Burn-up Fraction Research Articles

What next: Further implosion space exploration on the path to NIF extended yield capability

We present quantitative motivations and assessments of various proposed and ongoing directions to further improving yields and target gain of igniting indirect-drive implosions at the National Ignition Facility (NIF). These include increasing compression and confinement time, improving hohlraum and ablator efficiency, and further increasing peak power and laser energy. 1D hydroscaled simulations, augmented by analytic 1D theory, have been used to project yield improvements for each of these implosion optimization tracks, normalized to the best current performing 4 MJ shot. At current NIF capabilities of 2.2 MJ, 450 TW, we project several paths could reach 15 MJ yield levels. We also expect several key implosion physics questions will be addressed in attempting to reach this yield level. These include demonstrating to what extent lower adiabat designs leading to higher compression will increase gain and efficiency, and whether we can reduce residual kinetic energy and ablator-fuel mix that is probably limiting the current burn-up fraction. For an envisaged NIF upgrade to EL = 3 MJ at fixed 450 TW peak power, scaling capsule size and fuel thicknesses faster than pure hydroscaling should allow for yields that could reach up to 60–80 MJ, depending on the efficiency gains realized in increasing deuterium-tritium fuel thickness, reducing hohlraum losses, and switching to lower Z ablators. The laser-plasma instability and beam transmission scaling in these larger hohlraums is shown to be favorable if the spot size is increased with hohlraum scale.

Optimization of pellet design parameters to achieve deep fueling depth in EAST plasma with PAM code

Deep pellet fueling depth is necessary to achieve a high-density high confinement operation and to conduct some pellet-related researches in current devices, such as trigger of internal transport barrier, and to achieve high fusion power and tritium burn-up fraction in future fusion devices. The newly developed PAM code which can make a fast evaluation on pellet ablation and deposition is applied to optimize injection parameters to achieve deep pellet fueling. Systematic scans on pellet injection parameters including pellet injection positions, injection angles, sizes and speeds are performed for optimization purpose, while at the same time demonstrating flexibility and time efficiency of the PAM code. Dependences of the pellet fueling depth on these injection parameters are revealed by simulation results and analyzed. Simulation results indicate that pellet penetration contributes more to the deep pellet deposition than the ∇B -induced plasmoid drifts in low temperature plasmas, while deep pellet fueling in reactor relevant high temperature plasmas has to rely on plasmoid drifts. Though a shallow penetration is expected in high temperature plasmas, the ∇B -induced plasmoid drift is expected to be larger than that in relatively low temperature plasmas.

Design of the first fusion experiment to achieve target energy gain G>1.

In this work we present the design of the first controlled fusion laboratory experiment to reach target gain G>1 N221204 (5 December 2022) [Phys. Rev. Lett. 132, 065102 (2024)10.1103/PhysRevLett.132.065102], performed at the National Ignition Facility, where the fusion energy produced (3.15 MJ) exceeded the amount of laser energy required to drive the target (2.05 MJ). Following the demonstration of ignition according to the Lawson criterion N210808, experiments were impacted by nonideal experimental fielding conditions, such as increased (known) target defects that seeded hydrodynamic instabilities or unintentional low-mode asymmetries from nonuniformities in the target or laser delivery, which led to reduced fusion yields less than 1 MJ. This Letter details design changes, including using an extended higher-energy laser pulse to drive a thicker high-density carbon (also known as diamond) capsule, that led to increased fusion energy output compared to N210808 as well as improved robustness for achieving high fusion energies (greater than 1 MJ) in the presence of significant low-mode asymmetries. For this design, the burnup fraction of the deuterium and tritium (DT) fuel was increased (approximately 4% fuel burnup and a target gain of approximately 1.5 compared to approximately 2% fuel burnup and target gain approximately 0.7 for N210808) as a result of increased total (DT plus capsule) areal density at maximum compression compared to N210808. Radiation-hydrodynamic simulations of this design predicted achieving target gain greater than 1 and also the magnitude of increase in fusion energy produced compared to N210808. The plasma conditions and hotspot power balance (fusion power produced vs input power and power losses) using these simulations are presented. Since the drafting of this manuscript, the results of this paper have been replicated and exceeded (N230729) in this design, together with a higher-quality diamond capsule, setting a new record of approximately 3.88MJ of fusion energy and fusion energy target gain of approximately 1.9.

