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Aoki, Tadao; Sawada, Makoto; Oshita, Hironori; Matsuno, Yoshiaki*; Kitano, Akihiro; KAMEI, Michiru*
JNC TN4520 2004-001, 350 Pages, 2004/09
JNC-TN4520-2004-001.pdf:36.76MB
none
Oshu Genshiryoku Gakkai Genshiryoku Koho Tantosha Kokusai Koryu Kaigi, 0 Pages, 1998/00
*; *; Nakanishi, Seiji; Aoki, Tadao
PNC TN9410 86-004, 120 Pages, 1986/01
Thermal stratification tests of one-tenth scale models simulating the "Monju" reactor upper plenum were carried out to understand the therma stratification phenomena occuring in reactor upper plenum and to be reflected in the design of "Monju". The purpose of these tests is especially to obtain long-term data and to investigate the disappearance process of thermal stratification. The tests were carried out in following two phase. Phase I; The test model which simulated the whole length on one-tenth scala was used. Phase II; The test model which simulated the axial length only above the top of inner barrel on full scale was used to investigate the motion of stratification interface above the top of inner barrel. As mentioned hereunder, the test results were reflected in the design of "Monju". (1)The test data were utilized for the verification of "COPD", which is the plant dynamics code for "Monju". (2)The test data were utilized for the verification of multi-dimensional thermo-hydraulic analysis codes, which are utilized for the back-up of "COPD". (3)On the basis of the test results, the inner barrel of "Monju" was redesigned.
Aoki, Tadao
PNC TN951 84-04REV1, 38 Pages, 1985/03
PNC-TN951-84-04REV1.pdf:2.98MB
no abstracts in English
Yamaguchi, Katsuhisa; *; Aoki, Tadao
PNC TN941 85-56, 289 Pages, 1985/03
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.
two-equation turbulence model with COMMIX-1AMuramatsu, Toshiharu; Maekawa, I.*; Ninokata, Hisashi; Aoki, Tadao
PNC TN941 85-14, 73 Pages, 1985/02
COMMIX-1A is a single-phase three-dimensional thermal-hydraulic analysis code with finite difference method developed at U.S.Argonne National Laboratory. The code is provided with one-equation turbulence model in terms of turbulent kinetic energy, k. However the major shortcoming of the model is that the transport of turbulent length scale 
is not accounted for. Therefore the supplementary equation related to the turbulent length scale has been added to the original model. The dissipation rate of turbulent kinetic energy has been selected as the unknown variable of the additional equation. The model governed by the set of two equations is thus called "k- model". The incorporated k- model in the COMMIX-1A has been validated in the analyses of the following three turbulence experiments: (1)Pipe flow, (2)Expansion flow and (3)Buoyancy flow. In turbulent pipe flow of Re=3.9
10
, calculated velocity distribution agrees well within the error of 1 %, but distribution of k is underestimated by maximum 73% in the axial region. In the turbulent channel expansion flow case with backward facing step, calculated reattachment length is overestimated by 18 %. In the enclosed buoyancy driven recirculating flow case, calculated temperature, k and distributions have shown good agreement with those of the experiment with accuracy of second order in space.
*; *; *; Okada, Toshio; Yamaguchi, Katsuhisa; Aoki, Tadao
PNC TN941 85-13, 176 Pages, 1985/02
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
C) (Linear power of heater pin 40
520W/cm) (ReynoldS number 5,70041,000) 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 ...
Maekawa, I.*; Takahashi, Minoru; Muramatsu, Toshiharu; *; Ninokata, Hisashi; Aoki, Tadao
PNC TN952 84-11, 101 Pages, 1984/07
COMMIX-1A is a single phase multi-dimensional analysis code for thermal-hydraulics within components of nuclear power plants. The code was introduced from the Argonne National Laboratory in the U.S. in 1983. It has been extensively used for sodium and water analysis and now its improvement and validation programs are under way at the Reactor Engineering Section of OEC/PNC. This report is written based on the materials distributed at COMMIX-1A Workshop, 15th and 16th in May, 1984 which offered PNC engineers the chance to use the code and to get familiarity with it for the solution of thermal hydraulic problems of their own fields. This report consists of the following contents: (1) outline of COMMIX-1A, (2) capability and limitation of the code, (3) usage experiances of the code, (4) main input data description and (5) sample problems.
