• Title/Summary/Keyword: Pressurized water reactor

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Experimental Study of Chemical Effects on Head Loss across Containment Sump Strainer under Post-LOCA Environment (LOCA이후 원자로건물집수조 여과기의 수두손실에 대한 화학적 영향의 실험연구)

  • Ku, Hee-Kwan;Jung, Bum-Young;Hong, Kwang;Jung, Eun-Sun;Jeong, Hyun-Jun;Park, Byung-Gi;Rhee, In-Hyoung;Park, Jong-Woon
    • Journal of the Korea Academia-Industrial cooperation Society
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    • v.10 no.12
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    • pp.3748-3754
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    • 2009
  • An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

A Study on Electrodeionization for Purification of Primary Coolant of a Nuclear Power Plant (원자력 발전소의 일차 냉각수 정화를 위한 전기탈이온법의 기초연구)

  • Yeon, Kyeong-Ho;Moon, Seung-Hyeon;Jeong, Cheorl-Young;Seo, One-Sun;Chong, Sung-Tai
    • Journal of Radiation Protection and Research
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    • v.24 no.2
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    • pp.73-86
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    • 1999
  • The ion-exchange method for the purification of primary coolant has been used broadly in PWR(pressurized water reactor)-type nuclear power plants due to its high decontamination efficiency, simple system, and easy operation. However, its non-selective removal of metal and non-radionuclides shortens its life, resulting in the generation of a large amount of waste ion-exchange resin. In this study, the feasibility of electrodeionization (EDI) was investigated for the purification of primary cooling water using synthetic solutions under various experimental conditions as an alternative method for the ion exchange. The results shows that as the feed flow-rate increased, the removal efficiency increased and the power consumption decreased. The removal rate was observed as a 1000 decontamination factor(DF) at a nearly constant level. For the synthetic solution of 3 ppm TDS (Total Dissolved Solid), the power consumption was 40.3 mWh/L at 2.0 L/min of feed flow rate. The higher removal rate of metal species and lower power consumption were obtained with greater resin volume per diluting compartment. However, the flow rate of the EDI process decreased with the elapsed time because of the hydrodynamic resistivity of resin itself and resin fouling by suspended solids. Thus, the ion-exchange resin was replaced by an ion-conducting spacer in order to overcome the drawback. The system equipped with the ion-conducting spacer resolved the problem of the decreasing flow rate but showed a lower efficiency in terms of the power consumption, the removal rate of metal species and current efficiency. In the repeated batch operation, it was found that the removal efficiency of metal species was stably maintained at DF 1000.

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Structural Integrity Assessment of High-Strength Anchor Bolt in Nuclear Power Plant based on Fracture Mechanics Concept (원자력발전소 고강도 앵커 볼트의 파괴역학적 건전성평가)

  • Lim, Eun-Mo;Huh, Nam-Su;Shim, Hee-Jin;Oh, Chang-Kyun;Kim, Hyun-Su
    • Transactions of the Korean Society of Mechanical Engineers A
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    • v.37 no.7
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    • pp.875-881
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    • 2013
  • The failure of a bolted joint owing to stress corrosion cracking (SCC) has been considered one of the most important structural integrity issues in a nuclear power plant. In this study, the failure possibility of bolting, which is used to support the steam generator of a pressurized water reactor, owing to SCC and brittle fracture was evaluated in accordance with guidelines proposed by the Electric Power Research Institute, which are called the Reference Flaw Factor method. For this evaluation, first, detailed finite element stress analyses were conducted to obtain the actual nominal stresses of bolting in which either service loads or bolt preloads were considered. Based on these nominal stresses, the structural integrity of bolting was addressed from the viewpoints of SCC and toughness. In addition, the accuracy of the EPRI Reference Flaw Factor for assessing bolting failure was investigated using finite element fracture mechanics analyses.

Detection of Hydride Blisters in Zirconium Pressure Tubes using Ultrasonic Mode Conversion and Velocity Ratio Method (초음파 모드 변환 및 속도비 방법에 의한 지르코늄 압력관의 수소화물 블리스터 탐지)

  • Cheong, Yong-Moo;Lee, Dong-Hoon;Kim, Young-Suk
    • Journal of the Korean Society for Nondestructive Testing
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    • v.23 no.4
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    • pp.334-341
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    • 2003
  • When the pressure tubes(f are in contact with the calandria tube(CT) in the pressurized heavy water reactor(PHWR), the temperature difference between inner and outer wall of W results in a thermal diffusion of hydrogen (deuterium) and hydride blisters are formed on the outer surface of PT. Because the hydride blisters and zirconium matrix are acoustically continuous, it is not easy to distinguish the blisters from the matrix with conventional ultrasonic method. An ultrasonic velocity ratio method was developed to detect small hydride blisters on the zirconium pressure tube. Hydride blisters were grown in the PT specimen using a steady state thermal diffusion device. The flight times of longitudinal echo and reflected shear echo from the outer surface were measured accurately. The velocity ratio of the longitudinal wave to the shear wave was calculated and displayed using contour plot. Compared to the conventional flight time method of longitudinal wave, the velocity ratio method shows superior sensitivity to detect smaller blisters as well as better images for the blister shapes. Detectable limit of the outer shape of the hydride blisters was conservatively estimated as $500{\mu}m$, with the same specifications of ultrasonic transducer used in the actual PHWR pressure tube inspection.

