scholarly journals Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis

2011 ◽  
Vol 2011 ◽  
pp. 1-7 ◽  
Author(s):  
M. Pecchia ◽  
C. Parisi ◽  
F. D'Auria ◽  
O. Mazzantini
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
3D Model ◽  
Three Dimensional ◽  
Control Rod ◽  
Geometrical Complexity ◽  
Reactor Criticality ◽  
Reactor Cell ◽  
And Control ◽  
Control Rods

The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two-dimensional deterministic lattice physics code HELIOS. These results were implemented in the RELAP-3D model to perform safety analyses for the licensing process.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

scholarly journals Dynamic modeling of tunnel survey spatiotemporal data

2014 ◽  
Vol 20 (2) ◽  
pp. 354-375
Author(s):  
Xiaolong Li ◽  
Jiansi Yang ◽  
Bingxuan Guo ◽  
Hua Liu ◽  
Jun Hua
Keyword(s):  
Survey Data ◽  
Cross Section ◽  
Cross Sections ◽  
3D Modeling ◽  
3D Model ◽  
Three Dimensional ◽  
Spatiotemporal Data ◽  
Excavation Process ◽  
Special Cases ◽  
Time Stamps

Currently, for tunnels, the design centerline and design cross-section with time stamps are used for dynamic three-dimensional (3D) modeling. However, this approach cannot correctly reflect some qualities of tunneling or some special cases, such as landslips. Therefore, a dynamic 3D model of a tunnel based on spatiotemporal data from survey cross-sections is proposed in this paper. This model can not only playback the excavation process but also reflect qualities of a project typically missed. In this paper, a new conceptual model for dynamic 3D modeling of tunneling survey data is introduced. Some specific solutions are proposed using key corresponding technologies for coordinate transformation of cross-sections from linear engineering coordinates to global projection coordinates, data structure of files and database, and dynamic 3D modeling. A 3D tunnel TIN model was proposed using the optimized minimum direction angle algorithm. The last section implements the construction of a survey data collection, acquisition, and dynamic simulation system, which verifies the feasibility and practicality of this modeling method.

Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry

2018 ◽  
Author(s):  
Guo Chao ◽  
Liu Yu ◽  
He Hangxing ◽  
Liu Luguo ◽  
Wang Xiaoyu ◽  
...  
Keyword(s):  
Cross Sections ◽  
Three Dimensional ◽  
Expansion Method ◽  
High Order ◽  
Benchmark Problems ◽  
Weighting Method ◽  
Control Rod ◽  
Order Polynomial ◽  
Runge Kutta ◽  
Nodal Expansion Method

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry (HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated.

scholarly journals The Investigation of EM Scattering from the Time-Varying Overturning Wave Crest Model by the IEM

2016 ◽  
Vol 2016 ◽  
pp. 1-8
Author(s):  
Xiao Meng ◽  
Li-xin Guo ◽  
Tian-qi Fan
Keyword(s):  
Cross Sections ◽  
3D Model ◽  
Three Dimensional ◽  
Integral Equation Method ◽  
Wave Crest ◽  
Time Varying ◽  
Spike Generation ◽  
Horizontal Polarization ◽  
Em Scattering ◽  
Radar Cross Sections

Investigation of the electromagnetic (EM) scattering of time-varying overturning wave crests is a worthwhile endeavor. Overturning wave crest is one of the reasons of sea spike generation, which increases the probability of false radar alarms and reduces the performance of multitarget detection in the environment. A three-dimensional (3D) time-varying overturning wave crest model is presented in this paper; this 3D model is an improvement of the traditional two-dimensional (2D) time-varying overturning wave crest model. The integral equation method (IEM) was employed to investigate backward scattering radar cross sections (RCS) at various incident angles of the 3D overturning wave crest model. The super phenomenon, where the intensity of horizontal polarization scattering is greater than that of vertical polarization scattering, is an important feature of sea spikes. Simulation results demonstrate that super phenomena may occur in some time samples as variations in the overturning wave crest.

