scholarly journals Investigation of the Safety Features of Advanced PWR Assembly Using SiC, Zr, FeCrAl and SS-310 As Cladding Materials

2020 ◽  
Author(s):  
Sayed Mustafa
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Delayed Neutron ◽  
Strong Trend ◽  
Geometric Conditions ◽  
Reactivity Coefficients ◽  
Safety Parameters ◽  
Reactivity Coefficient ◽  
Safety Features

Abstract In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX code as cladding materials in advanced PWR assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 µm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

scholarly journals Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Sayed. Saeed. Mustafa
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Delayed Neutron ◽  
Pressurized Water ◽  
Strong Trend ◽  
Geometric Conditions ◽  
Reactivity Coefficients ◽  
Extended Code ◽  
Safety Parameters

AbstractIn this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

scholarly journals Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

2017 ◽  
Vol 2017 ◽  
pp. 1-12 ◽  
Author(s):  
Shengli Chen ◽  
Cenxi Yuan
Keyword(s):  
Power Distribution ◽  
Current System ◽  
Negative Temperature ◽  
New Combination ◽  
Fuel Cladding ◽  
Fuel Temperature ◽  
Service Lifetime ◽  
Reactivity Coefficients ◽  
Reactivity Coefficient ◽  
Control Rods

Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

scholarly journals A study on the core loading pattern of the VVER-1200/V491

2021 ◽  
Vol 7 (1) ◽  
pp. 21-27
Author(s):  
Vinh Thanh Tran ◽  
Viet Phu Tran ◽  
Thi Dung Nguyen
Keyword(s):  
Power Distribution ◽  
Nuclear Power ◽  
Safety Standards ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Burn ◽  
Loading Patterns ◽  
Safety Parameters ◽  
Safety Features ◽  
Effective Multiplication

The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant.However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER- 1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel andmoderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR.

scholarly journals ANALYSIS OF REACTIVITY COEFFICIENT CHANGE DUE TO BURN UP IN AP1000 REACTOR CORE USING NODAL3

2017 ◽  
Vol 19 (3) ◽  
pp. 131
Author(s):  
Iman Kuntoro ◽  
Surian Pinem ◽  
Tagor Malem Sembiring
Keyword(s):  
Reactor Core ◽  
Core Material ◽  
Inherent Safety ◽  
Temperature Coefficients ◽  
Fuel Burn ◽  
Reactivity Coefficients ◽  
Safety Parameters ◽  
Safety Parameter ◽  
Moderator Density ◽  
Reactivity Coefficient

One of the important things in reactor safety is the value of inherent safety parameter namely reactivity coefficient. These inherent safety parameters are fuel and moderator temperature coefficients of reactivity.  The objective of the study is to obtain the change of those reactivity coefficients as a function of fuel burn up during the cycle operation of AP 1000 reactor core. Fuel and moderator temperature coefficients of reactivity and in addition moderator density coefficient of reactivity were calculated using SRAC 2006 and NODAL3 computer codes. Cross section generation of all core material was done by SRAC 2006 Code. The calculation of core reactivity as a function of temperature and burn up were carried out using NODAL3 Code. The results show that all reactivity coefficients of AP 1000 reactor core are always negative during the operation cycles and the values are in a good agreement to the design. It can be concluded that the AP 1000 core has a good inherent safety of its fuelKeywords: reactivity coefficient, burn up, AP1000, NODAL3. ANALISIS PERUBAHAN KOEFISIEN REAKTIVITAS AKIBAT FRAKSI BAKAR TERAS REAKTOR AP1000 MENGGUNAKAN NODAL3.  Salah satu hal yang sangat penting dalam analisis kecelakaan pada reactor daya adalah koefisien reaktivitas untuk mengontrol daya reaktor. Penelitian ini bertujuan menentukan koefisien reaktivitas akibat perubahan fraksi bakar pada reaktor AP1000. Koefisien reaktivitas yang akan dihitung adalah koefisien reaktivitas bahan bakar dan moderator yang sering disebut inherent factor. Selain itu juga akan dihitung koefisien konsentrasi boron dan kerapatan moderator.  Semua koefisien reaktivitas ini dihitung saat terjadi perubahan fraksi bakar untuk mempertimbangkan produk fisi dan konsumsi bahan bakar. Perhitungan neutronik teras reactor disimulasi dengan menggunakan program SRAC2006 dan NODAL3. Perhitungan tampang lintang seluruh perangkat bahan bakar dan batang kendali reaktor AP1000 dilakukan dengan program SRAC2006. Perhitungan parameter neutronik sebagai fungsi temperature dan fraksi bakar dilakukan menggunakan program NODAL3. Perhitungan koefisien reaktivitas ditentukan berdasarkan perbedaan nilai reaktivitas. Hasil perhitungan menunjukkan bahwa koefisien reaktivitas teras reaktor AP 1000 selalu berharga negative untuk sepanjang siklus operasinya dan mendekati harga desain. Kesimpulan yang dapat ditarik adalah bahwa teras AP 10000 mempunyai keselamatan melekat yang baik.Kata kunci:  koefisien reaktivitas, fraksi bakar, AP 1000, NODAL3.

