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Currently only six countries worldwide are considered to have the capability necessary to export nuclear power plants, each with unique advantages: Korea, the United States, Japan, Russia, China, and France. Despite abundant reports documenting successful plant exportation, it is difficult to obtain official accounts that reveal the main factors that lead to successful export due to the secretive nature of nuclear power plant contracts. This study developed a model comparing the competitiveness between nuclear power plant supplier countries. The model 1) categorizes and classifies various criteria for nuclear power plant export and 2) assigns weighted values on each criterion based on an AHP analysis to 3) compare the competitiveness of the export-capable countries based on the developed framework. The analysis showed that China has the highest competitiveness followed by Russia, Korea, the United States, Japan, and France, in decreasing order. In this context, it is imperative for Korea to construct consumer oriented export packages and target countries with strong leadership, countries that disregard the nuclear fuel cycle, countries that require fast nuclear power plant adoption, or countries that are pro-American.

Bubble nucleation itself is less important safety issue for nuclear reactor, but it can easily lead to critical thermal-hydraulic events such as OFI (Onset of Fluid Instability) or CHF (Critical Heat Flux) when a research reactor operates under atmospheric conditions. Thus, the ONB (Onset of Nucleate Boiling) margin for normal operation in research reactor is recommended. In the IAEA-TECDOC-233 report (IAEA, 1980), the ONB margin for a research reactor is recommended as well. Although the ONB margin in a research reactor is emphasized for such reasons, only a few experiments have been performed for downward flow direction in a narrow, rectangular channel. In addition, several existing ONB prediction correlations are arguably applicable to the flow boiling condition in the narrow rectangular channel because most of them are developed based on Hsu’s model, which was developed in the pool boiling cases. In the study, ONB experiments for various inlet temperature conditions and mass flux conditions were performed with increasing heat flux step by step. Based on experimental data, the effect of inlet temperature and mass flux on the wall superheat and heat flux at ONB was investigated. In addition, existing ONB prediction correlations were evaluated for predicting wall superheat and heat flux at ONB based on the experimental data. A new ONB prediction correlation was then developed for better-evaluation and was compared with other correlations.

Nuclear Thermal Rocket (NTR) propulsion is a viable and meritorious option for human exploration into deep-space because of its high thrust, improved specific impulse, well established technology, bimodal capability, and enhanced mission safety and reliability. The NTR technology has already been investigated and tested by the United States of America and Russia and the former Soviet Union. The representative Nuclear Engine for Rocket Vehicle Applications (NERVA) type reactors traditionally used Highly Enriched Uranium (HEU) fuels, shaped in hexagonal fuel element geometries because of the importance of making a high power reactor with a minimum size. Although the HEU-NTR designs are the best choice in terms of rocket performance and technical maturity, they inevitably provoke nuclear proliferation obstacles not only for all research and development activities by civilians and non-nuclear weapon states but also for potential commercialization. To overcome the security issues due to HEU, the non-proliferative, small-size NTR engine with low thrust levels of 41 kN–53 kN (9.2 klb f ∼ 11.9 klb f ), Korea Advanced NUclear Thermal Engine Rocket utilizing a Low-Enriched Uranium fuel (KANUTER-LEU), is being designed for future generations. Its design goals are to make use of an LEU fuel for its fairly compact core, but to minimize the rocket performance sacrifice relative to the traditional HEU-NTRs. To achieve these goals, a new space propulsion reactor is conceptually designed with the key concepts of a high uranium density fuel with resistance against high heating and H 2 corrosion, a thermal neutron spectrum core, and a compact and integrated fuel element core design with protective cooling capability. In addition, a preliminary design study of neutronics and thermal-hydraulics was performed to explore the design space of the new LEU-NTR reactor concept. The result indicates that the innovative reactor concept has great potential, both to implement the use of an LEU fuel and to create comparable rocket performance, compared to the existing HEU-NTR designs.

