactor design is smaller, relies upon a significantly different passive design based on the natural processes of conduction, convection, gravity, and natural circulation to ensure safe shutdown, and the design is expandable with up to six NuScale Power Modules (NPMs) within the ov erall plant envelope. While the majority of individual NPM system designs are similar to traditional PWRs, a few systems vary from the large PWRs.

The primary and secondary coolant isotopic distribution is in Table A-2. The total calculated effluents are 850 Ci of gaseous effluent and 1,200 Ci of liquid effluent, with tritium being the largest contributor to both. The isotopic distribution totals are in Table A-3.

Effluent Release (GALE Replacement) Methodology and Results

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Executive Summary

The NuScale Power Plant (NPP) design is similar to large pressurized water reactors (PWRs) in the existing fleet with regard to normal radioacti ve effluent release calculations. The development of an alternate methodology is necessary because the existing PWRGALE code was developed in the 1980s for evaluation of the large PW Rs and does not appropriately address the US460 design. The US460 is smaller. A single NuScale Power Module (NPM) provides approximately eight percent of the electrical output of the example PW R in NUREG-0017 (Reference 7.2.1). Table 2-1 provides a comparison between the NPM and the example PWR in NUREG-0017.

relies upon a different passive design based on conduction, convection, gravity, and natural circulation.

is expandable with up to six NPMs within the overall plant envelope.

While the majority of individual plant system designs are similar to traditional PWRs, a few systems vary from larger PWRs, such as the use of integral helical coil steam generators (SGs).

In addition, there are some hard-coded parameters in the PWRGALE code that are not appropriate for the NPP design.

This technical report describes the methodology used to calculate normal operation, including anticipated operational occurrences (AOOs), r adioactive annual gaseous and liquid effluents to the environment from an operating NPP containing up to six NPMs. This report also includes specific in-plant source terms and results of effluent releases. The application of this methodology demonstrates compliance with regulat ory requirements, including a combination of site boundary isotopic concentrations and off-site dose consequence limits.

The methodology is realistic, yet conservative, us ing first principles-based calculations where appropriate, combined with recent nuclear in dustry experience where applicable, and lessons learned where available. Calculation of effluent s uses conservative yet realistically generated source terms by evaluating radionuclide transport throughout reactor and other radioactive plant systems and by evaluating effluent releases. This technical report documents the appropriate primary and secondary coolant concentrations of fission products, activated corrosion and wear products, and water activation products for the N PP. Source terms also include water activation products produced in the reactor pool, which is a unique design feature.

One important input parameter in this methodology is the assumed fuel failure fraction. Industry operating experience over the past 30 years show s long-term and continuing reductions in fuel failures. Because the annual fuel failure fraction in U.S. PWRs continues to decrease over time with the most recent data ((2(a),(c), showing a minimum value of ((

}}2(a),(c) and a maximum value of 66 rods per million (0.0066 percent), this analysis uses the maximum value of 66 rods per million. More than 90 percent of current U.S. nuclear power plan ts experience no fuel failures. The NPP design includes various design features that further mitigate fuel failure mechanisms. These design features further improve fuel performance. Based on the continued industry trend in fuel performance, the calculation of fission product re lated source term effluents uses a realistic yet conservative fuel failure fraction value.

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1.0 Introduction

1.1 Purpose

This report describes the methodology used to calculate the NuScale Power Plant (NPP) gaseous and liquid effluents to the envir onment during normal operations, including anticipated operational occurrences (AOOs). This report describes a conservative NuScale design-specific, alternative method to NUREG-0017 (Reference 7.2.1).

1.2 Scope

The scope of this report includes the methodol ogy and results of calculating normal gaseous and liquid effluent releases to t he environment associated with a single US460, assuming the combined effect of up to six operating NuScale Power Modules (NPMs), considering AOOs. The report discusses the differences and similarities between the NUREG-0017 methodology and assumptions and the methodology. This report includes specific in-plant source terms and applies to all radioactive plant systems. Releases from these systems through intended (e.g., letdown or discharge) or unintended (e.g., leakage) events may result in an off-site re lease of radioisotopes; this report explains and quantifies these releases. This report also disc usses the similarities and differences in the NPP compared to existing pressurized water reactor (PWR) designs as they relate to effluent releases.

This report does not include the calculation of site boundary radionuclide concentrations or doses to the public that result from the effluents. This report also does not include a discussion of the methodology used for the de termination of personnel protection design features of the NPP. The methodology to charac terize design basis events is out of scope for this technical report. This information and t he supporting calculations are addressed in the NuScale Standard Design Approval Application.

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1.3 Abbreviations Table 1-1 Abbreviations Term Definition AOO anticipated operational occurrence ANS American Nuclear Society ANSI American National Standards Institute BONAMI Bondarenko AMPX Interpolator (code) CES containment evacuation system CENTRM Continuous Energy Transport Module (code) CNV containment vessel CRUD corrosion and wear activation products CVCS chemical and volume control system DF decontamination factor EPRI Electric Power Research Institute GALE Gaseous and Liquid Effluents (NRC code implementing the methodology of NUREG-0017) gpd gallons per day gpy gallons per year GRWS gaseous radioactive waste system HCW high-conductivity waste HEPA high efficiency particulate air filter HVAC heating, ventilation and air conditioning IAEA International At omic Energy Agency LCW low-conductivity waste LRWS liquid radioactive waste system LWR light water reactor NEWT New Extended Step Characteristic -based Weighting Transport (code) NPM NuScale Power Module NPP Nuscale Power Plant NRC U.S. Nuclear Regulatory Commission OPUS ORIGEN-S Post-Processing Utility for SCALE (code) ORIGEN Oak Ridge Isotope Generation (code) ORIGEN-ARP Oak Ridge Isotope Gene ration-Automatic Rapid Processing ORIGEN-S ORIGEN-SCALE code PCA primary coolant activity PNNL Pacific Northwest Nuclear Laboratory PWR pressurized water reactor RBVS Reactor Building HVAC system RCS reactor coolant system RPV reactor pressure vessel RWB Radioactive Waste Building RXB Reactor Building SCALE Standardized Computer Analyses for Licensing Evaluation (modular code) SG steam generator TGB Turbine Generator Building TRITON Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion (code)

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2.0 Background

The NPP design houses up to six NPMs in a Reactor Building (RXB). The heating ventilation and air conditioning (HVAC) systems gather and process airborne releases in the RXB and in the Radioactive Waste Building (RWB) before being released as effluents. The processing provided by th e HVAC systems includes high-efficiency particulate air (HEPA) filters for particulates and charcoal filters for iodine removal associated with spent fuel pool releases. The RXB includes a separate, dedicated chemical and volume control system (CVCS) for each NPM for cleanup of primary coolant. There is also a common RWB located adjacent to the RXB that manages and processes radioactive waste for up to si x NPMs. The condenser air ejector systems remove gaseous releases from the main condensers and also monitor and release gaseous releases from the main condensers via the TGB to the environment.

There are important differences in the Nu Scale Power Plant US460 design that influence effluent releases. The design is an integral PWR that includes the reactor core, pressurizer, and two helical coil steam generators (SGs), which leads to the potential of direct activation of the secondary coolant bec ause of proximity of the SG to the reactor core. The primary coolant flow is solely natural circulation; a lower primary flow rate results in increased reactor coolant loop transit time and additional decay of activation products before they reach the secondary coolant. Also, each NPM consists of a reactor pressure vessel (RPV) surrounded by a high-p ressure containment vessel (CNV), which is evacuated to a low pressure under nor mal operations. There are up to six NPMs per plant located in a large, common, below gr ade reactor pool. The RXB encloses the NPMs and reactor pool. Performance of refueling operations is underwater in the refueling and spent fuel areas of the common reactor pool. During this time, the primary coolant water within the NPMs (after being cleaned up post shutdown by the CVCS) mixes with water in the reactor pool.

2.1 GALE Code Applicability

The development of an alternate methodology is necessary because NUREG-0017, the existing PWRGALE-86 code (Reference 7.2.1), is from the 1980s for evaluation of the large PWRs of that time and does not appropriately address the NPP design. The NUREG-0017 methodology used empirical data from existing large reactors and is still the current U.S. Nuclear Regulatory Commission (NRC)-endorsed effluent release code. The US460 design is smaller. A single NPM provides approximately 8 percent of the electrical output of the example PWR in NUREG-017 (Reference 7.2.1). Table 2-1 provides a comparison between the NPM and the example PWR in NUREG-0017. relies upon a different passive design based on conduction, convection, gravity, and natural circulation. contains up to six NPMs within the common reactor pool, RXB envelope, and radioactive waste management system.

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In an update to GALE in 2008, PWRGALE-08 incorporated equations and quantities from the American National Standards Institute/American Nuclear Society (ANSI/ANS) 18.1-1999 standard, Radioactive Source Term for Normal Operation of Light Water Reactors (Reference 7.2.2). The ANSI/ANS standard developed the calculation of radioactivity in the principal fluid streams of a light water reactor (LWR) based on historical data from the existing U.S. PWR fleet. Significant differences in NPP system parameters compared to a large PWR make direct scaling of most of this industry data an unsuitable extrapolation.

Another update to GALE in 2009, PWRGALE-09, incorporated a number of changes. The capacity factor was increased from 80 to 90 percent, although it was recognized in a 2012 Pacific Northwest National Laboratory (PNNL) report (Section 3.1.7, Reference 7.2.3) that this would still be too low for integral PWRs. The change in capacity factor, along with other hard-coded parameters in the GALE code, are not representative of the NPP design and cannot be changed as inputs. They c ould potentially be changed in the source code and recompiled, but recompiling would not address other applicability issues. Water activation product release rates decreased. As noted in a 2012 PNNL report (Section 3.1, Reference 7.2.3), NRC staff expressed concern that there were certain limits of applicability on the parameters built into the GALE code.

The PNNL report noted that there are five parameters that have narrow ranges of applicability to the empirical data. An attempt to adjust these parameters to better reflect the NPP design results in primary coolant concentrations outside the basis of the GALE code. These five parameter applicability ranges are also in Table 2-5 of NUREG-0017 Revision 1, along with one more parameter (steam flow) that represents the range of applicability for the secondary coolant system.

The latest update to the GALE code, GALE-PWR 3.2, was released in 2020. The GALE-PWR 3.2 Code comprehensively verified the applicability of the PWR-GALE Code from 1986, and added a graphical user interf ace. All of the inapplicability discussions above also apply to the GALE-PWR 3.2 Code as shown in Table 2-1. Table 2-1 (GALE) Applicability Range Parameter Units GALE NuScale Value Applicability Range Thermal power MWth 3000 - 3800 250 Primary coolant mass lb 500,000 - 600,000 100,000 Primary system letdown flow lb/hr 32,000 - 42,000 10,800 nominal (20,160 maximum) Shim bleed flow lb/hr 250 - 1,000 31 Letdown cation demineralizer flow lb/hr 7,500 0 Steam flow lb/hr 13,000,000 - 17,000,000 650,000

The NPP design is outside the range of these parameters, indicating that the GALE code is not appropriate for analysis of NPM coolan t activity concentrations or effluents. This report uses values from NUREG-0017, where appropriate, and explains with justification why using alternatives to values that are not appropriate yields an acceptable level of safety.

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2.2 Theory

Being unique and first-of-a-kind, NuScale does not rely on empirical effluent release data as the PWRGALE code does. The methodology separates effluents into three major phases: production (water activation, corrosion and wear activation products [CRUD], and fission products) transport (including removal mechanisms) release (liquid and airborne)

Production of radioactive isotopes (water activation, CRUD, and fission products) uses first-principles-based calculations where appropriate, combined with recent nuclear industry experience where applicable, and lessons learned where available, as appropriate in the development of source te rms (Section 3.0). This process ensures the realistic yet conservative generation of source terms for further evaluation.

