Moderator Temperature Coefficient 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.

AP-1000 fuel assembly performance with UO2 and UO2/U3Si2 fuels and silicon carbide cladding from the viewpoint of neutronics

ABSTRACT Accident Tolerate Fuels have been highly researched since the Fukushima disaster. To improve the safety of the AP-1000 fuel assembly, an advanced fuel, UO2/U3Si2, is considered. In addition, the silicon carbide (SiC) cladding that has the capability of reducing oxidation is compared with the baseline Zircaloy for AP-1000 fuel assembly along with the two different fuel cases (UO2 and UO2-U3Si2). The present study focuses on the neutronic performance of the UO2/U3Si2 fuel and thus the main neutronic parameters such as the fuel cycle length, pin power distribution, atom density evolution of main isotopes, and the reactivity feedback coefficients are evaluated. An increase of 6.72% in the criticality period is introduced when 70%UO2-30%U3Si2 fuel with SiC cladding is considered. The pin power peaking factors, fuel temperature coefficients, and moderator temperature coefficients of the proposed fuel result in very few changes. For instance, the maximum pin power distribution difference between 70%UO2-30%U3Si2 and UO2 fuel with SiC cladding is 0.001, which is negligible. Thus, its performance is as good as the traditional UO2 fuel from the neutronic aspects and controlling power 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.

Analysis of the LatticeNet neural network framework’s performance using prediction-calculated temperature coefficients in PWR assemblies

Various Machine Learning (ML) techniques have seen recent and growing interest in the creation of surrogate neutronics models as a potential way to avoid the computational expenses associated with conventional high-fidelity modeling. Artificial Neural Networks (ANNs) have been shown to be particularly useful for single-assembly predictions involving pin-wise power distributions and multiplication factors. In this paper, the LatticeNet neural network framework is investigated as a method to predict Doppler and moderator temperature coefficients for Pressurized Water Reactor (PWR) fuel assemblies, as well as differential boron worth. This approach uses the built-in tools developed alongside LatticeNet to construct two fully-connected network architectures capable of predicting k-eigenvalues from two inputs such as fuel enrichment and temperature when trained with data generated with CASMO-4E. A single network taking in all study parameters as inputs was then used to predict k-eigenvalues for fuel temperature, moderator temperature, and boron perturbation cases. The calculated temperature coefficients and differential boron worth values were compared with a bank of reference values to validate the efficacy of this method. Overall, differences in k-eigenvalues were within 0.017% in the worst case. The temperature coefficients saw mean errors of 1.85% and 1.69% for the two-parameter networks, respectively. The all-parameter network was then shown to predict 1100 points in 236 ms compared to the 4.95 min CASMO-4E took to generate them. Additionally, the differential boron worth achieved the lowest mean error of 0.30%; each of these values were within our acceptance criteria.

Study on neutronic behavior of VVER-1000 fuel assembly with duplex fuel rod design

Various studies suggested that duplex fuel rods have better neutronic characteristics than ordinary single-oxide fuel pellets. A duplex fuel rod consists of an inner uranium dioxide (UO2) layer and an outer thorium dioxide (ThO2) layer. This paper aims to study the neutronic behavior of the VVER-1000 fuel assembly with duplex fuel rod design. The duplex fuel rods were loaded into the fuel assembly in three configurations: outermost region, mid-region, and central region of the assembly. Three compositions with volume fractions of 75% UO2-25% ThO2, 70% UO2-30% ThO2, and 65% UO2-35% ThO2 were considered. A series of numerical calculations were conducted using the MCNP6 code and ENDF/B-VII.1 library. In general, there is an improvement in DTC and MTC value for fuel assembly using Duplex fuel rods. However, βeff and ℓ values tend to decrease while the value of Λ tends to increase which may affect fuel assembly response during transient scenarios. The results show that Configuration I can maintain its criticality for up to 15 MWd/kg for all fractions of UO2-ThO2 being used in duplex fuel, while Configuration II and Configuration III only reach below 15 MWd/kg when using duplex fuel rods. The Doppler temperature coefficient (DTC) and the moderator temperature coefficient (MTC) become stronger with the increasing ThO2 volume fraction within the VVER fuel assembly. Compared with non-duplex fuel, the amount of 239Pu was lower when using duplex fuel rods, but minor actinides was slightly higher at 40 MWd/kg. It can be concluded that the VVER-1000 with a duplex fuel rod could improve its safety characteristics besides that increasing fissile material content might be crucial to optimize VVER-1000 fuel management.

