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Hamada, Hirotsugu; Kurihara, Akikazu
JNC TN9400 2003-031, 96 Pages, 2003/05
JNC-TN9400-2003-031.pdf:3.34MB
The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1)The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in compatison with the previous method. (2)The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3)The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4)In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5)In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model.
Itooka, Satoshi*; Fujimata, Kazuhiro*
JNC TJ9440 2003-001, 286 Pages, 2003/03
JNC-TJ9440-2003-001.pdf:9.23MB
In Japan Nuclear Cycle Development Institute, the improvement of analysis method for overheating tube rapture was studied for the accident of sodium-water reactions in the steam generator of a fast breeder reactor and the evaluation of heat transfer condition in the tube were carried out based on study of critical heat flux (CHF) and post-CHF heat transfer equation in Light Water Reactors. In this study, the improvement of blow down model for the LEAP code was carried out taking into consideration the above-mentioned evaluation of heat transfer condition. Improvements of the LEAP code were following items. Calculations and verification were performed with the improved LEAP code in order to confirm the code functions. (1)The addition of critical heat flux (CHF) by the formula of Katto and the formula of Tong. (2)The addition of post-CHF heat transfer equation by the formula of Condie-Bengston Ⅳ and the formula of Groeneveld 5.9. (3)The physical properties of the water and steam are expanded to the critical conditions of the water. (4)The expansion of the total number of section and the improvement of the input form. (5)The addition of the function to control the valve setting by the PID control model.
Ishikawa, Hiroyasu; Ohno, Shuji; Miyahara, Shinya
JNC TN9400 2002-081, 46 Pages, 2003/01
JNC-TN9400-2002-081.pdf:2.21MB
Nitrogen gas can be an extinguisher or a mitigating material in the case of sodium leak and fire accident in an air atmosphere, which may occur at a liquid metal cooled nuclear power plant. However sodium combustion residuum sometimes reignites in the air atmosphere even at room temperature when it was produced by nitrogen gas injection to the burning sodium. In this study we have been investigating the cause of reignition and prevention measures. Experiments were carried out with small type test equipment, which can handle 1g order sodium fire and extinguishment. Sodium combustion residua, which were made by our equipment and sampled, were analyzed by X-ray diffraction and chemical analysis. The chemical analysis of reignitable residua showed that the residuum contained metallic sodium of about 40wt-% (61 mol-%) to 60-wt % (76mol-%) and most of the rest was sodium-monoxide (Na
O). Sodium-peroxide (NaO) was also included in less than 1wt-% of the residuum. Sodium or NaO cannot ignite by itself in the air atmosphere at room temperature in a few minutes. Therefore the reignition seems to be due to increase in the local temperature that is caused by oxidizing heat of Na and by adiabatic effect of NaO. It is important to deactivate this dispersed sodium on oxygen for prevention of the residuum reignition, hence it is considered as a rational measure to change the sodium to sodium-carbonate. Our experiments showed that the dispersed sodium on the exterior of residuum could be changed to carbonate by a mixture of carbon-dioxide (CO) gas (2 to 8vol-%), humidity (0.6 to 3vol-%) and nitrogen gas. The deactivated residuum did not reignite in the air atmosphere below 473K.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; ; Miyakawa, Akira; Okabe, Ayao;
JNC TN9400 2001-130, 235 Pages, 2002/03
JNC-TN9400-2001-130.pdf:7.05MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: (1)To evaluate the structural integrity of tube material, the strength standard for 2.25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200
C) creep data. This standard has been validated with the tube rupture simulation test data. (2)The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. (3)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. (4)The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. (5)The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system.
Nakagiri, Toshio; Sato, Isamu
JNC TN9400 2002-045, 75 Pages, 2002/02
JNC-TN9400-2002-045.pdf:2.43MB
Fission products (FPs) release behavior in the experiments (FP-1, FP-2) performed at JNC were evaluated from the viewpoint of source-term evaluation of FBR. Release fraction of FPs (Cs, Sb, Ru, Eu) in the experiments were calculated by NUREG-0772 model, and grain (hypothetical spheres) radius of fuel pellets and diffusion coefficient of FPs in grains for Booth model were calculated in the evaluation. Furthermore, the calculated values were compared with the values obtained in other experiments, The results of evaluation are summarized bellow. (1)The experiments (FP-1, FP-2) were performed in inert gas (Ar) condition, but FPs release fractions and diffusion coefficients in fuel grains almost agree with the values obtained in the other experiments which were performed in oxidized conditions. FPs release behavior in reduced condition containing such as sodium vapor are expected to be different from the behavior in FP-1 and FP-2. Some other experiments in reduced condition should be performed to investigate FPs release behavior in reduced conditions. Geometric grain radius obtained from metallographic images is used for the base data of diffusion coefficients evaluation, but the radius calculated from BET surface area is desirable. (2)Another release mechanism such as evaporation from the fuel pellet surface should be considered for more detailed evaluation of Cs composite and low volatile Ru and Eu release behavior, but, FPs chemical composition in fuel pellet, oxygen potential in carrier gas and FPs chemical composition in carrier gas, etc, should be evaluated in advance.
Takai, Toshihide; Nakagiri, Toshio; Hamada, Hirotsugu
JNC TN9400 2002-009, 41 Pages, 2002/02
JNC-TN9400-2002-009.pdf:1.07MB
A sodium pool bubbling test was conducted to investigate tbe hydrogen combustion behavior under sodium-concrete reaction accident. FRAT-1 test vessel (Volume of 3m
) of SAPFIRE (Safety Phenomenology Test on Sodium Leak, Fire and Aerosol) facility was used for the test. The sodium combustion pot of 
97mm was installed in FRAF-1 test vessel. Sodium was burned at the maximum hydrogen injection rate of 9.5
10
Nm/m
sec, which was determined value based on past sodium-concrete reaction experiments. And the recombination rate of hydrogen was estimated. From the results, the following conclusions were obtained: (1)The recombination rate of supplied hydrogen was 99% or more. (2)It is considered that the burning form of the supplied hydrogen is laminar nonpremixed name based on the molecular diffusion between hydrogen and oxygen. (3)Most of vapor generated by hydrogen combustion was seemed to consume by chemical reaction with sodium aerosol in the atomosphere.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Okabe, Ayao; Miyakawa, Akira
JNC TN9400 2001-099, 76 Pages, 2001/11
JNC-TN9400-2001-099.pdf:2.13MB
The JNC technical report "The Development and Application of overheating Failure Model of FBR Steam Generator Tubes" summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. (1)0n the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. (2)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. (3)Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure.