Temperature Coefficient Of Reactivity Research Articles

Analysis of the Atomic Ratio of H and Zr Effect on the Neutronics Parameters of ZrH Moderated Space Nuclear Reactor

Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial keff, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial keff increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.

Multipoint kinetics model with power reactivity defect for the axial offset control in the VVER-1200 nuclear reactor during the load following mode of operation

This article proposes the multipoint kinetics model consisting with different number of point kinetics model (two points, four points, six points, eight points, ten points) in the axial direction for the VVER-1200 nuclear reactor. Each node is coupled with others through the coupling coefficients determined from the diffusion approach. For more precise description of the dynamical modes of the reactor operation, the proposed model integrates the power reactivity feedback derived from temperature reactivity coefficients and Mann’s thermal hydraulic model which assumes one fuel node adjacent to two coolant nodes. On the model with four axial points, additionally was tested the influence of the different number of delayed neutrons groups on the accuracy results and model running time during the load-following mode of operation. Moreover, the novel model of the control rods is introduced, utilizing a combination of sign functions to sequentially influence all nodes during insertion or withdrawal. The computational results show that the accuracy of the proposed model is satisfactory, and general assumptions about transients align with their physical definitions. This research contributes to the advancement of the point-like nuclear reactor modeling for improvement of the automatic power controller synthesis.

Startup and Multiphysics Analysis of a Compact kW-Class Thermal Spectrum Microreactor

Self-regulating compact nuclear microreactor concepts are being developed for use in space and at terrestrial remote sites. The current emphasis is on the development of high-assay low-enriched uranium–fueled reactors that rely on metal hydrides to achieve size, specific weight, and power parity with the legacy highly enriched uranium systems. In the case of terrestrial applications, metal hydride–moderated systems can also improve economic feasibility. The neutronic design of metal hydride–moderated high-temperature reactor cores is complicated by the need to optimize and synchronize delayed spatial and temporal reactivity feedback from outer reflectors. Zebra is a thermal spectrum proof-of-principle core designed for demonstrating the reactor dynamics of hydride-moderated spectrum reactors, which is studied in this work. The reactor, including the fuel, yttrium hydride moderator, heat pipes, central control rod, structural supports, and the beryllium reflector, weighs ~425 kg. The design shares several features common with nuclear criticality test geometries routinely used in National Criticality Experiments Research Center (NCERC) experiments such that a prototype of the reactor can be readily assembled and tested. Additional features, such as heat pipes and hydrogen barrier clads, were added to minimize the potential for large thermal gradients that in turn could induce hydrogen loss or structural deformation during extended periods of operation at ~1000 K. In this work, normal and off-normal state performance of the Zebra rector core was analyzed using MCNP and Abaqus-based Reactor Multiphysics (MARM) software. Up-to-date nuclear, thermal, and mechanical performance data were used to characterize the performance of the yttrium hydride moderator. Steady-state analyses established that at a postulated power between 20 and 50 kW(thermal), the reactor core is nearly isothermal irrespective of quality of conduction coupling between heat pipes and fuel plates. Additionally, heat pipe failure modes, including simultaneous failure of all heat pipes in a quadrant of a reactor, were examined to bound the maximum credible temperature spike in the reactor core for extreme off-normal operating conditions. Finally, we detail the startup design challenges for the hydride-moderated thermal core, and analyze load-following cases to achieve “self-regulation” using the Dynamic Analysis of Reactor Transients module of MARM. The reactor is self-regulating with a reactivity temperature coefficient of ~−1 pcm/K at the operating point based solely on nuclear cross-section feedback. It can be further strengthened using additional spectral shift neutron absorbers and incorporating design features that enhance core expansion. This work captures that hydride-moderated systems are feasible for various self-regulating applications once systematic checks are verified in order to achieve a well-engineered core design.

