scholarly journals Thorium based fuel loading pattern for PWR reactor

2019 ◽  
Vol 108 ◽  
pp. 01028 ◽  
Author(s):  
Mikolaj Oettingen
Keyword(s):  
Monte Carlo ◽  
Numerical Model ◽  
Pressurized Water Reactor ◽  
Fuel Management ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Continuous Energy ◽  
Numerical Tools

The Monte Carlo Continuous Energy Burnup Code was used for numerical simulations of the pressurized water reactor (PWR) with mixed U-Th fuel. The thorium fuel was introduce to the core of the Westinghouse 4-loop PWR in order to investigate the time evolutions of fissile 233U, 235U, 239Pu, 241Pu and of fertile 232Th and 238U. The core loading pattern was designed using the Whole Assembly Seed and Blanket (WASB) approach. The following calculations show the feasibility of the proposed incore fuel management patter using the available numerical tools, developed numerical model and methods.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

Kinetic Monte Carlo simulations of cascades in Fe alloys

2000 ◽  
Vol 650 ◽  
Author(s):  
C. Domain ◽  
C.S. Becquart ◽  
J.C. van Duysen
Keyword(s):  
Monte Carlo ◽  
Kinetic Monte Carlo ◽  
Md Simulations ◽  
Atom Probe ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Displacement Cascades ◽  
Cu Content ◽  
Primary Damage ◽  
Eam Potential

ABSTRACTThe Pressurized Water Reactor vessel steels are embrittled by neutron irradiation. Among the solute atoms, copper play an important role in the embrittlement and different Cu-rich defects have been experimentally observed to form. We have investigated by Kinetic Monte Carlo (KMC) on rigid lattices the evolution of the primary damage. Since the point defects created by the displacement cascades have very different kinetics, their evolution is tracked in two steps. In a first step, we have studied their recombination in the cascade region and the formation of interstitial clusters using “object diffusion”. The parameters of this model are based on MD simulations, or on first principles calculations. In a second part, we have investigated the subsequent evolution of the primary damage with a model based on a vacancy jump mechanism. These simulations which rely on an adapted EAM potential show the formation of copper rich defects. Some of the potential's predictions that played a key role in the model were checked by ab initio calculations. The defects obtained from these simulations, subsequent to the primary damage created by displacement cascades, exhibit similarities with the ones observed by atom probe. The influence of temperature and Cu content on the final damage was investigated.

Application of Monte Carlo and discrete ordinate coupling method to pressurized water reactor cavity radiation streaming calculation

2012 ◽  
Vol 24 (12) ◽  
pp. 2946-2950
Author(s):  
郑征 Zheng Zheng ◽  
吴宏春 Wu Hongchun ◽  
曹良志 Cao Liangzhi ◽  
郑友琦 Zheng Youqi ◽  
张宏博 Zhang Hongbo ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Coupling Method ◽  
Pressurized Water Reactor ◽  
Discrete Ordinate ◽  
Pressurized Water ◽  
Water Reactor ◽  
Cavity Radiation

Towards Optimal In-Core Fuel Management of Thorium-Plutonium-Fuelled PWR Cores

2013 ◽  
Author(s):  
Nurjuanis Z. Zainuddin ◽  
Benjamin A. Lindley ◽  
Geoffrey T. Parks
Keyword(s):  
Nuclear Waste ◽  
Cycle Length ◽  
Feasibility Studies ◽  
Fuel Management ◽  
Loading Pattern ◽  
Burnable Poison ◽  
The Core ◽  
Mox Fuel ◽  
Operational Characteristics

Plutonium is a significant proliferation concern as well as a major contributor to the long-term toxicity of nuclear waste. Partial incineration in PWRs with uranium-MOX fuel is often considered to mitigate these concerns. Thorium-MOX is an alternative fuel with superior material properties and higher plutonium destruction rates, as shown in multiple feasibility studies. However, the core performance and operational characteristics (e.g. discharge burn-up, feasibility of controlling the core) are ultimately dependent on the core loading pattern (LP) and burnable poison (BP) design. In this paper, the LP for Th-Pu fuel of various compositions is optimized for (1) discharge burn-up, (2) radial form factor (RFF), (3) cycle length, (4) moderator temperature coefficient (MTC), and (5) reactivity swing over cycle. Maximizing the cycle length makes the discharge burn-up and reactivity swing worse due to placement of once- and twice-burnt fuel near the core periphery. It also makes the MTC less negative. The harder neutron spectrum of Th-Pu fuel compared to conventional U fuel favours the use of distributed integral burnable poisons to control the reactivity swing over the cycle. This leads to a significant amount of dissimilarity between LPs with relatively similar performance measures, and between optimal LPs for different Pu loadings in the fuel. The RFF can vary throughout the cycle but a careful placement of the assemblies can mitigate this. The cycle reactivity swing is controlled using enriched soluble boron, which makes the MTC worse, and this constrains feasibility for high Pu loading in the fuel.

Research on the Loading Pattern of Reactor Core With MOX Fuel in the First Cycle

2012 ◽  
Author(s):  
Tianqi Zhang ◽  
Shihe Yu ◽  
Xinrong Cao
Keyword(s):  
Cross Section ◽  
Reactor Core ◽  
Fuel Utilization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loading Pattern ◽  
Mox Fuel ◽  
Initial Cycle ◽  
Calculation Code

In order to research the performance of Pressurized Water Reactor (PWR) with 1/3 MOX fuel in the initial cycle, this paper serves Qinshan II reactor core as the reference core to design suitable MOX assemblies and study relevant core properties. The analyses documented within use assembly cross section calculation code CASMO-4 and core calculation code SIMULATE-3 studied by Studsvik. The purpose of this paper is to demonstrate that the Qinshan II reactor is capable of complying with the requirement for MOX fuel utilization without significant changes to the design of the plant. Several impacts on key physics parameters and safety analysis assumptions, introduced by MOX, are discussing in the paper.

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