EXCLUSION ZONES for SMALL MODULAR REACTORS a Thesis

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EXCLUSION ZONES FOR SMALL MODULAR REACTORS

A Thesis

Submitted to the Faculty of Graduate Studies and Research

In Partial Fulfillment of the Requirements

For the Degree of

Master of Applied Science

Industrial Systems Engineering

University of Regina

Bradley Edward Rudolph Lulik

Regina, Saskatchewan

March 2020

UNIVERSITY OF REGINA

FACULTY OF GRADUATE STUDIES AND RESEARCH

SUPERVISORY AND EXAMINING COMMITTEE

Bradley Edward Rudolph Lulik, candidate for the degree of Master of Applied Science in Industrial Systems Engineering, has presented a thesis titled, Exclusion Zones for Small Modular Reactors, in an oral examination held on March 27, 2020. The following committee members have found the thesis acceptable in form and content, and that the candidate demonstrated satisfactory knowledge of the subject material.

External Examiner: Dr. Irfan Al-Anbagi, Electronic Systems Engineering

Co-Supervisor: Dr. Esam Hussein, General Engineering

Co-Supervisor: Dr. David deMontigny, Industrial Systems Engineering

Committee Member: Dr. Adisorn Aroonwilas, Industrial Systems Engineering

Committee Member: Dr. Golam Kabir, Industrial Systems Engineering

Chair of Defense: Dr. Christopher Yost, Department of BIology

All Participated via ZOOM Abstract

The objective of this thesis is to estimate the size of the exclusion zone around a small modular reactor (SMR). The aim of such zone is to provide an atmospheric space sufficient to dilute any radioactive releases during an accident, to a level below the safe regulated radiation dose for the public. A hypothetical severe accident is considered for a generic SMR, and the whole-body radiation dose associated with the accident was estimated at various distances and reactor power levels. The results were verified against those of a more complex model for a typical CANDU reactor. The obtained results were then employed to estimate the radius of the exclusion zone, by determining the distance at which the dose is at or slightly below the permitted dose to a member of the public.

The method first estimates the quantity and type of radioactive materials available for release to the environment following a nuclear accident, known as the Source Term.

This thesis employed a simplified approach for estimating the Source Term, utilizing the magnitude of the fission product yields, radionuclide release fractions, and reactor thermal power.

The estimated Source Term values were then used as input to an atmospheric plume dispersion model, to determine the radiation dose at various distances after dilution. The HotSpot Health Physics code was employed to estimate the radiation dose, because it is a convenient and efficient tool for the many calculations associated with the numerous radionuclides that would be released during a postulated reactor accident.

In addition to the effect of atmospheric dilution of radionuclides, the thesis also examined how the size of the exclusion zone is influenced by technical regulations and standards, reactor design and safety features, and by the presence of engineered barriers.

Further, this thesis presents a survey of SMR designs currently in development and a review of their unique safety features.

iii

Acknowledgements

I wish to acknowledge my co-supervisor, mentor, and friend Dr Esam Hussein.

During my time at the University of Regina, I have had the privilege of learning from Dr

Hussein. I am immensely grateful for his continued guidance, support, encouragement, and patience. The Faculty of Engineering and Applied Science is blessed to have Dr

Hussein as Dean.

I would like to express my appreciation to my co-supervisor Dr David deMontigny. Over the past eight years, both as an undergraduate and graduate student, it has been a privilege to work with Dr deMontigny. His commitment to the quality of education being provided by the Faculty of Engineering and Applied Science is beyond compare.

To the Silvia Fedoruk Canadian Centre for Nuclear Innovation’s Board of

Directors, thank you for seeing the value in developing Saskatchewan’s technical capacity related to the siting of Small Modular Reactors. I am grateful for the monetary support that has allowed me to complete my work and appreciative of the opportunity to participate on this multidisciplinary project.

Dedication

To my wife, Justine, for her unconditional love, support, and patience throughout this endeavour and others.

To my parents, Debbie and Emil, for their love, encouragement, and continued interest in my studies.

Table of Contents

Abstract ...... ii

Acknowledgements ...... iv

Dedication ...... v

Table of Contents ...... vi

List of Tables ...... viii

List of Figures ...... ix

CHAPTER 1: INTRODUCTION ...... 1 1.1 Small Modular Reactors ...... 1 1.2 Canadian Regulations ...... 3 1.3 Exclusion Zone ...... 5 1.4 Thesis Objectives and Outline ...... 7

CHAPTER 2: SOURCE TERM ...... 9 2.1 Introduction ...... 9 2.2 Source Term ...... 10 2.3 Approximation ...... 11 2.4 Source Term Verification ...... 12 2.5 Sensitivity Analysis ...... 14 2.6 Conclusions ...... 18

CHAPTER 3: RADIOACTIVITY DISPERSION AND EXCLUSION ZONE ...... 19 3.1 Introduction ...... 19 3.2 HotSpot ...... 20 3.3 Verification ...... 23 3.4 Exclusion Zone for SMRs ...... 28 3.5 Sensitivity Analysis ...... 29 3.6 Conclusions ...... 33

CHAPTER 4: REDUCING EXCLUSION ZONE THROUGH DESIGN ...... 34 4.1 Introduction ...... 34 4.2 The Exclusion Zone ...... 35 4.3 Inherent and Passive Safety ...... 37 4.4 Reactor Material...... 39 4.5 Engineered Features and Barriers ...... 41 4.6 Conclusions ...... 43

CHAPTER 5: CONCLUSIONS ...... 45 5.1 Summary ...... 45 5.2 Conclusions ...... 47 5.3 Contribution to Knowledge ...... 48 5.4 Recommendations for Future Work ...... 48

REFERENCES ...... 51

Appendix A: Review of Small Modular Reactors ...... A-1 A.1 Introduction ...... A-1 A.2 Water Cooled Reactors ...... A-2 A.2.1 Light Water Reactors ...... A-2 A.2.2 Heavy Water Reactors ...... A-19 A.3 Gas Cooled Reactors ...... A-24 A.4 Molten Salt Reactors ...... A-28 A.5 Fast Neutron Spectrum Reactors ...... A-39 A.6 Summary ...... A-50

vii

List of Tables

Table 2-1: Fission product groups [20] ...... 11 Table 2-2: Published and simulated source term for significant radionuclides ...... 14 Table 2-3: Simulated source term for release fractions reduced by 25%...... 16 Table 2-4: Fission product yield range for significant radionuclides [23] ...... 17 Table 2-5: Simulated source term range for significant radionuclides ...... 18 Table 3-6: Significant Radionuclides [20] ...... 22 Table 3-7: Published dose for 24 hour accident and 24 hour generic large release ...... 25 Table 3-8: Published doses and HotSpot-simulated doses using Equation (2.1)...... 26 Table 3-9: Published doses and HotSpot-simulated doses using calibrated Equation (2.1)...... 27 Table 3-10: SMR dose at a variety of distances ...... 29 Table 3-11: Sensitivity to effective release height ...... 30 Table 3-12: Sensitivity to wind speed ...... 31 Table 3-13: Sensitivity to atmospheric stability ...... 32 Table A-1: Summary of light water cooled small modular reactors [38, 2, 40] ...... A-3 Table A-2: Main features of heavy water cooled SMRs [38, 2, 40] ...... A-20 Table A-3: summary of gas cooled SMRs [38, 2, 40] ...... A-24 Table A-4: Summary of Molten Salt SMRs [61, 62, 63, 64, 65, 66, 67] ...... A-29 Table A-5: Summary of Fast neutron Spectrum SMRs ...... A-40

viii

List of Figures

Figure 3-1: Procedure for the Verification of HotSpot ...... 23 Figure 3-2: Three levels of verification ...... 28 Figure A-1: A schematic of FBNR Fuel [41] ...... A-4 Figure A-2: A schematic of KLT-40S Fuel Assembly [42] ...... A-5 Figure A-3: An overview of a VBER-3000 SMR [43] ...... A-6 Figure A-4: An overview of a VVER-600 SMR [44] ...... A-8 Figure A-5: A schematic of VVER-640 SMR [47] ...... A-9 Figure A-6: A schematic of an IMR SMR [47] ...... A-11 Figure A-7: A schematic of a NuScale SMR [48] ...... A-12 Figure A-8: A schematic of a SMART SMR [49] ...... A-13 Figure A-9: A schematic of a ACP-100 SMR [2] ...... A-14 Figure A-10: A schematic of a mPower SMR [2] ...... A-15 Figure A-11: A schematic of Westinghouse SMR [2] ...... A-16 Figure A-12: A schematic of SMR-160 [2] ...... A-18 Figure A-13: Fuel Cycle for AHWR [50] ...... A-21 Figure A-14: Spherical Fuel Elements [56] ...... A-25 Figure A-15: PBMR Fuel Element Design [57] ...... A-26 Figure A-16: Prismatic HTR reactor schematic [58] ...... A-27 Figure A-17: MK1PB-FHR reactor schematic [61] ...... A-30 Figure A-18: ThorCon reactor schematic [69] ...... A-31 Figure A-19: IMSR-400 core schematic [70] ...... A-32 Figure A-20: MSTW optimal configuration principle [64] ...... A-33 Figure A-21: MSR-FUJI reactor schematic [65] ...... A-34 Figure A-22: SSR reactor core module [72] ...... A-35 Figure A-23: SmAHTR vessel schematic [73] ...... A-37 Figure A-24: SmAHTR vessel schematic [87] ...... A-42 Figure A-25: CLEAR-I reactor schematic [89] ...... A-43 Figure A-26: ALFRED vessel schematic [80] ...... A-44 Figure A-27: ELFR vessel schematic [90] ...... A-45 Figure A-28: PEACER core arrangement [91] ...... A-46 Figure A-29: BREST-OD-300 Schematic [83] ...... A-47 Figure A-30: SVBR-100 Schematic [84] ...... A-48 Figure A-31: G4M plant layout [85] ...... A-49

CHAPTER 1: INTRODUCTION

1.1 Small Modular Reactors

Interest in small modular reactors (SMRs) has recently grown, as they are viewed as a tool to reduce CO2 and other green-house-gas emissions if they replace traditional power generating facilities, such as aging coal-fired units. SMRs can also be integrated with renewable-energy electrical generators to create hybrid energy systems [1]. An SMR is distinguished from a conventional nuclear power station by a number of features. An

SMR has an electrical output of 30 MWe to 300 MWe [2]. Conventional power reactors produce higher power levels, e.g. the CANDU Bruce B Nuclear Generating Station in

Ontario is rated at 817 MWe per unit [3], while pressurized water reactors (PWRs) operating today produce between 300 to 1,660 MWe [4]. Another feature of SMRs is that they are to be produced in modules in a factory, to shorten construction timelines and lower initial capital expenditure [1]. Moreover, the smaller power levels permit the installation of SMRs in a serial manner, permitting units to be installed sequentially, as required, to meet demand [1].

The smaller size and the modular nature of SMRs allows for flexible application for a wide range of uses, whereas conventional commercial reactors are limited to electricity production. SMRs can serve off-grid locations, such as remote jurisdictions and industrial sites, while on-grid installation can be accommodated in smaller grids.

There is also a possibility to use SMRs for district heating and seawater desalination because of their small size [2]. Methanol production, petroleum refining, thermochemical hydrogen production, and coal gasification are other potential uses of SMRs.

While, many SMR designs have been proposed, the International Atomic Energy

Agency (IAEA) lists forty-eight SMR designs at varying stages of development [2]: eighteen land-based water-cooled, seven marine-based water-cooled, nine high- temperature gas-cooled, six fast neutron spectrum, and eight molten salt. Some of these designs incorporate a number of novel safety features. Appendix A reviews some of these designs and features. Of the forty-eight SMRs discussed, two Russian KLT-40S 35 MW floating reactors were launched in August 20191, the Indian IPHWR 200 have been in operation for many years, and the Chinese HTR-PM is currently under construction.

With the understanding that existing regulatory frameworks and licensing processes are largely devised for conventional power reactors [5], efforts are undergoing to revisit these regulations and examine their suitability for the emerging SMR designs, during normal operation and more importantly in the event of accidents. In the Canadian context, the Canadian Nuclear Safety Commission (CNSC) has released a discussion paper on the licensing of SMRs, indicating that existing regulations are mostly suited for the SMR technologies, and that the licensing process assesses risk regardless of the specific reactor technology or size [5]. However, the CNSC’s discussion paper also acknowledges that while “the licensing process is risk-informed and independent of reactor technology or size, [the] CNSC is interested in understanding where enhancements can be made” [5]. Below is a summary of the regulatory process in

Canada.

1 Russia launches 'floating Chernobyl' plant across Arctic, https://www.cnn.com/2019/08/23/europe/russia-arctic-floating-nuclear-power-station-launch- intl/index.html, accessed January 19, 2020.

1.2 Canadian Regulations

Licenses for nuclear facilities in Canada are granted by the Canadian Nuclear

Safety Commission (CNSC). The regulatory document to support an application for the licensing of SMR facilities is referred to as REGDOC-1.1.5 [6]. It addresses the licensing process for selecting and preparing a site, constructing a reactor, and operating it.

Additional regulatory documents for use in conjunction with REGDOC-1.1.5, include

REGDOC-1.1.1 [7], RD/GD-369 [8], and REGDOC-1.1.3 [9]. REGDOC-1.1.1 establishes requirements for site evaluation and preparation of new reactor facilities [7].

RD/GD-369 identifies the specific information required to support an application for construction [8]. REGDOC-1.1.3 sets out the requirements for an application to obtain a license to operate [9]. These three documents pertain to all reactor facilities and not

SMRs specifically. As such, REGDOC-1.1.5 is the focus of this section.

The CNSC’s approach to regulation is technology-neutral, meaning that applicants are required to make a compelling case in order to show that their design meets the intent of the requirements [6]. This is also known as a graded approach, which in part states that the level of justification provided by the applicant is commensurate with the potential risks to health, safety, security, and environment [6]. The CNSC determines if the potential risks are adequately addressed by the applicant when considering regulatory requirements, regulatory information, third-party research,

Indigenous perspectives, stakeholders, and supporting documentation provided by the applicant. The CNSC’s graded approach creates a critical component of the licensing process for SMRs. Alternative approaches to regulation are also permitted by the CNSC, if the approach will result in an equivalent level of safety [6].

The content of a license application to the CNSC addresses safety and control areas (SCAs) [6]. REGDOC-1.1.5 references fourteen SCAs: management systems, human performance management, operating performance, safety analysis, physical design, fitness for service, radiation protection, conventional health and safety, environmental protection, emergency management and fire protection, waste management, security, safeguards and non-proliferation, and packaging and transport [6].

Many of the SCAs are concerned with provisions in the design that reduce risk to the environment, community, and individual. Management systems ensure that the organization’s approach to meeting safety objectives are established in company processes and programs. Human performance management ensures that an adequate number of people with the skills, knowledge, tools, and procedures are available to carry out the required responsibilities [6]. Operating performance includes a review of license activity conduct to ensure effective performance of the reactor [6]. When documenting management systems, human performance management, and operating performance, the applicant is to directly address the “number and type of physical, engineered, or administrative barriers” [6]. This plays an important role in determining the probability for a release of radioactive materials to the environment. Safety analysis is the evaluation of potential hazards against “the conduct of a proposed activity or facility and considers the effectiveness of preventative measures and strategies in reducing the effects” [6].

Physical design refers to the ability of components, systems, and structures to maintain and meet the intended design parameters [6]. Radiation protection ensures that proper monitoring and controls are employed to maintain adequate protection that aligns with the relevant radiation protection regulations [6]. Safety analysis, physical design, and

4 radiation protection are directly concerned with SMR failure probability and the resulting consequences and health considerations. One regularity feature that may be revisited is the size of the exclusion zone, also called the emergency planning zone, surrounding a nuclear power plant, which is discussed below.