As mentioned in Section 2.1, the GALE code includes some hard-coded parameters that do not reflect the NPP design, such as the capacity factor. The method utilizes a higher, more conservative, and more appropriate capacity factor of 95 percent. The hard-coded radionuclide list in GALE omits a variety of nuclides, including environmentally mobile nuclides such as I-129 and Tc-99. The method uses a more comprehensive list of isotopics that carry forward throughout the evaluation of effluents. The isotopes reported in GALE (Reference 7.2.1) and ANSI/ANS-18.1-1999 (Reference 7.2.2), as well as the isotopes listed in the Design Control Document applications for the AP-1000 (Reference 7.2.9), U.S. EPR (Reference 7.2.10), US-APWR (Reference 7.2.11), and APR1400 (Reference 7.2.4) are the basis for the list of isotopics.

Calculations of radionuclide transport throughout the plant use guidance from NUREG-0017, especially with regard to the removal mechanisms appropriate to the system process and type of hardware. Unless there is a justified change, the methodology uses the assumed process parameters found in NUREG-0017 such as ion exchanger decontamination factors (DFs) in liquid process applications, and HVAC, HEPA, and charcoal iodine filtration efficiencies for particulates and iodines in airborne process applications. Although outside the scope of this technical report, the radioactive waste systems reduce radioactive effluent releases using similar processes and methods to those currently used at large PWRs, including filtration, resin absorption, liquid dilution, decay, and controlled liquid and gaseous releases.

The last phase of effluent evaluation is the release of radioactive materials from the plant site. The conservatively developed isotope activity levels, processed and reduced in quantity as appropriate, release to the en virons as normal operations effluents. Figure 2-1 shows the general locations of efflue nt releases. Liquid effluents consolidate in the liquid radioactive waste system (LRWS) and discharge in a controlled fashion while being mixed with the utility water system as a dilution source. Airborne releases from the RXB and RWB combine to be released through one plant exhaust stack. Airborne

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releases from the TGB, constituting a small fraction of total effluents, release directly and undergo monitoring through the secondary systems.

Figure 2-1 NuScale US460 Plant Layout

2.3 Regulatory Requirements

Application of the methodology presented in this report provides a basis to ensure compliance with regulatory requirements. While site boundary concentrations and off-site dose calculations are outside the scope of th is report, the radioactive effluent results presented in the Final Safety Analysis Report demonstrate compliance with 10 CFR 20 Appendix B, as well as with 10 CFR 20.1301-20.1302, Radiation Dose Limits for Members of the Public, (Reference 7.2.15) through site-specific, off-site dose calculations. In addition, effluent calculations demonstrate compliance with 10 CFR 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion As Low As Is Reasonably Achievable, for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents (Reference 7.2.16).

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3.0 Source Term Production

Production of source terms is the initial phase in determining plant radioactive effluents. There are three categories of radioactive isotopes generated as a result of reactor operations: water activation products (in waterborne elements). CRUD (corrosion and wear activation products). fission products (isotopes created in the fuel that migrate into the primary coolant).

3.1 Water Activation Products

The NPM containment vessel (CNV) is evacuated to a very low vacuum pressure (i.e., less than 1 psia) during operation (ver y little air surrounding the reactor vessel); therefore, air activation inside the CNV is insignificant. The CNV is partially immersed in the reactor pool, and there are several neutron activation reactions that can occur with stable isotopes in the primary coolant, secondary coolant, or reactor pool. These reactions produce activation products that can be a source of radioactive effluents. The evaluation of these activation products uses a first-principle physics model as shown in Equation 3-1:

RRx G==GEquation 3-1 g =1 g x,g Ng =1 g x.g

where,

RRx = number of reactions of type "x,"

g = neutron flux in energy group "g,"

G = maximum energy group,

x,g = microscopic cross-section for reaction "x" in energy group "g,"

N = number density of target atoms, and

x.g x,g N = macroscopic cross-section for reaction x in energy group "g."

To provide some conservatism, this methodology assumes no depletion of target isotopes in the primary coolant. Benchmarks to industry data in these calculated production values shown below are only for information and comparison purposes; downstream calculations do not use them.

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3.1.1 Tritium

Tritium is usually one of the major effluent release contributors for PWRs. Production of tritium in the primary coolant by fission neutron capture results in several different reactions. Of those, activation of soluble boron (Reference 7.2.12) produces the majority. In addition to the borated primary coolant, the analysis also evaluates the secondary coolant and borated reactor pool. Tritium production reactions are in Equation 3-2 through Equation 3-9.

10 +2 4 + 5 B 10 n f2 31 H Equation 3-2

10 + 8+ 5 B 10 n f4 Be 31 H Equation 3-3

10 + 4+ 5 B 10 n th2 73 Li Equation 3-4

7 + 1++ Equation 3-5 3 Li 10 n f0 n 42 31 H

7 + 5+ Equation 3-6 3 Li 10 n f2 He 31 H

11 + 9+ Equation 3-7 5 B 10 n f4 Be 31 H

6 + 4+Equation 3-8 3 Li 10 n th2 31 H

2 + 0+Equation 3-9 1 H 10 n th0 31 H

This calculation assumes the largest boron letdown curve calculated for any planned cycle to conservatively estimate the amount of tritium generated in the core. The analysis calculates tritium production based on all of the mechanisms in Equation 3-2 through Equation 3-9, and the production over an 18-month operating cycle is in Figure 3-1. There is an investigation of the buildup of deuterium, which offers a negligible contribution to the overall tritium production or concentrations.

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Figure 3-1 Time Dependent NuScale Isotopic Tritium Production Breakdown in Primary Coolant

RCS Tritium Total Production

200

180 Soluble Boron

160

140 Soluble Lithium

120

100 Deuterium & Fuel

80

60 Total RCS H3 40 Produced

20

0 0 0.2 0.4 0.6 0.8 1 1.2 1.4 Time (years)

The calculated tritium production from solubl e species (boron, lithium, and deuterium) is 120 Ci/yr per NPM in the primary coolan t, which is higher than the Electric Power Research Institute (EPRI) example plant value of 78 Ci/yr per NPM (Reference 7.2.12). The design also includes more water in the coolant per megawatt generated than a standard PWR. Combined with its higher capacity factor, the design has a substantial neutron flux for a longer period of time, in a larger relative amount of coolant, than a typical PWR. This results in more tritium production reactions with the coolant soluble species. Figure 3-2 shows a comparison between the relative contribution of the production from soluble species and the calculated values for a NPP. The relative difference is due to starting with a higher lithium concentration than in a typical PWR, to maximize the pH for minimization of CRUD production.

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Figure 3-2 Total NuScale Isotopic Tritium Production Breakdown in Primary Coolant

Ternary fission of uranium-235 also produces tritium. Only a small fraction of the total tritium produced in the fuel diffuses through the cladding into the coolant. The EPRI tritium management model (Reference 7.2.12, Table 7-2) provides primary coolant tritium production values from fission. Reference 7.2.12 also provides tritium generation values for an example plant. Scaling the tritium production rate for the NPP power output provides an estimate of 9.1 Ci/yr per NPM coming from within the core components, such as fuel pins. Because of the low neutron flux, there is negligible direct tritium production through activation in the reactor pool and secondary coolant. Therefore, the total tritium production is 126 (117 + 9) Ci/yr per NPM compared to the EPRI example plant value of 97 (78 + 19) Ci/yr per reactor prediction.

Tritium is a mobile radionuclide because it is chemically the same as protium (hydrogen with an atomic weight of one) and bonds with water, typically as H2O. Filtering does not remove it from the water, so it has a DF of one for cleanup systems. Tritium emits a beta particle with a half-life of 12.32 years. Therefore, it decays very little before release. Once the analysis generates a tritium source term, tritium transports throughout the plant systems until being released through both liquid and gaseous pathways. The total assumed release rate of tritium is approximately equal to its production rate.

Section 2.2.17.1 of NUREG-0017 lists a total value for tritium effluent release rates of 0.4Ci/yr/MWth. For a NuScale 250 MWth reactor, that equals 100 Ci/yr per module. A comparison of the NUREG-0017, EPRI, and NuScale values is in Figure 3-3.

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Figure 3-3 Comparison of GALE, Electric Powe r Research Institute, and NuScale Yearly Tritium Production

140 130 Core 120 Components, 9 110 100 Core 90 Components, 19 80 70 60 Solubles, 117 50 GALE Total, 100 40 Solubles, 78 30 20 10 0 NUREG-017 EPRI NuScale

3.1.2 Carbon-14

The reactor coolant during power operati on produces carbon-14 (or radiocarbon). The primary coolant, secondary coolant, and reactor pool produces carbon-14, taking several possible chemical forms. The chemistry of carbon-14 is complex, and there are only two significant production reactions involving isotopes dissolved in water in LWRs (including the NPP). These two reactions are in Equation 3-10 and Equation 3-11.

17 + 4+Equation 3-10 8 O 10 n th2 146 C

14 + 1+Equation 3-11 7 N 10 n th1 p 146 C

Nitrogen is both an impurity in the fuel or other core materials and dissolved in water as a gas or as a chemical compound (e.g., ammonia or hydrazine). The calculated potential production of carbon-14 from the two reactions in all three water sources is negligible in the pool and secondary coolant system due to the small neutron fluxes.

Carbon-14 beta decays with a half-life of 5,700 years, making decay negligible. Carbon-14 is typically a component of gaseous effluents.

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Section 2.2.25 of NUREG-0017 lists values of carbon-14 effluent release rates that vary between 0.58 Ci/yr and 46 Ci/yr with an average of 7.3 Ci/yr. With the NPPs much lower power production, smaller core volume, and smaller active fuel region, it produces less carbon-14, and the NPP is well below the carbon-14 average releases for large PWRs. Based on first-principle physics, the calculated carbon-14 production in the primary coolant is 1.2 Ci/yr per module, a fraction of the total yearly average effluent release of radionuclides.

3.1.3 Nitrogen-16

Oxygen-16 (99.76 percent of naturally occurring oxygen) in water can activate to form radioactive nitrogen (nitrogen-16). Nitrogen-16 is produced by neutron activation of oxygen via Equation 3-12.

16 + 1+Equation 3-12 8 O 10 n f1 p 167 N

The nitrogen-16 atoms combine with oxygen an d hydrogen in the coolant to form ions or compounds such as NO, NO2, NO3, N2, and NH4. Nitrogen-16 has a high formation rate and a short half-life of 7.13 seconds. Nitrogen-16 emits high-energy gamma rays (6.13MeV and 7.12 MeV).

Nitrogen-16 activity is high in the primary coolant in and near the active core; however, due to its short half-life, longer transit times through various plant systems, and off-site receptors, nitrogen-16 is not a significant contributor to radiation exposure beyond the primary coolant system and is, therefore, not a significant contributor to effluents. That is why NUREG-0017 Revision 1, Section 1.5.2.12.2 states that the GALE code does not consider nitrogen-16 as an effluent. Transit times are longer in the NPM than traditional large PWRs because of the slower natural circulation primary flow. The total reactor coolant system (RCS) loop transit time is approximately 46 seconds, which is more than six half-lives of nitrogen-16, which prevents buildup in the core. Section 4.1.1 calculates and discusses the nitrogen-16 concentration at various locations (e.g., at the bottom of the helical coil SG) within the RCS loop.

3.1.4 Argon-41

Neutron activation of argon-40 produces argon-41. Argon-40 is naturally found in air. The amount of argon in air is 0.934 percent (Reference 7.2.13 Table E-1), and Equation 3-13 shows the production of argon-41.

40 + 0+Equation 3-13 18 Ar 10 n th0 4118 Ar

Radioactive argon-41 is an inert gas that transforms into a stable isotope of potassium (potassium-41) through a relatively complex set of decay emissions. Argon-41 decay primarily produces both a 1.2 MeV beta particle and a 1.3 MeV

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gamma ray with a half-life of approximately 110 minutes (Reference 7.2.13 Table 5-1, B-1, and E-2).

In existing large PWRs, the activation of natural argon-40 in the air within the containment that surrounds the reactor vessel dominates production of argon-41. Primary and secondary coolant streams purge air before operating the plant, making production of argon in the coolant streams negligible in large PWRs. In the NPP design, there is little air surrounding the reactor vessel because the steel CNV, maintained at a low pressure (less than 1 psia) during power operation, surrounds the reactor vessel. Argon-40 is negligible inside containment. As a result, the main contributor of argon-41 effluent release from production is activation of argon-40 contained in air that dissolves in the water of the reactor pool surrounding the NPMs.

