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Journal Articles

Numerical visualization on a Large-scale bubbly flow in a vertical small duct

Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Akimoto, Hajime; Aoki, Takayuki*

Kashika Joho Gakkai-Shi, 25(Suppl.1), p.435 - 436, 2005/07

no abstracts in English

JAEA Reports

None

*; *; ; Aoto, Kazumi

JNC-TY9400 2000-010, 138 Pages, 2000/03

JNC-TY9400-2000-010.pdf:5.15MB

None

JAEA Reports

Damage evaluation of vessel model under thermal transient loading; Detection of damage by wavelet analysis for ultrasonic waveform

; *

JNC-TN9400 2000-018, 37 Pages, 2000/03

JNC-TN9400-2000-018.pdf:1.34MB

The damage evaluation for the vessel model on the cyclic thermal transient loading in sodium were performed by the ultrasonic detection method. The wavelet analysis that was an analysis method of the waveform was applied to detect the micro damage before a sign of the crack initiation. The time-frequency analysis by the wavelet transform was performed to evaluate the ultrasonic parameter for the micro damage. As the results, the ultrasonic echo was analyzed by some mother wavelet, and Gabor wavelet was reasonable. The analysis of ultrasonic echo by Gabor wavelet showed drop of the sound velocity at higher frequency than the peak frequency because of attenuation in the high frequency component. The difference of the peak frequency △fp between B1 and B2 echoes increased with the amount of damage, and △ fp was available as a parameter for the micro damage detection. The correlation between the sound velocity and the micro hardness for the amount of damage was also found, and each method suggested to available alternately. ln this study, it was indicated that an ultrasonic wave characteristic value that can detect damaged state before crack initiation was obtained from the wavelet analysis.

JAEA Reports

Post processing system for multi-dimensionaI thermal-hydraulic analyses

Miyake, Yasuhiro*; *; ; Kimura, Nobuyuki

JNC-TN9400 2000-016, 40 Pages, 1999/12

JNC-TN9400-2000-016.pdf:3.71MB

ln the conventional visualization system for the computational results, only Japanese (Nihongo) Line Printer (NLP) was available to print two dimensional cross sectional plots of vector and scalar fields. To evaluate the phenomena, an analyst had to print many plots on the NLP. This task makes difficult to check the computational results immediately after the calculation. Recently, as the visualization tools, we introduced Micro AVS and Field View which are utilized widely in the scientific and the industrial fields. ln order to show the numerical results on the visualization software, we constructed a post processing system which convert the results of the numerical code to "lntermediate files" which can be read by the visualization tools. As using this system, the examination of the numerical results can be executed on the display of the personal computer. Furthermore, the persuasive report and paper with high quality can be produced due to the color printing. As for the transient calculation, the change of the phenomena can be visually evaluated by using the animation function.

JAEA Reports

A Note on the representation of rate-of-rise of the thermal stratification interface in reactor plenum

Tokuhiro, Akira; Kimura, Nobuyuki

JNC-TN9400 2000-015, 26 Pages, 1999/09

JNC-TN9400-2000-015.pdf:1.43MB

The quantification of the rate-of-rise of the thermal stratification interface, a "thin" vertical zone where the temperature gradient is the steepest, is important in assessing the potential implications of thermally-induced stress problems in liquid-metal cooled reactors. Thermal stratification can likewise occur in confined volumes containing ordinary fluids (Pr$$geq$$1), where there is an input of thermal convective energy. In the prominent case of liquid metal reactors, there have been many studies on quantifying the rate-of-rise of a defined stratification interface, in terms of one or more of the following dimensionless groups, mainly: Richardson (Ri), Reynolds (Re), Grashof (Gr), Rayleigh (Ra) and/or Froude (Fr) numbers. Stratification is also a transient process in the volume in question. In the present work the anthors presents a derivation based on order-of-magnitude analysis (OMA), including an sensible energy balance, that produces a new representation more consistent than p

JAEA Reports

None

JNC-TN1400 99-017, 439 Pages, 1999/08

JNC-TN1400-99-017.pdf:14.06MB

no abstracts in English

JAEA Reports

Development of the flow control irradiation facility for JOYO

Soroi, Masatoshi;

