Critical Boron Concentration Research Articles

Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer

Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.

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.

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

This paper describes the effect of cross-section data generated by several codes on calculated neutronic parameters. The Pressurized Water Reactor Mixed Oxide and Uranium Oxide (PWR MOX/UO2) Core Transient Benchmark case was chosen because it has been used widely to validate neutronic codes. The cross-section data in this study will be generated by SRAC, Serpent, and HELIOS codes. The NODAL3 code will be used to calculate neutronic parameters from each cross-section. The neutronic parameters calculated by NODAL3 are the effective multiplication factor (keff), control rod worth, critical boron concentration, and power distribution under Hot Zero Power (HZP) conditions. The Power-Weighted Error (PWE) and Error-Weighted Error (EWE), as a measure of the relative error in fuel assembly power, are less than 5 %, indicating that the calculation is consistent with DeCART as a reference. The difference in calculated radial power peaking factor for all three cross-sections to reference data reaches 6.284 % (G-3), 8.438 % (G-3), and 10.998 % (C-7), respectively, for SRAC, Serpent, and HELIOS. The axial power distribution calculated by NODAL3 at the top and bottom of the reactor core has a relative error that peaked at 16.60 %, 13.86 %, and 10.20 %, respectively, for cross-sections provided by SRAC, Serpent, and HELIOS. Further improvements are needed for NODAL3 by applying various discontinuity factors to improve its performance.

Effect of two way thermal hydraulic-fuel performance coupling on multicycle depletion

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (Tbc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting Tbc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, Tbc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.

Steady-state analysis of ROSTOV-2 benchmark problem using SERPENT-2 and PARCS codes

Recently, a new benchmark problem has been introduced by the OECD/NEA on the Rostov-2 reactor to address the verification and validation of novel High-Fidelity Multi-Physics tools. In line with the first phase of this benchmark, the current research was dedicated to the steady state reactor physics analysis, using “traditional multi-physics” codes for further and future comparisons. A set of homogenized two-group cross-section libraries were developed using SERPENT-2 Monte Carlo code. Some sophisticated techniques have been employed for detailed consideration of radial as well as axial reflectors, spacer grids and assembly discontinuity factor evaluation. Since PARCS code requires the group constants in PMAXS format, the generated libraries were re-formatted using the GenPMAXS interface. The PMAXS library verification, as well as the steady state analysis have been conducted in three operational states of the reactor using PARCS nodal diffusion code, independent SERPENT-2 Monte Carlo simulation and some available reference data. A set of criticality calculations have been conducted to evaluate the total reactivity worth of different control rod groups, the group #10 integral reactivity curve, fuel temperature coefficient and the critical boron concentration in Hot Zero Power (HZP) and Hot Full Power (HFP) states. The maximum reactivity difference between PARCS and SERPENT-2 simulations was less than 100 pcm in all criticality calculations. In HFP state, the relative power distribution in radial and axial directions were calculated and compared. The maximum Power Peaking Factors (PPFs) and the location of the “hot assembly” from PARCS and SERPENT-2 are in good agreement with each other. Finally, the burn-up calculation was conducted with PARCS code to evaluate neutronic parameters at some specific burn-up values. The calculated reactor physics parameters including the critical boron concentration, the power peaking factors, the power and burn-up distributions at the Beginning of Experiment (BOE) were consistent with available reference data.

Calculation of boron concentration dependent reactivity versus moderator density curve with application on ATWS analysis for Maanshan PWR plant

A method used for determining the moderator density coefficient (MDC) based on the effective multiplication factor (keff) calculated from the SIMULATE-3 code is developed in this study. The calculation is performed at various water density and boron conditions for Maanshan unit 1 cycle 27. The calculated MDC data are fitted as a second-order polynomial function of density, modified with a factor of linear function of boron concentration. The fitted equation is then integrated and normalized to zero at rated density to represent the reactivity as a function of moderator density and boron concentration. A reactivity vs. density table under different boron concentrations can then be generated based on this function. The table generated is used as an input for the ATWS analysis to be performed using the RETRAN-02 thermal–hydraulic analysis code. As an illustration, a table based on a boron concentration of 1500 ppm is generated and used as an input for the analysis of the potential limiting ATWS case, i.e., Loss of Normal Feedwater Flow Event. The ASME emergency level limit of reactor coolant system pressure less than 220.63 bar is used as the acceptance criterion for the current ATWS analysis. It is known that the critical boron concentration at hot full power for each operating fuel cycle varies with the burnup through the cycle. Hence, the calculated ATWS results at different boron concentrations are equivalent to that calculated at different cycle burnup conditions. The cycle burnup point after which the calculated ATWS maximum is within the ASME limit can be determined based on the sensitivity analyses considering different boron concentrations.