A study, on the theoretical method to predict the critical heat flux (CHF) of saturated upward flow boiling water in vertical narrow rectangular channels, has been conducted. For the assessment of this CHF prediction method, 608 experimental data were selected from the previous researches, in which the heated sections were uniformly heated from both wide surfaces under the high pressure condition over 41 bar. For this purpose, representative previous liquid film dryout (LFD) models for circular channels were reviewed by using 6058 points from the KAIST CHF data bank. This shows that it is reasonable to define the initial condition of quality and entrainment fraction at onset of annular flow (OAF) as the transition to annular flow regime and the equilibrium value, respectively, and the prediction error of the LFD model is dependent on the accuracy of the constitutive equations of droplet deposition and entrainment. In the modified Levy model, the CHF data are predicted with standard deviation (SD) of 14.0% and root mean square error (RMSE) of 14.1%. Meanwhile, in the present LFD model, which is based on the constitutive equations developed by Okawa et al., the entire data are calculated with SD of 17.1% and RMSE of 17.3%. Because of its qualitative prediction trend and universal calculation convergence, the present model was finally selected as the best LFD model to predict the CHF for narrow rectangular channels. For the assessment of the present LFD model for narrow rectangular channels, effective 284 data were selected. By using the present LFD model, these data are predicted with RMSE of 22.9% with the dryout criterion of zero-liquid film flow, but RMSE of 18.7% with rivulet formation model. This shows that the prediction error of the present LFD model for narrow rectangular channels is similar with that for circular channels.

Similar to commercial nuclear reactors, research reactors are recommended to be designed with sufficient safety margins for thermalhydraulic issues such as critical heat flux (CHF) in both normal and accident conditions. In the MARS-KS system code, developed based on RELAP5, the 2006 Groeneveld CHF look-up table (Groeneveld et al., 2007) is adopted to predict CHF for plate-type fuel. The applicability of this table to CHF in rectangular channels for high pressure conditions of over 40 bar has been confirmed (Choi et al., 2015); however, its applicability to pressure conditions under 40 bar has yet to be clarified. While Sudo and Kaminaga previously suggested a rectangular CHF model for a wide experimental range, some notifications have indicated the inclusion of flow excursion data in their model. In this paper, a rectangular channel CHF database was constructed for wide experimental ranges and used to assess the 2006 Groeneveld CHF look-up table and Sudo-Kaminaga correlation to confirm their applicability to rectangular channels under low pressure. Then, with several parameters selected by the Random Forest method in the R package (Reynolds number, quality, aspect ratio, diameter ratio, and length over diameter), a new correction factor was suggested by applying a non-linear regression method from a statistical software package to correct the CHF values in the CHF look-up table for rectangular channels. Results confirmed that the CHF LUT with the newly suggested correction factor is capable to predict CHF in rectangular channels.

Deformed cladding in nuclear power plants can partially block the path of coolant flow and thus alter the heat transfer characteristics between the cladding surface and coolant. In this research, critical heat flux(CHF) in a sudden expansion flow path, was experimentally investigated. Flow boiling CHF experiments were carried out under atmospheric pressure with deionized water for vertical upward flow, with four expansion ratio cases investigated with mass flux ranging from 50 to 250 kg/m2 s. CHF was detected at two different locations. In experiments with either a low expansion ratio(e 1.5) and high mass flux(> 150 kg/m2 s), CHF was detected at the sudden expansion point. Quality at the sudden expansion point was almost zero, which implies that nucleated bubbles under the effect of sudden expansion flow induced CHF at the sudden expansion point. The reason for this transition of CHF location was suggested that local bubble stagnation induced by a balance between drag and buoyancy forces on the bubbles resulted in local flooding in the downward flow.

An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The accidents were simulated by MARS code considering several specific processes characterized from LOCA and SBO. For a high-pressure LOCA, water in an IPST could be injected into a steam generator using gravity after depressurization of the secondary circuit caused by opening atmospheric dump valves. For a low-pressure LOCA, the pressure in the RCS naturally decreased due to coolant loss. Additional depressurization was needed during LOCAs for which RCS criteria pressures at the time of estimation were in the range between high-pressure LOCA and low-pressure LOCA. This is referred to as medium-pressure LOCA, according to the range of the RCS pressure. The procedures of the PDHR strategy can be implemented after failure of the steps in emergency operating procedures, before entering the condition for the severe accident management. Even when external conditions make a given site extremely inaccessible, the PDHR strategy can allow for cooling the core in an EDMG by installing an IPSS beside each nuclear power plant. The proposed PDHR strategy can be applied in both current and future nuclear power plants by considering modifications for enhancing their safety.