Improving the reactor safety aspects by the implementation of (Th-U233-Pu) fuel in a PWR assembly

The aim of this work is not only to demonstrate how to introduce thorium as an alternative nuclear fuel into pressurized water reactors, but also to examine the several advantages that (Th, U233, Pu) has over (Th, U233) and (Th, Pu) fuels when (U233O2) is mixed with (Th-Pu)O2 fuel. To investigate these advantages, three thorium fuels, (Th, U233), (Th-U233-Pu) and (Th-Pu) are simulated and analyzed by the MCNPX 2.7 code. The reactor safety aspects of these fuels are compared with those of the reference UO2 fuel. The concentration of Th-233 and fissile isotopes is adjusted to obtain criticality (Keff = 1.03) at the same burnup value (35 GWd/tonne) for all fuel types. Based on the predicted results, the author revealed that: 1) 56 % of plutonium (238-242) is depleted in the case of (Th-U233-Pu), while 41 % of plutonium is depleted in the (Th-Pu) at 60 GWd/tonne. This result indicates that (Th-U233-Pu) is more proliferation-resistant than (Th-Pu). 2) In terms of improving the quantity of radioactive waste, (Th, U233, Pu) model produces 0.6 and 0.9 times the amount of Am (241-243) and Cm (242-244) in UO2, respectively. On the other hand, (Th, Pu) fuel produces about 7 times the amount of Am and 8 times the amount of Cm in UO2 fuel. 3) The fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) of (Th-U233-Pu) are more negative than that of UO2 and (Th-U233) at the early stages of fuel burnup (0–5) GWd/tonne.4) the total mass of fissile isotopes in (Th-U233-Pu) is greater than that of UO2 and (Th-U233) and as a result, the fuel life time is longer in the case of (Th-U233-Pu). 4) The conversion ratio provided by (Th-U233-Pu) is greater than that provided by UO2 and (Th-U233).This behavior is expected owing to the hardening of neutron spectrum in (Th-U233-Pu). 5) The peaking factors are lower in the case of (Th-U233-Pu) compared to (Th-Pu). 6) The concentration of xenon and samarium is noticeably lower for (Th-U233-Pu) in comparison with (Th-Pu) and consequently, the time required for restarting the reactor is accelerated.

Utilizing even Plutonium Isotopes as burnable absorbers for controlling the reactivity and power distribution in Pressurized Water Reactors

In this paper, a new approach is proposed for controlling the reactivity and power distribution in pressurized water reactor cores without the implementation of burnable absorbers. This approach depended on the homogeneous mixing of uranium oxide with even plutonium isotopes; Pu-238, Pu-240, and Pu-242. Three Plutonium fuels; (99% UO2 + 1% Pu-238), (99% UO2 + 1% Pu-240), and (98% UO2 + 0.5% Pu-238 + 1% Pu-240 + 0.5% Pu-242) were simulated using the MCNPX 2.7 code. The three plutonium fuels are compared with the other three erbium fuels; (99.9% UO2 + 0.1% 167Er2O3), (99.74% UO2 + 0.26% 167Er2O3), and (99.61% UO2 + 0.39% 167Er2O3), which are taken as reference fuels. The weight percent of UO2, plutonium isotopes and enriched erbium is adjusted to provide the same initial reactivity. This is important for holding a fair comparison between the two fuel types. The simulation demonstrated that even plutonium isotopes, like enriched erbium, can improve reactor safety parameters. It was found that Pu-240 added fuel can suppress the initial reactivity by 27% and extend the fuel cycle by 6.9% compared to the conventional UO2. The reason for fuel cycle extension is due to the production of Pu-241. Meanwhile, Pu-238 and Pu-even added fuels provide suppressing the initial keff by 13% and 35% and shortening the fuel cycle by 2.7% and 21%, respectively compared to the reference UO2 fuel. Moreover, the fuel temperature coefficient (FTC), moderator temperature coefficient (MTC) and void reactivity coefficient (VRC) of Plutonium fuels are found to more negative than those of erbium fuels at the later stages of burnup because of the spectrum hardening resulting from Pu-239 production.

Multiphysics assessment of a PWR-type SMR core using Spherical Cermet fuel through a neutronic and thermohydraulic coupled study

In this paper, a multiphysics assessment of the performance of a PWR-type SMR core using Spherical Cermet Fuel is carried out, through a coupled neutronic-thermohydraulic study using Serpent and Ansys CFX codes. The most important results related to the core performance are compared with those obtained when TRISO fuel is used. Correlations (Cp, k) were developed as function of temperature for the fuel region, which are unique and specific for the fuel of this study. The results obtained show that the neutronic-thermohydraulic coupled calculation converges from the fourth iteration, with Keff being the convergence indicator that converges more slowly. An extended fuel cycle of 1448 days was achieved with a uniform power distribution. With this new core configuration, a lower production of plutonium isotopes was obtained. Temperatures below the safe operation limit were obtained in all the studied zones, as well as basic negative safety parameters such as fuel temperature and moderator temperature coefficients, and the full coolant void coefficient.

