Core neutronics for Space reactors : analysis of HALEU configurations

LAUREA MAGISTRALEUna delle lacune tecnologiche più rilevanti che necessita di essere colmata al fine di supportare l'esplorazione di altri pianeti riguarda sistemi di generazione di potenza in grado di fornire per diversi anni una fonte di energia superiore a 1 kWe agli insediamenti planetari. Tra tutti i possibili sistemi di alimentazione, i reattori nucleari a fissione rappresentano la soluzione più interessante grazie alla loro elevata potenza specifica e alla capacità di produrre energia indipendentemente dalla posizione o dalle condizioni dell' ambiente esterno. Questo lavoro di tesi mira ad analizzare - dal punto di vista neutronico - diverse possibili configurazioni di reattore spaziale a base uranio a basso arricchimento (HALUE), caratterizzato da un contenuto di U-235 compreso tra il 5% e il 20%. Questa scelta consente al progetto di essere conforme alle politiche di anti-proliferazione che impediscono l'uso di uranio con un livello di arricchimento superiore a 20%. Il sistema proposto deve soddisfare i requisiti di sicurezza per l'approvazione del lancio ed essere ottimizzato in termini di massa, il che rappresenta un aspetto fondamentale al fine di ridurre i costi di lancio. La progettazione del reattore viene svolta prendendo come riferimento il design Kilopower proposto dalla NASA, unico reattore per applicazioni spaziali che negli utlimi 40 anni è stato testato con successo attraverso la dimostrazione sperimentale del Kilopower Reactor Using Stirling TechnologY (KRUSTY ) eseguito al Los Alamos National Laboratory (LANL). Il codice Monte Carlo Serpent di simulazione del trasporto di particelle è stato impiegato per svolgere l'analisi neutronica. Come primo passo, in Serpent viene sviluppato un modello di KRUSTY al fine dimostrare la capacità del codice di simulare reattori nucleari per applicazioni spaziali. I risultati del modello vengono confrontati sia con le simulazioni numeriche eseguite direttamente da LANL con un diverso codice neutronico, sia con i dati sperimentali raccolti durante la dimostrazione del corretto funzionamento di KRUSTY. Successivamente, sono state proposte tre possibili opzioni di concetti di reattore a base di HALEU: i) un reattore veloce simile KRUSTY, ma che impiega HALEU invece di HEU come combustibile; ii) un reattore termico il cui il materiale moderatore viene omogeneamente mescolato con il combustibile stesso (reattore omogeneoamente moderato); iii) un reattore termico in cui il combustibile e il materiale moderatore sono forma di dischi alternati (reattore eterogeneamente moderato). Ciascun concetto di reattore è supportato da un'analisi di ottimizzazione della massa e da un'analisi della sicurezza per garantire l'approvazione al lancio. Sul reattore veloce HALEU è stato effettuato uno studio su come il sistema raggiunge le condizioni criticità e una stima delle distribuzioni di densità di potenza. L'analisi svolta sui reattori termici omogeneamente moderati mira a dimostrare l'effetto benefico dell'aggiunta di moderatore all'interno del combustibile sull'economia neutronica del sistema, che si traduce in una riduzione della massa totale del sistema. Infine, viene studiato l'effetto dell'eterogeneità del combustibile, per provare l'esistenza di uno spessore ottimale dei dischi in termini di economia neutronica.One of the most critical technological gaps that need to be filled to support space exploration involving expedition to other planets concerns power generation systems capable of providing for several years a power source bigger than 1 kWe to planetary settlements. Among all the possible power supply systems, nuclear fission reactors represent the most attractive solution, thanks to their high specific power and the ability to produce energy regardless of their location or the external environment. This thesis work aims at analysing - from a neutronics perspective - different possible configurations for a space reactor employing High-Assay Low Enriched Uranium (HALUE), characterized by an enrichment level between 5% and 20%. This choice allows the design to be compliant with proliferation policies that prevent the use of uranium with an enrichment level higher than 20%. Furthermore, the proposed system has to meet safety requirements for launch approval and be optimized in terms of mass, which is of most importance for reducing launch costs. The design of the fission power system is carried out employing Kilopower reactor concept proposed by NASA as a reference, which is the only space reactor that has been recently tested with success through the experimental demonstration of the Kilopower Reactor Using Stirling TechnologY (KRUSTY) performed at Los Alamos National Laboratory (LANL). Serpent particle transport Monte Carlo code has been employed for the neutronic analysis. As first step, a model of KRUSTY is developed in Serpent to demonstrate the capability of the code to simulate nuclear reactors for space applications. The results are compared against both numerical simulation performed by LANL with a different neutronics code and with the experimental data collected from KRUSTY testing. Afterward, three possible design paths for HALEU reactor concept are investigated: i) a fast reactor just like KRUSTY, with the exception of using HALEU instead of HEU; ii) a homogeneously moderated thermal reactor, whose core is a homogeneous mixture of moderator and fuel; iii) a heterogeneously moderated thermal reactor, whose core is composed of separate layers of fuel and moderator. Each reactor concept is accompanied by a mass optimization analysis and safety analysis. The HALEU fast reactor is also supported by a study on how the system approaches criticality and estimations of power density distributions. The analysis performed on homogeneously moderated reactors is treated as an intermediate step to demonstrate the beneficial effect of adding moderator inside fuel, which turns into an improvement in neutron economy. Finally, the effect of heterogeneity is investigated to prove the existence of an optimal fuel cell pitch that permits to maximize reactor performance in terms of neutron economy