The previous empirical CHF correlations for upward flow in a vertical narrow rectangular channel which is uniformly heated from both wide sides were reviewed and analyzed by using the experimental data points at pressure condition over 40 bar. The new correlations, that is, Simple Correlation and New Correlation-A/B were derived and proposed by using simple regression and ACE algorithm, and it was shown that they have more improved prediction errors than the other previous correlations. The Look-Up Table (LUT) of AECL also estimates CHF as well as New Correlation—A/B even though LUT was generated from the CHF data points in circular channels. As a result of comparative assessments of LUT and the empirical correlation of low pressure condition, it is reasonably concluded that for wider pressure condition, LUT with proper correction factors is the most pragmatic and universal CHF prediction method for rectangular channel in this study.

질량전달계수 (m/s) Key Words: Annular Flow(환상유동), Critical Heat Flux(임계열유속), Droyout(드라이아웃), Liquid Film Dryout Model(액막건조모델), Narrow Rectangular Channel(얇은 사각유로), Saturated Boiling(포화비등) 초록: 산업적으로 얇은 사각유로에서의 임계열유속을 포함한 열수력 현상을 이해하고 이를 시스템 설계에 반영해야 될 필요성이 증대되고 있다. 포화비등조건에서 임계열유속이 발생하는 주요기구는 일반적으로 환상유동 영역에서 액막이 건조되는 것이다. 이러한 임계열유속을 예측하기 위하여 원형관에 대한 대표적인 액막건조모델을 고찰하고 환상유동 시작 경계조건을 상수로 가정하는 기존 모델의 한계를 살펴보았다. 균일한 열유속으로 가열되는 얇은 수직사각유로 상향유동에서의 임계열유속을 예측하기 위하여 환상유동을 단순 모델링하고, 새로운 액막건조모델을 적용하였다. 284개 실험데이타에 대한 예측성능을 확인한 결과 MAE 18.1%, RMSE 22.9% 예측오차로 실험데이타를 잘 예측할 수 있음을 확인하였다. Abstract: There is an increasing need to understand the thermal-hydraulic phenomena, including the critical heat flux (CHF), in narrow rectangular channels and consider these in system design. The CHF mechanism under a saturated flow boiling condition involves the depletion of the liquid film of an annular flow. To predict this type of CHF, the previous representative liquid film dryout models (LFD models) were studied, and their shortcomings were reviewed, including the assumption that void fraction or quality is constant at the boundary condition for the onset of annular flow (OAF). A new LFD model was proposed based on the recent constitutive correlations for the droplet deposition rate and entrainment rate. In addition, this LFD model was applied to predict the CHF in vertical narrow rectangular channels that were uniformly heated. The predicted CHF showed good agreement with 284 pieces of experimental data, with a mean absolute error of 18. 1 % and root mean square error of 22.9 %.

Flow instability and critical heat flux (CHF) especially for downward flow of water in a vertical narrow rectangular channel heated from both sides were experimentally investigated. The gap, width, heated width, and heated length of the channel were 2.35, 40, 30, and 350 mm, respectively. The flow boiling was developed as the wall heat flux was increased for the imposed mass flux of 500 and 1000 kg m −2 s −1 for high inlet subcooling (52–74 K) conditions under atmospheric pressure. In this paper, the unstable boiling flow was investigated by monitoring the channel pressure drop and high speed visualization for side view of boiling (HSV-SVB) with the back-lighting method. The results showed that an irremovable pressure drop fluctuation was involved in the flow boiling after the distinct initiating point called as the onset of pressure drop fluctuation (OPDF). We determined that coalescence of bubble (or vapor) layers on opposing heated surfaces triggered the OPDF, since an abrupt increase of pressure drop was occurred as the expansion of the liquid–vapor interface became bounded by the other layer on the opposite side. The fluctuation was amplified as the exit quality was increased and the flow regime passed through the unstable slug or churn flow. For a low inlet throttling condition, premature CHF was induced at the maximum fluctuation during those flow regimes. On the other hand, stable CHF was obtained in the re-stabilized flow regime as the annular flow was developed only for large inlet throttling conditions with an additional pressure drop across the valve of more than 0.4 bar for our experimental configuration.