Energy gain of wetted-foam implosions with auxiliary heating for inertial fusion studies

Low convergence ratio implosions (where wetted-foam layers are used to limit capsule convergence, achieving improved robustness to instability growth) and auxiliary heating (where electron beams are used to provide collisionless heating of a hotspot) are two promising techniques that are being explored for inertial fusion energy applications. In this paper, a new analytic study is presented to understand and predict the performance of these implosions. Firstly, conventional gain models are adapted to produce gain curves for fixed convergence ratios, which are shown to well-describe previously simulated results. Secondly, auxiliary heating is demonstrated to be well understood and interpreted through the burn-up fraction of the deuterium-tritium fuel, with the gradient of burn-up with respect to burn-averaged temperature shown to provide good qualitative predictions of the effectiveness of this technique for a given implosion. Simulations of auxiliary heating for a range of implosions are presented in support of this and demonstrate that this heating can have significant benefit for high gain implosions, being most effective when the burn-averaged temperature is between 5 and 20 keV.

Operational performance parameter range in ICF propulsion theory

This paper explores the operational performance parameter range for inertial confinement fusion propulsion systems under the assumption of best case and worst case mission scenarios. This includes the spacecraft thrust, jet power and specific power. We show the derivation of the key driving equations and then simplify these with approximations for ease of domain space analysis. This includes a consideration of minimum and maximum values for jet efficiency, momentum weighting factor, nozzle divergence angle, fuel burn-up fraction, capsule mass and detonation pulse frequency. It is estimated that the range of parameter values may be of order 0.001 N–100 MN thrust, 1000 W–1,000 TW jet power, 0.01 W/kg–10 GW/kg specific power. Although this range includes values that are likely outside the realistic design space of applicability and to show this data from published designs is also examined. The analysis is considered for reaction isotopes of low mass hydrogen and helium nuclei only and does not take into account the possibility of propellant expellant augmentation, vehicle staging, design optimality or enhanced fuel reactions.

Improvement of Few-Group Homogenized Cross-Sections for RSG-GAS In-Core Fuel Management Code Batan-FUEL

Abstract This paper presents the generation and the verification of the few-group homogenized cross section for Batan-FUEL code. This code is routinely used for fuel management in the G. A. Siwabessy Multipurpose Reactor (RSG-GAS). The Monte Carlo code Serpent 2 code, in conjunction with the latest nuclear data library ENDF/B-VIII.0, was used. Calculations using the existing newly generated few-group cross section data were carried out for the 88th core. The calculated core parameters such as excess reactivity and control rod worth are compared to the available experimental data. On the other hand, the fuel burnup fraction and radial power peaking factor (PPF) are compared to the results of Serpent 2. It was shown that the new cross section data have more consistency and a better agreement excess reactivity and control rod worth compared to the experimental data. Similarly, the U-235 burnup fraction and radial power peaking factor by the new cross section data are also shown to concur well with Serpent 2. The newly generated few-group cross section is recommended to replace the existing ones for the fuel management of RSG-GAS.

Self-consistent investigation of density fueling needs on ITER and CFETR utilizing the new Pellet Ablation Module

Self-consistent modeling using the stability, transport, equilibrium, and pedestal (STEP) workflow in the OMFIT integrated modeling framework (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as China Fusion Engineering Test Reactor (CFETR) Zhuang et al (2019 Nucl. Fusion 59 112010) will benefit from high-density operation (Greenwald limit fraction 0.7−1.3). Regimes with an operational density near the Greenwald limit will likely need peaked density profiles so that the pedestal density remains below the Greenwald limit. Peaked density profiles can be achieved with the help of pellet injection. A flexible Pellet Ablation Module (PAM), which predicts the density source based on a comprehensive analytical pellet ablation model, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. This workflow is applied to DIII-D and finds good agreement with experiments. On ITER the effect of pellet fueling is examined in an advanced inductive scenario, where a fusion gain of up to Q = 9 is predicted with strong central pellet fueling. On CFETR, with a mid-radius density source, an average of electrons s−1 are required to achieve the density and temperature profiles necessary for the 1000 MW advanced scenario with a tritium burn-up fraction of .

