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Journal Articles

Comparison between passive reactor cavity cooling systems based on atmospheric radiation and atmospheric natural circulation

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 151, p.107867_1 - 107867_11, 2021/02

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We compared the RCCS using atmospheric radiation with that using atmospheric natural circulation in terms of passive safety features and control methods for heat removal. The magnitude relationship for passive safety features is heat conduction $$>$$ radiation $$>$$ natural convection. Therefore, the magnitude for passive safety features of the former RCCS can be higher than that of the latter RCCS. In controlling the heat removal, the former RCCS changes the heat transfer area only. On the other hand, the latter RCCS needs to change the chimney effect. It is necessary to change the air resistance in the duct. Therefore, the former RCCS can control the heat removal more easily than the latter RCCS.

Journal Articles

A Study on sodium-concrete reaction in presence of internal heating

Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800$$^{circ}$$C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800$$^{circ}$$C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

JAEA Reports

Phase 1 code assessment of SIMMER-III; A Computer program for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu

JAEA-Research 2019-009, 382 Pages, 2020/03

JAEA-Research-2019-009.pdf:22.82MB

The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.

Journal Articles

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 133, p.830 - 836, 2019/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.

Journal Articles

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

Journal Articles

Decision-making process on selection of management policy for high-level radioactive waste; As an example of Switzerland and United Kingdom

Osawa, Hideaki; Hirose, Yukio*; Onuma, Susumu*; Otomo, Shoji*

Shakai Anzengaku Kenkyu, 9, p.145 - 160, 2019/03

We compared decision-making process on selection of management policy for high-level radioactive waste between Switzerland and UK by document review. The decision-making was performed extraction of options, establishment of comparative criteria, assessment and selection of options in turn. As comparative criteria, safety and intergenerational fairness etc. were considered to be important. The position of 'Deal with it now by waste disposal' emphasizing the responsibility of the present generation had a trade-off relation with that of 'Leave it until later by long-term storage' emphasizing the rights of future generations. The sense of values, recusing HLW repository as so-called troublesome facility against future and present generation, appeared to be mitigated through the consideration of trade-off relation described above. As a result, both countries have chosen waste disposal considering reversibility of decisions and retrievability of radioactive waste.

Journal Articles

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 122, p.201 - 206, 2018/12

 Times Cited Count:1 Percentile:74.25(Nuclear Science & Technology)

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.

Journal Articles

Evaluation of the effect of spent fuel layout on SFP cooling with MAAP5.04

Nishimura, Satoshi*; Satake, Masaaki*; Nishi, Yoshihisa*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

After the accident of Fukushima-unit 1 Nuclear Power Plant, Japanese utilities are newly requested by regulatory body to take prompt measures to enhance the safety of spent fuel pool. The most important objective of this new Japanese standards of regulation is keeping a water level in a Spent Fuel Pool (SFP) under any situations in order to prevent fuel failures due to increase of fuel temperature and to avoid the occurrence of re-criticality accidents. The utilities are considered to install several kinds of safety measures for SFP. For example, a spray injection and an alternate water injection to keep pool water level, and a spent fuel layout, such as 1 by 4, 1 by 8, checkerboard to enhance cooling of the spent fuel in SFP. The objective of the present study is to investigate the effect of spent fuel layout on SFP cooling with MAAP5.04.

Journal Articles

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

JAEA Reports

Research on demand of HTGR for investigation of introduction scenario and investigation on heat balance of HTGR

Fukaya, Yuji; Kasahara, Seiji; Mizuta, Naoki; Inaba, Yoshitomo; Shibata, Taiju; Nishihara, Tetsuo

JAEA-Research 2018-004, 38 Pages, 2018/06

JAEA-Research-2018-004.pdf:1.81MB

The demand of HTGR to investigate its introduction scenario and heat balance of HTGR have been researched. First, previous studies of HTGR demand were researched. Next, heat balance of GTHTR300, a commercial scale HTGR design, and its characteristics were researched. By using this information, installation number of HTGR to suit for demand in Japan are evaluated. In addition, heat balance evaluation code was developed in this study.

Journal Articles

Proposals of new basic concepts on safety and radioactive waste and of new high temperature gas-cooled reactor based on these basic concepts

Ogawa, Masuro

Nuclear Engineering and Design, 308, p.133 - 141, 2016/11

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

A new basic concept on safety; Not causing any serious catastrophe by any means and a new basic concept on radioactive waste; Not returning any waste that possibly affects the environment are proposed in the present study, aiming at nuclear power plants which everybody can accept, in consideration of the serious catastrophe that happened at Fukushima in 2011. In the present study, physical phenomena are used to continue confining, rather than confine. To continue confining is meant to apply natural correction to fulfill inherent safety function. Fission products must be detoxified to realize the new basic concept on radioactive waste, aiming at the final processing and disposal of radioactive wastes as same as that in the other wastes such as PCB. The New HTGR is proposed based on the new basic concepts. It is indicated that the New HTGR can response to social requirements for safety and environmental conservability against radioactive wastes, industrial requirements for economy, uranium resource sustainability and so on, and national requirements for non-proliferation and environmental protection against carbon dioxide.

Journal Articles

Analytical studies on fuel element failure propagation due to adventitious fuel pin failure in small to large size sodium-cooled fast reactors

Fukano, Yoshitaka

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. The four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified in the past study. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation were modeled into a safety assessment code which was applicable to arbitrary SFRs. Safety analyses on FEFPA of Japanese experimental fast reactor (JOYO), Japanese prototype fast breeder reactor (Monju), Japanese prototype fast breeder reactor with upgraded reactor core (Upgraded Monju) and Japan sodium-cooled fast reactor (JSFR) were performed using this methodology. Although analytical results were different owing to the different core designs in four SFRs, it was clarified in this study that FEFPA was highly unlikely in these SFRs. These results also suggest future possibility of long-term run-beyond-cladding-breach operation which would enhance the economic efficiency in SFRs.

