Safety aspects of Indian advanced reactors
K.K. Vaze,
Director
Reactor Design and Development Group
Bhabha Atomic Research Centre,
Trombay, Mumbai 400085 India
1
1
Post Fukushima Scenario
Fukushima Accident



On March 11th, 2011, a gigantic earthquake with a
magnitude 9 on the Richter scale shook Japan. The
earthquake triggered a tsunami, which was
exceptionally high, reached the Fukushima coast
about one hour after the earthquake.
All reactors in operation at Fukushima shut down
automatically. While the offsite external power source
was lost due to the earthquake, emergency diesel
generators (EDG) started up properly
Even though the earthquake was of a magnitude far
greater than anticipated, there is today no evidence
that it produced mechanical or structural damage
which would have, in the absence of the tsunami,
caused a severe accident. The seismic response
analysis and the visual investigations conducted so far
did not seem to show major damage to safety-related
equipment.
Fukushima Accident - contd

The majority of the damage was caused by the tsunami. At
Fukushima Daiichi it caused complete loss of AC power, loss
of ultimate heat sink and serious degradation of DC power
sources. This led to the loss of decay heat removal at three
NPP units, to severe reactor core damage, to the loss of
containment integrity and to significant radioactive releases
to the environment. In addition, the upper part of the fourth
unit reactor building was destroyed by hydrogen explosion
and the spent fuel pool structures of that unit suffered
mechanical damages.
Some reassuring thoughts as far India is concerned

Huge earthquakes and huge tsunamis are not
commonplace
Comparative Seismic Hazard
Status of Seismicity – Indian NPPs
• Criteria - No Active fault within 5 km
Site
Seismic Zone
Narora
IV
Rawatbhata
II
Kakrapar
III
Tarapur
III
Jaitapur
III
Kaiga
III
Kalpakkam
II
Kudankulam
II
Tsunamigenic locations for Indian coast
TARAPUR
KALPAKKAM
ONLY FAR FIELD
SOURCES
KUDANKULAM
TECTONIC PLATE BOUNDARIES
18 March 2011
How does this benefit us?
Fukushima
•
Earthquake knocked out Class 4 supply
•
Tsunami knocked out other supplies
India
•
EQ and tsunami don’t occur together
•
Ground motion due to an earthquake causing tsunami
is negligible
•
Earthquakes causing significant ground motion do not
cause tsunami
•
We get warning (~ 2 hrs)
Fukushima Accident
Lessons Learnt
The key criterion of success:
- recovery of power supply
- water feed for the decay heat removal
As prompt as possible!
 Availability of undamageable portable
engineering means for power and water
supply in the conditions of NPP isolation
 Accident prevention and accident mitigation:
- implementation of design fundamental;
- emergency preparation;
- Severe Accident management.
Source: Prevention and Mitigation —
Equal Priorities
Prof. Vladimir Asmolov, WANO President

ACCIDENT MANAGEMENT GOAL ACCIDENT MANAGEMENT
MEASURES
To prevent the core melting
The recovery of the core cooling
(To keep the integrity of the Ist
and IInd physical barriers – Fuel &
Clad)
To retain melt inside the RPV
(To keep the integrity of the IIIrd
physical barrier - RPV)
In-vessel cooling
Ex-vessel cooling
To prevent the containment failure Core catcher
(To keep the integrity of the IVth
Hydrogen management
physical barrier - Containment)
Filtered venting system
Source: Prevention and Mitigation — Equal Priorities
Prof. Vladimir Asmolov, WANO President
Genesis for development for advanced reactors
12
Securing energy for India’s future is a major challenge
World
Population
(billion)
OECD
Non-OECD
India
(developing world)
India
of our dream
6.7
1.18
5.52
1.2
Annual av. per
~2800
capita Electricity (kWh)
~9000
~1500
~780
5000
18.8
10.6
8.2
0.835
8.0
30
13
1.8
?
Annual
Electricity
Generation
(trillion kWh)
Carbon-di-oxide
Emission
(billion tons/yr)
17
1.6
(stabilised)
India alone would need around 40% of present global electricity
generation to be added to reach average 5000 kWh per capita
electricity generation
Dr. Kakodkar “Atoms for Prosperity
Global climate change is an immediate threat

Just ten years from now, greenhouse emissions from developing
nations will equal the emissions from the countries we now call
developed. After that, emissions from the developing world will be
the major driver of global climate change.

