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grade large tanksHayakawa, Masato; Shimoyama, Kazuhito; Miyakoshi, Hiroyuki; Suzuki, Shigeaki*
JAEA-Technology 2021-027, 33 Pages, 2022/01
JAEA-Technology-2021-027.pdf:3.64MB
At the Oarai Research and Development Institute of the Japan Atomic Energy Agency, experimental studies in various sodium environments are being conducted in connection with the research and development of sodium-cooled fast reactors such as the experimental fast reactor Joyo and the prototype fast reactor Monju. The dismantling of sodium test facilities and equipment that have achieved their purpose has been carried out sequentially, and a wealth of experience and technology has been accumulated. On the other hand, a large amount of metallic sodium used for research and testing is being reused for new testing facilities, and the large sodium tanks that contained the metallic sodium are being dismantled. In order to dismantle these tanks safely and efficiently, it is important to reduce the residual sodium inside the tanks (especially at the bottom) as much as possible before dismantling. Therefore, we have been working on the reduction of residual sodium at the bottom of several large sodium tanks of 100 m class. This report describes the technologies and experiences related to the reduction of residual sodium that have been carried out so far.
Hatakeyama, Nozomi*; Miura, Ryuji*; Miyamoto, Naoto*; Miyamoto, Akira*; Ara, Kuniaki; Shimoyama, Kazuhito; Kato, Atsushi; Yamamoto, Tomohiko
Journal of Computer Chemistry, Japan, 21(2), p.61 - 62, 2022/00
no abstracts in English
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12
Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.
Imamura, Hiroaki; Suzuki, Masashi*; Shimoyama, Kazuhito; Miyakoshi, Hiroyuki
JAEA-Technology 2019-005, 163 Pages, 2019/06
JAEA-Technology-2019-005.pdf:25.24MB
For the R&D of safety enhance in future fast reactor development, the constructed the large sodium test loop (mother loop) in advanced technology experiment sodium facility (AtheNa) was completed. The sodium test loop possesses the largest capacity of about 240 tons of the world's largest sodium and can supply impurity-controlled high temperature sodium to large structural and technology demonstration test sections. It is greatly expected as R&D such as future international cooperation. For the purpose of future R&D tests, this report compiled the design specifications, fabrication and performance confirmation results of sodium test loop.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Kikuchi, Shin
Nihon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00382_1 - 17-00382_11, 2018/03
Wastage on adjacent tubes (target-wastage) arise from water/steam leak in steam generators of sodium-cooled fast reactors (sodium-water reaction). Target-wastage is likely to be caused by liquid droplet impingement erosion (LDI) and Na-Fe composite oxidation type corrosion with flow (COCF) in an environment marked by high temperature and high-alkali (reaction jet) due to sodium-water reaction. In the previous study, the authors quantitatively evaluated the effect of material temperature and fluid velocity on COCF rate, and revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay. In this study, the applicability of new wastage correlations was confirmed for each tube in sodium-water reaction test with straight vertical tube bundle under practical steam generator operation condition. The authors established that the new wastage correlations were applicable to each tube of tube bundle in the above test, and the time progress of wastage was qualitatively investigated for the two penetrated tubes in the period including the water and/or steam blowdown.
Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu
JAEA-Technology 2017-018, 70 Pages, 2017/08
JAEA-Technology-2017-018.pdf:9.67MB
In case of the water leak into sodium in a SG of SFRs due to tube failure, reaction jet is formed by sodium-water reaction with exothermic heat. The reaction jet forms highly alkaline environment with high temperature and high pressure, which cause local thinning of adjacent heat transfer tubes (target wastage). In this report, for the purpose of elucidation of target wastage, the authors developed the experimental apparatus and experimental technique which enable the separate evaluation of wastage influence factors, including temperature, impingement velocity, reagent ratio and so on by using high temperature sodium hydroxide as major reaction product and sodium monoxide as secondary reaction product. In addition, the impingement corrosion experiments have been conducted by using high temperature reagents (NaOH and Na
O). Based on the corrosive data, authors quantitatively evaluated the influence factors of wastage and formulated the average corrosive equations.
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
JAEA-Technology 2016-030, 50 Pages, 2016/12
JAEA-Technology-2016-030.pdf:5.22MB
In case of tube failure of a steam generator in sodium-cooled fast reactors, the reaction jet with high temperature and high velocity under highly alkaline environment is formed by cited exothermic reaction (sodium-water reaction). When the high temperature reaction jet covers the adjacent tubes, the material strength of tube decreases in the high temperature condition, and the adjacent tube may be swollen and failed by inner pressure (overheating tube rupture). For evaluation of the overheating tube rupture, tube failure is judged by comparison the hoop stress loaded by inner pressure with stress strength standard defined as creep strength depending on tube temperature. Thus, it is important to confirm the validation of this failure criterion based on the findings obtained in the simulated experiment of overheating tube rupture. In this report, for consideration on the validation of the failure criteria and elucidation on the failure mode and strength characteristics of failure, the authors carried out the rapid heating rupture experiment for the thin single and double-walled 9Cr steel tubes at high temperature up to 1500 K by using TRUST-2 rig in the Japan Atomic Energy Agency.
