動力・エネルギー技術の最前線講演論文集 : シンポジウム
Online ISSN : 2424-2950
最新号
選択された号の論文の78件中1~50を表示しています
  • (15)チャギング防止ノズルとフィン付冷却管によるFCVSの高度化
    奈良林 直, 荒岡 勝政, 石川 慶浩
    セッションID: A111
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The FCVS under development in this research consists of a scrubber nozzle, a scrubber pool, a metal fiber and a silver zeolite, three-stage filters, and can constitute an ultra-high-performance filter vent system. However, it is necessary to consider the decay heat of the adsorbed radioactive materials and the heat of chemical reaction of hydrogen and CO with oxygen. Therefore, we conducted a cooling evaluation test using a cooling pipe with fins to prevent overheating in the filter, and conducted a test to evaluate the cooling effect that can maintain the silver zeolite at 800°C or less. The FCVS assumes automatic start-up by a safety relief valve (SRV) when the critical pressure of 2Pd, which is twice the design pressure of the containment vessel, is reached.

    When the FCVS starts up, the scrubber nozzle operates stably up to the saturation temperature, dissolving iodine and cesium iodide in water. For this reason, the decay heat of radioactive materials is removed by the water temperature rise in the scrubber pool and evaporation, so heat generation in metal fibers and silver zeolite filters is mainly due to chemical reactions of hydrogen and carbon monoxide with oxygen in the vent gas. As the test results, cooling with finned cooling pipes is effective and preventing overheating due to the hydrogen and CO with oxygen in the silver zeolite at the final stage.

  • (16) 銀ゼオライトAgXの共存ガス存在下での水素触媒性能の評価
    小林 稔季, 石川 慶浩, 遠藤 好司, 奈良林 直, 川原 康博, Suckow Detlef Joachim
    セッションID: A112
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The explosive gases such as hydrogen and CO will be generated in a severe accident (SA). We conducted joint research with the Swiss research institute PSI on the hydrogen catalytic properties of AgX and AgR under various conditions. Last year, we reported that the hydrogen catalytic reactivity is strong for AgX and weak for AgR. In this paper, we will report on the effects of coexisting gases such as CO. The experimental results shows that CO does not interfere with hydrogen catalysis reactivity. Furthermore, it is confirmed that AgX has an oxidizing effect on CO since CO concentration decreases and CO2 concentration increases when CO flows through AgX. As a result, we propose to utilize these characteristics as countermeasures against hydrogen and CO generated during SA.

  • (17) 銀ゼオライトAgXの水素触媒特性のPARへの応用
    石川 慶浩, 遠藤 好司, 奈良林 直, 川原 康博, Joachim Suckow Detlef
    セッションID: A113
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    When SA is generated, a large amount of hydrogen is generated due to the reaction between water vapor and zirconium. A PA is installed in the containment vessel as a countermeasure against hydrogen, but there are concerns about the effects of water vapor and methyl iodide.A silver zeolite AgX with excellent methyl iodide adsorption performance was developed and its performance was evaluated together with PSI. It was found that AgX can remove methyl iodide and hydrogen simultaneously. The properties of AgX are expected to play a supporting role for PAR.We also evaluated the hydrogen handling capacity of AgX under certain conditions and found that 1 kg of AgX can handle 0.207 kmol of hydrogen.

  • (18)銀ゼオライトXeAを用いた希ガス吸着性能評価
    椎野 朱里, 豊田 英晴, 川原 康博, 石川 慶浩, 奈良林 直, 遠藤 好司, 小林 三四郎
    セッションID: A114
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Various substances such as inorganic iodine, organic iodine, noble gases, and aerosols are generated when a severe accident occurs at a nuclear power plant. Radioactive noble gases are difficult to remove due to their poor chemical reactivity and low solubility in water, but because they have a short half-life, a hold-up system using activated carbon is applied. However, the adsorption of rare gases by activated carbon requires a long contact time and requires large equipment, so emergency response centers attached to nuclear facilities prepare air cylinders and retain the gas for two days as a countermeasure against radioactive noble gases. Rasa Industries has developed silver zeolite XeA as a rare gas adsorbent to replace activated carbon for rare gases. XeA has superior xenon retention performance compared to activated carbon used for noble gas adsorption, and it may be possible to make the conventional hold-up system more compact by replacing activated carbon with XeA. In addition, by replacing combustible activated carbon with noncombustible silver zeolite, fire safety can be enhanced. In applying XeA to a vent system, we fabricated a test device to evaluate various performances, evaluated it under the same conditions as activated carbon, and confirmed the superiority of XeA.

  • (19)銀ゼオライトを用いた高温試験装置の性能評価
    長塩 眞輝, 奈良林 直, 石川 慶浩, 川原 康博, 豊田 英晴
    セッションID: A115
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    To install an advanced Filtered Containment Venting System (FCVS) in nuclear power plants, a test device simulating the FCVS using high-temperature steam was manufactured. The equipment is branched into two systems; the normal pressure system is assumed to be installed in the CANDU reactor, and the high-pressure system is assumed to be installed in the PWR and BWR. We will give an overview of these devices. In addition, we will report the results of a test that evaluated the adsorption performance of organic iodine in the high-temperature, normal-pressure system that was previously completed.

