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Journal Articles

Chemical composition of insoluble residue generated at the Rokkasho Reprocessing Plant

Yamagishi, Isao; Odakura, Makoto; Ichige, Yoshiaki; Kuroha, Mitsuhiko; Takano, Masahide; Akabori, Mitsuo; Yoshioka, Masahiro*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.1113 - 1119, 2015/09

The characteristics of insoluble residues in fine suspension at the Rokkasho Reprocessing Plant were analyzed. The insoluble residues were washed with oxalic acid solution to dissolve zirconium molybdate residues. XRD profiles of unwashed residues showed the presence of a noble metal alloy, zirconium molybdate, and zirconia, but zirconium molybdate was not found after washing. More than 50% of the Sb-125 and Pu in thee residues was washed out as well. The noble metal alloy composed of Mo, Tc, Ru, Rh, and Pd occupied more than 90% of the total weight of 12 elements (Ca, Cr, Fe, Ni, Zr, Mo, Tc, Ru, Rh, Pd, Te, and U) found in the residues. In consideration of the chemical forms of 12 elements, the alloy-to-residue weight ratio was evaluated to be 64% and 78% with and without 18% of an unknown component, respectively.

JAEA Reports

Study to decrease design basis leak(DBL) in the FBR steam generators(SG); Feasibility test of tube protection sleeve to prolong failure propagation

*; *; Himeno, Yoshiaki; Kuroha, Mitsuo

PNC TN9410 89-123, 54 Pages, 1989/08

PNC-TN9410-89-123.pdf:1.65MB

In the present study, a tube protection sleeve was designed and manufactured as one of the positive measure against tube failure propagation. Its effectiveness was confirmed by water leak test in sodium. The tube protection sleeve manufactured was made of (1)a turn buckle, (2)a spacer and (3)a belt that are made from SUS304 steel. It was attached to a heat transfer tube by a belt of 30mm in width. Its attachment is able to be done easily in short time and is not necessary to weld. In the test, steam was fed to a tube attached by a tube protection sleeve in sodium. An artifitial hole is drilled in the initial leak tube. Sodium temperature was 505$$^{circ}$$C at the test. Results of the test revealed that the tube protection sleeve has enough function to postpone the failure propagation. Major results are as follows: (1)Sodium - water reaction occurred near both ends of the tube protection sleeve. Nevertheless, neighboring tube was not wasted until the failure of the sleeve. (2)At the water leak rate 10g/s, the secondary failure was delayed to six times in comparison to a tube having no sleeve. Therefore, it is possible to detect water leak(normal detection time is over 120sec) well in advance to the secondary failure. Effectiveness of the tube protection sleeve against the failure propagation was demonstrated by test. But, for its application to the SG component, several problem, such as durability and attachment property are still remained.

JAEA Reports

Basic experimental study on the development of acoustic water leak detection system (II)

Shimoyama, Kazuhito; Kuroha, Mitsuo; *

PNC TN9410 87-014, 103 Pages, 1987/01

PNC-TN9410-87-014.pdf:6.09MB

Acoustic type water leak detectors have promising potentiality in short detection time for minimising the extent of tube failure propagation caused by water leakage from a heat transfer tube of an LMFBR steam generator. Two different methods as follows were studied in this program : (1)The method to compare effective values between water leak sound and back ground noise using a single channel. (2)The method to detect and locate the leak using cross correlation signal processing of multi-channel. In the former one, it was estimated from acoustic signals obtained in the 50 MW Steam Generator Test Facility that the back ground noise levels of the Prototype and the Demonstration reactor were 0.0093G and 0.012G (G=gravity), respectively. The water leak rates equivalent to those back ground levels were evaluated as approximately 0.7 and 7 g/sec. In the latter one, first a detection and location software was developed in a off-line analysis, and secondly an on-line signal processing hardware was manufactured as a trial. In the off-line analysis, the influence of the internals on detection performance was examined by horizontal and vertical measurement. As the result, it revealed that back ground noise interfered the leak detection and location and that the potential depended on the leak positions even without noise. In the on-line analysis, leaks in a lower plenum were detectable with the same accuracy as the off-line analysis.

