Developing Technologies for Nuclear Fuel Cycles

Transcription

1 Developing Technologies for Nuclear Fuel Cycles Hitachi Review Vol. 48 (1999), No Radioactive Waste Treatment and Spent Fuel Storage Makoto Kikuchi, Ph.D. Takuma Yoshida Masami Matsuda, Ph.D. Hidetoshi Kanai OVERVIEW: A nuclear fuel cycle must be established so that uranium resources are efficiently utilized and electricity supplies are guaranteed in the future. Hitachi has been developing nuclear technologies for about 40 years, including advanced treatment technology for the nuclear fuel cycle in commercial nuclear power plants. This paper describes the new treatment as well as new s for spent fuel storage and reprocessing treatment. INTRODUCTION THE establishment of a nuclear fuel cycle is one of the most important issues from the viewpoints of longterm energy supply with high reliability, reduction of environmental impact due to radioactive disposal, and cost reduction for nuclear power generation. As a nuclear engineering and manufacturing company, Hitachi has been promoting research and development of new technologies for the nuclear fuel cycle. These technologies cover not only treatment and disposal of radioactive s from nuclear power plants but also uranium enrichment, spent fuel storage, spent fuel reprocessing, and decommissioning of old plants. These technologies will let Hitachi provide a Total Solution System to support safe, economical and environmentally clean nuclear power. Total Solution System When developing a Total Solution System, we must satisfy three important points for the nuclear fuel cycle technologies. They are high reliability, minimum load on the environment, and low cost as shown in Fig. 1. Hitachi s research and development program is managed with consideration of the longterm project plan of the Japanese government as well as international trends. Close cooperation with utilities, universities and other research institutes is also important in order to obtain the best practical results. The most important subject in nuclear technologies is always reliability and Hitachi gives it full consideration. The technologies developed by our research laboratories are tested in the Hitachi Nuclear Fuel Cycle Engineering Center, where full-scale demonstration plants are used for various tests. Several Uranium enrichment/ fuel fabrication LWR LLW disposal Total solution High reliability Minimum load on environment Low cost Spent fuel storage FBR : fast breeder reactor LLW: low level Uranium mining/ ore processing Fig. 1 Outline of Nuclear Fuel Cycle. Aiming at a total solution for nuclear fuel cycle. FBR cycle Fuel reprocessing TRU disposal LWR: light water reactor TRU : transuranium Fig. 2 Nuclear Fuel Cycle Engineering Center. Performance is demonstrated by full-scale pilot plants.

2 Developing Technologies for Nuclear Fuel Cycles 278 TABLE 1. Outline of R&D for Nuclear Fuel Cycle Technologies Fields Uranium enrichment Developing technology Laser uranium enrichment Notes Alternative technology LLW from NPS High-performance cement Continuous melting Laundry drain treatment by helical filter Environmental load reduction Economical improvement Spent fuel interium storage High-density spent fuel storage rack Metal cask Vault-type storage Increasing the storage capacity of operating plant Cost reduction of large interim storage facility Reprocessing Ion exchange process Advanced PUREX process Cl-gas free pyrochemical process Significant cost reduction from previous LLW: low-level NPS: nuclear power station TRU: transuranium PUREX: plutonium uranium reduction extraction TRU Decontamination & dismantling (D&D) TRU treatment & disposal Uranium-contaminated treatment System engineering Decontamination & inspection Recycle technology Stable solidification of long-half-life nuclides Remediation of contaminated facilities Significant cost reduction of D&D Application of construction technology s are installed in this engineering center as shown in Fig. 2. After the full-scale tests, an actual treatment test at a nuclear facility is normally carried out in order to clarify the performance under actual conditions. The s installed in the engineering center are kept for more than ten years so they can be modified and improved when the characteristics change after a few years operation. They are also useful for maintenance training. Outline of Developed Technologies Table 1 summarizes the nuclear fuel cycle technologies that Hitachi has developed. Emphasis has recently been placed on from spent-fuel reprocessing including TRU (transuranium) and decommissioning of old plants. This paper describes new technologies including a treatment for nuclear power plant, a spent-fuel storage, and a reprocessing treatment. WASTE TREATMENT FOR NUCLEAR POWER PLANTS Hitachi has been developing treatment technologies since about 1960 and has introduced many advanced s in power plants. Fig. 3 summarizes the concepts Hitachi has used. The fundamental goals in developing a radioactive treatment are (1) to reduce radioactivity release into the environment, (2) to reduce generation, (3) to reduce generated volume, (4) to obtain a stable package for final disposal, and (5) to provide a simple. In addition to these factors, it is also important to reduce total cost, including equipment and final Radioactivity release reduction Silver-alumina iodine filter Laundry drain helical filter Waste generation reduction Non-precoat type hollow fiber filter Non-regenerative demineralizer Establishment of rad treatment Volume reduction/stable packing High-performance cement Pelletizing/cement glass Continuous melting Radiation exposure reduction Remote control Advanced REDOX decontamination Final disposal Drum inspection Safety analysis Fig. 3 Concept of R&D for Rad Treatment Technologies. Establishment of total solution with reduction of radioactivity release or volume reduction, and final disposal. REDOX: reduction oxidation process Ensuring reliability Reduction of environmental load Economical improvement