1.3 Exclusion Zone

The Canadian Nuclear Safety Commission (CNSC) established that license applications, other than a license to abandon, shall contain provisions for an exclusion zone [10]. Class I Nuclear Facilities Regulations, an associated regulation of the Nuclear

Safety and Control Act (NSCF) [11], describe an exclusion zone as “a parcel of land within or surrounding a nuclear facility on which there is no permanent dwelling and over which a licensee has the legal authority to exercise control”. The Canadian perspective on exclusion zones is consistent with both the American and International perspectives. The

United States Nuclear Regulatory Commission (USNRC) defines an exclusion area as

“the area surrounding the reactor where the reactor licensee has the authority to determine all activities, including exclusion or removal of personnel and property” [12].

Similarly, the International Atomic Energy Agency (IAEA) describes the exclusion zone as “the area around the reactor that is controlled by the operating organization. In this area the operating organization has the full power to implement all necessary measures”

[13].

When determining an adequate size for the exclusion zone, there must be consideration of the evacuation needs, land usage needs, security requirements, and environmental factors [10]. Rather than prescribing the size of the exclusion zone, the designer must demonstrate to the regulator that the distance they are proposing is

5 acceptable and addresses the above factors. Evacuation needs refers to emergency response requirements. Land usage accounts for any potential future expansion of the reactor facilities. Security requirements considers threat assessment, external hazards, and available resources in the event of an incident. While evacuation needs, land usage needs, and security requirements are undoubtedly crucial to determining the size of an exclusion zone, they are not within the scope of this thesis as they do not impact the dose received by an individual at the boundary of the exclusion zone. The fourth, environmental factors, has a direct influence on the dose received by an individual at the boundary of the exclusion zone. Environmental factors, such as meteorological conditions and terrain, are to be carefully considered when determining the size of an exclusion zone.

The exclusion zone is an important component of defence in depth barriers, which are intended to mitigate radiological consequences following a postulated release of radioactivity, and ensures that regulated dose limits are not exceeded for members of the general public [10]. The Canadian Radiation Protection Regulations state that an individual at the boundary of the exclusion zone shall not receive an effective dose greater than 1 mSv over a one year period during normal operating conditions [14]. The effective dose limit for an anticipated operational occurrence (AOO) and design basis accident (DBA) shall be 0.5 mSv and 20 mSv, respectively, over a 30-day period following the analyzed event [10].

Prior to the current regulatory approach, Canadian nuclear power plant exclusion zones were defined as 914 meters (1000 yards) from the reactor building [10]. This was a conservative approach that accounted for uncertainties within the nuclear industry at the time [15]. As the industry’s ability to use atmospheric dispersion modeling to analyze

6 postulated severe accidents grew, the regulator allowed license applicants to propose alternative solutions [10]. Existing regulatory requirements for an exclusion zone are applied to large nuclear power plants. The objective of this thesis is to revisit the existing regulatory requirements of an exclusion zone in view of SMR’s reduced power output, and the likelihood that some of them may be installed in locations where the traditional one-kilometre exclusion zone are not practical; e.g. when used for district heating of a building complex or for producing process heat for an industrial site.

1.4 Thesis Objectives and Outline

The primary objective of this thesis is to addresses SMR exclusion zone sizing.

This is to be achieved in part through the development of a simple method for determining the whole-body dose from a hypothetical severe SMR accident at various distances and power levels.

This thesis also examines how the size of the exclusion zone is influenced by technical regulations and standards, reactor design and safety features, and by the presence of engineered barriers.

This thesis began by examining in Section 1.2, existing technical regulations and standards relating to dose limits and the reactor exclusion zone, the goal of which is to dilute any radioactive releases prior to reaching the public. A review of SMR designs currently in development and a survey of their unique safety features are reported in

Appendix A. A method for determining the whole-body dose from a hypothetical severe reactor accident at various distances and power levels is introduced in Chapter 2. The method first estimates the radioactivity (Source Term) of released radionuclides. Source

Term refers to the quantity and type of radioactive materials available for release to the environment following a nuclear accident [16]. Three levels of verification for the proposed method are discussed. In Chapter 3, the dispersion in a plume from a reactor following a postulated accident is modeled to estimate the radiation dose associated with the accident. The obtained results are verified against published data. Following independent verification of the empirical approach for Source Term estimation and dispersion modeling, the two methods are combined and verified against those of more complex models. The methodology is then applied to estimate the exclusion zone for a variety of SMRs. Results of the study are compared to existing technical regulations to ensure that the dose permitted at the boundary of the exclusion zone is not exceeded. Following review of the developed method for determining whole-body dose, innovations in SMRs are discussed in Chapter 4, including their inherent/passive safety systems, developments in reactor materials, and developments in engineered barriers, to determine how these features can further reduce the size of the exclusion zone or even eliminate the need for it. Chapter 5 provides the conclusions and recommendations for this thesis.

CHAPTER 2: SOURCE TERM

This chapter is adapted from two conference papers published in the proceedings of the 38th Annual Conference of the Canadian Nuclear Society [17] and of the 42nd

Annual CNS/CNA Student Conference [18].

2.1 Introduction

The Source Term refers to the quantity and type of radioactive materials available for release to the environment following a nuclear accident [16]. The released radioactivity is dispersed into the atmosphere, determining the radiation impact of the accident. The Source Term is often estimated using comprehensive computer codes, such as the Modular Accident Analysis Program (MAAP4-CANDU), which simulates the radiological consequences of postulated severe accident sequences [19]. The Canadian

Nuclear Safety Commission (CNSC) employed MAAP4-CANDU to approximate the

Source Term in a study that addressed a hypothetical nuclear reactor accident [20] in an

878 MWe (2809 MWth) CANDU unit. This chapter presents a simplified approach for

Source Term approximation, utilizing parameters such as: fission product yields, fuel composition, radionuclide release fractions, and reactor thermal power. The simplicity of the model enables readily calculating the Source Term for the parametric study employed in this work to determine the size of the exclusion zone.

In this chapter, the simplified empirical approach is presented and verified against

CNSC published data to confirm that the method is capable of accurately approximating the Source Term. The verified model is used for approximating a Small Modular

Reactor’s (SMR) Source Term following a hypothetical accident in Chapter 3.

2.2 Source Term

The Source Term is measured in becquerels (Bq), representing the inventory of radionuclides available for release into the environment. Section 50.2 of Title 10 of the

United States Code of Federal Regulations (10 CFR) [16] defines the Source Term as:

“…the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release.” The Source Term approximation must consider the fission product yield, the fractions of radionuclide release, airborne fractions, and the reactor thermal power (i.e. the reactor’s rate of heat generation).

Fission product yields depend on the fuel type, composition, and degree of burnup, since these factors determine the type and amount of fissionable materials present and each isotope has its own characteristic fission yield [21]. However, for the purpose of estimating the Source Term, these factors are not as important as is devising a refueling scheme, since the fission yields for all fissionable materials are reasonably close in value to each other [21].

Following a nuclear reactor accident, only a fraction of the inventory of the radioactive materials are released to the atmosphere [19]. The remaining portion might have decayed or remained within the confines of the reactor due to physical (defense in depth) barriers [10]. Table 2-1 lists the fission product groups, the number of radionuclides within each group, and their corresponding release fractions. This information can be obtained using the United States Department of Energy’s MACCS2 code [22]. A release fraction is the fraction of the total Source Term, for a particular

10 radionuclide, released to the atmosphere following a hypothetical postulated severe nuclear reactor accident.

Table 2-1: Fission product groups [20]

Fission Product Group Number of radionuclides Release Fraction Noble Gases 6 4.12×10-1 Halogens 5 1.52×10-3 Alkali Metals 4 1.52×10-3 Alkaline Earths 6 2.30×10-8 Refractory Metals 8 2.53×10-4 Lanthanides 12 8.51×10-9

2.3 Approximation

An empirical approximation was used in this thesis for simplicity. It is based on the basic expression given by Lamarsh [21]:

(2.1) where C0 is the Source Term measured in Bq, P is the reactor thermal power (MWth), i refers to a one of n radionuclides, Yi is fission product yield for i, Fpi is the fraction released for of i, and Fbi is the fraction that remains airborne. The constant of

3.1302x1016 is the total number of fissions occurring per section in the reactor when multiplied with reactor thermal power. Equation (2.1) ignores radioactive decay; overestimating the prompt fission products and underestimating the delayed ones produced by decay.

The radionuclide fission yields in Equation (2.1) depend on the fuel composition, which is design-dependent and changes with time due to burnup.

However, the difference in the fission yields for fissionable isotopes does not significantly differ from each other, and their accurate values are most significant when devising a refueling scheme. For Source Term calculation, the composition of an equilibrium core would be a good approximation. The Live Chart of Nuclides [23] is used in this thesis for calculating the fission yield. The radionuclides in Equation (2.1) were those identified in Table 2-1. These were for a typical CANDU reactor’s equilibrium core, but given the assumption that the fission products and their yields do not vary much from one fissionable isotope to another, this is a reasonable assumption.

Table 2-1 also provides the release factions for the fission products. The airborne fractions were initially assumed to be all equal to unity, i.e., all radionuclides were considered airborne. This is not a realistic assumption and is corrected for in the verification process of the dispersion model in Chapter 3.

2.4 Source Term Verification

For verification, the results provided by Equation (2.1) were compared against the

CNSC study [20], which modeled a large release of radionuclides to the atmosphere following a nuclear reactor accident for an 878 MWe (2809 MWth) CANDU unit. A large release is defined as the release of Cesium-137 in excess of 1.0×1014 Bq throughout the accident duration [10]. More than 40 radionuclides, in addition to Cesium-137, were considered in the CNSC study. The CNSC study employed MAAP4-CANDU to determine the Source Term, including the radionuclides listed in Table 2-1 [20]. The composition of the equilibrium core considered for verification consisted of four primary

12 nuclides: 5.65% of Uranium-238, 49.19% of Uranium-235, 43.64% of Plutonium-239, and 1.32% of Plutonium-241, representing an equilibrium CANDU core [24].

The total Source Term determined by applying the empirical approach was

6.97×1018 Bq, while the total Source Term listed in the CNSC published data is 4.50×1018

Bq. The two approaches are comparable in magnitude. The listed radionuclides in Table

2-2 were the ones deemed significant by the CNSC. Significance was likely determined by the greatest fission product yields, release fractions, airborne fractions, or radiological impact. In total, more than 40 radionuclides were considered for this thesis. The discrepancy between the published and simulated Source Term values in Table 2-2 can be attributed to assigning unity airborne fractions in the empirical model and ignoring radioactive decay. A calibration factor can be used to correct for the overestimation, as discussed further in Chapter 3. The underestimated Source Terms were for radionuclides with low values, and therefore their effect on the overall Source Term is negligible. The empirical method overestimated the overall Source Term value.

Table 2-2: Published and simulated source term for significant radionuclides

Radionuclide Published Source Empirical Source Release Fission Product Term (Bq) [20] Term (Bq) Fractions [20] Yield (%) [23] Ba-140 8.14×1011 1.18×1011 2.30×10-8 5.84186 Cs-134 3.21×1013 4.02×1011 1.52×10-3 0.00030 Cs-137 1.02×1014 8.50×1015 1.52×10-3 6.35827 Ce-141 2.40×1011 4.16×1010 8.51×10-9 5.55156 Ce-144 8.17×1010 3.48×1010 8.51×10-9 4.64968 I-131 3.93×1015 4.37×1015 1.52×10-3 3.26932 I-132 5.80×1011 6.33×1015 1.52×10-3 4.73788 I-133 2.79×1015 9.03×1015 1.52×10-3 6.75244 I-135 2.50×1014 8.50×1015 1.52×10-3 6.35500 Ru-103 1.00×1015 1.11×1015 2.53×10-4 4.98560 Ru-106 1.14×1014 5.01×1014 2.53×10-4 2.25011 Xe-133 1.99×1018 2.45×1018 4.12×10-1 6.76383 Total 4.50×1018 6.97×1018

As Table 2-2 shows, the total source term for the empirical approach is of the same order of magnitude as the published data and can be corrected using a calibration factor, as discussed in Chapter 3. This demonstrates that the simplified empirical approach is capable of reasonably approximating the Source Term.

2.5 Sensitivity Analysis

The sensitivity of the model of Equation (2.1) to changes is analyzed via the four

(4) variables that define the Source Term. The first variable is the reactor thermal power,

P, which is the driving force of the Source Term, i.e. the higher the power the larger the

Source Term. However, the Source Term should be evaluated at the nominal (maximum) reactor power, not a lower power used during the reactor operation. The second variable in Equation (2.1) is the airborne fractions, which were initially given the highest possible value of unity. This is not a realistic assumption and is corrected for as discussed in

Chapter 3. The other two factors are the release fractions of radionuclides, Fpi, and radionuclide fission yields, Yi. the effect of which on the empirical model of Equation

(2.1) is discussed below

The release fractions listed in Table 2-1 were obtained using the United States

Department of Energy’s MACCS2 code [22]. These values assume “total radionuclide release” and are used to “quantify the emissions from a hypothetical severe nuclear accident, in order to consider the implementation of emergency planning and to subsequently asses the human health and environmental consequences” [20]. These release fractions do not consider inherent/passive and active safety systems and additional barriers that can prevent radionuclides from being released to the enviroment.

The effect of these safety measures is discussed in greater detail in Chapter 4. Forsake of demonstrating the impact of reducing the release factors, let us assume that the release fractions listed in Table 2-1 can be reduced by 25%, as shown in Table 2-3.

Table 2-3: Simulated source term for release fractions reduced by 25%

Radionuclide Published Original Empirical Reduced Reduced Empirical Release Source Term Release Source Term Fractions [20] (Bq) Fractions [20] (Bq) Ba-140 2.30×10-8 1.18×1011 1.73×10-8 8.89×1010 Cs-134 1.52×10-3 4.02×1011 1.14×10-3 3.01×1011 Cs-137 1.52×10-3 8.50×1015 1.14×10-3 6.38×1015 Ce-141 8.51×10-9 4.16×1010 6.38×10-9 3.12×1010 Ce-144 8.51×10-9 3.48×1010 6.38×10-9 2.61×1010 I-131 1.52×10-3 4.37×1015 1.14×10-3 3.28×1015 I-132 1.52×10-3 6.33×1015 1.14×10-3 4.75×1015 I-133 1.52×10-3 9.03×1015 1.14×10-3 6.77×1015 I-135 1.52×10-2 8.50×1015 1.14×10-3 6.37×1015 Ru-103 2.53×10-4 1.11×1015 1.90×10-4 8.33×1014 Ru-106 2.53×10-4 5.01×1014 1.90×10-4 3.76×1014 Xe-133 4.12×10-1 2.45×1018 3.09×10-1 1.84×1018 Total 6.97×1018 5.23×1018

The 25% reduction in the release factors, as shown in Table 2-3, results in an overall lowering of the Source Term from 6.97×1018 Bq to 5.23E×1018 Bq. This reflects the expected linear and direct effect of the release fractions on the Source Term as expressed by Equation (2.1).

As indicated in Section 2.2, fission yields depend on the fuel composition, which is design-dependent and changes with time due to burnup. The cumulative fission yield, i.e. the total number of atoms produced over time after one fission, is provided not as a static number, but instead as a range. The results of Table 2-2 were completed using the mean cumulative fission yields from the IAEA’s Live Chart of Nuclides [23]. Table 2-4 lists the same mean values shown in Table 2-2, but also the low-range and high-range values, which were found using the IAEA Live Chart [23]. As can be seen, there is a

16 slight variance between the fission product yields for a number of significant radionuclides.