BUILDING RESILIENT COMMUNITIES: THE ROLE OF HYBRID NUCLEAR ENERGY SYSTEMS IN MODERNIZING THE U.S. ELECTRIC GRID

The U.S. electric grid is ill-prepared for the challenges of the next two decades, including outdated infrastructure, the need for decarbonization, increases in electricity demand, possible attacks from bad actors, and damage from natural disasters. Thus, it represents an enormous vulnerability for homeland security. This thesis explores the implementation of hybrid nuclear energy systems (HNESs) as the solution to these challenges with specific questions about which feasible technologies and grid architecture would increase resilience the most and what role homeland security leaders should play in their implementation. An analysis of history, economics, international comparisons, and technological advances reveals that HNESs would be most resilient if they were structured as continuously islanded integrated community energy systems (ICESs) in a distributed architecture and powered both by non-light-water small modular nuclear reactors running on tri-structural isotropic (TRISO) fuel and a mix of other distributed energy resources (DERs). Each ICES should be configured with medium voltage power electronics serving as a back-to-back direct-current tie interconnection with adjacent grids. Additionally, each ICES should be optimized utilizing (1) principles of common-pool resources and (2) a digital twin. This thesis further recommends that homeland security leaders educate themselves as modern utility customers to effectively navigate the current utility landscape.Distribution Statement A. Approved for public release: Distribution is unlimited.Civilian, City of Santa Fe Fire Departmen

MCNP analysis of fuel irradiation in varying regions of the advanced test reactor

In this study, irradiation testing principles are explored in support of current nuclear thermal propulsion and advanced nuclear fuel development. Using the radiation-transport tool MCNP, three types of cylindrical fuel specimens are modeled and analyzed. The three different fuel types selected for this study include UO2, UN, and UC. Cladding is omitted from the models allowing the behavior of the wrought fuel material to be the main phenomenon explored. Using MCNP tallies, details such as the average neutron flux and neutron current are obtained. Reaction rate calculations are also performed in MCNP to provide insight on details like the fission reaction rate. When combined with the known fast and thermal fluxes in the main flux trap, “B” positions, and “I” positions of the Advanced Test Reactor (ATR), the rate of fission reactions per second is calculated. The average neutron heating rate is also explored to provide insight on power generation in the fuel specimens. For a desired burnup, the required irradiation time is calculated for each fuel type and ATR irradiation testing location combination. The results confirm that the main flux traps reduce the total amount of irradiation time required to meet a desired burnup. “I” positions required an extended irradiation time due to their location outside of the ATR control drums. Furthermore, increased fuel enrichment decreased the irradiation time necessary to meet a specified burnup. Combined, this analysis provides a methodology for planning, modeling, and analyzing irradiation testing campaigns based on ATR testing location, fuel type, enrichment, and burnup. Future areas of work that would expand upon this effort are also described herein.Mechanical Engineerin

