Journal of Nuclear Energy Science & Power Generation TechnologyISSN: 2325-9809

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Review Article, J Nucl Ene Sci Power Generat Technol S Vol: 0 Issue: 1

Review of Characteristics of Post-Accident Waste Generated in Fukushima Daiichi Nuclear Power Plant Site and Issues to be Addressed in Processing and Disposal Stages

Masaki Tsukamoto1*, Daisuke Sugiyama2, Takeshi Yamamoto3, Motoi Kawanishi3 and Noriyuki Saito4
1Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, Tokyo, Japan
2Radiation Research Center, Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, Tokyo, Japan
3Civil Engineering Research Laboratory, Central Research Institute of Electric Power Industry, Chiba, Japan
4Tokyo Electric Power Company, Tokyo, Japan
Corresponding author : Masaki Tsukamoto
Nuclear Technology Research Laboratory, Central Research Institute of Electric Power Industry, 2-11-1 Iwadokita, Komae, Tokyo 201-8511, Japan
Tel: +81 3 3480 2111; Fax: +81 3 3480 3113
E-mail: [email protected]
Received: September 26, 2013 Accepted: November 24, 2013 Published: November 31, 2013
Citation: Tsukamoto M, Sugiyama D, Yamamoto T, Kawanishi M, Saito N (2013) Review of Characteristics of Post-Accident Waste Generated in Fukushima Daiichi Nuclear Power Plant Site and Issues to be Addressed in Processing and Disposal Stages. J Nucl Ene Sci Power Generat Technol S1. doi:10.4172/2325-9809.S1-002

Abstract

Review of Characteristics of Post-Accident Waste Generated in Fukushima Daiichi Nuclear Power Plant Site and Issues to be Addressed in Processing and Disposal Stages

The accident at Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Station has produced and will continue to produce various types and large amounts of waste contaminated by radionuclides. The literature and published internet information on possible types of waste produced are reviewed from the viewpoint of their characteristics. Issues associated with the waste were selected and analyzed for each stage of the future waste management, considering the characteristics and properties of the waste obtained so far. The stages considered are current (temporary) storage, processing including decontamination and solidification/packaging, storage up to disposal, transportation, and disposal.

Keywords: Fukushima daiichi nuclear power station; Accident; Waste; Radionuclide; Waste processing; Disposal

Keywords

Fukushima daiichi nuclear power station; Accident; Waste; Radionuclide; Waste processing; Disposal

Introduction

The accident at Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Station has produced and will continue to produce various types and large amounts of waste inside the station site (hereinafter referred to as “1F”) as well as off-site [1]. Such waste, which includes blocks of reinforced concrete and spent adsorbent (zeolite) columns, are contaminated by radionuclides that were dispersed by the explosions of hydrogen during the accident and/ or are contained in the highly contaminated water circulated to cool the reactor cores. Therefore, the main characteristic of the waste is its contamination by the radionuclides contained in burned nuclear fuel. In addition, most of the waste contains sodium chloride, owing to the injection of seawater shortly carried out after the accident, and boron other substances that may affect the performance of disposal systems. It is clear that the 1F waste has rather different characteristics and inventories from normal waste produced by the operation and decommissioning of nuclear power plants.
Basic investigations concerning the treatment of the secondary waste from the contaminated water facilities, i.e., zeolite in the columns that have adsorbed mainly radioactive cesium and the radioactive sludge, have already started. However, technical studies to establish suitable way of proper processing and disposing of all types of 1F waste are at in the initial stage. Full restoration of the site of the accident will require the safe disposal of the 1F waste in addition to the current attempts to treat the highly contaminated water, remove the fuel and fuel debris from the reactor cores, and so forth. However, it will be difficult to demonstrate the safe disposal of current and future 1F waste, in which the types and amounts of radionuclides contained are unknown.
On the other hand, the Agency for Natural Resources and Energy compiled and published a revised version of the “Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Station Units 1-4” in June 2013 [2]. This version states that “R&D for characterizing and analyzing waste properties will be promoted to explore processing and disposal methods” [3]. This review provides the first basic technical information to be used not only to realize the above aims but also to develop research and development programs sufficiently detailed enough to clarify the important basic studies that must be carried out.

