• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.23 no.11
• /
• pp.1390-1398
• /
• 1999

A numerical analysis for ROT sparger of PWR(Pressurized Water Reactor) is carried out. Computation is performed to investigate the flow characteristics as the change of design factor. As the result of this study, RDT sparger's flow resistance coefficient is K=3.53 at the present design condition if engineering mar&in is considered with 20%, and flow ratio into branch pipe is $Q_s/Q_i=0.41$. Velocity distribution at exit is not uniform because of separation in branch pipe. In the change of inlet flow rate and section area ratio of branch pipe for main pipe, flow resistance coefficient is increased as $Q_s/Q_i$ decreasing, but in the change of branch angle and outlet nozzle diameter of main pipe, flow resistance coefficient is decreased as $Q_s/Q_i$ decreasing. As the change rate of $Q_s/Q_i$ is the larger, the change rate of flow resistance coefficient is the larger. The change rate of pressure loss is the largest change as section area ratio changing. The optimal design condition of sparger is estimated as the outlet nozzle diameter ratio of main pipe is $D_s/D_i=0.333$, the section area ratio is $A_s/A_i=0.2$ and the branch angle is ${\alpha}=55^{\circ}$.

• Korean Journal of Materials Research
• /
• v.14 no.10
• /
• pp.713-718
• /
• 2004

To study the dynamic strain aging behavior of Zr-0.4Sn-1.5Nb-0.2Fe sample tube for nuclear fuel cladding in the range of pressurized water reactor (PWR) operation temperature, the tensile tests of the tube specimens, which had been finally heat-treated at $470^{\circ}C\;and\;510^{\circ}C$, had been carried out with the strain rate $1.67{\times}10^{-2}/s\;and\;8.33{\times}10^{-5}/s$ at the various temperatures from room temperature to $500^{\circ}C$. It was observed that the elongation of the specimens got shortened as the temperature increased from $200^{\circ}C\;to\;340^{\circ}C$. The specimens that were finally heat-treated at $470^{\circ}C$ showed a plateau more remarkably on the plot of yield strength-temperature than those heat-treated at $510^{\circ}C$. In the range of $310\sim400^{\circ}C$, the strain rate sensitivity of the specimens finally heat-treated at $510^{\circ}C$ was $30.4\%\sim33.7\%$ lower but the work hardening exponent index of the specimens was a little higher than that without dynamic strain aging effect.

• Applied Chemistry for Engineering
• /
• v.10 no.2
• /
• pp.218-224
• /
• 1999

Continuous ion exchange characteristics of the synthetic coolant contained Ni, Co and Ag ions of low concentration in acidic-oxidizing conditions have been studied to suggest the guideline for the optimum operation of mixed-bed demincralizer during the shutdown period of a pressurized water reactor (PWR). In the effect of the form of cation resins on the removal capacity of metal ions, the performance of a $H^+$-form resin was about 6% higher than that of a $Li^+$-form resin. Mixed-bed of cation and anion resins in comparison with nonmixed-bed of them, had no affected on the removal capacity of metal ions but very slightly increased the slope of breakthrough curves of metal ions. In the effect related to acidic-oxidizing conditions of the coolant, the addition of boric acid very slightly decreased the slope of breakthrough curves of metal ions, while the addition of hydrogen peroxide slightly decreased the removal capacity of metal ions.

