DISCUSSION
Severe Accident Assessment for
PFBR: A Designer’s Perspective
S C Chetal, P Chellapandi
A comment on “The Limits of
Safety Analysis: Severe Nuclear
Accident Possibilities at the PFBR”
by Ashwin Kumar and
M V Ramana (EPW, 22 October
2011), followed by a response by
the authors themselves.
S C Chetal (chetal@igcar.gov.in) is director,
Indira Gandhi Centre for Atomic Research, and
P Chellapandi (pcp@igcar.gov.in) is director,
Nuclear and Safety Engineering Group at the
IGCAR.
76
General Safety Features of PFBR
A
500 megawatt electric (MWe) Prototype Fast Breeder Reactor (PFBR)
designed and developed by the
Indira Gandhi Centre for Atomic Research
(IGCAR) (Chetal et al 2006) is under advanced stage of construction at Kalpakkam.
The PFBR possesses all the intrinsic and
engineered safety features, viz, two independent, fast acting, reliable shutdown
systems, decay heat removal capability by
natural circulation through dedicated
heat exchangers, warm roof concept to
minimise the risk of sodium aerosol deposits, application of leak before break justiication for the main vessel, sodium piping
and steam generators, provision of robotic
device for the main vessel in-service inspection, etc. The core is monitored by functionally diverse sensors. Two diverse parameters as far as possible are provided for
every design based event, which have the
potential to cross design safety limits. The
pump discharge head and speed are measured and used as trip parameters for the
protection against primary pipe rupture
and pump seizure respectively. Failure of
fuel is detected by monitoring the cover
gas ission product activity and delayed
neutron detection in the primary coolant.
Design Basis Events (DBE) as well as
Beyond Design Basis Events (BDBE) have
been systematically identiied and analysed using validated computer codes.
With the chosen reactor core parameters,
viz, lower height/diameter ratio (0.5),
smaller pin diameter (6.6 mm), annular
pellets and lower fuel volume fraction at
the mid-level of fuel assembly (0.32), sodium void coeficient under postulated core
disruptive accident (CDA) is limited to $2.4,
when both fuel melting and sodium boiling occur together and consequent energy
release is insigniicant (< 1 megajoules (MJ))
(Sathiyasheela and Srinivasan 2008). This
void coeficient is smaller compared to other
december 24, 2011
international values: $4 for the 600 MWe
Japanese demonstration fast breeder reactor (Endo et al 1994) and $5.9 for the
French SPX-1 (Stark et al 1991).
In order to accommodate the design basis
mechanical energy release, a containment
system has been conceived as two portions:
the upper and lower portions divided by
the top shield of reactor assembly. The
upper portion is the reactor containment
building (RCB) and the lower portion is
termed as the primary containment. The
primary containment consists of main
vessel welded to top shield, which is supported on the outer reactor vault. Surrounding the main vessel, there is a “safety
vessel” supported independently from
the inner reactor vault. The inter-vessel
space between main and safety vessels is
a leak tight boundary, illed with nitrogen.
RCB, top shield, main vessel along with
safety vessel are designed to meet the
speciied safety criteria relevant to structural integrity of containment system and
post-accident heat removal capability under
CDA (Bhoje et al 1990). To facilitate the
post-accident heat removal capability for a
long term, an in-vessel core catcher structure has been incorporated below grid
plate (Chellapandi et al 2011).
It is a general feeling that a heterogeneous
core would be the best choice for the sodium
void considerations. This has to be seen
comprehensively by taking into account
all the aspects. An organisation, while
studying the various design concepts, selects an option that can be designed and
operated with conidence and also considering the world over operating experience
and trend. This is the justiication for the
choice of homogeneous core instead of
heterogeneous one in PFBR.
