• Proceedings of the Korea Concrete Institute Conference
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• 2000.10a
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• pp.320-324
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• 2000

Nonlinear stress analysis of nuclear containment building is carried out using microscopic concrete material model. The present study mainly focuses on the evaluation of the ultimate pressure capacity of idealized containment building in nuclear power plant. For this purpose, an eight-node degenerated shell element it adopted and an imaginary opening in the apex of containment building is allowed in FE model. From numerical analysis, the adopted concrete material model performs well and has a good agreement with the result obtained by using ABAQUS. Finally, we propose the present study as a benchmark test for nonlinear analysis of containment building.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.231-235
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• 2016

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.37-41
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• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.23-30
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• 2005

From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutorn radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.244-251
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• 2015

The containment filtered venting system reduces the range of the contamination area around the nuclear power plant by strengthening the integrity of the containment building. In this study, the probabilistic assessment code MACCS2 was used to assess the effect of the CFVS to off-site. The accident source term was selected from a Probabilistic Safety Analysis report of SHINKORI 1&2 Nuclear Power Plant. The three source term categories from 19 STC were chosen to evaluate the effective dose and thyroid dose of residents around the power plant and the dose with CFVS and without CFVS were compared. The dose was calculated according to the distance from the nuclear power plant, so the damage scale based on the distance that exceeds the IAEA criteria for effective dose (100 mSv per 7 days) and thyroid dose (50 mSv per 7 days) were compared. The effective dose reduction rates of the STC-3, STC-4, STC-6 were about 95-99% in the whole range (0~35 km), 96-98% for the thyroid dose. There are similar results between effective dose and thyroid dose. After applying the CFVS, the damage scale that exceeds the effective dose criteria was about 1 km (mean). Especially, the STC-4 damage scale was decreased from 26 km (mean) to 1.2 km (mean) significantly. The damage scale that exceed the thyroid dose criteria was decreased to 2~3 km (mean). The STC-4 damage scale was also decreased significantly as compared to STC-3, STC-6 in terms of effective dose.

• 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.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.05a
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• pp.412-415
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• 2004

This study was performed to verify the behaviors of fiber Bragg grating (FBG) sensors attached to the containment structure in the nuclear power plant as a part of structural integrity test which demonstrates that the structural response of the non-prototype primary containment structure is within predicted limits plus tolerances when pressurized to $115\%$ of containment design pressure, and that the containment does not sustain any structural damage.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2003.03a
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• pp.118-125
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• 2003

Seismic damage assessment program for containment structure is developed. The program has been established through the combination of inelastic seismic analysis program and 3-D animation program. Damage indices at finite element level and structural level have been introduced for the seismic damage assessment. The seismic damage assessment program makes it possible to analyze in real-time the actual resistance capacity and damage level of containment structure. It will be expected that the program enables to establish the measures more quickly under the earthquake event.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.151-158
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• 2005

For the safety analysis of large structures such as nuclear containment buildings, we conventionally prefer to use analytical approach using finite element method rather than empirical test. Therefor, this paper is mainly focused to develop low-order solid finite element model with the elasto-plastic material model for the safety analysis of nuclear containment building. Drucker-Prager failure criteria in uncracked concrete and maximum tensile stress criteria in cracked concrete are used to model the constitutive behavior of concrete. The concrete material model takes into account the aspects of tensile strain, compression strength reduction of concrete and shear transfer to improve the accuracy of the finite element analysis. Finally, numerical simulation to compare the performance of the developed model with experimental results is employed. The numerical results in this study agree very well with the experimental data.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.05a
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• pp.758-761
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• 2004

Prestressing tendons of the nuclear containment building dome are arranged in a non-axisymmetric manner. However, simple axisymmetric modeling of the containment building is often employed to estimate the structural behavior for, e.g. the ultimate pressure, which requires the axisymmetric approximation of the actual tendon arrangements of the dome. A procedure is proposed that can devise the actual 3-dimensional tendon stiffness and prestressing effect into the axisymmetric model. A numerical example of the CANDU type is presented to verify the procedure and to estimate the amount of approximation.