Section 2.2.26 of NUREG-0017 lists values of argon-41 effluent release rates that vary between 0.02 Ci/yr and 208 Ci/yr with an average of 34 Ci/yr. With the NPPs lower power production, smaller fluxes, and the NPM submerged in water instead of air, there is substantially less argon-41 produced outside of the NPM, and the NPP is well below the argon-41 average releases for large PWRs.

The primary coolant can contain argon-40 as a tracer for leaks through the helical coil SG into the secondary side. If the tracing is performed, it achieves a desired argon-41 activity concentration in the primary coolant of 0.1 µCi/ml (Reference 7.2.23). This analysis assumes argon-40 addition.

3.2 Corrosion and Wear Activation Products

3.2.1 Mechanism Overview

The activated corrosion and wear products fo rm as a result of oxidation and wear of the materials of construction in the primary reactor coolant circuit that come in contact with the reactor coolant and activate by neutron interactions. When exposure of these alloys to the primary reactor coolant occurs at high temperature, oxygen diffuses into the base metal at the wetted surface and converts the elements in the alloy from the metallic state to an oxide state. In the proc ess, divalent metal ions release into water as soluble metal ions (Reference 7.2.21). Thus, a protective layer of corrosion products forms on the surface of an alloy, which separates it from the coolant. The ion conductivity of this layer is low; however, mass transfer still exists between the metal alloy and the primary coolant (Reference 7.2.22).

The activated corrosion and wear products can manifest itself in a solid phase, either as metal oxide films or as micrometer-sized particles of metal oxide (Reference 7.2.21). It can also exist as hy drolyzed species of metal oxides in the aqueous phase. Introduced species resulting from metallic corrosion are in the coolant, where they transport through convection onto other surfaces (Reference 7.2.22), including the surface of the fuel and the surface of in-core structure materials. Thus, they transform in to radioactive nuclides in the neutron flux, meaning that they activate. Neutron activation is possible when metal oxide species travel in the reactor core region or when they deposit on in-core surfaces.

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The activated corrosion products release fr om fuel surface deposits by erosion and spalling caused by hydraulic shear forces or dissolution. Some activated products release from in-core materials by dissolution and wear. They then transport by water to all parts of the primary system, where t hey can deposit on surfaces by the following mechanisms: turbulent diffusion, Brownian diffusion, inertial impaction, sedimentation, and thermophoresis. The pr oduction, transportation, solubility, and deposition have many complicated mechanisms. These include pH, temperature, materials of construction, flow rates and regimes, surface conditions, and chemistry. This complexity prohibits first-principle physics models of CRUD.

3.2.2 Modeling Corrosion and Wear Activation Products

There are developed models for the estima tion of radioactivity buildup and corrosion product transport in LWRs. These include empirical and semi-empirical models containing coefficients that derive from experimental data or plant design data. Some examples include: Japanese ACE, Korean CRUDTRAN, Czech DISER, Bulgarian MIGA, and French PACTOLE (Reference 7.2.22). These models use empirical data from the specific reactors whose behavior they model. For this reason, they are not applicable for reactors with different designs and geometries. In particular, the NPM has some characteristics that make it fundamentally different from other PWRs; therefore, none of the available models accu rately describes the NPM behavior with regard to the activated corrosion products transport and deposition. Differences in the NPM design and the existing fleet prec lude the use of these reactor-specific models.

Because there are no models available for the generation and transportation of corrosion and wear activation products, the model uses conservative empirical data. The ANSI/ANS-18.1-1999 standard provides a basis for determining the concentrations of radionuclides in the primary and secondary coolant of a nuclear power plant. Therefore, the standard calculates those values directly rather than calculating a production rate. This standard is for the purposes of calculating, through adjustment factors, radionuclide concentrati ons in support of the design and licensing process. The data contained in ANS18.1 is based on actual historical large PWR plant measurements, from a time when CRUD production was much higher in the industry. As such, it is a suitable and cons ervative standard for calculating anticipated corrosion and wear activation products in the primary coolant for the design. The calculated CRUD source term numbers in the primary are in Table 3-1. Table 3-1 CRUD Isotopic Primary Concentrations Isotope Primary Coolant Concentration (µCi/g) Na24 1.4E-02 Cr51 7.7E-04 Mn54 4.0E-04 Fe55 3.0E-04 Fe59 7.5E-05 Co58 1.1E-03 Co60 1.3E-04

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Table 3-1 CRUD Isotopic Primary Concentrations (Continued) Isotope Primary Coolant Concentration (µCi/g) Ni-63 6.6E-05 Zn65 1.3E-04 Zr-95 9.7E-05 Ag-110m 3.2E-04 W187 7.0E-04

Incorporating lessons learned from the industry decreased CRUD production over time. The design follows modern guidelines for the reduction of CRUD and employs design features that minimize CRUD production. The reactor uses the lowest possible cobalt and nickel materials appropriate for design conditions, along with lessons learned about RCS chemistry control (e.g., highest pH). As a result, values derived from the ANS standard for the NPP are conservative.

Additionally, the RPV and the CNV are stai nless steel, which is designed to survive the life of the plant in the borated water chemistry. As a result, the vessels should have minimal corrosion activation products.

3.3 Fission Products

The industry standard, Standardized Comp uter Analyses for Licensing Evaluation (SCALE), computer code develops spent fuel isotopic distribution and magnitude. To ensure conservative results, the methodology assumes a maximum peak burnup of 62 GWd/MtU for all fuel rods in the core. The fuel isotopic inventory per assembly at that burnup is in Table A-2 of Appendix A.

3.3.1 Software Use and Qualification

To further support the use of a first principles approach in the methodology, the SCALE 6.1 modular code package, developed by Oak Ridge National Laboratory, develops reactor core and primary coolant fi ssion product source terms. Specifically, the Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion (TRITON) and Oak Ridge Isotope Generation - Automatic Rapid Processing (ORIGEN-ARP) analysis sequences of the SCALE 6.1 modular code package, and ORIGEN-SCALE code (ORIGEN-S), run as a standalone module, generate radiation source terms for the fuel assemblies and various waste streams (Reference 7.2.14).

This industry standard commercial off-the-shelf software is used without modification by the methodology. This software has been extensively used in the evaluation of operating large LWRs. NuScales Softwar e Configuration Management Plan directs the use of the SCALE code package. The SCALE code is in compliance with ASME NQA-1 2008/2009A through the NuScale commercial grade dedication process.

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3.3.2 TRITON Code Sequence

The TRITON sequence of the SCALE code package is a multipurpose control module for nuclide transport and depletion, including sensitivity and uncertainty analysis. TRITON generates problem-dependent and burnup-dependent cross-sections. It performs multi-group transport calculations in one-dimensional, two-dimensional, or three-dimensional geometries. The ability of TRITON to model complex fuel assembly designs improves transport modeling accuracy in problems that have a spatial dependence on the neutron flux. In this case, TRITON generates burnup-dependent cross-sections for fuel assemblies for subsequent use in the ORIGEN-ARP depletion module.

The T-DEPL (time-depletion) sequence of the TRITON control module generates problem-dependent (i.e., assembly-specific) and burnup-dependent cross-sections. The model uses the Continuous Energy Transport Module (CENTRM)-based option of the T-DEPL sequence in which the Bo ndarenko AMPX Interpolator (BONAMI) processes microscopic cross-sections for the unresolved resonance energy range. The CENTRM code processes cross-sections from the continuous-energy library for the resolved resonance energy range. The CENTRM code uses a one-dimensional discrete ordinates calculation to generate point-wise fluxes, properly taking into account overlapping resonances from different isotopes. The multi-group cross-section module creates a problem-dependent multi-group library for the resolved resonance energy range using the weighting spectrum from CENTRM and combines it with the multi-group library processed by BONAMI. The Code to Read and Write Data for Discretized (CRAWDAD) solution and WORKER modules properly format the cross-section libraries at different stages of the processing.

The New Extended Step Characteristic-based Weighting Transport (NEWT) code module performs a two-dimensional, discrete ordinates transport calculation. The NEWT code post-processes the results of the transport calculation to generate region-averaged multi-group cross-sections and fluxes for each depletion material. The COUPLE module essentially couples NEWT and ORIGEN-S by collapsing the multi-group cross-sections into a one-group cross-section library for each depletion material using the fluxes from NEWT. The COUPLE module then combines the one-group cross-section library with decay data and energy-dependent fission product yields to produce a binary-formatte d ORIGEN-S nuclear data library. Finally, ORIGEN-S depletes each material using the normalized material power and the problem-and burnup-dependent nuclear data library. Decay intervals between depletion steps are also modeled by ORIGEN-S. The TRITON code models the complete depletion sequence by repeatin g the cross-section processing, transport calculations, depletion, and decay calculations for a user-specified series of depletion and decay intervals, using a predictor-corrector algorithm. The TRITON code saves each problem-dependent and burnup-dependent nuclear data library for future use with ORIGEN-ARP. After the final depletion step, TRITON can call the ORIGEN-S post-processing utility for SCALE (OPUS) module to post-process the ORIGEN-S time-dependent isotopic concentrations, produc ing an ASCII-formatted file of isotopic concentrations or source spectra for further analysis or plotting.

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3.3.3 ORIGEN (ORIGEN-ARP and ORIGEN-S) Code Sequences

The ORIGEN-ARP code is a SCALE depletion analysis sequence used to perform point-depletion and decay calculations with the ORIGEN-S module using problem-dependent and burnup-dependent cross-sections. The ORIGEN-ARP module prepares ORIGEN-S nuclear data libraries containing these cross-sections using interpolation in enrichment and burnup between pre-generated nuclear data libraries containing cross-section data that span the desired range of fuel properties and operating conditions. The ORIGEN-ARP sequence produces calculations with accuracy comparable to that of the TRITON sequence with a great savings in problem setup and computational time compared to repeated use of TRITON. There are many possible modeling variations in fuel asse mbly irradiation history. For depletion calculations involving fuel assemblies, the TRITON sequence generates ORIGEN-S nuclear data libraries, as described.

The ORIGEN-S module of SCALE 6.1 calculates the time-dependent isotopic concentrations of materials in a fuel assembly by modeling the fission, transmutation, and radioactive decay of fuel isotopes, fission products, and activation products in the assembly. The ORIGEN-ARP module sets up the input data for ORIGEN-S, ensuring use of the proper nuclear data library for each depletion or decay interval of the fuel assembly irradiation history.

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4.0 Radionuclide Transport, Removal Mechanisms, and Release

Transportation of radionuclides within the plant throughout the various systems, and selective removal of isotopes based on processing capabilities, is the second phase in determining plant radioactive effluents. Release of processed radionuclides into the environs through either liquid or gaseous effluent pathways, is the third phase (Section 4.8).

4.1 Primary Coolant Water System

Section 3.0 discusses the source term inputs to the primary coolant. The three inputs to the primary coolant are direct neutron activation in the water, CRUD, and fission products that leak and diffuse from failed or damaged fuel.

4.1.1 Water Activation Products

Because tritium cannot be removed from the primary coolant water, it does not reach an equilibrium value over a cycle during operation. Because the design facilitates recycling of primary water, three recycling modes calculate the tritium concentration in process streams: 1) no recycling of the primary coolant; 2) recycling of the primary coolant to the reactor pool; and 3) recycling of the primary coolant back to the CVCS as makeup. The first mode (no recycling) ma ximizes the tritium concentration in the liquid discharge effluent stream. Therefore, the liquid effluent calculation uses the letdown tritium concentration from no recycling.

The production rate (Figure 3-1), along with the cumulative water injection and bleed out of the primary coolant (Figure 4-1), develops a time-dependent balance of how much tritium is in the coolant versus how much has bled out of the coolant (Figure 4-2). The removal of primary coolant to control boron levels in the reactor and, subsequently, reactivity control in the core, forms the basis for letdown removal. Primary coolant is let down from the reactor to the LRWS via the CVCS.