In this study, an experimental work was conducted for investigating effects of wire nets on flow boiling critical heat flux (CHF). Attaching wire nets on a heating surface is an easier and more economical way to provide surface modification than building porous structures or deposition of nanoparticles. The working fluid was R-134a and flow condition was upward forced convection. Round tubes with inner diameter of 10.4 mm were utilized for test sections. The wire nets was rolled up and then covered the inner surface of the tube. It was attached onto the inner surface of round tube where R-134a flows in. The diameter of wire was 0.5 mm. Three different types of wire nets were used and their side lengths were 1.5 mm, 2.0 mm, and 2.5 mm respectively. The pressure conditions were 1.2 MPa and 1.5 MPa and mass flux conditions were 1000 kg/m 2 s, 2000 kg/m 2 s, and 4000 kg/m 2 s. Inlet subcooling condition was fixed by 40 kJ/kg. The test section was directly heated by electricity and the overall heating length of the test section was 1000 mm. CHF enhancements were observed for all conditions with wire nets. The maximum CHF enhancement ratios were 103% and 114% for 1.2 MPa and 1.5 MPa pressure conditions, respectively. In both cases, the maximum CHF enhancement was observed for a wire nets side length of 2.0 mm. Physical effects of wires nets to bubble detachments were considered as a main factor of the CHF enhancement. Rapid detachment of bubbles, additional bubble nucleation sites, and enhanced capillarity due to wire nets would delay formations of dry patches and enhance liquid re-flooding onto heating surface. Conclusively, higher heat flux would be needed for triggering CHF.

Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

Korean activities on fusion licensing and safety studies have been going through three phases. The first phase started with the licensing of the Korean Superconductor Tokamak Advanced Research (KSTAR) classified as a radiation-generating device on the basis of the existing nuclear regulations. To accommodate the specific characteristics of a tokamak device that is to use deuterium in fusion plasma physics studies, a set of instructions to examine the safety of KSTAR and to approve an operating permit were developed and issued by the regulatory body, Korean Institute of Nuclear Safety. The second phase is being investigated for the International Thermonuclear Experimental Reactor (ITER). To verify the design of the test blanket module that will be installed and tested in the ITER, safety studies were carried out using the experimental results and computational codes. To minimize occupational irradiation damages, a study on the tritium behaviors is being carried out as Korea will provide a part of the tritium systems of ITER. The last phase has initiated to recognize significant safety issues and to support future licensing process for a fusion DEMO plant. Relevant studies are being carried out in collaboration with the academies funded by the National Fusion Research Institute of Korea. These front-end studies include the conceptual analysis on the safety and radiological consequences, preliminary studies on the layout of research, and on the future regulatory requirements of the fusion DEMO plant. This paper summarizes these activities on fusion safety with the findings of the front-end studies to date. It also proposes a pathway for the novel licensing process of a fusion DEMO plant with the tokamak concept.

SiC nanofluid was used for critical heat flux (CHF) enhancement in the case of water pool boiling. Many kinds of nanofluid have been highlighted as a simple way to gain high thermal performance of fluids, and SiC is received attention these days as a promising material because of its relatively high thermal properties. In this study, SiC nanofluid was investigated to measure its thermal performance in water pool boiling experiment especially for CHF. The volume concentration of SiC nanofluid were 0.0001%, 0.001%, 0.01%. Several characteristic of SiC nanofluid, such as zeta potential, and contact angle which could be affect on thermal performance of the fluids had been measured. The experiments were conducted under atmospheric pressure. Two kinds of test sections with different dimensions were used for verifying CHF enhancement trends for SiC nanofluid. The CHF has been enhanced up to 105% for volume concentration 0.01%. CHF enhancement ratios of SiC nanoparticle deposited surfaces were well corresponded with the results of SiC nanofluid. CHF enhancement trend was interesting because it did not linearly dependent on nanoparticle concentration. The wettability change of SiC nanoparticle deposited surface was discussed as main reason of CHF enhancement variation.

