Japan: experience of radioactive waste (RAW) management and contaminated site clean-up

23 Japan: experience of radioactive waste (RAW) management and contaminated site clean-up H. R I N D O, K. TA K A H A S H I and M. TAC H I BA N A, Nuclear Cycle Backend Directorate, Japan DOI: 10.1533/9780857097446.2.723 Abstract: This chapter summarizes the current strategy and policy for radioactive waste management in Japan which has been hindered by a lack of public acceptance and of a final high level waste end-point (geological repository). Ongoing decommissioning of several nuclear facilities, including the Tokai-1 NPP, the Advanced Thermal Reactor ‘Fugen’ and the Plutonium Fuel Fabrication Facility (PFFF), are described. Key words: radioactive waste treatment, radioactive waste disposal, decommissioning and dismantling, nuclear facilities, policy and strategy, Japan.

23.1

Introduction

This chapter was written largely before the Fukushima accident, details of which and the clean-up programme are included in the next chapter. However, not only nuclear policy but also nuclear safety regulation in Japan is likely to change in the future.

23.1.1 Nuclear energy in Japan Japan has carried out nuclear power generation research since the middle of the 1950s. A test power reactor, the Japan Power Demonstration Reactor (JPDR), started operation in 1963 and Tokai-1 Nuclear Power Plant (NPP), the first commercial reactor, went into operation in 1966 with a generation capacity of 166 MWe. Currently, about 50 commercial nuclear reactors, predominantly boiling water reactors (BWRs), and pressurised water reactors (PWRs), are in operation, with a total generation capacity of 48,847 MWe. Prior to the Fukushima disaster, about 30% of Japan’s electricity came from nuclear power (Plate VIII between pages 448 and 449). Japan will continue to develop nuclear power as a mainstay of non-fossil energy, while placing the highest priority on safety1. 723 © Woodhead Publishing Limited, 2013

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The Framework for Nuclear Energy Policy (FNEP), which was established by the Japan Atomic Energy Commission (JAEC) as the basics for political measures regarding the use of nuclear power generation and radiation to be promoted by governmental agencies for the next 10 years, was approved by the Cabinet in October 20052. Prior to the events at Fukushima, nuclear energy was expected to continue to contribute to the pursuit of an optimum energy supply mix for Japan. The FNEP specified that nuclear power’s share of Japan’s total power generation should be maintained at 30–40% or more beyond 2030 and that the nuclear fuel cycle should be promoted3. Nuclear power generation is the key base-load power source. After Fukushima, in July 2011, the Energy & Environment Council (Enecan or EEC) was set up by the Cabinet Office to recommend on Japan’s energy future to 2050. It is chaired by the Minister for National Policy and will focus on future dependence on nuclear power. In September 2011, Japan’s prime minister said he expected the country to reduce its dependency on nuclear power in the medium and long term, and that the government would address the question of those new plants now under construction. He said that the national Basic Energy Policy would be revised from scratch, and that a new strategy and plan to 2030 would be created. He also stated that Japan’s ministerial-level Energy and Environment Council would ‘thoroughly review nuclear policy and seek a new form’. The review may recommend that nuclear power’s contribution to electricity be targeted at 0%, 15%, or 20–25% for the medium term – a 36% option was dropped.

23.1.2 Radioactive waste (RAW) management policy Radioactive waste (RAW) is generated by the research, development and utilization of nuclear energy at NPPs, nuclear fuel cycle facilities, test and research reactors, universities, institutes, and medical facilities, using accelerators, radioactive isotopes (RI) and nuclear fuel materials. It is essential that activities associated with research, development and utilization of nuclear energy also process and dispose of the RAW in such a way as to prevent any significant effects on the human environment now and in the future. The generation that has enjoyed the convenience and benefits of nuclear energy assumes the responsibility to expend all efforts for safe disposal of RAW for the next generation. There are four principles for the treatment and disposal of RAW: 1. 2. 3. 4.

The liability of generators, Minimization of radioactive waste, Rational treatment and disposal, Implementation based on mutual understanding with the people.

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Under these principles, it is important to appropriately classify the wastes and treat and dispose of each classification safely based on the recognition that the wastes may include materials with characteristics that take an extraordinarily long time for the radioactivity to drop to insignificant levels2. A near-surface disposal facility already operates for most of the low-level radioactive waste (LLW) generated at NPPs and is operated in Rokkasyo, Aomori-Ken by Japan Nuclear Fuel Limited (JNFL), as a private business, excluding part of the LLW. With regard to near-surface disposal of RI and research wastes, the Japan Atomic Energy Agency (JAEA) will conduct and promote disposal activity in cooperation with the government and other waste generators. As for the remaining LLW, JNFL plans to construct an intermediate depth disposal facility for NPPs and the Nuclear Waste Management Organization of Japan (NUMO) will geologically dispose of transuranic (TRU) wastes. Funds from the owner of the reprocessing plant and mixed oxide (MOX) fuel fabricator, etc., have been accumulating via a levy to pay for geological disposal of TRU wastes since 2009. However, the implementing body for subsurface disposal of LLW, RI and research wastes has yet to be decided. High-level radioactive waste (HLW), generated during reprocessing spent fuel (SF), is being vitrified and packaged prior to disposal in a geological repository. Research and development for that purpose had been conducted mainly by what was the Power Reactor and Nuclear Fuel Development Corporation (PNC), which was restructured as the JAEA in October 2005 through the Japan Nuclear Cycle Development Institute. The government worked to develop a disposal system taking into consideration these policy guidelines and scientific evidence, and enacted the Specified Radioactive Waste Final Disposal Act in June 2000. NUMO was created in October 2002 as an implementing body for disposal, as specified in the Act. In December 2002, NUMO started ‘open solicitation’, which encouraged municipalities to consider investigating the suitability of their local area for developing a deep repository for HLW. Meanwhile, electric utilities and others have been accumulating funds for the disposal of HLW.