Investigation on thermo-mechanical performance of fully ceramic microencapsulated fuel

• A finite element method (FEM) model for FCM fuel thermo-mechanical performances analysis was presented. • The thermo-mechanical behaviors, coupled with irradiation behaviors were developed. • A 3D modeling of FCM fuel pellet with typical packing fractions was realized by an optimized stochastic strategy. • Effects of thermal boundaries, burnup, particle spacing, no fuel zone width and other key parameters were well investigated. Fully ceramic microencapsulated (FCM) fuel is a promising ATF concept with extraordinary fission product inclusion capacity, excellent oxidation and corrosion resistance, and high thermal conductivity. The thermo-mechanical performances are of great significance when evaluating the safety and applicability of a nuclear fuel. In this study, a finite element method (FEM) model for FCM thermo-mechanical performances analysis was presented. The thermo-mechanical behaviors, coupled with irradiation behaviors (including irradiation-induced dimensional change (IIDC), heat generation, conduction, thermal expansion, creep, burnup, and gap/plenum pressure, etc.), were developed to investigate the thermal and mechanical characteristics of TRISO-based FCM Fuel. A simplified model of TRISO fuel particles was established and validated. After that, a full 3D modeling of FCM fuel pellet with typical packing fractions was realized by an optimized stochastic strategy. A numerical method was validated and successfully applied to realize the simulations of a FCM fuel pellet. Two different thermal boundaries, namely fixed particle power and fixed pellet power, were performed to investigate the irradiation behavior of FCM fuel. Moreover, the effects of burnup, particle spacing, no fuel zone width and other key parameters were well studied and investigated. The performed calculation showed that temperature and stress in matrix and TRISO are both positively correlated with the packing fraction. It's worth noting that compared with the cases of fixed particle power, cases of fixed pellet power may lead to higher fuel temperature and higher stress. The results also implied that from the perspective of probability evaluation, most of the TRISO particles in FCM fuel can maintain the structural integrity under different packing fractions, but some of the particles with higher stress may crack or damage, especially in the cases of high burnup and high packing fraction.

25-Pin metallic fuel performance benchmark case based on the EBR-II X430 experiments series

A metallic fuel benchmark case was developed for the fuel performance code BISON based on 25 uranium-zirconium and uranium-plutonium-zirconium pins of the Experimental Breeder Reactor II X430 experiment series. Results of the benchmarks were compared with measurements and calculations made at the time of the experiment as well as subsequent measurements reported in 2019. The comparisons were used to quantify the accuracy of the BISON predictions and to identify patterns in the BISON differences. BISON predicted burnup , plenum pressure, and fission gas release fractions accurately. BISON temperature predictions were somewhat cooler than the temperatures determined at the time of the experiment but appeared to be reasonably accurate considering uncertainties in the legacy temperature calculations, uncertainties in the legacy linear heat rate calculations, and the high sensitivities of the BISON-predicted temperatures to BISON inputs. Fuel axial elongation predictions had errors correlated to fuel composition; BISON tended to underpredict the elongation of binary fuels and overpredict the elongation of ternary fuels. BISON cladding radial dilation predictions were also significantly lower than legacy PIE measurements. Recommendations were made to improve the BISON fuel gaseous swelling model to account for fuel composition, to add additional capabilities to the coolant channel temperature model to ease benchmark development, and to continue developing benchmark cases based on a wide range of experiments in several reactors. Once a wide array of benchmarks is developed, an attempt can be made to enhance or calibrate BISON models to improve the cladding dilation predictions.

