scholarly journals MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE

2021 ◽  
Vol 247 ◽  
pp. 02028
Author(s):  
Wojciech Rydlewicz ◽  
Emil Fridman ◽  
Eugene Shwageraus
Keyword(s):  
Monte Carlo ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fast Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Complex Core ◽  
Nodal Diffusion ◽  
Full Core

This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

scholarly journals VERIFIKASI PROGRAM PWR-FUEL DALAM MANAJEMEN BAHAN BAKAR PWR

2015 ◽  
Vol 16 (1) ◽  
pp. 53 ◽  
Author(s):  
Surian Pinem ◽  
Tagor M. Sembiring ◽  
Mr Tukiran
Keyword(s):  
Power Distribution ◽  
Transport Theory ◽  
Neutron Transport ◽  
Fission Products ◽  
Diffusion Method ◽  
Fuel Management ◽  
The Core ◽  
Neutron Transport Theory ◽  
Fuel Assemblies ◽  
Nodal Diffusion

ABSTRAKVERIFIKASI PROGRAM PWR-FUEL DALAM MANAJEMEN BAHAN BAKAR PWR. Majemen bahan bakar dalam teras PWR tidak mudah karena jumlah perangkat bahan bakar dalam teras sebanyak 192 perangkat sehingga banyak kemungkinan penempatan bahan bakar dalam teras. Konfigurasi perangkat bahan bakar dalam teras harus tepat dan akurat sehingga reaktor beroperasi aman dan ekonomis. Untuk itu perlu dilakukan verifikasi program PWR-FUEL yang akan digunakan dalam manajemen bahan bakar PWR. Program PWR-FUEL didasarkan pada teori transport neutron dan diselesaikan dengan pendekatan metode difusi nodal banyak dimensi banyak kelompok dan metode difusi beda hingga (FDM). Tujuannya untuk memeriksa apakah program berfungsi dengan baik terutama untuk desain dan mana-jemen bahan bakar teras PWR. Verifikasi dilakukan dengan model pencarian teras setimbang pada tiga kondisi yaitu bebas boron, konsentrasi boron 1000 ppm dan konsentrasi boron kritis. Hasil perhitungan distribusi fraksi bakar rata-rata perangkat bahan bakar dan distribusi daya pada BOC dan EOC menunjukkan tren yang konsisten dimana perangkat bahan bakar dengan dengan daya yang tinggi pada BOC akan menghasilkan fraksi bakar yang tinggi pada EOC. Pada teras tanpa boron diperoleh faktor multiplikasi yang tinggi karena tidak adanya boron dalam teras dan efek produk fisi pada teras sekitar 3,8 %. Efek reaktivitas larutan boron 1000 ppm pada BOC dan EOC masing-masing 6,44 % dan 1,703 %. Distribusi fluks neutron dan kerapatan daya menggunakan metode NODAL dan FDM mempunyai hasil yang sama. Hasil verifikasi menunjukkan bahwa program PWR-FUEL berfungsi dengan baik terutama untuk desain dan pengolahaan bahan bakar dalam teras PWR. ABSTRACTTHE VERIFICATION OF PWR-FUEL CODE FOR PWR IN-CORE FUEL MANAGEMENT. In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron consentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is  6.44% and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

scholarly journals Validation of Doppler Temperature Coefficients and Component Power Distribution for the Advanced Neutronics Component Program KYLIN V2.0

2021 ◽  
Vol 9 ◽  
Author(s):  
Lei Jichong ◽  
Xie Jinsen ◽  
Chen Zhenping ◽  
Yu Tao ◽  
Yang Chao ◽  
...  
Keyword(s):  
Engineering Design ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Engineering ◽  
Temperature Coefficients ◽  
The Core ◽  
Fuel Assemblies ◽  
Design Requirements ◽  
Design Software ◽  
Good Agreement

