Reliability Assessment for a Safety-Related Digital Reactor Protection System Using Event-Tree/Fault-Tree (ET/FT) MethodRead the full article
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.
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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Manufacturing and Testing of an In Situ Stretching Sample Environment Equipment for Neutron Scattering Experiments
Neutron scattering technology is one of the most promising ways to observe microstructures of different materials. As a powerful microstructure characterization technology, neutron scattering is widely used in many disciplines. With the help of sample environment equipment, the microstructure detection for materials in various application scenarios can be further realized. In order to detect the microstructure changes of materials under different tensile conditions, an in situ stretching sample environment equipment for neutron scattering experiments was designed and manufactured. Stretching force holding test, sample breaking test, and vacuum maintaining test were carried out. In those tests, a tensile force holding test with no less than 5 hours, a breaking test with a screw bolt as the sample, and a vacuum leakage rate test with no less than 5 hours were obtained, respectively. Through analyzing values obtained, it is shown that the developed prototype of the sample environment equipment is able to meet the experiment requirements. The present prototype provides a reference for further development of sample environment equipment for different application scenarios in neutron scattering experiment.
Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes
This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.
Numerical Study of the Air Ingress Accident of the HTR-PM and Possible Mitigation Measures
The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct of a high-temperature gas-cooled reactor is an extremely hypothetical accident, which could cause the air to enter into the primary circuit and react with graphite in the reactor core. The performance of the HTR-PM plant under this extremely hypothetical accident has been studied by the system code TINTE in this work. The results show that the maximum fuel temperature will not reach the temperature design limitation, and the graphite oxidation will not cause unacceptable consequences even under some conservative assumptions. Moreover, nitrogen and helium injected from the fuel charging tube were studied as the possible mitigation measures to further alleviate the consequences of this air ingress accident. The preliminary results show that only the flow rate of nitrogen injected reaches a certain value, which can effectively alleviate the consequences, while for helium injection, both high and small flow rate can prevent or cut off the natural circulation and alleviate the consequences. The reason is that helium is much lighter than nitrogen, and the density difference between the coolant channel and the reactor core is small when helium is injected. Considering the injection velocity, the total usage amount, and the start time of gas injection, helium injected with a small flow rate is suggested.
Assessment of Lifetime Attributable Risk for Public Health Sustainability from the Fukushima Accident
The study aimed at reassessing the radiological risk from exposure to ionizing radiation from Fukushima Daiichi nuclear power plant accident. The estimated risks from the study were compared with World Health Organization (WHO) risk assessment estimates for validity and verification. A Radiation Risk Assessment Tool (RadRAT) was used to estimate Lifetime Attributable Risks (LAR) of cancer upon exposure. All solid cancers, leukemia, and thyroid cancer risks for ages of 1, 10, and 20 years (infants, children, and adults) in 100,000 persons at exposure were estimated. For solid cancers, LAR (10−2) was estimated in a range of 0.223∼0.668 and 0.345∼1.24 for males and females, respectively, whereas the LAR (10−2) for leukemia was estimated at 0.0155∼0.055 and 0.0118∼0.0375 for males and females, respectively. LAR (10−2) for thyroid cancer ranged from 0.0722∼0.545 and 0.0369∼0.265, respectively.
A New Precursor Integral Method for Solving Space-Dependent Kinetic Equations in Neutronic and Thermal-Hydraulic Coupling System
The accurate prediction of the neutronic and thermal-hydraulic coupling system transient behavior is important in nuclear reactor safety analysis, where a large-scale nonlinear coupling system with strong stiffness should be solved efficiently. In order to reduce the stiffness and huge computational cost in the coupling system, the high-performance numerical techniques for solving delayed neutron precursor equation are a key issue. In this work, a new precursor integral method with an exponential approximation is proposed and compared with widely used Taylor approximation-based precursor integral methods. The truncation errors of exponential approximation and Taylor approximation are analyzed and compared. Moreover, a time control technique is put forward which is based on flux exponential approximation. The procedure is tested in a 2D neutron kinetic benchmark and a simplified high-temperature gas-cooled reactor-pebble bed module (HTR-PM) multiphysics problem utilizing the efficient Jacobian-free Newton–Krylov method. Results show that selecting appropriate flux approximation in the precursor integral method can improve the efficiency and precision compared with the traditional method. The computation time is reduced to one-ninth in the HTR-PM model under the same accuracy when applying the exponential integral method with the time adaptive technique.
Analyzing Design Considerations for Disassembly of Spent Nuclear Fuel during Head-End Process of Pyroprocessing
We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.