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Shiina, Yasuaki; Komaki, Katsuya*; Tanaka, Gaku*; Hishida, Makoto*
JAERI-Tech 2001-076, 49 Pages, 2001/11
no abstracts in English
Endo, Hideo; Seki, Masayuki; ; *; *
JNC TN8410 2000-007, 89 Pages, 2000/03
JNC-TN8410-2000-007.pdf:6.28MB
(1)Outline of examination. Various test specimens were made to evaluate and confirm the weld strength properties of the oxide dispersion strengthened (ODS) cladding tube material (martensitic and ferritic steel), which had been manufactured in JFY 1997. The examination consisted of tensile tests (RT,650
C, 700C, 800C), internal pressure creep tests, internal pressure burst tests, and a rapid heating burst tests. (2)Examination results. The results of the tensile tests are as follows: (ferritic and martensitic) (a)All test specimens from RT to 700C failed in the tube. The weld zones had not failed. (b)The test specimens at 800C failed in the weld zones. There was little elongation. (ferritic) (a)The weld zone had fine grain structure and carbide precipitates. (martensitic) (a)Carbide had precipitated in the weld zone. From these results, the strength of weld zone decreased extremely at temperatures exceeding the endurance limit (700C) All of the internal pressure burst test specimens and the rapid heating burst test specimens failed in the tube and not the weld zone. (3)The quality assurance method of the test specimens. The weld reliability of the test specimens were confirmed by the process control of the welding conditions, by using control test specimens, and ultrasonic testing. Confirmation of the process control of the welding conditions; current wave, the voltage waveform, the accelerogram, and the displacement ripple in the welding process was recorded to assure an abnormal value had not occurred. (Process control of welding condition) The results the current waveform, voltage waveform, accelerogram, and the displacement waveform were excellent. (test specimens) The weld joint was excellent based on metallography examination. (Ultrasonic testing) The length of the weld joint was measured and found to be adequate. The reliability the weld joint can be assured by using the above-mentioned method.
Kimura, Nobuyuki; Tokuhiro, Akira; Kamide, Hideki
JNC TN9400 2000-027, 181 Pages, 2000/02
JNC-TN9400-2000-027.pdf:10.28MB
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing among jets causes thermal fatigue in structural components, is of importance for reactor safety. ln this study, a water experiment was performed using vertical and parallel triple jets, those are cold jet on center and hot jets on both side. The experimental parameter was discharged velocity of the triple-jet and local temperature and velocity were measured by ultrasound Doppler velocimetry and movable thermocouples. The objective is a quantification of the mixing process in the multiple-jet. Under isovelocity condition, the jets oscillated periodically and mixing among thejets was promoted by periodic oscillation. The periodic oscillation was dependent on the Strouhal number based on the discharged velocity. Under non-isovelocity condition, on the other hand, the jets did not oscillate periodically and mixing among the jets progressed more gentle compared with the case under isovelocity condition. The tempwrature fluctuation could be decomposed into coherent and random components using the phase averaging process. The rate of the coherent component in the temperature fluctuation increased and the rate of random component in temperature fluctuation decreased in proportion as the discharged velocity was increased.
; kasahara, Naoto; ; ; Kamide, Hideki
JNC TN9400 2000-010, 168 Pages, 2000/02
JNC-TN9400-2000-010.pdf:8.78MB
Thermal striping is significant issue of the structural integrity, where the hot and cold fluids give high cycle fatigue to the structure through the thermal stress resulted from the time change of temperatur distibution in the structure. In the sodium cooled fast reactor, temperature change in fluid easily transfers to the structure because of the high thermal conductivity of the sodium. It means that we have to take care of thermal striping, The thermal striping is complex phenomena between the fluid and structure engineering fields. The investigations of thermal striping are not enough to evaluate the integrity directly. That is the fluctuation intensity at the structure surface is assumed to be temperature difference between source fluids (upstream to the mixing region) as the maximum value in the design. 0therwise, the design conditions are defined by using a mockup experiment and large margin of temperature fluctuation intensity. Furthermore, such evaluation manners have not yet been considered as a design rule. Transfer mechanism of temperature fluctuation from fluid to structure has been investigated by the authors on the view points of the fluid and structure. Attenuation of temperature fluctuation was recognized as a dominant factor of thermal fatigue. We have devdoped a numerical analysis system which can evaluate thermal fatigue and crack growth with consideration of the attenuation of temperature fluctuation in fluid, heat transfer, and structure. This system was applied to a real reactor and the applicability was confirmed. Further verification is planned to generalize the system. For the higher cost performance of the fast reactor, an evaluation rule is needed, which can estimate thermal loading with attenuation and can be applied to the design. An idea of the rule is proposed here. Two methods should be prepared; one is a precise evaluation method where mechanism of attenuation is modeled, and the other is simple evaluation method where ...
