• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• Corrosion Science and Technology
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• v.23 no.3
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• pp.203-214
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• 2024

A heat exchanger is a device designed to transfer heat between two or more fluids. In a vehicle's thermal management system, Al heat exchangers play a critical role in controlling and managing heat for efficient and safe operation of the engine and other components. The fluid used to prevent heat exchangers from overheating the engine is mostly tap water. Heat exchange performance can be maintained at sub-zero temperatures using a solution mixed with antifreeze. Although the fluid flowing through the heat exchanger can reduce the temperature inside the engine, it also has various problems such as cavitation corrosion. Cavitation corrosion characteristics in tap water and corrosion characteristics were evaluated in this study when antifreeze was added for test specimens where AA4045 was cladded on the inner surface of AA3003 tubes of a fin-type heat exchanger. The cavitation corrosion resistance of AA3003 was found to be superior to that of AA4045 regardless of the test solution due to higher corrosion resistance and hardness of AA3003 than those of AA4045. The cavitation corrosion rate of Al alloys increased with the addition of antifreeze.

• Resources Recycling
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• v.21 no.5
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• pp.18-30
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• 2012

Zirconium is one of the most important material used as cladding of fuel rods in nuclear reactors because of its high dimensional stability, good corrosion resistance and especially low neutron-absorbing cross section. However, Hf free nuclear grade Zr sponge is commercially produced by only three countries including USA, France and Russia. So, Zr has been thoroughly managed as a national strategic material in Korea. Most of the zirconium is used for Korean nuclear industry as nuclear fuel cladding materials manufactured from Hf free Zr alloy raw material. Also, there are some other applications such as alloying element and detonator. In this review, zirconium production and recycling technologies have been reviewed and current industrial status was also analyzed. And recent achievements in innovative reduction technologies such as electrolytic reduction process and molten oxide electrolysis were also introduced.

• Fire Science and Engineering
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• v.30 no.4
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• pp.121-127
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• 2016

This study analyzed the structural and electrical characteristics of an iron fittings type Corrugated Stainless Steel Tubing (CSST) damaged by the Primary Current Injection Test System (PCITS). CSST consists of cladding, tube, nuts, clamp ring, flare cap, socket, and ball valve. For an evaluation of the dielectric withstand voltage, the area between the live part and non-live part of the CCST shall withstand a voltage of 220 V AC for one minute. For an evaluation of the insulation performance by 500 V DC, it is required that the insulation exceed more than $1M{\Omega}$ before the temperature rise test, $0.3M{\Omega}$ after the test. Although the average resistance of the product was $11.5m{\Omega}$, that of the product damaged at a current of 130 A by the PCITS was $11.50m{\Omega}$. Furthermore, parts of the cladding were melted and black smoke appeared when a current of 130 A applied for 10 s. After 60 s, most parts were heated and turned red. At 120 s, the parts that turned red had widened. Although it did not form a normal distribution because the P value was 0.019 with a confidential interval of 95%, it revealed outstanding characteristics with an AD (Anderson-Darling) value of 0.896 and a standard deviation of 0.5573.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.111-122
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• 1989

With the Zircaloy-4 tube oxidized in high temperature (1323 K) steam for 5, 10, 30 and 60 minutes, the expanding copper mandrel test was carried out over a temperature range of 673-l173k at $\varepsilon\;=\;3.0\times10^5S\;^1$. The oxidation parameters $(K_i)$ in the present study were linearly proportional to square root of time $(Ki= \delta_{kit})$ and their rate constants ($\delta_{ki}$) are 0.281, 2.82, and 2.313 for weight gain and thickness of Zr02 and $\alpha$(0) layer, respectively. Activation energy for high temperature (873-1073k) plastic deformation of Zircaloy-4 increases from 251 KJ/mol to 323 KJ/mol with increase in oxidation time from 5 minutes to 60 minutes due to the high strengthened Zr02. With the oxide layer thickness [K ; expressed in "Equivalent Cladding Reacted" (ECR,%)] and the yield stress obtained from the mandrel test, an empirical relation was derived as ($\sigma/C)^n=K^mexp$ (Q/RT) with n=6.9, m=5.7, C=0.155, 0.138, 0.051, and 0.046 MPa for Q=251, 258, 316, 323 KJ/mol, respectively.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.7 no.2
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• pp.113-119
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• 2006