Analysis of the whole process tritium transport based on fuel cycle modeling for CFETR

Abstract Tritium self-sustain is crucial for the successful operation of China Fusion Engineering Test Reactor (CFETR). In this work, tritium fuel cycle model is developed based on the practical tritium plant. The retention values caused by different purification technology which can’t be extracted during the fuel cycle is considered in several sub-systems like Isotope Separation System-Inner cycle(ISS-I). The typical inventory in each sub-system is calculated which resulting that the minimum initial startup inventory (Im) is about 3131 g and tritium breeding ratio that required to satisfy tritium self-sustain (TBRreq) is about 1.13 under typical parameters. The sensitivity relation of Im and TBRreq are analyzed. We found that the fractional burnup is a key factor to affect both.

РАСЧЕТНОЕ МОДЕЛИРОВАНИЕ ДОЖИГАНИЯ ТВС ЭНЕРГОБЛОКОВ № 1 И № 2 ЛЕНИНГРАДСКОЙ АЭС ПОСЛЕ ВЫВОДА ИЗ ЭКСПЛУАТАЦИИ

The RBMK-1000 reactor operates in a continuous refueling mode. The reactor core comprises FAs across the burn-up spectrum, from fresh fuel assemblies to the most burnt-up ones. Following the reactor shutdown for decommissioning, the majority of irradiated fuel assemblies (IFA) have the potential for further use (the so-called afterburning) at operating NPP units. Most of the irradiated fuel assemblies have a burn-up fraction, which is far from the specified threshold value. The neutron multiplication factor in the cell exceeds the core average value and, therefore, the IFA loading makes it possible to increase the reactivity margin to the desired value. Reuse of spent fuel assemblies in reactors at other power units offers a number of advantages. In economic terms, afterburning is able to reduce the use of fresh fuel assemblies (FFA). In terms of radioactive waste handling, afterburning is able to reduce the number of irradiated fuel assemblies, which require long-time storage in a spent nuclear fuel repository thus reducing the IFA and radioactive waste disposal loads. JSC NIKIET has proposed the use of irradiated fuel assemblies from units 1 and 2 of Leningrad NPP shut down after decommissioning in reactors at units 3 and 4 of the same NPP.

Analysis of the influence of the different impurity seeding on the burn-up fraction and plasma confinement in the EU DEMO reactor

Abstract This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar, Kr and Xe) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the impurity seeding on the burn-up fraction and plasma confinement in EU DEMO reactor. For the simulation with constant electron density at the separatrix, it is found that impurity seeding has a small influence on the burn-up fraction, which remains around 6 – 7%., but fuelling source is reduced from 2.7 to 2 × 1022 1/sec when moving from lowest to highest seeding level. The degradation of confinement for high Z seeding impurity, which is correlated to radiation in the core (due to seeded Kr and Xe) is observed. Better confinement for Ar at middle and high seeding level is observed in comparison to the case without seeding, because Ar radiation is small in the core (about 40-55 MW).

Theory of ignition and burn propagation in inertial fusion implosions

A detailed analytic model is presented here to investigate the physics of burn propagation in inertially confined plasmas. The onset of ignition and burn propagation occurs when alpha heating of the hot spot causes rapid ablation of shell mass into the hot spot. This allows large energy gains to be achieved since most of the fuel mass is located in the shell. Here, we first present a comprehensive review of previous analytic models that have been used to describe the physics of hot-spot evolution and ignition; we then show that a proper description of a propagating burn wave requires a comprehensive model of hot spot and shell evolution that includes proper mass conservation in the shell, fusion reactivity, and fuel depletion. The analytic theory is in good agreement with detailed radiation-hydrodynamic simulations that predict the onset of burn propagation as occurring when the yield enhancement caused by alpha heating is between 15- and 25-fold, fα ∼ 1.4, where fα = alpha energy deposited/hot-spot energy at bang time, and the hot-spot burnup fraction is approximately 2%. We show that the definition of ignition is not sensitive to the alpha-particle stopping power nor asymmetries provided that the absorbed fraction of alpha particles θα is correctly accounted for. Finally, we use the results of 2-D simulations to show that even when θα is small and unknown (as is true in hot spots with mid modes that have significant leakage of alpha particles into the surrounding cold bubbles), one can still relate the experimentally measureable parameter χα53 to the yield amplification and the burning-plasma parameter Qαhs = alpha energy deposited/total input work delivered to the hot spot.

