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This work covers the operational use of the Aerojet General Nucleonics Model 201 Reactor (AGN-201M) at the University of New Mexico (UNM) in Albuquerque, New Mexico. The AGN-201M is a research and test reactor that operates at 5-watts (thermal) and utilizes HALEU fuel encased in polyethylene fuel disks. Operational usage in hours and watt-hours for the 2021 and 2022 fiscal year is provided, as well as licensing data (including training usage), notable research projects, outreach, and educational use in the Nuclear Engineering program at UNM. All research projects described herein have published conference papers, journal articles, or in-progress conference papers, journal articles and benchmarks. [ABSTRACT FROM AUTHOR]

• The paper focus on the effects of non-uniform inflow on a small modular reactor coolant pump (SMRCP). • This paper delves into the vortex structure inside the SMRCP to evaluate the impact of non-uniform inflow on its performance. • Using the Q -criterion to identify complex vortex structures and analyzing the mechanism of non-uniform inflow. • The internal flow state of key hydraulic components of the SMRCP during one rotation cycle is analyzed. In small modular reactors, the extremely short distance between the steam generator (SG) and the reactor coolant pump (RCP) results in a non-uniform inflow at the pump inlet. Therefore, in order to investigate the effect of non-uniform inflow on the small modular reactors coolant pump (SMRCP), the internal vortex structure of the SMRCP was investigated in this study. Firstly, the uniform and non-uniform inflow models were constructed for the comparison of the study, and the CFD method was used to obtain the hydraulic characteristic curve and internal flow field information of the pump, and the CFD results were verified by experiments. The results show that the efficiency and head of the uniform inflow model is both higher than that of the non-uniform inflow model, and the non-uniform inflow has an effect on the performance of the pump. Secondly, by analyzing the velocity streamline distribution and vortex distribution under the pump steady calculation results, it is found that the non-uniform inflow will greatly change the flow state in the pump, and a large-scale reflux vortex is formed at the outlet of the casing under the non-uniform inflow, and the vortex structure changes the flow distribution in the casing. Finally, the vortex distribution of the impeller channel under transient calculation of the pump under different models for one week of impeller rotation, the turbulent kinetic energy distribution of the diffuser inlet and outlet and three-dimensional vortex distribution are extracted for comparison, and the analysis shows that the vortex structure of the SMRCP is slow to dissipate due to the influence of the non-uniform inflow, and the turbulence kinetic energy dissipation is serious, which influences the performance of the pump. This study provides new directions for optimizing the performance of the SMRCP under non-uniform inflow. [ABSTRACT FROM AUTHOR]

• The influence of geometric parameters including twist pitch and cross section sizes on friction factor and heat transfer coefficient is studied in the helical cruciform single rod. • Geometric dimensionless numbers are proposed to fit the friction factor and heat transfer coefficient of helical cruciform. • The proposed empirical correlations of frction factor and heat transfer coefficient are validated respectively based on numerical simulation. Helical cruciform fuel is a novel type of nuclear reactor fuel with self-supporting and large surface-volume ratio features, potential to increase the reactor power density. However, the shape of the fuel is so complicated that influence of geometric parameters on thermal hydraulic characteristics was not clearly researched. In this paper, geometric dimensionless numbers are proposed and validated based on the numerical simulation to modify friction factor and heat transfer coefficient correlations proposed through experiments of helical cruciform fuels. Twist pitch of the fuel has a great impact on flow resistance and heat transfer, resulting in the modified coefficient ranging from 0.94 to 1.04 for friction factor and from 0.022 to 0.024 for heat transfer coefficient respectively. Change in cross-section parameters has little influence on the fuel performance, leading to the unit cross-section modified factor for both friction factor and heat transfer coefficient. After fitting the correlations with the dimensionless number, several cases are established to validate the empirical correlations, with a maximum error below 10.0 % for friction factor equation and 8.5 % for heat transfer coefficient correlation when Reynolds number is between 2.7 × 104 and 2.7 × 105. This paper provides a guide for study of heat transfer and fluid flow in helical structures, and a reference for establishment of reactor system model using helical cruciform fuels. [ABSTRACT FROM AUTHOR]

• A new equivalent radiation source method, which equates the radiation effect produced by the unknown real radiation sources, is proposed in this paper. • A limited number of detection points can be planned to build the equivalent source model, which helps deal with the problems of limited coverage of radiation detection points and improves the detection efficiency. • The effectiveness and feasibility of the proposed method is verified by simulations of different source distributions and experiments using a nuclear detection robot. Reconstructing nuclear radiation field is essential for a safe and efficient treatment strategy, especially when a nuclear facility fails and there is a need to go inside. Since prior radiation source information may not be accurate in the abnormal situation, this paper proposes a new equivalent radiation source method, which equates the radiation effect produced by the unknown real radiation sources. A limited number of detection points can be planned to build the equivalent source model, which helps deal with the problems of limited coverage of radiation detection points and improves the detection efficiency. The effectiveness of the proposed method is verified by simulations of different obstacles and different source distributions. The feasibility is experimentally tested by detecting and reconstructing the radiation field around a nuclear irradiation facility using a nuclear detection robot. In addition, the influence of the parameters is also analyzed to guide the practical usage of the proposed method. [ABSTRACT FROM AUTHOR]

• This paper: Digitizes all experimental data for integral loss of coolant accident (LOCA) tests. • Evaluates the modeling capabilities and data required for modeling LOCA accidents. • Identifies gaps in reported data useful for modeling and simulation work. This paper compiles and digitizes the publicly reported data for integral loss of coolant accident (LOCA) testing so it can be used by both multiphysics modeling and simulation groups and experimental groups. Specifically, this paper compiles the reported rodlet geometries, burst conditions, reactor operating conditions, and instrumented measurements from several test campaigns, including the KfK's FR2, PBF's LOC, PHEBUS, ANL, Studsvik, Halden IFA-650, and ORNL SATS LOCA series. Openly accessible data for LOCA experiments, especially those using high burnup UO 2 or ATF concepts, are limited and inconsistent in instrumentation, resulting in highly empirical models that lack the necessary accuracy when predicting fuel performance under LOCA conditions. Available fuel performance codes that integrate models for high burnup UO 2 and ATF concepts under LOCA conditions include BISON, FAST, and FALCON. This paper assesses the inputs, boundary conditions, and experimental demands informing the thermomechanical model options for each fuel performance code. Following comparisons between experimental LOCA capabilities and current modeling demands, informed recommendations to bridge the gaps between experimentation and modeling capabilities include detailed reporting of fuel rod geometries, reactor conditions, and axially dependent power and temperature profiles in addition to continued LOCA testing with expanded instrumentation in advanced testing facilities. [ABSTRACT FROM AUTHOR]