23.2

Radioactive waste (RAW) management strategy

23.2.1 Sources, types and classification of radioactive waste In Japan, RAW is categorized as shown in Table 23.14. In May 2007, the Nuclear Safety Commission of Japan (NSC) issued a document which provides for upper bounds of concentration of radioactive elements in waste packages from power reactors and in TRU waste packages. The upper bounds of concentration of radioactive elements are so decided, that the

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Table 23.1 Classification of radioactive wastes in Japan

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Classification

Origin of

Disposal method

High-level radioactive waste

Reprocessing plant

Deep geological disposal (>300 m)

Nuclear power plant

Sub-surface disposal (50–100 m)

Low-level radioactive waste

Waste power reactor

Relatively higher level Lower level

Near-surface disposal with artificial barrier

Very low level

Near-surface disposal without artificial barrier

TRU

Reprocessing plant, MOX fuel fabrication plant

Deep geological, sub-surface and near-surface disposal

Uranium waste

Enrichment plant, fuel fabrication plant and conversion and refining plant

Not yet decided

RI and research waste

Research facility RI utilizing facility

Deep geological, sub-surface and near-surface disposal

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public exposure due to waste packages is well within the reference value, and that the upper bounds conform to the latest knowledge in the international community. Based on these concepts, disposal of RAW is categorized into Category 1 Waste disposal (geological disposal) and Category 2 Waste disposal (sub-surface disposal, near-surface disposal with artificial barrier and near-surface disposal without artificial barrier). Concerning the waste that does not need to be dealt with as RAW, the NSC has studied the clearance level of radionuclide concentrations and its calculation method, by reference to the ICRP document (Pub. 46, 1985) and IAEA-TECDOC-855, respectively.

23.2.2 Radioactive waste treatment In Japan, HLW and LLW are generated through nuclear power generation, nuclear industries (enrichment, fuel fabrication, reprocessing, etc.), utilization of RI and research. HLW is only generated from reprocessing plants and consists of liquid waste, which is mainly stripped liquid effluent from the extraction process, and solid waste, which is vitrified products of the liquid waste.5 LLW is generated from all nuclear facilities and consists of gaseous, liquid and solid wastes. Wastes from NPPs include gaseous waste (e.g., off-gas from the reactor system and off-gas from the ventilation system of the reactor building), liquid waste (e.g., effluents from the reactor cooling system, floor drains of the reactor building and detergent waste from laundry and hand-wash), and solid wastes (e.g., spent ion-exchange resin, paper, cloth, plastic sheets, tools, pipes, exchanged parts of equipment and filters). Wastes from nuclear fuel cycle facilities include gaseous waste (e.g., off-gas from each process, vessels, hot cells, glove boxes and building ventilation systems), liquid wastes (e.g., effluents from the off-gas scrubber, the acid recovery process and the solvent washing process, drain from analytical laboratory, solvent waste, floor drain and detergent waste), and solid wastes (e.g., hulls, gloves, vinyl bags, paper, cloth, plastic sheets, spent sampling jugs, tools, exchanged parts of equipment and filters). Wastes from enrichment and uranium fuel fabrication facilities include gaseous waste (e.g., off-gas from the building ventilation system), liquid waste (e.g., the floor drain and detergent waste), and solid waste (e.g., paper, cloth, tools pipes, exchanged parts of equipments and filters). Wastes from research facilities, and facilities in which RI are used for medical and industrial purposes, include gaseous wastes (e.g., off-gas from building ventilation systems), liquid wastes (e.g., chemical waste, spent liquid scintillate, the floor drain and detergent waste), and solid wastes (e.g., experimental instruments, syringes, paper, gloves, plastic sheets, tools and

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filters). The management of these types of waste in Japan is described in Section 23.2.4.

23.2.3 Storage and disposal of radioactive waste A large portion of Japan’s radioactive waste (about 50%) is stored in the radioactive waste management facilities at the nuclear facilities. About 1,690 canisters of vitrified product and about 380 m3 of liquid waste as HLW are stored in the reprocessing facilities at Tokai and Rokkasho (interim storage facility of the glass canisters), as of the end of March 2010. About 267,000 m3 of LLW (excluding used steam generators, spent control rods, disused channel boxes) are stored in all nuclear facilities in Japan as of the end of March 2008. Storage volume of LLW is made up of approximately 144,000 m3 NPP waste, approximately 25,000 m3 TRU waste, approximately 9,000 m3 uranium waste, approximately 65,000 m3 research waste, and approximately 24,000 m3 RI waste6. The JNFL near-surface disposal facility with engineered barrier systems in place at Rokkasho, Aomori-Ken is in operation for LLW from commercial NPPs and about 219,000 200 L drums have been disposed of as of the end of March 2010. About 1,670 tons of very low level wastes resulting from dismantling of the Japan Power Demonstration Reactor (JPDR) were disposed of at the near-surface disposal facility without engineered barriers at Tokai. This disposal facility has been on hold since October 1997.