JAEA Reports

Micro-Leak Behavior on LMFBR Monju Steam Generator Tube Materials -Studies of Micro-Leak Sodium Reaction, 3

Kuroha, Mitsuo; Shimoyama, Kazuhito

PNC TN9410 86-027, 38 Pages, 1986/03

PNC-TN9410-86-027.pdf:4.66MB

Behavior of a micro-leak as an initiator of the leak propagation has been experimentally studied using leak nozzles made of the 2.25Cr-1Mo steel and Type 321 austenitic stainless steel which were selected as the heat transfer tube materials of the Monju steam generators. Twenty-nine micro-leak tests have been carried out in three stagnant sodium pots of the SWAT-4 test rigs installed in PNC/OEC. The main test parameters were the leak rate in the range of 10$$^{-5}$$ to 10$$^{-2}$$ g/sec and the sodium temperature in the range of 460 to 505$$^{circ}$$C. The shapes of initial leak holes used for the tests were a circular type and a slit one. The test results showed that the self-wastage was enhanced due to both corrosion and erosion, while the self-plugging was occurred due to precipitation of the reaction products at the sodium side and the corrsion products at the steam side. Post-test examination of the self-enlarged holes revealed that the diameters of holes were in the range of 0.3 t0 0.85 mm that were insensitive to the changes in the leak rate, the material, the temperature, and the shape of the initial leak hole. The relationships among the average leak rate L$$_{R}$$, the self-wastage rate S$$_{W}$$, and the sodium temperature T was derived from the experimental data and was expressed in the form of the following equation for the above two materials. S$$_{W}$$ = Exp (a + b ln L$$_{R}$$ - c/T) a,b,c: constant

JAEA Reports

Wastage tests on Monju superheater tubu material SUS321

*; *; Kuroha, Mitsuo

PNC TN9410 86-023, 112 Pages, 1986/03

PNC-TN9410-86-023.pdf:6.08MB

It is essential to clarify wastage behavior of a heat transfer tube in a sodium-water reaction in order to analyze a water leakage incident in a steam generator of LMFBR Monju. There fore wastage tests in small and intermediate leak ranges were conducted for austenitic stainless steel JIS $$cdot$$ SUS321 of a Monju superheater tube material by use of Small Leak Sodium-Water Reaction Test Loop (SWAT-2) and Large Leak Sodiam-Water Reaction Test Rig (SWAT-1). In the tests, a water leak rate, a distance from a leak nozzle to a target tube, and a sodium temperature were varied as empirical parameters. Test Results are as follows: (1)In the small 1eak range (0.1$$sim$$10g/sec), the wastage rate of SUS321 depends on L/D and has maximum value at L/D of 20 to 30 ; where L ls distance from the nozzle to the target and D is a nozzle diameter. Since the maximun wastage rate of SUS321 is about half as high as that of SUS304, SUS321 is more resistive against wastage than SUS304. (2)In the intermediate leak range (30 and 150 g/sec), the wastage rate depends on L/D and has a peak at L/D of 20$$sim$$50. The maximum wastage rate is quarter as high as that of 2%Cr-1Mo Steel. (3)Empirical formulas were derived from these test results concerning the relation between the wastage rate and the parameters.

JAEA Reports

Long-term thermo-hydraulic analysis in large-scale sodium-water reaction (Analysis of SWAT-3 Runs 4, 5, 6 and 7 by SWAC-13E); Large-scale sodium-water reaction analysis (Report No.14)

*; *; Kuroha, Mitsuo; *; *; *; *

PNC TN941 85-53, 144 Pages, 1985/03

PNC-TN941-85-53.pdf:3.01MB

SWAC13E is a one-dimensional thermo-hydraulic computer program to analyze large scale sodium-water reaction accidents in an LMFBR steam generator. The code is the advanced version of SWAC13, the long-term hydraulic analysis module of SWACS; the energy conservation is taken into consideration in the new version to add the function to analyze the temperature behavior of the reaction. The present document covers the validation study of the code by using the large leak data of the Steam Generator Safety Test Facility (SWAT-3). The analytical parameters are as follows: (1)Model of relative velocity. (2)Void/droplet density. (3)The number of nodes where water leaks. (4)Reaction heat. It is concluded that the code can analyze the phenomena with a reasonable conservatism by choosing the proper value of the parameters.

