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Severe Accident Scenarios:
Indian Perspective
FR13, 4-7 Mar 2013, Paris
P.Chellapandi Indira Gandhi Centre for Atomic Research
Kalpakkam, India
Approach Towards CDA for Future FBRs
• Target reliability for shutdown system to be
considerably enhanced w.r.t PFBR: to be achieved
by introducing third level shutdown systems
On probabilistic basis
• Target reliability of DHR system to be enhanced w.r.t
PFBR: Innovative features to be introduced
• Combined analysis of reactor physics, energy
release and mechanical consequences, validated
with experiments
Assessment of mechanical consequences with
pessimistic assum on energy release
Two initiating events, that could lead to a CDA: (1) Loss of
flow without reactor shutdown leading to ULOFA
(Unprotected Loss of Flow Accident) and (2) Uncontrolled
withdrawal of control rods introducing reactivity ramp leading
to UTOPA: Unprotected Transient Overpower Accident .
In the UTOPA scenario, the in-pin motion of molten fuel
caused by fission gas pressure, called 'fuel squirting' or 'fuel
extrusion', introduces a lot of negative reactivity in the core
and hence stabilize the reactor at a higher power level
without causing any core melting / boiling / disruption.
Hence, UTOPA is not energetic
ULOFA is analyzed in detail, since it leads to a CDA that
could pose major threat to the structural integrity of primary
containment.
Severe Accident Scenario of PFBR
3 phases: Pre-disassembly, transition and Disassembly
Flow reduction immediately leads to coolant temperature rise
in core that gives positive reactivity, which however, is
dominated by the negative reactivity due to the radial thermal
expansion of core and the net reactivity is negative. This
results in decrease in power.
Power to flow ratio however increases subsequently, resulting
in high coolant temperature rise and voiding in the upper part
of the highly rated fuel channel.
Core voiding spreads radially outward and axially downward
increasing positive reactivity contribution. When the voiding
spreads into the central part of the core, the net reactivity
becomes positive, initiating a series of events: power
excursion, clad dry out, rapid temperature rise in the fuel and
clad and ultimately melting of fuel and clad.
ULOFA Scenario
The molten materials would be swept out of the core by the
shearing force of the coolant vapor (dominant force in the case of
fresh fuel) as well as by the accumulated fission gas pressure
(dominant in the case of irradiated fuel).
If there are sufficient negative reactivity introduced from Doppler
and fuel displacement, the core could become subcritical and the
accident terminates.
Else, with the high rate of positive reactivity addition, core attains
a super prompt critical condition and sub subsequently into
disruptive condition.
Consequence is large thermal energy release & vaporization of
significant portions of fuel and structural materials of the core.
Under an idealized condition, a mixture of molten materials at the
bottom with vapor phase at the top could be conceived at the end
of disassembly phase
ULOFA Scenario .. Contd….
Mechanical energy release depends upon the reactivity addition
rate in the disassembly phase, which in turn depends upon the
assumptions made on the sodium void propagation, fuel
displacement / slumping characteristics, reactivity feedback
mechanisms, cross section data, nature of temperature
distributions assumed for the disrupted core and cross section
data employed in the analysis
One of the important parameters influencing the coolant void
generation/propagation is flow halving time. With lower value, the
coolant voids could generate below the core top and spread
rapidly to the core centre, resulting in high positive reactivity rate
in the disassembly phase. With higher value, the coolant boiling
starts at the upper portion of active core, which introduces
negative reactivity due to the high neutron leakage.
Work Potential - contd..
Mechanical energy release when vapor phase expands
from its initial pressure to one atmosphere
Scenario Reactivity
addition rate
Energy
Release
Low flow halving time of 2 s, coherent core
lumping, absence of feedbacks, flat temperature
distribution across the core at the end of
disassembly phase and use of conservative cross
section data (CV2M cross section set),
200 $/s 1000 MJ
Conservative slumping model: active core zone
divided into three. The molten fuel from middle
one third occupies the core lower portion and fuel
from top one third occupies the middle portion.
65 $/s 100 MJ
A flat temperature distribution of the core at the
end of disassembly phase 50 $/s 268 MJ
Longer flow halving time of 8 s, incoherent core,
presence of all feedbacks, realistic temperature
distribution across the core and use of realistic
cross section data (ABBN cross section set),
10.5 $/s < 1 MJ
Work Potential – contd..
Synthesis of above results motivates to investigate the mechanical
consequences of a CDA over a wide range of work potentials
corresponding to the reactivity addition rates ranging from 25 $/s to 200 $/s.