Brazing is an important manufacturing process in the fabrication of Heavy Water Reactor fuel rods, in which bearing and spacer pads are joined to Zircaloy-4 cladding tubes. The physical vapor deposition(PVD) technique is currently used to deposit metallic Be on the surfaces of pads as a filler metal. Amorphous Zr-Be binary alloys which are manufactured by rapid solidification process are under developing to substitute the conventional PVD-Be coating. In the present study, brazed joint with PVD and amorphous alloys of $Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$ as filler metals are compared by mechanism, microstructure and hardness. The thickness of brazed joint with amorphous alloys became much smaller than that of PVD-Be. The erosion of base metal did not occur in the brazed joint with amorphous alloys. The brazing mechanism for PVD-Be seems to be Be diffusion into Zr-4 with capillary action resulting from eutectic reaction while that for amorphous alloys are associated with the liquid phase formation in the brazed joint. The brazed joint microstructure with PVD-Be consists of dendrite while that with amorphous alloys is globular. The $Zr_{0.7}Be_{0.3}$ alloy shows the smooth interface with little erosion in the base metal and is recommended a most suitable brazing filler metal for Zircaloy-4.

• Korean Journal of Materials Research
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• v.26 no.12
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• pp.681-687
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• 2016

Zirconium(Zr) nuclear fuel cladding tubes are made using a three-time pilgering and annealing process. In order to remove the oxidized layer and impurities on the surface of the tube, a pickling process is required. Zr is dissolved in HF and $HNO_3$ mixed acid during the process and pickling waste acid, including dissolved Zr, is totally discarded after being neutralized. In this study, the waste acid was recycled by adding $BaF_2$, which reacted with the Zr ion involved in the waste acid; $Ba_2ZrF_8$ was subsequently precipitated due to its low solubility in water. It is very difficult to extract zirconium from the as-recovered $Ba_2ZrF_8$ because its melting temperature is $1031^{\circ}C$. Hence, we tried to recover Zr using an electrowinning process with a low temperature molten salt compound that was fabricated by adding $ZrF_4$ to $Ba_2ZrF_8$ to decrease the melting point. Change of the Zr redox potential was observed using cyclic voltammetry; the voltage change of the cell was observed by polarization and chronopotentiometry. The structure of the electrodeposited Zr was analyzed and the electrodeposition characteristics were also evaluated.

• Journal of Mechanical Science and Technology
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• v.15 no.3
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• pp.291-299
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• 2001

Recently, severe wear on the shutdown rod cladding of Ulchin Nuclear Power Plant #1, #2 were observed by the Eddy Current Test(E.C.T.). In particular, the wear at the sixth card location was up to 75%. The test results indicated that the Flow Induced Vibration(F.I.V.) might be the cause of the fretting wear resulting from the contact between Rod Cluster Control Assemblies(RCCAs) and their spacing cards(guide plates) arranged in the guide tube. From reviewing RCCAs fretting wear repots and analyzing the general characteristics of F.I.V. mechanism in the reactor, geometric layout and flow conditions around the control rod, it is concluded that the turbulence excitation is the most probable vibration mechanism of RCCA. To identify the governing mechanism of RCCA vibration, an experiment was performed for a representative rod position in which the most serious fretting wear experienced among the six rod positions. The experimental rig was designed and set up to satisfy the governing nondimensional numbers which are Reynolds number and mass damping parameter. The vibration amplitude measurement by the non-contact laser displacement sensor showed good agreements in the frequency and the maximum wearing(vibration) location with Ulchin E.C.T. results and Framatome report, respectively. The sudden increase in the vibration amplitude was sensed around the 6th guide plate with mass flow rate variation. Comparing the similitude rod behaviour with the idealized response of a cylinder in flow induced vibration, it was found that he dominant mechanism of vibration was transferred from turbulence excitation to periodic shedding at the mass flow ate 90ι/min. Also the critical velocity of the vibration in RCCAs was determined and the vibration can be prevented by reducing the bypass flow rate below the critical velocity.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.416-424
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• 2018

Microstructure of oxide formed on Zr-Nb-Sn tube sample was intensively examined by scanning transmission electron microscopy after exposure to simulated primary water chemistry conditions of various concentrations of Zn (0 or 30 ppb) and dissolved hydrogen ($H_2$) (30 or 50 cc/kg) for various durations without applying desirable heat flux. Microstructural analysis indicated that there was no noticeable change in the microstructure of the oxide corresponding to water chemistry changes within the test duration of 100 days (pretransition stage) and no significant difference in the overall thickness of the oxide layer. Equiaxed grains with nano-size pores along the grain boundaries and microcracks were dominant near the water/oxide interface, regardless of water chemistry conditions. As the metal/oxide interface was approached, the number of pores tended to decrease. However, there was no significant effect of $H_2$ concentration between 30 cc/kg and 50 cc/kg on the corrosion of the oxide after free immersion in water at $360^{\circ}C$. The adsorption of Zn on the cladding surface was observed by X-ray photoelectron spectroscopy and detected as ZnO on the outer oxide surface. From the perspective of $OH^-$ ion diffusion and porosity formation, the absence of noticeable effects was discussed further.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3095-3105
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• 2022

Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.