• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1560-1569
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• 2022

Liquid fueled molten salt reactors (MSRs) have seen renewed interest because of their inherent safety features, higher thermal efficiency and potential for efficient thorium utilisation for power generation. Thorium fluoride is one of the salts used in liquid fueled MSRs employing Th-U cycle. In the present study, ThF₄ was prepared by hydro-fluorination of ThO₂ using anhydrous HF gas. Process parameters viz. bed depth, hydrofluorination time and hydrofluorination temperature, were optimized for the preparation of ThF₄ in a static bed reactor setup. The products were characterized with X-Ray diffraction and experimental conditions for complete conversion to ThF₄ were established which also corroborated with the yield values. Hydrofluorination of ThO₂ at 450 ℃ for half an hour at a bed depth of 6 mm gave the best result, with a yield of about 99.36% ThF₄. No unconverted oxide or any other impurity was observed. Rietveld refinement was performed on the XRD data of this ThF₄, and Chi2 value of 3.54 indicated good agreement between observed and calculated profiles.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3493-3505
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• 2023

Reactor containment buildings serve as the last barrier to prevent radioactive leakage due to accidents and their safety is crucial in overpressurization conditions. Thus, the Regulatory Guide (RG) 1.216 has mentioned the global strain as one of failure criteria in the free-field for cylindrical prestressed concrete containment vessels (PCCV) subject to internal pressure. However, there is a limit that RG 1.216 shows the free-field without the specific locations of failure criteria and also the global strain corresponding to only azimuth 135° has been mentioned in NUREG/CR-6685, regardless of the elevations of the structure. Therefore, in order to reevaluate the failure criteria of the 1:4 scaled PCCV, the high fidelity simulation model based on the experimental test was significantly validated in this study, and it was interesting to find that the experimental and numerical result was very close to each other. In addition, for the consideration of the material uncertainties, the Latin hypercube method was used as a statistical approach. Consequently, it was revealed that the radial displacements of various azimuth area such as 120°, 135°, 150°, 180° and 210° at elevations 4680 mm and 6,200 mm can represent as the global deformation at the free-field, obtained from the statistical approach.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4359-4372
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• 2022

Turbulent mixing in buoyant flows is an essential mechanism involved in many scenarios related to nuclear safety in nuclear power plants. Comprehensive understanding and accurate predictions of turbulent buoyant flows in the reactor are of crucial importance, due to the function of mitigating the potential detrimental consequences during postulated accidents. The present study uses URANS methodology to investigate the buoyancy-influenced flows in the reactor pressure vessel under the main steam line break accident scenarios. With a particular focus on the influence of turbulent heat flux closure models, various combinations of two turbulence models and three turbulent heat flux models are utilized for the numerical simulations of three ROCOM tests which have different characteristic features in terms of the flow rate and fluid density difference between loops. The simulation results are compared with experimental measurements of the so-called mixing scalar in the downcomer and at the core inlet. The study shows that the anisotropic turbulent heat flux models are able to improve the accuracy of the predictions under conditions of strong buoyancy whilst in the weak buoyancy case, a major role is played by the selected turbulence models with essentially a negligible influence of the turbulent heat flux closure models.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• Journal of Microbiology and Biotechnology
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• v.33 no.4
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• pp.419-429
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• 2023

Ultraviolet C (UV-C, 200-280 nm) light has germicidal properties that inactivate a wide range of pathogenic and spoilage microorganisms. UV-C has been extensively studied as an alternative to thermal decontamination of fruit juices. Recent studies suggest that the efficacy of UV-C irradiation in reducing microorganisms in fruit juices is greatly dependent on the characteristics of the target microorganisms, juice matrices, and parameters of the UV-C treatment procedure, such as equipment and processing. Based on evidence from recent studies, this review describes how the characteristics of target microorganisms (e.g., type of microorganism/strain, acid adaptation, physiological states, single/composite inoculum, spore, etc.) and fruit juice matrices (e.g., UV absorbance, UV transmittance, turbidity, soluble solid content, pH, color, etc.) affect the efficacy of UV-C. We also discuss the influences on UV-C treatment efficacy of parameters, including UV-C light source, reactor conditions (e.g., continuous/batch, size, thickness, volume, diameter, outer case, configuration/arrangement), pumping/flow system conditions (e.g., sample flow rate and pattern, sample residence time, number of cycles), homogenization conditions (e.g., continuous flow/recirculation, stirring, mixing), and cleaning capability of the reactor. The collective facts indicate the immense potential of UV-C irradiation in the fruit juice industry. Existing drawbacks need to be addressed in future studies before the technique is applicable at the industrial scale.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1791-1801
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• 2023

The multi-physics coupled methodologies that have been widely used to analyze the complex process occurring in nuclear reactors have also been used to the R&D of numerical reactors. The advancement in the field of computer technology has helped in the development of these methodologies. Herein, we report the integration of ADPRES code and RELAP5 code into the SALOME-ICoCo framework to form a multi-physics coupling platform. The platform exploits the supervisor architecture, serial mode, mesh one-to-one correspondence and explicit coupling methods during analysis, and the uncertainty analysis tool URANIE was used. The correctness of the platform was verified through the NEACRP-L-335 benchmark. The results obtained were in accordance with the reference values. The platform could be used to accurately determine the power peak. In addition, design margins could be gained post uncertainty analysis. The initial power, inlet coolant temperature and the mass flow of assembly property significantly influence reactor safety during the rod ejections accident (REA).

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4628-4636
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• 2023

Cables are indispensable in nuclear power plants for transmitting data measured by various types of detectors, such as self-powered neutron detectors (SPNDs). These cables will generate disturbing signals that must be accurately distinguished and eliminated. Given that the cable current is not very significant, previous research has focused on SPND, with little attention paid to cable evaluation and validation. This paper specifically focuses on the quantitative analysis of cables and proposes a theoretical model to predict cable noise. In this model, the reaction characteristics between irradiated neutrons and cables were discussed thoroughly. Based on the Monte Carlo method, a comprehensive simulation approach of neutron sensitivity was introduced and long-term irradiation experiments in a heavy water reactor (HWR) were designed to verify this model. The theoretical results of this method agree quite well with the experimental measurements, proving that the model is reliable and exhibits excellent accuracy. The experimental data also show that the cable current accounts for approximately 0.2% of the total current at the initial moment, but as the detector gradually depletes, it will contribute more than 2%, making it a non-negligible proportion of the total signal current.