CDA Scenario and Energy Release
The main parameter in the energy release
is reactivity insertion rate, which mainly
comes from sodium void effects. The studies
performed by Singh and Harish (2002)
have brought out the effect of reactivity
insertion rate on the mechanical energy
release and have indicated that the energy
release gets stabilised to 1,000 MJ beyond
$100/s. This implies that there is no need
of assessing critically the energy release
vol xlvi no 52
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DISCUSSION
mechanisms, once it is assumed to be
>1,000 MJ. In fact, this is the approach followed for the reactors of 1980s. However,
the authors in the same paper indicated
clearly that the realistic estimate for oxide
fuel is limited to $50/s, which would result
in an energy release of 23 MJ. Subsequently,
motivation to understand the realistic scenarios supported by tests data: TREAT
(Wright et al 1990), CABRI (Nissen et al
1986) and SEFOR (Caldarola et al 1972)
could have led to choose the value even
lower than $50/s for PFBR (Sathiyasheela
and Srinivasan 2008). This comes from the
improved understanding on the in-pin motion of molten fuel during pre-disassembly
phase, swept out effects of the core due to
shearing forces of the coolant and clad
vapours. The reactor physics analysis carried out with such improved understandings indicates that the whole core sodium
void coeficient is $2.4.
A few fundamental experiments, carried
out at IGCAR to quantify the mechanical
energy release due to molten fuel coolant
interaction effects have indicated that the
transient pressure due to Molten Fuel Coolant Interaction effects is insigniicant. In
spite of these data, an energy release of
100 MJ is pessimistically considered, for
which the RCB is designed. It has been
assessed through backward computations
that the 100 MJ of energy release corresponds to reactivity insertion rate of $66/s.
This is possible only through melting of almost the entire whole core and subsequent
gravitational fall of the molten core coherently, thereby increasing the effective density and reactivity. No other initiating
events such as gas and oil entries, sodium
voiding or uncontrolled withdrawal of all
the control rods can add such a reactivity
insertion rate. This clearly demonstrates
that 100 MJ is the pessimistic energy release
postulated for PFBR just to raise the conidence among designers and regulators.
Containment Design Aspects
Under normal operation, the pressure acting
on containment structures is insigniicant.
However, under a CDA, the main vessel and
top shield would be subjected to a static
equivalent pressure of about 2 bar (cumulative effect of short duration high peak
pressure) (Chellapandi et al 2002). The RCB
would be pressurised due to temperature
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december 24, 2011
rise consequent to chemical ire of sodium
ejected during CDA from the narrow
penetrations provided in the top shield
for facilitating the rotation of plugs during fuel handling operations. Thus, the
extent of temperature and pressure rise
in the RCB depends on the quantity of
sodium released to RCB from sodium pool
contained in the main vessel. The phenomenon of sodium release is explained
briely below. More details can be found
in Chellapandi et al (2010).
CDA results in the formation of a core
bubble constituted by vaporised sodium,
metal and fuel materials, having high
thermal energy. The process of rapid
expansion manifests as mechanical energy
release. The maximum work potential of
the bubble is the energy release when it
expands from its initial pressure till it
attains 1 atmosphere.
As far as mechanical energy release is
concerned, the primary containment is
subjected to transient forces in two consequent phases: direct impact pressure on
the radial and bottom portions causing
radial expansions of the reactor internals
(irst phase) and impact of accelerated
sodium at the bottom of the top shield
causing sodium leak through top shield
penetrations as well as local bulging of the
upper portion of the main vessel (second
phase). In these phases, the reactor internals absorb the maximum energy released
through core bubble expansion (~80%).
Remaining energy is associated with the
sodium release phenomenon through top
shield penetrations: since the top shield is
rigid structure, energy imposed on the top
shield is absorbed by the above core structures (control plug consisting of many
long and slender structures) to a large extent and further by a number of long ductile tie rods incorporated at the periphery
of the top shield support embedment, developing minimum strains on the bolts in
the top shield. Consequently, the lifting of
rotatable plugs is insigniicant due to sodium slug impact. These apart, the gaps in
the penetrations through which sodium
can leak out are kept to the minimum as
achievable by the manufacturing process.
Hence, the sodium release to RCB through
top shield penetrations is found to be less
than 350 kg and associated sodium ire
above top shield provides the basis for
vol xlvi no 52
deining the design pressure of 25 kPa for
RCB introducing adequate conservatisms
in the luid low and sodium ire simulations
(Chellapandi 2002).
