SRAC Code Research Articles

Validation of OpenMC Code for Low-cycle and Low-particle Simulations in the Neutronic Calculation

Validation of Low-Cycle and Low-Particle OpenMC Simulation Codes for Neutronics Calculations has been conducted. This study validates OpenMC, an evolving open-source neutron analysis code. Validation of Low-Cycle and Low-Particle Codes is crucial as it allows for effective calculations with minimal computational resources. Determining the convergence point of cycles and minimum particles in low-cycle and low-particle calculations enables maintaining calculation accuracy, thus providing sufficiently accurate results. This study demonstrates that a minimum of 15,000 particles, 100 cycles (30 inactive, 70 active), is required for low-cycle simulations. A comparison of k-eff calculation results with the SRAC code for MSR FUJI-12 at 7 burnup points (0-27 MWd/ton) yields a maximum error of 0.7%. These results validate the effectiveness of OpenMC in achieving accurate neutronic calculations with limited computational resources

Neutronics Analysis of SMART Reactor Fuel Cell and Fuel Assemblies with Two Absorber Materials: Gd2O3 – UO2 and Al2O3 – B4C Using SRAC Code and JENDL 3.3 and 4.0 Data Libraries

The SMART is a small modular nuclear reactor which has been developed by South Korea’s KAERI, utilizes 4.95% enriched uranium fuel and two types of absorber materials: Gd2O3 - UO2 and Al2O3 - B4C. This research focuses on the neutronic analysis of SMART at the fuel cell and fuel assembly levels using SRAC 2006 with JENDL 3.3 and 4.0 Nuclear Data Libraries. The study aims to compare the neutronic performance of the SMART fuel assembly using these two absorber materials. At the fuel level, three types of enrichment are considered: 2.82%, 3.25%, and 4.95%. Meanwhile, at the fuel assembly level, three types (A, B, and C) are investigated, based on the composition of fuel rods and absorber material rods. The results indicate that the SMART reactor with the KAERI design (enrichment of 4.95% and 3 types of fuel assemblies) can maintain critical conditions for 990 days (equivalent to 3 years) of operation for small modular reactor types. The two types of absorber materials are effective in absorbing excess positive reactivity from the fuel effectively. The use of absorbing materials greatly determines the neutronic characteristics and costs in the construction of nuclear reactors, It can be compared that using 2 absorbing materials can reduce the value of the effective multiplication factor and the infinitive neutron production in the core compared to using 1 absorbing material.

Validation of OpenMC Code Criticality Value Calculation for GFR Reactor with UN-PuN Fuel

The use of OpenMC code needs to be validated with other code, so that an accurate and valid criticality value is generated. The validation value is affected by the minimum number of particles used. Determination of the minimum number of particles is carried out by varying the number of particles so that they have a convergent value, which is in the range of 100-50000 particles. The results of determining the minimum number of particles are then used to find the minimum number of cycles consisting of active and inactive batches. This study uses the criticality value parameter in the form of an effective multiplication factor (k-eff) as a benchmark for code accuracy. The k-eff values generated by OpenMC and SRAC are then compared and the validation error value is searched. The error value is found by calculating the k-eff (Δk-eff) difference between the two codes. OpenMC code can be said to be validated if it has an error value of less than 1% against the calculation results of the SRAC code. The calculation results of determining the minimum number of particles show k-eff values in the range of 35000 - 50000 minimal changes significantly or not fluctuating. The calculation of k-eff and entropy in cycle 500 shows a convergent value at the active batch value of >30. Based on the data cycle, researchers used 100 active batches and 30 inactive batches to perform validation calculations. The results of the validation calculation using UN-PuN fuel with a plutonium material composition of 10% showed a maximum k-eff error value of 0.819%. The Δk-eff obtained is 0.008.

