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Ushiki, Hiroshi*; Okuda, Eiji; Suzuki, Nobuhiro; Takamatsu, Misao; Nagai, Akinori
JAEA-Technology 2015-042, 37 Pages, 2016/02
JAEA-Technology-2015-042.pdf:16.51MB
The reactor vessel of a sodium-cooled fast reactor (SFR) is filled with sodium coolant and cover gas (argon gas). In case of a cover gas boundary open (ie., in-vessel repair), installation of a temporary cover gas boundary and controlling the cover gas pressure slightly positive are required to prevent the cover gas release and the contamination of impurities, and during upper core structure (UCS) replacement in the experimental SFR Joyo from March to December 2014, a vinyl bag was installed as a part of the temporary cover gas boundary. However, because it has inferior thermal resistance, supply a cooling gas too much was required to maintain proper temperature for two months. On the basis of this requirement, a cover gas recycling system with precise pressure control was developed and adopted for UCS replacement. The system has a good pressure controllability and recyclability. The successful results of this system contributed to the certain promotion of UCS replacement. In addition, the insights and the experience gathered in this development are expected to improve the in-vessel repair techniques in sodium-cooled fast reactors.
; Nemoto, Masaaki; Saikawa, Takuya*; Sukegawa, Kazuya*
JNC TN9410 2000-008, 66 Pages, 2000/03
JNC-TN9410-2000-008.pdf:1.39MB
The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.
*;
PNC TN9410 96-102, 40 Pages, 1996/04
In designing the shield plug of LMFBR, it is important to evaluate the thermal response between the cover gas thermal-hydraulics and the temperature fields of the shield plug at the same time. Based on the experiments which were performed by OEC, the natural convection and the thermal radiation in the cover gas layer were calculated with the structure simulating the shield plug in a detail two-dimensional model. The calculations were carried out for 8 kinds of experimental RUNs using a FLUSH code. The main results were as follows: (1)For these 8 kinds of experimental RUNs, the velocity and the temperature distributions in the cover gas layer were presented. The radial and axial temperature distributions in the rotating plug were also presented, which were difficult to measure by the experiments. (2)The boundary surface temperature between the cover gas layer and the rotating plug had the same tendencies and the calculated average temperatures on the boundary surface had good agreements with the experimental data. The average relative deviations from experimental values were less than 1.3%. (3)The natural convection of the cover gas enhanced the temperature distributions in the structure. The effects of thermal radiation on the heat transfer was relatively small and it can be neglected when the temperature of the heated aluminum disk is less than 400
C.
Tanabe, Hiromi; Hamada, Hirotsugu
PNC TN9410 91-028, 14 Pages, 1991/01
A COMPUTER CODE,SWACS,WAS DEVELOPED TO ANALYZE A LARGE LEAK SODIUM-WATER REACTION EVENT IN AN LMFBR STEAM GENERATOR. THE JAPANESE PROTOTYPE REACTOR,MONJU,HAS A COVER GAS SPACE IN ITS STEAM GENERATOR BUT DIFFERENT DESIGNS ARE ALSOCONSIDERED FOR A FUTURE LARGER PLANT. THEREFORE,SWACS WAS MODIFIED TO ANALYZE THESODIUM-WATER REACTION EVENT UNDER SUCH VARIOUS DESIGNS. SO FAR THE CALCULATIONALMODULE OF AN INITIAL SPIKE PRESSURE AND ITS PROPAGATION TO IHTS WAS IMPROVED AND THERESULTS WERE COMPARED WITH THE DATA FROM LLTR AT ETEC, U.S.A. AND WATER-EXPLOSIVE SIMULATION TESTS AT PNC,JAPAN. THE COMPARISON REVEALED A FAIRLY GOODAGREEMENT BETWEEN THE TESTS AND THE ANALYSES. FOLLOWING THE VALIDATION STUDY,SWACS WAS USED FOR THE APPLICATION ANALYSIS TO COMPARE THE PRESSURE BEHAVIORBETWEEN THE COVER-GAS TYPE AND THE NO-COVER-GAS TYPE STEAM GENERATOR OF A FUTURELARGER PLANT. THE ANALYSIS CLARIFIED THE APPLICABILITY OF SWACS TO SUCH A DESIGN STUDYFROM A VIEWPOINT OF SUPPRESSING THE SWR PRESSURE.
