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Journal Articles

Development of a fast reactor for minor actinides transmutation, 1; Overview and method development

Takeda, Toshikazu*; Usami, Shin; Fujimura, Koji*; Takakuwa, Masayuki*

Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.560 - 566, 2015/09

The Ministry of Education, Culture, Sports, Science and Technology in Japan has launched a national project entitled "technology development for the environmental burden reduction" in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmutation and volume reduction of minor actinides with long-lived radioactivity and high decay heat contained in high level radioactive wastes by using sodium cooled fast reactors. We are developing MA transmutation core concepts which harmonize efficient MA transmutation with core safety. To accurately design the core concepts we have improved calculation methods for estimating the transmutation rate of individual MA nuclides, and estimating and reducing uncertainty of MA transmutation. The overview of the present project is first described. The method improvement is presented with numerical results for a minor-actinide transmutation fast reactor.

JAEA Reports

None

Shimoyama, Kazuhito; Usami, Masayuki; Miyake, Osamu; ; ; Tanabe, Hiromi

PNC TN9450 97-007, 81 Pages, 1997/03

PNC-TN9450-97-007.pdf:1.72MB

None

JAEA Reports

Investigation on the sodium leak accident of Monju; Sodium leak test simulating the Monju leak

Shimoyama, Kazuhito; Nishimura, Masahiro; Usami, Masayuki; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi

PNC TN9410 97-085, 163 Pages, 1996/11

PNC-TN9410-97-085.pdf:6.17MB

Sodium fire experiments were carried out two times using the Sodium Fire Test Rig (SOFT-1) in the Power Reactor and Nuclear Fuel Development Corp (PNC) as a part of works to research the cause of the accident in secondary main cooling system of Monju. The purposes of these experiments are to confirm the leak rate and leakage form of sodium from damaged thermometer, to confirm the damage to the piping insulating structure around the thermometer and to the flexible tube, and to compare the temperature history of the signal from the thermometer between the experiments and Monju. In the experiments 56($$pm$$2)g/sec was obtained as the leak rate under the condition of ensuring the leakage pass in the simulated thermometer. This leak rate was corrected to 53g/sec to take account of manufacturing error of the theemometer between the experiment and Monju. In calculation of this leak rate, it is assumed that the annulus size of thermometer well tip is a nominal distance and pressure value to the leakage sodium is 1.65kg/cm$$^{2}$$G, which was the maximum one during the leakage of Monju. Concerning the leakage form, connection condition between the thermometer and flexiblc tube affected the dropping style of the leaking sodium especially in its initial behavior. For the connection condition of the thermometer and flexible tube at the beginning of the experiments, the first experiment was started removing the connection to simulate the post accident observation results of Monju, while the second one was started in connected condition. In the second experiment, the connection condition became to be equal with the initial state of the first experiment 17 seconds after the beginning of thc leak ; the cap nut which fixed the flexible tube to the elbow connector came off. Until the connection came off, the typical leakage form was the dispersion from the elbow connector as a droplet and the flow penetrating the covering of the flexible tube as a streamline, while after the ...

JAEA Reports

Wastage characteristics of high-chrome steel heat transfer tube; Intermediate leak wastage tests

Shimoyama, Kazuhito; Hamada, Hirotsugu; Tanabe, Hiromi; Usami, Masayuki

PNC TN9410 93-212, 134 Pages, 1993/09

PNC-TN9410-93-212.pdf:5.99MB

A one-through unit type steam generator (SG) having the Mod.9Cr-1MO Steel for its heat transfer tube is considered to be promising for the development of large FBR SGs. Wastage data of the tube material was already obtained for the micro-/small leak region as formerly reported. Therefore, intermediate leak wastage tests were conducted in the range from 10 g/s to around 200 g/s by using the SWAT-1 test facility and the test results are summarized as follows: (1)The wastage resistivity of the Mod.9Cr-1Mo steel is between that of 2.25Cr-1Mo steel and austenitic stainless steel; namely, the Mod.9Cr-1Mo steel has about half the of wastage rate of the 2.25Cr-1Mo steel. An experimental wastage formula in the intermediate leak region was derived from the test data. (2)Almost all of the wastage profile of target tubes was toroidal type and it became about half the cross section area of the 2.25Cr-1Mo steel. An experimental formula on initial leak diameters versus equivalent secondary failure diameters was derived in the intermediate leak region. These test results would be applied to failure propagation analysis code LBAP which is to be used for the design of a one-through unit type SG.

