• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.719-724
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• 1998

A preliminary fuel channel analysis for 35% reactor inlet header (RIH) break in CANDU reactor loaded with the CANFLEX-RU fuel bundles has been performed. The predicted results are compared with those for the reactor compared with those for the reactor loaded with standard 37-element bundles. The maximum fuel centerline and sheath temperatures for the CANFLEX-RU bundle channel were lower by 338 and 122 $^{\circ}C$, respectively, than those for the standard bundle because of the Bower maximum linear power of the CANFLEX-RU bundle In spite of the 0.4 FPS higher power pulse of the CANFLEX-RU bundle case. Fuel integrity margin to fuel breakup for the CANFLEX-RU bundle is about 50 J/g higher than that for the standard bundle. The PT/CT contact for the CANFLEX-RU bundle occurred 2 s later than that for the standard bundle. The PT/CT contact temperature for the CANFLEX-RU bundle was 2 $^{\circ}C$ lower than that for the standard bundle. These provide the CANFLEX-RU bundle with the negligibly enhanced safety margin for the fuel channel integrity in CANDU 6 reactor, compared with the standard bundle.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.484-488
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• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.

• International conference on construction engineering and project management
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• 2017.10a
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• pp.8-13
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• 2017

Nuclear power plant (NPP) constructions are large scale projects that are executed for several years, and schedule control utilizing various schedules is a critically important factor. Recently Korea independently developed the Advanced Power Reactor (APR) 1400 and is building nuclear facilities applying this new reactor type. The construction of Shin-Kori NPP (SKN) Unit 3, which adopted the APR1400, was completed and commercial operation has begun, while, SKN 4, Shin-Hanul NPP (SHN) Units 1&2, and SKN 5&6 are currently under construction. Prior to the development of the APR1400, Korea built 24 reactors and accumulated the schedule data of various reactor types which provided the foundation for schedule reduction to be possible. However, as there is no schedule development and review system established based on the standard schedule data (standard activities, durations, etc.) by reactor type, the process for developing the schedule for new builds is low in efficiency consuming much time and manpower. Also all construction data has been accumulated based on schedule activities. But because the connectivity of activities between projects is low, it is difficult to utilize such accumulated data (causes for schedule delay, causes for design changes, etc.) in new build projects. Due to such reasons, issues continue to arise in the process of developing standard schedule activities and a standard schedule for nuclear power plant construction. In order to develop a standard schedule for NPP construction, i) the development of an NPP standard schedule activity list, ii) development of the connection logic of NPP standard schedule activities, iii) development of NPP standard schedule activity resources and duration, and iv) integration of schedule data need to be performed. In this paper, an analysis was made on the coherence of schedule activity descriptions of existing NPPs by applying the probabilistic methodology on activities with low connectivity due to the utilization of the numbering system of four APR1400 reactors (SHN 1&2 and SKN 3&4).This study also describes the method for developing a standard schedule activity list and connectivity measures by extracting same and/or similar schedule activities.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.576-582
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• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Nuclear Engineering and Technology
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• v.11 no.4
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• pp.287-293
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• 1979

Mixed standard-FLIP fuel loading patterns in the TRIGA Mark-III reactor were analyzed. It was judged that the mixed loading pattern with the standard fuel in the B-ring and the FLIP fuel in other rings was mostly desirable in view of fuel temperature, cooling condition with the natural convection, or effective thermal flux utilization in the central thimble. In addition, tile maximum useful flux in tile reactor beamports versus the loading patterns was evaluated.

• Journal of the Korea Safety Management & Science
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• v.16 no.2
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• pp.237-246
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• 2014

Researches on the elimination of sulfur and nitrogen oxides with catalysts and absorbents reported many problems related with elimination efficiency and complex devices. In this study, decomposition efficiency of harmful gases was investigated. It was found that the efficiency rate can be increased by moving the harmful gases together with SPCP reactor and the catalysis reactor. Calcium hydroxide($Ca(OH)_2$), CaO, and $TiO_2$ were used as catalysts. Harmful air polluting gases such as $SO_2$ were measured for the analysis of decomposition efficiency, power consumption, and voltage according to changes to the process variables including frequency, concentration, electrode material, thickness of electrode, number of electrode winding, and additives to obtain optimal process conditions and the highest decomposition efficiency. The standard sample was sulfur oxide($SO_2$). Harmful gases were eliminated by moving them through the plasma generated in the SPCP reactor and the $Ca(OH)_2$ catalysis reactor. The elimination rate and products were analyzed with the gas analyzer (Ecom-AC,Germany), FT-IR(Nicolet, Magna-IR560), and GC-(Shimazu). The results of the experiment conducted to decompose and eliminate the harmful gas $SO_2$ with the $Ca(OH)_2$ catalysis reactor and SPCP reactor show 96% decomposition efficiency at the frequency of 10 kHz. The conductivity of the standard gas increased at the frequencies higher than 20 kHz. There was a partial flow of current along the surface. As a result, the decomposition efficiency decreased. The decomposition efficiency of harmful gas $SO_2$ by the $Ca(OH)_2$ catalysis reactor and SPCP reactor was 96.0% under 300 ppm concentration, 10 kHz frequency, and decomposition power of 20 W. It was 4% higher than the application of the SPCP reactor alone. The highest decomposition efficiency, 98.0% was achieved at the concentration of 100 ppm.

• Journal of Energy Engineering
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• v.8 no.2
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• pp.285-292
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• 1999

The design parameter of the reactor vessel insulation for the Korea Standard Power Plant has been studied numerically. The heat loss from the reactor vessel through the insulation is analysed by using the computational fluid dynamics code, FLUENT. Parametric study has been performed on the air gap width between the reactor vessel wall and the inner surface of the insulation, and on the insulation thickness. Also evaluated is the performance degradation due to the chimney effect caused by gaps between the panels during the installation of the insulation system. From the analysis results, the optimal air gap width and the optimal insulation thickness are obtained.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.52-57
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• 1998

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-Zr $H_{1.6}$), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTBM 80+ and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-Zr $H_{l.6}$ fuel in the seed region without additional penalties in comparison with the standard RTR's U-Zr fuelr fuelel

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.559-566
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• 2015

The thermalehydraulic characteristics for the CANadian Deuterium Uranium Flexible (CANFLEX)-burnable poison (BP) fuel channel, which is loaded with a BP at the center ring based on the CANFLEX-RU (recycled uranium) fuel channel, are evaluated and compared with that of standard 37-element and CANFLEX-NU (natural uranium) fuel channels. The distributions of fuel temperature and critical channel power for the CANFLEX-BP fuel channel are calculated using the NUclear Heat Transport CIRcuit Thermohydraulics Analysis Code (NUCIRC) code for various creep rate and burnup. CANFLEX-BP fuel channel has been revealed to have a lower fuel temperature compared with that of a standard 37-element fuel channel, especially for high power channels. The critical channel power of CANFLEX-BP fuel channel has increased by about 10%, relative to that of a standard 37-element fuel channel for 380 channels in a core, and has higher value relative to that of the CANFLEX-NU fuel channel except the channels in the outer core. This study has shown that the use of a BP is feasible to enhance the thermal performance by the axial heat flux distribution, as well as the improvement of the reactor physical safety characteristics, and thus the reactor safety can be improved by the use of BP in a CANDU reactor.