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1
Syarifah, Ratna Dewi, and Alvi Nur Sabrina. "Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel." Computational And Experimental Research In Materials And Renewable Energy 2, no. 2 (2019): 64. http://dx.doi.org/10.19184/cerimre.v2i2.27368.
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A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and Americium in the world and also to reduce the effective multiplication factor (k-eff) value from the reactor. The results obtained that the addition of Neptunium and Americium causes the k-eff value was decreased at the beginning of burn-up time, but increase at the end of burn-up time. It was because Neptunium and Americium absorb neutrons at the beginning of burn-up time and turns into fissile material at the end of burn-up time. The addition of protactinium in the reactor causes the k-eff value to be decreased both at the beginning of the burn-up time and at the end of the burn-up time. It happens because Protactinium absorbs neutrons both at the beginning of the burn-up time and at the end of the burn-up time. Therefore protactinium is often called a burnable poison.
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Mabruri, Ahmad Muzaki, Ratna Dewi Syarifah, Indarta Kuncoro Aji, Artoto Arkundato, and Nuri Trianti. "Validation of OpenMC Code for Low-cycle and Low-particle Simulations in the Neutronic Calculation." JURNAL ILMU FISIKA | UNIVERSITAS ANDALAS 16, no. 2 (2024): 107–17. http://dx.doi.org/10.25077/jif.16.2.107-117.2024.
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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
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Surbakti, Tukiran, and Tagor Malem Sembiring. "NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 18, no. 1 (2016): 29. http://dx.doi.org/10.17146/tdm.2016.18.1.2329.
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Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing right now. The fuel of research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view. The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel. Key words: mini fuel, neutronics analysis, reactor core, safety analysis Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi. Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini akan dibahas analisis keselamatan uji bahan bakar mini U7Mo-Al dan U6Zr-Al ditinjau dari segi neutronik. Perhitungan dilakukan dengan menggunakan program komputer WIMSD-5B dan Batan-3DIFF. Hasil analisis menunjukkan bahwa kedua bahan bakar uji dapat diiradiasi dengan derajat bakar < 70 % selama 12 siklus operasi tanpa melampaui batas keselamatan neutronik. Kerapatan panas bahan bakar uji U7Mo-Al lebih besar dari bahan bakar U6Zr-Al. Kata kunci: Bahan bakar mini, analisis neutronik, teras reaktor, analisis keselamatan
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Blanco, J. A., P. Rubiolo, and E. Dumonteil. "NEUTRONIC MODELING STRATEGIES FOR A LIQUID FUEL TRANSIENT CALCULATION." EPJ Web of Conferences 247 (2021): 06013. http://dx.doi.org/10.1051/epjconf/202124706013.
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Framework • A detailed and highly flexible numerical tool to study criticality accidents has been developed • The tool implements a Multi-Physics coupling using neutronics, thermal-hydraulics and thermal-mechanics models based on Open FOAM and SERPENT codes • Two neutronics models: Quasi-Static Monte Carlo and SPN Objective: In this work a system composed by a 2D square liquid fuel cavity filled with a fuel molten salt has been used to: • Investigate the performance of the tool’s thermal-hydraulics and neutronics solvers coupling numerical scheme • Evaluate possible strategies for the implementation of the Quasi-Static (QS) method with the Monte Carlo (MC) neutronics code • Compare the QS-MC approach precision and computational cost against the Simplified P3 (SP3) method
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Bereznev, Valerij Pavlovich, Evgenij Fyodorovich Seleznyov, and David Serezhaevich Asatryan. "The «CORNER» neutronics calculation code." Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 2015, no. 1 (2015): 136–43. http://dx.doi.org/10.26583/npe.2015.1.15.
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Hossain, Md Imtiaj, Abdus Sattar Mollah, Yasmin Akter, and Mehraz Zaman Fardin. "Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code." Nuclear Energy and Technology 9, no. 4 (2023): 215–25. http://dx.doi.org/10.3897/nucet.9.91090.
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The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The keff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.
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Hossain, Md Imtiaj, Abdus Sattar Mollah, Yasmin Akter, and Mehraz Zaman Fardin. "Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code." Nuclear Energy and Technology 9, no. (4) (2023): 215–25. https://doi.org/10.3897/nucet.9.91090.
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The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The k<sub>eff</sub> values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.
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Nguyen, T. S., and G. B. Wilkin. "Monte Carlo Calculations Applied to NRU Reactor and Radiation Physics Analyses." AECL Nuclear Review 1, no. 2 (2012): 47–50. http://dx.doi.org/10.12943/anr.2012.00018.
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The statistical MCNP (Monte Carlo N-Particle) code has been satisfactorily used for reactor and radiation physics calculations to support NRU operation and analysis. MCNP enables 3D modeling of the reactor and its components in great detail, the transport calculation of photons (in addition to neutrons), and the capability to model all locations in space, which are beyond the capabilities of the deterministic neutronics methods used for NRU. While the simple single-cell model is efficient for local analysis in any site of NRU, the complex full-reactor model is required for calculations of the core physics and beyond-the-core radiation. By supplementing, adjusting or benchmarking the results from the existing NRU codes, the MCNP calculations provide greater confidence that NRU remains within the licence envelope.
