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C level structure by 
-ray spectroscopyHosomi, Kenji; Ma, Y.*; Ajimura, Shuhei*; Aoki, Kanae*; Dairaku, Seishi*; Fu, Y.*; Fujioka, Hiroyuki*; Futatsukawa, Kenta*; Imoto, Wataru*; Kakiguchi, Yutaka*; et al.
Progress of Theoretical and Experimental Physics (Internet), 2015(8), p.081D01_1 - 081D01_8, 2015/08
Times Cited Count：14 Percentile：66.24(Physics, Multidisciplinary)Level structure of the C hypernucleus was precisely determined by means of -ray spectroscopy. We identified four -ray transitions via the 
C
reaction using a germanium detector array, Hyperball2. The spacing of the ground-state doublet 
was measured to be 
(stat)
(syst)keV from the direct 
transition. Excitation energies of the 
and 
states were measured to be 
, keV and 
, keV, respectively. The obtained level energies provide definitive references for the reaction spectroscopy of 
hypernuclei.
Yamazaki, Dai; Soyama, Kazuhiko; Ebisawa, Toru*; Tamura, Itaru; Tasaki, Seiji*
Proceedings of ICANS-XVI, Volume 1, p.407 - 415, 2003/06
Drabkin energy filters extract neutrons of required wavelength using spatial neutron spin resonance. If they are applied to pulsed neutrons, they could sharpen pulse width and cut tail neutrons without reducing band-width or peak-intensity. Shaping of J-PARC coupled-moderator pulse by a Drabkin energy filter was simulated. In the simulation, discrete field gradient to Bx is introduced and total field B is varied in accordance with the dominant wavelength at each moment at the filter position. The results suggest that about 90 % or more of tail neutrons from coupled moderator are cut and a fliter with fields of 200 periods could reduce pulse width up to that of decoupled moderator. Effects of fluctuations of width of each half period were studied and it was found that fluctuation within 
1 % are tolerable for a flipper with field of 100 or 200 periods.
Yamashita, Kiyonobu; Takeuchi, Mitsuo; Fujimoto, Nozomu; Fujisaki, Shingo; Nakano, Masaaki*; Nojiri, Naoki; Tamura, Seiji*
Nihon Genshiryoku Gakkai-Shi, 41(1), p.35 - 38, 1999/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
; Tamura, Seiji*;
PNC TN941 84-161, 156 Pages, 1984/12
In the experimental fast reactor JOYO, following the MARK-1 core operation for 4 years, the whole core components were newly replaced in 1982 to establish an irradiation bed which is called MARK-2 core. The core has 6 control rods which are required both power control and scram functions. In March 1983, performance tests of the MARK-2 core were completed, and the rated power (100 MW thermal) operations have been continued since August 1983. During the MARK-2 power operation, it was found that the value of neutron flux fluctuation was larger than the expected from design condition. In order to make clear the phenomenon, reactor noise analysis was conducted. Neutron flux, reactivity signal, primary flow, load fluctuation of each control rod, acoustic signal of control rod driving mechanism were measured and analyzed to obtain their rms values, auto power spectral densities and coherences. This report deals with the estimation of control rods vibration modes as well as the identification of neutron flux noise source in the MARK-2 core. Main results are as follows. (1)Peak-to-peak value of neutron flux was about 4% of power level at the most, larger than the expected from design value. However, it was not large enough to reach alarm level (103 MW). (2)Main cause of neutron flux fluctuation is flow induced vibration of control rods. (3)Dominant driving force of control rods vibration is the pressure disturbance due to the large flow difference between control rod channel and surrounding fuel channels. It is different from the mechanism considered before. (4)In order to reduce neutron flux fluctuation, the reformation of vibration restriction mechanism is effective method.
