• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.388.1-388.1
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• 2004

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.2
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• pp.103-111
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• 2014

In this study, the probabilistic internal pressure fragility analysis was performed by using the non-linear finite element analysis method. The target structure is one of the containment buildings of typical domestic pressurized water reactors(PWRs). The 3-dimensional finite element model of the containment building was developed with considering the large equipment hatches. To consider uncertainties in the material properties and structural capacities, we performed the sensitivity analysis of the ultimate pressure capacity with respect to the variation of four important uncertain parameters. The results of the sensitivity analysis were used to the selection of the probabilistic variables and the determination of their probabilistic parameters. To reflect the present condition of the tendon pre-stressing force, the data of the pre-stressing force acquired from the in-service inspections of tendon forces were used for the determination of the median value. Two failure modes(leak, rupture) were considered and their limit states were defined to assess the internal pressure fragility of target containment building. The internal pressure fragilities for each failure mode were evaluated in terms of median internal pressure capacity, high confidence low probability of failure(HCLPF) capacity, and fragility curves with respect to the confidence levels. The HCLPF capacity was 115.9 psig for leak failure mode, and 125.0 psig for rupture failure mode.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.495-500
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• 1997

For hydrogen management in severe accidents with degraded nuclear core of PWR's, several experiments have been performed in the SNU hydrogen mixing facility. The objectives are understanding the extent of hydrogen mixing and analyzing the effects of factors which dominate uniform or non-uniform mixing at compartments in the containment building. The facility represents on a 1/11th linearly scaled model of the YGN unit 3&4, hydrogen was simulated by helium. Because there are the gaps between safety injection tank and compartment layers in the containment, the test facility was constructed in three dimentinal mode for analyzing of mixture behavior through the gaps. From the experimental results we could conclude that overall hydrogen concentration distributed uniformly in the free volume of the test compartment, but fluctuated in the gaps. This paper is focused on experimental result from several experiment.

• Journal of Radiation Protection and Research
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• v.17 no.2
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• pp.61-70
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• 1992

Photon, neutron and beta radiation fields were measured at PWR plants which are the representative types of nuclear power plant operated in Korea. The photon energy spectra were measured at locations in the auxiliary building during operation period and in the containment vessel(C/V) during shutdown period using a portable gamma spectrometer with a HPGe detector. The distribution of average energy was found to range from 440 to 780 keV in the C/V and from 280 keV to 760 keV in the auxiliary building, respectively. The average neutron energy measured at the five locations around the operation deck in the C/V in operation using a BMSS (Bonner Multi-Sphere Spectrometer) ranged from 20 keV to 210 keV. A computer code, BUNKI was used to unfold the spectrum. The beta energy spectra in the C/V and in the auxiliary building in annual outage were determined using 14 smear samples taken from the highly contaminated areas. The analysis showed that the representative corrosion product, $^{60}Co$ made main contribution to the beta energy field.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.5
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• pp.437-450
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• 2014

In this study, we developed a methodology to determine the reference parameter for an aircraft impact induced risk assessment of nuclear power plant (NPP) using finite element impact analysis of containment building. The target structure used to develop the method of reference parameter selection is one of the typical Korean PWR type containment buildings. We composed a three-dimensional finite element model of the containment building. The concrete damaged plasticity model was used for the concrete material model. The steels in the tendon, rebar, and liner were modeled using the piecewise-linear stress-strain curves. To evaluate the correlations between structural response and each candidate parameter, we developed Riera's aircraft impact force-time history function with respect to the variation of the loading parameters, i.e., impact velocity and mass of the remaining fuel. For each force-time history, the type of aircraft is assumed to be a Boeing 767 model. The variation ranges of the impact velocity and remaining fuel percentage are 50 to 200m/s, and 30 to 90%, respectively. Four parameters, i.e., kinetic energy, total impulse, maximum impulse, and maximum force are proposed for candidates of the reference parameter. The wellness of the correlation between the reference parameter and structural responses was formulated using the coefficient of determination ($R^2$). From the results, we found that the maximum force showed the highest $R^2$ value in most responses in the materials. The simplicity and intuitiveness of the maximum force parameter are also remarkable compared to the other candidate parameters. Therefore, it can be concluded that the maximum force is the most proper candidate for the reference parameter to assess the aircraft impact induced risk of NPPs.