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Zheng, Lei, Zhiyuan Feng, and Kan Wang. "ON-THE-FLY INTERPOLATION OF CONTINUOUS TEMPERATURE-DEPENDENT THERMAL NEUTRON SCATTERING DATA IN RMC CODE." EPJ Web of Conferences 247 (2021): 09012. http://dx.doi.org/10.1051/epjconf/202124709012.
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Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.
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Khan, Suhail Ahmad, and Umasankari Kannan. "Evaluation of DNBR with neutronics calculation in LWR systems." Nuclear Energy and Technology 9, no. 2 (2023): 77–83. http://dx.doi.org/10.3897/nucet.9.98452.
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The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.
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Khan, Suhail Ahmad, and Umasankari Kannan. "Evaluation of DNBR with neutronics calculation in LWR systems." Nuclear Energy and Technology 9, no. (2) (2023): 77–83. https://doi.org/10.3897/nucet.9.98452.
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The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.
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Ta, Duy Long, Huy Hiep Nguyen, Tuan Khai Nguyen, Vinh Thanh Tran, and Huu Tiep Nguyen. "Coulped neutronics/thermal-hydraulics calculation of VVER-1000 fuel assembly." Nuclear Science and Technology 6, no. 2 (2021): 31–38. http://dx.doi.org/10.53747/jnst.v6i2.153.
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This paper presents a computational scheme using MCNP5 and COBRA-EN for coupling neutronics/thermal hydraulics calculation of a VVER-1000 fuel assembly. A master program was written using the PERL script language to build the corresponding inputs for the MCNP5 and COBRA-EN calculations and to manage the coupling scheme. The hexagonal coolant channels have been used in the thermal hydraulics model using CORBRA-EN to simplify the coupling scheme. The results of two successive iterations were compared with an assigned convergence criterion and the loop calculation can be broken when the convergence criterion is satisfied. Numerical calculation has been performed based on a UO2fuel assembly of the VVER-1000 reactor.
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García, Manuel, Riku Tuominen, Andre Gommlich, et al. "SERPENT2-SUBCHANFLOW-TRANSURANUS PIN-BY-PIN DEPLETION CALCULATIONS FOR A PWR FUEL ASSEMBLY." EPJ Web of Conferences 247 (2021): 06016. http://dx.doi.org/10.1051/epjconf/202124706016.
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This work presents the results for a coupled neutronic-thermalhydraulic-thermomechanic pin-level depletion calculation of a PWR fuel assembly using Serpent2-SUBCHANFLOW-TRANSURANUS. This tool is based on a semi-implicit depletion scheme with pin-by-pin feedback, mesh-based field exchange and an object-oriented software design. The impact of including fuel-performance capabilities is analyzed, with focus on high-burnup effects. The treatment of the Doppler feedback to the neutronics is examined as well, in particular the use of radial fuel-temperature profiles or radially averaged values.
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Kuijper, J. C., and D. Muszynski. "Neutronics for the GEMINI+ HTGR." Journal of Physics: Conference Series 2048, no. 1 (2021): 012030. http://dx.doi.org/10.1088/1742-6596/2048/1/012030.
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Abstract Literally at the heart of the Euratom Horizon 2020 project GEMINI+ are the core neutronics (design) calculations. For these calculations on a relatively small (180 MWth) prismatic HTGR with cylindrical core, the 3-D monte-carlo particle transport and depletion code SERPENT version 2 (VTT, Finland) was selected, the main reasons being the flexibility and versatility of this code. This enables the modelling of all relevant details of the reactor without unnecessary approximations. A particularly useful feature of the SERPENT code is the multiphysics input capability. This allows to map a temperature field over the defined geometry, enabling the calculation of converged power and temperature distribution by means of iteration and data exchange between SERPENT and a (steady-state) thermal- hydraulics code. In this particular case the SPECTRA code (NRG, The Netherlands) was used to provide the temperature distribution. 4 to 5 iterations are sufficient to reach simultaneously converged distributions for power and temperature. The paper gives an overview of the performed analyses for the current (June 2020) design of the GEMINI+ HTGR, and results thereof. Neutronics features seem quite promising, but further improvements and therefore further investigations would be desirable.
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Yang, Wen, Jun Huang, Zhaofei Tian, and Guangliang Chen. "Uncertainty analysis and visualization for a coupled thermohydraulics-neutronics simulation by deterministic sampling." Journal of Physics: Conference Series 2313, no. 1 (2022): 012006. http://dx.doi.org/10.1088/1742-6596/2313/1/012006.