LEU+ loaded APR1400 using accident tolerant fuel cladding for 24-month two-batch fuel management scheme

In thiswork, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4–5 w/o UO2 fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er2O3) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd2O3) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO2 in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

Validation of DRAGON5 and DONJON5 computer codes with CNPGS Unit-3 cycle-1 design calculations

The necessity to replace outdated analogue nuclear system calculations as well as the desire to enhance and guarantee adequate levels of plant availability and safety have led to an increase in the usage of software based design and simulations of nuclear power plants in recent years. In this study, the computer codes DRAGON5 and DONJON5 are validated using design information for 300MWe Pressurized Water Reactor (PWR). The codes are used to perform design analysis of Unit-3 Cycle-1 of the Chashma Nuclear Power Generation Station (CNPGS). The analysis includes calculations for criticality, radial power distribution, reactivity coefficients, and control rod worth under various core conditions. The results are validated against the measured data. The difference in the critical boron concentration (CBC) for Hot Zero Power (HZP) and Hot Full Power (HFP) at All Rod Out (ARO) conditions are 16 pcm and 20 pcm respectively. It has been observed that, the core radial power distribution is in good agreement with the design data with maximum relative errors of 6.2% and 1.5% at HZP and HFP conditions respectively. The validation confirms that, the codes can be employed with confidence for the design calculations of 300MWe PWRs (CNPGS). Furthermore, the 1-D steady state thermal hydraulic analysis is also performed to predict the temperatures of the fuel and coolant at both the average and hot channels at the beginning of life (BOL) and the end of life (EOL). It is found that the peak to average values of fuel centerline temperature at BOL for average and hot channels are 1.084 and 1.17 respectively, while at EOL for average and hot channels are 1.059 and 1.069 respectively, which are in good agreement with the reference peak to average value of 1.6875.

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

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

Establishment of DeCART/MIG stochastic sampling code system and Application to UAM and BEAVRS benchmarks

In this study, a DeCART/MIG uncertainty quantification (UQ) analysis code system with a multi-correlated cross section stochastic sampling (S.S.) module was established and verified through the UAM (Uncertainty Analysis in Modeling) and the BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) benchmark calculations. For the S.S. calculations, a sample of 500 DeCART multi-group cross section sets for two major actinides, i.e., 235U and 238U, were generated by the MIG code and covariance data from the ENDF/B-VII.1 evaluated nuclear data library. In the three pin problems (i.e. TMI-1, PB2, and Koz-6) from the UAM benchmark, the uncertainties in kinf by the DeCART/MIG S.S. calculations agreed very well with the sensitivity and uncertainty (S/U) perturbation results by DeCART/MUSAD and the S/U direct subtraction (S/U-DS) results by the DeCART/MIG. From these results, it was concluded that the multi-group cross section sampling module of the MIG code works correctly and accurately. In the BEAVRS whole benchmark problems, the uncertainties in the control rod bank worth, isothermal temperature coefficient, power distribution, and critical boron concentration due to cross section uncertainties were calculated by the DeCART/MIG code system. Overall, the uncertainties in these design parameters were less than the general design review criteria of a typical pressurized water reactor start-up case. This newly-developed DeCART/MIG UQ analysis code system by the S.S. method can be widely utilized as uncertainty analysis and margin estimation tools for developing and designing new advanced nuclear reactors.

High-Performance and High-Fidelity GPU-Based Monte Carlo Solutions to the BEAVRS Benchmark

The BEAVRS benchmark is solved by PRAGMA, the graphics processing unit (GPU)–based continuous-energy Monte Carlo code. The solutions consist of the detailed simulation results for the two cycles that involve the reactivity and pin power distribution information for the zero-power physics tests and depletion. Primary results at hot zero power, such as the critical boron concentration at various rodded conditions, control rod bank worth, isothermal temperature coefficients, and assemblywise detector signal, are compared with the measured data. Core-follow calculations are performed with varied power, and the resulting boron letdown curves are compared with the measured one. Hot full-power depletion is also performed and the resulting pinwise power distributions of cycle 1 are compared with the nTRACER results. The comparison with the measured data and also with the nTRACER results demonstrates the high solution fidelity of PRAGMA. In all the calculations, PRAGMA uses a tremendously large number of histories, ranging from up to hundreds of millions per cycle, that are used to fully exploit the massive parallel computing capacity of GPUs. The execution time of the entire core-follow calculation with about 30 burnup steps takes less than 16 h on a single rack of computing nodes mounted with 24 gaming GPUs, which represents considerably high Monte Carlo core calculation performance.

