Loading Scheme Research of Small Modular Sodium Cooled Reactor in a Factory

2017 ◽  
Author(s):  
Hu Xiao ◽  
Chen Xiao Liang
Keyword(s):  
Neutron Source ◽  
Fast Reactor ◽  
Thermal Power ◽  
Reactor Vessel ◽  
Monte Carlo Code ◽  
Construction Time ◽  
Mcnp Simulation ◽  
The Core ◽  
China Experimental Fast Reactor ◽  
Loading Scheme

An innovative small modular sodium cooled fast reactor called S1 is designed by China Institute of Atomic Energy (CIAE). As an encapsulated nuclear source with thermal power of 3MW, S1 is characterized by small volume, light weight, high safety and sound reliability. The S1 systems adopt modularization, by which the core will be loaded in a factory and filled with sodium, then shipped to the assembled onsite, thus the construction time for S1 can be substantially reduced. In this paper, the Monte Carlo code MCNP is used to calculate the loading scheme of S1. Considering factory’s special conditions for loading and active chemical property of sodium, a special loading pattern is adopted for S1 — loading fuel first, then sodium filling. Ensure that the neutron source and detectors are well matched during loading, and detector counting rate is no less than 2cps when only neutron source but no fuel exists in the core. Three positions where the 252Cf neutron source is placed are studied in this paper: (1) at the center of the core; (2) at the middle of outside core plane; (3) at the bottom of outside the reactor vessel. Through MCNP simulation calculations and comparison of large resulting data, it finds the neutron source should be reasonably placed at the bottom of the reactor vessel where 252Cf strength is 105 s−1 neutrons, and the ex-core detectors are distributed symmetrically at the center of outside core plane; the most befitting moderator material of detector surface is methacrylate-C5H8O3. In this paper, 1/N extrapolation method is used during loading and kinds of loading schemes have been studied with reference to the principles of China experimental fast reactor (CEFR) and regulations of relevant research reactors, and 5-batch loading scheme is finally chosen as the optimal loading scheme. S1 is prepared for sodium filling at 250 °C. It shows that neutron flux variation of core can be more reliably monitored when the ex-core detectors are placed about 120cm away from the center core through MCNP simulation calculation. Such arrangement can also meet the monitoring requirements for loading and sodium filling.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor

2013 ◽  
Author(s):  
Jin Wang ◽  
Donghui Zhang ◽  
Wenjun Hu ◽  
Lixia Ren
Keyword(s):  
Fast Reactor ◽  
Nuclear Energy ◽  
Circuit Simulation ◽  
Thermal Power ◽  
Physical Models ◽  
Primary Circuit ◽  
Cooling Channel ◽  
Simulation Code ◽  
Wide Range ◽  
China Experimental Fast Reactor

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit. To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.

scholarly journals Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

2019 ◽  
Vol 2019 ◽  
pp. 1-6
Author(s):  
Toshio Wakabayashi ◽  
Makoto Takahashi ◽  
Naoyuki Takaki ◽  
Yoshiaki Tachi ◽  
Mari Yano
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Thermal Characteristics ◽  
Outer Core ◽  
Severe Accident ◽  
Monte Carlo Code ◽  
Core Concept ◽  
Neutron Absorber ◽  
The Core ◽  
Nuclear Data Library

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

Analysis of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Xiong Wenbin ◽  
Xie Qin ◽  
Li Huwei ◽  
Yang Sengai ◽  
Mao Huan ◽  
...  
Keyword(s):  
Monte Carlo ◽  
International Atomic Energy Agency ◽  
Fast Reactor ◽  
Physical Parameters ◽  
Energy Agency ◽  
Safety Review ◽  
The Core ◽  
Calculation Results ◽  
Design Values ◽  
China Experimental Fast Reactor

The whole core model of China experimental fast reactor (CEFR) is established according to the parameters of China experimental fast reactor, which are given by technical publication from the International Atomic Energy Agency (IAEA-TECDOC-1531), and the physical parameters of CEFR are simulated with the Monte Carlo N-particle code (MCNP4a). The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design values, which successfully demonstrate the acceptable fidelity of the MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.

Numerical Simulation of Safety Rod and its Drive Mechanism in Sodium-Cooled Fast Reactor During Scram Action

2018 ◽  
Author(s):  
Li Yan ◽  
Hu Wenjun ◽  
Ren Lixia
Keyword(s):  
Numerical Simulation ◽  
Fast Reactor ◽  
Simulation Method ◽  
Drive Mechanism ◽  
Pipe System ◽  
The Core ◽  
Hydraulic Method ◽  
Drop Time ◽  
China Experimental Fast Reactor ◽  
The Relationship

Safety rod and its drive mechanism is one of the shutdown systems in sodium-cooled fast reactor, which must be quickly inserted into the core to achieve emergency shutdown in the event of an accident. Therefore, it is necessary to study the falling process of safety rod. In this paper, the numerical simulation method is used to analyze the falling process of safety rod and its drive mechanism in China Experimental Fast Reactor. According to the flow path of the safety rod and its drive mechanism, the pipe system hydraulic method is used to model the safety rod and its drive mechanism and calculate the hydraulic force of the safety rod and its drive mechanism during the falling process. The relationship between time, displacement, velocity and acceleration is presented. The drop time of safety rod is calculated, which is compared with the experimental results. The factors that affect the drop process are analyzed and a sensitivity analysis is presented.