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Abstract Uncertainty quantification and sensitivity analysis by deterministic sampling is performed on a thermohydraulics-neutronics coupling simulation of a 5×5 rod bundle in pressurized water reactor. Thermohydraulics calculation is conducted by a CFD simulation and neutronics calculation is conducted by method of characteristic (MOC). The coupling simulation of CFD and MOC, together with large number of calculations needed by random sampling in uncertainty quantification, leads to high computational cost. Therefore, deterministic sampling is introduced in this study and proves its efficiency. Moreover, the calculated results usually have some distribution pattern on the selected section, using mass average value as output is unable to reflect the distribution feature. Therefore, a method realized by Python program is introduced to perform uncertainty quantification and sensitivity analysis of the outputs on the outlet section in fine-mesh level. It is found that the uncertainty of outputs and the inputs’ impact on outputs vary in different areas at the outlet section.
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Li, Bin, Bin Wu, Guangyao Sun, Lijuan Hao, Jing Song, and Ulrich Fischer. "APPLICATION OF SUPERMC3.2 TO PRELIMINARY NEUTRONICS ANALYSIS FOR EUROPEAN HCPB DEMO." EPJ Web of Conferences 247 (2021): 18006. http://dx.doi.org/10.1051/epjconf/202124718006.
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In the D-T fueled tokamak, the neutrons not only carry the approximately 80% energy released in the per fusion reaction, but also are the source of radioactivity in the fusion system. Therefore, high-fidelity neutronics simulation is required to support such reactor design and safety analysis. In the present work, taking European HCPB DEMO (Helium Cooled Pebble Bed demonstration fusion plant) developed by KIT (Karlsruhe Institute of Technology) as an example, the preliminary neutronics analysis covering the assessments of NWL (neutron wall loading), TBR (tritium breeding ratio), nuclear power generation, radiation loads on PFCs (plasma-facing components) and TFCs (toroidal field coils) has been carried out by using SuperMC in the case of both unbiased and biased simulations. The preliminary results indicate that the blanket scheme could satisfy the design requirements in terms of TBR and shielding of inboard blankets. Specially, a speed-up by ~164 times in the calculation for thick shielding region (TFC region) is achieved by using global weight windows generated via GWWG in SuperMC. In addition, compared to MCNP, SuperMC shows advantages in accurate and efficient modeling of complex system, efficient calculation and 3D interactive visualization.
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Kazansky, Yury A., and Gleb V. Karpovich. "Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility." Nuclear Energy and Technology 5, no. 4 (2019): 345–51. http://dx.doi.org/10.3897/nucet.5.48426.
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Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra. Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets. The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS. The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for 239Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries. The effects of various options of material arrangement on the values of keff were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the keff value by about 0.4%. The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV. The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%. The heterogeneous effect of α for 239Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations).
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Kazansky, Yury A., and Gleb V. Karpovich. "Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility." Nuclear Energy and Technology 5, no. (4) (2019): 345–51. https://doi.org/10.3897/nucet.5.48426.
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Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra. Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets. The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS. The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for <sup>239</sup>Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries. The effects of various options of material arrangement on the values of k<sub>eff</sub> were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the k<sub>eff</sub> value by about 0.4%. The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV. The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%. The heterogeneous effect of α for <sup>239</sup>Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations).
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Choe, Jiwon, Chirayu Batra, Vladimir Kriventsev, and Deokjung Lee. "Neutronic Analysis of Start-Up Tests at China Experimental Fast Reactor." Energies 15, no. 3 (2022): 1249. http://dx.doi.org/10.3390/en15031249.
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The China Experimental Fast Reactor (CEFR) is a small, sodium-cooled fast reactor with 20 MW(e) of power. Start-up tests of the CEFR were performed from 2010 to 2011. The China Institute of Atomic Energy made some of the neutronics start-up-test data available to the International Atomic Energy Agency (IAEA) as part of an international neutronics benchmarking exercise by distributing the experimental data to interested organizations from the member states of the IAEA. This benchmarking aims to validate and verify the physical models and neutronics simulation codes with the help of the recorded experimental data. The six start-up tests include evaluating criticality, control-rod worth, reactivity effects, and neutron spectral characteristics. As part of this coordinated research, the IAEA performed neutronics calculations using the Monte Carlo codes Serpent 2 and OpenMC, which can minimize modeling assumptions and produce reference solutions for code verification. Both codes model a three-dimensional heterogeneous core with an ENDF/B-VII.1 cross-section library. This study presents the calculation results with a well-estimated criticality and a reasonably good estimation of reactivities. The description and analysis of the core modeling assumptions, challenges in modeling a dense SFR core, results of the first phase of this project, and comparative analysis with measurements are presented.
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Tollit, Brendan, Alan Charles, William Poole, et al. "WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS." EPJ Web of Conferences 247 (2021): 06006. http://dx.doi.org/10.1051/epjconf/202124706006.
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The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions. The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.
