• Title, Summary, Keyword: Nuclear Hydrogen System

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Nuclear Hydrogen Production Technology Development Using Very High Temperature Reactor (초고온가스로를 이용한 원자력수소생산 기술개발)

  • Kim, Yong-Wan;Kim, Eung-Seon;Lee, Ki-yooung;Kim, Min-hwan
    • 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.

Investigation of a Hydrogen Mitigation System During Large Break Loss-Of-Coolant Accident for a Two-Loop Pressurized Water Reactor

  • Dehjourian, Mehdi;Sayareh, Reza;Rahgoshay, Mohammad;Jahanfarnia, Gholamreza;Shirani, Amir Saied
    • Nuclear Engineering and Technology
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    • v.48 no.5
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    • pp.1174-1183
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    • 2016
  • Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

SAFETY STUDIES ON HYDROGEN PRODUCTION SYSTEM WITH A HIGH TEMPERATURE GAS-COOLED REACTOR

  • TAKEDA TETSUAKI
    • Nuclear Engineering and Technology
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    • v.37 no.6
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    • pp.537-556
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    • 2005
  • A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

A Case Study on Quality Assurance Application of the Nuclear Hydrogen R&D (원자력수소 연구개발과제의 품질보증 적용 사례 연구)

  • Lee, Tae-Hoon;Lee, Ki-Young
    • Journal of the Society of Korea Industrial and Systems Engineering
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    • v.33 no.4
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    • pp.114-121
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    • 2010
  • Traditionally Nuclear Research and Development (R&D) result has been big influence on other industries and societies and it requires large scale investments and study period. So it is essential to apply Quality Assurance (QA) for systematic R&D management. This paper investigates QA System for U.S. Nuclear R&D and reviews QA elements. Based on this investigation, we applied QA requirements to Nuclear Hydrogen R&D project, and the scope of application be enlarged as R&D stage in progress. We also present QA system improvement way through consideration for Nuclear Hydrogen Project's QA application. As the need for QA in R&D is expected to increase in the future, it is necessary to prepare guidelines for R&D QA.

Preliminary Cost Estimates for Nuclear Hydrogen System Based on High Temperature Electrolysis (고온전기분해 이용 원자력수소 예비타당성 연구)

  • Yang, Kyeongjin;Lee, Taehoon;Lee, Kiyoung
    • 한국신재생에너지학회:학술대회논문집
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    • pp.228.2-228.2
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    • 2010
  • In this work, the hydrogen production costs of the nuclear energy sources are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of HTE process with Very High Temperature nuclear Reactor (VHTR) as a thermal energy source rather than the LUEC (Levelized Unit Electricity Cost). The general ground rules and assumptions follow G4-ECONS. Through a preliminary study of cost estimates, we wished to evaluate the economic potential for hydrogen produced from nuclear energy, and, in addition, to promptly estimate the hydrogen production costs for an updated input data for capital costs. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if $CO_2$ capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of $CO_2$. Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed.

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Preliminary Structural Sizing of the Co-axial Double-tube Type Primary Hot Gas Duct for the Nuclear Hydrogen Reactor (수소생산용 원자로에서 동심축 이중관형 1차 고온가스덕트의 예비 구조정산)

  • Song, Kee-nam;Kim, Y-W
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.4 no.2
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    • pp.1-6
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    • 2008
  • Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

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Development of a Compact Nuclear Hydrogen Coupled Components Test Loop (원자로수소생산을 위한 연결부품 실험용 소형 컴팩트 실험장치 개발)

  • Hong, S.D.;Kim, J.H.;Kim, C.S.;Kim, Y.W.;Lee, W.J.
    • Proceedings of the KSME Conference
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    • pp.2850-2855
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    • 2008
  • Very High Temperature Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR heat is transferred to a thermo-chemical hydrogen production process through an intermediate loop. Both Process Heat Exchanger and sulfuric acid evaporator provide the coupled components between the VHTR intermediate loop and hydrogen production module. A small scaled Compact Nuclear Hydrogen Coupled Components test loop is developed to simulate the VHTR intermediate loop and hydrogen production module. Main objective of the loop is to screening the candidates of NHDD (Nuclear Hydrogen Development and Demonstration) coupled components. The operating condition of the gas loop is a temperature up to $950^{\circ}C$ and a pressure up to 6.0MPa. The thermal and fluid dynamic design of the loop is dependent on the structures that enclose the gas flow, especially primary side that has fast gas velocity. We designed and constructed a small scale sulfuric acid experimental system which can simulate a part of the hydrogen production module also.