  • 丹野 颯人, 長坂 秀雄, 木倉 宏成
    セッションID: A121
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The 2011 Great East Japan Earthquake and Tsunami caused a severe accident at the Fukushima Daiichi Nuclear Power Plants, which resulted in the loss of all power and cooling functions. During this accident, a wetwell vent was used to vent gases out of the reactor containment vessel to prevent containment failure, but the pool scrubbing effect was less than expected and a large amount of radioactive materials were released into the environment. Learning from this accident, the Nuclear Regulation Authority (NRA) mandated the installation of a filtered containment venting system (FCVS). However, when this system is used, condensation due to contact between steam and pool water, resulting in fluid vibration. In this study, the effects of nozzle orientation, air flow rate, pool water temperature, and steam mass flow rate on fluid oscillations were investigated by measuring pressure transient with a pressure sensor and photographing the gas-liquid interface with a high-speed camera, assuming FCVS operating conditions. This paper reports the results obtained from the experiments.

  • 江連 俊樹, 秋元 雄太, 松下 健太郎, 田中 正暁
    セッションID: A122
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In upper plenums of sodium-cooled fast reactors, estimation of cover gas entrainment caused by vortex dimples on the free surface is an important thermal-hydraulic issue. For this reason, the authors have developed an evaluation method of gas entrainment with an evaluation tool named, “StreamViewer”. In this study, evaluation using StreamViewer was applied to a water experiment having a simplified hot pool geometry aiming at the validation of the evaluation method toward the application to the evaluation of a pool-type sodium cooled fast reactor. In StreamViewer, the three-dimensional distribution of pressure decrease along the vortex center line was calculated from the velocity distribution obtained by CFD analyses, and the free surface dimple depth was obtained from the hydraulic balance with the pressure distribution and the cover gas pressure. As the results, it was confirmed that the onset of gas entrainment could be predicted appropriately based on the above-mentioned calculation method, showing the validity of the method.

  • 松下 健太郎, 江連 俊樹, 今井 康友, 藤﨑 竜也, 田中 正暁
    セッションID: A123
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In the design of sodium-cooled fast reactors (SFRs), cover gas entrainment phenomena induced by the vortex dimple at the free surface in upper plena is an important thermal-hydraulic issue. The authors have developed an evaluation method of gas entrainment with an evaluation tool named “StreamViewer”. In this study, the modification of the evaluation model to improve quantitatively the prediction accuracy of the gas core length was investigated. In this model, lines of the vortex center (vortex center lines) which elongated from the suction port where the entrance of the gas to the heat transport system, for instance, the IHX inlet in the pool type SFRs, to the free surface in the plenum were to be identified, and the distribution of pressure decrease along the vortex center line was calculated to judge the possibility of gas entrainment in comparisons with the hydraulic head. This evaluation model was applied to the results of the water experiment with a rectangular open channel, where unsteady vortices are generated. It was confirmed that this model can identify the occurrence of gas entrainment.

  • - 動的画像処理によるガスコア長さの計測–
    遠藤 和紀, 小林 駿輔, Hamelberg Jasmine, 堺 公明, 松下 健太郎, 江連 俊樹, 田中 正暁
    セッションID: A124
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Assuming gas entrainment (GE) to the main coolant circulation system from cover gas, which is an inert gas to cover sodium coolant in a reactor vessel of the sodium-cooled fast reactor, there is a concern that reactivity disturbance will occur when bubbles pass through the reactor core. Conventionally, an evaluation method based on static vortex extension theory has been employed for the GE prediction. however, it is known that the method gives rather overestimation for the GE occurrence from the unsteady traveling vortex dimple at the wide liquid surface. In order to contribute to understand the phenomena, experimental data have been accumulated by the basic water experiment. In this study, measurement was performed for the length of a gas cores that grew while moving on the free liquid surface by dynamic image processing, and report the types of the GEs and the occurrence conditions.

  • 曽根原 正晃, 岡野 靖, 内堀 昭寛, 青柳 光裕, 大木 裕
    セッションID: A131
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Sodium combustion analysis is one of the most important issues in the development of safety measurements in sodium cooled fast reactors. However, conservative assumptions that lead to excessive safety margins and expensive designs were made in the previous safety evaluations. In order to evaluate the effects of sodium combustion in three dimensions, an analysis code AQUA-SF has been developed, which enables more detailed elucidation of sodium combustion phenomena and evaluation of the effectiveness of safety measures. In this paper, we examine the multidimensional effects of spray combustion for the SNL-T3 test as a benchmark analysis. In order to simulate the decrease of pressure and the temperature rise near the floor during the test, a new model is developed to take into account temporal cessation of sodium ignition, increased drag coefficient due to droplet deformation and liquid splash effect due to collision between the jet stream and the floor surface, and the results are compared with the test measurements.

  • 曽我部 丞司, 神山 健司, 飛田 吉春, 岡野 靖
    セッションID: A132
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.

  • 石田 真也, 田上 浩孝, 飛田 吉春, 岡野 靖, 山野 秀将, 久保 重信
    セッションID: A133
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The SIMMER-V code is being developed in cooperation between Japan and France to enable a consistent analysis of core disruptive accidents (CDA) from the initiating phase to transition phase and to be applicable to various types of cores in the evaluation of CDA in sodium-cooled fast reactors (SFRs). The main subject of the SIMMER-V development is to develop a new detailed fuel pin model to simulate the fuel pin behavior from accident initiation to fuel pin failure. This paper describes the development of verification of the fuel pin model, which is designed to be able to handle an axial heterogeneous core configuration and an annular fuel pellet. Along a verification matrix, in this study, the pin model addressing five physical phenomena (thermal behavior, fission gas release, fission gas swelling, mechanical deformation, and in-pin fuel motion) has been verified in comparison with existing initiating phase simulation code, thereby improving the reliability of SIMMER-V as a safety assessment tool for SFRs.