Table 2-4: Fission product yield range for significant radionuclides [23]

Radionuclide Fission Product Fission Product Fission Product Yield (%) Yield (%) Yield (%) (Low) (Mean) (High) Ba-140 5.76355 5.84186 5.92018 Cs-134 0.00022 0.00030 0.00038 Cs-137 6.27869 6.35827 6.43784 Ce-141 5.42073 5.55156 5.68240 Ce-144 4.60191 4.64968 4.69745 I-131 3.21395 3.26932 3.32468 I-132 4.66450 4.73788 4.81126 I-133 6.63057 6.75244 6.87430 I-135 6.12693 6.35500 6.58307 Ru-103 4.89832 4.98560 5.07287 Ru-106 2.18935 2.25011 2.31087 Xe-133 6.64179 6.76383 6.88588

Table 2-5 was calculated by altering the fission production yield, with the power levels and release fractions remaining the same as in Table 2-2. Significant radionuclides are listed in Table 2-4, but the totals represent all 40 radionuclides that were considered.

As can be seen, the total Source Term values change by approximately ±1.77×1017 Bq, which is equivalent to ± 2.5%. This indicates that the model is not very sensitive to the uncertainties in the fission yields. As discussed earlier, the fission yields for all fissionable materials are reasonably close to each other, as shown in Chapter 3 of

Lamarsh [21].

Table 2-5: Simulated source term range for significant radionuclides

Radionuclide Empirical Source Empirical Source Empirical Source Term (Bq) Term (Bq) Term (Bq) (Low) (Mean) (High) Ba-140 1.17×1011 1.18×1011 1.20×1011 Cs-134 2.96×1011 4.02×1011 5.07×1011 Cs-137 8.39×1015 8.50×1015 8.61×1015 Ce-141 4.06×1010 4.16×1010 4.25×1010 Ce-144 3.44×1010 3.48×1010 3.52×1010 I-131 4.30×1015 4.37×1015 4.45×1015 I-132 6.24×1015 6.33×1015 6.43×1015 I-133 8.86×1015 9.03×1015 9.19×1015 I-135 8.19×1015 8.50×1015 8.80×1015 Ru-103 1.09×1015 1.11×1015 1.13×1015 Ru-106 4.87×1014 5.01×1014 5.14×1014 Xe-133 2.41×1018 2.45×1018 2.50×1018 Total 6.79×1018 6.97×1018 7.15×1018

It should be emphasized that if different parameters are used in the model of

Equation (2.1), the verification process of Section 2.5 should be repeated.

2.6 Conclusions

This chapter demonstrated that a simplified empirical approach to determining reactor Source Term following a hypothetical postulated nuclear reactor accident provided adequate overall values, after calibrating against independent data obtained using more sophisticated modelling. Sensitivity analysis showed that as expected, the model is linearly sensitive to its variables, which facilitates the calibration process.

Therefore, this simplified empirical approach is used to approximate the Source Term in the remainder of this work. The validity of this simple model was enabled through calibration with a more complex model, as discussed in Section 3.3

CHAPTER 3: RADIOACTIVITY DISPERSION AND EXCLUSION

ZONE

This chapter is adapted from two conference papers published in the proceedings of the 38th Annual Conference of the Canadian Nuclear Society [17] and of the 42nd

Annual CNS/CNA Student Conference [25].

3.1 Introduction

Atmospheric dispersion modeling assists with the analysis of radionuclide releases following a postulated nuclear reactor accident. Exclusion zone sizing for nuclear reactors is enabled through dispersion modeling and indicates that radioactive materials are diluted to levels which do not exceed regulatory limits [20]. The dose received by an individual at varying distances from the reactor is typically determined with detailed simulations, using codes such as the United States of America Department of Energy’s MACCS2 code [22]. A reactor’s Source Term can also be predicted using codes capable of simulating severe accident sequences, such as MAAP4-CANDU [19].

This chapter presents a simplified approach to the above methods that can be applied to a variety of nuclear reactors, and more specifically Small Modular Reactors (SMRs), and enables the quick determination of the size of the exclusion zone for various SMR power levels.

The HotSpot Health Physics Code [26] is a tool which uses a Gaussian distribution for a first-order approximation of the dispersion of radionuclides in air. In this chapter, HotSpot is verified against published data to confirm that the simulation code is capable of accurately estimating radiation dose. The estimated doses were

19 determined as a function of downwind distance from the origin of the accident site and allows for the determination of the radius of the exclusion zone [20]. Following verification, the simulation code is used to approximate the radiation dose received by an individual following a hypothetical postulated severe accident for an SMR. The simulation code is then used to estimate the exclusion zones for generic SMRs at various power levels. The next section provides a description of the HotSpot simulation code.

3.2 HotSpot

The HotSpot simulation code, hereinafter referred to as HotSpot or code, serves as a conservative, but realistic, means for the estimation of the radiation dose associated with the atmospheric release of radionuclides [26]. Development and ongoing maintenance of HotSpot has been funded by the United States Department of Energy since the code was originally distributed in 1988. The initial purpose of the code was to equip emergency personnel with the ability to quickly examine an incident involving the release of radionuclides [26]. Overtime, HotSpot has grown to include a variety of atmospheric dispersion models, which are capable of approximating the doses received by an individual at varying distances from a radionuclide release [26]. The code is intended for near-surface releases, short-term durations, and short-range dispersion [26].

In this thesis, a general plume and normal (Gaussian) distribution were used for the analysis of radioactive materials and their respective impact on the atmosphere. The

Gaussian plume model “is the most widely used computational model for atmospheric diffusion assessment” [27]. HotSpot uses the Gaussian distribution to help determine the

20 time-integrated atmospheric concentration of radionuclides within the areas surrounding the reactor [26]:

(3.1) where is time-integrated atmospheric concentration (Bq-s)(m3), Q is Source Term (Ci),

H is effective release height (m), λ is radioactive decay constant (S-1), x is downwind distance (m), y is crosswind distance (m), z is vertical axis distance (m), sy is standard

deviation of integrated concentration distribution in crosswind direction (m), sz is standard deviation of integrated concentration distribution in vertical direction (m), s is average wind speed at the effective release height (m/s), and DF(x) is plume depletion factor [26].

The HotSpot code uses the time-integrated atmospheric concentration, along with a variety of additional inputs, to determine the dose received by an individual at varying distances from the origin of the accident [26]. This is the total effective dose equivalent

(TEDE) and includes the inhalation, submersion, ground shine, and resuspension components following a hypothetical nuclear reactor accident [26].

The fission products used to assemble the activities for the CNSC study were compiled through the use of MAAP4-CANDU [19], a code capable of simulating severe accident sequences, and REGDOC-2.5.2 [10], a regulatory document which provides designers with a reference point for defining severe accidents for reactor facilities. The

21 more than 40 radionuclides considered within the CNSC study [20] were classified by their similarities as shown in Table 2-1. The fission product groups in the Table allow for the determination of each individual radionuclides’ activity (Bq), which combine to form a reactor’s Source Term.

Although it is known that some radionuclides contribute to radiation dose more than others, each radionuclide associated with the CNSC study was assigned an activity.

A summary of the radionuclides and their respective activities, deemed to be the most critical, are listed in Table 3-6. HotSpot employs standard dose conversion factors

(coefficients) of activities for acute inhalation of radionuclides. The coefficients, described in federal guidelines report #11 (FGR-11) [28], were used in this work to convert our time-integrated atmospheric concentrations into doses.

Table 3-6: Significant Radionuclides [20]

Radionuclide Fission Product Release (Becquerel) Ba-140 8.14×1011 Cs-134 3.21×1013 Cs-137 1.02×1014 Ce-141 2.40×1011 Ce-144 8.17×1010 I-131 3.93×1015 I-132 5.80×1011 I-133 2.79×1015 I-135 2.50×1014 Ru-103 1.00×1015 Ru-106 1.14×1014 Xe-133 1.99×1018

3.3 Verification

The verification process consisted of three levels. The first level, which was discussed in Chapter 2, validated the simplified empirical approach for the source-term approximation against an independent simulation code. The second level validated the

HotSpot dispersion code against published data, using the source term found in the CNSC study [20]. The third level verified the combined use of the HotSpot code and the empirical source term. The latter level is used to find a calibration factor, which aligns the dose determined using the simplified approach with that provided by independent data.

HotSpot was verified against independent data for a hypothetical severe nuclear accident published in the CNSC study [20], which modeled a large release of radionuclides to the atmosphere following a hypothetical failure of a CANDU reactor at the Darlington Nuclear Generating Station. Figure 3-1 shows the verification process, which involves using the published activities, found in the CNSC study [20], as an input to HotSpot and comparing the simulated doses to the published doses, also found in the

CNSC study [20].

Figure 3-1: Procedure for the verification of HotSpot

The CNSC analyzed a variety of different accident conditions, including a 24- hour severe postulated nuclear accident followed by a 24-hour generic large release

(GLR), a 24-hour accident and 1-hour GLR, and a 24-hour accident and 72-hour GLR.

The same verification process discussed in this paper can be applied to any of the three

(3) scenarios in the CNSC study. However, the 24-hour accident and 24-hour GLR was selected as it demonstrates an emergency situation, which is realistic but still conservative. Table 3-7 lists the published dose found in the CNSC study, which corresponds to the published dose block in Figure 3-1.

Table 3-6 lists several of the published activities that were used to estimate the published doses in Table 3-7. Only activities for radionuclides that were deemed critical are listed in Table 3-6. The remaining activities can be found in the CNSC study. The data contained in the published dose column of Table 3-7 was estimated in the CNSC study using MACCS2 [22], a code based on a straight-line Gaussian plume model, similar to that of the HotSpot code. As such, the verification of HotSpot against the

CNSC published data is in fact a verification against MACCS2.

Where possible, the CNSC study was used to assist with defining input variables to ensure consistency. However, additional data was required, such as: effective release heights, damage ratios, wind speed and direction, atmospheric stability ratings, exposure length, weathering correction factor, resuspension factor, surface roughness, et cetera. All additional variable which were not discussed in the CNSC study were determined using

HotSpot default values [26].

The simulated dose column in Table 3-7 lists the dose that was determined using the HotSpot code and corresponds to the simulated dose in Figure 3-1. Similar to the

24 published dose column, the simulated dose column serves as the dose to an individual at varying distances from the origin of the accident. Table 3-7 corresponds to the last step in

Figure 3-1.

The published dose and simulated dose at 1 km from the accident origin were estimated to be approximately equal, as shown in Table 3-7. From 3 to 50 km, the published dose exceeds the simulated dose. Conversely, the simulated doses surpass the published doses between 70 and 90 km. The HotSpot code is intended for short distances

[26], however, these short distances are difficult to verify as the CNSC study does not examine distances less than 1 km. As the distances double in Table 3-7, the simulated dose approximately decreases with the inverse-squared distance, but the published data does not. This implies that the HotSpot code considers the source as a point source.

Nevertheless, the published and simulated results are on the same order of magnitude and are therefore comparable.

Table 3-7: Published dose for 24 hour accident and 24 hour generic large release

Distance from origin Published dose HotSpot Simulated (km) (mSv) [20] dose (mSv) 1 2.54×101 2.6×101 3 4.50×100 3.2×100 6 1.75×100 8.8×10-1 12 6.70×10-1 2.6×10-1 20 3.10×10-1 1.1×10-1 28 1.80×10-1 8.5×10-2 36 1.30×10-1 7.2×10-2 50 7.00×10-2 5.9×10-2 70 4.00×10-2 4.8×10-2 90 3.00×10-2 4.1×10-2

The empirical approach to calculate Source Term and the atmospheric dispersion model were verified independent of each other. The third level of verification combines both processes. Table 3-8 lists the published [20] and simulated doses, with the source term calculated using Equation (2.1). The two values are not significantly different, demonstrating the validity of the methods of Chapters 2 and 3. However, the simulated doses were adjusted by an “overall calibration” factor, determined by dividing the published dose by the simulated dose. This factor accounts for ignoring radioactive decay and using airborne fractions of unity in Equation (2.1).

Table 3-8: Published doses and HotSpot-simulated doses using Equation (2.1)

Distance from origin Published dose HotSpot Simulated (km) (mSv) [20] dose (mSv) 1 2.54×101 6.10×101 3 4.50×100 7.30×100 6 1.75×100 2.00×100 12 6.70×10-1 5.90×10-1 20 3.10×10-1 2.50×10-1 28 1.80×10-1 2.00×10-1 36 1.30×10-1 1.70×10-1 50 7.00×10-2 1.40×10-1 70 4.00×10-2 1.10×10-1 90 3.00×10-2 9.50×10-2

The exclusion zone for a CANDU reactor is typically equal to one km [2]. The exclusion zone radius for SMRs is expected to be smaller, given their lower power.

Therefore, the smallest distance in Table 3-8 (1 km) was used to determine the calibration factor. This resulted in a calibration factor of 0.42 (=25.4/61.0), which is used to adjust the overestimated simulated source term in subsequent calculations. It should be noted

26 though that the doses reported in the CNSC study [20] are for a hypothetical severe accident. Table 3-9 represents the adjusted doses using the 0.42 calibration factor.

Table 3-9: Published doses and HotSpot-simulated doses using calibrated Equation (2.1)

Distance from origin Published dose Simulated dose (km) (mSv) [20] (mSv) 1 2.54×101 2.6×101 3 4.50×100 3.1×100 6 1.75×100 8.5×10-1 12 6.70×10-1 2.5×10-1 20 3.10×10-1 1.0×10-1 28 1.80×10-1 8.3×10-2 36 1.30×10-1 7.0×10-2 50 7.00×10-2 5.7×10-2 70 4.00×10-2 4.7×10-2 90 3.00×10-2 4.0×10-2

Results from the three levels of verification discussed in Chapters 2 and 3 are summarized in Figure 3-2. The first level of verification, discussed in Chapter 2, concluded that the simplified empirical approach to determining reactor Source Term following a hypothetical postulated nuclear reactor accident provided adequate overall values. The second level of verification concluded that the HotSpot code is able to approximate dose limits comparable to those of the published doses. The third and final level of verification provided a calibrated value for the airborne fraction when employing the empirical approach for Source Term calculation, which was initially assumed to be unity.

Figure 3-2: Three levels of verification

3.4 Exclusion Zone for SMRs

The calibrated method presented above was applied to various SMRs with powers ranging from 50 to 300 MWe, adjusted to thermal power using a factor of 3.2, reflecting a typical efficiency of 30% converting heat to electricity. Table 3-10 lists the doses for all

SMRs considered in this study at distances ranging from 0.1 km to 1.0 km. These values were calculated over a 30-day period to match the CNSC dose limit of 20 mSv for a design basis accident [10]. The exclusion zone radius for each power level is the distance at which the dose is equal to or slightly below the 20 mSv limit. These values are highlighted (boldfaced) in Table 3-10. As can be seen, the radius of the exclusion zone for an SMR is estimated to vary from less than 0.4 km for a 50 MWe reactor to 0.8 km for a 300 MWe reactor. If a reactor is buried underground or encased in a thick concrete container, the exclusion zone can be further reduced. Chapter 4 discusses the effect of such an enclosure on the size of the exclusion zone. It should be noted though that the

28 doses reported in the CNSC study [20] are for a hypothetical severe accident, while the size of the exclusion zone is determined for a design basis accident (DBA). As such the doses reported in Table 3-10 are overestimated for a DBA.