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

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

McCARD Criticality Benchmark Analyses with Various Evaluated Nuclear Data Libraries

International Criticality Safety Benchmark Evaluation Project (ICSBEP) criticality analyses were conducted using the McCARD Monte Carlo code for 85 selected benchmark problems with 7 evaluated nuclear data libraries (ENDLs): ENDF/B-VII.1, ENDF/B-VIII.0, JENDL-4.0, JENDL-5.0, JEFF-3.3, TENDL-2021, and CENDL-3.2. Regarding the analyses, it was confirmed that the keff results are sensitive to the ENDL. It is noted that the new-version ENDLs show better performance in the fast benchmark cases, while on the other hand, there are no significant differences in keff among the different ENDLs in the thermal benchmark cases. The sensitivity of the keff results depending on the ENDL may impact nuclear core design parameters such as the shutdown margin, critical boron concentration, and power defects. This study and keff results will be a good reference in the development of new types of nuclear cores or new design codes.

Influence of Spacer Grids Homogenization on Core Reactivity and Axial Power Distribution

The paper presents the influence of spacer grid homogenization during cross section generation on core reactivity and axial power distribution. Homogenization calculation was performed at fuel assembly level using FA2D code. The first approach is to smear uniformly all centrally located spacer grids along 120 inches of fuel assembly and carry out 2D transport calculation. The second approach is to smear spacer grid within 6 inches of fuel assembly and perform homogenization calculation. That composition is then assigned to closest 6 in axial subdivision of the core calculation. The last analysed option is to do additional localization of spacer grids and carry out homogenization within 2 inches of fuel assembly height. The additional subdivision is afterward performed of the closest regular axial core subdivision in nodal core calculation. The core calculation was performed using modified PARCS 2.5 code for NPP Krško cycle 29. The normalized axial power distributions obtained by PARCS for three different ways of spacer grid homogenization are then compared to quantify the influence of modelling. Similar comparison was performed for critical boron concentration. As expected larger influence is present for axial power distribution (more details for fine localization), with some influence on axial power offset and global reactivity.

Polynomial interpolation cross-section parameterization method with the RMC Monte Carlo code

Due to the generality and flexibility of Monte Carlo methods in geometric modeling and the challenge for traditional deterministic transport methods in dealing with stochastic media, it is necessary to develop a cross-sectional library generation function using a Monte Carlo method as an interface for full core calculations. In this study, we proposed a cross-sectional parameterization method based on the RMC code for the generation of a few-group cross-sectional library. To perform realistic core calculations, a few-group neutron cross-sectional library of functions of burnup and thermal-hydraulic parameters was prepared in advance. The results of nodal diffusion code CORCA3D full core calculations with RMC B1 corrected cross-sectional sets agree well with those of RMC continuous-energy calculations. Meanwhile, with the consideration of T-H feedback, the results of the CORCA3D full core calculations show good agreement with RMC full core calculations, with a maximum difference in the critical boron concentration of 17 ppm.

Control of crystallization and magnetic properties of CoFeB by boron concentration

Controlling the crystallinity of CoFeB is the most essential issue for designing various spintronics devices. Here we show the microstructure and magnetic properties of MgO/CoFeB/MgO structures for various boron concentration. We present the effect of boron on the crystallinity of CoFeB into two categories: the critical boron concentration (5 ~ 6%) at which CoFeB crystallizes and the effect of remaining boron (0 ~ 5%) in the crystallized CoFeB. And the trends of the saturation magnetization, exchange stiffness, exchange length, domain wall energy and Gilbert damping constant according to the boron concentration are provided. Abrupt variation of properties near the critical boron concentration (5 ~ 6%) and a noticeable change in the crystallized CoFeB (0 ~ 5%) are confirmed, revealing a clear causal relationship with the structural analysis. These results propose that the crystallization, microstructure, and major magnetic properties of CoFeB are governed by the amount of boron, and emphasize the need for delicate control of boron concentration.

