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Saeki, Masakatsu
JAERI-Review 2004-011, 54 Pages, 2004/03
JAERI-Review-2004-011.pdf:5.21MB
In this review, experimental results on studies of radioactive iodine, tritium and neptunium are summarized. On studies of radioactive iodine, from various experimental results, the process establishing the formation mechanism of organic iodides was fully described. On the basis of formation mechanism, possibilities for the formation of organic iodides under nuclear reactor accidents and so on were also discussed. On studies of tritium, three topics were concisely described on the isotopic composition of tritium gases, the chemical forms and the diffusivity of tritium in various materials, and the sorption and desorption behaviors of tritium on surfaces of various materials. On the neptunium chemistry, the relationship between the structure of neptunium compound and the Moessbauer isomer shift and experimental results on neptunyl(VI) hydroxides were discussed.
Ohno, Shuji; Matsuki, Takuo*
JNC TN9400 2000-106, 132 Pages, 2000/12
Sodium fire analyses were performed on 7 kinds of sodium leak tests using a computer code ASSCOPS which has been developed to evaluate thermal consequences of sodium leak accident in an FBR plant. By the comparison between the calculated and the test results of gas pressure, gas temperature, sodium catch pan temperature, wall temperature, and of oxygen concentration, it was confirmed that the ASSCOPS code and the parameters used in the analysis give valid or conservative results on thermal consequences of sodium leak and fire.
Homma, Toshimitsu; Takahashi, Tomoyuki*; Yonehara, Hidenori*
JAERI-Review 2000-029, 83 Pages, 2000/12
JAERI-Review-2000-029.pdf:5.13MB
no abstracts in English
Noda, Kimio; Shinohara, Kunihiko; Kanamori, Masashi
JNC TN8410 2001-010, 35 Pages, 2000/10
JNC-TN8410-2001-010.pdf:3.85MB
We, the radiation control section of JNC have had two important experiences on the JCO critical accident and the JNC fire-explosion accident. Especially, at the critical accident in JCO, it was essential to take an action on the radiation protection activities for the evacuated neighboring inhabitants to the safety area. During the accident, we carried out the radiation protection activities, at the beginning of the accident, environmental monitoring of the surrounding area. Especially for the JCO accident, we took an action to terminate criticality, radiation shielding and monitoring, environmental monitoring, radiation survey of the residents, radiation monitoring of exhaust air.
; ; Ohno, Shuji;
JNC TN9400 2000-092, 247 Pages, 2000/08
JNC-TN9400-2000-092.pdf:20.29MB
Small-scale sodium pool combustion tests Run-F7-3 and Run-F8-1 were performed to investigate the corrosion of floor liner under high moisture condition. ln the both tests, which were performed using the 3m
FRAT-1 vessel at the SAPFIRE facility, the sodium of 507deg-C was leaked on the carbon steel catch pan about for 25 minutes with the flow rate of around 25 kg/h. The air in the vessel was ventilated with the flow rate of 5m/min containing the moisture of 25000-28000 vol.ppm. The duration of combustion was different in two tests by changing the starting time of argon gas injection into the vessel. As the results of post-test analysis such as observation of catch pan surface and chemical analysis of the deposits, it was confirmed that 'Na-Fe double oxidization type corrosion' was dominant in the both tests and that the catch pan and deposits were not under the condition leading to the occurrence of 'molten salt type corrosion.'
Tsukimori, Kazuyuki
JNC TN9400 2000-086, 103 Pages, 2000/08
JNC-TN9400-2000-086.pdf:3.67MB
ln order to evaluate the integrity of the floor liner of "MONJU" at sodium leakage accident, nonlinear finite element analyses have been conducted considering the effect of thinning of the liner due to molten salt type corrosion. Modeling by shell elements is appropriate since liner is composed of thin plates, however, it is difficult to deal with the very local strain behaviors. 0n the other hand, modeling by solid elements makes the numerical calculation impractical. If we extract the small part of the liner which includes local discontinuities, it is possible to evaluate local strain behaviors practically by using the solid element analysis model of the part. To realize this approach, the method to generate all the boundary displacements of the part model from the shell element analysis result of total structure is needed. The aims of this study are to develop the method to deal with the incompatibility between shell and solid elements at part model boundary, and to build numerical analysis circumstances including this method to make the detailed nonlinear finite element analyses of the floor liner of "MONJU" possible. Summary of the results is shown below, (1)The problem of the incompatibility between shell and solid elements was solved by introducing weighting function at 'T' and 'L' type corners and the interpolation function of 4-node rectangular plate bending element at the connection between liner plate and frame. (2)Software system was developed by using 'FINAS' and verified. (3)This approach was applied to one of the cases of the floor liner analyses of "MONJU" at sodium leakage accident. The analysis result showed that three-dimensional local strain behavior could be evaluated directly. ln addition, it was confirmed that the result by shell element analysis was conservative in evaluation of strain compared with that by solid element analysis in this case.