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Figure 4-1 Water Injection and Bleed in the Primary Coolant

Total RCS Coolant Discharge

1.80E+05 1.60E+05 1.40E+05 1.20E+05 1.00E+05 8.00E+04 6.00E+04 4.00E+04 2.00E+04 0.00E+00 0.00 100.00 200.00 300.00 400.00 500.00 600.00 Effective Full Power Days

Figure 4-2 Tritium Reactor Coolant System Balance

Mode 1 RCS H3

2.50E+02

2.00E+02

1.50E+02 Mode 1 Cumulative Discharge Ci 1.00E+02 Mode 1 RCS Tritum Activity(Ci) 5.00E+01 Total H3 Produced (Ci)

0.00E+00 0 100 200 300 400 500 600 Effective Full Power Days

The tritium inventory curve transforms into a concentration and the time weighted average to determine the average tritium concentration in the primary coolant. For

© Copyright 2023 by NuScale Power, LLC 21 Effluent Release (GALE Replacement) Methodology and Results

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comparison, Section 2.2.17.1 of NUREG-0017 lists an average tritium primary coolant concentration in PWRs of 1.0 µCi/ml. For normal operations with primary letdown, NuScale calculates an average concentrati on of 1.3 µCi/ml. In addition, the average concentration of primary coolant being let down from the RCS is 1.0 µCi/ml. Mode 1 (no recycling of primary coolant) forms the basis of these tritium concentrations.

NUREG-0017 does not list carbon-14 in the primary or secondary coolant, although it is included as a small contributor to gaseous effluent.

Table 2-2 of NUREG-0017 states that there is a nitrogen-16 primary concentration of 40 µCi/ml at the SG on the primary loop, where N-16 could leak into the secondary coolant.

With natural circulation in the NPM core, the coolant flow rate is slow enough that nitrogen-16 has a substantial amount of decay during its transit time through the primary system. Using the maximum full power coolant flow rate of 821 kg/s, the minimum, full power, total RCS transit time is approximately 46 seconds, more than six half-lives of nitrogen-16. Therefore, by the time the nitrogen-16 transits to the integral helical coil SG in 11 seconds, its concentration is 41 µCi/g, which is close to the NUREG-0017 value. Further, the nitrogen-16 concentration at the CVCS inlet with a transit time of 26 seconds is much smaller than the helical coil SG at 10 µCi/g.

The primary coolant can contain argon-40 for use as a tracer for SG leaks. This analysis assumes argon-40 addition to reach target argon-41 levels in the primary coolant of 0.1 µCi/ml (Reference 7.2.23).

4.1.2 CRUD

As discussed in Section 3.2, CRUD is evaluated in terms of primary coolant concentrations.

4.1.3 Fission Products

Fission product leakage into the primary coolant from the previously calculated fuel inventory uses a realistic yet conservative fuel failure fraction of 66 rods per million (discussed in Section 5.0) along with typical industry fission product isotopic escape coefficients (Reference 7.2.9, Reference 7.2.10, Reference 7.2.11), as shown in Table 4-1. These values are also conserva tive for the NPP design because escape rate coefficients are a function of linear heat generation rate and the NPM has a lower linear heat generation rate than larger PWRs. Table 4-1 Fuel Isotopic Escape Coefficients Isotope Value (s-1) Kr 6.5E-8 Xe 6.5E-8 Br 1.3E-8 Rb 1.3E-8 I1.3E-8

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Table 4-1 Fuel Isotopic Escape Coefficients (Continued) Isotope Value (s-1) Cs 1.3E-8 Mo 2.0E-9 Tc 2.0E-9 Ag 2.0E-9 Te 1.0E-9 Sr 1.0E-11 Ba 1.0E-11 Y1.6E-12 Zr 1.6E-12 Nb 1.6E-12 Ru 1.6E-12 Rh 1.6E-12 La 1.6E-12 Ce 1.6E-12 Pr 1.6E-12 Np 1.6E-12 Sb 1.6E-12 P1.6E-12

Table A-1 lists the maximum fuel inventory per assembly.

4.1.4 Primary Coolant Activity Concentrations

The primary coolant activity also includes the build-in of radioactive daughter products from the decay process. The equilibrium concentration of radionuclides in the primary coolant assumes a homogenized mixture of radionuclides throughout the entire water volume with the exception of nitrogen-16, as previously described.

The NPPs primary water volume-to-fuel ratio is much higher than a typical large PWR. Even assuming a proportional source term, this results in a lower concentration in the primary water due to greater dilution in the larger RCS volume.

The removal mechanisms of most of the radionuclides from the primary system are radioactive decay, purification (CVCS demineralizers) and letdown to the LRWS. The DFs for the mixed-bed demineralizers are 100 for halogens, 2 for cesium and rubidium, and 50 for other isotopes, per Section 2.2.18.1 of NUREG-0017. There is no specific degasification of the primary coolant, thus neglecting noble gas removal through the pressurizer is appropriate.

Although the concentration of individual is otopes in the primary coolant varies considerably over the operating cycle, t he conservative assumption is that maximum calculated equilibrium activity is present for the entire operating cycle, with the

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exception of some of the water activation products, which are treated separately. Equation 4-1 calculates the activity of the isotopes:

A cp = ------------------------------------A sp Equation 4-1 () p L U++

where,

A cp = activity of parent isotopes in the primary coolant,

A sp = activity generation rate of the source term parent isotopes,

p = decay constant of the parent nuclide,

L = letdown removal coefficient through LRWS degasifiers, and

U = removal coefficient for purification.

Equation 4-2 calculates the activity of the ingrowth of daughter product isotopes:

A cd = ------------------------------------A cp d f p Equation 4-2 () d L U++

where,

d = decay constant of the daughter nuclide, and

f p = branching fraction for the parent nuclide(s) that decay to the daughter isotope.

The list of radionuclide activity concentratio ns in the primary coolant is in Table A-2 in Appendix A.

4.2 Secondary Coolant Water System

Primary-to-secondary leakage determines the concentration of radionuclides in the secondary system. The Electric Power Research Institute (Reference 7.2.13) evaluated primary-to-secondary leakage in the industry and developed SG management guidelines, which NuScale follows. As operational experience with the NuScale helical coil SGs accumulates, modifications to EPRI guideli nes may occur to optimize the mitigation of potential leakage. The direct activation of the secondary water impurities is negligible due to the small flux at the bottom of the helical co il SGs, which is closest to the active core. The flux at the bottom of the helical coil SGs is several orders of magnitude less than the average active core flux.

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The methodology underestimates the total secondary coolant mass as 5.0E4 lbm for the effluent models. The NPP value for secondary coolant mass equates to the sum of water mass estimates for the various main component s of the secondary system (including both helical coil SGs and other components), and neglec ts the contribution of the fluids in the turbine and condenser as well as condensate polishing systems, structures, and components. This smaller mass is conservative because it overestimates the radionuclide concentrations.

One secondary side removal mechanism is cleanup through the demineralizers that have DFs of 100 for halogens, 10 for cesium and rubidium, and 100 for other isotopes, per Section 2.2.18.1 of NUREG-0017. Other secondary side removal mechanisms are liquid and gaseous leakage to the TGB (assumed to be upstream of the condensate polishers for conservatism), condensate air removal, and the turbine gland seal steam. The scaled leakage terms from the secondary system are from values provided in NUREG-0017 based on the low power level of each NPM (250 MW th) compared to a traditional large PWR with an assumed power level of 3400 MW th. Power scaling is appropriate because system capabilities scale to the size of the reactor. Main steam production is approximately proportional to core thermal power. Also, component sizing (e.g., pipe diameter) relates to core thermal power. This approach results in larger, more conservative values for the secondar y coolant radionuclide concentrations.

The secondary coolant sampling system drain rate, TGB floor drain rate, and steam leakage rate to the TGB are NUREG-0017 values linearly scaled to the power output of a NuScale core (250 MWth) from the nominal power output of a standard PWR (3400 MWth), as shown in Table 4-2.

Table 4-2 NUREG-0017 and Corresponding NuScale Parameters Parameter NUREG-0017 NuScale Module Primary-to-secondary leak rate (lb/day/NPM) 75 5.5 CVCS to RXB leak rate (lb/day/NPM) 160 11.8 TGB floor drains (gal/day/NPM) 7200 9.9 Secondary coolant sampling system drains (gal/day/NPM) 1400 1.9 Steam leak rate in TGB (lb/hr/NPM) 1700 125

The means to determine concentration of most of the radionuclides in the secondary coolant is similar to that of the primary coolant, because it shares the same basic governing equation. The main difference is t hat the production term for the secondary coolant is just the leakage of radionuclides fr om the primary into the secondary, given by Equation 4-3:

P s A p L PS=xEquation 4-3

where,

P s = production rate in the secondary coolant,

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A p = equilibrium activity of a radionuclide in the primary coolant, and

L PS = leak rate of coolant from the primary to the secondary.

This calculation leads to an equilibrium activity in the secondary coolant that is similar to Equation 4-1. The equation that models the secondary activity is in Equation 4-4:

C p L PSx A s = --------------------- d U+-Equation 4-4

where,

A s = equilibrium activity in the secondary coolant,

C p = equilibrium concentration in the primary coolant,

L PS = leak rate from the primary to the secondary,

d = decay constant for the radionuclide, and

U = cleanup constant for the radionuclide.

The concentration of radionuclides in the se condary coolant is the calculated secondary activity divided by the total mass of secondary coolant.

Because noble gases are not chemically reactive, cleanup systems do not generally remove noble gases from the coolant. Noble gases leave the secondary coolant quickly through gaseous removal mechanisms (primarily the condenser air removal system). Multiplying the concentration of the noble gas in the primary coolant by the primary-to-secondary leak rate and then dividing by the sum of the secondary flow rate and primary-to-secondary leak rate calculates the concentration of noble gases in the secondary coolant using Equation 4-5.

C Secondary C Primary=x-----------------------------------------L PS-Equation 4-5 L PS m+* ondary sec

Tritium, as an isotope of hydrogen, is chemically identical to hydrogen, preventing typical methods of cleanup from working on tritium, resulting in two important consequences. The first is that without cleanup or any other removal mechanism, the secondary coolant concentration of tritium reaches the same value as the primary coolant concentration. This modeling approach is not a reasonable approximation due to removal of tritium through leakage and decay. The second consequence is that tritium does not buildup in

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the cleanup systems. Therefore, tritium does not impact any shielding calculations for these systems because it is a weak beta emitter. The calculations in this document account for the eventual effluent release of tritium by considering the leakage rate of coolant out of the secondary system. The secondary coolant concentration is in Equation 4-6:

A Secondary C Primary=x-----------------L PS-Equation 4-6 d L+

where,

= activity of tritium in the secondary,A Secondary

= concentration in the primary,C Primary

= leak rate from primary to secondary,L PS

d = decay constant for tritium, and

L = leakage removal constant.

The total tritium concentration is the total tritium activity divided by the total mass of secondary coolant. For comparison, Table 2-3 of NUREG-0017 lists a tritium secondary coolant concentration of 1.0E-03 µCi/ml. The calculation determined a tritium activity concentration in the secondary coolant of 2.5E-03 µCi/ml. This tritium concentration relates to Mode 2, recycle of RCS to the pool.

A comprehensive list of radionuclide activity concentrations in the secondary coolant is in Table A-2 in Appendix A.

4.3 Chemical and Volume Control System

The radionuclide concentrations at the inlet to the CVCS are from the primary coolant system letdown at primary coolant concentrations. Demineralizers remove radionuclides in the coolant by an ion-exchange mechanism. Parameters that impact the removal of activity include the concentration of the isotope entering the demineralizer and the removal efficiency for each isotope, which is consistent with current designs of large PWRs.

Leakage from the CVCS that goes to drain collections assumes leakage before the demineralizers. The activity of the exiting water through letdown follows the guidance and DF values found in NUREG-0017 for process components such as isotope-specific DFs for demineralizers. The DFs for the CVCS mixed-bed demineralizers are 100 for

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halogens, 2 for cesium and rubidium, and 50 for all others. The activity of the coolant after passing through the demineralizers is in Equation 4-7:

C in C out = -------D f-Equation 4-7

where,

C out = Concentration levels on the outlet (µCi/g),

C in = Concentration levels on the inlet (µCi/g), and

D f = Decontamination factor for an isotope i in particulate filter or demineralizer.

Consistent with NUREG-0017, no credit is taken for CVCS filters.

