scholarly journals Fuel Loading Pattern Optimization with Constraint on Fuel Assembly Inventory Using Quantum-Inspired Evolutionary Algorithm

2018 ◽  
Vol 42 ◽  
pp. 01007 ◽  
Author(s):  
M. Rizki Oktavian ◽  
Alexander Agung ◽  
Andang Widi Harto
Keyword(s):  
Fuel Assembly ◽  
Evolutionary Algorithm ◽  
Cycle Length ◽  
Reactor Core ◽  
Combinatorial Problem ◽  
Weight Factor ◽  
Fuel Loading ◽  
Loading Pattern ◽  
Fitness Value ◽  
Fuel Assemblies

Nuclear fuel management was done by optimizing fuel loading pattern in a reactor core. Practically, performing fuel loading pattern optimization was difficult because of its combinatorial problem complexity which needed to be solved. Therefore, Quantum-inspired Evolutionary Algorithm (QEA) which could solve the combinatorial problem faster than conventional method was used. The main purpose of this research was to obtain an optimum fuel loading pattern of KSNP-1000 reactor core without altering fuel assembly inventories. KSNP-1000 core was modeled in SRAC code package using PIJ module for fuel pins and fuel assemblies’ lattices and CITATION module for fuel assemblies’ pattern in a quarter core symmetry. Optimization problem adaptation using QEA was made by presenting 52 fuel assemblies in Q-bit individuals with the length of 8 Q-bits. Q-bits were converted to corresponding bit values and then given weight which would be used as consideration to optimize the pattern. The optimization program was coupled with the SRAC neutronic code to obtain the values of effective multiplication factor (keff) and power peaking factor (PPF). The optimization was calculated based on fitness value which was a function of keff and PPF values with the particular weight factor. Using a rotation gate angle of Δθ=0.02π and a weight factor of w=0,041, fuel loading pattern optimization was done on 360 days burnup level. The optimization resulted in keff and PPF value of 1.11233 and 1.944 respectively. By calculating keff value on various burnup levels for the chosen core loading pattern, reactor cycle length obtained was 659 days with PPF at BOC was 2.19. Compared to the standard KSNP-1000 core which had 560 days of cycle length, the optimized core configuration increased 17.67% in cycle length.

scholarly journals Minimizing Power Peaking Factor of BEAVRS-based Reactor Using Polar Bear Optimization Algorithms

2021 ◽  
Vol 927 (1) ◽  
pp. 012004
Author(s):  
Amila Amatullah ◽  
Alexander Agung ◽  
Agus Arif
Keyword(s):  
Cycle Length ◽  
Polar Bear ◽  
Operating Time ◽  
Complex Problem ◽  
Fuel Loading ◽  
Loading Pattern ◽  
Maximum Value ◽  
Fuel Assemblies ◽  
The One ◽  
Optimum Solution

Abstract Fuel loading pattern optimization is a complex problem because there are so many possibilities for combinatorial solutions, and it will take time to try it one by one. Therefore, the Polar Bear Optimization Algorithm was applied to find an optimum PWR loading pattern based on BEAVRS. The desired new fuel loading pattern is the one that has the minimum Power Peaking Factor (PPF) value without compromising the operating time. Operating time is proportional to the multiplication factor (k eff ). These parameters are usually contradictive with each other and will make it hard to find the optimum solution. The reactor was modelled with the Standard Reactor Analysis Code (SRAC) 2006. Fuel pins and fuel assemblies are modelled with the PIJ module for cell calculations. One-fourth symmetry was used with the CITATION X-Y module for core calculations. The optimization was done with 200 populations and 50 iterations. The PPF value for the selected solution should never exceed 2.0 in every burn-up step. Out of 28 solutions, the best optimal fuel loading pattern had a maximum value PPF of 1.458 and a k eff of 0.916 at day 760 of calculated time (corresponding to a cycle length of 479 days). Therefore, the maximum PPF value was 27.1% lower than the safety factor, and the same operating time as the standard loading pattern has been achieved.

scholarly journals Application of Evolutionary Simulated Annealing Method to Design a Small 200 MWt Reactor Core

2021 ◽  
Vol 10 (4) ◽  
pp. 16-23
Author(s):  
Tran Viet Phu ◽  
Tran Hoai Nam ◽  
Hoang Van Khanh
Keyword(s):  
Simulated Annealing ◽  
Power Plant ◽  
Nuclear Power ◽  
Cycle Length ◽  
Reactor Core ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Assemblies ◽  
Full Power ◽  
Annealing Method

This paper presents the application of an evolutionary simulated annealing (ESA) method to design a small 200 MWt reactor core. The core design is based on a reference ACPR50 reactor deployed in a floating nuclear power plant. The core consists of 37 typical 17x17 PWR fuel assemblies with three different U-235 enrichments of 4.45, 3.40 and 2.35 wt%. Core loading pattern (LP) has been optimized for obtaining the cycle length of 900 effective full power days, while minimizing the average U-235 enrichment and the radial power peaking factor. The optimization process was performed by coupling the ESA method with the COREBN module of the SRAC2006 system code.

