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Journal Articles

Application of the modified neutron source multiplication method to the prototype FBR Monju

Truchet, G.*; Van Rooijen, W. F. G.*; Shimazu, Yoichiro*; Yamaguchi, Katsuhisa

Annals of Nuclear Energy, 51, p.94 - 106, 2013/01

 Times Cited Count:7 Percentile:47.83(Nuclear Science & Technology)

The Modified Neutron Source Method (MNSM) is applied to the prototype FBR Monju. Among Monju's particularities that have a big influence on the MNSM factors are: the presence of two californium sources near the core and the position of the detector, which is located far from the core outside of the reactor vessel. In order to evaluate the detector count rate, a propagation calculation has been conducted from the reactor vessel to the external detector. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). Results indicate a good agreement between the MNSM reactivity and the reactivity measured with other methods.

Journal Articles

Proton resonance elastic scattering in inverse kinematics on the medium heavy nucleus $$^{68}$$Zn

Imai, Nobuaki*; Hirayama, Yoshikazu*; Ishiyama, Hironobu*; Jeong, S.-C.*; Miyatake, Hiroari*; Watanabe, Yutaka*; Makii, Hiroyuki; Mitsuoka, Shinichi; Nagae, Daisuke*; Nishinaka, Ichiro; et al.

European Physical Journal A, 46(2), p.157 - 160, 2010/11

 Times Cited Count:3 Percentile:28.45(Physics, Nuclear)

Journal Articles

Gated multiple-sampling and tracking proportional chamber; New detector system for nuclear astrophysical study with radioactive nuclear beams

Hashimoto, Takashi; Ishiyama, Hironobu*; Ishikawa, Tomoko*; Kawamura, Takashi*; Nakai, Koji*; Watanabe, Yutaka*; Miyatake, Hiroari; Tanaka, Masahiko*; Fuchi, Yoshihide*; Yoshikawa, Nobuharu*; et al.

Nuclear Instruments and Methods in Physics Research A, 556(1), p.339 - 349, 2006/01

 Times Cited Count:33 Percentile:88.70(Instruments & Instrumentation)

A new type of three dimensional tracking and proportional gas counter has been developed. Adopting a gating-grid system, performance of the detector becomes stable under the injection rate of charged particles less than 4$$times$$10$$^4$$ pps. It is a useful detection system for astrophysical experiments using radioactive nuclear beams, since the efficiency is so high as 100 %.

Journal Articles

None

Yamaguchi, Katsuhisa

Nihon Genshiryoku Gakkai-Shi, 35(5), p.382 - 383, 1993/00

None

JAEA Reports

Key design parameter study (II) for large scale-up fast breeder reactor; Optimizing analysis of inherent negative reactivity feedback effect (I); Analysis on thermal transformation of core support plate

; Tanigawa, Shingo*; ; Yamaguchi, Katsuhisa; ; ;

PNC TN9410 88-141, 159 Pages, 1988/09

PNC-TN9410-88-141.pdf:10.2MB

The structural analyses of the core support plate have been applied to study thermal transfomation behaviors and the differences of the movement by changing analytical model, under anticipated transient without scram (ATWS) conditions of FBR. The analyses have been performed for 1000 MWe class loop type fast breeder reactor using a structural analysis code FINAS. The thermal-hydraulic results, which have been performed to ATWS conditions using a plant system code, were used as the thermal boundary conditions to the calculation. The scope of the analyses included a whole section of reactor vessel and the dead load of core assemblies was also considered. Following results were obtained from these studies. (1)The thermal transformation of a upper core support plate can be evaluated according to the free expansion behavior owing to the temperature change of core support plate itself. (2)The radial restriction due to core subassemblies has much influence on the axial bend of the core support plate. (3)There are some differences to the transformation results between by the whole model and by the one dimensional model during the thermal transient is large. Another analysis will be needed, however, about the reactivity change according to the displacement of the core structure.

JAEA Reports

Thermal-hydraulic analysis of plant dynamics test predictive analysis using SSC-L

; Haraguchi, Tetsuharu*; ; Tanigawa, Shingo*; Yamaguchi, Katsuhisa

PNC TN9410 88-107, 121 Pages, 1988/09

PNC-TN9410-88-107.pdf:4.84MB

In the studies using PLANDTL, it would be planned to valid the thermal-hydraulic analysis codes which were developed each for whole system, plenum and subassembly, and also to evaluate the reactor plant in the future using these codes. SSC-L is to be as the main code in these studies and is used for design analysis through test analysis. In the first step of this study, model development and modification of SSC-L has been achieved for PLANDTL and predictive analyses have been applied as to validate the models and examine the design of PLANDTL. The estimated transient curves have been obtained about flow rate and temperatures at subassembly and loop of PLANDTL. As a result, the design conditions have been given to be able to perform the programmed tests. It have been validated that the conditions of tests would be within the design value, and the characteristics of PLANDTL and operational conditions have been obtained from the predictive analyses using design data of the plant. The modification and validation of SSC-L will be applied using the results of various kinds of functional tests, and test analyses will be performed in future.

