Research reactors play a crucial role in advancing scientific understanding, medical diagnostics, and nuclear technology development. This review article provides a comprehensive examination of research reactors, emphasizing classification and highlighting the unique features of periodic pulsed reactors. The classification of research reactors is elucidated, distinguishing between stationary and pulsed reactors. Stationary reactors serve diverse purposes, while pulsed reactors offer distinct advantages in certain applications as time-of-flight research and neutron generation lifetime calculation. Particular attention is given to research fast periodic pulsed reactors in Russia, showcasing their significant contributions to scientific research and technological innovation. Detailed discussions include the IBR-1, IBR-30, IBR-2, and NEPTUNE reactors, elucidating their design, capabilities, and scientific achievements. Furthermore, the article explores the utilization of Neptunium-237 as a promising nuclear fuel for fast pulsed reactors, highlighting advantages such as low neutron generation time, high neutron flux, and the ability to burn up part of dangerous nuclear waste. Overall, this review consolidates essential knowledge on research reactors and underscores the importance of ongoing research and development efforts, particularly in harnessing innovative fuels like Neptunium-237, to meet the evolving demands of science and technology.

A self-sustaining reactor system with both a high conversion ratio (CR) and fuel discharge burnup (B d ) is essential for enhancing the utilization of nuclear fuel and improving its natural resources. The mechanical spectral shift control (SSC) method can improve the efficiency of nuclear fuel utilization by removing, or at least greatly lowering, soluble and burnable neutron poisons during reactor operation. The implementation of the proposed mechanical SSC method involved the on-power variation of the moderator-to-fuel volume ratio (V M /V F ). This was achieved by means of movable natural uranium spectral shift control rods (NU-SSCRs) distributed uniformly within the fuel assembly. In this work, the neutronic characteristics of the VVER-1000 fuel assembly benchmark with mechanical SSC regulation have been investigated, compared to the traditional poison method described in the OECD/NEA benchmark model. We adopted three different sequences of NU-SSCRs withdrawal strategies at various burnup points, each associated with different V M /V F , during one batch fuel loading pattern. The findings revealed a notable improvement of approximate to 40 % in the discharge burnup across the three different withdrawal strategies, alongside the attainment of a CR value of 0.82 compared to the benchmark poisoned model. Additionally, the mechanical SSC regulation has effectively suppressed the excess reactivity at the beginning of the cycle compared to boric acid (H 3 BO 3 ) and gadolinium oxide (Gd 2 O 3 ), which are incorporated in the benchmark model by 754 pcm. The results showed a higher accumulated mass of plutonium in the fuel assembly of 3.3, 5.2, and 6 kg for the three different sequences of withdrawing strategies, respectively, at the end of the cycle. Furthermore, the void reactivity coefficient has been investigated for the three different withdrawal strategies in both partially voided (20 % void) and completely voided (100 % void) systems.

The performance of the spectral shift control (SSC) method is evaluated and compared to the conventional poison method in the VVER-1000 fuel assembly design. The SSC method can be implemented by gradually adjusting the ratio of heavy water to light water moderator (D2O/H2O) during the fuel cycle. In this study, the efficiency of using the SSC design with or without a thermal absorber (gadolinium) is investigated. We apply the SSC with both 12 burnable absorber rods containing 4.0 wt.% Gd2O3 (Case 1) and without Gd2O3 (Case 2). The neutronic calculations indicate that the discharge burnup is enhanced by 60% and the conversion ratio (CR) is increased by 64.4% at the beginning of the cycle (BOC) compared to the benchmark data. The breeding of Pu239 and Pu241 is extended to 33.7% and 29.5%, respectively, for the SSC design case (2), and better utilization of U-235 and U-238 has been achieved compared to BM.

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.

The spectral shift control (SSC) is a promising method to decrease the fuel cycle cost and enhance the utilization of fuel resources (U, Pu) in nuclear reactors. In the chemical SSC method, reactivity control is accomplished by varying the ratio of heavy water to light water moderator (D2O/H2O). In the current work, we apply the chemical SSC method to analyze the VVER-1000 LEU fuel assembly from the OECD benchmark. Additionally, we compare the chemical SSC method to the conventional poison method presented in the benchmark. The results showed that the discharge burnup was improved by 60% and the conversion ratio value reached 0.83. The k(infinity) was suppressed at the beginning of cycle by -8% compared to the benchmark model. Furthermore, the dedication of 12 burnable absorber rods with 4.0 wt% Gd2O3 would save the D2O quantity during the adjustment process resulting in reducing the capital cost of the SSC method. However, it's found that the physical changes associated with applying the SSC method including a high contribution of Pu-239 and Pu-241 to power generation, the low effect of Xe-135 and Sm-149 on the reactivity system, and reduction of control rod worth should be considered in the safety analysis of the SSC design. (C) 2021 Elsevier Ltd. All rights reserved.

