Sensitivity and uncertainty analysis with respect to nuclear data has become an important part of analyzing new systems, known as Generation IV, to identify the sources of uncertainties. While there is a rise in interest in the field, the number of tools for the analysis is limited, and they are generally proprietary or specific, or they require a license to be used. This raises the entry threshold and complicates the analysis. To mitigate this issue, a Python-based open-source tool for Sensitivity And UNcertainty Analysis (SAUNA) is presented, which leverages the capabilities of well-established Python packages. It includes all the basic needs from preparing the covariances to uncertainty propagation and computing target accuracy requirements to make the process seamless. Furthermore, interoperability-related features are implemented, such as importing covariances from ENDF-, COVERX-, ABBN-, COMMARA-formatted covariances and sensitivities from Serpent or the “.sdf” files of SCALE.

In the absence of operational experience with Generation IV reactors and limited experimental data, developing test tasks is crucial for ensuring the safety and reliability of novel reactor designs. This study presents a test task for coupled neutronic and thermal-hydraulic analysis of a lead-cooled fast reactor (LFR) fuel unit cell in the BREST-OD-300 core. Neutronic simulations using the Monte Carlo-based MCU-FR code identified an optimal axial discretization of 44 layers to reduce uncertainties in power density distribution. The resulting axial power density profiles were used in CFD thermal-hydraulic modeling with LOGOS software to assess the impact of different turbulent Prandtl number correlations on axial temperature profiles. The study shows that the choice of turbulent Prandtl number introduces up to 1% model-uncertainty in temperature predictions due to differences in turbulence modeling assumptions. In addition, the application of fully coupled neutronic and thermal-hydraulic feedback results in a deviation of above 2% in average fuel temperature compared to the non-coupled solution. This test task is aimed at enhancing the accuracy of reactor simulations and safety assessments while contributing to the development of a benchmark suite for LFR reactors, supporting cross-verification of computational tools and methods.

Nuclear data uncertainties are considered a main limiting factor in proving safety of advanced reactors. A reasonable amount of available information should be applied to sensitivity and uncertainty analyses for this, and this work considers the influence of usually omitted angular distribution uncertainties. A sensitivity and uncertainty analysis of MOX3600 is conducted using Serpent. The angular distribution influence is assessed by applying covariances of the state-of-the-art libraries. The sensitivity analysis shows that quantifying the influence of higher-order Legendre moment uncertainties on the eigenvalue and void reactivity is of minor importance for MOX3600. The uncertainty analysis shows the P1-coefficients for 238U and 16O introduce uncertainties to the eigenvalue similar to 56Fe and should be accounted for. A library comparison reveals TENDL-2021 and ENDF/B-VIII.0 introduce uncertainties of a few hundred pcm to the eigenvalue, and the data should be reassessed, while the other libraries give an uncertainty of several tens of pcm.

A systematic error is present in reactor calculations performed with the aid of the MCU-code. This paper introduces an algorithm that in calculations by the series method makes it possible to determine the number of neutron generations in the NBAT series and the number of neutrons in the NTOT generation, which reduces the systematic error in the calculation of functionals and their variances. The described procedure includes determining the NBAT-dependence of the standard deviation, repeated calculations of the required functional to refi ne and correct NBAT, and calculations to refi ne NTOT. The algorithm is demonstrated on the example of calculating the effective multiplication factor and fission reaction rates for the VVER-1000 unit cell model with the fuel column divided into 10 layers along the height. It is shown that in order to significantly reduce the systematic error of the local rate of fission reactions it is necessary to play more than 1000 particles in each series and variances in more than 200 series in each neutron generation.

