Burnup Cycle Research Articles

Closing the nuclear fuel cycle: Strategic approaches for NuScale-like reactor

The present study proposes the potential implementation of eight different closed fuel cycle strategies for a NuScale-like reactor core using its own spent fuel as a reusable source of fissile material for energy production. For that, the spent fuel composition after three burnup cycles of approximately 12 MWd/kgU of NuScale-like initial core and five years of cooling in a spent fuel pool was theoretically reprocessed by GANEX or UREX+ methods. After reprocessing, these two new fuel compositions were spiked in a mixture of thorium (Th) or depleted uranium (DpU), and afterwards inserted into specific batch positions of the core. Therefore, the proposed NuScale-like core configurations contain fuel assemblies loaded with conventional uranium-based fuel and others loaded with reprocessed fuel, resulting in the following combinations: UO2 and GANEX spiked with Th, UO2 and GANEX spiked with DpU, UO2 and UREX+ spiked with Th, UO2 and UREX+ spiked with DpU. The main idea is to comprehend the advantages of adopting the closed nuclear fuel cycle for a NuScale-like reactor by comparing the reference case and the cases containing reprocessed fuel. The results exhibited that all instances in which the core was simulated with reprocessed fuel improved the feedback coefficient, maximum excess of reactivity varying the boron concentration in the coolant, and power peak factor (PPF). Furthermore, the closed nuclear fuel strategies also demonstrated savings of about 17.50% in terms of separating work units (SWU) due to plutonium and uranium recycling, and a potential burnup extension of approximately 43%. The Serpent code version 2.1.32 developed by VTT and ENDF/B-VII.0 nuclear data library has been used to perform the simulations.

On excess reactivity control of the advanced high temperature reactor (AHTR)

The Advanced High Temperature Reactor (AHTR) is a large Fluoride salt cooled High temperature Reactor (FHR) with a thermal power of 3400 MWt, utilizing hexagonal fuel elements with TRISO fuel spheres embedded in graphite fuel plates. While its cycle length is relatively short (less than one year), its specific power is high, resulting in high cycle burnup and consequently high initial excess reactivity that needs to be controlled. This is achieved by using both Eu2O3 burnable poison spheres and molybdenum-hafnium-carbon control rods. This work details studies developing and implementing these two reactivity control features over representative AHTR fuel cycles. Various Eu2O3 densities are considered to evaluate both the achieved reactivity control and the residual reactivity penalty of burnable poison at end of cycle. Control rod studies are performed to find integral and partial insertion reactivity worth for control rod banks at various radial locations. Axial power impacts are also evaluated. Focus then shifts toward control rod movement strategies. A manually generated insertion scheme leading to a critical configuration is first investigated to serve as a reference point for automated schemes to follow. Next, automated insertion and withdrawal strategies are evaluated based on their radial power peaking performance. Control rod insertions into the core locations with the highest local power are shown to be more stable than control rod withdrawals from core locations with the lowest local power. Finally, a methodology is introduced for finding a critical control rod insertion configuration. The methodology is intended to be used over a depletion sequence where the positions of the control rods are automatically adjusted to maintain criticality over cycle. In summary, this work presents a possible implementation of AHTR reactivity control and introduces automated control rod movement strategies for use in multi-physics depletion analyses.

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Comparative analysis of different FeCrAl alloys in pressurized water reactors

Discussing and researching issues related to their safety is essential to guarantee the continuity of fission systems in generating energy for existing and new electrical matrices. Commercial fission reactors currently in operation have undergone a series of redesigns. One aspect much discussed in recent years is related to research about new fuels and cladding materials. Therefore, in this work, the feasibility of replacing Zircaloy-4 with FeCrAl-based alloys (C06M, C35M, C36M, Kanthal APMT) was evaluated considering these materials' physical–chemical, thermal, and neutronic properties. Moreover, thermal–hydraulic and neutronic simulations have been presented to compare the cladding behavior in a PWR core in steady state and transient operation conditions. The results showed, in a general way, that the analyzed alloys have equivalent properties from a physical–chemical and thermal point of view. Despite the similarity, the Kanthal APMT alloy presented the worst performance, and the C35M alloy showed more advantageous characteristics. Considering the similarities in the composition of the C06M, C35M, and C36M alloys, which have the same elements in their constitution, the alloy with the best thermal performance has the lowest aluminum content. Considering the neutronic behavior, the Kanthal APMT alloy, with more Cr percentile, presents the lowest neutron effective multiplication factor, keff. Axial and radial power distribution did not change significantly after the cladding replacement. Nevertheless, modifications must be made to keep the criticality during the fuel cycle burnup for all FeCrAl alloys.