• Decay Heat Removal is an important aspect of nuclear Reactor Design. • Need to keep fuel and structure temperatures within limits. • Designs to ensure proper decay heat removal in sodium cooled fast reactors. • Needed under loss of grid power and station power. • Studies and experiments in support of decay heat removal. Shutdown systems and decay heat removal systems form the backbone of the nuclear plant protection system. While the former ensures safe shutdown of the fission reaction, the latter is essential to remove the heat from the decay of the fission products during earlier fissions. Thus, heat continues to be generated even after shutdown. Residual or decay heat removal (DHR) systems are needed to ensure that fuel clad temperatures do not rise beyond limits after the reactor shutdown. In fast reactors which have higher power density in the core, decay heat removal becomes very important to keep the core temperatures and structures within acceptable limits as otherwise fuel failure can result with its attendant consequences of adding radioactivity to the primary coolant. Also, in situations of loss of off-site and on-site power, this heat needs to be removed. Toward this, one has to ensure that the design is amenable to natural circulation cooling in different situations. The primary coolant system in Sodium cooled fast Reactors (SFR) can easily be configured to provide natural circulation shutdown heat removal. Different reactors designed/built/operated have resorted to different ways of decay heat removal. This paper traces the evolution of different decay heat removal options used by different fast reactor designs in different countries. Also presented is a brief review of some of the studies carried out on the efficacy of the different DHR systems besides parametric investigations carried out in different countries, backed up by experiments. The paper makes important observations on the type of DHR system to be adopted based on the above studies. [ABSTRACT FROM AUTHOR]

[Display omitted] • This paper demonstrates the Kartini Reactor's educational utility. • A case study was conducted by investigating reactor criticality. • The study achieved reactor criticality by inserting 5 fuel elements, totaling 2524.57 g. • The critical mass determination was refined by plotting the inverse of counts against the total mass of U-235. • This approach utilized both polynomial and linear trend lines to analyze the data. Research reactors serve critical roles across multiple domains—from physics to medicine—while distinctly not being used for electrical power generation. The Kartini Research Reactor located in Indonesia exemplifies the integration of research reactors in education and scientific research, particularly highlighting its educational applications through the Internet Reactor Laboratory (IRL). Established in 1978 and operational since 1979, the Training Research and Isotope Production by General Atomic (TRIGA) Mark II reactor is instrumental in advancing nuclear science with facilities like the Lazy Suzan/Rotate Rack and Beam Port. The IRL significantly enhances its educational process, enabling remote access to reactor physics experiments, thus facilitating global university participation in nuclear engineering education. This capability allows for the real-time visualization of reactor operations and detailed presentation of operational data and experiment results, enriching the learning experience. This paper demonstrates the Kartini Reactor's educational utility through a case study investigating reactor criticality and determining the critical fuel mass under safe operational parameters. Employing telecommunication hardware and software, the study achieved reactor criticality by inserting 5 fuel elements, totaling 2524.57 g. The critical mass determination was further refined by plotting the inverse of counts against the total mass of U-235. This approach utilized both polynomial and linear trend lines to analyze the data. While the polynomial analysis provided upper and lower thresholds for critical mass, the linear regression offered a precise critical mass value of 2428.62 g, showcasing a deviation of only 4% from the experimental findings. [ABSTRACT FROM AUTHOR]

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm2) and are amenable to longer "shaped" transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (∼2E15 n/cm2). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (∼1E15 n/cm2). An evolution of the UO 2 system, termed "ternary ceramic" fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (∼7E15 n/cm2), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (∼6E15 n/cm2). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. Like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations. [ABSTRACT FROM AUTHOR]

• Developing of integral VVER1000 model using coupled code TRACE5-P05/PARCS with 3D neutron kinetics model. • Investigation of VVER 1000 integral model capabilities using coupled model of TRACE5-P05/PARCS with 3D kinetics. • Comparison of "Main steam line break" calculated results simulated by coupled 3D TRACE/PARCS code and 1D RELAP5 code. • Comparison between 3D TRACE/PARCS coupled code with 1D RELAP5 code are carried out and minor differences are investigated. This paper presents a comparative analysis of the Main Steam Line Break (MSLB) in a VVER-1000 reactor simulated with RELAP5 using Point Kinetics and the coupled code TRACE5-P05/PARCS using 3D kinetics. In the MSLB-scenario, it is assumed that the main steam line break of 580 mm inner diameter is located between the steam generator (SG) and the steam isolation valve (SIV), outside the containment. In a MSLB, a non-symmetric overcooling of the primary coolant takes place leading to a positive reactivity insertion. Hence, the main safety concern is to assess if the core may become critical despite SCRAM and it there is a considerable power increase (return-to-power). This paper will discuss the capabilities of different computational approaches to simulate the VVER-1000 plant behaviour during a MSLB; one approach based on 1D thermal hydraulics and Point Kinetics while the other one based on 3D thermal hydraulics of the reactor pressure vessel (RPV) and 1D thermal hydraulics for the remaining plant components based on a 3D neutron kinetics model. The analyses are performed for Beginning of Cycle (BOC) conditions i.e., with a fresh core loading when the plant is operated at nominal power. The neutron kinetic parameters for the RELAP5 Point Kinetics model were generated PARCS for the BOC assuming a boron concentration of 1630 ppm. The respective 2 energy group homogenized cross section libraries in PMAXS-format were generated by KIT using the SERPENT2 code. The investigations were performed in the frame of CAMIVVER-project, which focus was the assessment and development of reliable neutron physical and system thermal hydraulic models for safety evaluations of VVER-1000 reactors. The comparative analysis for the MSLB has shown that both applied codes are able to qualitatively predicts the plant behaviour under MSLB-conditions in similar manner. Differences are caused by the different approach to represent the core and RPV followed by RELAP5 and TRACE5.05/PARCS as expected. [ABSTRACT FROM AUTHOR]

• ASTEC code calculations have been performed, within the first phase of reactor calculations of R2CA project (task 2.3). • LOCA DBA and DEC-A scenarios have been simulated. • Two cases have been considered for each scenario, characterized by some differences on initial/boundary conditions, accident management and assumptions. • Obtained results are discussed with a focus on the estimated number of failed fuel rods, fission products release in the environment and radiological consequences. • A reassessment of DBA and DEC-A calculation is made, within the task 2.5 of R2CA that contemplates a second run of reactor calculations, by using improved models and calculation chains. The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project. More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project. This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios. [ABSTRACT FROM AUTHOR]

• System level LBE thermal-hydraulics analysis code is developed. • Coupling effect of LBE oxidation corrosion and thermal-hydraulics is considered. • Influence of oxidation layer on heat transfer characteristics of LBE reactor is conducted. For Lead-Bismuth Eutectic (LBE) alloy cooled reactor, oxidation corrosion is very important and would affect the thermal hydraulic characteristics of the reactor. In this paper, based on LETHAC code developed by Xi'an Jiaotong University (XJTU) for system level thermal hydraulic analysis LBE reactor, a simulation module for oxidation corrosion is developed and added. The influence of oxidation layer on the heat transfer characteristics of the reactor is conducted. Based on LESMOR reactor designed by XJTU, a simulation calculation is carried out under the long-term operation conditions of the reactor considering the growth of oxidation layer. The distribution of the oxide layer in the reactor, the change of temperature appreciation caused by the oxide layer over time and the distribution in the reactor are obtained. The presence of oxide layer will affect the fluid temperature and wall temperature of reactor. The longer the run time, the greater the impact. This paper could provide solid analysis methodology reference for the design of LBE reactor. [ABSTRACT FROM AUTHOR]