Kurihara, Akikazu; Umeda, Ryota; Kikuchi, Shin; Shimoyama, Kazuhito; Ohshima, Hiroyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 14(4), p.235 - 248, 2015/11
Sodium-water reaction would take place due to a breach of heat transfer tube in steam generator (SG) of sodium-cooled fast reactor (SFR), and the reaction jet may cause wear to the neighboring tubes by thermal and chemical effects, which is so-called target-wastage. Accordingly, failure propagation caused by target-wastage may potentially detract the secondary cooling system integrity. In previous study, a great number of target-wastage experiments have been carried out for candidate materials under practical SG operation conditions. Target-wastage rate was derived from macroscopic boundary factors of reaction jet. However, this mock-up approach is not versatile, and does not befit for large-scale SG design. Therefore, target-wastage should be focused for safety assessment of the various SG design. In this study, experiment apparatus and technique on composite oxidation type corrosion with flow (COCF), which is integral part of target-wastage, were constructed to figure out the separation effect of local wastage factors under the high temperature sodium hydroxide (NaOH) and sodium monoxide (NaO) environment mainly generated by SWR. The authors quantitatively evaluated the effect of material temperature and fluid velocity on COCF rate, and diffusion coefficient of Mod.9Cr-1Mo steel into NaOH-NaO. Besides, it was revealed that COCF was sodium-iron composite oxidation type corrosion from metallographic observation and element assay.
Abe, Yuta; Shimoyama, Kazuhito; Kurihara, Akikazu
JAEA-Technology 2014-026, 40 Pages, 2014/07
JAEA-Technology-2014-026.pdf:33.12MB
In case of water leak from a penetrated crack on a tube of steam generator in the sodium cooled fast reactor (SFR), self-wastage, that increases the size of leak, may take place by corrosion related to chemical reaction between sodium and water. For the safety evaluation of the accident, JAEA has been developing the analytical method of self-wastage using the multi-dimensional sodium-water reaction code. This report describes the development of new experimental rig (SWAT-2R). SEAT-2R enables to examine corrosion effecting factors that were ambiguous in the previous studies. The report includes description of development of micro-leak test piece, examination of experimental procedure. The results will provide fundamental data for validation of the self-wastage analytical method.
Kurihara, Akikazu; Umeda, Ryota; Shimoyama, Kazuhito; Abe, Yuta; Kikuchi, Shin; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu, B, 79(808), p.2640 - 2644, 2013/12
Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively the heat transfer coefficient between reaction jet and adjacent tubes which is one of the major influencing factor. The authors carried out the sodium-water reaction test (SWAT-1R) under the simulated operation condition of a real plant, and measured the correlation between heat transfer coefficient and void fraction around an adjacent tube. The authors confirmed that thermal environment around an adjacent tube was inferable from measured data, and heat transfer correlation equation proposed by Hamada et al. was applicable to the operation condition at elevated pressure and temperature.
Kurihara, Akikazu; Ohshima, Hiroyuki; Shimoyama, Kazuhito; Umeda, Ryota
Nihon Kikai Gakkai Rombunshu, B, 77(776), p.964 - 968, 2011/04
Sodium reacts chemically with water in case of unexpected heat transfer tube failure in a steam generator (SG) of sodium-cooled fast breeder reactors (FBRs) and exothermic reaction produces reaction field with high temperature and high corrosive action (sodium-water reaction). Adjacent tubes are damaged due to erosive and corrosive environment of the reaction field (wastage). Therefore, it is integral to evaluate such sodium-water reaction phenomena with high accuracy for the safety assessment of FBRs. For the purpose of understanding the wastage mechanism, an experiment was carried out in which sodium hydroxide (NaOH) as the main reaction product collided with the tube material under high temperature conditions simulating the reaction field. We confirmed that the erosion-corrosion rate of tube material has a tendency to increase as the temperature and velocity of NaOH are raised.
