Compromising Safety: Design Choices and Severe Accident Possibilities in India's Prototype Fast Breeder Reactor

NOTES AND REFERENCES

• Fast breeder reactors are thus termed because they are based on energetic (fast) neutrons and because they produce (breed) more fissile material than they consume
   Google Scholar
• Bhabha , H. J. and Prasad , N. B. . “A Study of the Contribution of Atomic Energy to a Power Programme in India” . paper presented at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy . Geneva. pp. 89 – 101 . R. Chidambaram and C. Ganguly, “Plutonium and Thorium in the Indian Nuclear Programme,” Current Science 70(1) (1996): 21–35. While uranium can be used to fuel reactors, thorium cannot fission at low energies and so cannot fuel a reactor directly. But it can be converted to the isotope uranium-233 through neutron absorption followed by radioactive decay. Uranium-233, in turn, is fissile and can fuel reactors
   Google Scholar
• Suresh , K. V. Kumar . 2002 . “Fast Breeder Test Reactor. 15 Years of Operating Experience” . paper presented at the Technical Meeting on Operational and Decommissioning Experience with Fast Reactors . March 11–15 2002 , Cadarache. pp. 15 – 27 .
   Google Scholar
• Prasad , R. 2001 . “India: FBTR Passes 53-Day Continuous Operation Test,” . The Hindu , 22 March
   Google Scholar
• DAE . 2006 . “Atomic Energy in India: A Perspective” , Department of Atomic Energy, Government of India .
   Google Scholar
• Bidwai , Praful . 1983 . “The Fast Breeder Reactor: DAE's Strange Nuclear Priorities,” . The Times of India , 31 August
   Google Scholar
• DAE . 1991 . “Performance Budget 1990–91” , Mumbai : Department of Atomic Energy .
   Google Scholar
• Hibbs , Mark . 1990 . “India's New Breeder Will be on Line by 2000, Iyengar Says,” . Nucleonics Week , 31 ( 42 ) : 11
   Google Scholar
• Subramanian , T. S. 2004 . “A Milestone at Kalpakkam,” . Frontline , 19 November
   Google Scholar
• Grover , R. B. and Chandra , Subash . 2006 . “Scenario for Growth of Electricity in India,” . Energy Policy , 34 ( 17 ) : 2834 – 2847 . M. R. Srinivasan, R. B. Grover, and S. A. Bhardwaj, “Nuclear Power in India: Winds of Change,” Economic and Political Weekly XL(49) (2005): 5183–5188
   Web of Science ®Google Scholar
• Another organization, the Atomic Energy Regulatory Board (AERB), has apparently reviewed some of the safety features of the PFBR, but the reviews are not publicly available
   Google Scholar
• A reduction of coolant density has three effects. The reduced coolant absorbs fewer neutrons, the mean energy of neutrons is higher, and there is more leakage. In a fast reactor, higher neutron energy results in more Pu-239 fissions and therefore the first two effects increase reactivity. Leakage effects are important only near the periphery of the core, and therefore become less important as a whole as the volume of the core increases
   Google Scholar
• IAEA . 2006 . “Fast Reactor Database: 2006 Update” , Vienna : International Atomic Energy Agency .