A theory and analysis of the impact of gas in the dynamical behavior of the molten salt research reactor leading to the computation of the "gas coefficients of reactivity"

Background Molten salt reactors, and other types of circulating, liquid fueled, nuclear reactors contain a certain amount of gas entrained in their liquid nuclear fuel. This gas induces an effect on the nuclear and dynamical behavior of the reactor as a whole. Gas voids respond to variation in temperature and pressure differently than liquids. When the gas voids in the reactor working fluid expand, the nuclear fuel is pushed from the core. Likewise, when the gas voids contract, more nuclear fuel enters into the core. Methods This paper examines the interplay of gas void fraction and reactivity in a molten salt reactor, and attempts to elucidate the dynamical response of the void fraction and the reactivity of the system to perturbation in system temperature, pressure, and gas quantity. A theory is presented that aims at describing the relationship between reactivity and gas behavior. This theory is then applied to the Molten Salt Research Reactor (MSRR) design, a facility currently under construction at Abilene Christian University campus. Results A result of this paper is the temperature and void fraction parameterized gas coefficients of reactivity for the Molten Salt Research Reactor. Conclusions The presence of voids accounts for 5-30% of the total temperature coefficient of reactivity, demonstrating their non-trivial contribution. Additionally, the study emphasizes the importance of considering gas content in MSR physics, especially in the context of pressure transients and system reactivity during pump trips. The initial system pressure, particularly in designs like the MSRR operating at sub-atmospheric pressures, is crucial due to its influence on reactivity changes during rapid pressure increases.

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

Potential of mixed oxide fuel for higher fissile conversion ratio in light water reactors

The prospect of adopting various mixed oxide fuels to increase the conversion ratio (CR) of conventional LWRs has been explored in this study. The Westinghouse AP1000 fuel assembly (FA) was taken as the reference design. While the geometry of the FA was kept the same, five assembly models were constructed utilizing different MOX fuels in the fuel rods. Depleted UO2 was mixed with PuO2 for the first model, the next two models featured a combination of UO2 with ThO2, the fourth one utilized a mixture of PuO2 and ThO2, and the final model employed a combination of UO2, PuO2, and ThO2. Depletion analysis of these fuel assembly models was then carried out using the lattice physics code Dragon 5. The burnup-dependent infinite multiplication factor and reactivity were determined and a significant improvement in cycle length was observed at the end of the burnup period because of fertile to fissile conversion. A higher fast fission rate was obtained for plutonium-based fuel models compared to the reference, whereas the thorium-based fuel models demonstrated higher thermal fission rates. The linear reactivity model (LRM) was utilized to calculate fuel cycle performance for a three-batch refueling scheme. The three fuel models utilizing PuO2 achieved improved discharge burnup and cycle length, and required lower charge fuel mass to obtain the same thermal power. All of the five proposed models attained higher conversion ratios compared to the reference, while the highest increment in CR was obtained for the last model utilizing all three oxides. The CR of conventional UO2 fuel in AP1000 was 0.64, whereas it was calculated to be 0.73 for Model 5. Key safety parameters including the fuel and moderator temperature coefficients of reactivity and the effective delayed neutron fraction of the models were also determined. Finally, the isotopic evolution of minor actinides and fission product poisons of the models were analyzed as a function of burnup and compared with the reference.

Neutronic characteristic analysis of the GAMA microreactor

The GAMA Microreactor is a low-power nuclear power plant dedicated for remote area with a power output of 300 kWe. The primary characteristics of the GAMA Microreactor are its system compactness, design simplicity, and safe operation without moving parts. The fuel is uranium hydride powder, which works simultaneously as a moderator, contained within a stainless-steel vessel and surrounded by a graphite reflector. Liquid sodium-filled heat pipes are embedded in the fuel powder to extract fission-generated heat via passive coolant flow due to capillary action. As a part of the design process, neutronic parameters of the GAMA Microreactor have been determined using SCALE6.2 and OpenMC codes. Among the analyzed parameters are the excess reactivity as a function of burn-up, shutdown margin, control rod worth, temperature coefficient of reactivity (TCR), fuel density coefficient of reactivity (DCR), and reactivity coefficient due to hydrogen dissociation. From the calculation results, the GAMA Microreactor has a maximum excess reactivity of 0.049 at the operational temperature (700 °C) and 0.0788 at ambient temperature. The TCR value is negative (−2.267 pcm/°C) at the beginning of the cycle (BOL) and −4.9618 pcm/°C at the end of the cycle (EOC). Hydrogen dissociation imposes negative reactivity on the reactor both at the BOL (−0.2631 Δk/k-%dissociation) and EOC (−0.3524 Δk/k-%dissociation). Therefore, the GAMA Microreactor maintains its inherent safety characteristics over a considerably long operation time.