JAEA Reports

Wastage characteristics of welds of high-chrome steel heat transfer tube of steam generator; Micro-Leak and small leak sodium-mater reactions

Shimoyama, Kazuhito; ; Usami, Masayuki; Tanabe, Hiromi; Yoshida, Eiichi; *

PNC TN9410 91-288, 72 Pages, 1991/07

PNC-TN9410-91-288.pdf:2.3MB

For the development of a once-through unit type steam generator for an FBR, materials having a high mechanical strength at elevated temperatures and high resistivity against stress corrosion cracking (scc) are needed as its heat transfer tubes. High-chrome ferritic steels such as Mod.9Cr-1Mo, 9cr-2Mo, and 9Cr-1Mo-Nb-V steel are considered to meet these requirements. As previously reported, there was not large difference of resistivity among these materials and it was confirmed that high-chrome ferritic steels were more wastage -resistant than the 2.25Cr-1Mo steel which is a representative ferritic steel as tube materials. Since wastage data of weld portion at which the initial leak is apt to occur had been insufficient in order to totally evaluate the wastage resistivity of the high-chrome ferritic steels wastage tests in the micro-leak and small leak ranges were conducted for the tube-tube weld point of Mod.9Cr-1Mo steel. Major results are as follows : (1)In the micro-leak tests, there are not large differences in the self-wastage resistivity between the weld and the base metal. And the wastage resistivity is independant of the location of the leak starting point (a weld metal, a bond of weld, or a heat affected zone). (2)In the small leak tests, there are not large differences in the wastage resistivity between the weld and the base material. In the results, the empirical formula of wastage rate that derived from the base materials of the high-chrome ferritic steel can also be applied to the heat transfer tubes including the weld materials.

JAEA Reports

Integrity confirmation test for duplex-wall heat transfer tubes in case of its inner-wall leak

Hamada, Hirotsugu; *; Himeno, Yoshiaki; *

PNC TN9410 89-146, 85 Pages, 1989/08

PNC-TN9410-89-146.pdf:2.61MB

Steam wastage tests of the duplex-wall heat transfer tubes for the steam generators were conducted by placing its emphasis on the investigation of the possible occurrence of a subsequent failure on an outer-wall of the tube in case of its inner-wall leak. Based on the limitation from the test rig, the test tubes, each of which has an artificial crack on its outer-wall instead of on its inner-wall, were manufactured and were subjected to the test. In the test, a super-heated and pressurized steam as conceptual plant design or a nitrogen gas was fed to the tube and was impinged against the inner-wall through its crack for 24 hours which are conservative enough to evaluate test results. The test with a nitrogen gas was to obtain reference data. Before and after the test, equivalent hydraulic diameter of the crack was determined by measuring a pressure drop due to a flowing helium assuming that the crack can be regarded as an orifice. Then, changes in the equivalent diameter of the crack were determined. After the test, the tubes were subjected to the post-test metallurgical examination. Results thus obtained are as follows: (1)In all tubes, equivalent diameter of the cracks decreased after the test. Some cracks were even plugged by stream corrosion products. No enlargements of the crack, therefore, was found. (2)Post-test metallurgical examination showed no evidence of a steam wastage. Only steam corrosion products were found in the gap between the inner and outer walls. In conclusion, within the extent of the present test, failure possibility of the duplex-wall tube following a generation of an initial crack on an inner-wall is negligible.

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

Wastage properties of high-chrome steels as heat transfer tube material for steam generator; Small leak wastage tests

*; *; Himeno, Yoshiaki

PNC TN9410 88-129, 116 Pages, 1988/10

PNC-TN9410-88-129.pdf:10.05MB

From an economical point of view, a unit-type steam generator is being considered as the most promising one for a demonstration FBR. Material nominated for the heat transfer tubes of that type steam generator is such high chrome ferritic steels as 9Cr type steal. However, wastage data of the FBR steam generator are insufficient for high chrome steels, so that it is necessary to construct wastage database for selecting a tube material for the unit type steam generator. Therefore, small leak wastage tests were conducted for Mod.9Cr-1Mo, 9Cr-2Mo, and 9Cr-1Mo-Nb-V steels in SWAT-2 Test Loop. In the tests, (a)water leak rate, (b)leak nozzle to target distance, and (c)sodium temperature, were varied as empirical parameters. Test results are as follows: (1)There are not large differences in the resistivity to the wastage among the three 9Cr materials. (2)The high-chrome steels are twice more wastage-resistant than the 2.25Cr-1Mo steel. (3)Wastage rates of the high-chrome steels are about the same as those of austenitic stainless steels at the water leak rate below 0.5g/s. (4)A common empirical formula of the wastage rate is obtained for the three materials; Mod.9Cr-1Mo, 9Cr-2Mo and 9Cr-1Mo-Nb-V steels.

