Cost Saving When Using Enhanced Conductivity Nuclear Fuel Containing BeO in WWER-1000 Reactors

2014 ◽  
Author(s):  
Tomáš Zahrádka ◽  
Radek Škoda
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Fuel ◽  
Fuel Pellet ◽  
Internal Stresses ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Adverse Conditions ◽  
Advantages And Disadvantages ◽  
Fuel Pin ◽  
Water Reactors

Current pressurized water reactors utilize sintered UO2 that has a number of advantages and disadvantages. Uranium Dioxide’s low thermal conductivity results in a large thermal gradient within the fuel pellet corresponding to higher centerline temperatures compared to other potential fuel forms. These gradients result in non-uniform thermal expansion leading to large internal stresses resulting in cracking of the pellet and fuel-clad interaction, which can lead to loss of the integrity of the fuel pin. Higher fuel temperatures also increase the release of fission gases. Fuels with higher thermal conductivity may alleviate or reduce the severity of these adverse conditions. It is shown that higher thermal conductivity can be obtained by adding BeO to the basic UO2 matrix. This paper focuses on WWER1000 hexagonal fuel geometry. Improvements when using 10% of BeO, as proposed in this paper, reduce the centerline nuclear fuel temperature by 234°C and improve the fuel economy while reducing its cost by 7%. The study was done for NPP Temelín which has two units WWER1000/320.

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

2017 ◽  
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.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

2002 ◽  
Author(s):  
Hyun-Jong Joe ◽  
Barclay G. Jones
Keyword(s):  
Negative Impact ◽  
Nucleate Boiling ◽  
Water Radiolysis ◽  
Trace Impurities ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Fuel Pin ◽  
Negative Effect ◽  
Water Reactors ◽  
Coolant System

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.

scholarly journals Thermal Conductivity Measurement of Zr-ZrO<sub>2</sub> Simulated Inert Matrix Nuclear Fuel Pellet

2013 ◽  
Vol 03 (02) ◽  
pp. 46-50 ◽  
Author(s):  
Dong-Joo Kim ◽  
Young Woo Rhee ◽  
Jong Hun Kim ◽  
Jang Soo Oh ◽  
Keon Sik Kim ◽  
...  
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Fuel ◽  
Conductivity Measurement ◽  
Thermal Conductivity Measurement ◽  
Fuel Pellet ◽  
Inert Matrix

scholarly journals Effect of compaction pressure on the thermal conductivity of UO2-BeO-Gd2O3 pellets

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Denise Das Mercês Camarano ◽  
Ana Maria Matildes Dos Santos ◽  
Fábio Abud Mansur
Keyword(s):  
Thermal Conductivity ◽  
Thermal Diffusivity ◽  
Compaction Pressure ◽  
Beryllium Oxide ◽  
Laser Flash ◽  
Immersion Method ◽  
Pressurized Water ◽  
Fuel Cycles ◽  
Fuel Pellets ◽  
Water Reactors

The (U,Gd)O2 fuels are used in pressurized water reactors (PWR) to control the neutron population in the reactor during the early life with the purpose of extended fuel cycles and higher target burnups. Nevertheless, the incorporation of Gd2O3 in the UO2 fuel decreases the thermal conductivity, leading to premature fuel degradation. This is the reason for the addition of beryllium oxide (BeO), which has a high thermal conductivity and is chemically compatible with UO2. Pellets were obtained from powder mixtures of the UO2, Gd2O3 and BeO, being the oxide contents of the beryllium equal to 2 and 3wt%, and the gadolinium fixed at 6wt%. The pellets were compacted at 400, 500, 600, and 700 MPa and sintering under hydrogen reducing atmosphere. The purpose of this study was to investigate the effect of BeO, Gd2O3 and compaction pressure on the thermal conductivity of the UO2 pellets. The thermal diffusivity and conductivity of the pellets were determined from 298 K to 773 K and the results obtained were compared to UO2 fuel pellets. The thermal diffusivity was determined by Laser Flash and Thermal Quadrupole methods and the thermal conductivity was calculated from the product of thermal diffusivity, the specific heat capacity and density. The sintered density of the pellets was determined by the xylol penetration and immersion method. The results showed an increase in the thermal conductivity of the pellets with additions of BeO and with the compaction pressure compared to the values obtained with UO2 pellets.