Yang Zumao*;
JNC TN9400 2000-009, 81 Pages, 2000/02
JNC-TN9400-2000-009.pdf:47.3MB
It is important to study thermal stratification and striping phenomena for they can induce thermal fatigue failure of structures. This presentation uses the AQUA code, which has been developed in Japan Nuclear Cycle Development Institute (JNC), to investigate the characteristics of these thermal phenomena in water, liquid sodium, liquid lead and carbon dioxide gas. There are altogether eight calculated cases with same Richardson number and initial inlet hot velocity in thermal stratification calculations, in which four cases have same velocity difference between inlet hot and cold fluid, the other four cases with same temperature difference. The calculated results show : (1) The fluid's properties and initial conditions have considerable effects on thermal stratification, which is decided by the combination of such as thermal conduction, viscous dissipation and buoyant force, etc., and (2) The gas has distinctive thermal stratification characteristics from those of liquid because for
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.
; Ohno, Shuji; Miyake, Osamu; ; Seino, Hiroshi
PNC TN9520 97-001, 185 Pages, 1997/12
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 as the user's manual of ASSCOPS version 2.0. ASSCOPS is an integrated code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratoly in the U.S. The experimental studies conducted at PNC have been reflected in the ASSCOPS improvement. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (volume and structure surface area and thickness, etc.), and the atmospheric initial conditions, such as gas temperature, pressure, and gas composition. ASSCOPS calculates the time histories of atmospheric pressure and temperature changes along with those of the structural temperatures.
PNC TN9410 94-205, 114 Pages, 1994/07
Thermal striping phenomena are characterized by random temperature fluctuations and observed in the region immediately above the core exit of Liquid Metal Fast Breeder Reactors (LMFBRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Therefore the in-vessel components located in the core outlet region, such as upper core structure (UCS), flow guide tube, C/R guide tube, etc., must be protected against the random thermal processes which might induce high-cycle fatigue. In this regard the MONJU reactor adopted ALLOY 718 as a coating material to protect these components against the fatigue. In this study, reduction measures of coolant temperature fluctuations have been investigated using a multi-dimensional thermohydraulic analysis code AQUA with the Algebraic Stress turbulence Model (ASM). From the analysis, it was confirmed that the area to be protect in-vessel components against the thermal fatigue can be decreased to 1/10 of the total area of ALLOY 718 adopted in the MONJU reactor by the tuning of the support plate height and the flow rate ratio between F/A and C/R assemblies.
; ; ; Isozaki, Tokuju;
PNC TN8450 93-004, 144 Pages, 1993/06
Generally speaking, two kinds of water plumes which have different temperature, salinity and so on are not easily mixturred and have special boundary. So, it is possible to predict ocurrence of special boundary of sea water from observation of temperature, salinity of sea water. Not only marine environmental monitoring such as radioactivity analysis of sea water, sea-bed sediment, marine products but also observation of water current, water direction, surface and ventical observation of water temperature, salinity and so on in Tokai coastal area are carried out with some observation instruments by Environmental Protection Section, Tokai Works, PNC. The results of vertical observation from 1990 to 1991 are recorded in this report.
PNC TN9410 92-105, 65 Pages, 1992/04
A thermal striping phenomenon characterized by a random temperature fluctuation occurs in the region immediately above the FBR core due to the temperature difference of the core outlet coolant between subassemblies. In this study, a direct numerical simulation code DINUS-3(Direct NUmerical Simulation using 3rd order upwind scheme) has been developed based on the third order upwind scheme and investigated applicability of the DINUS-3 code to temperature fluctuation analysis. From the analysis of von Karman vortex streak behind a rectangular obstacle, the following results have been obtained: (1)Change of the vortex frequency (the strouhal number St) with increase of the Reynolds number Re can be estimated by the DINUS-3 code. (2)A stationary random turbulence fluctuation including a buffer region between the transition and the turbulent regions can be predicted using the DINUS-3 code. And the followings became clear after the analysis of a nonisothermal parallel jet experiment using water. (1)A temperature fluctuation phenomenon including complicated frequency components can be simulated well using the DINUS-3 code. (2)Calculated dominant frequency has shown good agreement with the experiment. From the analysis, it is concluded that the DINUS-3 code based on the third order upwind sheme has a sufficiently high potential in providing good interpretation of experimental results related to the temperature fluctuation phenomena such as thermal striping.
