Fuel Volume Fraction Research Articles

Fuel element microreactor integrating a square UO2 fuel rod with an internal heat pipe

In this study we propose the Fuel Element Micro Reactor (FEMR), with 308 square UO2 fuel rods, each one containing inside a heat pipe. We selected Mercury as the working fluid for the heat pipes and operate the microreactor at average temperature around 900 K or 637 °C. Heat pipes introduce empty regions into the core that increase neutron leakage and severely reduce core reactivity. Square fuel lattices allow greater fuel volume fractions and favor increasing core reactivity. The adoption of a 20 cm BeO reflector and Zircaloy as fuel cladding further increases core reactivity. The FEMR core thermal power is 2.3 MW, the power density distribution is rather flat with peaks occurring at the core-reflector interface and the peak fuel temperature is 1160 K. The reactor has a negative temperature isothermal coefficient of reactivity. The reactivity control system with 8 rotational drums and a central cruciform absorber rod meets the criteria of stuck rod and diversity of engineering principles. The average fuel discharge burnup is 8.96 MWd/kgU and the fuel cycle length without refueling is 8.7 years. The ratio between thermal power and total Uranium mass at beginning of life is 2.46 kW/kgU.

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

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

A comparative analysis of gas-cooled fast reactor using heterogeneous core configurations with three and five fuel variations

GFR or Gas-cooled Fast Reactor is one type of fast generation-IV that uses a very high cooling temperature. Thus, it is necessary to have the right reactor core design so that the power distribution of neutrons produced reaches a safe and even limit point. The use of a uniform (homogeneous) reactor core can produce peaking power. This is very avoidable because it will cause a reactor accident. In this study, researchers tried to compare the results of the analysis for two heterogeneous reactor core designs including the configuration of 3 fuel variations and 5 fuel variations using UN-PuN fuel. This study aims to determine the keff value produced by both types of fuel variations during 5 years of burn-up and determine the characteristics of neutron flux, fission rate, and fission product during 15 years of burn-up. This study was started by calculating the homogeneous and heterogeneous core of 3 and 5 fuel variations with neutron transport simulation involving OpenMC. The calculation results show that the heterogeneous core configuration of 5 fuel variations for the keff value is more optimal than 3 fuel variations, because it has the smallest excess reactivity value. The neutron flux and fission rate characteristics for 5 fuel variations are more evenly distributed when compared to 3 fuel variations to maintain neutron lifetime and reactor life in operation. Burn-up residual plutonium material and minor actinide waste for 5 fuel variations have less mass than 3 fuel variations. The results of neutronic analysis of GFR reactors with heterogeneous reactor core designs for 5 fuel variations are better in terms of reactor criticality, neutron power distribution, and waste produced. Finally, optimization of the UN-PuN fuel volume fraction of 60 % provides the optimal keff value

Inverted fuel geometry implementation for a fast reactor: Potential improvements in neutronic and thermalhydraulic performance

This paper presents a detailed procedure for implementing the inverted fuel geometry to a fast reactor to improve its safety system and economy. The study is starting from the fuel unit cell, fuel assembly, reactor core, and burnup analysis. A proposed multivariable graph (v, ΔP, TFmax–Dco, VF) introduced at the fuel unit cell level provides comprehensive thermalhydraulic and neutronic parameters in a single graph, allowing for an efficient optimization process. The fuel unit cell study reveals that the inverted fuel design has a higher fuel volume fraction and lower core pressure drop than conventional pin-typed fuel. This is beneficial for the reactor economy and enhances the reliability of the safety system. With the inverted fuel design, the primary loop can save pumping power by up to 40 % and provides an excess driving force for natural circulation. The male–female axial grid structure separating the fuel assembly potentially eliminates coolant flow path restriction and fretting issues. The core is named the Inverted Core Fast Reactor (IC-FR), an LBE-cooled fast SMR designed to generate 60 MWth for at least 40 years of full-power operation without refueling and fuel shuffling. IC-FR is a transportable reactor and has load following capability that can be deployed for many applications, including marine and land-based applications, and stand alone or mixing power grid. The burnup study of IC-FR reveals that the balance of neutron leakage and fissile inventory yields a small reactivity swing (<1$) for 40 years. This study extensively utilizes Monte Carlo (MC) code MCS for neutronic calculation. Owing to the high computational expense of MC code, the approaches to optimize the MC usage are also presented.

