• Nuclear Engineering and Technology
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• 제49권4호
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• pp.855-864
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• 2017

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2173-2182
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• 2020

Control valves are widely used to regulate fluid flux in nuclear power plants, and there are more than 1500 control valves in the primary circuit of one nuclear power plant. With their help, the flux can be regulated to a specific level of water or steam to guarantee the energy efficiency and safety of the nuclear power plant. The flux characteristics of the control valve mainly depend on the valve core shape. In order to analyze the effects of valve core shapes on flux characteristics of control valves, this paper focuses on the valve core shapes. To begin with, numerical models of different valve core shapes are established, and results are compared with the ideal flux characteristics curve for the purpose of validation. Meanwhile, the flow fields corresponding to different valve core shapes are investigated. Moreover, relationships between the valve core opening and the outlet flux under different valve core shapes are carried out. The flux characteristics curve and equation are proposed to predict the outlet flux under different valve core openings. This work can benefit the further research of the flux control and the optimization of the valve core for control valves in nuclear power plants.

• 한국재난정보학회 논문집
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• 제8권4호
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• pp.328-335
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• 2012

건설공사의 표준품셈에 의한 자원기반 적산방식은 정부고시 노임단가와 기자재의 공시가 격에 표준적이고 보편적인 공법 및 공종에 대한 단위작업당 재료량, 노무량, 장비사용시간 등을 수치로 표시한 품을 곱하여 견적하였다. 그러나 자원기반 적산방식은 실제 거래가격을 적절히 반영하지 못하는 한계점이 발생하였다. 따라서 본 연구는 이러한 한계점을 극복하기 위하여 새로운 시도로 수로터널 현장을 대상으로 Cutter 점검/교환, TBM 정비, TBM 점검/급유, 후속설비, 광차, 운영교대, 갱내보강/낙반보강, 안전/회의의 작업 공종 별로 TBM 굴진 생산성을 분석하고, 이를 근거로 순굴진속도를 분석하여 실적 공사비를 분석하였다. 본 연구에서 제시하는 실적공사비 산정 접근법은 TBM 수로터널 현장에서 실적 공사비를 예측하는데 유용한 도구로 활용할 수 있을 것으로 확신한다.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.17-36
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• 2010

In this study, we compare the mass release rates of radionuclides(1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. the total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fission-product radionuclides, $^{129}I$, $^{79}Se$, and $^{36}Cl$, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would by sited. the footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.785-796
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• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.867-874
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• 2009

Solubility of a radionuclide is important for defining the release source term of a radioactive waste in the safety and performance assessments of a radioactive waste repository. When the pH and redox potential of the KURT groundwater were changed by an electrical method, the concentrations of uranium and thorium released from $UO_2$(cr) and $ThO_2$(cr) at alkali pH(8.1 ${\sim}$ 11.4) and reducing potential (Eh < -0.2 V) conditions were less than $10^{-7}mole/L$. Unexpectedly, the concentration of tetravalent thorium is slightly higher than that of uranium at pH = 8.1 and Eh= -0.2 V conditions, and this difference may be due to the formation of hydroxide-carbonate complex ions. When $UO_2$(s) and $UO_2$(am, hyd.), and $ThO_2$(s) and $Th(OH)_4(am)$ were assumed as solubility limiting solid phases, the concentrations of uranium and thorium in the KURT groundwater calculated by the PHREEQC code were comparable to the experimental results. The dominating aqueous species of uranium and thorium were presumed as $UO_2(CO_3)_3^{4-}$ and $Th(OH)_3CO_3^-$ at pH = 8.1 ${\sim}$ 9.8, and $UO_2(OH)_3^-$ and $Th(OH)_4(aq)$ at pH = 11.4.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.323-334
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• 2013

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.246-252
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• 2018

A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet-clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.