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GOTHIC-3D APPLICABILITY TO HYDROGEN COMBUSTION ANALYSIS

  • LEE JUNG-JAE;LEE JIN-YONG;PARK GOON-CHERL;LEE BYUNG-CHUL;YOO HOJONG;KIM HYEONG-TAEK;OH SEUNG-JONG
    • Nuclear Engineering and Technology
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    • v.37 no.3
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    • pp.265-272
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    • 2005
  • Severe accidents in nuclear power plants can cause hydrogen-generating chemical reactions, which create the danger of hydrogen combustion and thus threaten containment integrity. For containment analyses, a three-dimensional mechanistic code, GOTHIC-3D has been applied near source compartments to predict whether or not highly reactive gas mixtures can form during an accident with the hydrogen mitigation system working. To assess the code applicability to hydrogen combustion analysis, this paper presents the numerical calculation results of GOTHIC-3D for various hydrogen combustion experiments, including FLAME, LSVCTF, and SNU-2D. In this study, a technical base for the modeling oflarge- and small-scale facilities was introduced through sensitivity studies on cell size and bum modeling parameters. Use of a turbulent bum option of the eddy dissipation concept enabled scale-free applications. Lowering the bum parameter values for the flame thickness and the bum temperature limit resulted in a larger flame velocity. When applied to hydrogen combustion analysis, this study revealed that the GOTHIC-3D code is generally able to predict the combustion phenomena with its default bum modeling parameters for large-scale facilities. However, the code needs further modifications of its bum modeling parameters to be applied to either small-scale facilities or extremely fast transients.

A SMALL MODULAR REACTOR DESIGN FOR MULTIPLE ENERGY APPLICATIONS: HTR50S

  • Yan, X.;Tachibana, Y.;Ohashi, H.;Sato, H.;Tazawa, Y.;Kunitomi, K.
    • Nuclear Engineering and Technology
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    • v.45 no.3
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    • pp.401-414
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    • 2013
  • HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's $950^{\circ}C$, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to $750^{\circ}C$ for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to $900^{\circ}C$ for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

Preliminary numerical study on hydrogen distribution characteristics in the process that flow regime transits from jet to buoyancy plume in time and space

  • Wang, Di;Tong, Lili;Liu, Luguo;Cao, Xuewu;Zou, Zhiqiang;Wu, Lingjun;Jiang, Xiaowei
    • Nuclear Engineering and Technology
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    • v.51 no.6
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    • pp.1514-1524
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    • 2019
  • Hydrogen-steam gas mixture may be injected into containment with flow regime varying both spatially and transiently due to wall effect and pressure difference between primary loop and containment in severe accidents induced by loss of coolant accident. Preliminary CFD analysis is conducted to gain information about the helium flow regime transition process from jet to buoyancy plume for forthcoming experimental study. Physical models of impinging jet and wall condensation are validated using separated effect experimental data, firstly. Then helium transportation is analyzed with the effect of jet momentum, buoyancy and wall cooling discussed. Result shows that helium distribution is totally dominated by impinging jet in the beginning, high concentration appears near gas source and wall where jet momentum is strong. With the jet weakening, stable light gas layer without recirculating eddy is established by buoyancy. Transient reversed helium distribution appears due to natural convection resulted from wall cooling, which delays the stratification. It is necessary to concern about hydrogen accumulation in lower space under the containment external cooling strategy. From the perspective of experiment design, measurement point should be set at the height of connecting pipe and near the wall for stratification stability criterion and impinging jet modelling validation.