  • ~米国高速実験炉EBR-II適用による機能確認~
    吉村 一夫, 堂田 哲広, 中峯 由彰, 藤崎 竜也, 井川 健一, 飯田 将来, 田中 正暁
    セッションID: A134
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In Japan Atomic Energy Agency (JAEA), a virtual plant model of the sodium-cooled fast reactor (SFR) plant composed in a computer is being developed to reduce the development cost, by replacing the costly experiments to the numerical simulations with analyses of the physical phenomena accounting for the interaction between components under various plant conditions. To establish the methodology to construct the virtual plant model and perform the coupled analysis, the results of the numerical analysis of a ULOHS test conducted in the U.S. experimental fast reactor named EBR-II was examined. In the virtual plant model of EBR-II, the upper plenum in the reactor vessel and the cold plenum, the core, and other components in heat transport system were modeled by using a multi-dimensional computational fluid dynamics (CFD) code, a coupled method of core thermal hydraulics, thermal deformation of the fuel assemblies, and neutronics analysis codes, and a plant dynamics analysis code, respectively, in the framework of the multi-level simulation system developed in JAEA. Through the numerical analysis of the ULOHS test, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.

  • 部門設立からの歩みと出版の意義:動力エネルギーの人類への役割
    小泉 安郎
    セッションID: A211
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The Power and Energy Systems Division (PESD) of the Japan Society of Mechanical Engineers was established in 1990. PESD decided to publish books of thermal and nuclear power generation as one of 30th anniversary projects. It can be said that we are at a turning point, such as, reducing carbon dioxide emissions to zero, and coping with the accident of the Fukushima Daiichi Nuclear Power Plants. Eight volumes are published by Elsevier.

  • (Vol. 1動力エネルギー工学の基礎)
    大川 富雄
    セッションID: A212
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    To understand the thermal and nuclear power generation, basic knowledge on thermodynamics, fluid dynamics, heat transfer, combustion, and nuclear physics is indispensable. In addition, knowledge on the history and practical experience of power and energy systems engineering is also definitely useful in developing improved systems. In the first volume of the memorial publication for the 30th anniversary of JSME Power & Energy Systems Division, the essence of such knowledge was described by the experts in each field. The major issues in the field of power generation and future prospects were also discussed.

  • ナトリウム冷却高速炉の開発 -「常陽」「もんじゅ」から実証炉へ-
    大野 修司, 前田 誠一郎
    セッションID: A213
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. The book is a collection of the past experience of design, construction, and operation of the experimental reactor “Joyo” and the prototype reactor “Monju”, the latest knowledge including related research and development activities and technology for SFR designs, and the future prospects of SFR development in Japan, looking back the history of development of fast reactors started in the early 1960s. The development of sodium-cooled fast reactor in Japan, which contributes to energy security and high-level waste reduction, is reaching to the stage where demonstration reactor will be deployed based on the experience of “Joyo” and “Monju” design, construction and operation. The present report introduces outlines of experiences, results and activities accumulated through these reactors and R&Ds for demonstration reactor.

  • (ナトリウム冷却高速炉の開発-社会実装に向けた熱流動・安全性研究)
    田中 正暁, 内堀 昭寛, 岡野 靖, 横山 賢治, 上羽 智之, 江沼 康弘, 若井 隆純, 浅山 泰
    セッションID: A214
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. The book is a collection of the past experience of design, construction, and operation of the experimental reactor “Joyo” and the prototype reactor “Monju”, the latest knowledge including related research and development activities and technology for SFR designs, and the future prospects of SFR development in Japan, looking back the history of development of fast reactors started in the early 1960s. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. As a new way forward to achieve a social implementation of SFR, this introductory paper also provides an overview of an integrated evaluation system named “ARKADIA” to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.

  • 高速炉の運用条件を考慮した規格基準類の開発
    岡島 智史, 高屋 茂, 若井 隆純, 浅山 泰
    セッションID: A215
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. Codes and standard for SFRs have been and is being developed considering differences of operation conditions between SFRs and existing LWRs. In this part, outline of the development of the codes and standards for SFRs is explained. The explained fields contain elevated temperature design and materials, seismic design, and operation and maintenance.

  • (BWR炉心燃料の研究開発)
    師岡 愼一, 西村 章, 吉本 佑一郎
    セッションID: A222
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In Chapter 3 of this book, the step-by-step research and development of BWR reactor core and fuel assembly are described to realize uranium savings by high burnup fuel and MOX fuel. This presentation described the reactor internal and coolant flow paths in a reactor pressure vessel, advances in fuel development, thermal-hydraulic design, proving test on thermal-hydraulic performance, new design criteria (Post-Boiling Transition standard), and the issue in developing the future fuel assembly.

  • (原子炉定格熱出力一定運転と高精度流量計によるアップレートと課題)
    森 治嗣
    セッションID: A223
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Under the basic concept of maximizing the use of existing nuclear power plants on the premise of maintaining and improving safety, advanced utilization of light water reactors, such as power uprate to make use of reactor thermal power output maximize and increase electrical power output, has already implemented mainly in Europe and the United States. Compared to the construction of new nuclear power plants, the less risk exists in terms of licensing and investment issues, and not-a-few countries are now implementing reactor power uprate as an effective means of reducing greenhouse gas emissions and providing energy security.