; ; Tamura, Seiji;
Proceedings of 3rd International Conference on Liquid Metal Engineering and Technology in Emergy Production, 0 Pages, 1984/00
None
; ; Tamura, Seiji*
PNC TN941 83-54, 53 Pages, 1983/04
In order to develop one of the on-line plant anomaly monitoring system, applying reactor noise analysis technique, a modeling of transfer function for the indisial response and spacial dependent neutron fluctuation in the core was studied using WS (weight sequence) model technique. It was proven that the reactor kinetics and space dependent neutron fluctuation Were well identified by means of the WS model technique and also the WS model may detect the anomaly experienced in JOYO. The WS model method has advantages, if compare to the ordinary frequency analysis technique. The access time is shorter and the volume of program is smaller. The results suggested that the WS model has a potential to be utilized in the on-line anomaly monitoring system.
Muramatsu, Toshiharu; ; Tamura, Seiji*; ;
PNC TN941 83-06, 70 Pages, 1983/02
Flowrate seasurement of subassemblies of the reactor core was conducted under zero power and 250
C conditions for the final core configuration of MK-I operation (79 fuel core S/A). The purpose of the mesurement is to comfime the change of flow distribution in reactor core due to core volume enlargement, As consequences, (1)Flowlate measurement of Subassemblies under the high flowrate condition of the main loops brought up followings: (i)Flowrates for core fuel assemblies decreased about 6.3% than the 70 fuel S/A core measurement. (ii)Flowrates of inner blanket fuel assemblies increased about 4.0% than the 70 fuel S/A core. (iii)Flowrates of all outer blanket fuel assemblies decreased about 2.5% than the 70 fuel S/A core. (2)All subassemblies flowrate under the low flowrate condition of the main loops was measured lower than the estimated values. (3)The flowrate indication at the center channed increased about 3.0% after 115 hours after steeped in sodium. This change was compasated to evaluate the subassembly flowrate as flowmeter drift.
Muramatsu, Toshiharu; Tamura, Seiji*
PNC TN941 82-269, 85 Pages, 1982/10
The reactor operation support system has been developed to increase safety and efficiency of nuclear plant operation. An artificial intelligence program system (DIAMIN SYSTEM) is under way for the purpose. The organization of the DIAMIN SYSTEM may be devided into following 4 subsystems. (1)Natural Language Processing (2)Plant Control Processing (3)Plant Simulation Processing (4)Knowledgement Data Base. It was confirmed that practical use of the DIAMIN SYSTEM, from verification tests at experimental fast reactor "JOYO". The detected abnormal plant condition terms in verification test are as followings, (1)Neutron flux high (at Power regulation) (2)Cooling system all down (at Natural Circulation test) (3)Secondary flow A low (at motor trouble)
Tamura, Seiji*; Hikichi, Takayoshi*; Rindo, Hiroshi
PNC TG033 82-01(3), 13 Pages, 1982/01
Monitoring of fuel failure in the experimental fast reactor JOYO is provided by two different methods, which are cover gas monitoring (FFDCGM) by means of a precpitator, and delayed neutron monitoring (FFDDNM) by means of neutron detectors.
; Shinohara, Yoshikuni; ; ; ; Nishihara, Hideaki*; ; ; Tamura, Seiji*;
Nihon Genshiryoku Gakkai-Shi, 24(3), p.188 - 198, 1982/00
Times Cited Count：0 Percentile：0.02(Nuclear Science & Technology)no abstracts in English
Muramatsu, Toshiharu; ; ; ; Tamura, Seiji*
PNC TN941 81-75, 53 Pages, 1981/04
Eddy current type flow meter has been installed above center of the core in "JOYO", to measure outlet sodium velocity of center channel. This information is applied, together with outlet sodium temperatures subassemblies, in monitoring the malfunction of the center subassembly. This paper presents a study of flow measurements from the initial 50 MWt power ascention to the third cycle operation of 75 MWt power. The results of the tests are as followings. (1)If an unbalance of the secondary coils is adjusted to zero, lineality of velocity vs. primary main flow and temperature dependency above the temperature of 350C are obtained within error of 6%. (2)The unbalance signal must be adjusted prior to every cycle operation, because of its drift during the power operation. (3)On constant power operation, the flow signal drifts 2% of full scale. (4)It was confirmed that added signal detected temperature change with almost zero delay time. (5)The flow velocity obtained by cross-correlation method was approximately 20% less than the flow signal reading.
