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*; *; *; Daigo, Yoshimichi
PNC TN941 83-158, 241 Pages, 1983/11
Failure propagation Tests have been carried out by use of the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center since 1979 in order to select a design basis leak (DBL) for sodium-water reaction accidents in Japanese prototype LMFBR Monju steam generators. Here reported are results of Runs 16 and 17 that were conducted to obtain data under the initial water leak rate of a few kilogramms per second. Temperature was measured at various locations in the tube bundle during the tests and the values of outer surface heat transfer coefficient in the sodium-water reaction zone were obtained. Based on these test results, the possibility of tube failure due to overheating seems to be very small. Post-test examination of the test internals disclosed that the wastage was not so severe.
*; *; *; Daigo, Yoshimichi
PNC TN941 83-38, 144 Pages, 1983/03
Sodium-water reaction tests were conducted at Oarai Engineering Center, PNC by using SWAT-1 (Large Sodium Water Reaction Test Rig) in the range of medium steam leak rate. These tests include, (1)determination of wastage rate on weld joint of heat transfer tube, (2)evaluation of toroidal westage on cooling heat transfertube, (3)evaluation of westage due to steam leak into cover gas phase, and (4)determination of heat transfer rate from jet flame to heat transfer tube, all of which have not been studied in the early tests. From the present tests results, the following conclusions were yield. (1)No difference was observed in wastage rate between weld metal and base metal of 2.25Cr-1Mo tube material. (2)The maximum secondary leak size of torroidal shape damage should be considered 1/2 DEG failure. (3)In case of water leak event in cover gas space, the ring shape wastage was examined in all tubes at sodium free surface level, but the wastage rate is much smaller than the direct impingement wastage in sodium. (4)The value of outer surface heat transfer coefficient was obtained.
*; *; *; Daigo, Yoshimichi
PNC TN941 82-99, 362 Pages, 1982/04
The Failure Propagation Tests have been carried out using the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center since 1979, in order to select a design basis leak (DBL) for sodium-water reaction accidents in Japanese prototype LMFBR MONJU's steam generators. Here reported are results of third Failure Propagation Tests (SWAT-3 Run-14 and Run-15) conducted in September 1980 and April 1981. Aim of these tests are to grasp how tube failures develop from the occurrence of initial leak to the completion of the water dump. The significant results are as follows: (1) In Run-14 test whose initial leak rate was 18 g/sec, the failure propagations occurred at 94, 145, and 168 sec, and additional two propagations occurred during tho water dump. (2) In Run-15 test, at first no propagation occurred during about four hours injection at the leak rate of 1.2 g/sec, therefore, as it was increased up to 14 g/sec by changing the water conditions, propagations occurred at 50, 195, 224 and 253 sec which was during the dump. (3) The maximum water leak rate is 2.7 kg/sec during the water dumping in Run-15, which suggests that the water leak rate during the water dump can exceed that before dump. (4) Numbers of the failed tubes are four in both tests. The maximum size of penetration is 18 mm 
4.1 mm which was generated at the fifth propagation in Run-15 and is fish-mouth type. (5) Mechanism of the propagations is wastage in the small leak region, but in the range of the leak rate above hundred g/sec, there are some cases of high temperature effect added to wastage effect.
*; Kuroha, Mitsuo; *; Daigo, Yoshimichi; *; *
PNC TN941 82-98, 87 Pages, 1982/04
Various tests were conducted in SWAT-2 to investigate the hydrogen behavior both in the cover-gas and in the sodium by using the In-cover gas and the In-sodium hydrogen meters. The results of the hydrogen injection tests, the water injection tests and the bubble rising velocity measurment tests are presented in this report. The conclusions that were drawn from these tests are the followings. (1)Traveling velocity of hydrogen gas in the cover gas was dominated by the buoyancy and the convection rather than by the diffusion. (2)The background hydrogen partial pressure in the cover gas did not become lower than 10
Torr. This lower limit was two to three orders of magnitude higher than the hydrogen pressure in sodium. (3)The experimentally determined bubble rising velocity in sodium agreed well with the theoretical prediction. (4)The detection rate of cover gas hydrogen that was not disolved into sodium was determined to be about 20Z, and that of hydrogen disolved into sodium to be about 80% with the present test vessel. (5)The hydrogen disolution rate during the water injection tests was higher than that of the hydrogen injection tests.
*; Miyake, Osamu; Daigo, Yoshimichi; *
PNC TN941 82-100, 48 Pages, 1982/04
The Computer code LEAP II had been developed in order to analize failure propagation phenomena by the sodium-water reaction in the steam generator of LMFBR. Here reported is verification analysis of the LEAP code by using Runs 14 and 15 test results of Steam Generator Safety Test Facility (SWAT-3). The main results are as follows: (1)As the results of parametric survey, the effects of the significant parameters such as time mesh, jet division number, etc. to the code were understood. (2)In comparison with the test results of Runs 14 and 15 of SWAT-3, the LEAP code can estimate the phenomena conservatively enough.
