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Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Hydrogen permeation through heat transfer pipes made of Hastelloy XR during the initial 950$$^{circ}$$C operation of the HTTR

Sakaba, Nariaki; Ohashi, Hirofumi; Takeda, Tetsuaki

Journal of Nuclear Materials, 353(1-2), p.42 - 51, 2006/07

 Times Cited Count:6 Percentile:55.2(Materials Science, Multidisciplinary)

The permeation of hydrogen isotopes through the Hastelloy XR high-temperature alloy adopted for the heat transfer pipes of the intermediate heat exchanger in the HTTR, is one of the concerns in the hydrogen production system, which will be connected to the HTTR in the near future. The hydrogen permeation between the primary and secondary coolant through the Hastelloy XR was evaluated using the actual hydrogen concentration observed during the initial 950$$^{circ}$$C operation of the HTTR. The hydrogen permeability of the Hastelloy XR was estimated conservatively high as follows. The activation energy E$$_{0}$$ and pre-exponential factor F$$_{0}$$ of the permeability of hydrogen were E$$_{0}$$ = 65.8 kJ/mol and F$$_{0}$$ = 7.8$$times$$10$$^{-9}$$m$$^{3}$$(STP)/(m$$ast$$s$$ast$$Pa$$^{0.5}$$), respectively, in the temperature range from 707K to 900K.

Journal Articles

Structural integrity assessments of helium components in the primary cooling system during the safety demonstration test using the HTTR

Sakaba, Nariaki; Tachibana, Yukio; Nakagawa, Shigeaki; Hamamoto, Shimpei

Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4499 - 4511, 2005/08

Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactor candidates. The coolant flow reduction test by running down gas circulators, which is one of the safety demonstration tests, is a simulation test of anticipated transients without scram. During the coolant flow reduction test, temperature of the high-temperature helium components and chemistry in the primary circuit are changed rapidly. This paper describes the structural integrity assessments of helium components, e.g. helium pipes, heat exchangers, during the coolant flow reduction test. From the result of this evaluation, it was found that the helium components were kept their structural integrity during temperature and chemistry transient condition in the coolant flow reduction test from the reactor power at 30%. It was also confirmed by this assessment that the coolant flow reduction test will be able to perform with its enough safety margins from the reactor power at 100%.

JAEA Reports

Evaluation of heat exchange performance for intermediate heat exchanger in HTTR

Tochio, Daisuke; Nakagawa, Shigeaki

JAERI-Tech 2005-040, 39 Pages, 2005/07

JAERI-Tech-2005-040.pdf:1.88MB

In High Temperature Engineering Test Reactor (HTTR) of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary pressurized water cooler and the intermediate heat exchanger. Heat exchangers in main cooling system of HTTR should satisfy two conditions, achievement of reactor coolant outlet temperature 850 $$^{circ}$$C/950 $$^{circ}$$C and removal of reactor generated heat 30 MW. That is, heat exchange performance should be ensured as that in heat exchanger designing. In this report, heat exchange performance for Intermediate heat exchanger (IHX) in main cooling system is evaluated with rise-to-power-up test and in-service operation data. Moreover, the applicability of IHX thermal-hydraulic design method is discussed with comparison of evaluated data with designed value.

JAEA Reports

Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro*; et al.

JAERI-Tech 2005-031, 174 Pages, 2005/06

JAERI-Tech-2005-031.pdf:20.71MB

Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001, and a lot of operational test data on heat exchanges were obtained in these tests.In this report specifications, structures and heat transfer formulae of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Evaluation codes were newly made to evaluate heat transfer characteristics from measured test data. Overall heat-transfer coefficient obtained from the experimental data were compared and evaluated with the prospective value calculated with heat transfer formulae. As a result, heat transfer performance and thermal efficiency of these heat exchangers were confirmed to be appropriate.

Journal Articles

Heat exchanger performance in main cooling system on high temperature test operation at High Temperature Gas-Cooled Reactor 'HTTR'

Tochio, Daisuke; Nakagawa, Shigeaki; Furusawa, Takayuki*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.147 - 155, 2005/06

High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at JAERI achieved the reactor outlet coolant temperature of 950$$^{circ}$$C for the first time in the world at Apr. 19, 2004. To remove of generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value.

JAEA Reports

Management techniques of the JRR-4 heat exchanger

Horiguchi, Hironori; Oyama, Koji; Ishikuro, Yasuhiro; Hirane, Nobuhiko; Ito, Kazuhiro; Kameyama, Iwao

JAERI-Tech 2005-001, 38 Pages, 2005/02

JAERI-Tech-2005-001.pdf:2.79MB

After JRR-4 heat exchanger was renewed in made of stainless steel from carbon steel, it was examined how to manage the heat exchanger. The main subject is the cleaning technology of the heat exchanger. The recovery of old heat exchanger cooling performance has been by only chemical cleaning. Now we use chemical and dry cleaning as a new technique. It helps prevent of corrosions of secondary pipes and reduce of management costs. This report describes the performance management and cleaning technology of the JRR-4 heat exchanger and the management of the JRR-4 coolant.