While energy conservation, windmills, and solar panels may help,
we cannot hope to rely on such measures alone to meet our
world’s expanding appetite for more energy.
John Ritch, Director General of the World Nuclear Association, 15th Pacific
Basin Nuclear Conference, Sydney, 15-20 Oct. 2006
1979
2003
Source:
http://www.nasa.gov/centers/goddard/
news/topstory/2003/1023esuice.html
Comparison of sea-ice from 1979 and 2003.
14
Safety Goals for Advanced Reactors
15
CNS Extraordinary Meeting Summary Report
The
displacement of people and the land
contamination after the Fukushima Daiichi accident
calls for all national regulators to identify provisions
to prevent and mitigate the potential for severe
accidents with off-site consequences.
Nuclear power plants should be designed,
constructed and operated with the objectives of
preventing accidents and, should an accident occur,
mitigating its effects and avoiding off-site
contamination.
The Contracting Parties also noted that regulatory
authorities should ensure that these objectives are
applied in order to identify and implement
appropriate safety improvements at existing plants.
Dr. Kakodkar

An essential goal for nuclear safety is “Never Again” should
there be any significant off site emergency

Dual level design basis

Design Basis
• Risk Lowered to an acceptable level
• No impact in public domain

Extreme Event
• Maximum potential
• No significant off-site emergency

Extra margin between design and ultimate load capacity should be
sufficient to cope with this
Can the nuclear community set for itself an ambitious goal
to meet the challenge of the numbers?
“Four decades from now, in any country of the world, it
should be possible to start replacing fossil fuelled power
plants, at the same urban or semi-urban site where these
are located, with advanced NPPs that would, more
economically, deliver at least twice the power that was
being produced by the replaced plants”
R.K. Sinha, “The IAEA’s Contribution to the Peaceful Use of
Nuclear Power”, Nuclear Power Newsletter, Vol. 3, No. 3,
Special Issue, Sept. 2006
18
Level of safety goals increases with multi-fold
increase in deployment of nuclear reactors
Special Siting
Criteria, Risk
approach
Special Siting
Criteria
(may/may not);
CDF, LERF
Safety
Goals
Siting criteria
Dose Criteria
Reactors
under
operation
(existing
technology)
Advanced
reactors
under
construction
Advanced
future
Reactor
Systems
Number of reactors in operation
19

Strategy for safety
measures and features of
nuclear installations is
two-fold:
 To prevent accidents
• Preventing the
degradation of plant
status and
performance
 If prevention fails, limit
their potential
consequences and
prevent any evolution to
further serious conditions
Monitored Process Parameter
Achievement of safety goals through enhanced
levels of Defence-In-Depth
Mo
nit
ore
d
Pr
oc
es
s
par
am
ete
r
Level 4 of DiD
Design Basis Safety limit
Level 3 of DiD
Safety system setting
Level 2 of DiD
Operational limit
Level 1 of DiD
Alarm setting
Steady state operation
Time
CHALLENGE
20
Passive and Inherent Safety Features are Instrumental in
Meeting New Safety Criteria

The conventional reactors or so called “Traditional ones” have seen an
extensive use of “active” engineering safety systems for reactor
control and protection in the past.
• These systems have certain potential concerning termination of events or accidents
that are effectively coped with by a protective system limited by the reliability of the
active safety systems or prompt operator actions.

Since the reliability of active systems can not be improved above a
threshold and that of the operator’s action is debatable, there is growing
concern about the safety of such plants due to the large uncertainty
involved in Probabilistic Safety Analysis (PSA) particularly in analyzing
human faults.
• In view of this, a desirable goal for the safety characteristics of an innovative reactor is
that its primary defence against any serious accidents is achieved through its design
features preventing the occurrence of such accidents without depending either on the
operator’s action or the active systems.