Muramatsu, Toshiharu; Maekawa, I.*; Ninokata, Hisashi; Aoki, Tadao
PNC TN941 84-98, 69 Pages, 1984/07
The NJS3D computer code has been developed to analyze sodium temperature fluctuation at the outlet region of FBR subassemblies and applied to Experimental Fast Reactor JOYO. Sodium temperature fluctuation causing high cycle thermal fatigue in the upper core structure is known as thermal striping. This code development work is in the preliminary stage and the code should serve as prototypical evaluation tool for the integrities of structures. The code verification study has been carried out making use of the following experiments: (i)Simple tube model, and (ii)7-assemblies mock-up model. As for (i), the comparison of the analysis with the experiment is good. The mesh effects on temperature fluctuation are negligible for mesh scale 1.5 mm/mesh. As for (ii), temperature fluctuation with flow ratio 
= 1.06 is overestimated, and one with flow ratio = 0.39 is underestimated. The code has been applied to JOYO MK-II irradiation core and the maximum temperature fluctuation in the CMIR upper region is predicted to be 97C.
Takahashi, Minoru; Maekawa, I.*; Tamura, Shinji*; Muramatsu, Toshiharu; Hasegawa, Y.*; Ninokata, Hisashi; Aoki, Tadao
PNC TN952 84-08, 157 Pages, 1984/06
COMMIX-1A is a single-phase three-dimensional thermal-hydraulic analysis code with finite difference method developed at U.S. Argonne National Laboratory. ICE version (Ver. 3.0) of COMMIX-1A was released from ANL to PNC through U.S. NRC in 1981. Then, graphics package compatible with version 10.2 of COMMIX-1A was released to PNC in November, 1982, and SIMPLER version (Ver. 12.0) of COMMIX-1A (fixed version) was released in January, 1983. The SIMPLER version and the graphics program have been modified and used for the following analyses at the Reactor Engineering Section: (1)analysis of MONJU upper plenum thermal stratification tests in sodium and in water, (2)post evaluation of JOYO Mark-I natural circulation test, and (3)analysis of water test for thermal hydraulics in pool-type LMFBR hot plenum. After the modification and the experience of COMMIX application at RES, we have made the code accessible for general users in PNC. We have prepared this COMMIX-1A input manual in Japanese so that the user can set up input data and operate the code with great ease for the interim use. The present document consists of the following contents: (i)COMMIX-1A input description both in Japanese and in English, (ii)graphics program input description in Japanese, and (iii)JCL for execution of COMMIX-1A and the graphics program.
Furuhashi, Ichiro*; Nakanishi, Seiji; Aoki, Tadao
PNC TN941 84-50, 195 Pages, 1984/03
A two-dimensional thermal-hydraulic analysis code "NAGARE-2D" has been developed based on the existing codes and the results of comparative validation against experimental data. Major improvements and capabilities added to the conventional codes are as follows: (1)Introduction of pressure drop term to the momentum equation. (2)Automatic treatment of boundary conditions at walls. (3)Introduction of thermal capacity of wall and heat transfer coefficient at wall surface to the enthalpy formula, and change in the convection heat transfer equation. (4)Addition of automatically calculating function of turbulent kinematic viscosity coefficient in K- turbulent model. (5)Addition of heatf lux print out function. (6)Improvement on calculation efficiency by minimizing calculation steps. From these improvements, the application of the code can be extended to wider range of engineering problems.