Journal Articles

Experimental evaluation of tritium permeation through stainless steel tubes of heat exchanger from primary to secondary water in ITER

Nakamura, Hirofumi; Nishi, Masataka

Journal of Nuclear Materials, 329-333(Part1), p.183 - 187, 2004/08

 Times Cited Count:20 Percentile:20.43(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Permeability of hydrogen and deuterium of Hastelloy XR

Takeda, Tetsuaki; Iwatsuki, Jin*; Inagaki, Yoshiyuki

Journal of Nuclear Materials, 326(1), p.47 - 58, 2004/03

 Times Cited Count:17 Percentile:24.57(Materials Science, Multidisciplinary)

Permeation of hydrogen isotope through a high-temperature alloy used as heat exchanger and steam reformer pipes is an important problem in the hydrogen production system connected to be a High-Temperature Engineering Test Reactor (HTTR). An experiment of hydrogen (H$$_{2}$$) and deuterium (D$$_{2}$$) permeation was performed to obtain permeability of H$$_{2}$$ and D$$_{2}$$ of Hastelloy XR, which is adopted as heat transfer pipe of an intermediate heat exchanger of the HTTR. Permeability of H$$_{2}$$ and D$$_{2}$$ of Hastelloy XR were obtained as follows. The activation energy E$$_{0}$$ and pre-exponential factor F$$_{0}$$ of the permeability of H$$_{2}$$ was E$$_{0}$$=67.2$$pm$$1.2 kJ/mol and F$$_{0}$$=(1.0$$pm$$0.2)$$times$$10$$^{-8}$$m$$^{3}$$(STP)/m$$^{-1}$$/s$$^{-1}$$/Pa$$^{-0.5}$$, respectively, in the pipe temperature ranging from 843K-1093K.

Journal Articles

Heat removal performance of auxiliary cooling system for the High Temperature Engineering Test Reactor during scrams

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki*

Annals of Nuclear Energy, 30(7), p.811 - 830, 2003/05

 Times Cited Count:1 Percentile:88.18

no abstracts in English

JAEA Reports

Structural integrity assessment of intermediate heat exchanger in the HTTR based on results of rise-to-power test

Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki

JAERI-Tech 2002-091, 45 Pages, 2002/12

JAERI-Tech-2002-091.pdf:1.77MB

no abstracts in English

Journal Articles

Multi-dimensional thermal-hydraulic analysis for horizontal type PCCS

Arai, Kenji*; Kurita, Tomohisa*; Nakamaru, Mikihide*; Fujiki, Yasunobu*; Nakamura, Hideo; Kondo, Masaya; Obata, Hiroyuki*; Shimada, Rumi*; Yamaguchi, Ken*

Proceedings of 10th International Conference on Nuclear Engineering (ICONE 10) (CD-ROM), 7 Pages, 2002/00

no abstracts in English

Journal Articles

Experimental investigation of thermal-hydraulic performance of PCCS with horizontal tube heat exchangers; Single U-tube test

Nakamura, Hideo; Anoda, Yoshinari; Tabata, Hiroaki*; Obata, Hiroyuki*; Arai, Kenji*; Kurita, Tomohisa*

JAERI-Conf 2000-015, p.177 - 184, 2000/11

no abstracts in English

Journal Articles

Single U-tube Testing and RELAP5 code analysis of PCCS with horizontal heat exchanger

Nakamura, Hideo; Kondo, Masaya; Asaka, Hideaki; Anoda, Yoshinari; Tabata, Hiroaki*; Obata, Hiroyuki*

Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.336 - 343, 2000/00

no abstracts in English

Journal Articles

Feasibility study on the applicability of a diffusion-welded compact intermediate heat exchanger to next-generation high temperature gas-cooled reactor

Takeda, Takeshi; Kunitomi, Kazuhiko; *; *

Nucl. Eng. Des., 168, p.11 - 21, 1997/00

 Times Cited Count:37 Percentile:7.5(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Application of neutron radiography to visualization of cryogenic fluid boiling two-phase flows

Takenaka, Nobuyuki*; Asano, Hitoshi*; Fujii, Terushige*; Ushiro, Toshihiko*; Iwatani, Junji*; Murata, Yutaka*; Mochiki, Koichi*; Taguchi, Akira*; Matsubayashi, Masahito; Tsuruno, Akira

Nuclear Instruments and Methods in Physics Research A, 377(1), p.174 - 176, 1996/07

 Times Cited Count:1 Percentile:76.03(Instruments & Instrumentation)

no abstracts in English

Journal Articles

Measurement of dynamic behavior of void fraction in tube-banks of a simulated fluidized-bed by neutron radiography

*; *; *; *; Matsubayashi, Masahito; Tsuruno, Akira

Fifth World Conf. on Neutron Radiography, 0, p.610 - 616, 1996/00

no abstracts in English

Journal Articles

High accuracy heat transfer correlation on shell side of heat transfer tubes for pressurized water cooler in high temperature use

Kunitomi, Kazuhiko; Takeda, Takeshi; *; ; *; *; *

Nippon Genshiryoku Gakkai-Shi, 38(8), p.665 - 672, 1996/00

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

no abstracts in English

38 (Records 1-20 displayed on this page)