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.

Conceptual Core Design of HAPPY200 Reactor

2018 ◽  
Author(s):  
Xiaosheng Li ◽  
Linsen Li ◽  
Lianghui Peng ◽  
Xiaosong Chen ◽  
Zhaocan Meng ◽  
...  
Keyword(s):  
Temperature Coefficient ◽  
Power Distribution ◽  
Reactor Core ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Control Rod ◽  
The Core ◽  
Passive Safety System ◽  
Reactivity Coefficients ◽  
Full Power

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.

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 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.

The Super Equivalence Method in Monte Carlo Based Homogenization

2014 ◽  
Author(s):  
Mancang Li ◽  
Kan Wang ◽  
Dong Yao
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Surface Current ◽  
Reaction Rates ◽  
Reactor Physics ◽  
Sph Method ◽  
Equivalence Theory ◽  
Spe Method ◽  
Improved Accuracy

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.

scholarly journals Evaluation of PBMR-400 Core Design Steady State Condition with Serpent and AGREE

2021 ◽  
Vol 2048 (1) ◽  
pp. 012024
Author(s):  
H Ardiansyah ◽  
V Seker ◽  
T Downar ◽  
S Skutnik ◽  
W Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Steady State Condition ◽  
Light Water ◽  
Water Reactor ◽  
The Core ◽  
Core Simulation ◽  
Full Core

Abstract The significant recent advances in computer speed and memory have made possible an increasing fidelity and accuracy in reactor core simulation with minimal increase in the computational burden. This has been important for modeling some of the smaller advanced reactor designs for which simplified approximations such as few groups homogenized diffusion theory are not as accurate as they were for large light water reactor cores. For narrow cylindrical cores with large surface to volume ratios such the Ped Bed Modular Reactor (PBMR), neutron leakage from the core can be significant, particularly with the harder neutron spectrum and longer mean free path than a light water reactor. In this paper the core from the OECD PBMR-400 benchmark was analyzed using multigroup Monte Carlo cross sections in the HTR reactor core simulation code AGREE. Homogenized cross sections were generated for each of the discrete regions of the AGREE model using a full core SERPENT Monte Carlo model. The cross sections were generated for a variety of group structures in AGREE to assess the importance of finer group discretization on the accuracy of the core eigenvalue and flux predictions compared to the SERPENT full core Monte Carlo solution. A significant increase in the accuracy was observed by increasing the number of energy groups, with as much as a 530 pcm improvement in the eigenvalue calculation when increasing the number of energy groups from 2 to 14. Significant improvements were also observed in the AGREE neutron flux distributions compared to the SERPENT full core calculation.

scholarly journals EFFECT OF THE UNIFORM FISSION SOURCE METHOD ON LOCAL POWER VARIANCE IN FULL CORE SERPENT CALCULATION

2021 ◽  
Vol 247 ◽  
pp. 04024
Author(s):  
Yurii Bilodid ◽  
Jaakko Leppänen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Local Power ◽  
Axial Distribution ◽  
Source Method ◽  
Power Variance ◽  
Full Core ◽  
Statistical Variance

One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method. In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.

scholarly journals Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

2014 ◽  
Vol 2014 ◽  
pp. 1-8
Author(s):  
Po Hu ◽  
Paul P. H. Wilson
Keyword(s):  
Steady State ◽  
Supercritical Water ◽  
Power Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Core Model ◽  
Core Flow ◽  
The Core ◽  
Extended Code ◽  
Reference Design

This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

scholarly journals Flattening of the Power Distribution in the HTGR Core with Structured Control Rods

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 7377
Author(s):  
Michał Górkiewicz ◽  
Jerzy Cetnar
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Reactor Performance ◽  
Fission Rate ◽  
Control Rod ◽  
Structured Design ◽  
The Core ◽  
Stable Core ◽  
Control Rods ◽  
Burnup Calculation

Control rods (CRs) have a significant influence on reactor performance. Withdrawal of a control rod leaves a region of the core significantly changed due to lack of absorber, leading to increased fission rate and later to Xe135 buildup. In this paper, an innovative concept of structured control rods made of tungsten is studied. It is demonstrated that the radial division of control rods made of tungsten can effectively compensate for the reactivity loss during the irradiation cycle of high-temperature gas-cooled reactors (HTGRs) with a prismatic core while flattening the core power distribution. Implementation of the radial division of control rods enables an operator to reduce this effect in terms of axial power because the absorber is not completely removed from a reactor region, but its amount is reduced. The results obtained from the characteristic evolution of the reactor core for CRs with a structured design in the burnup calculation using the refined timestep scheme show a very stable core evolution with a reasonably low deviation of the power density and Xe135 concentration from the average values. It is very important that all the distributions improve with burnup.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

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