;
JNC TN9400 2000-060, 168 Pages, 2000/05
JNC-TN9400-2000-060.pdf:4.09MB
Optimal vertical isolation characteristics were studied for the structural concept of vertical seismic isolation system, which uses a common deck and a set of large coned dish springs. Four kinds of earthquake wave and three kinds of artificial seismic wave were used. The earthquake response analysis of a base isolated building was carried out considering some ground conditions and some vertical vibration characteristics of the building isolator. Floor response and acceleration time history at the vertical isolation level were arranged. Using the acceleration time history as a seismic input, the earthquake response analysis of the vertical isolation system according to single degree of freedom model was carried out. Linear analysis and non-linear analysis were made. ln the linear analysis, vertical isolation frequency was examined within 0.8 to 2.5 Hz, and damping ratio was examined within 2 to 60%. ln the non-linear analysis, it was examined within vertical isolation frequency 0.5 to 5Hz, which depended only on the rigidity of the coned disk spring, rigidity ratio of the damping devise 1 to 20 and yield seismic intensity of the damping devise 0.01 to 0.2. As the optimal vertical isolation characteristics of the system, the criterion of largest relative displacement, maximum acceleration and maximum value of the floor response acceleration between 5 to 12Hz was set, the combination region of the appropriate parameter were examined. ln case of largest relative displacement 50mm, acceleration response magnification of 0.75, floor response magnification of 0.33 were used as a criterion, from the result of the linear analysis, vertical frequency was set at 0.8 to l.2 Hz, and by combining the damping ratio over 20 %, it was proven that appropriate vertical isolation characteristics were obtained. The result of the non-linear analysis showed that the combination of the coned disk spring of vertical frequency 0.8 to 1.0 Hz and the damping element of rigidity ...
Kakehi, Isao; Nakabayashi, Hiroki
JNC TN9400 2000-051, 237 Pages, 2000/04
JNC-TN9400-2000-051.pdf:8.14MB
In this study, we have proposed the concept of safety systems (solutions of safety problems) in pyrochemical reprocessing systems (lt consists of pyrochemical reprocessing methods and the injection casting process for the metal fuel fabrication, or vibro-packing process for the oxide fuel fabrication.) which has different concept from the existing PUREX reprocessing method and pellet fuel fabrication process. And we performed its safety evaluations. FoIlowing the present Japanese safety regulations for reprocessing facilities, we pointed out functions, design requirements and equipments relating to its safety systems and picked up subjects. For the survey of safety evaluations, we first selected anticipated events and accident events, and second by evaluated 6the correspondence of the limitation of the public exposure to the accidents above, by using two parameters, the safety design parameter (the filter performance to confine radioactive matelials) and the leak inventory of radioactivities, and last by picked up its problems. ln addition to the above evaluations we performed basic criticality analyses for its systems to utilize these results for the design and evaluation of the criticality safety management system. Thus this study specified the concept of safety systems for pyrochemical reprocessing processes and then issues in order to establish safety design policies (matters which must consider for the safety design) and guides and to advance more definite safety design.
; ; *;
JNC TN9400 2000-043, 23 Pages, 2000/03
ln the feasibility study on commercialized fast breeder reactor (FBR) cycle systems conducted in JNC, it is required for candidate FBR plants that the level of safety should be enhanced so as to assure: (1)Comparative or superior safety level to that of light water reactors (LWRs), and (2)releaf of the public from anxiety about potential nuclear hazard. Adopting Passive safety characteristics is one of the measures. To attain the above safety objective, we considered implication of the basic safety principles for nuclear power plants that were created by the international nuclear safety advisory group of IAEA. The way to relieve from the anxiety was also taken into account. Then a definite safety objective was set from the standpoint of prevention of core disruptive accident (CDA). Furthermore, as a definite safety goal relating to reactor coresafety, elimination of re-criticality issues under CDA was set by considering characteristics of FBR in comparison with those of LWR. To examine measures for elimination of re-criticality issues, we developed a quick method to estimate possibility of re-criticality under CDA, by drawing a map about criticality characteristics under CDA in various degraded cores. Then hopeful measures were proposed for elimination of re-criticality issues in sodium-cooled FBR with mixed-oxide fuel. Molten fuel discharge behavior of their measures was preliminarily analyzed. We concluded that discharge capability of "a subassembly with an internal duct" was effective, and that "partial removal of axial blanket" was also effective as one of the measures though it has small effect on core performance.
*; Takemura, Morio*
JNC TJ9440 2000-005, 157 Pages, 2000/03
With use of the two-dimensional discrete ordinates code DORT and the standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the representative configurations in the Radial Shield Attenuation Experiment of the JASPER were performed. The results were compared with those obtained with use of traditional method DOT3.5/JSDJ2 for the previous JASPER experimetal analyses. In general, the change of the cross section library gives higher results and the change of the transport code gives lower results. Finally the new analysis method gives better agreement with the experimental results and also less deviations of calculational errors between various detectors. Experimental analyses for the thick concrete configulation in the Gap Streaming Experiment of the JASPER was also performed with the new analysis method, after solving the poor agreement found in last year with the original JASPER experimental analyses. The same tendency due to the library change was confirmed with the above mentioned analyses of the Radial Shield Attenuation Experiment. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments were continued through the above-mentioned experimental analyses and related informations were added for repletion of the database preserved in a computer disk holding previously accumulated data. Input data descriptions were made for auxiliary routines needed for the experimental analyses and their sample data were compiled and stored in the database.