Procedure to Survey Issues

It is possible to categorize the 1F waste using information on the decommissioning of normal nuclear power plants, materials used in current treatments to remove radioactivity and to prevent the explosion of contaminated materials, and so forth, although these are many types of 1F wastes with a wide range of properties. On the other hand, the types and concentrations of radionuclides in the 1F waste will remain unknown until sufficient data has been accumulated by radioactivity measurement and/or estimation. The amount of waste cannot be easily estimated because some of the waste, such as spent zeolite, sludge, and containers/tanks, will be produced in the future. Under the above limitation, the following four approaches are important and desirable in surveying issues that must be considered when developing processing and disposal technologies:
• Finding possible factors affecting the performance of waste disposal system.
• Analyzing conditions under which the waste is stabilized (contained/solidified).
• Finding conditions of safe waste disposal, i.e., by parametric safety analysis.
• Summarizing the issues by evaluating their importance, relationship, difficulty of implementation, etc.
Currently, evaluation by the above approaches may lead to results with a large uncertainty since it is difficult to estimate the amount of each type of waste and its inventory with high accuracy. Thus, the issues are surveyed as follows in this review:
• Categorization of each type waste considering its origin and properties.
• Collection of published information on the categorized waste.
• Consideration of processes required from the current storage to disposal.
• Extraction of issues associated with each process and surveying of possible techniques to resolve them.
• Surveying of useful information obtained from ordinary waste management and previous accidents around the world to support the above.
A parametric safety analysis was carried out to extract the important radionuclides under various geological conditions and the performance of engineered barrier systems. Details of the parametric safety analysis will be published elsewhere.

Categorization of the 1F Waste

Table 1 summarizes the main types of 1F waste for which information can be obtained. In the table, the types of waste were categorized considering the place of generation, the main contamination procedure, and when it is generated. The types of radionuclide contained in the waste, and the amount of chemicals possibly affecting the performance of disposal systems, such as sodium chloride, were not considered even though they are important factors in evaluating the safety of waste disposal. This is because it is not yet possible to evaluate their concentration in each type of waste. Thus, it is still difficult to estimate the inventory of each type of waste.
Table 1: Main types of 1F waste.
Dose rate and volume of typical 1F waste are summarized in International Experts’ Symposium on the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Plant Unit 1-4 [1] with pictures of some wastes. It is mentioned that sampling of rubble, felled trees and zeolite in cesium adsorption vessels is challenged [1]. However, actual data of the samples have not been published yet.
The main types of waste are described in the following. Secondary waste from the treatment of the highly contaminated water includes spent zeolite in the columns, sludge, and concentrated treated water with thick sodium chloride concentration by reverse osmosis membrane (RO) and evaporation processes. Such waste contains high concentrations of radioactive cesium and other radionuclides. The safe and stable storage of this waste is currently important issue. Operation of the facilities set up to treat the contaminated water after the RO process (Advanced Liquid Processing System, ALPS) will generate other new types of waste including several types of adsorbent and sludge [4].
Rubble, including reinforced concrete and its fragments, and crushed instruments, etc., has also been generated by the hydrogen explosions and the removal of the upper level of the reactor buildings to treat the spent fuels and the reactors. Deforestation carried out to obtain places to build treatment facilities and tanks has generated contaminated felled (cut down) trees, mainly pine. The rubble and trees have been stored in suitable facilities in an appropriate manner according to their radioactivity [5], and it is assured that such waste will continue to accumulate in the future. In Environmental Dosage Reduction Measure (TEPCO) [5], methods and capacity of storage is tabulated for each storage place with picture views of some of the storage places. The removal of the spent fuel in the pools of the reactor buildings for on-site storage is being considered [6].
Fuel debris is currently kept covered with water while a suitable method of removal is being considered [6]. The fuel debris is scheduled to be removed after around 20-25 year [6]. Part of the activated instruments near the reactors might have melted to form debris, which will be removed during or after the removal of fuel debris.
Soil on the ground has mainly been contaminated through the air, and part of the underground soil has probably been contaminated. The contaminated soil will be taken for treatment/disposal when tasks to restore the ground are carried out.
Secondary waste has also been generated by the treatment of the contaminated seawater in the port in front of the intake cannel. Contaminated seabed soil in front of the intake channel has been covered with suitable materials to prevent it from migration. The seabed soil with the covering materials is not decided to be taken yet.
The spent fuel and fuel assembly will be removed from the spent fuel pools and stored in an appropriate manner until its method of treatment has been decided. Clothes for the workers and other instruments used in on-site restoration are contaminated by both air and highly contaminated water. The parts of these wastes that can be burned will be burned in a new furnace.

Processes Considered to be Needed for 1F Waste Management

The right side of Figure 1 shows a schematic of the general processes being considered to treat the various types of waste. The process required for 1F waste management depends on the type of waste. Thus, possible options for each process were examined on a case by case basis (not shown in the figure). On the left side, the scheme employed to extract issues for each process is illustrated.
Figure 1: Schematic of the procedure followed to extract issues (left) and the processes involved in the treatment of waste (right).
The starting point of the general process is to categorize and determine the properties of the waste. The wastes needs to be suitably classified, and materials with different radioactivity concentrations should be separated before disposal. The on-site reuse and recycling of large amounts of waste with very low radioactivity concentrations should be discussed. However, criteria for classifying waste to be reused and recycled have not yet been discussed and must be determined in the future. Various treatments such as concentration, stabilization, and volume reduction are effective for safe waste disposal. Wastes will probably be stored in current storage facilities for a longer period than expected. Some waste will require another storage process before it is stabilized or solidified for disposal. Issues that may arise in each process have been analyzed through the scheme.

