Burnable Poison Research Articles

Optimization of Fuel Enrichment and Reactor Core Pitch Dimension in Hydride Microreactor using NSGA-II Method and OpenMC Calculation

Abstract This study focuses on the design optimization of a hydride microreactor, aimed at supporting the equitable development of electricity in remote regions of Indonesia. Its compact design, inherent safety properties, and independence from the conventional electric grid make this microreactor concept suitable for electrifying even the most inaccessible areas. Current research investigates various aspects of the hydride microreactor design, employing a standard reactor core pitch dimension of 2.4 cm and fuel enrichment of 12%. It is crucial to operate the reactor using specific concentrations of Gadolinia burnable poison to enable a shutdown in the event of a stuck rod. By optimizing only the pitch dimension and fuel enrichment, the reactor can be simulated relatively quickly without the need for burnup calculations. Ideal configuration aims for lower enrichment, an extended fuel cycle, and a relatively low power peaking factor (PPF). The optimization process utilizes the Non-dominated Sorting Genetic Algorithm – II, with the neutron multiplication factor (keff) and PPF as primary objectives. Fuel cycle and feedback reactivity are the sub-objectives that are used to identify the optimal solution from the Pareto front. The calculation process of keff, PPF, fuel cycle, and feedback reactivity coefficients for specific pitch dimensions and fuel enrichment are done by utilizing the OpenMC Monte Carlo code. The optimal configuration is obtained through four generations of NSGA-II, featuring a reactor core pitch of 2.1 cm and fuel enrichment of 10.4%. This configuration yields a keff value of 1.0712049 and a PPF value of 2,0728 at the Beginning of Life (BOL). Remarkably, this configuration enables safe operation without using burnable poison for a continuous period of 12 years, generating 1 MW of thermal power. Emphasizing the reduction of fissile material is also essential to enhance the inherent safety of the reactor.

Assemblies design and modeling analysis of a new fine mesh neutronics/thermal-hydraulics coupling benchmark for plate-type PWR core

In order to support the verification of neutronics/thermal-hydraulics coupling calculation method or simulation codes at the fine mesh for plate-type pressurized water reactor (PWR) with high parameters (which the ratio of power to mass flow rate is greater than 235 kW/kg, the core outlet enthalpy exceeds 1500 kJ/kg), a set of coupling calculation of plate-type PWR based on high parameters (COPHP) is design by Nuclear Power Institute of China (NPIC). Multiple industry research teams, including Tsinghua University, Xi'an Jiaotong University, and Huazhong University of Science and Technology, participated in the production of COPHP benchmark. This article provides a detailed explanation of the completed assembly design and modeling calculations, and provides the keff, pin-by-pin wise fine mesh relative power distribution, and deviation results calculated for 10 conditions of 6 assemblies using RMC, OpenMC, and KYLIN V2 software. The results show that: 1) when using the same cross-sectional library, the OpenMC calculation results are in good agreement with the RMC results. Taking ENDF/B-VII.1 as an example, the maximum deviation of keff in the entire burnup process of all assemblies is −157pcm, the maximum deviation of relative power is −1.13 %, and the maximum power weight error(PWE) is 0.226 %. 2) Compared between KYLIN V2 and the RMC by using ENDF/B-VII.1, the maximum deviation of keff is −468pcm, the maximum deviation of relative power is −1.49 %, and the maximum PWE is 0.316 % when calculating the standard assemblies with all control rod out(ARO) condition and two burnable poison assemblies. The maximum deviation of keff is −795pcm, the maximum deviation of relative power is 1.81 %, and the maximum PWE is 0.369 % when calculating the standard assemblies with all control rod inserted(ARI) conditions.

Analysis of initial core and time dependent fuel burnup for high temperature testing reactors (HTTRs)

Abstract HTTRs are high temperature testing reactors which are considered generation IV nuclear power reactors. The reactor core startup and fuel burnup time dependent are simulated and analyzed. For the present computer model core startup, fuel loading with approach to criticality, control rod critical positions, scram reactivity and control rod worth, different core excess reactivity are evaluated. The results are compared with both typical measurements and other previously published methods. Moreover, the time dependent core model and fuel burn up are simulated to calculate the core multiplication factor, cycle length, discharged burnup, core isotopic evolution with time and burnable poisons behavior.

Research on spherical harmonics method based on the MOC neutron transport code OpenMOC

In the Method of Characteristics (MOC) neutron transport code OpenMOC, the scattering source term is processed with an isotropic scattering approximation. However, in the actual process of neutron transport, the probability of neutron scattering in various directions is not uniform. By expanding the scattering cross-sections with Legendre polynomials and the angular flux terms with spherical harmonics, anisotropic scattering treatment under the flat source approximation is incorporated in the MOC code OpenMOC. This study selects VERA benchmark problems 1–3, which include different temperatures and burnable poison absorber rods, as well as several problems containing MOX fuel assemblies to validate the higher-order scattering functionality. Additionally, a sensitivity analysis is conducted about the impact of the number of particle histories on simulation time and keff calculation accuracy in OpenMOC. The results show that, compared to the zeroth-order approximation and the transport-corrected approximation, anisotropic scattering treatment achieves higher calculation accuracy in most cases.