Energy Absorbing Potential
of PFBR Primary Containment
A parametric study was performed on
mechanical release without linking to the
design basis value (100 MJ). The analysis
indicated that higher the energy release,
larger is the vessel deformation. This
allows the liquid sodium free level to fall
down in the irst phase, thereby limiting
the magnitude and duration of sodium
slug impact force on the roof slab. This
has been demonstrated in our simulated
tests involving many scaled down models
employing low density chemical explosive. This has also been demonstrated
experimentally and numerically that the
main vessel alone can absorb more than
1,200 MJ of energy before failure, thanks
to the high ductility of austenitic stainless
steels used as structural material in PFBR
(TBRL Report 2002).
The core bubble in the reactor environment can release a mechanical energy of
70 to 80 MJ, due to the constraints imposed by the reactor internals that do not
allow the bubble to attain 1 atmosphere.
However, in the simulated tests to assess
the mechanical consequences (structural
integrity and sodium release), the mass of
low density chemical explosive is such that
it released an energy of 110 MJ. This implies
that the applied mechanical energy for
assessment of mechanical consequences is
about 30% higher (TBRL Report 2002).
In view of the above, it is concluded that
the main vessel has potential to absorb
more than 1,200 MJ of energy and the
sodium release to RCB would not exceed
design basis leak (350 kg), even if the
mechanical energy release exceeds beyond
the design basis value (100 MJ). This substantiates the perceptions of designers that
there may not be any need of RCB for PFBR;
instead a simple coninement would sufice
(Paranjpe 1992; Paranjpe 1991).
Economy without Sacriice
of Safety
While designing the experimental reactors
or small size reactors, the “hell for strong”
concepts are generally adopted without
77
DISCUSSION
giving high emphasis on economy. The
economics with due concern for safety
forms the fundamental basis for any
engineering activity or industry to succeed
and providing adequate conservatism is
the essence of design. To achieve targeted
commercial exploitation, particularly for
the large size reactors, precise assessment
of design basis loads is essential for the
designers worldwide and adequate experimental knowledge needs to be incorporated in the safety evaluation. In this
respect, it is worth to comment on the conservatism embedded in SNR-300 or SPX-1 in
the deinition of CDA energy release. During the design phase, high conservative
energy release values were deined for
these reactors (370 MJ (Hennies 1989) and
800 MJ respectively (Guezence et al). Subsequently, the reassessment of CDA for SNR
300 with experimental indings showed
that design basis of 370 MJ is far conservative and mechanical energy release of
more than 100 to 150 MJ still could not be
substantiated by any physically conclusive
line of arguments (Hennies 1989). In the
case of SPX-1, subsequent to its construction,
it was shown that even with very permissible assumptions, CDA energy overestimate
is by a factor of 2 or 3 for SPX-1 (Guezence).
Hence, the design values adopted for
SNR-300 and SPX-1 should not be the basis
for new reactors. In fact, they are not the
basis for new reactor in western Europe.
For the next fast reactor in France (SPX-2),
such conservative approach that was
adopted for SPX-1 was not followed: the
energy release for SPX-2 was estimated
only 150 MJ (Dell Beccaro et al 1989).
Accordingly, the design of PFBR conceived in 1992 was reviewed critically including the issues related to reactor safety
based on accumulated numerical and experimental data from in-house and international resources (Bhoje et al 2001). Revision of design basis loads is based on the
comments received from extensive multitier review mechanisms (about 100 Project
Design Safety Committee Meetings and
numerous review input from 17 specialist
groups). Among them, the RCB design
pressure is one important parameter. The
design pressure of 25 kPa approved by the
safety committee has undergone investigations by multidisciplinary experts backed
up with numerous high quality data. In
78
this exercise, we have established a few
international benchmark data particularly
in the domain of the mechanical consequences of CDA (Chellapandi et al 2010).