Issues Considered for Main Waste

Issues examined in accordance with the above-mentioned procedure are described for each types of waste, along with their treatment processes and their important characteristics and properties.
Spent zeolite vessels
Properties: Two types of “spent zeolite vessels” have been generated from two treatment facilities; 1) Cesium adsorption apparatus and 2) 2nd Cesium adsorption apparatus in the highly contaminated water treatment system [7]. The main target of the two facilities is the removal of radioactive cesium. One facility employed two other types of adsorption vessels to remove technetium and iodine and the other facility consists of an adsorption vessel and lead shielding. Spent vessels are stored in a storage area after removing the water contained. The vessels have a ventilation system to exhaust hydrogen produced by the radiolysis of water. Oil in the highly contaminated water is removed by prefiltering, although part of the oil dissolved in the water may remain in the spent vessel.
Estimating the inventory of the radioactive cesium, the main radionuclide, and other radionuclides is a major issue. It is difficult to take out the spent zeolite from the vessels for radioactivity measurement because they are closely packed and because of the very high level of radiation. However, radioactivity measurement of the contaminated water at the in-let and out-let of the operated vessels has been attempted. The difficulty in radioactivity measurement resulted in some radionuclides exhibiting a lower concentration than that actually detected, which implies the need for a very conservative inventory due to the rather high detection limits. Efforts to upgrade the measurement are still being continued. Another approach to estimating the cesium concentration distribution in the vessels have been examined in which experimental adsorption isotherms for a zeolite sample [8] and a code simulating the time dependence of the cesium concentration in a vessel using the double-porosity model [9] are combined. Both the experimental and simulation approaches are being examined to estimate the inventory of spent zeolite vessels.
Sodium chloride was contained in the spent vessels used in the early stage of treatment. Boron and organic substances might also be contained, which should be considered in subsequent processes such as solidification and disposal. The effects of decay heat and gaseous radionuclide migration should also be considered in subsequent processes.
Issues associated with main processes: It is scheduled that the spent zeolite vessels will be held in long-term storage, for which the migration of hydrogen and gaseous radionuclides is an issue. The former may possibly be absorbed in hydrogen-absorbing alloys. The latter can be removed using a suitable filter system. Sodium chloride might also cause the corrosion of the vessels. The early termination of long-term storage would require the removal and stabilization of the spent zeolite particles in the vessels. In the process of stabilization, the selection of an appropriate method is important. Two methods are possible: one is drying and encapsulation of the spent zeolite and the other is solidification after drying the spent zeolite. In the former case, the spent vessels themselves may be a possible container for the zeolite waste. The solidification of zeolite has been investigated and several options are possible [10].
Although the removal of the zeolite is an important issue, its storage and transportation after stabilization do not appear to involve major issues. The type of container used in each process is an issue, i.e., whether one container should be used throughout the processes up to disposal or whether several containers should be used, which involves consideration of the properties of the containers used in the processes.
For the safe disposal of zeolite waste, sufficient information to carry out a safety analysis, such as an inventory, the radionuclide concentration, and the amount of affecting substances, is important. In the case of disposing of dried spent zeolite in a vessel, each spent vessel possibly may have radioactive properties, i.e., a concentration and distribution in the vessels. The effective classification of spent vessels to be applied to appropriate disposal systems must be carried out. The disposal of lead shield is another issue. Decay heat will affect the disposal system configuration and the storage period until disposal.
Information from previous accidents: Investigations carried out after the Three Mile Island Unit 2 accident (hereafter “TMI”) provide useful technical information on the treatment of 1F spent zeolite and its vessels. A submerged demineralizer system (SDS) was developed to treat 2,300 m3 of high-level contaminated water at the bottom of the reactor building. Some highly radioacitive SDS vessels were examined to investigate the zeolite performance and monitoring the disposal system [11,12]. Vitrification of the spent zeolite by the in-can method was investigated [13] using three SDS vessels. The other vessels were transported to the Hanford site, where a monitoring system for the vessels was set up. After that, the spent zeolite in some of the vessels was moved to high isolation containers (HICs) in Battelle Columbus Laboratory. Finally, they were disposed of at the Barnwell site. This indicates the possibility of 1F zeolite vitrification and the disposal of spent zeolite in HICs.
The use of iron enriched basalt has also been studied as a solidification method, which was applied to solidify TMI fuel debris [14]. It was demonstrated that iron enriched basalt with a high zeolite content can be synthesized [15] and that the amount of cesium evaporation is small when it is melted [16]. In addition, large-scale solidification using simulated waste has been examined [17]. Note that the effect of sodium chloride contained in the 1F zeolite should be clarified when applying the above experiences of TMI.