Ohtaki, Akira
PNC TN9410 90-062, 145 Pages, 1990/04
It is important to control an operation to avoid water flow in which cause of SCC for a superheater which is made by astinetic stainless steel. During the superheater war㎜ing operation at the 50MW Steam Generator Test Facility, temperature fluctuation at the inlet tube sheet are easily occurred, and existence of saturated water with in warmming steam is regarded as the cause of it. A test simulating operating conditions of "Monju" superheater warmming was carried out to comfirm start up condition of "Monju". Test results show that a part of war㎜ing steam was cooled down to saturated temperature in main steam pipe which temperature is still lower than the saturated temperature, then the saturated water flows in the tube sheet which temperature is as high as sodium inlet temperature, and cause the temperature fluctuation. In order to confirm the effect of pre-heating the main pipe using trace line, a trace line was adapted at the main steam pipe, and the same superheater warmming simulation test was performed. Test results show that trace line was effective to pre-heat the main steam pipe exceeding temperature, and the temperature fluetuation was considerably decreased.
Muramatsu, Toshiharu
PNC TN9410 90-029, 76 Pages, 1990/04
A thermal striping phenomenon characterized by a random temperature fluctuation occurs in the region immediately above the FBR core due to the temperature difference of the core outlet coolant between sub-assemblies. In this study, we have investigated the thermal striping phenomena in sodium through implementing the algebraic stress turbulence model (ASM) in AQUA. From the analysis based on the ASM, the following analytical results have been obtained: (1)A calculated intensity of the temperature fluctuation was slightly overestimated by the ASM in the 1st row of the inner core region and the 2nd row of the radial blanket region, whereas a calculated mean temperature has agreed well with the experiment, (2)For a geometry with a back-up control rod in the inner core region, the calculated intensity of the temperature fluctuation was larger than that of the above two cases. The following information on turbulence modeling has been obtained: (1)A qualitative characteristics of the intensity of temperature fluctuation can be estimated by the ASM. (2)An assumption that the temperature fluctuation phenomena can be considered one of the turbulence phenomena, has been confiremed in evaluating the qualitative characteristics. (3)In order to evaluate the intensity of the temperature fluctuation, it is more appropriate to use the ASM that includes second order momentum of turbulence, than to use the k-
turbulence model.
*; Tanigawa, Shingo*; *; Yamaguchi, Katsuhisa; *; *; *
PNC TN9410 88-141, 159 Pages, 1988/09
The structural analyses of the core support plate have been applied to study thermal transfomation behaviors and the differences of the movement by changing analytical model, under anticipated transient without scram (ATWS) conditions of FBR. The analyses have been performed for 1000 MWe class loop type fast breeder reactor using a structural analysis code FINAS. The thermal-hydraulic results, which have been performed to ATWS conditions using a plant system code, were used as the thermal boundary conditions to the calculation. The scope of the analyses included a whole section of reactor vessel and the dead load of core assemblies was also considered. Following results were obtained from these studies. (1)The thermal transformation of a upper core support plate can be evaluated according to the free expansion behavior owing to the temperature change of core support plate itself. (2)The radial restriction due to core subassemblies has much influence on the axial bend of the core support plate. (3)There are some differences to the transformation results between by the whole model and by the one dimensional model during the thermal transient is large. Another analysis will be needed, however, about the reactivity change according to the displacement of the core structure.
*; *; Nakanishi, Seiji; *; *
PNC TN9410 88-111, 134 Pages, 1988/08
In order to rationalize the fuel handling system of Large Scale-up Fast Breeder Reactor, the equipment design of the Ex-vessel Storage (EVS) type fuel handling system was conducted in 1986. It can be considered that EVS type is coordinate with in-Vessel Storage (IVS) type from a viewpoint of materials and advantageous to Large Scale-up Fast Breeder Reactor for (1) reduction of reactor vessel diameter, (2) removal of the fuel handling process from sodium environment to water, and (3) realizationof ex-vessel NIS. As concerns sodium pot EVS type, the heat removal analysis at normal and transient conditions was performed to confirm the adequacy of the results gained from the equipment design in 1986 and to investigate requirements for the system. It was confirmed from the analysis results that the design was adequate and the requirements were extracted.
Takizuka, Takakazu
JAERI-M 85-055, 209 Pages, 1985/04
no abstracts in English