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The Indian Energy Resource Position Explains Our Strategy For Deployment Of Nuclear Energy If the level of our per capita electricity consumption is raised.

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Presentation on theme: "The Indian Energy Resource Position Explains Our Strategy For Deployment Of Nuclear Energy If the level of our per capita electricity consumption is raised."— Presentation transcript:

1 The Indian Energy Resource Position Explains Our Strategy For Deployment Of Nuclear Energy
If the level of our per capita electricity consumption is raised to the level of a developed country (~5000 kWh/person/year) and only a single energy resource is to be used: Domestic extractable coal reserves will last for < 13 years. Uranium in open cycle will last for ~ 0.5 year Uranium in closed cycle with FBRs will last for ~ 73 years Known reserves of thorium in closed cycle with breeder reactors will last for > 250 years Entire renewable energy (including hydroelectric capacity) will be sufficient for < 70 days/ year Total solar collection area (based on MNES estimate 20 MW/km2) needed will be at least ~ sq. km. It is obvious that for long term energy security nuclear energy based on thorium has to be a prominent component of Indian energy mix.

2 Comparison of Fuel Characteristics
Calorific value of fossil fuels (kcal/kg) Domestic Coal: 4000, Imported Coal: 5400, Naphtha: , LNG: 9500 Indian uranium-ore contains only 0.06% of uranium (Canada’s 18%), but this provides 20 times more energy per tonne of mined material than coal when uranium is used in once through open cycle in PHWRs 1200 to 1400 times more energy per tonne of mined material than coal when used in closed cycle based on FBRs 1000 MWe Nuclear Power Plant needs movement of 12 trucks (10 Te/truck) of uranium fuel per year 1000 MWe Coal Power Plant needs movement of 3,80,000 trucks (10 Te/truck) of coal per year

3 Nuclear Energy Sources

4 Review of Atomic Structure
Nucleus Contains protons and neutrons Small Size Relatively large mass Extremely large density Large amount of stored energy Orbiting Electrons Large size Low density Orbit nucleus near speed of light Small amount of energy relative to nucleus Responsible for chemical bonds

5 Compound Nucleus in an excited state of high internal energy
Neutron Nucleus n Radiation Fission Fragments ~200 MeV of Energy Compound Nucleus in an excited state of high internal energy Fast-n The Fission Reaction The fast neutrons have a low probability of inducing further fissions (but used as such in fast reactors), and hence generating more neutrons thus sustaining a chain reaction. So in thermal reactors, we need to slow down the neutrons (i.e., thermalise or moderate them), which we do by using a moderator such as water (Heavy Water or Light water). Combustion of 1 atom of C => 4 eV; Fission of 1 atom of U => 200 MeV

6 Classification of Reactor Systems
Nuclear reactors operating on fission are broadly classified into two types Classification of Reactor Systems Thermal Reactors Fission is sustained primarily by thermal neutrons ( E ~ eV). Moderator (Ordinary water, heavy water, graphite, beryllium) is required to slow down the high energy fission neutrons. Large core. Very high fission cross-section for thermal neutrons, less fuel inventory. Fast Reactors Fission is sustained primarily by fast neutrons (E ~ 1 MeV) No moderator used. Compact core. High core power density – liquid metal or helium gas as coolant. Higher number of neutrons available for capture in fertile material. Breeding possible. In thermal reactors, the fission is caused by thermal neutrons having energy less than eV. This type of reactor uses natural uranium as fuel. The neutrons generated during fission posses very high energy which are slowed down with the help of a moderators to reduce the energy of neutrons less than eV. In fast reactors, fission is basically caused by neutron possessing energy more than 1 MeV. Another important process that is taking place in the fast reactor is breeding of fissile material.

