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The Prototype Fast Breeder Reactor (PFBR) is pool-type, sodium-cooled reactor. PFBR has eight steam generators (SGs) in its secondary circuit. The availability of SG units in the secondary circuit ...

Cold trap is a purification unit used in sodium cooled Fast Spectrum Reactors (FSRs) for maintaining the oxygen and hydrogen level in sodium within acceptable limits. It works on the principle of crystallization and precipitation of oxides and hydrides of sodium in a wire mesh, when the temperature of sodium is reduced below the saturation temperature. The sodium hydride gets accumulated in the secondary cold trap as a consequence of the continuous diffusion of hydrogen in sodium and with time the trap is fully loaded and becomes inoperable. The removal of these impurity deposits at intervals by keeping the cold trap in same location of the loop is known as in situ regeneration. After regeneration cold trap is qualified by gamma radiograph technique to ensure adequate removal of impurities before bringing the cold trap back to service in sodium for purification. The numerical results predict the impurity deposition pattern in the wire mesh region of cold trap. The mathematical model has been validated with experimental data obtained from model cold trap. This paper discusses the methodologies developed for qualification of regeneration of cold trap using radiography and image processing techniques for assessing impurity deposition before and after regeneration.

An ultrasonic under-sodium scanner has been developed for deployment in Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India. Its purpose is to scan the above-core plenum for detection, if any, of displacement of sub-assemblies. During its burn-up in the reactor, the head of a Fuel Sub-Assembly (FSA) may undergo a lateral shift from its original position (called `bowing') due to the fast neutron induced damage on its structural material. A simple scanning technique has been developed for measuring the extent of bowing in-situ. This paper describes a PC-controlled mock-up of the scanner used to implement the scanning technique and the results obtained of scanning a mock-up FSA head under water. The details of the liquid-sodium proof transducer developed for use in the PFBR scanner and its performance are also discussed.

The Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam, India, is a 500 MWe sodium cooled pool type reactor. The core of the PFBR consists of 1758 free standing subassemblies supported on the grid plate. The entire core is divided into 15 different flow zones and the flow rate required through each zone is calculated based on the fission heat generation. The coolant sodium flows from the bottom of the subassembly to top and the design of the subassembly for each flow zone is quite complex. There are 181 fuel subassemblies in PFBR core with 217 fuel pins in each subassembly, vertically held in the form of bundle within a hexagonal wrapper tube. The pins are separated by spacer wires wound around the pins helically. Analytical prediction of subassembly pressure drop, vibration and determination of inception of cavitation for this complex geometry is very difficult. So experiments were conducted extensively to get a more accurate evaluation of the design and for its qualification for the use in PFBR, which is designed for 40 years of operation. Pressure drop and cavitation experiments were carried out in water on full scale (1:1) subassemblies of all flow zones. The overall pressure drop of the subassembly determines the ratings of the pump. Cavitation of the pressure drop devices lead to erosion damage of fuelpins and may also result in reactivity fluctuation due to sodium-void effect. So it is essential to confirm that the subassembly is not cavitating in the operating regime of the reactor. Subassembly can vibrate in cantilever mode due to the turbulence in the flow and can result in reactivity fluctuation, reactor control problem and can even lead to the failure of the fuel pins. So vibration measurements were carried out in water on the maximum rated subassembly. This paper discusses various experiments carried out on PFBR subassembly, the similarity criteria followed, instrumentation, results and conclusion.

Wire type leak detectors working on conductivity principle are used for detecting sodium leak in the secondary sodium circuits of fast breeder reactors. It is required to assess the performance of these detectors and confirm that they are meeting the requirements. A test facility by name LEENA was constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam to test the wire type leak detector lay out by simulating different sodium leak rates. This test facility consists of a sodium dump tank, a test vessel, interconnecting pipelines with valves, micro filter and test section with leak simulators. There are three different test sections in the test set up of length 1000 mm each. These test sections simulate piping of Prototype Fast Breeder Reactor (PFBR) secondary circuit and the wire type leak detector layout in full scale. All test sections are provided with leak simulators. A leak simulator consists of a hole of size one mm drilled in the test section and closed with a tapered pin. The tapered pin position in the hole is adjusted by a screw mechanism and there by the annular gap of flow area is varied for getting different leak rates. Various experiments were conducted to evaluate the performance of the leak detectors by creating different sodium leak rates. This paper deals with the details of wire type leak detector layout for the secondary sodium circuit of PFBR, performance requirement of leak detection system as per codes, description of test facility, experimental procedure, test results of various experiments conducted and details of measurements made on contact resistance of detector after sodium leak. Paper also reviews the experiment conducted in CEA, Cadrache and compares with results of present experimental study.

