Unprotected Overpower Transient Analysis of a Million Kilowatt Traveling Wave Reactor Core

2017 ◽  
Author(s):  
Pengrui Qiao ◽  
Jian Zhang ◽  
Chao Lin
Keyword(s):  
Fast Reactor ◽  
Traveling Wave ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Nuclear Nonproliferation ◽  
Utilization Rate ◽  
Fuel Cladding ◽  
Technological Level ◽  
Long Lifetime

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the unprotected overpower transient (UTOP) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the UTOP accident, the core of the million kilowatt TWR demonstrates a good safety performance.

Unprotected Loss of Flow Analysis of a Million Kilowatt Traveling Wave Reactor Core

2017 ◽  
Author(s):  
Wenjun Hu ◽  
Pengrui Qiao
Keyword(s):  
Fast Reactor ◽  
Traveling Wave ◽  
Nuclear Reactor ◽  
Flow Analysis ◽  
Reactor Core ◽  
Travelling Wave ◽  
Nuclear Nonproliferation ◽  
Utilization Rate ◽  
Technological Level ◽  
Long Lifetime

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the Unprotected loss of flow (ULOF) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the ULOF accident, the core of the million kilowatt TWR demonstrates a good safety performance.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

scholarly journals Improving Nuclear Safety of Fast Reactors by Slowing Down Fission Chain Reaction

2014 ◽  
Vol 2014 ◽  
pp. 1-18 ◽  
Author(s):  
G. G. Kulikov ◽  
A. N. Shmelev ◽  
V. A. Apse
Keyword(s):  
Cross Sections ◽  
Fast Reactor ◽  
Reactor Core ◽  
Large Contribution ◽  
Delayed Neutrons ◽  
Slowing Down ◽  
Radiogenic Lead ◽  
Radiative Neutron ◽  
Long Lifetime ◽  
Radiative Neutron Capture

Light materials with small atomic mass (light or heavy water, graphite, and so on) are usually used as a neutron reflector and moderator. The present paper proposes using a new, heavy element as neutron moderator and reflector, namely, “radiogenic lead” with dominant content of isotope 208Pb. Radiogenic lead is a stable natural lead. This isotope is characterized by extremely low micro cross-section of radiative neutron capture (~0.23 mb) for thermal neutrons, which is smaller than graphite and deuterium cross-sections. The reflector-converter for a fast reactor core is the structure capable of transforming some part of prompt neutrons leaked from the core into the reflected neutrons with properties similar to those of delayed neutrons, that is, sufficiently large contribution to reactivity at the level of effective fraction of delayed neutrons and relatively long lifetime, comparable with lifetimes of radionuclides-emitters of delayed neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in the fast reactor. This can improve the fast reactor safety and reduce some requirements to the technologies used to fabricate fuel for the fast reactor.

Neutronics Design Study on Ignition Zone of Travelling Wave Reactor

2013 ◽  
Author(s):  
Jian Zhang ◽  
Hong Yu ◽  
Zhi Gang ◽  
Keyuan Zhou ◽  
Yun Hu ◽  
...  
Keyword(s):  
Traveling Wave ◽  
Nuclear Reactor ◽  
Travelling Wave ◽  
Reactor Operation ◽  
Geometry Model ◽  
Nuclear Transmutation ◽  
Neutron Poison ◽  
Distribution Of Power ◽  
Distribution Of Power Density ◽  
Ignition Zone

Traveling wave reactor is a kind of nuclear reactor that can convert fertile material into fissile fuel as it runs using the process of nuclear transmutation. In the ignition stage of traveling wave reactor, the core performance is especially complex, since the fissile fuel and fertile material is put in different regions at the beginning. And the distribution of power density will change severely with burn-up during the reactor operation. It is an important part of the traveling wave reactor study to optimize the design of the ignition stage. In this paper, based on a two-dimensional RZ geometry model, some schemes with different sizes and compositions of the ignition zone, middle ignition zone position design and burnable neutron poison addition are simulated and analyzed. Finally, an optimized core design with multi-zone configuration and burnable neutron poison addition is shown. Some design outlines are introduced for further study.