7 Slowing down (thermalisation or moderation) of fission neutrons facilitates lower critical mass, but leads to some loss of neutrons through absorption in the moderator Variation of fission cross-section (barns) of U-235 with neutron energy (eV) Cross-section: The effective target presented by a nucleus for collisions leading to nuclear reactions . 1 barn = cm2 Energy distribution of fission neutrons peaks at ~ 0.7 MeV with average energy at ~ 1.9 MeV. Thermal Reactors Fast Reactors

8 Interaction of Radiation
with Matter Radiation deposits small amounts of energy, or "heat" in matter alters atoms changes molecules damage cells & DNA similar effects may occur from chemicals Much of the resulting damage is from the production of ion pairs

9 Ionization The process by which a neutral atom acquires a positive or negative charge electron is stripped from atom - The neutral atom gains a + charge = an ion + Alpha Particle highlight "IONIZING RADIATION:" When ion pair is created the atom or molecule is biologically active Biological damage can result

10 Ionization - - - - - Ionization by a Beta particle: ejected electron
Colliding Coulombic Fields The neutral absorber atom acquires a positive charge -

11 Types of Radiation Mass (amu) Charge Travel Distance in Air 4.0000
Alpha +2 few centimeters Beta Plus 0.0005 +1 few meters few meters Beta Minus 0.0005 -1 Gamma 0.0000 many meters 0.0000 X-Rays many meters Neutron 1.0000 many meters

12 Gamma Interactions Interactions called "cataclysmic" - infrequent but when they occur lot of energy transferred Three possibilities: May pass through - no interaction May interact, lose energy & change direction (Compton effect) May transfer all its energy & disappear (photoelectric effect)

13 Compton Effect - - - - - - - -
An incident photon interacts with an orbital electron to produce a recoil electron and a scattered photon of energy less than the incident photon Before interaction After interaction Scattered Photon - - - - - - - - Incoming photon Collides with electron Electron is ejected from atom



16 Seven Basic Components of Nuclear Reactor

17 Nuclear Fuel

18 Zircaloy-4, Optimised Zircaloy
STRUCTURAL MATERIALS Selection Criteria: Low neutron absorption cross section Low cost Adequate tensile strength Adequate creep strength Adequate ductility after irradiation Corrosion resistance Materials: Reactor Cladding BWR Zircaloy-2 / Zircaloy-4 PWR Stainless Steel 304 Zircaloy-4 PHWR Zircaloy-4, Optimised Zircaloy Zirlo LMFBR Type 316SS (20% CW) Alloy D9 (20% CW) (Modified 9Cr-1Mo) HTGR Graphite

19 2100 (0.2% light water as impurity) Metallic Beryllium (Be)
MODERATOR MATERIALS To slow down and moderate fast neutrons from fission Materials with light nuclei are most effective Materials Moderating Ratio Light water(H2O) 70 Heavy water (D2O) 2100 (0.2% light water as impurity) 12000 (100% heavy water) Metallic Beryllium (Be) 150 Graphite 170 Beryllium oxide 180 {Moderating ratio = macroscopic scattering cross section / absorption cross section} REFLECTOR MATERIALS To cut down the neutron leakage losses from core Desired properties same as moderators Water Heavy Water Beryllium Graphite Thermal Reflectors

20 PWR (Clad in CW 304 SS/Inconel 627)
CONTROL MATERIALS Selection Criteria: Neutron absorption cross section Adequate mechanical strength Corrosion resistance Chemical and dimensional stability (under prevailing temperature and irradiation) Relatively low mass to allow rapid movement Fabricability Availability and reasonable cost Materials: Boron, Cadmium, Gadolinium, Hafnium, Europium B4C BWR (Clad in 304 SS) 80% Ag-15%In+5%Cd PWR (Clad in CW 304 SS/Inconel 627) LMFBR

21 Coolant Material

22 SHIELDING MATERIAL To protect personnel and equipment from the damaging effects of radiation Good moderating capability Reasonable absorption cross section Cost and space availability Neutron, a,b and g shielding Both light and heavy nuclei are preferred WATER PARAFFIN POLYETHYLENE Pb, Fe, W Boral (B4C in Al matrix) Concrete