   Google Scholar
• This is a natural scale because a reactivity increase of that magnitude would allow the reactor to be critical on only prompt neutrons; accordingly very rapid rates of power increase would be possible. The resultant timescales would be too short for control rods to be inserted and stabilize the reactor
   Google Scholar
• It is possible to obtain a positive void coefficient even when thermal neutrons are responsible for most of the fission. This happens when the coolant plays an important role in neutron absorption, but does not contribute much to reducing neutron energies to the thermal range. This is true to some extent in Pressurized Heavy Water Reactors, but is more prominent in the graphite moderated RBMK design, as exhibited dramatically in the 1986 Chernobyl accident
   Google Scholar
• Beynon , T. D. 1974 . “The Nuclear Physics of Fast Reactors,” . Reports on Progress in Physics , 37 : 951 – 1034 . The Doppler effect results from increased absorption of neutrons by fertile uranium-238 nuclei. When the fuel in a reactor becomes hotter, the average speed of the nuclei increases, thereby generating a wider range of relative neutron speeds or energies. The bulk of the fuel is uranium-238, which resonantly absorbs neutrons at specific energies. As the random motion of the uranium-238 nuclei increases, the probability of neutrons having an energy corresponding to one of these resonant energies increases. This increases the number of neutrons absorbed by uranium-238 nuclei, reducing the number of neutrons available to cause fission, reducing the reactivity. As the ratio of uranium-238 to fissile nuclei increases, the Doppler coefficient becomes increasingly negative, at 1029
   Web of Science ®Google Scholar
• There are also advantages of using a sodium coolant. First, because operations at low pressures are possible, a breach in the coolant pipes would not by itself lead to boiling of the remaining coolant. Second, high thermal conductivity allows for more reliable decay heat removal under accident conditions
   Google Scholar
• Ivanenko , V. N. and Zybin , V. A. 1996 . “Fast Reactor Sodium Systems Operation Experience and Leak-before-Break Criterion” . paper presented at the Technical Committee Meeting on Evaluation of Radioactive Materials Release and Sodium Fires in Fast Reactors, IWGFR—92, O-arai . November 11–14 1996 , Ibaraki (Japan). pp. 255 – 269 . The Russian BN-600 reactor has had numerous sodium leaks and fires; N. N. Oshkanov, M. V. Bakanov, and O. A. Potapov, “Experience in Operating the BN-600 Unit at the Belyi Yar Nuclear Power Plant,” Atomic Energy 96(5) (2004): 315–319
   Google Scholar
• Hans , A. Bethe and Tate , J. H. 1956 . “An Estimate of the Order of Magnitude of the Explosion When the Core of a Fast Reactor Collapses” , United Kingdom Atomic Energy Agency .
   Google Scholar
• There was a reason to ignore these feedbacks. Early reactors were small and had high enrichments. Because the Doppler absorption is primarily from fertile nuclei, the corresponding reactivity feedback effects were small
   Google Scholar
• Ibid
   Google Scholar
• Wilson , Richard . 1977 . “Physics of Liquid Metal Fast Breeder Reactor Safety,” . Reviews of Modern Physics , 49 ( 4 ) : 893 – 924 .
   Web of Science ®Google Scholar
• Bell , C. R. 1981 . “Multiphase, Multicomponent Hydrodynamics in HCDA Analysis: Present Status and Future Trends,” . Nuclear Engineering and Design , 68 : 91 – 99 . The uncertainties arise from significant core motions, large-scale fluid dynamics, and variable neutronic states, and a variety of possible flow regimes, especially when many of the core components have failed or melted. In addition, feedbacks are often nonlinear, such as the relationship between liquid temperatures and vapor pressures, or between material motions and reactivity effects
   Web of Science ®Google Scholar
• Wilson . 911 op. cit
   Google Scholar
• Jackson , J. F. and Nicholson , R. B. 1972 . “Venus-II: An LMFBR Disassembly Program” , 88 – 96 . Argonne, IL : Applied Physics Division, Argonne National Laboratory .
   Google Scholar
• KAERI . 1997 . “Review of Core Disruptive Accident Analysis for Liquid-Metal Cooled Fast Reactors” , Korea Atomic Energy Research Institute .
   Google Scholar
• Theofanous , T. G. and Bell , C. R. 1984 . “Assessment of CRBR Core Disruptive Accident Energetics” , Los Alamos National Laboratory .
   Google Scholar
• Badham , V. and Chan , C. K. 1979 . “A Look at Alternative Core-Disruptive Accidents in LMFBRs—Part II: Neutronic and Fuel Element Behavior,” . Nuclear Engineering and Design , 55 : 1 – 7 . The authors estimate that coherent core collapse leads to a reactivity insertion of around 0.3, or approximately $75, and that this occurs in around 0.35 seconds
   Web of Science ®Google Scholar
• Om , Pal Singh and Harish , R. 2002 . “Energetics of Core Disruptive Accident for Different Fuels for a Medium Sized Fast Reactor,” . Annals of Nuclear Energy , 29 : 673 – 683 .