Neutronic and fuel cycle performance analysis of fluoride and chloride fuels in Molten Salt Fast Reactor (MSFR)

The signification of fast breeder reactors as an inexhaustible source of energy has been elucidated by many researchers. Molten Salt Reactors (MSRs) show great promise in fuel cycle performance. Fuel selection for MSRs is a key parameter due to its effect on many fundamental parameters of neutronic, safety, and fuel cycle through the neutron spectrum changes. choosing the fuel for MSRs has many degrees of freedom due to the variety of salts, carrier salts, and fissile and fertile isotope types. The foremost objective of this paper is a comparison of performance of fluoride and chloride fuels with the Th-U cycle for Molten Salt Fast Reactor (MSFR) concept. For this purpose, six fuels with start-up fissile isotopes of 233U, TRansUranium (TRU), and TRU/enrU are modeled by Monte Carlo Serpent 2 code. According to the MSFR design, the fuel compositions of three fluoride fuels are chosen from different conceptual designs. Chloride fuels, however, are selected based on the limit of the maximum melting point of salts (565 °C), criticality at the beginning of the cycle (BOC), and inherent safety criteria (temperature reactivity coefficients). The most important evaluated parameters include the energy spectrum, absorption and fission reaction rate, kinetic parameters, temperature reactivity coefficients, power density distribution, breeding ratio, and inventory changes of important isotopes during the burnup cycle. The results reveal neutron spectrum is less fast with fluoride salt; in contrast neutron leakage of chloride fuels are more than fluoride fuels due to their harder spectrum and transparency. Furthermore, the high leakage reduces the neutron budget required for breeding 232Th to 233U and the low density of chloride fuels causes a lower loading of fissile and fertile isotopes in the reactor. Therefore, using chloride fuels leads to larger core volume and need a proper reflector to reduce neutron leakage; Anyway, the U-Pu cycle shows better indications of the breeding for chloride fuels than the Th-U cycle. By the way corrosion and radioactivity of generated elements in chloride fuel circuit are serious problems.

Calculation of boron concentration dependent reactivity versus moderator density curve with application on ATWS analysis for Maanshan PWR plant

A method used for determining the moderator density coefficient (MDC) based on the effective multiplication factor (keff) calculated from the SIMULATE-3 code is developed in this study. The calculation is performed at various water density and boron conditions for Maanshan unit 1 cycle 27. The calculated MDC data are fitted as a second-order polynomial function of density, modified with a factor of linear function of boron concentration. The fitted equation is then integrated and normalized to zero at rated density to represent the reactivity as a function of moderator density and boron concentration. A reactivity vs. density table under different boron concentrations can then be generated based on this function. The table generated is used as an input for the ATWS analysis to be performed using the RETRAN-02 thermal–hydraulic analysis code. As an illustration, a table based on a boron concentration of 1500 ppm is generated and used as an input for the analysis of the potential limiting ATWS case, i.e., Loss of Normal Feedwater Flow Event. The ASME emergency level limit of reactor coolant system pressure less than 220.63 bar is used as the acceptance criterion for the current ATWS analysis. It is known that the critical boron concentration at hot full power for each operating fuel cycle varies with the burnup through the cycle. Hence, the calculated ATWS results at different boron concentrations are equivalent to that calculated at different cycle burnup conditions. The cycle burnup point after which the calculated ATWS maximum is within the ASME limit can be determined based on the sensitivity analyses considering different boron concentrations.