• That means, the plant can be designed with adequate passive and
inherent safety features to provide protection for any event that may
lead to a serious accident.
Such robustness in design contributes to a significant reduction in the
conditional probability of severe accident scenarios arising out of initiating
events of internal and external origin.
21
Example of Applications Passive Systems and Inherent
Safety Features in Defence-In-Depth in AHWR
22
The Indian Advanced Heavy Water Reactor
(AHWR-Pu)
AHWR is a 300 MWe vertical pressure tube type, boiling light water cooled
and heavy water moderated reactor using 233U-Th MOX and Pu-Th MOX fuel.
Major design objectives



•Design validation through
extensive experimental
programme.
•Pre-licensing safety
appraisal by AERB
•Site selection in progress.
•Detailed engineering
consultancy in progress
23


65% of power from Th
Several passive features

7 days grace period

No radiological impact
Passive shutdown system to
address insider threat scenarios.
Design life of 100 years.
Easily replaceable coolant
channels.
AHWR-Pu is a Technology
demonstrator for the closed thorium
fuel cycle
AHWR-LEU extends the AHWR
technologies with LEU-Th MOX Fuel
for the global market
Top Tie Plate
Water
Tube
Fuel
Pin
Displacer
Rod
Bottom Tie Plate
AHWR Fuel
assembly
AHWR incorporates several technolological solutions to a higher level
of safety and security against both internal and external threats
Control
room
and
auxiliary
systems
Instrumentation
& control signals
Pneumatic supply
Electrical power
(Class 1 to 4)
External events
Malevolent
act
Control and
S/D systems
Turbine
Core
Condenser
Ultimate
heat sink
(Cooling
tower or
sea)
Pump

No unacceptable radiological impact outside the plant boundary
with
(a)
(b)
(c)
(d)
24
Failure of all active systems, and
Failure of external infrastructure to provide coolant, power and other services,
and
Malevolent acts by an insider, one of the consequences of which is the failure of
instrumentation signal initiated shutdown actions, and
Inability of plant operators to manage the events and their consequences, for a
significantly long time.
Some important passive safety features of AHWR
–1/4
Heat removal from core under both normal full power operating condition as
well as shutdown condition is by natural circulation of coolant.
25
Some important passive safety features of AHWR
–2/4
Passive Containment
Cooling
(Th-Pu) MOX
Fuel pins
Passive
Containment
isolation
Central Tube for
ECCS water
(Th-233U) MOX
Fuel pins
AHWR FUEL CLUSTER
Passive injection of cooling water, initially from accumulator and later from the
overhead GDWP, directly into fuel cluster.
26
Some important passive safety features of AHWR
–3/4
Passive Poison Injection in moderator
during overpressure transient
Passive Poison Injection System actuates during very low probability event of failure of
wired shutdown systems (SDS#1 & SDS#2) and non-availability of Main condenser
27
Some important passive safety features of AHWR
–4/4
Use of
moderator
as heat
sink
Water in
calandria
vault
Flooding of reactor cavity
following LOCA
28
Fukushima and AHWR







AHWR has been assessed for TMI as well as
Chernobyl type of accidents
Critics comments: It is easy to become wise after
the event (TMI, Chernobyl)
Fukushima type event (Extended SBO) was
anticipated even before it happened
Practically no change required in AHWR design to
meet Fukushima event
GDWP and passive systems adequate to cater to
the extended SBO
No impact in public domain, No need of
evacuation
No need of exclusion zone, sterilized zone
Prolonged Station Black Out in AHWR
Decay heat removal by Isolation Condensers
A strong earthquake with/without Tsunami causing
prolonged SBO for several days. Reactor tripped on
seismic signal.
7
6
GDWP Level
5
Level (m)
4
Gravity Driven Water Pool is intact.
3
2
Heat is removed by Isolation Condensers
1
0
-1
320
300
0
10
20
30
40
50
60
70
80
90
100
110
280
Time(days)
Clad Surface Temperature
0
Temperature ( C)
260
240
220
200
180
160
2.8
140
2.6
120
0
2.4
20
30
40
50
60
70
80
90
100
110
Time (days)
2.2
Pressure (bar)
10
Containment Pressure
2.0
1.8
1.6
1.4
GDWP water removes decay heat for ~110 days with
periodic containment venting allowed after 10 days.
1.2
1.0
0.8
0
20
40
60
Time (days)
80
100

Level 1 DID:

