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Management options for Fukushima corium
Progress in Nuclear Energy xxx (2015) 1e7
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Management options for Fukushima corium Susie M.L. Hardie a, *, Ian G. McKinley a, Steve Lomperski b, Hideki Kawamura c, Tara M. Beattie d €fernstrasse 11, 5405 Baden-Da €ttwil, Switzerland MCM Consulting, Ta Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439, USA c Obayashi Corporation, Nuclear Facilities Division, Inter City B, Konan 2-15-2, Minato-ku, Tokyo 108-8502, Japan d MCM Consulting, Orchard Street Business Centre, 13-14 Orchard Street, Bristol BS1 5EH, UK a
b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 17 February 2015 Received in revised form 13 July 2015 Accepted 24 July 2015 Available online xxx
The loss of core cooling for units 1e3 during the accident at Fukushima Dai-ichi caused major fuel damage. Although full details are not yet available, fuel melting produced corium within the reactor pressure vessels that has, to an unknown degree, melted through into the primary containment. The present priority is cooling the damaged reactors and managing contaminated water, but planning of longer term decommissioning has already begun. Management of highly damaged fuel and corium will be of primary concern, with the main options being recovery for reprocessing or packaging for direct disposal. Although the latter option may have significant cost advantages, it presents some novel safety challenges for both operational and post-closure phases. Concerns include criticality management and modelling of long-term dissolution of materials having highly variable composition. Further R&D is required to fill knowledge gaps e of which the most sensitive may involve determination of the extent to which small “hot particles” of corium have been produced. © 2015 Elsevier Ltd. All rights reserved.
Keywords: Fukushima Dai-ichi Nuclear accident Core meltdown Corium characterisation Damaged fuel Diagnostics Severe accident codes
1. Introduction The 2011 earthquake off the pacific coast of Japan generated a devastating tsunami that triggered an unprecedented series of reactor severe accidents at the Fukushima Dai-ichi nuclear power plant (denoted here as “1F”). An overview of the progress of this incident and its consequences are described in detail elsewhere (Hatamura et al., 2012; Kurokawa et al., 2012). This paper focuses entirely on planning the decommissioning of the damaged reactors and, in particular, management of the corium produced as a result of core melting in units 1e3. The main technical issues involved are outlined in Fig. 1 although, as discussed below, socio-political and communication issues must also be taken into account before major actions are implemented. Worldwide, there have been a number of accidents involving reactor core damage, most of which had little radiological significance (McKinley et al., 2011). Three-Mile Island (TMI) is probably most relevant for corium management, with defueling completed in 1990 (USNRC, 2009), after which the corium was transported to
* Corresponding author. E-mail address: [email protected] (S.M.L. Hardie).
the Idaho National laboratory where it sits on a concrete plinth awaiting final disposal (IAEA, 1991, 1992; EPRI, 1990, 1992). Other reactors that suffered major core damage were either simply sealed, e.g. Windscale, UK; Chernobyl, Ukraine or decommissioned, with damaged fuel either reprocessed, e.g. Lucens, Switzerland (ENSI, 2012) or stored for eventual direct disposal, e.g. SRE (Sodium Reactor Experiment), USA. In addition, reactor severe accident experiments have been conducted for decades to study a wide range of phenomena. These include fuel rod dryout and degradation, e.g. Steinbruck et al., 2010; Toth et al., 2010, in-vessel (RPV) retention and cooling of corium (Bechta et al., 2001; Kang et al., 2006), vapour explosions (Kim et al., 2010; Magallon and Huhtiniemi 2001) exvessel corium spreading (Cognet et al., 2001; Journeau et al., 2003) and corium/concrete interactions and coolability (Journeau et al., 2009; Lomperski and Farmer, 2007). This paper considers management of the corium waste rather than accident phenomena and progression. Loss of instrumentation and hydrogen explosions have obscured the extent of core damage. Even now, high radiation fields and contamination limit our ability to inspect and characterise the reactor cores and corium debris. Severe accident codes, e.g. MELCOR and MAAP, have used available data to produce the core
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Fig. 1. Technical issues to be considered when developing a management plan for 1F corium.