The measurement of oxygen and metal ratio of simulated spent fuels by wet and dry chemical analysis (습식 및 건식법에 의한 모의 사용후핵연료의 O/M비 측정)

  • Choi, Ke Chon;Lee, Chang Heon;Kim, Won Ho
    • Analytical Science and Technology
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    • v.16 no.2
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    • pp.117-124
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    • 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.

Electrochemical Reduction Process for Pyroprocessing (파이로프로세싱을 위한 전해환원 공정기술 개발)

  • Choi, Eun-Young;Hong, Sun-Seok;Park, Wooshin;Im, Hun Suk;Oh, Seung-Chul;Won, Chan Yeon;Cha, Ju-Sun;Hur, Jin-Mok
    • Korean Chemical Engineering Research
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    • v.52 no.3
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    • pp.279-288
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    • 2014
  • Nuclear energy is expected to meet the growing energy demand while avoiding CO2 emission. However, the problem of accumulating spent fuel from current nuclear power plants which is mainly composed of uranium oxides should be addressed. One of the most practical solutions is to reduce the spent oxide fuel and recycle it. Next-generation fuel cycles demand innovative features such as a reduction of the environmental load, improved safety, efficient recycling of resources, and feasible economics. Pyroprocessing based on molten salt electrolysis is one of the key technologies for reducing the amount of spent nuclear fuel and destroying toxic waste products, such as the long-life fission products. The oxide reduction process based on the electrochemical reduction in a LiCl-$Li_2O$ electrolyte has been developed for the volume reduction of PWR (Pressurized Water Reactor) spent fuels and for providing metal feeds for the electrorefining process. To speed up the electrochemical reduction process, the influences of the feed form for the cathode and the type of anode shroud on the reduction rate were investigated.

Adsorption Characteristics of Co(II), Ni(II), Cr(III) and Fe(III) Ions onto Cation Exchange Resin - Application to the Demineralizing Process in a Primary Coolant System of PWR (양이온교환수지에 대한 Co(II), Ni(II), Cr(III), Fe(III) 이온의 흡착 특성 - 원자로 일차 냉각재 계통내 탈염 공정에의 적용)

  • Kang, So-Young;Lee, Byung-Tae;Lee, Jong-Un;Moon, Seung-Hyeon;Kim, Kyoung-Woong
    • Journal of Radiation Protection and Research
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    • v.27 no.1
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    • pp.27-35
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    • 2002
  • Characteristics of Amberlite IRN 77, a cation exchange resin, and the mechanisms of its adsorption equilibria with Co(II), Ni(II), Cr(III) and Fe(III) ions were investigated for the application of the demineralizing process in the primary coolant system of a pressurized water reactor (PWR). The optimum dosage of the resin for removal of the dissolved metal ions at $200mgL^{-1}$ was 0.6 g for 100 mL solution. Most of each metal ion was adsorbed onto the resin in an hour from the start of the reaction. Each metal adsorption onto the resin could be well represented by Langmuir isotherms. However, in the case of Fe(III) adsorption, continuous formation of Fe-oxide or -hydroxide and its subsequent precipitation inhibited the completion of the equilibrium between the metal and the adsorbent Cobalt(II) and Ni(II), which have an equivalent electrovalence, were adsorbed to the resin with a similar adsorption amount when they coexisted in the solution. However, Cr(III) added to the solution competitively replaced Co(II) and Ni(II) which were already adsorbed onto the resin, resulting in desorption of these metals into the solution. The result was likely due to a higher adsorption affinity of Cr(III) than Co(II) and Ni(II). This implies that the interactively competitive adsorption of multi-cations onto the resin should be fully considered for an efficient operation of the demineralizing ion exchange process in the primary coolant system.

Changes in Mechanical Properties and Magnetic Parameters of Neutron Irradiated Mn-Mo-Ni Low Alloy Steels (중성자에 조사된 Mn-Mo-Ni 저합금강의 기계적 및 자기적 성질 변화)

  • Jang, Gi-Ok;Ji, Se-Hwan;Park, Seung-Sik;Kim, Byeong-Cheol;Kim, Jong-O
    • Korean Journal of Materials Research
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    • v.8 no.11
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    • pp.1020-1025
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    • 1998
  • Irradiation-induced changes in mechanical properties and magnetic parameters were measured and compared to explore possible correlations for Mn-Mo-Ni low alloy steel surveillance specimens which were irradiated to a neutron fluence of $2.3\times10^{19}n/cm^2$(E>1.0MeV) in a typical pressurized water reactor environment at about $288^{\circ}C$. For mechanical property parameters, microvickers hardness, tensile and Charpy impact test were performed and Barkhausen noise amplitude, coercivity, remanence, maximum induction were measured for magnetic parameters. respectively. Results of mechanical property measurements showed an increase in yield and tensile strength, microvickers hardness. 41J indexed $RT_{NDT}$ and a decrease in upper shelf energy irrespective of base and weld metals. However, in the case of tensile properties, the changes in weld metal were negligible compared to the base metal. In the case of magnetic measurements, it is found that magnetic remanence, BN amplitude. BN energy have dropped significantly but coercivity(H,) has increased rapidly after irradiation. In this study. the measurements conducted on surveillance specimens of Mn-Mo-Ni low alloy steel showed that there were strong correlations between mechanical properties and magnetic properties.

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A Study on Radiation Safety Evaluation for Spent Fuel Transportation Cask (사용후핵연료 운반용기 방사선적 안전성평가에 관한 연구)

  • Choi, Young-Hwan;Ko, Jae-Hun;Lee, Dong-Gyu;Jung, In-Su
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.17 no.4
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    • pp.375-387
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    • 2019
  • In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h-1 and 0.065 mSv·h-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.