Simplified Reactor Model for Microreactor Coupled with Helium Closed Brayton Cycle

Microreactors comprise a new actively developing class of very small advanced reactors that have the potential to be an alternative to carbon-intensive energy technologies. A microreactor based on high-temperature gas reactor (HTGR) technology is a very promising advanced reactor with inherent safety, and it can couple with a closed Brayton cycle for higher efficiency. Since dynamics characteristics are fundamental to analyzing a power generation system and a reactor is the main source of the dynamics characteristics of a system, it is necessary to study a microreactor model suitable for system analysis. The main goal is to simulate the performance of the previously mentioned integrated system, focusing on the details of the power conversion unit while still ensuring acceptable calculation times. Hence, a simplified reactor model is needed that could supply sufficiently accurate values of pressure drop and heat transfer across the core. In this paper, by simplifying the physical processes in a microreactor, a dynamic model described by differential algebraic equations is obtained based on the lumped parameter modeling methodology and the basic conservation of fluid mass, momentum, and energy. Coupling thermal hydraulics with neutron kinetics, the temperature coefficient of reactivity and xenon poisoning are considered. Finally, the model is programmed and calculated using Modelica language. The transient responses of the main parameters under typical perturbations are analyzed, and the results show that the responses are correct. Because of the effect of reactivity feedback, fluctuations of the main parameters caused by microperturbations eventually tend to stabilize. In addition, the effects of negative reactivity introduced by xenon poisoning under two typical dynamic processes are analyzed. In power regulation, excess reactivity is required to compensate for the negative reactivity introduced by 135Xe. The model and results can properly predict the systematic parameters and serve as a basis for system analysis of microreactor coupling with the helium closed Brayton cycle.

Neutronic design of small modular long­life pressurized water reactor using thorium carbide fuel at a power level of 300–500 MWth

This study presents the neutronic design of a small modular long­life Pressurized Water Reactor (PWR) using thorium carbide fuel with 233U fissile material. The target optimization for this study is a reactor designed to operate for 20 years, with excess reacti­vity throughout the reactor operational cycle consistently below 1.00 % dk/k. The analysis involves dividing the reactor core into three fuel regions with 233U enrichment levels ranging from 3 % to 8 %, with a 1 % difference for each fuel region. To achieve optimum conditions, 231Pa was randomly added to the fuel. The fuel volume fraction in this design varied from 30 % to 65 %, with a 5 % incremental variation. Power level variations are also studied within the 300–500 MWth with increments of 50 MWth. Calculations were performed using the Standard Reactor Analysis Code (SRAC) program with the PIJ and CITATION modules for cell and core calculations utilizing JENDL­4.0 nuclide data. Neutronic calculations indicate that the fuel with a 60 % volume fraction achieves optimum conditions at a power level of 300 MWth. The best performance was observed with a fuel volume fraction of 65 %, reaching optimum conditions across power levels ranging from 300 to 500 MWth. For the fuel with the best performance, the power density distributions for low and high power levels follow the same pattern radially and axially. The power peaking factor (PPF) for all fuel configurations approaching the optimum conditions remains below two, a safe limit for the PWR. Other neutronic safety parameters, such as the Doppler coefficient and void fraction coefficient, also stay within the safe limits for the PWR, with both values remaining negative throughout the reactor operational cycle

Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators

Graphite plays a crucial role as a moderator and reflector in high-temperature gas-cooled reactors (HTGRs). However, under high-irradiation conditions, graphite exhibits microcracking within the operational period. Studies have been investigated the potential of composite-based materials for replacing graphite in HTGRs. This study focused on a magnesium oxide (MgO)-based composite material as the host matrix and a homogeneously distributed beryllium oxide (BeO) as the entrained moderating phase and investigated the feasibility of the MgO–BeO as the new moderator in a small HTGR to achieve high burnup performance. In this study, the conceptual design of the HTR50S was selected as one of the candidates for the small HTGR concept. Burnup calculations and safety evaluation in HTR50S design were performed. The Monte Carlo MVP 3.0 and MVP-BURN codes were used in this study for neutronic calculations. Results revealed that a high burnup of 80 GWd/t can be achieved using a fuel composition of 6 kg heavy metal per fuel block with 17 wt% of 235U enrichment. Furthermore, a negative temperature coefficient of reactivity was achieved during the operation period.