damage estimates shown in Table 1 (JAEA, 2014). The three BWR units contained a total of 1496 fuel assemblies, with 32 of them MOX in unit 3. Each assembly has 60 zirconium alloy-clad fuel rods. The fuel burnup histories vary between reactors, the original location of fuel assemblies in the core and location along the fuel rods (due to differing fuel loading patterns). The 1F reactors discharge fuel after 4 cycles with a burnup of 39.5 MW d kg1 (GW d/tHM) at which point, the remaining three quarters of the load would have burnups of about 10, 20 and 30 MW d kg1 (NEI, 2012). Whilst there is yet no direct evidence that corium reached the primary containment of any of the three reactor units, as illustrated in Fig. 2 (TEPCO, 2014), accident progression simulations generated by reactor severe accident codes clearly indicate corium breach for at least unit one (Yamanaka et al., 2014; Gauntt et al., 2012). The extent to which melt through has occurred is, however, unknown and probably varies significantly between reactors. In addition, there is considerable uncertainty in what little data is available, for example a recent press release has suggested possibly more melt through in unit 3 than had been previously reported (TEPCO, 2014). More recently, preliminary data from a cosmic-ray muon radiography installation at unit one suggests that most or all of the core has melted and relocated (IRID, 2015). Though this measurement technique has low spatial resolution, it can remotely map the disposition of reactor internals using the density difference between reactor fuel and structural materials (Miyadera et al., 2013; Takamatsu et al., 2015). 2. Characterisation of 1F corium Corium is a somewhat vaguely defined term applied to the mixture of nuclear fuel and structural materials produced during a reactor core melt accident (EPRI, 2014). Its composition depends on the original type of fuel (UO2 or MOX, in this case), burnup, the design and materials in the fuel assembly, the temperature profile of the incident, and the extent to which molten fuel reacts with
Table 1 Characteristics of units 1e3 at 1F. Unit
Power (MW)
Fuel load (tHM)
No. of assemblies
% melt (JAEA)
% melt (TEPCO)
1 2 3
460 784 784
77 107 107
400 548 548
100 70 64
100 57 63
other materials. Before fuel melting, cladding cracks at about 1200 C, its oxidation begins at about 1300 C (releasing hydrogen from steam). The zirconium cladding melts at about 1850 C and reacts with uranium oxide to form a molten eutectic, which would release volatile fission products such as iodine and caesium. However, any bulk UO2 not in contact with zircaloy will begin to melt at about 2800 C. In the case of TMI, molten fuel interactions were restricted to core components (control rods, fuel assembly, instrumentation, etc.) and the inner wall of the pressure vessel. The material in the case of 1F is likely to be much more complex due to melt through and reaction with the concrete base of the primary containment (Fig. 2). Such liquid fuel/concrete reaction is exoenergetic and would result in a complex range of solid products, further complicated by quenching reactions when the core and containment were flooded, initially with sea water. Corium is inherently heterogeneous and will contain varying quantities of uranium and plutonium along with activation and fission products. Physical forms would include metallic phases, mixed oxides, and aluminosilicates, chlorides and carbonates from reactions with concrete and sea water. As our focus here is on the management of corium and coriumcontaminated materials, we define terminology as the following: 1. “corium” is the main bulk or mass of melted core material that has interacted with other materials such as concrete or steel 2. “corium-contaminated materials” are structure surfaces such as the RPV that have been coated or spattered with corium 3. “fine particles” of fuel debris produced during exoenergetic reactions, which may not be confined solely to surfaces in the RPV or primary containment but may be mobile and transported significant distances (e.g. turbine buildings, water filters, ground water etc.). Fine particles may exist in the form of aerosols (when transported in the gaseous phase), or colloids or suspended particles (when transported in the liquid phase) After solidification, corium properties will change over the years as it interacts with cooling water e initially including seawater or recycled water with relatively high chloride content. Transformation of corium and corium contaminated material may be confined to surface layers, but, for thin layers or finely dispersed particulate material, complete alteration may occur with loss of soluble elements and erosion of fine-grained reaction products by water flow.
3. Decommissioning approach There can be advantages in delaying decommissioning to allow decay of shorter-lived radionuclides e.g. 80 years in the case of Windscale (The Engineer, 2011). However the current strategy is to initiate 1F decommissioning as soon as practicable, within the next few decades, and thus planning has already begun. Although the management of corium is only a small component of the required work, it does present some special challenges due to its heterogeneity and potential for localised risks of criticality, flammability and release of highly active fine particles. Planning of decommissioning is inherently linked to waste storage and disposal concepts. The extreme variants for corium contaminated material and corium would be: a) Emplacement in transport casks for shipping to off-site storage and final conditioning/packaging/disposal b) Conditioning, packaging and direct disposal performed on or near site.
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Fig. 2. Schematic overview of 1F core melting where, following melt through the reactor pressure vessel (RPV), liquid fuel has reacted with concrete of the primary containment and caused significant erosion of this barrier, but has not come close to the steel shell of the primary containment vessel (TEPCO, 2014). *This image is for reference purposes, and is not quantitatively accurate in terms of the actual size of fuel debris, etc.
In the former case, decommissioning must extract and reduce large masses for safe loading into transportation containers and manageable handling at the storage site. In the latter, it may be possible to handle large units and dispose of them with minimum conditioning and packaging. There are a number of options between these two extremes e e.g. variants of the Studsvick “rip & ship” concept (Lidar et al., 2013), where large units are transported to a centralised waste management centre.