International Perspective
The design experience shows that a larger
safety margin to the conservative limits is
possible. This, in turn, is relected also in
the practical experience with the prototype plants: over 1,00,000 FBR fuel-pins
up to a burn-up of about 1,00,000 MWd/t
were irradiated worldwide (Von Heinz
Vossebrecker 1999), in which the failure
rate was generally lower than by an order
of magnitude, below the limit of 1%,
depending upon the design. Regarding
reactivity insertion rate due to sodium
voiding for the large core, results of analysis
carried out by Hummel et al (1976a) using
a more recent version of the SAS code
can be referred, where it is noted that a
maximum ramp rate of $30/s results from
sodium voiding even for a 4,000 MWt reactor design with a $7 sodium void contribution. Hummel et al (1976a and 1976b)
performed similar calculations for CRBRP
and found a maximum voiding rate of
$25/s; they suggested that a loss-of-low
accident would not lead to prompt critical
conditions with a possibility of subsequent
large energy releases. It was shown by
theory and experiments at SEFOR between
1969 and 1972 that large mixed-oxide
fuelled cores always have a strong negative power coeficient and a good control
stability against reactivity or coolant-low
oscillations (Caldarola et al 1972). In addition, it was concluded that the strong negative Doppler coeficient together with the
negative structural and fuel expansion coeficients predominate over the positive
sodium void coeficient in central parts of
LMFBR cores. Further it is conirmed that
the oxide fuel is favourable (some in-pile
demonstration exists) by clad vaporisation.
The cladding boiling point is roughly equal
to the melting point of the fuel (uranium
plutonium) suggesting that steel vapour
from clad boiling can provide an effective
dispersal mechanism. Explosive sodium
vapour formation is not likely to be
involved for oxide-fuel/sodium systems.
Fauske (1976) summarised the technology
status during CRBRP licensing. In contrast
to FFTF (~0 $), CRBRP had a signiicantly
december 24, 2011
positive coolant void reactivity worth (~3 $).
Regulators agreed with the applicant that
CDA energetics would be excluded from
the plant design basis (Strawbridge et al
1985). Ultimately, a construction permit
was issued for CRBRP, but after its construction was abandoned for reasons not
related to safety (Cahalan 2005).
With reference to mechanical energy
release, it is worth to look at the FRAX
calculations performed for EFR (Bernard
2005), which indicates that the mechanical
energy release is 1,000 MJ with the assumption of coherent core and it is insigniicant
for the realistic core. The most likely estimate was lower than 150 MJ for this 1,500
MWe MOX reactor. It is also worth to note
the conclusions of Richard (1977), where it
is stated that the mechanical energy itself is
not the number of interest for any given
reactor, rather it is instantaneous forces
and pressures which could raise the reactor
vessel head and/or violate the containment.
The energy available to raise the vessel head
is about one-tenth to one-ifth of the energy
of expansion to one atmosphere. This justiies that sodium release is not strongly
linked to mechanical energy in CDA.
The above few extracts raise conidence
on the CDA analysis and design provisions
adopted for PFBR, which are all of course
based on widely accepted practices followed by designers of sodium cooled fast
reactors worldwide.
Epilogue
Sodium cooled fast reactors have several
inherent and engineered safety features.
Further for PFBR, robust safety features
have been incorporated and qualiied
based on extensive theoretical and experimental testing and evaluations, which
have been reviewed elaborately through
multi-tier review process under the Atomic
Energy Regulatory Board. With these
justiications, it can be stated with high
conidence that there is no concern on
core meltdown accident and associated
mechanical consequences such as failure
of containment system in PFBR.
References
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EFR Safety Approach”, Workshop on Severe
Accident in Sodium Fast Reactors, Cadarache,
8 and 9 December.
Bhoje, S B et al (1990): “Safety Criteria for PBFR”,
AERB Document.
vol xlvi no 52
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DISCUSSION
– (2001): “Cost Competitiveness of Breeder Reactor”,
Indira Gandhi Centre for Atomic Research Newsletter, October issue.
Caldarola, L, D D Freeman, P Greebeler et al (1972):
“SEFOR Experimental Results and Application to
LMFBR’s”, Proc Int Conf Engineering of Fast Reactors for Safety and Reliable Operation, Karlsruhe,
1312-30.
Cahalan, J (2005): “FFTF and CRBRP Licensing”,
Workshop on Severe Accident in Sodium Fast
Reactors, Cadarache, 8 and 9 December.
Chellapandi, P, S C Chetal and Baldev Raj (2010):
“Structural Integrity Assessment of Reactor
Assembly Components of a Pool-Type Sodium Fast
Reactor in a Core Disruptive Accident–I: Development of Computer Code and Validations”, Nuclear
Technology, (172): 1-15.
