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Mine, Tatsuya*; Mihara, Morihiro;
JNC TN8430 2000-009, 35 Pages, 2000/07
JNC-TN8430-2000-009.pdf:0.88MB
In the geological disposal system of TRU wastes, nitrogen generation by denitrifying bacteria could provide significant impact on the assessment of this system, because nitrate contained in process concentrated liquid waste might be electron acceptor for denitrifying bacteria. In this study, the activities and tolerance of denitrifying bacteria under disposal condition were investigated. pseudomonas denitrificans as denitrifying bacteria was used. The results showed that Pseudomonas denitrificans had activity under reducing condition, but under high pH condition (PH
9.5), the activity of Pseudomonas denitrificans was not detected. It is possible that the activity of Pseudomonas denitrificans would be low under disposal condition.
Hamada, Hirotsugu; Tanabe, Hiromi
PNC TN9410 96-027, 41 Pages, 1995/12
If a sodium-water reaction jet was formed due to water leakage in an FBR steam generator(SG), neighboring tubes would suffer from overheating. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such a severe overheating condition. So far, an analytical model using the structural integrity analysis code, FINAS, has been prepared and validated by the explosive torch overheating test data. This report presents the results on the overheating failure analysis of the under-sodium leak in the PFR superheater(SH), 1987. In the SH with slow steam dump system in 1987, neighboring overheated tubes are failed about 3 seconds after the SH isolation, which is shown both by the leak in the PFR and its analysis. For the SH in which a fast steam dump system was installed after the leak of 1987, the analysis shows no tube failure due to the fast steam depression and cooling effect inside. These results indicate that the FINAS model adequately predicts the overheating failure and the specific SH design and operation possibly result in further growth of the leak. It is concluded that steam blow effect is extremely important and the analysis model presented here is useful for the overheating failure evaluation of the SGs.
Hamada, Hirotsugu
PNC TN9410 95-262, 35 Pages, 1995/09
Neighboring tubes in an FBR Steam Generator (SG) would suffer from overheating if a sodium-water reaction jet were formed due to water leakage in the SG. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such an overheating condition. An analytical model using the structural integrity analysis code, FINAS, has been prepared to evaluate the overheating failure and here an explosive torch overheating test was analyzed to validate the FINAS model. These experiments and analysis indicate that the overheating failure is closely associated with heat transfer coefficients (HTCs) of outer and inner tube wall and that the FINAS model conservatively predicts the overheating failure within acceptable accuracy. For making progress in further tests like an explosive torch test and its code validation, it would be required that sodium-water reaction experiments should be performed to provide the data on the HTCs, high pressurized and superheated steam should be supplied in the explosive torch test, and that a multidimensional analytical model should be developed to closely predict the temperature distribution in the axial(z-) and circumferential(
-) directions on the tube wall.
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PNC TN9410 94-093, 52 Pages, 1994/05
Sodium-water reaction under the high pressure condition of several hundred times of atmospheric pressure was analysed based on the occurence assumption of a hypothetical sea water leakage accident through a pressure hull with the rupture of a bellows type accumulator concerning a deep sea reactor. The sodium-water reaction analysis code of SWACS/REG4 was used in the analysis. The basic calculation case of the analysis adopted the water leakage rate of 0.128 kg/s by considering the accumulator rupture time of 1 s and the water leakage rate used in the heat transfer pipe rupture accident analyses of FBR heat exchangers which had been performed by the code. The calculational result clarified no existence of pressure wave in the upper plenum of a reactor vessel, which was due to the attenuation of the wave in the long slender pipe of 1.5 m in length and 2 cm in inner diameter between the accumulater rupture point and the reactor vesssel. In the analysis, survey calculations were also performed by changing the parameter values of the pressure, the water leakage rate and the gas space volume remaining in the pressure hull.
*; *; *; *; *; *; *
PNC TJ9164 94-006, 133 Pages, 1994/03
This report gives an applicability of SWAT-3 facility and contents of the reconstruction in order to confirm a DBL (Design Basis Leak) for the demonstration reactor SG. (1).Test Cndition and test case. Evaluation of the wall temperature for adjacent heat transfer tubes under the sodium-water reaction event was performed. (a)As the effect of tube rupture due to overheating, failure of upper part of the helical coil was severer than one of the lower part. (b)The wall temperature depends on the water side condition. (c)Reference test condition, whici is water leak rale about 1 kg/s, failure of upper part of the helical coil and 30% partial load, was selected. A total of ten test cases were decided. (2).System and Components Design. (a)Large leak sodium-water reaction analyses including water injection rate analysis and quasi-steady pressure analysis were performed. The maximum water leak rate of 1 DEG was 7.2 kg/s and the water leak rate at the quasi-steady state was 3.1 kg/s. The maximum pressure was 18.1kg/cm
a at the piping between the reaction vessel and IHX, the pressure was within the design condition of SWAT-3 facility. (b)Based on the results of the large leak sodium-water reaction analyses, a reaction vessel, water heaters and a dump tank were designed and their design specification were clarified. The reaction vessel was a scale of one third of the demonstration reactor SG and it was designed to be able to conduct the water injection test twice with one test unit. (c)A system and piping diagram, and many kinds of list (Piping list, Valve list, instrumentlist) were made up. (3).Reconstruction scope and arrangement plan. The reconstruction scope and a layout for the components and piping were clarfied and the arrange ment plans were made up. (4)Reconstruction period. The recoastruction period and man power for the design, fabrication, inspection and installation were studied and the reconstruction schedule was made up.
*; *; *; *
PNC TJ9124 94-009, 164 Pages, 1994/03
In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. In this study, the general plan for the next models to evaluate the reasonable DBL have been designed; a)overheating tube bursting models (structural/fractural dynamics), b)unsteady heat conduction analysis models, c)blow down analysis models and d)reaction zone temperature distribution analysis models. Then blow down analysis models were developed to evaluate the overheating tube bursting and analysis code was preliminarily designed in which the module construction of this code and link of each modules were described. Furthermore, easy coupling of this code and LEAP in future was fully considered.
Shimoyama, Kazuhito; Kuroha, Mitsuo; *
PNC TN9410 87-014, 103 Pages, 1987/01
Acoustic type water leak detectors have promising potentiality in short detection time for minimising the extent of tube failure propagation caused by water leakage from a heat transfer tube of an LMFBR steam generator. Two different methods as follows were studied in this program : (1)The method to compare effective values between water leak sound and back ground noise using a single channel. (2)The method to detect and locate the leak using cross correlation signal processing of multi-channel. In the former one, it was estimated from acoustic signals obtained in the 50 MW Steam Generator Test Facility that the back ground noise levels of the Prototype and the Demonstration reactor were 0.0093G and 0.012G (G=gravity), respectively. The water leak rates equivalent to those back ground levels were evaluated as approximately 0.7 and 7 g/sec. In the latter one, first a detection and location software was developed in a off-line analysis, and secondly an on-line signal processing hardware was manufactured as a trial. In the off-line analysis, the influence of the internals on detection performance was examined by horizontal and vertical measurement. As the result, it revealed that back ground noise interfered the leak detection and location and that the potential depended on the leak positions even without noise. In the on-line analysis, leaks in a lower plenum were detectable with the same accuracy as the off-line analysis.