; ; Tamura, Seiji*
PNC TN941 81-03, 82 Pages, 1981/01
Reactor noise tests were carried out in 75 MW Power Up Test. The main purpose of this analysis is to investigate the existance of the peak in APSD of neutron flux signal and the low correlations of neutron flux signals from different position power range flux monitors in the frequency range 5 
10
7 10
Hz. The anormalous reactivity phenomenon, observed at the first 75MW power-up, accompanied the power coefficient change of 50 MW power level, thereby it was predicted that reactor noise characteristics have aloso changed. These changes were investigated, and the phenomena of neutron flux signal were examined. The primary points of the results are as follows. (1)By the quantitative analysis and the investigations of reactor noise characteristics changes, it became almost clear that the phenomena of the low correlation between different locations is caused by the spacial dependency effect of neutron flux fluctuations. (2)Coherence functions of neutron flux signals and transfer functions of reactor inlet temperature - neutron flux have changed with the power coefficient changes, thereby it became almost clear that the intensity of spacial dependency effect have also changed with the anormalous reactivity phenomenon. (3)It was estimated that the peak in APSD of neutron flux signal is the inherent phenomenon of the reactor and caused by the movements of core composition elements. Based on these estimation, it was assumed that the peak is a resonant phenomenon between thermal bowing of the subassembly wrapper tube and the neutron flux, which accompanies the movements of core subassemblies.
; ; Tamura, Seiji*; Doi, Motoo*; ; Yamamoto, Hisashi*
PNC TN941 80-168, 53 Pages, 1980/12
Heat balance measurements and calculations were performed for the JOYO experimental fast reactor. Some pertinent results of this tests are presented below. (1)The heat removal rates calcalated using the air flowrates and 
Ts from the DHXs differed from heat removal rates obtained using measured primary sodium flowrates and Ts. The heat removal rates determined from the DHX data are about 11% larger and about 4% larger than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. The air flowrates and outlet temperatures from the DHX are measured on an 88 grids in the outlet area of the air cooler, using Pitot tubes and thermocouples. (2)The heat removal rates obtained using measured secondary sodium flowrates and DHX Ts are almost same as and about 8% smaller than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. (3)The core heat generation rate (including blanket effects, etc.) calculated using the measured individual subassembly outlet temperatures, and subassembly flowrates, is about 7% larger than the value calculated using the primary sodoum flowrate and the reactor T. The reason for this discrepancy is conjectured to be due to the fact that the subassembly average outlet temperature is lower than the measured value, since the thermocouple is located in the center of the subassembly outlet channel. In both the A and B Loops of JOYO, the error between the heat removal rates from the measured DHX and the primary or secondary sodium flowrate and temperature data, and in the B Loop between the heat removal rates obtained from the primary sodium data and the secondary flowrate and DHX T, are greater than the error band of the instrumentation. The eauses of this unbalance in the heat removal are being investigated. THis report is based on the meetings held in February and March 1980 in the Reactor Technology Section ...
Yamamoto, Hisashi*; Sekiguchi, Yoshiyuki*; Hirose, Tadashi*; Sanda, Toshio*; Tamura, Seiji*; ;
PNC TN941 80-179, 402 Pages, 1980/10
On April 24th, 1977, the initial criticality of JOYO was achieved and on July 5th, 1978, the reactor output reached rated power of 50 MW for the first time. The 75MW power ascension test was started in July, 1979, followed by two cycles of rated power operations, and the 100 hour nominal power continuous operation was completed in February, 1980. Through the tests for the core, plant it self, radiation shield and plant monitoring, the results proved satisfactory operation characteristics at 75MW. This report presents the summary of an the results obtained in the Test of MK-I core.