Electromagnetic Pumps have been used for pumping liquid sodium in auxiliary circuits such as fill and drain and purification circuits of sodium cooled fast breeder reactors. Despite their low efficiency these pumps are used in fast reactors because of their high reliability and low maintenance due to absence of moving parts. Besides, EM Pumps can be used for pumping impure sodium. IGCAR has developed electromagnetic pumps of various capacities and successfully used them in experimental facilities. This paper deals with the design, development and performance testing of a large electromagnetic pump called Annular Linear Induction Pump (ALIP). This 170 m3/h capacity ALIP is for use in PFBR Secondary Sodium Fill and Drain Circuit (SSFDC) and was introduced in the sodium circuit of SGTF for testing its performance. In this type of pump, a linearly traveling magnetic field is generated by means of circular windings placed spatially apart in slots and excited by 3-phase supply. This traveling field induces circulating currents in liquid sodium which generates secondary magnetic field. Interaction of primary magnetic field and secondary magnetic field produces pumping force on liquid sodium. The pump duct is made of SS316L. Pump winding is made up of copper with class H insulation. The pump is designed for 360 V and for a maximum sodium temperature of 450 °C. The pump is a reflux type of pump with both inlet and outlet on the same side. The pump was tested in the cold leg of Steam Generator Test Facility (SGTF) and its performance characteristics were obtained. The efficiency of the pump was also calculated and compared with the theoretical value. The successful testing and operation of the pump in SGTF has indicated sound design and indigenous manufacturing capability. This paper describes the design data of the pump and details of the testing carried out in IGCAR

Two independent sodium loops under common name INSOT facilities were constructed in Fast Reactor Technology Group, IGCAR for conducting material testing of PFBR components in dynamic sodium. One loop is utilized for in-sodium Low Cycle Fatigue (LCF) and Creep - Fatigue Interaction (CFI) studies and the second loop for in sodium creep studies. The loop components and piping were constructed using AISI type 316LN/316L material. The sodium loops were constructed by adopting stringent quality control measures stipulated in relevant international codes. Suitable creep and fatigue testing chambers were designed and developed indigenously. The test chambers were qualified for sodium service before commencement of experiments by conducting trial runs. The sodium purity was achieved by on-line purification method prior to admitting into the test chambers and the purity was maintained during testing. Characterization of sodium in the loops during test run were carried out by foil equilibration technique, electro-chemical carbon meter and chemical analysis of sodium sample drawn from the loop using state of art sampling techniques. A versatile PLC based instrumentation system has been successfully implemented for monitoring and controlling the process parameters. Safety interlocks were provided by hardware logic to ensure reliability.

Cold trap is a purification unit used in sodium system of FBR's for maintaining the oxygen/hydrogen level in sodium within acceptable limits. It works on the principle of crystallization and precipitation of oxides/hydrides of sodium in a wire mesh, when the temperature of sodium is reduced below the saturation temperature. The cold traps presently used have lower effectiveness and get plugged prematurely. The plugged cold traps are cleaned and then put back into service. Frequent cleaning of cold trap results in the long down time of the sodium system. New design of cold trap has been conceived to overcome the above problems. The mathematical modeling for the new design was carried out and validated with experimentally tested results for its effectiveness. This paper shares the experience gained on the new design of cold trap for FBR's.

Sodium is used as a coolant in Liquid Metal Fast Breeder Reactor (LMFBR). Sodium flow measurement is of prime importance both from the operational and safety aspects of a fast reactor. Various types of flowmeters namely permanent magnet, saddle type and eddy current flowmeters are used in FBRs. From the safety point of view flow through the core should be assured under all operating conditions. This requires a flow sensor which can withstand the high temperature sodium environment and can meet the dimensional constraints and be amenable to maintenance. Eddy current flowmeter (ECFM) is one such device which meets these requirements. It is meant for measuring flow in PFBR primary pump and also at the outlets of the fuel sub-assemblies to detect flow blockage. A simulation model of ECFM was made and output of ECFM was predicted for various flowrates and temperatures. The simulation model was validated by testing in a sodium loop. This paper deals with the design, simulation and tests conducted in sodium for the eddy current flowmeter for use in the Prototype Fast Breeder Reactor (PFBR).