The paper presents an assessment of the effect caused by technological uncertainties on keff using the example of the test problem for a MET1000 fast sodium reactor with metal fuel. It is proposed to perform this assessment using nuclear sensitivity factors by random sampling or direct perturbation without requiring multiple calculations. The TSUNAMI-3D module of the SCALE program was used to analyze the sensitivity to nuclear data uncertainty. The obtained constant and technological uncertainties based on ENDF/B VII.1 equal 1.15 and 0.6

There is a rise in sensitivity and uncertainty analyses with respect to nuclear data, which are a basis for a neutron transport simulation. The analyses are primarily concerned with the state-of-the-art reactors, known as Generation IV. They provide assessments of nuclear data performance, key nuclear data, and their uncertainty influence on functionals such as the eigenvalue, reactivity coefficients, delayed neutron fraction, etc. In this work, recent advances in sensitivity and uncertainty analysis with corresponding transport codes are reviewed. Furthermore, the available results of the analyses are compared and discussed. As a result, this review summarizes current analyses performed for the advanced systems in recent years. The gaps in the Generation IV system analyses are noted including SCWRs, MSFRs of different types, and other systems. Besides, the gaps in nuclear data of state-of-the-art libraries are discussed. Based upon this, these gaps can become a part of analyses for identifying future nuclear data needs.

The excess reactivity in WWER-type pressurized water reactors is compensated using strong neutron absorbers. This leads to useless neutron consumption and reduce the breeding ratio and fuel burnup. In this work, one of the methods of spectral regulation of reactivity margin for fuel burnup was considered, namely, variation of the water-to-fuel ratio by inserting hollow cylindrical zirconium rods between fuel elements in the fuel assembly. Calculations were performed for a thorium–uranium-233 fuel. The range of change in the water-to-fuel ratio was estimated as a function of the diameter of inserted hollow zirconium rods. A comparison was made with the results of similar calculations for a uranium fuel at equal (3.7

This paper analyzes the methods and technologies for assessing the method of formation, composition, characteristics and features of corium, which is a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of an accident accompanied by partial or complete core melting. The study is based on data from the study of corium formed as a result of the accident at the Fukushima Daiichi nuclear power plant, which are in the public domain and are the result of the work of many scientific organizations around the world. Corium research is one of the main issues in the framework of improving nuclear safety in the future and is one of the objectives of the successful procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant. Without a detailed analysis of the neutronic, materials science, gravimetric and other characteristics of the corium, as well as the creation of a complex model of the corium that combines these data, it is impossible to organize an efficient and safe process for removing nuclear materials from the damaged units of the Fukushima Daiichi nuclear power plant. The objective of this work is to combine the existing research results into a data set that allows modeling of the corium using neutronic calculation codes and includes such data as the size, density and morphology of corium samples and their approximate nuclide composition. Such modeling allows not only to perform tasks related to increasing the level of safety in the implementation of the procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant, but also to serve as an international benchmark for modeling a mixture containing nuclear materials.

Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pu(n, gamma), 238U(n, gamma), and 238U(n, n'). The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.

This chapter examines Small Modular Reactors (SMRs), which are modular-type nuclear reactors with installed capacities ≤ 300 MWel with claimed features of “modularity” in design, production, and/or construction, and Small- and Medium-size Reactors (S&MRs), with installed capacities ≤ 300 MWel (Small) and > 300–700 MWel (Medium-size), many having claimed features of “modularity” in design, production, and/or construction. The requirements and objectives for any and all new nuclear reactors of any and all sizes are given as: safer than previous “generations”; having low financial risk exposure and capital cost; ease and speed of build; readily licensable; simple to operate and secure; assured fuel supply and sustainability; providing social value and acceptance; and still being competitive. Existing SMRs and S&MRs are tabulated by type, country, and status. Although many SMR designs and concepts have been proposed, Russia is the first country in the world to develop, design, and put into operation two SMRs, and Russian technology is examined in detail in this chapter, with numerous diagrams and photos of various systems provided.

Virtual reality (VR) technology is now being adopted in many industries, including entertainment, medicine, science, and engineering. In the nuclear field, the primary purposes of VR are: reducing radiation dose rates, security of nuclear facilities, visualization of physical processes, and training of personnel. Additionally, VR is a much cheaper alternative to expensive and license-requiring experimental nuclear facilities. This work focuses on reconstructing the workroom with the Uranium-Water Subcritical Assembly (UWSA) located at the National Research Nuclear University MEPhI to determine the optimal uranium–water ratio associated with this assembly in virtual reality. The creation of the virtual analog using Unreal Engine 4 was introduced to integrate the physical model into the virtual environment. The neutronic model of the UWSA was obtained by the MCU code. A similar model was generated by the Serpent code for verification purposes. Additional functions such as neutron flux visualization, radiation dose rate distribution visualization, and dose accumulation mechanics were introduced into the project to improve the quality of education. Visualization of both neutron flux in the assembly and gamma radiation distribution in the workroom was performed using particle systems and volumetric fog based on calculated and experimental data. Operating experience feedback was introduced to prevent or minimize difficulties that may occur in the future by learning from events that have already occurred.

