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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.
*; *; 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.
Doi, Motoo*; Endo, Masayuki*; *; *; *; Wada, Hozumi*; Hirose, Tadashi*
PNC TN941 80-65, 269 Pages, 1980/05
The purpose of this test is to confirm that the heat transfer characteristics of the intermediate heat exchangers (IHX) and the dump heat exchangers (DHX) satisfy their design values. Each primary and secondary heat transport systems has two loops (A&B) and each secondary loop has two DHX's as terminal heat exchangers. The IHX/DHX heat transfer characteristics are measured under normal steady-state operating conditions (at 25, 40, 50, 65 and 75MWt), and also under special steady-state conditions (low sodium temperature, low secondary sodium flow rate) The major test results were : (1)The heat transfer coeffients of the IHX's and the DHX's measured in the reactor power range of 50 to 75 MWt closely matched predicted values. (2)The pressure drop in the DHX air flow, duct based on measured blower outlet pressure was less than the predicted value. This report presents the results of these plant tests.
*; *; *; *; *; Hirose, Tadashi*; *
PNC TN941 79-161, 50 Pages, 1979/10
A radionuclide-analysis and a determination of activity of the waste argon and nitrogen gas were carried out during the power ascension test. The sampling of the gas was carried out by means of stainless steel vaccum vessels or vinyl bags and gamma activities were analyzed with a multi channel pulse height analyzer (MCA) and a vibrating reed electrometer (VRE). 
Ar was observed as the major part of gamma activity. In addition to Ar, some long-lived radionuclides were observed in the decay measurement with VRE. Concerning long-lived radionuclides, it was confirmed that some amounts of 
H existed as a part of activity and some amounts of 
C were considered to be in the waste gas by judging several charactiristic informations of the gas. The concentration of radionuclides in the waste gas were also evaluated. No relationship between the concentrations and a reactor power were found. Above results, however, indicated such general tendency that the concentration of this long-lived radionuclides released from the argon and nitrogen waste gas are H and Ar. The values of concentrations of the radionuclides and release rates of the waste gas were well below the control criteria in this test period.
Hirose, Tadashi*; *; Nanashima, Takeshi*; *; *; Endo, Masayuki*
PNC TN941 79-119, 31 Pages, 1979/08
The isothermal reactivity coefficient of JOYO reactor was measured for the minimum critical core configuration and for the initial power up core configuration. The isothermal reactivity coefficient was obtained at every 10C increase or decrease of primary coolant temperature by measuring excess reactivity of the reactor core. The average rate of sodium temperature, change was about 
6C/hr. The temperature dependency of latch position and hold position of regulating rods was also investigated, which had an infuluence on the critical point. Results are as follows; (1)The measured value of isothermal reactivity coefficient for 64 (minimum critical mass core) and 70 (Initial core) core sub-assemblies were -3.67 
10
%
K/K/C and -3.7710%K/K/C respectively which were in good ageement with design value of -3.63 10%K/K/C. (2)Analysed value for 70 core sub-assemblies configulation was -3.76 10%K/K/C (analysed by CITATION code with JFSVII micro cross-section).
Hirose, Tadashi*; Endo, Masayuki*; Nanashima, Takeshi*; Doi, Motoo*; Enomoto, Toshihiko*; Suzuki, Yukio*; Sekiguchi, Yoshiyuki*; Yamamoto, Hisashi*
PNC TN941 79-91, 81 Pages, 1978/12
This system is used to remove reactor decay heat in cases where the main cooling system is inoperable for unexpected reasons, a lower than normal sodium level exists in the R/V or during in-service inspection in the R/V. The purpose of this test is to verify that the design heat removal rate (2.6MWt) can be achieved by the Auxiliary cooling system. With the sodium level lowered below main cooling system outlet nozzles and the coolant temperature (DHX outlet temperature) controlled at 250 C, the reactor power was increased first to approx. 1MW (1.16 MWt actual) and then to 2 MW (2.16 MWt actual) to provide the "decay heat". At both steps, steady-state conditions were verified and test data were recorded, from which the heat removal rate at design conditions was calculated. (Testing was terminated after the second step to maintain the calculated distortion of the partially-filled R/V within prescribed limits.) Test Results : At the second test condition (reactor power = 2.16 MWt) the R/V inlet Na temperature of 267 C corresponded to a 72% open DHX inlet vane setting. Extraporating this to the design condition (R/V inlet temperature = 370C), a 100% DHX vane opening would permit the removal of 3.1 MWt decay heat.