Comparison of the neutronic properties of the (Th-233U)O2, (Th-233U)C, and (Th-233U)N fuels in small long-life PWR cores with 300, 400, and 500 MWth of power

Abstract The neutronic characteristics of (Th-233U)O2, (Th-233U)C, and (Th-233U)N have been compared in small long-life pressurized water reactors (PWRs). Neutronic calculations were carried out at 300 MWth, 400 MWth, and 500 MWth with two cladding types: zircaloy-4 and ZIRLO (Zr low oxygen). They were performed using the Standard Reactor Analysis Code (SRAC) and JENDL-4.0 nuclide data, dividing the reactor core into three fuel zones with varying 233U enrichment levels, ranging from 3% to 9% and fluctuating by 1%, employing the PIJ module at the fuel cell level and the CITATION module at the reactor core level. In addition, 231Pa was added as burnable poison (BP). The (Th-233U)N fuel demonstrated superior criticality compared to the other fuel types, as it consistently achieves critical conditions throughout the reactor’s operating cycle with excess reactivity <1.00% dk/k for several fuel configurations at the 300 MWth and 400 MWth power levels. Moreover, the (Th-233U)N and (Th-233U)C fuels exhibited similar and flatter power density distribution patterns compared to the (Th-233U)O2 fuel. The power peaking factor (PPF) value was relatively higher for (Th-233U)O2 fuel than the other two fuels. The (Th-233U)N fuel exhibited the most negative Doppler coefficient, followed by (Th-233U)C and (Th-233U)O2 fuels. Analysis of burnup levels revealed that the (Th-233U)O2 fuel achieved significantly higher burnup than the other two fuels.

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

Passive compact molten salt reactor (PCMSR) is a design concept of a molten salt reactor (MSR) currently under development in Universitas Gadjah Mada, Indonesia. It is designed as a thermal breeder reactor using thorium fuel cycle. However, our previous study shows that the original PCMSR design was incorrectly modelled, primarily overestimating its thorium breeding capability. To improve PCMSR neutronic design, we modified the core configuration by the addition of radial fuel channel layers previously nonexistent in original PCMSR core design in various configurations. Neutronic parameters of modified PCMSR geometries in the beginning of life (BOL) were simulated using MCNP6.2 radiation transport code with ENDF/B-VII.0 library. All variations of fuel layer addition show improvement in both temperature coefficient of reactivity (TCR) and breeding ratio (BR), with TCR values became more negative and BR values are larger than unity, ensuring proper breeding capability. Configuration Inner Core-Outer Blanket (IC-OB) achieves the largest BR and lowest doubling time (DT), whilst its TCR is an improvement from the original design. Therefore, IC-OB fuel layer configuration can be applied to redesign the original PCMSR and used in various design optimization scenarios.

Preliminary analysis of the core design for Indonesia Micro Reactor (IMR-13)

The development of microreactor technology can meet the energy needs of specific regions of Indonesia’s underdeveloped, foremost, and outermost areas. One of the latest microreactor designs is the Indonesia Micro Reactor (IMR-13), which has seed and blanket fuel and 5MWt power generation. The IMR-13 reactor has characteristic design features for operation for as many as ten years without refuelling. The reactor core comprises integral fuel elements consisting of a heat pipe, fuel matrix, fuel cladding, and graphite moderator. There are two fuel element types: seed fuel elements with 19.75 % low-enriched uranium oxide and blanket fuel elements with thorium oxide – graphite reflectors on the bottom, sides, and top surround the reactor core. Side reflectors have moving parts that function to regulate reactor power. There are no active components in this reactor. Heat transfer occurs by natural circulation in the heat pipe to the molten salt pool and then from the molten salt pool to the steam generator. This research focuses on the neutronic aspect of the reactor, particularly in calculating the seed and blanket ratio to allow the reactor to operate for ten years with minimal power peaking factor. The selected variant model has 979 seeds and 294 blankets. The variation model has a value of excess reactivity at the beginning of the life of 937 pcm. Each value of the shutdown margin for system first and system second sequentially is 3020 pcm and 2994 pcm. The temperature reactivity coefficient is negative, implying inherent safety. At the end of life, the burnup value was achieved at 4.97 GWd/MTU with an excess reactivity of 271 pcm. A parameter analysis of the burnup characteristics will follow in the selected variant model. The parameters examined were the core power distribution, delayed neutrons, the fast fission factor, conversion ratio, and decay heat.

Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly

Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters, and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.