Currently, there is a lack of computer power to perform high-precision reactor core analysis. In full-scale simulation of nuclear reactor cores using the stochastic Monte Carlo method, there are a number of factors that increase the excessive computational load and make calculations difficult. Among them is the large flux attenuation, which can be observed in deep penetration problems. In order to improve the efficiency of Monte Carlo calculations, various reduction techniques are used, which make it possible to reduce the statistical uncertainty of the functional evaluation without increasing the number of simulated histories. This article is devoted to the study and testing of techniques for reducing the variance in the deep penetration problem. A test problem is formulated the solution of which will make it possible to demonstrate the possibility of using various techniques of nonanalog simulation. In order to determine the quantitative efficiency of variance reduction techniques, the FOM characteristic is considered, which is a function of the relative error in a flux estimate and the computational time of the simulation. The article deals with the techniques of nonanalog simulation implemented in the MCU and OpenMC codes. As part of the study, a module of the OpenMC code has been developed which makes it possible to automatically generate weight windows. It is shown that variance reduction techniques increase the calculation efficiency by several times. In particular, the weight window technique in OpenMC increased the efficiency of neutron flux estimation by seven times, while the number of simulated histories remained unchanged. The formulated recommendations can be used in the future for the calculation of full-scale models of cores of innovative nuclear reactors.