23.2.4 Waste treatment practices7–10 NPPs Off-gas waste from NPPs contains mainly short-lived noble gas nuclides. Off-gas treatment is aimed at decay of short-lived nuclides and the removal of aerosol radionuclides. The off-gas treatment system consists of a hydrogen recombiner unit, an activated charcoal retention unit and a filtration unit. The off-gas from the ventilation system is passed through the charcoal filter and the high efficiency particulates air (HEPA) filter to eliminate iodine and aerosol, respectively. The treated off-gas is discharged through a stack after verification that it is under the regulatory limit. Treatment of the liquid waste from NPPs aims where possible at recycling in the plant system, removing the radioactivity in controlled liquid discharges and eliminating process effluents. The liquid waste treatment system in a BWR is composed of the low conductivity subsystem, the high conductivity subsystem, the detergent waste subsystem, and the solidification subsystem. The liquid waste from processes in BWR is divided into the low conductivity effluent, which has relatively high purification, and the high conductivity effluent, which has relatively low purification. The low

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conductivity subsystem collects and processes liquid wastes typically from the equipment drains in the primary cooling system. This waste is filtered through a hollow fibre membrane for removal of insoluble material and demineralized by mixed ion exchange resin for removal of soluble chemicals, and then returned to condensate storage prior to further use as reactor coolant. The high conductivity subsystem collects and processes liquid wastes from floor drains and effluents from regeneration of the resins. These wastes are concentrated by evaporation, and fed to the solidification subsystem. The distillate is demineralized on the mixed ion-exchange resin, and then returned to condensate storage or discharged to the ocean after verification under regulatory limit. The detergent subsystem collects and processes detergent waste from personnel hand-wash and laundry operations. These wastes are filtered, and then discharged to the ocean. The solidification subsystem collects concentrated waste in a dedicated tank. This waste is mixed with cement or bitumen, and solidified in 200 L drums. The latest solidification subsystem adopts the dry-pelletizing method in which the concentrated waste is dried with a film evaporator, and dried powder mixed with binder is pelletized by a granulator. This method gives a high waste reduction volume compared to cementation/bituminization. The liquid waste treatment system for PWR is similar to that for BWRs. PWR employs the recoverable effluent subsystem corresponding to the low conductivity subsystem of BWR. The recoverable effluent containing boric acid from the primary coolant system is demineralized, and then treated by evaporation to separate water and boric acid solutions for further use. Other subsystems are similar to those in BWRs. Treatment of the solid waste from NPPs is aimed at stabilization and volume reduction for storage and conditioning prior to disposal. The solid waste treatment system is constructed typically of the pre-treatment subsystem, the incineration subsystem for combustible material, the compaction subsystem for incombustibles and the conditioning subsystem. The pre-treatment subsystem is composed usually of cutting and segregation. Large wastes are cut into small pieces appropriate for compaction/packing. Wastes are sorted into combustible, incombustible, compressible and incompressible wastes. Combustible wastes are burned typically in an excess airtype incinerator, and the incinerated ash is placed in a 200 L drum for storage. The compaction subsystem usually makes waste compacts of compressible and incombustible wastes with a compressing force between 50 kN and 3 MN. For higher reduction ratios, the super-compactor or the melting system is adopted in some NPPs. Waste compacts and incompressible wastes are placed in 200 L drums, and then filled with mortar in the conditioning subsystem for disposal. Figure 23.1 shows a typical treatment flow for BWR wastes.

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Radioactive waste management and contaminated site clean-up Air ejector

Hydrogen recombiner

Activated charcoal retention unit

Ventilation

Charcoal filter

HEPA filter

Low conductivity tank

Hollow fibre filter

Demineralizer

High conductivity tank

Evaporator

Demineralizer

Gaseous waste

Liquid waste

HEPA filter To stack

To condensate storage

Cementation unit Detergent waste collector

To the ocean Filter

Combustibles

Incinerator To storage

Solid waste

Cutting/ segregation

Compressible

Compactor/ Super-compactor

Mortar filler

Incombustibles Incompressible

Mortar filler

23.1 Typical waste treatment flow for BWR waste.

Nuclear fuel cycle facilities11,12 Treatment of the gaseous waste from nuclear fuel cycle facilities removes aerial radioactive particles and gaseous radioactive nuclides before discharge of the gaseous effluent into the environment. In reprocessing plants, the gaseous waste is filtered through scrubbers, Ag-zeolite/Ag-silica-gel filters for iodine and HEPA filters, and then discharged through a stack after radioactivity measurement to ensure it is below regulatory limits. In MOX fuel fabrication plants, only aerial radioactive particles are generated. The off-gas is passed through HEPA filters, and discharged through a stack after measuring radioactivity. Treatment of the liquid waste from reprocessing plants removes radioactivity in controlled liquid discharges and eliminates process effluents. The liquid waste treatment system is composed typically of the high active liquid subsystem, the intermediate active liquid subsystem, the low active liquid subsystem, the solidification subsystem and the solvent waste subsystem. The high active liquid subsystem collects and processes typically raffinate from the separation/extraction process. This waste is concentrated by evaporation, vitrified and stored in a dedicated interim facility. The intermediate active liquid subsystem collects and processes effluents from the acid recovery process, the solvent washing process, the off-gas scrubber, etc. This waste is concentrated by evaporation, and then the distillate is fed to the low active liquid subsystem.