Fuel Assembly Bowing and Core Restraint Design in Fast Reactors

2014 ◽  
Author(s):  
J. J. Grudzinski ◽  
C. Grandy
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Transient Behavior ◽  
Restraint System ◽  
The Core ◽  
Fuel Elements ◽  
Over Time

The reactivity of a fast spectrum nuclear reactor core is sensitive to changes in the fuel position. The core is formed by a hexagonal array of fuel assemblies which contain the fuel elements. The main structural components of the assemblies are thinwall hexagonal ducts. The fuel elements represent negligible stiffness in the fuel assembly compared to the ducts such that the ducts determine the location of the fuel. Thermal gradients across the fuel assembly cross sections create differential thermal expansion which causes the assemblies to bow. This bowing is proportional to the power to flow ratio such that it can become an important part of the reactivity change during reactor transients such as during reactor start-up, transient overpower (TOP), and unprotected loss of flow without scram (ULOF). In addition to these short term transients, thermal and fast neutron flux gradients within the core cause the assembly ducts to swell and bow over time due to irradiation creep and swelling. These latter effects produce permanent bowing of the ducts which change the reactivity over time and more importantly affect the mechanical forces required to refuel the core as the bowing is greater that the duct-to-duct clearance. Understanding these bowing responses is important to the understanding of both the transient behavior of a fast reactor as well as the refueling loads. Through proper design of the core restraint system, the bowing response can be engineered to provide negative feedback during the above mentioned transients such that it becomes part of the inherent safety of a fast reactor. Similarly, the opposing effects of creep and swelling can be manipulated to reduce the permanent core bowing deformations. We provide a discussion of the key features of analyzing and designing a core restraint system and provide a brief survey of the past work.

Study on Ultra-Long Life, Small U-ZR Metallic Fuelled Core With Burnable Poison

2002 ◽  
Author(s):  
Kenji Tsuji ◽  
Hiromitsu Inagaki ◽  
Akira Nishikawa ◽  
Hisato Matsumiya ◽  
Yoshiaki Sakashita ◽  
...  
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Major Part ◽  
Reactor Core ◽  
Core Concept ◽  
Safety Requirements ◽  
Burnable Poison ◽  
The Core ◽  
Long Lifetime ◽  
Long Life

A conceptual design for a 50MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up.

Machine Learning-Based Methodology for Assessment of Doppler Reactivity of Sodium-Cooled Fast Reactor

2021 ◽  
Vol 7 (4) ◽  
Author(s):  
Đorđe Petrović ◽  
Konstantin Mikityuk
Keyword(s):  
Machine Learning ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Fourth Generation ◽  
Fuel Temperature ◽  
Horizon 2020 ◽  
Classical Methodology ◽  
Artificial Neural Network Ann

Abstract In order to close nuclear fuel cycle and address the problem of sustainability, advanced nuclear reactor systems of the fourth generation are in the focus of the research for many years. With a simple goal of supporting this research, machine learning-based methodology for the assessment of the Doppler reactivity has been developed and applied to the European Sodium Fast Reactor (ESFR) in the frame of the ESFR-Safety Measures Assessment and Research Tools (SMART) Horizon-2020 project. In the scope of this study, a database of the precise Monte Carlo (MC) calculations was prepared and used to train artificial neural network (ANN) as a surrogate model to assess the Doppler reactivity across the range of reactor conditions that could occur throughout the life of the reactor core, in fast, yet accurate manner. The database was generated for all the combinations of several core parameters carefully predefined in order to account for both operational and accidental states of the core. Subsequently, Doppler reactivity change as a function of the above-mentioned parameters was assessed by herein developed methodology, as well as by widely used logarithmic dependence of the Doppler reactivity on the fuel temperature and compared to the results of the precise MC simulations. This study proves that, if certain computational resources are allocated to the database generation and ANN training, newly developed methodology yields similar or even more accurate results compared to the classical methodology and at the same time provides a tool for parameterization and interpolation of Doppler reactivity not only on the fuel temperature but also on the other parameters characterizing core of the sodium-cooled fast reactor (SFR).

scholarly journals Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

2008 ◽  
Vol 2008 ◽  
pp. 1-9 ◽  
Author(s):  
Enrico Zio ◽  
Francesco Di Maio
Keyword(s):  
Confidence Interval ◽  
Order Statistics ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Fuel Cladding ◽  
Complete Group ◽  
Bootstrap Technique ◽  
Safety Margins ◽  
Group Distribution

In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Advances in the analysis of fast reactor core and coolant circuit structures

1992 ◽  
Vol 134 (1) ◽  
pp. 37-58
Author(s):  
Y.W. Chang ◽  
D.T. Eggen ◽  
A. Imazu ◽  
M. Livolant
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Coolant Circuit

scholarly journals Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor

2017 ◽  
Vol 2017 ◽  
pp. 1-7
Author(s):  
Mazhyn K. Skakov ◽  
Nurzhan Ye. Mukhamedov ◽  
Alexander D. Vurim ◽  
Ilya I. Deryavko
Keyword(s):  
Heat Capacity ◽  
Thermal Diffusivity ◽  
Fast Reactor ◽  
Thermophysical Properties ◽  
Heat Conductivity ◽  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Full Scale ◽  
Temperature Fields ◽  
First Time

For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel.

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