PNC-TN9410 98-050, 57 Pages, 1998/05

PNC-TN9410-98-050.pdf:1.58MB

This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor "JOYO". FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. The main results and problems of the development of FLORA are as follows; (1)The results of the development: (a)The neutron detector in FLORA can detect the delayed neutron which is emitted from failed fuel. (b)Out-of-pile A-LIP tests in sodium conditions has been completed. (The length of the tested A-LIP is half the actual size.) Out-of-pile test results showed that the A-LIP achieved a 300$$ell$$/min flow rate and 265kPa pressure in 550$$^{circ}$$C sodium. This pump performance satisfied the FLORA requirements. (c)By controlling the sodium flow rate from 40 to 100% using the A-LIP, we can control the fuel cladding temperature satisfactorily. (2)The problems: (a)In the development of the process detector, it is necessary to miniaturize the neutron detector and test the effect of neutron irradiation and high temperatures on the permanent magnet in the flow meter. (b)The problem which is left about A-LIP is its influence on neutron irradiation. For this purpose, we have to irradiate a small size A-LIP and test its characteristics and electric isolation. (c)To get more accurate results concerning the efficiency of the A-LIP, we have to ...

JAEA Reports

None

; Matsumoto, Mitsuo;

PNC-TN1410 97-039, 99 Pages, 1997/10

PNC-TN1410-97-039.pdf:2.25MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-031, 638 Pages, 1997/08

PNC-TN1410-97-031.pdf:12.12MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-010, 462 Pages, 1997/02

PNC-TN1410-97-010.pdf:17.56MB

no abstracts in English

JAEA Reports

Implementation of an MRACnn System on an FBR Building Block Type Simulator

Ugolini; Yoshikawa, Shinji; Ozawa, kenji

PNC-TN9410 95-253, 13 Pages, 1995/10

PNC-TN9410-95-253.pdf:0.5MB

This report presents the implementation of the a model reference adaptive control system based on the artificial neural network technique (MRAC$$_{nn}$$) in a fast breeder reactor (FBR) building block type (BBT) simulator representing the Monju prototype reactor. The purpose of this report is to improve the control of the outlet steam temperature of the three evaporators of the Monju prototype reactor. The connection between the MRAC$$_{nn}$$ system and the BBT simulator is achieved through an external shared memory accessible by both systems. The MRAC$$_{nn}$$ system calculates the demand for the position of the feedwater valve replacing the signal of a PID controller collocated inside the heat transport system model of the Monju prototype reactor. Two series of simulation tests havc been performed, one with one loop connected to the MRAC$$_{nn}$$ system (leaving the remaining two connected to the original PID controller), and the other with three loops connected to the MRAC$$_{nn}$$ system. In both simulation tests the MRAC$$_{nn}$$ system performed better than the PID controller, keeping the outlet steam temperature of the evaporators closer to the required set point value through all the transients.

JAEA Reports

Development of program for arbitrary real time simulation; (1)Design of prototype program

kasahara, Naoto;

PNC-TN9410 95-211, 32 Pages, 1995/08

PNC-TN9410-95-211.pdf:1.54MB

In order to optimize elevated temperature structural systems in fast reactor plants, where main loading is thermal stress induced by transient operation of circuit, authors proposed new design frame by applying design by analysis concept to both structural design and system design. A key technology in this design frame is an integrated analysis method for both thermo-mechanical behaviors of structures and plant thermo-hydraulic dynamics, developing of a prototype code of which, named PARTS (Program for Arbitrary Real Time Simulation), was started this year. For the purpose to achieve flexible coupling of several codes, authors designed three categories of calculation parts (objects): (1)thermo-hydraulics of coolant, (2)thermo-mechanical behavior of structures, and (3)material strength. These calculation parts can be handled and connected easily on the PARTS-Workbench. Real time simulation is planed to be accomplished by parallel processing of individual parts calculation, and by prediction of neural-network which learned past calculation results. Object oriented languages, Smalltalk and C++, were adopted for implementation of calculation parts. The PARTS-Workbench was programed by visual Basic and Visual Smalltalk for considering user customization. In the next phase of study, parallel processing function and neural network parts will be incorporated in the PARTS code. Prototype of this code is going to be completed until F.Y.1996. and be applied to study of thermal mitigation structures.

