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Inadequate Basis for Safety of the PFBR

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.

DISCUSSION

Severe Accident Assessment for PFBR: A Designer’s Perspective

S C Chetal, P Chellapandi

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:

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.

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 justification 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 fission 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 identified 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 coefficient 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 insignificant (< 1 megajoules (MJ)) (Sathiyasheela and Srinivasan 2008). This void coefficient is smaller compared to other

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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, filled with nitrogen. RCB, top shield, main vessel along with safety vessel are designed to meet the specified 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 confidence and also considering the world over operating experience and trend. This is the justification 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

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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 coefficient 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 insignificant. 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 confidence among designers and regulators.

Containment Design Aspects

Under normal operation, the pressure acting on containment structures is insignificant. 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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rise consequent to chemical fire 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 briefly 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 (first 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 insignificant 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 fire above top shield provides the basis for

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defining the design pressure of 25 kPa for RCB introducing adequate conservatisms in the fluid flow and sodium fire 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 first 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 confinement would suffice (Paranjpe 1992; Paranjpe 1991).

Economy without Sacrifice of Safety

While designing the experimental reactors or small size reactors, the “hell for strong” concepts are generally adopted without

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 definition of CDA energy release. During the design phase, high conservative energy release values were defined for these reactors (370 MJ (Hennies 1989) and 800 MJ respectively (Guezence et al). Subsequently, the reassessment of CDA for SNR 300 with experimental findings 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 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 reflected 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-flow 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 coefficient and a good control stability against reactivity or coolant-flow oscillations (Caldarola et al 1972). In addition, it was concluded that the strong negative Doppler coefficient together with the negative structural and fuel expansion coefficients predominate over the positive sodium void coefficient in central parts of LMFBR cores. Further it is confirmed 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 significantly

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 insignificant 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-fifth of the energy of expansion to one atmosphere. This justifies that sodium release is not strongly linked to mechanical energy in CDA.

The above few extracts raise confidence 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 qualified 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 justifications, it can be stated with high confidence that there is no concern on core meltdown accident and associated mechanical consequences such as failure of containment system in PFBR.

References

Bernard Carluec (2005): “LMFBR Severe Accident 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.

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– (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 Verification, 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 of Fast Reactor Safety, Guernsey, BNES, London, (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

    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 figure is 100 $/s, though sometimes even 200 $/s is used (KAERI

    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 influences the subsequent safety analysis and

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    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 difficult by uncertainties,

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    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 efficiency 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

    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 confidence would not be justified, 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 flow 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 finalised 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 difficult choice it faced between safety and economics. However, it did neither but repeatedly expressed excessive confidence in the safety of its design.

    The difficulty 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 justified.

    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 confidence 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 nuc lear reactor, including the PFBR. This requires also that the RCB is designed for a sufficiently 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.

    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 defined 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 difficult.

    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.

    References

    Bell, C R (1981): “Multiphase, Multicomponent Hydrodynamics in HCDA Analysis: Present Status and Future Trends”, Nuclear Engineering and Design

    68: 91-99. KAERI (1997): Review of Core Disruptive Accident Analysis for Liquid-metal Cooled Fast Reactors, Korea Atomic Energy Research Institute. Kumar, Ashwin and M V Ramana (2008): “Compromising Safety: Design Choices and Severe Accident

    Possibilities in India’s Prototype Fast Breeder Reactor”, Science and Global Security, 16: 87-114.

  • (2009): “Reply by Authors”, Science & Global Security 17 (2-3): 197-200, http://www.tandfonline.com/ doi/abs/10.1080/ 08929880903451397.
  • (2011): “The Limits of Safety Analysis: Severe Nuclear Accident Possibilities at the PFBR”, Economic & Political Weekly, Vol XLVI, No 43, 22 October, 44-49.
  • Paranjpe, S R, Om Pal Singh and R Harish (1992): “Influence of a Positive Sodium Void Coefficient of Reactivity on the Consequences of Transient Overpower and Loss-of-Flow Accidents in a Medium-sized Fast Reactor”, Annals of Nuclear Energy, 19 (7): 369-75.

    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.

    Theofanous, T G and C R Bell (1984): Assessment of CRBR Core Disruptive Accident Energetics, Los Alamos National Laboratory.

    Wirtz, Karl (1978): Lectures on Fast Reactors, La Grange Park, IL: American Nuclear Society.

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