Two-phase flow excursion instability especially for a downward flow in a narrow rectangular channel heated on both sides was experimentally investigated. Three test channels with different gap sizes (2.5, 3.3 and 4.1 mm) for the same channel width (40 mm), the heated width (30 mm) and the length (350 mm) were adopted for the experiment. For several imposed heat flux values (ranging 0–1000 kW m −2 ) and inlet subcooling temperature (43–75 K), subcooled flow boiling was investigated while reducing the fluid mass flux starting from a sufficient high-mass-flux condition up to the flow excursion point when vigorous boiling occurs. The minimum mass flux conditions in which a stable flow is sustainable were identified via 47 flow instability data points and were compared with relevant correlations. The results implied flow excursion points that were close to the onset of a significant void. The visualization of the flow excursion using a high-speed camera was achieved, clearly demonstrating that the flow excursion is triggered by the coalescence of facing bubbles (for Pe Pe > 14000) on opposing heated surfaces. Deviation from a prediction of earlier correlations with the present data was revealed, and new empirical correlations that reflect the gap size effect which predicts our data with fairly good accuracy are suggested.

The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively.

Design and operation options that can reduce both the initiating event frequency and the accident mitigation probability were addressed in an integrated framework to cope with station blackout. The safety, engineering cost, water delivery cost and testing/maintenance cost of each option were quantitatively evaluated to calculate the cost variation and to find an optimal point in the reference reactor, OPR1000. Design variables that represent additional emergency water supply, diverse emergency diesel generator, and surveillance test period modification were investigated. Based on these design variables, we applied the developed formula to quantify cost items, which were presented as changes of the economics and the safety. A case study was provided to illustrate the change of the total cost. Different risk aversion factors that represent different attitudes of the public were also investigated. The result shows that the costs and benefits of various complicated options can be effectively addressed with the proposed risk-informed decision making framework.

The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the containment outside into integrated passive safety tanks (IPSTs). The coolant in the IPSTs is used for various functions in accident scenarios. Also, potential problems for the realistic installation of the IPSS are proposed and the solutions to these problems are schematically described. IPSS is the design for the passive safety enhancement in preparation for a loss of AC power. Consequently, it is designed for the supplementation and enhancement of current nuclear power plants, not as a replacement. The specific optimization design for each current or future reactor will be studied as further works.

The concept and assessment approach of nuclear safety in nuclear power plants (NPPs) have been evolved with the technological progress and the lessons learned from the major events. Recently, studies on the integrated approach of deterministic and probabilistic method have been done. In this study, a combined deterministic and probabilistic procedure (CDPP) is proposed for safety assessment of the beyond design basis accidents (BDBAs). In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. To verify applicability of the methodology, performance of the APR-1400 emergency core cooling system is assessed against large break loss of coolant accident (LOCA), under the premise that LOCAs for any breaks larger than transition break size would be regarded as BDBA. In addition, discussions are made for analysis results for allowable NPP changes of emergency diesel generator start time extension and power uprating. It is concluded that the proposed CDPP is applicable to safety assessment of BDBAs in NPPs without significant erosion of the safety margin.

