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DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗

DOE OSTI2018

Development of Novel Functionally Graded Transition Joints for Improving the Creep Strength of Dissimilar Metal Welds in Nuclear Applications

Steam generators of the small modular Very High Temperature Reactor (VHTR) will require use of Alloy 800 and 2.25Cr-1Mo steel in different regions depending on temperature and corrosion resistance requirements. As a result, Dissimilar Metal Welds (DMWs) between these two materials are critical for design, development, and manufacturing of the VHTR. DMWs are also used in many other applications in the power generation and petrochemical industries. Experience has demonstrated that failures of such DMWs can occur prematurely well below the expected creep life of either base metal, and these failures have significant consequences in terms of plant reliability, safety, and economic impact. These DMWs failures are caused by the abrupt change in composition that occurs at the weld interface between the two alloys. In this research, the concept of developing and optimizing Graded Transition Joints (GTJ) was developed through an integrated modeling and experimental approach. With GTJs, the composition is gradually graded between the two alloys, thus eliminating the sharp change in composition that causes failure. Microstructural modeling was first conducted to identify the optimum grade length required for improved resistance to carbon diffusion that is largely responsible for premature failure. Next, GTJs were fabricated and characterized in detail to understand how the microstructure evolves during fabrication and long term aging. Finally, the creep behavior of the GTJs was investigated and compared to that of conventional DMWs to demonstrate the superior creep strength of the GTJs. The results of this research provide significant insight into design, fabrication, microstructural evolution, and resultant performance of GTJs that can be used for reducing or eliminating DMW failures in a number of high temperature applications. The results also demonstrate that the GTJs can be readily fabricated with either conventional or additive manufacturing processes, and detailed guidelines are provided on the length of grades needed for optimal performance. The results of this research benefits the general public by improving the reliability of power generation and petrochemical plants that rely on DMWs for safe and reliable operation to deliver power and other products.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Webinar: Microreactors—Nuclear Power in a Small, Modular Package

Power Point presentation on Nuclear Reactor to children.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2023

PRO-X Fuel Cycle Transportation and Crosscutting Report: Avoidability Task

As the frequency and quantities of nuclear material shipments escalate internationally to meet the increased demand for small modular (SMR) and advanced (AR) reactors, the risks and costs associated with shipping activities also likely to increase with them. The primary objective of this study is to evaluate possibilities for risk reduction via avoidability—or, avoiding or reducing the need for nuclear shipments, where possible, either by reducing the frequency or quantities of materials contained in shipments.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2023

The research challenges in security and safeguards for nuclear fission batteries

This paper discusses the nuclear security and safeguards research challenges presented by the development and deployment of nuclear fission batteries. These are defined as easily transportable and deployable nuclear systems which are designed to operate either unattended or autonomously. We start by defining the current landscape of domestic and international safeguards and security and discuss how it can be affected by the introduction these new nuclear systems. We then specifically discuss the technology gaps and technologies to be developed to facilitate their practical deployment. We find that, as expected, we can leverage conclusions from existing security and safeguards studies for small modular rectors and develop bespoke target set analyses to inform security postures based on probabilistic risk assessment. We find also that specific fission battery security economic analysis tools are needed and security by design must be applied early. However, the most important finding is that these new systems will require a new comprehensive set of cyber tools covering transportation, installation, operation, maintenance security of the fission battery systems. Altogether, we find that the development of the necessary security and safeguards requirements at the design phase of such a technology will be of great benefit in the smooth deployment of this modern nuclear energy system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Parametric Study of Factors that Affect Calculated Dose from TRISO Fueled Microreactor Transportation Accident

Microreactors are very small nuclear reactors with a power output up to 50 megawatts electric (MWe) which also meet the standards of an “advanced nuclear reactor.” Microreactors are designed to be factory-built, modular in nature, and may be transportable. These compact small-sized reactors in some design configurations are small enough to be transported by truck which could help solve certain energy challenges. This report refers to these reactors as transportable microreactors or transportable nuclear power plants (TNPPs). Transportable microreactors are being considered for a wide range of independent operation applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Development of a reactor with carbon catalysts for modular-scale, low-cost electrochemical generation of H 2 O 2