Neutronic analysis of the BYU molten salt micro reactor

Background: With the rising need for reliable and clean energy, researchers at Brigham Young University created a molten salt microreactor (MSMR) concept to help meet the world’s growing energy demands. The MSMR is rated at 45 MWth and is uniquely capable of passively dissipating all decay heat using conduction only. Methods: This work presents the results of the simulated neutronic behavior of the MSMR using the Monte Carlo neutronics code OpenMC with the ENDF/B-VII.1 cross-section library at steady state conditions. The specific characteristics discussed in this work are: keff, power profile, axial and radial peak power factors, Doppler coefficient of reactivity, and the moderator temperature feedback coefficient. Results: At steady state operating conditions, keff was calculated to be 1.07091 ± 0.00087 indicating that this configuration can achieve criticality. The axial peaking power factor was found to be 1.003 and the radial peaking power factor was found to be 1.45. Feedbacks were calculated by offsetting the temperature of the fuel and the moderator respectively. Doppler coefficient of reactivity to be -12.585 pcm/K and the moderator temperature coefficient of reactivity to be -0.34332 pcm/K. Conclusions: It is worth noting that the largest contributor of the negative reactivity comes from the fuel salt and not the moderator, further demonstrating the safety of the reactor. Future work will include a full neutronic optimization of the core, fuel burn up analysis, and exploration of a suitable burnable neutron poison.

Parametric neutronic analysis of different cladding options for ThO2 pellet of advanced dual-cooled annular PWR assembly

This study presents a parametric neutronic evaluation of different candidate claddings for a purposed candidate fuel (Th-233U-235U)O2 in an advanced dual-cooled annular PWR assembly. The investigated claddings include Zircaloy-2/4, zirconium carbide (ZrC), silicon carbide (SiC), ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and austenitic 310 (SS-310), compared to the by design cladding for the AP1000 assembly, Zirlo™. Neutronic calculations were performed using the deterministic code DRAGON, and the parameters included infinite multiplication factor, criticality periods, neutron flux spectra, radial power distributions, rim effects, and reactivity feedback coefficients.The results show that the candidate fuel with Zirlo™ achieves a higher criticality period than UO2 fuel in both solid and annular configurations. The candidate fuel with SS-310 cladding exhibits a higher criticality period penalty of about 12.60% compared to Zirlo™, APMT, and FeCrAl claddings show a criticality period penalty of about 10.94% and 8.79%, respectively. The candidate fuel with ceramic claddings, SiC and ZrC, can achieve slightly higher criticality periods compared to Zirlo™ of about 1.11% and 0.24%, respectively, due to their low thermal neutron absorption cross-sections. Regarding the power profile, the annular (Th-233U-235U)O2 fuel exhibits a flatter behavior compared to UO2 fuel. However, the claddings with lower capture cross-sections exhibit slightly increased fission power through the fuel pellet. The fuel temperature coefficient for candidate fuel exhibits more negative values as the capture cross-section of the cladding increases, while the molybdenum and zirconium-containing claddings exhibit more favorable moderator temperature coefficients.

Investigating the possibility of extending the BWR cycle length for 15 years of operation by mixing highly enriched UO2 fuel with burnable absorbers

This paper investigates the possibility to extend the cycle length of boiling water reactor bundles to 15 years of operation with three different burnable poisons; gadolinium, erbium, and boron carbide. This can be carried out by mixing highly enriched UO2 fuel (15–19.9% U-235) with high concentrations of Gadolinium oxide (3–14% Gd2O3) or Erbium oxide (2–4% Er2O3).The boron carbide B4C was modeled as (Al2O3-B4C) rods in the bundle guide tubes. MCNPX code 2.7 was used to evaluate infinite multiplication factor (K-inf), power distribution, peaking factor, void reactivity coefficient, fuel cycle length, depletion of U-235, and fissile inventory ratio for the three designs at 40% void. The MCNPX simulation showed that introducing gadolinium rods at the bundle periphery has the advantage of lowering reactivity swing throughout the exposure range. The uniform distribution of erbium in all fuel rods contributed to the flattening of peaking factor at all the burnup stages. For the B4C design, the author found that the assembly with B4C–Al performs best in terms of reactivity flattening when five of the B4C–AL2O3 rods are positioned in the central region of the assembly. Furthermore, the fuel temperature coefficient is more negative for gadolinium design at all burnup stages. On the other hand, the boron model delivers the lowest control rod worth. Finally, the moderator temperature coefficient is more negative for erbium and WABA designs due to the enhanced thermal neutrons capture by the effect of the strategic arrangement of WABA rods and the uniform distribution of erbium.

