scholarly journals Transient Analysis of a Subcritical Reactor Core with a MOX-Fuel using the Birth-and-Death Model

2020 ◽  
Author(s):  
Dr Tamara Korbut ◽  
Andrey Kuzmin ◽  
Eduard Rudak ◽  
Maksim Kravchenko
Keyword(s):  
Transient Analysis ◽  
Reactor Core ◽  
Mox Fuel ◽  
Subcritical Reactor

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.

scholarly journals PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE

2021 ◽  
Vol 247 ◽  
pp. 07002
Author(s):  
Tsutomu Okui ◽  
Akifumi Yamaji
Keyword(s):  
Temperature Gradient ◽  
Transient Analysis ◽  
Large Temperature ◽  
Control Rod ◽  
Mox Fuel ◽  
Fuel Layer ◽  
Axial Temperature ◽  
Short Period ◽  
Full Power ◽  
Supercritical Water Cooled Reactor

The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 229-235
Author(s):  
Y. Alzahrani ◽  
K. Mehboob ◽  
F. A. Abolaban ◽  
H. Younis
Keyword(s):  
Reactor Core ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
The Monte Carlo Method ◽  
Burnable Absorber ◽  
Effective Multiplication Factor ◽  
Reactivity Coefficient ◽  
Increasing Temperature ◽  
Effective Multiplication

Abstract In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

Structure Design of Jordan Subcritical Reactor

2013 ◽  
Author(s):  
Chengzhi Yao ◽  
Yuerong Fan ◽  
Jinshan Zhang ◽  
Haifen Han ◽  
Dayong Yi ◽  
...  
Keyword(s):  
Neutron Source ◽  
Reactor Core ◽  
Structure Design ◽  
History Of ◽  
Production Research ◽  
Design Characteristics ◽  
Operation Platform ◽  
Fuel Elements ◽  
And Training ◽  
Subcritical Reactor

The subcritical reactor is a kind of small nuclear facility, which may realize the chain reactor reaction through introducing an external neutron source to meet with the production, research, teaching, testing and training purposes. Meanwhile, the subcritical reactor will not reach the critical status, which has advantages of simple structure, safety, reliability. Jordan subcritical reactor is a light water moderated reactor, it employs the Pu-Be as external neutron source, and the neutron multiplication factor in the range of 0.94～0.95. Jordan subcritical reactor mainly consists of fuel elements, reactor core and reactor core vessel assembly, operation platform, water loop system, neutron source driving system, which can be used for the purposes of teaching, training, testing and research. This paper reviews the international application history of subcritical reactor and its status of research and development, describes the design purpose and requirement of Jordan subcritical reactor. The detailed structure design of Jordan subcritical reactor is illustrated. Furthermore, the structure design characteristics and some difficulties of Jordan subcritical reactor are mentioned in detail, such as the choose of structure materials, lay out of Jordan subcritical reactor, control of machining precision and sealing of pipes. Then, the possible solutions of these problems are presented. Now, the manufacture and installation of Jordan subcritical reactor has been completed, which fulfill the anticipated design requirement.

scholarly journals Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

2014 ◽  
Vol 29 (4) ◽  
pp. 259-267
Author(s):  
Fiifi Asah-Opoku ◽  
Zhihua Liang ◽  
Ziaul Huque ◽  
Raghava Kommalapati
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Monte Carlo Code ◽  
Oxide Fuel ◽  
Pressurized Water ◽  
Plutonium Isotopes ◽  
Unequal Distribution ◽  
Mox Fuel ◽  
Enriched Uranium

Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX), uranium oxide fuel (UOX), and commercially enriched uranium or uranium metal (CEU) - are used in this simulation and their impact on the effective multiplication factor (Keff) and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

Code Development for the Neutronics/Thermal-Hydraulics Coupling Transient Analysis of Molten Salt Reactors

2017 ◽  
Author(s):  
Tianliang Hu ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Kun Zhuang
Keyword(s):  
Molten Salt ◽  
Cross Section ◽  
Cross Sections ◽  
Transient Analysis ◽  
Reactor Core ◽  
Dynamics Simulation ◽  
Monte Carlo Code ◽  
Section Data ◽  
Dynamics Simulations ◽  
Data Library

A code system has been developed in this paper for the dynamics simulations of MSRs. The homogenized cross section data library is generated using the continuous-energy Monte-Carlo code OpenMC which provides significant modeling flexibility compared against the traditional deterministic lattice transport codes. The few-group cross sections generated by OpenMC are provided to TANSY and TANSY_K which is based on OpenFOAM to perform the steady-state full-core coupled simulations and dynamics simulation. For verification and application of the codes sequence, the simulation of a representative molten salt reactor core MOSART has been performed. For the further study of the characteristics of MSRs, several transients like the code-slug transient, unprotected loss of flow transient and overcooling transient have been analyzed. The numerical results indicated that the TANSY and TANSY_K codes with the cross section library generated by OpenMC has the capability for the dynamics analysis of MSRs.

Reactor core transient analysis of an innovative high-level nuclear waste transmuter with metal fuel

2017 ◽  
Vol 41 (9) ◽  
pp. 1322-1334 ◽  
Author(s):  
Youqi Zheng ◽  
Mingtao He ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Xunzhao Li ◽  
...  
Keyword(s):  
Nuclear Waste ◽  
Transient Analysis ◽  
Reactor Core ◽  
Metal Fuel ◽  
High Level Nuclear Waste ◽  
High Level

scholarly journals Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Fuel Temperature ◽  
Safety Level ◽  
The Core ◽  
Mox Fuel ◽  
Calculation Results ◽  
Reactivity Coefficient ◽  
Fuel Elements

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>

scholarly journals Effect of MOX fuel and the ENDF/B-VIII on the AP1000 neutronic parameters calculations by using MCNP6

2019 ◽  
Vol 34 (4) ◽  
pp. 325-335
Author(s):  
Sonia Reda ◽  
Ibrahim Gomaa ◽  
Ibrahim Bashter ◽  
Esmat Amin
Keyword(s):  
Reactor Core ◽  
Delayed Neutron ◽  
Optimization Criteria ◽  
Mox Fuel ◽  
235U Enrichment ◽  
Reactivity Coefficients ◽  
Effective Multiplication Factor ◽  
Neutronic Parameters ◽  
Ap1000 Reactor ◽  
Effective Multiplication

The present work studies the effect of introducing MOX fuel on Westinghouse AP1000 neutronic parameters. The neutronic calculations were performed by using the MCNP6 code with the ENDF/B-VII.1 library and the new release of the ENDF/B-VIII, the AP1000 core with three 235U enrichment zones (2.35 %, 3.40 %, and 4.45 %). The obtained results showed that the simulated model for the AP1000 core satisfies the optimization criteria as a Westing- house reference. The results which included: effective multiplication factor, keff, delayed neutron fraction, beff, excess reactivity, rex, shutdown margin, temperature reactivity coefficients, whole core depletion, neutron flux, power peaking factor and core power density, were calculated and compared with the available published results. The keff in the cold zero power was found to be 1.20495 and 1.20247 with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which matches the value of 1.205 presented in the AP1000 Design Control Document for the UO2 fuel core. On the other hand, keff in the cold zero power was found to be 1.19988 and 1.19860 for MOX core with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which show good reception and confirm the safety of the design and efficient modeling of AP1000 reactor core.

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