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Yang, W., Q. Zeng, Ch Chen, et al. "SHIELDING DESIGN AND NEUTRONICS CALCULATION OF THE GDL BASED FUSION NEUTRON SOURCE ALIANCE." Problems of Atomic Science and Technology, Ser. Thermonuclear Fusion 44, no. 2 (2021): 164–66. http://dx.doi.org/10.21517/0202-3822-2021-44-2-164-166.
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TAKEDA, Toshikazu. "Improvement of Neutronics Calculation Methods for Fast Reactors." Progress in Nuclear Science and Technology 2 (October 1, 2011): 289–93. http://dx.doi.org/10.15669/pnst.2.289.
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Nguyen, Kien Cuong, Ton Nghiem Huynh, Vinh Vinh Le, and Ba Vien Luong. "Validation of neutronics libraries through benchmarks and critical configurations of The Dalat Nuclear Research Reactor using low enriched uranium fuel by monte carlo method." Nuclear Science and Technology 3, no. 4 (2013): 20–28. http://dx.doi.org/10.53747/nst.v3i4.323.
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From evaluated data sources like ENDF, JENDL and JEFF, neutronics data libraries forMCNP computer code have been produced, including neutron scattering cross section library S (β,α) in thermal energy range, by using NJOY computer code. The evaluation and validation of these neutronics data libraries have been carried out through calculation of some parameters such as effective multiplication factor and reaction cross sections of benchmark problems, VVR-M2 fuel type as well as the critical configurations of the Dalat Nuclear Research Reactor loaded with low enriched Uranium fuel. After implementing about analysis and evaluation of the calculated results with abovementioned libraries, the library provides results that consistent with experimental data can be used in core and fuel management calculation for the Dalat Nuclear Research Reactor (DNRR).
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Luciano, Nicholas P., Brian J. Ade, Kang Seog Kim, and Andrew J. Conant. "MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS." EPJ Web of Conferences 247 (2021): 10031. http://dx.doi.org/10.1051/epjconf/202124710031.
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MPACT is a state-of-the-art core simulator designed to perform high-fidelity analysis using whole-core, three-dimensional, pin-resolved neutron transport calculations on modern parallel computing hardware. MPACT was originally developed to model light water reactors, and its capabilities are being extended to simulate gas-cooled, graphite-moderated cores such as Magnox reactors. To verify MPACT’s performance in this new application, the code is being formally benchmarked using representative problems. Progression problems are a series of example models that increase in complexity designed to test a code’s performance. The progression problems include both beginning-of-cycle and depletion calculations. Reference solutions for each progression problem have been generated using Serpent 2, a continuous-energy Monte Carlo reactor physics burnup calculation code. Using the neutron multiplication eigenvalue ke_ as a metric, MPACT’s performance is assessed on each of the progression problems. Initial results showed that MPACT’s multigroup cross section libraries, originally developed for pressurized water reactor problems, were not sufficient to accurately solve Magnox problems. MPACT’s improved performance on the progression problems is demonstrated using this new optimized cross section library.
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Choe, Jiwon, Chirayu Batra, Vladimir Kriventsev, and Deokjung Lee. "MONTE CARLO SIMULATION OF NEUTRONICS START-UP TESTS AT CHINA EXPERIMENTAL FAST REACTOR (CEFR)." EPJ Web of Conferences 247 (2021): 10008. http://dx.doi.org/10.1051/epjconf/202124710008.
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China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release some of the neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activities through the coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”, launched in 2018. This benchmark aims to perform validation and verification of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data. The six physics start-up tests considered for this CRP include evaluation of the criticality, control rod worth, void reactivity, temperature coefficient, swap reactivity, and foil irradiation. Twenty-nine participating research organizations are performing independent blind calculations during the first phase of the project. As a part of this coordinated research, IAEA performed neutronics calculations using Monte Carlo code SERPENT. Two kinds of 3D core models, homogenous and heterogeneous, were calculated using SERPENT, with ENDF/B-VII.0 continuous energy library. Preliminary results with a reasonably good estimation of criticality, as well as theoretically sound results of other five test cases, are available. The paper will discuss the core modelling assumptions, challenges and key findings of modelling a dense SFR core, preliminary results of the first phase of the CRP, heterogeneity impact analysis between homogenous core models and heterogeneous core models and future work to be performed as a part of this four-year project.
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Manturov, G., M. Nikolaev, and V. Koshcheev. "NUCLEAR DATA FOR REACTOR NEUTRONICS CALCULATIONS - ROSFOND DATA LIBRARY AND ABBN-RF GROUP DATA SYSTEM." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (2021): 5–24. http://dx.doi.org/10.55176/2414-1038-2021-2-5-24.