JAEA Reports

Key Technological Design Study of a Large LMFBR(2); System Dyrawics Analysis for Mitigating ATWS Consequences of a 1000MWe Loop-Type LMFBR

Yamaguchi, Katsuhisa

PNC TN9410 87-161, 19 Pages, 1987/11

PNC-TN9410-87-161.pdf:2.49MB

A system dynamics analysis was applied to a 1000 MWe loop-type liquid-metal fast breeder reactor (IMFBR) to examine influence of possible innovative reactor designs on mitigating consequences of anticipated transients without scram (ATWS). Theanalysis included all the reactivity feedbacks having been employed in current analyses of hypothetical core disruptive accidents (HCDAs). In addition, the present analysis stressed inherent responses of the reactor system by including structural reactivity feedbacks due to axial expansion of control rod driveline (CRD) and radial expansion of reactor core driven by the expansion of the core support plate (CSP). An upper-core flow chimney was considered to make the CRD expansion feedback effective. The flow coastdown rate of the primary pump and the initial position of the control rod (CR) were treated as parameters. ATWS initiators examined were unprotected loss-of-flow (ULOF), loss-of-heat-sink (ULOHS) and transient overpower (UTOP). The ULOF accident was mitigated and peak sodium temperature was suppressed below boiling point by using the primary pump having a 40 s halving time of flow coastdown. The halving time could be shortened to 10 s by assuming that the CR was initially inserted into the active core by about 250 mm. The CRD expansion feedback controlled the earlier transient, and the CSP expansion feedback became dominant in the latter phase. The ULOHS consequence was eompletely enveloped in that of ULOF accident. The sodium temperatures in the primary system became lower than the ULOF case by about 100 $$^{circ}$$C. The UTOP accident conceivable from the current plant design, i.e., the reactivity insertion of 60 ¢ with the rate of 1 to 3 ¢/s, suppressed sodium temperatures and fuel melt fractions below 650 $$^{circ}$$C and 25 5, respectively.

JAEA Reports

Key Design Parameter Study(1) for Large Scale-up Fast Breeder Reactor

Yamaguchi, Katsuhisa

PNC TN9410 87-160, 19 Pages, 1987/11

PNC-TN9410-87-160.pdf:2.42MB

A system dynamics analysis was applied to a 1000 MWe loop-type liquid-metal fast breeder reactor (LMFBR) with mixed oxide fuel to examine inherently safe, passive shutdown performance of the reactor under anticipated transients without scram (ATWS) conditions. The analysis included all of the reactivity feedbacks having been employed in current analyses of hypothetical core disruptive accidents (HCDAs). In addition, the present analysis stressed inherent responses of the reactor system by including structural reactivity feedbacks due to axial expansion of control rod driveline (CRD) and radial expansion of the reactor core driven by the expansion of the core support plate (CSP). The ATWS initiator examined was an unprotected loss-of-flow (ULOF) accident caused by a site power blackout. Some design options were considered: An upper-core flow chimney was introduced to make the CRD expansion feedback effective. The flow coastdown rate of primary pumps was selected to have a halving time of 10 seconds. The initial position of the control rods (CRs), the pony-motor flow level, and so on, were treated as parameters. By assuming that the CRs were initially inserted into the active core by about 0.2 m and that a pony-motor has a capacity of driving more than 20 % of the rated flow, the ULOF consequence was mitigated and the peak sodium temperature was suppressed below boiling point. The CRD expansion feedback controlled the short-term transient, and the CSP expansion feedback became dominant in the later phase. The asymptotic temperature of the primary heat transport system against the upset was successfully lowered below 650 $$^{circ}$$C by optimizing the design parameters.

