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*; *; *; *; *; *; *
PNC TN241 85-12, 292 Pages, 1985/03
Computer codes for safety analysis, including systems codes, which are used for the evaluation of LMFBR plants, have been developed along with the progress of the construction and the operation of experimental fast reactor "JOYO". On the other hand, a large number of data has been accumulated at O-arai Engineering Center and other laboratories both in Japan and abroad with the advance of safety R&D on Liquid Metal Fast Breeder Reactor (LMFBR). Various models of analysis have been proposed and many computer codes have been developed which are based on these models. This paper describes the essential part of the models and the functions of the codes, thus developed, which are used for the safety evaluation of LMFBR plant such as the prototype fast breeder reactor "MONJU". Analize codes for sodium leakage accident were modified, so that detailed conditions could be taken into account.
Hirose, Tadashi*; *; *; *; Yamamoto, Hisashi*
PNC TN941 82-81, 153 Pages, 1982/03
The purpose of this test is to establish the power-up procedure to the rated power. The first stage rated power was 50 MWt and it was achieved on July 5, 1978. The second stage rated power was 75 MWt and it was achieved on July 16, 1979. This report describes the results of these power-up tests. The results were; (1)The optimum heat up rate from warm stand-by to hot stand-by is about 20
C/hr, corresponding to about 1.5mm/5min of the regulation rod handling speed. (2)The optimum power-up rate from hot stand-by is about 5 MW/20min, corresponding to about 1mm/2min of the regulation rod handling speed. (3)The optimum reactor power to start the main blowers is about 10 MWt.
*; *; *; *; *; *; *
PNC TN241 81-28, 292 Pages, 1981/11
Computer codes for safety analysis, including systems codes, which are used for the evaluation of LMFBR plants, have been developed along with the progress of the construction and the operation of experimental fast reactor "JOYO". On the other hand, a large number of data has been accumulated at O-arai Engineering Center and other laboratories both in Japan and abroad with the advance of safety R&D on Liquid Metal Fast Breeder Reactor (LMFBR). Various models of analysis have been proposed and many computer codes have been developed which are based on these models. This paper describes the essential part of the models and the functions of the codes, thus developed, which are used for the safety evaluation of LMFBR plant such as the prototype fast breeder reactor "MONJU".
Doi, Motoo*; *; *; *
PNC TN941 81-78, 32 Pages, 1981/04
Decay heat measurement by auxiliary cooling system was conducted to measure the decay heat after 75 MWt 1st cycle operation of the Joyo experimental fast reactor. Auxiliary primary pump was started after plant condition reached warm stand-by, then main primary and secondary pumps were stopped after parallel operation with auxiliary loop and primary loops. Test results: (1)Decay heat was approximately 470 kW. It contains the heat which lost from the auxiliary piping in the reactor vessel. (2)Auxiliary cooling system had enough capacity to remove the decay heat at this test condition. (3)After main primary pumps were stopped, temperature difference between hot leg and cold leg of the main primary loops soon reached the limitation of this test. So this test was stopped and plant condition was reverted to warm stand-by.
*; *; Tamura, Seiji*; Doi, Motoo*; *; Yamamoto, Hisashi*
PNC TN941 80-168, 53 Pages, 1980/12
Heat balance measurements and calculations were performed for the JOYO experimental fast reactor. Some pertinent results of this tests are presented below. (1)The heat removal rates calcalated using the air flowrates and 
Ts from the DHXs differed from heat removal rates obtained using measured primary sodium flowrates and Ts. The heat removal rates determined from the DHX data are about 11% larger and about 4% larger than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. The air flowrates and outlet temperatures from the DHX are measured on an 8
8 grids in the outlet area of the air cooler, using Pitot tubes and thermocouples. (2)The heat removal rates obtained using measured secondary sodium flowrates and DHX Ts are almost same as and about 8% smaller than the rates obtained from the primary sodium data for the A Loop and B Loop of JOYO, respectively. (3)The core heat generation rate (including blanket effects, etc.) calculated using the measured individual subassembly outlet temperatures, and subassembly flowrates, is about 7% larger than the value calculated using the primary sodoum flowrate and the reactor T. The reason for this discrepancy is conjectured to be due to the fact that the subassembly average outlet temperature is lower than the measured value, since the thermocouple is located in the center of the subassembly outlet channel. In both the A and B Loops of JOYO, the error between the heat removal rates from the measured DHX and the primary or secondary sodium flowrate and temperature data, and in the B Loop between the heat removal rates obtained from the primary sodium data and the secondary flowrate and DHX T, are greater than the error band of the instrumentation. The eauses of this unbalance in the heat removal are being investigated. THis report is based on the meetings held in February and March 1980 in the Reactor Technology Section ...
*; *; Hirose, Tadashi*
PNC TN941 80-205, 224 Pages, 1980/11
A decay heat removal test by in-vessel natural circulation was performed in september, 1978, at the end of the "JOYO" 50MW start up test period. Prior to the testing, preliminary analyses were performed. The test condition, limitations, and test schedule were decided based on the preliminary analyses results. A thermal stress analysis of the reactor vessel and the leak jacket was performed to verify its integrity by using the test result. The maximum heat removel capability was also predicted using the test result.