    There are two types of increasing power output to gain larger electric power more efficiently, which have been carried out to date: one is to increase electric power output without changing equipment and rated thermal power generated by fission in a core, which is called as the constant-rated reactor thermal power operation; the other is to uprate the licensed rated-thermal-power of the reactor with changing equipment to achieve a large increase in electric power output. Current presentation outlines increase in electric power output by constant-rated reactor thermal power operation and by uprate.

  • (過酷事故に対する格納容器強化)
    岩城 智香子, 横堀 誠一
    セッションID: A224
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In Vol.4 "Boiling Water Reactor" of JSME series in thermal and nuclear power generation, progress in reinforcement of containment vessel design considering severe accidents is described extensively. Also, regarding the core catcher, which is one of the measures to cool the molten core to enhance the safety of the containment vessel, its current technique of each plant and related R&Ds are described. This paper introduces some of them. Recently, the development of innovative light water reactors with improved safety based on current light water reactor designs has accelerated. This paper also introduces the application of containment vessel reinforcement technology to the innovative light water reactors.

  • (福島第一原子力発電所の事故の教訓を生かしたパッシブ冷却系)
    奈良林 直
    セッションID: A225
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Many lessons can be learned from the Fukushima Daiichi NPS accident in Chapter 4. First, if an isolation condenser (IC) had continued to operate, the accident would have terminated soon. Reactor core isolation cooling (RCIC) steam turbines also stopped because loss of battery power in Units 2 and 3, and temperature and pressure in each primary containment vessel (PCV) were so high that the accident management and water injection took too long. After the loss of emergency core cooling system (ECCS) and IC core cooling, fuels in the core melted. Leak of fission product and hydrogen began because of the damage to the O-ring seals between PCV upper flanges due to high temperature. Hydrogen explosion occurred in the upper floor in the reactor building at Units 1, 3, and 4. The New Regulatory Requirements, based on the concept of “defense in depth,” for Commercial Nuclear Power Reactors came into force on July 8, 2013. It is hoped that the lessons learned from this accident will improve the safety of nuclear power plants worldwide.

  • ((1)日本における高温ガス炉の研究開発)
    武田 哲明, 橘 幸男, 佐藤 博之, 大橋 弘史, 久保 真治, 篠崎 正幸, 坂場 成昭, 西原 哲夫
    セッションID: A231
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    A high-temperature gas-cooled reactor (HTGR) is a nuclear reactor that can supply high-temperature heat energy of 750°C-950°C by using a spherical fuel coated with ceramics such as carbon and silicon carbide, inert helium gas as a coolant, and graphite as a moderator. HTGR can be used in various ways such as hydrogen production and process heat supply as well as power generation. At present, a light water reactor is mainly used for nuclear power generation, but HTGR is attracting attention mainly for the following two reasons: the first reason is that the nuclear heat should be used for fields other than power generation, because fossil fuels and renewable energy have a limit to meet the increasing energy demand in the future. The second reason is excellent safety. Among all reactor types, HTGR has excellent inherent safety and its safety has been demonstrated in experimental reactors. This volume 5 provides the latest research on HTGR development and utilization, beginning with an analysis of the history of HTGRs.

  • (2)我が国の高温ガス炉技術に基づく高温ガス炉実証炉開発
    橘 幸男, 野口 弘喜, 角田 淳弥, 大橋 弘史, 佐藤 博之, 坂場 成昭
    セッションID: A232
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    For the High-Temperature Gas-cooled Reactor (HTGR) demonstration reactor, establishment of an annular core, enlargement of isolation valves, etc., establishment of codes and standards, establishment of technology to connect the hydrogen production system to the HTGR, etc. are required. In FY2022, JAEA started a project to connect hydrogen production system to the HTTR. JAEA will advance safety evaluations to ensure sufficient safety, and confirm the hydrogen production technology using nuclear heat for the first time in the world. JAEA has been requested by the UK and Poland for technical cooperation for the introduction of HTGRs, and has started technical cooperation to make the best use of the technology for the future demonstration reactor in Japan. This report introduces the current status of development for the HTGR demonstration reactor.

  • カーボンニュートラルに向けた火力発電の技術動向と将来展望
    藤井 智晴
    セッションID: A234
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In order to achieve carbon neutrality in 2050, it is essential to reduce the amount of carbon dioxide emitted from thermal power generation. For this reason, research and development are underway to further improve the efficiency of thermal power generation and expand the use of decarbonized fuels such as hydrogen, ammonia, and biomass. In addition, in order to ensure Japan's energy security, it will be necessary to continue to generate thermal power using a certain amount of fossil fuels, but in this case, it will be necessary to use and storage the captured carbon dioxide. Furthermore, variable renewable energy such as solar power and wind power is expected to continue to increase in the future, and thermal power generation is expected to play a role as an adjustment power to respond to this output fluctuation. Based on this background, this paper describes current technological trends and future prospects for thermal power generation.

