scholarly journals New Neutron Imaging Facility development at the Penn State Breazeale Nuclear Reactor

2021 ◽  
Vol 253 ◽  
pp. 04012
Author(s):  
Alibek Kenges ◽  
Kenan Unlu ◽  
Daniel Beck
Keyword(s):  
Neutron Beam ◽  
Nuclear Reactor ◽  
Current System ◽  
Thermal Flux ◽  
Reactor Power ◽  
Neutron Imaging ◽  
Preliminary Results ◽  
Exit Surface ◽  
Penn State ◽  
American Society

Preliminary results of characterization experiments for the New Neutron Imaging Facility (NIF) that is being developed at the Penn State Breazeale Nuclear reactor are presented in the following sections. The methodology of neutron beam characterization described in the American Society for Testing and Materials (ASTM) documents for the neutron imaging systems have been followed to improve the NIF at Penn State to a Category I facility by ASTM designation of quality. Preliminary results showed that our system is capable of producing images of high quality, corresponding to Category I; however, further experiments are needed for full declaration of our facility as such. Additionally, the effective collimation ratio (L/D ratio) of our current system is ∼110 with the capability of improvement to ∼150. The thermal flux at the exit surface of the neutron beam is equal to 5.4 × 106n cm−2s−1 at 1MWth reactor power, which corresponds to the industry comparable value.

Neutron Transport Characteristics of a Nuclear Reactor Based Dynamic Neutron Imaging System

2006 ◽  
Author(s):  
Anas M. Khaial ◽  
Glenn D. Harvel ◽  
Jen-Shih Chang
Keyword(s):  
Neutron Flux ◽  
Neutron Beam ◽  
High Speed ◽  
Thermal Neutron Flux ◽  
Nuclear Reactor ◽  
Imaging System ◽  
Neutron Transport ◽  
Image Plane ◽  
Neutron Imaging ◽  
Beam Tube

An advanced dynamic neutron imaging system has been constructed in the McMaster Nuclear Reactor (MNR) for nondestructive testing and multi-phase flow studies in energy and environmental applications. A high quality neutron beam is required with a thermal neutron flux greater than 5.0×106 n/cm2-s and a collimation ratio of 120 at image plane to promote high-speed neutron imaging up to 2000 frames per second. Neutron source strength and neutron transport have been experimentally and numerically investigated. Neutron source strength at the beam tube entrance was evaluated experimentally by measuring the thermal and fast neutron fluxes, and simple analytical neutron transport calculations were performed based upon these measured neutron fluxes to predict facility components in accordance with high-speed dynamic neutron imaging and operation safety requirements. Monte-Carlo simulations (using MCNP-4B code) with multiple neutron energy groups have also been used to validate neutron beam parameters and to ensure shielding capabilities of facility shutter and cave walls. Neutron flux distributions at the image plane and the neutron beam characteristics were experimentally measured by irradiating a two-dimensional array of Copper foils and using a real-time neutron radiography system. The neutron image characteristics — such as neutron flux, image size, beam quality — measured experimentally and predicted numerically for beam tube, beam shutter and radiography cave are compared and discussed in detail in this paper. The experimental results show that thermal neutron flux at image plane is nearly uniform over an imaging area of 20.0-cm diameter and its magnitude ranges from 8.0×106 – 1.0×107 n/cm2-sec while the neutron-to-gamma ratio is 6.0×105 n/cm2-μSv.

Temperature Transients in a Triple Heat Exchanger

1984 ◽  
Vol 198 (3) ◽  
pp. 145-154
Author(s):  
P C Chiu ◽  
E H K Fung
Keyword(s):  
Heat Exchanger ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Solar Heating ◽  
Inlet Temperature ◽  
Exchange Processes ◽  
Flow Experiments ◽  
Heating Plant ◽  
Reactor Power Plant ◽  
Secondary Fluid

A triple heat exchanger, so called because there are three heat exchange processes taking place in it, was built to simulate the system behaviour of a nuclear reactor power plant or a solar heating plant which is characterized by the two circulating loops of the fluid flow. Experiments were carried out to study the temperature transients under disturbances in secondary fluid inlet temperature and power output from immersion heaters. Numerical results were obtained from the weighted residual formulation of the proposed dynamic model and they were shown to be in general agreement with the two sets of experimental responses.

NSGA-II algorithm-based LQG controller design for nuclear reactor power control

2022 ◽  
Vol 169 ◽  
pp. 108931
Author(s):  
Jiaoshen Xu ◽  
Hui Tang ◽  
Xin Wang ◽  
Ge Qin ◽  
Xin Jin ◽  
...  
Keyword(s):  
Power Control ◽  
Nuclear Reactor ◽  
Controller Design ◽  
Reactor Power ◽  
Nsga Ii

The Characteristics Study of Helium-Xenon Mixture in Closed Brayton Cycle for Space Nuclear Reactor Power

2018 ◽  
Author(s):  
Xie Yang ◽  
Lei Shi
Keyword(s):  
Physical Properties ◽  
Pressure Loss ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Loss Coefficient ◽  
Working Fluid ◽  
System Efficiency ◽  
Pressure Loss Coefficient ◽  
Molar Weight ◽  
Xenon Mixture

Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

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