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.

The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). The reactivity of the SD-TMSR core is controlled through two systems of control assemblies: (1) the Control Safety Devices (CSD) and (2) the Diverse Safety Devices (DSD). In the present work, the control rods are natural B4C and B4C-90 (with 90% weight content of the main absorbing 10B isotope). Since the numbers and distribution of control assemblies in SD-TMSR have not been studied previously, we proposed a unique distribution as a starting point for this analysis. The distribution of these 25 fuel assemblies with control rods in the SD-TMSR core and their numbering scheme were presented in this paper. Additionally, excess reactivity, control rod worth, and shutdown margin were calculated using Monte Carlo code Serpent2. Analysis results showed that the proposed placement of the control rods will make it possible to compensate for excess reactivity during fuel burnout and emergency shutdown of the reactor.

Recent studies on Molten Salt Reactors (MSRs) showed that the excess reactivity at the beginning of the operation is large for many fueling strategies and must be compensated by a reactivity control system. The current work introduces a reliable safety system based on control rods in addition to the online feed system reactivity control in the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR). Three different initial fissile loadings are considered: U-233, reactor-grade Pu, and transuranic (TRU) elements as a startup fuel. We applied six different absorbing materials to investigate the main operational and safety parameters in the SD-TMSR: natural B4C, enriched B4C with 90% B-10, HfB2 , HfH1.62 , Eu2O3, and Gd2O3 . The present work focuses on control rod design, integral and differential control rod worth, shutdown margin, and shadowing effects at steady-state. We employed the SERPENT-2 Monte-Carlo code to calculate the reactivity worth and analyze the performance of the reactivity control system. We showed that U-233 and reactor-grade Pu startup cores maintain adequate shutdown margin with all considered absorbers. Finally, this paper proposes a design of control rod clusters that compensate the excess reactivity of the SD-TMSR loaded with different initial fissile material. (C) 2020 Elsevier Ltd. All rights reserved.

The long-lived minor actinides (MA); americium, neptunium, and curium are main contributors to the long-term radiotoxicity of used fuel. Thus, the transmutation of these MAs is considered as an alternative to direct burial. Until now, no unambiguous internationally recognized quantitative criterion for the effectiveness of MA transmutation has been developed, although this would be highly desirable. The absolute and relative decrease in the total mass of MA is completely inadequate, since they ignore the accumulation of higher radiotoxic MA from the transmuted nuclide. In this paper, we propose a new criterion for the efficiency of MA transmutation in nuclear reactors and demonstrate its efficiency when comparing two molten salt reactors; Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and Small Molten Salt Fast Reactor (SMSFR). In addition, the proposed criterion takes into account the mass of all useful MA, short-lived MA, and short-lived fission products (FPs). We present a new approach to loading MA in SD-TMSR and SMSFR. The total change in the mass of actinides and FPs during irradiation was calculated using SERPENT-2 Monte Carlo code. The results show that the transmutation efficiency of 241Am (a major candidate for transmutation) in SD-TMSR is much higher than in SMSFR. After 1500 days of irradiation, the transmutation efficiency reaches 82.6% for SD-TMSR, but for SMSFR it reaches 52.5%.

Molten Salt Reactor (MSR) is one of the six innovative systems, which has been selected by the international Generation IV forum as a candidate for the next-generation reactors. In this paper, the transmutation of241Am, the main contributor to the waste toxicity in the nuclear fuel cycle, in both thermal- and fast-spectrum molten salt reactors is compared. The Am-cylinder is irradiated in the center of the thermal (SD-TMSR) and fast (SMSFR) spectrum MSRs, respectively, and during a long period of time. The total neutron fluxes in SD-TMSR and SMSFR are 4.1×1014 and 1.8×1015 n.cm-2.s-1, respectively. The overall change in actinides and fission products (FPs) mass during the irradiation has been investigated using direct Serpent2 calculations. Under fast spectrum irradiation, transmutation leads to produce short-lived FPs (by fission) rather than the accumulation of heavier actinides. The ratio between the maximum and initial released energy from all isotopes in the Am-cylinder has been calculated in this paper.