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The low active liquid subsystem collects and processes floor drain liquids, the detergent waste, etc. These wastes are distilled, then filtered and discharged into the ocean after activity measurement. The concentrated waste is fed to the solidification subsystem. The solidification subsystem collects and processes the concentrated wastes from the intermediate active liquid subsystem and the low active liquid subsystem. These wastes are adjusted to appropriate pH and concentrated by flocculation/ultrafiltration, and then solidified with cement in the Tokai reprocessing plant of the JAEA Tokai Research and Development Center. In the spent fuel reprocessing plant in Rokkasho village (Rokkasho reprocessing plant), these wastes are dried with a film evaporator and pelletized with a granulator. The processed solid wastes are stored in the facilities. The solvent waste subsystem collects and processes waste solvent from the solvent washing process. This waste is solidified with epoxy resin at Tokai and hydrothermally solidified after pyrolysis at Rokkasho. Treatment of the solid waste is implemented with the aim of volume reduction for storage, because a disposal facility is not yet available in Japan for TRU waste. The combustible wastes are incinerated, and the incinerated ash is placed in 200 L drums for storage. The non-combustible wastes are placed directly in appropriate containers, and stored at the facilities. In Tokai MOX fuel fabrication plant, plastics and polyvinyl chloride (PVC) are also incinerated in a dedicated incinerator. Enrichment plants Treatment of the gaseous waste from enrichment plants removes fluoride and radioactive particles before discharge into the environment. The off-gas from centrifuges is typically filtered with NaF filters, alumina filters, and HEPA filters, and then discharged through a stack after radioactivity measurement. In the enrichment plants, small amounts of liquid waste are generated from floor drainage and detergent wastes. These wastes are typically treated by flocculation using a flocculate agent such as polyaluminum chloride, and discharged into the environment. Treatment of the solid waste is aimed at volume reduction for storage. Combustible wastes are incinerated, and then placed in 200 L storage drums. Incombustible wastes are placed directly in appropriate containers, and stored at the facilities. Research facilities13 Treatment of the gaseous waste from research facilities is performed to remove radioactive particles before discharge into the environment. The

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off-gas from the ventilation system is passed through a HEPA filter, and then discharged through a stack after radioactivity measurement. Small amounts of liquid waste are generated from chemical drains, floor drains and detergent waste. These wastes are treated to remove radionuclides by flocculation or evaporation, and discharged into the environment. Pre-treatment such as neutralization is performed before treatment as needed. In many small laboratories, solid wastes are placed in containers for storage without treatment. However, large institutes, in which many solid wastes are generated, treat the solid waste to reduce its volume for storage. The solid waste treatment system consists typically of pre-treatment, incineration and super-compaction. The pre-treatment is composed of cutting and segregation. Large wastes are cut into small pieces appropriate for compaction/packing. Solid wastes are sorted into combustible wastes and non-combustible wastes. The non-combustible wastes are further sorted into compressible and incompressible wastes. The combustible wastes are burned in excess air-type incinerators, and then stored. The compressive wastes are compressed with 20 MN force at the super-compactor. The waste compacts and incompressible wastes are placed in 200 L storage drums. Medical, industrial and research laboratories using RI (RI waste) Radioactive liquid and solid wastes from utilization of RI are exclusively collected and treated by the Japan Radioisotope Association. Organic liquid waste and flammable solid wastes are treated by incineration, and stored in suitable containers. Inflammable solid wastes are compressed with a compactor, and placed in 200 L drums. Technological development Development of treatment technology for liquid wastes is mainly aimed at solidification of the concentrated liquid waste to reduce the product volume ratio. Cementation technology was originally adopted while bituminization, plastic solidification with unsaturated polyester resin, and improved cementation technologies were developed later. The latest solidification treatment adopts the dry-pelletizing method combining a film evaporator and a granulator. The resulting granules are mixed with Portland cement and solidified in 200 L drums. This method gives a high waste volume reduction compared to the early cementation method. Technology development of solid waste treatment has largely focused on incineration technology and volume reduction technologies for miscellaneous waste. Incineration for chlorine-containing materials including PVC using an incinerator with a water-cooled cylindrical chamber has been

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developed. This incinerator has an off-gas system as a countermeasure against dioxin. An incinerator with an ash melting system has also been developed with a high temperature chamber to burn and melt simultaneously or an incinerator and a separate melting furnace. This type of incineration system can treat flammable and non-flammable materials producing slug granules with high volume reduction ratio. For non-combustible wastes, super-compaction technology and melting technology have been developed. A super-compaction system with over 10 MN compressive force has been developed using a triaxial compressive or a uniaxial compressive machine with a drawing mould for direct encapsulation to 200 L storage drums. Both high-frequency induction or plasma heating furnaces are used in NPP for melting. The high-frequency induction furnaces use a disposable crucible, which can be placed directly in a 200 L drum. Some treatment technologies such as vitrification, steam reforming, etc., for iodine filters and ion-exchange resins are under development.