Integrated simulation of U-10Mo monolithic fuel swelling behavior

A separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to consider various processes occurring in RTR fuels during irradiation. To accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions versus fission density) predicted using the phase-field method. The remainder of the fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration, detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

Design study of 300MWth GFR with UN-PuN fuel using SRAC-COREBN code

Research related to 300MWth with UN-PuN fuel optimization on GFR using 3D core (X-Y-Z) with the code SRAC-COREBN has been carried out. UN-PuN fuel optimization was carried out by studying neutronic parameter analysis on variations in the addition of Pu-239 percentages in homogeneous and heterogeneous core configurations, as well as on variations in fuel volume fractions. The design of the reactor core is cylindrical pancake with a diameter of 300 cm and a height of 100 cm. Neutronic calculations are carried out in two steps, the first is the calculation of burn-up per fuel pin using the PIJ code system on SRAC and the JENDL4.0 nuclide library. The second calculation is the calculation of burn-up at the reactor core level using the COREBN code. The results of the core burn-up calculation analyzed were neutronic parameters, such as k-eff, excess reactivity, and power density. The optimal design based on the neutronic parameter analysis that has been carried out is a fuel volume fraction of 64% with a heterogeneous core configuration of 3 types of fuel percentage, fuel 1- fuel 2- fuel 3 = 7.5%-8%-8.5%. The maximum k-eff value is 1.0031841 with a maximum excess reactivity of 0.32%. The maximum power density for the X and Y directions is 70.78 Watt/cc; while for the Z direction it is 67.97 Watt/cc.

Comparative Study of UO2 and (Th,U-233)O2 Performance in Small Long-Life PWR Fuel Cell

A study was performed comparing the performance of UO2 and (Th,U-233)O2 fuel in small long-life PWR. The neutronic calculation carried out by PIJ module of SRAC2006 was done to a fuel cell in 10 years of operation. The calculation was conducted by varying the enrichment of U-235 in UO2 and U-233 in (Th,U-233)O2 for 1% - 20% and also by varying the fuel volume fraction for 40%, 45%, 50%, 55%, and 60%. The performance was observed by comparing the enrichment needed by each fuel type to gain criticality in 10 years, the infinite multiplication factor (k-inf) value, and the conversion ratio (CR) value. The calculation results showed that 60% fuel volume fraction gave critical conditions with the lowest infinite multiplication factor and highest conversion ratio for both fuel types. While in terms of fissile nuclide enrichment needed, (Th,U-233)O2 had better performance than UO2, because only 5% U-233 was needed in (Th,U-233)O2 while UO2 needed 9% U-235 to gain criticality in 10 years of operation.

Parametric Study of Thorium Fuel Utilization on Small Modular Pressurized Water Reactors (PWR)

Small Modular Pressurized Water Reactor (PWR) can be used to overcome the problem of electricity shortages in remote areas with fewer populations. These reactors usually use uranium-based fuel. Currently, the use of thorium-based fuels for this type of reactor has received constant attention, leading to several investigations of fuel performance. This study aims to determine the performance of thorium-based fuel in the PWR core at a power level of 300 MWt. The analysis was carried out by varying the percentage of U-233 as fissile material in Thorium (Th-233U)O2 fuel by 3%-8% and Pa-231 variation as burnable poison 0.2%-5%. In addition, analysis of changes in the volume fraction of fuel ranging from 40%-60%, with the type of cladding used in the form of Zirlo. The calculation system uses the SRAC computer code with PIJ and CITATION modules for cell and core calculations, using JENDL-4.0 nuclide data. Parameters analyzed from the fuel cell calculation results are reactor criticality, excess reactivity, and power density profile. The analysis results show that the best fuel performance is found in fuel volume fraction of 60%, U-233 configuration on the fuel of 4-5-6%, and the addition of 2.5% Pa-231. This fuel configuration can achieve cycles of up to 19 years with excess reactivity at BOL of 2.57% dk/k and excess reactivity at EOL of 0.87% dk/k.

