• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.109-118
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• 2006

Various indirect power cycle options for a helium cooled gas cooled fast reactor (GFR) with particular focus on a supercritical $CO_2(SCO_2)$ indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The balance of plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and $SCO_2$ recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of $550^{\circ}C$, (2) advanced design with turbine inlet temperature of $650^{\circ}C$ and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect $SCO_2$ recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR &proximate-containment& and the BOP for the $SCO_2$ cycle is very compact. Both these factors will lead to reduced capital cost.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3585-3593
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• 2024

Radioisotopes tend to play important roles in many fields, such as industry, healthcare, agriculture, aerospace, etc. Radioisotope production is mainly through accelerators or research reactors, and high flux research reactor is one of the most effective approaches for radioisotope production. The physical basis of preparing radioisotope relies on nuclear reactions occurring in the reactor core, which includes fission, (n,γ), (n,α), and (n,p) reaction, etc. Among them, fission and (n,γ) reaction are most important in the nuclear reactor. For example, the ⁹⁹Mo could be generated by uranium fission and extracting from the fission products, or through the radiative capture reaction from enriched ⁹⁸Mo. As for the fission method, the irradiation target is gradually transitioning from high enriched uranium (HEU) target to low enriched uranium (LEU) target due to the requirement of non-proliferation. In this paper, studies on the impacts of different fission targets on radioisotope productions are conducted. Moreover, an optimized study on the radiative capture method is performed to improve the production efficiency. It is concluded that it is advantageous to use radiative capture method to generate radioisotopes in high flux reactor, which helps to improve the specific activity with environmental friendliness.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.12
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1813-1824
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• 2022

This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from -0.3$ to -0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.

• Journal of Korea Technology Innovation Society
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• v.6 no.4
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• pp.462-479
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• 2003

The policy trends of use and the development of nuclear power in the United States, France, United Kingdom. Germany, Russia, China, Japan and Korea are briefly investigated. Nuclear power technology has been developed as the national policy in the nuclear-advanced countries. 50 years has passed since the declaration of "Atoms for Peace" by USA President Eisenhour in December 1953. Recently, it appears to revitalize the nuclear power program in world major countries in order to recover the shortage of electric power and to curb the excess emission of carbon dioxide as well as to secure competitiveness in electricity markets. Advanced countries are making new initiatives for the development of the fourth generation nuclear power system. Furthermore, wide-ranged use and development of nuclear power technologies are expected in district heating in commercial sectors, power in the space exploration, and propulsion power of large tankers and spaceships. High temperature gas cooled nuclear power reactor will be applied for mass production of hydrogen energy in the future.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1349-1359
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• 2016

Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX ($T_{co}$) and the hot outlet temperature of the intermediate heat exchanger ($T_{ho2}$) by manipulating the hot-side flow rates of the heat exchangers ($F_h/F_{h2}$) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX ($T_{co}$) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.520-525
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• 2024

The optimization design problem of nuclear reactor radiation shield is a typical multi-objective optimization problem with almost 10 sub-objectives and the sub-objectives are always demanded to be under tolerable limits. In this paper, a design method combining multi-objective optimization algorithms with paralleling discrete ordinate transportation code is developed and applied to shield design of the Savannah nuclear reactor. Three approaches are studied for light-weighted and compact design of radiation shield. Comparing with directly optimization with 10 objectives and the single-objective optimization, the approach by setting sub-objectives representing weight and volume as optimization objectives while treating other sub-objectives as constraints has the best performance, which is more suitable to reactor shield design.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.1116-1120
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• 2004

In the nuclear power plant, there are several cylindrical vessels such as reactor vessel, pressuriser and so on. The vessels are usually constructed by welding large rolled plates, forged sections or nozzle pipes together. In order to assure the integrity of the vessel, these welds should be periodically inspected using sensors such as ultrasonic transducer or visual cameras. This inspection is usually conducted under water to minimize exposure to the radioactively contaminated vessel walls. The inspections have been performed by using a conventional inspection machine with a big structural sturdy column, however, it is so huge and heavy that maintenance and handling of the machine are extremely difficult. It requires much effort to transport the system to the site and also requires continuous use of the utility's polar crane to move the manipulator into the building and then onto the vessel. Setup beside the vessel requires a large volume of work preparation area and several shifts to complete. In order to resolve these problems, we have developed an underwater mobile robot guided by the laser pointer, and performed a series of experiments both in the mockup and in the real reactor vessel. This paper introduces our robotic inspection system and the laser guidance of the mobile robot as well as the results of the functional test.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.225-229
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• 2001

In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is $5.89 {\times} 10^{17}\;n/cm^2$ and that on the radial surface of reactor vessel is $4.49 {\times} 10^[16}\;n/cm^2$. These results meet the requirement, $1.0 {\times} 10^{20}\;n/cm^2$, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.