3 Hitachi Review Vol. 48 (1999), No disposal fees. In order to achieve these goals, several new technologies and new s have been developed. These include a drying and pelletizing for volume reduction and a hollow fiber filter for liquid treatment and for non-regeneration of condensate demineralizer. Slim Rad System 1) The new radioactive treatment has been in operation since 1996 at the site of the first ABWR (advanced boiling water reactor). Slim Rad is a total which consists of off-gas, liquid and solid treatment subs as well as a package inspection sub for final disposal. Slim Rad was designed by utilizing our operating experience with the ABWR, upgrading component capability, optimizing the overall treatment, and applying newly developed technologies. An outline of the is shown in Fig. 4. The main features of Slim Rad are listed below. (1) Optimized : The tank volume and treatment capacity are about 60% those of the previous. (2) Waste generation and volume reduction: Average generation per plant (1,350 MWe) is about 100 drum/year, which is 1/8 that of the previous plant (Fig. 5). (3) New technologies: High-performance cement solidification is applied to solidify all s such as liquid, incinerated ash, and miscellaneous solid. High volume reduction efficiency and low radioactivity release rate to the environment are obtained. A helical filter treats laundry and shower drain. This compact filter can remove not only radioactive species but also organic contaminants such as detergent and oil. A sponge metal catalyst, instead of an alumina catalyst, is applied to the off-gas recombiner unit and the unit size is reduced to 1/3 the size of the previous unit. The package inspection is applied, as Recombiner Charcoal bed Off-gas Boron rack Iodine filter Stack Solid Low conductivity liquid High conductivity liquid Hot shower drain Laundry drain Spent fuel pool Sludge Non-combustible Hollow fiber filter Tank vent Sludge Evaporator Conc. Combustible Helical filter Spent charcoal Non-regenerative demineralizer Compactor Spent resin Concentrated liquid Auto sampling Sample tank Spent resin Incinerator Heater-drain pump-up Spent resin Sample tank Incinerator Hollow fiber filter Sludge Turbine Condenser High-performance cement Inspection Re-use Disposal Discharge : Waste : Applied technology Fig. 4 Outline of Slim Rad System. Slim Rad is a total which consists of off-gas, liquid and solid treatment subs, as well as the package inspection sub for final disposal.

4 Developing Technologies for Nuclear Fuel Cycles 280 Amount of drums (drums/year plant) 1, (About 800) Non-combustible Bead resin Powdered resin Concentrated Previous plant Non-precoat filter (Hollow fiber filter) Resin incineration Combustible Non-regeneration of demineralizer Super compactor (About 400) ABWR (About 100) Slim Rad Fig. 5 Waste Generation Reduction by Slim Rad System. Average generation per plant (1,350 MWe) is about 100 drums/year, which is 1/8 that of the previous plant. an option, for final disposal. This compact and fully automatic measures such items as radioactivity concentration, compressive strength, and surface contamination, which must be measured according to disposal criteria. High-performance Cement Solidification2, 3) The most important technology in the rad treatment is the volume reduction sub. Figs. 6 and 7 outline the concept and the technique of the new solidification process. One simple, similar to a conventional cement solidification unit, can treat liquid, spent ion exchange resins, incinerated ash, and miscellaneous solid, which are the major components of low and intermediate level radioactive generated in nuclear power plants. Our developed technique not only efficiently reduces large volumes, but also provides long-term stability of the final packages for land disposal. And it is implemented by using a simple treatment. In the present research work we aimed at developing an advanced solidification using a new cement so as to increase loading and decrease radioactivity release into the environment. Features of high-performance cement High-performance cement consists of slag cement, reinforcing fiber, natural zeolite, and lithium nitrate (LiNO 3 ). The fiber allows loading to be increased from 20- to 55-kg dry resin/200 L. The natural zeolite, whose main constituent is clinoptilolite, reduces cesium leachability from the form by 90%. Lithium nitrate prevents alkaline corrosion of the aluminum, contained in ash and miscellaneous solid, and reduces hydrogen gas generation. (1) High loading The spent resin content in a cementitious form is typically kept below 20-kg dry resin/200 L because the form tends to swell and crack in water when resin content is higher. Fundamental experiments to quantitatively examine the deterioration mechanism were performed. The form was found to crack in water when the swelling pressure of the ion exchange resin exceeded the tensile strength of the cement. It was also confirmed that tensile strength could be doubled by adding carbon fiber. Applicable to various s Meeting disposal criteria Cost reduction Meeting disposal criteria Concept One applicable for all kinds of Simple High volume reduction Stable form Developed Centralized volume reduction High performance cement Fig. 6 Concept of Advanced Solidification System. Advanced solidification consists of two newly developed s, centralized volume reduction and high performance cement. Liquid Spent resin Incinerated ash Miscellaneous solid Liquid/solid separator Mixer 200 L drum High-performance cement Water Inspection Simple Land disposal (Rokkasho) Fig. 7 High-performance Cement Solidification. Single facility can be applied to various s. Also this satisfies disposal criteria and volume reduction requirements.