Core design optimization of a 200 MWt pressurized water SMR using evolutionary simulated annealing

The paper presents the design optimization of a pressurized small modular reactor (SMR) core using evolutionary simulated annealing (ESA). The core was designed for the power output of 200 MWt based on the typical 17 × 17 fuel assemblies of a traditional PWR. The core consists of 37 fuel assemblies of F235, F340 and F445 types corresponding to the 235U enrichment levels of 2.35, 3.40 and 4.45 wt%, respectively. The design target is to attain the fuel cycle length of about 900 effective full power days (EFPDs) with a minimum core average 235U enrichment, while flattening the radial power distribution. Lattice and core physics calculations were performed using the SRAC code system. The COREBN module of the SRAC code system was coupled with the ESA method for performing the core burnup calculations during the optimization process. A fitness function was chosen to combine the objectives of the cycle length, the core average enrichment and the radial power peaking factor (PPF). A survey was conducted for determining the optimal control parameters of the ESA method and the weighting factors in the fitness function associated to the design objectives. The optimal core was selected from the ESA search consisting of 21 assemblies of F235, 8 of F340 and 8 of F445 type, respectively. The core average 235U enrichment is 3.031 wt% with the cycle length of about 840 EFPDs. The PPF of the selected core is 1.428 which appears at the centered assembly.

Effect of Protactinium-231 (Pa-231) Addition on Effective Multiplication Factor Value of ThN-U233N Fueled PWR Reactor

This study aims to determine the optimum percentage of addition of protactinium-231 in a ThN-U233N fueled PWR reactor.In ThN-U233N, uranium-233 is used as the fissile material. Pa-231 as burnable poisson is added to ThN-U233N fuel to reduce the keff value. In this research conducted benchmarking keff values ​​of OpenMC and SRAC codes, homogeneous and heterogeneous core configurations, and the addition of Pa-231 in ThN-U233N fuel. Neutronic analysis was performed using the OpenMC code. In the homogeneous core configuration, the enrichment percentage of U-233 is varied by 1%-15%. Meanwhile, the heterogeneous core configuration uses 3 variations of ring geometry and fuel. The results of the benchmarking of the OpenMC and SRAC code keff values ​​show a maximum error of 1,906%. The homogeneous core configuration produces the optimum value for U-233 enrichment of 3%. Meanwhile, the optimum value of the heterogeneous core configuration is achieved in ring geometry F1 = F2 = F3 = 3 rings with enrichment of U-233 at F1 = 0,5%, F2 = 3%, F3 = 5,5% and addition of Pa-231 of 1.5%. The optimum calculation results has an excess reactivity value 8.37%.

Optimization of 200 MWt HTGR with ThUN-based fuel and zirconium carbide TRISO layer

Abstract TRISO fuel particle using ZrC has better strength and resistance to high temperatures than SiC. Previous studies show that the ZrC layer, as a substitution of SiC within the TRISO layer of coated fuel particles, has an insignificant difference in the performance of the neutronic aspect. Further neutronic studies are required to obtain the best combination of thorium-based fuel with ZrC coating for HTGR. This study analyzed the neutronic performance of three types of thorium-based fuels, oxide, carbide, and nitride, for HTGR. The reactor design refers to the High-Temperature Test Reactor with some axial and radial fuel configuration adjustments. This reactor is designed to operate at 200 MWt and has been modified to use a ZrC layer as a substitute for the SiC layer on the coated fuel particles. The neutronic study is carried out using SRAC2006 code with JENDL 4.0 nuclear data library. Neutronic parameters analyzed include multiplication factor, power peaking factor, and neutron spectrum. Neutronic analysis results show that thorium nitride fuel’s multiplication factor (k eff) is better than other compared fuel types with k-eff 1.050, higher than thorium carbide, 1.004. At the same time, thorium oxide has been sub-critical. The power-peaking value of all materials is close to the ideal peaking value that is one. Other neutronic aspects, such as the neutron spectrum for three compared fuel types, have a similar trend.

