Dynamic Simulation Research on Primary Coolant System of Sodium Cooled Fast Reactor

2012 ◽  
Author(s):  
Yang Yu ◽  
Yun Guo
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Cold Pool ◽  
Primary Coolant ◽  
Simulation Research ◽  
Flow And Heat Transfer ◽  
Comparison Of The Results ◽  
Pool Type

The Chinese Experimental Fast Reactor (CEFR) is a 65MWt/20MWe sodium cooled fast reactor. It is a pool-type reactor where the reactor and other internals such as pumps and intermediate heat exchangers (IHX) are immersed in a sodium pool. In this paper a one-dimensional dynamic code was developed to model the primary sodium circuit which included the reactor core, IHX, pumps, hot and cold pool etc. Moreover, the model of the property of sodium flow and heat transfer correlations was collected and compiled. This paper also discusses the mathematical models of various components of the primary sodium circuit, the numerical techniques to solve the models, the thermal-hydraulic studies of some design basis events such as the loss of primary pump or secondary pump accident etc, the comparison of the results of the code with that of the safety analysis report. Studies were conducted simulating both full and low power operating conditions. The dynamic code has been validated, and the results show that it has a benign response to some typical accidents. Finally, the model and code derived in this paper could be used in the safety analysis of pool-type sodium cooled fast reactor, and adopted in the development of CEFR simulation platform.

Development and Application of the Neutronics/Thermal-Hydraulics Coupling Code for Safety Analysis of EBR-II Loss of Heat Sink Tests Without Scram

2016 ◽  
Author(s):  
Daogang Lu ◽  
Chao Guo ◽  
Danting Sui
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Fuel Cycle ◽  
Transient Behavior ◽  
Safety Analysis ◽  
Cold Pool ◽  
Thermal Hydraulics ◽  
Closed Fuel Cycle ◽  
Zone Method ◽  
Gen Iv

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.

Assessing Advanced Test Reactor Process States Using Acoustic Measurement Infrastructure

2017 ◽  
Author(s):  
Vivek Agarwal ◽  
James A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Signal Transmission ◽  
National Laboratory ◽  
Reactor Core ◽  
Operating Conditions ◽  
Acoustic Measurement ◽  
Primary Coolant ◽  
Test Reactor ◽  
Operation Cycle ◽  
Measurement Infrastructure

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement). This paper demonstrated different in acoustic signature of the ATR under different operating conditions. In particular, ATR acoustic baseline is captured during typical operation cycle and during power axial locator mechanism operation cycle. The difference in two acoustic baseline is significant and highlights salient difference that are critical in the design and development of acoustically telemetered sensors.

Safety Analysis of Fast Reactor Core with Uranium-Free Fuel for Actinide Transmutation

2003 ◽  
Vol 143 (3) ◽  
pp. 281-289 ◽  
Author(s):  
Igor Krivitski ◽  
Mikhail Vorotyntsev ◽  
Valentin Pyshin ◽  
Ludmila Korobeinikova
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Safety Analysis ◽  
Actinide Transmutation

Concept of Pressurized Water Lead-Bismuth-Cooled Fast Reactor (PLFR)

2014 ◽  
Author(s):  
Minoru Takahashi
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Point Of View ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Lead Bismuth Eutectic ◽  
Innovative Concept

The innovative concept of pressurized water lead-bismuth-cooled fast reactor (PLFR) has been proposed and studied based on the previous LFR concept: PBWFR. Primary pumps and steam generators that contact lead-bismuth coolant are eliminated. A feedwater is directly injected into the primary coolant of hot lead-bismuth eutectic (LBE) at the outlet of the reactor core under the pressure of 14 MPa as PWR. The specifications of PLFR system are discussed and presented from thermal-hydraulic point of view.

Modelling Frozen Salt Films in a Molten Salt Fast Reactor

2018 ◽  
Author(s):  
Gregory M. Cartland-Glover ◽  
Stefano Rolfo ◽  
Alex Skillen ◽  
David R. Emerson ◽  
Charles Moulinec ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Molten Salt ◽  
Fast Reactor ◽  
Energy Production ◽  
Nuclear Energy ◽  
Numerical Models ◽  
Reactor Core ◽  
Separated Flow ◽  
Flow And Heat Transfer ◽  
Innovative Solutions

Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. The techniques used to model thermal hydraulics of a molten salt fast reactor when frozen salt wall technology is applied to the core vessel wall are presented here for 2D numerical models of a hyperboloid reactor core region with a heat exchanger was applied in Code_Saturne. A 3D simulation of the fluid flow and heat transfer with 16 recirculation loops containing the heat exchangers is also presented. It was found that there is strong cooling in separated flow regions in the external heat exchanger, which freezes where the porous model is applied.

Design of a Discharge Header and a Working Platform for a Research Reactor With a CFD Model

2013 ◽  
Author(s):  
Kyoungwoo Seo ◽  
Hyungi Yoon ◽  
Dae-young Chi ◽  
Seonghoon Kim ◽  
Juhyeon Yoon
Keyword(s):  
Research Reactor ◽  
Cold Water ◽  
Cooling System ◽  
Reactor Core ◽  
Water Layer ◽  
Hot Water ◽  
Primary Coolant ◽  
Radiation Level ◽  
Reactor Pool ◽  
Pool Type

Most research reactors are designed as an open-pool type and the reactor is located on the bottom of the open-pool. The reactor in the pool is connected to the primary cooling system, which is designed for adequate cooling of the heat generated from the reactor core. One of the characteristics of an open-pool type research reactor is that the primary coolant after passing through the reactor core and the primary cooling system (PCS) is returned to the reactor pool. Because the primary coolant contains many kinds of radionuclides, the research reactor should be designed to protect the radionuclides from being released outside the pool by a stratified stable water layer, which is formed between a hot water layer and cold water near the reactor and prevents the natural circulation of water in the pool. In this study, additional components such as a discharge header and a working platform inside the pool were developed to help diminish the radiation level to the pool top. To discharge coolant stably inside the reactor pool, a discharge header was installed at the end of the pool inlet pipe. Many holes were made in the discharge header to discharge the coolant slowly and minimize the disturbance of the hot water layer by the flow inside the pool. The working platform was also equipped inside the reactor pool to remove the convective flow near the pool top. The commercially available CFD code, ANSYS CFD-FLEUNT, was used to specifically design the discharge header and working platform for satisfying the requirement of the pool top radiation level. The computations were conducted to analyze the flow and temperature characteristics inside the pool for several geometries using an SST k-ω turbulent model and cell modeling, which were conducted to isolate the root cause of these differences and the given inlet conditions. The discharge header and working platform were designed using the CFD results.

Analysis code development for the direct reactor auxiliary cooling system of the pool-type sodium-cooled fast reactor

Kerntechnik ◽  
2018 ◽  
Vol 83 (3) ◽  
pp. 232-236 ◽  
Author(s):  
D. L. Zhang ◽  
P. Song ◽  
S. Wang ◽  
X. Wang ◽  
J. Chen ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Pool Type ◽  
Code Development

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
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