Modular fuzzy control for nonlinear multipoint based core model of TMI PWR under variable fuel cycle conditions

This paper provides design of modular fuzzy controller based on reactor power for multipoint kinetic model of TMI PWR, under transients and variable operational fuel cycle conditions. The validated four-point kinetic model is developed considering six groups of delayed neutron precursors, reactivity feedbacks and diffusion among selected nodes. Fuel burn-up effects and power variations are modeled using the variable thermal coefficients. The ability of adaptive multi-module fuzzy controller for reactor power is evaluated and compared at start and end of burn-up cycle. The power dependant nonlinearity of reactor core and time dependant variations of reactivity characteristics are handled with modular approach of TS fuzzy control. The performance monitoring and control improvements are achieved by online parametric identification of process model. The four modules are distributed on the power base line using the triangular mapping. Design of fuzzy controller is based on the optimized gain requirement at different power levels, error and error change. Control output is mapped as biased linear function of selected input variables using TS type fuzzy logic inference. Safety analysis is performed with Lyapunov function criteria for multipoint model. Transient analysis is performed for power tracking and accident handling ability of proposed control design and compared with the PID controller. The power variation among the adjacent nodes is reduced up to 5% under load following transients with the proposed model. Modular fuzzy approach found proficient in handling the changing operational conditions and variable dynamics of the system.

Comprehensive Comparison of Nuclear Materials in BNPP Fuel Assemblies

There is an obvious effort to increase the burnup of used fuel assemblies in the Bushehr WWER-1000 Nuclear Power Plant (BNPP) in order to improve fuel utilization. The outcomes of this research could result in an increase in the BNPP reactor cycle length, which would lead to improved fuel consumption. Considering the lack of uranium resources and the planning to use new types of fuel in the BNPP, the use of integrated burnable absorber (IBA) materials is of great importance. An analysis of the performance of various IBAs, including Gd2O3-UO2, Er2O3-UO2, and Dy2O3-UO2, as well as the standard (proposed by the designer) burnable absorber (BA) (CrB2Al) in the BNPP, and their impact on fuel neutronic characteristics has been performed. Five fuel assemblies: one without a BA fuel rod and four each containing standard BA gadolinia, erbia, and dysprosia fuel pins were investigated. The neutronic properties of BAs were evaluated by the infinite multiplication factor, reactivity swing, and power peaking factor dependence on fuel burnup. Gadolinia, with a concentration of 5%, has the greatest effect on initial reactivity with 10 893 pcm and the lowest effect on the reactivity swing with 0.277 Δk among the other BAs, which leads to selecting the most appropriate BA for improving reactor core stabilities and enhancing operational safety. The gadolinium IBA extends the cycle burnup by about 1 GWd/tonne U compared to the standard BA. At the beginning of the cycle, erbium has a more uniform power distribution than the standard BA; however, at the end of the cycle, gadolinia has a more uniform power distribution.

Burnup Characteristics of Transuranic Fuel With Burnable Poison in Dual Cooled Annular Rod of VVER-1000

Abstract The potentiality of Transuranic (TRU) fuel as dual cooled annular fuel rods with different types of burnable absorbers (BAs)- integrated burnable absorbers (IBAs), burnable poison rods (BPRs), coated absorbers, etc. in the hexagonal assembly of VVER-1000 was studied. Annular 7, Annular 8, and Annular 9 models were taken for various combinations of TRU fuel and BAs. Planned models were simulated in Monte Carlo particle simulation code OpenMC and lattice physics deterministic code Dragon Version5. Burnup-dependent multiplication factors (MFs) were simulated for 3000 effective full power days (EFPDs). Reactivity was calculated by taking 3% neutron leakage. The study showed that both the MF and reactivity for TRU fuel are significantly higher than conventional UO2 fuel. Linear reactivity model (LRM) was applied to find out cycle burnup, discharge cycle burnup, and cycle length. High cycle burnup, discharge burnup, and cycle length have been observed for TRU fuel compared to UO2 fuel. Burnup-dependent atomic concentration graphs showed that slight burn of Np-237, constant concentration for Cm-244, a slight increase for Am-242, and linear burnout of BAs- Gd-155, Gd-155, and Er-167. A lower concentration of Xe-135 has been observed for TRU fuel. Pin power distribution and energy-dependent neutron flux for different models and BAs combinations are also included in this study.

Validation of the CORD-2 System for the NPP Krško Nuclear Core Design Calculations

The CORD-2 package intended for core design calculations of PWRs has been recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krško core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krško core.