Leaching of Used Candu Fuel: Results from a 19-Year Leach Test Under Oxidizing Conditions

1996 ◽  
Vol 465 ◽  
Author(s):  
S. Stroes-Gascoyne ◽  
L. H. Johnson ◽  
J. C. Tait ◽  
J. L. McConnell ◽  
R. J. Porth
Keyword(s):  
Grain Boundary ◽  
Cross Sections ◽  
Peak Power ◽  
Physical State ◽  
Fuel Pellet ◽  
Fuel Surface ◽  
Pellet Surface ◽  
Fuel Bundle ◽  
Dissolution Behaviour ◽  
Congruent Dissolution

ABSTRACTA fuel leaching experiment has been in progress since 1977 to study the dissolution behaviour of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH2O, ∼2 mg/L carbonate) and tapwater (∼50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of 137Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of 137Cs and 90Sr leached are slightly larger in tapwater than in DDH2O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH2O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH2O is not prevalent, and in tapwater appears to be limited to the outer %0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

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.

scholarly journals Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Tomasz Kozlowski ◽  
Hyun Chul Lee ◽  
Deokjung Lee
Keyword(s):  
Comparative Analysis ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Computer Code ◽  
Anisotropic Scattering ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Modeling Approach ◽  
Significant Difference

A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics

2013 ◽  
Author(s):  
A. A. Mishin ◽  
V. V. Galchenko
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Pellet ◽  
Coolant Temperature ◽  
Radial Temperature ◽  
Group Constant ◽  
Temperature Distributions ◽  
Active Core ◽  
Core Characteristics

The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.

Population-Based Algorithms for Optimization of the Reactive Power Distribution and Selection of the Cable Cross-Section in the Power-Supply Systems

2015 ◽  
Vol 792 ◽  
pp. 230-236 ◽  
Author(s):  
Vadim Manusov ◽  
Elena Tretyakova ◽  
Pavel Matrenin
Keyword(s):  
Cross Sections ◽  
Transmission Lines ◽  
Power Distribution ◽  
Reactive Power ◽  
Optimization Algorithms ◽  
Population Based ◽  
Optimal Arrangement ◽  
Power Sources ◽  
Active Power Losses ◽  
Stochastic Optimization Algorithms

The article is devoted to optimization of the reactive power sources arrangement in the industrial power grid to reduce the active power losses in transmission lines and select cross-sections of cable of the lines. To reduce the losses installing of additional reactive power sources in junctions of the grid are proposed. Thus, a problem of the optimal arrangement of the reactive power sources is appeared. This problem was solved by two stochastic optimization algorithms such as the Genetic and the Particle Swarm optimization algorithms. The cost-effectiveness of the suggested method is shown.

scholarly journals Challenges Faced While Designing Body in White

2021 ◽  
Vol 9 (9) ◽  
pp. 49-56
Author(s):  
Aditya Dhobale
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Cross Section ◽  
Cross Sections ◽  
Element Analysis ◽  
Computer Aided ◽  
Safety Parameters ◽  
Body In White ◽  
White Sheet ◽  
Metal Design

Abstract: Construction of Body in White (BiW) revolves around plenty of challenges. Ranging from BiW fixtures to curbing weight of Body in White sheet metal design. This paper discusses about all the design aspects in BiW manufacturing in automobile and confronting challenges that occurs. At present, lots of existing theories are being applied and efforts to improve the same are being made. This paper provides a path on how components can be developed and make necessary improvements. CAE (Computer Aided Engineering) tools have been used for FEA (Finite Element Analysis) and also an example of stress analysis of automotive chassis is given. An outcome depending on behaviour of loads acting on frame is drawn. The importance of hollow tubes, tubes of different- cross sections to counter weight and ease the designing of BiW frame have been proposed. This paper also provides insight on safety parameters with current construction of tubular frame chassis. Other solutions such as hybrid tubes, foam padding and plastic trim have been pointed out in this paper. Keywords: CAE, FEA, manufacturing, loads, tubes, cycle-time, cross-section.

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