Tamura, Kazuo*; Iriya, Yoshikazu*
JNC TJ9440 2000-004, 22 Pages, 2000/03
JNC-TJ9440-2000-004.pdf:2.35MB
In the probabilistic safety assessment(PSA), the fault tree/event tree technique has been widely used to evaluate accident sequence frequencies. However, event tansition which operators actually face can not be dynamically treated by the conventional technique. Therefore, we have made the dynamic analysis program(DYANA) for event transition for a liquid metal cooled fast breeder reactor. In the previous development, we made basic model for analysis. However, we have a probrem that calculation time is too long. At the current term, we made parallelization of DYANA usig MPI. So we got good performance on WS claster. It performance is close to ideal one.
*; *
JNC TJ9440 2000-002, 90 Pages, 2000/03
JNC-TJ9440-2000-002.pdf:1.43MB
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. In this work 9 sequences were analyzed and integrated into an input file for preparing the functions for DYANA using the analytical model and input data which developed for Super-COPD in the previous work. These sequences could not analyze in the previous work, which were categorized into the PLOHS (Protected Loss of Heat Sink) event.
Ohno, Shuji; Matsuki, Takuo*; ; Miyake, Osamu
JNC TN9520 2000-001, 196 Pages, 2000/01
JNC-TN9520-2000-001.pdf:5.13MB
ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature.
Ijima, Takashi; ; Matsumoto, Mitsuo; *
JNC TN3410 2000-002, 93 Pages, 2000/01
JNC-TN3410-2000-002.pdf:2.54MB
Fugen Nuclear Power Station ("Fugen") is a prototype Advanced Thermal Reactor (ATR), it has been demonstrated the plutonium utilization by loading many Mixed Oxide Fuels (MOX) since the reactor start up March 1979, and no fuel defect had been occurred, The MOX fuel assemblies has the high reliability and has been loaded more than 700 fuel assemblies. This is the largest in the world as a thermal neutron reactor. However, "Fugen" is planning to stop its operation in the year 2003, because the role of the Fugen almost finished. Therefore, we are going to summarize the ATR project including the Plutonium utilization experience. This paper is summarized as part of the experience.
Miura, Akihiko;
JNC TN8410 99-044, 189 Pages, 1999/10
This report includes several results that were made by calculation with several methods to clarify the cause of the fire and explosion incident. In the early times, we didn't have exact information of chemicaI property, reaction rate and any physical constants that we needed. But because the only data that indicate the cooling process of bituminized product was reported, we made heat-transporting calculation with taking this data. Based on the theory of the thermal hazard evaluation that was called Semenov theory or Frank-Kamenetskii theory, the amount of heat generation was estimated using the heat transporting calculation. Common theories were introduced in first section. In the second section, several results of heat transporting calculation were indicated. Calculations were made as follows. First, the model of bituminized product that was filled in the drum was created with the data of cooling process. Second, when the heat was generated in the drum, time-dependent temperature distribution was calculated. And last, judging from the balance of heat generation and heat radiation the critical heat rate was estimated.
Omori, Eiichi; ; ; ; Maki, Akira; Yamanouchi, Takamichi
JNC TN8410 2000-003, 93 Pages, 1999/10
JNC-TN8410-2000-003.pdf:4.92MB
Radioactive materials in aqueous solution at a nuclear fuel reprocessing plant causes radiolytic generalion of several gases including hydrogen. Hydrogen accumulating in equipment can be an explosion hazard. In such plants, though the consideration in the design has been fundamentally made in order to remove the ignition source from the equipment, the hydrogen concentration in the equipment should not exceed the explosion threshold. It is, therefore, desired to keep the hydrogen concentration lower than the explosion threshold by diluting with the air introduced into equipment, from the viewpoint which previously prevents the explosion. This report describes the calculation of hydrogen generation, evaluation of hydrogen concentration under abnormal operation and consideration of possible improvement at Tokai Reprocessing Plant. The amount of hydrogen generation was calculated for each equipment from available data on radiolysis induced by radioactive materials. Taking into consideration for abnormal condition that is single failure of air supply and loss of power supply, the investigation was made on the method for controlling so that the hydrogen concentration may not exceed the explosion threshold. Possible means which can control the concentration of hydrogen gas under the explosion threshold have been also investigated. As the result, it was found that hydrogen concentration of most equipment was kept under the explosion threshold. It was also shown that improvement of the facility was necessary on the equipment in which the concentration of the hydrogen may exceed the explosion threshold. Proposals based on the above results are also given in this report. The above content has been described in "Examination of the hydrogen produced by the radiolysis" which is a part of "Review of Design Data for Safety Assessment of Tokai Reprocessing Plant" (JNC TN8410 99-002) published in February 1999. This report incorporates the detail evaluation so that operation ...