Research and Development of Advanced Nuclear System      
500

500
Water in downcomer (Exp.)
Water in heater tube (Exp.)
Sodium (Exp.)
Experimental range (Exp.)
Water in downcomer (NETFLOW)
Water in heater tube (NETFLOW)
Sodium (NETFLOW)

400

Superheater
490
Temperature (℃)

Temperature (℃)

450

350

300

480470Sodium (Exp.)
Experimental range (Exp.)
Sodium (NETFLOW)

Evaporator
460

250

Effective heat transfer coiling region
Effective heat transfer coiling region

200
2Reactor exit (Exp.)
IHX primary exit (Exp.)
IHX secondary inlet (Exp.)
IHX secondary exit (Exp.)
Reactor exit (NETFLOW)IHX primary exit (NETFLOW)IHX secondary inlet (NETFLOW)IHX secondary exit (NETFLOW)

500
450

Temperature (℃)

1400

34
Position (m)

56600
500
Primary (Exp.)
Secondary (Exp.)
Primary (NETFLOW)
Secondary (NETFLOW)

400

350

300

300

200

250

100200

0

2000

4000
6000
Time (sec)

8000

7Top010000

Flow rate (kg/s)-10
Bottom450-0.50
Bottom0.5
1Position (m)1.52Top     


Simulation of sodium temperature by the shell and water / vapor
temperature in tubes of steam generators at rated output using theNETFLOW code. Agrees well with actual boiling behavior in theevaporator.

  
  


 
A turbine trip test was simulated by the NETFLOW code, modeling thefirst through the third loop of Monju. Important plant parameters areshown. Good agreement was obtained between test and calculation
results over 10000 seconds.

                                                                                                                   

Mochizuki, H., Verification of NETFLOW Code using Plant Data of Sodium Cooled Reactor and Facility, Nuclear Engineering and Design, vol.237,
issue 1, 2007, p.87-93.

24 JAEA R&D Review 2006