5 Hitachi Review Vol. 48 (1999), No These findings led to the development of fiber reinforced cement, which allows the loading to be increased from 20- to 55-kg/200 L. (2) Prevention of alkaline corrosion of aluminum Solidification of incinerated ash and miscellaneous solid presents a unique problem. These s include aluminum materials which are vulnerable to corrosion under highly alkaline conditions. This corrosion is accompanied by hydrogen gas generation in the cement paste. This hydrogen results in lowered uniformity and reduced compressive strength of the forms. It is therefore necessary to develop a treatment which prevents alkaline corrosion of the aluminum and reduces hydrogen gas generation. After surveying many inhibitors, we found that addition of lithium nitrate (LiNO 3 ) forms a Li-Al double salt on the aluminum surface and, thus stops aluminum corrosion and hydrogen gas generation. (3) Low leachability When is in a cementitious form, leachability of radiocesium is relatively high compared with other radioactive nuclides such as Co-60, Sr-90, and C-14. Cs adsorbents were surveyed and a natural zeolite, whose main ingredient is clinoptiolite, was selected. The Cs leachability was reduced by 90% by adding the zeolite to cement. The leachability index was thus improved from 9.3 to Continuous Melting System In order to more efficiently reduce the volume of miscellaneous non-combustible such as metallic pipes and insulators, several types of melting s using plasma and induction heating have been developed. In particular, we have developed a continuous melting which is more compact and more efficient. The original had a capacity of 4 ton/hour and was used for recycling scrap metal from discarded home appliances. Fig. 8 shows an outline of the. It contains graphite (about 10-cm diameter rods) which is heated to about 1,500 C indirectly by an induction coil surrounding the furnace. The is fed continuously from the top of the furnace and is melted by the heat of the graphite, then the molten material flows down from the bottom outlet. The features of this are as follows: (1) Insulators and concrete can not be melted by a normal induction heating furnace. However, in this, indirect heating from the graphite is able to melt these s. (2) Since the heating efficiency is about 50%, which Miscellaneous Induction heating coil Melter Refractory Graphite Outlet (heating medium) Receiver Off-gas Secondary combustor Ceramic filter High efficiency particulate air filter Stack Fig. 8 Continuous Melting System. Continuous melting was developed for non-combustible. is twice that of other methods, treatment speed is higher and the is more compact. (3) The major problem of the melting is the life of the refractory. Since most of the molten runs through the graphite without contacting the refractory, the life of the refractory is expected to be one year. (4) In this, little molten stays in the furnace, so steam explosion is unlikely to occur due to containing water or organic materials. Helical Filter for Laundry Waste Treatment Laundry produces the largest amount of liquid from nuclear power plants. It has very low radioactivity but contains organic contaminants Laundry Discharge to ocean Laundry drain Filtrate Activated charcoal Charcoal treatment vessel Helical filter Spent charcoal Incinerator Fig. 9 Laundry Drain Treatment System. In this laundry drain treatment, neither radioactive materials nor COD materials, such as detergents are discharged to the ocean.