   Web of Science ®Google Scholar
• “Energetics of Core Disruptive Accident for Different Fuels for a Medium Sized Fast Reactor,” 678
   Google Scholar
• Jackson , J. F. and Nicholson , R. B. 1980 . “Venus-II: An LMFBR Disassembly Program,” . In NEA, Venus-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions , Nuclear Energy Agency . http://www.nea.fr/abs/html/nesc0511.html [cited 21 October 2008])
   Google Scholar
• Jackson and Nicholson . “Venus-II: An LMFBR Disassembly Program,” 8
   Google Scholar
• This is redrawn from a figure in Singh and Harish, op. cit
   Google Scholar
• 679 Ibid
   Google Scholar
• Table 6 of Ibid. shows 2130 MJ thermal energy and 23 MJ mechanical energy for an insertion rate of 50 $/s
   Google Scholar
• Berthoud , Georges . 2000 . “Vapor Explosions,” . Annual Review of Fluid Mechanics , 32 : 573 – 611 .
   Web of Science ®Google Scholar
• IAEA . 1996 . “Fast Reactor Fuel Failures and Steam Generator Leaks: Transient and Accident Analysis Approaches” , Vienna : International Atomic Energy Agency .
   Google Scholar
• Berthoud . 594 op. cit.
   Google Scholar
• Wilson . 915 op. cit.
   Google Scholar
• Thomas , B. Cochran . 1974 . The Liquid Metal Fast Breeder Reactor: An Environmental and Economic Critique , 183 Washington, D.C. : Resources for the Future, Inc. .
   Google Scholar
• Chellapandi , P. 2003 . “Analysis for Mechanical Consequences of a Core Disruptive Accident in Prototype Fast Breeder Reactor” . paper presented at the 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17) . August 17–22 2003 , Prague (Czech Republic). pp. 1 – 8 .
   Google Scholar
• One concern about that claim is that the analysis of the reactor vessel's integrity does not incorporate the possibility of corrosion
   Google Scholar
• Lee , S. M. 2002 . “Status of Fast Reactor Development in India: April 2001–March 2002” . paper presented at the Technical Working Group on Fast Reactors . April 22–26 2002 , Karlsruhe, Germany. pp. 333 – 385 . 343 – 344 .
   Google Scholar
• Chellapandi . op. cit
   Google Scholar
• It is assumed that the elongation increases linearly for higher energies
   Google Scholar
• Gluekler , E. L. and Huang , T. C. 1979 . “Response of Secondary Containment to Presence of Sodium and Hydrogen,” . Nuclear Engineering and Design , 55 : 283 – 291 . A pool fire can also occur in the primary vessel if the cover gas is lost and air takes its place, as has been assumed in some analyses, for example, KAERI, “Preliminary Safety Analysis for Key Design Features of Kalimer” (Korea Atomic Energy Research Institute, 2000)
   Web of Science ®Google Scholar
• IGCAR . 2003 . Design of Prototype Fast Breeder Reactor , Indira Gandhi Centre for Atomic Research . www.igcar.ernet.in/broucher/design.pdf [cited 10 March 2006])
   Google Scholar
• Chetal , S. C. 2006 . “The Design of the Prototype Fast Breeder Reactor,” . Nuclear Engineering and Design , 236 : 852 – 860 .
   Web of Science ®Google Scholar
• Energy here refers to the mechanical energy, which is a small fraction (a few percent) of the total heat energy released
   Google Scholar
• It is assumed that the density of sodium in the reactor vessel remains constant, and the pressure in the containment building throughout the leakage remains much smaller than the pressure inside the reactor vessel, while the velocity is much higher
   Google Scholar
• Krishnan , L. V. and Garg , D. D. 1980 . “Scaling Laws for Contained Explosions,” . Nuclear Engineering and Design , 56 : 405 – 412 .