Optimization and burnup calculations of BNPP's reactor core with the new generation fuels (TVS-2M) by artificial neural network

In recent years, many types of research focused on optimizing, modernizing, and developing the fuels for VVER-1000s. A new generation and developed fuel are TVS-2M. Using TVS-2M has many benefits, such as burnup and cycle length increase compared to former fuel assemblies. However, the main issue is to find the best arrangement of the fuel assemblies to increase burnup and safety features. An artificial neural network (ANN) is a powerful tool to predict core burnup for different reactor core configurations in nuclear power plants. One can use ANNs to find the best Fuel arrangement. This article uses a multi-layer neural network to determine the optimum TVS-2M fuel arrangement, which can calculate the burnup of different fuel arrangements. The method utilizes some core parameter data to train a neural network to set up a real-time predicting system for burnup and actinides concentration predictions for different fuel assembly arrangements. The MCNPX code output applies as training data on a multi-layer perceptron neural network. After ensuring the high accuracy of the designed neural network, 20 different arrangements are inserted into the neural network to calculate the burnup parameter for them and select the best arrangement. At last, trained ANN can calculate the actinides concentration and other desired parameters. The results showed that replacing UTVS with TVS-2M FAs, in addition to increasing the length of the operation cycle from 289.6 days to 338.7 days, can increase the cycle's burnup, thus increasing the economic efficiency of the power plant. Also, the results confirm the power of the developed neural networks and approve the use of these networks instead of executing time-consuming codes like MCNPX.

Neutronic and fuel cycle performance of LEU fuel with different means of excess reactivity control: Impact of neutron leakage and refueling scheme

Low enriched uranium (LEU) fuel with different Integral Burnable Absorbers (IBAs) and Burnable Poison Rods (BPRs) in AP1000 assembly was investigated. These include IBAs of Gd2O3, Er2O3, AmO2, PaO2, and BPRs of Al2O3-B4C. The conventional uranium dioxide (UO2) fuel and two accident tolerant fuels (ATFs) namely UN-U3Si2 and U3Si2 were taken as the base fuel materials. Nine fuel compositions of these fuels with and without burnable poison were simulated in the lattice physics code Dragon, version 5. The burnup-dependent infinite multiplication factor was determined and significant variation was found due to the addition of poison. Reactivity was calculated at different levels of neutron leakage (0–5%). More positive reactivity at the early burnup stage and less negative reactivity at end of life were found for PaO2. The Linear Reactivity Model (LRM) was used to calculate the fuel cycle parameters. Cycle burnup and discharge burnup were determined considering different refueling batch number at each leakage scenario. Both parameters are found to decrease with neutron leakage while a large batch number results in a higher discharge burnup. Reduced discharge burnup was observed for each IBA and BPR except PaO2. Cycle length was also calculated for each of the composition and cycle length penalty was positive only for PaO2. The fuel temperature coefficient of reactivity, pin power distribution, and energy-dependent neutron flux were determined using the Monte Carlo code OpenMC. The burnup-dependent isotopic compositions were also studied.

An approach to minimize reactivity penalty of Gd2O3 burnable absorber at the early stage of fuel burnup in Pressurized Water Reactor

The high capture cross-section (σc) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high σc as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd2O3 and Er2O3 respectively. Models comprise nuclear fuel with a homogeneous mixture of Er2O3, AmO2, SmO2, and HfO2 with Gd2O3 as well as the coating of PaO2 and ZrB2 on the Gd2O3 pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er2O3 and SmO2 showed expected results namely higher values of these parameters than the reference design of Gd2O3 at the early burnup period. The highest value was found for the model of PaO2 and Gd2O3 followed by ZrB2 and HfO2. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd2O3 and absorber coating thickness.

Neutronic and fuel cycle performance of VVER-1000 for dual cooled annular fuel with coated burnable poison

The feasibility of the dual cooled annular fuel rod with coated burnable absorber in the hexagonal assembly of the VVER-1000 reactor was studied. The models consist of a 4.5% enriched dual cooled UO2 fuel rod with a coating of Gd2O3, Er2O3, PaO2, AmO2, and ZrB2 as the burnable absorber. Designed models along with the reference designs were simulated in lattice physics code Dragon and Monte Carlo code OpenMC. Burnup analysis was conducted for 1620 EFPD. The multiplication factor, as well as reactivity, was determined for fuel burnup. Improvement in the multiplication factor and reactivity is observed for dual cooled annular fuel with coated burnable poison. Designed models also showed a significant increase in thermal fission reaction rate compared to the reference design. Linear reactivity model (LRM) was used to calculate cycle burnup, discharge burnup, and cycle length for three and four batch refueling schemes. No degradation is observed in fuel cycle performance using dual cooled annular fuel with coated burnable absorber except for the model of gadolinium. The other absorbers showed better results than gadolinium in both neutronic and fuel cycle performance. PaO2 used model showed the most promising result among them. The value of k-infinity was found to be 1.27089 at BOL and the cycle burnup was around 15 MWD/kg. Also, no cycle length penalty was observed for PaO2. Burnup-dependent atomic concentration was studied. A lower concentration of fission product poison was also noticed in designed models. The analysis of relative pin power distribution and energy-dependent neutron flux is also included in this study.