23 Pressurized Heavy Water Reactor

24 PHWR Calandria PHWR Fuel Assembly

25 Typical Pressurized Water Reactor

26 Boiling Water Reactor In a closed system, back reactions occur, and an equilibrium is established; H2O H /2O2 is the overall effect. If H2 is added to the system, back reactions are promoted, and radiolysis is effectively suppressed. radiation

27 Zircaloy Development

28 Zircaloy Development ---- contd

29 Principal Materials of Construction – BWRs
Reactor Vessel: Fuel cladding; Zircaloy 2 (Sn %, Fe 0.2%, *Ni 0.08%, Cr 0.15%, balanced with Zr) * Ni-free less susceptible to hydriding fuel assembly accessories; Alloy-600 (Inconel) (Ni 72%, Cr 14-17%, Fe %, C 0.15%) Alloy-X750 (Inconel) (Ni 70%, Cr 14-17%, Fe %, C 0.08%, Nb+Ta %, Ti %, Al %) neutron-absorbing control rods; B4C powder in * SS or Inconel sheath BWR Normal Water Chemistry High [O2]  IGSCC (InterGranular Stress Corrosion Cracking) of sensitized stainless steel, increases nodular corrosion – local oxide nodules and IASCC (Irradiation-Assisted Stress Corrosion Cracking). ~

30 Principals Materials of Construction For PHWR
Fuel: Natural U as UO2 Fuel sheath: Zircaloy-4 Pressure tube: Zr – 2.5 wt.% Nb End fitting: Type 403 SS (Cr 12%, Ni < 0.5%, Mn < 0.5%, balanced with Fe) Feeders, headers, S.G. heads, piping: Carbon steel Steam generator tubing: Inconel-600 at Bruce NGS (Cr 15%, Ni 72%, balanced with Fe) Alloy-800 (Incoloy) at Darlington NGS and CANDU-6s (Cr 21%, Ni 32%, balanced with Fe) Alloy-400 (Monel) at Pickering (Cr 0%, Ni 70%,Fe 2%, Cu 28%) Dissolved D2: Minimizes radiolysis (keeps [O2] low); should not contribute to the hydriding (deuteriding) of Zr – 2.5 Nb pressure tubes. H2 (not D2) usually added; exchanges rapidly with D in D2O Principal Material of Construction – Moderator Calandria vessel: Type 304 SS, Calandria tubes: Zircaloy-4, Piping: Type 304 SS Moderator HXs: Incoloy -800

31 PFBR Reactor Assembly 01 Main Vessel 02 Core Support Structure 03
Core Catcher 04 Grid Plate 05 Core 06 Inner Vessel 07 Roof Slab 08 Large Rotating Plug 09 Small Rotating Plug 10 Control Plug 11 CSRDM / DSRDM 12 Transfer Arm 13 Intermediate Heat Exchanger 14 Primary Sodium Pump 15 Safety Vessel 16 Reactor Vault

32 PFBR Core Configuration

33 Core Structural Materials
Though the desire is to have only fuel in the core, structural material form 25% of the total core To support and to retain the fuel in position Provide necessary ducts to make coolant flow through & transfer/remove heat Clad tubes- 50%, Wrapper-40%, spacer wire-10% For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order pins (both inner & outer core Fuel SA) Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes These structural materials see hostile core with max temperature and neutron flux

34 Need for special core structural Materials for FBR
More Residence Time Increased residence time of the fuel which in turn demands extended service of the core structural material (540 Days for the 100 GWd/t B.U and even more) More Temperature Utilising the high energy spectrum for more breeding leave very high power densities in the core (compact core) which makes the core struct. Matls to see high temperatures (clad mid wall temp °C) High Neutron Flux Low Fission C.S of fuel & Neutron absorption C.S is low for the Stainless Steels at high neutron energy levels and also with higher enrichments, very high neutron flux levels (~1015 n/cm2/sec) Thermal reactors – Fuel fissile fraction dictates the residence time Fast Reactors – Core Structural materials dictates and demands development of special core struct matls.