23.2.5 Radioactive waste disposal14 High-level radioactive waste In line with the Specified Radioactive Waste Final Disposal Act, final disposal facilities are planned for the geological disposal of HLW and are scheduled to start operation in the 2030s through the following three-step selection process: selection of preliminary investigation areas, selection of detailed investigation areas, and selection of the final disposal facility areas. When local governments wish to volunteer for ‘areas to be investigated as to the feasibility of constructing final repository of HLW’, it is important that the implementor (NUMO), the government and the utility companies give sufficient understanding and awareness to the local residents about the advantages and disadvantages of the final repository and various sectors of the local community, including local government. The government, research and development (R&D) institutions and NUMO, while giving due consideration to their own roles and in close partnership, are expected to consistently promote R&D into HLW geological disposal. NUMO is expected to safely implement the final HLW disposal project and systematically perform technical development to improve the economics and efficiency of the disposal activities. R&D institutions, led by the JAEA, through utilization of underground research facilities, continue to conduct research on underground geology, basic R&D towards improved reliability of geological disposal technology and safety assessment methods, and for safety regulations. While being aware of overseas knowledge and experience, it is important to develop and maintain an advanced knowledge base that supports final

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repository projects and safety regulations, as well as to appropriately reflect it in NUMO’s final disposal projects. To this end, the government and R&D institutions work together to survey the entire Japanese waste management programme systematically and efficiently. R&D institutions such as JAEA, Radioactive Waste Management Funding and Research Center, etc., need to cooperate with the government and NUMO in activities to improve the understanding and awareness of society at large. Furthermore, it is necessary for the government to develop specific rules concerning safety regulations based on the progress of these R&D activities. Geological disposal of radioactive wastes containing transuranium elements Some LLW containing TRU elements needs to be disposed of geologically. If some TRU waste targeted for geological disposal can be buried together with HLW (co-disposed), the number of repository sites may be reduced, improving economic efficiency. Based on assessment of the influence of TRU and HLW co-disposal on the integrity of the disposal site, the government should then consider necessary measures, including the nature of an implementing body and its own involvement. LLW from overseas reprocessing consigned by Japan will gradually be returned from France and the UK. French reprocessing firms suggest changing the solidification method from embedding in bitumen to vitrification, while UK reprocessing companies will embed the LLW in cement for geological or disposal with institutional control. In the latter case, the waste returned to Japan is HLW (vitrified waste) with equivalent levels of radioactivity to the LLW exported. In light of these suggestions, it is expected that the number of shipments can be reduced and storage facilities in Japan for LLW awaiting final disposal can be downsized. Thus, the government, in response to discussions with the operators, will assess the benefits of waste treatment by the new solidification methods, suggested by France, and of the conversion indexes of waste, as suggested by the UK. If these suggestions are found to be acceptable, the government should discuss the institutional issues. Radioactive wastes for disposal with institutional control Methods of disposal with institutional control include near-surface disposal without artificial barriers, near-surface disposal with artificial barriers and sub-surface disposal with artificial barriers. Near-surface disposal with artificial barriers is already used for LLW generated in commercial nuclear reactor facilities. Near-surface disposal without artificial barriers is being partly implemented while the reactor operators improve safety regulations

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on the remainder of the wastes. Operating entities are conducting studies and tests on sub-surface disposal with artificial barriers. Based on the results, it will be necessary to discuss the establishment of the framework, including safety regulations. The current status of management of other LLW is described in Section 23.1.2.

23.3

Spent fuel management strategy, practice and issues15

In accordance with the basic principle in the Framework for Nuclear Energy Policy, the ‘Act for Deposit and Administration of Reserve Funds for Reprocessing of Spent Fuel from Nuclear Power Generation’ was established requiring operators to place funds for spent fuel reprocessing in a fund administration corporation. The objective of ‘the Act’ is to ensure the proper implementation of spent fuel reprocessing, disposal of radioactive wastes generated from reprocessing and decommissioning of the reprocessing facilities. The reserve fund held by the 10 utility companies at the end of March 2007 was a 1,390 billion yen. As a part of its waste management plans, the Ministry of Economy, Trade and Industry (METI) designated the ‘Radioactive Waste Management Funding and Research Center’ as a nonprofit ‘fund administration corporation’ (October, 2005) that is supervised by the METI through supervisory orders and on-the-spot inspection. SF generated in power reactors is sent to reprocessing facilities after a period of on-site cooling and storage. SF has historically been reprocessed overseas in accordance with contracts with British and French companies, with the exception of a portion reprocessed by the Tokai reprocessing plant of the JAEA. However, considering the national need, JNFL constructed the Rokkasho reprocessing plant, based on operational experience accumulated at the Tokai reprocessing plant and on overseas technologies and experience. The plant underwent active testing using SF in 2008 and started operation in 2008. Storage of SF in the plant storage facility started in 1999, and export of SF to foreign reprocessing plants ended in July 2001. The Law for Regulation of Nuclear Source Material, Nuclear Fuel Material and Nuclear Reactors (Reactor Regulation Law) was amended in 1999 to incorporate provisions on interim SF storage. Tokyo Electric Power Company and Japan Atomic Power Company (JAPC) jointly established the ‘Recycle Fuel Storage Company’ to prepare for commercial operation of the first interim fuel storage facilities planned for 2010. In March 2007, the company applied for a licence to construct and operate the Recycle Fuel Storage Center at Mutsu city, Aomori Prefecture, and the licence application is now undergoing the safety examination.

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SF from research reactor facilities has been, and is to be, returned to the US, UK or France, or is to be reprocessed or stored in Japan.