*; Miyake, Osamu; Daigo, Yoshimichi; *
PNC TN952 82-04, 74 Pages, 1982/02
A computer code, LEAP (Leak Enlargement and propagation), was developed in order to analyze failure propagation phenomenon in an LMFBR steam generator. This code calculates a water leak rate change from the initiation of water leakage to the termination of sodium-water reaction by the water dump. It consists of such components as; the operational condition of the steam generator, the calculation of the water leak rate, the decision of jet impingement, target wastage, the penetration of the target tube, the self-enlargement of the leak nozzle, the calculation of the leak detection schedule, the calculation of the cover gas pressure, and time step control. Here reported are calculation models, analytical methods, the source program organization, and the input/output forms. This program was developed originally in 1978 by Kawasaki Heavy Industries and has been modified in some points to reflect the results of research and the development of the sodium-water reaction tests.
*; *; Daigo, Yoshimichi
PNC TN941 82-42, 235 Pages, 1982/02
Failure propagation tests have been carried out using the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center since 1979, in order to establish the method of safety design of the LMFBR Monju prototype steam generator with reference to preventing sodium-water reaction accidents. A main object of these tests is to understand how the failure propagation to heat transfer tubes around progresses owing to the water leakage from the initial nozzle. Here reported are results of the second failure propagation tests conducted in May 1980. Three injection tests (SWAT-3 Run-11 through 13) were executed using three tube bundle test units manufactured according to the unification of the MONJU steam generater specifications. Though water did not inject in Run-11 because of malfunction of the injection device, the failure propagations took place in Run-12 and Run-13 whose initial injection rates were 87 g/sec and 900 g/sec, respectively. The main test results are as follows; (1).The second failure occurred at 74 seconds and the third one at 145 seconds in Run-12; eight gas-filled tubes failed from 70 to 175 seconds and a water-filled tubes failed at 158 seconds. (2).The maximum size of failure in Run-12 was 18 mm 5.8 mm of third failure tube, while two gas-filled tubes in Run-13 have a hole equivalent to one double-ended guillotine (DEG) failure. (3).Failure mechanism is wastage in the region of low water leak rate. In the case that the water leak rate is over about 1 kg/sec, overheating effect appears; however, the failure time of this case is as long as that of wastage case.
Kuroha, Mitsuo; Takeda, Kunio; Iitsuka, Shoji; Sasaki, Shuichi; Okada, Toshio; Isozaki, Mikio; Daigo, Yoshimichi; Sato, Minoru
PNC TN941 81-49, 204 Pages, 1981/05
PNC type in-sodium hydrogen meters have been developed as leak detectors for LMFBR MONJU steam generators. In order to confirm the long-term reliability and the durability of the meters, the four meters were installed in three sodium loops at the O-arai Engineering Center, and they had been tested over a long time in flowing sodium. A period of the test was from oct. 1977 to Feb. 1980. They are called type II. The dynamic chamber of the vacuum system can separate from the static one, and be also connected with it using one flexible tube. Important findings from the test are; (1) The operating time of two meters exceeded 10,000 hours, and the total of all meters was about 35,000 hours. No trouble had been experienced in the sodium systems and the nickel membranes of them during the period, which had the good durability. Air leaks, however, occured three times in the two vacuum systems. (2) Any secular changes had hardly happened in the permeability of hydrogen through the nickel membrane and the relationship between ion pump current and hydrogen pressure. (3) The pumping speeds had decreased with increasing the amount of absorbed hydrogen. The decreasing rates differed among four ion pumps, and those of two pumps were considerably large at the beginning of absorbing hydrogen. (4) The calibration curves, which describe the relationship between hydrogen concentration in sodium and ion pump current, had changed with time. The largest cause was the decrement of the pumping speeds. (5) The UHV gauges were superior to the ion pumps from the point of the signal-to-noise ratio as the hydrogen sensor.
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 500
C 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.