6 Developing Technologies for Nuclear Fuel Cycles 282 such as detergents and oil. Towards our goal of a compact filter with large capacity, we have developed a new filtering with a helical filter. The is outlined in Fig. 9. Activated charcoal is added (0.1% concentration) to the laundry tank. And this charcoal adsorbs the detergents and radioactive material. The mixture is filtered by the helical filter and the filtrate is discharged to the ocean. At the end of the operation, activated charcoal is dried by steam (< 30 wt% moisture content) and sent to an incinerator for final volume reduction. This helical filter provides the two functions of filtering and drying in the same unit, so the is more compact. SPENT FUEL STORAGE SYSTEM 4) The spent fuels routinely generated from nuclear power plants are tentatively stored on site before reprocessing; this is called interim storage. Because of delays to the operation schedule of new reprocessing plants in Japan, a centralized large-scale interim storage is under consideration. As a safe and efficient storage, we have developed a high-density boron rack, which increases the storage capacity of existing pools on site. A dry storage using metal casks and a vault is also available for on-site and offsite storage. New Boron Rack Most of the spent fuel is currently stored in the pool of a nuclear power plant. But because more spent fuel has recently been generated, nuclear power plants have had to increase the storage capacity of their pools by using a high-density rack. In order to meet such demands, Hitachi has developed a new boron rack, which has a 40% more capacity than a conventional stainless steel rack. Since boron adsorbs neutrons from spent fuel, higher density storage is possible, while avoiding criticality. Fig. 10 shows the new boron rack. A new stainless steel, which contains about 1% boron, was developed for the rack. Material tests as well as performance tests including seismic analysis were performed. This new boron rack features not only higher capacity, but also lighter weight and better seismic behavior; thus, it can be fitted to the existing pools in both BWR and PWR plants easily. Metal Cask An outline of the metal cask is shown in Fig. 11. The primary functions of the cask are to act as a radiation shield and to confine radioactivity. The casks Fig. 10 High-density Spent Fuel Storage Rack. By using stainless steel with added boron, storage density of spent fuel can be increased. 2nd lid 1st lid Basket Main body (Forging metal, γ-ray shielding) Upper trunnion (for hanging) Support rod Heat plate Neutron shielding (resin) Outer case Lower trunnion (for emplacement) Fig. 11 Metal Cask. Metal cask contains the spent fuel in a dry condition, and shields γ rays and neutrons by its main body. containing spent fuels are arranged inside the storage building and the heat generated by radionuclide decay is removed by natural convection. The benefit of the cask is that storage capacity can be increased gradually as more casks are added, so the is very economical and relatively small. High-efficiency Vault A high-efficiency vault storage was developed in order to reduce costs of a large-scale storage facility. Figs. 12 and 13 outline the highefficiency vault storage. It consists of a canister which contains about 40 spent fuel bundles, a storage tube which contains canisters in two layers, a shielding lid on the top of the storage tube, and a canister

7 Hitachi Review Vol. 48 (1999), No Shielding lid Storage tube Canister (2 layers) Performance comparison (in case of 3,000 ton-uranium scale facility) Item This Previous Storage density (ton/m 2 ) Capacity (SF/can) Site area (m 2 ) , ,700 Fig. 13 Plastic Model of Vault-type Storage Facility. High-efficiency vault storage was developed to achieve better cost reduction for large-scale storage facilities. Fig. 12 High-performance Vault-type Storage Facility. By adopting 2 layered canisters, storage density and handling can be improved. which is below the limit required for building concrete (65 C), even in a canister containing 40 bundles of high burn-up spent fuel. handling machine. Decay heat is continuously removed by natural convection using a stack. The canister storage room is divided into two layers by the middle roof in order to improve homogeneity of cooling air and to make the structure earthquake-proof. Cooling air is taken from the inlet and discharged from the stack after it has horizontally flowed through the storage tubes. The storage has the following features: (1) A large canister with two-layer storage in the storage tube increases storage density more than 50% (Fig. 12) and reduces canister handling work. (2) Two-layer containment by the canister and storage tube prevents radioactivity leakage and provides corrosion resistance from salt in the air. (3) The module structure allows the storage capacity to be easily increased by simply adding another module unit consisting of a storage tube and a stack. The initially installed handling machine is used for the expanded part. This module-type reduces the initial investment and provides flexibility for future capacity expansion. The most important function of this is to efficiently remove heat. So both simulation and experiments using a large-scale pilot plant (1/5 actual plant size) were used to optimize the design parameters such as density and arrangement pattern of receiving vessels. Fig. 14 shows typical simulation results of cooling air flow and temperature distribution. The maximum temperature in the storage room is 54 C, REPROCESSING PLANT WASTES We have recently developed treatment and disposal techniques for s not only from nuclear power plant but also from nuclear facilities such as reprocessing plant and uranium handling facilities. These s contain long half-life radioisotopes, while primary isotopes contained in power plant are cobalt-60 and cesium-137. As one example of treatment for a reprocessing plant, lowtemperature vitrification of radioiodine is described here. Spent iodine adsorbent is generated from the dissolver off-gas cleaning in the spent fuel reprocessing plant. Radioiodine (I-129) is one of the most important radionuclides in the safety assessment Temperature [ C] 60 Fig. 14 Heat Removal Simulation of High-performance Vault Storage Facility. Cooling air flow and temperature distribution at real size has been simulated. The result indicates sufficient heat removal capacity