Outfield overpressure characteristics and flame behavior of fuel vapor explosion in tubes with one weakly covered end

An experiment system of tubes with one weakly covered end was established, and the outfield characteristics of fuel vapor explosions were studied through experiments. The results are as follows. After the rupture of the weakly covered end, three outfield overpressure peaks appeared due to the rupture of the weakly covered end (P1), the venting of the flame (P2), and external explosion (P3). The value of overpressure peaks showed a downward trend with an increase in distance, and the maximum overpressure was formed due to the external explosion. The value of overpressure peaks were related to the L/D ratio of the tube. With an increase in the L/D ratio, P1 and P3 in the near field showed an upward linear trend, while P2 and P3 in the far field showed a quadratic trend. In the case of the same L/D ratio, the maximum overpressure largely depended on the fuel volume fraction and external distance. The flame shape showed a changing process of jet flame - radial tension - mushroom cloud-shaped flame. The mushroom-shaped flame was formed mainly due to the Kelvin-Helmholtz instability and vortex effect, and the value of axial flame parameters were greater than those of radial ones. With an increase in the L/D ratio, the axial flame parameters showed an upward linear trend, the radial flame parameters showed a downward linear trend, and their ratio showed an upward linear trend.

Ameliorating the positive temperature feedback coefficient for an MSR fueled with transuranic elements

Transuranic (TRU) elements can be used as an alluring fuel in a graphite-moderated Molten Salt Reactor (MSR) with liquid fuel and online reprocessing. However, the positive temperature feedback coefficient (TFC) stemming from the resolvable fission resonances of Pu-239 and Pu-241 at thermal neutron energies should be addressed. The aim of this study is to propose an effective solution to ameliorate the positive TFC, by adding Er-167 into the fuel salt or adjusting the minor actinides (MA) ratio in TRU. Er-167, with a strong capture resonance cross section of about 10,000 barns at 0.5 eV, is one of the most promising poisons. After adding Er-167 into the fuel salt with hectogram scale, the fuel TFC is decreased significantly to below −10 pcm/K and the total TFC to −2 pcm/K correspondingly. Similar to Er-167, some MA isotopes, especially Np-237 (at 0.5 eV) and Am-241 (at 0.3 eV, 0.6 eV and 1.1 eV) reveal significant capture resonance cross sections (>1000 barns). In this way, increasing the MA ratio in TRU is another available approach to obtain a negative TFC. The simulated results show that adjusting the MA ratio up to 50% can ameliorate the TFC efficiently. Combing the negative TFC and the salt solubility limit, only a narrow region of the fuel volume fraction lower than 25% can meet the two requirements, and the eligible region becomes smaller for a lower MA ratio.

Analisis Fraksi Volume Bahan Bakar Uranium Karbida Pada Reaktor Cepat Berpendingin Gas Menggunakan SRAC Code

Analysis of fuel volume fraction with uranium caride fuel in Gas Cooled Fast Reactor (GFR) with SRAC Code is has been done. The calculation used SRAC Code (Standard Reactor Analysis Code) which is developed by JAEA (Japan Atomic Energy Agency), and the data libraries nuclear used JENDL 4.0. There are two calculation has been used, fuel pin cell calculation (PIJ Calculation) and core calculation (CITATION Calculation). In core calculation, the leakage is calculated so the calculation more precise. The CITATION calculation use two type of core configuration, i.e. homogeneous core configuration and heterogeneous core configuration. The power density value of two type core configuration is quite difference. It is better use heterogeneous core configuration than homogeneous core configuration, because the power density of heterogeneous core configuration is flatter than the other. From the analysis of fuel volume fraction, when the volume fraction is increase, the k-eff value is increase. And the optimum design after has been analysis for fuel volume fraction, that is the fuel volume fraction is 49% with a heterogeneous core configuration of three types of fuel percentages, for Fuel1 9%, Fuel2 12% and Fuel3 15%. This reactor is cylindrical, has a core diameter of 240 cm and a core height of 100 cm.

On the impact of fuel injection angle in Euler–Lagrange large eddy simulations of swirling spray flames exhibiting thermoacoustic instabilities