Sludge

Properties: Most of the “sludge” was generated by the operation of the decontamination apparatus in the early stages of treating the highly contaminated water, and about 600 m3 of sludge has been accumulated and is in temporary storage. The sampling of the sludge for chemical and radiochemical analyses has not been attempted owing to its very high radioactivity. However, it is expected to contain sodium chloride, boron, organic substances, and materials resulting from their decomposition, which may affect the subsequent treatment and disposal.
The most noteworthy property of the sludge is that it contains ferrocyanide, which has been used for immobilizing cesium. The production of cyan gas and cyanide is also an issue when ferrocyanide decomposes. Ferrocyanide ions are very stable in air and difficult to biologically decompose [18,19]. However, the possibility of decomposition owing to the effects of microbes, high temperature, light, and so forth [20-23] during long-term storage or during solidification treatment should be noted. The decomposition of ferrocyanide will be an issue in any subsequent burning/melting processes. Ferrocyanide ions also affect the chemistry (complexation) and mobility of important radionuclides during the disposal.
Issues associated with main processes: Issues that may arise during long-term storage are similar to those of the spent zeolite vessels, since the sludge has similar properties to the spent zeolite, i.e., decay heat, hydrogen generation, corrosion of the container, high radiation, and the necessity of stabilization technology in the case of the early termination of storage. Ferrocyanide decomposition is added to the above issues, and solidification is recommended to stabilize the waste. Solidification with cementitious materials is a possible stabilization method although the high concentration of sodium chloride and boron may inhibit solidification.
Hydrogen may be generated by the radiolysis of the water contained in solidified waste. Burning and melting are expected to be employed to remove ferrocyanide and cyanide before subsequent treatment. In particular, it is desirable that possible effects on the disposal system performance and safety are minimized. Suitable treatment and stabilization technologies should be found during the storage period. Treatment and stabilization will change the inventory of the waste, although that of the sludge itself is still unknown. Evaluation of the inventory of the waste form used for disposal will be important as well as the amount of impurities remaining in the waste form, which may affect the disposal system performance.
Information from previous accidents: The La Hague reprocessing plant has considerable experience of treating sludge including ferrocyanide used for cesium immobilization [24]. The formation of bitumen from dried sludge is no longer carried out because of the problems of the degradation of bitumen and the generation of hydrogen due to radiation. Instead, cementitious solidification and vitrification of the dried sludge [25] and the use of grouting technology in a container (DRYPAC) [26], which have also been used at the Sellafield reprocessing plant [27], have been examined. The effect of sodium chloride is once again an issue when applying French or British technologies to treat 1F sludge.
Concentrated waste
Properties: “Concentrated waste” is defined as evaporated liquid waste containing a high concentration of sodium chloride generated by evaporating the contaminated water from the on-site sanitization apparatus with a RO method for the treated highly contaminated water by zeolite vessels. About 5,500 m3 of concentrated waste has been produced and is being stored in tanks. The evaporation system is not under operation since it is planned that the water treated with the sanitization apparatus will in future be treated by the ALPS facilities to sufficiently decrease the concentrations of radionuclides remaining in the treated water of the current treatment facilities at an even lower level. There is a possibility for the waste to be treated again by ALPS. Thus, disposal of the waste is assumed to survey issues in this review.
The inventory and the amount of impurities in the concentrated waste are still unknown, although high radioactivity and a high concentration of impurities are assumed owing to the RO and evaporation processes. In particular, radioactive strontium and alpha emitters are a concern. Boron, which is difficult to trap using the treatment systems, may be concentrated in the waste.
Issues associated with main processes: The problem of corrosion of the tanks should be considered owing to the high sodium chloride concentration and radiolysis during storage in the tanks. This problem should be investigated to enable the stabilization of waste by a suitable method as well as the spent zeolite and the sludge. The solidification of concentrated waste is a possible means of preventing the liquid waste from dispersing owing to corrosion of the tanks and of stable storage, although this makes it difficult to remove the radionuclides contained within. The effect of radiation on workers and the corrosion of containers are issues since all the radionuclides in the liquid waste would be concentrated in the solid waste. Magnesium and sulfate ions in the seawater may form salts, possibly affecting the container and disposal system performances. The additional removal of sodium chloride and radionuclides from the concentrated waste may also be a feasible strategy.
