Yang, Z. H.*; Kubota, Yuki*; Corsi, A.*; Yoshida, Kazuki; Sun, X.-X.*; Li, J. G.*; Kimura, Masaaki*; Michel, N.*; Ogata, Kazuyuki*; Yuan, C. X.*; et al.
Physical Review Letters, 126(8), p.082501_1 - 082501_8, 2021/02
A quasifree (,) experiment was performed to study the structure of the Borromean nucleus B, which had long been considered to have a neutron halo. By analyzing the momentum distributions and exclusive cross sections, we obtained the spectroscopic factors for and orbitals, and a surprisingly small percentage of 9(2)% was determined for . Our finding of such a small component and the halo features reported in prior experiments can be explained by the deformed relativistic Hartree-Bogoliubov theory in continuum, revealing a definite but not dominant neutron halo in B. The present work gives the smallest - or -orbital component among known nuclei exhibiting halo features and implies that the dominant occupation of or orbitals is not a prerequisite for the occurrence of a neutron halo.
Momiyama, Satoru*; Doornenbal, P.*; Scheit, H.*; Takeuchi, Satoshi*; Niikura, Megumi*; Aoi, Nori*; Li, K.*; Matsushita, Masafumi*; Steppenbeck, D.*; Wang, H.*; et al.
Physical Review C, 96(3), p.034328_1 - 034328_8, 2017/09
no abstracts in English
Lee, J.*; Liu, H.*; Doornenbal, P.*; Kimura, Masaaki*; Minomo, Kosho*; Ogata, Kazuyuki*; Utsuno, Yutaka; Aoi, Nori*; Li, K.*; Matsushita, Masafumi*; et al.
Progress of Theoretical and Experimental Physics (Internet), 2016(8), p.083D01_1 - 083D01_7, 2016/08
no abstracts in English
Nemoto, Takahiro; Kaneshiro, Noriyuki*; Sekita, Kenji; Furusawa, Takayuki; Kuroha, Misao; Kawakami, Satoru; Kondo, Masaaki
JAEA-Technology 2015-006, 36 Pages, 2015/03
The High-Temperature engineering Test Reactor (HTTR) has been developed for establishing and upgrading the technical basis of HTGR.HTTR facilities have their structures, systems and a lot of components including reciprocating gas compressors, commonly used to extract and/or discharge reactor coolant helium gas contained in primary/secondary coolant systems. From the fact of the operational experiences of these compressors, seal-oil leakage has been frequently observed, although rod-seal mechanisms with complicated structures are equipped and improved for preventing coolant helium gas. So, we tried to change the rod-seal materials which might be a primary reason of frequent seal-oil leakage, that resulted in decreasing a mass and frequently of seal-oil leakage. It is confirmed that it is important to select adequate materials of rod seal for sliding speed of the piston of the compressor to prevent seal-oil leakage. Additionally, the procedure to estimate seal-oil leakage for each compressor is discussed. This report describes the results of investigation for improvement on seal-oil leak tightness of the compressors in HTTR facilities.
Shimizu, Atsushi; Furusawa, Takayuki; Homma, Fumitaka; Inoi, Hiroyuki; Umeda, Masayuki; Kondo, Masaaki; Isozaki, Minoru; Fujimoto, Nozomu; Iyoku, Tatsuo
Journal of Nuclear Science and Technology, 51(11-12), p.1444 - 1451, 2014/11
JAEA has kept up a data-base system of operation and maintenance experiences of the HTTR. The objective of this system is to share the information obtained operation and maintenance experiences and to make use of lessons learned and knowledge into a design, construction and operation managements of the future HTGR. More than one thousand records have been registered into the system between 1997 and 2012. This paper describes the status of the data-base system, and provides suggestions for improvement from four experiences: (1) performance degradation of helium compressors; (2) malfunction of reserved shutdown system in reactivity control system; (3) maintenance experiences of emergency gas turbine generators; and (4) experiences of the Great East Japan Earthquake. These experiences are extracted from the system as important lessons learned to be expected to apply for design, construction and operation managements of future HTGR.
Shimazaki, Yosuke; Homma, Fumitaka; Sawahata, Hiroaki; Furusawa, Takayuki; Kondo, Masaaki
Journal of Nuclear Science and Technology, 51(11-12), p.1413 - 1426, 2014/11
Kondo, Masato*; Tsubouchi, Masaaki
Optics Express (Internet), 22(12), p.14135 - 14147, 2014/06
We investigated liquid-sheet jets with controllable thickness for application to terahertz (THz) spectroscopy. Slit-type and colliding-jet nozzles were used to generate optically flat liquid jets. The thickness of the liquid sheet was determined precisely by spectral interference and THz time-domain-spectroscopy methods. By adjusting the collision angle of the colliding-jet nozzle, we could control the thickness of the liquid sheet from 50 to 120 m.