The development of small-scale, decentralized reactors for H 2 O 2 production that can couple to renewable energy sources would be of great benefit, particularly for water purification in the developing world. Herein, we describe our efforts to develop electrochemical reactors for H 2 O 2 generation with high Faradaic efficiencies of >90%, requiring cell voltages of only ~1.6 V. The reactor employs a carbon-based catalyst that demonstrates excellent performance for H 2 O 2 production under alkaline conditions, as demonstrated by fundamental studies involving rotating-ring disk electrode methods. Finally, the low-cost, membrane-free reactor design represents a step towards a continuous, modular-scale, de-centralized production of H 2 O 2 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

SCALE Analysis of a Fluoride Salt-Cooled High-Temperature Reactor in Support of Severe Accident Analysis

As part of a US Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with SCALE and MELCOR, SCALE was used for the modeling and simulation of a fluoride salt-cooled high-temperature reactor (FHR). Based on the preconceptual design for a small modular 236 MWth FHR developed by the University of California, Berkeley (PB-FHR-Mk1), a SCALE model of the PB-FHR-Mk1 reactor core was developed. The reactor was modeled at equilibrium state with different fuel compositions in different regions of the reactor. An iterative approach was used to interpolate and mix the burnup-dependent fuel compositions obtained through the depletion calculation of a core slice model. After demonstrating the applicability of SCALE’s multigroup (MG) approach for the simulation of the PB-FHR-Mk1, the resulting equilibrium core was studied in terms of the power profile, the flux profile, temperature reactivity coefficients, and the xenon reactivity. Furthermore, the tritium production rate in the salt coolant was determined, and the dependence of the one-group cross sections on the burnup and location in the reactor core was studied. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the equilibrium core, a zone-wise power profile, temperature feedback coefficients, the tritium production rate, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Material Control & Accountancy for Molten Salt Reactors (FY2021 Report)

There is significant domestic and international interest, investment, and research and development momentum to pursue advanced nuclear reactor technologies. Molten salt reactor (MSR) concepts display the largest variability in fuel type and design features among the current advanced concepts. MSRs have been proposed with various core designs, sizes (power), and fuel cycles. Salt-fueled molten salt systems represent the only advanced reactor type with fuel that is not in a solid form during operation. These “liquid-fueled” MSRs are unique from perspectives of fuel fabrication, spent irradiated fuel and waste components, licensing, and material control and accountability (MC&A) including the potential of fissile material holdup. The liquid fuel salt is the defining distinction in comparison to other advanced reactors that propose TRI-structural ISOtropic particle fuel pebbles, various coolant options (e.g., molten salts or metals, high temperature gas), or small modular alternatives using solid fuel variants including both light water reactors and non-light water reactors. MSRs are appealing to the nuclear energy industry because of the diverse reactor characteristics they can support including various neutron energy spectra, fueling requirements, fuel cycles, and/or fuel utilization. However, because of the significant deviation and diversity of a salt-fueled system compared to traditional solid fuel light water-cooled reactors (LWRs), the history, regulatory licensing framework, modeling capabilities, and supporting engineering technology are either lacking or, in some cases, nonexistent. Therefore, the research community is actively supporting advanced MSR development on many of these fronts in particular to assist MSR vendors with licensing requirements. ORNL is leading the research and development of respective MC&A approaches for salt-fueled MSRs. This report summarizes the research performed at Oak Ridge National Laboratory (ORNL) under the US Department of Energy, Office of Nuclear Energy, Advanced Reactor Safeguards (ARS) program to investigate safeguards and security by design concepts, licensing and regulatory considerations, and dynamic system-level modeling to understand radioisotope concentrations for salt-fueled MSRs. The report builds upon the previous research and literature, identifies the MC&A challenges inherent to a salt-fueled MSR, reviews current regulatory frameworks for LWRs and their applicability towards salt-fueled MSRs, summarizes the status and progress of an MSR dynamic modeling tool, and discusses a prospective MC&A approach based on the Molten Salt Demonstration Reactor (MSDR) model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

