Replacement Zircaloy for Silicon Carbide as Fuel Cladding Material in PWR – A Neutronic Evaluation

2015 ◽  
Vol 1769 ◽  
Author(s):  
Rochkhudson B. de Faria ◽  
Felipe Torres ◽  
Fabiana B. A. Monteiro ◽  
Claubia Pereira
Keyword(s):  
Silicon Carbide ◽  
Nuclear Reactors ◽  
Temperature Tolerance ◽  
Multiplication Factor ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
High Temperature Tolerance ◽  
Literature Reviews ◽  
Cladding Material

ABSTRACTSilicon carbide (SiC) has a potential to replacement zircaloy as fuel cladding material due to its high temperature tolerance, chemical stability and low neutron affinity. These characteristics may improve the economic and safety of nuclear reactors. Previous work has examined the possible use of SiC as a fuel cladding material in a PWR (Pressurized Water Reactor) environment. However, the advantage thermo mechanical and neutronic analysis replacement zircaloy cladding is not clear. Literature reviews has been done to predict the thermo mechanical behavior of SiC in high temperatures. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This codes system is widely accepted and used worldwide for safety analysis and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR. It was used the CSAS6 and TRITON modules. The goals are to evaluate the behavior of the infinite multiplication factor (kinf) and neutron flux using SiC as a fuel cladding material.

scholarly journals Effect Of -Quenching On Oxidation Resistance Of Zirconium Alloyzrnbmoge For Fuel Cladding Material

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
B. Bandriyana
Keyword(s):  
Oxidation Resistance ◽  
Growth Characteristic ◽  
Optical Microscope ◽  
Magnetic Suspension ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Corrosion Properties ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Cladding Material

<p>Effect of b-quenching of Zr-2.5Nb-0.5Mo-0.1Ge alloy used for advanced fuel cladding material of Pressurized Water Reactor (PWR) was investigated. The aim of this research isto improve the mechanical and corrosion properties through modificationof the alloy with regard to high reactor burn up. The quenching process was conducted by heating the sample at temperature of 950 <sup>o</sup>Cand soaking 2.5 hours,followed by quenching in water at room temperature and then continued with annealing process at 500 and 600<sup>o</sup>C. The change of hardness and oxidation resistance were characterized using optical microscope and scanning electron microscope (SEM). The effect on the oxidation resistance was investigated by the high temperature oxidation test using the MSB (Magnetic Suspension Balance) at 700 <sup>o</sup>C for 5 hours. The hardness increased from 217 VHN to 265 VHN after quenching due to grain refinement and precipitation hardening. The oxidation rate followed the typical parabolic growth characteristic. The formation of thin layer was considered to be a stable oxide ZrO<sub>2</sub>that influenced the oxidation characteristic and increasing the hardness of the alloy.</p>

PWR Fuel Element Neutronic Analysis with Burnable Poison Rods Using Zircaloy and Hi-Nicalon Type S Claddings

2016 ◽  
Vol 1814 ◽  
Author(s):  
R. B. de Faria ◽  
J. G. Mantecón ◽  
A. R. Hamers ◽  
A. L. Costa ◽  
A. Fortini ◽  
...  
Keyword(s):  
Silicon Carbide ◽  
Fuel Element ◽  
Temperature Tolerance ◽  
Burnable Poison ◽  
High Temperature Tolerance ◽  
Nuclear Plants ◽  
Reactivity Coefficients ◽  
Great Relevance ◽  
The Impact ◽  
Cladding Material

ABSTRACTThe alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of rising the burning and maintaining the safety of nuclear plants, is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and a low absorption cross-section for thermal neutrons. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with burnable poison variations, the impact on multiplication factor and reactivity coefficients to both claddings: zircaloy and silicon carbide. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. This code system is widely accepted and used worldwide for safety analysis, and criticality of nuclear reactors has been utilized to model a typical fuel element of a PWR.

High Temperature Oxidation of the Hot Rolled ZrNbMoGe Alloy for Fuel Cladding Material of Pressurized Water Reactor

2015 ◽  
Vol 1123 ◽  
pp. 356-359 ◽  
Author(s):  
Bandriyana ◽  
A.K. Rivai ◽  
J.H. Prajitno ◽  
A. Dimyati
Keyword(s):  
High Temperature ◽  
Oxide Layer ◽  
High Temperature Oxidation ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Water Reactor ◽  
Hot Rolled ◽  
Cladding Material

