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; Kamide, Hideki;
PNC TN9410 98-083, 118 Pages, 1998/07
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a demonstration fast reactor. The test facility is consisted of components from a reactor vessel to a steam generator (SG). Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the demonstration fast reactor with two primary cooling loops and two into one secondary loop. The secondary piping length of the test facility is longer than the 1/3 scale of the demonstration fast reactor. The tests facility has the branch and junction of the secondary piping because of two primary loops and one SG. There is a possibility of flow and temperature unbalance if a buoyancy force were large and pressure loss were small. Therefore, dynamics analyses of the thermal transition tests had been done in which the secondary piping length. To examine the unbalance occurred or not, the natural circulation analysis had been performed providing different heat transfer area of the IHX or presser loss of the primary loop between A loop and B loop. It was shown from the analyses that the temperature response during the transition was delayed in the test model compared to the real reactor. Main cause of the delay was due to the real scaled SG. Other parameters, the length of piping etc., were not very influential to the response. The analysis such predicted that there wasn't large difference of global behaviors between the loops. Therefore, it was shown that there would be no problem, if the difference were made between the loops due to a manufacturing error.
*; ; ; *; *
PNC TN9410 91-132, 85 Pages, 1991/06
A series of high temperature fracture toughness tests of 9Cr type steels was put in practice on a three-years scheme in 1988. In this experimental study, Mod.9Cr-1Mo steel, 9Cr-2Mo steel and 9Cr-1Mo-Nb-V steel are tested, and these fructure toughness are investigated on the basis of J integral. In a first year, high temperature fracture toughness test method was established using R-curve method and unloading compliance method, and J
values of Mod. 9Cr-1Mo steel were measured at RT, 400
C, 500C, 550C and 600C. In a second year, fracture toughness tests of 9Cr-2Mo steel, 9Cr-1Mo-V-Nb steel and these aged materials were performed, and several factors which affected fracture toughness values of 9Cr type steels were clarified. In this report, the test results in a second year are summarized. In a last year, fracture toughness tests of weldment and thick plate of 9Cr type steel are going to be carried out. Furthermore, an effect of crack direction on J value will be studied.
*; *
PNC TN9410 87-037, 101 Pages, 1987/04
Pressure behaviors in a large scale sodium-water reaction event were analyzed using the SWACS code to clarify the effect of the cover gas region of the steam generator on the pressures as a part of the design study of the Large LMFBR plant of Japan. The non-cover-gas type steam generator handled here comes from "Key Technology Design Study (II) (1985)" and the cover-gas-type was designed by adding cover gas region to the top of the former one. An initial spike pressure analysis, i.e. a short term analysis, revealed that due to the lack of a pressure attenuation effect in sodium free surface, pressures in hot-leg pipes and IHX in the non-cover-gas type were higher than those of the cover-gas type. The maximum pressures were 27.0 and 16.2 kg/cm
a in the gasless and gas type, respectively. Even in the gasless type, however, it is possible to reduce the short-term pressures to a certain degree by installing the rupture discs close enough to the steam generators reaction zone. The results of a quasi-steady pressure analysis, a long term analysis, clarified that the rupture disc bursting pressure almost corresponds to the peak pressure in the gas type design while a quasi-steady pressure build-up was minor due to initial disc burst in the gasless type. From such analyses, it reveald that the SWACS code had a sufficient potential for applying to a design evaluation study of the Large LMFBR plant design.
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.
