• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.866-872
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• 2024

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.119-132
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• 2001

This paper provides the basic technical and safety criteria to guide establishing the reference HLW geological repository system that has been developing based on the recommendations from the international organizations such as IAEA and ICRP as well as the comparison of the regulations of several leading countries in HLW disposal. The proposed criteria and guidelines were categorized by the basic principles and general criteria for the radiological safety and the functional criteria of the repository system components. They would be useful for the development of the national regulations and criteria for HLW disposal in the future. They, of course, will be revised based on the deep geological investigation in Korean Peninsular which will be implemented in the future.

• Journal of Radiation Protection and Research
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• v.23 no.3
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• pp.185-196
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• 1998

현재의 방사선방호 원칙과 체계는 국제방사선방호위원회 (ICRP)의 권고에 기반을 두고 있 다. ICRP 의 권고는 대부분 히로시마/나카사키 원폭피해 생존자들에 대한 역학조사 및 수명 연구결과 그리고 관련 방사선생물학 연구결과를 바탕으로 전세계 5개 인구집단(일본, 미국, 푸에리토리코, 영국과 중국)에 대한 방사선위험계수의 예측 및 평가결과에 근거를 두고 있다. 이 저선량 방사선의 (확률적 영향) 위험계수는 인체피폭 방사선량과 그 영향간에는 선형 비례관계가 있으며 영향유발의 문턱값이 존재하지 않는다는 가정인 '선형 무문턱값 선량-영향 모델 (Linear No-Threshold Dose-Effect Model, 이른바 LNT 모텔)' 을 도입하여 유도된 것이다(譯者 밑줄). 그러나 이 LNT 가정의 과학적 근거와 정당성에 대한 비난이 원자력산업계나 일부 과학자들에 의해 제기된 이래, 최근에는 미국 보건물리학회 (HPS)에서 'LNT 가정이 선량과 영향의 관계를 단순화하며 낮은 선량의 위험음 과대평가한다'는 성명서를 발표하기도 했다. 이후 이에 대한 논쟁이 다시 시작되어, 1997 년에 스페인의 Sevill에서는 IAEA와 WHO의 공동주최와 UNSCEAR의 협조로 '저준위 방사선 영향에 대한 국제회의'가 개최되기도 하였으나 아직 어느 쪽에도 유리한 결론이 단정적으로 나지 않았으며, 실질적인 대안이 없는 현실에서 이 LNT 가정은 여전히 방사선방호의 철학적 기초로 남아 있다(譯者 밑줄). 한편, 저선량 방사선의 영향에 대해서는 우리나라에서도 '방사선방어학회, ‘98 년 춘계 심포지움' 및 '원자력학회, '98 년 춘계 학술발표회 워크??????'에서 한양대학교의 이재기 교수에 의하여 소개, 논의된 바 있다.이 논문은 이러한 논의의 후속으로 역자중 일인이 위원으로 있는 OECD/NEA 방사선방호위원회 (CRPPH)가 최근에 ('98.7.) 발간한 보고서를 번역, 주해한 것으로, 과학지식의 진보에 따라 방사선방호분야에서 관심이 되는 주제들에 대한 위원회의 검토의견을 소개하고 있다. 따라서 이 논문이 국내의 방사선방호분야 관계자들에게 최신정보 습득과 지식함양에 좋은 도움이 되기를 기대한다.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.552-561
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• 2010

This study investigates the solubility of neptunium (Np) in the deep natural groundwater of the Korea Atomic Energy Research Institute Underground Research Tunnel (KURT). According to a Pourbaix diagram (pH-$E_h$ diagram) that was calculated using the geochemical modeling program PHREEQC 2.0, the redox potential and the carbonate ion concentration both control the solubility of neptunium. The carbonate effect becomes pronounced when the total carbonate concentration is higher than $1.5\;{\times}\;10^{-2}$ M at $E_h$ = -200 mV and the pH value is 10. Given the assumption that the solubility-limiting stable solid phase is $Np(OH)_4(am)$ under the reducing condition relevant to KURT, the soluble neptunium concentrations were in the range of $1\;{\times}\;10^{-9}$ M to $3\;{\times}\;10^{-9}$ M under natural groundwater conditions. However, the solubility of neptunium, which was calculated with the formation constants of neptunium complexes selected in an OECD-NEA TDB review, strongly deviates from the value measured in natural groundwater. Thus, it is highly recommended that a prediction of neptunium solubility is based on the formation constants of ternary Np(IV) hydroxo-carbonato complexes, even though the presence of those complexes is deficient in terms of the characterization of neptunium species. Based on a comparison of the measurements and calculations of geochemical modeling, the formation constants for the "upper limit" of the Np(IV) hydroxo-carbonato complexes, namely $Np(OH)_y(CO_3)_z^{4-y-2z}$, were appraised as follows: log $K^{\circ}_{122}\;=\;-3.0{\pm}0.5$ for $Np(OH)_2(CO_3)_2^{2-}$, log $K^{\circ}_{131}\;=\;-5.0{\pm}0.5$ for $Np(OH)_3(CO_3)^-$, and log $K^{\circ}_{141}\;=\;-6.0{\pm}0.5$ for $Np(OH)_4(CO_3)^{2-}$.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.12
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• pp.654-662
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• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2311-2320
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• 2022

