Temperature adjusted cross section libraries used for criticality calculations

Abstract The change in the temperature of the nuclear reactor components (fuel, moderator, coolant, and structural materials) is considered to be a significant source of reactivity variation. This change must be taken in account during criticality calculations for safety analysis of the reactor. Hence, the exact representation of temperature in the calculations is very important. In this paper, two PWR assemblies are simulated, solid 16 ⨯ 16 and annular 12 ⨯ 12 fuel assemblies. The infinite multiplication factor and its temperature dependent parameters are calculated for both fuel assemblies. Adjusted temperature dependent libraries are created using makxsf code to exactly represent the different temperature values used in the calculations. It is shown that the results obtained using adjusted cross section libraries are more reliable. The two fuel assembly types follow the same behavior despite the differences in their geometrical configuration. The introduction of annular fuel has a very small effect on the investigated neutronic parameters because the moderator to fuel ratio is preserved.

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

Nuclear and Radiation Safety ◽

10.32918/nrs.2019.1(81).08 ◽

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.

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Features of Burnup Calculations in the SERPENT 2 Software Package Using the Example of a BWR Fuel Assembly

KnE Engineering ◽

10.18502/keg.v3i3.1618 ◽

2018 ◽

Vol 3 (3) ◽

pp. 182

Author(s):

Pham Bui Dinh Lam ◽

Kolesov V.V.

Keyword(s):

Fuel Assembly ◽

Software Package ◽

Multiplication Factor ◽

Spent Fuel ◽

Detailed Model ◽

Fuel Assemblies ◽

Neutron Multiplication ◽

Phase Iiib ◽

Nuclide Composition

In this paper, we used the data from “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Nuclide Composition and Neutron Multiplication Factor of BWR Spent Fuel Assembly” ([1]) for the verification of the SERPENT 2 code. The results obtained which were compared with the results of other authors, which were also given in “OECD/NEA Burnup Credit Criticality Benchmark Phase IIIB: Burnup Calculations of BWR Fuel Assemblies for Storage and Transport” ([2]). Investigations of the influence of the detailed model of pins and pins with gadolinium, as well as various methods of burn-up calculations were also carried out.

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PWR Fuel Assembly Modal Testing and Analysis

Flow-Induced Vibration ◽

10.1115/pvp2003-2084 ◽

2003 ◽

Cited By ~ 1

Author(s):

Bruno Collard ◽

Ste´phane Pisapia ◽

Sergio Bellizzi ◽

Fre´de´ric Witters

Keyword(s):

Fuel Assembly ◽

Flow Rate ◽

Nuclear Reactor ◽

Experimental Tests ◽

Mechanical Tests ◽

Full Scale ◽

Modal Testing ◽

Pressurized Water ◽

Fuel Assemblies ◽

Non Linear

Pressurized Water Reactor (PWR) seismic or Lost Of Coolant Accident (LOCA) loads could result in impacts between nuclear fuel assemblies or between fuel assemblies and the core baffles. Forces generated during these shocks are often the basis for the determination of the maximum loads and of the spacer grid and fuel rod design. The knowledge of the fuel assembly kinematics is essential to compute these maximum loads, and this requires experimental tests. Our study aims at characterizing the behavior of a full-scale fuel assembly subjected to various excitations. The effect of the assembly environment (air, still water and water under flow) is studied. The French Nuclear Reactor Directorate experimental facility HERMES T allows hydraulic and mechanical testing of full-scale fuel assemblies. It is designed for flow rate up to 1200 m3/h and temperature up to 170°C. Specific excitation devices allow mechanical tests with amplitudes of motion up to 20 mm. Laser vibrometry, displacement transducers and tracking camera apparatus measure the fuel assembly displacement. To identify this Multi Degree Of Freedom (MDOF) system (assembly or assembly + fluid), two dependent problems have to be addressed: the linear or non-linear model selection, and the estimation of the corresponding parameters. Under different environments and excitation types, it is shown that the mechanical system is strongly non-linear. The damping term, essentially fluid, increases with flow rate and with motion amplitude, while the stiffness decreases with amplitude. The main results, the measuring and identification methods and the extrapolation to the reactor thermohydraulic conditions are presented and discussed.

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Validation of SCALE Model of VVR-M Reactor

Nuclear Power and the Environment ◽

10.31717/2311-8253.20.3.5 ◽

2020 ◽

Vol 18 ◽

pp. 42-47

Author(s):

V. I. Borysenko ◽

◽

V. V. Goranchuk ◽

Keyword(s):

Physical Model ◽

Nuclear Reactor ◽

Physical Characteristics ◽

Multiplication Factor ◽

The Core ◽

Fuel Assemblies ◽

Neutron Multiplication ◽

Research Nuclear Reactor ◽

Effective Neutron ◽

Calculation Code

The peculiarities of development of neutron-physical model of the VVR-M research nuclear reactor in the SCALE calculation code are considered in the article. Models of separate core elements, which influence neutron-physical characteristics of VVR-M, have been developed. Simulation was performed using the CSAS6 control module. Validation of the VVR-M neutron-physical model, built in the SCALE calculation code, has been carried out by comparing the calculated value of the effective neutron multiplication factor with the critical reactor state at the beginning of seven fuel loads with the number of fuel assemblies in the core from 72 to 129. The model is developmed to determine the effective neutron multiplication factor in the reactor, as well as other neutron-physical characteristics, such as neutron spectrum, neutron flux density in various cells of the reactor. Thus, it is possible to conduct numerical experiments to determine the most optimal locations of research channels in the core of the VVR-M, to conduct physical experiments on the irradiation of the research samples, detectors, structural materials, etc. In the article, the simplifications accepted at construction of neutron-physical model of research nuclear reactor VVR-M in SCALE calculation code are presented. The main elements of the model are described: fuel assemblies, beryllium displacer, control rods.

