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Fukaya, Masaaki*; Hata, Koji*; Akiyoshi, Kenji*; Sato, Shin*; Takeda, Yoshinori*; Miura, Norihiko*; Uyama, Masao*; Kaneda, Tsutomu*; Ueda, Tadashi*; Toda, Akiko*; et al.
JAEA-Technology 2014-040, 199 Pages, 2015/03
JAEA-Technology-2014-040.pdf:37.2MB
The researches on engineering technology in the Mizunami Underground Research Laboratory (MIU) project consists of (1) development of design and construction planning technologies, (2) development of construction technology, (3) development of countermeasure technology, (4) development of technology for security, and (5) development of technologies for restoration and/or reduction of the excavation damage. The researches on engineering technology such as verification of the initial design were being conducted by using data measured during construction as a part of the second phase of the MIU plan. Examination about the plug for reflood test in the GL-500m Access/Research Gallery-North as part of the development of technologies for restoration and/or reduction of excavation damage were carried out. Specifically, Literature survey was carried out about the plug, based on the result of literature survey, examination of the design condition, design of the plug and rock stability using numerical simulation, selection of materials for major parts, and grouting for water inflow from between rock and plug, were carried out in this study.
Imaizumi, Tomomi; Miyauchi, Masaru; Ito, Masayasu; Watahiki, Shunsuke; Nagata, Hiroshi; Hanakawa, Hiroki; Naka, Michihiro; Kawamata, Kazuo; Yamaura, Takayuki; Ide, Hiroshi; et al.
JAEA-Technology 2011-031, 123 Pages, 2012/01
JAEA-Technology-2011-031.pdf:16.08MB
The number of research reactors in the world is decreasing because of their aging. However, the planning to introduce the nuclear power plants is increasing in Asian countries. In these Asian countries, the key issue is the human resource development for operation and management of nuclear power plants after constructed them, and also the necessity of research reactor, which is used for lifetime extension of LWRs, progress of the science and technology, expansion of industry use, human resources training and so on, is increasing. From above backgrounds, the Neutron Irradiation and Testing Reactor Center began to discuss basic concept of a multipurpose low-power research reactor for education and training, etc. This design study is expected to contribute not only to design tool improvement and human resources development in the Neutron Irradiation and Testing Reactor Center but also to maintain and upgrade the technology on research reactors in nuclear power-related companies. This report treats the activities of the working group from July 2010 to June 2011 on the multipurpose low-power research reactor in the Neutron Irradiation and Testing Reactor Center and nuclear power-related companies.
*; Iida, Hiromasa; *; Adachi, Junichi*; ; Ebisawa, Katsuyuki*; *; Fukaya, Kiyoshi; *; *; et al.
JAERI-M 87-219, 336 Pages, 1988/01
no abstracts in English
Hiroi, Hiroshi*; *; *; *; *; *
PNC TN941 79-04, 274 Pages, 1979/10
Large leak sodium-water reaction tests have been carried out using SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype, LMFBR Monju is steam generator against large leak accident. This report gives the results of SWAT-3 run-5 test. The heat transfer tube bundle of the evaporator used in run-5 test was designed and fabricated by MITSUBISHI HEAVY INDUSTRIES, LTD. The water injection rate into the evaporator was 15 kg/sec, which corresponds to test scale of 5 tubes failure in actural size system according to iso-velocity modeling. Measurements were made of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 24.6 kg/cm
a nearest to injection point, and the maximun quasi-steady pressure in evaporator was 7.6 kg/cma. The rupture disc of evaporator was bursted at 0.23 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
Hiroi, Hiroshi*; *; Daigo, Yoshimichi; *
PNC TN941 79-155, 367 Pages, 1979/08
Large leak sodium-water reaction tests have been carried out using the SWAT-3 facility in PNC O-arai Engineering Center to obtain data on the safe design of the prototype LMFBR Monju's steam generator with reference to preventing large leak accident. This report gives the results of SWAT-3 run-7 test. The heat transfer tube bundle of the evaporator used in Run-7 test was designed and manufactured by TOSHIBA/IHI. Main purpose of this test is to clarify the sodium-water reaction phenomena occured in the downcommer region. Water was injected into the evaporator at the rate of 10.6 kg/sec, which corresponds to a test scale of 2 tube failure in an actural size system according to iso-velocity modeling. Measurements were taken of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 19.5 kg/cma closest to the injection point, and the maximum quasi-steady pressure in the evaporator was 5.8 kg/cma. The rupture disc of the evaporator burst 0.613 sec. after water was injected, and the pressure relief system functioned well. No secondary tube failure was observed.