Table 3-10: SMR dose at a variety of distances

Distance from 50 MWe 100 MWe 150 MWe 200 MWe 250 MWe 300 MWe origin (km) 30-day Dose (mSv) 0.1 1.8×102 3.6×102 5.4×102 7.2×102 9.0×102 1.1×103 0.2 5.0×101 1.0×102 1.5×102 2.0×102 2.5×102 3.0×102 0.3 2.3×101 4.5×101 6.8×101 9.1×101 1.1×102 1.4×102 0.4 1.3×101 2.6×101 3.9×101 5.1×101 6.4×101 7.7×101 0.5 8.3×100 1.7×101 2.5×101 3.3×101 4.1×101 5.0×101 0.6 5.8×100 1.2×101 1.7×101 2.3×101 2.9×101 3.5×101 0.7 4.3×100 8.5×100 1.3×101 1.7×101 2.1×101 2.6×101 0.8 3.3×100 6.6×100 9.8×100 1.3×101 1.6×101 2.0×101 0.9 2.6×100 5.2×100 7.8×100 1.0×101 1.3×101 1.6×101 1.0 2.1×100 4.2×100 6.4×100 8.5×100 1.1×101 1.3×101

3.5 Sensitivity Analysis

The sensitivity of the HotSpot code to changes is analyzed via three (3) variables that were determined using HotSpot’s default values. As discussed previously, the CNSC study was used to define most input variables for the HotSpot simulations. However, additional variables were found using HotSpot recommendations, including effective release heights, wind speed, and atmospheric stability.

The first variable is the effective release heights. The rise of radioactive plumes is impacted by velocity and temperature differential between the surrounding air and stack effluent [26]. As the effective release height increases, the integrated concentrations at the ground level decrease [26]. The value was assumed to be 10 meters for the purposes

29 of this thesis. For the sake of demonstrating the impact of reducing/increasing the effective release hieght, let us assume that the original value of 10 meters was reduced to either 0 meters (ground-level) or increased to 20 meters. Using the simulated dose limits shown in Table 3-9 as a reference, Table 3-11 lists simulated dose limits using 0, 10, and

20 meter effective release heights.

Table 3-11: Sensitivity to effective release height

Distance from origin Simulated dose Simulated dose Simulated dose (km) 0-meter effective 10-meter effective 20-meter effective release height release height release height (mSv) (mSv) (mSv) 1 2.8×101 2.6×101 2.4×101 3 3.4×100 3.1×100 2.9×100 6 9.4×10-1 8.5×10-1 8.1×10-1 12 2.7×10-1 2.5×10-1 2.4×10-1 20 1.1×10-1 1.0×10-1 9.9×10-1 28 9.2×10-2 8.3×10-2 7.9×10-2 36 7.8×10-2 7.0×10-2 6.7×10-2 50 6.3×10-2 5.7×10-2 5.5×10-2 70 5.2×10-2 4.7×10-2 4.5×10-2 90 4.4×10-2 4.0×10-2 3.8×10-2

At 1 km, for example, the simulated dose varies approximately ±2.0 mSv (± 8%) when using 0 meters and 20 meters as the effective release height rather than 10 meters.

This indicates that the model is moderately sensitive to change as the effective release height is altered.

The second variable is wind speed. Wind speed was selected to be 3.0 meters per second throughout this thesis. To demonstrate how increased/decreased wind speed impacts the simulated dose, let us assume that the value of 3.0 meters was decreased to

0.1 meters and increased to 6 meters. Using the simulated dose limits shown in Table 3-9

30 as a reference, Table 3-12 lists simulated dose limits using 0.1, 3 and 6 meters per second.

Table 3-12: Sensitivity to wind speed

Distance from origin Simulated dose Simulated dose Simulated dose (km) 0.1 meters per 3 meters per 6 meters per second second second (mSv) (mSv) (mSv) 1 5.6×102 2.6×101 1.3×101 3 5.7×101 3.1×100 1.6×100 6 1.4×101 8.5×10-1 4.3×10-1 12 3.6×100 2.5×10-1 1.3×10-1 20 1.3×100 1.0×10-1 5.3×10-2 28 1.0×100 8.3×10-2 4.2×10-2 36 8.2×10-1 7.0×10-2 3.6×10-2 50 6.3×10-1 5.7×10-2 2.9×10-2 70 4.8×10-1 4.7×10-2 2.4×10-2 90 4.0×10-1 4.0×10-2 2.1×10-2

As shown, varying wind speed can have a significant impact on doses. In general, meterological conditions have a significant impact on the model. While meterological conditions were not a primary focus of this thesis, average values were selected to represent normal conditions. The model was then calibrated using published CNSC data.

The third variable is atmospheric stability. Hotspot interprets atmospheric stability using a matrix that considers various meterological conditions, including wind and solar

[26]. These categories from least severe to most severe are: moderately stable, slightly stable, neutral, slightly unstable, moderately unstable and extremely unstable. For the purposes of this thesis, extremely unstable was used, which correlates to a wind speed of

3 meters per second when the sun is high in the sky. To demonstrate how this impacts the simulated dose, let us assume that slightly unstable was instead selected, which also

31 correlates to a wind speed of 3 meters per second, but when the sun is low in the sky.

Table 3-13 lists the simulated dose limites for both extremely unstable and slightly unstable.

Table 3-13: Sensitivity to atmospheric stability

Distance from origin Simulated dose Simulated dose (km) Extremely Slightly unstable unstable (mSv) (mSv) 1 2.6×101 1.4×102 3 3.1×100 1.9×101 6 8.5×10-1 6.2×100 12 2.5×10-1 2.2×100 20 1.0×10-1 1.1×100 28 8.3×10-2 7.2×10-1 36 7.0×10-2 5.3×10-1 50 5.7×10-2 3.6×10-1 70 4.7×10-2 2.4×10-1 90 4.0×10-2 1.8×10-1

As shown in Table 3-13, similar to varying wind speed, varying atmospheric stability can have a reasonably significant impact on dose limits. As atmospheric stability increases, so does the “intensity of turbulence, and subseqently, the diffusion process”

[29]. To compensate for the effects of the atmospheric stability parameter, the model was calibrated in section 3.3 against published data for a hypothetical severe nuclear reactor accident.

There are several other variables the HotSpot dose calculations are sensitive to, including damage ratios (fraction of the Source Term that is actually impacted in the release scenario [26]), weathering correction factors (dose rate reduction as a function of time after surface contamination [30]), resuspension factors (ratio of the air concentration

32 of radionuclides to the ground concentration [26]), surface roughness (“the surface roughness height is approximately equal to the physical height divided by 10, e.g., a surface roughness height of 3 cm would be associated with a field of objects with an average physical height of 30 cm” [26]), et cetera. Where possible, the CNSC study was used to specify HotSpot input variables to ensure consistency. For the remaining variables, such as surface roughness, HotSpot recommended values were used [26]. If input values deviate from the ones employed in the validation process of section 3.3, the code should be re-calibrated to correct for the effect of the selected parameters.

3.6 Conclusions

The results of this Chapter demonstrate that HotSpot is able to approximate the exclusion zone and dose following postulated severe accidents for nuclear reactors.

Independent data was used to verify that the HotSpot code is capable of atmospheric dispersion modeling, compared to codes such as MACCS2. The method allows for a quick estimation of the exclusion zone radius for Small Modular Reactors (SMRs). In the next chapter, methods to further reduce the exclusion zone radius for SMRs are explored.

CHAPTER 4: REDUCING EXCLUSION ZONE THROUGH DESIGN

4.1 Introduction

In Chapters 2 and 3, a method was developed to determine the exclusion zone as a function of reactor power, under conservative assumptions. It was shown, as one would expect, that the size of the exclusion zone decreases with reactor power, which makes the size of the exclusion zone smaller than that required for conventional larger reactors.

Given that SMRs may be built nearby populated areas, e.g. to provide power off-grid in an isolated community, the question arises: is it possible to reduce or eliminate exclusion zones for SMRs?

A recent discussion paper [5] by the Canadian Nuclear Safety Commission

(CNSC) seems to indicate that the regulators are open to the idea of significantly reducing the size of the exclusion zone. Referring to exclusion zones as Emergency planning zones (EPZ), it states [5]:

"There are no legislative or regulatory requirements for EPZ sizing in Canada and

therefore no restrictions currently in place on minimum EPZ size. EPZ and other

planning actions should be undertaken in relation to the risks associated with the

specific technology. As such, results from safety analyses (i.e., the probabilistic

safety analysis) in combination with the protection strategy used by offsite

planners will determine the EPZ size. This is consistent with the overall

methodologies documented by the IAEA."

Given the above regulatory position and the inherent and passive safety of SMRs, which allows for decreased likelihood of a release of radionuclides, this chapter examines

34 the possibility of reducing or eliminating the exclusion zone for SMRs, taking advantage of their inherent safe nature. The chapter starts by re-examining the concept of the exclusion zone, including how different regulatory jurisdictions view the exclusion zone and how its radius is determined. Examples of inherent and passive safety features in

SMR designs that can assist in reducing the size of the exclusion zone are then given, followed by a discussion of suitable materials that can be used to retain radioactive material within the reactor, further lowering the need for a large exclusion zone. Finally, employing engineered barriers in several designs, in lieu of the exclusion zone, is presented. The chapter concludes by commenting on whether the exclusion zone can be eliminated entirely.

4.2 The Exclusion Zone

In Chapter 2, the quantity of released radionuclides from an SMR following a severe reactor accident was estimated. The results were then used to model the atmospheric dispersion and approximate the required exclusion zone radius in Chapter 3.

Current regulatory and legislative requirements do not prescribe a method for emergency planning zone sizing or a minimum size [5]. Instead, “results from safety analyses in combination with the protection strategy used by offsite planners” are together used to determine the appropriate size for an emergency planning zone [5]. That is, the risk associated with the selected technology is used to determine an appropriate size [5]. The safety analysis involves the full range of potential accidents, as well as their respective probabilities, and the degree of passive and inherent safety features employed within the design [5]. The Canadian description of an exclusion zone is “a parcel of land within or surrounding a nuclear facility on which there is no permanent dwelling and over which a

35 licensee has the legal authority to exercise control” [11]. The American definition is “the area surrounding the reactor where the reactor licensee has the authority to determine all activities, including exclusion or removal of personnel and property” [12]. The

International definition is “the area around the reactor that is controlled by the operating organization. In this area the operating organization has the full power to implement all necessary measures” [13]. The three definitions, from three (3) independent regulatory bodies, do not suggest that the exclusion zone can be eliminated. Instead, each regulatory body provides guidance for exclusion zones that ensures radioactive releases are diluted to a level that does not exceed regulatory limits. For instance, from the Canadian perspective, during normal operating conditions, an individual at the boundary of the exclusion zone shall not receive an effective dose greater than 1 mSv over a one (1) year period [14]. The CNSC also defines acceptable dose criteria for an anticipated operational occurrence (AOO) and design-basis accident (DBA). The AOO and DBA dose limit shall be 0.5 mSv and 20 mSv, respectively, over a 30-day period following the analyzed event. Therefore, one can conclude that current regulations do not suggest that an exclusion zone can be eliminated entirely, but they do allow a licensee to justify a smaller exclusion zone radius. While it is understood that an exclusion zone cannot be eliminated, research reactors built within university campuses do not have exclusion zones, e.g. the McMaster Nuclear Reactor [31]. The McMaster Nuclear Reactor is a 5

MWe research reactor located on the McMaster University campus. The reactor is within a reinforced concrete building and the core is submerged in pool water, which is used to passively cool and provide shielding. These features and the small size of the reactor mean that in the unlikely event of an accident, radioactive materials will be contained.

The CNSC does not define the acceptable dose criteria for a severe nuclear reactor accident for a nuclear power plant (NPP). Therefore, in this thesis, the dose limit for a design basis accident (DBA) was used in lieu of the beyond design basis accident

(BDBA). The DBA is defined as “accident conditions for which an NPP is designed according to established design criteria, and for which damage to the fuel and the release of radioactive materials are kept within regulatory limits” [32]. While the probability of a

DBA remains low, it is still greater than that of a BDBA, which is defined by the CNSC as “accident conditions less frequent and more severe than a design basis accident” [32].

The CNSC does define a BDBA but does not assign an acceptable dose limit. For example, following a DBA, the CNSC has determined that the acceptable dose criteria is

20 mSv over a 30 day period. In the previous chapter, the methodology for approximating

SMR exclusion zones was verified using data from a published CNSC study that modelled the hypothetical failure of a CANDU reactor. The data represented a release following a BDBA and was compared against the regulatory dose limit for a DBA in order to approximate the size of the exclusion zone. While conventional reactors, such as the CANDU reactors, share many similarities with SMRs, there is value in completing a similar study that hypothesizes a DBA for an SMR. Due to the many different varieties of

SMRs, including their fuel compositions, safety systems, and configurations, such studies will inevitably be carried out, but they are not reported in the open literature at the time of writing this thesis.

4.3 Inherent and Passive Safety

Inherent safety refers to “the achievement of safety through the elimination or exclusion of inherent hazards through the fundamental conceptual design choices made

37 for the nuclear plant”, while passive safety means reliance is “placed on natural laws, properties of materials and internally stored energy” [33]. Typically, passive safety refers to the ability of a reactor to dissipate heat via passive heat removal systems that function independent of mechanical means or external input [34], while inherent safety allows a reactor to fail safe by reducing power and decay heat levels without human or computer intervention [35]. There are many different inherent and passive safety systems serving the different varieties of SMRs. Almost all SMR designs claim a negative temperature coefficient of reactivity [2], which reduces the reactor power as temperature increases; lowering the fuel temperature and preventing it from melting and releasing its radioactive content. Several SMRs, such as the NuScale design [2], are surrounded by water pools that serve as a large heat sink to remove heat in the event of an accident. Many SMR designs have an integrated design, in which the primary coolant system or most of its components are installed within the reactor pressure vessel; eliminating the possibility of a large pipe break with the coolant leaving the reactor [2]. For example, in the SMART system, the reactor pressure vessel houses the core, steam generators, coolant pumps, and control rod mechanisms [2]. Several reactors operate at atmospheric pressure and are not subject to depressurization that result in coolant boiling. For instance, the DHR reactor is a pool type reactor designed to operate at a low temperature and at atmospheric pressure, which precludes the undesirable effects of depressurization events that may take place in high-pressure systems [2]. Many SMRs have passive residual heat removal systems that cool the reactor during shutdown, and emergency core cooling systems that inject coolant into the core if needed. For example, the SC-HTGR design employs a passive decay heat removal system that relies on natural circulation and thermal radiation to transfer heat

38 from the reactor core to the surrounding reactor vessel [2]. The Lead-Cooled Fast Reactor

Amphora-Shaped (LFR-AS-200) and the Lead-Cooled Fast Reactor Transportable Long-

Lived (LFR-TL-X) SMRs use heat removal systems consisting of air coolers, which are passively actuated by core thermal expansion as a result of temperature rise [2].