Yamamoto, Hisashi*; Sekiguchi, Yoshiyuki*; Hirose, Tadashi*; Sanda, Toshio*; Tamura, Seiji*; *; *
PNC TN941 80-179, 402 Pages, 1980/10
On April 24th, 1977, the initial criticality of JOYO was achieved and on July 5th, 1978, the reactor output reached rated power of 50 MW for the first time. The 75MW power ascension test was started in July, 1979, followed by two cycles of rated power operations, and the 100 hour nominal power continuous operation was completed in February, 1980. Through the tests for the core, plant it self, radiation shield and plant monitoring, the results proved satisfactory operation characteristics at 75MW. This report presents the summary of an the results obtained in the Test of MK-I core.
*; *; Yamamoto, Hisashi*
PNC TN941 80-171, 105 Pages, 1980/10
Initial Criticality testing of the JOYO experimental fast reactor was started in March, 1977 and low power testing was completed in November, 1977. This report describes the main results obtained during these sequences of tests, and compares design values (Preliminary calculated values) with the experimental results. The test results confirmed that the measured values for Safety Rank A items satisfied the requirements stated in the safety analysis report and the construction permit report. Safety Rank A items are reactor safety parameters such as : reactivity worth of control rod, reactivity insertion rate, scram time, sodium void reaetivity, etc. This report summarized the results of the two papers : "Start up Test Progress Situation - Critical and Low Power Tests -" and "Summary of JOYO Start up Test Results (Critical and Low Power Tests)", which were written during the testing periods.
*; *; *; *
PNC TN941 80-12, 233 Pages, 1980/02
A temperature and ternal stress analysis of the JOYO rotating plug was perfomed to verify its integrity at 50 MW power operating conditions. The "MSC/NASTRAN" finite element analysis computer program was used for the analyses. Thermal analyses were performed to determine the temperature profile of the rotating plug which was used as a thermal load for the thermal stress analysis. The thermal analysis results were compared with the observed temperature profile and the caluculations were repeated until the caluculated results were acceptable as the thermal load for the themal stress analysis. The analytic model for the themal stress analysis was divided into two pieces because of the computer memory capacity required to store the thermal loads. One model consisted of the large rotating plug, and the other was the small rotating plug. Maximum stress intensity (represented by the difference of the principal stresses) was 21 kg/mm
, and occured at the lower shield cylinder of the large rotating plug. This thermal stress exceeded the yield stress of the material, but did not exceed the stress intensity limit for combined primary and secondary stresses.
Yamamoto, Hisashi*; Sekiguchi, Yoshiyuki*; Tamura, Seiji*; *; *; Sasaki, Makoto; *
PNC TN941 79-164, 185 Pages, 1979/09
The 75MW power ascension test of "JOYO" was started in July, 1979, followed by the successful 50MW power ascension test and two cycles of rated power operations. In the 75MW test reactor power was raised up to 65MW at first in order to confirm the characteristics of the plant stability and safety. On July 16th, 1979, the reactor out put reached the rated power of 75MW for the first time. Through the tests for the core, plant it self, radiation shield and plant monitoring, the results proved satisfactory operation characteristics at 75MW. The power ascension test was completed, except for the 100 hour nominal power continuous operation which is scheduled in January, 1980. This report presents the summary of the main results obtained in the test conducted untill August 1979.
*; *; *
PNC TN952 77-04, 230 Pages, 1977/05
A computer code -PLANT76- has been developed to analyze the plant dynamics behaviors of the Japan Experimental Fast Reactor "JOYO", which include such postulated transient accidents as the loss of power, the pump failure or the reactivity incertion. A original program of this code which was so called "Plant" was developed at JAERI. The main modification of this code from the original code is that the primary and secondary coolant flow rate are calculated from the momentum conservation equations with each time step, instead of being provided from the input data. This report is written as user's guide of computer code. A detailed descriptions of a basic equation, analytical model, input and output form, and plotting routine, are also given in this report. Besides, two sample calculations of the primary and secondary flow coast down curves are presented in this report comparing with the experimental results. The above results show that this code was applicable to the JOYO plant behavior analysis. The detail analysis of the dynamic behaviors with this code will be presented at another report.
*; *
PNC TN941 77-76, 70 Pages, 1977/01
This report presents the preliminary analytical results of the dynamic behavior of JOYO plant with the PLANT 76 code. These items considered here are as follows, (1)Primary flow coast down (2)Secondary flow coast down (3)Loss of off-site electrical power (4)Loss of electrical power to one primary pump (5)Loss of electrical power to one secondary pump (6)Manual reactor shutdown The analyses of the primary and the secondary coastdowns were made to compare the analytical results with the preoperational test data of JOYO. It was found out that the analytical flow coast down curves were good agreement with the experimental data. The calculation of the other items (3) through (6) were performed to predict the behavior of the plant dynamic test of JOYO which would be scheduled to be performed in the next year.
