• Nuclear Engineering and Technology
• /
• v.53 no.4
• /
• pp.1236-1249
• /
• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1686-1704
• /
• 2021

This study addresses the numerical simulation of the shield building of an AP1000 nuclear power plant (NPP) subjected to a large commercial aircraft impact. First, a simplified finite element model (F.E. model) of the large commercial Boeing 737 MAX 8 aircraft is established. The F.E. model of the AP1000 shield building is constructed, which is a reasonably simplified reinforced concrete structure. The effectiveness of both F.E. models is verified by the classical Riera method and the impact test of a 1/7.5 scaled GE-J79 engine model. Then, based on the verified F.E. models, the entire impact process of the aircraft on the shield building is simulated by the missile-target interaction method (coupled method) and by the ANSYS/LS-DYNA software, which is at different initial impact velocities and impact heights. Finally, the laws and characteristics of the aircraft impact force, residual velocity, kinetic energy, concrete damage, axial reinforcement stress, and perforated size are analyzed in detail. The results show that all of them increase with the addition to the initial impact velocity. The first four are not very sensitive to the impact height. The engine impact mainly contributes to the peak impact force, and the peak impact force is six times higher than that in the first stage. With increasing initial impact velocity, the maximum aircraft impact force rises linearly. The range of the tension and pressure of the reinforcement axial stress changes with the impact height. The perforated size increases with increasing impact height. The radial perforation area is almost insensitive to the initial impact velocity and impact height. The research of this study can provide help for engineers in designing AP1000 shield buildings.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.1964-1969
• /
• 2021

We developed a Dual-PPACs detector for fast neutron measurements that consists of two sets of PPAC: conventional PPAC and fission PPAC. A²³⁸U(U₃O₈) coating is placed in the fission PPAC's anode, which is used as the neutrons conversion layer. An experiment was performed to measure neutron time-of-flight (TOF) in which ²⁵²Cf spontaneous fission source was used. An excellent time resolution of 164ps has been observed at 6 mbar in isobutene gas. With the excellent time resolution of Dual-PPACs detector, exact neutron energy can be extracted from the timing measurement. The experimental detection efficiency was 1.9 × 10^-7, consistent with the efficiency of 2.5 × 10^-7 given by a Geant4 simulation. Ultimately, the results show that the Dual-PPACs detector is a suitable candidate for measuring fast neutrons in the future CiADS system.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2207-2220
• /
• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Nuclear Engineering and Technology
• /
• v.53 no.1
• /
• pp.44-60
• /
• 2021

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1037-1040
• /
• 2019

For the past 50 years, graphite has been widely used as a moderator, reflector and fuel matrix in different kinds of gas-cooled reactors. Resulting in approximately 250,000 metric tons of irradiated graphite waste. One of the most significant long-lived radioisotope from graphite reactors is carbon-14 ($^{14}C$) with a half-life of 5730 years, this makes it a huge concern for deep geologic disposal of nuclear graphite (NG). Considering the lifecycle of NG a number of waste management options have been developed, mainly focused on the achievement the radiological requirements for disposal. The existing approaches for recycling depend on the cost to be economically viable. In this new study, an affordable process to remove $^{14}C$ has been proposed using samples taken from the Nuclear Power Plant in Latina (Italy) which have been used to investigate the capability of organic and inorganic solvents in removing $^{14}C$ from exfoliated nuclear graphite, with the aim to design a practicable approach to obtain graphite for recycling or/and safety disposed as L& LLW.

• Nuclear Engineering and Technology
• /
• v.51 no.5
• /
• pp.1417-1427
• /
• 2019

In this study, we evaluated the performance of a commercial pixelated cadmium zinc telluride (CZT) detector for spectroscopy and identified its feasibility as a Compton camera for radiation monitoring in a nuclear power plant. The detection system consisted of a $20mm{\times}20mm{\times}5mm$ CZT crystal with $8{\times}8$ pixelated anodes and a common cathode, in addition to an application specific integrated circuit. The performance of the various radioisotopes $^{57}Co$, $^{133}Ba$, $^{22}Na$, and $^{137}Cs$ was evaluated. In general, the amplitude of the induced signal in a CZT crystal depends on the interaction position and material non-uniformity. To minimize this dependency, a drift time correction was applied. The depth of each interaction was calculated by the drift time and the positional dependency of the signal amplitude was corrected based on the depth information. After the correction, the Compton regions of each spectrum were reduced, and energy resolutions of 122 keV, 356 keV, 511 keV, and 662 keV peaks were improved from 13.59%, 9.56%, 6.08%, and 5%-4.61%, 2.94%, 2.08%, and 2.2%, respectively. For the Compton imaging, simulations and experiments using one $^{137}Cs$ source with various angular positions and two $^{137}Cs$ sources were performed. Individual and multiple sources of $^{133}Ba$, $^{22}Na$, and $^{137}Cs$ were also measured. The images were successfully reconstructed by weighted list-mode maximum likelihood expectation maximization method. The angular resolutions and intrinsic efficiency of the $^{137}Cs$ experiments were approximately $7^{\circ}-9^{\circ}$ and $5{\times}10^{-4}-7{\times}10^{-4}$, respectively. The distortions of the source distribution were proportional to the offset angle.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.987-995
• /
• 2019