Elimination of the hazard
of loss of coolant flow:
• Heat removal from
the core under both
normal full power
operating condition
as well as shutdown
condition is by
natural circulation
of coolant.
F
Initiating Events are
caused by Failures of
the Level 1 of D.I.D
10-2
10-6
10-7
Level 1 Defence in Depth : prevention of abnormal operation and system failure
Passive Systems in Defense-In-Depth of AHWR
Events
managed by
Lev 2 of D.I.D
Events
managed by
Lev 3 of D.I.D
Events
managed by
Lev 4 of D.I.D
AOO
UNACCEPTABLE DOMAIN
AC
SPC
C

Reduction of the extent of overpower transient:
• Slightly negative void co-efficient of reactivity.
• Low core power density.
• Negative fuel temperature coefficient of reactivity.
• Low excess reactivity
31
Passive Systems in Defense-In-Depth of AHWR
(Contd.)


Level 2: Control of abnormal operation and detection of failure
• An increased reliability of the control system achieved with the use of
high reliability digital control using advanced information technology.
• Increased operator reliability achieved with the use of advanced displays and
diagnostics using artificial intelligence and expert systems.
• Large coolant inventory in the main coolant system.
Level 3: Control of accidents within the design basis
• Increased reliability of the ECC system, achieved through passive injection
of cooling water directly into a fuel cluster through four independent
parallel trains.
• Increased reliability of a shutdown, achieved by providing two independent
shutdown systems. Further enhanced reliability of the shutdown, achieved by
providing a passive shutdown device
• Increased reliability of decay heat removal, achieved through a passive
decay heat removal system, which transfers the decay heat to GDWP by
natural circulation.
• Large inventory of water inside the containment (about 8000 m3 of water
in the GDWP) provides a prolonged core cooling meeting the requirement of
grace period.
32
Passive Systems in Defense-In-Depth of AHWR
(Contd.)

Level 4: Control of severe plant conditions, including prevention of
accident progression and mitigation of consequences of severe
accidents
• Use of moderator as heat sink.
• Presence of water in the calandria vault
• Flooding of reactor cavity following a LOCA.

Level 5: Mitigation of radiological consequences of significant release of
radioactive materials
• The following features help in passively bringing down the
containment pressure and eliminates any releases from the
containment following a large break LOCA:
• Double containment;
• Passive containment isolation
• Core catcher
• Filtered vent
33
Peak Clad Temp v/s frequency of occurrence – a
quantitative probabilistic safety criteria
Large Break LOCA without ECCS
Decrease in coolant inventory
Increase in coolant inventory
Increase in heat removal
Increase in system pressure/Decrease in heat removal
Decrease in coolant flow
Reactivity anamolies
Operational occurances/transients
Multiple failure events
Wires system failure events
1300
1200
BDBEs
0
Temperature( C)
1100
200 % LOCA
1000
900
DBEs
800
700
600
500
AOO & NO
400
300
200
1E-111E-10 1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 0.01
0.1
1
10
Frequency
34
Core Damage Frequency Per Year
Ref: Lecture on Near Term Advanced Nuclear Reactors and Related MIT
Research, by Prof. Jacopo Buongiorno, MIT, USA, June 16, 2006.
AHWR
~ 1x10-8
35
Severe Accident Management
Incorporation of Hard vent
• Hard Vent system is designed to prevent the over
pressurization of the containment beyond design
pressure occurring due to failure of multiple safety
systems because of an extreme event such as
prolonged SBO with non-availability of GDWP water
or large seismic event causing cracks in GDWP
along with LOCA.
• Also retains the radio-activity in the scrubber and
minimize activity release beyond the containment
boundary.
• Scrubber tank contains water + NaOH solution (ph
= 8.5).
• NaOH combines with Iodine whereas Cs which is in
form of CsI, CsOH, CsO2, Cs2CO3 is soluble in
water.
3 I2 + 6 NaOH = 3 H2O +5 NaI + NaIO3
• A 4 inch Dia pipe is provided at the top of primary
containment for venting, which will be connected to
scrubber tank.
To Stack
From
Containme
nt
Passive Autocatalytic ReCombiner System
(PARCS)
Postulated Accidents
DBA : Single failure (LB LOCA): No hydrogen generation
BDBA : Multiple failure (LBLOCA and non-availability of
Wired Shutdown System)
~ 30 kg in 300 s.
Prolonged SBO + non-availability of GDWP ~ 450 Kg in 2 hr
starting after 40hrs of transient (~5000 m3 at ambient)
• Peak H2 generation rate ~ 0.3 kg/s
 The released hydrogen will be combined by Passive Autocatalytic
Recombiners (PARCS) located at several locations in the
containment designed in such a way to reduce the hydrogen
concentration in the containment below the flammability limits.
 Experiments are being carried out for demonstration of hydrogen
removal using PARCS
Recombination rate ~ 0.1 kg/hr/m2 (for 2 - 4% H2 conc.)
Overall box size : 1000 x 400 x 1000 (L X B X H)
(8.29 m2 of Catalyst Deposited area)
Estimated Conversion rate : 0.83 kg/hr
No. of Recombiners for one Plant ~ 100
(Total Conversion Rate = 83 kg/hr)
Design of Core Catcher
Water from GDWP