Several stabilization technologies including solidification with cementitious materials are possible for concentrated waste and dried sodium chloride containing radionuclides. The effect of boron is also an issue in the solidification of waste with cementitious materials. The release of sodium chloride from waste is expected to affect the disposal system performance since sodium and chloride ions prevent barrier materials from sorbing radionuclides.
Information from previous accidents: There is a lack of information on treating concentrated sodium chloride solution. The use of a ceramic membrane filter (NaSICON) to electrochemically remove sodium chloride from liquid waste has been examined at Hanford [28]. The selective removal of cesium and strontium from a liquid with a high sodium concentration has been applied to a power plant [29] and is still being investigated [30,31].
Waste generated in ALPS facilities
Properties: The ALPS facilities consist of two types of precipitation apparatus and followed by various types of adsorption vessels to remove 62 radionuclides except for tritium to decrease their concentrations via precipitation/co-precipitation processes and the combination of several types of adsorbent [32]. The operation of ALPS will induce the precipitation of iron hydroxide and calcium carbonate containing radionuclides such as alpha emitters, Co-60, Mn-54, and other major elements found in seawater, i.e., Mg and Ca. In addition, spent adsorbents such as activated carbon, synthesized minerals, and chelating resin, as forth will be produced as secondary waste [32]. Ferrocyanide is planned to be used to trap cesium in ALPS.
The precipitates will contain water and be stored in HICs as sludge [4]. The spent adsorbents are also planned to be stored in HICs. HICs containing such waste will be stored until the treatment method for each type of waste for disposal has been clarified. The sodium chloride concentration in the waste should be low because the treated water will have already been processed by the sanitization system with the RO methods. The high water content in the sludge will be an issue in subsequent processes since it is difficult to remove water from the sludge owing to the properties of the precipitates, which are very fine or aggregated hydrates.
Radioactivity will accumulate in each adsorbent vessel in accordance with operation of ALPS. It will be possible to estimate the inventory for each type of waste by sampling and radioactivity analysis and by employing well-controlled system in which radionuclide migration behavior is considered.
Issues associated with main processes: HICs have the issue of sufficient performance during the storage period. The leakage of radionuclides and hydrogen gas from HICs should be examined. The use of other type containers for disposal should be considered, although HICs might be able to be a final form for disposal. One type of adsorbent vessel is planned to be used for storage without the transfer of its contents to HIC. The treatment of the spent adsorbent is another issue. ALPS will produce many types of waste with different properties, which should be suitably treated for safe disposal considering their characteristics and existing information on treatment technologies. In particular, ferrocyanide will contain toxic substances and affect to the disposal system performance when they decompose as well as chelating resin, which should be evaluated in the safety analysis of the disposal.
Rubble
Properties: “Rubble” including reinforced concrete and metallic substances, has been taken from the ground around the reactor buildings and the upper part of the operating floor of some units. Iron rods have been separated from the concrete as much as possible. Concrete blocks, and iron rods and other metallic substances are piled up in storage facilities and classified according to their radiation levels. Their size and radiation levels have large distributions. The radioactivity measurement of concrete specimens is being carried out, although results have not been published yet. Most specimens were found to be contaminated by radionuclides distributed by the hydrogen explosions and contain sodium chloride owing to the injection of seawater as well as organic substances included in the covering reagents used to prevent the radionuclides from dispersing.
They may also contain radionuclides in the highly contaminated water owing to mixing when the water was injected to cool the reactors. Most radionuclides exist on the surface of the concrete rubble, although part of them may have penetrated into the pores of the concrete by diffusion. Large distribution of radionuclide concentration in the concrete rubble makes it difficult to carry out representative sampling to estimate an inventory of the waste in addition to analyzing beta and alpha emitters.
More rubble will accumulate in the future, resulting in a huge amount of such waste. Its classification according to radionuclide concentration is necessary in order to select an appropriate treatment and disposal methods considering the possibility of reuse and recycling. Thus, the most important issue is to estimate an inventory by a suitable method. Information on places where rubble is being generated may help when estimating the inventory, although it is currently difficult to obtain such information. Toxic metals such as lead and aluminum should be separated and suitably treated, even though their amount of is expected to be small, to prevent chemical toxicity and hydrogen generation. The decomposition of materials covering metallic wires such as polyvinyl chloride should also be considered.
Issues associated with main processes
 