Option (a) above is being carried out at TMI, but it should be noted that the boundary conditions are somewhat different from those of 1F. In the latter case, there is corium melt through the RPV, major external damage, and contamination throughout the reactor buildings. There is also a need to decommission all 6 units: reactors 1e3 with major core damage, reactor 4 with explosion damage and major contamination (during volatile releases and subsequent flows of cooling water) along with undamaged units 5 and 6. A
Fig. 3. Example of an integrated decommissioning plan for the Reactor Buildings (RB) and Turbine Buildings (TB) for units 1e6, with assumption that TBs 5 & 6 are refit as waste management facilities capable of handling large components at least as large as RPVs.
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Fig. 4. Options for high corium content material: H12 variant (upper) and larger CARE variant (lower).
comprehensive approach to decommissioning would be advantageous from the point of view of worker dose minimisation, environmental impact and cost. It would involve development of local, centralised waste handling, decontamination, conditioning and packaging facilities. One example of such an approach is illustrated in Fig. 3. In Japan, the options for disposal depend on the concentrations of alpha and beta/gamma emitters in the waste: surface/near surface disposal is practiced for wastes with up to 1 PBq kg1 b/g and 1 TBq kg1 a activity, above these levels intermediate depth or deep geological disposal is planned. Based on expected alpha activity levels, deep disposal would be required for most corium and corium-contaminated materials. 4. Corium management options Corium should be reasonably localised and could, in principle, be segmented in-situ. This would be done using tele-operated equipment in air, under water, or after coating with immobilising agents such as resins. The first option may be technically easier, but has greater risks in terms of fire or production of high-alpha dust. In all cases, great care is required to ensure that changes in geometry or the presence of neutron reflectors cannot give rise to criticality excursions. In the absence of better information, a number of inventory scenarios can be developed. A reasonably conservative one assumes that all fuel has melted in each unit and that about 80% of this (240 tHM) has formed corium debris within or below the RPVs. Of this, it is assumed half has higher fuel content (50% by weight,
Fig. 5. Option for disposal of lower corium content waste.
giving a corium mass of 240 t) while the other half has less (20% fuel by weight, corium mass of 600 t). Such corium could, in principle, be reprocessed. Although this would recover fissionable material, the main benefit of reprocessing would be to produce waste forms similar to those already generated in normal reprocessing operations. Nevertheless, due to the complexity and heterogeneity of corium relative to LWR fuel, reprocessing using a conventional approach would be difficult and probably best carried out at a small plant (e.g. the pilot plant at Tokai) rather than the commercial facility at Rokkasho. An alternative might be to integrate this material into an investigation of advanced reprocessing technology like pyroprocessing to garner technological benefit from the necessity of dealing with corium. Packaging of corium for direct disposal would certainly be much simpler, cheaper and involve less risk to operators. The problems here are associated with lack of knowledge on the long-term behaviour of this heterogeneous material. Nevertheless, extensive national and international experience in developing concepts for geological disposal of spent fuel, high-level vitrified waste from reprocessing, and higher-toxicity intermediate-level waste (termed TRU waste in Japan) forms a basis for confidence that this can also be achieved for corium. Indeed, the fact that the most problematic volatile “instant release fraction” (IRF) or soluble radionuclides will have been mainly lost from the corium should, in principle, ease development of a safety case. For the special case of corium contained within the RPVs, an option worth considering is the removal of the entire RPV as a unit for direct disposal e effectively equivalent to concepts being developed for handling large components during decommissioning (e.g. NEA, 2012). This is particularly appropriate if on-site disposal of these units was possible (IAEA, 1999). If corium is segmented in a more conventional decommissioning approach, it may be packaged for disposal in the type of “H12” steel overpacks already developed in Japan for HLW or, to improve flexibility to manage larger blocks, massive steel casks of the type used in the CARE concept (Fig. 4: more background on variants in NUMO, 2004). Perhaps it should be noted here that although Japan is a particularly seismically active country, a cavern on-site, like any other geological disposal facility, will be sited below 300 m. Seismic movement at this depth is a fraction of what it is on the surface, where large ground motions can occur. Closed geological repositories are thus effectively immune to earthquakes and, as the 1F site is east of the volcanic front, is not at risk from igneous intrusions. Potential perturbations must be considered during the operational phase; however, a repository can be designed to be resilient against such natural hazards (Umeki et al., 2015). The former would be emplaced in tunnels and the latter in larger caverns. In both cases, the amount of corium within a package would be specified to avoid criticality concerns, with risks further reduced through boronated cement mortar or lowtemperature boronated glass to fill voids. For these options, if it is assumed that all corium is loaded with 1 t per package for H12 overpacks and 10 t per package for CARE casks, the material would fill either 840 overpacks or 84 casks. Structures contaminated with corium fines may be less localised and found throughout the RPV and lower parts of the primary containment. In-situ removal of surface corium may be very difficult. Direct sectioning would thus result in large volumes of material with potentially high alpha content, requiring packaging for geological disposal. However, if large surface-contaminated units are taken to a central facility, it would be possible to locate and remove high alpha surface layers. This could reduce volume of high-alpha waste even if some remaining material is still classified as radioactive waste. Disposal of lower corium content material could be based on the packages and disposal systems developed for
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Fig. 6. Corium management options.