Tanzawa, Sadamitsu; ; *
JAERI-M 90-159, 80 Pages, 1990/09
no abstracts in English
; ; Murakami, Yoshio
Shinku, 28(5), p.351 - 353, 1985/00
no abstracts in English
IEEE Transactions on Nuclear Science, NS-31(1), p.757 - 760, 1984/00
no abstracts in English
Kuroha, Mitsuo; *; *; *; Daigo, Yoshimichi; *; *
PNC TN941 81-51, 70 Pages, 1981/02
An in-cover gas hydrogen concentration meter was designed and manufactured as the water leak detector for sodium heated steam generators. This meter consists of a thin nickel membrane, a vacuum system, and an electric heater. The nickel membrane is used under the internal pressure in order to prevent its buckling. This meter was installed in the Small Sodium-Water Reaction Test Loop (SWAT-2) at PNC-Oarai Engineering Center. The specification, the construction and some of the test results are described in this report. The following results were obtained; (1)This hydrogen meter was very compact and capable for the leak detector and the hydrogen concentration meter. (2)The test result indicated that this meter was applicable to hydrogen concentration from 3 Vppm up to 10000 Vppm at nickel membrane temperature of 500C and cover gas pressure of 1 kg/cm
G. (3)Hydrogen permeability through the nickel membrane in sodium vapour was similar to that in sodium.
Okuda, Eiji; Ushiki, Hiroshi; Suzuki, Nobuhiro; Sasaki, Jun; Takamatsu, Misao
no journal, ,
no abstracts in English
Hashidate, Ryuta; Morioka, Tatsuya; Sawazaki, Hiromasa; Shiotani, Hiroki; Obata, Ikuhito; Uekura, Ryoichi
no journal, ,
We have been evaluating mainly the data acquired by the central computer so far, but we focused on evaluation using preservation data from this time. In this report, we report on the results of examination of the shaft sealing mechanism which constitutes the boundary of the cover gas including sodium vapor from the viewpoint of seal gas effect considered in the design stage for sodium vapor and sealability of the cover gas boundary by using the Monju plant data and conservation data for 20 years.
Suzuki, Shigeaki*; Hayakawa, Masato; Shimoyama, Kazuhito; Umeda, Ryota; Yoshida, Eiichi; Miyakoshi, Hiroyuki
no journal, ,
We are conducting dismantling inspection of large-scale sodium component that we have been using in sodium environment for a long time. We got the cover gas area sodium deposition rate of the large tank used for several decades. As a result of examination based on the data and past findings, it was possible to derive the recommended value "1.0e-10 g/cm/s" for the cover gas region sodium adhesion rate of the plant operating in the low temperature range (150 to 200C). For disassembling large sodium equipment, it is possible to estimate and evaluate sodium adhesion amount based on this recommended value and operation history. With this evaluation, it became possible to increase the reliability of disassembly technology for safety measures and safety management related to sodium fires etc.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Tanaka, Masaaki
no journal, ,
Japan Atomic Energy Agency (JAEA) is developing AI-aided advanced reactor life cycle optimization method, ARKADIA, in order to support the development of various reactor systems, including sodium-cooled fast reactors (SFRs). One of the important thermal-hydraulic issues in SFR design is the suppression of cover gas entrainment caused by dip vortices in the free surface of the upper plenum of the reactor, and JAEA is developing gas entrainment evaluation system centered on the gas entrainment evaluation tool, StreamViewer, in ARKADIA. In this report, the overview of gas entrainment evaluation system in ARKADIA and the development status of StreamViewer are reported. In addition, the results of applying StreamViewer to the analysis results of the water flow test system and confirming that it is possible to evaluate the gas entrainment phenomenon are reported.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
no journal, ,
In the sodium-cooled fast reactors, evaluation of cover gas entrainment (GE) by dip vortices at the free surface in the upper plenum of reactor vessel is important. In this study, GE evaluation model, including the identification of vortex center lines between free surface and suction port from the flow velocity distribution obtained by three-dimensional analysis, the calculation of pressure decrease distribution along vortex center line, and the comparison of pressure decrease and the hydrostatic pressure, was investigated. The analyses for GE experiments in a rectangular flow channel were performed by changing the inlet flow velocity condition, and this evaluation model was applied to the flow velocity distribution obtained from the analysis results. As a result, it was confirmed that the phenomenon in which the occurrence of GE increased as the inlet flow velocity increased was reproduced as in the experimental results.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
no journal, ,
In sodium-cooled fast reactors, the cover gas is entrained into the coolant due to the dimple vortices generated in the free surface in the upper plenum of the reactor vessel, and gas entrainment (GE) phenomenon may affect core behavior. It is necessary to develop evaluation methods of GE phenomenon. In this study, a GE evaluation model, PVL model (Pressure Vortex Line model) was constructed. The PVL model includes the identification of vortex center lines from the flow velocity distribution obtained by 3-D analysis, the calculation of the 3-D distribution of pressure decrease along the vortex center line and the evaluation of gas core length of vortex by comparing the pressure decrease and the hydrostatic pressure. The PVL model was applied to the analysis results of advective vortex experiments in a rectangular channel system to evaluate GE. As the result, it was confirmed that it was possible to judge the GE occurrence of advective vortices by the application of PVL model.