Prototype Fast Breeder Reactor (PFBR) is 500 MWe sodium cooled reactor under construction at Kalpakkam, India. To improve the reliability of shutdown, Diverse Safety Rods (DSRs) are used in-addition to normal Control and Safety rods. During reactor operating condition, the DSR is parked above the active core and held in its top position by an electromagnet. In the event of a scram signal from the safety logic, the electromagnet holding the DSR is de-energised. Hence the DSR is released into the active core and at the end of travel DSR gets deposited in its bottom position. Because of the mechanical constraints, hard wired connectivity is not permitted from the DSR subassembly to the instrumentation outside the reactor. Hence an inductively coupled Eddy Current Position Sensor (ECPS) has been conceptualized to detect that the DSR has reached its bottom most position and to measure the drop time. Results of feasibility study on laboratory model have been reported earlier. Testing of a 1:1 scale engineering model of ECPS is reported in this paper. Results obtained from the high temperature sodium testing of ECPS indicate a clearly measurable change in pick up voltage with sensitivity of 11 % at 675 Hz. The ECPS is in advanced stage of implementation in DSRDM of PFBR.

In Prototype Fast Breeder Reactor (PFBR), Kalpakkam, India, there are three Diverse Safety Rod Drive Mechanisms (DSRDM) in the control plug, which hold the diverse safety rods (DSR). During normal operation, DSR is held outside the active core region by an electromagnet. On receiving Safety Control Rod Accelerated Motion (SCRAM) signal, the electromagnet de-energizes and drops the DSR, which falls under gravity in sodium. The free fall time is required to be monitored during each SCRAM action. An innovative eddy current based position sensor is designed for this purpose. It is not possible for the sensor to have any wired connectivity with the instrumentation. To circumvent this problem an indirect excitation through inductive coupling is incorporated in the design of the sensor. This paper describes the design and testing of the sensor configuration on a one-to-one model. The results have established the feasibility of using inductively coupled eddy current position sensor for DSR in PFBR.

Wire type leak detectors working on conductivity principle are used for detecting sodium leak in the secondary sodium circuits of FBRs. It is required to assess the performance of these detectors and confirm that they are meeting the requirements. A test facility by name LEENA was constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam to test the wire type leak detector lay out by simulating sodium leaks of different rates. This test facility consists of a sodium dump tank, a test vessel, interconnecting pipelines with valves, micro filter and test section with leak simulators. There are three different test sections in the test set up of length 1000 mm each. These test sections simulate piping of Prototype Fast Breeder Reactor (PFBR) secondary circuit and the leak detector layout in full scale. All test sections are provided with leak simulator. A leak simulator consists of a hole of size one mm drilled in the test section and closed with a tapered pin. The pin position is adjusted by a screw mechanism and there by the annular gap of flow area is varied for getting different leak rates. Test facility was commissioned and 20 experiments were attempted at 350°C to 550°C. Out of 20 experiments 11 experiments were successfully completed and 9 experiments were terminated in between due to the choke in the simulator hole. From the experimental data it is found that sodium leak rate of 200 g/h and above can be detected within 6 hours. A relationship between leak rate and detection time was established from the experimental results and found that sodium leak rate of 100g/h is likely to be detected in 11.4 hours. This paper deals with the details of wire type leak detector layout for the secondary sodium circuit of PFBR, performance requirement of leak detection system as per codes, description of test facility, experimental procedure and test results. Paper also reviews the experiment conducted in CEA, Cadrache and compares with results of present experimental study.Copyright © 2009 by ASME

Kamini (Kalpakkam Mini) reactor is a U233 fuelled, demineralised light water moderated and cooled, beryllium oxide reflected, low power (30 kW) nuclear research reactor. This reactor functions as a neutron source with a flux of 1012 n/cm2 s−1 at core centre with facilitates for carrying out neutron radiography, neutron activation analysis and neutron shielding experiments. There are two beam tubes for neutron radiography. The length/diameter ratio of the collimators is about 160 and the aperture size is 220 mm×70 mm. Flux at the outer end of the beam tube is ∼106–107 n/cm2 s. The north end beam tube is for radiography of inactive object while the south side beam tube is for radiography of radioactive objects. The availability of high neutron flux coupled with good collimated beam provides high quality radiographs with short exposure time. The reactor being a unique national facility for neutron radiography has been utilized in the examination of irradiated components, aero engine turbine blades, riveted p...