4.4 Reactor Material

Judicious choice of the fuel, cladding, and coolant enables some SMRs to retain radionuclides within the system, lowering the magnitude of the Source Term and in turn the size of the exclusion zone. For example, TRISO (tristructural-isotropic) coated fuel particles are employed by high-temperature gas-cooled reactors (HTGR) [36]. As shown in Figure A-15 in the SMR review section (Appendix A), each TRISO fuel particle is coated with a ceramic layer of silicon carbide, which improves the retention of fission products within the fuel particles [36]. Forming the fuel in small particles also increases the heat transfer (by increasing the surface-to-volume ratio of the fuel), reducing the probability of temperature rise leading to fuel melting and the associated release of fission products [36]. Several HTGR designs, including the Pebble Bed Modular Reactor

(PBMR), employ spherical graphite pebbles, each comprised of thousands of TRISO coated fuel particles [36]. Another barrier against the release of radioactive materials is provided by the fuel cladding, i.e. the sheath in which the fuel is enclosed. Maintaining the integrity of the cladding material during reactor operation is essential to preventing the release of fission products from the fuel, which is most often comprised of corrosion- resistant material, such as zirconium or steel [36]. However, new cladding materials are being developed, e.g., the designers of the energy multiplier module (EM2) reactor are

39 currently developing a silicon carbide cladding material that has a high melting point and meets design criteria for an accident condition [36]. In the integrated modular water reactor (IMR) design, a Zr–Nb alloy is used as a cladding to maintain fuel integrity at a high temperature [2]. In the UNITHERM design, the gap between the fuel-containing matrix and its zirconium cladding is filled with silumin (a high-strength aluminum alloy) to improve the fuel’s resistance to radiation damage [2].

The type of reactor coolant also has an effect on the source term, hence the size of the exclusion zone. For example, a reactor with a coolant that operates at low pressure is more immune to the effect of depressurization-induced accidents. In this regard, the

Sustainable Proliferation-Resistance Enhanced Refined Secure Transportable

Autonomous Reactor (SUPERSTAR) uses lead, a low-pressure coolant, which eliminates the need for the reactor to have a significant pressure retention capability, and reduces the risks associated with release of radioactive materials to the atmosphere following an accident [2]. Coolants that are not subject to phase change, e.g. boiling, are not prone to the adverse effects of such changes. For example, the HTGR employs helium for its coolant, which prevents variations in decay heat removal capacity as it is not prone to phase change [2]. The higher temperature tolerances of helium allows the coolant to operate more effectively at increased temperatures than conventional water-cooled reactors, thus reducing the risks associated with increased fuel temperatures [2]. Coolant phase change can also be avoided by using a coolant with a high boiling point. For example, the Super Safe, Small, and Simple liquid metal cooled reactor (4S-LMR) and the Small, Sealed, Transportable, Autonomous Reactor (SSTAR) and Secure,

Transportable, Autonomous Reactor – Liquid Metal variant (STAR-LM) are sodium

40 cooled and lead-cooled, respectively, both of which have high boiling points, which makes it difficult for reactor over pressurization to occur [2].

The choice of the reactor core structure material can also assist in reducing the

Source Term. Reactors that employ graphite as a structure material, such as the high- temperature gas-cooled reactor pebble-bed module (HTR-PM), High Temperature

Modular Reactor (HTMR-100), and Liquid Fluoride Thorium Reactor (LFTR), have an increased heat capacity, resulting in an improved ability to dissipate heat via conduction and radiation. This assists with maintaining the integrity of the reactor during accidents, retaining the fission products within [2].

In summary, the lower power of SMRs reduces the amount of radioactivity that would be released following a severe accident. Moreover, the associated reduction in the physical size of SMRs enables the introduction of further engineered barriers to the release of radioactivity. Both of these aspects lead to reducing the risk associated with severe reactor accidents, which reduces the need for a large exclusion zone.

4.5 Engineered Features and Barriers

SMR designs incorporate design features that improve reactor safety [37]. Many

SMRs, including the NuScale and mPower, have an integrated design in which all primary reactor systems are contained within a single vessel [37]. This increases the overall size of the pressure vessel, but yields a higher amount of water per unit of power than a conventional power plant, which reduces the rate at which reactor temperature increases and provides operators with more time to respond following a disturbance [37].

The larger reactor vessels per unit power also facilitate improved natural convection

41 cooling, which serves to remove decay heat from the reactor core and vessel [37]. SMRs are able to more effectively manage decay heat as the overall pressure vessel surface area increases considerably in comparison to conventional systems [37]. Further, integrated designs such as the NuScale and mPower are intended to minimize the number of penetrations between the primary (nuclear) side and the secondary (heat exchanger) side.

In the event of a leak, it will be contained within the vessel. Another safety feature is introduced in the International Reactor Innovative and Secure (IRIS) SMR design, in which the relative pressurizer volume increased. The pressurizer serves as an expansion

(surge) tank to accommodate pressure variances in the primary coolant circuit, which provides additional time for operators to respond to reactor disturbances [37].

Some SMR designs are equipped with special barriers to prevent the release of radionuclides to the atmosphere following a loss of coolant accident [2]. For example, the

Advanced Heavy Water Reactor (AHWR) includes a double containment cylindrical concrete structure that is roofed by two (2) concrete domes and is the first SMR design to actively claim that an exclusion zone is not required [2]. Similarly, the LFR-AS-200 and

LFR-TL-X both employ concrete containment structures and, together with additional inherent and passive safety systems, strive to eliminate the need for an emergency preparedness zone [2]. Concrete is one example of an engineered barrier, but water, metal, lead, and soil are all materials that could be used. It is also possible to combine multiple materials together. For example, the VBER-300 has an inner steel containment and an outer reinforced concrete containment structure that can accommodate a beyond design basis accident [2]. Similarly, the CANDU 300 containment structure employs the use of reinforced concrete and a steel liner to improve leak-tightness [38]. An epoxy

42 lining can be used in lieu of a steel liner/containment [33]. Locating a reactor below grade adds an additional barrier, as in the AHWR design. Similarly, water is employed in several reactor designs, e.g. the China Advanced Passive Pressurized Water Reactor

(CAP200) and NuScale [12].

The introduction of engineered barriers allows for a more reliable method of preventing the dispersion of radioactivity that is not influenced by external factors, such as weather conditions. Further, as engineered barriers are added, the exclusion zone may be further reduced. From the perspective of sustainability, there are elements of economic, environmental, and social impact associated with a reduced exclusion zone.

The ability to minimize an exclusion zone frees land for normal uses and development around a reactor. Engineered barriers reduce the quantity of radioactive materials released to the environment following an accident. Reducing the size of the exclusion zone should also increase the public’s comfort with the technology, removing the apprehension around an installation surrounded by a large isolation zone. Minimizing the size of the exclusion zone also allows members of the general public to familiarize themselves with the reactor facility, which may result in them becoming more comfortable with the technology.

4.6 Conclusions

One of the shared design objectives of many SMRs is to employ new technology and innovation in lieu of conventional exclusion zones, or to reduce their size. Passive and inherent safety features, innovations in reactor materials, and engineered barriers have allowed SMR designers to justify the reduction or the elimination of exclusion

43 zones. In order to justify a minimal exclusion zone to the regulator, provisions must be made such that if an SMR fails, no release of radioactive materials to the atmosphere occurs. The AHWR claims that no exclusion zone is required on account of its advanced safety systems and double containment. Designers of the LFR-AS-200 and LFR-TL-X state that their “ultimate goal is the elimination of the need of an emergency preparedness zone” [2]. The SC-HTGR claims that regulatory limits are met at an exclusion boundary of a few hundred meters [2]. The 4S-LMR “foresee no measures needed beyond the plant boundary in response to any severe accidents. [36]. The SSTAR and STAR-LM designers envision that the exclusion zones may be “reduced in size as a result of inherent safety features and the expected low probability of radioactive material release relative to light water reactor designs with a similar power level” [36]. Designers are considering requesting the regulator to have the IRIS’ emergency planning zone largely reduced or eliminated [36]. The AHWR, LFR-AS-200, LFR-TL-X, SC-HTGR, 4S-LMR, SSTAR,

STAR-LM, and IRIS’ overall objective is to reduce the possibility of radioactive materials being released to the environment, thus justifying a reduced exclusion zone.

Passive and inherent safety systems result in a reduction of the release probability for radioactive materials to the environment. The source term that is available for release to the environment, and that remains airborne following release to the atmosphere, is also reduced as the release fractions become lower. While it is understood that current regulatory requirements do not explicitly allow the elimination of the exclusion zone, they seem to be willing to consider its minimalization [5]. As such, detailed analysis is required for each SMR reactor type, so that conclusions specific to each reactor can be made.

CHAPTER 5: CONCLUSIONS

5.1 Summary

The attention paid to small modular reactors (SMRs) has recently grown, as they are viewed as a credible alternative to fossil fuel plants which are to be decommissioned to reduce CO2 and other green-house-gas emissions. SMRs have an electrical output of

30 MWe to 300 MWe [2] and are produced in modules in a factory to shorten construction timelines and the initial capital expenses [1]. The smaller electrical output and the modular nature of SMRs allows for flexibility with applications. SMRs can service off-grid locations, be used for district heating, seawater desalination, methanol production, petroleum refining, thermochemical hydrogen production, and coal gasification [2]. Therefore, an SMR may be installed in close proximity to occupied or populated areas, such as an isolated community, an industrial operation, or an office complex. This necessaires re-examination of the traditional one-kilometer exclusion zone employed in traditional power plants, within which no permanent dwelling is allowed.

The objective of this thesis was to address the sizing of an SMR’s exclusion zone.

Chapter 1 examined the definition of the exclusion zone and the associated national and international regulations and standards. It was shown that the purpose of the exclusion zone is to provide atmospheric space to allow any radionuclides released from a reactor during normal or accidental conditions to be diluted to a level that does not deliver an unacceptable dose to the public residing at the periphery of the zone. This dose limit is defined by the Canadian regulations as an effective dose greater than 1 mSv over a one-year period during normal operating conditions [14], and 0.5 mSv and 20 mSv over a 30-day period following, respectively, an anticipated operational occurrence

(AOO) and design basis accident (DBA) [10]. To be conservative, this work used the whole-body dose from a hypothetical severe reactor accident, instead of the design basis accident to determine the size of the exclusion zone. This requires estimating the Source

Term (the quantity and type of radioactivity released during a sever accident [16]) and the degree of atmospheric dispersion following such release.

In Chapters 2 and 3, a simplified method for determining the whole-body dose from a hypothetical severe reactor accident was presented. The simplicity of this method allowed the determination of the dose at various distances and power levels. Chapter 2 presented and validated an empirical approach for estimating the Source Term. Since there are many SMR designs [2], this work considered a generic reactor to estimate the

Source Term, with the view that the distribution of fission produces is not strongly dependant on the type of fuel used. Chapter 3 modelled the dispersion in a plume from a reactor following a postulated accident to estimate the radiation dose received at varying distances. Three levels of verification for the method described in Chapters 2 and 3 were performed. Following verification, the methodology was then applied to estimate the exclusion zone for SMRs at various power levels.

The radiation dose estimated by the methods of Chapters 2 and 3 did not take into account any innovations in reactor designs to limit the radiation dose. These design and safety features were discussed in Chapter 4, including inherent safety features, new reactor materials and additional engineered barriers. These features should further reduce the size of the exclusion zone and may even eliminate it altogether. In doing so, a review of SMR designs currently in development and a survey of their unique safety features were performed and reported in Appendix A.

The methodology presented in this thesis has several limitations. The empirical approach to determine Source Term ignores radioactive decay; overestimating the prompt fission products and underestimating the delayed ones produced by decay. For the

HotSpot code, if input values deviate from the ones employed in the validation process, the code should be re-calibrated to correct for the effect of the selected parameters. The methodology has been calibrated using published data for a hypothetical severe accident, which is a conservative approach as the size of the exclusion zone is determined for a less-severe design basis accident.

5.2 Conclusions

It was demonstrated in Chapter 2 that a simplified empirical approach to determining reactor Source Term following a hypothetical postulated nuclear reactor accident provided adequate overall values. Following completion of a sensitivity analysis, the model was shown to have a linear sensitivity to its variables, which was to be expected.

It was shown in Chapter 3 that HotSpot, though a health physics code designed to provide “emergency response personnel and planners with a fast, field-portable set of software tools for evaluating incidents involving radioactive material” [26], can be used to approximate the radiation dose following postulated severe accidents for nuclear reactors. By using the empirical approach described in Chapter 2 as an input to the

HotSpot code, the exclusion zone radius for SMRs of differing power levels was quickly estimated. It was found that the calculated exclusion zones for SMRs are less than the

914 meters (1,000 yards) traditionally employed in Canadian nuclear power plants.

Passive and inherent safety features, innovations in reactor materials, and engineered barriers can result in a reduction of the release probability for radioactive materials to the environment, as discussed in Chapter 4. This will allow SMR designers to justify a reduction or elimination of the exclusion zone. While it is understood that current regulatory requirements do not explicitly allow the elimination of the exclusion zone, they seem to be willing to consider its minimization [5].

5.3 Contribution to Knowledge

This thesis built on previously available information regarding Source Term estimation and exclusion zone sizing for conventional nuclear reactors, while contributing new knowledge pertaining specifically to SMRs. In doing so, the use of a simplified empirical relation to estimate the Source Term was validated against published data from a detailed analysis. Further, this is the first time HotSpot, a health physics code, was used to assess the dose following postulated severe accidents for nuclear reactors, instead of more complex codes often used to complete atmospheric dispersion modeling, such as the United States of America Department of Energy’s MACCS2 code [22]. The empirical approach to estimate Source Term and use of HotSpot combine to create a simplified tool that enables the quick examination of different conditions and provides a way to perform a parametric study to determine the exclusion zone for various SMR power levels. It should be noted though that these simplified models need to be recalibrated when significantly different parameters and conditions are employed.

5.4 Recommendations for Future Work

Moving forward from this thesis, there are several recommendations for future work, including:

1) A similar study to the CNSC study [20] should be completed for a design based

accident (DBA). The CNSC study [20] used to calibrate the models in this work

included data for a hypothetical severe accident, i.e. beyond design basis accident

(BDBA), while the exclusion zone is defined for a design basis accident (DBA)

[10]. As such, proposed exclusion zone sizes for SMRs at varying power levels

presented in Chapter 3 are overestimated.

2) It is recommended that the release fractions contained in the method developed in

Chapters 2 and 3 be adapted to account for new fuel compositions, since the one

used in this work resembled those used for a CANDU reactor [20].

3) The study should be repeated for specific SMR designs, in to order to incorporate

the effect of fuel type and composition, reactor materials devised to retain fission

products, and special barriers to the release of radioactivity. In Chapter 3, the

methodology for approximating SMR exclusion zones was verified using data

from a published CNSC study [20] that modelled the hypothetical failure of a

CANDU reactor. While conventional reactors, such as the CANDU reactors,

share many similarities with SMRs, the difference in nature and design can be

sufficiently significant to alter the value of the Source Term at a given power

level.

4) It is recommended that various engineered barriers be evaluated to determine their

effectiveness in reducing the quantity of radioactive materials released to the

environment following a reactor accident. The software used within this work,

HotSpot, is only suited for modeling atmospheric dispersion. Therefore, different

49 modelling methods will need to be used to measure the effectiveness of engineered barriers.

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Appendix A: Review of Small Modular Reactors

A.1 Introduction

There are currently more than fifty SMR designs and technologies at varying stages of development, with interest continuing to grow amongst the international community. The smaller physical size and power output allows for SMRs to be incrementally employed, as required, to meet electrical demand [39]. The small size is also attractive for jurisdictions with smaller populations, such as the Province of

Saskatchewan, as conventional nuclear reactors are not practical due to the large power outputs, physical size, and capital expenditure. There are a number of applications for

SMRs beyond traditional power generation, including seawater desalination, district heating, and hydrogen generation [40]. Another potential application is to use an SMR as a replacement for existing coal-fired power plants that are being decommissioned [40].

The modularity of SMRs enables major reactor components to be mass-manufactured in a factory setting and then assembled on site. This reduces the capital expense associated with construction and allows for a more efficient construction timeline [2]. The proceeding sections provide an overview of the technology, including: the nature of modular construction, a description of prominent designs, and proposed applications.

