• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.437-442
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• 2006

European Material Test Reactors (MTRs) have provided essential support for nuclear power programs over the last 40 years. MTRs are now ageing in Europe and they cannot ensure the securing of experimental capability for the next decades. In this context, a new Material Testing Reactor, named Jules Horowitz Reactor -JHR-, operated as an international user-facility, is under development in Europe. The European MTRs context and the JHR objectives and status will be presented. Emphasis will be put on experiments in the field of nuclear fuels and materials irradiation which are developed in the framework of European and international collaboration.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.285-292
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• 2016

Reactor neutrinos have been detected in the past 50 years by various detectors for different purposes. Beginning in the 1980s, neutrino physicists have tried to use neutrinos to monitor reactors and develop an optimized detector for nuclear safeguards. Recently, motivated by neutrino oscillation physics, the technology and scale of reactor neutrino detection have progressed considerably. In this review, I will give an overview of the detection technology for reactor neutrinos, and describe the issues related to further improvements in optimized detectors for reactor monitoring.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.299-305
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• 2015

Nuclear hydrogen production technology is being developed for the future energy supply system. The sulfur-iodine thermo-chemical hydrogen production process directly splits water by using of the heat generated from very high temperature gas-cooled reactor, a typical Generation IV nuclear system. Nuclear hydrogen key technologies are composed of VHTR simulation technology at elevated temperature, computational tools, TRISO fuel, and sulfur iodine hydrogen production technology. Key technology for nuclear hydrogen production system were developed and demonstrated in a laboratory scale test facility. Technical challenges for the commercial hydrogen production system were discussed.

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.155-159
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• 2016

Background: Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. Materials and Methods: To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% $^{235}U$ contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. Results and Discussion: We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. $^{241}Pu$) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate Conclusion: Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1017-1023
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• 2019

The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink $MATLAB^{(R)}$ model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• Journal of Mechanical Science and Technology
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• v.18 no.12
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• pp.2080-2094
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• 2004

A model predictive controller is designed to control thermal power in a nuclear reactor. The basic concept of the model predictive control is to solve an optimization problem for finite future time steps at current time, to implement only the first optimal control input among the solved control inputs, and to repeat the procedure at each subsequent instant. A controller design model used for designing the model predictive controller is estimated every time step by applying a recursive parameter estimation algorithm. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), was used to verify the proposed controller for a nuclear reactor. It was known that the nuclear power controlled by the proposed controller well tracks the desired power level and the desired axial power distribution.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.309-316
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• 2024

The U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.20 provides guidance on the comprehensive vibration assessment program (CVAP) to be performed on reactor internals during preoperational and startup tests. The purpose of the program is to identify loads that could cause vibration in the reactor internals and to ensure that these vibrations do not affect their structural integrity. The structural vibrational analysis program involves creating finite element analysis models of the reactor internals and calculating their structural responses when subjected to vibration loads. The appropriateness of the structural analysis methodology must be demonstrated through benchmarks or any other reasonable means. Although existing structural analysis methodologies have been proven to be appropriate and are widely used, this paper presents the development of an improved new structural analysis methodology for APR1400 reactor internals using scaled model tests.

• Strategy21
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• s.41
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• pp.261-293
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• 2017

Nuclear power is a way of attaining an enormous amount of energy with relatively small amount of resources and after it has been introduced to the submarine since 1954, there are approximately 150 of nuclear powered submarine currently on a mission around the world. This is due to the maneuverability, mountability and covertness of nuclear submarines. However, there are other tasks, not only the high level of nuclear technology that are needed to be dealt with in order to construct nuclear powered submarine. The biggest task of all is to secure the enriched uranium. Accordingly, this research is about the way of enriching and securing the nuclear fuel that are used in the nuclear submarine with the characteristics, merits and demerits of the nuclear submarine. Due to the fact that the pressurized water reactor in South Korea is the reactor that was originally built for the development of nuclear powered submarine, many parts is designed to be suitable for the submarine propulsion. However, in order to apply this to submarine it is needed to consider additional requests such as the position of reactor, accident-coping system, radioactive covering, reactor output adjustment and ship's pitch and roll in order to apply this to submarine. Nuclear submarines have much higher speed based on the powerful propulsion in comparison with diesel-electric submarine and also have bigger loading area. Besides, there is no need to snorkel and they also have advantages in covertness with the multi-noise proof system. The nuclear technology in South Korea has seen the dramatic development since 1962 and in 1998 reached to the level that we have succeeded in the localization of nuclear plant and exported the world-class one-piece small-sized reactor (SMART) to UAE. To operate these reactors, we import the whole quantity of low-enriched uranium and having our own uranium enrich facility is not probable because of the budget and international regulations. With the ROK/US nuclear agreement revised on 2015 November, the enrichment of uranium that are available without special permission has changed up to 20%. According to the assumption that we use the 20% enrichment of Uranium on U.S. virginia class submarine, it is necessary to change the fuel after 11 years and it will cause additional cost of 1 billion dollars. But the replace period by the uranium's enrichment rate is not fixed so that it is possible to change according to the design of reactor. Therefore, I would like to make a suggestion on two types of design concepts of nuclear submarine that can be operated for 30 years without nuclear fuel change by using the 20% enriched uranium from ONNp.First of all, it is possible by increasing the size of reactor by 3 times and it results in the 1,000t increase of the weight. And secondly, it is by designing the one piece reactor to insert devices such as steam turbine, condenser into the inside of nuclear core like the Rubis class submarines of France.