• This paper introduces the coupling methodology of the thermo-mechanical code, FRAPTRAN/CUPID. • The paper discusses the verification and validation calculations of the coupled FRAPTRAN/CUPID. • The paper presents an OPR1000 SLB calculation as a reactor application. This paper introduces the coupling methodology of the thermo-mechanical code, FRAPTRAN/CUPID, and discusses the verification and validation calculations of the coupled FRAPTRAN/CUPID. In the verification calculation, which utilized 3 × 3 fuel rod channels, the coupled calculation produced consistent results compared to the standalone CUPID calculation. The validation calculations, based on two international and domestic LBLOCA tests (OECD-Halden IFA-650.5 and ICARUS-RT-20-02), demonstrate that the coupled FRAPTRAN/CUPID code effectively simulates the thermal–hydraulic behaviors of nuclear fuel rods during LBLOCA. It also provides valuable thermo-mechanical information regarding the deformation and burst of the nuclear fuel rods during LBLOCA. Finally, the paper presents an OPR1000 SLB calculation as a reactor application, utilizing the FRAPTRAN/MASTER/CUPID/MARS framework. [ABSTRACT FROM AUTHOR]

• Numerical study of turbulent copper–water nanofluid flow in a heat exchanger. • The heat exchanger in power plant was in the form of a tube with a large number of turbulators. • Reduced entropy production by turbulators length and space. • Larger hole diameter decreases thermal (98.7%), frictional (72.5%), and total entropy (98.5%). This paper presents a numerical investigation of turbulent nanofluid flow in a power plant heat exchanger using conical rings. In this paper focus on varying conical ring hole diameters and their impact on key parameters. Utilizing artificial intelligence techniques, the finite element method (FEM), and the RNG k-epsilon model, analyzed pressure drop, thermal entropy, frictional entropy, and total entropy while varying distances between turbulators (80 mm to 180 mm) and turbulator lengths (45 mm to 60 mm). The results show that the smallest turbulator length and diameter, with an inter-turbulator distance of 145 mm, generate the highest frictional entropy (72 % increase). Smaller distances between turbulators with larger turbulator dimensions result in the smallest pressure drop. Increasing hole diameter significantly reduces thermal (98.7 %), frictional (72.5 %), and total entropy (98.5 %). These findings offer insights for optimizing power plant heat exchanger design using nanofluids and conical turbulators. [ABSTRACT FROM AUTHOR]

• The simulation and experimental errors of less than 20 per cent. • The effects of system pressure, inlet Re number, heat flow density on the void fraction, surface heat transfer coefficient, pressure drop are investigated. • Comparison with some scholars' empirical relational formulas to find conforming prediction formulas. • Analysing the effect of geometric parameters on non-uniform distribution of wall temperature. Due to the structural characteristics of the helical tube, centrifugal force complicates the flow boiling phenomenon inside the tube. How to accurately describe and predict the parameter distributions involved in helical tube flow and heat transfer is a concern for scientists. Based on the OpenFOAM, this paper combines the wall boiling model, the interphase model, other closed models with the Eulerian two-fluid model, analysed changes of void fraction, surface heat transfer coefficient, pressure drop with working conditions in the single-phase to nuclear boiling section in the tube. The results show that this solver and the corresponding empirical relational model have the ability to accurately simulate the boiling of the flow in helical tube; In nuclear boiling section at the working conditions of P =4–8 MPa, q = 200–350 kW/m2, Re = 66827–89103, The degree of gas-phase buildup near the inner wall surface of the spiral tube decreases with the increase of Re number, increases with the increase of heat flux and pressure, and the ratio of friction pressure drop to total pressure drop decreases with the increase of Re number, heat flux, and pressure by a maximum of 1.4 %, 4.26 %, and 17.35 %. This paper can provide a reference for adding new models and developing new solvers in the OpenFOAM to simulate boiling in helical tube flows. [ABSTRACT FROM AUTHOR]

• A scoping analysis tool is developed to estimate fuel cycle cost for various PWR core sizes. • Pursuing higher specific power and optimal burnups are the main fuel cycle cost reduction drivers. • Shorter fuel assemblies lead to higher fabrication costs but unlock potential for power uprates. • Spent fuel volumes are linked to core discharge burnup which is independent from reactor size. The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined. [ABSTRACT FROM AUTHOR]

• Introduces the Fuzzy Analytic Hierarchy Process (AHP)-VIKOR technique for risk ranking in the nuclear power industry. • Emphasizes the need for thorough examination and sequential assessment of each component's risks to minimize failures and risks. • Utilizes expert evaluations based on Severity, Occurrence, and Detection criteria to calculate risk weights and rankings. • Provides a higher-level risk analysis that incorporates multiple performance parameters. The critical systems in industries such as nuclear power plants rely on various preventive methods to minimize failures or risks through efficient strategies and equipment. However, in many businesses, maintenance tasks are carried out infrequently, improperly, and without consideration for the overall state of the plant or its equipment. To choose the appropriate risk preventive approach, a thorough examination of each component's risks in a sequential manner becomes imperative. This paper introduces the Fuzzy Analytic Hierarchy Process (AHP)-VIKOR technique, a Multicriteria Decision Making Approach employed to rank the various risks prevalent in the nuclear power industry. By identifying the proper sequence of risks, this approach aims to reduce the occurrence of unfortunate mishaps, along with minimizing recovery time and costs. Five experienced researchers and experts assessed the impact of risk based on three risk criteria: Severity, Occurrence, and Detection. Utilizing the opinions and judgments of these experts, the Fuzzy AHP-VIKOR approach was employed to calculate the weight of each performance criterion and the ranking of each risk or hazard. The suggested method is designed to assist supervisors in resolving discrete problems characterized by incommensurable and conflicting criteria. This study contributes to the industry by providing a higher risk analysis incorporating various performance parameters. The paper concludes by presenting a result with the priorities of all risks in the industry using the fuzzy AHP-VIKOR method. [ABSTRACT FROM AUTHOR]

Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, the Japan Atomic Energy Agency (JAEA) has been modifying a critical assembly called "STACY." The first criticality of the modified STACY is scheduled for spring 2024. This paper reports the consideration results of the core configurations of the modified STACY for the first criticality and inspections. The specifications of these core configurations were determined in advance, we tried to make them with a simplified computational model that considers the reactivity effect around the core. At the first criticality, two types of grid plates with different neutron moderation conditions (their hole spacings are 1.50 cm and 1.27 cm) were prepared. On the other hand, there is a limitation on the number of available UO 2 fuel rods. The core configurations for the first criticality satisfying these experimental constraints were designed by computational analysis. A cylindrical core configuration with a 1.50 cm pitch grid plate close to the optimum moderation condition needs 253 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered core configurations with 2.54 cm intervals by using doubled pitches of the grid plate. It will need 213 fuel rods for the criticality to be reached. In addition, the experimental core configuration was considered with steel/concrete simulant rods to simulate fuel debris conditions. This paper shows six core configurations with different conditions, and all of them satisfy the regulatory requests. [ABSTRACT FROM AUTHOR]

Timely and accurate detection of defective fuel rods is critical as the release of radioactive fission products from defective fuels can lead to primary circuit contamination and radiation exposure. Due to the complexity of the physical phenomena, models for fault diagnosis can be difficult to construct and recently data driven surrogate models have being increasingly used to detect and characterize defective fuel rods: they make use of a computational database to learn from and make predictions about new unknown data. In this paper, we present a method for the elaboration of an anomaly detector based on neural networks, taking into account the fact that physical computation can be CPU intensive and thus overcome this issue. A physical model for fission products release and coolant activity calculation was built and used to generate a surrogate activity model that enables the generation of a bigger database in small amount of CPU times. Then using this bigger computational database, a recurrent autoencoder was trained for anomaly detection. The network classifies the defect status with 100% accuracy and a good time precision. A sensitivity analysis with lower activity increase at defect onset and addition of noise was conducted in order to better understand the limits of this method. Such methods can be useful for operators of the existing as well as future reactors to make timely predictions of defective fuel rods and avoid operational and economic setbacks for power plants. The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). [ABSTRACT FROM AUTHOR]