Sacrificial Concrete
(300 mm depth)
High porosity concrete
(300 mm depth)
Sacrificial concrete layer
mixes with the melt,
reduces its temperature,
solidus
temperature
(typically from 2800oC to
Riser Tubes
1500oC)
and helps in
7.4
m
(
100mm)
Water pool
spreading the melt over
(500 mm depth)
large surface area
Poison added in sacrificial Structure of core catcher
concrete
prevents
Sacrificial concrete
recriticality
composition
High porosity concrete
layer below the sacrificial
concrete helps in flooding Design objective of the core catcher
water from below
• Retention of the melt in the
cavity
Riser tubes inject water
within the melt-concrete
• Quenching it within 30 minutes
mixture
• Stabilize it for substantial
The downcomers supply
period of time (several days)
water to the water pool
from GDWP passively
39
Indian High Temperature Reactor Programme
40
Indian High Temperature Reactor Programme
Compact High Temperature Reactor (CHTR)Technology Demonstrator
• 100 kWth, 1000 °C, TRISO coated particle fuel
• Several passive systems for reactor heat removal
• Prolonged operation without refuelling
Innovative High Temperature Reactor for
Hydrogen Production (IHTR)
• 600 MWth , 1000 °C, TRISO coated particle fuel
• Small power version for demonstration of
technologies
• Active & passive systems for control & cooling
• On-line refuelling
Indian Molten Salt Breeder Reactor (MSBR)
• Large power, moderate temperature, and based on
233U-Th fuel cycle
• Small power version for demonstration of
technologies
• Emphasis on passive systems for reactor heat
removal under all scenarios and reactor conditions
41
Status: Design of most of the
systems worked out. Fuel and
materials under development.
Experimental
facilities
for
thermal
hydraulics
setup.
Facilities for design validation
are under design.
Status:
Optimisation
of
reactor physics and thermal
hydraulics design, selection of
salt and structural materials in
progress. Experimental facilities
for molten salt based thermal
hydraulics
and
material
compatibility studies set-up.
Status: Initial studies being
carried out for conceptual
design
41
Technology for fuel kernel by sol-gel technique is well established –
Focus is on technologies for TRISO coating and fuel compact




Initial trials with
zirconia kernels
completed
Fabrication trials of
TRISO fuel using
natural UO2 kernel
carried out
Fuel compact
prepared by two
different techniques
High packing density
(45-50%) achieved
OPyC
Zirconia
SiC
IPyC
Buffer
PyC
X-ray radiographic image of TRISO
particle with Zirconia kernel
Radiograph and tomograph of fuel compact
made by different technique
SEM images of particle with Nat. UO2 kernel
Fuel Compacts
42
Fabrication of C/C composite tubes and coating
with SiC
High Temperature Fluidized bed Coater
(Inset shows fluidized bed
distributor assembly)
Fluidised bed based SiC
coating method developed
Cooling
tower
• High density C-C composite
fuel tube samples
fabricated in collaboration
with National Physical
Laboratory, New Delhi
• Pre-form was made using
high strength carbon fibers
• Pre-form subjected to
multiple cycles of resin
impregnation and hot isostatic pressing with
intermediate machining
cycles
43
Machining trials of graphite
components (AFD)
Sample
with
graphite
fixtures
and
graphite
susceptor
Induction
heating
system
Ar rotameter
Fluidized Bed Heated graphite
Distributor being dipped in
fluidized bed
Thermal hydraulic studies for liquid metal (Pb-Bi)
Liquid Metal Loop
(2009)
Major areas of development
• Analytical studies and development of computer codes
• Liquid metal loop for experimental studies
• Loop at 550 °C in operation since 2009
• Loop at 1000 °C under commissioning
• Steady state and transient experiments carried out
• In-house developed code validated
• Experimental and analytical studies for freezing and de-freezing of
coolant
YSZ
based
• Test bed for development of instrumentation –level probes,
oxygen
sensor, EM pump and flowmeters
oxygen sensor
5
10
4
Ress
10
Comparison of steady state
correlation [Vijayan, 2002]
with experimental data
3
10
Experimental
Correlation (Vijayan,2002)
2
10
44
8
2.0x10
8
4.0x10
8
6.0x10
8
8.0x10
44
Sufficient time margin before shutdown or passive
alternate heat removal system needs to act