Implementation of Erbia With ZrB2 in 17×17 Assemblies for a 4-Loop Core With a 24-Month Cycle

2002 ◽  
Author(s):  
James E. Platte ◽  
Ernesto Pitruzzella ◽  
Youssef Shatilla ◽  
Baard Johansen
Keyword(s):  
Fuel Assembly ◽  
Temperature Coefficient ◽  
Cycle Length ◽  
Zirconium Diboride ◽  
Fuel Rods ◽  
Fuel Assemblies ◽  
Burnable Absorber

There are many types of burnable absorbers currently used in power reactors. They are used to provide reactivity and power peaking control. Westinghouse reactors most commonly use Zirconium Diboride Integral Fuel Burnable Absorbers (ZrB2) while Combustion Engineering reactors most commonly use Erbia Integral Fuel Burnable Absorbers (Erbia) in Combustion Engineering reactors. This paper documents the study to determine the effect of placing Erbia and ZrB2 within a Westinghouse 17×17 fuel assembly, and the effect of these ZrB2/Erbia assemblies on the physics characteristics of a representative Westinghouse 4-loop, 24 month cycle length design. The study consisted first of producing optimal within-assembly burnable absorber configurations where ∼25% of the ZrB2-bearing fuel rods within an assembly were replaced with Erbia-bearing fuel rods. This ratio was selected in order to provide an effective balance between potential peaking factor improvements and the known Erbia disadvantage of increased residual absorber penalty compared with ZrB2. The optimal patterns were selected as the ones that most reduced the assembly-wise cumulative peak-to-average rod power during the depletion compared with existing all-ZrB2 BA configurations with the same BA rod quantity loading. The second part of this study consisted of substituting various quantities of these ZrB2/Erbia feed fuel assemblies in a representative Westinghouse 4-loop, 24 month cycle core design to study the effect on power peaking factors, moderator temperature coefficient (MTC), and cycle length.

scholarly journals Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Materials ◽  
2019 ◽  
Vol 12 (3) ◽  
pp. 494
Author(s):  
Alexander Vasiliev ◽  
Jose Herrero ◽  
Marco Pecchia ◽  
Dimitri Rochman ◽  
Hakim Ferroukhi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Design Parameters ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
National Cooperative ◽  
Criticality Safety

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

scholarly journals Adaptation of HELIOS Models for WWER-440 Fuel Assemblies to Application by the TRACE/PARCS Program

2019 ◽  
pp. 46-51
Author(s):  
I. Ovdiienko ◽  
O. Kuchyn ◽  
M. Ieremenko ◽  
P. Vlasenko
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Neutron Transport ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Safety Analysis ◽  
Neutron Transport Equation ◽  
Fuel Assemblies ◽  
Adaptation Procedure

The preparation of a few-group neutron cross-section library is an important step in implementation of the computer packages that are based on solution of the neutron transport equation in the few-group diffusion approximation into the safety analysis practices. The accuracy of modelling the physical neutron kinetic processes in the reactor core directly depends on the quality of few-group cross-section library. It is important to note that such cross-section library should be prepared in the format applied in the computer package and with use of a spectral code that models the fuel assembly quite adequately. The best option for preparing the few-group neutron crosssection library for the PARCS few-group diffusion code, which is being introduced into SSTC NRS safety analysis practices as a part of the TRACE/PARCS coupled neutron kinetic/thermal hydraulic package, is to adapt the previously developed and validated models of fuel assemblies for the HELIOS spectral program. The adaptation procedure for HELIOS models for WWER-440 including the fuel follower and transition part forming the input file structure required for correct work of the GenPMAXS program is presented. The approaches to the choice of reference states and branch parameters in the PARCS code format are presented. The results from correctness analysis of the adaptation of the HELIOS WWER-440 fuel assembly computer models are presented. The results are based on a comparative analysis of the fuel assembly multiplication properties obtained by the HELIOS model that was developed for preparation of the cross-section libraries for the DYN3D program (validated and widely used at SSTC NRS at present), and by the HELIOS model that was adapted for the GENPMAX program.

scholarly journals Method for Contactless Determination of the Height of A Fuel Assembly by Means of A 3-D Reconstruction of A Collinear Stereopair Images

2018 ◽  
Vol 7 (3.13) ◽  
pp. 51
Author(s):  
S Kravtsov ◽  
K Rumyantsev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Vision System ◽  
Reactor Core ◽  
System A ◽  
Fuel Assemblies ◽  
Nuclear Power Unit ◽  
Head Height ◽  
Set Of Points

A method for determining the head height of fuel assemblies in the reactor core of a nuclear power unit using a 3-D reconstruction of a stereopair of collinear images is considered. The method is based on the principle of statistical evaluation of the height of a set of points for a 3-D reconstruction of the contour of the head of the fuel assembly. To obtain a stereopair of images, it is suggested to use a collinear digital stereo-vision system. A model experiment was carried out. The results are compared with the known method for determining the height of the heads of fuel assemblies, based on an estimate of the height of the centers of gravity of the contours of fuel assembly heads. The proposed method shows a higher accuracy in solving the problem of determining the heights of fuel assembly heads in comparison with the known method.  