Perhitungan Desain Konsep Reaktor Cepat Berpendingin Karbondioksida Superkritis dan Berbahan Bakar Uranium Metalik Alam

This study discusses the design concept of the fast reactor core with supercritical carbon dioxide as coolant and natural metallic uranium as fuel that has been burnup for one hundred years. Calculations and simulations are performed using the SRAC code. The PIJ-SRAC module is for calculating the fuel cell level, and the CITATION-SRAC module is for calculating the reactor core level. Radial modified CANDLE is applied as a fuel arrangement strategy in the reactor core. The survey parameters were calculated in the form of burnup level, infinite multiplication factor (kinf), the density of fissile and fertile atoms, conversion ratio, and ef ective multiplication factor (kef ). The calculated reactor core thermal power varies from 400 MWt, 450 MWt, and 500 MWt. Based on calculation and simulation results, a reactor with thermal power of 450 MWt with geometry RZ (R = 150 cm and Z = 150 cm) is the best design. The calculation results show that the kef value from the beginning of the combustion cycle is around 1.002662 with a reactivity value of 0.2654%; until the end of the cycle, the kef value is 1.041057 with a reactivity value of 3.9437%.

Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of UN fuel reached when used heterogeneous core configuration case four with percentage of U-235 in F1=5.5 %, F2=7 % and F3=8.5 % also with the addition of Np-237 0.2 % and fuel fraction 56 %. It has a maximum excess reactivity value 12.56 % %∆k/k. And then, the optimum configuration of ThN fuel reached when used heterogeneous core configuration case three with percentage of U-233 in F1=2 %, F2=4 % and F3=6 % with the addition of Pa-231 0.5 % and fuel fraction 53 %. It has a maximum excess reactivity value 7.67 % %∆k/k. The comparison of optimum design of UN and ThN fuel shows that the ThN fuel has the k-eff value closer to critical than UN fuel. Therefore, in this study, ThN fuel is more suitable for use in PWR reactors because it has a small excess value and can operate for 10 years without refueling

Design study of 208Pb-Bi eutectic-cooled reactor with natural uranium as fuel cycle input with radial fuel shuffling

In the CANDLE burnup scheme, the burnup moves in the axial direction continuously. Meanwhile, the Modified CANDLE (MCANDLE) burnup scheme divides the core into several equal volume parts in the axial or radial direction. It can increase the utilization of natural uranium by using natural uranium as fuel cycle input. The use of 208Pb-Bi eutectic as a coolant can increase the criticality of the reactor. The core was divided into six discrete regions in radial directions. The refueling scheme was optimized to obtain an excellent reactor design. SRAC code system was used for neutronic calculations and JENDL 4.0 was used for the nuclear data library. A scheme that located natural uranium in the central region and nearby to the highest burnup level fuel has an excess reactivity lower than 0.5% Δk/k and PPF lower than two during operation time. The scheme also has a burnup of natural uranium spent fuel ∼ 48% FIMA.

Comparison of neutronic aspects in high‐temperature gas‐cooled reactor using ZrC and SiC Triso particle with 50 and 100 MWt power

The use of zirconium carbide (ZrC) as a substitute for silicon carbide (SiC) has been proposed to improve the performance of coated particles for high-temperature gas-cooled reactor (HTGR) fuel. Irradiation test on ZrC has proved that it has excellent characteristics for HTGR fuel compared with SiC, such as fission product retention capabilities and better resistance on fission product corrosion. Neutronic analysis on 30 MWt HTGR using ZrC and SiC has been performed in many previous studies. The research presented in this paper was conducted as a further continuation of the aforementioned studies. The present study was directed to analyze tristructural isotropic (TRISO)-coated particles with ZrC as a substitute for the SiC layer on an HTGR with different power rates. Here, we used high-temperature test reactor design, an experimental scale prismatic HTGR built and operated in Japan. The power was varied at the range of 50-100 MWt, while the reactor geometry was modified by applying an additional fuel layer in the axial and radial directions. Neutronic analysis of the reactor was investigated by calculating the k-eff, k-inf, power peaking factor, burn-up level, neutron spectrum, and Doppler effect using ZrC and SiC layers for TRISO-coated fuel particles. The result shows the coated fuel particle using the ZrC layer has the same performance as SiC with an insignificant difference in the values on neutronic aspects calculation. Neutronic calculations are performed using the SRAC code with the Japanese Evaluated Nuclear Data Library 4.0 nuclear data library.