Case-1




Case-2

~40 min
~58 min
250% step
increase in
power
LOCA
No heat sink
Similar to
case-1, but
with a 300%
“spike” in
power before
stabilizing at
250%
Sufficient time available to activate
primary and/or secondary shutdown
system, or passive gas-gap filling
system
45
Negligible rise in peak temperatures after
shutdown due to decay heat
Minimum temperatures well above freezing point of coolant even after 1 hour
46
Innovative High Temperature Reactor (IHTR) for
commercial hydrogen production




600 MWth, 1000 °C, TRISO
coated particle fuel
Pebble bed reactor concept with
molten salt coolant
Natural circulation of coolant for
reactor heat removal under
normal operation
Current focus on development:





Pebble
Reactor
Vessel
Reactor physics and thermal
hydraulic designs – Optimisation
Thermal and stress analysis
Code development for simulating
De-Fuelling
pebble motion
Chute
Experimental set-up for tracing path
of pebbles using radio-tracer
technology
Side Reflector
Pebble feeding and removal
systems
•Hydrogen: 80,000 Nm3 /hr
•Electricity: 18 MWe, Water: 375 m3/hr
•No. of pebbles in the annular core ~150000
•Packing fraction of pebbles ~60%
•Packing fraction of TRISO particles ~ 8.6 %
233U Requirement 7.3 %
•47
Fuelling
pipe
Core Barrel
Support
TRISO coated
particle fuel
Coolant
Outlet
Coolant
Pebble
Retaining
Mesh
Pebbles and
Coolant
Central
Reflector
Bottom
Reflector
Coolant
Inlet
Thermal hydraulic studies and material
compatibility studies for molten salt coolant
Molten salt
loop
EXPANSION
TANK
SAFETY TANK
Major areas of development
• Analytical studies and development of computer codes
• Molten salt natural circulation loop for experimental
studies
• Molten fluoride salt corrosion facility using FLiNaK
• Experiments being carried out upto 750 °C mainly
on Inconel materials
Molten salt corrosion test
facility
FILTER
COOLER
HEATER
CONTROL
VALVE
MELT TANK
48
48
Design features of Indian HTRs leading to inherent
safety

TRISO coated fuel particles: Retention of fission products up to 1600
°C

High thermal inertia of ceramic core and low power density

Sufficient margin between reactor operation and boiling point of the
coolant

Negative temperature coefficient of the core and coolant

Natural circulation of liquid metal / molten salt coolant in single
phase
• Low pressure of the system

Passive removal of heat under normal operation and postulated
accident scenarios
• High temperature heat pipe for CHTR

