Article of the Year 2020
Supercritical CO2 Brayton Cycle Design for Small Modular Reactor with a Thermodynamic Analysis SolverRead 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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Development and Experimental Validation for Quantifying the Moisture Carryover in a Moisture Separator Using an Air/Water Test Facility
The moisture carryover (MCO) of the primary separator in a steam generator is the most important design parameter to ensure high efficiency in a steam generator. There is an inherent limitation to experimentally evaluate the MCO under the prototype conditions. In this study, the air/water test facility was constructed based on the similarity law, and a new isokinetic system was developed to quantify the MCO. Several experiments were performed for the mass quality ranging from 0.315 to 0.382. The accuracy and versatility of the experimental method was verified experimentally using a full and half scale of separators. The test results were compared with the prototype results. It was proved to be a reliable experimental method for evaluating the MCO of the moisture separator.
Study on the Sensitivity and Uncertainty of Nuclear Data to the Sodium-Cooled Linear Breed-and-Burn Fast Reactor Using SCALE6.2 Code
Previously, the neutronics design of a small and compact linear breed-and-burn fast reactor (B&BR) was completed. The reactor produces 400 MWth power, and it can operate with excess reactivity of less than 1$ for more than 50 years without refuelling. As the blanket fuel, the spent nuclear fuel (SNF) from existing light water reactors (LWRs) is used to reduce the burden from the problematic long-lived isotopes in SNF. However, by loading massive nuclides at the initial core, the impact of nuclear data uncertainty on the reactivity calculation results of SNF-fuelled B&BR at the beginning of life (BOL) is expected to be significant because these nuclides have different credentials in evaluated nuclear data libraries. In this study, the impact of nuclear library uncertainty from ENDF/B-VII.0 and ENDF/B-VII.1 on reactivity calculation of B&BR is evaluated using the continuous-energy TSUNAMI-3D module in the SCALE6.2 code package. The uncertainty of reactivity calculation results of B&BR caused by the inaccuracy of two libraries is significant (more than 2000 pcm), mainly from the uncertainty of 235,238U and 56Fe cross section. The energy-dependent sensitivity profiles show that they are significant at the fast energy range. The uncertainty of coolant void reactivity (CVR) is about 18%, and that of fuel temperature coefficient (FTC) is about 15% of the reactivity effect. The top five contributions for CVR accounted for elastic scattering of 238U, capture of 235,238U, and elastic scattering of 23Na and 56Fe. Meanwhile, the top contributors for FTC were accounted for elastic scattering of 238U and 56Fe, capture of 235U, and elastic scattering of 94Zr and 57Fe. It is highly recommended to improve the accuracy of those isotopes’ cross sections at the high energy range to provide a more reliable reactivity calculation for the fast system.
Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1
Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.
Uranium Decontamination from Waste Soils by Chlorination with ZrCl4 in LiCl-KCl Eutectic Salt
The dissolution behavior of U, contained in the soils, was examined through chlorination with ZrCl4 to reduce the U concentration to clearance levels. Natural soils, composed of Si, Al, and approximately 2 ppm U, acted as surrogates for the contaminated soils. A salt mixture of LiCl-KCl-ZrCl4 was prepared in an Al2O3 crucible at 500°C, and SiO2 or natural soils were loaded for the chemical reactions. The reaction of SiO2 and Al2O3 with ZrCl4 was monitored by cyclic voltammetry, and no obvious change was observed. The results showed that SiO2 and Al2O3 were stable against ZrCl4. The reaction of natural soils with ZrCl4 indicated that the U content decreased from 2 to 1.2 ppm, while ∼0.4 ppm U appeared in the salt. Thus, the U, in the soils, dissolved into the salt by chlorination with ZrCl4. Therefore, based on these results, a new method to remediate U-contaminated soil wastes by chlorination with ZrCl4, followed by electrorefining of U, is suggested.
Inerting Strategy for a Demonstration-Scale Hot Cell Facility Based on Experiences from Pilot-Scale Argon Cell Facility Operation and CFD Analysis
Pyroprocessing is being developed at Korea Atomic Energy Research Institute (KAERI), and in recent years, all process equipment required for integrated processes have been examined in the PyRoprocess-integrated Inactive DEmonstration (PRIDE) facility. Based on the successful operation of a pilot-scale facility, a conceptual design for this scale-up facility was actualized. Implementing a “demonstration-scale” hot cell facility is challenging as it is intended to supersede PRIDE and satisfy the increased requirements of larger-scale facilities. This study focused on an inerting strategy for a larger-scale (demonstration-scale) hot cell facility to achieve conditions equivalent to those in a pilot-scale gas-tight argon cell facility. The study applies the inerting strategy to a demonstration-scale hot cell facility beyond that of the currently existing pilot-scale hot cell facility and performs computational fluid dynamics (CFD) simulation with various flow rates to determine an appropriate approach for inerting the target facility. To this end, practical constraints on the simulation are introduced based on experiences from the existing pilot-scale facility. The results show that the purging flow rate should be accurately predicted, and a variable flow rate should be applied to achieve hot cell inerting effectively. The required purging time and amount of inerting source are essential factors in the larger-scale hot cell facility. The study results can be helpful in the design of large hot cell facilities operated under inert conditions.
Research and Implementation of SVDU Simulator Based on Emulation Technology
The safety video display unit (SVDU), as the display machine of the reactor protection system, performs the functions of displaying the reactor’s safety parameters and sending safety control commands. In order to meet the needs of nuclear power safety-level digital control system (DCS), like designing verification, operator training, and accidental drills, for the SVDU in the NASPIC platform developed independently by China National Nuclear Corporation, a virtual embedded system technology based on the micro x86 industrial host is proposed to make software simulation to the SVDU physical controller with exactly the same hardware appearance of the original one. The SVDU stimulator realized by this research can achieve 100% simulation of the logical functions of the SVDU physical equipment and the synchronized upgrading function between stimulator and the real equipment. With the development of multiple engineering application requirements, such as configuration verification and operator training, this stimulator has been applied in several virtual security level DCS projects.