Journal Articles

PNC status report on leak detector development for LWFBR steam generators

Kuroha, Mitsuo;

IAEA/IWGFR Speciliasts Meeting, 0 Pages, 1983/11

Chemical and acoustic type leak detectors have been developed for detecting a small sodium-water reaction in an LMFBR steam generator.This paper presents a summary of the development.(1)Test results on PNC type in-sodium hydrogen meters including a description ofthe structure,the long-term reliability and the durability,and the improved meter with an office,(2)Development of in-cover gas hydrogen meters,(3)Hydrogen detection tests and analyses,(4)Operating experience of electrochemical in-sodium oxygen meters,and (5)Basic studies on acoustic characteristics of the sodium-water reaction.

JAEA Reports

Experiments of hydrogen behavior in the cover gas and sodium space of the LMFBR's SG; Studies of leak detector developments on LMFBR's SG (4)

*; Kuroha, Mitsuo; *; ; *; *

PNC TN941 82-98, 87 Pages, 1982/04

PNC-TN941-82-98.pdf:3.26MB

Various tests were conducted in SWAT-2 to investigate the hydrogen behavior both in the cover-gas and in the sodium by using the In-cover gas and the In-sodium hydrogen meters. The results of the hydrogen injection tests, the water injection tests and the bubble rising velocity measurment tests are presented in this report. The conclusions that were drawn from these tests are the followings. (1)Traveling velocity of hydrogen gas in the cover gas was dominated by the buoyancy and the convection rather than by the diffusion. (2)The background hydrogen partial pressure in the cover gas did not become lower than 10$$^{-2}$$ Torr. This lower limit was two to three orders of magnitude higher than the hydrogen pressure in sodium. (3)The experimentally determined bubble rising velocity in sodium agreed well with the theoretical prediction. (4)The detection rate of cover gas hydrogen that was not disolved into sodium was determined to be about 20Z, and that of hydrogen disolved into sodium to be about 80% with the present test vessel. (5)The hydrogen disolution rate during the water injection tests was higher than that of the hydrogen injection tests.

JAEA Reports

Study of micro-defect self-wastage phenomena on LMFBR prototype steam generators' tube; Studies of micro-leak sodium-water reactions (2)

Kuroha, Mitsuo; *; *; *; *

PNC TN941 82-101, 185 Pages, 1982/04

PNC-TN941-82-101.pdf:8.79MB

A series of micro steam leak tests has been carried out to accumlate the micro defect self-wastage data on LMFBR prototype MONJU steam generators' tubes, whose materials are planned to be 2.25Cr-1Mo and 321 stainless steels. The SWAT-2 test loop was used for the first stage of tests, and the SWAT-4 test rigs manufactured exclusively for the tests have been used since the following tests at the O-arai engineering center in PNC. The six test pieces, each of which had a micro crack nozzle through the thickness, were tested under the various conditions; the sodium temperatures were 470$$^{circ}$$C and 505$$^{circ}$$C, the steam pressure was about 130 kg/cm$$^{2}$$g throughout the sixt tests, and the average steam leak rates were in the range of 2$$times$$10$$^{-5}$$ to 4$$times$$10$$^{-2}$$ g/sec before the self-enlargements. The Main results obtained from the six tests are as followed; (1)It was obserbed in the case of 2.25Cr-1Mo steel nozzle that though the average leak rate of about 2 $$times$$10$$^{-5}$$ g/sec was very small, the leakage continued without any self-plugs until the self-wastage was completed through the thickness and finally increased the leak rate, which can cause wastage damage on adjacent tubes. (2)Self-wastage rates were apparently dependent on materials, steam leak rates and sodium temperatures. However, the increased leak rates were not strongly affected by those test parameters, and whose maxium values were apt to be in the range from to 10 g/sec in spite of the different test parameters. (3)The metallurgical structure of the stainless steel nozzles enlarged in size at the self-wasted sections due to the sodium-steam reactions, however, the composition did not changed at all. The surfaces of the self-wasted sections were rugged near the sodium and steam sides, but comparatively even midway between the two sides.