Journal Articles

Flow pattern and dryait under sodium boiling conditions at decary power levels

Yamaguchi, Katsuhisa; ;

Nuclear Engineering and Design, 99, p.247 - 263, 1987/02

 Times Cited Count:18 Percentile:83.35(Nuclear Science & Technology)

Quasi-steady low-heat-flux sodium boiling experiments were conducted examine the heat removal capability of liquid-metal fast breeder reactor(LMFBR) fuel subassemblies under sodium boiling conditions. The Sodium Boiling and Fuel Failure Propagation Test Facility, SIENA, and two 37-pin bundles, 37F and 37G, were used for this study.

JAEA Reports

Pre-test analysis of the plant dynamics test using SSC-L

; ; Yamaguchi, Katsuhisa

PNC TN9410 86-019, 39 Pages, 1986/03

PNC-TN9410-86-019.pdf:2.19MB

The Sodium Boiling Experiment at Decay Heat Levels (DHB) has been performed to study the transient thermal-hydraulics of the liquid metal fast breeder reactor (LMFBR) fuel subassemblies with emphasis on the so-called second temperature peak which might be caused during the hypothetical loss-of-pipe-integrity (LOPI) accident.Furthermore, it is planned to construct a new sodium loop, Plant Dynamics Test Loop (PLANDTL), to conduct a simulation experiment of the whole plant dynamics including the first and second temperature peaks. It is required to establish the analytical means using the safety analysis codes available, preparing not only for the pre-test and post-test analyses of these experiments but also for the safety assessment of the LMFBR plant on the basis of the usage experiences. SSC-L ls one of the dynamics analysis codes for evaluating the plant responses to various kinds of accidents including LOPI. It has been checked to verify its ability to become being modified in its analytical models upon requests and being improved to make it handy. The post-test analysis of the DHB experiment was performed using SSC-L, thus devised, and, based on the exercise, the pre-test analysis was conducted concerning the standard experimental case which would be tested with PLANDTL.

JAEA Reports

Sodium Boiling Experiments at Decay Power Levels(4); Summary Assessment of the Low-Flow and Low-Heat Flax Sodium Boiling Expeiments at PNC

Yamaguchi, Katsuhisa; ; Aoki, Tadao

PNC TN941 85-56, 289 Pages, 1985/03

PNC-TN941-85-56.pdf:9.23MB

The objective of the Sodium Boiling Experiments at Decay Power Levels is to examine the heat removal capability of reactor fuel subassemblies under sodium boiling condition, which is a matter of arguments in analyzing the accidents like the loss of piping integrity and the loss of shutdown heat removal system. Prior to progressing the test program, a survey study was conducted to fix the scope within which an advanced investigation should be required to analyze the event sequence following after sodium boiling at low flow. The study focused on the results of the past low-flow and low-heat-flux boiling tests performed with the SIENA Facility, with special attension to summarizing the critical (dryout) conditions of the two-phase flow heat transfer. The corresponding data for water cases were also examined. The topical results are as follows: (1)The dryout phenomenon reproducible under annular flow condition at relatively high flow is well predicted to occur by the criterion that the exit quality equals to 0.5. (2)Even if the mismatching ratio of power to flow is increased at low flow range, the slug flow pattern is sustained, repeating the void expansion and contraction synchronized with the unstable flow oscillation. In this case, the extra-superior heat removal capability is expected due to strong heat sink around the voided region. The tendencies of the dryout quality data at annular flow on several parameters are resembled to those of water data, from which one can reach the conclusion that the dryout criterion confirmed here would be reasonable for the sodium flow cases. The forthcoming experiment should be, therefore, concentrated on examining the factors influential to the flow pattern transition and on generalizing the dryout data base at annular flow having less coolable nature.

JAEA Reports

Experimental studies on the coolant mixing effect in a "Joyo" irradiated fuel Assembly; Experimental results for the 7-pin bundle in the A-Type irradiation fuel assembly