JAEA Reports

Study of thermohydraulic behavior within the fuel bundle under a loss of flow condition

M.E.Kab*;

PNC-TN9410 92-018, 58 Pages, 1992/01

PNC-TN9410-92-018.pdf:1.31MB

This report describes the result of the analysis of unprotected Loss of Flow (LOF) ansient experiment conducted at the PLANt Dynamics Test Loop (PLANDTL) experimentalfility by Super System Code (SSC) and SubAssembly Boiling EvolutioN Analysis (SABENA)ode. This report also describes the effect of the modification we made in SSC with t recent void fraction and two-phase friction multiplier models during the analysis othe experiment. After the analysis, it was found that the two-fluid two-phase flow mel of SABENA 1-D is better than the homogeneous model of SSC in predictiong the therhydraulic behavior within the simulated fuel bundle test section of thePLANDTL facily in case of high quality sodium boiling experiment. Moreover, it wasalso revealed tt the two-fluid one dimensional model is not accurate enough in predicting the onsetf boiling and axial evolution of boiling region inside the heatedchannel.

JAEA Reports

0peration experience report of experimental fast reactor JOYO; A special level monitoring for reactor vessel in the occurrance of the abnormal 1evel incident

; ; ; ; Ozawa, kenji; ; Terunuma, Seiichi

PNC-TN9410 91-187, 41 Pages, 1991/07

PNC-TN9410-91-187.pdf:1.0MB

A reactor vessel in JOYO provides three induction type level meters which is defined in the safety protection system. They have two kinds of measuring range and display the sodium level below to the discharge nozzle of the primary cooling system. One is from 350mm about the normal sodium level to 1,600mm below it and other two sets are from 350mm above to 350mm below it. This report describes a special monitoring method of sodium level in the occurrence of the abnormal sodium level incident which reaches it more than 1600㎜ below the normal sodium level in the reactor vessel. The special monitoring method uses the discharge sodium pressure of the primary auxiliary cooling pump. A discharge sodium pipe from the primary auxiliary cooling pump is located in the bottom of the reactor vessel and it's discharge pressure is correlated with the reactor vessel sodium level which works back pressure to the pump. Therefore, it was assumed that abnormal sodium level which reaches it more than 1600mm below the normal sodium level can be monitored using this discharge sodium pressure. A verification test was conducted to measure the correlation of the discharge sodium pressure and the reactor vessel sodium level. Main results obtained from this test were as follows. (1)Validity of this special level monitoring method was confirmed in the sodium level range from normal to 3,390㎜ below it and in case of sodium level changing which is decreased at the rate of 47.5m$$^{3}$$/h by this test during the system sodium drain work. (2)A correlation equation is obtained using parameters of discharge sodium pressure, flow and temperature of the primary auxiliary cooling system to gain sodium level of reactor vessel. (3)Parametor chart of the reactor vessel sodium level was made using multi regressive analysis.

JAEA Reports

Investiation on presence of inner barrel for large fast breeder reactor

*

PNC-TN9410 90-147, 115 Pages, 1990/10

PNC-TN9410-90-147.pdf:4.05MB

In-vessel thermohydraulics analysis was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to evaluate effects of an inner barrel on a large fast breeder reactor. Then four thermohydraulics phenomena, a thermal stratification, a main loop temperature transient, a circumferential temperature distribution and a sodium surface velocity were discussed. Through the analysis using the multi-dimensional code AQUA and the discussion, the following have been effects of the inner barrel as obtained: [Thermal Stratification] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of an axial temperature distribution can be neglected from a structural design. [Main Loop Temperature Transient] An inner barrel is required. Because a cold shock with maximum temperature transient -2.0$$^{circ}$$C/s occurred at a outlet nozzle when an inner barrel was not equipped. [Circumferential Temperature Distribution] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of the temperature distribution can be neglected from a structural design. But further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface velocity] An inner barrel is unnecessary. From the above results, it is concluded that an inner barrel is unnecessary if the cold shock is improved by a increase of effective mixing region on a design.