This work is interested in verifying and analyzing the advanced neutronics assembly program KYLIN V2.0. Assembly calculations are an integral part of the two-step calculation for core design, and their accuracy directly affects the results of the core physics calculations. In this paper, we use the Doppler coefficient numerical benchmark problem and CPR1000 AFA-3G fuel assemblies to verify and analyze the advanced neutronics assembly program KYLIN V2.0 developed by the Nuclear Power Institute of China. The analysis results show that the Doppler coefficients calculated by KYLIN V2.0 are in good agreement with the results of other well-known nuclear engineering design software in the world; the power distributions of AFA-3G fuel assemblies are in good agreement with the results of the RMC calculations, it’s error distribution is in accordance with the normal distribution. It shows that KYLIN V2.0 has high calculation accuracy and meets the engineering design requirements.

scholarly journals Predicting Ex-core Detector Response in a PWR with Monte Carlo Neutron Transport Methods

2020 ◽  
Vol 225 ◽  
pp. 03007
Author(s):  
Tanja Goričanec ◽  
Domen Kotnik ◽  
Žiga Štancar ◽  
Luka Snoj ◽  
Marjan Kromar
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Reactor Core ◽  
Response Prediction ◽  
Detector Response ◽  
Design Calculation ◽  
Monte Carlo Code ◽  
Control Rod ◽  
The Core ◽  
Fuel Assemblies

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

scholarly journals Thorium application in the medium-sized sodium-cooled fast reactor

2019 ◽  
Vol 137 ◽  
pp. 01030
Author(s):  
Eeshu Raaj Saasthaa Arumuga Kumar ◽  
Piotr Darnowski ◽  
Mihir Kiritbhai Pancholi ◽  
Aleksandra Dzido
Keyword(s):  
Fast Reactor ◽  
Waste Incineration ◽  
Computer Code ◽  
Fuel Cycles ◽  
Mixed Oxide Fuel ◽  
The Core ◽  
Long Time ◽  
The Impact ◽  
Thorium 232 ◽  
Full Core

The report presents an analysis of the medium-sized Sodium-Cooled Fast Reactor (SFR) core with Thorium-based Mixed-Oxide fuel. The introduction of Transuranics (TRU) to the fuel was to allow long-lived nuclear waste incineration. The studied core is based on the modified Advanced Burner Reactor (ABR) 1000MWth core design, which was analysed in the OECD/NEA “Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core simulations with SERPENT 2.1.31 Monte Carlo computer code and ENDF library were performed, including static criticality and fuel burnup calculations for five fuel cycles. The core inventories at the Beginning of Cycle (BOC) and End of Cycle (EOC) were studied, and the impact of thorium fuel was assessed. The proposed core design is a burner reactor which uses thorium fuel. The excess core reactivity stays positive for long time despite large net consumption of transuranic elements as new fissile Uranium 233 is constantly breed from Thorium 232. Breeding of uranium allows longer fuel cycles.

scholarly journals On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Luigi Lepore ◽  
Romolo Remetti ◽  
Mauro Cappelli
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Signal To Noise Ratio ◽  
Reactor Power ◽  
Fast Reactors ◽  
Flux Monitoring

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Development of a Method for Evaluation of Pin-wise Power Distribution in Fuel Assemblies of Fast Reactors

1997 ◽  
Vol 34 (10) ◽  
pp. 983-991
Author(s):  
Mitsuaki YAMAOKA ◽  
Masatoshi KAWASHIMA ◽  
Takashi YAMAGUCHI ◽  
Hirofumi TAKASHITA
Keyword(s):  
Power Distribution ◽  
Fast Reactors ◽  
Fuel Assemblies

scholarly journals SPX Benchmark Part I: Results of Static Neutronics

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk ◽  
Liang Zhang ◽  
Evgeny Nikitin ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Thermal Conditions ◽  
Reaction Rates ◽  
Second Phase ◽  
Reference Solution ◽  
Transient Behaviour ◽  
The Core ◽  
Reactivity Coefficients

Abstract In the paper, the specification of a new neutronics benchmark for a large Sodium cooled Fast Reactor core and results of modelling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and results of the reference solution. Different core geometries and thermal conditions from cold “as fabricated” up to full power were considered. The reference Monte Carlo solution of Serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modelling with seven other solutions using deterministic and Monte Carlo methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for evaluation of transient behaviour of the core.

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