Hiroi, Hiroshi*; *; *; *; *; *
PNC TN941 79-118, 282 Pages, 1979/06
Large leak sodium-water reaction tests have been carried out using the SWAT-3 facility in PNC O-arai Engineering Center to obtain data on the safe design of the prototype LMFBR Monju's steam generator with reference to preventing large leak accident. This report gives the results of SWAT-3 run-4 test. The heat transfer tube bundle of the evaporator used in Run-4 test was designed and manufaetured by TOSHIBA/IHI. Main purpose of this test is to clarify sodium-water reaction phenomena occured in the upper coil region, that is, the place near by sodium surface. Water was injected into the evaporator at the rate of 9.0 kg/sec, which corresponds to a test scale of 5 tube failure in an actural size system according to iso-velocity modeling. Measurements were taken of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 14.7 kg/cma closest to the injection point, and the maximum quasi-steady pressure in the evaporator was 5.4 kg/cma. The rupture disc of the evaporator burst 0.536 sec. after water was injected, and the pressure relief system functioned well. No secondary tube failure was observed.
*; Daigo, Yoshimichi; Miyake, Osamu; *; *
PNC TN952 78-07, 48 Pages, 1978/10
Water leak detectors-hydrogen concentration meters are equipped with secondary cooling system of LMFBR to detect small leak of water from heat transfer tube in steam generator. Leak rate region to be able to detect the water leaks before secondary tube failure is decided by using the concept of so-called "safety map". Computer cade "SWAC-10-MJ/1" provides the safety map for secondary cooling system of proto-type reactor. This paper is written for user's manual of the code. SWAC-10-MJ/1 code was made through the modification of original code "SWAC-10" which had been developed for 50MW Steam Generator Test facility. Major differences from SWAC-10 code are as follows ; (1) Secondary cooling system of proto-type LMFBR is selected for calculation object (2) Wastage rate equations proposed by PNC are accepted. (3) Effect of enlargement of initial water leak hole due to self-wastage is introduced. (4) Hydrogen diffusion process in Ni membrane of hydrogendetector is introduced.
Kanegae, Naomichi*; *; Daigo, Yoshimichi; *; *
PNC TN941 78-91, 214 Pages, 1978/10
Seven PNC Type In-Sodium Hydrogen Meter Type-II were designed and manufactured to confirm that these meters will be applicable to the "MONJU" plant, and they have been installed and operated without any troubles in several testrigs of O-arai Engineering Center, Power Reactor & Nuclear Fuel Development Corporation. These meters were improved from the Type-I reported previously, and a new vacuum system named as "Separation Type of Dynamic and Static Chamber" was developed and applied to the Type-II. In this new system, the dynamic and the static chamber are separated from each other in the vacuum system, so that the operation method become very simpler and it becomes possible to calibrate these meters with more accurately and shorter time. This paper explains the basic design specifications, the detail constructions and the test results of in-sodium tests including the calibration and the response characteristics in case of sodium-water reactions were occured. The following results were obtained. (1)The separation type of dynamic and static chamber is effective to simplify the operation method and to improve the accuracy of calibration of hydrogen meter. (2)The optimum design method of hydrogen meter reported preveously was established by several experiences of designing, manufacturing and operating of the Type-II and also the Type-I. (3)Several infomations applicable to the hydrogen meter of the MONJU plant were obtained, and a prospect that this Type-II hydrogen meter will be able to be used in MONJU plant was obtained. Now, long term operation test in sodium is under conducting for confirming life time or characteristics of operation time dependency of the Type-II meters in O-arai Engineering Center.
*; *; Hiroi, Hiroshi*; *; *; *
PNC TN941 78-154, 210 Pages, 1978/10
Large Leak sodium-water reaction tests have been carried out using SWAT-3 facility in PNC O-arai Engineering Center to obtain the data on the safe design of the prototype LMFBR Monju's steam generator against large leak accident. This report describes the resulting data of Run-6 test. The heat transfer tube bundle of the evaporator used was fabricated by HITACHI/BABCOCK HITACHI. The water injection rate into the evaporator was 9.4 kg/sec, which corresponds to test scale of 5.7 tubes failure in actual size system according to iso-velocity modeling. Pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, etc were measured during water injection test. Initial spike pressure was 12 kg/cma nearest to injection point, and the maximun quasi-steady pressure in evaporator was 5.1 kg/cma. The rupture disc of evaporator was bursted at 0.58 sec after water injected, and the pressure relief system was well functioned. No secondary tube failure was observed.