This study deals with the fundamental problem of combustion dynamics in gas turbine combustors operating with liquid fuel. In this framework the present work proposes the study of an academic liquid fueled combustor sensitive to thermoacoustic instabilities, simulated via high-fidelity Large Eddy Simulations. The experimental setup addressed is SICCA-spray from EM2C laboratory featuring both stable and unstable flames depending on the combustion chamber length. The proposed analysis, based on the Euler-Lagrange modeling approach, studies the impact of the spray injection angle θ on both the stable flame and the triggering of the longitudinal combustor acoustic mode when using a longer quartz tube. For the liquid injection modeling, the FIM-UR semi-empirical model is adopted with three different θ values: θ=35∘,45° and 60°. In stable conditions, the spray angle is proven to have a negligible impact on the flame anchoring point, however, the mean flame length and fuel distribution are found to be slightly modified by the velocity at which droplets enter the combustion chamber. For the thermoacoustically unstable conditions, two well-established stable limit cycles with the same frequency and similar amplitudes are obtained when fuel is injected at θ=45∘ and 60°. Contrarily, the system stabilizes when θ=35∘ pointing to the importance of the dynamics of the liquid film layer formed inside the injector for this setup. Likewise, this liquid film layer dynamics and its modeling appear critical as already suggested by previous studies on the same configuration. The detailed analysis of the thermoacoustically unstable different predictions is then performed through the investigation of the spatial fields of the local Rayleigh index obtained following the novel extension in the frequency domain of the Rayleigh criterion complemented by the application of Dynamic Mode Decomposition. It confirms that the injection angle of the liquid spray has a significant effect on the thermoacoustic response of the system. Indeed the influence of θ on the dynamics of the liquid fuel when entering the combustion chamber is proven to have an impact on the synchronization mechanism governing the liquid phase with respect to acoustics sustaining the observed limit cycles. More specifically, couplings at the liquid phase level are evidenced by introducing two novel indices correlating the fluctuations of liquid fuel volume fraction and evaporation rate with pressure.

Neutronics analysis for MSR cell with different fuel salt channel geometries

The neutronic properties of molten salt reactors (MSRs) differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity. Based on the Monte-Carlo N particle transport code, the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study. The results show that the infinite multiplication factor (k∞) first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction (FVF) conditions. For the same FVF and average chord length, when the average chord length is relatively small, the k∞ values for different fuel salt channel shapes agree well. When the average chord length is relatively large, the k∞ values for different fuel salt channel shapes differ significantly. In addition, some examples of practical applications of this study are presented, including cell selection for the core and thermal expansion displacement analysis of the cell.

ENRICHMENT ZONING STUDY FOR THE VERSATILE TEST REACTOR

The primary mission of the Versatile Test Reactor (VTR) is to provide peak fast flux in excess of 4.0 x 1015n/cm2-s to support fuel and material testing. To achieve a high fast flux, it is beneficial to maximize the flux peaking in the center of the core. With a single enrichment zone, a highly peaked flux distribution produces a highly peaked power distribution. Coolant inlet orifices can be designed to handle the peaked power distribution but orifice design can be simplified if a more even radial power distribution can be achieved. An approach to reduce the power peaking factor is to use enrichment zoning, which would improve coolant flow homogeneity. Several alternative VTR core configurations are considered with two enrichment zones (15 wt% Pu and 20 wt% Pu). These alternative configurations require more assemblies to maintain reactivity than the reference VTR core, which leads to failure to achieve the design criterion for experimental fast flux with the target core power. Configurations using 20 wt% Pu with different fuel assembly designs having smaller and larger fuel volume fractions are also analyzed. The case having a larger fuel volume fraction reduces the number of fuel assemblies required for criticality, which keeps the experimental flux higher. Configurations with volume fraction zoning can slightly decrease the peaking factor while maintaining the desired fast flux, although some thermal hydraulic limits may not be satisfied. Volume fraction zoning configurations may offer benefits, but determining the feasibility of these configurations requires further thermal hydraulic design and analysis work beyond the scope of the present work.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

Ignition of wildland fuels by idealized firebrands

Experiments were carried out in the Idealized-Firebrand Ignition Test (I-FIT), a bench scale apparatus specifically designed to test the ignition of forest fuel layers from a representative firebrand. A cylindrical heater was used to model the firebrand, which allowed to control the incident radiative heat flux on the specimen, from the critical heat flux up to 25 kW/m2, for five different porosities of the fuel layer. Experimental ignition delay times were interpreted based on a theoretical model of the radiative heating of the fuel layer. Radiative heat transfer within the fuel layer was modeled by using the P1 approximation. In the limit of small ignition delay times an analytical expression was derived to correlate the inverse of the ignition time to the incident heat flux. This analytical expression is used to obtain the ignition temperature and effective properties for the forest fuel layers, namely the product of the fuel volume fraction by solid fuel density and solid heat capacity. Analytical solutions were found to be consistent with experimental data and a correlation relating the inverse of the non-dimensional time-to-ignition to the non-dimensional heat flux is provided.