JAEA Reports

Basic test on sodium fire protection systems (4); Water simulation test of a leak from IHTS pipe

Himeno, Yoshiaki; *; *

PNC TN9410 86-088, 37 Pages, 1986/08

PNC-TN9410-86-088.pdf:3.3MB

In the safety evaluation of the Monju design basis sodium leak accident, a large leak from an IHTS pipe whose leak hole is equivalent to (1/4)$$cdot$$Dt (where, D is the pipe diameter and t is the pipe wall thickness) is postulated. To investigate the possible occurrence of a spray fire in the event of a sodium leak, a water simulation test has been conducted. Three test sections (straight, elbow, and T pipes) of the full-mockup Monju ITHS pipes with thermal insulation jackets (the jackets consisted of an inner and an outer jackets) were manufactured for this purpose. The test was divided into two parts. One was with the test sections equipped with their inner jacket only and the other with the test sections equipped with both inner and outer jackets. Subjects investigated in the test were (a) leakage flow pattern, (b) fraction of spray flow among a total leak flow, (c) droplet diameters of a spray flow, and (d) pressure drop coefficient across a leak hole and the thermal insulation jackets. From the test results, the following conclusions were drawn. (1)Test sections with the inner jackets only. More than 50% of a leak flow is in the form of a spray. Number mean diameter of the spray droplets determined is several millimeters. (2)Test sections equipped with both inner and outer jackets. A spray formation is almost perfectly suppressed. Since an integrity of the thermal insulation jackets during a leak has already been confirmed, the results indicates that a spray fire does not occur in the event of an real sodium leak. (3)Pressure drop coefficient across a leak hole and thermal insulation jackets Pressure drop coefficient determined is 2.3-2.8. While, the pressure drop coefficient used in the safety evaluation of the Monju sodium leak accident is 1.0. Comparison of these two numbers indicates that the safety margin of the Monju safety evaluation in regard to a sodium leak flow is about 50%. It is also Confirmed that the most part of the pressure ...

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

The Sodium-water reaction product removal test by use of cold trap; SWAT-3 RECT-II test

*; *; *

PNC TN941 85-127, 92 Pages, 1985/08

PNC-TN941-85-127.pdf:3.25MB

RECT-II (the Removal test of reaction products by cold trap) was conducted by use of SWAT-3 (the Steam Generator Safety Test Facility) at PNC in order to construct the post-accident operation of steam generators of the prototype FBR Monju and a larger plant following it. In prior to the test, some amount of the sodium-water reaction products (SWRP) generated in the water injection test (Run 18) was remained in the sodium system. An objective of the test is to confirm the purifying method to remove SWRP by hot sodium circulating through a cold trap (CT). A meshless type cold trap was selected to avoid choking by impurities and to enable efficient SWRP removal. RECT-II started on April 4, 1984 and terminated on April 26 when the plugging temperature decreased to 187$$^{circ}$$C. Major results obtained in the test are as follows: (1)Post-test observation revealed that the SWRP having remained at the bottom of the evaporator and the sodium outlet pipe were completely removed through the purification operation. (2)Hence, it is concluded that after the hot draining the SWRP of 14 kg-H$$_{2}$$0 remained in the sodium system out of that generated by the 42 kg-H$$_{2}$$0 injection and that almost all of the former was removed through the operation. (3)However, some amount of the hydrocarbon-oxide and SWRP in the slit articles simulating crevice and stagnant region still remained after the operation. Then it is concluded that it is insufficient to remove SWRP in crevice and stagnant region by the circulation of hot sodium. (4)A mass transfer coefficient of oxygen is evaluated as 2 $$times$$ 10$$^{-4}$$ [g/(mm H ppm)] if the cross section of the evaporator and inner surface of the 8 inch horizontal pipe are assumed to be the entire surface area of SWRP. (5)Since the choking of the cold trap degrades the efficient SWRP removal, it is essential to develop a cold trap which hardly chokes and easily regenerates even after choking; one of answers for this request is a ...

JAEA Reports

Design of electric power supply system for spent fuel storage facility in JOYO

*; *; *; *; *; *; Endo, Akira

PNC TN941 77-205, 45 Pages, 1977/11

PNC-TN941-77-205.pdf:1.22MB

The design of electric power supply system for the spent fuel storage facility was conducted in consideration of the following. (1)Insurance of safety condition of essential equipment and machines under the power failure. (2)stability power supply. (3)Exchangeability of electric appliances. This document was to be compiled with the design condition and its basis of electric power supply system for the spent fuel storage facility.

Journal Articles

None

; ; ; Usami, Masayuki

PATRAM '95 (PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIALS), , 

None

Journal Articles

None

Uruwashi, Shinichi; ; Usami, Masayuki;

PATRAM'95, , 

None

Oral presentation

Development of a comprehensive nuclear material accountancy system at JAEA

Takeda, Hideyuki; Usami, Masayuki; Hirosawa, Naonori; Fujita, Yoshihisa; Kodani, Yoshiki; Komata, Kazuhiro*

no journal, , 

no abstracts in English

15 (Records 1-15 displayed on this page)
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