  • ボイラ技術の史的展開-我々は次世代に何を残すべきか
    小澤 守
    セッションID: A235
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Steam power has been continuously a leading technology since the beginning of the 18th century. The development of steam power has been closely related to advancements in societal factors such as industries, economy and trading, together with the development in iron and steel technology, construction, and machine tools, as well as the scientific knowledge. The present book of Vol. 2 in “JSME Series in Thermal and Nuclear Power Generation,” is dedicated to describing the history of boiler development and the state of the art in the boiler technology. The existing conventional power boilers are being successively replaced by the USC (Ultra-Supercritical) technology, and the development of even more efficient A-USC (Advanced Ultra-Supercritical) boilers is currently underway. In Japan, the new technology referred to as IGCC (Integrated Gasification Combined Cycle) shows promising potential for clean coal usage and is progressing successfully.

  • 山野 秀将, 栗坂 健一, 高野 和也, 菊地 晋, 近藤 俊樹, 梅田 良太, 白倉 翔太
    セッションID: A241
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Next generation innovative reactors have a new value of their flexibility with variable renewable energy. A sodiumcooled fast reactor (SFR) can make flexibility by coupling a thermal energy storage (TES) system with molten salt. New challenging items for the SFR coupled with TES are to develop safety design approach and a heat exchanger between sodium and molten salt. On that account, a three-year project has been performed to develop 1) a safety design approach and risk assessment methodology of the SFR with TES, 2) a performance evaluation technology of a heat exchanger between sodium and molten salt as well as heat transfer improvement measures, and 3) an evaluation technology of chemical reaction characteristic between sodium and molten salt as well as safety improvement measures. This paper describes the project overview and progress in JFY2022.

  • -HTTR-熱利用試験計画-
    石井 克典, 守田 圭介, 野口 弘喜, 青木 健, 水田 直紀, 長谷川 武史, 永塚 健太郎, 野本 恭信, 清水 厚志, 飯垣 和彦, ...
    セッションID: A242
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    JAEA initiated the HTTR heat application test project coupling a hydrogen production facility to the HTTR (high temperature engineering test reactor), a research reactor of High Temperature Gas-cooled Reactor (HTGR) which can supply high temperature heat of 950°C. The project aims to establish “coupling technologies” between HTGR and hydrogen production achieving large-scale, stable and economically competitive carbon-free hydrogen production using the HTGR heat. Important considerations towards establishment of coupling technologies are development of system technologies for HTGR hydrogen production systems and components required for coupling between two facilities. This paper explains a system concept of the HTTR heat application system which can maintain safe and stable operation of the HTTR against temperature transients induced by abnormal events in a hydrogen production plant with the results of operational scheme as well as heat and mass balance of the system. Development plans for hot gas duct, high temperature isolation valves and helium gas circulators are also presented.

  • (1)基本概念と炉心および格納容器冷却系
    奈良林 直, 木倉 宏成, 笠井 和彦
    セッションID: A243
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Nuclear power generation is a stable basic power source that does not emit CO2 on the premise of ensuring safety, and has recently been re-evaluated as an attractive option from the viewpoint of energy security and environmental protection. Factors such as the recent sluggish power demand, power grid capacity limits, and initial investment limits to avoid risks do not favor large-scale plant output. In order to globalize nuclear power generation to mitigate the greenhouse effect, we need a small modular reactor (SMR) that can be easily adopted in any country and can be modularized and manufactured in factories with short construction periods. The concept of the reactor introduced in this section has a simplified BWR (LSBWR) configuration with a low output, long operating cycle, and comprehensive safety features, which was presented in 1999 at the annual meeting of the JSME and ICONE11 by Narabayashi et, al. To be economically competitive, the LSBWR design includes system and structural simplifications, modularity for short construction times, and increased availability. Comprehensive safety features are not intended to be evacuated by reliable equipment or systems such as lower core layout, IVR features, and hybrid ECCS including passive system. The concept proposed here is to provide flexibility for different site conditions and power demands, reduce investment risk and promote public acceptance. Finally, the author also introduces a new SMR named GX-BWR, which uses a reactor internal recirculation pump (RIP) for the purpose of load follow with fluctuating renewable energy and enhance facilitates for stable grid control.

  • 有馬 博史, 末広 翔一, 西口 正尚
    セッションID: B111
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The plate heat exchangers (PHEs) with titanium plate are used on the Ocean Thermal Energy Conversion (OTEC) system. Because the system is using an ammonia as a working fluid and a seawater as heat sources. The OTEC system is required to improve the thermal performance of the heat exchangers. The author proposed that the application of microchannel heat exchanger as an alternative to titanium PHEs to improve the performance of OTEC. The microchannel heat exchanger has high heat transfer performance due to the boiling bubbles confined on narrow channel and the bubble formed narrow liquid film on the heated surface. Then, the heat transfer dominates evaporative heat transfer. However, there is few researches of the ammonia boiling in the microchannel heat exchanger. In this study, the unique designed microchannel heat exchanger was manufactured and the heat transfer performance of ammonia convective boiling in the heat exchanger was measured. The result of that the comparison of overall and boiling heat transfer at different mass flow rate of working fluid, hot water temperature and mass flow temperature of hot water conditions were reported.