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The work is devoted to one of the most important scientific and technical problems in reactor physics related to the development and verification of codes and nuclear data that provide reliable and highly accurate calculations of neutron-physical characteristics of fast reactors and radiation shielding, including nuclear fuel cycle, criticality and radiation safety parameters. The designed neutronics characteristics of fast reactors should be based on certified, qualified sets of codes and nuclear constants: the calculation tools should be related to the modern state of scientific knowledge and computational techniques, and used nuclear physics constants should be adequate to the most reliable evaluations of nuclear data. In connection with rapid development of the computing engineering and all greater introduction in practice of calculation Monte Carlo codes, the methodical constituent of calculation error falls down substantially. In these terms the nuclear constant’s constituent of error of calculations becomes fully qualificatory. A situation is intensifyed by the fall-off of financing of experimental works, why in this connection the amount of fast critical stands in the world diminishes sharply. The paper consider the state of art of the constant’s providing system CONSYST/ABBN, created on the basis of the national library of neutron data files ROSFOND and libraries of multigroup constants ABBN-93 and ABBN-RF. One of the most important problems under consideration here also is the methodical and software for estimating of errors of the calculated physical characteristics. System of codes and nuclear data for reactor neutronics calculations is based on the unified methodological basis that ensures the transparency of the procedure for obtaining the data used in the calculations, the reliability of their verification and the obtaining of guaranteed accuracy of the calculated physical reactor characteristics.
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Yuan, Baoxin, Jie Zheng, Jian Wang, et al. "Numerical Calculation Scheme of Neutronics-Thermal-Mechanical Coupling in Solid State Reactor Core Based on Galerkin Finite Element Method." Energies 16, no. 2 (2023): 659. http://dx.doi.org/10.3390/en16020659.
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It is of practical significance to study the multi-physical processes of solid state nuclear systems for device design, safety analysis, and operation guidance. This system generally includes three multi-physical processes: neutronics, heat transfer, and thermoelasticity. In order to analyze the multi-physical field behavior of solid state nuclear system, it is necessary to analyze the laws of neutron flux, temperature, stress, and other physical fields in the system. Aiming at this scientific goal, this paper has carried out three aspects of work: (1) Based on Galerkin’s finite element theory, the governing equations of neutronics, heat transfer, and thermoelasticity have been established; (2) a neutronics-thermal-mechanical multi-physical finite element analysis code was developed and verified based on benchmark examples and third-party software for multi-physical processes; (3) for a solid state nuclear system with a typical heat pipe cooled reactor configuration, based on the analysis code developed in this work, the neutronics-thermal-mechanical coupling analysis was carried out, and the physical field laws such as neutron flux, temperature, stress, etc., of the device under the steady-state operating conditions were obtained; and (4) finally, the calculation results are discussed and analyzed, and the focus and direction of the next work are clarified.
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HUANG, Liangsheng, Liqun HU, Luying NIU, et al. "Application of local Monte Carlo method in neutronics calculation of EAST radial neutron camera." Plasma Science and Technology 24, no. 2 (2022): 025601. http://dx.doi.org/10.1088/2058-6272/ac42bb.
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Abstract The Local Monte Carlo (LMC) method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera (RNC) diagnostic system on the experimental advanced superconducting tokamak (EAST), and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained. Compared with the results calculated by the global variance reduction method, it is shown that the LMC calculation is reliable within the reasonable error range. The calculation process of LMC is analyzed in detail, and the transport process of radiation particles is simulated in two steps. In the first step, an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation. The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary. Based on the characteristic that only a local geometric model is needed in the second step, it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.
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Mahlers, Y. P. "VVER-1000 neutronics calculation with ENDF/B-VII data." Annals of Nuclear Energy 36, no. 8 (2009): 1224–29. http://dx.doi.org/10.1016/j.anucene.2009.04.002.
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Zeng, Qin, Hongli Chen, Zhongliang Lv, Lei Pan, Haoran Zhang, and Wei Li. "Impact Analysis of the Model on CFETR Neutronics Calculation." Journal of Fusion Energy 35, no. 4 (2016): 683–88. http://dx.doi.org/10.1007/s10894-016-0090-1.
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Kovalishin, A. A., A. V. Moryakov, and K. F. Raskach. "Neutronics Calculation of Fast Reactor Using Modern Computing Systems." Atomic Energy 124, no. 2 (2018): 75–81. http://dx.doi.org/10.1007/s10512-018-0378-5.
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Popykin, A., N. Zhylmaganbetov, and A. Smirnova. "ABOUT THE REQUIREMENTS AND RECOMMENDATIONS OF THE REGULATORY BODY TO ERROR ESTIMATION OF NEUTRONICS CALCULATIONS." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 2 (2019): 127–35. http://dx.doi.org/10.55176/2414-1038-2019-2-127-135.