The management of high-level nuclear waste (HLW) and spent nuclear fuel (SNF) is an important aspect when considering any nuclear fuel cycle scenario. To address the challenges of nuclear waste disposal, tools are essential for assessing the thermal characteristics of geological disposal facilities (GDFs) across different nuclear fuel cycles, as often strict temperature limitations are in place that ultimately dictate repository footprint. This paper provides a comprehensive overview of BADGER, a code designed for modelling the thermal behaviour of repositories containing HLW and SNF. BADGER has unique capabilities in efficient thermal modelling and the ability to model the accommodation of diverse waste types within a single repository. It can also model how the evolution in time of different waste streams impact repository footprint. Notably, BADGER assesses the repository impact of different fuel cycles when coupled with a fuel cycle analysis code like ORION, calculating peak temperatures, optimal canister spacing and repository footprint. In our analysis, we demonstrate BADGER's capabilities with two illustrative UK fuel cycle scenarios: one open and one closed. For the open cycle, the repository footprint was found to be 14.32 km 2 , whereas for the closed cycle, it was 5.53 km 2. This paper explores the features, underlying theory, verification and two illustrative UK fuel cycle scenarios, to highlight BADGER's capabilities in evaluating various fuel cycles' impact on repositories. • A comprehensive overview of BADGER, a code designed for modelling the impact of nuclear waste on repository footprint. • Illustrates BADGER's capabilities with two UK fuel cycle scenarios (one open and one closed). • Highlights the integration of BADGER with fuel cycle analysis codes (such as ORION) for comprehensive fuel cycle evaluations. • Outlines strategies in BADGER to enhance computational efficiency, enabling repository layout optimisation studies. [ABSTRACT FROM AUTHOR]

Nuclear data evaluations available in existing nuclear data libraries are derived based on differential measurements that includes experimental effects such as target temperature, time-of-flight resolution, data normalization, self-shielding, multiple scattering, etc. Measurements are often made at temperatures corresponding to room temperature, 293.6 K, to avoid complexity in the experimental setup and costs of carrying out measurements for temperatures other than room temperature. This paper investigates the impact of experimental effects on the evaluation of a set of resonance parameters that fit the experimental differential data and its use in integral benchmark calculations. Given the importance of the temperature in integral benchmark results, the impact of the Doppler effect will be examined. Very seldom are experimental differential data available for temperatures below or above room temperature. Nuclear data measurements and evaluation needs are driven by reactor applications; consequently, the majority of data evaluations in nuclear data libraries are for temperatures above room temperature. Recently there has been a demand for nuclear data for low temperatures, below room temperatures, for criticality safety applications. Currently, calculations in response to low-temperature needs are based on extrapolating the existing data from the nuclear data libraries to temperatures below 293.6 K. For temperatures above 293.6 K, common practice is to process the data library to temperatures different from the temperature it was evaluated and use them in practical applications. Although this is an acceptable practice, care should be taken to understand whether the validity of the nuclear data can be extended to low and high temperatures. Issues in connection with temperature effects for low and high temperature nuclear data and their impact on practical applications are addressed in the paper. Given that experimental data for low- and high-temperatures are scarce, the results of the presented approach are based on data simulations. Simulated data for 235U in the resonance region, in particular the resolved resonance region, were used as part of the studies and demonstration. Furthermore, temperature effects were also investigated for thermal neutron scattering data, S (α , β) , for light water. However, the thermal scattering data are not based on simulation, but are the result of measurements carried out at the Spallation Neutron Source. A continuous-energy nuclear data library was used in Monte Carlo calculations to assess the impact in integral benchmark results. [ABSTRACT FROM AUTHOR]

• Neutron physics, thermal–hydraulic and water chemistry are coupled to predict corrosion product growth and CIPS risk. • Boron hideout mass shows a parabolic tendency as CRUD depositions thicken and bulk coolant boron concentration decreases. • Power tends to shift to 0 % axial offset at EOC after the most negative axial offset of −5.07 % with 5.03 mg boron hideout. CRUD depositions on fuel cladding are the main cause of power shift and localized corrosion in nuclear power plants. This paper is the third of a three-part study concerning the prediction of CIPS risks during fuel burnup. In this paper, based on the coupling modules of CRUD growth and internal heat and mass transfer in Part 2, a multi-physics coupling method for high-fidelity simulation of CRUD growth and CIPS phenomenon is developed to better understand coupled physics and respective feedback mechanisms. Based on OpenMC, Fluent and self-developed CRUD-related code, this coupling method considers nuclide depletion, coolant flow, heat transfer and water chemistry, in which fine radial CRUD mesh in neutronic module and influences of CRUD surface characteristics on heat transfer in thermal–hydraulic module are introduced. The coupling method is applied to predict CIPS risk during a 360-day cycle depletion process reaching burnup of 22.04 MWd/kgHM. Corrosion products deposit on cladding to thicken CRUD, while flow erosion gets enhanced washing away more particles on CRUD surfaces, resulting in a slow-down net CRUD growth rate, reaching maximum values of 8.60 g in mass and 37.10 μm in thickness at EOC. As CRUD depositions thicken, boron hideout mass increases, while subsequently bulk coolant boron concentration decreases and boron diffuses out of CRUD depositions causing a decrease in boron hideout mass. The highest boron hideout mass and most negative AO both occur near MOC time. At EOC, net mass reduction of boron hideout and bottom-favored nuclide depletion contribute toward pushing power distribution closer to 0 % axial offset. The results of this study provide a precise method for understanding CIPS risk and its influencing factors to further predict and alleviate CRUD-related safety issues. [ABSTRACT FROM AUTHOR]

1 1 Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (https://www.energy.gov/doe-public-access-plan). Safeguards verification measurements of nuclear material content in fresh fuel salt for liquid-fueled molten salt reactors (MSRs) are likely to be required as part of nuclear material accountancy for International Atomic Energy Agency safeguards. This paper presents a comprehensive review and evaluation of 18 potential candidate techniques to quantify total uranium and 235 U for input accountancy measurements for liquid-fueled MSRs. As part of an overall screening and down-selection effort to identify the most promising techniques for further development for an MSR feed monitoring system, this paper defines eight figures of merit (FOMs): reasonably achievable measurement uncertainty, measurement time required, capital cost, burden upon the facility operator, maintenance intensity, technological maturity, human capital requirements for operation, and whether the technique introduces a path for potential material removal. Each candidate technique is then evaluated across these FOMs to identify the techniques with the highest potential for future development for fresh fuel accountancy measurements in MSRs. Our findings indicate that no single technique or combination thereof currently has the requisite technological maturity for immediate implementation in nuclear material accountancy at a liquid-fueled MSR facility. While several promising techniques are identified, there is a critical lack of experimental data for most systems in the context of molten salt applications. • We propose criteria to screen feed accountancy techniques in molten salt reactors. • We evaluated 18 candidate techniques to identify the most promising for development. • Presently, no technique is sufficiently mature for MSR feed accountancy measurements. • Critical gaps in measurement data of actinide-bearing molten salts must be addressed. [ABSTRACT FROM AUTHOR]