Evaluation of tritium burnup fraction for CFETR scenarios with core-edge coupling simulations

A key mission for the next-step fusion tokamak device China Fusion Engineering Test Reactor (CFETR) is demonstrating tritium self-sufficiency, which requires a sufficiently high tritium burnup fraction (fburnup) in order to match a practically achievable tritium breeding ratio (TBR) with the blanket design constraints. Core-edge coupling simulations are performed to investigate the dependence of fburnup on different controlling parameters for CFETR scenarios. Core plasma profiles with a range of pedestal densities are simulated by consistent iterative calculations of equilibrium, transport, auxiliary heating and current drives within the OMFIT framework. The core-SOL integrated COREDIV code is then used to evaluate fburnup with the OMFIT modelled core plasma parameters as input. According to the simulations, fburnup can be effectively increased with a higher pedestal density on account of the fusion power increasing faster than the fueling source required to maintain steady-state. Higher can also increase the fburnup due to increase of fuel recycling. Deeper pellet fueling deposition and lower ratio of particle to thermal diffusivities D/χ can both increase the effective particle confinement time and thus fburnup. However, the effect of helium and other impurities (Ar and W) is shown to reduce fburnup for comparable impurity and main ion transport. Based on our analysis, using present pellet fueling technology, achieving fburnup > 3% for CFETR will be very challenging. This is a lower limit for the required TBR (>1) to match the achievable TBR for tritium self-sufficiency. Furthermore, our study suggests that if fueling can penetrate deeper than r/a < 0.8 under optimistic conditions, the required burnup fraction could be attainable. The modelling results thus provide important suggestions and implications for the optimization of CFETR scenarios and development of advanced fueling systems.

Burnup Calculation Study of Pebble Bed Equilibrium Core

Reaktor Daya Eksperimental (RDE) is a high temperature gas-cooled reactor (HTGR) producing a 10 MW thermal with a pebble bed fuels which being developed by BATAN. The purpose of this paper is to study the burnup distribution and characteristic in the equilibrium core for different multipass recirculation method. Understanding the reactor physics, in particular the burnup, is important for optimum design and safety analysis of nuclear reactor. Related with the design approval phase of RDE, this study provides a design data which needed for the RDE’s safety analysis report, e.g. fuel pebble performance analysis. Analysis in this study was performed using PEBBED diffusion code. PEBBED is designed to solve the neutronics and thermalhydraulics parameter cases for high temperature pebble bed reactors for different fuel recirculation including once-through-then-out (OTTO) and multipass scheme. The reactor core calculation was performed by applying multipass scheme with variation from 5 passes up to 15 passes. With energy and power as inputs, the calculations produce the burnup fraction at the end of the cycle and the fast and thermal neutron fluxes in 8 energy groups. The calculation results showed that the lowest 5 passes fuel recirculation pattern has the highest and lowest minimum discharged burnup value. This is related to the average power distribution in the core which means more passes will flatter the burnup distribution including the power distribution and also seen in the value of flux and power peaking factor produced. power peaking factor produced.

Fuel Burn-up and Radioactivity Inventory Analysis for New In-core Fuel Management of the RSG-GAS Research Reactor

A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new fuel management pattern is done so that the replacement of fuel in every operating cycle is more efficient. The new fuel replacement pattern required various safety analyzes, among others related to the fuel fraction and radioactivity inventory of the RSG-GAS. Both of those are the main elements in dose acceptance analysis for workers and communities around the reactor when the reactor operates both in normal or in abnormal conditions. Fuel burn-up and inventory analysis was performed using ORIGEN2 computer program. Inputs for ORIGEN2 are the fuel mass, the time required for one operating cycle as well as the peak power factor in each fuel in the specified fuel management pattern. The result for the 97th Core configuration (T97) is that the average burn-up fraction of each cycle is 6.79%, and the maximus fuel fraction is 52.36% for Fuel Elemet and 56.52% for Contro Elemet. It is also obtained that the largest and dangerous human inventory activity at the end of cycle (EOC) for Iodine radionuclide group is I-131 for 5.18E+04 Ci, and for the alkali metal radionuclide group Cs-137 of 7.65E+03 Ci.

