放射性岩棉玻璃固化过程中的关键工艺参数

针对放射性岩棉的玻璃固化配方,分别进行了高温粘度及低温粘度研究,对按照优化工艺参数制得的放射性岩棉玻璃固化体进行性能验证与评价。结果表明:优化配方玻璃熔融体高温粘度曲线方程为η=1.27×10-8×e29 794.11/T,相关系数达到0.999 1,预测优选熔化温度为1 181℃、成型操作前期温度范围为1 034~914℃、成型操作后期温度范围为914~619℃;优选退火温度范围为544~574℃;按照优化工艺参数制得的玻璃体均匀性好,密度满足玻璃固化体要求,玻璃化程度高,机械强度较高,表明研究所得的工艺参数为适用于放射性岩棉配方的优化结果,为等离子体高温焚烧装置的优化设计及放射性岩棉玻璃固化配方的工程应用提供了参考依据。     

Vitrification of a simulator of Savannah River site (USA) wastes with high iron and aluminum content on bench and commercial facilities with a cold crucible

Bench and commercial-facility experiments have been performed on cold-crucible vitrification of a simulator of high-level wastes from the Savannah River site (USA). The wastes contained up to 29 mass% Fe₂O₃ and 26 mass% Al₂O₃. The specific product flow reached 1700 and 2450 kg/(m²·day) with specific energy consumption 14–16 and 9–10 kW·h/kg, respectively. The crucibles did not undergo any appreciable corrosion during the period of the work performed and are reusable. The product consisted of a borosilicate matrix, containing up to 10 vol.% crystalline phase of spinel. The method of induction melting in a cold crucible is especially effective for crucibles with a large diameter, since the specific productivity increases and the specific energy consumption on the vitrification of high-level wastes decreases. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 291–295, May, 2008.     

Synthesis and Use of Borates of Polyatomic Alcohols for the Preparation of Liquid High-Level Wastes for Inclusion into Borophosphate Glass

Boro-organic reagents suitable for fluxing a solution of liquid high-level wastes in order to immobilize them in borophosphate glass are synthesized by dissolving boric acid and sodium tetraborate in ethylene glycol and glycerin. The properties of the reagents obtained are investigated, borophosphate glasses are synthesized, and it is shown that the vitrification process using them presents no danger of an explosion. A technological scheme for preparing liquid high-level wastes for inclusion in borophosphate glass is proposed. __________ Translated from Atomnaya Energiya, Vol. 99, No. 4, pp. 276–282, October, 2005.     

Heat transfer in high-level waste management

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed.     

Compaction of Radioactive Thermal-Insulation Materials and Construction Debris by Melting in a Cold Crucible

A technology is proposed for reprocessing radioactive thermal-insulation materials and construction debris, produced at nuclear power plants and radiochemical facilities, by melting the wastes in a induction furnace with a cold crucible. The results show that the final volume of the thermal-insulation wastes can be decreased by a factor of 40 and that of the construction debris by a factor of 2.5. The high hydrolytic stability of the final materials produced as a result of melting (the ¹³⁵Cs and ⁹⁰Sr leach rate is less than 1·10⁻⁷ g/(cm²·day) allows these wastes subsequently to be stored and buried. Recommendations are developed for an apparatus-technological scheme and for electrotechnical equipment for a commercial facility. __________ Translated from Atomnaya Energiya, Vol. 99, No. 3, pp. 167–171, September, 2005.     

低放废物近地表岩洞处置实践

利用辽宁兴城退役铀矿山废矿井的＋43m中段处置退役的镭厂低放废物.对该处置设施基本情况、需处置的废物来源、整备设施及运输、废物处理处置方法及环境影响进行了分析.该处置设施已接受8 747m3废物、尚有2 000m3的容量空间,待剩余空间接受其他来源的废物后实施处置场封闭.     

Investigation of the behavior of uranium,plutonium, and radiogenic americium in glassy compositions during localization of model high-level liquid wastes

Conclusions The experiments showed that for solidification of α-containing high-level wastes, to increase the incorporation of plutonium and americium into the phosphate and borosilicate matrices, it is necessary to develop a technological regime for conducting the vitrification that would reduce substantially the residence time of the melt in the high-temperature region, that would ensure the presence of strong convection fluxes, which would not permit sedimentation of the heterogeneous phases formed in the melt (or the possibility of their periodic removal), in the melter and thereby would guarantee homogeneity of the melt of the compositions obtained throughout the entire technological process. Apparently, the most promising technological process, which largely meets the requirements listed, is vitrification of wastes in IPKhT-type melters (induction melter with a cold crucible), which is now being actively studied. State Science Center of the Russian Federation, A. A. Bochvar All-Union Scientific-Research Institute of Standardization in Machine Engineering. Translated from Atomnaya énergiya, Vol. 79, No. 2, August, 1995.     

