Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
*; *
PNC TJ9381 94-001, 84 Pages, 1994/02
The study in last year examined the realization of 1000MWt natural circulation ATR. In this study. The sensitivity analysis is did making for the design. The results are as below. (1)The influence of riser pipe diameter. The natural circulation on riser pipe 3B is analyzed. 3B pipe is cheaper than 5B riser pipe that is a base condition. The results are that fuel dryout is happen by 1.8MW channel power on 3B, but is not happen on 5B riser pipe. (2)The influence of the riser pipe number reduction and the riser pipe diameter change. The case of connection 2 pressure tubes by Y piece pipe is analyzed. In this case, the riser pipe number is half. The cases that riser pipe is 5B and 4B are analyzed. The results are as below. (a)Natural circulation flow lower by riser pipe number is half. (b)Fuel dryout is happen by 1.8MW channel power on 4B, but is not happen on 5B. (3)The influence of the middle position header connection and the riser pipe diameter change, The case of connection 50 pressure tubes by the middle position header is analysed. The cases that riser pipe is 24B and 32B are analyzed. The results are as below. (a)In 2 cases, natural circulation flow lower as compared with base case. (b)In 2 cases, fuel dryout are not happen by 1.8MW channel power. (4)The influence of pressure tube diameter. The case of pressure tube diameter widening is analyzed. So, the pressure loss of pressure tube domain govern the flow. (a)Natural circulation flow increase by pressure tube diameter widening. (b)The flow swing is by pressure tube diameter widening. (c)In analysis cases, fuel dryout is not happen. We get 1.8MW channel power on the condition that riser pipe is 4B, riser pipe is connected by Y piece pipe, pressure tube diameter is 122.0mm.
*; *; Uno, Tetsuro*
PNC TN9410 86-029, 68 Pages, 1986/02
A new test facility "Air-Cooling Thermal Transient Test Facility" (ATTF) was constructed at O-arai Engineering Center. This test facility is utilized, in the first place, for evaluating the strength of outlet tube-sheets of steam generators of FBR Plants. The objectives of the tube-sheet model tests are as follows. The first is to investigate and evaluate the strain concentration in the plastic region. The second is to confirm the adequacy of the design criteria for the prototype reactor MONJU. The third is to confirm the safety margin for failure incorporated in the design evaluation methods. ATTF can impose severe thermal loadings (only cold shock) on the test specimens. The facility produces compressed air (Max. 35kg/cm
G) by two large-sized compressors, and stores it in a storage tank (about 60m
). After a test specimen is heated up to the aimed temperature the compressed air passes through the test specimens quickly by opening the valve to apply cold shock and is released in the atmosphere. Each main loop pipe is 8 inches in diameter and the flow rate is max. 10kg/s in compressed air. The most severe down thermal transient condition is from 550
C to 150C (for tube-sheet model) in about 4 min. The test section can be modified for various kinds of structures, which should be air-tight and have the maximum pressure of 8kg/cmG. The facility is operated automatically by two sequencer controllers. One of the main features of ATTF is the adoption of compressed air instead of sodium as coolant. By using compressed air, various kinds of sensors which can not be used in the sodium environment can be used in ATTF; particularly strain gages can be used effectively to obtain strain distribution for thermal transient condition, and the location as well as the mode of failure of test specimens can be recognized easily through the detection of crack initiation and the observation of crack growth. ATTF is expected to be a powerful ...
*; *; Kuroha, Mitsuo; *; *; *; *
PNC TN941 85-53, 144 Pages, 1985/03
SWAC13E is a one-dimensional thermo-hydraulic computer program to analyze large scale sodium-water reaction accidents in an LMFBR steam generator. The code is the advanced version of SWAC13, the long-term hydraulic analysis module of SWACS; the energy conservation is taken into consideration in the new version to add the function to analyze the temperature behavior of the reaction. The present document covers the validation study of the code by using the large leak data of the Steam Generator Safety Test Facility (SWAT-3). The analytical parameters are as follows: (1)Model of relative velocity. (2)Void/droplet density. (3)The number of nodes where water leaks. (4)Reaction heat. It is concluded that the code can analyze the phenomena with a reasonable conservatism by choosing the proper value of the parameters.