23.4

Decommissioning strategy, practice and issues16–20

23.4.1 Decommissioning strategy The basic policy for decommissioning commercial NPPs was established by the JAEC in 1982. It states that retired commercial NPPs should be dismantled as early as possible after shutdown and the site should be effectively re-used for next generation NPP. The Framework for Nuclear Energy Policy issued by the JAEC states that it is the operator’s own responsibility, but under government regulation, to carry out decommissioning of a nuclear facility, ensuring safety, while obtaining local communities’ understanding and cooperation. The regulatory policy for dismantling or decommissioning reactor facilities has been discussed by the NSC. To ensure safety during decommissioning of commercial NPPs, the regulation was implemented by applying existing provisions in the Reactor Regulation Law by the operators. To date, decommissioning of reactor facilities has been implemented at facilities such as the JPDR of the JAEA and the Tokai-1 NPP of JAPC, the development and application of dismantling technologies have progressed, and know-how for decommissioning has been accumulated. The NSC examined the idea of ideal safety regulation, based on the experience of decommissioning of nuclear facilities. It also took into consideration the features of nuclear facilities post-termination and the level of potential risks. The Decommissioning Safety Subcommittee has investigated the appropriate regulation systems for decommissioning, based on regulatory experiences of decommissioning reactor facilities. The Decommissioning Safety Subcommittee proposed the decommissioning regulations as: •

replacing dismantling notification by licensee, to approval of the licensee’s decommissioning plan by the regulatory body, • implementation of decommissioning as approved in the decommissioning plan, • completion of decommissioning is confirmed by the regulatory body and after confirmation of the completion of decommissioning, the operating licence becomes ineffective, • the regulatory activities during the decommissioning process should be changed in accordance with the changes of functions of facilities and safety operation activities as the decommissioning proceeds.

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On the basis of such recognition, the Reactor Regulation Law was amended in 2005. A licensee applying for approval of decommissioning has to submit a decommissioning plan that describes, for example dismantling methods, radiation controls, safety assessment and the financial plans. The regulatory body approves the decommissioning plan after examining its conformity with technical standards. At the final stage of decommissioning, the licensee submits a document that describes the implementation status of dismantling, management of contaminated materials and the final distribution of contamination and requests the regulatory body’s confirmation. The decommissioning is completed after the regulatory body confirms that the measures for radiation hazard prevention are no longer necessary and management of contaminated materials is completed.

23.4.2 Decommissioning practice and issues Nuclear facilities in the process of being decommissioned in Japan include Tokai-1 NPP of the JAPC, the Advanced Thermal Reactor ‘Fugen’ of the JAEA and the Plutonium Fuel Fabrication Facility (PFFF) of the JAEA. Hamaoka Nuclear Power Station Reactor’s No. 1 and 2 of Chubu Electric Power Company shut down in January 2009 and their decommissioning plans were approved by the METI in November 2009. Tokai-1 Tokai-1 NPP (GCR) is a graphite-moderated, gas-cooled reactor. Tokai-1 NPP started commercial operation in 1966. However, it has disadvantages from an economic standpoint because the carbon dioxide GCR has a relatively low power output for the volume of the reactor. This raises the cost of electricity generation compared with light water reactors. Tokai-1 NPP was finally shut down in 1998 after it was defuelled and all fuel elements were shipped off-site for reprocessing by 2001. The reactor area was stored in a safe condition for 10 years after final shutdown to reduce radioactivity. During safe storage, conventional facilities outside the reactor area are removed to secure a transportation route for dismantling wastes, and also to create space for new waste conditioning facilities. Starting with peripheral equipment outside the reactor area, Tokai-1 NPP is being demolished in stages. Equipment in the reactor area will be dismantled and removed after being securely stored until radioactivity has decayed to an allowable level. Finally, the site would be able to be re-used for a new NPP. Dismantling activities initiated in 2001 and during the first five years, conventional facilities, such as the turbine system were removed. Cartridge cooling pond (CCP) water was also drained and the CCP was cleaned up for clearance activities.

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Four steam raising units (SRUs) have been removed since 2006. The SRUs are 24.7 m in height, 6.3 m in diameter and within the SRUs are radioactively contaminated and complicated structures. Jack-down methods and a remote dismantling system were developed for workers’ safety and to minimize the extent of the contaminated areas. The SRUs are removed with the system remotely in turn from bottom while lifting them with large jacks. The system enables cutting and holding not only of the SRU body but also other internal parts of the SRUs. Figure 23.2 shows images of cutting work on the SRU. The jack-down method prevents the need to work in high places and restricts the radiation controlled area to the bottom area of the SRU. Fugen Fugen NPP (ATR, 165 MWe) is a heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. Fugen NPP began operating in March 1979, finally shut down in March 2003, and its decommissioning plan was approved in February 2008. Fugen NPP dismantling was separated into the following four periods. 1. Spent fuel transfer period. SF will be transported to Tokai reprocessing plant, and heavy water will be transported to Canada for re-use at

Device configuration

Tier transfer

Primary cutting

Secondary cutting

23.2 Cutting working images of the SRU.

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CANDU reactors. Less contaminated equipment such as turbines will be dismantled, while some related systems for SF storage remain operational. 2. Peripheral facilities dismantling period. After SF transportation is complete, the related SF storage systems are dismantled and the peripheral reactor equipment will be dismantled to enable installation of remotely operated dismantling machines. 3. Reactor core dismantling period. The reactor core by dismantled by remote operation underwater, and it is expected that the exposure dose in the dismantling activities will be minimized to the equivalent dose of an annual inspection during the plant operation. In this period, both dismantling of all contaminated equipment and decontamination of buildings will be carried out to release the radiation-controlled area. 4. Building demolition period. In this period, both released buildings and non-contaminated buildings will be demolished by conventional methods. After the approval of the programme, decommissioning was initiated. SF has been transferred to the Tokai reprocessing plant, and heavy water has been transported to Canada. Dismantling of the turbine facility was started in parallel. Two of five feedwater heaters and main steam lines were dismantled. Experience of cutting technologies and relevant data such as total manpower have been accumulated for future work. The reactor core of Fugen NPP has a complicated configuration arising from its pressure tube-type structure. The pressure tube and the calandria tube are made of zirconium alloy which can be combustible in powder form. Also, they have been highly activated during operation. It is thus planned to dismantle the core structure underwater for shielding radiation, to prevent airborne dust and for fireproof cutting.