Evaluating Hydrogen-based Moderators in High-Temperature Gas-cooled Reactors with 5 wt% Enriched Uranium Annular Fuel Rods

Abstract Small Modular Reactors (SMRs) based on high temperature gas-cooled reactor (HTGR) technology are being developed for providing high-temperature process heat and high-efficiency (>40%) electrical power generation. However, most of the HTGR-SMR concepts require high assay low enriched uranium (HALEU) fuel, with enrichments typically above 10 wt% 235U/U, to get sufficiently high burnup levels and fuel lifetime. The need for HALEU is primarily a consequence of the low volumetric density of fissionable material in tri-structural isotropic (TRISO) fuel particles, and also the use of graphite as a moderator/reflector. A previous study has shown that a modified prismatic HTGR fuel assembly with hydrogen-based moderator (7LiH) and cylindrical fuel elements of 5 wt% 235U/U enriched uranium can greatly reduce fuel consumption of a HTGR. However, such a design concept could lead to positive temperature reactivity coefficients (TRCs), making reactor control more challenging, with reduced passive safety. The purpose of this study is to evaluate variations of the hydrogen-based moderator in this alternative fuel assembly concept to identify configurations that achieve negative TRCs, thus improving passive safety characteristics. Calculation results demonstrate that negative TRCs can be achieved with reduced hydrogen mass such that natural uranium consumption is substantially less than that of the TRISO fuel concept, with comparable or higher core life. This study also explores the option of using 7LiOH and NaOH as hydrogen-based moderators, instead of 7LiH, thus allowing operation at higher temperatures, where hydrogen TRCs are lower.

Verification and Geometry Optimization of a One Fluid Molten Salt Reactor (OFMSR) with Fixed Volume

Abstract Thermal molten salt reactors can be designed in many configurations. This paper investigates the optimal geometry of a one fluid molten salt reactor (OFMSR) in a virtual one-and-half fluid configuration with a fixed fuel salt volume. Two primary configurations were studied, axial blanket (three models) and radial blanket (two models). Neutronic calculations were performed using MCNP6.2 and Serpent-2 reactor physics codes with ENDF/B-VII.0 continuous neutron library. The analysis comprises criticality calculation, temperature coefficient of reactivity (TCR), breeding ratio (BR), and kinetic parameters. The results imply a good agreement between MCNP and Serpent calculations. TCR values show a different pattern between axial and radial blanket configuration. Whilst the correlation between TCR and BR is inversely correlated in axial blanket, it is linear in radial blanket configuration. Overall, radial blanket configuration seemed to show better neutronic performance than axial blanket configuration, with comparably strong negative TCR and large BR.

A proposed refueling scheme for the modular pebble bed reactor

In this paper, a suggested design of the fuel pebble and loading pattern in the MIT Modular Pebble Bed HTGR (MPBR) using thorium (232Th) is presented. Using the MCNP5 code model for the MPBR reactor, different compositions of Thorium Low Enriched Uranium (ThLEU) with different Heavy Metal (HM) loading and 235U enrichment were investigated to achieve the maximum thorium utilization with relatively minimum reactivity insertion in case of water ingress accident. Using pure Low Enriched Uranium (LEU) pebbles and ThLEU pebbles with the same HM loading and 235U enrichment, different pebbles loading patterns were studied. One loading pattern was selected to achieve the criticality with enough excess reactivity compensating for the poisoning effect of xenon build-up at the startup of the reactor operation and the negative effect of fuel and reflector temperature coefficients of reactivity. This loading pattern aims also to reduce the radial power peaking in the core and to minimize the axial Power Peaking Factor (PPF) in the central hot channel, with a minimum number of pure LEU fuel batches. A calculation procedure was developed that combines fuel pebble depletion and full core calculations to investigate the impact of a proposed online refueling scheme for the MPBR reactor and to simulate the multicycle refueling mechanism. Using SERPENT 1.1.7, the nuclide concentrations were calculated as a function of pebble burnup for both LEU and ThLEU pebbles. A multicycle online refueling scheme was proposed and the average number of passes required for LEU and ThLEU pebbles to reach their maximum burnup was estimated.