Impact of control rod insertion during burnup on PWR fuel assembly isotopic composition

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 173-181
Author(s):  
R. M. Refeat ◽  
H. K. Louis
Keyword(s):  
Safety Analysis ◽  
Second Phase ◽  
Spent Fuel ◽  
Reactor Operation ◽  
Control Rod ◽  
Two Phases ◽  
And Control ◽  
Criticality Analysis ◽  
Control Rods ◽  
Criticality Safety

Abstract Criticality analysis of spent fuel assumes that the fuel material is unburned which means that it is in its most reactive condition. In fact, this is not the real situation for fuel as it is burned during reactor operation causing reduction in the reactivity. Considering the reduction in reactivity during spent fuel calculations is the Burn-up Credit concept (BUC). In addition, the control rods radial and axial positions have an effect on the reactivity which can be considered in the criticality safety analysis. This paper studies the effect of burnup and control rods (CRs) movement on reactivity and isotopes inventory. Calculations are carried out in two phases, first kinf is calculated for different burnup profiles with control rods are either fully withdrawn or fully inserted. In the second phase keff is calculated for different control rods insertion levels. For both phases, burnup calculations are performed for a UO2 assembly then multiplication factor calculations of burned UO2 assemblies in cold state are done. The burnup calculations are performed using MCNP6 code and ENDF/B-VII library for different burnup levels up to 45 GWd/tU. The results obtained can be taken in consideration in criticality safety analysis performed for the spent fuel to improve the economic efficiency for manufacture, storage and transportation of fissile materials.

scholarly journals Identification of Limiting Case Between DBA and SBDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System

2012 ◽  
Author(s):  
Raul Gonzalez ◽  
Alessandro Petruzzi ◽  
Francesco D’Auria ◽  
Oscar Mazzantini
Keyword(s):  
Three Dimensional ◽  
Injection System ◽  
Neutron Kinetics ◽  
The Third ◽  
Geometrical Complexity ◽  
Break Area ◽  
The Republic ◽  
Under Construction ◽  
Limiting Case ◽  
Control Rods

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void Coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D© system code. Within the framework of the third Agreement “NA-SA – University of Pisa” a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

scholarly journals Control blade history reactivity and pin power effects evaluated with Westinghouse BWR nodal core simulator POLCA8

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Petri Forslund Guimaraes
Keyword(s):  
Cross Sections ◽  
Critical Condition ◽  
Prediction Accuracy ◽  
Predictive Accuracy ◽  
Boiling Water ◽  
Simultaneous Presence ◽  
Cbh Model ◽  
Control Rod ◽  
Water Reactor ◽  
Control Rods

The so-called “Control Blade History” (CBH) effect has always posed a serious challenge for any nodal core simulator in performing Boiling Water Reactor (BWR) core analyses. In this paper a method to handle such CBH effects is proposed based on the idea of interpolating lattice physics data between two extreme cases with regard to CBH, namely, the case with the control rod always inserted during depletion and the case with the control rod never inserted during fuel irradiation. In POLCA8, the latest upgrade of the Westinghouse BWR nodal core simulator POLCA, one applies the methodology to macroscopic cross sections, discontinuity factors, pin powers and detector constants. Overall, the proposed CBH model performs very well in terms of predictive accuracy of reactivity and pin powers although simultaneous presence of control rods (CRs) and burnable absorbers (BAs) still poses a challenge due to some observed interference of their impact on reactivity. Applying the CBH model for pin power reconstruction is particularly promising and provides excellent prediction accuracy in the vicinity of the CR and at the point of CR withdrawal being the most challenging and critical condition with regard to CBH.

Utilization Research and Development of Hydride Materials in Fast Reactors

2014 ◽  
Vol 94 ◽  
pp. 23-31 ◽  
Author(s):  
Kenji Konashi ◽  
Kunihiro Itoh ◽  
Tsugio Yokoyama ◽  
Michio Yamawaki
Keyword(s):  
Cross Sections ◽  
Fast Neutrons ◽  
High Hydrogen ◽  
Neutron Absorber ◽  
Control Rod ◽  
Irradiation Test ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Hafnium Hydride ◽  
Control Rods

Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. An application of the hafnium hydride has been investigated as a neutron absorber in the Fast Breeder Reactors (FBRs). Fast neutrons are efficiently moderated by hydrogen in Hf hydrides and are absorbed by Hf. Since three isotopes of Hf have large cross sections, increase in the life of control rod is considered by Hf hydride. Results of design study of the core with Hf hydride control rods shows that the long lived hafnium hydride control rod is feasible in the large sodium-cooled FBR. Results of irradiation test conducted in BOR-60 has demonstrated the integrity of the capsules during irradiation. Na bonded capsule has an advantage in confinement effect of hydrogen compared with He bonded one. An application of hydride technique to transmutation target of MA was also discussed. MA hydride target is able to enhance the transmutation rate in FBR.

scholarly journals MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS

2021 ◽  
Vol 247 ◽  
pp. 02007
Author(s):  
Tung Dong Cao Nguyen ◽  
Hyunsuk Lee ◽  
Xianan Du ◽  
Vutheam Dos ◽  
Tuan Quoc Tran ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Fast Reactor ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Metallic Fuel ◽  
Sodium Fast Reactor ◽  
The Difference ◽  
Nodal Diffusion ◽  
Core Problem

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

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