• Corrosion Science and Technology
• /
• v.15 no.2
• /
• pp.84-91
• /
• 2016

The transport and deposition of corrosion products in pressurized water nuclear reactor (PWR) steam generators have led to corrosion (SCC, denting etc.) problems. Lancing, mechanical cleaning and chemical cleaning have been used to reduce these problems. The methods of lancing and mechanical cleaning have limitations in removing corrosion products due to the structure of steam generator tubes. But high temperature chemical cleaning (HTCC) with EDTA is the most effective method to remove corrosion products regardless of the structure. However, EDTA in chemical cleaning aqueous solution and chemical cleaning environments affects the integrity of materials used in steam generators. The nuclear power plants have to perform the pre-test (also called as qualification test (QT)) that confirms the effect on the integrity of materials after HTCC. This is one of the series studies that assess the effect, and this study determines the effects of 20 % EDTA aqueous solution on defective tubes in high temperature chemical cleaning environments. The depth and magnitude of defects in steam generator (SG) tubes were measured by eddy current test (ECT) signals. Surface analysis and magnitude of defects were performed by using SEM/EDS. Corrosion rate was assessed by weight loss of specimens. The ECT signals (potential and depth %) of defective tubes increased marginally. But the lengths of defects, oxides on the surface and weights of specimens did not change. The average corrosion rate of standard corrosion specimens was negligible. But the surfaces on specimens showed traces of etching. The depth of etching showed a range on the nanometer. After comprehensive evaluation of all the results, it is concluded that 20 % EDTA aqueous solution in high temperature chemical cleaning environments does not have a negative effect on defective tubes.

• Nuclear Engineering and Technology
• /
• v.21 no.1
• /
• pp.1-11
• /
• 1989

Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.229-236
• /
• 2018

Accident-tolerant fuels (ATFs) are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding). This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc.) serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS) technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD), laser coating, or Chemical vapor deposition techniques (CVD), the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions ($500^{\circ}C$ steam, $1200^{\circ}C$ steam, and Pressurized water reactor (PWR) pressurization test) and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX), or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing.

• Nuclear Engineering and Technology
• /
• v.56 no.10
• /
• pp.4068-4076
• /
• 2024

In the event of severe accidents in pressurized water reactor (PWR) nuclear power plants, the potential leakage of radioactive aerosols through containment cracks poses a considerable radioactive hazard to the public. Understanding aerosol transport and retention in cracks helps reduce the conservatism and uncertainty of radioactive hazard assessment. Concrete cracks are recognized as a pivotal pathway for the leakage of radioactive aerosols, and several researchers have undertaken experimental investigations concerning the aerosol transport and retention in concrete cracks. However, the majority of these studies have rather low gas flow Reynolds numbers. In this work, an experimental setup is built to study aerosol transport and retention in concrete cracks under high Reynolds number flow. The TiO₂ aerosol with a mass median diameter of 1 ㎛ and two concrete crack specimens are used in experiments. The results of gas flow experiments indicate that the Reynolds number is capable of reaching 10547. Combining the flow experimental data and Suzuki's formula, the equivalent heights of these two crack specimens are approximated as 303.67 ㎛ and 231.48 ㎛. The experimental results indicate a notably high retention rate of aerosols, exceeding 0.8. Furthermore, under high Reynolds number flow, the retention rate varies over a relatively narrow range, with the larger the equivalent height of the crack resulting in a lower retention rate. The experimental results match well with the mechanistic analysis based on inertial deposition theory, demonstrating the rationality of the inertial deposition theory.

• Analytical Science and Technology
• /
• v.16 no.2
• /
• pp.117-124
• /
• 2003

Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

• Journal of the Korean Institute of Intelligent Systems
• /
• v.3 no.2
• /
• pp.18-35
• /
• 1993

The water level control system of the steam generator in a pressurized water reactor and its control problems are analysed. In this work the stable control strategy during the low power operation and transient states is studied. To solve the problem, a fuzzy logic control method is applied as a basic algorithm of the controller. The control algorithm is based on the operator's knowledges and the experiences of manual operation for water level control at the compact nuclear simulator set up in Korea Atomic Energy Research Institute. From a viewpoint of the system realization, the control variables and rules are established considering simpler tuning and the input-output relation. The control strategy includes the dynamic tuning method and employs a substitutional information using the bypass valve opening instead of incorrectly measured signal at the low flow rate as the fuzzy variable of the flow rate during the pressure control mode of the steam generator. It also involves the switching algorithm between the control valves to suppress the perturbation of water level. The simulation results show that both of the fine control action at the small level error and the quick response at the large level error can be obtained and that the performance of the controller is improved.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.10 no.11
• /
• pp.3260-3268
• /
• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.