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The article discusses the issues of determining the error of neutron-physical calculation taking into account its specificity, the features of the code implementation of various calculation methods and the requirements of regulatory documents in the field of atomic energy. According to the requirements of Russian regulatory documents, the design and safety analysis of nuclear facilities must be carried out on the basis of a conservative approach. A conservative approach is an approach to design and building of nuclear power plants (NPP), when by choosing values of parameters and characteristics of NPPs, NPP sites and (or) other methods, more unfavorable results are obtained during accident analyze of NPP. In order to implement the principle of conservatism in neutron-physical calculations, in addition to determining the desired quantity, it is necessary to determine the error in the calculation of this quantity. The safety of nuclear facilities is mainly analyzed through calculations using codes. Russian regulatory documents require codes to be verified and validated, the result of verification and validation is the certification passport of the code, which contains the errors calculated by the code values. They are established as a result of verification (validation) of the code. It is noted that it is necessary to take into account measurement errors in experiments, including those carried out at operating power units, for code verification. It is noted that in the Safety Guide RB-061-11, concerning the neutron-physical calculation, recommendations are given to determine the error, the article presents their analysis and synthesis.
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GOTO, Minoru, Shigeaki NAKAGAWA, Kuniyoshi TAKAMATSU, and Tetsuaki TAKEDA. "ICONE15-10205 VALIDATION OF NEUTRONICS CALCULATION CODES FOR VHTR NUCLEAR DESIGN USING HTTR EXPERIMENTAL DATA." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_97.
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Cattaneo, Paolo, Roland Lenain, Elsa Merle, Cyril Patricot, and Didier Schneider. "NUMERICAL OPTIMIZATION OF A MULTIPHYSICS CALCULATION SCHEME." EPJ Web of Conferences 247 (2021): 06008. http://dx.doi.org/10.1051/epjconf/202124706008.
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This work concerns the numerical optimization of a multiphysics calculation scheme. The considered application is a 5x5 Pressurized Water Reactor (PWR) assemblies mini-core surrounded by radial and axial reflectors. The scenario adopted for the analysis is steady-state nominal conditions and fission products set to the equilibrium concentration. The neutronics is modelled at the pin-cell scale and the thermal-hydraulics at the subchannel level. Depending on the scenario, the damped fixed-point algorithm might not be sufficiently robust or efficient enough. For this reason, a technique based on the partial convergence of the solvers is tested. In every multiphysic iteration, a maximum number of iterations is imposed for both the neutronics and the thermal-hydraulics solvers. In combination with that, the solver restarts from the results of the last calculation. In this way, if the method is convergent, the initialization progresses towards the fixed-point solution. The results show that such a technique improves both the robustness and the speed of the algorithm. Within this approach, the range of relaxation factors that makes the algorithm converge is significantly broadened and the importance of this parameter on the global performance is reduced. The computing time also decreases by a factor between 10 and 20. Furthermore, this gain does not strongly depend on the exact imposed maximum number of iterations. Some preliminary observations are also reported in respect with the application of such a technique to the Anderson acceleration method.
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Chambon, Amalia, Esben Klinkby, Leif Emås, and Bent Lauritzen. "Assessment of shutdown dose rates at the ESS target cooling system using SCALE6.2." Journal of Neutron Research 22, no. 2-3 (2020): 309–18. http://dx.doi.org/10.3233/jnr-190136.
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The production of high-energy neutrons at the European Spallation Source through the spallation process may cause an erosion of the tungsten target. The eroded particles could be released into the target helium cooling system which contains four kind of filters. Among them, the auxiliary filters called “getters” are designed to capture volatile elements and remaining dust. In this work, the ORNL’s SCALE6.2 modelling and simulation suite for nuclear safety analysis is applied to assess shutdown dose rates and determine if added shielding and/or robotic arms are needed for their maintenance. SCALE6.2 is well adapted to treat this problem as it allows for isotope selection regarding source term calculation. Dose rates are determined by an ORIGEN2 source term and a MAVRIC shielding sequence calculation. As SCALE6.2 is non-standard software for ESS, the results are verified against MCNP, which is the baseline tool for neutronics analysis at ESS. Dose rate calculations show that additional shielding and/or robot arm are not needed to remove the getters after 3 months of cooling time, following 5400 h of operation at 5 MW beam power. At a distance of 1 mm from the getter, the dose rate is 0.2 mSv/h in the most conservative estimation.
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Andrianov, Andrei, Olga Andrianova, Ilya Kuptsov, and Anastasia Uvarova. "COMPARISON OF THE CALCULATION ACCURACY OF THE NEUTRONICS CHARACTERISTICS OF A HEAVY LIQUID METAL COOLED FAST REACTOR MODEL USING THE VARIOUS EVALUATED NEUTRON DATA LIBRARIES." Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika 3 (January 9, 2024): 109–24. https://doi.org/10.26583/npe.2024.3.09.