A New Concept for a Higher Burn-Up Fraction Improvement in DEMO Reactor

The possibility of improving the fuel burn-up fraction by recirculating a part of the helium and fuel particles and separating helium atoms from fuel atoms is investigated for the DEMOnstration Power Plant conventional divertor configuration. It is shown that the burn-up fraction can be considerably improved by a continuous exhaust gas reinjection.

Phase and Chemical Transformations of the β–Radioactive Intermetallics in Fast-Neutron-Irradiated Uranium–Plutonium Nitride

Thermodynamic analysis of the chemical and phase compositions of uranium–plutonium nitride (U0.8Pu0.2)N0.995 irradiated by fast neutrons shows that a complex structure forms in it. It consists of a solid solution containing uranium, plutonium, americium, neptunium, zirconium, yttrium, and lanthanide nitrides and individual condensed phases (U2N3, CeRu2, Ba3N2, CsI, Sr3N2, LaSe, (Mo + Tc), U(Ru, Rh, Pd)3. The contents and compositions of these phases are calculated at a temperature of 900 K and a burn-up fraction up to 14%. The β– decay of the metallic radionuclides is found to change the phase and chemical compositions of the irradiated (U0.8Pu0.2)N0.995 nitride. The kinetics of the transformations of the radioactive intermetallics U109Pd3 → U + 3109Ag, 141Ce103Ru2 → 141Pr103Rh2, and 144Ce106Ru2 → 144Nd106Pd2 is calculated.

A methodology for performing sensitivity analysis in dynamic fuel cycle simulation studies applied to a PWR fleet simulated with the CLASS tool

Fuel cycle simulators are used worldwide to provide scientific assessment to fuel cycle future strategies. Those tools help understanding the fuel cycle physics and determining the most impacting drivers at the cycle scale. A standard scenario calculation is usually based on a set of operational assumptions, such as reactor Burn-Up, deployment history, cooling time, etc. Scenario output is then the evolution of isotopes mass in the facilities that composes the nuclear fleet. The increase of computing capacities and the use of neutron data fast predictors provide new opportunities in nuclear scenario studies. Indeed, a very high number of calculations is possible, which allows testing a high number of operational assumptions combinations. The global sensitivity analysis (GSA) formalism is specifically well adapted for this kind of problem. In this new framework, a scenario study is based on the sampling of operational data, which become input variables. A first result of a scenario study is the highlight of relations between operational input data and outputs. Input variable subspace that satisfy optimization criteria on an output, such as plutonium incineration or stabilization, can also be determined. In this paper, a focus is made on the methodology based on GSA. This innovative methodology is presented and applied to a simple fleet simulation composed of a PWR-UOx fuel and a PWR-MOx fuel. Calculations are done with the fuel cycle simulator CLASS developed by the CNRS/IN2P3 in collaboration with IRSN. The design of experiment is built from five fuel cycle input sampled variables. Sensitivity indices have been calculated on plutonium and minor actinide (MA) production. It shows that the PWR-UOx Burn-Up and the fraction of PWR-MOx fuel are the most important input variables that explain the plutonium production. For the MA production, main drivers depend strongly on isotopes. Sensitivity analysis also reveals input variable subspace responsible of simulation crash, what led to an important improvement of the model algorithms. An equilibrium condition on the plutonium mass in the stockpile used for building MOx fuel has been applied. The solution is represented as a subspace in the PWR-UOx Burn-Up and PWR-MOx fraction input space. For instance, achieving a plutonium equilibrium in a stockpile fed by a PWR-UOx that operates at 40 GWd/t requires a PWR-MOx fraction between 9 and 14%. This study also provides data related to plutonium incineration induced by the utilization of the MOx.

FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor) type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN) and the IAEA (International Atomic Energy Agency). In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U 3 Si 2 -Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92), all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.