*; *; *; Daigo, Yoshimichi
PNC TN941 81-05, 235 Pages, 1981/01
Failure propagation tests have been carried out using the Steam Generator Safety Test Facility (SWAT-3) in PNC O-arai Engineering Center to establish the method of safety design of the LMFBR Monju's prototype steam generator with reference to preventing sodium-water reaction accidents. The main object of these tests is to understand how the failure propagation to heat transfer tubes around progresses owing to the water leakage from the initial nozzle. Here reported are the data of the first three failure propagation tests conducted in october 1979. Three injection tests (SWAT-3 Run-8 through 10) were executed whose initial water leak rates were 36 g/s, 6.8 g/s and 570 g/s respectively. (1)Failure propagation progressed in about each one minute in Run-8 and 10, but it did not occur in Run-9. (2)The maximum size of penetration holes is about 5.7 mm
for water tubes, and is 18 mm 33 mm for gas tubes. (3)The main mechanism of failure seemed to be the wastage. (4)There were some bowing and buldging tubes as well as wastaged tubes in Run-8 and 10. (5)The wastage rate was less than 710 mm/s in accordance with results of intermediate wastage tests.
Cr-1Mo steel nozzles for LMFBR steam generators; Studies of micro-leak sodium-water reactions (1)Kuroha, Mitsuo; *; Daigo, Yoshimichi; *
PNC TN941 80-135, 67 Pages, 1980/08
Experimental study on self-enlargement of a micro-defect was carried out using SWAT-2 test loop in order to establish the counter-plan for the micro-water leak in the LMFBR steam generators. The leak nozzles were made of 2.25Cr-1Mo steel which will be used as the heat transfer tube in the evaporator. The sodium temperature was fixed at 480C and the initial leak rates were chosen within the range of 1.610
g/s to 2.310
g/s. Main results of the experiments are as follows: (1)The self-wastage rate S
(mm/sec) is dependend on the water leak rate L(g/sec), and the relation is expressed as following equation. S = 0.0173 L
(2)The post test micrographs show that the self-enlargements started from the surface of sodium side, developed toward the water side, and finally the leak rates increased suddenly. (3)It is a characteristic of the self-wastaged holes that enlarged diameters of the sodium side are several times as large as that of the water side. (4)The minimum diameters of the enlarged nozzles were measured within 0.45 mm to 1.3mm, and the enlargement ratios of nozzle diameters increased as the water leak rates decreased.
Kuroha, Mitsuo; *; Daigo, Yoshimichi; *
PNC TN941 79-188, 58 Pages, 1979/10
One of PNC designed in-sodium type hydrogen meters which have been developed to use as the water leak detectors for the Monju's steam generators was modified in the vaccum system in order to measure the pumping speed of the ion pump during the operation. The modification was to increase the pressure drop between the ionization vaccum gauge and the ion pump by installing a orifice in front of the ion pump. As the result, it could be attained to estimate accurately the permeability factor K for the hydrogen through Ni membrane. The factor of K obtained is 110
cm Torr
/sec, and the pressure dependence in the low pressure range is rather smaller than published datas. It is found through the disccusions that the relationship between P
(partial pressure of hydrogen in sodium) and the measured pressure P
(pressure in vaccum side) is not expressed empirically by the half-power low, but followed by the pressure dependence of K reported. The best pair of the Ni membrane area and the orifice conductance is disccusd to dynamic vaccum system of the PNC designed in-sodium hydrogen meter. Also, it is concluded for the in-sodium hydrogen meter with a orifice using in the Monju's operation conditions that less than a quater of Ni membrane area adopted at present is usefull, and more than one year of the ion pump operation period without the change of the calibration curve can be attained.
*; *; *; Daigo, Yoshimichi; *
PNC TN941 79-141, 198 Pages, 1979/09
When a large leak sodium-water reaction accident occurs in a steam generator of LMFBR, pressure relief piping might be received thermal shock or blocked by frozen sodium, if it is not preheated. Then, the thermal transient tests were performed using the large leak sodium water reaction test rig SWAT-1. The results are summarized as follows; (1)Four tests were executed. The water injection rate of two tests was equivalent to that of several DEG (double-ended guilotine) failure of heat transfer tubes considering the difference of evapourator inner diameters between "Monju" and SWAT-1, and in other two tests the injection ratio was equivalant to less than that of 1 DEG. (2)Flow pattern in the pressure relief piping of two large injection rate tests was as follows, void fraction was as low as that of sodium single-phase flow in its early stage of 0.2
0.3 sec., and rapidly increased to about 0.9. In case of the small injection rate tests, the stratified flow had continued for 23 sec., it was followed by hydrogen gas single-phase flow. (3)In the large injection rate tests the maximum value of heat flux was about 110
[kcal/(mh)], and that of heat transfer coefficient was 310
[kcal/(mhC)] except in its very initial stage. In case of the small tests, they were lower. (4)In the large injection rate tests, stain of outer piping surface was about 8001,50010
, which agrees with the calculation using above value as heat transfer coefficient. (5)Possibility of blockage by frozen sodium is seemed to be very little in SWAT-1 test rig.