The Single-fluid Double-zone Thorium Molten Salt reactor (SD-TMSR) has a promising future. Inherent safety, liquid fuel, and possibility of online reprocessing and refueling are unique potentials of the SD-TMSR. In the present study, the Monte Carlo code Serpent2 has been chosen to model the full core of the SD-TMSR. In addition, online reprocessing and refueling has been modeled based on subroutine provided by Serpent2. During the burnup time (10 years), the total mass of the fuel inside the core has been checked and found to be almost constant. Furthermore, we controlled the reactivity by adjusting the feed rate of fissile and/or fertile materials. The change of Keff, breeding ratio, Th and U-233 refill rates with burn-up time have been investigated. Additionally, the build-up of U-233 during 10 years has been calculated.

Thermal-spectrum molten salt reactor (MSR) concepts usually adopt graphite as a neutron moderator. Although graphite as a moderator has many advantages, it also has drawbacks, including a relatively short lifespan (it has to be replaced), positive temperature feedback, and loss of impermeability due to expansion. Replacing graphite with heavy water in MSRs can effectively solve the problems introduced by graphite. The SD-HWMSR is a Single-fluid Double-zone Heavy Water-moderated Molten Salt Reactor. Optimization of SD-HWMSR’s fuel channel pitch and radius has been conducted in this work. As a result, the SD-HWMSR with a high breeding ratio (1.07465 ± 0.00060), low initial 233U loading (1.43 t), and negative temperature and void reactivity coefficients is put forward. The SERPENT-2 is used to analyze the neutronics parameters of the reactor design. The current work investigates the change in the multiplication factor, breeding ratio, and the accumulation of significant nuclides in the core. The suitable 232Th and 233U refill rates needed to maintain criticality and enable analysis of the whole core of SD-HMMSR are thoroughly determined in this study. Uranium-233 and thorium-232 both have a maximum refill rate of 3.49 and 3.20 kg/d, respectively. While the average refill rate for 233U and 232Th during the 60 years of operation is 1.90 and 2.35 kg/d, respectively. The net production of 233U rises with time, and by the end of the 60 years, it is around 1.98 t. The doubling time for the SD-HWMSR is 31 years, which is consistent with previous results.

BackgroundHepatitis C virus (HCV) is a major health problem worldwide (Omran et al., 2018).The highest prevalence of HCV infection is present in Egypt, with 92.5% of patients infected with genotype 4 (Smith et al., 2014).In 2015, an Egyptian demographic survey study stated that 6.3% of the population have been tested HCV Antibody (HCV Ab) positive and the rate of infection increases with age reaching 27.6% in those aged (55-60 years) (Ministry of Health and Population [Egypt], 2015; Blach et al., 2017).Egypt launched a large program for controlling HCV in the country which utilized a wide network of specialized viral hepatitis treatment centers covering

The long-lived Minor Actinides (MAs):Np- 237, Am-241, Am-243, Cm-243, Cm-244, and Cm-245 are responsible for the effective dose and heat generation after direct disposal in deep geological structures. Thus, long-lived MAs represent a major burden of nuclear power. The long-lived MAs have not yet been utilized as nuclear fuel. Therefore, the transmutation of these MAs is proposed as an alternative to the direct final disposal. In the current work, we analyze and compare the MAs transmutation performance in the critical Singlefluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and Small Molten Salt Fast Reactor (SMSFR). We study the variation of the K-eff and core reactivity with different MAs loadings, the neutron spectrum shift, the time evolution of MAs and major nuclides inventories, and the transmutation ratio (TR). The TR of the long-lived MAs is calculated by using the SERPENT-2 Monte-Carlo code. The total neutron flux in the SD-TMSR and SMSFR can reach 4.1 x 10(14) and 1.8 x 10(15) n.cm(-2).s(-1), respectively. The results show that the SD-TMSR consumes about 50% of the generated Pu isotopes in the fuel salt, however, the SMSFR consumes about 86.5% of the generated Pu isotopes. During burnup, we apply the online reprocessing and refueling, therefore, the core is maintained critical and the total fuel mass in the core and blanket is almost constant. The results demonstrate that both reactors effectively transmute( 237)Np, (241)AM,(243)AM, and Cm-241, meanwhile, the SMSFR has a higher TR than the SD-TMSR. The TR of the total MAs reaches 54.84% and 87.97% in the SD-TMSR and SMSFR, respectively. (C) 2020 Elsevier Ltd. All rights reserved.

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/U-233 fuel cycle with U-233 used as startup fissile material. Since U-233 does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GWd/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and U-233. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of U-233 are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactor grade Pu helps transition to the thorium fuel cycle within a relatively short time years) compared to 26 years for U-233 startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time years) without any external feed of U-233. (C) 2020 Elsevier Ltd. All rights reserved.