Coupling supercritical carbon dioxide (S-CO2) Brayton cycle with Gen-IV reactor concepts could bring advantages of high compactness and efficiency. This study aims to design proper simple and recompression S-CO2 Brayton cycles working as the indirect cooling system for a mediate-temperature lead fast reactor and quantify the Brayton cycle performance with different heat rejection temperatures (from 32&#xb0;C to 55&#xb0;C) to investigate its potential use in different scenarios, like arid desert areas or areas with abundant water supply. High-efficiency S-CO2 Brayton cycle could offset the power conversion efficiency decrease caused by low core outlet temperature (which is 480&#xb0;C in this study) and high compressor inlet temperature (which varies from 32&#xb0;C to 55&#xb0;C in this study). A thermodynamic analysis solver is developed to provide the analysis tool. The solver includes turbomachinery models for compressor and turbine and heat exchanger models for recuperator and precooler. The optimal design of simple Brayton cycle and recompression Brayton cycle for the lead fast reactor under water-cooled and dry-cooled conditions are carried out with consideration of recuperator temperature difference constraints and cycle efficiency. Optimal cycle efficiencies of 40.48&#x0025; and 35.9&#x0025; can be achieved for the recompression Brayton cycle and simple Brayton cycle under water-cooled condition. Optimal cycle efficiencies of 34.36&#x0025; and 32.6&#x0025; can be achieved for the recompression Brayton cycle and simple Brayton cycle under dry-cooled condition (compressor inlet temperature equals to 55&#xb0;C). Increasing the dry cooling flow rate will be helpful to decrease the compressor inlet temperature. Every 5&#xb0;C decrease in the compressor inlet temperature will bring 1.2&#x0025; cycle efficiency increase for the recompression Brayton cycle and 0.7&#x0025; cycle efficiency increase for the simple Brayton cycle. Helpful conclusions and advises are proposed for designing the Brayton cycle for mediate-temperature nuclear applications in this paper.

Pan Wu

Chuntian Gao

Yanping Huang

Dan Zhang

Jianqiang Shan

Supercritical CO2 Brayton Cycle Design for Small Modular Reactor with a Thermodynamic Analysis Solver

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls&#x2009;&#x2b;&#x2009;pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube&#x2009;&#x2b;&#x2009;brass pellets and zircaloy-4 tube&#x2009;&#x2b;&#x2009;simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Young-Hwan Kim

Yung-Zun Cho

Jin-Mok Hur

Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

Atmospheric dispersion modelling and radiological safety analysis have been performed for a postulated accident scenario of a generic VVER-1000 nuclear power plant using the HotSpot Health Physics code. The total effective dose equivalent (TEDE), the respiratory time-integrated air concentration, and the ground deposition concentration are calculated considering site-specific meteorological conditions. The results show that the maximum TEDE and ground deposition concentration values of 3.69E&#x2009;&#x2013;&#x2009;01&#x2009;Sv and 3.80E&#x2009;&#x2b;&#x2009;06 kBq/m2 occurred at downwind distance of 0.18&#x2009;km from the release point. This maximum TEDE value is recorded within a distance where public occupation is restricted. The TEDE values at distances of 5.0&#x2009;km and beyond where public occupation is likely to be found are far below the annual regulatory limits of 1&#x2009;mSv from public exposure in a year even in the event of worse accident scenario as set in IAEA Safety Standard No. GSR Part 3; no action related specifically to the public exposure is required. The released radionuclides might be transported to long distances but will not have any harmful effect on the public. The direction of the radionuclide emission from the release point is towards the north east. It is observed that the organ with the highest value of committed effective dose equivalent (CEDE) appears to be the thyroid. It was followed by the bone surface, lung, red marrow, and lower large intestine wall in order of decreasing CEDE value. Radionuclides including I-131, I-133, Sr-89, Cs-134, Ba-140, Xe-133, and Xe-135 were found to be the main contributors to the CEDE.

K. Gyamfi

S. A. Birikorang

E. Ampomah-Amoako

J. J. Fletcher

Radiological Safety Analysis for a Hypothetical Accident of a Generic VVER-1000 Nuclear Power Plant

A method for solidifying spent tributyl phosphate and kerosene (TBP/OK) organic liquids in a phosphate acid-based geopolymer (PAG) was investigated. The TBP/OK emulsion containing tween 80 (T80), TBP/OK organic liquids, and H3PO4 was prepared. The TBP/OK emulsion was mixed with metakaolin to obtain solidified TBP/OK forms (SPT). The compressive strength of the SPT was up to 59.19&#x2009;MPa when the content of TBP/OK was 18&#x0025;. The loss of compressive strength of SPT was less than 10&#x0025; after immersion and less than 25&#x0025; after freeze-thaw treatment. The final setting time was 40.0&#x2009;h, and the shrinkage of SPT was nearly 3&#x0025;. The leaching test indicated that the release of TBP/OK from hardened SPT was limited. Characterization of SPT suggested that solidification of TBP/OK using PAG occurred by physical encapsulation.

Teng Dong

Shuibo Xie

Jingsong Wang

Guodong Zhao

Qingchun Song

Solidification and Stabilization of Spent TBP/OK Organic Liquids in a Phosphate Acid-Based Geopolymer

This paper focuses on fluid forces acting on a confined cylinder subjected to axial flow in application to fuel assembly dynamic behavior. From the literature, it is difficult to estimate the damping induced by the flow. Therefore, it is proposed to study numerically the damping fluid forces on a cylinder for various parameters. It has been observed that it increases with the smaller confinement and with the presence of an obstacle and decreases when the Reynolds number increases. Larger values correspond to a greater contribution of pressure forces. Dynamic simulations are compared to the steady ones and give different values, but the order of magnitude and general trend remain the same. Therefore, steady simulations are suitable to have a rough estimation of drag coefficients in dynamics.

Guillaume Ricciardi

Numerical Investigation of Fluid Forces Acting on a Confined Cylinder with Obstacle Subjected to Axial Flow

About this journal

Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.

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Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.

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Science and Technology of Nuclear Installations

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Science and Technology of Nuclear Installations