TRU waste (JAEA/FEPC, 2007). As indicated in Fig. 5, waste is packaged into large concrete boxes, which could be packed with boronated cement mortar. The large concrete boxes are emplaced in a large cavern. If about 20% of the original fuel (60 tHM) is dispersed with other material with an average content of 5% by weight, this would amount to 1200 t of waste. A loading of 1 t per package would generate 1200 packages. A particularly difficult challenge is associated with hot particles e even if they correspond to only a small fraction of the total melted fuel inventory (<<1%). These could be potentially distributed throughout the reactor and turbine buildings, the cooling water treatment/water storage system, and perhaps even the groundwater flow system. No direct measurements of such particles have been made to date and so location and quantity are both uncertain, but it is prudent to assume they exist. Because of the extremely high beta/gamma radiation fields in contaminated zones, the
presence of such particles would be much harder to detect than in other environments with low background radiation levels (e.g. Dennis et al., 2007). Although the integral inventory of activity in this form may be small, their mobility and extremely high radiotoxicity calls for careful management. 5. R&D requirements The planning of 1F decommissioning is still at a very early stage and a wide range of options are open for corium management (Fig. 6). A better understanding of the technical issues involved is necessary to better define the best path to final waste disposal. A starting point is better definition of the corium inventory, requiring development of localisation/characterisation technology that can be applied in-situ e underwater, in confined spaces and high radiation fields. This may require developing exotic techniques, e.g. muon tomography for localising massive corium (Bacon
Fig. 7. Integrated on-site management option (from McKinley et al., 2013).
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et al., 2014), Laser Induced Breakdown Spectroscopy (LIBS) for characterisation (Saekia et al., 2014), and testing by conventional sampling and laboratory analysis. Possibly more critical in terms of developing an overall management plan is determination of the inventory and distribution of hot particles. It is not currently clear how this would be done but, as a starting point, sampling and analysis of filters from the water treatment plant would be advisable. To form a basis for the assessment of on- and off-site management variants, an evaluation of the technology for handling large radioactive packages is required, along with a cost/benefit analysis from the perspective of decommissioning all 6 reactors. Due to the present emphasis on reprocessing spent fuel in Japan, the option of reprocessing damaged fuel/corium will certainly be considered and hence an assessment of reprocessing technology for corium is needed. This would complement development of a more detailed concept and safety case for direct disposal options, which may require a programme of experimental studies to produce the required database for quantitative modelling of the long-term performance of geological disposal systems (possibly for both on and off-site implementation options). Finally, it should be emphasised that assessment of management options requires not only technical input, but additionally has socio-political dimensions that require efforts to communicate the essence of decisions to be made, to all stakeholders. This should be integrated with a long-term effort to build acceptance in all communities that will host any of the required facilities. 6. Conclusions and a look to the future Decommissioning of 1F will involve many challenges and the management of corium will certainly be one of the more difficult ones. Reactor dismantling and waste handling will be costly and require development of new technologies as well as establishment of waste treatment and disposal guidelines appropriate to accident conditions rather than adopting over-prescriptive regulations for other sources of waste. However, effective planning and coordination can considerably reduce costs, environmental impact and risks to workers. Indeed, given that decommissioning of older reactors will be a major industry in the coming decades, technological spinoffs may be valuable. A much more politically sensitive area where optimisation is possible is waste disposal. From a purely technical point of view, there are considerable benefits if all wastes are disposed of on-site (or in a facility that is accessed directly from the site e Fig. 7). As discussed further elsewhere (McKinley et al., 2013), the 1F site could even be developed as a national waste management facility e which would be compatible with the recent decision to set up a local centre of excellence in decommissioning technology. It should be emphasised that this is not analogous to the situation in Chernobyl e where the reactor site is considered as a national waste disposal facility. There is no concept for remediation of the Chernobyl exclusion zone whereas decontamination of evacuated areas around 1F is well underway. A decommissioning and disposal centre based at the 1F site would be a basis for revitalisation of the local economy and repopulation of the region devastated by the tsunami. Acknowledgements We would like to thank Dr. Dave McGinnes (AXPO) for information regarding burn-up history of 1F LWRs, Dr. Lake Barrett (L. Barrett Consulting) for information on corium management at TMI and Dr. Hironori Ohba (JAEA) for information on LIBS.