• Designed and constructed a fault data acquisition system for nuclear power plant electric gate valves. • Based on the study of the fault mechanism of electric gate valves, analyzed the occurrence locations of faults in electric gate valves, completed fault settings, and provided theoretical guidance for designing relevant experimental platforms for other scholars. • Analyzed the changes in acceleration signals and AE signals when different types of faults occur in electric gate valves, selected corresponding sensors, and completed the arrangement of sensors and signal acquisition. • To effectively improve the accuracy of fault classification and the precision of fault severity assessment, this paper improved the original algorithms and developed a fault diagnosis system for nuclear power plant electric gate valves based on the SAE algorithm and a fault severity assessment system based on the Bi-LSTM algorithm. Electric gate valve failure is a common type of fault in nuclear power plants. Among all factors leading to reactor shutdown in nuclear power plants, valve failures account for a significant proportion. Due to the difficulty in obtaining valve failure data in nuclear power plants, to effectively obtain such data and provide data support for the development of subsequent fault diagnosis algorithms, this paper adopts an experimental research method. It designs and constructs an electric gate valve failure simulation test bench, obtains experimental data under various states of electric gate valves, and develops a fault diagnosis system for electric gate valves in nuclear power plants based on this. The experimental results show that the data generated in this experiment can well achieve the purpose of classifying valve failure types and evaluating the degree of failure. Moreover, the developed fault diagnosis system exhibits high diagnostic accuracy and low error in evaluating the degree of failure. [ABSTRACT FROM AUTHOR]

Modeling and quantifying dependency is a foundational process in human reliability analysis (HRA) for complex engineering systems. However, existing HRA dependency modeling and quantification paradigms typically rely on simple correlation-driven, summative dependency levels and probability adjustments, with little consideration of the underlying causal mechanisms that drive the HRA dependency. This leads to challenging quantification, limited technical insight, and significant variability in numerical results. In this paper, we propose a data-informed framework for conceptualizing and modeling HRA dependency to overcome these challenges. Our methodology leverages the HRA dependency idioms and the SACADA HRA database to build objective and traceable formative HRA dependency models. SACADA data is used to create the model nodes, and the HRA dependency idioms guide the creation of the causal model structure implemented in a BN. The method is demonstrated on a LOCA case study. A companion paper details the methodology for quantifying the model using HRA data. • Proof-of-concept algorithm for building causal Bayesian Networks of HRA scenarios. • Demonstrates use of SACADA HRA data and HRA dependency idioms for BN construction. • Demonstrates the algorithm on a case study of a simulated HRA scenario. [ABSTRACT FROM AUTHOR]

Human reliability analysis (HRA) identifies the causal factors impacting the occurrence of human failure events and quantifies the human failure event probabilities based on those causal factors, and requires understanding the dependency structures that exist between failure events and causal factors. Many HRA methods incorporate the dependency framework established in the Technique for Human Error Rate Prediction (THERP), which uses simple multipliers on human error probabilities (HEPs) resulting from considering only a few factors. Accordingly, those HRA methods have a limited ability and accuracy when characterizing dependency. This paper presents a methodology for using HRA data to quantify Bayesian networks built from the HRA dependency idioms. The result is an objective and traceable human reliability model that enforces formative, rather than summative, dependency. The results show that baseline HEPs in this model are reasonable, and the data-informed dependency is a significant improvement for the field. • Develops quantification process for causal HRA Bayesian Network using SACADA data. • Demonstrates quantification process on case study model from the companion paper. • Demonstrates reusability of idioms and BN submodels for creating multiple HRA models. • Demonstrates that BN quantification results align with HRA expectations. [ABSTRACT FROM AUTHOR]

• Typical CRUD depositions are obtained via accerlated deposition method in rod array channels. • Effective CRUD thermal conductivity decreases as heat flux increases with 1.3866 W/(m × K) on average. • CRUD growth process is a combination of soluble precipitation and particulate aggregation. • Fouling resistance of corrosion product deposition behavior has been experimentally measured. CRUD depositions on fuel cladding are the main cause of power shift and localized corrosion in nuclear power plants. This paper is the first of a three-part study concerning the formation mechanism of CRUD depositions and its related heat transfer issues. In this paper, CRUD depositions are obtained under subcooled boiling conditions in 2 × 2 rod array channels via accelerated deposition method to explore corrosion product deposition and its feedback on heat transfer. Imbalance of deposition and erosion at initial stage causes fouling resistance to increase first and then decrease. CRUD growth is proposed as a relatively pure fouling process combining soluble precipitation and particulate aggregation. Through high-resolution characterizations, boiling chimney diameters span from 3 μm to a dozen microns. Principal components are nickel ferrite and nickel elemental with Fe/Ni ratio of 1.9:1. Static contact angle decreases to less than 30°. Effective CRUD thermal conductivity decreases with the increase of heat flux with 1.3866 W/(m × K) on average. The results of this study provide a precise method for understanding corrosion product deposition and its impact on heat transfer to further establish accurate CRUD-related models and predict CRUD-related safety issues. [ABSTRACT FROM AUTHOR]

Monte Carlo (MC) codes coupled to depletion solvers are increasingly used to provide high fidelity fuel cycle modeling capabilities. These coupled depletion-MC tools produce accurate results in general but can experience nonphysical spatial oscillations when time steps are large or when a system's dominance ratio approaches unity. Two substepping techniques have been developed previously to remedy and dampen these spatial oscillations without needing to reduce step sizes. The first approach relied on higher-order techniques to account for spectral changes within steps (extrapolation and interpolation techniques). The second approach used the first order perturbation (FOP) theory to account for the change in the one-group spatial flux distribution within steps. This paper develops a hybrid depletion methodology which, in a way, combines how the flux is handled in both substepping techniques. Specifically, the multigroup (MG) MC Shift code is used to update the flux distribution within steps rather than a one-group FOP solver. A fully reflected pincell is investigated, which is not spatially dependent in the MG representation. Thus, the analysis in this paper is an initial demonstration of hybrid depletion. An upcoming companion paper will focus on how the hybrid depletion dampens spatial oscillations. The hybrid depletion approach is verified to be consistent with previous constant extrapolation depletion (CED) methods. This paper finds that the hybrid CED exhibits some error in the eigenvalue and one group constants within macro steps. To address this discrepancy, a simple interpolation scheme (CELI) is investigated. This work found that CELI sufficiently addresses the discrepancy in spectrum for macro steps up to 100 days. Overall, this work demonstrates that the hybrid depletion method can significantly reduce the number of high fidelity MC executions in a MC-coupled depletion with an acceptable eigenvalue error. [ABSTRACT FROM AUTHOR]

• Introducing the Gaussian-Cauchy hybrid mutation into SSA to enhance optimization, shown effective in comparisons with SSA. • Using MDI indicators and K-L divergence in feature extraction for better fault signal representation from sensor data. • Improving SSA for optimizing random forest model parameters, achieving 96.375% accuracy in electric gate valve fault diagnosis. Electric gate valve (EGV) is an essential equipment within nuclear power plant (NPP). This paper presents an advanced fault diagnosis (FD) approach, leveraging Variational Modal Decomposition (VMD), Mutual Dimensionless Indicator (MDI) and the Random Forest (RF) optimized through Improved Sparrow Search Algorithm (ISSA), aimed at improving the accuracy of fault diagnosis and optimizing the FD model during EGV failure events. To commence, we employ the VMD algorithm for modal decomposition of raw electric gate valve signals. This process yields several Intrinsic Mode Function (IMF) components with diverse frequencies, enabling the capture of the underlying dynamics of the signals and facilitating a more comprehensive analysis of the fault conditions. We subsequently apply the K-L divergence to identify key IMF components that closely resemble the original signals. These selected key IMF components serve as the foundation for extracting dimensional indicators (DI) and mutual dimensionless indicators (MDI) as signal features. Furthermore, the Improved Sparrow Search Algorithm (ISSA) is enlisted to optimize the maximum feature count and the number of decision trees in the Random Forest (RF) algorithm. Ultimately, the optimized RF algorithm is deployed for fault diagnosis. Our paper offers a comparative analysis, pitting the VMD method against Empirical Mode Decomposition (EMD) and Local Mean Decomposition (LMD). Additionally, we compare our proposed fault diagnosis model with traditional RF algorithm and the SSA-RF algorithm. Through rigorous experimentation, our results achieved an average fault diagnosis accuracy of up to 96.375%. [ABSTRACT FROM AUTHOR]