*; Aoki, Tadao; *
PNC TN941 83-44, 82 Pages, 1983/04
In sodium test on the two models of 22-in isolation valves to be applied to the secondary cooling system of MONJU was carried out. The results of initial performance test was reported in the perious paper PNC SN941 79-57. Present paper describes the long-term performance of the same valve models (about 10,000 hrs for the butterfly valve and about 15,000 hrs for the gate valve) in sodium at 400 C. Results obtained are as follows: (1)The leak test done in sodium and the in-water leak test done after dismantling showed that the butterfly valve satisfies the allowable leak rate of SG isolation valve. The values of allowable leak rates (provisional) are 1.5/sec for ACCS operation and 10 /min for sodium-water reaction in SG. (2)The continuous actuation tests in sodium at 400 C showed that both valves satisfy the allowable actuating time of 60 sec. (3)A test was carried out to assure the integrity of freeze seal under possible high ambient temperature on the failure of air conditioning facility and it was found that the shaft seal function is maintained under the temperature as high as about 80 C. (4)Reliable function of sodium leak detector was demonstrated. (5)Replacement of shaft real part may be needed on the maintenance campaign of MONJU. An effective method using oil-pressure jack is proposed for this replacement.
Aoki, Tadao; *; *; *; *; *
PNC TN960 83-01VOL3, 237 Pages, 1983/02
PNC-TN960-83-01VOL3.pdf:4.62MB
None
Aoki, Tadao; *; *; *; *; *
PNC TN960 83-01VOL2, 302 Pages, 1983/02
PNC-TN960-83-01VOL2.pdf:9.75MB
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Aoki, Tadao; *; *; *; *; *
PNC TN960 83-01VOL1, 110 Pages, 1983/02
PNC-TN960-83-01VOL1.pdf:3.17MB
None
; *; *; Aoki, Tadao; *; *
PNC TN941 82-216, 62 Pages, 1982/09
This report describes the performance test of pony motor system mounted on the MONJU primary prototype pump. The test was conducted under various operation modes, and the following results were obtained: (1)The transition from the main motor to the pony motor was achieved very smoothly. (2)The transmission of the pony motor system worked well with its speed and torque kept constant. (3)start-up of a pump with a pony motor was accomplished very smoothly. (4)Apony motor current was about 6365% of it rated operation. (5)A characteristics of pony motor system in high temperature atomosphere remained as same as that in the room temperature. (6)The amplitude of pony motor bearing vibration during operation was a little.
*; *; *; Aoki, Tadao; *; *
PNC TN941 82-178, 34 Pages, 1982/08
It was already confirmed that SID (Sodium lionization Detector) and DPD (Differential Pressure Detector) type sodium leak detectors for LMFBR "MONJU" had enough detection sensitivity. But an endurance test was required for putting them to practical use. Especially, life time of filament for the sodium aerosol ionization in the SID had to be examined. So, endurance performance of SID and DPD were investigated during about one year using an experimental equipment which simulates nitrogen atmosphere of reactor primary loop and sodium aerosal sampling tube. Results oftained in this test are as follows ; (1)SID filament did not break during one year operation, and little change of detection sensitivity was observed. (2)Neither performance deterioration nor measurement circuit trouble was experienced on the DPD.
*; *; *; Aoki, Tadao; *; *
PNC TN941 82-153, 44 Pages, 1982/07
Research and development of the Ultrasonic Flowmeter (USFM) for the large pipe sodium loop in the Prototype LMFBR "MONJU" have been conducted so far, and sodium flow measurement test using 12-inches USFM was already finished. But cold leg pipe of secondary loop of "MONJU" is 22B (primary loop; 24B), in which USFM is to be installed, so 24-inches USFM was made and some performance were measured in still sodium as the first step of some tests, and following results were obtained. (1)The ultrasonic pulse height enough for the operation of the USFM electronic circuit and long term stability was obtained in 200 550C sodium temperature. (2)The electronic circuit parameters which could collect temperature dependency of USFM output within 
0.2% (200550C) were decided by the measurement of ultrasonic transmission time. (3)The zero point drift of USFM output was measured in the constant and transient sodium temperature, and the stability more than 0.2% (FS 6 m/s) was obtained. From these results, 24-inches USFM is found to have the possibility of utilization, so temperature collection accuracy of USFM Scale Factor and other performance must be confirmed by the sodium flow test for the application in "MONJU".
; *; *; *
Hitachi Hyoron, 11 Pages, 1980/10