• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.130-136
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• 1998

In the PANBOX 2 system an excore detector module simulating the excore signal responses during a short term transient is implemented in order to simulate the reaction of the flux detector and control system upon rapid power changes as it occurs e. g, in rod drop events. This module has been verified in the past by comparison calculations with the PANBOX 1 system. This report describes additional PANBOX 2 validation calculations which have bee compared with experiment data measured at german plant KKG, cycle 1, for a rod drop event. In general, the PANBOX 2 results are in very good agreement with the KKG experiments. Therefore it is concluded that the excore detector model of PANBOX 2 is successfully validated.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.401-413
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• 2022

This paper provides a simple method by which to estimate the cross-section stress profiles for nozzles designed according to ASME Code Section III. Further, this method validates the effectiveness of earlier work performed by the authors on standard nozzles. The method requires only the geometric information of the pressure vessel and the attached nozzle. A PWR direct vessel injection nozzle, a PWR outlet nozzle, a PWR inlet nozzle and a BWR recirculation outlet nozzle are selected based on their corresponding specific designs, e.g., a varying nozzle radius, a varying nozzle thickness and an outlet nozzle boss. A cross-section stress profile comparison shows that the estimates are in good agreement with the finite element analysis results. Differences in stress intensity factors calculated in accordance with ASME BPVC Section XI Appendix G are discussed. In addition, a change in the dimensions of an alternate nozzle design relative to the standard values is discussed, focusing on the stress concentration factors of the nozzle inside corner.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3320-3325
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• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• Asian Pacific Journal of Cancer Prevention
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• v.14 no.12
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• pp.7731-7736
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• 2013

Aim: To compare the agreement of screening breast mammography plus ultrasound and reviewed mammography alone in asymptomatic women. Materials and Methods: All breast imaging data were obtained for women who presented for routine medical checkup at National Cancer Institute (NCI), Thailand from January 2010 to June 2013. A radiologist performed masked interpretations of selected mammographic images retrieved from the computer imaging database. Previous mammography, ultrasound reports and clinical data were blinded before film re-interpretation. Kappa values were calculated to assess the agreement between BIRADS assessment category and BIRADS classification of density obtained from the mammography with ultrasound in imaging database and reviewed mammography alone. Results: Regarding BIRADS assessment category, concordance between the two interpretations were good. Observed agreement was 96.1%. There was moderate agreement in which the Kappa value was 0.58% (95%CI; 0.45, 0.87). The agreement of BI-RADS classification of density was substantial, with a Kappa value of 0.60 (95%CI; 0.54, 0.66). Different results were obtained when a subgroup of patients aged ${\geq}60$ years were analyzed. In women in this group, observed agreement was 97.6%. There was also substantial agreement in which the Kappa value was 0.74% (95%CI; 0.49, 0.98). Conclusions: The present study revealed that concordance between mammography plus ultrasound and reviewed mammography alone in asymptomatic women is good. However, there is just moderate agreement which can be enhanced if age-targeted breast imaging is performed. Substantial agreement can be achieved in women aged ${\geq}60$. Adjunctive breast ultrasound is less important in women in this group.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2785-2796
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• 2023

How to generate the precise broad group cross section is important for the fast reactor design. In this study, a fast reactor multi-group cross-section generation code MGGC2.0 are developed in-house for processing ultrafine group MATXS format library. Validation and verification are performed for MGGC2.0 code by applying the benchmarks of ICSBEP handbook, and the results of MGGC2.0 agree well with that of MCNP. The consistent P_N method with critical buckling search is in good agreement that condensed with TWODANT flux and flux moment for the inner core and outer core region. For the radial blanket and reflector, two region approximation method has been applied in MGGC2.0 by using collision Probability Method neutron flux solver. The RBEC-M benchmark was used to verify the power distribution calculation, and the relative error of power distribution comparison with the reference are less than 0.8% in the fuel region and the maximum relative error is 5.58% in the reflector region. Therefore, the precise broad cross section can be generated by MGGC2.0 for fast reactor.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2591-2603
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• 2023

Simulating the Corrosion-Related Unidentified Deposit (CRUD) on the surface of fuel assemblies is necessary to predict the axial offset anomaly and the localized corrosion induced by the CRUD during the operation of nuclear power plants. A new CRUD model was developed to predict the formation of the CRUD deposits, considering the deposition and erosion mechanisms. The heat transfer and capillary flow within the CRUD were also considered to evaluate the boiling amount within the CRUD layer. This model predicted a CRUD deposit thickness of 44 ㎛ during a one-cycle operation of the Seabrook nuclear power plant. The CRUD deposition tended to accelerate and decelerate during the simulation, by being related to boiling mechanism on the deposits surface. Additionally, during a three-cycle operation corresponding to the refueling period, the CRUD deposition was saturated at a thickness of 80 ㎛, which was in good agreement with the suggested thickness for CRUD buildupin pressurized water reactors. Surface boiling on the thin CRUD deposits enhanced the acceleration of the deposition, even when the wick boiling properties were not favorable for CRUD deposition. To ensure the certainty of the simulation results, sensitivity analyses were conducted for the porosity, chimney density, and the constants employed in the proposed model of the CRUD.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1002-1012
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• 2024

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.230-235
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• 1996

The new concept of dry area formation based on Poisson distribution of active nucleation sites and the concept of the critical active site density is presented. A simple statistical model is developed to predict the change of slope of the boiling curve up to critical heat flux (CHF) quantitatively. The predictions by the present model are in good agreement with the experimental data. Also it turns out that the present model well explains the mechanism on how the surface wettability influences CHF.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.407-412
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• 1997

The predictive results between a dynamic food-chain model (DYNACON) and an equilibrium model (NRC model) were compared to show the physical validity of DYNACON. Although the mathematical formulations and transport processes of radionuclides in the environment are different between two models, the comparative study shows good agreement for deposition events that occur during the growing season of plants.

• Mass Spectrometry Letters
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• v.6 no.3
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• pp.75-79
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• 2015

Suitable analytical procedures for the bulk analysis of ultra-trace amounts of uranium and plutonium have been developed using multi-collector inductively coupled mass spectrometry (MC-ICP-MS). The quantification and determination of the isotopic ratios of uranium and plutonium in three simulated swipe samples, a swipe blank, and a process blank were performed to validate the analytical performance. The analytical results for the simulated swipe samples were in good agreement with the certified values, based on the measurement quality goals for the analysis of bulk environmental samples recommended by the International Atomic Energy Agency (IAEA)