The aim of this contribution is to deeply introduce the ALCYONE Fuel Performance Code. ALCYONE is being implemented within the PLEIADES numerical framework and is dedicated to the multiphysics behavior of fuel rods in pressurized water reactors (PWRs). It has been a joint development between CEA, EDF and Framatome since 2005 and has been used since then for numerous studies covering normal, off-normal and accidental conditions. However, a detailed description of ALCYONE's physical and numerical foundations has never been published. In this sense, this paper is intended to be the reference presentation paper for ALCYONE. On the first hand, it provides a comprehensive description of the modeling features and associated multiphysics and multiscale computational schemes. On the other hand, focuses are made on advanced modeling features showing enhanced capabilities of simulation of ALCYONE. • Detailed description of ALCYONE, the PLEIADES PWR Fuel Performance Code. • ALCYONE's physical models and numerical features are presented. • ALCYONE's simulations cover normal, off-normal, and accidental conditions. • Verification, validation and uncertainty quantification process is discussed. • Insights on some advanced modeling capabilities and numerical coupling schemes. [ABSTRACT FROM AUTHOR]

• Modeling and analysis of the entire small pressurized water reactor nuclear power plant were conducted using the MELCOR code. • This paper analyzed the total release quantity and distribution patterns of radioactive aerosols and inert gases. • This paper analyzed the impact of the activation of the containment spray system on the release and distribution of radioactive aerosols and inert gases. The small-sized pressurized water reactor (PWR) characterized by its high inherent safety and wide applicability, has become a promising type of reactor with broad prospects for development. Currently, there is a limited amount of research dedicated to the analysis of accident source terms specific to small-sized pressurized water reactors, this study adopts the ACP100 as a reference to establish a model of a small-sized pressurized water reactor power plant using the severe accident analysis code MELCOR. This study investigated the release and distribution characteristics of radioactive materials under the scenario of a large break loss of coolant accident in the heat pipe section. Specifically, it analyzed the impact of opening and shutting down the containment spray system on the release and distribution of radioactive materials. The results indicate that in the absence of the containment spray system, the majority of radioactive aerosols are primarily distributed in liquid form within the containment, with a small portion depositing on other components and dispersing in the gas phase; inert gases are primarily distributed in the gaseous phase within the containment. With the containment spray system engaged, there is an increase in the total quantity of radioactive aerosols and the concentration of radioactive aerosols in the gaseous phase significantly decreases, while the concentration of liquid-phase radioactive aerosols dramatically increases. The containment spray system exhibits strong capabilities for the removal and retention of radioactive aerosols. While the total release quantity remains largely unchanged, the time to reach the peak release of inert gases is significantly shortened The containment spray system significantly affects the release rate of inert gases. [ABSTRACT FROM AUTHOR]

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures). In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years. This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA's Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided. The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident. This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation. • Multi-physics coupled simulations for Sodium Fast Reactors (SFR). • Prospective simulations of Sodium Fast Reactors (SFR) accidental transients. • Sodium boiling studies and simulations. • Experimental validation on the Fast Flux Test Facility (FFTF). [ABSTRACT FROM AUTHOR]

• This paper describes PIV measurement on the behavior of jets passing through a grid-type obstacle. • The experiment clearly captured the jet fragmentation, and the jet recoupling after passing through the obstacle. • The meandering flow was suppressed, and the number of vortex structures entraining the surroundings was fewer due to the rectifying effect. • The experimental data shown in this paper is useful for CFD validation. During a severe accident in a nuclear power plant, jets released from the primary system exhibit complex thermohydraulic behavior due to buoyancy effects and impingement on internal obstacles such as inner walls and floors. Thus, the obstacle-influenced jets are of interest in recent research activities. This paper describes an experimental investigation of the behavior of jets passing through a grid-type obstacle. The flow field was acquired by a particle image velocimetry system. The experiment captured the jet fragmentation by the grid-type obstacle and the jet recoupling after passing through the obstacle. The mean velocity field obtained by postprocessing indicates a "Rectifying effect," with the axial velocity increasing at the center and the magnitude of the radial velocity decreasing. The meandering flow was suppressed due to this effect. In the near grid-obstacle region, the axial turbulence intensity was relatively large at the edge of each fragmented region due to shear stress. Moreover, the spatial distribution of the radial turbulence fluctuation became more complex. Further investigation is required to clarify the budget of the transport equation for turbulence fluctuation. In addition, the experimental data shown in this paper is useful for computational fluid dynamics validation. [ABSTRACT FROM AUTHOR]

This paper investigates the effect thermal feedback has on the stochastic nuclear reactor dynamics of low neutron source nuclear power plant (NPP) start-ups. Stochastic mathematical and computational models are required to determine the probability of a stochastic power surge occurring during nuclear reactor start-up that would damage the nuclear fuel. The aim is to design the nuclear reactor, the nuclear fuel, and the operational start-up procedures in a manner that minimises the probability of a stochastic power surge occurring, which damages the nuclear fuel, to a prescribed level of probabilistic risk (1 0 − 8 – 1 0 − 5 ). Recently, the Pál-Bell equations have been used for such low neutron source nuclear reactor start-up simulations. However, the stochastic nuclear reactor start-up models, based upon the Pál-Bell equations, cannot accommodate changes in the macroscopic neutron cross-sections arising from feedback processes. An alternative approach that could, in principle, include thermal feedback processes is the forward master equations. However, these are complex to implement for multidimensional and multi-group stochastic nuclear reactor dynamics problems. In addition, time-dependent analog Monte Carlo models could be used but these are computationally prohibitive for most nuclear reactor start-up simulations. This is due to the stringent requirements on the statistical accuracy of the safety probability associated with stochastic power surges. Therefore, this paper uses an alternative Itô-calculus approach to compute the stochastic properties required for low neutron source NPP start-up. The Itô-calculus approach is an approximate mathematical method, compared to the more general Pál-Bell and Monte Carlo methods, for low neutron source nuclear reactor start-up and fast burst systems. Therefore, the implementation of the Itô calculus method is first validated against the Caliban fast burst nuclear reactor experimental wait-time results to understand the accuracy of the method. The implementation of a simple feedback model is also verified against the point neutron kinetics equations. The neutron population CDF calculated using the Pál-Bell equations without thermal feedback, and the safety probabilities computed with, and without, including thermal feedback mechanisms are then analysed and assessed. These results demonstrate that the Itô-calculus approach can be used to gain useful insight into the behaviour of stochastic nuclear kinetics and dynamics during low neutron source NPP start-up. Furthermore, the results suggest that the safety probabilities computed using the Pál-Bell method are not affected by neglecting thermal feedback mechanisms which is an important result from a nuclear reactor safety perspective. [ABSTRACT FROM AUTHOR]

