Safety Practices in Chemical and Nuclear Industries

Transcription

1 Lecture 10 Safety Practices in Chemical and Nuclear Industries Fast Breeder Reactor (FBR) & Sodium-Water Reaction (SWR) Dr. Raghuram Chetty Department of Chemical Engineering Indian Institute of Technology Madras Chennai

2 Fast Breeder Reactor (FBR) Liquid metal cooled reactor or Liquid Metal Fast Breeder Reactor (LMFBR) is an advanced type of nuclear reactor where the primary coolant is a liquid metal FBR can produce nuclear fuel more than consumption. Uses the fast neutrons from 235 U fission on surrounding 238 U to produce 239 Pu In years, enough Pu is produced to power another reactor No moderators No water coolant U must be at 15%-30% enrichment to generate power with fast neutrons while breeding Pu.

3 Fission in Light Water Reactor (LWR) Moderator Water, Graphite Moderation Control Fast neutron 10,000 km/s (1/30 of light speed) 2.2 km/s 10,000 km/s Some fast neutrons are absorbed in 238 U and produce 239 Pu. Courtesy: Google Images

4 Comparison Between FBR and LWR Fast Breeder Reactor Pu (~20%) U (~80%): MOX Fuel Light Water Reactor Slightly enriched uranium ( 3~4% 235 U) Liquid sodium Coolant Water None Moderator Water Fast neutron Fission Thermal neutron Core is small Temp. is more than 500 o C Atmospheric pressure Reactor Large core Temp: ~300 High-pressure

5 FBR around the world DFR, PFR (Therso) Super Phénix BN-600, 800(Belouarsk) (Crays-Malville) BOR-60(Dimitrovgrad) BN-350(Aktau) Phénix (Marcoule) CEFR(Beijing) Monju (Tsuruga) Joyo (O-arai) FFTF (Hanford) EBR (Idaho falles) FBTR, PFBR (Kalpakkam) :Closed :in operation Courtesy: Google Images

6 FBR Programme in India Future FBR 1000 MWe Pool Type Indegenous Beyond 2025 CFBR 500 MWe Pool Type UO2-PuO2 Indigenous From Ø FBTR 40 MWt 13.5 MWe Loop type PuC UC Since 1985 PFBR 1250 MWt 500 MWe Pool Type UO2-PuO2 Indigenous From MFTR 120 MWe From 2025

7 Fast Breeder Reactor at Kalpakkam The 500 MWe Prototype Fast Breeder Reactor (PFBR) under construction at Kalpakkam.

8 Schematic of PFBR at Kalpakkam Further details at

9 Liquid Metal Cooled Reactor The liquid metals used typically need good heat transfer characteristics. Fast neutron reactor core tend to generate a lot of heat in a small space when compared to other types of reactor. A low neutron absorption is desirable in any reactor coolant, and very important for a fast reactor. It has safety advantages because the reactor is not kept under pressure, and they allow much higher power density than traditional coolants.

10 Liquid Metal Fast Breeder Reactor Sodium is used as coolant in LMFBR because of the following properties High heat capacity Boiling point (882 C) is much higher than the reactor's operating temperature Sodium does not corrode steel reactor parts It requires low pumping power compared to water.

11 Liquid Metal Fast Breeder Reactor A disadvantage of sodium is its chemical reactivity, which requires special precautions to prevent and suppress fires. If sodium comes into contact with water it explodes, and it burns when in contact with air In addition, neutrons cause it to become radioactive. Hence the study of sodium-water reactions is important concern to LMFBR operations.

12 LMFBR Courtesy: Google Images Overview of Liquid Metal Fast Breeder (LMFBR)

13 LMFBR supe rhea ter Second Sodium loop (non-radioactive) The heat from the reactor is transferred through the primary and secondary cooling system to the evaporator and superheater which generates steam to drive the turbine generator. In this plant, liquid metal sodium is used as the coolant of the primary and secondary cooling system. Courtesy: Google Images

14 Fuel Arrangement in FBR 238 U + Pu (~20%) Blanket fuel of 238 U is placed in the peripheral, upper and lower region of the core in order to breed effectively. Enrichment is high in the outer region of the core in order to flatten the power distribution. Driver fuel Small core Short fuel Small diameter fuel in comparison to LWR Courtesy: Google Images