4.4 Reactor Pool and Spent Fuel Pool

The activity of the reactor pool (including the refueling area of the common reactor pool) and the connected spent fuel pool in the NPP is dependent on the primary coolant activity within an NPM at the time of module disassembly for refueling. When an NPM is shut down after an operating cycle, the CVCS cleans up the primary coolant. The cleanup time period assumes sufficient cleaning of the primary coolant after a chemically-induced CRUD burst and an iodine spike to meet two dose rate targets. The first target maintains the accessible areas above and around the pool under a dose rate of 2.5 mRem/hour. The second target maintains the doses one meter above the pool below 5 mRem/hour per EPRI guidelines (Reference 7.2.19). When NPM disassembly occurs for refueling, the cleaned primary coolant releases into the refueling area of the pool.

Direct neutron activation of surrounding reac tor pool water products from operating NPMs is negligible compared to the contribution fr om the primary coolant during refueling, due to the small flux in the pool. At the outside of the CNV, the largest neutron flux is at the core centerline and is several orders of magnitude lower than that of the active core. Additionally, it drops off very quickly because the pool is borated to 2000 ppm.

An evaluation for potential activation consid ers inadvertent impurities introduced into the pool. Resin backwash and breakthrough, lubric ating oils, and hydraulic fluids have the potential for introduction into the pool in small quantities. They are hydrocarbon chemicals that do not introduce any new radioisotopes into the effluent stream. The postulated impurities either float on the top of the pool or sink to the bottom. In either case, they would not be close to the active co re except for a very brief transit period while sinking. Therefore, there is negligible neutr on flux available for activation. These small quantities dilute throughout a very large pool water mass, making their concentrations negligible to radioisotope production.

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The activity released from a disassembled NPM in the refueling area of the pool is assumed to instantly mix homogenously throughout the entire pool volume (reactor pool and spent fuel pool). This modeling approach is conservative for effluent release because it does not take into account pool water cleanup during the time it takes the released activity to mix throughout the pool. During an event, the activity releases near the bottom of the refueling area of the pool and mixes both vertically and horizontally. By the time the released activity diffuses to the top of the pool, where it can become airborne (becoming an effluent source), there is some pool cleanup system removal as well as some decay. The concentration of the pool reaches a peak concentration for a short period before removal by radioactive decay, pool cleanup, and evaporation reduces the pool activity.

The pool purification system reduces the activity of the pool water to pre-refueling conditions, so that subsequent reloads do not result in a continuous buildup of radionuclides in the pool over time. The concentration is governed by Equation 4-8:

Nt() N o - FR =xexp t+---------------xx-Equation 4-8 M

where,

N = concentration of the given radionuclide,

= decay constant for the given radionuclide,

FR = flow rate of the water through the cleanup system,

= efficiency of the cleanup system, between zero (no effect) and one (perfectly efficient),

M = mass of water, and

t = time.

The exception to this treatment of radionuclides is tritium, which is difficult to remove from the water through cleanup. Tritium continues to build up to an equilibrium concentration in the pool due to losses from evaporation and decay, and is governed by Equation 4-9.

Production rate Ci---------- Tritium Pool Inventory ()= ------------------------------------------------------------------------------------------------------------------yearEquation 4 evaporation rate g 365.25 days--------x-----------

+ ----------------------------------------------------------------------------------------------dayyear pool mass g()

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The second mode of recycling primary water directly to the pool maximizes the tritium concentration in the pool, which also maximizes the tritium in the gaseous effluent stream, due to pool evaporation. Therefore, the gaseous effluent calculation uses the tritium concentration in the pool from recycling primary water to the pool.

4.5 Airborne Activity

Evaporation from the RXB reactor pool and spent fuel pool is the main source of airborne activity in the NPP. NUREG-0017 identifies numerous locations and sources of airborne radioactive material in a PWR as the main contributors of the gaseous effluent releases from normal operation and AOOs. The methodol ogy evaluates a design-specific AOO of an inadvertent emergency core cooling system actuation that results in pressurizing the CNV. Primary coolant leaks, pool evaporation, and secondary coolant leaks use partition factors of 1 for gases and tritium, 0.01 for halogens, and 0.005 for other nuclides taken from NUREG-0017, page 2-10, Table 2-6. The pool evaporation partition coefficient for iodine is 2000 based on a pool temperature of 120 degrees F and a pH of 5 (Reference 7.2.5). For conservatism, these values are steam/water partition factors designated for U-tube SGs and used for pool evaporation. These values are conservative because more radionuclides become airborne from pressurized steam than from pool evaporation due to the excess energy acting as a driving force of both the pressure and the energy from the higher temperatures. Prim ary coolant leaks into the RXB contribute to airborne activity using a 40 percent flash fr action, where 60 percent of the leak remains in liquid form and 40 percent leaves as steam per Table 2-26 of NUREG-0017.

4.5.1 Waste Gas Processing System

Section 4.6 discusses the gaseous radioactive waste system (GRWS), which includes the waste gas processing system. Potential leakage from this system may result in airborne contamination. The system evaluation occurs at locations where the potential for airborne radioactivity exists.

4.5.2 Steam Generator Blowdown System

The NPM helical coil SG is an integral, once-through, helical coil design. Because the secondary coolant circulates on the inside of the SG tubes, the NPM helical coil SG does not have the capability to blowdown, and therefore does not have a blowdown system.

4.5.3 Condenser Air Ejector Exhaust

Each NPM has a dedicated secondary system with independent condenser air ejector systems. The condenser air ejector systems exhaust is a source of noble gases as well as halogens at an average release rate of 125 Ci/yr/NPM per µCi/g of secondary coolant. This value is linearly scaled by reactor thermal power from 1700 (Ci/y releases per µCi/g of primary coolant) from Table 2-22 of NUREG-0017 (Reference 7.2.1). The condenser air removal system maintains a vacuum on the condenser to remove gases. Removed gases pump through water separator tanks and vent to the atmosphere. This report determines the annual release rate for

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halogens and noble gases based on primary-to-secondary coolant system leak rates as well as leak rates out of the condenser air removal system. The condenser air removal system and gland seal steam sy stem exhausts have direct, unfiltered pathways out of the TGB to the atmosphere.

4.5.4 Containment Purge Exhaust

The NPP design uses a steel CNV surrounding the RPV. Section 2.2.6 of NUREG-0017 attributes three percent of the primary coolant inventory of noble gases as leakage to containment every day. For the NPP, the containment evacuation system (CES) manages the CNV air and maintains the CNV under evacuated conditions. The CES normally vents to the Re actor Building HVAC system (RBVS). If the CES radiation monitors detect high radiation, the exhaust flow redirects to the GRWS for processing. The CES removes RPV leakage (0.47lbm/hr/NPM) into the CNV and routes RPV leakage via the GRWS decay beds for normal effluent. This method uses the low volumetric flow rate of gases leaving the CNV via the CES vacuum pump. Therefore, there is a need for sensitive radiation monitoring detection. Additionally, due to the benefit of the integral and natural circulation features of the NPM, there is less opportunity for gas leaks from the RCS.

4.5.5 Ventilation Exhaust Air from the Radioactive Waste Building and the Reactor Building

Sources of airborne radionuclides include primary leakage from the CVCS. Section 2.2.6 of NUREG-0017 attributes 160 lbm/day/NPM leak rate of primary coolant into the Auxiliary Building. Assuming a NPP has six times the primary leak rates of a larger PWR is overly conservative and unrealistic. The NPMs are much smaller and have less inventory. The methodology linearly scales the 160 lbm/day/reactor leak rate value by thermal power to 11.8 lbm/day per NPM, for a total plant leakage of 70.6 lbm/day. The total plant leakage of 70.6 lbm/day forms the basis for the effluent airborne inventory in the RXB from primary leaks from the CVCS. The NPM RXB functions similar to the Auxiliary Building of a large PWR, in terms of release pathways from the CVCS. Upon a high radiatio n signal in the RXB, the ventilation flow routes through HEPA and charcoal filters before release. The normal operation effluent calculations do not credit both HEPA and charcoal filtration.

4.5.6 Steam Leakage from Secondary System

Assumed steam leakage from the secondary system occurs in the TGB at the rate of 125 lbm/hour/NPM, for a total plant leak rate of 750 lbm/hour. This leak rate is linearly scaled by reactor thermal power to the 1700 lb/hr/reactor leak rate per RXM from NUREG-0017.

4.5.7 Reactor Pool Evaporation

In the RXB, evaporation from the reactor pool has the capability to release radioactive contaminants into the RXB airspace, which are then available for release to the environment. The pool source term rises during refueling events because the cleaned

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post-CRUD-burst primary coolant mixes with the pool water, as described in Section 4.4. The time-weighted average assumed pool source term over a year evaporates into the RXB airspace, which then goes through the RBVS and out the plant exhaust stack. The calculated total reactor pool evaporation rate is 1300 lbm/hour, with the drydock included. The total evaporation rate includes evaporation from the drydock for conservativism in gaseous effluent releases. The total pool water volume calculation does not include the drydock water volume for conservativism in the pool radionuclide concentrations.

4.5.8 Inadvertent Emergency Core Cooling System Actuation Anticipated Operational Occurrence

An AOO that is NuScale-specific is a single inadvertent emergency core cooling system actuation that floods the CNV with pr imary water, resulting in pressurization of the CNV. The CNV leaks an assumed 0.2 weight percent per day into the pool or the airspace under the bioshield. For the pur pose of evaluating the effluent consequence of these AOOs, the CNV leakage is an assumed steady state gas leak into the region below the bioshield for 30 hours. Thirty hours is the period of time it takes the NPM to depressurize following an accident, based on containment transient thermal-hydraulic calculations. This leakage quantification uses the same method as the primary coolant leaks. This release is calculated to be 140 mCi into the RXB airspace.

4.6 Gaseous Radioactive Waste System

Up to six NPMs in a single NPP share the GRWS. The GRWS processes gaseous waste from degasification of the primary system letdown and the CES upon actuation of a high radiation signal through decay beds before di scharge through the filtered plant exhaust stack.

4.6.1 Activity Input to the Guard Bed

The guard bed is the first charcoal bed to receive gaseous input from the LRWS degasifiers and the CES after the gas passes through a gas cooler and a moisture separator. For effluent release, the guard bed does not collect or delay any radionuclides, so the assumed input goes directly into the decay beds.

4.6.2 Activity Input to the Decay Beds

The charcoal decay beds delay noble gases from being released long enough to decay, thus reducing the amount released as gaseous effluent from the plant. There are two trains of decay beds in the GRWS, an A and B train. Each decay bed train has a charcoal mass of 4600 pounds. The adsorption coefficients and delay times for each bed are in Table 4-3.

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Table 4-3 Charcoal Decay Bed Information Element Adsorption Coefficient Holdup Time (cm3/g) (days/train) Argon 8.9 0.44 Krypton 60 2.9 Xenon 1400 69

Radionuclides present in the gaseous stream that collect in the beds decay over time. In some cases, these radionuclides decay to daughter products that are also radioactive. The calculation of daughter products is taken into account for the beds and evaluates parent radionuclides that buildup up to an equilibrium activity.

Because the charcoal filters collect at l east 90 percent of iodine species from the gaseous stream, an assumed 90 percent of the chemically similar bromine species also collect.

Parent-to-daughter decay chains produce halogens. One-half of the halogen production is volatile in a gaseous form. Fifty percent of the daughter halogens produced in the bed are non-gaseous and st ay in the bed. The volatile fraction of halogen production collects at a 90 percent effi ciency by the charcoal bed, resulting in a 45 percent (0.5*0.9=0.45) retention. A total of 95 percent of the daughter halogen production is retained in the bed and five percent releases to the next bed.

Accounting and treatment of the total daily production rate of noble gas daughter products is an additional incoming activity (i.e., it is added to the system with the input source streams from the LRWS and CES).

4.7 Liquid Radioactive Waste System

Up to six NPMs in a single plant share the LRWS. The LRWS processes liquid waste from primary system letdown and other sources such as RXB floor drains, hot machine shop waste, spent resins, and other contami nated inputs resulting from plant operations.

Radioactive waste process streams calc ulate decay of radionuclides, including development of daughter products, taking into account the time for fluid collection and processing operations to complete.