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The paper analyses the neutron data and covariance matrixes, which are crucial for the calculation prediction of the neutronics characteristics of fast neutron reactors with uranium-plutonium fuel and currently available in the state-of-the-art versions of evaluated nuclear data libraries: ENDF/B-VIII, JENDL-5, TENDL 2021, JEFF 4T1. The newly obtained covariance data are compared to the data that was presented in the Russian evaluated nuclear data library BROND 3.1. A simplified model of a fast neutron reactor with mixed dense nitride uranium-plutonium fuel and a heavy liquid metal coolant was applied to calculate the spread of neutronics characteristic values and their uncertainties due to neutron data using various evaluated nuclear data libraries for the following characteristics: effective multiplication factor, effective delayed neutron fraction, Doppler reactivity coefficient, breeding ratio and gain, reactivity margin for fuel burnup and other fuel burnup characteristics. It has been observed that, on the whole, the uncertainties of the reactor functionals have decreased when estimated using state-of-the-art versions of the evaluated neutron data libraries, compared to previous releases of those libraries. The article also examines the changes in target accuracies for predicting the main neutronics characteristics of fast neutron reactors over the past decades, as well as evaluates the requirements for neutron data uncertainties to achieve the recently declared target accuracies. The main findings of this study are presented in two aspects: firstly, in terms of the effects evaluated neutron data from different libraries have on the accuracy of calculations for the primary neutronics characteristics of fast reactors; and secondly, in terms of the potential for improving the accuracy of predicted neutron-ics characteristics of fast reactors by considering the results of reactor physics measurements.
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Wang, Xinyan, Yuxuan Liu, William Martin, and Shane Stimpson. "IMPLEMENTATION OF 2D/1D GAMMA TRANSPORT AND GAMMA HEATING CAPABILITY IN MPACT." EPJ Web of Conferences 247 (2021): 02038. http://dx.doi.org/10.1051/epjconf/202124702038.
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This paper presents the most recent progress on the development of gamma transport capability for the CASL neutronics code MPACT: (1) 3D gamma transport and (2) explicit gamma heating capabilities. The 3D gamma calculation capability was implemented by leveraging the 2D/1D solver originally developed for neutron calculations. The results were verified by MCNP6 on a small assembly with 5 × 5 pins. Generally, errors were lower than 0.5% on each axial mesh as long as MPACT was running with enough axial meshes. The gamma heating calculation considered the energy deposition from photoelectric absorption, Compton scattering, and pair production. Verification with MCNP6 for both 2D and 3D benchmarks showed that the errors of energy depositions are comparable with those of gamma fluxes, proving the proper implementation of the energy deposition.
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Vuiart, Romain, Aimeric Eustache, Sarah Eveillard, and Géraud Prulhière. "PRATIC: A soluble-boron-free, pressurized water cooled, SMR core benchmark." EPJ Nuclear Sciences & Technologies 10 (2024): 25. https://doi.org/10.1051/epjn/2024026.
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Current nuclear reactor research is actively exploring small modular pressurized water reactors (PWRs), particularly soluble-boron-free (SBF) configurations. SBF designs utilize control rods and gadolinium-poisoned fuel rods to manage reactivity. Additionally, small modular reactors (SMRs) commonly integrate steel reflectors to minimize neutron leakage. However, the compactness of SMRs and the adoption of these technical solutions result in notable fluctuations in neutron flux within the core during normal cycle operation. Hence, comprehensive analysis is essential, especially concerning reactor performance under normal and accident conditions, the reliability of neutronics calculation assumptions, etc. Research into these issues requires reactor core neutronics benchmarks that are consistent with industrial concepts so that the analysis results can be applied to real reactors. In this context, this article introduces PRATIC, a SBF-PWR SMR core neutronics benchmark designed to match the global performances of industrial concepts. The development of PRATIC was conducted using a deterministic calculation scheme coupling the APOLLO2 and CRONOS2 codes. PRATIC features a thermal power of 350 MWth, an equilibrium cycle length of 1.9 years, and an average discharge burnup of about 34 GWd/t, while maintaining controlled power distributions. The article elaborates on the design assumptions for PRATIC, then details the reactor core and its equilibrium cycle. Access to the PRATIC modeling data is available via a GIT repository, accessible upon request via email at pratic@cea.fr.
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Read, Nathaniel, and Eugene Shwageraus. "APPLICATION OF TWO STAGE METHOD OF CHARACTERISTICS / SP3 METHODOLOGY TO TRISO-FUELLED LEU SPACE REACTOR IN WIMS 11." EPJ Web of Conferences 247 (2021): 01009. http://dx.doi.org/10.1051/epjconf/202124701009.
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In order to minimise the mass of a 1MWe LEU space fission power system design, a rapid neutronics analysis tool is sought. A two-stage deterministic analysis routine has been constructed using a core-plane method of characteristics calculation followed by a full-core SP3 calculation, within the ANSWERS© code WIMS11. This is compared to a faster route that skips the core-plane calculation and also the Monte Carlo code Serpent. Results suggest sufficiently good agreement for the WIMS-based methods to be useful in a full system mass-minimising optimisation routine.
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Wilson, DJ, and AIM Ritchie. "Neutron moisture meter : the dependence of their response on soil parameters." Soil Research 24, no. 1 (1986): 11. http://dx.doi.org/10.1071/sr9860011.