A study of improved MHR 50/100 toward a higher level of inherent safety

A new concept of the Mitsubishi small-sized High temperature gas-cooled modular Reactors (MHR-50/100) had been developed and published in papers. The study results of the first design concept show that the MHR-50/100 can achieve the inherent safety level set as a design target in case of water ingress during steam generator tube rupture accident. And more specifically, the reactor was shown to remain stable during long-term station black out (SBO) with protection of only passive devices during a depressurization accident and with additional motion of steam dump system during a water ingress accident. Recently greater requirements for safety of future nuclear plants including the MHR-50/100 have been expected. This study has thus made a key design improvement for the MHR-50/100 in order to secure the inherent safety aspect without reliance on active steam dump system in case of a water ingress accident. The innovative technologies listed below have been created and investigated to achieve the improved MHR-50/100 design; (1) Design improvement of steam generator, (2) Heat balance optimization of steam cycle, (3) Control system design of differential pressure between primary helium gas and water/steam, (4) Study on operation procedure during a water ingress accident. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Thermal Energy Storage Selection for Near Term Nuclear Integration

Thermal energy storage (TES) integration with existing and future light water reactors (LWRs) has the capability to increase the deployment options and uses for the large amounts of heat produced in nuclear reactor cores. The present method of exclusively producing electricity has become less economical with increases in renewable energy penetration leading to times of low or negative electricity prices. Some response to this new grid reality has been to begin load following nuclear reactors, an unideal choice due to the relatively small modular cost of producing nuclear energy compared to the revenue that could be received by using all of the heat produced. To support research in this area, Idaho National Laboratory (INL) has joined with Utah Associated Municipal Power Systems (UAMPS) on the Carbon Free Power Project (CFPP) to contract the Joint Use Modular Plant (JUMP). CFPP is planned to be a 12-module NuScale operating site built on INL property, with the first module dedicated to JUMP for research, development, and demonstration purposes. The JUMP module coming online around the beginning of fiscal year 2027 in the current project timeline. One primary goal of JUMP is to demonstrate TES integrated with an operating LWR in a real-world environment. TES integration will allow for the demonstration of nuclear power’s capability to participate in grid load following without reactor cycling, process heat integration, and electrical ancillary market opportunities. The goal of this research is to support both SMR deployment opportunities as well as support LWR retrofit capabilities. In a discussion limited to load following or energy arbitration, “charging” and “discharging” can be simply defined. TES charging refers to time when energy is removed from the nuclear reactor power conversion cycle and directed to be stored in the TES storage medium resulting in less available energy available for grid output. Discharging refers to the process of releasing energy stored in the TES medium and having an increase in grid. This paper quantifies energy storage technologies for use with nuclear reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Sensitivity Analysis of Passive Heat Removal Capability of FHR Based on Efficient Surrogate Model and GenFlow