Neutronic investigation of enriched gadolinium, natural gadolinium and Zirconium-erbium as burnable poisons in the advanced PWR assemblies by MCNPX code

In Pressurized Water Reactors, burnable absorbers are implemented for controlling the excess reactivity and avoiding the high power peaking factor. A feasibility study was carried out for investigating the effect of enriched gadolinium (157Gd2O3), natural gadolinium (152 - 160Gd2O3) and Zirconium-Erbium (ZrEr2) as burnable absorbers on the safety parameters of advanced PWR assembly. Three poisoned assemblies are modeled and simulated by the well-known MCNPX code. To hold a fair comparison between these assemblies, the author adjusted the concentration of fissile materials and burnable absorbers in each assembly to provide approximately the same initial reactivity. The MCNPX simulation showed that the enriched gadolinium model achieved a number of advantages over Zirconium-Erbium as: 1) prolonging the fuel cycle length by 1.04%, 2) increasing the negativity of moderator temperature coefficient and fuel temperature coefficient by 10.3% and 9.3% at the BOL and this contributes to core safety, 3) minimizing residual reactivity penalty at the higher burnup steps and 4) more hardening of neutron spectrum at the BOL. On the other hand, the Zirconium – Erbium layer presented a number of benefits over enriched gadolinium as 1) reducing the peaking factor by 1.3%, 2.6% and 2% at 20, 30 and 50 GWd/ton, 2) increasing the reactivity worth by 1.6% at the BOL, 3) increasing the power flattening by 7.6% at the EOL, and this is considered an enormous economic benefit. Therefore, the modeling of enriched gadolinium and zirconium erbium in this study is suggested for improving the neutronic and safety parameters of PWR cores.

Investigation on Neutronic Behavior of Pebble Bed Reactor for TRU Transmutation

One alternative effort to eliminate the accumulation of transuranic (TRU) elements generated from LWR spent fuel is by utilizing TRU as nuclear fuel in pebble bed reactor. This work was aimed to investigate the neutronic behavior of a pebble bed reactor used for TRU transmutation. The reactor geometry is adopted from the HTR-Modul. MCNP6 multipurpose radiation transport code with ENDF/B-VII.1 neutron library was used to calculate the neutronic aspects. The results indicated that the initial effective multiplication factor (keff) values decrease as the amount of TRU fuel pebbles within the core increases, concurrent with the fuel burnup. Increasing the TRU fuel pebble ratio resulted in weakened Doppler temperature coefficients (DTCs) and improvement in moderator temperature coefficients (MTCs). The calculated total mass of plutonium and minor actinide transmutation demonstrated that a pebble bed reactor with 100% TRU pebble can reduce TRU mass up to 55.47% of its initial mass load.

Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors, the influences of geometric parameters on the temperature coefficient of reactivity (TCR) at an assembly level were characterized. A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size. The results show that the fuel salt temperature coefficient (FSTC) is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region. Depending on the fuel salt channel spacing, the graphite moderator temperature coefficient (MTC) can be negative or positive. Furthermore, an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC. As the fuel salt volume fraction increases, the negative FSTC first weakens and then increases, owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing. Meanwhile, the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates. Thus, the negative TCR first weakens and then strengthens, mainly because of the change in the fuel salt density coefficient. As the assembly size increases, the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient, whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback. Then, the negative TCR weakens. Therefore, to achieve a proper negative TCR, particularly a negative MTC, an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.

Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one.

Conceptual design of the double tube burnable poison for the next generation small modular reactor

In this study, a new application of an innovative burnable poison (BP), which is called Double Tube Burnable Poison (DTBP), is introduced. Two design cincepts of the DTBP were tested in a small modular reactor (SMR) to achieve a longer operating cycle with low soluble boron concentration. These two concepts can be loaded in different locations in the fuel assembly, which increases the degree of freedom of nuclear design. The flexibility of this design allows it to increase its poison effect with the ability to control the depletion speed of its absorber materials. The DTBP pins were combined with the Erbia pins to achieve a flat k-infinite letdown curve with a significant reduction of the excess reactivity. The assembly and core calculations were performed by DeCART2D and MASTER codes. The results showed that the combination of DTBP and Erbia core achieved a cycle length of more than 3.5 Effective Full Power Years. Also, the highest value of Critical Boron Concentration was reduced to 309 ppm within the design limitations of the moderator temperature coefficient and the maximum pin power peaking. The performance of this option is the best among the other options including the conventional BP designs.