Study of a nuclear thermal thruster for interplanetary manned missions

This study delves into the development and analysis of nuclear space propulsion systems for future missions, drawing valuable insights from past experiences and the historical context of these systems. Nuclear Thermal Propulsion (NTP) emerges as a promising option, offering several advantages over alternative propulsion systems, such as reduced propellant mass per payload and the potential for shorter travel times to Mars and near-Earth objects. The extensive research conducted demonstrates that NTP engines possess key attributes, including energy efficiency, high specific impulse, and extended operational capabilities, making them particularly well-suited for Mars missions. Furthermore, the study places emphasis on the meticulous selection of materials for the primary components of the NTP system. Factors such as high-temperature strength, corrosion resistance, and pressure resistance are carefully considered in this analysis, ensuring the optimal performance and reliability of the system. Besides, in order to compute the maneuver and the mission to Mars, it is necessary compute the specific impulse ��� of the engine. Therefore, to ascertain the most accurate approach for calculating specific impulse (���), two methodologies, Chemical Equilibrium and Frozen Composition, are compared using the CEARUN software. The results highlight the distinct characteristics of each method, with Chemical Equilibrium assuming certain properties and Frozen Composition incorporating various factors, leading to variations in the calculated ���. Once the ��� is selected, the mission to Mars is carefully modelled and analyzed utilizing computational tools such as MATLAB. This comprehensive analysis encompasses diverse propulsion maneuvers and trajectory optimization, shedding light on the potential benefits and considerations associated with the implementation of NTP systems in space missions. In conclusion, NTP systems have the inherent capacity to significantly reduce travel durations in space missions owing to their augmented thrust capability and high specific impulse, ultimately resulting in a more efficient utilization of propellant. However, it is important to underscore that fully harnessing the benefits of NTP systems in future space missions necessitates further research and optimization efforts. By continuously exploring and refining these systems, it is possible unlock their full potential and propel humanity towards new frontiers in space exploration

The National Criticality Experiments Research Center and its role in support of advanced reactor design

The National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS) in the Device Assembly Facility (DAF) and operated by Los Alamos National Laboratory (LANL) is the only general purpose critical experiments facility in the United States. Experiments from subcritical to critical and above prompt critical are carried out at NCERC on a regular basis. In recent years, NCERC has become more involved in experiments related to nuclear energy, including the Kilopower/KRUSTY demonstration and the recent Hypatia experiment. Multiple nuclear energy related projects are currently ongoing at NCERC. This paper discusses NCERC’s role in advanced reactor design and how that role may change in the future