  • 斎藤 海希, 金井 大造
    セッションID: B112
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Pool scrubbing is a critical measure employed during severe accidents at nuclear power plants to mitigate the release of fission products (FP) into the environment. It involves passing gas containing FP through a liquid filter before its release during depressurization. Accurately predicting the efficiency of FP removal through pool scrubbing is essential for realistic Probabilistic Risk Analysis (PRA) and requires a deep understanding of gas-liquid two-phase flow phenomena. This study aims to develop novel techniques for obtaining comprehensive insights into these complex phenomena. In this work, a new approach utilizing multi-view imaging is being developed to acquire three-dimensional (3D) two-phase flow data. The study also focuses on the development of a machine-learning based two-phase flow simulation and a methodology to incorporate the acquired 3D flow data into the calculation. By integrating this data-driven approach, the developed techniques will contribute to the improvement of the PRA methodologies by providing a thorough understanding of gas-liquid two-phase flow phenomena.

  • (ブロック式小型熱交換器の凝縮伝熱流動特性評価)
    千葉 皓太, 岩城 智香子, 佐藤 正幸, 片山 義紀, 椎原 克典, 荒木 翔太, 中野 秀士, 田中 徹
    セッションID: B113
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In the development of a compact heat exchanger with additive manufacturing, the thermal-hydraulic characteristics of a block type heat exchanger were experimentally evaluated. Three types of block type heat exchangers with different internal microchannel structures were manufactured by using powder bed fusion type metal 3D printer. As a result, it was confirmed that the heat exchange rate could be improved by staggered arrangement of the micro channels and by reducing the diameter of the channels. The performance prediction of the block type heat exchanger with microchannels is expected to be possible by modifying the heat exchange correlation based on the existing heat transfer correlation equation.

  • 齊藤 大輔, ヨセフス・アルディーノ クルニアント・プライトノ, 三輪 修一郎, 武居 昌宏
    セッションID: B114
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    One of the essential parameters for the safety and optimal control of industrial facilities is the void fraction α, which represents the percentage of gas phase volume in a unit volume of a pipeline. Wire mesh sensor is a technique to directly measure the void fraction distribution at high speed at several kHz or faster. However, it is highly invasive and not easy to maintain. In this study, Void fraction estimation method that combines multi-electrode impedance measurement and machine learning (MCV-ML) is proposed. The MCV-ML is composed of four steps which are 1) in-situ void fraction measurement by wire-mesh sensor (WMs), 2) Simulated voltage by electrical simulation, 3) Measured voltage by MCV system, and 4) training ML model and estimated void fraction . Experiments on the bubbly to slug flow were conducted in a vertical, inclined, and horizontal pipe. As a result, the MCV-ML predicts the with an averaged normalized cross correlation of 0.70. The trained machine learning model clearly estimated slag but failed to capture non-slag bubbles.

  • - PIVによる実験体系内の速度分布の評価–
    小林 駿輔, 遠藤 和紀, Hamelberg Jasmine, 堺 公明, 松下 健太郎, 江連 俊樹, 田中 正暁
    セッションID: B121
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Assuming gas entrainment (GE) to the main coolant circulation system from cover gas, which is an inert gas to cover sodium coolant in a reactor vessel of the sodium-cooled fast reactor, there is a concern that reactivity disturbance will occur when bubbles pass through the reactor core. Conventionally, an evaluation method based on static vortex extension theory has been employed for the GE prediction. however, it is known that the method gives rather overestimation for the GE occurrence from the unsteady traveling vortex dimple at the wide liquid surface. In order to contribute to understand the phenomena, experimental data have been accumulated by the basic water experiment. In this study, the velocity distributions were measured under the conditions where GE occurs by particle image velocity (PIV) measurement in an experimental system to observe the gas cores that grow from the unsteady traveling vortex dimple.

  • WANG BICHENG, FU YAN, 梅原 裕太郎, 森 昌司
    セッションID: B122
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    To promote the "Basic Hydrogen Strategy" first formulated by Japan [1], hydrogen storage technology based on toluene and methylcyclohexane (MCH) reversible reaction has received widespread attention due to its convenience in hydrogen transport. Superheated vaporization is a necessary process for the catalytic dehydrogenation of MCH, and a novel method with a simple structure to generate superheated water steam quickly and efficiently have been proved in our previous study. However, the start-up response and efficiency of this porous media steam generator has not yet been clarified while using organic fluid. As MCH is toxic to human body, ethanol, which has similar physical properties to MCH, was used in our experiment. Compared to the water, ethanol with lower latent heat has better performance in generating superheated vapor rapidly and efficiently, which also proves that the proposed vapor generator is suitable for the MCH reaction. Considering both heat transfer and fluid flow in the porous media, the energy utilization efficiency and surface temperature were analyzed using one-dimensional models.

  • 塩澤 龍一, 舩谷 俊平
    セッションID: B123
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Cavitation occurrence has been reported in automotive transmission under certain conditions. When cavitation occurs in automotive transmissions, there is concern about the effects on the physical properties of the transmission oil due to the local transfer of heat from the cavitation bubble collapse to the transmission oil(1). Although there is no previous study of heat transfer to oil by cavitation bubble collapse, the details of this phenomenon are still unknown. In this study, we measured the temperature rise phenomenon caused by acoustic cavitation in the transmission oil. For the measurements, we used the non-contact visualization system to measure temperature distribution in colored oil based on the two-color LIF method(2), which was established in our previous study. This system uses Pirromethene597 as the fluorescent particle because of its temperature dependence of the fluorescence spectrum and its ability to dissolve in oil. The excitation wavelength of Pirromethene597 is longer than that of transmission oil, so the fluorescence of the oil itself can be reduced. In addition, this system is effective for high-precision temperature measurement in colored oil because it does not require positional calibration by extracting two wavelengths from a color image taken by a single CMOS camera. As a result, we succeeded in capturing the temperature change caused by acoustic cavitation bubble collapse in transmission oil as a visualized image and the generation of a high temperature field in a small area due to the cavitation bubble collapse was observed. In addition, the feasibility of this measurement system was demonstrated.