35 Clad and Wrapper Materials of FBRs
Prolonged service at high temperature ( oC) High neutron doses in the range (~ dpa) High energy neutron irradiation leads to displacement of atoms (vacancies and interstitials) Agglomeration of vacancies Void formation and swelling

36 Differential swelling due to gradients in temperature and flux leads to distortion and bowing of wrapper Swelling of clad will reduce sodium coolant flow between the fuel pins and increases the local temperature leading to possible failure of clad Radiation induced segregation and changes in microchemistry Irradiation induced precipitation (Alpha Prime, G- phase, M6C, Chi-phase, Laves phase) Coarsening of existing precipitates and/or dissolution

37 Irradiation Induces Changes in Mechanical Properties
Radiation Hardening Loss of Ductility Loss of Fracture Toughness Increase in DBTT Decrease in Upper Shelf Energy Irradiation Creep and Swelling leads to dimensional changes

38 Core Components - Structural Material Aspects
Radiation damage is major consideration Effects of Irradiation on Materials Void Swelling Irradiation Creep Irradiation Embrittlement Helium Embrittlement Increase in Ductile Brittle Transition Temp Considerations in Material selection High Swelling Resistance Adequate end of life Creep Strength and Ductility Compatibility with Sodium Compatibility with Fuel material and Fission products Corrosion Resistance Burnup of 200 GWd/t can be achieved by using advanced materials

39 Design Aspects – Bowing & Dilation
Fresh Core Bowed Core Dilation (Swelling + Creep) Bowing

40 Bowing - Influence in Design
SA Handling Limit Misalignment Handling Machine capacity Core SA Temperature Monitoring Reactivity Change Limit to burnup dictated by structural material deformation

41 Fuel Pin Deformation Pin Bundle – Wrapper Interaction Easy assembling
Pressure Drop Flow induced vibration of pin Pin – Spacer Wire Interaction Spacer wire – tightening & Loosening Strains on clad and wrapper

42 Oxide Dispersion Strengthened Alloys
Development of suitable clad material is essential for achieving high burn-up of fuel in Fast Breeder Reactors. Austenitic stainless steels swell significantly beyond 120 dpa Conventional Ferritic/Martensitic steels possess high swelling resistance ( < 2% swelling upto 200 dpa) compared to Aust. SS Ferritic steels posses poor thermal creep strength above 550oC ODS alloys serve as one of the alternatives with the potential of having advantage of ferritic steel and able to push operating temperatures to 650oC and beyond.

43 Temperature (K)


45 Accident cause 11/03/2011, 2:46 p.m. local time (7 hours earlier Romanian time) near the Japanese island of Honshu was an earthquake of 9 on the Richter scale. The quake had an impact on section of north-east coast of Japan where they are located a series of nuclear power plants (NPP). Nuclear reactors have been shut down properly.

46 Event description Units 4-6 in shut down status for periodic maintenance and refuelling Units 1-3 were stopped automatically after the quake Reactor buildings and the containment successfully resist to the earthquake All reactor were dissconnected from the external AC supply Backup sources (diesel generators) started At approximately one hour after the earthquake tsunami hit the site destroyed fuel tanks of the diesel generators flooded the diesel generator building (10m protection wall was not sufficient) Mobile generators were sent to the site in a short time but they ran out of fuel Hydrogen Explosion Unit 1 Evacuation of population from the area of 20km Daiichi NPP and 10km Daina NPP (approx person On-site radioactivity increased

47 Event description Lowering the internal pressure led to hydrogen explosion at unit 3 Injection of sea water into the reactor vessel without cooling units at unit 1-3 Variable on-site radioactivity Increased radioactivity at Onagawa NPP (north of Daiichi) revealed that comes from Daiichi NPP