전자후방산란회절 특성화를 통한 원자로급 지르칼로이 핵연료 피복재의 수소화물 상호연결성 연구

학위논문(박사) -- 서울대학교대학원 : 공과대학 에너지시스템공학부, 2023. 8. 이유호.핵연료 피복재에 영향을 미치는 다양한 열화 메커니즘 중 수소유발 취성은 가장 주요한 열화 인자로 알려져 있습니다. 수십 년 동안 수소화물 취성 및 지르코늄 합금의 기계적 특성 저하에 대한 광범위한 연구가 수행되었으나 지르코늄 합금의 전자후방산란회절(EBSD) 특성화의 어려움은 수소화물 형태와 거동에 대한 더 깊은 이해를 방해했습니다. 이 연구는 Zircaloy 의 EBSD 분석에 대한 그러한 한계를 극복했으며, 이를 통해 이전에 탐구되지 못했던 Zircaloy 재료의 입자 수준 특성과 수소화물 상호 연결성을 포괄적으로 이해할 수 있게 되었습니다. 본 연구는 다음과 같은 통합 접근법을 활용합니다. 첫째, 수소화물 특성화를 위해 EBSD 특성화 기술을 활용하였습니다. 둘째, 수소화물 취성과 관련된 기계적 특성을 분석하기 위해 고리 압축 실험, 강도 및 경도 측정과 같은 실험들을 수행했습니다. 셋째, 수소화물 결합 이론을 이용하여 실험 결과를 이론적으로 해석하였으며, 열역학 모델을 개발하였습니다. 마지막으로, 핵연료 코드 시뮬레이션을 사용하여 연료 거동을 예측하고 사용후핵연료 규제에 미치는 영향을 분석하기 위해 크리프 변형과 같은 추가적인 열화 거동을 조사했습니다. 본 연구는 Zircaloy-4 및 ZIRLO 합금 피복재의 수소 유발 취성에 미치는 물질 미세구조, 특히 수소화물 연결성의 영향을 조사하였습니다. EBSD 분석과 열역학 모델링을 통해 두 합금에서 급격한 연성-취성 전이 현상의 차이가 주로 Zircaloy-4 튜브의 약간 큰 결정립 지름에 기인한다는 것을 밝혀냈습니다. 결정립의 크기가 커지면 입계 수소화물 석출을 위한 가능한 위치가 줄어들고, 상호연결된 수소화물 네트워크의 형성이 방지되어 수소화물 취화에 대한 저항성을 향상시킵니다. 또한, 소둔 온도가 증가함에 따라 입계 수소화물 연결성이 감소하는 잠재적인 메커니즘을 확인했습니다. 주요 원인은 입내 수소화물 형성으로 인해 입계 수소화물의 연결성이 낮아지며, 결정립계 에너지의 균질화에 의해 에너지적으로 수소화물 석출이 선호되는 결정립계의 집중도가 낮아지는 것이었습니다. 수소 함량이 높은 피복재에서는 소둔을 통한 수소화물 연결성 감소가 연성 증가에 크게 기여함을 기계적 평가를 통해 입증하였습니다. 또한, ZIRLO 합금의 소둔을 통해 재결정화와 결정립 성장을 조사하였으며, 장기적인 소둔시에 450°C 에서도 상당한 결정립 성장이 됨을 관찰하였습니다. 또한 본 연구에서는 SNF 취급 및 운송 중 피복재에 대한 충격 에너지를 평가하기 위한 충격 시험 장치를 개발하였습니다. 광범위한 EBSD 분석을 통해 강화된 재료 미세구조에 대한 수소화물 석출의 민감도에 대한 이해를 바탕으로 핵연료 피복재의 용접영역에서 억제된 수소화물 석출이 설명되었습니다. 본 연구를 통해 개발된 열역학 모델링은 수소화물 석출에 영향을 미치는 주요 인자로 매트릭스 강성, 부적합 변형률, 입계 배향각 등 EBSD 분석을 통해 얻은 데이터를 활용합니다. 또한 이 연구는 수소화물 핵형성과 관련된 다양한 요인들을 고려하여 수소화물 핵형성 거동을 이해하기 위한 프레임워크를 제공합니다. 이 프레임워크는 수소화물 침전의 강화 또는 지연 효과에 영향을 미치는 요인에 대한 포괄적인 이해를 제공하는 성과를 거두었습니다. 더불어, 수소화물의 석출 거동에 대한 증진된 이해를 바탕으로 코드 시뮬레이션을 수행하여 사용후핵연료의 현재 거동에 대한 규제적 시사점을 검증하였습니다. 본 연구는 FRAPCON-SNUSF 코드를 사용하여 건식 저장 상태에서 다양한 연소도의 사용후핵연료의 거동을 시뮬레이션 하였습니다. 시뮬레이션 결과는 현재의 건식 저장 온도 규제 한계인 400°C 가 적절하며, 현재 연소도 60 MWd/kgU 에서 최대 20°C 이상의 안전 여유를 제공한다는 것을 시사합니다. 이 연구는 사용후핵연료 관리에서 수소화물 취성에 대한 이해를 기반으로 건식 저장 관리 전략에 미치는 영향을 조사하였습니다. 전반적으로, 본 연구는 국내 원자력산업에서 중요한 이슈인 사용후핵연료 피복재의 기계적 건전성 평가에 기여하였습니다.Among various degradation mechanisms affecting nuclear fuel cladding, hydrogen-induced embrittlement is known to be major degradation factor. For decades, extensive studies have been conducted on hydride embrittlement (HE) and mechanical property degradation of zirconium alloys. However, difficulties in EBSD characterization of zirconium alloys have hampered a deeper understanding of hydride morphology and behavior. This study overcomes the limitations of Zircaloy's EBSD analysis, and this breakthrough has led to a comprehensive understanding of the previously unexplored grain-scale characteristics of Zircaloy materials and hydride interconnectivity. This study utilizes the following integrated approaches. Firstly, EBSD characterization technology was utilized to conduct hydride characterization. Secondly, tests such as ring compression, strength, and hardness measurements were performed to analyze the mechanical properties associated with hydride embrittlement. Thirdly, the experimental results were theoretically interpreted using the hydride interlinked theory, leading to the development of a thermodynamic model. Lastly, nuclear fuel code simulation was employed to forecast fuel behavior, while also examining additional