Acoustic techniques find wide application in Liquid Metal Fast Breeder Reactors (LMFBRs) for ensuring its high reliability, safety and plant availability. Various surveillance methods based on acoustic technique can be employed in these reactors to detect deviations from normal operating conditions. This could be used for the measurement of drop time of Diverse Safety Rods (DSRs) in the core, detection of in-sodium water leaks in Steam Generators, cavitation detection in sodium pumps and reactor core components. An active R&D program is being pursued in these areas at Indira Gandhi Centre for Atomic Research. Acoustic measurement technique has been used to determine the drop time of Diverse Safety Rods in sodium. 3 nos of Diverse Safety Rods (DSRs) are provided in Prototype Fast Breeder Reactor (PFBR) for its safe shut down in case of a SCRAM. An online drop time measurement system using acoustic technique is planned to detect the proper insertion of DSRs into their corresponding Subassemblies. Experiments were conducted during the performance testing of DSRs in sodium using accelerometer instrumented wave-guide system and free fall time and braking time of DSR have been measured. For detection of in-sodium water leaks in Steam Generators, acoustic method serves as a supplementary monitoring technique with an intermediate sensitivity and instantaneous response. To develop an acoustic leak detection system for Steam Generators of Prototype Fast Breeder Reactor, preliminary studies on the behavior of micro leak and its propagation has been carried out in Sodium Water Reaction Test Rig, injecting steam into sodium. Acoustic technique was employed to characterize the onset of leak. Cavitation in LMFBRs can occur in fuel subassemblies, pressure drop devices, pumps etc. It is important to minimize cavitation to reduce the risk of damage from erosion. Acoustic technique was extensively used in qualifying Prototype Fast Breeder Reactor components against cavitation phenomenon. This paper discusses the various experiments carried out towards the development of the acoustic surveillance methods for FBRs, instrumentation involved, results obtained from experiments and brief details of the future programme.

Once-through Steam Generator (SG) is a critical component of Liquid Metal Fast Breeder Reactor (LMFBR) plant. It is a counter current heat exchanger, in which heat is transferred from the hot sodium flowing on the shell side to water/steam in tube side. High pressure subcooled water enters the SG tube from bottom, gets heated up to saturation, goes through nucleate boiling, dry out and post dry out heat transfer, getting converted to saturated steam and finally gets superheated. For this the process design needs to be carried out accurately. A computer code DESOPT has been developed for the process design of straight vertical, serpentine and helical geometries and validated against reported designs in literature. Recently a test facility to test a 5.5 MWt sodium heated steam generator has been commissioned. The predictions of the code have been compared with the measurements and found satisfactory. This paper brings out different heat transfer mechanisms in SG and describes the one-dimensional code, its validation based on literature and in-house tests and presents the results of comparison between predicted and actual operation at different part loads.

The Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam, India. PFBR has eight numbers of sodium heated once through Steam Generators (SG) each with 157 MWt capacity. Parts like steam generator tube sheet will see high temperature gradients during normal operation as well as in transient conditions. Thermal baffles are provided to protect the steam generator tub e sheets from the high temperature gradients during transient and steady state conditions. In Steam Generator Test Facility (SGTF) of IGCAR a 5.5MWt capacity SG was tested for its thermal hydraulic performance. Effectiveness of the thermal baffles provided in the model SG of PFBR is studied numerically and the performance of the same has been predicted. Further Transient experiments were conducted in Steam Generator Test Facility for evaluating the same. This paper brings out the details of numerical studies and the experimental evaluation of the effectiveness of thermal baffles provided to protect the steam generator tube sheets.

Neutron Radiography has emerged in recent years as an important non-destructive test method for post-irradiation examination of irradiated objects. A swimming pool type reactor (30 KW) in the basement of hot cells, having neutron flux of 3×1010 at reflector surface is to be utilised for neutron radiography. The examination is carried out by lowering the objects in front of neutron beam from one of the cells through a containment and guide tube which is an integral part of the hot cell extending to the basement below it thus avoiding transfer of irradiated objects from hot cells for the purpose of neutron radiography. This approach has many advantages, but it puts severe constraints on the design and safety aspects of neutron radiography rig. A moving cage holds the object and is lowered in the containment and guide tube for neutron radiography. Cassette feed mechanism and controls have been devised such that it is possible to take six shots without human entry to the neutron radiography area.