During temporary and long-term storage, it is important to find the best way to handle the rubble as part of waste management and to investigate suitable technologies required in the subsequent treatment. The classification of waste requires radioactivity measurement methods, and separation methods suitable for rubble with a large radioactivity distribution are also needed. Decontamination and reuse/recycling are useful strategies for sufficiently decreasing the amount of waste to be disposed of to disposal facilities. Technological issues may arise in the selection of the best decontamination method. The reuse and recycle strategy should include a concrete plan to reuse and recycle the waste. The processes of classification, treatment such as decontamination, reuse/recycling, and disposal should be strategically considered as a single package.
Another issue is that radioactivity (radiation) measurement of the rubble should be carried out on-site, where the background radiation is expected to remain high in the future. Measurements of gamma emitters will require shielding apparatus, and alpha and beta emitters must be measured in facilities. A conveyer belt system, which has been used to separate contaminated soil [33], is a possible method of separating concrete specimens. The measurements and separation must be automatically performed because of the high radioactivity of the rubble. The rubble may have an issue in establishing homogeneous radioactivity in its waste forms to be disposed of when considering drums or concrete/metal boxes as a container considered in existing studies in Japan [34]. Stabilization of the waste, i.e., solidification, and impurities affecting the disposal system performance are also issues. The effect of the radioactive heterogeneity of the waste form on the uncertainty of disposal safety should be evaluated in addition to that of the impurities.
Trench- or pit-type subsurface disposal systems may be suitable for the waste. On the other hand, a monolith-type disposal system [35], in which other types of waste are disposed of, and melting of a large amount of the rubble to obtain radioactive homogeneity, should also be considered for the disposal of concrete and metal specimens with a wide range of shape, sizes and radioactivity contents.
Information from previous accidents: After the Chernobyl accident, waste generated by the decommissioning of the building was stored in a shelter, in facilities built in the site, and in the Chernobyl Exclusion Zone (CEZ). Part of the waste was disposed in trench- and concrete-pit-type disposal facilities that were filled with concrete and covered with sand and clay as an engineered barrier [36] depending on their radioactivity. Many types of waste material were mixed in these facilities. Near the shelter, highly radioactive waste that could not be placed in the above disposal facilities is stored in facilities with a concrete filling.
Felled trees
Properties: Many trees, mainly pine, have been felled to install storage tanks and instruments on-site and are being accumulated in the storage/facilities after separating the branches, leaves, roots, and trunk. These “felled trees” are mainly contaminated with radioactive cesium through the air after the accident. The injection of seawater and the treatment with covering reagents resulted in impurities such as sodium chloride and various organic substances. Off-site investigations have revealed that radioactive cesium is mainly concentrated in the leaves and branches and that the center part of the trees contains little cesium. Additionally felled trees are expected to contain less cesium than before owing to the removal of the leaves and their decontamination.
Issues associated with main processes: During storage, the accumulated trees have decomposed by microbial activities, resulting in heat and gas generation. The volume reduction of trees by burning them in a furnace appears to be a possible method of treating the waste, although the risk of workers being exposed to a high radiation dose during the burning process and from the resulting ashes is an issue. Stabilization of the highly radioactive ashes is another issue.
Decomposition will change the properties of the waste, i.e., substances similar to soil and organic fertilizer will be produced. Mixing soil from the ground and covering sandbags will affect to the burning process of trees. Melting may be a means of stabilizing the ashes. The migration behavior of cesium during the burning and melting processes must be evaluated to ensure the safety of workers and safe disposal as well as the performances of the apparatus used in the processes. On the other hand, composting is a possible method of volume reduction without burning, resulting in the production of organic materials including microbes and insects, which should be suitably treated. The same issues to rubber are possible in the disposal stage.
Information from previous accidents: Investigations of the decontamination of waste after the Chernobyl accident included an investigation of the feasibility of pulp fabrication from waste [37], and the evaluation of dose using a model assuming the use of the ashes obtained by burning contaminated trees as fertilizer [38]. Solidification of the ashes of burned waste generated after the Chernobyl accident was also studied [39]. Ash solidification has been investigated in the field of management of radioactive waste produced in researches [40] and off-site waste.
Fuel debris
Properties: “Fuel debris” is planned to be removed in the future after suitable tasks have been completed. There is a lack of information on the condition of the fuel debris, its position in the reactor buildings, its heterogeneity, and the amount of fuel debris mixed with the control rods and concrete. A maximum inventory can only be estimated using the amount of installed fuel and the burning history for each reactor. Sampling the fuel debris will be attempted before its removal. There is a possibility that small fragments of the fuel debris have been circulating in the highly concentrated water treatment system and have been adsorbed somewhere in the system. The effects of the injection of seawater and borated water on the properties of the fuel debris should be considered. The presence of sodium chloride and boron may affect the subsequent treatment and disposal system performance.
Issues associated with main processes: The fuel debris will be placed in “fuel debris containers” and transported outside the reactor buildings during the removal step. The functions of the containers should be discussed at an early stage if they are to be used for the subsequent treatment and disposal.
In the treatment process, an issue will be whether reprocessing or direct disposal with stabilization if necessary should be carried out. The former requires the development of technology to resolve the fuel debris and produces a vitrified waste when employing normal reprocessing technology. A dry process can also be used for reprocessing. Direct disposal will require clarification of the performance of the fuel debris as a waste, i.e., its radionuclide release characteristics.