4.7.1 Overall Liquid Radioactive Waste System Flow and Parameters

The processing equipment for the LRWS is site-s pecific. As such, there is a simplified LRWS processing skid model. For the presented method and conclusion to be applicable, the specifications shown in Table 4-4 and Table 4-5 must be maintained by site-specific LWRS process equipment.

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Table 4-4 Processing Paths for Liquid Radioactive Waste Low-Conductivity Waste (LCW) High-Conductivity Waste (HCW) Liquid Processing Path Liquid Processing Path Combined component as Representative components modeled in planned LCW skid Granulated activated charcoal filter (2x) Pre-conditioning filter (2x) LCW liquid processing path as needed LCW filters, IX, accumulators Solids collection filter NA Accumulator vessels (3x) NA Ion exchanger vessels (5x) NA Reverse Osmosis (RO) Skid Reverse osmosis skid NA Polishers downstream of RO Polishing IX vessels (4x) NA

Table 1-4 of NUREG-0017 provides DFs for common treatment systems for PWR liquid waste. The removal DFs applied to the systems listed above are in Table 4-5.

Table 4-5 Decontamination Factors Used in Liquid Radioactive Waste System Processing for Effluent Release (Reference 7.2.1) Treatment System Decontamination Factor Anion Cs, Rb Other Nuclides LCW filters, IX, accumulators 100 17 100 LCW Reverse osmosis 10 (liq uid wastes - all nuclides) LCW polishers downstream of RO 10 10 10 HCW granulated activated carbon filter 0 0 0 Carbon bed for gaseous radioactive waste treatment 90% for iodines Evaporators (radwaste) 1000 for a ll except iodine, 100 for iodine

Liquid radioactive waste treatment of effluent source terms do not credit the granular activated charcoal beds in the LRWS.

The expected liquid waste inputs are shown in Table 4-6.

Table 4-6 Expected Liquid Waste Inputs LRWS Input Source Expected Expected Activity Input Rate LCW collection tank RXB and RWB equipment drains 2.9E+04 gpy0.001 primary 80 gpd coolant activity (PCA) Other equipment drains 1.1E+04 gpy 30 gpd 0.093 PCA Normal letdown (6 operating modules) 1.9E+05 gpy 520 gpd CVCS outlet Additional CVCS letdown streams 3.8E+04 gpy 100 gpd CVCS outlet Degasification before shut down (6 times per year) PCA through 3.0E+03 gpy evaporator, modeling PZR venting

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Table 4-6 Expected Liquid Waste Inputs (Continued) LRWS Input Source Expected Expected Activity Input Rate LCW Total 2.7E+05 gpy HCW collection tank RXB and RWB floor drains 7.3E+04 gpy (via oil separator) 200 gpd 0.1 PCA RXB reactor component cooling water drain tank (via oil separator) 3.6E+01 gpy 0.001 PCA Hot machine shop, decontamination room sump (via oil separator) 9.0E+04 gpy 0.01 PCA RXB chemical drain tank (hot lab sink) 4.4E+03 gpy (via oil separator) 12 gpd 0.05 PCA RXB chemical drain tank (CES sample tank & floor drains) 2.2E+04 gpy (via oil separator) 60 gpd CES liquid Pump seal leaks 8.1E+03 gpy (via oil separator) 22 gpd 0.1 PCA Valve packing leaks 4.8E+03 gpy (via oil separator) 13 gpd 0.1 PCA Groundwater and condensation 2.5E+05 gpy (via oil separator) 680 gpd 0.001 PCA Equipment area decontamination (outside hot machine shop) 1.5E+04 gpy (via oil separator) 40 gpd 0.01 PCA Pool inlets to HCW 2.9E+05 gpy Pool source term CVCS outlet sources into HCW 3.5E+04 gpy CVCS outlet Secondary coolant sampling drai ns 4.2E+03 gpy Secondary coolant Condensate polisher rinse and tran sfer 3.6E+04 gpy Secondary coolant Condensate polisher regeneration solu tions 1.0E+04 gpy Secondary coolant TGB floor drains 2.2E+04 gpy Secondary coolant HCW Total 8.6E+05 gpy

4.7.2 Activity Input to Liquid Radioactive Waste Collection Tanks

The LRWS collection tanks are two 16,000 gallon HCW tanks and two 16,000 gallon LCW tanks. The difference between HCW and LCW streams is that LCW is within a system boundary, whereas HCW comes through the floor or equipment drain system. In addition to a radiological component, the HCW may contain non-radiological contaminants such as dirt and oil.

Although the volume of the tanks is 16,000 gallons, the limit on total fill volume of the tanks is 14,400 gallons to prevent spilling and sloshing of liquid. This methodology uses the 14,400 gallon volume as the batch volume transferred to the liquid radioactive waste processing skids for treatment. Once a tank fills, the contents go through the processing equipment. The radionuclide content sums up from all incoming streams.

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4.7.3 Activity Input to the Oil Separators

The oil separators receive input from the following sources: RXB floor drain sump RXB reactor component cooling water drain tank Hot machine shop decontamination room sump RWB floor drain sump RXB chemical drain tank

The oil separators process these liquids before entry to the HCW collection tanks.

4.7.4 Low-Conductivity Waste Sample Tanks

The LCW sample tanks receive treated low-conductivity liquid radioactive waste after it processes through the LCW processing skids.

4.7.5 High-Conductivity Waste Sample Tanks

The HCW sample tanks receive treated high conductivity liquid radioactive waste after it processes through the HCW processing skids. To determine the radionuclide content, the sample tank fills with HCW liquid that processes through the HCW granulated activated charcoal filter and optional LCW processing skid.

4.8 Plant Effluent Release

Effluent releases from the NPP consider the sum of individual liquid and gaseous releases. Liquid and gaseous effluents tracking and tabulation are by isotope. Once the radionuclides have left the plant, the analysis of site boundary concentrations and doses are the same as if the effluents were derived from GALE.

4.8.1 Gaseous Effluent Release

During normal operations, gaseous effluent releases come from the GRWS through the gaseous charcoal decay beds and from building exhausts (both processed and direct). The sum of these gaseous effluent release pathways constitutes the total annual gaseous effluent release from the plan t. The following is a list of the modeled gaseous effluent pathways from a NPP: GRWS

- Degasifier letdown
- RPV leakage via CES RBVS
- Pool evaporation
- Containment vessel leakage AOO

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- Primary system leaks TGB
- Condenser air removal system
- System steam leaks, including from the gland seal steam condenser

In the TGB, the gland seal steam condenser and system leaks combine together into a single leakage term. The GRWS normally receives fission product gases from the primary coolant letdown (degasification) and processes them through decay beds before releasing them to the environment through the plant exhaust stack. Section 4.6 describes the added decay times allow for a r eduction in total activity coming from the plant.

Section 4.5.7 describes how the concentrati on of radionuclides in the reactor pool water spikes during refueling events and then decreases as the water is cleaned up before the next refueling event. As a result, the airborne concentration in the airspace above the reactor pool water exhibits a simi lar behavior. While RXB ventilation design is based on the peak activity concentrations, the gaseous effluent from reactor pool evaporation is based on a time-weighted annual average reactor pool water source term, pool water evaporation rate, airspace ventilation rate, and ventilation system filter efficiencies. An average airborne concentration estimates the annual off-site dose from pool evaporation in the following Equation 4-10:

A () PK= Equation 4-10

where,

A () = activity in the system at equilibrium (µCi),

P = production term by which activity is added to the system (µCi/hr), and

K = total removal rate of activity from the system (1/hr).

Then, the total airborne activity is divided by the volume of the airspace.

C RXB Air A =() V airEquation 4-11

where,

C RXB Air = airborne equilibrium concentration (µCi/ml), and

V air = volume of the airspace (ml).

The evaporated pool water releases to the environment via the RXB ventilation system at a constant rate equal to the pool room exhaust flow rate. Section 4.5.5

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describes another contribution of airborne ac tivity to the RXB ventilation system: primary system coolant leaks into the RXB originating from the CVCS. The plant exhaust stack monitors and releases airborne radionuclides captured by the building ventilation.

To account for a design-specific AOO, NuSc ale includes the gaseous effluent from an inadvertent emergency core cooling system actuation, as described in Section 4.5.8. Section 4.5 describes additional sources of gas eous effluent from the TGB, including secondary coolant steam leaks and the condenser air removal systems, which are direct (unfiltered) ground releases. The tota l gaseous effluent release from the plant is in Table A-3 of Appendix A.

4.8.2 Liquid Effluent Release

Liquid radioactive waste collects, and the HCW and LCW go to collection tanks in the RWB for processing. The collection tanks collect plant waste from normal reactor letdown, drains, resin backwash, and other contaminated liquids. The LRWS processes, samples, and discharges the liquids through a common release point through the utility water system. Section 4.7 describes the LRWS input volumes and processing parameters.

An adjustment factor to account for AOOs adds an additional 0.071 Ci per year release to the cumulative non-tritium liquid effluent releases. This value scales linearly with reactor thermal power (250 MWth

6 vs. 3400 MWth) from the 0.16 Ci per year value from NUREG-0017.

The total liquid effluent release from the plant is in Table A-3 of Appendix A.

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5.0 Fuel Failure Fraction

The GALE code is based on empirical (operating) data. Therefore, NUREG-0017 does not specify a fuel failure fraction. The methodology employs a first-principles calculation that determines fission-product related contributions to effluents by assuming a realistic and conservative fuel failure fraction. The industry-reported fuel failure fraction is an equivalent release value that represents the effects from several failure mechanisms. The evaluation of fuel isotopic inventory uses a NuScale-assumed fuel failure fraction, with radionuclide release, buildup and removal, equilibrium concentrations in the primary coolant, and forms the basis for determining liquid and gaseous contributions.

Without operating history, the methodology us es available industry operating experience based on the similarities between the NuScale core and fuel design compared to existing PWRs. The design uses the same 17 x 17 PWR fuel assemblies, shorter in length, with Framatome M5TM cladding and low enriched uranium-235 uranium dioxide pellets in helium-backfilled and pressurized fuel rods. The se lection of a fuel failure fraction is based on recent PWR fuel performance observed in the operating fleet for similar type fuel.

5.1 US Pressurized Water Reactor Fuel Failure History

The numerical basis for fuel failure rates comes from EPRIs PWR Fuel Rod Failure Rate Analysis report (Reference 7.2.20). Table A-4 presents data from the Fuel Reliability Database report. Table 5-1 presents a summary of the data from this report representing the years 2007-2016 highlighting the highest and lowest values of failed fuel fraction. This data represents fuel failures determined with the outage method as described in the International Atomic Energy Agency (IAEA) Review of Fuel Failures in Water Cooled Reactors, IAEA Nuclear Energy Series No. NF-T-2.1 (Reference 7.2.24), as the number of failed pins during an operating cycle divided by the total number of pins with a refueling outage that year. This data includes only oper ating PWRs within the U.S. with zirconium clad fuel.

Table 5-1 Fuel Failure Values (Reference 7.2.20) ((

}}2(a),(c)

Table 5-1 shows that the lowest data point in the most recent ten years of U.S. PWR data is (( }}2(a),(c) reported in (( }}2(a),(c) (Reference 7.2.6). The highest data point is 66 rods per million (0.0066 percent) reported in (( }}2(a),(c) (Reference 7.2.7). For comparison, NUREG-0017 was published in 1985 and based on data from the 1970s (Reference 7.2.1). The ANSI/ANS

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ANS-18.1-1999 standard for primary and secondary coolant concentrations was published in 1999 based on industry data of that time. The PWRGALE-09 code was benchmarked against operational reactor data from 2005 to 2010 (Reference 7.2.8). The average fuel failure fraction from that time period was ((

}}2(a),(c).

Table A-4 lists the fuel failure data for U.S. PWRs with zirconium-alloy cladding for the years 2002 to 2016.

5.2 Fuel Failure Fraction Conclusions

The NPP fuel design is based on a standard design Framatome 17 x 17 fuel assembly, which is approximately half the length of current large PWRs. The fuel uses the same fabrication techniques, quality assurance, and testing as the fuel assemblies fabricated by Framatome and irradiated in large PWRs. Thus the utilization of fuel performance from the operating PWR fleet is applicable for this purpose.