degradation mechanisms like creep deformations, with the aim of analyzing their impact on spent fuel regulation. This study investigates the effect of material microstructure, especially hydride connectivity, on hydride embrittlement of Zircaloy-4 and ZIRLO alloy claddings. EBSD analysis and thermodynamic modeling attribute the difference in the abrupt DTB transition in the two alloys primarily to the slightly larger grain diameter of the Zircaloy-4 tube. Increasing grain size reduces the sites available for intergranular hydride precipitation, preventing the formation of an interconnected hydride network, improving brittle resistance. Potential mechanisms for decreasing intergranular hydride connectivity with increasing annealing temperature were identified, including the formation of intergranular hydrides and homogenization of intergranular energies. Reduced hydride connectivity has been demonstrated to significantly increase the strain energy density of claddings with high hydrogen content by improving cladding ductility. The recrystallization and grain growth of ZILRO alloy through annealing are investigated, and significant grain growth is observed even at 450 °C during long-term annealing. Additionally, this study develops an impact test device to evaluate the impact energy to cladding during SNF handling and transportation. Based on our understanding of the sensitivity of hydride precipitation to material microstructure enhanced by extensive EBSD analysis, suppressed hydride precipitation in the weld zone region of fuel cladding was explained. The thermodynamic modeling developed through this study utilizes data obtained through EBSD analysis, such as matrix stiffness, misfit strain, and grain boundary orientation angle, as major factors influencing hydride precipitation. This study provides a framework for understanding hydride nucleation behavior by considering several factors involved in hydride nucleation: nucleation sites such as grain size and grain boundary density, grain boundary energy, and critical nucleation energy based on matrix stiffness and misfit strain. This framework provides a comprehensive understanding of the factors that influence the enhancing or retarding effect of hydride precipitation. In addition, this study also verifies the regulatory implications of the current behavior of spent nuclear fuel through code simulation based on the understanding of the precipitation behavior of hydrides. FRAPCON-SNUSF, a modified version of FRAPCON-4.0, was used to simulate the behavior of spent fuel under dry storage at various burnups. At different burnup levels, the margin of safety is examined in terms of hoop stress thresholds for creep rupture and hydride reorientation. Simulation results suggest that the current dry storage temperature limit of 400 °C is adequate and provides a safety margin of up to +20 °C at the current burnup level of 60 MWd/kgU. By conducting these research studies, a holistic comprehension of various aspects concerning commercial nuclear fuel rods can be acquired. This includes an extensive understanding of the entirety of fuel rods utilized in commercial settings, encompassing tubes and end cap weld zone. Furthermore, these researches enable a deeper insight into the phenomena of hydride precipitation and embrittlement, elucidating their correlation