Iwamoto, Yukiharu*; Kondo, Manabu*; Minamiura, Hirotaka*; Tanaka, Masaaki; Yamano, Hidemasa
Journal of Fluid Science and Technology (Internet), 7(3), p.315 - 328, 2012/08
LDV measurements in a 90 degree elbow of which the curvature radius coincides with its inner diameter were examined for the cases of inflow from a long pipe, short pipe and swirl generator. Ensemble averaged flow distribution at the Reynolds number of 320000 based on the inner pipe diameter and bulk velocity shows that shortening the upstream pipe length to 4.9D from 10D induces the flow separation downstream of the elbow.
Iwamoto, Yukiharu*; Kondo, Manabu*; Ogawa, Shota*; Tanaka, Masaaki; Yamano, Hidemasa
Nihon Kikai Gakkai Rombunshu, B, 78(792), p.1383 - 1387, 2012/08
LDV measurements in a 90 degrees elbow which curvature radius coincides with the diameter have been conducted. This paper especially focuses on a result of the deflected inflow, comparing with a result of the short pipe. The result shows that the deflected inflow reinforced a convex velocity distribution occurring near the curvature inside in the downstream region, concluding that the deflected inflow promotes the secondary flow of Prandtl's first kind in the elbow. Its Strouhal number increases to 0.6 from 0.5, compared with the short pipe case. Results of frequency analyses are also shown for other cases that we have been examined. Dominant Strouhal numbers in most of the cases become 0.5, except for 0.6 in cases of the inflow from the long pipe and deflector. This frequency shift might be related with the boundary layer size and the local flow velocity, since the corresponding fluctuation is caused by vortex shedding from the boundary layer at the elbow inside.
Iwamoto, Yukiharu*; Kondo, Manabu*; Yasuda, Kazunori*; Sogo, Motosuke*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Tochio, Daisuke; Hamamoto, Shimpei; Inoi, Hiroyuki; Shimazaki, Yosuke; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Emori, Koichi; et al.
JAEA-Technology 2010-038, 57 Pages, 2010/12
In HTTR, in-service operation is conducted through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. To make practical use HTGR system, it must be demonstrated to supply stable heat to heat utilization system for long-term. In HTTR, high-temperature/parallel-loaded long-term operation had been performed from January 2010. As the result, it was demonstrated to supply stable heat to heat utilization system for 50 days with HTTR, moreover, various long-term operation data were gained. This paper reports the characteristics of the high-temperature long-term operation for HTTR obtained from the operation.
Tochio, Daisuke; Nojiri, Naoki; Hamamoto, Shimpei; Inoi, Hiroyuki; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Fujimoto, Nozomu
JAEA-Technology 2009-005, 47 Pages, 2009/05
HTTR is now conducted in-service operation through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. In order to demonstrate to supply stable heat to heat utilization system for long-term, HTTR was conducted rated/parallel-loaded 30-days operation. This paper reports the characteristics of long-term operation for HTTR.
Kondo, Masaaki; Kimishima, Satoru*; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi
JAEA-Technology 2008-062, 46 Pages, 2008/10
The reactor containment of HTTR is tested to confirm leak-tight integrity of itself. "Type A test" has been conducted in accordance with the standard testing method in JEAC4203 since the preoperational verification of the containment was made. Type A tests are identified as basic one for measuring containment leakage rate, it costs much, however. Therefore, the test program for HTTR was revised to adopt an efficient and economical alternatives including "Type B and Type C tests". In JEAC4203-2004, following requirements are specified for adopting alternatives: upward trend of leakage rate by Type A test due to aging should not be recognized; criterion of combined leakage rate with Type B and Type C tests should be established; the criteria for Type A test and combined leakage rate test should be satisfied; correlation between the leakage rates by Type A test and combined leakage rate test should be recognized. Considering the performances of the tests, the policies of corresponding to the requirements were developed, which were accepted by the regulatory agency. This report presents an outline of the tests, identifies issues on the conventional test and summarizes the policies of corresponding to the requirements and of implementing the tests based on the revised program.