The long-term industry trend on improved fuel performance is well defined and highlights the continuing improvement. Therefore, to replace the GALE empirical data, the methodology uses a realistic and conservative fuel failure fraction based on industry performance of 0.0066percent (66 rods per million) for fission product-related effluent releases. The following supports this value. In U.S. PWRs, the fuel failure fraction has decreased and continues to decrease over time, with the most recent data ((

}}2(a),(c) and a maximum value of the most recent ten years of available data (( }}2(a),(c) of 66 rods per million (0.0066 percent).

The method uses a realistic and conservati ve fuel failure fraction based on industry performance of 0.0066 percent with conservative escape rate coefficients for the purpose of evaluating fission product-related effluent releases. The realistic and conservative fuel failure fraction is initiated at the beginning of cycle and is sustained for the duration of the fuel cycle.

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6.0 Summary and Conclusion

The NPP design is similar to large PWRs in the existing fleet with regard to effluent releases (production, process, and release). Du e to differences associated with a smaller, passive NPP design, the GALE code is not representative of the NPP design and does not accurately estimate NPP effluent releases. This results in the need to develop an alternate GALE replacement methodology. The effluent release methodology described in this report is based on compliance with applicable regulations, with no exemptions needed.

The methodology is realistic and conservative, using first-principles-based calculations where appropriate, combined with recent nuclear industry experience and lessons learned. Table 6-1 presents a summary of the effluent release methodology. Liquid and gaseous effluents use realistic and conserva tive source terms. The calculation of effluents includes the design-specific treatment of liquid and gaseous radioactive source terms such as filtration, resin absorption, holdup, dilution, and decay.

Table 6-1 Primary Contributors and Methodology Employed for Effluents Primary Contributors NuScale Methodology Water activation products

H-3 (tritium)
C-14 (radiocarbon) Calculations based on first-principles physics
N-16
Ar-41 Activated corrosion and wear products (CRUD)
Recent large PWR operating data* Lessons learned

Fission products (failed-fuel related)

Calculations based on first-principles physics* Recent large PWR operating data

The primary and secondary coolant isotopic distribution is in Table A-3. The total effluents are 850 Ci of gaseous effluent and 1,200 Ci of liquid effluent, with tritium being the largest contributor to both. The gaseous effluent is evaluated in Mode 2 (RCS recycled to pool), and the liquid effluent is evaluated in Mode 1 (no recycling of RCS), which maximizes the respective effluent releases.

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7.0 References

7.1 Source Documents

7.1.1 American Society of Mechanical Engineers, Quality Assurance Program Requirements for Nuclear Facility Applications (QA), ASME NQA-1-2008, ASME NQA-1a-2009 Addenda, as endorsed by Regulatory Guide 1.28, Rev. 4, New York, NY.

7.1.2 U.S. Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Facilities, Appendix B, Part 50, Chapter 1, Title 10, Energy, (10 CFR 50 Appendix B).

7.2 Referenced Documents

7.2.1 U.S. Nuclear Regulatory Commission, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluen ts from Pressurized Water Reactors (PWRGALE Code), NUREG-0017, Rev. 1, April 1985.

7.2.2 American National Standard Institute/American Nuclear Society, Radioactive Source Term for Normal Operation of Light Water Reactors, ANSI/ANS-18.1-1999, LaGrange Park, IL.

7.2.3 Pacific Northwest National Laboratory, Applicability of GALE-86 Codes to Integral Pressurized Water Reactor Designs, PNNL-21386, May 2012.

7.2.4 Korea Electric Power Corporation (KEPKO) and Korea Hydro & Nuclear Power Co., Ltd. (KHNP), APR1400 Design Control Document Revision 0, December 2014, NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML15006A059

7.2.5 Electric Power Research Institute "Nuclear Power Plant Related Iodine Partition Coefficients," EPRI-NP-1271, December 1979

7.2.6 Idaho National Laboratory, Bragg-Sitton, S., Light Water Reactor Sustainability Program, Advanced LWR Nuclear Fuel Cl adding System Development: Technical Program Plan, INL/MIS-12-25696, Rev. 1, December 2012.

7.2.7 Electric Power Research Institute, The Path to Zero Defects: EPRI Fuel Reliability Guidelines, EPRI, Palo Alto, CA, 2008.

7.2.8 Geelhood, K.J., Benchmarking of GALE-09 Release Predictions Using Site Specific Data from 2005 to 2010, PNNL-22076, November 2012.

7.2.9 Westinghouse, AP1000 Design Control Document Revision 19, June 2011, NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML11171A500.

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7.2.10 AREVA NP Inc., NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML13220A883.

7.2.11 Mitsubishi Heavy Industries, LTD., US-APWR Design Control Document Revision 4, September 2013, NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML13262A304.

7.2.12 Electric Power Research Institute, EPRI Tritium Management Model Project Summary Report, (EPRI #1009903), Palo Alto, CA, November 2005.

7.2.13 Electric Power Research Institute, Steam Generator Management Program: PWR Primary-to Secondary Leak Guidelines, (EPRI #1022832), Rev. 4, Palo Alto, CA, September 2011.

7.2.14 Oak Ridge National Laboratory, SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge, Tennessee, June 2011.

7.2.15 U.S. Code of Federal Regulations, St andards for Protection against Radiation, Part 20, Chapter 1, Title 10, Energy, (10 CFR 20).

7.2.16 U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy, (10 CFR 50).

7.2.17 American Nuclear Society, 2014 Performance Indicators issued for U.S. Power Reactors, Nuclear News, June 2015.

7.2.18 U.S. Nuclear Regulatory Commission, Radioactive Effluents from Nuclear Power Plants-Annual Report 2010 - Final Report, NUREG/CR-2907, Vol. 16, May 2018. NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML18151A529

7.2.19 Electric Power Research Institute, EPRI Pressurized Water Reactor Primary Water Chemistry Guidelines, 3002000505, Vol. 1, Rev. 7, Palo Alto, CA, 2014.

7.2.20 Electric Power Research Institute, PWR Fuel Rod Failure Rate Analysis, FRP_2018_1 Final Letter Report, Palo Alto, CA, February 2018.

7.2.21 Castelli, R. A., Nuclear Corrosion Modelling, The Nature of CRUD, Elsevier, Oxford, 2010.

7.2.22 International Atomic Energy Agency, Modelling of Transport of Radioactive Substances in the Primary Circuit of Water-Cooled Reactors, IAEA-TECDOC-1672, Vienna, Austria, March 2012.

7.2.23 Electric Power Research Institute, Steam Generator Management Program: PWR Primary-to-Secondary Leak Guidelines Revision 4," EPRI Technical Report 1022832, Palo Alto, Ca, September 2011.

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7.2.24 International Atomic Energy Agency, Review of Fuel Failures in Water Cooled Reactors, IAEA Nuclear Energy Series No. NF-T-2.1, June 2010.

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Appendix A Summary Tables Table A-1 Maximum Fuel Inventory per Assembly (Ci) Radionuclide Assembly Activity (Ci) Noble Gases Kr83m 2.0E+04 Kr85m 4.0E+04 Kr85 4.1E+03 Kr87 7.7E+04 Kr88 1.0E+05 Kr89 1.2E+05 Xe131m 2.8E+03 Xe133m 1.2E+04 Xe133 3.8E+05 Xe135m 8.8E+04 Xe135 1.1E+05 Xe137 3.4E+05 Xe138 3.1E+05 Halogens Br82 1.1E+03 Br83 1.9E+04 Br84 3.2E+04 Br85 4.0E+04 I129 1.5E-02 I130 1.1E+04 I131 1.9E+05 I132 2.8E+05 I133 3.8E+05 I134 4.2E+05 I135 3.6E+05 Rubidium, Cesium Rb86m 8.4E+01 Rb86 6.7E+02 Rb88 1.0E+05 Rb89 1.3E+05 Cs132 1.4E+01 Cs134 9.3E+04 Cs135m 1.2E+03 Cs136 2.1E+04 Cs137 4.8E+04 Cs138 3.4E+05 Other FPs P32 2.9E+01 Co57 2.2E-01 Sr89 1.4E+05 Sr90 3.3E+04 Sr91 1.8E+05 Sr92 2.0E+05

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Table A-1 Maximum Fuel Inventory per Assembly (Ci) (Continued) Radionuclide Assembly Activity (Ci) Y90 3.4E+04 Y91m 1.1E+05 Y91 1.9E+05 Y92 2.1E+05 Y93 2.4E+05 Zr97 3.0E+05 Nb95 2.9E+05 Mo99 3.4E+05 Mo101 3.3E+05 Tc99m 3.0E+05 Tc99 5.9E+00 Ru103 3.7E+05 Ru105 3.0E+05 Ru106 2.3E+05 Rh103m 3.7E+05 Rh105 2.8E+05 Rh106 2.6E+05 Ag110 9.7E+04 Sb124 5.6E+02 Sb125 4.1E+03 Sb127 2.2E+04 Sb129 6.4E+04 Te125m 9.6E+02 Te127m 3.6E+03 Te127 2.2E+04 Te129m 1.1E+04 Te129 6.2E+04 Te131m 4.2E+04 Te131 1.6E+05 Te132 2.7E+05 Te133m 1.7E+05 Te134 3.2E+05 Ba137m 4.5E+04 Ba139 3.3E+05 Ba140 3.1E+05 La140 3.3E+05 La141 2.9E+05 La142 2.8E+05 Ce141 3.0E+05 Ce143 2.7E+05 Ce144 2.5E+05 Pr143 2.6E+05 Pr144 2.5E+05 Np239 5.2E+06

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Table A-2 Primary and Secondary Coolant Radionuclide Activity Concentrations (Mode 1) Radionuclide Primary Activity (µCi/g) Secondary Activity (µCi/g) Noble Gases Kr83m 7.7E-04 2.7E-10 Kr85m 3.2E-03 1.1E-09 Kr85 1.6E-01 5.7E-08 Kr87 1.8E-03 6.2E-10 Kr88 5.1E-03 1.8E-09 Kr89 1.2E-04 4.2E-11 Xe131m 1.3E-02 4.5E-09 Xe133m 1.1E-02 4.0E-09 Xe133 8.3E-01 2.9E-07 Xe135m 1.1E-03 3.9E-10 Xe135 2.3E-02 8.1E-09 Xe137 3.9E-04 1.4E-10 Xe138 1.3E-03 4.7E-10 Halogens Br82 2.1E-05 7.6E-12 Br83 1.2E-04 4.3E-11 Br84 5.7E-05 1.8E-11 Br85 6.9E-06 1.2E-12 I129 3.5E-10 1.3E-16 I130 1.7E-04 6.2E-11 I131 4.4E-03 1.6E-09 I132 2.1E-03 7.2E-10 I133 6.7E-03 2.4E-09 I134 1.2E-03 4.1E-10 I135 4.3E-03 1.5E-09 Rubidium, Cesium Rb86m 5.2E-09 4.5E-16 Rb86 3.0E-05 1.2E-11 Rb88 5.1E-03 1.7E-09 Rb89 2.4E-04 7.5E-11 Cs132 6.0E-07 2.3E-13 Cs134 4.3E-03 1.7E-09 Cs135m 3.6E-06 1.3E-12 Cs136 9.4E-04 3.7E-10 Cs137 2.2E-03 8.7E-10 Cs138 1.9E-03 6.8E-10 Other FPs P32 8.4E-11 3.0E-17 Co57 6.4E-13 2.3E-19 Sr89 3.8E-06 1.4E-12 Sr90 5.9E-07 2.1E-13 Sr91 2.0E-06 7.1E-13 Sr92 1.1E-06 3.7E-13