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.

Long term reactivity operation for a high temperature testing reactor (HTTR) loaded with uranium nitride fuel

A High Temperature Testing Reactor (HTTR) is one of the fourth generation nuclear reactors, the reactor uses UO2 as typical fuel with different axial and radial enrichments. In the present research, uranium nitride (UN) is proposed as a fuel for the reactor for the same core enrichment distribution. The core multiplication factor, cycle length, discharge burnup are calculated and the results of UO2 fuel are compared with UN fuel results. Neutron spectrum, fissile isotopes behavior and Breeding capability are compared for the two fuel types. Kinetic parameters, Control rods and Burnable poisons worth are also evaluated. The results indicated that uranium nitride (UN) has a longer fuel cycle, higher breeding capability and harder thermal flux compared to UO2 fuel.

The Effect of Burnable Poison Np-237 on the Conceptual Design of the GFR-2400 MWt Using MCANDLE Burnup Strategy

This study was used to design the Gas-cooled Fast Reactor (GFR) with 2400 MWt power as well as to implement a modified CANDLE (MCANDLE) burning strategy. Metallic uranium with Neptunium-237 at a concentration of 1% to 10% was used as burnable poison in the fuel while helium was applied as a coolant. Neptunium has been produced in conventional reactors by the decay product of plutonium obtained from U-238. This isotope has a relatively high absorption cross-section. As a nuclear waste, it is very useful to utilize as a burnable poison to reduce waste in the future and extend reactor operation lifetime with negligible reactivity. Moreover, the reactor was designed to operate for 100 years of burnup. Nuclear Fuel cell level calculations were conducted using the PIJ module and the reactor core was analyzed using the CITATION module contained in the SRAC. The core height was found to be 420 cm while the diameter was 300 cm and it was designed to have a refueling period of ten years. The survey parameters evaluated include burnup level, factor multiplication, conversion ratio, atomic density, and power distribution. The calculation conducted at the fuel cell level showed that the maximum value in the 76th year for the infinite multiplication factor was 1.30097. It was also discovered that the reactor core for the effective multiplication factor parameter with a fuel fraction of 55% was 1.094 without the addition of burnable poison.

Separation and determination of percentage level niobium in gadolinium zirconate using bifluoride fusion and spectrophotometry

A simple and reliable method for the determination of percentage level niobium (3.5-8 %) in highly refractive gadolinium zirconate (Gd2Zr2O7) doped with niobium, a burnable poison material of nuclear reactor has been proposed. A fusion procedure using ammonium bifluoride (ABF) with Gd2Zr2O7 was developed. Fusion conditions such as time, temperature and ABF to sample ratio were optimised for realizing quantitative recovery of Nb. The fused masses obtained under different conditions were characterized using SEM-EDX and XRD to comprehend the conversion of the refractive oxide form to soluble fluoro compounds and their identification. The real samples fused under optimized conditions were dissolved in 2 % tartaric acid solution and the same was used for quantifying Nb employing spectrophotometry using 5-bromo pyrogallol red as chromogenic agent. Precision of the measurements was better than 3 % (RSD) for Nb(V) at 0.5 ppm and the accuracy of the method is 99.8 %.

New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation

Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the Reactor Monte Carlo (RMC) code, offering multiple modes with varying levels of fidelity and computational requirements. The most precise mode is utilized in coupling simulations between RMC and the subchannel thermal-hydraulic analysis code SUBCHAN, incorporating an explicit moderator heating fraction in the coupling interface. The new treatment is verified against references from MCNP, Serpent, and OpenMC for three light water reactor (LWR) assembly cases, and great agreement is achieved. Energy deposition in different materials and components is emphasized in Kilowatt Reactor Using Stirling TechnologY (KRUSTY) modeling, and the results obtained using different modes are compared. The RMC-SUBCHAN coupling calculations for the three LWR assembly cases, employing the most accurate model, reveal a maximum increase of 94.6 K in the control rod centerline temperature, with a normalized energy deposition of 35.9% in the control rod regions. In the assembly case with gadolinium (Gd) burnable poison, a temperature increase of 7.3 K is observed in the Gd rod centerline, while the coolant outlet temperature decreases by 1.6 K due to the reduced explicit moderator heating fraction of 2.1%, compared to the constant 2.6% in the previous coupling scheme.

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.