with the microstructural characteristics of the materials involved. Based on understanding hydride embrittlement via EBSD characterization, this study investigates the impact on dry storage management strategies. Overall, this study contributed to the evaluation of the mechanical integrity of SNF cladding, a key issue in the Korean nuclear power industry.Chapter 1. Introduction 1 1.1. Study background 1 1.1.1. Necessity for interim storage of spent nuclear fuel and preparation for integrity evaluation 1 1.1.2. Hydride embrittlement, major degradation mechanism of the cladding 3 1.1.3. EBSD technique on Zircaloy 5 1.2. Purpose of research 6 Chapter 2. Understanding of hydride embrittlement mechanism via hydride connectivity characterization 11 2.1. Hydride embrittlement resistance of Zircaloy-4 and ZIRLO alloy cladding tubes and its implications on spent fuel management 11 2.1.1. Introduction 11 2.1.2. Experimental setup 14 2.1.3. Results and discussion 22 2.1.4. Conclusions 43 2.1.5. Supplementary 45 2.2. Recrystallization and grain growth of ZIRLO alloy in 400 500 C and effect on hydride embrittlement 49 2.2.1. Introduction 49 2.2.2. Experimental procedure 50 2.2.3. Results 53 2.2.4. Discussion 64 2.2.5. Conclusions 85 2.2.6. Supplementary 87 2.3. Impact test on spent nuclear fuel cladding 91 2.3.1. Introduction 91 2.3.2. Experimental setup 93 2.3.3. Results and discussion 103 2.3.4. Conclusion 113 Chapter 3. Hydride precipitation thermodynamic model development and understanding on the hydride embrittlement resistance 115 3.1. Introduction 115 3.2. Experimental setup 117 3.3. Results 120 3.3.1. Hydride formation analysis (Hydride-free weld zone) 120 3.3.2. Microstructural characterization (EBSD) 122 3.4. Discussion: thermodynamic interpretation 126 3.4.1. Theoretical background 126 3.4.2. Misfit strain 129 3.4.3. Zr-Matrix stiffness (shear modulus) 131 3.4.4. Theoretical calculation of 134 3.4.5. Grain boundary energy at the WZ and the tube 135 3.4.6. Theoretical calculations of hydride nucleation rates in the WZ and tube 138 3.5. Conclusions 143 3.6. Supplementary 144 Chapter 4. Revisiting the current SNF regulations for dry storage via FRAPCON-SNUSF simulation 165 4.1. Introduction 165 4.2. Background 167 4.3. SNF behavior at a discharge burnup of 60 MWd/kgU 169 4.3.1. FRAPCON-SNUSF code 169 4.3.2. Hydrogen concentration and hydride precipitation 171 4.3.3. Cladding temperature 173 4.3.4. Creep rate 174 4.3.5. RIP and cladding hoop stress 175 4.3.6. Cladding hoop strain and diameter deformation 178 4.3.7. Analysis of the safety margin at a discharge burnup of 60 MWd/kgU 180 4.4. Analysis of the SNF behavior and safety margin at higher discharge burnups (~80 MWd/kgU) 183 4.4.1. Spent fuel behavior at higher discharge burnups (~80 MWd/kgU) 184 4.4.2. Analysis of the safety margin at higher discharge burnups (~80 MWd/kgU) 188 4.5. Conclusions 190 4.6. Supplementary 192 Chapter 5. Conclusions 199 Bibliography 203 Appendix 229 Appendix for Policy research 229 Appendix A.1. Analysis of supply characteristics of enriched uranium and LEU+/HALEU supply and demand trends: Korea's LEU+ and HALEU Supply Strategy 229 Appendix A.2. Highly Enriched Uranium (HEU) Supply Plan for Next-Generation Reactors: Promoting Cooperation with the US 273 Abstract in Korean (국문 초록) 299 Acknowledgement 303박