; ; ; Okada, Toshio; Yamaguchi, Katsuhisa; Aoki, Tadao

PNC TN941 85-13, 176 Pages, 1985/02

PNC-TN941-85-13.pdf:6.47MB

The subchannel coolant mixing effect in the "Joyo" A-Type irradiated fuel assembly, has been experimentally investigated by using the sodium test 100p, the mixing test loop, and a mockup test section (7 heater pin-bundle; pin diameter=6.5 mm, wire diameter=0.9 mm, pin pitch=7.45 mm, wire wrapping pitch=209 mm). The radial and axial temperature profiles were measured with calibrated Chromel-Alumel thermocouple (6.5 mm in outer diameter) attached on the pin and rapper tube surfaces. The experiments were performed under the following conditions: (Inlet sodium temperature 370$$^{circ}$$C) (Linear power of heater pin 40$$sim$$520W/cm) (ReynoldS number 5,700$$sim$$41,000) The subchannel sodium temperature profiles measured were compared with the calculated results by the SWIRL and COBRA-IV codes, while the cladding temperature profiles measured were compared with the results by the SPOTBOW code. These analyses led to the following conclusion: (1)The mixing coefficients Cs(1), Cs(2) and Cs(3) of the SWIRL code were fitted satisfactorily by the data and the values for the 7-pin bundle were 0.55, 2.30 and 0.74 in order at the Reynolds number of 30,000. The mixing coefficient Cs(1) was less sensitive to the calculated subchannel temperatures than the others, Cs(2) and Cs(3). The mixing coefficient Cs(3) decreased with the increasieng Reynolds number, while the others kept almost constant. (2)The forced cross flow parameters DUR1, DUR2 and DUR3 of the COBRA-IV code were also fitted wen and the following values were obtained for the 7-pin bundle at the Reynolds number of 30,000: DUR1=0.012, DUR2=0.05 and DUR3=0.08. The sensitive forced cross flow parameters DUR(1) was less sensitive to the calculated subchannel temperature than the others, All of them decreased with increasing Reynolds number. (3)The circumferential temperature distribution caluculated by the SPOTBOW code for the cladding of a specified heater pin was higher than that measured. However, there was ...

Journal Articles

Boiling and Dryout Conditions in Disturbed Cluster Geometry and Their Application to the Liquid-Metal Fast Breeder Reactor Local Fault Assessment

Yamaguchi, Katsuhisa; Nakamura, H.; Haga, Kazuo

Nuclear Science and Engineering, 88(3), p.464 - 474, 1984/11

 Times Cited Count:6 Percentile:57.12(Nuclear Science & Technology)

The effect of a local cooling disturbance caused by an edge-type blockage in a liquid-metal fast breeder reactor(LMFBR) fuel subassembly was investigated with a series of out-of-pile local blockage experiments with water and sodium. The heat exchange layer model first developed for central-type blockage cases applied well to the present edge-type cases. An empirical formula was developed for estimating maximum temperatures in various subassembly, and the conclusion was reached that a middle size edge-type blockage could lead to sodium boiling. The critical heat flux data of Power Reactor and Nuclear Fuel Development Corporation and Kernforschungszentrum Karlsruhe were correlated with the boiling inception heat flux for various core flow velociti

JAEA Reports

A Data processing program for transient sodium boiling and fuel failure propagation Tests (III); SISCO-A Computer code to analyze preservation data file

Onda, Kaoru*; Yamaguchi, Katsuhisa

PNC TN952 83-08, 258 Pages, 1983/12

PNC-TN952-83-08.pdf:205.61MB

A Data processing program SISC0, which has capabilities of checking the information of experimental data measurements with little labor and of analyzing the data conveniently using the checked information, was developed to conduct the accurate data processing of the test results of low-flow and low-heat-flux sodium boiling runs and the other runs obtained from the experiments carried out with the Transient Sodium Boiling and Fuel Failure Propagation Test facility. It is devised to use, to the full extent, the utility libraries of the Interactive Data Analysis System AXEL installed in the FACOM M-190/200 computer system at the O-arai Engineering Center. The data managements of the SISCO code are composed of two sub-systems : one is the BSISCO having functions of creating or updating the data bases with batch jobs to prepare for subsequent data analyses, and another is the ISISCO which is in charge of analyzing the data with TSS job under the support of AXEL system. In onder to respond to various kinds of requirements from analysts, seven option routines and thirteen command-macro procedures are available in respective sub-systems. The SISCO code supports following two types of data handling jobs : (i)data calibration job ; to fix and service the information of measurements, which is to be compiled into a preservation data file for the data reproduction, by the statistical analyses of calibration data and the displays of the results, (ii)data analysis job ; to obtain the graphic outputs of the multi-data analyses applied to the specified preservation data files. The data analysis of the SISCO code is so designed that one can use a certain command-macro as a stand-alone utility program appliciable to any test data comprehensively. It is possible to yield (a) summary traces of the test results containing the sub-frames of time series data plots of key signals in one figure, (b) multi-traces of the axial temperatures at required transient times, (c) analytical ...