A-1

A.2 Water Cooled Reactors

Of the countries currently developing SMRs, nearly all have committed to some form of water cooled reactor. In total, fifteen water cooled SMRs are being designed [39].

This is consistent with conventional nuclear power plants, where water cooled reactors account for more than 95% of all operating reactors [40]. Water cooled reactors can be classified into two (2) major categories: light water cooled and heavy water cooled.

Moreover, light water cooled reactors can be further differentiated as pressurized water or boiling water [2]. The following section provides brief summaries of light water reactors currently under development.

A.2.1 Light Water Reactors

Light water reactors and pressurized water reactors consist of two coolant loops.

The primary loop removes energy that is produced via fission and transfers it to the secondary coolant system, where the energy is then used to heat water and produce steam

[39]. Most SMRs are designed using an integrated design, which means that the steam generators, pressurizer, and primary coolant system are located within the reactor pressure vessel. Ultimately, this results in the elimination of the potential for coolant to be released from the reactor vessel [39]. While conventional safety measures such as control rods, shut down rods, and poison injection are still used by reactors, light water

SMRs also employ natural circulation, which reduces the reliance on mechanical elements during accident conditions [39]. Table A-1 provides a summary of the main features of light water SMRs.

A-2

Table A-1: Summary of light water cooled small modular reactors [39, 2, 41]

Reactor name: FBNR KLT-40S VBER-300 VVER-600 IMR NuScale

Electric power 72 70 325 600 350 540 (MWe):

Thermal power 218 300 917 1,600 1,000 1,920 (MWth):

Fuel type: CERMET UO2 UO2 UO2 & Gd UO2 UO2

Enrichment (%): 5 13 4.95 <5 4.8 4.95

Fuel configuration: Sphere Pellet Pellet Pellet Pellet Pellet

Neutron spectrum: Thermal Thermal Thermal Thermal Thermal Thermal

Development stage: Concept Construct Concept Concept Concept Develop

Country of origin: Brazil Russia Russia Russia Japan USA

Reactor Name: SMART ACP-100 mPower W-SMR SMR-160 DMS

Electric power 100 100 195 225 160 300 (MWe):

Thermal power 330 310 575 800 525 840 (MWth):

Fuel type: UO2 UO2 UO2 UO2 UO2 UO2

Enrichment (%): 4.8 2.4 to 4.0 <5 <5 <4.95 <5

Fuel assembly type: Pellet Pellet Pellet Pellet Pellet Pellet

Neutron spectrum: Thermal Thermal Thermal Thermal Thermal Thermal

Development stage: Concept Concept Develop Concept Concept Concept

Country of origin: Korea China USA USA USA Japan

Fixed Bed Nuclear Reactor (FBNR): The FBNR is a Brazilian light water cooled and moderated thermal reactor with a 72 MWe (218 MWth) capacity [42]. The reactor fuel consists of 5% enriched UO2 spheres covered in zirconium cladding, which is referred to as CERMET fuel [2]. Figure A-1 shows a schematic of a CERMET fuel element. The spherical shaped fuel also serves as an inherent safety feature as this allows

A-3 for the core to be emptied of fuel elements following a reactor accident, which in turn results in loss of criticality [42]. The Federal University of Rio Grande do Sul (FURGS) is currently in the conceptual design phase for the FBNR [42].

Figure A-1: A schematic of FBNR Fuel [42]

KLT-40S: The KLT-40S is a Russian light water cooled and moderated thermal reactor with a 70 MWe (300 MWth) capacity [43]. The KLT-40S originates from the

KLT-40, which is a marine propulsion plant that has demonstrated failure-free operation for approximately 300 reactor years [2]. The reactor consists of approximately 14% enriched UO2, which is termed low enriched Uranium by the IAEA [43]. The fuel is contained within hexahedral shrouded fuel assemblies as shown in Figure A-2, positioned in a triangular lattice [43]. The KLT-40S’s design incorporates a number of inherent and passive safety features, which strives to minimize the likelihood of a release of radioactive materials to the atmosphere [43]. Safety systems include a negative reactivity coefficient, which assists with suppressing a rapid increase of power levels resulting from supercriticality (i.e. excursion) [43]. The KLT-40S’s fuel composition has high thermal conductivity, which results in a relatively low fuel temperature and eliminates the risks

A-4 associated with a core-meltdown [43]. Further, adequate natural circulation within the primary systems minimizes the reliance on mechanical systems to remove reactor decay heat, thus preventing over pressurization of the reactor core [43]. Finally, the KLT-40S has a high heat storage capacity, which accommodates increases in reactor pressure [43].

Afrikantov OKB Mechanical Engineering is currently supervising the construction of a

KLT-40S unit [43].

Figure A-2: A schematic of KLT-40S Fuel Assembly [43]

VBER-300: The VBER-300, as schematically shown in Figure A-3, is a Russian light water cooled and moderated thermal reactor with a 325 MWe (917 MWth) capacity

[44]. The reactor consists of 5% enriched UO2 fuel in the form of pellets [44]. The

VBER-300 design is a result of nearly 6,500 reactor-operating years of experience with marine propulsion systems. As a result, the main design is quite similar to that of a marine-based reactor and is intended to be used as either a ground-based reactor or marine-based [2]. The VBER-300 includes inherent safety features that are capable of

A-5 safely shutting down the reactor and limiting the amount of energy released [44]. For example, the VBER-300 has a negative reactivity coefficient that assists with suppressing reactor excursion, which is a rapid increase of power levels resulting from supercriticality

[44]. The VBER-300 also employs natural circulation, which allows for decay heat removal following reactor shutdown regardless of if mechanical systems are not functioning. The inherent safety features result in the simplification of conventional safety requirements. The manufacturer claims that the inherent safety features are stable against any disturbances, including personnel errors and terrorist interference [44].

Afrikantov OKB Mechanical Engineering is currently completing the conceptual design for the VBER-300 [44].

Figure A-3: An overview of a VBER-3000 SMR [44]

VVER-600: The VVER-600 is a Russian light water cooled and moderated thermal reactor with a 600 MWe (1,600 MWth) capacity [45]. The reactor consists of 5% enriched UO2 and UO2Gd fuel, which is contained in a hexahedral fuel assembly [2]. The

A-6 addition of gadolinium oxide (Gd2O3) is unique to the VVER-600 and is used as an integrated burnable absorber. Similar to other SMRs, passive safety systems that are used to manage beyond design basis accidents are employed. These include a core passive flooding system and a steam generator passive heat removal system [45]. Following a loss of coolant accident, the passive flooding system will flood the pressure vessel of the

VVER-600 with water from the hydroaccumulators [46]. The hydroaccumulators are separated from the reactor by check valves. When the reactor pressure drops below the design threshold the cheque valves open, allowing the water supply to enter the reactor

[46]. The VVER-600 consists of several natural circulating circuits, each connected to the stream generator. As heat from the reactor core is released, it generates steam. This steam then condenses, which rejects the heat to the ambient air [46]. Passive systems are to operate in tandem with the active safety systems to maintain an acceptable safety level

[45]. Figure A-4 shows a schematic of the equipment arrangement for the VVER-600.

The VVER-600 is currently in the conceptual design phase, which is being managed by

Gidropress [45].

A-7

Figure A-4: An overview of a VVER-600 SMR [45]

VVER-640: The VVER-640 is a Russian light water cooled and moderated thermal reactor with a 645 MWe (1,800 MWth) capacity [47]. The reactor consists of

UO2 pellets, with initial enrichment level for the fuel averaging at 1.72% [2]. The low enrichment results in fuel-cost reduction. The VVER-640 is equipped with passive safety systems that remove all decay heat from the reactor [2]. The ECCS hydro-accumulator injects cooling water into the reactor core during a loss of coolant accident [47]. As the primary reactor vessel’s pressure decreases, check valves allow cooling water to enter the core [47]. Following a loss of coolant accident or failure of active safety systems, decay heat is removed from the steam generator via the passive heat removal system and the containment heat removal system. Both systems are designed for long-term removal of reactor decay heat through the removal of heat from the steam systems [47]. It is the

A-8 intention to have all design basis accidents controlled via the passive systems, where active systems are only employed to further mitigate an accident scenario [47]. Figure A-

5 is a schematic for the VVER-640 reactor that is currently being designed by Gidropress

[47].

Figure A-5: A schematic of VVER-640 SMR [48]

Integrated Modular Water Reactor (IMR): The IMR is a Japanese light water cooled and moderated thermal reactor with a 350 MWe (1,000 MWth) capacity [48], a

A-9 schematic of which is shown in Figure A-6. The reactor consists of UO2 fuel, with an enrichment averaging 4.80% [2]. While there are 349 fuel rods with the uranium dioxide fuel, there are also 32 rods that contain gadolinium, which assists with balancing reactivity within the core by managing fuel burn-up [48]. The IMR is equipped with a number of inherent safety features, such as: the hybrid heat transport system to eliminate loss of flow and an in-vessel control rod drive mechanism to eliminate control rod ejection [48]. The hybrid heat transport system is able to employ natural circulation through the use of reactor decay heat and the steam generator. As heat is generated, the coolant begins to boil at the bottom of the reactor, which results in the coolant bubbling and rising via a channel to the top of the reactor, where it is then cooled via the steam generator [48]. The in-vessel control rod drive mechanism is contained entirely within the reactor vessel. This eliminates the possibility of the control rod assembly being ejected due to difference in reactor vessel and ambient air pressure following a mechanical failure [48]. The conceptual design phase has been completed by the designer, Mitsubishi

[48].

A-10

Figure A-6: A schematic of an IMR SMR [48]

NuScale: The NuScale Power Modular and Scalable Reactor, as shown in Figure

A-7, is an American light water cooled and moderated thermal reactor with a 540 MWe

(1,920 MWth) capacity [49]. The reactor consists of UO2 ceramic pellets with an enrichment of less than 4.95% [2]. The fuel is encased in an array of fuel assemblies that are shorter in length than that of other SMRs and allow for a compact reactor size [49].

This is one of the characteristics that distinguish the NuScale from other SMR designs.

Another is that the reactor has a smaller fuel inventory than other SMRs, which would result in an accidental release of radioactivity well below that of conventional nuclear reactors [49]. The core is cooled via natural circulation and the reactor has a modular containment rather than the traditional cast in-place design [49]. NuScale is equipped with two redundant passive safety systems and does not require external power to actuate

A-11

[2]. First, the NuScale reactor is submerged within a pool of water, which serves as a heat sink capable of absorbing the decay heat produced by the reactor core for more than 30 days [49]. Second, vent valves are opened, which allows for steam to be released from the vessel and condense along the sides of the vessel [49]. The condensate travels to the sump and recirculation valves are opened [49]. This allows for natural circulation of the liquid through the vent valves and recirculation valves [49]. The NuScale is currently being developed by NuScale Power.

Figure A-7: A schematic of a NuScale SMR [49]

System-Integrated Modular Advanced Reactor (SMART): The SMART, as shown in Figure A-8, is a Korean light water cooled and moderated thermal reactor with a 100 MWe (330 MWth) capacity [50]. The reactor consists of 5% enriched UO2 fuel contained in rod bundles [50]. SMART offers enhance safety and reliability through the implementation of inherent safety features and passive safety systems. The inherent

A-12 safety features include an integral coolant system and improvements to the natural circulation of the reactor [2]. The passive safety features introduced in the SMART design include the installation of a residual heat removal system that mitigates loss of coolant accidents [50]. During an accident condition, the passive residual heat removal system will prevent the over pressurization of the primary system via natural circulation, which removes decay heat produced in the reactor core via the steam generator.

Conceptual designs for the SMART are currently being completed by KAERI.

Figure A-8: A schematic of a SMART SMR [50]

ACP-100: The ACP-100 is a Chinese light water cooled and moderated thermal reactor with 100 MWe (310 MWth) capacity [2], a schematic is shown in Figure A-9.

The UO2 fuel is enriched between 2.4% to 4% and contained within squared fuel assemblies [2]. The ACP-100 is to be buried underground to mitigate the likelihood of a

A-13 radioactive release to the atmosphere [2]. The reactor is designed to eliminate the possibility of a loss of coolant accident due a number of passive safety systems, including a decay heat removal system and passive emergency core cooling system [2]. The decay heat removal system is designed to prevent a reactor core meltdown following a beyond design basis accident. The decay heat is removed via a heat sink that is naturally circulated [2]. Following a loss of coolant accident, the passive emergency core cooling system condenses the steam located within the reactor vessel. This results in a release of heat from the steam, which is removed from the reactor vessel via conduction [2]. Basic designs for the ACP-100 have been completed and are being led by the China National

Nuclear Corporation [2].

Figure A-9: A schematic of a ACP-100 SMR [2]

mPower: The mPower reactor, as shown in Figure A-10, is an American light water cooled and moderated thermal reactor with 195 MWe (575 MWth) capacity [2].

The UO2 fuel pellets are enriched up to 5% [2]. The reactor is equipped with a steel

A-14 containment vessel located entirely underground [2]. The containment structure is able to passively cool via a water tank located on the outside of the structure. As the steam condenses in the reactor core, the heat is transferred through the dome via conduction [2].

The emergency core cooling system serves to depressurize the reactor, reactor inventory control during an accident condition, and remove decay heat. [2]. During an accident condition, the reactor will use valves to regulate pressure between the reactor core and containment structure [2]. Following this, check valves will allow cooling water to be injected into the reactor core [2]. The injection of cooling water allows for passive cooling of the reactor, without need for external intervention [2]. The mPower is currently under development by Generation mPower, LLC [2].

Figure A-10: A schematic of a mPower SMR [2]

A-15

Westinghouse SMR: The Westinghouse SMR is an American light water cooled and moderated thermal reactor with 225 MWe (800 MWth) capacity [2]. The reactor employs 89 fuel assemblies using UO2 pellets enriched to less than 5% [2]. The SMR design is based on Westinghouse’s larger AP-1000 reactor, which further builds on the concepts of simplicity and passive safety [39]. The SMR is entirely modular which limits the size of primary components and allows for ease of transportation [2]. The passive safety features employ gravity and natural circulation in order to mitigate accidents [2].

This means that the reactor itself is not reliant on external power to perform safety functions [2]. Conceptual designs for the Westinghouse, shown schematically in Figure

A-11, have been completed by the Westinghouse Electric Company LLC [2].

Figure A-11: A schematic of Westinghouse SMR [2]

SMR-160: The SMR-160, also referred to as the Holtec SMR, is an American light water cooled and moderated reactor with 160 MWe (525 MWth) capacity [2],

A-16 shown schematically in Figure A-12. The reactor has 4.95% enriched UO2 fuel pellets and employs a square array of fuel assemblies [2]. The SMR-160 is described as having a simplistic design, with fewer valves, pumps, heat exchangers, and instrumentation [2].

This results in a lower likelihood of mechanical system failures. There are a number of unique safety features within the passive core cooling systems, which include: a primary heat removal system, the secondary decay heat removal system, the automatic depressurization system, and the passive make-up water system. The primary heat removal system rejects heat from the primary coolant to the secondary loop via heat transfer [2]. The automatic depressurization system is designed to regulate the reactor’s cooling system pressure, which allows for make-up cooling water to be injected into the reactor vessel during an accident condition [2]. While the reactor is intended to be used for electricity production, it is also capable of hydrogen generation, district heating, and seawater desalination [2]. Another potential application is to use the SMR-160 as a replacement for existing coal-fired power plants that are being decommissioned [2].