Shimoyama, Kazuhito
JNC TN9410 2004-009, 46 Pages, 2004/03
JNC-TN9410-2004-009.pdf:4.08MB
In the water leak accident of a steam generator designed for a sodium cooled reactor in the Feasibility Study, the localization of tube failure propagation by using an advanced water leak detector will be required from the viewpoints of the safety and economical efficiency of the plant. So far, the conventional knowledge and analytical tools have been used in the investigation and evaluation of water leak phenomenon; nevertheless, there was neither test data nor the study of quantitative evaluation on the corrosion behavior, so-called wastage-resistant characteristics, of 12Cr steel tube material in sodium-water reactions. Wastage tests for the 12Cr steel tube material were conducted in small water leaks by use of the Sodium-Water Reaction Test Rig (SWAT-1R), and the data of wastage rate were obtained in the parameter of water leak rate under the constant sodium temperature and distance between leak and target tubes. The test results lead to the following conclusions: (1) The wastage-resistibility of 12Cr steel is 1.6 times greater than that of 9Cr steel and is 2.7 times greater than that of 2.25Cr-1Mo steel. (2) The wastage-resistibility of 12Cr steel increases in smaller water leaks; especially in water leak rates of 1 g/sec or less, it is more excellent than that of SUS321 stainless steel used as Monju superheater tube material. (3) Based on the correlation of wastage rate for the 9Cr steel, the correlation for the 12Cr steel has been obtained to be used for the evaluation of tube failure propagation. As the correlation of wastage rate for the 12Cr steel is based on the correlation for the 9Cr steel, it gives enough conservatism in smaller water leaks. To serve in accurately evaluating the tube failure propagation in smaller water leaks, it is necessary to obtain new correlation of wastage rate for the 12Cr steel based on the data in the wide range of water leak rates.
Nishimura, Masahiro; Shimoyama, Kazuhito; Kurihara, Akikazu; Seino, Hiroshi
JNC TN9400 2003-014, 167 Pages, 2003/03
JNC-TN9400-2003-014.pdf:11.65MB
Sodium-water reaction tests are carried out using the Sodium Water Reaction Test Rig.(SWAT-1R). The objective of this study is to obtain the experimental knowledge of following items. (1)Thermal characteristic of sodium-water reaction jet (2)Effects on heat transfer tubes by high temperature reaction jet Those are necessary to establish a systematic evaluate method for the failure propagation of FBR steam generator heat transfer tubes. Then we performed parametric three HT tests (series of test confirm thermal influence on tubes), and the parameter is water leak rate. Following items are clarified in this report. (1) (a)No difference was observed about the maxim temperature of reaction jet in each test. HT-1:1161
C, HT-2:1013C, HT-3:1164C (b)The reaction jet became stable within ten seconds, and the jet size depended on water leak rate. (2) (a)The thermal data for calculating the heat transfer coefficient were obtained and it's evaluation method was established. (b)The thermal influence of heat transfer tubes suffered by more than 1100C high temperature jet in 20 seconds are almost equal to that of standardization test materials suffered by 900C in 20 seconds.
Shimoyama, Kazuhito; *
JNC TN9410 2001-020, 105 Pages, 2001/08
JNC-TN9410-2001-020.pdf:7.59MB
The Steam Generator Safety Test Facmty, SWAT-3, was constructed in order to confirm the safety of the steam generator of a fast reactor against a water leakage from a heat transfer tube. And to clarify the range of the part in which the repair is required before restart by grasping the damage of the steam generator. This report describes the technical experience of extraction of sodium and removal of sodium-water reaction products from the SWAT-3, which was disassembled in 1996 through 1999. It was the first experience in Japan to disassemble the large-scale sodium facilities containing reaction products. The disassembly work was performed safely and efficiently as planned procedures. Especially, through the disassembly of the dump tank that contained a large amount of reaction product deposits, numerous knowledge and experience were obtained and some special disassembly techniques were developed. Main results are introduced in the following (1)To prevent the blockage of the piping by sodium-water reaction products in the cover gas space division in the tank, it was clarified that the double piping with the electric heater was effective. (2)Welding Technique between a pipe and a sodium tank surface was established. (3)It was verified that the fusion cutting of tank main body was possible by covering sodium-water reaction products with dry sand. No significant corrosion was observed in the structural material of the dump tank whose internal surfaces had been contacted with the reaction products for a long time. Even small scratches or marks during manufacturing procedure were observed to remain in the inspection. The obtained knowledge and experience will provide useful information for planning and actual works of the similar procedures and safety management relating to future disassembly of the present sodium-water reaction test facilities and sodium loops containing certain impurities, also for planning of post accident procedures of FBR plants.
Shimoyama, Kazuhito; Usami, Masayuki; Miyake, Osamu; ; ; Tanabe, Hiromi
PNC TN9450 97-007, 81 Pages, 1997/03
None
Shimoyama, Kazuhito; Nishimura, Masahiro; Usami, Masayuki; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi
PNC TN9410 97-085, 163 Pages, 1996/11
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(
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
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 ...
Miyahara, Shinya; Sagawa, Norihiko; Shimoyama, Kazuhito
Journal of Nuclear Science and Technology, 33(2), p.128 - 133, 1996/00
None
Nakagiri, Toshio; ; Ohno, Shuji; ; *; Koyama, Shinichi; Shimoyama, Kazuhito
PNC TN9510 94-001, 246 Pages, 1994/05
None
Shimoyama, Kazuhito; Hamada, Hirotsugu; Tanabe, Hiromi; Usami, Masayuki
PNC TN9410 93-212, 134 Pages, 1993/09
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.