Possibility of curium as a fuel for VVER-1200 reactor

In this research, curium oxide (CmO2) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO2 is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO2 and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO2 is competent like UO2 as a fuel due to higher fission cross-section of 243Cm and 245Cm isotopes and neutron capture cross-section of 244Cm and 246Cm isotopes. As a result, when some or all of the UO2 of fuel rods or IBA rods are replaced by CmO2, we get a similar k-inf like the reference even with lower enrichment UO2 fuels. These studies show that the use of CmO2 as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO2 with inert material ZrO2 in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

Study of neutronic performance analysis of Sodium-Cooled Fast Reactor (SFR) design based on the variations of output power using radial fuel suffling strategy has been performed. SFR is a type of the generation IV reactor that is widely researched for application on a commercial scale. The reactor design uses natural uranium as a fuel and Sodium as a coolant. The research has been carried out by using the SRAC code and JENDL-3.2 as a library with two dimensional R- Z suffling strategy of cylinder core for variations of power output 300, 350, 400, 450 and 500 MWTh. The neutronic parameters such as a multiplication factor (k-eff and k-inf) and burn-up analysis are observed. The SFR core is separated into 10 regions having the same volume fraction in the radial direction. At beginning of burn-up process, the reactor core is only filled with natural uranium fuel called fresh fuel and prepared for the one cycle of 10 years of burn-up time. The burn-up result in the first region is shifted into the second region, the burn-up result in the second region is shifted into the third region, and so on until burn-up result into the tenth region. The burn-up result in the tenth region is removed from the reactor core, then the first region can be filled with fresh natural uranium fuel and so on up to 10 times fuel cycles of reactor operation. The neutron calculation results indicate that the multiplication factors (k-eff and k-inf) in a critical condition are occurred on the 300 MWTh of output power. Overall, the output power of 300 MWTh has requirements and a greater chance of being operated for SFR designed.

Neutronic analysis of VVER-1000 MOX fuel assembly with burnable absorber Gadolinia and Erbia

This paper represents the neutronic characteristics and feasibility of using Gadolinia and Erbia in integral burnable absorber (IBA) rods for the proposed MOX fuel assembly of VVER-1000 reactor by Nuclear Energy Agency (NEA). For comparison the 4.0% Gd2O3 IBA is considered as a reference. The assembly was burned to 40 GWd/t, and by neutronic calculation, the optimum percentage of Gd2O3 and Er2O3 was determined using SRAC-2006 code and JEFF-3.1 data library. The burnup dependent reactivity of a MOX assembly containing 12 UO2-Gd2O3-Er2O3 (U-Gd-Er) IBA rods with 0.9% Gd2O3, and 5.0% Er2O3 was found almost identical to the reference assembly with 4.0% Gd2O3 in IBA rods. However, the assembly with U-Gd-Er IBA rods has a flatter pin-wise power distribution and lower power peaking factor (PPF). Long-lived fission products, poisons, and minor actinides are also investigated for choosing Gd2O3 and Er2O3 as IBA over only Gd2O3 for the VVER-1000 MOX fuel assembly.

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.

Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

Use of americium as a burnable absorber for VVER-1200 reactor

The objective of this research is to the use of americium (AmO2) as a burnable absorber effectively instead of conventional gadolinium (Gd2O3) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO2 as a burnable absorber. From these studies, it is found that AmO2 can decrease the excess reactivity like Gd2O3 without changing the criticality life span and enrichment of 235U. A homogeneous mixture of the 0.20% AmO2+ 4.95% enriched UO2 fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO2+UO2 in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.

Study on Fissile Material Usage for 50 MWt High Temperature Gas Reactor using (Pu-U)O2 Fuel

One of the most potential nuclear power plants is High Temperature Gas-cooled Reactor (HTGR). It is a reactor system designed using TRISO (Tristructural-Isotropic) coated fuel and has very high outlet temperature to produce electricity and many co-generation application processes. The present study aims to analyze fissile material usage for 50 MWt prismatic-type HTGR using (Pu-U)O2 fuel. Calculations were performed using neutron diffusion method that applied in SRAC2006 code, and nuclear data based on JENDL 4.0. Selected neutronic parameters are analyzed to study the effect of the number of fissile enrichment and number of Coated Fuel Particle (CFP) to the reactor system. The results of the parametric survey show that for prismatic-type of 50 MWt HTGR, 10% of enrichment and 8950 of number CFP give the best performance in term of fissile material usage during reactor operation.

Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006