JAEA Reports

Experimental study of local sodium boiling detection by use of temperature and flow fluctuations at the outlet of subassembly

; ; Haga, Kazuo*; Yamaguchi, Katsuhisa;

PNC TN941 83-97, 71 Pages, 1983/06

PNC-TN941-83-97.pdf:2.53MB

Out-of-pile local boiling experiments were carried out with a wire spacer type 91-pin bundle. A half edge part of its cross-sectional area was blocked by a 5mm thick stainless-steel plate. The purpose of the experiments is to study the feasibility of detecting coolant boiling accident within a LMFBR fuel subassembly using the temperature and flow fluctuations at the outlet of individual subassembly. Initial conditions of the experiment are as follows: (1)Bundle inlet temperature : 400 $$sim$$ 500 $$^{circ}$$C (2)Heat flux : 72 $$sim$$ 93 W/cm$$^{2}$$ (3)Coolant flow velocity at the normal bundle section : 0.64 $$sim$$ 1.13 m/s. Local boiling was generated immediately behind the blockage by increasing heat flux gradually, with inlet sodium temperature and coolant flow rate being held to be constant. Specifications of the instrumentation system for measuring temperature and flow fluctuations are listed below: (1)Thermocouples $$cdot$$ Chromel-Alumel thermocouples, Grounded-type (O.D. 0.3mm) and ungrounded-type (O.D. 4.8mm) $$cdot$$ Time constant (63.2%) 10 msec and 2.14 sec respectively (2)Flowmeter $$cdot$$ Eddy-current type temperature and flowmeter $$cdot$$ Exciting frequency : 425 Hz (3)Fluctuation measuring circuit : $$cdot$$ Maximum gain : 60 dB $$cdot$$ Frequency characteristics : flat in the frequency range from 0.01 to 15 Hz, 20 dB/dec attenuation slope below 0.01 Hz and 200 dB/dec attenuation slope above 15 Hz. The following subjects have been studied. (1)To investigate how boiling informations are transferred. (2)Anomaly detection methods by use of the temperature and flow fluctuations. (3)Feasibility of detecting local boiling accident by fluctuation signals. These studies yielded the following conclusions. (1)Spectral peak arround 4 Hz in the power spectral density functions of the temperature fluctuations was observed at the downstream bundle section of the blockage. This frequency peak is consistent with that of void formation and collapse and may ...

JAEA Reports

Temperature rise due to fission gas release in locally blocked LMFBR fuel subassembly simulators

Haga, Kazuo*; Yamaguchi, Katsuhisa; Namaekawa, F.*

PNC TN941 82-189, 10 Pages, 1982/09

PNC-TN941-82-189.pdf:0.47MB

The combined thermal effects of local blockage and fission product gas release in an LMFBR fuel subassembly were investigated in the present study. The temperature rise in the wake region was measured in out-of-pile experiments using three sets of electrically heated 37-pin bundles cooled by sodium. From the test results the influence of the main factors which govern the cooling capability of the blocked bundle, such as gas release rate, sodium flow rate, spacer type and blockage location, was examined. The temperature rise due to gas release in the central blocked wire-spacer bundle was almost identical to that in the edge blocked grid-spacer bundle. From these results the wake temperature rise in the case of gas release under reactor operating conditions was estimated.

Oral presentation

Research and development using fast breeder prototype reactor Monju

Yamaguchi, Katsuhisa

no journal, , 

no abstracts in English

Oral presentation

Isobaric analog resonances of $$^{69}$$Zn by thick target inverse kinematics proton resonance scattering

Imai, Nobuaki*; Hirayama, Yoshikazu*; Ishiyama, Hironobu*; Jeong, S.-C.*; Miyatake, Hiroari*; Watanabe, Yutaka*; Makii, Hiroyuki; Mitsuoka, Shinichi; Nagae, Daisuke; Nishinaka, Ichiro; et al.

no journal, , 

The isobaric analog resonance of $$^{69}$$Zn was investigated by resonance elastic scattering of p($$^{68}$$Zn,p) with a 5.5 MeV/nucleon $$^{68}$$Zn beam and a thick CH$$_2$$ target at JAEA tandem accelerator facility. Excitation function of the differential cross section of proton elastic scattering was measured around 0 degree in the laboratory frame. Angular momentum, and proton and total widths of the resonance determined using $$R$$-matrix calculation are in good agreement with earlier measurements performed in normal kinematics.

Oral presentation

The Modified neutron multiplication method for Monju

Yamaguchi, Katsuhisa; Truchet, G.*; Van Rooijen, W. F. G.*; Shimazu, Yoichiro*

no journal, , 

Oral presentation

Fuel failure and Failed fuel detection at "JOYO" and "MONJU"

Yamaguchi, Katsuhisa

no journal, , 

no abstracts in English

22 (Records 1-20 displayed on this page)