JAEA Reports

Preliminal study of micro-defect self-wastage on 2 $$frac{1}{4}$$Cr-1Mo steel nozzles for LMFBR steam generators; Studies of micro-leak sodium-water reactions (1)

Kuroha, Mitsuo; *; ; *

PNC TN941 80-135, 67 Pages, 1980/08

PNC-TN941-80-135.pdf:4.85MB

Experimental study on self-enlargement of a micro-defect was carried out using SWAT-2 test loop in order to establish the counter-plan for the micro-water leak in the LMFBR steam generators. The leak nozzles were made of 2.25Cr-1Mo steel which will be used as the heat transfer tube in the evaporator. The sodium temperature was fixed at 480$$^{circ}$$C and the initial leak rates were chosen within the range of 1.6$$times$$10$$^{-5}$$ g/s to 2.3$$times$$10$$^{-1}$$ g/s. Main results of the experiments are as follows: (1)The self-wastage rate S$$_{R1}$$ (mm/sec) is dependend on the water leak rate L$$_{R1}$$(g/sec), and the relation is expressed as following equation. S$$_{R1}$$ = 0.0173 L$$_{R1}$$$$^{0.58}$$ (2)The post test micrographs show that the self-enlargements started from the surface of sodium side, developed toward the water side, and finally the leak rates increased suddenly. (3)It is a characteristic of the self-wastaged holes that enlarged diameters of the sodium side are several times as large as that of the water side. (4)The minimum diameters of the enlarged nozzles were measured within 0.45 mm to 1.3mm, and the enlargement ratios of nozzle diameters increased as the water leak rates decreased.

JAEA Reports

Investigations of the vaccum system with a orifice in the hydrogen concentration meter; Studies of leak detector developments on LMFBR's SG

Kuroha, Mitsuo; *; ; *

PNC TN941 79-188, 58 Pages, 1979/10

PNC-TN941-79-188.pdf:2.06MB

One of PNC designed in-sodium type hydrogen meters which have been developed to use as the water leak detectors for the Monju's steam generators was modified in the vaccum system in order to measure the pumping speed of the ion pump during the operation. The modification was to increase the pressure drop between the ionization vaccum gauge and the ion pump by installing a orifice in front of the ion pump. As the result, it could be attained to estimate accurately the permeability factor K for the hydrogen through Ni membrane. The factor of K obtained is 1$$times$$10$$^{-4}$$ cm$$^{2}$$ Torr$$^{0.5}$$/sec, and the pressure dependence in the low pressure range is rather smaller than published datas. It is found through the disccusions that the relationship between P$$_{NH}$$ (partial pressure of hydrogen in sodium) and the measured pressure P$$_{N}$$ (pressure in vaccum side) is not expressed empirically by the half-power low, but followed by the pressure dependence of K reported. The best pair of the Ni membrane area and the orifice conductance is disccusd to dynamic vaccum system of the PNC designed in-sodium hydrogen meter. Also, it is concluded for the in-sodium hydrogen meter with a orifice using in the Monju's operation conditions that less than a quater of Ni membrane area adopted at present is usefull, and more than one year of the ion pump operation period without the change of the calibration curve can be attained.

JAEA Reports

Second phase test of the 1MW steam generator; Progress of operation, dismantling and cleaning

*; Morimoto, Makoto*; *; *; *; Kuroha, Mitsuo; *

PNC TN941 77-104, 101 Pages, 1977/06

PNC-TN941-77-104.pdf:3.89MB

After finishing first phase test, the 1MW Steam Generator was modified and second phase test was started from February 1973 and finished the test by June 1975. Then the steam Generator was dismantled, cleaned, and structural materials of the steam generator were externally inspected. This is to report the progress of operation of second phase test, repairing of facility and to describe the details of dismantling and cleaning of the steam generator addng the explanations of external inspection with photographs. Items clarified by the second phase test, followed by dismantling and cleaning were as follows : (1)Modifications based on the first phase test results were effective to improve the performance of the steam generator. (2)It was demonstrated that the cleaning of steam generator by flushing mixture of steam and argon gases was safe and effective. (3)During the external inspection after dismantling, they were observed slight deformations, and in way of the tube supports, wearing of tubes, which was supposed to be caused by fretting, was also observed.

Journal Articles

None

Kuroha, Mitsuo; ; ; Tanabe, Hiromi; Miyake, Osamu;

Joki Hasseiki No Anzensei Ni Kansuru IAEA Semmonka Kaigi, , 

None

13 (Records 1-13 displayed on this page)
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