Professor Pavel Leonidovich KirillovProfessor Pavel L. Kirillov died on Oct. 8, 2021, on his 95th year after a life as a husband, father, and an internationally renowned scientist, researcher, and educator in the field of nuclear engineering, thermalhydraulics, heat transfer, and two-phase flow. He was passionate and dedicated in everything that he did and leaves an incredible legacy to the profession.He was born on Aug. 20, 1927 in Russia, and received his M.A.Sc. degree in thermal physics in 1950 (Moscow Power-Engineering Institute (MPEI) (Московский Энергетический Институт (МЭИ)), Faculty of Physics and Power Engineering (Физико-Энергетический Факультет), Ph.D. and Doctor of Technical Sciences degrees—in 1959 and 1969, respectively.Professor P. L. Kirillov was a Fellow of the International and National Engineering Academies; member of the Russian Nuclear Society and ASME; member of Scientific Councils of the Institute of Atomic Energy by the name of I. V. Kurchatov (ИАЭ им. И.В. Курчатова) (1985–1990) and A.I. Leypunsky Institute for Physics and Power Engineering (IPPE) (Физико-Энергетический Институт (ФЭИ)) (from 1975); member of the Journal Boards of the Atomic Energy (Атомная Энергия) (from 1977) and the ASME Journal of Nuclear Engineering and Radiation Science (from 2014).After graduating from the MPEI (МЭИ) in 1950, Pavel Kirillov has joined the IPPE (ФЭИ) (Obninsk, Russia), currently, State Scientific Centre of the Russian Federation—Leypunsky Institute for Physics and Power Engineering, Joint-Stock Company (IPPE JSC) (Акционерное общество «Государственный научный центр Российской Федерации – Физико-энергетический институт имени А.И. Лейпунского» (АО «ГНЦ РФ—ФЭИ»)) as a junior scientist in 1950 (he participated in construction and operation of the world's first nuclear power plant in Obninsk, AM-1 (“Atom Peaceful”–1 in Russian abbreviations (Атом Мирный)), which was commissioned at the IPPE on June 27, 1954), and worked there on various positions: Senior scientist (1953–1954); head of laboratory (1954–1969); head of branch (1969–1975); director of thermal-physics division (1975–1995); deputy director of thermal-physics division (1995–2010); advisor of the director of thermal-physics division (from 2010) and of the general director of IPPE JSC.He was an associate professor (1959–1965); professor (1965–1972); chair of the thermal-physics department (1972–1985); and chair of the nuclear-power-plant department at the Obninsk Branch of the Moscow Engineering Physics Institute (MEPhI) (Обнинский филиал Московского Инженерно-Физического Института (МИФИ)) (1985–1992). Professor Pavel Kirillov has prepared a large number of undergraduate and master-degree students; and 15 Ph.D. candidates.Knowledgeable, friendly and technically informed, Dr. P.L. Kirillov was a role model and mentor to numerous generations of researchers/scientists in nuclear engineering, thermalhydraulics, heat transfer, and two-phase-flow fields. He is definitely one of the most admired and ingenious researchers in these fields. His many researches and achievements include contributions in such special areas as molten-metals nuclear-reactor coolants; supercritical water; research (BR-10 (Fast Reactor (sodium-cooled)) and BOR-60 (fast experimental reactor (sodium-cooled)); power (BN-350 and BN-600 (fast sodium power reactors)), transportation (lead-bismuth-cooled), and spacecraft (BUK and TOPAZ) nuclear reactors. Professor Kirillov is well respected among his colleagues in the nuclear-engineering community all over the world despite being heavily focused on Russian developments to which he made major contributions. His 2009 text on “Hydrodynamic Calculations” (in Russian) covered and demonstrated his encyclopedic knowledge of fluid flow and heat transfer. The earlier 2007 major text “Thermophysical Properties of Materials for Nuclear Engineering” (in English) sets the standard for excellence, breadth and depth with not only essential basic data and tabulations, but includes fundamental design information for all types of reactors.For his outstanding work, Professor P. L. Kirillov was awarded with the following honored titles: Honored Scientist of Science and Engineering of the Russian Federation (1988) (Заслуженный деятель науки и техники РСФСР) and Honored Worker of the Atomic Industry of the Russian Federation (2016) (Заслуженный работник атомной промышленности Российской Федерации); and with three state orders and a number of state and jubilee medals.During his work at the IPPE (ФЭИ) and Obninsk Branch of MEPhI (МИФИ), Professor Kirillov has published over 350 technical publications including handbooks, reference books, textbooks, papers, inventions, and reports (see selected publications listed below). His professional contributions and critical thinking continued unabated and he was fully involved in the series of articles summarizing the status of nuclear energy in the world and its future prospects.Professor Pavel Leonidovich Kirillov was a respected technical leader, mentor, and friend to innumerable students, researchers, scientists, and engineers, and he will be sadly missed by all who had the privilege to know him. He was an outstanding contributor in every aspect of his prolific work and career in the true traditions of technical excellence and critical thinking, and his irreplaceable loss is deeply felt worldwide.The official information on Professor Kirillov can be found on the following website.1On behalf of the ASME J. NERS Editorial Board, colleagues, students, and friends from around the world:

The traditional methods for controlling reactor excess reactivity are expensive and have defects during reactor operation. By altering the fuel structure and fissile material distribution, this work presents a new technique for managing excess reactivity at the start of the fuel life. MCNPX code edition 2.7 was used to investigate the neutronic characteristics of the suggested blanket-seed assembly models and compare them with the reference PWR fuel assembly. Instead of U-238, Th-232 was suggested as a fertile nuclide. The impact of the proposed models on the infinite multiplication factor, fuel component concentrations, reactor-grade plutonium, minor actinides, radial flux of thermal neutrons, and the radial distribution of power has been analyzed. The safety parameters, including the control rods worth, Doppler effect, moderator temperature reactivity coefficients, and the effective delayed neutron fraction, have been studied to reach the most optimum fuel structure.

Fuel debris removal is the most challenging part of damaged nuclear power station decommissioning. It is important to carry out nuclear safety calculations accurately and quickly enough. Here, it was clarified that modern codes based on the Monte Carlo method were capable of performing neutronic analysis with the same accuracy and without significant differences in the results. The benchmark calculations were performed using three codes: MVP, Serpent, and MCU. In this study, the comparison focused on multiplication factor, neutron fluxes and reaction rates relative difference, and calculation time of many fuel debris particles system. Then the calculation results were used when codes comparing. It was shown that the calculation results showed good agreement between all codes. It was assumed that minor differences in the thermal range of neutron fluxes can be caused by different thermal neutrons scattering treatment for all codes. The study also showed that solving such problems requires significant computing power and time. It has been proven that the statistical geometry model in the MVP and the explicit stochastic geometry model in the Serpent have the possibility to provide solutions with the same accuracy, but much faster.