*; *; *
PNC TN941 85-127, 92 Pages, 1985/08
RECT-II (the Removal test of reaction products by cold trap) was conducted by use of SWAT-3 (the Steam Generator Safety Test Facility) at PNC in order to construct the post-accident operation of steam generators of the prototype FBR Monju and a larger plant following it. In prior to the test, some amount of the sodium-water reaction products (SWRP) generated in the water injection test (Run 18) was remained in the sodium system. An objective of the test is to confirm the purifying method to remove SWRP by hot sodium circulating through a cold trap (CT). A meshless type cold trap was selected to avoid choking by impurities and to enable efficient SWRP removal. RECT-II started on April 4, 1984 and terminated on April 26 when the plugging temperature decreased to 187C. Major results obtained in the test are as follows: (1)Post-test observation revealed that the SWRP having remained at the bottom of the evaporator and the sodium outlet pipe were completely removed through the purification operation. (2)Hence, it is concluded that after the hot draining the SWRP of 14 kg-H
0 remained in the sodium system out of that generated by the 42 kg-H0 injection and that almost all of the former was removed through the operation. (3)However, some amount of the hydrocarbon-oxide and SWRP in the slit articles simulating crevice and stagnant region still remained after the operation. Then it is concluded that it is insufficient to remove SWRP in crevice and stagnant region by the circulation of hot sodium. (4)A mass transfer coefficient of oxygen is evaluated as 2 
10
[g/(mm H ppm)] if the cross section of the evaporator and inner surface of the 8 inch horizontal pipe are assumed to be the entire surface area of SWRP. (5)Since the choking of the cold trap degrades the efficient SWRP removal, it is essential to develop a cold trap which hardly chokes and easily regenerates even after choking; one of answers for this request is a ...
Hiroi, Hiroshi*; Miyake, Osamu; *
PNC TN941 82-37, 170 Pages, 1982/02
Blowdown tests of high temperature and pressure water from a long pipe were carried out to validate the computer code SWAC-11 which is used for the calculation of the water leak rate from a failed heat transfer tube in an LMFBR's SG. The steady leak rate, and transients of pressure and thrust force of the pipe were measured. Especially, the short term transient of thrust force can be obtained by a new measuring method using the spring-mass model. These data were compared with calculation results of SWAC-11. As for steady data, the Moody's model of the critical flow and the effect of the two-phase multiplier were studied. Major conclusions are as follows: (1)The calculation results of SWAC-11 almost agreed with the steady data. But in detail, SWAC-11 inclined to predict 10 
15% less than experimental data of water leak rate and thrust force in the case of high pressure saturated water. This discrepancy will be reduced by introducing the Thom's correlation as the two-phase multiplier. (2)The calculation results of SWAC-11 also agreed with the experimental data after 5 msec since the blowdown was initiated. (3)The flow model of SWAC-11 can be applied to the blowdown of the subcooled water. (4)The thrust force (F) immediately after the blowdown is the sum of the wave force and the blowdown force. The relation of F, the initial pressure P
, and cross section S can be given by the expression, F/S
P=1.36. (5)Compared with calculation results in detail, the profiles of experimental data were found to be more complicated. This tendency was observed markedly in the case of the subcooled water blowdown. (6)Test results of superheated steam blowdown agreed with SWAC-11 predictions as for unsteady data as well as steady data.
*; *; *
PNC TN941 81-52, 296 Pages, 1981/02
Recently surveillance systems for nuclear power plants are increasingly required for the improvement of plant safety and availability. In order to establish the surveillance system of the prototype fast breeder reactor "MONJU", some techniques have been developed and applied to the 50MW Steam Generator Test Facility ty at OEC. As the first stage of the development, information display techniques for the plant operators and some anomalous state detection techniques are discussed in this paper. The operators can obtain such plant informations as digital and graphic outputs by cathode ray tubes (CRTs) and print out by a lineprinter and typewriters. Also the operators are informed of results of anomalous diagnosis by annunciator alarms moment by moment. Application tests of the anomalous state detection techniques have been carried out. These techniques include a cross check technique of multi-measuring system, a automatic detection system of a small scale sodium-water reaction, a differential alarm and prediction method of the time of anomalous occurrance and a display method of degree of superheat of evaporator (EV) outlet steam. It was concluded by our evaluation of the test results that those techniques are applicable to the "MONJU" design without major modification. We will develop new techniques and improve these systems to make them applicable to "MONJU", considering the "man-machine system", using this test facility.