• This paper investigates the core characteristics of a reactor loaded with ATF. • Cr-coated and LAS doped fuel pellet ATF were considered. • The overall impact on core performance is not significant exculding cycle length. Following the Fukushima Daiichi accident, efforts have been made in a variety of disciplines to respond to nuclear accidents. From the perspective of nuclear fuel, various approaches have been attempted, and development has been made as an Accident Tolerant Fuel (ATF). Advanced fuel technology, often known as advanced fuel concepts, is ATF. The ATF is a type of fuel designed to be more resistant to damage in the event of a loss-of-coolant accident (LOCA) or other type of incident. Some ATF designs make use of materials that can better withstand high temperatures and pressures, which can help prevent fuel melting and release of radioactive materials. Currently, we are in the stage of developing a method to apply Cr coating on zircaloy cladding or a new cladding made of SiC. Additionally, some minerals like Cr, Mn, or La 2 O 3 -Al 2 O 3 -SiO 2 (LAS) are doped into pellet materials to alter their characteristics in order to create ATF fuel. In this paper, the characteristics of a core loaded with Cr-coated clad, and LAS doped fuel pellet ATF was investigated. The multiplication factor, power distribution, peaking factor, moderator temperature coefficient, and cycle length of the ATF-loaded core were calculated and compared to the results of the ATF-free core. The neutron absorption increases when the ATF is loaded due to the Cr coating and LAS doped fuel pellet, resulting in a difference when compared to the reference case. However, this study confirmed that the difference attributable to ATF was not significant except for the cycle length. To compensate for the reduced cycle length due to ATF, the cycle length and peaking factor were evaluated for the core with increased fuel enrichment and compared with the results of the conventional core. [ABSTRACT FROM AUTHOR]

• This paper explores the capabilities of machine-learning based fault detection and diagnosis. • The scope of the research is narrowed to the identification of transient events. • The K-Nearest Neighbors model achieved the highest accuracy of 97% in the primary system. The increasing interest in artificial intelligence and automation within the nuclear industry stems from the hope of elevating the safety and reliability of nuclear power plants by minimizing the impact of human errors. This paper explores the capabilities of machine-learning based fault detection and diagnosis (FDD) models in accurately classifying transient events in a nuclear power plant using the Classification Learners Application from MATLAB. The scope of the research is narrowed to the identification of transient events of the same nature occurring at various locations within the plant. To accomplish this, the Generic Pressurized Water Reactor (GPWR) from the Western Services Cooperation (WSC) was used to simulate 14 transients in the primary system and 10 transients in the secondary system, resulting in two distinct datasets of 108,000 and 79,200 observations. Nine types of classifiers, including Decision Trees, Discriminant Analysis, Support Vector Machines, Logistic Regression, Nearest Neighbors, Naive Bayes, Kernel Approximation, Ensembles, and Neural Networks, with a total of 33 predefined models, were trained and validated. The K-Nearest Neighbors model achieved the highest accuracy of 97% in the primary system, while the Efficient Linear Discriminant and Logistic Regression models achieved the highest average accuracy of 99% in the secondary system. The assessment of the top-performing models was conducted through a comprehensive analysis employing key classification model evaluation metrics, such as the confusion matrix, accuracy, precision, recall, and the F1 score. In an effort to optimize the models, the study integrated feature selection using the minimum Redundancy Maximum Relevance (mRMR) algorithm. Additionally, the investigation explored different validation schemes, including both holdout and k-fold cross-validation schemes, resulting in a substantial reduction in training time without significantly compromising the overall accuracy of the models. The optimized models in this study demonstrated remarkable prediction accuracies, consistently exceeding 94%. Despite the complexity of the models and the intricacies involved in the training process, the training times ranged from 10 to 1800 s, reflecting the efficiency and robustness of the implemented optimization techniques. The results of this study underscore the potential of machine learning models in identifying transient events in nuclear reactors, showcasing their promising capabilities while maintaining low computational and execution costs. This suggests their efficacy in optimizing safe and efficient operational practices within nuclear facilities. [ABSTRACT FROM AUTHOR]

The prestressed concrete containment vessel (PCCV) is a critical safety measure for nuclear power plants. In the analysis of PCCV, the parameters of PCCV are usually determined through experience and a small number of experiments, which are highly accidental, and this paper provides a new idea for clarifying the scope of PCCV material parameters and clarifying some parameters that are important to PCCV based on the Bayesian updating method. Finite element models consisting of multilayer shell elements were established based on the 1:4 PCCV experiment conducted at Sandia National Laboratories (SNL) in this paper. The model was used to identify the most influential 5 parameters during the late loading stage, and a Bayesian model updating method was applied to adjust these parameters. The finite element model, incorporating the modified posterior parameters, effectively predicted the displacements of measurement points and failure mechanisms of the model. • A rapid model updating method applied in the complex models is proposed. • Bayesian model updating approach first is applied in the nuclear containment vessel. • The uncertainty analysis process of nuclear containment vessel is showed in detail. [ABSTRACT FROM AUTHOR]

• Predicted the occurrence position and power of the rod bundles' CHF. • Compared the influence of cladding materials' properties on the rod bundles' CHF. • Provided a detailed analysis of the impact of grid spacers on the rod bundle' CHF. The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation methods. In addition, most simulation studies have not considered the solid region of the cladding, so there is little research on the effect of cladding materials on CHF. The role of grid spacers in rod bundle CHF is the focus of this paper, and the influence of cladding materials is another discussion direction. In this paper, following meticulous mesh generation and computational model validation, a study on the CHF of the 5 × 5 rod bundle with multiple grid spacers and different cladding materials is carried out by using CFD software. The verification results indicated that the physical model established in this study can reasonably predict the generation location and critical power value in the 5 × 5 rod bundle with grid spacers. Furthermore, the study identified that the grid spacer exerts a significant influence on coolant flow within the channel, while the weakening of the mixing effect of the grid spacer on the coolant flow is the main factor to lead the boiling crisis upstream of the grid spacer. By comparing the effects of thermal properties of different materials on CHF, it was found that when the thermal conductivity ratio of the two materials is 1.08, CHF is almost identical under the condition of a reference heat flux of 0.1 MW/m2 as a step size. In addition to thermal properties, further attention may be paid to the influence of surface characteristics of materials on CHF in future work. [ABSTRACT FROM AUTHOR]

• This paper studies the space nuclear energy propulsion system using fission fragments for propulsion. • The uniform magnetic field model and the magnetic collimation structure model are proposed for analysis. • RMC and MATLAB were used to simulate the critical characteristics and propulsion performance of the system, respectively. • Using Am-242 m as fuel, the calculation finds that the system can achieve criticality with good safety. • The I SP of the magnetic collimation system can reach 6*105s, which is 40 % higher than that of the uniform model. Deep space exploration holds immense significance for the future of space exploration, and it requires propulsion systems with high specific impulse characteristics to accomplish long-duration space missions. This paper focuses on studying a space nuclear propulsion system based on the direct propulsion of fission fragments. Am-242 m is utilized as fuel, and two system models, namely uniform magnetic field and magnetic collimation system, are proposed. The system's critical and propulsion characteristics are analyzed using RMC and MATLAB, respectively. The calculation results demonstrate that the system with magnetic collimation exhibits superior comprehensive performance, with a specific impulse of approximately 6 × 105s. By adjusting the system materials and parameters, the k eff of the system is optimized to reach about 1.02, while the utilization rates of fission fragments and energy are approximately 20 % and 10 %, respectively. The results also indicate that the fission fragment direct propulsion system possesses excellent safety features and can play a crucial role in deep space exploration. Furthermore, the paper highlights the exploration of propulsion systems utilizing Pu239 and U235 in the future, as well as the investigation of enhancing the fission fragment escape rate and energy utilization rate. [ABSTRACT FROM AUTHOR]

The steady state neutron transport equation can be studied resorting to different eigenvalue formulations, useful to investigate the criticality state of a nuclear system. In this respect, the knowledge of the eigenvalue spectra featuring the main eigenproblems is fundamental, from a practical standpoint, to guide the eigenvalue solvers searching for the dominant eigenpairs. The study of the eigenvalue spectrum shape on the complex plane is also relevant to better understand the physico-mathematical aspects of the neutron transport numerical approximations. This paper investigates the features of the eigenvalue spectra with respect to various parameters like modelling approximations, departure from criticality and spatial heterogeneity in a one-dimensional slab geometry framework. The sensitivity of the energy spectrum characterising the fundamental eigenfunction is then evaluated with respect to the model adopted and to perturbations in the input parameters. • The Neutron transport equation can be arranged into different eigenvalue problems for approaching the critical state. • The knowledge of the eigenvalue spectra featuring the main eigenproblems is fundamental for both theoretical and practical evaluations. • This paper investigates the features of the eigenvalue spectra with respect to various parameters like modelling approximations, departure from criticality and spatial heterogeneity in a one-dimensional slab geometry. • The sensitivity of the energy spectrum characterising the fundamental eigenfunction is evaluated with respect to the model adopted and to perturbations in the input parameters. [ABSTRACT FROM AUTHOR]

• The work proposes an optimization algorithm combining grey wolf optimization (GWO) and particle swarm optimization (PSO) to compute the hyperparameters of support vector machine (SVM) classifiers to improve the classification accuracy of SVM classifiers. • This paper proposes a boundary data extraction method combining classification and regression algorithms for the classification boundary of adaptive sampling methods. • Combining the above two methods this paper proposes an optimized adaptive sampling method and establishes a calculation process to achieve the improvement of the calculation accuracy of the adaptive sampling method. The risk-informed safety margin characterization (RISMC) method utilizes a probabilistic safety margin (PSM) comparison between a load and capacity distribution, rather than a deterministic comparison between two values. The RISMC method can be effective in exploiting the potential safety margins and improving the economics of nuclear power plants (NPPs). However, the high computational cost of the coupling calculation involved in RISMC significantly limits its application. This paper presents an optimized adaptive sampling (OAS) method to efficiently calculate the PSM of NPP accidents. The OAS method uses the support vector machine (SVM) method to predict the limit surface and grey wolf optimization combined with particle swarm optimization to tune the SVM hyperparameters. Additionally, OAS uses the Gaussian regression algorithm to modify the limit surface. The OAS method was firstly tested on three synthetic test functions and then generalized to station blackout (SBO) accidents of China Pressurized Reactor (CPR1000). The SBO results showed that the OAS method reduces the relative error between calculated and true values to about 1 % and improves the calculation efficiency by 5 times. [ABSTRACT FROM AUTHOR]

• The effects of temperature variation on the thermal consolidation characteristics of unsaturated soil are investigated. • This study takes into account the effects of a semi-permeable boundary, which encompass various different boundary scenarios. • In order to thoroughly investigate these characteristics, the study employed numerical modeling techniques. • The paper also considers the influence of depth-dependent linear initial conditions on the thermal consolidation process. This paper focuses on investigating the one-dimensional thermal consolidation of unsaturated soil. Through the utilization of the Laplace transformation, decoupling method, and inverse Laplace transformation, the semi-analytical solutions for excess pore pressures and ground settlement are deduced, particularly in the context of a single-sided semi-permeable boundary. To further investigate the thermal consolidation characteristics in unsaturated soils, numerical modeling is employed, encompassing both single-sided semi-permeable and symmetric semi-permeable boundaries. Consequently, the proposed solution is subjected to a rigorous comparison and analysis against existing published solutions and numerical results, thereby establishing a remarkable degree of consistency. A comprehensive parametric analysis is conducted to discuss the thermal consolidation behavior, revealing the temperature effect can lead to negative pore-water pressure at the end of dissipation and a rebound phenomenon of the ground settlement. Notably, the variation in temperature affects the final settlement of the soil. Furthermore, alterations in top permeability parameters influence the dissipation of excess pore pressures within specific depth ranges, wherein the influence depth increases as the boundary permeability strengthens. Of particular interest is the applicability of the current solution, which accommodates the one-dimensional thermal consolidation model for unsaturated soils characterized by varying permeability topsides and depth-dependent linear initial conditions. [ABSTRACT FROM AUTHOR]

• Two coupled parallel frameworks for shell and tube heat exchangers are proposed. • Fully three-dimensional parallel computation on both sides of the heat exchanger is realized. • One-to-one method achieves 85% parallel efficiency in 14 cores computation. • One-to-many method improves computational efficiency by about 20 times over serial in 28 cores computation. Shell and tube heat exchanger is widely used as a common type of heat exchanger. When the size and the number of heat exchanger tubes are enormous, it is difficult to use the direct modeling method for numerical simulation of heat exchangers. The simplified modeling method based on porous media utilizing user-defined function (UDF) has been widely applied. However, when the three-dimensional (3D) modeling is required on the tube side, the parallelization problem still restricts the computational efficiency of numerical simulation. In this paper, a one-to-one parallel method was first developed to solve the problem of coupled parameter transfer between the primary and secondary sides in the parallel framework, which was tested to achieve a parallel efficiency of 85% under 14 cores. However, the method acquires complicated data transfer between nodes, large memory occupation, and the modeling of the tube side is different from the actual structure under certain circumstances. To solve the problems, a one-to-many parallel method is proposed, and the computational efficiency is tested to be enhanced by about 20 times under 28 cores compared with single core, with a parallel efficiency of 70%. The two methods proposed in this paper are of great significance to the parallel transformation of the future 3D full-size code for shell and tube heat exchangers development and the calculation efficiency improvement. [ABSTRACT FROM AUTHOR]

• 4 types of high-temperature nuclear loop with movable equipment elevated were compared. • Influence evaluation of U-type layout design parameters was made. • Application of U-type layout to the TMSR-LF1 was conducted. High temperatures could make nuclear power more efficient but also pose a huge challenge to the layout design. In this paper, based on the characteristics of the Thorium Molten Salt Reactor-Liquid Fuel 1 (TMSR-LF1) high-temperature nuclear loop, the schema design and optimization for four typical types of Linear-type, L-type, Z-type, and U-type loop layouts with moveable equipment were carried out. Then, the influencing factors for relevant design parameters of the proposed U-type layout were discussed. Finally, the compact layout with moveable equipment of the TMSR-LF1 loop was determined, and the finite element calculation results considering the creep-fatigue effect of the designed loop pipes met the requirements of the ASME-III-5-HCB code. Moreover, the experimental displacement of the molten salt pump in commissioning condition was basically consistent with the calculated results, which verified the rationality of the proposed loop layout. The research in this paper has practical engineering implications for the design of a high temperature nuclear loop. [ABSTRACT FROM AUTHOR]