Searching for optimal accident tolerant fuel for the VVER-1200 reactor from the neutronic point of view

ABSTRACT In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2.

Neutronic Evaluation and Optimization of the Centrifugal Nuclear Thermal Rocket Concept

The centrifugal nuclear thermal rocket is a concept for a liquid-fueled nuclear system that would allow for a much higher specific impulse than the more traditional solid-fueled nuclear thermal propulsion designs. Although some preliminary neutronics analyses have been done on conceptual designs, this work seeks to perform a more systematic analysis and optimization of design parameters and to investigate additional neutronics properties such as power distributions and reactivity coefficients. This work used OpenMC for neutronics analysis and Dakota for the parametric study and optimization. Inter- and intra-fuel element power distributions were calculated, and a strong radial dependence was noted within fuel elements that may pose a challenge to thermal constraints. A positive moderator temperature coefficient of 3.78 0.16 pcm/K was calculated for the reference model, which may pose a challenge for system design and control. The optimization study of reflector size, fuel spacing, fuel mass, and fuel element radius indicated many trade-offs in the design considerations, and that the baseline model can be significantly improved in all respects. Positive reactivity feedback can be minimized by reducing moderation, and peaking factors can be reduced by limiting the amount of fuel per fuel element, which also minimizes the system mass.

Advanced fuel and burnable absorbers designed for long-cycle operation of BNPP

The present study is based on the facts that are happening for the Bushehr Nuclear Power Plant (BNPP) and coincides with the operational changes and step-by-step modifications of the BNPP in the near future. Bushehr reactor core will change from BNPP core to TVS-2M core. In other words, a new fuel, called the TVS-2M, is being used by the Russians to upgrade WWER-1000 reactors to reduce power generation costs, increase cycle length and fuel consumption. The neutronic analysis performed in this work is in line with this change. It provides a feasibility study of BA change and increase of cycle length in BNPP reactor without changing the initial structure of the reactor core. This research provides a new core configuration that can achieve the stated goals to a large extent without changing the BNPP core to the TVS-2M core. This research provides a feasibility study of Integrated burnable absorber (IBA) change by maintaining the initial neutronic conditions and a new high-performance configuration aiming for a long-cycle operation in Iran's nuclear power reactor. In Water-Water Energetic Reactors (WWER) reactors, two alternatives to CrB2Al (current burnable absorber in BNPP) are gadolinia (Gd2O3-UO2), and erbia (Er2O3-UO2). The benefits of the combinational IBA concept are expected by compensating weak points of each other. The main contribution of this research is the use of two cylindrical coaxial IBA layers with the same volume of gadolinium pins plus erbium (separated by a thin layer of zirconium alloy (Zr + 1% Nb)) and gadolinium pins. A neutronic evaluation has been conducted to investigate the variations of effective multiplication factor versus burnup, reactivity swing, power peaking distribution, moderator temperature coefficient (MTC), critical boron concentration (CBC), and cycle length. The result shows that the cycle length can be extended from 296 days in the BNPP core to 366 days in the proposed core, which is comparable to the cycle length of the TVS-2M core. The average BA concentration in the proposed core is 5%, which is lower than the average BA concentration of 5%–8% in the TVS-2M core. Regarding the gain of 73.90 days in cycle length, the revenue is estimated to be around 19,503,000 USD, making it worth the implementation considering the number of gadolinium and erbium pins in the proposed core.

A Novel Core Design with Movable Moderator for a Fluoride Salt–Cooled High-Temperature Reactor

The Fluoride salt–cooled High-temperature Reactor (FHR) is a Generation IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance inherent safety. In this study, an FHR core design with 165 MW of thermal output [MW(thermal)] is proposed. The reactor core employs tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe [lithium-beryllium fluoride (2 7LiF-BeF2)] as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from four aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that a larger fuel channel pitch would have a relatively harder neutron spectrum and yield a relatively longer cycle length, lower PPF, and better fuel temperature coefficient and moderator temperature coefficient (MTC). In addition, burnable poison (BP) (Er2O3) can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. The preliminary design of control blades is also presented in this paper. Furthermore, on the basis of the proposed 165-MW(thermal) core, we propose a novel core design that incorporates “fuel inside radial moderator (FIRM)” assemblies, movable moderator, and movable BP. This new design can extend the fuel cycle length by approximately 45 days for an 18-month fuel cycle. In addition, improvements were also found in PPF, discharge burnup, and temperature coefficients.