Code Scaling, Applicability and Uncertainty Evaluation Methodology (CSAU) is used for quantifying reactor safety margins and works as an alternative method for licensing nuclear power plant. In the framework of CSAU, sensitivity and uncertainty analysis (SA and UA) should be performed. However, for large-scale complicated thermal hydraulics analysis simulated by code (e.g. RELAP5, CFD and subchannel analysis, the sensitivity and uncertainty analysis is really time-consuming. Furthermore, sometimes it is not feasible to perform SA and UA. Specifically, the computation time for Monte Carlo sampling or advanced LHS sampling can be neglected, but the computation time for thermal hydraulics code makes up main part of time during SA and UA process. In order to accelerate the speed of SA and UA, thermal hydraulics code can be improved based on advanced mathematic algorithm or using parallel method. However, the development for new mathematic method needs long time and strict verification and validation process is needed for a new code, which usually takes a couple of years or even longer time to have an approved code that can be used in nuclear industry. An alternative way to improve the speed of SA and UA is using surrogate model. The surrogate model is derived from the input as well as output data of the original code, assuming the code as a black box. The most common method is using response surface method (RSM) as a surrogate for the code, the input parameters are then sampled from their assumed distribution and the corresponding value of the response surface function is calculated. However, a couple of input parameters need hundreds or even thousands of original code calculations. The large number of runs is an obvious disadvantage of this method. Besides, the second disadvantage is an appropriate response surface function must be selected (e.g. quadratic, cubic, quartic, with or without cross product terms, or using a non-polynomial form) to avoid the overfit and underfit problem. This process is time consuming and expensive. Polynomial chaos expansion method is a reasonably fast and attractive alternative, stochastic model. The chaos expansion provides an efficient high-order method of including non-linear effects in stochastic analysis. Usually, the information about the model parameters and their probability distribution is limited and the dependence between them is not known. However, if arbitrary polynomial chaos is used, instead of assuming probability distribution for input parameters which may introduces additional subjectivity into the analysis, the available information can directly and most purely be used in stochastic analysis based on arbitrary data distribution. Thus, it is not necessary to know the exact probability density functions to perform SA and UA based on arbitrary polynomial chaos (aPC). The aPC has been applied for uncertainty quantification in many fields (e.g. in solid mechanics, fluid flows, thermal sciences), and become widely used in nuclear industry recently. Mk1 PB-FHR is a small modular fluoride-salt-cooled, high-temperature reactors (FHRs) that is a concept design by University of California, Berkeley (UCB). It has inherently safe features like natural-circulation-driven systems (e.g. direct reactor auxiliary cooling system (DRACS)) to passively remove decay heat from the reactor core during design basis and beyond design. It is necessary to perform SA and UA, since the licensing process of this new kind of reactor is under way and there are many design parameters that have not been fixed. Current nuclear commercial codes, such as RELAP5, have mainly been developed for light water reactors, which require the use of comprehensive two-phase flow models. This capability is not required to simulate single-phase flow phenomena in FHRs, whose baseline coolant, flibe, remains liquid up to 1430 deg. C at atmospheric pressure. The FHR advanced natural circulation analysis (FANCY) code has been developed specifically for FHR technology modeling and validation work has been performed by using CIET data in UCB. The code has been renamed General Flow (GenFlow) because it has been further developed to have the capability to model complicated pipework system based on object-oriented concept for user friendly purpose. Besides, it can be used to simulate both the force circulation as well as natural circulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Preliminary System Interface Failure Mode Analysis for Two Advanced Reactors

This study explores two distinct Balance of Plant (BOP) configurations: the Rankine cycle for a sodium-cooled fast reactor (SFR) and the Brayton cycle for a gas-cooled reactor (GCR). As representative designs, the Power Reactor Innovative Small Module (PRISM) by GE Hitachi Nuclear Energy was selected for the SFR, while the Gas Turbine Modular Helium Reactor (GT MHR) by General Atomics was chosen for the gas-cooled reactor. Both configurations were adapted to deliver high-quality heat for industrial applications. A Failure Modes and Effects Analysis (FMEA) was conducted for each system to identify critical failure modes affecting key components. This study marks the first phase of a two-step design optimization approach, integrating FMEA into the design process. Based on the analysis, design modifications and mitigation strategies were proposed to enhance system resilience. The second phase, to be detailed in a subsequent report, will focus on the role of the control system in mitigating these failures. The FMEA serves as a foundation for defining the control topology, ensuring system resilience against component failures that could compromise essential functions, such as electricity and heat production.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2 ) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Investigation of neutron noise in a micro-scale, natural circulation molten salt fission battery system

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. A specific novel micro-MSR, natural circulation, battery design concept involves placing all primary components within a single reactor vessel containment without the need of primary forced pumping. This small modular, integral design presents potential cost savings while producing safety, reliable, and transportable carbon-free power for decades. This investigation evaluates the neutron noise induced by density and flow fluctuations in a natural circulation MSR battery concept being developed at the University of Idaho. This study finds numerical solutions to the one-dimensional, one-group, coupled diffusion equations for a bare, homogeneous core, and evaluates neutron flux and delayed neutron precursor concentration noise due to core flow and fuel salt density perturbation sources. Noise analysis shows both point kinetic and space-dependent behavior is present despite the small size, low-flow, closely-coupled nature of the natural circulation MSR concept. Analysis shows that the low flow of the convective system results in noise behavior typically representative of a larger forced-circulation MSR system. Finally, noise techniques may also be useful in MSR core diagnostics as a non-invasive, low-cost radiological option to traditional monitoring schemes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