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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Management options for Fukushima corium Susie M.L. Hardie a, *, Ian G. McKinley a, Steve Lomperski b, Hideki Kawamura c, Tara M. Beattie d €fernstrasse 11, 5405 Baden-Da €ttwil, Switzerland MCM Consulting, Ta Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439, USA c Obayashi Corporation, Nuclear Facilities Division, Inter City B, Konan 2-15-2, Minato-ku, Tokyo 108-8502, Japan d MCM Consulting, Orchard Street Business Centre, 13-14 Orchard Street, Bristol BS1 5EH, UK a
b
a r t i c l e i n f o
a b s t r a c t
Article history: Received 17 February 2015 Received in revised form 13 July 2015 Accepted 24 July 2015 Available online xxx
The loss of core cooling for units 1e3 during the accident at Fukushima Dai-ichi caused major fuel damage. Although full details are not yet available, fuel melting produced corium within the reactor pressure vessels that has, to an unknown degree, melted through into the primary containment. The present priority is cooling the damaged reactors and managing contaminated water, but planning of longer term decommissioning has already begun. Management of highly damaged fuel and corium will be of primary concern, with the main options being recovery for reprocessing or packaging for direct disposal. Although the latter option may have significant cost advantages, it presents some novel safety challenges for both operational and post-closure phases. Concerns include criticality management and modelling of long-term dissolution of materials having highly variable composition. Further R&D is required to fill knowledge gaps e of which the most sensitive may involve determination of the extent to which small “hot particles” of corium have been produced. © 2015 Elsevier Ltd. All rights reserved.
Keywords: Fukushima Dai-ichi Nuclear accident Core meltdown Corium characterisation Damaged fuel Diagnostics Severe accident codes
1. Introduction The 2011 earthquake off the pacific coast of Japan generated a devastating tsunami that triggered an unprecedented series of reactor severe accidents at the Fukushima Dai-ichi nuclear power plant (denoted here as “1F”). An overview of the progress of this incident and its consequences are described in detail elsewhere (Hatamura et al., 2012; Kurokawa et al., 2012). This paper focuses entirely on planning the decommissioning of the damaged reactors and, in particular, management of the corium produced as a result of core melting in units 1e3. The main technical issues involved are outlined in Fig. 1 although, as discussed below, socio-political and communication issues must also be taken into account before major actions are implemented. Worldwide, there have been a number of accidents involving reactor core damage, most of which had little radiological significance (McKinley et al., 2011). Three-Mile Island (TMI) is probably most relevant for corium management, with defueling completed in 1990 (USNRC, 2009), after which the corium was transported to
* Corresponding author. E-mail address: [email protected] (S.M.L. Hardie).
the Idaho National laboratory where it sits on a concrete plinth awaiting final disposal (IAEA, 1991, 1992; EPRI, 1990, 1992). Other reactors that suffered major core damage were either simply sealed, e.g. Windscale, UK; Chernobyl, Ukraine or decommissioned, with damaged fuel either reprocessed, e.g. Lucens, Switzerland (ENSI, 2012) or stored for eventual direct disposal, e.g. SRE (Sodium Reactor Experiment), USA. In addition, reactor severe accident experiments have been conducted for decades to study a wide range of phenomena. These include fuel rod dryout and degradation, e.g. Steinbruck et al., 2010; Toth et al., 2010, in-vessel (RPV) retention and cooling of corium (Bechta et al., 2001; Kang et al., 2006), vapour explosions (Kim et al., 2010; Magallon and Huhtiniemi 2001) exvessel corium spreading (Cognet et al., 2001; Journeau et al., 2003) and corium/concrete interactions and coolability (Journeau et al., 2009; Lomperski and Farmer, 2007). This paper considers management of the corium waste rather than accident phenomena and progression. Loss of instrumentation and hydrogen explosions have obscured the extent of core damage. Even now, high radiation fields and contamination limit our ability to inspect and characterise the reactor cores and corium debris. Severe accident codes, e.g. MELCOR and MAAP, have used available data to produce the core
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Fig. 1. Technical issues to be considered when developing a management plan for 1F corium.