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the nuclear power plant (NPP) Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 libraries are considered. In the first part of this work the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few 100 pcm in multiplication factor for a fresh fuel due to differences in 235U and 238U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout, and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1%. Some differences in isothermal temperature coefficient were observed, however, they only marginally affect core power defect going from zero to full power.

Development of the Unified Bayesian Inference method for improving the pressurized water reactor simulation

Nuclear data is an important input in nuclear power plant core simulation. With the development of advanced simulation methods and the gradual improvement of computing power, the measurement uncertainty of nuclear data becomes the main source of the uncertainty of core physical response. In this paper, firstly, the sensitivity and uncertainty analysis of the nuclear data for the first cycle core loading of Unit 1 of the Hainan nuclear power plant are performed by using the Reactor Monte Carlo (RMC) code. The results show that the uncertainty of the eigenvalue keff introduced by the nuclear data is 519 pcm, and the uncertainty of critical boron concentration is 57 ppm. Secondly, the UBI method based on Bayesian inference is proposed in the actual simulation of nuclear power plants. By combining the measurement results of core critical boron concentration with the calculation results of the NESTOR software package, the multi-group nuclear data library used by the NESTOR software package is adjusted, and the posterior estimation of the nuclear data and calculated values is obtained. The results show that the accuracy and precision of the first cycle simulation of Unit 1&2 are greatly improved.

Validation of PWR Neutronics Code Package TORCH V2.0 With Nuclear Power Plant Measurements

This article presents the verification and validation (V and V) of PWR neutronics code package TORCH V2.0 with nuclear power plant (NPP) measurements. The advanced nuclear power engineering design software, TORCH V2.0, was developed by the Nuclear Power Institute of China (NPIC), China National Nuclear Corporation (CNNC). Based on the two-step calculation scheme, TORCH V2.0 mainly contains lattice physics code for assembly homogenization, link calculation code for few-group constant parameterization, and core simulation code for few-group core calculation. The calculation modules of each code were already verified against various benchmark problems, whereas this article focuses on the V and V of linked code system. The measured values of the reactor startup physics test and NPP operation from six PWR NPPs (Daya Bay NPP, Ling Ao NPP, Fangjiashan NPP, Qinshan NPP, Hainan Changjiang NPP, and Fuqing NPP) were utilized to perform the comparison and analysis of V and V. Compared parameters of the reactor startup physics test include critical boron concentration, control rod integral value, boron differential value, and isothermal temperature coefficient. Compared parameters of the NPP operation contain critical boron concentration, assembly-wise power distribution, hot spot factor, and nuclear enthalpy rise factor. The results show that the software TORCH V2.0 has reliable calculation ability and can be applied in the PWR nuclear power engineering design which is based on square fuel assembly.

DONJON5/CLASS coupled simulations of MOX/UO2 heterogeneous PWR core

Most fuel cycle simulation tools are based either on fixed recipes or assembly calculations for reactor modeling. Due to the high number of calculations and extensive computational power requirements, full-core computations are often seen as not viable for this purpose. However, this leads to additional hypotheses and modeling biases, thus limiting the realism of the resulting fuel cycle. For several applications, the current modeling method is sufficient, but precise calculations of discharged fuel composition may require further refinements. CLASS (Core Library for Advanced Simulation Scenarios) is a dynamic fuel cycle simulation code developed since 2012 with reactor models based on neural networks to produce nuclear data and physical quantities. Past work has shown a first coupling between CLASS and DONJON5 to quantify neural networks approach biases. This work assesses the applicability of 3D full-core diffusion calculations using the DONJON5 code coupled with nuclear scenario simulations involving a realistic PWR core at equilibrium cycle conditions. DONJON5 interpolates burnup dependent diffusion coefficients and cross sections generated beforehand by DRAGON5, a deterministic lattice calculation tool. Whereas previous studies considered only homogeneous reactors (i.e. homogeneous assembly in terms of composition and enrichment as well as homogeneous core), the present contribution focuses on the integration of full-core calculations in CLASS for fuel cycles involving a MOX/UO2 PWR core (i.e. 1/3 MOx–2/3 UOx). The DONJON5 model considered in this work describes a core with critical boron concentration at each time step partially loaded with MOx heterogeneous assemblies composed of three enrichments. In fuel cycle calculations, the main issue is to adapt, in the fabrication stage, the fresh fuel composition for the reactor with regards to the isotopic composition of the available stocks. This work presents a fuel loading model based on power peaking factors minimization that respects irradiation cycle length, 235U enrichment as well as Pu concentration and fissile quality, hence, ensuring a more uniform power distribution in the core.