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A multigroup diffusion theory calculation based on a nuclear reactor neutronics code is used to determine the response of a neutron moisture meter to changes in soil parameters such as dry soil density, soil water content, thermal neutron absorption cross-section and neutron scattering crosssection. Empirical equations which fit the results can be used to estimate the response at values of the soil parameters other than those used in the calculations. These equations can also be used to estimate the accuracy with which the parameters must be known to achieve a required accuracy in the derived soil water content.
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Rabir, Mohamad Hairie, Aznan Fazli Ismail, and Mohd Syukri Yahya. "Neutronics calculation of the conceptual TRISO duplex fuel rod design." Nuclear Materials and Energy 27 (June 2021): 101005. http://dx.doi.org/10.1016/j.nme.2021.101005.
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Cepraga, D. G., G. Cambi, F. Carloni, M. Frisoni, and D. Ene. "Neutronics and activation calculation for ITER generic site safety report." Fusion Engineering and Design 63-64 (December 2002): 193–97. http://dx.doi.org/10.1016/s0920-3796(02)00130-8.
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Takeda, Toshikazu, W. F. G. van Rooijen, Katsuhisa Yamaguchi, Masayoshi Uno, Yuji Arita, and Hiroyasu Mochizuki. "Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor." Science and Technology of Nuclear Installations 2012 (2012): 1–13. http://dx.doi.org/10.1155/2012/351809.
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This paper discusses the objectives and results of a multiyear R&D project to improve the modeling accuracy for the detailed calculation of the Japanese Sodium-cooled Fast Reactor (JSFR), although the preliminary design of JSFR is prepared using conventional methods. For detailed design calculations, new methods are required because the JSFR has special features, which cannot be accurately modeled with existing codes. An example is the presence of an inner duct in the fuel assemblies. Therefore, we have developed new calculational and experimental methods in three areas: (1) for neutronics, we discuss the development of methods and codes to model advanced FBR fuel subassemblies, (2) for fuel materials, modeling and measurement of the thermal conductivity of annular fuel is discussed, and (3) for thermal hydraulics, we describe advances in modeling and calculational models for the intermediate heat exchanger and the calculational treatment of thermal stratification in the hot plenum of an FBR under low flow conditions. The new methods are discussed and the verification tests are described. In the validation test, measured data from the prototype FBR Monju is partly used.
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Luong, Ba Vien, Vinh Vinh Le, Ton Nghiem Huynh, and Kien Cuong Nguyen. "Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor." Nuclear Science and Technology 4, no. 1 (2014): 10–25. http://dx.doi.org/10.53747/jnst.v4i1.209.
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The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.
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Irka, F. H., Z. Suud, D. Irwanto, S. N. Khotimah, and H. Sekimoto. "Neutronics performances of gas-cooled fast reactor for 300-600 MWt Output Power with Modified CANDLE burn-up scheme in radial direction." Journal of Physics: Conference Series 2072, no. 1 (2021): 012013. http://dx.doi.org/10.1088/1742-6596/2072/1/012013.
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Abstract Gas-Cooled Fast Reactor-GFR is a Generation IV reactor that is helium-cooled and has a closed fuel cycle. Due to the target operation on 2022-2030, this reactor type still needs further research and development technologies. We investigated the neutronics performances of a GFR balance type core with some modification of CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) burn-up scheme in the radial direction. The output power varied from 300 to 600 MWt. The neutronics calculation was performed using SRAC 2002 with JENDL 4.0 nuclear data library. The analysis indicate the reactor could operate critically for ten years without refueling with burn-up level 20% HM.
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Sood, Avneet, R. Arthur Forster, B. J. Archer, and R. C. Little. "Neutronics Calculation Advances at Los Alamos: Manhattan Project to Monte Carlo." Nuclear Technology 207, sup1 (2021): S100—S133. http://dx.doi.org/10.1080/00295450.2021.1956255.
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FURUTA, Kazuo, Yoshiaki OKA, and Shunsuke KONDO. "Accuracy of Multi-Group Transport Calculation in D-T Fusion Neutronics." Journal of Nuclear Science and Technology 24, no. 4 (1987): 333–39. http://dx.doi.org/10.1080/18811248.1987.9735810.
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Andrianov, Andrey A., Olga N. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, and Anastasiya A. Spiridonova. "A computer code for optimizing the neutronics model parameters based on results of reactor physics experiments." Nuclear Energy and Technology 9, no. 4 (2023): 289–96. http://dx.doi.org/10.3897/nucet.9.117198.
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The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.
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Andrianov, Andrey A., Olga N. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, and Anastasiya A. Spiridonova. "A computer code for optimizing the neutronics model parameters based on results of reactor physics experiments." Nuclear Energy and Technology 9, no. (4) (2023): 289–96. https://doi.org/10.3897/nucet.9.117198.
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The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code's application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.
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Shahid, Izza, Nadeem Shaukat, Amjad Ali, et al. "Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION." Science and Technology of Nuclear Installations 2022 (January 5, 2022): 1–13. http://dx.doi.org/10.1155/2022/6319628.
Full textAbstract:
A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.