damage estimates shown in Table 1 (JAEA, 2014). The three BWR units contained a total of 1496 fuel assemblies, with 32 of them MOX in unit 3. Each assembly has 60 zirconium alloy-clad fuel rods. The fuel burnup histories vary between reactors, the original location of fuel assemblies in the core and location along the fuel rods (due to differing fuel loading patterns). The 1F reactors discharge fuel after 4 cycles with a burnup of 39.5 MW d kg1 (GW d/tHM) at which point, the remaining three quarters of the load would have burnups of about 10, 20 and 30 MW d kg1 (NEI, 2012). Whilst there is yet no direct evidence that corium reached the primary containment of any of the three reactor units, as illustrated in Fig. 2 (TEPCO, 2014), accident progression simulations generated by reactor severe accident codes clearly indicate corium breach for at least unit one (Yamanaka et al., 2014; Gauntt et al., 2012). The extent to which melt through has occurred is, however, unknown and probably varies significantly between reactors. In addition, there is considerable uncertainty in what little data is available, for example a recent press release has suggested possibly more melt through in unit 3 than had been previously reported (TEPCO, 2014). More recently, preliminary data from a cosmic-ray muon radiography installation at unit one suggests that most or all of the core has melted and relocated (IRID, 2015). Though this measurement technique has low spatial resolution, it can remotely map the disposition of reactor internals using the density difference between reactor fuel and structural materials (Miyadera et al., 2013; Takamatsu et al., 2015). 2. Characterisation of 1F corium Corium is a somewhat vaguely defined term applied to the mixture of nuclear fuel and structural materials produced during a reactor core melt accident (EPRI, 2014). Its composition depends on the original type of fuel (UO2 or MOX, in this case), burnup, the design and materials in the fuel assembly, the temperature profile of the incident, and the extent to which molten fuel reacts with
Table 1 Characteristics of units 1e3 at 1F. Unit
Power (MW)
Fuel load (tHM)
No. of assemblies
% melt (JAEA)
% melt (TEPCO)
1 2 3
460 784 784
77 107 107
400 548 548
100 70 64
100 57 63
other materials. Before fuel melting, cladding cracks at about 1200 C, its oxidation begins at about 1300 C (releasing hydrogen from steam). The zirconium cladding melts at about 1850 C and reacts with uranium oxide to form a molten eutectic, which would release volatile fission products such as iodine and caesium. However, any bulk UO2 not in contact with zircaloy will begin to melt at about 2800 C. In the case of TMI, molten fuel interactions were restricted to core components (control rods, fuel assembly, instrumentation, etc.) and the inner wall of the pressure vessel. The material in the case of 1F is likely to be much more complex due to melt through and reaction with the concrete base of the primary containment (Fig. 2). Such liquid fuel/concrete reaction is exoenergetic and would result in a complex range of solid products, further complicated by quenching reactions when the core and containment were flooded, initially with sea water. Corium is inherently heterogeneous and will contain varying quantities of uranium and plutonium along with activation and fission products. Physical forms would include metallic phases, mixed oxides, and aluminosilicates, chlorides and carbonates from reactions with concrete and sea water. As our focus here is on the management of corium and coriumcontaminated materials, we define terminology as the following: 1. “corium” is the main bulk or mass of melted core material that has interacted with other materials such as concrete or steel 2. “corium-contaminated materials” are structure surfaces such as the RPV that have been coated or spattered with corium 3. “fine particles” of fuel debris produced during exoenergetic reactions, which may not be confined solely to surfaces in the RPV or primary containment but may be mobile and transported significant distances (e.g. turbine buildings, water filters, ground water etc.). Fine particles may exist in the form of aerosols (when transported in the gaseous phase), or colloids or suspended particles (when transported in the liquid phase) After solidification, corium properties will change over the years as it interacts with cooling water e initially including seawater or recycled water with relatively high chloride content. Transformation of corium and corium contaminated material may be confined to surface layers, but, for thin layers or finely dispersed particulate material, complete alteration may occur with loss of soluble elements and erosion of fine-grained reaction products by water flow.
3. Decommissioning approach There can be advantages in delaying decommissioning to allow decay of shorter-lived radionuclides e.g. 80 years in the case of Windscale (The Engineer, 2011). However the current strategy is to initiate 1F decommissioning as soon as practicable, within the next few decades, and thus planning has already begun. Although the management of corium is only a small component of the required work, it does present some special challenges due to its heterogeneity and potential for localised risks of criticality, flammability and release of highly active fine particles. Planning of decommissioning is inherently linked to waste storage and disposal concepts. The extreme variants for corium contaminated material and corium would be: a) Emplacement in transport casks for shipping to off-site storage and final conditioning/packaging/disposal b) Conditioning, packaging and direct disposal performed on or near site.
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Fig. 2. Schematic overview of 1F core melting where, following melt through the reactor pressure vessel (RPV), liquid fuel has reacted with concrete of the primary containment and caused significant erosion of this barrier, but has not come close to the steel shell of the primary containment vessel (TEPCO, 2014). *This image is for reference purposes, and is not quantitatively accurate in terms of the actual size of fuel debris, etc.
In the former case, decommissioning must extract and reduce large masses for safe loading into transportation containers and manageable handling at the storage site. In the latter, it may be possible to handle large units and dispose of them with minimum conditioning and packaging. There are a number of options between these two extremes e e.g. variants of the Studsvick “rip & ship” concept (Lidar et al., 2013), where large units are transported to a centralised waste management centre.