A-17

Figure A-12: A schematic of SMR-160 [2]

DMS: The DMS is a Japanese light water cooled and moderated reactor with a

300 MWe (840 MWth) capacity. The DMS uses UO2 fuel pellets that are enriched up to

5% [2]. Similar to other SMRs, natural circulation of the coolant is used to remove heat produced in the core, which eliminates the need for recirculation pumps [2]. The pressure vessel is simplified, which results in the volume of the unit being significantly reduced.

The reduced reactor pressure vessel size results in the reduction of the construction, manufacturing, and transportation timelines. The reactor is currently being designed by

Hitachi-GE Nuclear Energy [2]. While all SMRs are intended to produce electricity, the

DMS is also suitable for remote regions with less developed grid infrastructure. The

A-18

DMS is also capable of district heating, oil sand extraction, steam assisted gravity drainage, and desalination [40].

A.2.2 Heavy Water Reactors

Heavy water reactors account for nearly 50 nuclear power plants worldwide and are the second most common reactor design [39]. Heavy water is used as the moderator and the fuel is natural uranium due to the moderator’s low absorption of neutrons [39].

The coolant for a heavy water reactor is dependent on the design, but is either heavy water or light water [39]. In heavy-water SMR designs, natural circulation is used to remove heat during normal operating conditions, as well as during accident conditions.

Similar to the above described light water cooled reactors, this results in a reduced reliance on mechanical systems, such as pumps. The calandria, which contains the heavy water moderator, is at a near-ambient temperature and atmospheric pressure and is equipped with a series of passive cooling system. The passive cooling systems will be discussed in greater detail below. This results in decreased leakage of tritium, which is produced by neutron absorption of the heavy-water’s deuterium, to the atmosphere. The following provides a summary of the main features of heavy water SMRs, see also Table

A-2.

A-19

Table A-2: Main features of heavy water cooled SMRs [39, 2, 41]

Reactor name AHWR IPHWR Electric power (MWe) 304 236 Thermal power (MWth) 920 755 Fuel type (ThU) MOX & UO2 (ThPu) MOX Enrichment (%) 2.5-3.5 U and 0.7 2.5-4 Pu Neutron spectrum Thermal Thermal Development stage Design In operation Country of origin India India

Advanced Heavy Water Reactor (AHWR): The AHWR is a light water cooled and heavy water moderated reactor with a 304 MWe (920 MWth) capacity. The thermal reactor is currently being designed by the Bhabha Atomic Research Centre in India [51].

As shown in Figure A-13, the reactor uses Th233U and Th/Pu mixed oxide fuels [51]. The plutonium requirements for the reactor is satisfied by reprocessing the spent fuel of pressurized heavy water reactors. The Th233U mixed oxide fuel is reprocessed from the recovered Thorium and Uranium [51]. The reactor fuel assembly consists of three (3) rings, each with different levels of enrichment: ring 1 uses a Th233U mixed oxide fuel with 3.0% enrichment; ring 2 uses a Th233U mixed oxide fuel with 3.75% enrichment; and ring 3 uses a Th/Pu mixed oxide fuel with an upper half enrichment of 4.0% and lower half of 2.5% [51]. A design objective of the AHWR is to implement passive and active safety features in lieu of an exclusion zone [51]. Following reactor shutdown, passive removal of heat from the steam drums is completed using the pool, which acts as a heat sink. Further, if a loss of coolant accident were to occur, the emergency core cooling system will employ accumulators to regulate pressure between the containment

A-20 structure and the reactor core. This allows for cooling water to be injected into the core, which serves as a means to passively cool the core for approximately 7 days [51].

Figure A-13: Fuel Cycle for AHWR [51]

The primary containment is surrounded by the secondary containment and both are designed to withstand the effects of a Loss-of-Coolant Accident (LOCA), which results from a loss of coolant following a break in the primary heat transport system [52,

32]. The primary containment is 48 meters wide, 75.5 meters high, 21 meters below grade, and contains a number of advanced safety systems that are intended to control the release of radioactivity and remove heat from the containment atmosphere during an accident condition [52]. The primary containment is 40” thick and is constructed of prestressed cement concrete with an epoxy liner. Prestressed cement concrete is made of heavy concrete, which has increased density and allows for the shielding of neutron and gamma radiation [53]. The use of prestressed concrete also eliminates the tensile stresses at the inside face by using high tensile strength steel wires [54]. This allows for increased tensile stress resistance and minimizes the risk of cracking, whereas heavy concrete alone

A-21 is often employed to resist compressional forces rather than tension [54]. The pre-stressed concrete containment structure resembles the design of the CANDU 600 reactor. The secondary containment is 57 meters wide, 72 meters high, 21 meters below grade, 32” thick, and is constructed of reinforced cement concrete. [52] Reinforced concrete is permeable and, unlike prestressed concrete, it is normal for small cracks to exist due to tension. As a result, a steel liner is installed to eliminate the risks associated with radionuclide leakage. [54]. The steel liner is considered to be a superior form of barrier, with a higher reliability over the epoxy lining employed by prestressed concrete. The reinforced concrete containment structure resembles the design of the CANDU 300 reactor. While the safety systems employed by the AHWR resemble those of the

CANDU 300 and 600 reactors, the unique design characteristic is the use of double containment, which is not employed by either CANDU reactor. Due to the smaller size of the AHWR when compared to the CANDU, it is feasible to construct the double concrete containment structure. The double concrete containment and advanced safety systems are designed to eliminate or minimize the need for an exclusion zone, combining design characteristics from both the CANDU 300 and 600. SMRs are defined as a reactor less than 350 MWe, making the AHWR a large SMR. If the AHWR containment structure is able to effectively eliminate the required exclusion zone then this same containment structure and equivalent advanced safety systems can be applied to smaller reactors to achieve the same result [17].

Indian PHWR (IPHWR): The IPHWR is an Indian heavy water cooled and moderated reactor with natural UO2 fuel and a capacity of 236 MWe (755 MWth) [55].

The IPHWR reactor is in operation and was designed by the Nuclear Power Corporation

A-22 of India (NPCIL) [55]. There are 16 PHWR units of similar size in India, providing considerable operating experience [55]. India’s first two (2) units, referred to as RAPS-1 and RAPS-2, are of Canadian design and are modelled after the Douglas Point CANDU reactor [55]. In fact, most of the equipment used to construct RAPS-1 was imported from

Canada and construction of the unit was made possible through collaboration between

Indian and Canadian officials [55]. Through the use of an emergency core cooling system

(ECCS), high pressure heavy water (D2O) or medium pressure light water (H2O) is injected into the reactor in case of a major loss-of-coolant accident. In the event that the accident demands additional water supply, the ECCS activates an additional low-pressure recirculation pumps and heat exchangers. As a last resort, fire water can be injected into the reactor.

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A.3 Gas Cooled Reactors

Gas cooled reactors account for approximately 3% of all operating nuclear power plants worldwide [56]. The fuel is in the form of tri-structural isotropic particle fuel

(TRISO), rather than conventional pellets [39]. Tri-structural isotropic particle fuel is spherical and has a plutonium or uranium core and a graphite and silica coating [39]. Gas cooled reactors employ Helium (He) or carbon dioxide (CO2) as the coolant with graphite often being used as the moderator [39]. Neither helium nor carbon dioxide experience phase change, avoiding the changes in cooling capacity associated with boiling of water coolants [2]. This allows the fuel to tolerate higher temperatures without damage [2]. The following provides an overview of gas cooled SMRs, with Table A-3 serving as a summary.

Table A-3: summary of gas cooled SMRs [39, 2, 41]

Reactor name: HTR-PM PBMR P-HTR Electric power (MWe): 211 165 150 Thermal power (MWth): 500 400 350

Fuel type: UO2 TRISO UO2 TRISO UO2 TRISO Enrichment (%): 8.5 9.6 15.5 Neutron spectrum: Thermal Thermal Thermal Development stage: Demonstrate Conceptual Conceptual Country of origin: China South Africa USA

High Temperature GCR - Pebble-Bed Module (HTR-PM): The HTR-PM is a helium cooled and graphite moderated thermal reactor with a 211 MWe (500 MWth) capacity [57]. The SMR is currently being designed by Tsinghua University in China

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[57]. The HTR-PM uses a UO2 tri-structural isotropic particle fuel with an enrichment level of 8.5%, which is shown in Figure A-14.

Figure A-14: Spherical Fuel Elements [57]

The HTR-PM demonstrates a variety of inherent safety features [57]. For example, the HTR-PM structure employs a large quantity of graphite, which has a high heat capacity. As such, during accident conditions, decay heat will be rejected from the reactor core to the reactor vessel via conduction and radiation [57]. Further, the reactivity coefficient is negative, which will result in the reactor suppressing rapid increases in reactor power levels [57]. The intent of the HTR-PM design is to eliminate the possibility of an accident that results in the release of radionuclides to the atmosphere

[57].

Pebble Bed Modular Reactor PBMR: The PBMR is a helium cooled and graphite moderated thermal reactor with a capacity of 165 MWe (400 MWth) [58].

Originating in South Africa, the PBMR is currently being designed by Pebble Bed

Modular Reactor (Pty) Limited [58]. UO2 tri-structural isotropic particle fuel is used,

A-25 shown in Figure A-15, which has an enrichment level ranging between 5% to 20% and temperature limits exceeding 1,600 ℃ [58]. Even as temperatures approach 1,600 ℃, the fission products are contained within the fuel particle, which is one of the unique safety features of the PBMR. The higher temperature allowance also results in improved thermodynamic efficiency [58]. The reactor’s design allows for the passive rejection of decay heat from the reactor core to the reactor cavity cooling system, which is a heat sink, via convection and conduction. This passive feature reduces the reliance on traditional mechanical equipment [58]. In fact, the only major mechanical equipment in the primary reactor system is the circulator and steam generator [58].

Figure A-15: PBMR Fuel Element Design [58]

Prismatic Modular High Temperature GVR (Prismatic HTR): The Prismatic

HTR, shown in Figure A-16, is a helium cooled and graphite moderated thermal reactor with a capacity of 150 MWe (350 MWth) [59]. The Prismatic HTR uses UO2 fuel enriched to 15.5% [59]. It includes a number of inherent safety characteristics, including

A-26 that the helium coolant does not become radioactive. [59]. Helium also has improved thermodynamic properties when compared to other gases, such as CO2, higher specific heat, and a lower neutron cross section [60]. The fuel also contains a graphite core which allows for increased heat capacity, increased structural stability at higher temperatures, and decreased thermal response [59]. The Prismatic HTR is equipped with two (2) active heat removal systems, with a back-up passive heat removal system surrounding the reactor vessel [59]. In the event that the passive heat removal system fails, the reactor temperature will not exceed design limits due to conduction of heat from the core to the surrounding walls and earth, and the thermal radiation from the vessel [59].

Figure A-16: Prismatic HTR reactor schematic [59]

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A.4 Molten Salt Reactors

Molten salt reactors were originally developed in the 1950s [61], but are re- emerging as candidates for SMRs [61]. Molten salt allows for increased operating temperatures, which result in improved thermal efficiencies [61]. The increased temperatures are ideal for electricity generation, but also for a number of process heat applications [61]. Molten salt reactors also operate at lower coolant pressures, mitigating the risks associated with, and the likelihood of, a loss of coolant accident, thereby serving as a distinct safety characteristic [2] [61]. Although dependent on the particular design, a number of designs don’t require solid fuel, eliminating the risks associated with manufacturing and disposal [2] [61]. The reactor is rather flexible and can also be used with a variety of fuel types, including uranium-plutonium and thorium-uranium cycles

[61]. The proceeding sections describe the features of a number of molten salt SMR designs currently under development, with Table A-4 serving as a summary.

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Table A-4: Summary of Molten Salt SMRs [62, 63, 64, 65, 66, 67, 68] Reactor name: MK1PB-FHR ThorCon IMSR-400 MSTW Electric power (MWe): 100 250 194 115 Thermal power (MWth): 236 557 400 270 Fuel type: TRISO NaF-BeF2- UF4 in Sodium- ThF4-UF4 diluent actinide fluorides fluoride Enrichment (%): 19.8 19.7 Start-up: 2-3 1.1 Makeup: 5- 19 Neutron spectrum: Thermal Thermal Thermal Thermal Development stage: Concept Design Design Concept Country of origin: USA International International Denmark

Reactor name: MSR-FUJI SSR-U SmAHTR LFTR Electric power (MWe): 200 300 N/A 250 Thermal power (MWth): 450 750 125 600 Fuel type: Molten salt Molten TRISO Uranium-233 with thorium fluoride salt derived from and uranium Thorium Enrichment (%): 2.0 <15 8 N/A Neutron spectrum: Thermal Thermal Thermal Thermal Development stage: Concept Concept Design Concept Country of origin: Japan UK USA USA

Mark 1 Pebble-Bed Fluoride-Salt-Cooled High Temperature Reactor

(MK1PB-FHR): This is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 100 MWe (236 MWth) [62]. The reactor, schematically shown in

Figure A-17, uses tri-structural isotropic particle fuel coated uranium fuel particles enriched to 19.8% [62]. Similar to the PBMR, the reactor fuel is spherical with the internal tri-structural isotropic particle fuel particles encased in several graphite coatings

[62]. The design incorporates a number of passive safety features. The reactor core heat removal system is controlled using passive check valves [62]. Following a loss of coolant accident, the check valves will allow for water to enter the reactor core [62]. This results

A-29 in natural circulation of the water. Further, the control rods will insert if the reactor coolant temperature exceeds 615 °C due to buoyancy [62]. As temperature increases, the water’s viscosity and density decreases, which results in reduced buoyancy to maintain the position of the control rods [62].

Figure A-17: MK1PB-FHR reactor schematic [62]

ThorCon: The ThorCon, shown in Figure A-18 is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 250 MWe (557 MWth) [63]. Using a liquid fuel, UF3 and UF4, the reactor’s enrichment level is 19.7% [63]. ThorCon is marketed as a “walkaway safe” reactor, meaning that if the reactor overheats it will automatically shut down and then drain the fuel to a tank where it is then passively cooled [69]. The fission products of greater concern, including Sr-90 and Cs-137, are

A-30 chemically bound to the salt [69]. This means that draining the fuel from the reactor also reduces the possibility of harmful radionuclides being released into the atmosphere following an accident [69]. One of the ThorCon’s unique design features is that the operator is unable to prevent draining of the fuel or cooling of the reactor following overheating [69]. The reactor is buried 15 meters below grade, contains three (3) physical containment barriers, and operates at a positive pressure to mitigate risks associated with a rupture of the primary loop [63]. Similar to other SMRs, the Thorcan reactor is modular and manufactured in blocks which are then shipped to site for assembly.

Figure A-18: ThorCon reactor schematic [70]

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Integrated Molten Salt Reactor-400 (IMSR-400): The IMSR-400, shown in

Figure A-19, is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 194 MWe (400 MWth) [71]. The IMSR’s safety philosophy is designed around removing the possibility of radioactive materials being pushed into the environment, which is achieved by the reactor operating at a low pressure [71]. The

IMSR does not rely on operator intervention, mechanical components, or coolant injection during an accident condition [71]. This is achieved through a fully passive core complete with a containment cooling system [71]. One distinct innovation of the IMSR design is the integration of primary reactor components [64]. The IMSR’s major components are each mounted within a single vessel, allowing the core to be manufactured in a controlled facility, where it is then shipped to site and assembled.

Following the end of useful life, the sealed core is safely replaced [64].

Figure A-19: IMSR-400 core schematic [71]

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Molten Salt Thermal Wasteburner (MSTW): The MSTW is a molten salt cooled and graphite moderated thermal reactor with a capacity of 115 MWe (400 MWth)

[65]. Under abnormal operating conditions, control of the MSTW reactor does not require any active measures, meaning that the reactor does not require human intervention [65].