Durability of a reinforced concrete designed for the construction of an intermediate-level radioactive waste disposal facility

The National Atomic Energy Commission of the Argentine Republic is developing a nuclear waste disposal management programme that contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The repository is based on the use of multiple, independent and redundant barriers. The major components are made in reinforced concrete so, the durability of these structures is an important aspect for the facility integrity. This work presents an investigation performed on a reinforced concrete specifically designed for this purpose, to predict the service life of the intermediate level radioactive waste disposal facility from data obtained with several techniques. Results obtained with corrosion sensors embedded in a concrete prototype are also included. The information obtained will be used for the final design of the facility in order to guarantee a service life more or equal than the foreseen durability for this type of facilities.     

Trial Operation of the Advanced Volume Reduction Facilities for LLW at JAEA

The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level radioactive solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former has cutting installations for large size wastes and the latter has melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750m³ and the volume reduction ratio is from 1.7 to 3.7. The WVRF has been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005.     

高放废物玻璃固化体的辐照效应研究进展

本文概述了国内外针对玻璃辐照效应所开展的研究及其进展。介绍了α、β、γ射线在玻璃中产生的微观缺陷类型、相对应的表征手段、不同微观损伤与吸收剂量的关系及其在不同温度条件下的稳定性。介绍了玻璃宏观性质随辐照剂量的改变以及饱和现象,提出了宏观性质的改变和无序度的联系来解释实验结果。最后,介绍了玻璃固化体研究所面临的问题和挑战。     

1980 National Symposium on Atomic Energy,Tokyo, Japan

Reviewed is the as-titled conference, held in October 23 to 28, 1989, in Kyoto, Japan, sponsored by American Society of Mechanical Engineers, Japan Society of Mechanical Engineers, and Atomic Energy Society of Japan, with other co-sponsoring societies. In the Conference, all the aspects of radioactive waste management were discussed, including policies and general strategies for waste management, solidification and incineration of low-, and intermediate-level wastes (LLW and ILW), shallow-land burial disposal of LLW/ILW and its safety assessment, decommissioning, storage and transportation of spent fuels, vitrification of high-level wastes (HLW), deep geologic disposal of HLW and its safety assessment by natural analogs, and partitioning and transmutation. The significance of the Conference is found in that the informations of Japan's waste management activities, which do not always reach audience in other part of the world, were conveyed in detail in the Conference. More papers were presented for LLW/ILW than for HLW, which might reflect the fact that there are still more activities for LLW/ILW than for HLW in Japan.     

Properties of phosphate glass waste forms containing fluorides from a molten salt reactor

Radioactive fluoride wastes are generated during the operation of molten salt reactors(MSRs) and reprocessing of their spent fuel.Immobilization of these wastes in borosilicate glass is not feasible because of the very low solubility of fluorides in this host.Alternative candidates are thus an active topic of research including phosphatebased glasses,crystalline ceramics,and hybrid glass-ceramic systems.In this study,mixed fluorides were employed as simulated MSRs waste and incorporated into sodium aluminophosphate glass to obtain phosphate-based waste form.These waste forms were characterized by X-ray diffraction,Raman spectroscopy,and scanning electron microscopy.Leaching tests were performed in deionized water using the product consistency test A method.This study demonstrates that up to 20 mol%of simulated radioactive waste can be introduced into the NaA1 P glass matrix,and the chemical durability is much better than that of borosilicate.The addition of Fe_2O_3 in the NaAlP glass matrix results in increases of the chemical durability at the expense of fluoride loading(to 6.4 mol%).Phosphate glass vitrification of radioactive waste containing fluorides is a potential method to treat and dispose of MSR wastes.     

Epoxy Compounds for Immobilizing Radioactive Wastes

The radiation resistance of epoxy compounds, solidified by crystalline hardening agents - metaphenyl-enediamine and phthalic anhydride - is investigated. It is shown that under conditions of γ irradiation (E = 1.33 MeV) the temperature of vitrification of the compounds depends on the dose (doses up to 1500 Mrad were investigated) and temperature 20–160°C, and the radiation gas release depends on the vitrification temperature and irradiation. It is shown that epoxy-phthalic anhydride compound is best as a matrix for immobilizing solid radioactive wastes at high temperatures (80–130°C).The experiments showed that the proposed compound can be recommended for immobilizing solid radioactive wastes.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 348–351, May 2005.     