Option (a) above is being carried out at TMI, but it should be noted that the boundary conditions are somewhat different from those of 1F. In the latter case, there is corium melt through the RPV, major external damage, and contamination throughout the reactor buildings. There is also a need to decommission all 6 units: reactors 1e3 with major core damage, reactor 4 with explosion damage and major contamination (during volatile releases and subsequent flows of cooling water) along with undamaged units 5 and 6. A
Fig. 3. Example of an integrated decommissioning plan for the Reactor Buildings (RB) and Turbine Buildings (TB) for units 1e6, with assumption that TBs 5 & 6 are refit as waste management facilities capable of handling large components at least as large as RPVs.
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Fig. 4. Options for high corium content material: H12 variant (upper) and larger CARE variant (lower).
comprehensive approach to decommissioning would be advantageous from the point of view of worker dose minimisation, environmental impact and cost. It would involve development of local, centralised waste handling, decontamination, conditioning and packaging facilities. One example of such an approach is illustrated in Fig. 3. In Japan, the options for disposal depend on the concentrations of alpha and beta/gamma emitters in the waste: surface/near surface disposal is practiced for wastes with up to 1 PBq kg1 b/g and 1 TBq kg1 a activity, above these levels intermediate depth or deep geological disposal is planned. Based on expected alpha activity levels, deep disposal would be required for most corium and corium-contaminated materials. 4. Corium management options Corium should be reasonably localised and could, in principle, be segmented in-situ. This would be done using tele-operated equipment in air, under water, or after coating with immobilising agents such as resins. The first option may be technically easier, but has greater risks in terms of fire or production of high-alpha dust. In all cases, great care is required to ensure that changes in geometry or the presence of neutron reflectors cannot give rise to criticality excursions. In the absence of better information, a number of inventory scenarios can be developed. A reasonably conservative one assumes that all fuel has melted in each unit and that about 80% of this (240 tHM) has formed corium debris within or below the RPVs. Of this, it is assumed half has higher fuel content (50% by weight,
Fig. 5. Option for disposal of lower corium content waste.
giving a corium mass of 240 t) while the other half has less (20% fuel by weight, corium mass of 600 t). Such corium could, in principle, be reprocessed. Although this would recover fissionable material, the main benefit of reprocessing would be to produce waste forms similar to those already generated in normal reprocessing operations. Nevertheless, due to the complexity and heterogeneity of corium relative to LWR fuel, reprocessing using a conventional approach would be difficult and probably best carried out at a small plant (e.g. the pilot plant at Tokai) rather than the commercial facility at Rokkasho. An alternative might be to integrate this material into an investigation of advanced reprocessing technology like pyroprocessing to garner technological benefit from the necessity of dealing with corium. Packaging of corium for direct disposal would certainly be much simpler, cheaper and involve less risk to operators. The problems here are associated with lack of knowledge on the long-term behaviour of this heterogeneous material. Nevertheless, extensive national and international experience in developing concepts for geological disposal of spent fuel, high-level vitrified waste from reprocessing, and higher-toxicity intermediate-level waste (termed TRU waste in Japan) forms a basis for confidence that this can also be achieved for corium. Indeed, the fact that the most problematic volatile “instant release fraction” (IRF) or soluble radionuclides will have been mainly lost from the corium should, in principle, ease development of a safety case. For the special case of corium contained within the RPVs, an option worth considering is the removal of the entire RPV as a unit for direct disposal e effectively equivalent to concepts being developed for handling large components during decommissioning (e.g. NEA, 2012). This is particularly appropriate if on-site disposal of these units was possible (IAEA, 1999). If corium is segmented in a more conventional decommissioning approach, it may be packaged for disposal in the type of “H12” steel overpacks already developed in Japan for HLW or, to improve flexibility to manage larger blocks, massive steel casks of the type used in the CARE concept (Fig. 4: more background on variants in NUMO, 2004). Perhaps it should be noted here that although Japan is a particularly seismically active country, a cavern on-site, like any other geological disposal facility, will be sited below 300 m. Seismic movement at this depth is a fraction of what it is on the surface, where large ground motions can occur. Closed geological repositories are thus effectively immune to earthquakes and, as the 1F site is east of the volcanic front, is not at risk from igneous intrusions. Potential perturbations must be considered during the operational phase; however, a repository can be designed to be resilient against such natural hazards (Umeki et al., 2015). The former would be emplaced in tunnels and the latter in larger caverns. In both cases, the amount of corium within a package would be specified to avoid criticality concerns, with risks further reduced through boronated cement mortar or lowtemperature boronated glass to fill voids. For these options, if it is assumed that all corium is loaded with 1 t per package for H12 overpacks and 10 t per package for CARE casks, the material would fill either 840 overpacks or 84 casks. Structures contaminated with corium fines may be less localised and found throughout the RPV and lower parts of the primary containment. In-situ removal of surface corium may be very difficult. Direct sectioning would thus result in large volumes of material with potentially high alpha content, requiring packaging for geological disposal. However, if large surface-contaminated units are taken to a central facility, it would be possible to locate and remove high alpha surface layers. This could reduce volume of high-alpha waste even if some remaining material is still classified as radioactive waste. Disposal of lower corium content material could be based on the packages and disposal systems developed for
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Fig. 6. Corium management options.