15 Safety Criteria Category of conditions 1 Normal Event frequency Plant criteria Safety criteria > 1 /year High availability Radiological release ALARA 2 Incidental > 10-2 /year Able to return to power at short term after rectification Radiological release lower than the limit (50 µsv/y) 3 Accidental > 10-4 /year Able to restart after inspection, repair, requalification Radiological release lower than the limit (1 msv/y) 4 Hypothetical > 10-6 /year Plant restart not required To maintain core coolability, limited change of core geometry (< 50 msv/y) Design Extension Condition (DEC) > 10-7 /year Loss of plant investment Releases lower than the targets (no need off-site provisions)

16 LMFBR Reactor Safety Features Does not induce increased reactivity upon LOCA Negative core reactivity Secondary sodium isolation loop (non-radioactivity) Failure of one module causes reactor shutdown (loss of criticality) Two independent reactor shutdown systems Forced and natural convection safety grade decay heat removal system (SGDHRS) through three or four independent loops.

17 Safety Features Control & Safety Rod Drive Mechanism Diverse Safety Rod Drive Mechanism Shutdown System & Decay Heat Removal Systems

18 Passive Shutdown System Diverse Safety Rod Drive Mechanism Liquid Poisson Injection System Temperature Sensitive Electro Magnet in Diverse Safety Rod Drive Mechanism (DSRDM) to minimize failure of shutdown system due to instrumentation failure Additional Liquid Poisson Injection System

19 Potential Failure Events from FBR Na leakage (primary/secondary Na, fuel storage Na) Na fouling/clogging, mixture of foreign materials in Na Sodium-water reaction (small scale, medium/large scale) failure of instrument/control systems failure or malfunction of equipment leakage of vapor, water, chemicals or oil structural deformation (vessels, piping, fuel assembly, heat exchanger) radiation leakage (gas, liquid, solid) fire (electricity, turbine oil, controlled area, welding) failure of electrical system (diesel, power-generating facility, electric motor)

20 Severe Accident Scenarios Initiating Events Total Instantaneous Blockage (TIB) of a subassembly (SA) Unprotected Loss of Flow (ULOF) or Unprotected Transient Over Power (UTOP) Failure of Post Accident Heat Removal Systems Consequences Melt propagation from one SA to neighboring SAs Bulk core melting Criticality events Mechanical energy release Recriticality Vessel integrity

21 Core Disruptive Accident Scenario Fuel attains gradual transition from solid to liquid phase, resulting in core boiling. Criticality condition can recur. Transient response from initiation to neutronic shutdown with essentially intact geometry and a gradual core melt down Vapour explosion, Recriticality, Post accident heat removal

22 Sodium Reactivity with Water: Sodium-water chemical interactions take place in two stages. In the first stage, the reaction proceeds at a high rate with release of gaseous hydrogen: Na + H 2 O = NaOH + ½ H kj/mole In the second stage, chemical interaction takes place between the products of the first stage and excess sodium: 2 Na + NaOH = Na 2 O + NaH Na + ½H 2 = NaH In steam generator, sodium and water are separated by thin tube. A single crack or hole in tube can lead to sodium-water interaction. NaOH and Na 2 O is corrosive agent and can further increase the water leak rate.

23 Na-H 2 O Reaction The pressure of sodium side remains low (around 2-3 bars), while the water (steam) side pressure is high (around 150 bars). The sodium and water are separated by the heat transfer tube walls. Therefore if there is a hole, weld defect or crack by thermal vibration in the heat transfer tube, the high-pressure water/steam leaks into the sodium part, resulting in a sodium-water reaction. This reaction is so rapid and violent that the safety system of the steam generator and secondary heat transfer system is confronted with a dangerous state.

24 Types of Sodium Water Reaction Na-H 2 O reaction can be classified according to water leak rate as: Micro Leak Small Leak Intermediate Leak Large Leak

25 Micro Leak Leak rate is less than 0.1 gm/sec and the diameter is less than 0.7 mm. Leak rate is so small that no wastage of adjacent tubes is occurred. As the hole becomes enlarge due to corrosion, self wastage of tube occurs. Self Wastage - a micro leak may enlarge quite suddenly after a period of time at a constant leak rate. The enlargement of the leak may be enough to cause an increase in leak rate of several orders of magnitude.