Multiphysics analysis of a metal hydride moderated megawatt heat pipe reactor with burnable poisons

With the development of nuclear energy, microreactors have received increasing interest among researchers in recent years. In this paper, a megawatt heat pipe reactor with metal hydride moderators and burnable poisons is proposed. The hydrogen stability of the reactor under accident conditions, including reactivity insertion accidents, loss of power conversion unit heat sink accidents, and heat pipe failure accidents are analyzed. In this work, Gd2O3 is introduced as a burnable poison in the form of mixing with the UO2 fuel. According to the results of the burnable poison design, the 0.1% mass fraction of Gd2O3 is selected as the burnable poison loaded in the core. Safety analysis indicates that the introduction of burnable poisons can be beneficial during the positive reactivity insertion accident as it can reduce the excessive reactivity at BOL, thus reducing power peak and preventing hydrogen dissociation in ZrH1.4 rods. However, during the loss of PCU heat sink accident, ZrH1.4 rods will dissociate regardless of the presence of burnable poisons, whereas YH1.7 rods show better hydrogen stability. Moreover, in the event of the heat pipe failure accident, ZrH1.4 rods are more susceptible to dissociation than YH1.7 rods. As a result, the YH1.7+BP core is a better choice compared with other designs proposed in this paper as it provides a relatively high temperature margin.

Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor

Abstract 200 MWth lead-cooled fast reactor (LOTUS) reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at Vietnam National University (VNU) University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refueling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the startup core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers are withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refueling. Good neutronic behavior of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add to the burnable poison rods.

Performance analysis of an advanced polymeric burnable poison for PWR fuel assembly

Performance analysis of an advanced polymeric burnable poison for PWR fuel assembly

Study on damage of Gd2O3-CeO2 under electronic energy loss: comparison between bulk-like and nanostructure.

To understand the physical phenomena responsible for radiation damage of the materials used in nuclear reactors, and thus study their operation life and/or efficiency, it is required to simulate the conditions by exposing the materials to energetic ions. Ceria (CeO2) has been proposed as one of the inert matrices for the transmutation of minor actinides in the futuristic inert matrix fuel (IMF) concept. The inert matrix should also contain burnable poison to compensate for the initial reactivity of fuel. In this context, gadolinium (Gd) is an excellent burnable poison with a high neutron absorption cross-section. In view of this, Gd2O3-CeO2 nano-powders were synthesized and sintered at 800 °C and 1300 °C to obtain different grain sizes and morphologies. FESEM and TEM were carried out to study the grain size of pristine pellets. The sintered pellets were irradiated with 80-MeV Ag ions (electronic energy loss (Se) regime) at room temperature to emulate the effect of fission fragments. For analysis of the effect of grain size on the irradiation-induced structural degradation at different fluences, GIXRD and Raman spectroscopy were performed. Significantly large damage has been observed for the smaller grain-sized samples (sintered at 800 °C) as compared to the large grain-sized sample (sintered at 1300 °C). Neither of the samples amorphized under the present experimental conditions as indicated by the presence of the Raman-active T2g mode (centred at 462 cm-1) and all the XRD peaks of fluorite cubic structure up to the highest fluence employed (1 × 1014 ions cm-2). X-ray photoelectron spectroscopy results demonstrate that Ce4+ to Ce3+ and vacancy-related isolated clusters are the main defects produced in the systems. The radiation tolerance behaviour of the samples is understood with the help of thermal spike simulation, which indicates higher transient lattice temperatures with longer duration in the smaller grain-sized sample upon irradiation. Gd-doped ceria thus possesses good radiation stability in the Se regime, indicating its potential for application in IMFs.

Accelerating split-exponential track length estimator on GPU in TRIPOLI-4®

In the context of computing 3D volumetric scores for nuclear applications, combiningMonte Carlo methods andHigh Performance Computing is key to achieve precise, computationally tractable simulations andmeet the industrial deadlines. The next-event “split exponential track-length estimator” (seTLE) is typically adapted to the estimation of a tally in a mesh. We take advantage of the computing capabilities of Graphics Processing Unit (GPU) to mitigate the CPU-intensive tasks involved in the application of the seTLE: the estimation of cross section, the sampling of many out-going pseudo-particles at collision, the ray tracing across the geometry and the accumulation of scores in volumes. We discuss the performance improvements brought by porting each of these steps to GPU. We discuss the influence of the many parameters involved, and select a working point providing the best trade-off between the acceleration factor and the GPU utilization. We show on two examples thatwe obtain an acceleration factor of 8.56 on average over the entire map for the shielding application (up to 50 behind the concrete shield), and of 6.30 on average for the criticality application (up to 16 in the burnable poison tubes), compared to standard TLE on CPU.