The compensation for the excess reactivity in the pressurized water reactors WWER is realized with high neutron absorber materials. The traditional excess reactivity regulation methods lead to unfeasible neutron utilization and reduce the breeding coefficient and fuel burnup. In the current work, the change of moderator-to-fuel ratio is investigated as one of the spectral regulation methods for excess reactivity control and its effect on the fuel burnup. Cylindrical Zirconium rods (Zr rods) are used to fulfill the moderator-to-fuel ratio change. The Zr rods are placed between fuel rods in WWER-1000 fuel assembly. The current work calculations are performed for the thorium fuel cycle (Th-U233). The change of the Zr rods diameter leads to the variation in moderator-to-fuel ratio. A comparison between the Zr rods as a reactivity regulator in WWER-1000 fuel assembly for both fuel cycles UO2 and Th-U233. The concentration of the fertile and fissile fuel components for both fuel cycles has been analyzed. The fissile isotopes accumulation coefficient can reach 0.75 with the decrease of the moderator-to-fuel ratio in the Th-U233 fuel cycle. The primary safety parameters such as the Control rods worth, Doppler Effect reactivity coefficient, and Moderator Temperature reactivity Coefficient have been studied at different moderator-to-fuel ratio values. The safety parameters in the Th-U233 fuel cycle have higher values more than the UO2 fuel cycle with the insertion of Zr rods. From the comparison between the Zr rods effect in both fuel cycles, it is clearly shown that Zr rods in the UO2 fuel cycle have a more influential role in regulating the WWER-1000 core reactivity compared with its effect in the Th-U233 fuel cycle.

In the current work, we introduce a new approach compared to solid-fuel reactors to load the minor actinides (MAs) into the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The proposed approach merges the advantages of both homogeneous and heterogeneous approaches. Among MAs nuclides, Am-241 and Np-237 are selected for transmutation due to their long half-life. We simulate two separate tanks; Pu + U tank and FPs tank. In this study, a tank is a right cylinder with a volume of 1.0 m(3). The Pu + U tank is used to store Pu and U isotopes extracted from the central channel of the SD-TM SR. However, the FPs tank is used to store all fission products (FPs) produced from the transmutation process in the central channel. The overall change in the actinides and FPs mass during the irradiation has been calculated using direct SERPENT-2 calculations. The results show that the transmutation ratio of Am-241 and Np-237 reaches 98.5% and 93.2%, respectively after 1500 days of irradiation. We notice that the major isotope in the Pu + U tank is Pu-238. Under Am-241 irradiation, our proposed approach offers approximate to 0.3 kg of Pu-238 after 1 year of operation. However, under Np-237 irradiation, 2.5 times more Pu-238 can be extracted after the same period of operation. The produced Pu-238 can be used in the radioisotope thermoelectric generators (RTG, RITEG) and radioisotope heater units.

In the design and operation of nuclear power plants, one of the most important tasks is to assess the radiation protection of the reactor. Currently, the most widespread are deterministic (method of discrete ordinates) and stochastic computational methods for evaluating functionals. At large attenuations of the neutron flux (by 5-15 orders of magnitude) the deep penetration problems require large computational costs. The most accurate simulation of radiation transfer is achieved by using precision programs that implement the Monte Carlo method with a continuous energy dependence of the cross sections. A detailed description of the geometry and the use of continuous cross sections for particle interactions in calculations lead to high computational costs. To improve computational efficiency, there are variance reduction techniques (non-analog modeling). In this paper the possibility of using non-analog modeling in MCU-FR program by calculating the protection of the fast reactor full-scale model with a lead coolant is considered. The volume-integral neutron fluxes were estimated at points located in a long distance from the center of the reactor core. Analysis results were shown the significant reduction of the variance in the reactor shielding by using the non-analog Monte Carlo method.