The main outcomes of the experiments H2P6 performed in the thermal-hydraulics large-scale PANDA facility at PSI in the frame of the OECD/NEA HYMERES-2 project are presented in this article. The experiments of the H2P6 series consists of two PANDA tests characterized by the activation of three (H2P6_1) or one (H2P6_2) cooler(s) in an initially stratified and pressurized containment atmosphere. The initial stratification is defined by a helium-rich region located in the upper part of the vessel and a steam/air atmosphere in the lower part. The activation of the cooler(s) results i) in the condensation of the steam in the vicinity of the cooler(s), ii) the corresponding activation of large scale natural circulation currents in the vessel atmosphere, with the result of iii) the re-distribution and mixing of the Helium stratification initially located in the upper half of the vessel and iv) the continuous pressure decay. The initial helium layer represents hydrogen generated in a postulated severe accident. The main question to be answered by the experiments is whether or not the interaction of the different, localized cooler units would be important for the application of numerical methods. The paper describes the initial and boundary conditions and the experimental results of the H2P6 series with the suggestion of simple scaling laws for both experiments in terms of i) the temperature difference(s) across the cooler(s), ii) the transient steam and helium content and iii) the pressure decay in the vessel. The outcomes of this scaling indicate that the interaction between separate, closely localized units does not play a prominent role for the present experiments. It is therefore reasonable to model several units as one large component with equivalent heat transfer area and total water flow rate.

• Journal of Radiation Protection and Research
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• v.33 no.4
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• pp.127-134
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• 2008

Potential economic impact of a hypothetical severe accident at a nuclear power plant(Uljin units 3/4) was estimated by applying the Delphi method, which is based on the expert judgements and opinions, in the process of quantifying uncertain factors. For the purpose of this study, it is assumed that the radioactive plume directs the inland direction. Since the economic risk can be divided into direct costs and indirect effects and more uncertainties are involved in the latter, the direct costs were estimated first and the indirect effects were then estimated by applying a weighting factor to the direct cost. The Delphi method however subjects to risk of distortion or discrimination of variables because of the human behavior pattern. A mathematical approach based on the Bayesian inferences was employed for data processing to improve the Delphi results. For this task, a model for data processing was developed. One-dimensional Monte Carlo Analysis was applied to get a distribution of values of the weighting factor. The mean and median values of the weighting factor for the indirect effects appeared to be 2.59 and 2.08, respectively. These values are higher than the value suggested by OECD/NEA, 1.25. Some factors such as small territory and public attitude sensitive to radiation could affect the judgement of panel. Then the parameters of the model for estimating the direct costs were classified as U- and V-types, and two-dimensional Monte Carlo analysis was applied to quantify the overall economic risk. The resulting median of the overall economic risk was about 3.9% of the gross domestic products(GDP) of Korea in 2006. When the cost of electricity loss, the highest direct cost, was not taken into account, the overall economic risk was reduced to 2.2% of GDP. This assessment can be used as a reference for justifying the radiological emergency planning and preparedness.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4348-4358
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• 2022

The thermal-hydraulics phenomena in a containment during an accident will necessarily include radiative heat transfer (i) within the gas mixture due to the high radiative absorption and emission of steam and (ii) between the gas mixture and the surrounding structures. The analysis of some previous PANDA experiments (PSI, Switzerland) demonstrated the importance of the proper modelling of radiation for the benefit of numerical simulations. These results together with dedicated scoping calculations conducted for the present experiments indicated that the radiative heat transfer is considerable, even for a very low amount of steam (≈2%). The H2P2 series conducted in the large-scale PANDA facility at the Paul-Scherrer-Institut (PSI) in the framework of the OECD/NEA HYMERES-2 project is intended to enhance the understanding of thermal radiation phenomena and to provide a benchmark for corresponding numerical simulations. Thus, the test matrix was tailored around the two opposite extremes: either gas compositions with small steam content such that radiative heat transfer phenomena can be neglected. Or gas mixtures containing larger amounts of steam, so that radiative heat transfer is expected to play a dominant role. The H2P2 series consists of 5 experiments designed to isolate the radiation phenomena from convective and diffusive effects as much as possible. One vessel with a diameter of 4 m and a height of 8 m was preconditioned with different mixtures of air / steam at room and elevated temperatures. This was followed by the build-up of a stable helium stratification at constant pressure in the upper part of the vessel. After that, helium was injected from the top into the vessel which leads to an increase of the vessel pressure and a corresponding elevation-dependent and transient rise of the gas temperature. It is shown that even the addition of small amounts of steam in the initial gas atmosphere considerably impacts the radiative heat transport throughout all phases of the experiments and markedly influences i) the monitored gas peak temperature, ii) the temperature history during the compression and iii) the following relaxation phase after the compression was stopped. These PANDA experiments are the first of its kind conducted in a large scale thermal-hydraulic facility.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3874-3897
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• 2023

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High-Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next-generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds-averaged Navier-Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.