Vision of Nuclear Power Deployment in Latin America and the Caribbean: A Focus on Small Modular Reactors and the Regional Experience of Central Argentina de Elementos Modulares

Increasing the number of nuclear power reactors in the Latin American and Caribbean region presents technical, financial, regulatory, and environmental challenges. Focused on fostering economic stability, growth, and human capacity development, the deployment of small modular reactors (SMRs) emerges as a key aspect in the region’s energy landscape. The emergence of SMRs represents an opportunity for multidisciplinary cooperation among different sectors. To comprehensively address the challenges related to the protection of nuclear facilities in the region, the Tlatelolco Treaty and the Non-Proliferation Treaty should be strengthened as legally binding instruments to enforce the safety and safeguarding principles integral to the peaceful applications of nuclear energy. Financial, technical, human resource, and digital challenges must also be addressed to enable opportunities for integrating nuclear technology into the region’s power grids. This article serves as a call of action to collaborate on harnessing the power of nuclear energy while championing gender diversity and international cooperation for the benefit of the region’s socioeconomic landscape. Drawing from the exemplary Central Argentina de Elementos Modulares (CAREM) project in Argentina, which showcases the integration of SMRs into the national energy mix, this paper underscores the multiple aspects that the SMR market needs to address for deployment in the region. This study provides an overview of the region\u27s energy and nuclear landscape, highlighting challenges and opportunities in various sectors. It provides introductory elements for planning a strategy for the use of nuclear power in the region, emphasizing joint cooperation among Latin American and Caribbean countries to pave the way for a sustainable and prosperous future driven by a clean, safe and reliable energy solution. In particular, CAREM-25 is a 25 MWe prototype SMR based on pressurized water reactor technology and was designed and launched by the National Atomic Energy Commission of Argentina (CNEA). The project is under construction and has an estimated criticality date of late 2027. The Argentine case also provides outstanding examples of women in leadership positions, such as General Manager of the CAREM project Dr. Sol Pedre and former President of the CNEA Dr. Adriana Serquis

A Multivariant Comparative Analysis Of Small Modular Nuclear Reactors And Their Role In Net Zero 2050

Climate Change is one of the world’s most urgent issues, and the ideas to reduce the emission of Greenhouse Gases (GHG) from manmade sources needed to help resolve this plight. Globally, the production of electricity accounts for nearly 40% of all GHG emissions in 2022. The substitution of nuclear energy can reduce that number greatly. SMRs, or small modular nuclear reactors, are smaller than conventional nuclear reactors. They can be made on-site, assembled, and transported to their final destination. SMRs offer many advantages over traditional reactors. They are more cost-effective, safer, and more flexible.SMRs could be used to replace fossil fuel-based power generation which is a significant contributor to global greenhouse gas emissions. SMRs can generate electricity with nuclear power and help reduce carbon emissions. This could be a step towards achieving net zero emissions by 2050. This means they can be used in a wider variety of applications, such as in remote areas or in microgrids. There are challenges that come with the deployment and use of SMRs. There are regulatory and licensing hurdles, public approval, and financing. Many countries and companies have begun to invest in SMR technology despite these obstacles. This is a way to meet future energy needs while reducing carbon emissions. This dissertation will rank various SMRs using the Technique for Order of Preference by Similarity to Ideal Solution (TOPSIS) method and will examine how the SMRs are able to aid in reducing GHG emissions by replacing high GHG emitters. With this substitution we will discuss quantitatively how these changes will be able to help the world meet the Paris Agreement’s Net Zero 2050 aims

Investigation of the impacts of deploying reactors fueled by high-assay low enriched uranium