Hiroi, Hiroshi*; *; Daigo, Yoshimichi; *
PNC TN941 79-155, 367 Pages, 1979/08
Large leak sodium-water reaction tests have been carried out using the SWAT-3 facility in PNC O-arai Engineering Center to obtain data on the safe design of the prototype LMFBR Monju's steam generator with reference to preventing large leak accident. This report gives the results of SWAT-3 run-7 test. The heat transfer tube bundle of the evaporator used in Run-7 test was designed and manufactured by TOSHIBA/IHI. Main purpose of this test is to clarify the sodium-water reaction phenomena occured in the downcommer region. Water was injected into the evaporator at the rate of 10.6 kg/sec, which corresponds to a test scale of 2 tube failure in an actural size system according to iso-velocity modeling. Measurements were taken of pressure, strain, temperature, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on. Initial spike pressure was 19.5 kg/cma closest to the injection point, and the maximum quasi-steady pressure in the evaporator was 5.8 kg/cma. The rupture disc of the evaporator burst 0.613 sec. after water was injected, and the pressure relief system functioned well. No secondary tube failure was observed.
*; Daigo, Yoshimichi; Miyake, Osamu; *; *
PNC TN952 78-07, 48 Pages, 1978/10
Water leak detectors-hydrogen concentration meters are equipped with secondary cooling system of LMFBR to detect small leak of water from heat transfer tube in steam generator. Leak rate region to be able to detect the water leaks before secondary tube failure is decided by using the concept of so-called "safety map". Computer cade "SWAC-10-MJ/1" provides the safety map for secondary cooling system of proto-type reactor. This paper is written for user's manual of the code. SWAC-10-MJ/1 code was made through the modification of original code "SWAC-10" which had been developed for 50MW Steam Generator Test facility. Major differences from SWAC-10 code are as follows ; (1) Secondary cooling system of proto-type LMFBR is selected for calculation object (2) Wastage rate equations proposed by PNC are accepted. (3) Effect of enlargement of initial water leak hole due to self-wastage is introduced. (4) Hydrogen diffusion process in Ni membrane of hydrogendetector is introduced.
Kanegae, Naomichi*; *; Daigo, Yoshimichi; *; *
PNC TN941 78-91, 214 Pages, 1978/10
Seven PNC Type In-Sodium Hydrogen Meter Type-II were designed and manufactured to confirm that these meters will be applicable to the "MONJU" plant, and they have been installed and operated without any troubles in several testrigs of O-arai Engineering Center, Power Reactor & Nuclear Fuel Development Corporation. These meters were improved from the Type-I reported previously, and a new vacuum system named as "Separation Type of Dynamic and Static Chamber" was developed and applied to the Type-II. In this new system, the dynamic and the static chamber are separated from each other in the vacuum system, so that the operation method become very simpler and it becomes possible to calibrate these meters with more accurately and shorter time. This paper explains the basic design specifications, the detail constructions and the test results of in-sodium tests including the calibration and the response characteristics in case of sodium-water reactions were occured. The following results were obtained. (1)The separation type of dynamic and static chamber is effective to simplify the operation method and to improve the accuracy of calibration of hydrogen meter. (2)The optimum design method of hydrogen meter reported preveously was established by several experiences of designing, manufacturing and operating of the Type-II and also the Type-I. (3)Several infomations applicable to the hydrogen meter of the MONJU plant were obtained, and a prospect that this Type-II hydrogen meter will be able to be used in MONJU plant was obtained. Now, long term operation test in sodium is under conducting for confirming life time or characteristics of operation time dependency of the Type-II meters in O-arai Engineering Center.
Kanegae, Naomichi*; *; Daigo, Yoshimichi; *
PNC TN941 78-85, 187 Pages, 1978/01
This report describes the PNC In-sodium Hydrogen meter Type-1 which is manufactured for establishment of the optimum design methods of hydrogen meter reported previously. This type-1 meter is now under operating without any troubles in the Small Sodium-Water Reaction Test Facility (SWAT-2) of PNC-Oarai Engineering Center. This report explains the design specifications, the detail constructions and the test results of in-sodium and in-gas tests and also explains several new informations obtained during designing and manufacturing Type-1 meter, as follows, (1)Three basic objects were established, the first one is to make this meter including sodium system, simple and compact, the second one is that this meter is to have a function of concentration monitor and the last one is that this meter is to have a function of leak detector. (2)Several informations such as, methods of leak tests after welding of nickel membrain, a new type vacuum gauge for detecting low vacuum region, the baking conditions of the vacuum side and the self-pumping effects of vacuum gauges were obtained. Based on these results, advanced meters called as Type-2 is now under designing and manufacturing and a long term operation test in sodium will be held for confirming several characteristics of these meters on the assumption that these meters will be used in the MONJU Plant.
