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 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.

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 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.

scholarly journals Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

2022 ◽  
Vol 2022 ◽  
pp. 1-13
Author(s):  
Izza Shahid ◽  
Nadeem Shaukat ◽  
Amjad Ali ◽  
Meer Bacha ◽  
Ammar Ahmad ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Distribution ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Pressurized Water ◽  
Control Rod ◽  
Operational Conditions ◽  
The Core ◽  
Safety Parameters

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

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.

Prediction of safety parameters of pressurized water reactor based on feature fusion neural network

2022 ◽  
Vol 166 ◽  
pp. 108803
Author(s):  
Yinghao Chen ◽  
Dongdong Wang ◽  
Cao Kai ◽  
Cuijie Pan ◽  
Yayun Yu ◽  
...  
Keyword(s):  
Neural Network ◽  
Feature Fusion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Parameters

A Neutron Amplifier: Prospects for Reactor-Based Waste Transmutation

2004 ◽  
Author(s):  
Anatoly Blanovsky
Keyword(s):  
Neutron Source ◽  
Power Distribution ◽  
Wave Model ◽  
System Size ◽  
Reverse Direction ◽  
Fast Neutrons ◽  
Beam Power ◽  
Primary System ◽  
Pressurized Water ◽  
Wide Range

A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 1015 n/s (0.7MW D-T or 2.5MW electron beam power) is needed to control the HFSR that produces 300MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

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