  • 岡本 孝司
    セッションID: B131
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Online-maintenance is the superior tool to improve safety of the Nuclear Power Plant. Without the Risk consideration, including Probabilistic Risk Assessment, online-maintenance cannot be performed. Thus, people at the NPP should understand the Risk. Application of online-maintenance has few de-merit, but lots of risk reduction.

  • 奈良林 直
    セッションID: B132
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    It is very important to enhance safety and stable operation of nuclear power plants through maintenance technologies and human activities. Recent years, the capacity factor of Japanese plant is very low compared with the European and American plants. During a decade, the capacity factor of Japanese plants getting worth and worth. The old type of maintenance during annual inspection made to slowdown the development of new type maintenance, such as on-line monitoring and maintenance. Therefore, we should learn from oversea-cases to develop a lot of maintenance technologies and conduct the daily maintenance activities and obtain the public trust.

  • 石橋 文彦, 岡本 孝司
    セッションID: B134
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    “On-Line Maintenance (OLM)” with effective utilization of IRIDM (Integrated Risk-Informed Decision Making) can contribute both plant safety enhancement and plant operability improvement simultaneously in nuclear power plants (NPPs). At the stage of planning and implementing OLM in NPPs, all workers should recognize the risks, and establish risk management actions. These processes are very important to enhance plant safety. OLM can contribute leveling maintenance work loads which used to be accumulated during outage. The more experienced workers can lead maintenances, the more technical transfer from experienced worker to young worker will be made. As OLM has already had excellent results in the United States and Europe, we shall promote implementation of OLM in NPPs in Japan.

  • 更田 豊志, 滝吉 幸嗣, 小野 達也, 村松 健
    セッションID: B135
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    The Nuclear Regulatory Authority has so far effectively disallowed the implementation of on-line maintenance(OLM), and it is considered important for the introduction of OLM to strengthen the risk-informed technical basis for each LCO, remedial action, and AOT. We will discuss the steps to be taken by regulators and operators with regard to the implementation of OLM.

  • (上流外乱が流体温度変動特性に与える影響)
    三好 弘二
    セッションID: B211
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Thermal fatigue cracking may initiate at a tee pipe where high and low temperature fluids flow in. In this study, the penetration depth and fluid temperature fluctuations in a branch pipe of a mixing tee were investigated under flow patterns where the main pipe flow impinges on the branch pipe wall and penetrates into the branch pipe. The test section consists of a horizontal main pipe with an inner diameter of 150 mm and a vertical branch pipe with an inner diameter of 50 mm. A 45° elbow was installed at the upstream position on the branch pipe side in order to study the effect on the penetration depth and temperature fluctuations. Fluid temperature distribution along the branch pipe was measured with eight sheathed thermocouples. The maximum penetration depth into the branch pipe increased when the upstream elbow was installed on the branch pipe side. The fluid temperature fluctuations also increased, especially in the range of relatively small momentum ratios where the hot mainstream intermittently penetrated into the branch pipe.

  • (高温・高圧条件での最大侵入深さの推定)
    歌野原 陽一, 三好 弘二
    セッションID: B212
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Thermal fatigue cracks may be initiated at T-junction pipes where high- and low-temperature fluids flow in and mix. The authors previously conducted visualization studies of flow fields in a T-junction pipe. As a result, flows from the main pipe penetrated into the branch line intermittently depending on the momentum ratio between main and branch lines. In the present study, numerical simulations were carried out to reproduce flow fields observed in the visualization studies and estimate the temperature on the inner surface of the blanch line. The large eddy simulation (Dynamic Smagorinsky) was carried out. Inner diameters were Dm = 60 mm (main) and Db = 20 mm (branch). The simulation results of the maximum penetration depth for experimental conditions were agreed well with experimental data quantitatively. Temperature fluctuation appeared on the inner surface of the branch line. As the momentum ratio increased, the peak value of the temperature fluctuation appeared at the deeper point on the branch inner surface. When the temperature difference between the main and branch pipes increased from 30°C to 150°C, the simulation results of the maximum penetration depth increased by a factor of 1.2 to 1.5. The possible reasons were increase of the buoyancy and the decrease of the viscosity which caused the decrease of the wall friction.

  • 平野 庫一郎, 山縣 貴幸, 森田 良
    セッションID: B213
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In this study, the effect of the pipe diameter ratio on the pipe-wall thinning at T-junction piping system is investigated. Mass transfer coefficient is experimentally measured by plaster dissolution method in a test model with different pipe diameters for the main and branch pipes. The experimental results show that the mass transfer coefficient is higher on the wall opposite the branch pipe connection, and the single peak is observed at the connection side.