JAEA Reports

The Second periodic safety review report of Tokai Reprocessing Plant

Shirai, Nobutoshi; Miura, Yasushi; Tachibana, Ikuya; Omori, Satoru; Wake, Junichi; Fukuda, Kazuhito; Nakano, Takafumi; Nagasato, Yoshihiko

JAEA-Technology 2016-007, 951 Pages, 2016/07

JAEA-Technology-2016-007-01.pdf:11.93MB
JAEA-Technology-2016-007-02.pdf:4.7MB

The periodic safety review of TRP is to confirm the safety activities and get effective additional measures the facility safety and its reliability. We implemented 4 items; for (1) evaluation of safety activity implementation, we confirmed we are adequately expanding its safety activities by the necessary documents and schemes. For (2) evaluation of status of safety activities reflecting the latest technical knowledges, we confirmed we reflect latest knowledges for improvement of safety and reliability. For (3) technical evaluation about aging degradation, we can keep the safety of the facilities important to safety and the sea discharge line, under assumption of the present maintenance, because of "focuses for aging degradation". For (4) planning measures about a 10-years-plan that the operator shall implement to keep the facility condition, by the technical evaluation, we found no additional safety plans into maintenance strategies.

Journal Articles

Design approach for mitigation of air ingress in high temperature gas-cooled reactor

Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 8 Pages, 2016/06

This paper intends to propose a practical solution to protect HTGR from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR to study the effectiveness of the proposed design concept using simplified lumped element models. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.

Journal Articles

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*

Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12

 Times Cited Count:6 Percentile:64.92(Physics, Fluids & Plasmas)

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.

Journal Articles

Hydrophobic platinum honeycomb catalyst to be used for tritium oxidation reactors

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*; Noguchi, Hiroshi*; Edao, Yuki; Taniuchi, Junichi*

Fusion Science and Technology, 68(3), p.596 - 600, 2015/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

We have newly developed the hydrophobic platinum honeycomb catalysts applicable to tritium oxidation reactor since the honeycomb-shape catalyst can decrease the pressure drop. Two types of hydrophobic honeycomb catalyst have been test-manufactured. One is the hydrophobic platinum catalyst on a metal honeycomb. The other is the hydrophobic platinum catalyst on a ceramic honeycomb made of silicon carbide. The fine platinum particles around a few nanometers significantly improve the catalytic activity for the oxidation tritium at a tracer concentration. The hydrogen concentration in the gaseous feed slightly affects the overall reaction rate constant for hydrogen oxidation. Due to the competitive adsorption of hydrogen and water molecules on platinum surface, the overall reaction rate constant has the bottom value. The hydrogen concentration for the bottom value is 100 ppm under the dry feed gas. We have experimentally confirmed the activity of these honeycomb catalysts is as good as that of pellet-shape hydrophobic catalyst. The results support the hydrophobic honeycomb catalysts are applicable to tritium oxidation reactor.

Journal Articles

Accumulation of gadolinium isotopes in used nuclear fuel

Suyama, Kenya; Kashima, Takao

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.273 - 282, 2015/09

In the technical development of the criticality safety control of the fuel debris of Fukushima accident in Japan, there have been a discussion on a possibility of adopting BUC with FP. The Expert Group on Burnup Credit Criticality Safety (EGBUC) under the Working Party on Nuclear Criticality Safety (WPNCS) in OECD/NEA Nuclear Science Committee had carried out an international burnup calculation benchmark "Phase-IIIB" and "Phase-IIIC" for BWR fuel assemblies. In these benchmarks the difference of the calculation results of $$^{155}$$Gd among the participants obtained keen interests because it showed rather larger difference among the participants. Authors has been carried out additional analyses on the accumulation of the gadolinium isotopes in the used nuclear fuel during the burnup. Without cooling time, the assembly-averaged amount of $$^{155}$$Gd against the burnup value depends on the burnout property of gadolinium in the burnable poison rods. However, after few year cooling time, $$^{155}$$Gd increase drastically by the decay of $$^{155}$$Eu. In this case, the amount of gadolinium isotopes in the burnable poison rods has less importance. It means that the adopted parameters and data concerning the $$^{155}$$Eu generation have much more importance than the burnup treatment of the burnable poison rods for better prediction of $$^{155}$$Gd.

Journal Articles

Development of hydrophobic platinum catalyst for the effective collection of tritium in fusion plants

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*

Isotope News, (736), p.12 - 17, 2015/08

We have successfully developed a new hydrophobic platinum catalyst for collecting tritium at nuclear fusion reactors. Catalysts used to collect tritium are called hydrophobic precious metal catalysts. In Japan, hydrophobic precious metal catalysts manufactured from polymers have been used for heavy water refinement.However, this catalyst has issues related to embrittlement to radiation and thermal stability. These technological issues needed to be solved to allow for its application to nuclear fusion reactors requiring further enrichment from highly-concentrated tritiated water. We developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The catalyst created with this method was also confirmed to have achieved the world's highest exchange efficiency, equivalent to 1.3 times the previously most powerful efficiency. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water.

Journal Articles

New reactor cavity cooling system using novel shape for HTGRs and VHTRs

Takamatsu, Kuniyoshi; Hu, R.*

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12

continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 ($$^{circ}$$C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.

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