8 AgI Spent iodine adsorbent AgI separation Silica gel Silver phosphate Developing Technologies for Nuclear Fuel Cycles 284 CONCLUSIONS Some examples of new technologies related to the nuclear fuel cycle have been reported. Implementation of the nuclear fuel cycle in the electrical power industry will become more important, and Hitachi will continue to devote more effort to develop new technologies to offer the total solution. Melting Waste form Fig. 15 Outline of Low Vitrification Process Using New Glass System. The new glass can contain iodine homogeneously as one of the glass formation materials by low-temperature vitrification. of disposal because it has a long half life (10 7 years) and low distribution coefficient in geological materials. In order to safely dispose s containing radioiodine, an encapsulation technique which keeps the iodine release rate from the repository as low as possible must be developed. Low-temperature Vitrification 5) Although a vitrification process using borosilicate glass is one of the most popular processes for producing low leachable forms, its application to radioiodine s would result in volatilization of iodine because of the high melting temperature (above 1,000 C) of borosilicate glass. It is also difficult to ensure long-term safety regarding from cracking because their glass matrixes do not chemically bond with iodine. The AgI-Ag 2 O-P 2 O 5 glass, which was developed as a new solid electrolyte, can contain iodine homogeneously as one of the glass formation materials. And its melting temperature is relatively low (below 400 C). A flow chart of the developed vitrification process using the new glass is shown in Fig. 15. The process consists of two steps: (1) separation of radioiodine as AgI from spent iodine adsorbent and (2) vitrification of the separated AgI with silver phosphate. We have demonstrated that the new glass can homogeneously contain up to 60 mol% AgI, leading to high volume reduction efficiency (approximately 1/25 of spent iodine adsorbent), without volatilization of iodine. The leach rate of the product is in the order of 10-8 to 10-9 g/cm 2 /d, which is as low as that of borosilicate glass. REFERENCES (1) M. Kikuchi et al., Advanced Radioactive Waste Treatment System SLIM RAD, Proc. of Internat. Symp. on Radiation Safety Management, Taejon, Korea (1997), pp (2) M. Kikuchi et al., Advanced Solidification System Using High Performance Cement, Proc. of Radioactive Waste Management and Environmental Remediation - ASME 1995, pp (3) M. Matsuda et al., Solidification of Spent Ion Exchange Resin Using New Cementitious Materials (I), J. Nucl. Sci. Technol. 29, (1992) p (4) M. Oda et al., Heat Removal Experiment of the High Performance Vault Type Dry Storage System for Nuclear Spent Fuel, Proc. of WM 99, Tucson, USA (1999) (5) T. Nishi et al., Low Temperature Vitrification of Radioiodine using AgI-Ag 2 O-P 2 O 5 Glass System, Proc. of MRS 98, Boston, USA (1998), to be published. ABOUT THE AUTHORS Makoto Kikuchi Joined Hitachi, Ltd. in 1975, and now works at the Nuclear Systems Division. He is currently engaged in the R&D of nuclear fuel cycle technology. Dr. Kikuchi is a member of Atomic Energy Society of Japan, Chemical Society of Japan and the Society of Chemical Engineering of Japan, and can be reached by at kikuchi@cm.hitachi.hitachi.co.jp. Takuma Yoshida Joined Hitachi, Ltd. in 1986, and now works at the Nuclear Systems Division. He is currently engaged in the R&D of radioactive management technology. Mr. Yoshida can be reached by at ta_yoshida@cm.hitachi.hitachi.co.jp. Masami Matsuda Joined Hitachi, Ltd. in 1979, and now works at the Power & Industrial Systems R&D Laboratory. He is currently engaged in the R&D of nuclear fuel cycle technology. Dr. Matsuda is a member of the Atomic Energy Society of Japan and can be reached by e- mail at m_matsuda@cm.hitachi.hitachi.co.jp. Hidetoshi Kanai Joined Hitachi, Ltd. in 1981, and now works at the Nuclear Systems Division. He is currently engaged in the design of spent fuel storage. Mr. Kanai can be reached by at kanai@cm.hitachi.hitachi.co.jp.