– (2010): “Structural Integrity Assessment of Reactor Assembly Components of a Pool-Type Sodium
Fast Reactor in a Core Disruptive Accident–II:
Analysis for a 500 MW (electric) Prototype
Fast Breeder Reactor”, Nuclear Technology, (172):
1-13.
– (2011): “Safety Features of 500 MWe Prototype
Fast Breeder Reactor”, Nuclear Safety Systems:
Measurement Modelling Assessment and Veriication, Chapter 7a, Woodhead Publication
(in press).
Chellapandi, P, K Velusamy, S E Kannan, Om Pal Singh,
S C Chetal and S B Bhoje (2002): “Core Disruptive
Accident Analysis in Prototype Fast Breeder Reactor”,
1st National Conference on Nuclear Reactor Safety,
Mumbai, India, 25-27 November.
Chetal, S C, V Balasubramaniyan, P Chellapandi,
P Mohanakrishnan, P Puthiyavinayagam, C P Pillai,
S Raghupathy, T K Shanmugham and C Sivathanu
Pillai (2006): “The Design of the Prototype Fast
Breeder Reactor”, Nuclear Engineering and
Design, 236: 852-60.
Dell Beccaro, R et al (1989): “HCDA Consequences in
the SUPER PHENIX 2 Containment”, Tran 12th
Structural Mechanics in Reactor Technology,
Vol E, 287-90.
Endo H et al (1994): “A Study of the Initiating Phase
Scenario of Unprotected Loss of Flow in a 600
MWe MOX Homogeneous Core”, IAEA TCM on
Material Coolant Interactions and Material Movement and Relocation in LMFRs, O-arai, Ibaraki,
Japan.
Fauske, H K (1976): “The Role of Core-Disruptive
Accidents in Design and Licensing of LMFBRs”,
Nuclear Safety, September-October, Vol 17, No 5,
pp 550-67.
Guezence, J Y, M Butaye and M Natta: “A Comparison
of the Safety Features of PWR and LMFBR in
France”, (II): 635-44.
Hennies, H H (1989): “The Fast-Neutron-Breeder
Fission Reactor – Safety Issues in Reactor Design
and Operation” (Lonlies: Report of Kernforschungszentrum Karlsruhe GmbH, Karlsruhe, Germany).
Hummel, H H, Kalimullah and P A Pizzica (1976a):
“Limits on Sodium Voiding Reactivity Addition
Rate for Very Large LMFBRs”, Nucl Sci Eng 59,
440 and also ANL Report 76-77.
– (1976b): “Studies of Unprotected LOF Accidents
for CRBR”, ANL Report 76-51.
Nissen, K L, W Pfrang, D Struwe, W Vath, M H Wood,
M Cranga, C Melis Struzik and J C Sato I (1986):
“Interpretation of Selected CABRI Loss of Low
Experiments”, Procs Conf on Science and Techn
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(1): 115-20.
Paranjpe, S R (1991): “An Update on Indian Fast
Breeder Programme” in International Conference
Inadequate Basis for Safety
of the PFBR
M V Ramana, Ashwin Kumar
W
e welcome the response from
S C Chetal and P Chellapandi of
the department of atomic energy
(DAE) to concerns we have raised about
the safety of the Prototype Fast Breeder
Reactor (PFBR) in a core disruptive accident (CDA) (Kumar and Ramana 2011).1
However, there are persistent disagreements that we outline below. It is important to take into account the consequences
of a CDA because safety systems, including
multiple ones, can fail.
The authors’ response is based on the
assumption that in a CDA, reactivity of the
core increases mainly from the formation
of voids in the sodium coolant once it
boils. This is inadequate, because it ignores
the increase in the reactivity if large parts
of the core collapse. The assumed reactivity increase in an accident critically inluences the subsequent safety analysis and
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december 24, 2011
the resulting assessments of the adequacy
of the containment, and therefore this is a
matter of concern.
The calculations of energy released from
a collapse of the core depend sensitively on
the reactivity insertion rate (Wirtz 1978).