Steam jet condensation is of great importance to pressure suppression containment and automatic depressurization system in nuclear power plant. In this paper, the condensation processes of sonic steam jet in a quiescent subcooled pool are recorded and analyzed, more precise understanding are got in direct contact condensation. Experiments are conducted at atmospheric pressure, and the steam is injected into the subcooled water pool through a vertical nozzle with the inner diameter of 10 mm, water temperature in the range of $25-60^{\circ}C$ and mass velocity in the range of $320-1080kg/m^2s$. Richardson number is calculated based on the conservation of momentum for single water jet and its values are in the range of 0.16-2.67. There is no thermal stratification observed in the water pool. Four condensation regimes are observed, including condensation oscillation, contraction, expansion-contraction and double expansion-contraction shapes. A condensation regime map is present based on steam mass velocity and water temperature. The dimensionless steam plume length increase with the increase of steam mass velocity and water temperature, and its values are in the range of 1.4-9.0. Condensation heat transfer coefficient decreases with the increase of steam mass velocity and water temperature, and its values are in the range of $1.44-3.65MW/m^2^{\circ}C$. New more accurate semi-empirical correlations for prediction of the dimensionless steam plume length and condensation heat transfer coefficient are proposed respectively. The discrepancy of predicted plume length is within ${\pm}10%$ for present experimental results and ${\pm}25%$ for previous researchers. The discrepancy of predicted condensation heat transfer coefficient is with ${\pm}12%$.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1696-1707
• /
• 2019

This paper conducts a qualitative policy analysis of current challenges to safety culture and security culture in Southeast Asia and emerging best practices in Northeast Asia that are aimed at strengthening both cultures. It analyses lessons, including strengths and limitations, that can be derived from Northeast Asian states, given the long history of nuclear energy in South Korea, China and Japan. It identifies and examines best practices from Northeast Asia's Nuclear Security Centres of Excellence in terms of boosting nuclear security culture and their relevance for Southeast Asia. The paper accentuates the important role of the State in adopting policy and regulatory frameworks and in institutionalising nuclear education and training programmes to deepen the safety-security cultures. Best practices in and challenges to developing a nuclear safety culture and a security culture in East Asia are examined using three frameworks of analysis (i) a comprehensive nuclear policy framework; (ii) a proactive and independent regulatory body; and (iii) holistic nuclear education and training programmes. The paper argues that Southeast Asian states interested in harnessing nuclear energy and/or utilising radioactive sources for non-power applications must develop a comprehensive policy framework on developing safety and security cultures, a proactive regulatory body, and holistic nuclear training programmes that cover both technical and human factors. Such measures are crucial in order to mitigate human errors that may lead to radiological accidents and nuclear security crises. Key lessons from Japan, South Korea and China such as best practices and challenges can inform policy recommendations for Southeast Asia in enhancing safety-security cultures.

• Journal of Digital Convergence
• /
• v.17 no.4
• /
• pp.65-73
• /
• 2019

This study performed Panel Regression Analysis (PRA) with the debt ratio as a dependent variable and the ROE (return on equity), sales volume, current ratio, total capital, and Shanghai Containerized Freight Index (SCFI) as an independent variable. According to the GLS analysis, the current ratio to liabilities ratio was negative, and for sales, the ratio of liabilities was positive. Capital totals also had a negative impact on the debt ratio. However, ROE, unlike the hypothesis, had negative effects on the liability ratio, and the SCFI index was not significant. As implications of this research, the company confirmed that its sales increased as the debt ratio of global shipping companies rose, achieving economies of scale. However, it was confirmed that the actual size of the economy through the injection of other capital would help increase sales but not affect net profit. Shipping companies should expand their business power and secure large container vessels to secure credibility of shippers. In the future research, an analysis considering exchange rate, global economic growth rate, and manufacturing production index is needed.