   Web of Science ®Google Scholar
• It is not clear what the leakage areas are in the DAE analysis. A spray fire requires unobstructed paths to the containment, which is more likely when the reactor vessel has suffered greater damage. Therefore, the leakage area for spray fires is likely to be higher. The calculations here do not require the magnitude of these areas, only their possible dependence on CDA energy
   Google Scholar
• The containment atmosphere has a volume of 87000 m3; IAEA, “Fast Reactor Database: 2006 Update.” At 30 o C and atmospheric pressure, around 24,000 kg of oxygen is present in this volume. Stoichiometrically, to burn 23 g of sodium, 32 g of oxygen are needed. Hence, the oxygen in the containment is sufficient to burn over 17 tons of sodium
   Google Scholar
• In older analyses, other values for the overpressure have been given. One paper presented at a conference in 1998 mentioned that “preliminary analysis indicate(s) a pressure rise of ∼ 10 kPa resulting from the sodium spray fire inside the containment.” S. B. Bhoje et al., “Impact of LMFBR Operating Experience on PFBR Design” (paper presented at the Technical Committee Meeting on Unusual Occurrences During LMFBR Operation, Vienna, 1998): 180–196. Another paper quotes an overpressure of 30 kPa resulting from the burning of 1.5 tons of sodium. S. B. Bhoje, “Status of Fast Reactor Development in India: April 1996–March 1997” (paper presented at the 30th Meeting of the International Working Group on Fast Reactors, Beijing, 14–17 May 1997): 79–109, 93
   Google Scholar
• During formation of sodium monoxide, 871 KJ is released for combustion of 4 moles (or 23∗4 grams) of sodium. This yields the aforementioned figure. Formation of peroxide results in a higher energy release per unit mass
   Google Scholar
• Yamaguchi , Akira and Tajima , Yuji . 2002 . “Numerical Investigations of Mass and Heat Transfer in Sodium Pool Combustion,” . Numerical Heat Transfer, Part A , 41 : 697 – 709 . The authors of this paper have chosen the maximum temperature difference between the pool and the flame of around 330 K. See. This gives the smallest temperature and pressure loading on the containment because as the pool temperature increases and becomes closer to the flame temperature, the fraction of the heat produced that is transferred to the containment atmosphere increases
   Web of Science ®Google Scholar
• “Numerical Investigations of Mass and Heat Transfer in Sodium Pool Combustion,”
   Google Scholar
• Cherdron , W. and Jordan , S. . “Thermodynamic Consequences of Sodium Leaks and Fires in Reactor Containments” . paper presented at the Specialists' Meeting on Sodium Fires . pp. 57 – 77 . Obninsk : USSR . This lies within the range of burning rates that have been found in the FAUNA test facility
   Google Scholar
• Chellapandi . op. cit. The burning time (15 minutes) is estimated by measuring from the plot given in the DAE paper, and is an upper bound taking into account measurement uncertainties
   Google Scholar
• Inhomogeneity could lead to local hot spots in the containment and a higher chance of failure
   Google Scholar
• This coefficient has been calculated by assuming heat transfer by natural convection, and that the height of the containment building is 1.5 times both the length and the width
   Google Scholar
• This is obtained by subtracting the area occupied by the intermediate heat exchanger, rotating plugs, and sodium pump from the area of the vessel head
   Google Scholar
• Chellapandi , P. , Jalaldeen , S. and Bhoje , S. B. 1992 . “Assessment of Structural Integrity of PFBR Reactor Assembly under HCDA” . paper presented at the National Symposium on Safety of Nuclear Power Plants and Other Facilities, Bhabha Atomic Research Centre . March 11–13 1992 , Trombay. III.30; G. Vaidyanathan et al., “Safety Considerations in the Design of PFBR” (paper presented at the Conceptual Designs of Advanced Fast Reactors: Technical Committee Meeting, Kalpakkam (India), 3–6 October 1995): 159–166
   Google Scholar
• The sodium inventory in the primary circuits is around 1,200 tons. Of this, a significant fraction could escape into the containment atmosphere during a CDA if the pressure vessel is ruptured. However, the amount of sodium that can be burnt is limited by the oxygen in the containment, which can sustain only up to 17 tons of sodium. But even this is much larger than the sodium leakage calculated from the scaling relationship, which is 3,360 kg of sodium leak for a CDA with 1,200 MJ mechanical energy. If one assumes that of the sodium entering the containment building from a ruptured primary vessel, about 15 tons were to burn as a spray fire, then the overpressure on the containment can exceed 150 kPa. Similarly high overpressures have been calculated in some accident scenarios for the much larger CRBRP containment. See Gluekler and Huang, op. cit., 287
   Google Scholar
• NRC . 1975 . “Reactor Safety Study (RSS), Study Director: N. V. Rasmussen” , Washington, D.C. : U.S. Nuclear Regulatory Commission .
   Google Scholar
• APS Study Group . 1985 . “Radionuclide Release from Severe Accidents at Nuclear Power Plants,” . Reviews of Modern Physics , 57 ( 3 ) : S1 – S154 . at S92
   Web of Science ®Google Scholar
• Pannerselvan , A. S. 1999 . “Close to a Critical Mess,” . Outlook , 8 November
   Google Scholar
• Madan , M. Mohan . 1994 . “Kaiga Questions: A Gaping Hole in Safety Standards,” . Frontline , 17 June : 84 – 85 .