26 Small Leak Leak rate is in the range of 0.1 gm/sec to 10 gm/sec, leak size can between 0.7 mm to 1 mm. As leak rate and size of leak is large enough, it can damage the tube opposite to it. Impingement wastage occurs due to erosion and corrosion. The secondary tube can be failed within few minutes.

27 Intermediate Leak Leak rate is more than small leak (10 gm/sec < leak rate < 2 kg/sec). Where leak hole diameter can vary from 1 mm to 7 mm. Due to Na-Water reaction heat and hydrogen is produced, and this overheating can lead to multiple tube failure

28 Large Leak Here leak rate is greater than 2 kg/sec and hole diameter is more than 7 mm, as leak rate is too high, pressure increases rapidly and multiple tube wastage is occurred. To reduce pressure Surge tank and Rupture disc are provided in heat exchanger.

29 Leak Detection System In the steam generator of a fast reactor, the high pressure steam and hot sodium are separated by a steel wall. Any defect in the tube walls can cause steam to leak into sodium. Since the sodium water reaction is highly exothermic and caustic producing, the leaks can expand rapidly and lead to explosions. The best means of detecting these leaks at the very inception is to monitor the sodium at the steam generator outlet for hydrogen/oxygen concentration.

30 Leak Detection System The leak detection system must be sensitive to small leaks that may occur in the sodium to steam/water tube boundary, and provide a reliable and responsive signal to the plant operator. Additionally, provisions are required for prompt plant corrective action to minimize potential damage that may occur as a result of a steam/water to sodium leak. The device currently in use in fast reactors, measure hydrogen flux. The hydrogen detector senses a change in hydrogen concentration in sodium by measuring a change in the rate of hydrogen diffusion through a nickel membrane which is immersed in the sodium. An ion pump continuously pulls a vacuum on the back side of the membrane and hydrogen flux through the membrane is determined by measurement of the ion pump current.

31 Leak Detection System The oxygen meter is an electrochemical cell for measuring oxygen activity. This device has an electrical output which is proportional to the difference in the oxygen activity between an air reference electrode and the activity of oxygen in the sodium. Two oxygen meters / hydrogen sensors are provided in a single module so that if one malfunctions, the leak detector continues to operate until a convenient replacement time occurs. The leak detector can be utilized to detect steam/water to sodium leaks by two methods: (1) detecting a strong signal in a single pass (2) detecting a gradual buildup of hydrogen and oxygen concentration with loop circulation time or a rate of rise (ROR) of concentration.

32 Strong Signal Detection Strong Signal Detection - The detector has a sensitivity sufficient to resolve a signal from approximately 3 ppb change in concentration in a hydrogen background of 100 ppb. A sensitivity of 6 ppb in the sodium stream is chosen as the basis for detectability of an assured leak signal. The water leak rate can be related as follows: Conc. of H 2 = Leak rate/ Na flow rate The shortest time the detector can detect a leak (first pass) is about one minute and the longest is about five minutes.

33 Rate of Rise Detection Rate of Rise (ROR) Detection - A lower level of detection is possible when the concentration is building up in each pass. The rate of rise (ROR) of hydrogen build up would also indicate a leak when the concentration reached a detectable level. The ROR technique can identify a leak on the order of 0.9 x 10-5 kg/sec, with a detection time of about 10 minutes. The time for a leak signal to be translated into a control room signal depends on sodium flow rate, leak rate, location, time constant of the detector module, and the hydrogen background in the sodium stream. The time available for corrective action depends on the wastage rate of the Steam Generator tube resulting from the leak.

34 Plant Operator Actions Time available for corrective action depends on how damaging a leak is in terms of self wastage and wastage on the adjacent tube. By considering the combined factors of detector capability and wastage damage potential, plant corrective measures can be formulated. At the present, the approach to establishing plant operator actions is in the formative stages.

35 Plant Operator Actions The reference approach is as follows: A leak signal should be confirmed by at least one other detection device before corrective action is taken. Corrective action should be initiated promptly upon receipt of a confirmed leak signal. Shutdown procedures should be based upon the objective of minimizing damage to the steam generator and intermediate loop. Shutdown procedures should provide for more rapid actions if leakage rate increases during shutdown process.