Two different methodologies for neutron flux calculation for the isotopic depletion problem in subcritical systems driven by an external neutron source in a coarse-mesh framework

The isotopic depletion calculation is typically solved by using the quasi-static approximation. In this approximation, the neutron flux is assumed as constant along the time steps in which the burn-up cycle is divided. This work proposes two different calculation methodologies to obtain the neutron flux, used in the isotopic depletion calculation in subcritical systems driven by an external neutron source, in a coarse mesh framework. The so-called depletion calculation methodology with Coarse Mesh Finite Difference (CMFD), and depletion calculation methodology with Pseudo-Harmonic Expansion (PHE). In these methodologies, the neutron flux at the beginning of each burn-up step is calculated with CMFD and the PHE methods, respectively. In order to verify the applicability of both methodologies to an ADS-type reactor, it is simulated a 45 MW small-reactor with burn-up cycle of 90 days, and an 800 MW large-reactor with burn-up cycle of 600 days. The subcriticality adjustment mechanism is based on the effective multiplication factor, solving the neutron diffusion equation with the Nodal Expansion Method (NEM). In addition, the averaged power density and the intensity of the neutron source are evaluated throughout the burn-up cycle. The results obtained by both methodologies are in good agreement with each other.

Separation of the 235U and 239Pu Prompt Energy Spectra in NEOS-II

The NEOS searches for sterile neutrinos by detecting reactor antineutrinos at a very short baseline in Korea. The NEOS detector (1-ton Gd-LS) is deployed at the tendon gallery of the Hanbit reactor unit 5 (2.8 GW thermal power), 23.7 m away from the reactor core. In NEOS-I, we measured the prompt energy spectrum from inverse beta decay using 180 days of reactor-on data and observed the “5 MeV excess”. To understand the origin of the “5 MeV excess”, NEOS-II has taken 368 (120) days of reactor-on (-off) data from September 2018 to October 2020, covering a whole burnup cycle of the reactor. We tested a method for the extraction of the prompt energy spectra for 235U and 239Pu from the whole burnup cycle using pseudo data.

Study on the effect of various burnable poisons on pebble bed reactor with OTTO fuelling schemes using Serpent 2 Monte Carlo code

Pebble bed reactor with a once-through-then-out fuelling scheme has the advantage of simplifying the refueling system. However, the core upper-level power density is relatively higher than the bottom, producing an asymmetric core axial power distribution. Several burnable poison (BP) configurations are used to flatten the peak power density and improve power distribution while suppressing the excess core reactivity at the beginning of the burnup cycle. This study uses HTR-PM, China’s pebble bed reactor core, to simulate several burnable poison (BP) configurations. Serpent 2 coupled with Octave and a discrete element method simulation is used to model and simulate the pebble bed reactor core. It is found that erbium needs a large volumetric fraction in either QUADRISO or distributed BP to perform well. On the other hand, gadolinium and boron need a smaller volumetric fraction but perform worse in radial power distribution criteria in the fuel sphere. This study aims to verify the effect of BP added fuel pebbles on an OTTO refueling scheme HTR-PM core axial power distribution and excess reactivity.