TRU utilization and MA transmutation in thorium-based fluorinated molten salt fast reactor

In this paper, a thorium-based fluorinated molten salt fast neutron reactor is taken as the research object, and the utilization of thorium resources, reuse of transuranic waste (TRU), and transmutation of minor actinide (MA) in fast reactor are the research purpose. The fast reactor adopts a multi-layer core structure, in which the fuel is divided into an inner fission fuel zone and an outer breeding fuel layer, respectively loaded with fluorinated molten salts LiF+ThF4+XF4 (where X is 233U, TRU, or 233U+MA) and LiF+ThF4. In addition to serving as nuclear fuel, these two molten salts also act as coolants in the primary circuits.Through the cross sections of nuclear reactions, transmutation chain of actinide nuclides, and the particle population density formula of actinide nuclides in nuclear reaction equilibrium, the possible nuclear reactions of actinide nuclides in thorium-based fast reactors are analyzed. The calculated results show that the thorium-based fast reactor has the following characteristics: fast neutron energy spectrum, negative temperature reactivity coefficient, effective utilization of thorium resources, the ability to breed fissle fuel such as 233U, production of high-purity radioisotope 238Pu material, and the ability to incinerate Pu and MAs in the TRU spent fuel.In the 233U fuel scheme, the 233U fuel is breeded, breeding ratio of fissile fuels is 1.21, the reaction products contain very little Np, and almost no Pu, Am, and Cm long-lived radionuclides. In the TRU fuel scheme, the annual capacity to incinerate MAs and Pu is 1072.6 kg, while producing a large amount of 233U and 233Pa fuel (together 1162.5 kg/year).In the 233U+2.0 wt% TRU-MAs fuel scheme, 127.35 kg of MA waste can be incinerated annually, and their transmutation rate is 15.95%, which is equivalent to the annual production of MA nuclear waste from 4.9 conventional 1000 MWe light water reactors. Here, TRU-MAs is the mixtured MA fuels after separation of Pu from the TRU fuel. At the same time, radioactive fuel 238Pu, which can be used for space nuclear power, is produced with an annual yield of 87.89 kg and an isotopic abundance of 87.57%. If the added TRU-MAs fuel is replaced with 2.0 wt% 237Np fluorinated salt, the annual yield and isotope abundance of 238Pu are further increased to 130.24 kg and 95.4%.

Characterization of neutronic parameters and radiation shielding design for an aqueous homogeneous reactor

In this research, an aqueous homogeneous reactor is simulated and its neutronic parameters are investigated. Due to the negative reactivity caused by temperature change and void formation, the cold excess reactivity is considered equal to 4 βeff. Also, calculations of radiation shielding and neutronic characterization of the reactor are performed. For different heights of uranyl sulfate fuel, the effective multiplication factor (keff) of the reactor is calculated under cold zero power and hot full power conditions to assess the required excess reactivity. The amount of fuel burnup and changes in the keff are investigated for 500 days, finally 200 operating days are chosen as the working duration of the reactor. The control rods worth (CRW) for different diameters of boron carbid and reactor safety parameters including shot down margin (SDM) and safety reactivity factor (SRF), temperature and void coefficients of reactivity are analyzed. In order to design a suitable radiation shield for this reactor, two different shields are considered. The first shield is a combination of polyethylene thicknesses and barite concrete, and the second shield is made only of barite concrete. The total thickness of the first and second shield was 230 and 180 cm, respectively. Due to the problems of using polyethylene (e.g. problems related to machining, shield dimensions and low melting temperature), the second shield is chosen as the suitable radiation shield for this reactor. Using this radiation shield, the total dose rates of neutron and gamma rays satisfy the ICRP dose limits.

ВВЭР-1000 ЯДРОЛЫҚ РЕАКТОРЫНДАҒЫ ЖЫЛУФИЗИКАЛЫҚ ПРОЦЕСТЕРДІ КОМПЬЮТЕРЛІК МОДЕЛЬДЕУ

The priority use of water-water reactors in the nuclear power industry is due to their advantages. Water is ideally suited as a moderator and refrigerant, providing compactness, high power and simple reactor design. Water-water reactors have high resistance and self-regulation due to the negative temperature coefficient of reactivity. Their activity is caused by short-lived nuclides, which facilitates biological protection and access to equipment. Water effectively removes heat. However, there are also disadvantages: high neutron absorption by water, the requirement of uniform distribution of water, corrosion activity, expensive cleaning system and temperature limitation. An increase in unit power is possible due to the equalization of heat generation in the core. This requires an accurate thermohydraulic calculation. As a result of the study, a theoretical thermohydraulic calculation of heat transfer in the heat - generating set of the VVER-1000 reactor was obtained for the weighted average and maximum values of heat release in the heat-generating set. Computer simulation of heat transfer was also carried out for the distribution of moderately loaded energy in the heat-generating set. The results obtained were compared and analyzed. As a result of the calculation, the distribution graph and temperature field of the heat-releasing element of the coolant, fuel and reactor in the shell of the VVER-1000 heat-releasing kit were obtained.