Water Confinement Effects in Response of Fuel Assembly to Faulted Condition Loads

2004 ◽  
Author(s):  
S. J. Shah ◽  
B. Brenneman ◽  
G. T. Williams ◽  
J. H. Strumpell
Keyword(s):  
Fuel Assembly ◽  
Reference Frame ◽  
Reactor Core ◽  
Axial Flow ◽  
Pressure Gradients ◽  
Baffle Plate ◽  
Horizontal Lift ◽  
Velocity Modeling ◽  
Fuel Assemblies ◽  
Large Flow

It has been established by other authors [1] that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect — the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to always reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers [2,3] that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than “ground”. The large flow-induced damping really acts in a relative reference frame rather than an absolute or inertial reference frame, and thus it becomes a flow-induced coupling between the fuel and baffles. This has a significant effect on the fuel assembly motions and tends to reduce the relative displacements and impact forces between fuel assemblies and baffle walls.

scholarly journals MULTIPLE CHOICES OF REACTOR CORE NUCLEAR DESIGN FOR ACP100’S APPLICATION IN DIFFERENT SCENARIOS

2021 ◽  
Vol 247 ◽  
pp. 19002
Author(s):  
Wang Liangzi ◽  
Ju Haitao ◽  
Li Qing ◽  
Qin Dong ◽  
Wang Lianjie ◽  
...  
Keyword(s):  
Cycle Length ◽  
Fuel Cycle ◽  
Safety Margin ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Safety Standards ◽  
Reactor Technology ◽  
Fuel Assemblies ◽  
Compact Design

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology; it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design; thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for daily boron adjustment and relative waste storage; choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate the core, through large amount of optimization, both choices achieve a 24-month fuel cycle length; the average discharge exposure of fuel assemblies reach about 40000MWd/tU, which is competitive among SMRs, especially for boron-free ones; and sufficient reactivity control ability and safety margin is validated to fully meet the reactor safety requirements.

scholarly journals ADVANCED HOMOGENIZATION METHODS FOR PRESSURIZED WATER REACTORS

2018 ◽  
Vol 19 ◽  
pp. 14 ◽  
Author(s):  
Pavel Suk
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Testing Method ◽  
Infinite Lattice ◽  
Fuel Assemblies ◽  
Macroscopic Cross Section ◽  
Water Reactors ◽  
Homogenization Methods

Macroscopic cross section generation is key part of core calculation. Commonly, the data are prepared independently without a knowledge of fuel loading pattern. The fuel assemblies are simulated in infinite lattice (with mirror boundary conditions). Rehomogenization method is based on combination of actual neutron flux in fuel assembly with macroscopic data from infinite lattice. Rehomogenization method was implemented into the macrocode Andrea and tested on a reference cases. Cases consist of fuel cases, cases with strong absorber, cases with absorption rods, or cases with reflector assemblies. Testing method is based on a comparisons of homogenized and rehomogenized macroscopic cross sections and later on a comparisons of relative power of each fuel assembly. Above that there is comparison of eigenvalue.

Confinement Dependency on Added Stiffness Effect on a Fuel Assembly Under Axial Flow

2017 ◽  
Author(s):  
G. Ricciardi
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Axial Flow ◽  
Dynamical Behaviour ◽  
Seismic Excitation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Spacer Grids ◽  
The Core ◽  
Fuel Assemblies

Safety measures are required to insure the drop of control rods and that the core is cooled when the fuel assemblies of a Pressurized Water Reactor (PWR) are subjected to a seismic excitation. A way to insure these two criteria is to prevent the spacer grids from buckling. The reactor core made of fuel assemblies is subjected to an axial water flow to cool the reactor. The flow modifies the dynamical behaviour of the fuel assemblies. Tests made on a real fuel assembly highlighted an added stiffness effect under axial flow. In previous studies simulations were compared to experiment involving by-passes significantly larger than the distance between two fuel assemblies in a PWR core. Thus, one could wonder if the observations made on a fuel assembly with large by-passes are representative of core geometry. Simulations using a fluid-structure model of the core to a seismic excitation have been proposed. A parametric study has been conducted to observe the effect of confinement on the added stiffness effect for several confinements and bulk velocities. Simulations showed that the added stiffness reaches a maximum for a confinement around 20 mm, and that the added stiffness should be negligible in a real core configuration.

Sign in / Sign up

Export Citation Format

Share Document