Design Conceptual of 800MWt Long Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

A long-life pressurized water reactor (PWR) has been reviewed as an innovative reactor design that can fulfill electricity demand. This study aimed to find out the optimum design of 800MWt long life PWR using Thorium-Uranium dioxide (Th-U)O2 fuels with Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. An established computer code of SRAC 2006 with JENDL 4.0 as data nuclear library had been used for the analysis. A two-dimensional R-Z geometry and fuel volume fraction of 40% were used for core geometry analysis. The different fraction of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 had been carried out. The result of this study was a design of PWR 800MWt using Uranium dioxide fuel of 60% with enrichment 11%-12%-13% Uranium-235 and the addition of 0,025% Gd2O3 and 1,0% Pa-231 that could operate for ten years without refueling. The reactor could produce a power density of 45,4 watts/cc with excess reactivity about 3,6% dk/k. This study is expected to be a reference for a long-life pressurized water reactor using the Thorium-Uranium fuel cycles.

Conceptual Analysis of (Th,U-233)O2 Fueled Small Long-Life PWR with Np-237 and Pa-231 as Burnable Poison

Conceptual analysis of (Th,U-233)O2 fueled small long-life PWR with Np-237 and Pa-231 as burnable poison has been performed. The fuel used is (Th,U-233)O2 with Th-232 and U-233 as fertile-fissile nuclides pair with 8% U-233 and addition of Pa-231 and Np-237 as burnable poison (BP). This study aims to investigate optimum design of fuel and core configuration and the performance of reactor. The calculation is done by SRAC2006 code developed by JAEA and is based on JENDL-4.0 library. Cell calculation conducted by PIJ module shows that optimum design is reached at 60% of fuel volume fraction and 6% of burnable poison. Core optimization calculation by CITATION module for two dimensional (R – Z) core geometry shows that reactor could reach 60 years of lifetime at 150 MWt power level which correspond to 25.36 power density. Calculation for heterogen reactor shows that reactor with BP combination of 8% of Pa-231 at center zone and 6% of Pa-231 at outer zone has better performance with 60 years of operation period than reactor with combination of 8% of Np-237 at center zone and 6% of Pa-231 at outer zone that reaches 46 years lifetime.

Neutronic Analysis of Small Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

The requirement for electricity increases with the growth of the human population. The existing power plants have not been able to fulfill all electricity requirements, especially in remote areas. The small long-life pressurized water reactor (PWR) is one of the solutions and innovations in nuclear technology that can produce electrical energy for a long time without refueling. This study aimed to analyze the neutronic of small long-life PWR that using Thorium-Uranium dioxide ((Th-U)O2) fuels with enriched Uranium-235 (U-235) and the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. The SRAC Code with the JENDL-4.0 nuclear data library had been used for the calculation method. In this study, the geometry of the two-dimensional (R-Z) reactor core with different fuel volume fraction was analyzed. Moreover, variations of the Uranium-235, Gadolinium, and Protactinium-231 fractions in the fuels were carried out. The result in this study was a PWR 420 MWt design using 60% Uranium dioxide fuel with enriched Uranium-235 of 10%-11%-12% and the addition of 0,0125% Gadolinium and 1,0% Protactinium-231 as the burnable poisons that could operate for thirteen years without refueling. The small long-life PWR design could produce a power density of 85,1 watts/cc with the reactivity for less than 4,6% dk/k.

Performance of Thorium Uranium Nitride (Th, U233)N Fuel for 500 MWth GFR Long-Life Without Refuelling use FIITB-CHI Code

Performance of Thorium Uranium Nitride Fuel for 500 MWth GFR Long-Life Without Refueling use FI-ITB-CHI code has been done. Gas Cooled Fast Reactor use fast neutron spectrum, and belong to one of fast breeder reactor type. Thorium is three up to four times more abundant than uranium and is widely distributed in nature can be obtained in many countries. Natural thorium does not contain any fissile material and is made up of the fertile Th- 232 isotope only. Thorium fuel and fuel cycles are particularly relevant to countries having large thorium like Indonesia. The calculation use FI-ITB-CHI Code which has been verified by SLAROM code. The core calculation use R-Z geometry dimensional. The multi group diffusion calculation is carried out every year to get neutron flux distribution and power distribution. Then burn up calculation is carried out for each spatial mesh. The reflector radial-axial width is 50 cm. Parametric survey with various percentage of plutonium show the critical condition reach when 11% percentage of Pu and 60% fuel volume fraction. When the fuel volume fraction is increase, the reactor can be operated longer. The optimum value of heterogeneous core configuration when use 8% percentage of plutonium for F1, 10% for F2 and 14% for F3. The reactor can be operated almost 25 years without refueling.