This is the rate at which rearrangement of
the fuel increases (“inserts”) the reactivity
within the core. The DAE’s studies of the
PFBR have shown that when the entire core
participates in an accident, the resulting
reactivity insertion rates are much higher
than what the authors now claim are the
maximum possible rates in the event of a
CDA (Singh and Harish 2002). This weakens their argument that a mechanical energy release of 100 megajoules (MJ), for
which the PFBR is designed, is pessimistic.
The modelling of severe accidents using
“mechanistic models”, such as that of the
DAE, is made dificult by uncertainties,
vol xlvi no 52
on Fast Reactors and Related Fuel Cycles, 1a.4-11.4-9, Atomic Energy Society of Japan, Kyoto,
PNC and JAPC.
Richard, Wilson (1977): “Physics of Liquid Metal Fast
Breeder Reactor Safety”, Reviews of Modern Physics,
American Physics Society, (49), 4: 893-924.
Sathiyasheela, T and G S Srinivasan (2008): “Analysis
of ULOF, UTOP Events and CDA Energy Release”,
PFBR/01117/DN/1022/R-A.
Singh, Om Pal and R Harish (2002): “Energetics of
Core Disruptive Accident for Different Fuels for a
Medium Sized Fast Reactor”, Annals of Nuclear
Energy, 29: 673-83.
Stark, H, D Barnes and U Wehmann (1991): “Core
Optimisation of European Fast Reactor FR 91”,
International Conference on Fast Reactors and
Related Fuel Cycles, Vol 1, Kyoto, Japan.
Strawbridge, L E and G H Claire (1985): “Exclusion of
Core Disruptive Accidents from the Design Basis
Accident Envelope in CRBRP”, proceedings of the
International Meeting on Fast Reactor Safety,
Knoxville, Tennessee, American Nuclear Society,
CONF-850410, pp 317-27, 21-25 April.
TBRL Report (2002): “Investigation of Mechanical
Consequences of a Core Disruptive Accident in
Fast Breeder Reactor Based on Simulated Tests on
Scaled Down Models”, TBRL/IGCAR/TRIG/1997,
Terminal Ballistic Research Laboratory.
Von Heinz Vossebrecker (1999): “Special Safety Related
Thermal and Neutron Physics Characteristics
of Sodium Cooled Fast Breeder Reactors”,
Translation No IGC/GER/2, Ref : Warme Band 86,
Heft 1: 1-17.
Wright, A E, D S Dutt and L J Harrison (1990): “Fast
Reactor Safety Testing in Treat in the 1980s”,
International Fast Reactor Safety Meeting,
Session (2), Vol II: 233-43.
arising partly from the complexity of the
reactor core once its fuel has melted (Bell
1981). Therefore, studies of the energetics
of fast reactor accidents typically start by
specifying an initial reactivity insertion
rate; a typical igure is 100 $/s, though
sometimes even 200 $/s is used (KAERI
1997; Theofanous and Bell 1984).2 As has
been demonstrated previously using an
illustrative model of the effects of gravitational collapse of the entire PFBR core, reactivity insertions rates well above $100/s
are possible (Kumar and Ramana 2008).
Up until about 150 $/s, DAE’s studies themselves show that mechanical energy release is very sensitive to the reactivity
insertion rate (Singh and Harish 2002). A
reactivity insertion rate of even 100 $/s
would mean an energy release from a
CDA of 650 MJ, if one assumes the same
thermal to work energy conversion eficiency as DAE, much more than what the
reactor containment building (RCB) is
capable of withstanding.
Sodium Voiding
This puts a different complexion on the
matter of sodium voiding. It is not enough
to argue, as the authors have done, that
79
DISCUSSION
the contribution of sodium voiding alone
to the reactivity insertion rate is too small
to produce 100 MJ of energy in a CDA. The
question is whether it is possible to be certain that only a fraction of the core can
melt and collapse, leading to no more than
100 MJ of mechanical energy being produced in the accident. Such conidence
would not be justiied, because there are
several omissions in the DAE’s studies,
which undermine any claims that accident
progression is well understood (Kumar
and Ramana 2008; Kumar and Ramana
2009). For example, published studies of
loss of low accident progression in the
PFBR have ignored the effects of the
boiling of coolant (Paranjpe, Singh and
Harish 1992).