In this study, pool boiling CHF experiments under atmospheric pressure with SA508 test heater was performed. SA508 is the material of the reactor pressure vessel in a nuclear power plant. Therefore, the CHF behavior of SA508 material is important for the reactor pressure vessel integrity at accident, but there are few studies about that. SA508 material showed very different CHF behavior from other cases (stainless steel). It showed very higher CHF value and the test heater surface was changed significantly. This test heater surface change is due to corrosion of SA508, and the rate of corrosion increases with boiling time. The corrosion product is analyzed as magnetite (Fe3O4), and magnetite has been used as one of nanofluid for CHF enhancement. The size of magnetite produced on the test heater surface is 20–140 nm, that is, it is nanoscale. Therefore, the CHF enhancement mechanism of SA508 material is similar to the case using other test heater (like stainless steel) with magnetite nanofluid. Additives like TSP, Al2O3 and CNT nanoparticles are also tested, but they did not show the CHF enhancement effect. In case of TSP, it showed CHF decrease effect due to preventing effect on SA508 corrosion.

Advanced fuel cycle includes SFR and pyro-processing, which are actively researched in various fields of nuclear energy. Focus of previous work (Lee et al., Nuclear Engineering and Technology, 2011, 43, 383–390) was mainly on the comparison of uranium utilization with implicitly expressed reprocessing terms. However, in some cases, the result can be distorted by too much favor on SFR system and thus reliable cost analysis cannot be performed. In this study, the key factors – Recycling cost (SFR capital cost and Pyro-processing cost), Uranium price, and Direct disposal cost – were analyzed by explicitly expressing spent nuclear fuel (SNF) and its related terms, and the qualitative inter-dependency between each costs were discussed. Furthermore, comparison between the direct disposal and recycling was conducted in two aspects, which are spent nuclear fuel accumulation and uranium utilization.

KALIMER-600 (Korea Advanced Liquid Metal Reactor), which is a pool type SFR (Sodium-cooled Fast Reactor), was developed by KAERI (Korea Atomic Energy Research Institute). DBE (Design Basis Events) for KALIMER-600 has been analyzed in the one dimension. In this study, the one dimensional analyzed sodium hot pool is modified to a three dimensional node system, because the one dimensional analysis cannot represent the phenomena of the inside pool of a big size pool with many compositions, such as UIS (Upper Internal Structure), IHX (Intermediate Heat eXchanger), DHX (Decay Heat eXchanger), and pump. The results of the multi-dimensional analysis compared with the one dimensional analysis results in normal operation, TOP (Transient of Over Power), LOF (Loss of Flow), and LOHS (Loss of Heat Sink) conditions. First, the results in normal operation condition show the good agreement between the one and multi-dimensional analysis. However, according to the sodium temperatures of the core inlet, outlet, the fuel central line, cladding and PDRC (Passive Decay heat Removal Circuit), the temperatures of the one dimensional analysis are generally higher than the multi-dimensional analysis in conditions except the normal operation state, and the PDRC operation time in the one dimensional analysis is generally longer than the multi-dimensional analysis. In particular, we can see the reverse and mixing flow phenomena in LOHS better than TOP and LOF, which affect the peak temperature in the one dimensional analysis at 330 s but is not observed in the multi-dimensional analysis. In addition, the different temperatures at the central line of fuel and cladding are shown due to different mass flow in the core. The two PDRCs show the almost same results in one and multi-dimensional analysis, respectively. However, in the comparison between the one and multi-dimension, the heat removal capability of PDRC in the multi-dimensional analysis is better than the one dimensional analysis at the beginning of calculation. The aspects change at the ending of calculation due to the different mass flow in PDRC and over flow sodium temperature. In wrap up, the sodium temperatures of the core in the one dimensional analysis are generally higher than the multidimensional analysis. The temperature peak point in the one dimensional analysis under LOHS does not show in the multi-dimensional analysis, because the reverse and mixing flow pattern between the hot pool and the core outlet region. Even thought there are some different results between the two different dimensional analyses, we can say that the KALIMER-600 system has the safety margins.

This study describes critical heat flux (CHF) experiments using a 2-D curved test section with trisodium phosphate (TSP: Na 3 PO 4 ) and boric acid (BA: H 3 BO 5 ). The CHF values of TSP solution, BA solution, and TSP + BA solution were enhanced by as much as 50% for all experimental conditions except the condition of 150 mm radius with BA solution. The enhancement can be explained by decreased contact angle (enhancement of wettability). This CHF enhancement could provide additional thermal margin for the IVR-ERVC strategy.

Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters.

The effects of the tragedy at Japan’s Fukushima power plant will continue to reverberate over the upcoming weeks, months, and years. And, as the writers in this symposium explain, the consequences of the disaster go beyond Japan—like Chernobyl and Three Mile Island, Fukushima will reshape nuclear agendas and policies in countries around the world. In this Global Forum, leading experts reflect on the current and future implications of Fukushima for their own countries—the United States, the European Union, and South Korea. Mark Cooper (2011) writes from the United States; Caroline Jorant (2011) from the European Union; and Soon Heung Chang from South Korea. In August, this forum will continue as a Roundtable at www.thebulletin.org.

This study describes flow boiling critical heat flux (CHF) experiments using Al 2 O 3 nanofluid and Al 2 O 3 nanoparticle deposited tubes. The flow boiling CHF of Al 2 O 3 nanofluid with a plain tube (NFPT) and de-ionized water with an Al 2 O 3 nanoparticle deposited tube (DWNT) were enhanced up to about 80% for all experimental conditions. There was no big difference in the CHF results between NFPT and DWNT; these results indicate that the CHF enhancement of Al 2 O 3 nanofluid is surely caused by deposition of nanoparticles on the test section tube inner surface. After the flow boiling CHF experiments, the inner surfaces of the test section tube were explored by FE-SEM, which revealed the deposition of Al 2 O 3 nanoparticles on the heated surfaces.

Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.

In this study, the effects of tri-sodium phosphate (TSP) and boric acid on CHF enhancement were studied. Both TSP and boric acid are used to control pH in nuclear power plants. TSP is a kind of surfactant, and several surfactants, include TSP, have been reported to have an effect on enhancement of heat transfer. Nothing has yet been reported for the case of boric acid. CHF experiments were performed with mass flux ranging from 100-500 kg/m2 s and inlet subcooling temperature of 50 °C under atmospheric pressure. The test section was a vertical circular SS316 tube having an inner diameter of 10.98 mm. Its heated length was 224 mm, and it was heated by a heat flux control system using DC electricity. Fluids in the test loop were plain water, TSP solutions, and boric acid solutions. TSP solutions had three concentrations (0.2, 0.4, 0.6%), and boric acid solutions had four concentrations (0.2, 0.4, 0.6, 0.8%). In the case of TSP, 21.4% enhancement of CHF was observed at the inlet subcooling temperature of 50 °C and extremely low mass flux (100 kg/m2 s). In the case of boric acid, 12.4% enhancement of CHF was observed at inlet subcooling temperature of 50 °C and extremely low mass flux 100 kg/m 2 s.

An experimental study was performed using R-134a in an annulus channel to investigate the heat transfer characteristics in the post-CHF region. The experiments were conducted under the outlet pressures of 1.1–2.0 MPa, the mass fluxes of 100–400 kg/m 2 s and the inlet temperatures of 25–51 °C. Experiments were conducted in three cases of the annular channel test without spacers, the test with I-type spacer grids, and the test with split-swirl type spacer grids. The empirical correlation of the heat transfer in the post-CHF region was developed with 301 data point obtained in the annular test without spacer grids. The heat transfer coefficient was calculated based on the heater-rod temperature and the saturated vapor property. The average and root-mean-square (RMS) errors of the predictions were 0.17 and 3.4%, respectively. I-type spacer grids and split-swirl mixing vanes (with blockage-area ratios of 4.0 and 5.8%) were tested, which showed the enhancing effects on the heat transfer at the downstream location of the spacers and the results from the spilt-swirl-type mixing vane tests were more effective than the I-type grid tests.