48 Event description Cooling with seawater stopped at Unit 2 (unknown cause), variable water level in the reactor Hydrogen Explosion at Unit 2 Cooling with sea water stopped at all units due to lack of fule and water source Fire then explosion in the spent fuel storage pool at unit 4 (relatively fresh fuel) Restart seawater injection in the reactor wessel at all units Significant radioactive emission Housing on the area of km Risk of melting the core and damage of the containment at Unit 2

49 Event description Fire in spent fuel storage pool at Unit 4, cooling water evaporation Water level decrease at Unit 5, taking water from Unit 6 Unsuccessfull attempts to feed with cooling water and boric acid the spent fuel storage pool at Unit 4 Possible melting (at least partially, 50%) of the core at Units 1 and 3 Fill with water the reactor vessel of the Unit 2 Lowering water levels in the spent fuel pool at Units 3 and 4 Increasing temperature in the spent fuel pool at unit 5 and 6 Cooling with water canons from the police departement

50 Event description 17.03. Radioactivity observed outside of the site
Fukushima: μSv / h (30 km from the NPP) In two places increasing dose 80 to 170, and 26 to 95 μSv/h Other directions 1-5 μSv/h Begining actions to connect a cable for AC supply to unit 2 Continue attempts for cooling Unit 4 with water from helicopters (without succes) then with water canons One of the diesel generators from Unit 6 supplies Unit 5 for cooling spent fuel storage pool and the reactor wessel

51 These tubes are manufactured from different grades of
stainless steels such as D9 - Fuel clad tubes - Reflector clad tubes - Blanket clad tubes - Inner Boron Carbide clad tube - CSR clad tubes - DSR clad tubes - DSR absorber pin bundle sheath - Hexagonal Wrapper tube 316LN - Axial shielding pin clad tubes Stainless Steel Seamless Tubes plant (SSTP), NFC was set up in late 70’s was first plant of its kind which pioneered manufacture of seamless Steel tubes in a wide range of sizes and grades. With this experience, Nuclear Fuel Complex (NFC), Hyderabad has taken up the challenging task of producing these tubes.

SNO APPL. AREA ALLOYS CHALLENGES PROCESSING METHOD 1 CORE D9 (316 Ti) 20% CW D9 Strict Ultrasonic Acceptance Criteria Narrow chemistry Mn control Low gas limit VIM + VAR 2 STEAM GENERATOR 9Cr1Mo (modified) Grade 91 filler wire and welding electrode Restriction on ‘S’ level Control N/Al N control in range Heavy Forgings with complex profile AF + VIR + ESR - Control ESR melt parameters 3 STRUCTURAL SS 316 L(N) & SS 304 L(N) Close range of carbon and nitrogen

53 MDN 316 Ti – D9 Chemical Composition Alloying Elements Element
ASTM A 771 Grade S38660 LMFR – D9 C 0.030 – 0.050 0.035 – 0.05 Si  0.50 – 1.00 0.5 – 0.75 Mn 1.65 – 2.35 Cr 12.5 – 14.5 13.5 – 14.5 Ni 14.5 – 16.5 14.5 – 15.5 Mo 2.0 – 2.5 Ti 0.10 – 0.40 4.5 (C +N) – 6.0 (C+N) Fe Balance Alloying Elements

54 MDN 316 Ti – D9 Chemical Composition Tramp Elements Element ASTM A 771
Grade S38660 LMFR – D9 P  0.040  0.02 S  0.010  0.01 N  0.005 Al  0.050 Nb Ta  0.020 0.020 Co As  0.030 V  0.05 B* 10 – 20 PPM Cu  0.04 Tramp Elements