TRU waste (JAEA/FEPC, 2007). As indicated in Fig. 5, waste is packaged into large concrete boxes, which could be packed with boronated cement mortar. The large concrete boxes are emplaced in a large cavern. If about 20% of the original fuel (60 tHM) is dispersed with other material with an average content of 5% by weight, this would amount to 1200 t of waste. A loading of 1 t per package would generate 1200 packages. A particularly difficult challenge is associated with hot particles e even if they correspond to only a small fraction of the total melted fuel inventory (<<1%). These could be potentially distributed throughout the reactor and turbine buildings, the cooling water treatment/water storage system, and perhaps even the groundwater flow system. No direct measurements of such particles have been made to date and so location and quantity are both uncertain, but it is prudent to assume they exist. Because of the extremely high beta/gamma radiation fields in contaminated zones, the
presence of such particles would be much harder to detect than in other environments with low background radiation levels (e.g. Dennis et al., 2007). Although the integral inventory of activity in this form may be small, their mobility and extremely high radiotoxicity calls for careful management. 5. R&D requirements The planning of 1F decommissioning is still at a very early stage and a wide range of options are open for corium management (Fig. 6). A better understanding of the technical issues involved is necessary to better define the best path to final waste disposal. A starting point is better definition of the corium inventory, requiring development of localisation/characterisation technology that can be applied in-situ e underwater, in confined spaces and high radiation fields. This may require developing exotic techniques, e.g. muon tomography for localising massive corium (Bacon
Fig. 7. Integrated on-site management option (from McKinley et al., 2013).
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et al., 2014), Laser Induced Breakdown Spectroscopy (LIBS) for characterisation (Saekia et al., 2014), and testing by conventional sampling and laboratory analysis. Possibly more critical in terms of developing an overall management plan is determination of the inventory and distribution of hot particles. It is not currently clear how this would be done but, as a starting point, sampling and analysis of filters from the water treatment plant would be advisable. To form a basis for the assessment of on- and off-site management variants, an evaluation of the technology for handling large radioactive packages is required, along with a cost/benefit analysis from the perspective of decommissioning all 6 reactors. Due to the present emphasis on reprocessing spent fuel in Japan, the option of reprocessing damaged fuel/corium will certainly be considered and hence an assessment of reprocessing technology for corium is needed. This would complement development of a more detailed concept and safety case for direct disposal options, which may require a programme of experimental studies to produce the required database for quantitative modelling of the long-term performance of geological disposal systems (possibly for both on and off-site implementation options). Finally, it should be emphasised that assessment of management options requires not only technical input, but additionally has socio-political dimensions that require efforts to communicate the essence of decisions to be made, to all stakeholders. This should be integrated with a long-term effort to build acceptance in all communities that will host any of the required facilities. 6. Conclusions and a look to the future Decommissioning of 1F will involve many challenges and the management of corium will certainly be one of the more difficult ones. Reactor dismantling and waste handling will be costly and require development of new technologies as well as establishment of waste treatment and disposal guidelines appropriate to accident conditions rather than adopting over-prescriptive regulations for other sources of waste. However, effective planning and coordination can considerably reduce costs, environmental impact and risks to workers. Indeed, given that decommissioning of older reactors will be a major industry in the coming decades, technological spinoffs may be valuable. A much more politically sensitive area where optimisation is possible is waste disposal. From a purely technical point of view, there are considerable benefits if all wastes are disposed of on-site (or in a facility that is accessed directly from the site e Fig. 7). As discussed further elsewhere (McKinley et al., 2013), the 1F site could even be developed as a national waste management facility e which would be compatible with the recent decision to set up a local centre of excellence in decommissioning technology. It should be emphasised that this is not analogous to the situation in Chernobyl e where the reactor site is considered as a national waste disposal facility. There is no concept for remediation of the Chernobyl exclusion zone whereas decontamination of evacuated areas around 1F is well underway. A decommissioning and disposal centre based at the 1F site would be a basis for revitalisation of the local economy and repopulation of the region devastated by the tsunami. Acknowledgements We would like to thank Dr. Dave McGinnes (AXPO) for information regarding burn-up history of 1F LWRs, Dr. Lake Barrett (L. Barrett Consulting) for information on corium management at TMI and Dr. Hironori Ohba (JAEA) for information on LIBS.
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Please cite this article in press as: Hardie, S.M.L., et al., Management options for Fukushima corium, Progress in Nuclear Energy (2015), http:// dx.doi.org/10.1016/j.pnucene.2015.07.017