Vitrification of high-level radioactive waste considering the behavior of platinum group metals

To demonstrate a method using liquid metal for the removal of PGMs during the vitrification process of high-level radioactive waste, removal of Pd was performed using Cu from molten glass containing fission products such as Nd, Sr, Zr, Mo, Te and Ni. Almost all the Pd, 93%, was extracted into liquid Cu and removed as a separable Cu–Pd metal button from the glass. Tellurium and Ni were also extracted 42 and 5.6%, respectively. Nearly 100% of the other elements, especially the heat generating elements such as Sr and Cs, for which Na was used as a substitute of Cs remained in the glass.     

Characteristics of Cast Stone cementitious waste form for immobilization of secondary wastes from vitrification process

The high-temperature in vitrification process of radioactive wastes could cause radioactive technetium (⁹⁹Tc) in secondary liquid wastes to become volatile. Solidified cementitious waste forms at low temperature were developed to immobilize radioactive secondary waste. This research focuses on the characterization of a cementitious waste form called Cast Stone. Properties including compressive strength, surface area, phase composition, and technetium leaching were measured. The results indicate that technetium diffusivity is affected by simulant type. Additionally, ettringite and AFm (Al₂O₃–Fe₂O₃–mono) main crystalline phases were formed during hydration. The Cast Stone waste form passed the qualification requirements for a secondary waste form, which are compressive strength of 3.45 MPa and technetium diffusivity of 10^?9 cm²/s. Cast Stone was found to be a good candidate for immobilizing secondary waste streams.     

A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na₈(AlSiO₄)₆Cl₂, chlorapatite, Ca₅(PO₄)₃Cl, and spodiosite, Ca₂(PO₄)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700-800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.     

德国放射性废物地质处置的研究历程

本文简要介绍了德国放射性废物地质处置及相关研究的历程和现状,包括中低放废物和高放废物的处置情况,同时从技术层面分析了德国高放废物处置库场址评价所面临的问题。希望对于我国放射性废物地质处置的研究有所启示。     

Radioactive waste management practices in India

Concern for the environment and establishment of radiation protection goals have been among the major priorities in planning of India's nuclear energy programme. In the Indian nuclear fuel cycle, right from inception, a closed loop option has been adopted where spent fuel is reprocessed to recover plutonium and unused uranium. The emphasis has been to recover actinides, individual fission products and recycle them back to the fuel cycle or use them for various industrial applications. The development of innovative treatment processes for low and intermediate level wastes in recent times has focused on volume reduction as one of the main objectives. In the case of high-level liquid waste, vitrification in borosilicate matrix is being practiced using induction heated metallic melters at industrial scale plants at Tarapur and Trombay.Currently, there are seven operating near surface disposal facilities co-located with power/research reactors in various parts of the country for disposal of low and intermediate level solid wastes. These are routinely subjected to monitoring and safety/performance assessment. An interim storage facility is operational for the storage of vitrified high-level waste overpacks for 30 years or more. Nation wide screening of potential regions and evaluation of rock mass characteristics is in progress for ongoing geological repository programme. Preliminary design and layout of an underground research laboratory/repository has also been initiated.A research programme is underway for long-term evaluation of vitrified waste product under simulated repository conditions. Research is also directed towards development of advanced technologies for waste processing as well as conditioning in vitreous and ceramic matrices. The Department of Atomic Energy with participation of the Indian industry has developed all essential remote-handling gadgets required for operation and maintenance of waste management system and assemblies including decommissioning.     

陶瓷电熔炉在动力堆高放废液玻璃固化中适用性分析

随着我国核电事业发展和核燃料循环体系日益完善,玻璃固化动力堆高放废液的需求已提上日程。为探讨陶瓷电熔炉技术在我国后续动力堆高放废液玻璃固化项目中的适用性,本文从源项、熔炉技术特点和熔炉更换解体三个方面进行了分析,认为陶瓷电熔炉技术可以用于玻璃固化动力堆产生的高放废液。     

加速器质谱与环境科学

介绍了近年来国内外用加速器质谱计在环境科学方面的研究工作,包括测量城市空气污染、监测来自核设施的泄漏和有关全球环境的测定。为了适应多学科应用研究的需要,经过对装置进行大规模的改造和对高精度14^C测年方法的研究,北京大学加速器质谱计^14C测量精度已经达到好于0．5％的水平。在国家重大项目“夏商周断工程”中,测定了大量样品并结合考古已经达到好于0．5％的水平。在国家重大项目“夏商周断代工程”中,测定了大量样品并结合考古信息给出了比较可靠的年表。