Sekita, Kenji; Kuroha, Misao; Emori, Koichi; Kondo, Masaaki; Ouchi, Hiroshi; Shinozaki, Masayuki
JAEA-Technology 2008-002, 49 Pages, 2008/03
Graphite structures are used as one of the HTTR core internal structures. Graphite structures have high heat resistant property but its mechanical strength degrades easily by oxidization. To prevent the oxidization of graphite structures, impurity concentrations in the coolant of helium are controlled strictly. The helium sampling system is installed to measure the impurity concentrations in the helium. At gas compressor in helium sampling system, seal-oil leak at rod seal mechanism was occurred. The causes are degradation of seal material and contaminant abrasion powder of grand-packing. As these countermeasure, material of seal material was changed and contaminant was decreased. As the result long term operation is enabled. Moreover, reliable data can be obtained and efficient impurity control is enabled due to renewal of data acquisition control computer of gas chromatograph mass spectrometer and improvement of liquid nitrogen trap.
Yamano, Hidemasa; Kondo, Teppei*; Sugaya, Masaaki*; Kamiyama, Kenji
JAEA-Research 2007-054, 89 Pages, 2007/06
The SIMMER-III code and its three-dimensional code SIMMER-IV have been developed to evaluate the consequence of core disruptive accidents in liquid-metal cooled fast reactors. The present study has extended the number of nodes in a structure model of SIMMER code from a conventional fixed two-node model to a multi-node model. The number of nodes can be specified automatically or manually. The model also treats the effect of nuclear heating and axial heat transfer models. The model was validated by basic verification calculations. The model alleviates the limitation of conventional model significantly as well as improves the reliability and accuracy of fast reactor safety analyses. This study is expected to contribute to the design study of recriticality-free concept.
Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Sakaba, Nariaki; Kimishima, Satoru; Kuroha, Misao; Noji, Kiyoshi; Aono, Tetsuya; Hayakawa, Masato
JAEA-Testing 2006-002, 55 Pages, 2006/07
The leakage rate test for the reactor containment vessel of HTTR is conducted in accordance with the absolute pressure method provided in Japan Electric Association Code(JEAC4203). Although leakage test of a reactor containment vessel is, in general, performed in condition of reactor coolant pressure boundary to be opened in order to simulate an accident, the peculiar test method to HTTR which use the helium gas as reactor coolant has been established, in which the pressure boundary is closed to avoid the release of fission products into the environment of the reactor containment vessel. The system for measuring and calculating the data for evaluating the leakage rate for containment vessel of HTTR was developed followed by any modifications. Recently, the system has been improved for more accurate and reliable one with any useful functions including real time monitoring any conditions related to the test. In addition, the configuration of containment vessel boundary for the test and the calibration method for the detectors for measuring temperature in containment vessel have been modified by reflecting the revision of the Code mentioned above. This report describes the method, system configuration, and procedures for the leakage rate test for reactor containment vessel of HTTR.
Aono, Tetsuya; Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Kuroha, Misao; Ouchi, Hiroshi
JAEA-Testing 2006-004, 39 Pages, 2006/06
The High Temperature Engineering Test Reactor (HTTR) has an emergency air purification system(EAPS). The system keeps the service area negative pressure condition and exhausts the filtered air to prevent fission products release to environment in accident condition. The EAPS is one of the engineered safety features which is started automatically when radioactivity in the service area increase or might increase. The performance of the EAPS should satisfy the analytical condition for public dose evaluation in the severest accidents of the HTTR. The performance should be confirmed by function tests. The function tests are divided into many tests corresponding to each assumed phenomenon. The confirmation of the performance of the system was carried out effectively by the tests. Moreover, the stable operation of the system can be achieved by improvements of the method of leak tight tests of exhaust filter unit. The report describes the outline of EAPS system, maintenance works and improvement of the system.
Sakaba, Nariaki; Iigaki, Kazuhiko; Kondo, Masaaki; Emori, Koichi
Nuclear Engineering and Design, 233(1-3), p.135 - 145, 2004/10
The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radio-activity and will maintain a correct pressure in the service area. The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept the design pressure well below its allowable limitation by the emergency air purification system which filter efficiency of particle removal and iodine removal were well over the limited values. The obtained data demonstrates that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.
Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kamiyama, Kenji; Kondo, Satoru; Morita, Koji*; Fischer, E. A.; Brear, D. J.; Shirakawa, Noriyuki*; Cao, X.; et al.
JNC TN9400 2003-071, 340 Pages, 2003/08
An advanced safety analysis computer code, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III Version 3.A are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 3.A, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliability of LMFR safety analyses.
Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kondo, Satoru; Morita, Koji*; Sugaya, Masaaki*; Mizuno, Masahiro*; Hosono, Seigo*; Kondo, Teppei*
JNC TN9400 2003-070, 333 Pages, 2003/08
An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 2) coupled with the three-dimensional neutronics model. With the completion of the SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV.This report describes SIMMER-IV Version 2.A, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users.