Uranium refining and conversion plant (URCP) The uranium refining and conversion plant (URCP) at Ningyo-toge was constructed in 1981 to demonstrate refining and conversion of yellow cake (or uranium trioxide) to uranium hexafluoride via uranium tetrafluoride. There are two different types of refining processes in the URCP. One is the wet process for converting natural uranium and the other is the dry process for reprocessed uranium. Dismantling of the dry process facilities began in March 2008. The basic strategy concerning plant dismantling was to optimize the total labour costs and minimize the radioactive wastes generated. The basic schedule for dismantling is as follows.

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• Phase 1: removal of large equipment or processes involving uranium hexafluoride, • Phase 2: removal of the greater part of the utilities connecting the main process of URCP, • Phase 3: removal of equipment of main process, • Phase 4: removal of ventilation systems. The majority of equipment will be dismantled, except for building decontamination, by 2013. A large amount of fluidization media had been stored in tanks held underground in the URCP. The fluidization media is composed of small aluminum pellets which absorbed uranium oxides or unreacted uranium tetrafluoride used for the fluorination reaction. They therefore contain high levels of uranium and thorium as progenies of U-232. Among its progenies, Tl-208 is a high gamma emitter, so some external exposure will arise in handling the fluidization media.

Uranium enrichment demonstration plant (UEDP) The uranium enrichment demonstration plant (UEDP) in Ningyo-toge was used to demonstrate uranium enrichment by the gas centrifuge (GCF) method, and was operating continuously from 1988 to 2001. As a result, significant uranium was deposited in the equipment mainly as intermediate uranium fluorides. System chemical decontamination using IF7 gas was proposed as an efficient decontamination method. The secondary waste characteristic of IF7 treatment is IF5 and minor adsorbent. In addition, IF5 is easy to convert to IF7 and re-use for system decontamination. The IF7 treatment technique is performed at room temperature and very low pressures such as 10–45 hPa. Secondary reaction is insignificant in IF7 treatment except for the reaction between IF7 gas and the intermediate uranium fluoride. The weights of uranium deposited in the cascades were approximately 700 kgU per cascade before IF7 treatment. The IF7 treatment period for each cascade is 60 days applying the near-optimal processing conditions. More than 96% of uranium was recovered from the cascade system. As a result, the U radioactivity of the main parts of the GCF fell to 1.0 Bq/g and below.

Plutonium fuel fabrication facility (PFFF)21,22 The plutonium fuel fabrication facility (PFFF) operated from 1972 to 2002 for fabricating MOX fuels for Fugen and the experimental Joyo fast breeder reactor (FBR). The decommissioning and dismantling (D&D) project for the PFFF is divided into the following four phases:

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• Phase 1 (up to 2010): stabilization and shipment of nuclear material in the facility. Choose decontamination and volume reduction techniques. • Phase 2 (2010–2015): D&D planning and adaptability tests. • Phase 3 (2015–2020): size reduction of equipment and glove box. R&D programme carried out. • Phase 4 (2020–2035): re-use of buildings for waste storage. An issue relating to the accumulation of special nuclear material became apparent in the 1990s in this facility. Eight glove boxes in the facility had to be replaced by those with an improved automated fuel fabrication system and residuals recovery system. In order to dismantle these glove boxes, it was necessary to have a more durable containment structure than that of the plastic enclosure, commonly used at the time. To circumvent these issues, the glove box dismantling facility, a centralized decommissioning workshop to dismantle glove boxes, was developed. The purpose of the workshop is to safely dismantle the after-service glove boxes and recover the fuel residuals from the glove boxes. The basic concepts of the workshop are as follows: 1. The workshop has the functionality of a glove box. To prevent the spread of contamination, the level of the internal pressure is kept around 300 Pa in gauge pressure negative to the surrounding room pressure. 2. The workshop is installed in a room in the basement of the plutonium fuel production facility (PFPF) and used for glove box dismantling repeatedly. 3. Remote-controlled devices are installed in the workshop to reduce the radiation dose to which workers are exposed. The activity undertaken was of both remote and hands-on type size reduction. The data and knowledge will be reflected in the planning of the D&D project for PFFF. Technological developments to reduce secondary waste generation are being carried out. Dismantled equipment is cut and wrapped in plastic sheets and packing tape, and stored in 200 L drums. The amount of packaging material (secondary waste) sheets may be about 20% of the volume of dismantled materials. In addition, the packaging activities are performed by workers wearing airline suits. These suits are also secondary wastes. In addition, waste treatment facilities will remove the packaging materials from the dismantled equipment which must then be sorted. To reduce waste treatment work and the amount of secondary waste, a direct in-drum system for RAW management has been developed. The direct in-drum system can be stored directly in a double-skin drum without packaging. In addition, RAW stored in the drums can easily be retrieved

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Radioactive waste management and contaminated site clean-up Direct in-drum system

Glove

Lid of drum

Incombustible (primary waste)

Double cover type drum Radioactive waste is stored in drum. Radioactive waste is retrieved from drum.

23.3 Direct in-drum system.

from them. To prevent the leakage of radioactivity during storage and retrieval of RAW, the lid of the double-skin drum and of the direct in-drum system are connected by a gasket. The direct in-drum system (Fig. 23.3) is attached to the glove to be used for waste storage.