   Google Scholar
• Ibid
   Google Scholar
• Lyman , Ed . 2008 . “Can Nuclear Plants Be Safer?,” . Bulletin of the Atomic Scientists , September/October : 34 – 37 . The numerator is a product of two measures of the ability of the containment, its design pressure, and its (large) volume, to withstand an accident. The choice of the denominator stems from the expectation that the energy that would potentially be released during an accident would be very roughly proportional to the power rating of the reactor. A different metric that has also been used is just the ratio of the containment volume to the thermal power, but this does not take into account the design pressure
   Web of Science ®Google Scholar
• APS Study Group . S94 op. cit.
   Google Scholar
• Bhardwaj , S. A. 2006 . “The Future 700 MWe Pressurized Heavy Water Reactor,” . Nuclear Engineering and Design , 236 : 861 – 871 .
   Web of Science ®Google Scholar
• IAEA . “Fast Reactor Database: 2006 Update.” This is the maximum coolant void coefficient and includes only regions with a positive void coolant reactivity worth
   Google Scholar
• Waltar and Reynolds . op. cit.
   Google Scholar
• Troyanov , M. F. 1990 . “The Present Status of Fast Breeder Reactors in the USSR,” . Philosophical Transactions of the Royal Society of London, Series A, Mathematical and Physical Sciences , 331 ( 1619 ) : 313 – 321 .
   Web of Science ®Google Scholar
• Paranjpe , S. R. 1992 . “Report on the Specialists' Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5–7, 1991,” . Nuclear Safety , 33 ( 4 ) : 506 – 513 .
   Google Scholar
• Moreover, the contribution of rapid core voiding to the reactivity of the core in a CDA cannot be neglected, as was clear in a previous section
   Google Scholar
• Paranjpe, op. cit
   Google Scholar
• Hohenemser , Christoph . 1988 . “The Accident at Chernobyl: Health and Environmental Consequences and the Implications for Risk Management,” . Annual Review of Energy , 13 : 383 – 428 .
   Google Scholar
• Paranjpe, op. cit
   Google Scholar
• Chetal . op. cit
   Google Scholar
• Paranjpe , S. R. 1991 . “An Update on Indian Fast Breeder Programme” . paper presented at the International Conference on Fast Reactors and Related Fuel Cycles . October 28–November 1 1991 , Kyoto. pp. 1.4-1 – 1.4-9 .
   Google Scholar
• Suchitra , J. Y. and Ramana , M. V. “The Costs of Power: Plutonium and the Economics of India's Prototype Fast Breeder Reactor” (in preparation)
   Google Scholar
• Bhoje , S. B. October 2001 . “Cost Competitiveness of Breeder Reactor,” . In Indira Gandhi Centre for Atomic Research Newsletter October ,
   Google Scholar
• Bhoje , S. B. 2002 . “Prototype Fast Breeder Reactor,” . Nu-Power , 16 ( 1–2 ) : 1 – 5 .
   Google Scholar
• AERB . 2003 . “Annual Report 2002–2003” , 12 Mumbai : Atomic Energy Regulatory Board . For example, the PFBR has a design pressure of only 25 kPa although the AERB recommended that the design pressure of the containment building “should not be less than 30 kPa.”
   Google Scholar
• Grover and Chandra . op. cit
   Google Scholar
• Graham , John . 1971 . Fast Reactor Safety , 175 – 176 . New York : Academic Press . Wirtz, op. cit., 138
   Google Scholar
• Wirtz . op. cit
   Google Scholar
• IAEA . “Fast Reactor Database: 2006 Update,” 24
   Google Scholar
• Waltar and Reynolds . op. cit
   Google Scholar
• Saito , T. and Suzuki , K. . “Role of Fission Products in Whole Core Accidents” . paper presented at the Specialists' Meeting on Role of Fission Products in Whole Core Accidents . 28 June–1 July 1977 , AERE Harwell (United Kingdom). pp. 41 – 55 .
   Google Scholar
• Eriksson , M. 1995 . “Inherent Safety of Fuels for Accelerator-Driven Systems,” . Nuclear Technology , 151 : 314 – 333 .
