Profile SFR-42 SGTF INDIA EXPERIMENTAL FACILITIES FOR THE DEVELOPMENT AND DEPLOYMENT OF LIQUID METAL COOLED FAST NEUTRON SYSTEMS

Transcription

1 Profile SFR-42 SGTF INDIA EXPERIMENTAL FACILITIES FOR THE DEVELOPMENT AND DEPLOYMENT OF LIQUID METAL COOLED FAST NEUTRON SYSTEMS GENERAL INFORMATION NAME OF THE FACILITY ACRONYM COOLANT(S) OF THE FACILITY LOCATION (address) OPERATOR CONTACT PERSON STEAM GENERATOR TEST FACILITY SGTF Sodium Fast Reactor Technology Group (FRTG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India FRTG, IGCAR K.K.Rajan, Director, Fast Reactor Technology Group, (name, address, institute, Indira Gandhi Centre for Atomic Research, Kalpakkam-function, telephone, ) , India, STATUS OF THE FACILITY In operation Start of operation (Date): 2003 MAIN RESEARCH FIELD(S) Zero power facility for V&V and licensing purposes Design Basis Accidents (DBA) and Design Extended Conditions (DEC) Thermal-hydraulics Coolant chemistry Materials Systems and components Instrumentation & ISI & R TECHNICAL DESCRIPTION Description of the facility - 1 -

2 Steam Generator Test Facility (SGTF) is set up to optimize the design of once through steam generators for Indian fast breeder reactors. At present experiments are carried out on a 23 m long, 19 tube steam generator which is a model of the steam generator used in the 500 MWe Prototype Fast Breeder Reactor (PFBR) being constructed at IGCAR. SGTF consists of an oil fired heater of capacity 5.7 MWt to heat the sodium from 355 C to the rated temperature of 525 C. An annular linear induction type electromagnetic pump of 170 m 3 /h capacity is used for maintaining sodium flow in the system. The main sodium circuit consists of the fired heater, electromagnetic sodium pump, surge tank, buffer tank, sodium to air heat exchanger and steam generator (SG). It also has online purification system, cover gas system and reaction products discharge circuit. Steam generator produces superheated steam at 172 bar and 493 C. The high pressure, high temperature steam from steam generator is depressurized and de superheated in three stages before condensing in the condenser and the condensate is pumped back to steam generator. A dedicated hydrogen leak detection system is installed to detect the failure of steam generator tubes which results in sodium-water reaction. The facility was commissioned in the year 2004 and the steam generator has so far operated in power for 10,000 hours. ACCEPTANCE OF RADIOACTIVE MATERIALS - No Scheme/Diagram - 2 -

3 FIG. 1. Scheme of the SGTF facility - 3 -

4 FIG. 2. Comparison of SGTF model SG and PFBR SG - 4 -

5 3D Drawing/photo FIG. 3. View of the SGTF facility - 5 -

6 Parameters table Coolant inventory Power Test sections Steam generator Coolant chemistry measurement and control (active or not, measured parameters Instrumentation 15 tonne Furnace oil fired heater of capacity 5.7 MW Characteristic dimensions 19 tube model steam generator Tube OD :17.2 mm Tube ID : 12.6 mm Tube length : 23 m Shell : 184 mm OD, 10 mm thick Shell and tube Material : Mod 9 Cr-1 Mo Static/Dynamic experiment Dynamic Temperature in the test section Water/Steam side : 235 to 493 C Sodium side : 355 to 525 C Operating pressure and design pressure Shell side Operating pressure : 5 bar (g) Design pressure : 105 bar (g) Tube side Operating pressure : 172 bar (g) Design pressure : 180.3bar (g) Flow range (mass velocity etc) Sodium flow : 105 m 3 /h Water/steam flow : 8,820 kg/h Coolant is not active Coolant purity is maintained by cold trapping, and monitored using online plugging indicator and periodical sampling and analysis I. Thermocouples for temperature measurement II. Wire type and spark plug type leak detectors, and sodium aerosol detectors for sodium leak detection III. Resistance type discontinuous and mutual inductance type continuous level probes for monitoring sodium level - 6 -

7 COMPLETED EXPERIMENTAL CAMPAIGNS: MAIN RESULTS AND ACHIEVEMENTS Endurance testing of steam generator was carried out by operating SG at rated conditions continuously for a month. The operation was smooth that validated the design of once through steam generator. Experiments were carried out and the flow stability mapping giving stable operating region was obtained. Performance of thermal baffles in protecting steam generator tube sheets under various plant transients was experimentally tested and found adequate. Heat transfer area margin available on the steam generator was experimentally obtained. Feasibility of using acoustic based system for detecting steam generator tube leak was tested and a similar system for PFBR SG is being developed. Assessment of flow induced vibration of steam generator tube bundle was studied and found within limits proving the adequacy of tube support system. Operation of steam generator with one tube plugged condition was demonstrated ruling out any thermal loading on surrounding tubes. PLANNED EXPERIMENTS (including time schedule) i. Assessment of pressure drop characteristics of steam generator tubes under various loads. ii. Assessing the performance of SG thermal baffles under plant transients with sympathetic safety actions as envisaged in PFBR. iii. Further studies on acoustic based steam generator tube leak detection system for the fine tuning of system. TRAINING ACTIVITIES Training activities can be considered with IGCAR Kalpakkam for the operation of experimental campaign under the supervision of IGCAR qualified staff. REFERENCES 1. NUCLEAR ENGINEERING & DESIGN Steam generator test facility A test bed for steam generators of Indian sodium cooled fast breeder reactors,, 248 (2012)