*; *; *; Hiroi, Hiroshi*; *; *
PNC TN941 77-91, 89 Pages, 1977/01
To obtain the data for the safe design of the steam generator system of the prototype fast reactor Monju against the postulated large leak sodium-water reaction, experiments have been carried out using SWAT-3 test facility. In this report in order to estimate the behaviour of the forces acting on the pressure relief pipes during the large leak sodium-water reaction, the results of the SWAT-3 Run-3 test concerning the sodium flow in the pipes and resulting forces were arranged and examinedi Results are as follows ; (1)Sodium expulsion nearly finished within three soconds after the rupture of the disc in the pressure relief pipes. During this term the flow was two-phase consisted of iquid sodium and hydrogen/cover gas, and the maximum pressure was 4.4 kg/cma, the minimum void fraction 0.75, the maximum sodium flow rate 600 kg/sec. (2)Forces exerted on the piping are such as (a)force due to movement of a evaporator (b)Fluid-induced force associated with the momentum change through a bend (c)force due to movement of a pressure relief tank (d)force due to thermal expansion of the piping (3)The change of the force (2) - (a) is analogous to step function. (4)Maximum value of the force (2) - (b) is 2.9 ton. (5)The force (2) - (c) is consisted of two components with frequency of 2.5 Hz and 3.1 Hz.
*; *; *
PNC TN941 77-82, 86 Pages, 1977/01
Pressure wave propagation in a FBR secondary loop was studied using l/12.5 scale model of the prototype LMFBR "MONJU" for the purpose of establishing the analysis method for the pressure wave propagation through FBR secondary circuit in case of a sodium-water reaction. The test were conducted using water as the coolant fluid instead of sodium. The first wave of input pressure had 2.5 
7 msec rise fime and 2 7 kg/cm peak value. The test results were compared with the calculation results by SWAC-5K and SWAC-5H codes which were developed for analyzing the pressure wave propagation in a FBR secondary circuit. Following conclusions were obtained from this study; (1)The agreements between the experimental results and calculated results of the transmitted pressure were good in an the measuring points except in and near IHX. (2)In and near IHX, the calculated wave forms were similar to the experimental results, but the calculated peak values were higher than the experimental results, i.e. calculated results give safe side values. (3)It would be appropriate to analyse the pressure wave propagation of "MONJU" secondary circuit by SWAC-5K or SWAC-5H code. The calculation models and ideas used in this report could also be applied to the "MONJU" analysis.
*; *; *
PNC TN941 77-44, 57 Pages, 1977/01
Pressure wave propagation in and through a steam generator was studied using 1/12.5 scale model of the MONJU steam generator for the purpose of establishing the analyzing method of pressure-wave propagation through the FBR secondary circuit in case of a sodium-water reaction. The tests were conducted using water as the coolant flnid instead of sodium. The first wave of input pressure had 1.5 msec rise time and 2.27.0 kg/cm peak value. The test results were compared with the calculation results by SWAC-5H and SWAC-5K codes which were developed for analyzing the pressure wave propagation in an FBR secondary circuit. Following conclusions were obtained from this study; (1)Peak value of the first wave was decreased from that of the input wave to about 1/15 in the lower plenum of SG and to about 1/30 in the upper plenum of SG. (2)The agreements between the experimental results and calculated resutls were fairly good as a whole. (3)Following models were used for the calculation of pressure propagation in SG; (a)One-dimensional variable cross-section model (b)Average cross-section model (c)Down-comer cross-section model (Omitting the tube bundle path and center pipe path) All of the above models showed a fairly good agreement with the experimental results and there were not much differences between the calculated results by each model. (4)The above calculation models can be applied for the calculation of pressure wave propagation in and through a superheater and reheater which have the similar configuration with this steam generator model. (5)The SWAC-5 codes can be used for the analysis of pressure wave propagation in the actual FBR secondary circuit.
*; Hiroi, Hiroshi*; *; *
PNC TN941 77-170, 73 Pages, 1977/01
For the analysis of largc scale sodium/water reaction accident a computer program, SWAC-13 is developed to predict the behaviour of quasistatic pressure and sodium flow in the secondary cooling system composed of steam generators (EV, SH), IHX, and pipes, and pressure relief system composed of reaction product tank and pipes using network model by implicit method. Here reportted are modeling, the method of calculation, an outline of this code as users' manual and an example of simulation of SWAT-3 test. This code is written in JIS-FORTRAN computer language and requires 190 k bites core memories. It takes about 50 minutes for FACOM-230/58 to compute 10,000 steps for 29 links.
*; *; *; *; *; *; *
PNC TN941 76-84, 63 Pages, 1976/08
PNC and CRIEPI have jointly perfomed a simulation experiment for pressure wave propagation in LMFBR steam generators during sodium/water reaction. The purpose of this work was to contribute to the safe design of the steam generators and from the results of this project the following has been established: (1)Analytical method has been devised to obtain the equivalent cross-sectional area for a system with inner structures, and (2) Structural dynamic response of the shell has been measured for given pressure wave form. By the simulation experiments, the equivalent cross-sectional areas have been established for Monju SG in its various design alternatives and, moreover, verification was made of the equivalent cross-sectional area for straight pipe intervals. In item (2) above, dynamic strain of the vessel wall was measured for the given simulated pressure pulse.