36 Current Plant Operator Action Plan Leak Diameter Hole / m Detection Method Action Micro Leak x 10-3 Not detectable Continue operation Small Leak ~0.050 x 10-3 Intermediate Leak ~0.10 x 10-3 Rate of rise First pass Initiate orderly system shutdown, depressurize steam side; about 4 h for evaporator, 2 h for superheater. Initiate orderly system shutdown, depressurize steam side; about 1 h for evaporator, 12 min for superheater. Large Leak ~0.18 x 10-3 First pass Initiate rapid loop shutdown, depressurize steam side; about ten min. for evaporator, 2 min for superheater.

37 Sodium Fire Protection

38 Na Leak Liquid sodium is used as a coolant in Fast Breeder Reactor (FBR) systems. In the rare case of failure of a sodium bearing component, sodium can leak out and react with oxygen in the air and catches fire, when oxygen concentration in the air is more than 5% and the sodium temperature is more than 200 o C. Design provisions to defend against such sodium leaks and the resultant fires play an important role in the safe operation of a fast reactor.

39 Characteristics of Sodium Sodium is alkali metal which is soft and has metallic color. Weight of sodium is 0.97 times of water at 20 ºC. Melting point is 98ºC. Boiling point is 882ºC at atmospheric pressure. Sodium is lighter than water Soft & can cut by a knife Liquid sodium

40 Sodium Fire When sodium reacts with water, hydrogen gas, NaOH (which is corrosive) and heat are produced. Leak speed of sodium is generally slow due to atmospheric system pressure. However, sodium reacts with air as shown below, and produces lot of white alkali aerosol.

41 Sodium Inventory Primary Na system ~ 1150 t Secondary sodium system ~ 430 t, Safety grade decay heat removal (SGDHR) system ~ 120 t Reserve sodium 50 t (Non Radioactive) Total sodium inventory ~ 1750 t

42 Design Consideration Na is Radioactive or Non - Radioactive Location of system / component Quantity of Na leak Accessibility for intervention Type of sodium fire Sodium reacts with water in the concrete and produces heat. Which can lead to damage of concrete structure.

43 Design Strategies Make Na leak improbable Minimise Na leak by: Early Na leak detection Safety actions Minimise contact with air / concrete Dealing with sodium fire by: Fire fighting Controlled ventilation of Na aerosols Cleaning and disposal

44 Dealing with Sodium Fire Major systems in well separated buildings Use of partition walls / barriers to prevent spread of sodium fire. If sodium is radioactive, only passive system is required to avoid Na fire (No direct fighting) For non radioactive Na passive & active system of fire fighting Well defined approach route for fire fighting persons Direct Na fire fighting using dry chemical powder (DCP) e.g. sodium bicarbonate or lithium carbonate. Inert gas such as nitrogen or argon could be injected to extinguish Na fire. Storage of adequate quantity of dry chemical powder at strategic locations (3 times more) Application of DCP by shovel for small fires, portable/ mobile extinguishers for medium and large fires.

45 Example of a leak location Leak location IHX Reactor Evaporator Super heater Courtesy: Google Images Air cooler Pump for secondary system

46 Design basis to prevent Sodium fire To reduce the quantity of sodium coming out Double envelope and Leak collection tray are used. To prevent the spreading of sodium fire well barriers or partition walls are used (which can withstand for three hours). To prevent sodium concrete interaction, a thin metal liner on concrete wall is used. Sodium circuit is protected from accidental dropping of objects and other missile object by suitable design.

47 Na Leak Collection Tray The leak collection tray (LCT) mainly consists of two sloping plates (angle 20 o C) forming a funnel like structure supported on the sodium hold-up vessel. These sloping plates with V-shape orientation rapidly guide the leaked sodium to a central drainpipe, which ends at 20 mm above the bottom surface of the hold-up vessel. The drained sodium is accumulated in the hold-up vessel with limited exposure to air. A vent pipe of smaller diameter is provided on the slopping plates to facilitate the easy draining of the leaked sodium. Details of Prototype of LCT

48 Research on FBR The fire risks of liquid sodium have been, and continue to be, investigated, as the new generation of FBR systems is designed and developed. Safety remains at the forefront of nuclear industry thinking, and it is all too aware of the radioactive and safety risks posed by reactor leaks and fires at nuclear sites. Drawing on past experience, a range of new experimental and numerical simulations and research projects are being undertaken, in order to design and enhance reactor safety systems and accident management procedures.