scholarly journals Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

2021 ◽  
Vol 11 (1) ◽  
pp. 9-15
Author(s):  
Van Khanh Hoang ◽  
Vinh Thanh Tran ◽  
Dinh Hung Cao ◽  
Viet Ha Pham Nhu
Keyword(s):  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Model ◽  
Reactivity Coefficients ◽  
Fuel Pin ◽  
Safety Parameters ◽  
Long Life

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

Distributions of radionuclides on and in spent nuclear fuel claddings of pressurized water reactors

1990 ◽  
Vol 174 (1) ◽  
pp. 45-52 ◽  
Author(s):  
T. Hirabayashi ◽  
T. Sato ◽  
C. Sagawa ◽  
N.M. Masaki ◽  
M. Saeki ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors

Main Evolutions of the RCC-C Design and Construction Code for Fuel Assemblies Since 2015

2018 ◽  
Author(s):  
Marc Ton-That ◽  
Christine Vauglin ◽  
Gilbert Trillon
Keyword(s):  
Nuclear Fuel ◽  
Electric Utility ◽  
Original Text ◽  
Nuclear Operator ◽  
Pressurized Water ◽  
Development Organization ◽  
Fuel Assemblies ◽  
Core Components ◽  
Mechanical Components ◽  
Water Reactors

AFCEN is a French Standard Development Organization which publishes codes for design, construction and in-service inspection rules for Pressurized Water Reactors. The fields covered by theses codes are mechanical components, in-service surveillance of mechanical components, electrical equipments, nuclear fuel, civil works and fire protection. AFCEN was initially founded by electric utility EDF and nuclear steam supply system manufacturer FRAMATOME. AFCEN has more than 60 institutional members, representing more than 650 experts who contribute to the development and continuous improvement of codes. The RCC-C code, which is dedicated to PWR fuel assemblies and associated core components, set forth generic requirements to be fulfilled by the suppliers and by the manufacturers for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. The RCC-C is intended to be used in the frame of contractual relations between a customer (nuclear operator) and a nuclear fuel supplier. The first edition was published in 1981. Over the years, many changes have been made to the original text but the structure hasn’t been much modified. Because of this, the text was becoming less coherent for the users and was lacking also minimal explanations. A redesign of the code was scheduled for the 2015 edition to address those problems. With the involvement of fuel vendors FRAMATOME, WESTINGHOUSE, and French nuclear operator EDF, the text was restructured and clarified. New requirements were implemented and the set of both design and manufacturing rules was strengthened to reflect fuel vendors’ practices and operator expectations. This article explains the main modifications that were implemented since the 2015 edition, and also outlines the prospects for future changes taking into account the latest regulatory requirements and evolutions of the industrial practices.

Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR

2017 ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chunyu ◽  
Li Aolin ◽  
Yuan Cenxi
Keyword(s):  
Mechanical Performance ◽  
Three Dimensional ◽  
Fission Products ◽  
General Purpose ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Three Dimensional Modeling ◽  
Water Reactors

The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.

scholarly journals Assessment of the Radiotoxicity of Spent Nuclear Fuel from a Fleet of PWR Reactors

Energies ◽  
2021 ◽  
Vol 14 (11) ◽  
pp. 3094
Author(s):  
Mikołaj Oettingen
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Hybrid Energy ◽  
Cycle Simulation ◽  
Water Reactors

The paper presents the methodology for the estimation of the long-term actinides radiotoxicity and isotopic composition of spent nuclear fuel from a fleet of Pressurized Water Reactors (PWR). The methodology was developed using three independent numerical tools: the Spent Fuel Isotopic Composition database, the Nuclear Fuel Cycle Simulation System and the Monte Carlo Continuous Energy Burnup Code. The validation of spent fuel isotopic compositions obtained in the numerical modeling was performed using the available experimental data. A nuclear power embarking country benchmark was implemented for the verification and testing of the methodology. The obtained radiotoxicity reaches the reference levels at about 1.3 × 105 years, which is common for the PWR spent nuclear fuel. The presented methodology may be incorporated into a more versatile numerical tool for the modeling of hybrid energy systems.

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