Work Potential – contd..
1
10
100
1000
10000
25 50 75 100 125 150 175 200
Reactivity - $/S
Me
ch
an
ica
l E
ne
rgy
- M
J
Important Consequences of a CDA in SFR
a b c Deformations of
vessels Sodium ejection to
RCB Post Accident Heat
Removal
• Preliminary Deformations of Components (0 - 50 ms)
• Upward Motion of Sodium Slug (50 – 100 ms)
• Sodium Slug Impact on top shield and development of
Transient Forces on Reactor Vault (100-150 ms).
• Sodium Release to RCB during quasi-static state (150 –
900 ms)
• Post Accident Heat Removal Condition depending upon
coolability of core bubble (> 900 ms)
Mechanical Energy Release Scenario
0
50
100
150
200
250
300
350
0 25 50 75 100 125 150 175 200 225 250
Time - ms
En
erg
y r
ele
ase -
MJ
Strain energy absorbed by the vessel
100
MJ
1000
MJ
500
MJ
0
10
20
30
40
50
60
70
0 100 200 300 400 500 600 700 800 900 1000
Work potential - MJE
ne
rgy
re
lea
se
- %
Energy balance in the vessel
Total energy
absorbed
upper portion
bottom portion
Sequence of Mechanical Loadings and
Energy Absorptions
Local deformations, get saturated at high mechanical energy release and
the vessel absorbs the energy uniformly, enhancing its energy absorbing
potential.
-0.2
0.0
0.2
0.4
0.6
0.8
1.0
0 2 4 6 8 10 12 14 16 18 20
Developed length from bottom - m
Dis
pla
ce
me
nt
- m
Radial displacement in the vessel
1000 MJ
100 MJ
0
2
4
6
8
10
12
14
16
0 100 200 300 400 500 600 700 800 900 1000
Work potential - MJ
Str
ain
- %
Peak strain at top
Averaged strain
Peak to averaged strain
Membrane strains in the vessel
Main Vessel Deformations
Deformation becomes more uniform compared to lower work
potential cases. This is a favorable feature that the energy absorbing
potential is not linear and the vessel can absorb higher energy
without undergoing rupture locally upon application of higher energy
by the core bubble.
0
1
2
3
4
5
6
7
8
0 100 200 300 400 500 600 700 800 900 1000
Work potential - MJ
Imp
ac
t p
re
ss
ure
- M
Pa
Peak impact pressure on top shield Sodium slug impact scenario
-30
-20
-10
0
10
20
30
0 50 100 150 200 250
Time - ms
Imp
ac
t v
elo
cit
y -
m/s
1000 MJ
500 MJ
200 MJ
100 MJ
Slug Impact Loadings and Their Effects
In view of short duration of impact loadings and high mass inertia of top
shield structures, it has high potential to absorb higher impact loads and
hence, the integrity of top shield would not be of concern and do not
decide the acceptable work potential.
1000 MJ 600 MJ 800 MJ 900 MJ
100 MJ 200 MJ 400 MJ 500 MJ
Sodium Slug Impact w.r.t Sodium Release to RCB
Sodium slug decelerates to get separated from the top shield. Hence
the quasi-static pressure in the cover gas is tending towards
saturation with higher work potential
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0 100 200 300 400 500 600 700 800 900 1000
Work potential - MJ
Pre
ss
ure
- k
Pa
Quasi-static cover gas pressure
0
5
10
15
20
25
0 100 200 300 400 500 600 700 800 900 1000
Work potential - MJP
re
ss
ure
- k
Pa
Pressure rise in RCB
Pressure rise in Reactor Containment Building
Since quasi-static pressure saturates, sodium release to RCB saturates
and hence containment loadings would attain saturation at higher work
potentials.
11 Tests on 1/13th scale mockups to
demonstrate the structural integrity of
DHX and to simulate sodium leak
Integrity of Main Vessel, Top Shield and DHXs
Main vessel capacity = 1200 MJ
DHX capacity = 500 MJ
Maximum sodium leak = 275 kg for 100 MJ
• Parametric study on mechanical energy release
values in the range 100 – 1000 MJ indicates that
primary containment has high potential to
withstand the transient forces generated by
energy release even more than 1000 MJ.
• Sodium ejection into the RCB through top shield
penetrations under sodium slug impact
phenomenon is limited with higher energy
• The deformations of decay heat exchangers
immersed in the sodium pool could limit the
acceptable work potential.
• For PFBR, this value is found to be 500 MJ from
simulated experimental study.
Conclusions
Thank You