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Analysis of coolant flow at the outlet of the physical model of the fuel assembly II. of the nuclear reactor VVER 440 / V213

MATEC Web of Conferences ◽

10.1051/matecconf/202032801010 ◽

2020 ◽

Vol 328 ◽

pp. 01010

Author(s):

Peter Mlynár ◽

František Világi ◽

Zdenko Závodný ◽

František Urban ◽

František Ridzoň

Keyword(s):

Velocity Profile ◽

Fuel Assembly ◽

Physical Model ◽

Nuclear Reactor ◽

Coolant Flow ◽

Coolant Temperature ◽

Cfd Simulations ◽

Central Tube ◽

Fuel Assemblies ◽

The Mean

To safely and efficiently load the fuel assemblies of the VVER 440 / V 213 nuclear reactor, the relation between the temperature of the coolant at the outlet of the fuel assembly, measured by a thermocouple in the assembly’s axis, and the mean coolant temperature, present in the plane of the thermocouple, must be analysed. Based on the analysis of the coolant flow at the output of the physical model of the fuel assembly I. [1] and published CFD simulations [2,3,4] it was shown, that a special attention has to be paid to the influence of the water flow in the central tube on the temperature and velocity profile of the coolant at the thermocouple’s plane in the fuel assembly. For this reason, an experimental device with a physical model of the fuel assembly II. of the nuclear reactor VVER 440 / V 213 was designed, manufactured, and operated at the Faculty of Mechanical Engineering STU in Bratislava.

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Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident

Materials ◽

10.3390/ma14081818 ◽

2021 ◽

Vol 14 (8) ◽

pp. 1818

Author(s):

Di-Si Wang ◽

Bo Liu ◽

Sheng Yang ◽

Bin Xi ◽

Long Gu ◽

...

Keyword(s):

Fuel Assembly ◽

Steam Generator ◽

Metal Flow ◽

Maximum Temperature ◽

Steam Generator Tube ◽

Driven System ◽

Fuel Assemblies ◽

Accelerator Driven System ◽

The Impact ◽

Steam Bubble

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

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Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

Volume 9: Student Paper Competition ◽

10.1115/icone25-66744 ◽

2017 ◽

Cited By ~ 1

Author(s):

Zhixiong Tan ◽

Jiejin Cai

Keyword(s):

Silicon Carbide ◽

Fuel Assembly ◽

Power Distribution ◽

Safety Margin ◽

Severe Accident ◽

Pressurized Water Reactors ◽

Fuel Pellet ◽

Pressurized Water ◽

Neutronic Parameters ◽

Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

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Application of CFD Model for Inlet Flow Region of 17×17 Fuel Assembly

Volume 4: Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition ◽

10.1115/icone14-89503 ◽

2006 ◽

Cited By ~ 1

Author(s):

Milorad B. Dzodzo ◽

Bin Liu ◽

Pablo R. Rubiolo ◽

Zeses E. Karoutas ◽

Michael Y. Young

Keyword(s):

Fuel Assembly ◽

Fluid Dynamic ◽

Flow Distribution ◽

Flow Region ◽

Fretting Wear ◽

Jet Flows ◽

Cfd Model ◽

Lateral Velocity ◽

Fuel Assemblies ◽

Inlet Conditions

A numerical investigation was performed to study the variation in axial and lateral velocity profiles occurring downstream of the inlet nozzle of a typical Westinghouse 17×17 PWR fuel assembly. A Computational Fluid Dynamic (CFD) model was developed with commercial CFD software. The model comprised the lower region of the fuel assembly, including: the Debris Filter Bottom Nozzle (DFBN), P-grid, Bottom Inconel grid, one and half grid span, as well as the lower core plate hole. The purpose of the study was to obtain insight into the flow redistribution resulting from the interaction of the jet arising from the lower core plate hole and the fuel assembly structure. In particular the axial and lateral velocities before and after the nozzle were studied. The results, axial and lateral velocity contours, streamlines and maximum axial and lateral velocity distributions at various elevations are presented and discussed in relation to the potential risk of high turbulent excitation over the rod and the resulting rod-to-grid fretting-wear damage. The CFD model results indicated that the large jet flows from the lower core plate are effectively dissipated by DFBN nozzle and the grids components of the fuel assembly. The breakup of the large jets in the DFBN and the lower grids helps to reduce the steep velocity gradients and thus the rod vibration and fretting-wear risk in the lower part of the fuel assembly. The presented CFD model is one step towards developing advanced tools that can be used to confirm and evaluate the effect of complex PWR structures on flow distribution. In the future the presented model could be integrated in a larger CFD model involving several fuel assemblies for evaluating the lateral velocities generated due to the non-uniform inlet conditions into the various fuel assemblies.

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