• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.211-216
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• 1996

Since the middle of the eighties of this century a research program both for heating reactor and investigation of heating reactor thermal-hydraulics has been carried out in Institute of Nuclear Energy Technology(INET) of Tsinghua university in China. This kind of heating reactor is a light water cooled and integrated natural circulation reactor with low system pressure and low quality at the exit of core. Because of relatively long riser and low system pressure. a little change of the quality at the exit of the core will result in a relatively large variation of void fraction in the riser. Two full scale test loops. HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 heating reactors in geometry and operation parameters respectively, and some test results from the HRTL-200 test facility are shown in this paper. The range of studied system pressure is from 1.0MPa to 4.0MPa, the largest heat flux is about 50 W/cm2, and the quality at the exit of test section is less than 5%.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.2
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• pp.167-175
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• 2014

If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Membrane Journal
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• v.1 no.1
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• pp.100-100
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• 1991

A novel system to thicken the concentrated colloidal solution from membrane water treat-ment processes was developed. A hollow-fiber microfilter(hydrophilic polyethylene nominal pore size 0.1 μm total surface area 0.42 m2) was installed in an acrylic housing that has an aeration port 5 cm below the membrane and a clarifier in the bottom. The concentrate was uniformly supplied from the top of the housing. Bacuum filtration caused downward flow of concentrate and as a result thickening interface. The addition of poly-aluminum chloride (PAC) resulted in rapid increase of trans-membrane pressure (TMP) and in no improvement of the filtered water turbidity and thickening process. Two types of con-centrate and concentrate turbidity had little effect on the increase of TMP and concentrate thickening. It was observed that for the same height of membrane housing membrane surface area to housing volume (A/V) ratio had significant effect on the increase of TMP. When the housing volume was increased ten times the increasing rate of TMP was three times faster as compared to the original housing. A hydraulic model successfully simulated the formation and sedimentation of thickening interface.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.663-667
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• 2017

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below $300^{\circ}C$. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor ($K_{IH}$) to initiate DHC as a function of temperature in Zry-4 for temperatures between $227^{\circ}C$ and $315^{\circ}C$. The experimental technique used in this study was the pin-loading testing technique. To determine the $K_{IH}$, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around $300^{\circ}C$, there was a sharp increase in $K_{IH}$ indicating the upper temperature limit for DHC. The value for $K_{IH}$ at $227^{\circ}C$ was determined to be $2.6{\pm}0.3MPa$ ${\surd}$m.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.14-20
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• 2011

The end of design lifetime for Wolsong unit 1 will be reached on 20th November in 2012. So the license renewal documents for the continuous operation of Wolsong unit 1 is under reviewing now. Major components of primary system such as pressure tubes, feeder pipes including delayed neutron monitoring system tubing are being replaced and many components of secondary system are also being repaired. In this paper, the assessment on the wear degradation of delayed neutron monitoring system tubing(on the other hand, DN tube was called) was performed for the ageing management of the same component. The wear defects of this component was one of causes that resulted in heavy water leakage accidents. Therefore design specifications of Wolsong uint 1 and heavy water leakage accidents of pressurized heavy water reactors were reviewed and causes of wear defect for DN tubes were analyzed. Wear propagation equations based on the heavy water leakage history were made and the proper repairing time was possible to be expected if the continued operation was considered. Finally design change items of DN tubes that were conducted for the long term operation of Wolsong unit 1 are introduced.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Proceedings of the Membrane Society of Korea Conference
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• 1998.10a
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• pp.15-18
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• 1998

Membrane reactors are systems which combine a chemical reactor with a membrane separation process allowing to carry out simultaneously conversion and product separation. The catalyst can be immobilized on the membrane or simply compartmentalized in a reaction space by the membrane. Membrane reactors are today investigated to produce optically pure isomers and/or resolve racemic mixture of enantiomers. The interest towards these systems is due to the increasing demand of enantiomerically pure compounds to be used in the pharmaceutical, food, and agrochemical industries. In fact, enantiomers can have different biological activities, which often influence the efficacy or toxicity of the compound. On the basis of current literature there are basically two schemes on the use of membrane technology to produce enantiomers. In one case, the membrane itseft is intrinsically enantioselective: the membrane is the chiral system which selectively separates the wanted isomer on the basis of its conformation. In the other, a kinetic resolution using an enantiospecific biocatalyst is combined with a membrane separation process; the membrane separates the product from the substrate on the basis of their relative chemical properties (i.e. solubility). This kind of configuration is widely used to carry out kinetic resolutions of low water soluble substrams in biphasic membrane reactors [Giomo, 1995, 1997; Lopez, 1997]. These are systems where enzyme-loaded membranes promote reactions between two separate phases thanks to the properties of enzymes, such as lipases, to catalyse reactions at the org ic/aqueous interface; the two phases are maintained in contact and separated at the membrane level by operating at appropriate transmembrane pressure. A schematic representation of biphasic membrane reactor is shown in figure 1, while an example of enantiospecific reaction and product separation carried out with these systems is reported in figure 2.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.19-30
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• 2017

Austenitic Stainless Steels (ASSs) are widely used as structural materials in the pressurized water reactors (PWRs) because of their superior mechanical properties and corrosion resistance. However, it is well known that ASSs are susceptible to the environmental assisted cracking (EAC) such as environmental assisted fatigue (EAF) during the long term operation. There have been extensive tests and researches to understand the extent and the mechanisms of environmental effects. In this paper, the world-wide EAF test results of ASSs are introduced including those of Korean test programs. The suggested EAF mechanisms of ASSs are also discussed. Finally, the areas of further research to resolve the issue of EAF are suggested.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.