  • 内山 雄太
    セッションID: B214
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In order to evaluate the occurrence conditions and rate of progression of flow-induced aging degradation phenomena such as pipe wall thinning and vibration fatigue in steam piping system, it is essential to accurately understand the steam flow conditions inside piping. However, the flow structure (flow pattern and sound velocity) of water-steam two-phase flow under high steam quality conditions is not sufficiently clear at present, and the occurrence and progression of degradation phenomena may not be properly evaluated. In this study, the flow pattern of water-steam two-phase flow under high steam quality conditions was visually observed to clarify the flow pattern, and the possibility of determining the flow pattern using existing flow regime map was demonstrated through comparison with existing flow regime map of water-air two-phase flow.

  • (ひずみ速度不感帯に関する一考察)
    堤 一也
    セッションID: B221
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In hot water simulating LWR coolants, the fatigue life of stainless steel depends strongly on temperature and strain rate and decreases with decreasing strain rate. On the other hand, it has been experimentally confirmed that a certain strain range on the compression side does not contribute to the reduction of fatigue life and has been reported as the insensitive strain region. Although the evaluation method of environmental effects is prescribed in the JSME Standard Environmental Fatigue Evaluation Method, it is expected that the amount of reduction in fatigue life can be predicted more rationally by considering this region.

    In the Fatigue Evaluation Subcommittee and the Fatigue Evaluation Techniques Working Group of the JSME, we are studying a method to estimate the value of this region easily and safely, and also incorporate it into fatigue assessment. This paper describes the outline of this activities.

  • 千田 格, 廣田 圭一, 齊藤 和宏, 西岡 剣, 吉田 洋輔, 原 悠, 畠山 和輝, 馬場 俊樹
    セッションID: B222
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    Some damage was found at the forks of low-pressure turbine blades caused by steam-flashback vibration at nuclear power plant. As countermeasure against vibration during operation, laser peening for turbine blade was investigated and effectiveness was evaluated by material tests. 12 Cr stainless steel test pieces with pin hole were fabricated and laser peening was performed on the surface. As the result of material evaluation, it was confirmed that compressive residual stress was induced to the material surface by laser peening and fatigue strength was improved about 30% compared to unpeened test pieces. In the near future, laser peening will be applied to turbine blades to improve plant reliability.

  • 加藤 優一, 竹田 陽一
    セッションID: B223
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In order to clarify the crack growth behavior of Alloy 625 at elevated temperatures, the crack growth rate (CGR) was evaluated in a superheated steam environment at 750°C under cyclic loading. A compact tension specimen with a width of 50 mm and a thickness of 12.5 mm was used for the test. Three types of loading waveforms with the same cyclic period of 0.0003 Hz were applied. One was a triangular waveform in which loading and unloading rates were the same. The others were sawtooth waveforms with a fast loading rate (F-S) and with a slow loading rate (S-F). The CGR per unit cycle was 1.42 times higher for F-S and 0.87 times higher for S-F than the one obtained under the triangular waveform. It was confirmed that even at the same loading frequency, the waveform difference influences the CGR. Under the F-S waveform in which the faster CGR was exhibited, the thicker oxide film was formed around the crack, suggesting that an increase in unloading time leads to more oxide film formation. This is an acceleration factor for the crack growth rate under the same loading frequency.

  • 渡辺 瞬, 湯淺 朋久, 森田 良
    セッションID: B224
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    When a leakage of high temperature fluid occurs in a power plant due to deterioration or damage of piping, thermal effects on the human body can be expected due to splashing of the fluid. In this study, we modeled human skin as three layers (epidermis, dermis and subcutaneous fat) and attempted to evaluate heat conduction in the skin when exposed to high temperature fluid. It was found that the temperature of the subcutaneous fat gradually increased depending on the exposure time when the fluid was at a high temperature of 100°C or higher. The severity of the burn index calculated by the temperature distribution in the area in question increased rapidly soon after the skin was exposed to the high temperature fluid, and was determined to be the most serious “Third degree heat injury” in this analysis. In the future, we will evaluate the PBI (Prognostic Burn Index) using the obtained burn index and attempt to quantify the survival rate at the time of burn injury.

  • 日隈 幸治, 峯村 武宏, 西村 達仁, 西 優弥
    セッションID: B225
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In domestic nuclear power plant, realization of stable operation, improvement of capacity factor, and reduction of maintenance cost are the most important challenge after restart for recovery of investment in safety equipment related to new regulatory requirements, achieving carbon neutral, energy security, and so on. As one of the measures, maintenance optimization for power plant components is mentioned. However, there are few applications of methods for quantitatively evaluating inspection frequency, maintenance methods, and cost for optimization. In this paper, aiming reduction of failure risk and maintenance cost during equipment life, we propose a method for predicting failure risk and maintenance cost during long-term operation by associating each failure rate of initial failure, accidental failure, and wear and tear failure estimated from actual failure cases with equipment operation(operation/maintenance) plan and results.

  • 小久保 知己, 鈴木 研悟
    セッションID: C111
    発行日: 2023年
    公開日: 2024/03/25
    会議録・要旨集 認証あり

    In a liberalized market, investment in power generation facilities depends on competition in the electricity market. Therefore, new policies are being considered to encourage facility investment to ensure a stable electricity supply. However, few studies have dealt with short-term market competition and long-term power supply investment. One of the reasons for this is the large amount of computation time required. Therefore, this study investigates a method to reduce the simulation time of ABM, which represents short-term wholesale power transactions over a year. By increasing the time increments handled by the electricity auction, the market price and estimated profit were slightly increased. There was little impact on the power supply mix. By using the original increment width, the computation time of the simulation was reduced to 22% without significantly affecting the market price and estimated profit.

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