Various treatment technologies to stabilize the fuel debris in the subsequent stages including disposal have to be investigated. Suitable technologies for treating the fuel debris should be available before choosing between reprocessing and direct disposal. Simulated and real fuel debris samples are planned to be investigated to obtain information on their properties. Direct disposal will require the design of a disposal system considering criticality and hydrogen generation. The recovery of dispersed fine fuel debris is also an issue in obtaining as accurate inventory as possible.
Information from previous accidents: The fuel debris was stored in the case of TMI. The fuel debris removed from the reactor building of TMI 2 is currently stored in three types of container [41]. Technologies for drying at high temperatures [42] and a portable vacuum water-removal technique [43] were investigated to remove pore water from the fuel/concrete debris. The effect of sodium chloride on container corrosion is an issue in the storage of 1F fuel debris. High-temperature metallurgical technology was examined and Li reduction using synthetic “corium” which was a mixture of (U, Zr)O, Zr, Fe, and Cr was successfully carried out [44].
Analysis of the safety of direct disposal of the fuel debris was performed and disposal at the Yucca Mountain site was once planned [45]. The dissolution rate of the fuel debris was set to 100 [46] and 1,000 times [47] that of the spent fuel, considering the increased specific surface area due to the existence of small pieces of debris. For stabilization, the use of iron enriched basalt was investigated [14] and applied to the fuel debris as well the spent zeolite from TMI. A basic study on the solidification and leaching of uranium dioxide containing 15% Zircaloy was also carried out [15]. The experience obtained from TMI can be referred to when considering the performance of waste forms of the 1F fuel debris.
At Chernobyl, the fuel debris was stored in the shelter together with all the remaining fuels and graphite as forth without any treatment [36].
Activated waste
Properties: “Activated waste” consists of control rods, channel boxes, tie plates, spacer, and other materials in the reactor pressure vessels, and vessels themselves, that did not form debris. The inventory of activated radionuclides for each type of waste can be calculated. However, it is difficult to estimate the radionuclide component from the contamination of the fuel debris and the highly contaminated water. It is necessary to clarify the amounts of boron, oil, and other impurities by measurement. The corrosion of these materials due to sodium chloride may have resulted in the adsorption and absorption of some radionuclides, making inventory estimation difficult. It will be possible to decontaminate these materials when they are removed after the complete removal of the fuel debris. The management of these many types of material is important to maintain the accurate inventories needed in the safety analysis. It is also necessary to distinguish between the activated radionuclides and those contaminating the surface in the safety analysis because both have different release rates.
Issues associated with main processes: After the removal of the activated waste, appropriate separation of the waste and the recording of its properties including its inventory are important for obtaining accurate data for safety analysis. If necessary, decontamination technology should be chosen considering the generation of secondary waste from the decontamination process and the branching of the inventory during the process. Obtaining an inventory of each type of waste is important in selecting a disposal method for the waste.
Information from previous accidents: At Chernobyl, the waste corresponding to the activated waste was stored in facilities called “cascade walls” near the shelter along with other waste [36]. The ordinary decommissioning process and technologies applied in Japan can be used when considering the 1F waste characteristics.
Decommissioning waste
Properties: “Decommissioning waste” is defined as waste generated during the decommissioning of buildings after completion of the fuel debris removal. Large amount of reinforced concrete, metals, and instruments will be involved. Contamination is widely dispersed in the buildings. Reactor buildings and part of the turbine buildings are immersed in highly contaminated water containing radionuclides included in the spent fuel and the fuel debris as well as the impurities. Most of the radionuclides are expected to be on the surface of the concrete and the instruments. Although some of them may have penetrated the concrete through cracks and pores during the long term immersion, it is thought that the radionuclide concentration in the highly contaminated water has gradually decreased.
The other part of the waste is contaminated with radionuclides through the air. It is considered that the waste will be generated in a more predicable manner than the rubble waste. Therefore, appropriate observation and decontamination of the radionulides contained in the waste are expected to significantly decrease the amount of radioactive waste in combination with a reuse/recycle strategy. Technologies for decontamination and estimation of the thickness of the concrete to be decontaminated are issues.
Issues associated with main processes: Decommissioning of the buildings will be carried out after 20-25 years once the fuel debris has been removed. It is important to discuss and make plans to ensure appropriate handling of the decommissioned waste during the management of waste.
Contaminated soil
Properties: The soil of the ground surface around the buildings is contaminated by radionuclides dispersed by the hydrogen explosions. Bushes and fallen leaves covering the ground will be mixed with the soil when separation is difficult. Part of the underground soil may have been contaminated through groundwater. Most of the contaminated soil is expected to remain where it is, although radionuclide measurements have been carried out at a few points on-site as well as at locations near the nuclear power plant [48]. Only background levels of radioactive cesium and plutonium isotopes have been observed in the surface soil. Sodium chloride and other impurities are also contained in the surface soil as well as other major types of 1F waste. There is no data for underground soil.
Issues associated with main processes: The soil will be removed or treated as waste in accordance with the progress of the restoration of the plant, i.e., when the decontamination and removal of soil are required to construct new apparatus. Techniques to effectively estimate the radionuclide concentration in the soil and in the contaminated area underground are needed by evaluating observed data or radiation rate data.
The contaminated soil will be suitably treated applying the technology for treating the soil will be discussed in the near future, also considering on-site reuse and recycling. Therefore, technologies are issues when planning separation and treatment for decontamination to reduce and control the volume of soil to be disposed of and reused, similarly to the case of rubble. In addition, a suitable disposal concept should be discussed for the contaminated soil containing impurities.
Information from previous accidents: Near the Hanford site, the installation of engineered barriers and purification of the groundwater have been performed to prevent the spread of radionuclide contamination [49]. Environmental remediation has also been carried out for contaminated soil at other nuclear facilities around the world [33]. At Chernobyl, the contaminated soil waste is stored with other types of waste in the facilities near the shelter and in the facilities located in the CEZ [36].