Advanced gas-cooled reactors technology for enabling molten-salt reactors design – Optimisation of a new system

Molten salts present significant advantages as coolants over traditional light water or gas. Salt-cooled nuclear reactor can potentially i allow an increase in core power density and simplify the reactor safety case. Traditionally, molten salt reactors assume molten salt being both the fuel and the coolant. This, however, creates major challenges for the design and operation due to requirements for chemistry control and corrosion behaviour of the core structural materials. The Fluoride salt-cooled High-temperature Reactors (FHR) on the other hand, have conventional solid fuel geometry surrounded by moderator and coolant. Furthermore, some FHR variations share many common characteristics with the Advanced Gas-cooled Reactors (AGR). Thus, using the knowledge accumulated during many decades of successful AGR operation can speed up the FHR development and deployment. However, replacing carbon-dioxide coolant by molten salt significantly changes the reactor performance. Hence, a new core design is needed. In this work, a Multi-Objective Particle Swarm Optimisation is used to identify the most favourable configurations for a new system layout. Initially, the optimisation targeted the beginning of cycle parameters such as criticality and Coolant Temperature Coefficient (CTC). The present work is the next step in this analysis. Each configuration is examined with respect to its thermal-hydraulic performance to assess the power uprate potential which is limited by multiple temperature constraints (e.g. fuel centreline and cladding temperatures as well as the coolant freezing/boiling). The estimated maximum power was then used in the fuel burnup calculations, from which a discharge burnup and cycle average CTC were obtained. As a result of the optimisation process, several families of possible solutions were identified, which form an optimal Pareto front. The most attractive configurations in terms of achievable power density, however, were not necessarily on the Pareto front. Most of the identified design options had only a small amount of graphite moderator or no graphite at all, relying entirely on moderation in the salt. Increasing the graphite volume was consistently found to worsen most of the optimisation parameters. The newly identified design options have the potential to achieve power density which is higher than that of a typical AGR by up to a factor of five, while maintaining negative CTC through the burnup cycle.

Core calculations for small modular reactor during burnup cycle

ABSTRACT Nuclear, modular water (AVB) reactor is a small modular reactor with special capabilities and it can be used as a source of energy for various applications. In this research, the core calculations of this reactor have been performed during the burnup cycle with deterministic and Monte Carlo approaches by the computational codes WIMSD, PARCS, and MCNPX. Due to the temperature and pressure limits for coolant in this reactor, the PMAXS generator and the developed PARCS code, PARCS-MOD, have been used in a deterministic approach. Quantities include the effective multiplication factor, the relative power distribution, the fuel, and moderator temperature distribution, reactivity feedbacks, and concentrations of some isotopes during the cycle are calculated. In addition, to comparing the results of the two approaches, some general results were compared with the results published in the documents, which we see an acceptable difference in this regard. The maximum value of for states BOC, MOC and EOC are −1.39, −1.85, and −1.87 (pcm.K−1), respectively. The relative difference between the calculated relative power is less than 2.6%. The relative difference between the coolant average temperature and the fuel average temperature calculated with the reference values is 0.21%, and 0.14%, respectively. The maximum axial coolant density is 650 (kg.m−3). Also in this research, the dead time of the ABV reactor has been calculated which is zero.

Neutronic characteristics and feasibility analysis of micro-heterogeneous duplex ThO2-UO2 fuel pin in PWR

The shortage of natural uranium resources, nuclear waste disposal and nuclear proliferation issues are important constraints to the development of nuclear energy. The development and utilization of thorium fuel is an effective way to solve these problems. This paper analyses the neutron physical properties of micro-heterogeneous duplex fuel and compares them with UO2 fuel in terms of burnup depth, reactivity coefficients and radioactive waste. Additionally, the best grid of duplex fuel is explored as well.Results from this study show that the ThO2-UO2 duplex fuel can match the discharge burnup of UO2 fuel with the 235U enrichment greater than 5% but less than 20%. The reactivity was calculated at different levels of thorium dioxide fuel, and the rate of decrease in reactivity is lower for duplex fuel than for uranium dioxide fuel. Fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) are within the acceptable range specified for pressurized water reactors. Production of both weapons-grade plutonium and minor actinides are greatly reduced and the effect of fission nuclides breeding is significant. When the volume ratio of moderator to fuel is 3.27, the duplex fuel performs the best in burnup and the cycle burnup is greatly increased. It also has good performance on safety and nonproliferation issues under such grid condition.Physical calculations of assembly size were performed for 20% ThO2 duplex fuel at a volume ratio of moderator to fuel of 3.27 and compared them with the UO2 assembly calculations. Duplex fuel has higher cycle burnup and economy. Moreover, The negative FTC and MTC and the high control rod value of duplex fuel enable the reactor to be safe. Duplex fuel has a lower power peaking factor and a more uniform power distribution which are good for reactor operation.