JAEA Reports

Outline of air-cooling thermal transient test facility

*; *; Uno, Tetsuro*

PNC-TN9410 86-029, 68 Pages, 1986/02

PNC-TN9410-86-029.pdf:12.61MB

A new test facility "Air-Cooling Thermal Transient Test Facility" (ATTF) was constructed at O-arai Engineering Center. This test facility is utilized, in the first place, for evaluating the strength of outlet tube-sheets of steam generators of FBR Plants. The objectives of the tube-sheet model tests are as follows. The first is to investigate and evaluate the strain concentration in the plastic region. The second is to confirm the adequacy of the design criteria for the prototype reactor MONJU. The third is to confirm the safety margin for failure incorporated in the design evaluation methods. ATTF can impose severe thermal loadings (only cold shock) on the test specimens. The facility produces compressed air (Max. 35kg/cm$$^{2}$$G) by two large-sized compressors, and stores it in a storage tank (about 60m$$^{3}$$). After a test specimen is heated up to the aimed temperature the compressed air passes through the test specimens quickly by opening the valve to apply cold shock and is released in the atmosphere. Each main loop pipe is 8 inches in diameter and the flow rate is max. 10kg/s in compressed air. The most severe down thermal transient condition is from 550$$^{circ}$$C to 150$$^{circ}$$C (for tube-sheet model) in about 4 min. The test section can be modified for various kinds of structures, which should be air-tight and have the maximum pressure of 8kg/cm$$^{2}$$G. The facility is operated automatically by two sequencer controllers. One of the main features of ATTF is the adoption of compressed air instead of sodium as coolant. By using compressed air, various kinds of sensors which can not be used in the sodium environment can be used in ATTF; particularly strain gages can be used effectively to obtain strain distribution for thermal transient condition, and the location as well as the mode of failure of test specimens can be recognized easily through the detection of crack initiation and the observation of crack growth. ATTF is expected to be a powerful ...

JAEA Reports

None

PNC-TN243 81-08, 14 Pages, 1981/11

PNC-TN243-81-08.pdf:0.34MB

no abstracts in English

Oral presentation

Effect of dielectric characteristics on transient boiling phenomena induced by microwave heating

Yamaki, Tatsunori*; Abe, Yutaka*; Kaneko, Akiko*; Kanagawa, Tetsuya*; Kitazawa, Toshihide*; Segawa, Tomoomi; Kawaguchi, Koichi; Yamada, Yoshikazu

no journal, , 

For practical application of high-volume production of microwave heating denitration method, it is required to avoid the transient boiling phenomena of overflow and flushing during microwave heating and to optimize the design condition of vessel shape and microwave outpot. The boiling phenomena and flow structure of KCl aqueous solution by microwave heating were measured with the KCl concentration as a parameter. Flushing phenomena does not become difficult to occur and flow structure becomes disordered to be created the vortex structure according to increase of the KCl concentration. From the results of measuring the temperature distribution of side cross-section surface of the KCl jelly, near the center is mainly heated in the case of the water jelly. On the other hand, around the jelly is mainly heated in the KCl jelly. It is clarified that the generation condition of flushing and the boiling phenomena are significantly influenced by the difference of water and KCl solution.

Oral presentation

Development of multiscale numerical simulation method for thermal transient phenomena of sodium-cooled fast reactors, 1; Outline of simulation method development

Tanaka, Masaaki; Hiyama, Tomoyuki; Murakami, Satoshi*; Doda, Norihiro; Ohshima, Hiroyuki

no journal, , 

In order to improve the accuracy of the numerical estimation method for the thermal transient phenomena in the sodium-cooled fast reactor has been conducted by using code coupling technology with the system analysis code for plant dynamics analysis, the multi-dimensional code for analysis of thermal-hydraulics in the plenum, and numerical estimation code of structural integrity for the local region, in viewpoint of enhancement of safety measures in sodium-cooled fast reactor. In this presentation, outline of simulation method development is introduced.

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