Other waste
The other types of waste described in the following are considered to be less important because of their lower radioactivity and previous experience of dealing with similar radioactive waste. This is in contrast to fuel debris and other types of previously discussed waste, for which there is no experience of processing and disposal. The characteristics and noteworthy issues of other types of waste are described below.
Some of the spent fuel taken from the spent fuel pools in accordance with the schedule in the roadmap may have been affected by contamination with impurities during the cooling operations after the accident. The effect of such impurities should be considered in the storage and the subsequent treatment and disposal even if most of them can be removed by washing. Investigation of the dissolution of spent fuel and borosilicate glass in brine or seawater carried out in the world [50,51] will necessary to ensure the safe disposal safety of the 1F spent fuel.
Secondary waste such as used activated carbon, zeolite, and chelating resin has been generated in the adsorption vessels by the purification of the seawater filling the turbine buildings of Units 5 and 6. This waste may have very similar characteristics to the secondary waste of the highly contaminated water, even though its radioactivity is not high owing to the treatment of the slightly contaminated seawater. The low radioactivity concentration of this waste is expected to make its processing and disposal simple and straightforward. Seawater in the port in front of the intake canal is also being treated in the same manner as the slightly contaminated seawater, resulting in the generation of waste possibly with a rather higher radioactivity concentration, which may cause hydrogen production due to radiolysis. There is an issue of the effect of sodium chloride during the storage of waste. Zeolite particles contained in the sand bags used to adsorb radioactive cesium dispersed by the action of rain in the nuclear power station can be treated in the same way as the above described zeolite waste.
Seabed soil covered with cement and bentonite clay remains in the port in front of the intake canal to prevent the diffusion of radioactive materials. Note that the likelihood of the dispersion of soil particles containing radioactivity must be minimized when removing the soil from the sea bottom. Seawater contained in the seabed soil may affect metallic containers and facilities, which should be considered when planning long-term storage and disposal. Technologies applied in the treatment of contaminated soil at U. S. nuclear facilities are worth examining for their applicability to the seabed soil.
Tanks used to store various types of treated water are being replaced. These tanks must be disposed of or reused on-site. Various types of instruments, tools, and workers’ clothes are accumulating, which will continue with the progress of restoration. Burnable waste will be treated by a furnace on-site. The ash waste and unburnable waste will be considered by referring the techniques that have been applied to the waste from ordinary nuclear power plant operations.
Common issues
There are some issues common to all types of 1F waste from the viewpoints of their properties and the process of contamination, i.e., through the air or by highly contaminated water. It is worth considering the issues described for each type of waste from a crosssectional viewpoint. Inventory estimation is a typical common issue for all types of waste. The difficulty of inventory estimation depends on the type of waste through factors such as the important radionuclides contained, the distribution range of radioactivity in the waste, and the existence of measurement techniques suitable for determining radioactivity of the waste.
The presence of chemical impurities mixed with the wastes is also a common issue. The effects of chemical impurities on the processing and disposal are different for each type of waste because of their contents, cleaning techniques, and their effectiveness, and the required performances of natural and engineered barrier systems. Possible methods of estimating the inventories and the effects of impurities should be attempted periodically using applicable data and information with decreasing uncertainty in a stepwise manner.
The corrosion of waste containers is an issue for all the processes owing to the high content of sodium chloride in most of the liquid and solid waste as well as the possibility of hydrogen gas generation by radiolysis. The functions of the containers are also an issue, i.e., whether one type container should be used for all processes whether waste should be transferred to a suitable container for each process.
Separation and categorization are important for waste with a wide radioactivity concentration distribution. Suitable technologies for such waste are expected to help control the volume of radioactive waste to be disposed of.
Waste treatment involving decontamination, stabilization, and solidification should be effectively performed considering the ease of handling each waste during the processing and disposal. Various treatment technologies are required for each type of waste. Generation of secondary waste by treatment and obtaining an inventory of the treated waste are also issues.
A suitable disposal concept and system design for each types of waste must be selected, although generic approach to disposal safety analysis will be continued for a while. The configuration of engineered barrier systems and the effect of chemical impurities are to be discussed and evaluated using appropriate parameters. Concepts investigated in other countries are also to be examined for some types of 1F waste that have very different characteristics from normal radioactive waste from nuclear power generation.

Conclusion

Research and development on the management of post-accident waste in Fukushima Daiichi Nuclear Power Plant are in the initial stage, although the management of very limited types of waste is already under investigation. The waste is being accumulated and should be managed in a suitable manner. On the other hand, information on the different types of waste is not sufficient, and more time is required to define inventories and accurately determine the amount of waste. However, it will be worthwhile reviewing information on up-todate on-site waste characteristics and properties in the power plant as well as the experiences obtain from ordinary radioactive waste management and from previous nuclear accidents in order to discuss the future research and development of waste management strategies.
The technical information reviewed and the issues examined in this review are expected to be useful for making detailed research and development plans and improving them by analyzing the issues in detail. Work carried out to restore of the site of the accident may change in future, affecting the characteristics of the waste and possibly generating new types of waste. Information on the waste should be updated frequently with the progress of the restoration to improve the effectiveness of research and development and to find valid methods and technologies for the processing and disposal of each type of 1F post-accident waste.

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