• 한국태양에너지학회 논문집
• /
• 제34권1호
• /
• pp.48-57
• /
• 2014

The transient and synchronization behaviors of a TA (thermo acoustic) laser pair were investigated experimentally for various crossing angles and different separation distances between the laser openings. Sound waves generated by the lasers were measured and analyzed at or near the focusing point by means of microphones, SPL meters, and a commercial software called Signal-Express. The two TA lasers were acoustically coupled through the air mass between their openings, and the only mode-locking operation that could be achieved was the one that was nearly $180^{\circ}C$ out of phase. The time to achieve synchronization was found to be dependent upon the initial mistuning of the frequencies and the crossing angle between the laser axes. The synchronization process could also be accelerated by turning on the laser with the lower power input first.

• Nuclear Engineering and Technology
• /
• 제51권6호
• /
• pp.1532-1539
• /
• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
• /
• 제49권3호
• /
• pp.466-475
• /
• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• 설비공학논문집
• /
• 제10권4호
• /
• pp.398-410
• /
• 1998

Two-phase flow phenomena are observed in many industrial facilities and make much importance of optimum design for nuclear power plant and the liquid transportation system. The particular flow pattern depends on the conditions of pressure, flow velocity, and channel geometry. However, the research on drag reduction in two-phase flow is not intensively investigated. Therefore, experimental investigations have been carried out to analyze the drag reduction and void fraction by polymer addition in the two-phase flow system. We find that the polymer solution changes the characteristic of two-phase flow. The peak position of local void friction moves from tile wall of the pipe to the center of the pipe when polymer concentration increase. And then we predict that it is closely related with the frau reduction.

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.1013-1019
• /
• 2021

This paper reviewed the temporal changes in the public opinions on nuclear energy in Korea with a big data analysis of nuclear energy-related articles and their comments posted on the portal site NAVER. All articles that included at least one of "nuclear energy," "nuclear power plant (NPP)," "nuclear power phase-out," or "anti-nuclear" in their titles or main text were extracted from those posted on NAVER in January 2010-December 2019. First, we performed annual word frequency analysis to identify what words had appeared most frequently in the articles. For that period, the most frequent words were "NPP," "nuclear energy," and "energy." In addition, "safety" has remained in the upper ranks since the Fukushima NPP accident. Then, we performed sentiment analysis of the pre-processed articles. The sentiment analysis showed that positive-tone articles have been reported more frequently than negativetone over the entire analysis period. Last, we performed sentiment analysis of the comments on the articles to examine the public's intention regarding nuclear issues. The analysis showed that the number of negative comments to articles each month-irrespective of positive or negative tone-was always larger than that of positive comments over the entire analysis period.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1705-1716
• /
• 2021

Bangladesh, as a newcomer country, is expecting to start her nuclear power journey by 2022. Due to evident reasons, newcomer nuclear countries face several key challenges concerning the development of national nuclear safety infrastructure. The paper investigates the status of the 7 key safety infrastructure issues out of the 19 and readiness of the supportive organizations, laboratories, and workforces following the International Atomic energy Agency's status evaluation guide at milestone 3 and foreign countries' practice. Much progress has been achieved at phase 3 regarding the establishments of a few Acts, a regulator, and an operator. However, comprehensive regulatory frameworks, skilled workforces, establishments of a few supportive organizations, and laboratories for managing environmental radioactivity, radiological accidents, and radioactive wastes are yet to ready. Several suggestions are made for establishing and expediting radiation monitoring laboratories, a radiological emergency management center, a radioactive waste management company, and technical support organizations for the safety infrastructure. To avoid perceived risks, policymakers and competent authorities need to emphasize creating an optimized safety infrastructure before commissioning and operating the 1^st nuclear power plant safely, securely, and cost-sustainably.

• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.248-257
• /
• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 추계학술발표회논문집(1)
• /
• pp.295-300
• /
• 1996

To investigate the effects of the side-orifice on the pressure drop for single-phase flow, a series of experiments have been carried out with 16 different downstream test sections with various combinations of side-orifice shapes, different numbers of side-orifices, and different arrangements of the side-orifice using water as a working fluid. From the measurements of the pressure drop and the flow rate, the pressure loss coefficient of the side-orifice(s) has been evaluated. Based on the total number of 529 present data, an empirical correlation for the pressure loss coefficient has been developed in terms of Reynolds number and geometric parameters, such as area ratio, equivalent diameter, leading edge, and average width of side-orifice.

• Journal of Mechanical Science and Technology
• /
• 제17권9호
• /
• pp.1297-1315
• /
• 2003

This study deals with the free vibration of two identical circular plates coupled with a bounded or unbounded fluid. An analytical method based on the finite Fourier-Bessel series expansion and Rayleigh-Ritz method is suggested. The proposed method is verified by the finite element analysis using commercial program with a good accuracy The case of bounded or unbounded fluid is studied for the effect on the vibration characteristics of two circular plates. Also, the effect of gap between the plates on the fluid-coupled natural frequencies is investigated.

• Nuclear Engineering and Technology
• /
• 제33권3호
• /
• pp.315-326
• /
• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.