Simulation Calculation of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

2017 ◽  
Author(s):  
Xiong Wen-bin ◽  
Cao Jian ◽  
Xie Qin ◽  
Wang Zhan-yong ◽  
Wang Chuang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Nuclear Safety ◽  
Core Model ◽  
Physical Parameters ◽  
Safety Review ◽  
The Core ◽  
Simulation Calculation ◽  
Calculation Results ◽  
Design Value ◽  
China Experimental Fast Reactor

The whole core model of CEFR is established according to the parameters of China Experimental Fast Reactor which are given by IAEA-TECDOC-1531, and the physical parameters of CEFR are simulated with the MCNP4a program. The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design value, which successfully demonstrates the reliability of this MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.

Transient Analysis of Crossduct Break Scenarios Using the CATHARE2 Code for the 75MW ALLEGRO Demonstrator

2014 ◽  
Author(s):  
Gusztáv Mayer ◽  
Fabrice Bentivoglio
Keyword(s):  
Fast Reactor ◽  
Transient Analysis ◽  
Early Stage ◽  
Thermal Power ◽  
The Road ◽  
The Core ◽  
Nitrogen Injection ◽  
Stage Of Development ◽  
Wide Range ◽  
On The Road

The helium cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV International Forum. Since no gas cooled fast reactor has ever been built, a medium power demonstrator reactor — named ALLEGRO — is necessary on the road towards the 2400MWth GFR power reactor. The French CEA completed a wide range of studies on the early stage of development of ALLEGRO, and later the ALLEGRO reactor have been developed in several European Union projects in parallel with the GFR2400. The 75 MW thermal power ALLEGRO is recently being developed in the frame of European ALLIANCE project. As a result of the collaboration between CEA and MTA EK new improvements were done in the CATHARE modeling of ALLEGRO. In particular, the capability of simulation of breaks located in the crossduct (concentrically arranged pipes with the hotduct located inside the colduct) has been developed. A first scenario of hotduct break has been simulated, that does not lead to the depressurization of the system because of the crossduct technology. Nevertheless this transient leads to a high bypass of the core. Then a scenario of full rupture of the hotduct and the colduct has been tested, leading to beyond design state with depressurized situation combined with a large bypass of the core. However this study shows that the peak cladding temperature can be kept below the cladding melting point using nitrogen injection. In this paper the CATHARE model implemented for the crossduct rupture scenario and the results of the simulation are presented and discussed.

scholarly journals MONTE CARLO SIMULATION OF NEUTRONICS START-UP TESTS AT CHINA EXPERIMENTAL FAST REACTOR (CEFR)

2021 ◽  
Vol 247 ◽  
pp. 10008
Author(s):  
Jiwon Choe ◽  
Chirayu Batra ◽  
Vladimir Kriventsev ◽  
Deokjung Lee
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Impact Analysis ◽  
Physical Models ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Preliminary Results ◽  
Research Activities ◽  
Start Up ◽  
China Experimental Fast Reactor

China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release some of the neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activities through the coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”, launched in 2018. This benchmark aims to perform validation and verification of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data. The six physics start-up tests considered for this CRP include evaluation of the criticality, control rod worth, void reactivity, temperature coefficient, swap reactivity, and foil irradiation. Twenty-nine participating research organizations are performing independent blind calculations during the first phase of the project. As a part of this coordinated research, IAEA performed neutronics calculations using Monte Carlo code SERPENT. Two kinds of 3D core models, homogenous and heterogeneous, were calculated using SERPENT, with ENDF/B-VII.0 continuous energy library. Preliminary results with a reasonably good estimation of criticality, as well as theoretically sound results of other five test cases, are available. The paper will discuss the core modelling assumptions, challenges and key findings of modelling a dense SFR core, preliminary results of the first phase of the CRP, heterogeneity impact analysis between homogenous core models and heterogeneous core models and future work to be performed as a part of this four-year project.

Design Analysis of Corium Immobilization and Cooling of the Reactor Vessel in Severe Accidents at NPP With Commercial Unit V-320

2013 ◽  
Author(s):  
Roman A. Stremedlovskyi
Keyword(s):  
Soviet Union ◽  
Thermal Power ◽  
Design Analysis ◽  
Reactor Vessel ◽  
Reactor Plant ◽  
Commercial Unit ◽  
Pressurized Water ◽  
The Core ◽  
Technical Ability ◽  
Severe Accidents

Keeping the melt inside of the reactor vessel is the key strategy of the management of severe accidents for the exploited, as well as for the newly designed reactors with pressurized water. The concept has been substantiated for emergency cooling of VVER-440 reactor at Loviisa NPP in Finland [1]. However, for more powerful reactor, such as VVER-1000, up to now it was considered that this concept did not work. This work analyzes the problem of the interaction of molten corium with the reactor vessel. The work is representing the results of studies that were carried out as part of research with a goal to validate design functions of the system of containment the melt inside the VVER vessel. The VVER or WWER is a series of PWR designs originally developed in the Soviet Union, and nowadays in Russia. The reactor plant of VVER-1000 type of V-320 series is a reactor plant with a water-cooled and water-moderated power reactor for NPP. It has an electric power of 1000 MW and a thermal power of 3000 MW. The plant consists of four loops, each one with a horizontal SG and a MCP. In the work the design analysis of the accident with melting of the core of the reactor plant V-320 was conducted on codes RELAP/SCDAPSIM/MOD3.4 and MELCOR. The main processes that determine the state of the core melting in the reactor vessel were analyzed. The main results of the calculation of the molten corium interaction with the wall of the reactor vessel were represented. The results of the preliminary calculated estimation of the melt behavior and the technical ability to hold the core melt in the vessels of PWR reactors were represented. Based on the carried out analysis and implementation of technological and design study, the recommendations were made with a purpose to improve the safety of NPP with a reactor plant V-320 using the systems of melt restraining and cooling of the reactor vessel.

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