Geometrical shape and self-shielding effect of burnable poison particles on pin-in block type HTGR neutronic performance

In our previous works, two different spherical burnable poison particles (BPPs) as B4C and Gd2O3 in pin-in block type HTGR core had utilized to suppress the excess reactivity and to control long-term reactivity during the burnup period. In the present work, we performed the neutronic analysis of a prismatic HTGR operating at 850 °C with thermal power of 100 MW containing spherical and cylindrical BPPs and then studied the self-shielding effect of BPPs and shape effect. The calculations were performed when the surface area (1) or volume (2) of cylindrical BPPs equals to that of the spherical BPPs. The calculations showed that the neutronic parameters were slightly better for the second case than the first one, such as the excess reactivity of the reactor core at the beginning of the cycle were more suppressed, the core lifetime were more extended, and the fuel-burning were more efficiently. The neutron spectrum in each region of the cylindrical BBPs slightly differs than that of the spherical BPPs. Therefore, the self-shielding effect of BPPs on reactor core performance depends on the particle's geometrical shape.

Investigation of reactivity control strategies for a micro-transportable gas-cooled thermal reactor

Micro nuclear reactors have gained increasing attention in recent years due to their high efficiency and long lifetime, making them an attractive solution for meeting the energy needs of remote and off-grid locations. Previous work has proposed a conceptual design for a micro solid nuclear reactor cooled by a helium-xenon mixture. The main aim of this study is to examine the effects of introducing burnable poisons on the neutronic performance of the designed micro nuclear reactor and to improve the design of the sliding reflector segment for reactivity control. The computational results indicate that the inclusion of burnable poisons can decrease the initial excess reactivity and radial power peaking factor, while not having a significant impact on the lifetime of the core. The use of 0.4%wt Gd2O3 and 0.75%wt ZrB2 as burnable poisons results in a reduction of 6.0 $ and 7.4 $ in the maximum reactivity of the core, respectively, while the core lifetime is reduced by 169 EFPD and 545 EFPD, respectively. By partitioning the sliding reflector, a linear introduction of reactivity can be achieved, and the influence of introducing burnable poisons on the reactivity control ability of the sliding reflector is relatively small.

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

This paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO2. The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB2 Integral Fuel Burnable Absorber (IFBA) and B4C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd2O3-doped UO2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

Comprehensive feasibility assessment of cycle length extension for CNPGS Unit-3

The core cycle length is crucial parameter for effective and economic operation of a nuclear power plant. Nuclear industry has great emphasis on improving the feasibility and sustainability of nuclear power generation. The cost of the electricity can be reduced by increasing the cycle length without violating the safety margins. Various methods and strategies have been examined to extend a Pressurized Water Reactor (PWR) core cycle length with optimized loading patterns. The objective of this work is to analyse the feasibility of the system to extend the cycle length of Chashma Nuclear Power Generating Station (CNPGS) unit-3, currently operating in Pakistan from 12 months to ∼18 months. To achieve the objective, fuel enrichment is increased from 3.4% to 4.1% with Gd2O3 as burnable poisons to suppress the excess reactivity. The multi-cycle analyses showed promising results and the core cycle length has been extended to ∼18-month for CNPGS unit-3 keeping power peaking factor, temperature coefficients and discharge burnup within design limits. The method can be confidently applied at any cycle of the power plant. Additionally, fuel and moderator temperatures at average and hot channels at beginning of life (BOL) and the end of life (EOL) are predicted using the 1-D steady state thermal hydraulic analysis. The peak to average temperatures at BOL and EOL are found well below the reference value (1.6875).

Loading method of Li rods for tritium production using High-Temperature Gas-Cooled reactor for fusion reactors

An initial tritium inventory is required to start fusion DEMOnstration Power Station (DEMO) reactors. However, a method to supply adequate tritium has not been determined yet. Tritium production via 6Li(n,α)T reaction by loading Li rods into the burnable poison (BP) holes of a high-temperature gas-cooled reactor (HTGR) has been proposed to address this problem (Matsuura et al., Nucl. Eng. Des. 243 (2012) 95 - 101). In previous preliminary studies, Li rods loaded in all BP holes were assumed to have the same design. This study evaluated whether the performance of Li rods can be improved for future optimization by adjusting the Li rod arrangement and the amount of Li compounds in them. The amount of tritium produced for gas turbine high-temperature reactor 300 (GTHTR300) was evaluated, while the total amount of Li compounds was maintained and the amount of loaded Li compounds changed depending on the layers and fuel regions. The maximum amount of tritium produced did not increase during the evaluations when reactor feasibility was satisfied. This implies that it is possible to reduce the number of Li rods while maintaining the amount of tritium produced for optimization, thereby reducing the costs of manufacturing Li rods and tritium recovery. GTHTR300 can produce 800 g of tritium in 360 days of operation using 2160Li rods. The results showed that the same amount of tritium could be produced by loading 720Li rods with the same number of fuel blocks. In addition, the effective multiplication factor, burn up, and power density of GTHTR300 were not significantly influenced during the operation.