; ; Tamura, Seiji*
PNC TN941 80-146, 60 Pages, 1980/08
In the experimental fast reactor "JOYO", in-core acoustic monitoring has been continued, in order to detect in-core abnormal condition like sodium boiling at an early stage. Three in-core acoustic detectors are situated in the core addressed [5A2], [5C2] and [5F2]. The following facts were derived from the measurements and the monitoring by these detectors through the first 75 MW duty cycle operation. (1)The background noise at several ten kHz range, which is suitable range for sodium boiling detection, is mainly composed of electromagnetic noise from the primary sodium flow control system. The power spectral density of the noise has a broad peak at about 22 kHz. The signal level of the noise are independent of reactor power and primary sodium flow (primary sodium pump in operation). (2)Since this in-core acoustic detection system has poor SN ratio, the analogy of the result of the local sodium boiling test conducted at off-site may resulted in difficulty for detection of local sodium boiling by this system. (3)Insulation resistance of the detectors became lower as sodium temperature became higher. Over about 330C, however, the resistance of [5A2] in-core acoustic detector suddenly increased more than 300 times of its normal value. (4)Abnormal core condition has not been found until now according to the in-core acoustic monitoring system.
; Tamura, Seiji*; ; ; Muramatsu, Toshiharu
PNC TN941 80-145, 60 Pages, 1980/08
Anomaly reactivity monitoring program which is one of functions of Joyo data reduction system was tested under various plant conditions during power ascension testing period. The program is based on a reactivity balance of measured reactivity and known reactivity, and is to detect unknown (or anomaly) reactivity. As consequences of the testings, following results were obtained. For each change of power level and coolant temperature, reactivity change of more than 10cent was induced, the program, however, indicated no trace of any anomaly reactivity. While for change of coolant flowrate of 20%, approximatly 3cent error reactivity was detected due to, probably, mismatch of thermal constants in the program. Through these testings, it was cleared that the modeling of the program is well simulating the reactor kinetics within 2cent of error except for the change of flowrate. In normal operation, since flowrate is held constant, the error due to flowrate is insignificant.
Tamura, Seiji*; Muramatsu, Toshiharu; Sanda, Toshio*;
PNC TN941 80-03, 27 Pages, 1980/01
Thermocouple temperature sensars are installed above center region of the core, to monitor outlet coolant temperatures of 115 subassemblies. This paper presents a study of temperature measurements obtained during the initial 50MW power ascension and the first 2 cycle operations of the reactor. According to the study of data obtained under various plant conditions, (1)A standard deviation of temperature data under 250C isothermal was 0.59C and no remarkable change during the period was seen. (2)Temperature distribution at 50MW operation was compaired with calculated temperature distribution, and it was found that the former was higher than the latter as much as 6C and 20C in core region and blanket region respectively. It is estimated that the difference is due to flowrate change with the reactor power. (3)Through the observation of temperature distribution under unequal temperature in A and B loop inlets, it was found that coolant from two loops is not well mixed in the lower plenum of the vessel. (4)On the measured data under 370C and 250C reactor inlet temperature operations, same tendancy of flowrate change was seen. The testing to study flowrate dependency on reactor power and temperature is to be contineued.
; ; ; Tamura, Seiji*; Sanda, Toshio*; Yamamoto, Hisashi*
PNC TN941 80-02, 70 Pages, 1980/01
Neutron flux, control rod (CR) load cell signals, CR acoustic signals and reactivity signal were measured and analyzed to study the effect of CR vibration induced by primary sodium flow during the low power and the power ascension tests of the Experimental Fast Reactor "Joyo". The measured signals were recorded by a multi-channel data recorder under various operating conditions of primary sodium flow, CR position and reactor power. Then the recorded signals were analyzed by a spectrum analyzer to obtain their power spectral density functions (PSD), coherence functions and r.m.s. values. Following characteristics were determined. (1)Impact sound detected on housing of CR drive mechanism became obvious as the sodium flow rate reached 100%. Its frequency was about 2Hz. (2)The auto PSDs of the fluctuation of the measured signals had evident spectra at about 2Hz. Also the coherence functions between them showed more significant at about 2Hz. (3)The r.m.s. value of the fluctuation of reactivity signal decreased as a control rod was withdrawn. The r.m.s. value at 50Mw was about 0.1 cent. Based on these results, the frequency of CR vibration due to the flow was about 2Hz. The r.m.s. value of reactivity fluctuation at 50MW mainly caused by the vibration was about 0.1 cent.