Chemical inertness of the lead based coolant with air/water
49
Molten Salt Breeder Reactor (MSBR)
This concept is attractive to India because of large
thorium reserves and possibility of breeding 233U in
thermal spectrum – For the third stage of Indian Nuclear
Power Programme
50
Schematic of Indian MSBR
1.
2.
3.
4.
Design guidelines
Heat removal by
natural circulation of
molten salts
Avoid moderator to
reduce solid high level
waste generation
Ability to tolerate
outage of reprocessing
plant
Enhanced safety as
compared to current
reactors for possible
deployment near
population centres
Fertile salt
drain tank
Fissile salt
drain tanks
Turbine
IHX
Redox control
(Fertile Salt)
Helium bubbling
and Redox control
(Fuel Salt)
Condenser
Pump
Fuel
Salt
Fertile
Salt
Selection of
salts, materials
and conceptual
design in
progress
Coolant salt
drain tank
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Inherent safety features of MSBR (1/2)
 Continuous
addition of fuel to maintain criticality
 Less initial reactivity
 Fission products, including xenon and krypton, are
continuously taken out of the system,
 No excess reactivity reaquired for xenon override
 No danger of their release under accident condition
 Entire fuel salt inventory can be dumped into smaller
subcritical dump tanks, through freeze valves,
 Reducing the chances of any untoward incidents.
 The molten salt has a high boiling point (~1400°C),
hence there is a very low vapor pressure
 Normal operating temperatures ~ 700 to 800 C
52
Inherent safety features of MSBR (2/2)
 The
density of fuel salts decreases with increase in
temperature,
 With increase in temperature fuel salt is pushed out of the
core leading to reduction of reactivity
 No scenario for ‘fuel melt down’
 Modification of existing safety codes required for defining CDF
 Molten fluorides are simple ionic liquids
 Stable to the irradiation
 Do not undergo any violent chemical reactions with air or
water
 Fuel has no burnup limits
 Life is dictate by life of moderator and structural materials
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Accelerator Driven Systems
54
BARC is developing technologies for Accelerator Driven System (ADS)
mainly for Thorium utilization and waste transmutation

Major Role:



Accelerator-driven Sub-critical reactor system
High conversion sub-critical blanket with thorium for producing 233U
Incineration of minor actinides and some fission products
Turboelectrical
generation
plant
Steam
generator plant
Sub-critical
reactor core
Accelerator-driven Sub-critical reactor system
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Generation of fissile materials from thorium by spallation
reaction using high energy proton accelerators
High Energy & High Current Proton Beam from
Accelerator (Cyclotron/LINAC)
ADS
Concept,
and subsystems
Beam Channel
Collimator
Coolant: Pb,
LBE, Na, Heavy
Water etc.
Spallation target region:
liquid lead, LBE,
solid W, etc
Fuel
Window
(Solid:
W-Rh,
SS etc)
Breeding
Th-232 to
U-233
Spallation
•Inherently
safe, flexible
fuel cycle
•Higher burnup
•Reduced
doubling time
for ADSbreeders
•Intense, lowenergy-cost
neutron source
•Fissile factory
for U-233 from
Th-232
•Suitability for
transmutation
& burning
nuclear waste
233U  Fission fragments
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ADS for Transmutation & with Th-fueled reactor
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Summary

In the Indian context, large scale deployment of
nuclear reactors is required, with possible deployment
near population centres

Enhanced level of safety is one of the primary goals for
advanced reactors under design in BARC
 Defence-in-depth
 Passive safety devices
 PSA studies
 Margin assessment
 Advanced materials
 Advanced Reactor concepts
58
Thank You
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Modification to Strengthen “Severe Accident
Prevention Features”

Improving availability of onsite power supply
Providing back up emergency DG (air cooled) at a higher location
- Providing a smaller/mobile DG to power essential loads and charge
station batteries
-

-
-

-

Improving steam generator heat sink
Securing FFW diesel engines pumps from external flood and margins w.r.
t earthquake evaluated
Additional diesel engine operated pumps to transfer deaerator storage
tank inventory to steam generator
Provision of hook up connections outside reactor building, qualified for
maximum anticipated earthquake and flood
Provision for Passive Decay Heat Removal (PDHR) system for 700 MWe
Improving onsite water storage for one month
SBO period
Augmentation of water inventory
Sources of water near stations are identified for fire tenders
Hook upto Primary Heat Transport System
/ECCS
Injection into PHT system for making up leakage during SBO
- Injection into PHT for unsuccessful long term ECCS operation
-
Other measures

Introduction of Seismic Trip (already exists
in NAPS & KAPS)

Strengthening provision for monitoring of
critical parameters under prolonged loss of
power

Creation of an emergency response facility
capable of withstanding severe flood,
cyclones & earthquake

Provision for Tsunami early warning system