23.5

References

1. Website of Federation of Electric Power Companies of Japan (http://www.fepc. or.jp). 2. Framework of Nuclear Energy Policy, Japan Atomic Energy Commission, October 11, 2005. 3. The Challenges and Directions for Nuclear Energy Policy in Japan – Japan’s Nuclear Energy National Plan, Ministry of Economy, Trade and Industry (METI), December 2006. 4. Upper bounds of concentration of radioactive elements in waste packages, Japan Nuclear Safety Commission, May 2007. 5. Website of Ministry of Economy, Trade and Industry (METI), (http://www. enecho.meti.go.jp/rw/gaiyo03.html). 6. Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste, National Report of Japan for the third Review Meeting, Government of Japan, October 2008. 7. Nagasaki S and Nakayama S (2011), Engineering of Radioactive Waste, Ohmsha, (in Japanese). 8. Akiyama T, Asakura Y, Ando H, Ishihara T, Emura S, Ouchi J, Ohashi H, Ohashi M et al. (1997), Radioactive Waste Management, Japan Atomic Industrial Forum (in Japanese). 9. Ishihara T, Emura S, Kamiyama H, Koizumi T, Suzuki K et al. (1994), Guide Book of Radioactive Waste Management, 1994 edition, Japan Atomic Industrial Forum (in Japanese).

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10. Hayashi M (1992), ‘Generation of Operational Waste in Nuclear Power Reactor and Transition of Treatment’, Radioactive Waste Management Center Topics, No. 22, pp. 1–7. 11. AESJ (1990), Management and Safety of Radioactive Waste, Atomic Energy Society of Japan (in Japanese). 12. JAIF (2005), Management of Radioactive Waste from Nuclear Facilities, Japan Atomic Industrial Forum (in Japanese). 13. Ohuchi J, Miyamoto Y, Ikeda S, Ogata Y, Takeda K, Yokoyama K, Yoshioka M and Tanimoto K (1996), ‘Research and Development on Treatment of Radioactive Waste’, PNC Technical Review, December, No. 100, pp. 215–233. 14. JAEC (1982), ‘The Long-Term Program for Development and Utilization of Nuclear Energy’, Japan Atomic Energy Commission. 15. Government of Japan (2008), ‘Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste’, National Report of Japan for the third Review Meeting, October, Government of Japan. 16. NSC (1985), ‘Basic Philosophy to Assure Safety for the Dismantling Nuclear Reactor Facilities’, Nuclear Safety Commission (revised in August 2001). 17. NES (1997), ‘Aiming at Decommissioning of Commercial Nuclear Power Facilities’, Nuclear Energy Subcommittee, Advisory Committee for Natural Resources and Energy. 18. DSS (2001), ‘Philosophy for Safety Assurance and Safety Regulation on the Decommissioning of Commercial Power Reactor Facilities’, Decommissioning Safety Subcommittee, Nuclear and Industrial Safety Subcommittee, Advisory Committee for Natural Resources and Energy. 19. NSC (2005), ‘An Ideal Safety Regulation System for Nuclear Facilities Post Termination’, Nuclear Safety Commission. 20. OECD (2006), ‘The NEA co-operative programme on decommissioning, a decade of progress’. 21. Kitamura A, Okada T, Kashiro K, Yoshino M and Hiroshi H (2007), ‘Waste Handling Activities in Glovebox Dismantling Facility’, Proceedings of Global 2007, 531–536, Idaho, September 9–13. 22. Iemura K, Nakai K, Watahiki M, Kitamura A, Kazunori S and Aoki Y (2011), ‘Decommissioning Plutonium Fuel Fabrication Facility’, Journal of the RANDEC, 2–9, 43 (in Japanese).

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Tokyo Electric Power Co. – Kashiwazaki Kariwa 1

2

3

4

5

6

Electric Power Development Co. – Ohma 7

Houkuriku Electric Power Co. – Shika 1

2

The Japan Atomic Power Co. – Tsuruga 1

2

3

Hokkaido Electric Power Co. – Tomari

Tohoku Electric Power 1 Co. – Higashidori

1

Tohoku Electric 1 Power Co. – Maki

3

1

2

3

Tokyo Electric Power Co. – Fukushima Daiichi

4

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1 2 3 6 4 5 Tokyo Electric Power Co. – Fukushima Daini

The Kansai Electric Power Co. – Mihama 3 1 2 The Kansai Electric Power Co. – Chi 3 4 1 2

1

2

3

4

The Japan Atomic Power Co. – Tokai Closed (Mar.1998)

The Kansai Electric Power 1 2 3 4 Co. – Takahama The Chugoku Electric Power Co. – Shimane 3 1 2 The Chugoku Electric Power Co. – Kaminoseki 1 2 Kyushu Electric Power Co. – Genkai 3 4 1 2

2

Tohoku Electric Power Co. – Onagawa

The Japan Atomic Power Co. – Tokai Daini

Chubu Electric Power Co. – Hamaoka 1

2

3

4

5

Shikoku Electric Power Co. – lkata

Kyushu Electric Power Co. – Sendai 1 2

1

2

3

Number of Units Total Output (MW)

Output scale Under 500 MW Under 1 000 MW Over 1 000 MW

Operaling Under construction Preparing for construction

Operational

Under construction Preparing for construction Total

52 3 8 63

Plate VIII (Chapter 23) Map of Japan showing location of its NPP fleet (prior to shutdowns caused by Fukushima).

45.742 3.838 10.315 59.895