TR-123242-NP Revision 1

Table A-2 Primary and Secondary Coolant Radionuclide Activity Concentrations (Mode 1) (Continued) Radionuclide Primary Activity (µCi/g) Secondary Activity (µCi/g) Y90 1.4E-07 5.2E-14 Y91m 1.1E-06 3.6E-13 Y91 5.6E-07 2.0E-13 Y92 9.0E-07 3.2E-13 Y93 4.3E-07 1.5E-13 Zr97 6.3E-07 2.2E-13 Nb95 1.6E-06 5.6E-13 Mo99 1.1E-03 4.0E-10 Mo101 4.3E-05 1.3E-11 Tc99m 1.0E-03 3.7E-10 Tc99 2.1E-08 7.6E-15 Ru103 1.1E-06 3.9E-13 Ru105 3.6E-07 1.3E-13 Ru106 6.8E-07 2.4E-13 Rh103m 1.1E-06 3.6E-13 Rh105 7.3E-07 2.6E-13 Rh106 6.8E-07 3.2E-14 Ag110 4.8E-06 1.9E-13 Sb124 1.6E-09 5.8E-16 Sb125 1.2E-08 4.3E-15 Sb127 6.1E-08 2.2E-14 Sb129 7.6E-08 2.7E-14 Te125m 1.7E-06 6.2E-13 Te127m 6.6E-06 2.4E-12 Te127 2.6E-05 9.4E-12 Te129m 1.9E-05 6.8E-12 Te129 2.7E-05 9.3E-12 Te131m 6.2E-05 2.2E-11 Te131 3.1E-05 9.8E-12 Te132 4.6E-04 1.6E-10 Te133m 3.9E-05 1.3E-11 Te134 5.6E-05 1.8E-11 Ba137m 2.1E-03 3.3E-10 Ba139 1.0E-06 3.6E-13 Ba140 5.6E-06 2.0E-12 La140 1.6E-06 5.8E-13 La141 3.2E-07 1.1E-13 La142 1.5E-07 5.3E-14 Ce141 8.6E-07 3.1E-13 Ce143 6.5E-07 2.3E-13 Ce144 7.3E-07 2.6E-13 Pr143 7.6E-07 2.7E-13 Pr144 7.2E-07 2.2E-13 Np239 1.4E-05 4.9E-12

TR-123242-NP Revision 1

Table A-2 Primary and Secondary Coolant Radionuclide Activity Concentrations (Mode 1) (Continued) Radionuclide Primary Activity (µCi/g) Secondary Activity (µCi/g) Corrosion Activation Products - CRUD Na24 1.4E-02 4.9E-09 Cr51 7.7E-04 2.8E-10 Mn54 4.0E-04 1.4E-10 Fe55 3.0E-04 1.1E-10 Fe59 7.5E-05 2.7E-11 Co58 1.1E-03 4.1E-10 Co60 1.3E-04 4.7E-11 Ni63 6.6E-05 2.3E-11 Zn65 1.3E-04 4.5E-11 Zr95 9.7E-05 3.5E-11 Ag110m 3.2E-04 1.2E-10 W187 7.0E-04 2.5E-10 Water Activation Products H3 1.3E+00 2.4E-03 C14 2.6E-04 9.4E-11 N161 4.1E+01 1.5E-05 Ar41 1.4E-01 5.0E-08 1 N16 concentration values represented are for the t op (entrance) of the SG region. N16 values vary throughout the NPM primary coolant volume due to decay during transit from low primary coolant flow rate.

Table A-3 Gaseous and Liquid Yearly Effluent Release Values for a NuScale Power Plant (with Six Operating Modules) Gaseous Effluent (Ci/yr) Liquid Effluent Radionuclide Plant Exhaust Turbine Generator Total (Ci/Yr)1 Stack Releases Building Releases Noble Gases Kr83m 9.1E-03 4.2E-03 1.3E-02 - Kr85m 3.8E-02 1.8E-02 5.6E-02 - Kr85 1.4E+02 8.9E-01 1.5E+02 - Kr87 2.1E-02 9.7E-03 3.1E-02 - Kr88 6.1E-02 2.8E-02 8.9E-02 - Kr89 1.4E-03 6.4E-04 2.0E-03 - Xe131m 7.0E-01 6.9E-02 7.7E-01 - Xe133m 5.6E-01 6.3E-02 6.2E-01 - Xe133 1.6E+01 4.6E+00 2.0E+01 - Xe135m 6.7E-02 6.1E-03 7.3E-02 - Xe135 3.0E-01 1.3E-01 4.3E-01 - Xe137 4.6E-03 2.1E-03 6.7E-03 - Xe138 1.6E-02 7.2E-03 2.3E-02 - Halogens Br82 1.0E-06 2.8E-08 1.0E-06 2.3E-09

TR-123242-NP Revision 1

Table A-3 Gaseous and Liquid Yearly Effluent Release Values for a NuScale Power Plant (with Six Operating Modules) (Continued) Gaseous Effluent (Ci/yr) Liquid Effluent Radionuclide Plant Exhaust Turbine Generator Total (Ci/Yr)1 Stack Releases Building Releases Br83 5.8E-06 1.6E-07 5.9E-06 - Br84 2.7E-06 6.9E-08 2.8E-06 - Br85 3.2E-07 4.3E-09 3.3E-07 - I129 1.7E-11 4.7E-13 1.7E-11 1.9E-12 I130 8.2E-06 2.3E-07 8.5E-06 2.1E-11 I131 5.3E-04 5.9E-06 5.4E-04 1.7E-05 I132 9.8E-05 2.7E-06 1.0E-04 3.8E-07 I133 3.5E-04 9.0E-06 3.5E-04 5.8E-08 I134 5.7E-05 1.5E-06 5.9E-05 - I135 2.0E-04 5.6E-06 2.1E-04 5.3E-14 Rubidium, Cesium Rb86m 1.2E-10 1.3E-12 1.2E-10 - Rb86 1.5E-06 3.5E-08 1.6E-06 5.0E-06 Rb88 1.2E-04 5.0E-06 1.3E-04 - Rb89 5.5E-06 2.2E-07 5.7E-06 - Cs132 2.8E-08 7.0E-10 2.9E-08 3.3E-08 Cs134 2.3E-04 5.0E-06 2.3E-04 1.4E-03 Cs135m 8.4E-08 3.9E-09 8.8E-08 - Cs136 4.7E-05 1.1E-06 4.8E-05 1.2E-04 Cs137 1.2E-04 2.6E-06 1.2E-04 7.2E-04 Cs138 4.5E-05 2.0E-06 4.7E-05 - Other Fission Products P32 2.7E-12 8.9E-14 2.7E-12 3.4E-13 Co57 2.1E-14 6.8E-16 2.1E-14 4.5E-15 Sr89 1.2E-07 4.1E-09 1.3E-07 3.6E-08 Sr90 1.9E-08 6.3E-10 2.0E-08 4.4E-09 Sr91 4.7E-08 2.1E-09 4.9E-08 1.3E-14 Sr92 2.5E-08 1.1E-09 2.6E-08 - Y90 6.6E-09 1.5E-10 6.7E-09 4.1E-09 Y91m 2.5E-08 1.1E-09 2.6E-08 8.1E-15 Y91 1.8E-08 5.9E-10 1.9E-08 3.5E-09 Y92 2.1E-08 9.5E-10 2.2E-08 - Y93 1.0E-08 4.5E-10 1.1E-08 5.6E-15 Zr97 1.5E-08 6.7E-10 1.6E-08 1.2E-12 Nb95 3.9E-05 1.7E-09 3.9E-05 7.7E-07 Mo99 3.2E-05 1.2E-06 3.4E-05 6.8E-07 Mo101 1.0E-06 3.8E-08 1.1E-06 - Tc99m 3.0E-05 1.1E-06 3.1E-05 6.6E-07 Tc99 7.0E-10 2.3E-11 7.2E-10 1.6E-10 Ru103 3.5E-08 1.2E-09 3.6E-08 6.3E-09 Ru105 8.4E-09 3.8E-10 8.8E-09 - Ru106 2.2E-08 7.2E-10 2.3E-08 4.9E-09 Rh103m 3.4E-08 1.1E-09 3.6E-08 6.3E-09

TR-123242-NP Revision 1

Table A-3 Gaseous and Liquid Yearly Effluent Release Values for a NuScale Power Plant (with Six Operating Modules) (Continued) Gaseous Effluent (Ci/yr) Liquid Effluent Radionuclide Plant Exhaust Turbine Generator Total (Ci/Yr)1 Stack Releases Building Releases Rh105 2.0E-08 7.8E-10 2.0E-08 8.5E-11 Rh106 2.2E-08 9.7E-11 2.2E-08 4.9E-09 Ag110 4.0E-05 5.7E-10 4.0E-05 1.9E-07 Sb124 5.2E-11 1.7E-12 5.4E-11 1.0E-11 Sb125 3.9E-10 1.3E-11 4.1E-10 8.8E-11 Sb127 1.8E-09 6.5E-11 1.9E-09 6.7E-11 Sb129 1.8E-09 7.9E-11 1.8E-09 - Te125m 5.6E-08 1.9E-09 5.8E-08 1.1E-08 Te127m 2.2E-07 7.1E-09 2.2E-07 4.5E-08 Te127 6.8E-07 2.8E-08 7.1E-07 4.4E-08 Te129m 6.1E-07 2.0E-08 6.3E-07 1.1E-07 Te129 7.4E-07 2.8E-08 7.7E-07 6.8E-08 Te131m 1.6E-06 6.6E-08 1.7E-06 3.5E-09 Te131 7.6E-07 2.9E-08 7.9E-07 7.9E-10 Te132 1.3E-05 4.8E-07 1.4E-05 3.7E-07 Te133m 9.2E-07 3.9E-08 9.5E-07 - Te134 1.3E-06 5.5E-08 1.4E-06 - Ba137m 1.1E-04 9.7E-07 1.1E-04 6.8E-04 Ba139 2.4E-08 1.1E-09 2.5E-08 - Ba140 1.8E-07 5.9E-09 1.8E-07 2.1E-08 La140 7.3E-08 1.7E-09 7.5E-08 2.3E-08 La141 7.5E-09 3.4E-10 7.8E-09 - La142 3.6E-09 1.6E-10 3.7E-09 - Ce141 2.8E-08 9.1E-10 2.9E-08 4.8E-09 Ce143 1.7E-08 6.9E-10 1.8E-08 5.4E-11 Ce144 2.4E-08 7.7E-10 2.4E-08 5.2E-09 Pr143 2.4E-08 8.1E-10 2.5E-08 3.2E-09 Pr144 2.3E-08 6.5E-10 2.4E-08 5.1E-09 Np239 3.8E-07 1.5E-08 4.0E-07 5.8E-09 Corrosion Activation Products - CRUD Na24 3.3E-04 1.5E-05 3.5E-04 1.0E-08 Cr51 6.7E-03 8.2E-07 6.7E-03 2.7E-05 Mn54 3.6E-03 4.2E-07 3.6E-03 1.7E-05 Fe55 2.7E-03 3.2E-07 2.7E-03 1.3E-05 Fe59 6.6E-04 8.0E-08 6.6E-04 2.9E-06 Co58 1.0E-01 1.2E-06 1.0E-01 4.0E-04 Co60 1.2E-03 1.4E-07 1.2E-03 5.9E-06 Ni63 6.0E-04 7.0E-08 6.0E-04 3.0E-06 Zn65 1.2E-03 1.3E-07 1.2E-03 5.5E-06 Zr95 8.7E-04 1.0E-07 8.7E-04 3.9E-06 Ag110m 2.9E-03 3.4E-07 2.9E-03 1.4E-05 W187 1.3E-03 7.4E-07 1.3E-03 3.6E-08 Water Activation Products

TR-123242-NP Revision 1

Table A-3 Gaseous and Liquid Yearly Effluent Release Values for a NuScale Power Plant (with Six Operating Modules) (Continued) Gaseous Effluent (Ci/yr) Liquid Effluent Radionuclide Plant Exhaust Turbine Generator Total (Ci/Yr)1 Stack Releases Building Releases H3 6.7E+02 7.4E+00 6.8E+02 1.2E+03 C14 2.4E-01 2.8E-07 2.4E-01 - N16---- Ar41 4.0E+00 7.7E-01 4.8E+00 -

1. The total does not include an adjustment factor to ac count for AOOs that adds an additional 0.071 Ci per year release to the cumulative n on-tritium liquid effluent releases.

Table A-4 Fuel Failure Data for U.S. Pressurized Water Reactors with Zirconium-Alloy Cladding (Reference 7.2.20) ((

}}2(a),(c)

ML23304A359

Text

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