   Google Scholar
• Lallement , R. , Tuzov , A. and Bagley , K. Q. 1990 . “Fast Breedor Reactor Fuel Performances,” . Philosophical Transactions of the Royal Society of London, Series A, Mathematical and Physical Sciences , 331 ( 1619 ) : 343 – 354 . Saito and Suzuki, op. cit
   Web of Science ®Google Scholar
• Jagannathan , Venkatachari . 2003 . Powerful Leap: IGCAR Press Release http://www.igcar.ernet.in/press_releases/press6a.htm [cited 8 August 2004])
   Google Scholar
• IGCAR . 2005 . IGCAR Milestones , Indira Gandhi Centre for Atomic Research . [cited 24 February 2007]
   Google Scholar
• Chetal . op. cit
   Google Scholar
• Saito and Suzuki . op. cit
   Google Scholar
• Waltar and Reynolds . 558 op. cit.
   Google Scholar
• Paranjpe , S. R. , Pal Singh , Om and Harish , R. 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 – 375 . S. R. Paranjpe, Om Pal Singh, and R. Harish, “Sodium Void Coefficient and Fast Reactor Safety” (paper presented at the International Conference on Fast Reactors and Related Fuel Cycles, Kyoto, 28 October–1 November 1991): 13.6-1–13.6-9
   Web of Science ®Google Scholar
• Maeda , Koji , Katsuyama , Kozo and Asaga , Takeo . 2005 . “Fission Gas Release in FBR Mox Irradiated to High Burnup,” . Journal of Nuclear Materials , 346 : 244 – 252 .
   Web of Science ®Google Scholar
• Eriksson . op. cit
   Google Scholar
• IAEA . “Fast Reactor Fuel Failures and Steam Generator Leaks: Transient and Accident Analysis Approaches.”
   Google Scholar
• Aoyama , T. 2002 . “Operational Experience and Upgrading Program of the Experimental Fast Reactor Joyo” . paper presented at the Technical Meeting on Operational and Decommissioning Experience with Fast Reactors . March 11–15 2002 , Cadarache (France). pp. 244 – 249 .
   Google Scholar
• Kawakita , Takashi , Aoi , Sadanori and Hojuyama , Takeshi . 1991 . “Study on Large FBR Core Optimization to Enhance Core Safety” . paper presented at the Passive and Active Safety Features of LMFRs: Meeting of the Technical Working Group on Fast Reactors . November 5–7 1991 , O-arai, Japan.
   Google Scholar
• Juan , J. Carbajo . 2001 . “A Review of the Thermophysical Properties of Mox and UO2 Fuels,” . Journal of Nuclear Materials , 299 : 181 – 198 .
   Web of Science ®Google Scholar
• According to the DAE the intrinsic density of the fuel is 90% of its theoretical density, which corresponds to 10% porosity volume
   Google Scholar
• Baldev , R aj . 2002 . “Development of Fuels and Structural Materials for Fast Breeder Reactors,” . Sadhana , 27 ( 5 ) : 527 – 558 .
   Google Scholar
• Paranjpe , S. R. 1991 . “Core (PFBR) Safety Characteristics” . paper presented at the Passive and Active Safety Features of LMFRs: Meeting of the Technical Working Group on Fast Reactors . November 5–7 1991 , O-arai, Japan. pp. 176 – 190 . Om Pal Singh et al., “Energetics of a Hypothetical Core Disruptive Accident for Different Fuels for a Medium Sized Fast Reactor” (paper presented at the International Conference on Fast Reactors and Related Fuel Cycles: Current Status and Innovations Leading to Promising Plants FR1991, Kyoto (Japan), 1991): P3.3-1–P3.3-10
   Google Scholar
• Vasudeva Rao , P. R. 2006 . “Oxygen Potential and Thermal Conductivity of (U, Pu) Mixed Oxides,” . Journal of Nuclear Materials , 348 : 329 – 334 .
   Web of Science ®Google Scholar
• Carbajo . op. cit
   Google Scholar
• Paranjpe , Singh and Harish . op. cit
   Google Scholar
• The assumption is that loss of coolant flow occurs only due to a failure of power supply to the coolant pumps and this loss does not occur instantaneously but is characterized by a flow halving time constant
   Google Scholar
• Paranjpe , Singh and Harish . op. cit
   Google Scholar
• The boiling point of sodium is between 960°C and 1050°C, depending on the pressure
   Google Scholar
• Paranjpe , Singh and Harish . 373 op. cit.
   Google Scholar