An analysis on the two-phase flow in a helical wire inserted tube using commercial CFD code, CFX11.0, was performed in bubbly flow and annular flow regions. The analysis method was validated with the experimental results of Takeshima. Bubbly and annular flows in a 10 mm inner diameter tube with varying pitch lengths and inserted wire diameters were simulated using the same analysis methods after validation. The geometry range of p / D was 1–4 and e / D was 0.08–0.12. The results show that the inserted wire with a larger diameter increased swirl flow generation. An increasing swirl flow was seen as the pitch length increased. Regarding pressure loss, smaller pitch lengths and inserted wires with larger diameters resulted in larger pressure loss. The average liquid film thickness increased as the pitch length and the diameter of the inserted wire increased in the annular flow region. Both in the bubbly flow and annular flow regions, the effect of pitch length on swirl flow generation and pressure loss was more significant than that of the inserted wire diameters. Pitch length is a more dominant factor than inserted wire diameter for the design of the swirl flow generator in small diameter tubes.

The critical heat flux (CHF) is one of the most important thermal hydraulic parameters in heat transfer system design and safety analyses. CHF enhancement allows higher limits of operation conditions such that heat transfer equipment can be operated safely with greater margins and better economy. The application of nano-fluids is thought to have strong potential for enhancing the CHF. In this study, zeta potentials of Al2O3 nano-fluids were measured and flow boiling CHF enhancement experiments using Al2O3 nano-fluids were conducted under atmospheric pressure. The CHFs of Al2O3 nano-fluids were enhanced up to ∼70% in flow boiling for all experimental conditions. Maximum CHF enhancement (70.24%) was shown at 0.01 vol% concentration, 50 °C inlet subcooling, and a mass flux of 100 kg/m2 s. Inner surfaces of the test section tube were observed by FE–SEM and the zeta potentials of Al2O3 nano-fluids were measured before and after the CHF experiments.

This paper summarized a development history and lessons of Korean nuclear power infrastructures from the beginning of the nuclear power program in 1956 to the localization of complete scope of PWR technology in 1990. The objective of this paper is to show the guideline on the issues that the development of a national infrastructure for nuclear power using the realistic experiences in order to help the developing countries newly starting nuclear power program as a long-term energy supply option. Development strategies and lessons learned from the successful Korean experience have been presented based on milestones structure of IAEA in order to help decision makers, advisers, senior managers and national planners of nuclear power program. Lessons for national nuclear power programs include considerations before launching a program, preparation and decision making, and the construction of the first nuclear power plant. Scope of these lessons includes knowledge and human resources management, financial and industrial infrastructure development, nuclear safety, legislative and regulatory experiences, fuel cycle and waste management, international cooperation. Fourteen lessons learned either positive or not are derived from the Korean case and are suggested for incorporation in the IAEA's efforts in support of developing countries’ development of nuclear infrastructure and planning.

Main objective of this study is to investigate an applicability of a steady-state heat transfer correlation to pressure transient sequences and an effect of the pressure transient rates on the overall heat transfer rates under the supercritical pressures. Heat transfer rates are brought in line for both the pressure increasing and the pressure decreasing transients. And effects of pressure transient rates on heat transfer rates are trivial. As for an applicability of steady-state heat transfer correlation to the pressure transient sequences, the heat transfer correlation always overestimates the Nusselt number measured in the pressure transient heat transfer experiments by average 30%.

Global concerns for the security of energy have steadily been on the increase and are expected to become a major issue over the next few decades. Urgent policy response is thus essential. However, little attempt has been made at defining both energy security and energy metrics. In this study, we provide such metrics and apply them to four major energy sources in the Korean electricity market: coal, oil, liquefied natural gas, and nuclear. In our approach, we measure the cost of energy security in terms of supply disruption and price volatility, and we consider the degree of concentration in energy supply and demand using the Hirschman-Herfindahl index (HHI). Due to its balanced fuel supply and demand, relatively stable price, and high abundance, we find nuclear energy to be the most competitive energy source in terms of energy security in the Korean electricity market. LNG, on the other hand, was found to have the highest cost in term of energy security due to its high concentration in supply and demand, and its high price volatility. In addition, in terms of cost, we find that economic security dominates supply security, and as such, it is the main factor in the total security cost. Within the confines of concern for global energy security, our study both broadens our understanding of energy security and enables a strategic approach in the portfolio management of energy consumption.