Furthermore, sensitivity of energy released to the reactivity insertion rate
should itself urge the consideration of the
entire range of potential insertion rates
when designing safety systems such as the
containment. Therefore, it is not appropriate to assume much lower rates, when the
underlying calculations are uncertain and
also omit processes that could worsen
the accident.
We are not privy to all the safety-relevant information in possession of IGCAR
engineers and scientists. For example,
Chetal and Chellapandi cite an internal
DAE report which evidently shows that
there has been a downward revision in the
magnitude of the DAE’s estimate of the reactivity addition. The only offered explanation, i e, improved understanding of old
experimental data (note that the references cited are from 1972, 1986, and 1990),
also implies that the reactor design was
inalised before reaching a full understanding of accident possibilities.
While new information about fast reactor
physics is desirable especially in light of
the DAE’s plans for many such reactors,
they also illustrate the incomplete and
evolving nature of knowledge of severe
accident behaviour. Together with the
possibility of high reactivity insertion
rates in the PFBR from collapse of the core,
this should have led the DAE to suspend
judgment about the design until safety
could be assured. Alternately, it might have
chosen a design while fully acknowledging
its dilemma of incomplete knowledge, and
the dificult choice it faced between safety
80
and economics. However, it did neither but
repeatedly expressed excessive conidence
in the safety of its design.
The dificulty of establishing an upper
bound on the reactivity insertion rate
from core collapse has consequences for
the integrity of the RCB in a CDA. As
explained previously (Kumar and Ramana
2011), sodium burns in air and raises the
pressure of its surroundings. Sodium
escape into the RCB can happen through
a variety of pathways, in spite of the
deformation of the pressure vessel produced by more energetic accidents, which
the authors point out. Estimates of the
amount of sodium escaping would only
increase when higher insertion rates, and
consequently higher mechanical energy
releases in a CDA, are considered (Kumar
and Ramana 2008). Therefore, the assertion “the sodium release to RCB would not
exceed design basis leak (350 kg), even if
the mechanical energy release exceeds
beyond the design basis value (100 MJ)” is
not justiied.
Furthermore, it is possible for enough
sodium to be released so that the pressures generated in the RCB are high
enough to cause it to rupture, releasing radioactive material into the environment
(the total amount of sodium in the primary
vessel is approximately 1,200 tonnes, and
escape of only a small fraction of this is
necessary for the RCB’s rupture).3 Therefore, we hope that the DAE will not allow
its conidence in the integrity of the pressure vessel in a CDA to prevent enhancements to the safety of the RCB.
Building a strong reactor containment
building is necessary because despite the
best efforts of designers and operators,
severe accidents might occur at any nuclear
reactor, including the PFBR. This requires
also that the RCB is designed for a suficiently severe accident; otherwise it
merely creates the illusion of safety without actually achieving it.
M V Ramana (ramana@princeton.edu)
is with the Program on Science and
Global Security, Woodrow Wilson School
of Public and International Affairs,
Princeton University, Princeton, US.
Ashwin Kumar (ashwink@cmu.edu) is a PhD
scholar at the department of engineering and
public policy, Carnegie Mellon University,
Pittsburgh, US.
december 24, 2011
Notes
1 We use DAE as an umbrella term to denote both
the department of atomic energy as well as the
associated institutions like the Indira Gandhi
Centre for Atomic Research.
2 A “$” worth of reactivity is deined in nuclear
reactor engineering as the increase in reactivity
equivalent to the fraction of delayed neutrons;
this is a natural scale for measuring the consequences of reactivity increases because an increase of this magnitude would make the reactor
critical on prompt neutrons alone, making
control dificult.
3 Extrapolation of the DAE’s estimate that 350 kg of
sodium would escape into the RCB in a 100 MJ
CDA, suggests 3,300 kg of sodium would escape
in a 1,200 MJ accident (Kumar and Ramana
2008). The threat posed by the large inventory of
sodium to the RCB’s integrity is more general, and
we hope that the DAE has rigorously examined
other pathways for the escape of sodium, as during an earthquake.
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