The United States is considering the deployment of advanced reactors that require uranium enriched between 5-20% 235U, often referred to as High Assay Low Enriched Uranium (HALEU). At the present, there are no commercial facilities in the US to produce HALEU, prompting questions of how to create a dependable supply chain of HALEU to support these reactors. HALEU can be produced through two primary methods: downblending High Enriched Uranium (HEU) and enriching natural uranium. The amount of HEU available and impurities present in the HEU limit downblending capabilities. The Separative Work Unit (SWU) capacity and amount of natural uranium available limit enriching natural uranium capabilities. To understand the resources necessary to commercially produce HALEU with each of these methods, one can quantify the material requirements of transitioning to HALEU-fueled reactors. In this dissertation, we model the transition from Light Water Reactors to different advanced reactors, considering once-through and closed fuel cycles to determine material requirements for supporting these fuel cycles. Material requirements of interest across this work include the mass of enriched uranium, mass of HALEU, feed uranium, SWU capacity, and the mass of used fuel sent for disposal. We use CYCLUS and publicly-available information about Light Water Reactors, the X-energy Xe-100, the Ultra Safe Nuclear Corporation Micro Modular Reactor, and the NuScale VOYGR to model potential transition scenarios and demonstrate the methodologies developed in this work. To more accurately model the closed fuel cycles, we develop a new CYCLUS archetype, called OpenMCyclus, that couples with OpenMC to dynamically model fuel depletion in a reactor and provide more accurate used fuel compositions. The results of this transition analysis show how the characteristics of the advanced reactors deployed drive the materials required to support the fuel cycle. Closing the fuel cycle reduces the materials required, but the reduction in materials is driven by the amount of material available for reprocessing. To gain more insight into how transition parameters not considered in the transition analysis affect material requirements, we perform sensitivity analysis on one of the once-through transitions by coupling CYCLUS with Dakota. The results of the sensitivity analysis highlight some of the trade-offs between different reactor designs. One such tradeoff is the increased HALEU demand but decreased used fuel discharged when increasing the Xe-100 deployment and decreasing the VOYGR deployment. Additionally, these results identify the Xe-100 discharge burnup as consistently being one of the most impactful input parameters for this transition, because of how the deployment scheme in this work affects the number of Xe-100s built no matter which advanced reactor build share is specified. To identify potential transitions that minimize material requirements, we then use the CYCLUS-Dakota to optimize a once-through transition using the genetic algorithms in Dakota. In single-objective problems to minimize the SWU capacity required to produce HALEU and minimize the amount of used nuclear fuel, the algorithm finds solutions that are consistent with the results of the sensitivity analysis. The results cannot be taken at face value, because the algorithm did not fully converge and the genetic algorithms do not enforce the applied linear constraint for the advanced reactor build shares to sum to 100%. However, the results provide guidance on how to adjust the input parameters to optimize the transition for a minimal HALEU SWU or the used fuel mass. Parameter adjustments include maximizing the number of Light Water Reactors that receive license extensions to operate for 80 years. Similar results occur when using this method for a multi-objective problem to minimize both the HALEU SWU capacity and the used fuel mass. Finally, we use neutronics models of the Xe-100 and Micro Modular Reactor reactor designs to evaluate the steady-state reactor physics performance of downblended HEU in these two designs. We compare the performance of the downblended HEU to nominally enriched fuel, based on the k-eff, βeff, energy- and spatially-dependent neutron fluxes, as well as the fuel, moderator, coolant, and total reactivity temperature feedback coefficients. The differences in the fuel compositions leads to differences in each of the metrics. However, these differences are within error of the results of the nominally enriched fuel, or would not prevent the reactor from meeting stated design specifications or operating in a safe state. The work completed in this dissertation develops and demonstrates a methodology for modeling fuel cycle transitions and understanding the effects of deploying HALEU-fueled reactors in the US. The effects investigated in these example scenarios include various materials and resources required to support these reactors, and how the parameters of the transition affect these requirements. The information generated from this new methodology can be used to develop the necessary infrastructure and supply chains for support a transition to HALEU-fueled reactors. Furthermore, this work explores how the HALEU production method (enriching compared with downblending) affects reactor performance.Submission original under an indefinite embargo labeled 'Open Access'. The submission was exported from vireo on 2024-03-01 without embargo termsThe student, Amanda Bachmann, accepted the attached license on 2023-11-14 at 09:59.The student, Amanda Bachmann, submitted this Dissertation for approval on 2023-11-14 at 10:21.This Dissertation was approved for publication on 2023-11-20 at 16:44.DSpace SAF Submission Ingestion Package generated from Vireo submission #19921 on 2024-03-01 at 13:14:3