The MSTW is designed to operate at an optimized configuration, whereby any deviation would lead to the system moving away from the optimum and result in a reactor shutdown if not alleviated [65]. This is further demonstrated in Figure A-20, which indicates that the core is over-moderated to ensure negative temperature and void reactivity coefficients [65]. Negative temperature coefficient means that the reactor deviates down from criticality (reactivity) when the temperature increases, while a negative void coefficient means that the reactivity decreases as the void content increases.

Figure A-20: MSTW optimal configuration principle [65]

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Molten Salt Reactor-FUJI (MSR-FUJI): The MSR-FUJI is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 207 MWe (450 MWth)

[66]. The MSR-FUJI design is based on previous molten salt reactor designs developed or operated at the Oak Ridge National Laboratory [72]. There are a number of special features included in the MSR-FUJI design, including a modular design to allow for multi- module plants and a lifetime core operation that does not require on-site refueling. The reactor’s characteristics allow for the possibility of severe accidents to be minimized. For instance, the reactor operates at a low pressure and is inert at increased temperatures, which minimizes the possibility and consequences of high-pressure rupturing and eliminates the possibility of a steam and/or hydrogen explosion [66]. In the event of an emergency, the molten salt is drained to an emergency drain tank designed to ensure that a re-criticality accident does not occur. In fact, the molten salt only reaches criticality within the reactor’s graphite core. Figure A-21 is a schematic of the MSR-FUJI.

Figure A-21: MSR-FUJI reactor schematic [66]

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Stable Salt Reactor (SSR-U): The SSR-U is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 300 MWe (750 MWth) [67]. Figure A-22 shows the reactor module for the SSR.

Figure A-22: SSR reactor core module [73]

The SSR-U’s design philosophy is to reduce plant costs by simplifying the design and eliminating hazards rather than containing them [67]. The designer intends to achieve this through the implementation of a number of features that combine the safety and operational benefits of conventional reactors with molten salt reactors [67]. One of the features is to virtually eliminate the radioactive material that may be released following an accident, which would minimize the liability associated with the release of radioactive materials to the atmosphere [67]. For reactors that employ conventional fuels, the release of Cesium and Iodine to the atmosphere remains a major concern to human health and safety [73]. However, for molten salt fuel, these fission products are in the form of stable

A-35 salts that are unable to be dispersed as gasses through the air [73]. The designer has theorized that this has the potential to reduce the volatile radioactive material by up to six orders of magnitude when compared with a traditional oxide fueled reactor [73]. Through this, it is believed that the exclusion zone can be reduced or possibly eliminated entirely

[73].

Small fluoride salt-cooled High Temperature Reator (SmAHTR): The

SmAHTR, shown in Figure A-23, is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 125 MWth [74]. The reactor employs tri-structural isotropic particle fuel, enriched to 8%. The coated fuel is coated with a series of protective layers and is embedded with graphitic material, which has an increased thermal failure point and serves as a barrier to assist with preventing the release of fission products to the atmosphere [75]. The SmAHTR design shares a number of key design features of other SMRs, including a large negative reactivity coefficient, natural convection for passive decay heat removal, and numerous inherent safety design features

[75]. The SmAHTR is ideal for remote applications as the passive decay heat removal system is non-reliant on off-site power [75]. The inherent safety features combined with a reduced source term will allow the designer to justify a smaller emergency preparedness zone to the regulator [75]. The core and primary components are all contained within the reactor vessel, eliminating the likelihood of a loss of coolant accident [75]. In fact, there are several barriers preventing the release of radioactive materials to the atmosphere, including the coated particle fuel mentioned above, graphite moderator, reactor guard vessel, and the physical containment structure [75].

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Figure A-23: SmAHTR vessel schematic [74]

Liquid Fluoride Thorium Reactor (LFTR): The LFTR is a fluoride salt cooled and graphite moderated thermal reactor with a capacity of 250 MWe and 600 MWth [68].

It uses U-233 fuel derived from Thorium [68]. The LFTR has four (4) main design principles demonstrated. The first which is an inherently safe reactor that operates at a low-pressure [68]. The reactor is self-controlling as any increase to operating temperature results in a decrease to the fuel salt density, which will inherently stabilize the reactor without the need for human intervention or mechanical systems [68]. Further, the fuel is not pressurized, which eliminates the ability for large amounts of radionuclides to be

A-37 dispersed following an accident [68]. The second is a simplistic design that is intrinsically stable and self-regulating [68]. Similar to other fluoride salt reactors, the fission products of greatest radiological concern are retained in the overall salt mixture and are not dispersed to the atmosphere following a nuclear reactor accident [68]. The third is the design of an efficient nuclear reactor and the fourth is to produce far less waste [68]. The

LFTR is able to capture the energy content of thorium at an efficiency approaching

100%, by utilizing thorium fuel in a thermal (low energy) neutron spectrum where the fission probability is highest, which results in a high operating efficiency and minimizes waste products [68].

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A.5 Fast Neutron Spectrum Reactors

Since 1950, approximately 20 fast neutron spectrum reactors have been designed and operated [76]. Fast reactors use Uranium-238 more deliberately and offer a more efficient use of uranium resources [76]. Further, fast reactors burn actinides, which are otherwise high-level nuclear wastes [76]. Currently being developed are lead and lead- bismuth cooled and gas cooled fast reactors, but sodium cooled remains the most mature fast-reactor technology [77]. Fast reactors present a number of key advantages over thermal (slow-neutron) reactors with respect to waste management and sustainability, with increased efficiency resulting in a reduction of radioactive waste [78]. Fast reactors use no moderator and instead rely on fast neutrons to cause fission. In an effort to avoid neutron moderation and ensure a highly efficient heat transfer medium, the coolant is a molten liquid (lead or lead-bismuth). The following sections discuss fast neutron spectrum SMRs, with Table A-5 serving as a summary.

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Table A-5: Summary of Fast neutron Spectrum SMRs [79, 80, 81, 82, 83, 84, 85, 86, 87] Reactor name: EM2 CLEAR-I ALFRED Electric power (MWe): 240 N/A 125 Thermal power (MWth): 500 10 300 Fuel type: U-Pu-MA UO2 MOX Enrichment (%): 1 19.75 30 Neutron spectrum: Fast Fast Fast Development stage: Concept Design Concept Country of origin: USA China Italy

Reactor name: ELFR PEACER BREST-OD-300 Electric power (MWe): 630 300 300 Thermal power (MWth): 1,500 850 700 Fuel type: MOX U-TRU-Zr PuN-UN Enrichment (%): 13.5 Neutron spectrum: Fast Fast Fast Development stage: Concept Concept Design Country of origin: Italy Korea Russian

Reactor name: SVBR-100 G4M MYRRHA Electric power (MWe): 101 25 N/A Thermal power (MWth): 280 70 100 Fuel type: UO2 Uranium Nitride MOX Enrichment (%): 16.5 19.75 Neutron spectrum: Fast Fast Fast Development stage: Design Concept Concept Country of origin: Russia USA Belgium

Energy Multiplier Module (EM2): This is a helium cooled fast neutron spectrum reactor with a capacity of 240 MWe (500 MWth) [79]. EM2, shown in Figure

A-24, utilizes uranium carbide pellets with an approximate enrichment of 14.5% [2]. The pellets have a high thermal conductivity and melting point [2]. The EM2’s design includes provisions to include passive safety systems that allow for sustained protection in the event of a severe accident [2]. Further, the reactor strives to optimize fuel

A-40 utilization in an effort to minimize waste and optimize efficiency [2]. The safety design includes three barriers against the release of radionuclides [2]. The first barrier is the fuel cladding that encompasses the pellets, which is able to maintain its strength at temperatures far exceeding normal operating conditions [2]. The second barrier is the primary vessel, which encompasses the reactor [2]. The barrier serves as a traditional concrete barrier to prevent the release of radioactive materials to the atmosphere following an accident [2]. The design leakage rate for the barrier is less than 0.2% per day [2]. The third barrier is the below-grade containment, which will serve as an additional physical barrier to the release of radioactive materials [2]. In addition to physical barriers, there exists passive safety systems, including the direct reactor auxiliary cooling system [2]. The cooling system is able to operate during normal shutdown and accident conditions and will passively remove heat without need for external power [2].

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Figure A-24: SmAHTR vessel schematic [88]

China LEAd-based Research Reactor (CLEAR-I): This is a lead bismuth eutectic alloy cooled fast neutron spectrum reactor with a capacity of 10 MWth [80]. It uses uranium dioxide fuel enriched to 19.75% [89]. Lead bismuth has high thermal tolerance and low chemical reactivity properties, meaning that the CLEAR-I has a negative temperature reactivity coefficient and passive heat removal capacity [89]. The primary cooling system is designed to rely entirely on the passive heat removal system through natural circulation [89]. During an accident condition, there exists an air-cooling system to assist with removing decay heat [89]. The air will continue to naturally circulate [89]. In addition to the passive cooling system, there is a double-containment structure erected around the reactor core, which can be seen in Figure A-25 [89].

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Figure A-25: CLEAR-I reactor schematic [90]

Advanced Lead Fast Reactor European Demonstrator (ALFRED): This is a lead cooled fast neutron spectrum reactor with a capacity of 125 MWe (300 MWth) [81].

ALFRED uses mixed oxide fuel with maximum plutonium enrichment of 30% [81]. The reactor itself is a pool-type, with all components being easily removable [81]. The reactor is equipped with two redundant shutdown systems [81]. The first consists of rods that are passively inserted by buoyancy from the bottom of the core and the second consists of rods that are passively inserted via depressurization of a pneumatic system from the top of the core [81]. The reactor is also equipped with two redundant passive heat removal systems [81]. Figure A-26 shows a schematic of the reactor [81].

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Figure A-26: ALFRED vessel schematic [81]

European Lead Fast Reactor (ELFR): ELFR, shown in Figure A-27, is a lead cooled fast neutron spectrum reactor with a capacity of 630 MWe (1,500 MWth) [82].

The ELFR reactor shares many similarities with ALFRED. ALFRED was designed to demonstrate that a lead fast reactor is capable of being used in commercial power plant applications [91]. As such, the ELFR also uses a mixed oxide fuel, has two (2) redundant safety rod systems, and contains two redundant passive heat removal systems [91].

Similar to ALFRED, the first safety rod system is inserted upwards using the buoyancy force of the reactor core [91]. The second safety rod system is inserted downward through the use of a pneumatic system [91]. The first heat removal system is composed of four isolation condenser systems connected to four steam generators [82]. The second system is composed of four isolation condenser systems connected to four dip coolers

[82].

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Figure A-27: ELFR vessel schematic [91]

Proliferation-resistant Environment-friendly Accident-tolerant Continuable and Economical Reactor (PEACER): This is a lead bismuth eutectic alloy cooled reactor with a capacity of 300 MWe (850 MWth) [83]. The reactor, shown in Figure A-

28, employs uranium transuranic zirconium alloy fuel [92]. The lead-bismuth coolant has a high temperature tolerance, which allows the coolant to remain a liquid over a wide range of temperatures and pressures. This allows for the natural circulation of coolant to continue at increased temperatures and pressures [92]. One of the distinguishing features of the PEACER is the reactor vessel air cooling system, which will provide passive cooling capabilities for decay heat [83].

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Figure A-28: PEACER core arrangement [92]

BREST-OD-300: This is a lead cooled fast neutron spectrum reactor with a capacity of 300 MWe (700 MWth) [84]. The reactor, shown in Figure A-29, uses uranium nitride fuel enriched to 13.5% [84]. The BREST-OD-300 has ambitious design objectives, including the elimination of reactor accidents that require evacuation [84].

This will be accomplished through an innovative design that employs a number of inherent safety systems. In fact, the reactor employs a number of natural properties that allow for a high level of inherent safety, including: lead, uranium nitride, and the core.

For example, lead coolant is inert when coming into contact with either air or water, minimizing the risks associated with a loss of coolant accident [84]. Further, the BREST-

OD-300 passively removes heat from the primary circuit via natural circulation [84].

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Figure A-29: BREST-OD-300 Schematic [84]

SVBR-100: This is a lead bismuth eutectic allow cooled fast neutron spectrum reactor with a capacity of 101 MWe (280 MWth) [85]. The reactor, shown in Figure A-

30, uses uranium dioxide fuel enriched to 16.5% [85]. The SVBR-100 meets the IAEA standards for inherent safety and prevention of severe accidents [85]. In doing so, the reactor contains both passive and active emergency safety systems and a passive residual heat removal system [85]. For example, the lead bismuth coolant is chemically inert, which contributes towards mitigating the likelihood of a loss of coolant accident [85].

Further, all primary reactor equipment is located within a single vessel and the reactor operates at near atmospheric pressure, which further assists with mitigating an accident, including radioactive releases to the atmosphere [85].

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Figure A-30: SVBR-100 Schematic [85]

Gen4 Module (G4M): This is a lead bismuth eutectic alloy cooled fast neutron spectrum reactor with a capacity of 25 MWe (70 MWth) [86]. The design philosophy of the Gen4 Module reactor, which is shown in Figure A-31, includes ensuring the protection of the facility and surrounding environment [86]. This is achieved through a sealed core, reactor simplicity, mechanical components, and separation between the power producing sections and conversion sections [86]. While the reactor does not employ emergency safety systems, residual heat is removed passively from the core through heat transfer during the naturally circulation of coolant in the primary and secondary loops [86]. Further, the reactor contains two independent control system to assist with control of reactivity [86].

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Figure A-31: G4M plant layout [86]

Multi-purpose hYbrid Research Reactor for High-tech Applications

(MYRRHA): This is a lead bismuth eutectic alloy cooled fast neutron spectrum reactor with a capacity of 100 MWth [87]. The primary cooling system consists of two pumps and four heat exchangers [87]. The secondary cooling system consists of a water cooling system for the primary heat exchangers [87]. The tertiary cooling system is an air cooling system [87]. In the event of a loss of flow, the primary, secondary, and tertiary cooling systems are designed to manage decay heat passively via natural convection [87]. As heat is generated, the coolant increases in temperature, which results in the coolant rising to the top of the reactor core, where the fluid is cooled via convection from the core via the primary, secondary, and tertiary cooling systems.

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A.6 Summary

Thirty-six prominent SMRs are detailed above. Readers can refer to the IAEA’s

Advanced Reactors Information System [41] and Advances in Small Modular Reactor

Technology Developments [2] for additional information. The review of SMRs included a summary of the passive and inherent safety systems that allow for improved safety.

While there are numerous safety systems in the thirty-six SMRs discussed above, among the most common for water-cooled, gas-cooled, molten-salt and fast-neutron spectrum reactors are the decay heat removal system and the emergency core cooling system, both of which minimize the probability of a loss of coolant accident via the passive removal of decay heat from the reactor core when mechanical components are not operational.

Additionally, there are a number of SMRs that employ multiple containment structures, which will assist with preventing the release of radionuclides to the atmosphere following an accident. As a result, the inherent/passive safety systems and multiple containment structures contribute towards the reduction and/or elimination of the exclusion and emergency planning zones.

The Canadian Nuclear Association commissioned a report to discuss if SMRs are able to play a role in Canadian energy production [93]. This review discusses that, “as part of the design process, an emergency planning zone should be as small as possible.

Utilizing a dose/distance approach, it may be possible to show that SMRs could have a planning zone that is coincident with the site boundary” [93]. The inherent/passive safety systems described above, combined with the resulting doses/distances and other innovative design features of SMRs, allows for the designer to engage with the regulator to justify the elimination or reduction of the exclusion and emergency planning zones.

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