55 MDN 316L (N) Element ASTM A 276 Grade 316 L LMFR 316L (N) C < 0.03 0.024 – 0.03 P < 0.045 S < 0.01 Si < 0.75 < 0.5 Mn < 2.00 1.6 – 2.0 Cr 16.0 – 18.0 17.0 – 18.0 Ni 10.0 – 14.0 12 – 12.5 Mo 2.00 – 3.00 2.3 – 2.7 N < 0.10 0.06 – 0.08 Ti - < 0.05 Nb Cu < 1.0 Co < 0.25 Fe Balance A Extra Low Carbon Austenitic Stainless Steel with restricted range of nitrogen Special requirement: Chemistry as per ASTM with following restrictions PROCESS MODIFICATION MELTING IN VACUUM INDUCTION REFINING FURNACE FOR LOW CARBON, LOW LEVEL OF IMPURITIES & NARROW RANGE OF ALLOYING ELEMENTS AND ESR TO ENSURE LOW SULPHUR.

56 MDN 9Cr1Mo A Chromium Molybdenum Ferritic Steel that have been exclusively formulated and are stringent compared to ASME requirements for PFBR application. Special requirement Control of sulphur in specific range (0.005 – 0.010), especially for tube sheet and tubes for wettability during autogenous welding Nitrogen to aluminium ratio greater than 2 PROCESS MODIFICATION: MODIFICATION OF SLAG TO ENSURE SULPHUR WITHIN THE SPECIFIED RANGE DURING ESR. THERMOMECHANICAL TREATMENT TO ENSURE FINE GRAINS WATER QUENCHING DURING HARDENING TO ACHIEVE UNIFORM HARDENABLITY


58 Manufacturing Process includes Hot extrusion followed by Cold working i.e. Pilgering and/or Drawing
The Manufacturing facilities at NFC include ton capacity Hot Extrusion Press 1200 ton capacity Piercing press Cold Pilger Mills (Two Roll & Three Roller type) Triple Tube Draw Bench Annealing Furnaces ( Air & Bright annealing) D9 Fuel clad Tubes are being manufactured at SSTP.

59 Controls imposed by these specifications include
Restricted chemistry Micro structural control Ultrasonic and eddy current tests with very low % of wall thickness as defect standard Tight dimensional tolerances Narrow band of cold work. Mechanical and corrosion properties

60 Dimensions of Fuel Clad Tubes
Outside diameter ( true) mm / Inside diameter (true) mm / Wall Thickness, (min) mm Length (nominal) mm Straightness / 500 or better

61 Forged rounds Billet machining Hot extrusion De-glassing ID & OD conditioning Ultrasonic Testing Cold working

62 Solvent / vapor Degreasing
Pilgering Alkali Degreasing Annealing Sink Drawing Solvent / vapor Degreasing Bright annealing Final Fixed Plug drawing with 20 % Cold work with controlled dimensions

63 Longitudinal Defects observed
Degreasing Straightening Numbering Eddy current Testing Ultrasonic Testing Longitudinal Defects observed Attributed to sinking of thin walled tubes

64 Steam Generator Tubes Size : 17.2 mm OD x 2.3 mm WT x 23100 mm long
Grade : 9Cr-1Mo Critical requirements : Close Control of dimensions over entire length Cleanliness of the tubes internally and externally before HT Ultrasonic Testing with continuous wall thickness measurement Eddy Current Testing Pressure Testing and drying Special Heat Treatment cycles consisting of Normalizing followed by Tempering Mechanical properties at room & high Temperature after simulation heat treatment Handling and transportation of long length of 23.1 m Special vacuum sealed Packing Qty produced and supplied : 5000 Nos for 9 steam generators

65 Gas Heater Tubes Size : 38.1 mm OD x 2.6 mm WT x 2550 - 4660 mm long
Grade : 9Cr-1Mo Critical requirements : Close Control of dimensions Ultrasonic Testing for flaw and continuous wall thickness measurement Eddy Current Testing Pressure Testing and drying Special Heat Treatment cycles in Bright Annealing furnace Normalizing & Tempering Mechanical properties at room & high Temperature and after simulation heat treatment Qty produced : 1000 Nos

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