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 ~ 31000 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.
1
Comparison of Fuel Characteristics
• Calorific value of fossil fuels (kcal/kg)
Domestic Coal: 4000, Imported Coal: 5400, Naphtha: 10500, 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
Nuclear Energy Sources
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

The Fission Reaction
n
Neutron
Nucleus
Compound Nucleus in an
excited state of high internal
energy
Fission
Fragments
~200 MeV
of Energy
Fast-n
Radiation
• 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).
5
Nuclear reactors operating on fission are broadly classified into
two types
Classification of Reactor Systems
Thermal Reactors
Fast Reactors
 Fission
is sustained primarily by
thermal neutrons ( E ~ 0.025 eV).
 Fission
 Moderator
 No
(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.
6
is sustained primarily by
fast neutrons (E ~ 1 MeV)
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.
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 crosssection (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 = 10-24 cm2
Energy distribution of fission
neutrons peaks at ~ 0.7 MeV with
average energy at ~ 1.9 MeV.
Thermal
Reactors
Fast
Reactors
7
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
Ionization
The process by which a neutral atom acquires a positive
or negative charge
-
Alpha Particle
+
+
-
electron is
stripped from
atom
The neutral atom
gains a + charge
= an ion
-
Ionization
Ionization by a Beta particle:
Beta Particle
-
-
-
Colliding
Coulombic Fields
The neutral absorber atom
acquires a positive charge
ejected electron
-
Types of Radiation
Mass
(amu)
Charge Travel Distance in Air
Alpha
4.0000
+2
few centimeters
Beta Plus
0.0005
+1
few meters
Beta Minus
0.0005
-1
few meters
Gamma
0.0000
0
many meters
X-Rays
0.0000
0
many meters
Neutron
1.0000
0
many meters
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)
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
Seven Basic Components of Nuclear Reactor
Nuclear Fuel
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
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
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
B4C
PWR (Clad in CW 304
SS/Inconel 627)
B4C
LMFBR
Coolant Material
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
Pressurized Heavy Water Reactor
PHWR Fuel Assembly
PHWR Calandria
Typical Pressurized Water Reactor
Boiling Water Reactor
In a closed system, back reactions occur, and an equilibrium is established;
radiation
1
H 2O
H2 + /2O2
is the overall effect.
If H2 is added to the system, back reactions are promoted, and radiolysis is
effectively suppressed.
Zircaloy Development
Zircaloy Development ---- contd
Principal Materials of Construction – BWRs
Reactor Vessel:
Fuel cladding; Zircaloy 2 (Sn 1.2-1.7%, 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 610%, C 0.15%)
Alloy-X750 (Inconel) (Ni 70%, Cr 14-17%, Fe 59%, C 0.08%, Nb+Ta 0.7-1.2%, Ti
2.25-2.75%, Al 0.4-1.0%)
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).
~
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:
Moderator HXs:
Type 304 SS, Calandria tubes:
Incoloy -800
Zircaloy-4, Piping: Type 304 SS
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
PFBR Core Configuration
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
39277 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
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 600-700°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.
Clad and Wrapper Materials of FBRs
• Prolonged service at high temperature (300-700 oC)
• High neutron doses in the range (~ 150-200 dpa)
• High energy neutron irradiation leads to
displacement of atoms
(vacancies and interstitials)
• Agglomeration of vacancies
• Void formation and swelling
• 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, Gphase, M6C, Chi-phase, Laves phase)
• Coarsening of existing precipitates and/or dissolution
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
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
Design Aspects – Bowing & Dilation
Fresh Core
Bowed Core
Bowing
Dilation
(Swelling + Creep)
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
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
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
resistance ( < 2% swelling upto 200 dpa) compared to Aust. SS
swelling
 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.
Temperature (K)
THANK YOU
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.
Event description 12.03.
•
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. 200 000
person
• On-site radioactivity increased
Event description 13.03.
• 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
Event description 14-15.03.
• 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 20-30 km
• Risk of melting the core and damage of the containment at
Unit 2
Event description 16.03.
• 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
Event description 17.03.
• Radioactivity observed outside of the site
– Fukushima: 3-170 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
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.
PROCESSING OF ADVANCED ALLOYS AT MIDHANI
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
Narrow chemistry
AF + VIR + ESR
MDN 316 Ti – D9
Chemical Composition
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
1.65 – 2.35
Cr
12.5 – 14.5
13.5 – 14.5
Ni
14.5 – 16.5
14.5 – 15.5
Mo
1.50 - 2.50
2.0 – 2.5
Ti
0.10 – 0.40
4.5 (C +N) – 6.0 (C+N)
Fe
Balance
Balance
Alloying
Elements
MDN 316 Ti – D9
Chemical Composition
Element
ASTM A 771
Grade S38660
LMFR – D9
P
 0.040
 0.02
S
 0.010
 0.01
N
 0.005
 0.005
Al
 0.050
 0.050
Nb
 0.050
 0.050
Ta
 0.020
0.020
Co
 0.050
 0.050
As
 0.030
 0.030
V
 0.05
 0.050
B*
 0.0020
10 – 20 PPM
Cu
 0.04
 0.04
Tramp Elements
MDN 316L (N)
Element
C
ASTM A 276
Grade 316 L
< 0.03
LMFR
316L (N)
0.024 – 0.03
P
< 0.045
< 0.03
S
< 0.03
< 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
PROCESS MODIFICATION
N
< 0.10
0.06 – 0.08
Ti
-
< 0.05
Nb
-
< 0.05
Cu
-
< 1.0
Co
-
< 0.25
MELTING
IN
VACUUM
INDUCTION REFINING FURNACE
FOR LOW CARBON, LOW LEVEL
OF IMPURITIES & NARROW
RANGE OF ALLOYING ELEMENTS
AND ESR TO ENSURE LOW
SULPHUR.
Fe
Balance
Balance
A Extra Low Carbon
Austenitic Stainless Steel
with restricted range of
nitrogen
Special
requirement:
Chemistry as per ASTM
with
following
restrictions
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
Manufacturing Process includes Hot extrusion followed by
Cold working i.e. Pilgering and/or Drawing
The Manufacturing facilities at NFC include
3750 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.
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
Dimensions of Fuel Clad Tubes
Outside diameter ( true) mm
Inside diameter (true) mm
Wall Thickness, (min) mm
Length (nominal) mm
Straightness
6.60 + / - 0.02
5.70 + / - 0.02
0.43
2555
0.25 / 500 or better
Forged rounds
Billet machining
Hot extrusion
De-glassing
ID & OD conditioning
Ultrasonic Testing
Cold working
Pilgering
Alkali Degreasing
Annealing
Sink Drawing
Solvent / vapor Degreasing
Bright annealing
Final Fixed Plug drawing with 20 % Cold work with
controlled dimensions
Degreasing
Straightening
Numbering
Eddy current Testing
Ultrasonic Testing
Longitudinal Defects observed
Attributed to sinking of thin walled tubes
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
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