The Zr-2.5Nb-0.5Mo-0.1Ge alloy were developed for application as fuel cladding material in an advanced Pressurized Water Reactor (PWR) with higher burn-up at higher service temperature. Oxidation behaviour of the alloy at high temperature similar to condition related to the operating with Lost of Cooling Accident (LOCA) conditions was investigated. The sample ingot was synthesized in an arc melting furnace followed by hot rolling at 850 °C down to 3 mm of thickness. Each one alloy ingot and one alloy sheet were subjected to the high temperature oxidation test in the Magnetic Suspension Balance (MSB) at 500 °C. Light optical microscope (LIOM) and Scanning Electron Microscope (SEM) equipped with X-ray Diffraction Spectroscopy (EDS) were used to characterize the microstructure of the oxide layer. The Vickers hardness tester was used to evaluate the hardness of the alloy matrix before and after oxidation processes. The results concluded that both samples showed oxidation rate characteristic which follows the parabolic phenomenon. However the hot rolled sample had lower rate. The oxide layer was indicated as ZrO2 .

scholarly journals Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

2015 ◽  
Author(s):  
Sharon M. Robinson ◽  
Marc Rhea Chattin ◽  
Joseph Giaquinto ◽  
Robert Thomas Jubin
Keyword(s):  
Reactor Fuel ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Tritium Content ◽  
Pressurized Water ◽  
Water Reactor

Study on the Small Pressurized Water Reactor Based on Fully Ceramic Microencapsulated Fuel

2021 ◽  
Author(s):  
Jin Feng Huang
Keyword(s):  
Thermal Conductivity ◽  
Silicon Carbide ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Carbide Matrix ◽  
Silicon Carbide Matrix

Abstract After Fukushima nuclear power plant disaster, the efforts to overcome these defects of PWRs were carried out, such as replacing the cladding and fuel materials. One of these feasible efforts is using Fully Ceramic Microencapsulated (FCM) fuel replacement traditional UO2 pellets fuel into PWR. The FCM fuels are composed of Tri-structural-isotropic (TRISO) particles embedded in silicon carbide matrix. The TRISO fuel can hold its containment integrity and without fission production releases under the design temperature limit of 1600 °C. Furthermore, the silicon carbide matrix will benefit the thermal conductivity, radiation damage resistance, environmental stability and proliferation resistance. Consequently, the safety of the reactor could be significantly improved with FCM fuel instead of the conventional UO2 pellet fuel in PWR. We also analyzed the temperature distribution for the FCM fuel compared the traditional UO2 pellets, the calculation indicated that the centerline temperature is lower than UO2 pellets due to FCM higher thermal conductivity. The calculation demonstrated that, utilizing FCM replacement of conventional UO2 fuel rod is feasible and more security in a small pressurized water reactor. In this paper, a small pressurized water reactor utilizing FCM fuel is considered. A 17 × 17 fuel assemblies with Zircalloy cladding was applied in conceptual design through a preliminary neutronics and thermal hydraulics analysis. The reactor physics is accomplished, including the refuel cycle length, the effective multiplication factor, power distribution analysis being discussed. The Soluble Boron Free (SBF) concepts are introduced in small PWR, as a result, it makes the nuclear power plant more simpler and economical. FCM fuel loading has a very high excess reactivity at the beginning of reactor core life, and it is important to flat reactivity curve during operation, or to minimize excess reactivity during the core life. Consequently, conventional burnable poison configurations were introduced to suppress excess reactivity control at beginning of cycle.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

High-temperature tolerance of the piezoresistive effect in p-4H-SiC for harsh environment sensing

2018 ◽  
Vol 6 (32) ◽  
pp. 8613-8617 ◽  
Author(s):  
Tuan-Khoa Nguyen ◽  
Hoang-Phuong Phan ◽  
Toan Dinh ◽  
Abu Riduan Md Foisal ◽  
Nam-Trung Nguyen ◽  
...  
Keyword(s):  
Silicon Carbide ◽  
High Temperature ◽  
High Stability ◽  
Temperature Tolerance ◽  
Harsh Environments ◽  
Harsh Environment ◽  
Piezoresistive Effect ◽  
High Temperature Tolerance ◽  
Chemical Inertness ◽  
Silicon Based

4H-silicon carbide based sensors are promising candidates for replacing prevalent silicon-based counterparts in harsh environments owing to their superior chemical inertness, high stability and reliability.

scholarly journals Completion of a Chlorination Test Using 250 grams of High-Burnup Used Fuel Cladding from a North Anna Pressurized Water Reactor

2018 ◽  
Author(s):  
Emory D. Collins ◽  
Tom D. Hylton ◽  
Guillermo Daniel DelCul ◽  
Barry B. Spencer ◽  
Rodney Dale Hunt ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Burnup

Research on Algorithm of Sump Level Operator Assisted Support Program for PWR Nuclear Power Plant

2018 ◽  
Author(s):  
Liu Qiaofen ◽  
Xiao Sanping ◽  
Liu Yu ◽  
Liu Xichao ◽  
Jiang Xulun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactors ◽  
Support Program ◽  
Leakage Flow ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Calculation Results ◽  
Leak Before Break

Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.

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