Hirose, Tadashi*; Endo, Masayuki*; Nanashima, Takeshi*; *; *; Yamamoto, Hisashi*
PNC TN941 79-90, 47 Pages, 1978/12
The purposes of this test are to : (1)verify the procedure for decreasing reactor power from 50 MWt to shutdown, (2)remove the decay heat after shutdoun by the main heat-transport system using the normal procedure, (3)confirm that the rate of temperature decrease does not exceed -50C/hr during the cooldown. At the end of a full power (50MWt) run of 100hours, record steady state plant data. By insertion of one regulating rod, decrease reactor power until PRM indicates 80% (40MWt). Verify that a steady-state condition has been achieved (at least 15 minutes) and repeat data recording. Following same procedure, decrease power stepwise to 30 MWt, 20MWt and 10MWt. Stop the DHX Blowers at 10MWt. Decrease reactor power with one regulatiag rod to 1MWt as indicated on IRM, then shut down the reactor by slow scram. By manual operation of DHX vanes and dampers, decrease coolant temperature as rapidly as possible to the "warm standby" condition (The maximum allowable cooldown rate of 50C/hr must not be exceeded). Test results : (1)The main coolant temperature decrease rate averaged 34C/hr (maximum of 48C/hr) during the cooldown from full power to reator shutdown. During this period, the insertion rate of one regulating rod was approx. 1mm/minute. (2)The average rate of main coolant temperature decrease from shut down to "warm standby" was 35C/hr in manual control.
Nanashima, Takeshi*; Endo, Masayuki*; Enomoto, Toshihiko*; Hirose, Tadashi*; Yamamoto, Hisashi*
PNC TN941 79-128, 204 Pages, 1978/12
The purpose of this test is to confirm that the heat transfer characteristics of the intermediate heat exchangers (IHX) and dump heat exchanger (DHX) satisfy the design values. Each primary and secondary heat transport system has two loops (A&B) and each has an IHX. Each secondary loop has two DHX's (a total of 4 ). However, only one DHX (unit 2B) is instrumented to provide the required test data. The IHX/DHX heat transfer characteristics are measured under steady-state operating conditions at least four times (at 15, 25, 40 and 50 MWt). Test results : (1) The heat transfer coefficient for the A-loop IHX is approximately 60% of the expected value. The B-loop IHX value is approx 120% of expected. (2) The DHX 2B air-side heat transfer coefficients measured in the range of 25 to 50 MWt (Reactor power) closely mateh predicted values. The pressure drop in the DHX air flow duct based on measured blower outlet pressure is less than the predicted value.
Yamamoto, Hisashi*; *; Hirose, Tadashi*; Sanda, Toshio*; *
PNC TN941 78-143, 310 Pages, 1978/01
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 testing was conducted as criticality test, low power test and power raise test, through the testing about 40 different test items were conducted to assure characteristics of both reactor core and plant system. All tests were conducted as scheduled, and completed on September 16th, 1978. Results of the tests proved that JOYO is operated at 50 MW with safe and stable characteristics. This report presents brief description of test results from the initial criticality through the power up test.
Yamamoto, Hisashi*; *; *; *; Sanda, Toshio*; *; Hirose, Tadashi*
PNC TN908 78-04, 315 Pages, 1976/12
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