模拟非α中低放废液大体积浇注水泥固化的可行性研究

本文对在我国高放废液分离流程中产生的中低放非α废液的大体积浇注水泥固化,在实验室小试条件下进行了可行性研究。结果表明,经浓缩、脱硝及中和处理后,非低α中放废液可用大体积浇注水泥固化进行处理和处置。推荐的固化配方为：水灰比１．０￣１．１,盐灰比０．１８￣０．２０,及占水泥重量０．５％（ｗｔ）的水泥高效缓凝减水剂。固化体第４２ｄＣｓ的浸出率为１０＾－３￣１０＾－５ｃｍ／ｄ,Ｓｒ的浸出率在３．１×１０＾     

高放废液碱矿渣水泥固化体研究

介绍了用碱矿渣水泥固化高放废液新工艺。该工艺以碱矿渣水泥为基体,掺加适量沸石和硅灰,无需加温加压,利用静态爆破剂水化时产生的膨胀压,在限容下使水泥基体致密。其抗压强度、耐热性和浸出率等性能均优于其它水泥。当废物氧化物包容量为25％时,固化体抗压强度可达65～100MPa,孔隙率可小于10％,Cs和Sr的浸出率可分别达到10~(-5)g·cm~(-2)·d~(-1)与10~(-7)g·cm~(-2)·d~(-1)水平。高放废液碱矿渣水泥固化体与玻璃固化体性能相当;而固化工艺比玻璃法简单。此外,还探讨了固化机理,核素离子在碱矿渣水泥固化体中的形态。     

放射性废物处理与处置标准在中放废液大体积浇注水泥固化工程中的应用探讨

本文扼要阐述了放射性废物处理与处置标准在中放废洼失体积浇注水泥固化工程中的应用,针对GB 14569.1-93<低、中水平放射性废物固化体性能要求水泥固化体>和GB 7023-86<放射性废物固化体长期浸出试验>放射性废物处理与处置标准在中放废液太体积浇注水泥固化工程应用中所存在的问题进行了探讨,并提出了建议.     

含^60Co和^152Eu放射性废液的水泥固化配方研究

针对含60Co3.8×105Bq/L、152Eu6.67×105Bq/L、总放射性活度为2×107Bq的放射性废液进行了水泥固化配方及工艺试验研究。结果表明水泥浆流动度和初凝时间随水灰比增大而增大,而固化体的抗压强度则随其增大而降低。优选配方的水泥固化体各种性能均满足中低放废液固化体性能要求水泥浆流动度≥130mm;水泥固化体28d抗压强度＞7MPa;42d浸出率60Co为1.84×10-4cm/d、152Eu为2.76×10-5cm/d（剂灰比0.15）,60Co为5.47×10-4cm/d、152Eu为1.55×10-4cm/d（无添加剂）;总β的累积浸出分数（42d）分别为1.7×10-2cm（剂灰比0.15）和3.5×10-2cm（无添加剂）。     

中放废液大体积浇注水泥固化配方研究

本文研究了后处理厂中放蒸残液和元件脱壳的偏铝酸钠废液大体积浇注水泥固化的特殊工艺配方。实验用525普通硅酸盐水泥,模拟废液中分别含~(134)Cs 和~(85)Sr,其放射性浓度均为3.7×10(?)Bq/L。实验结果表明,选用加 DH 型水泥添加剂的配方可满足大体积浇注固化池内水泥浆“自流式”流平的技术要求,水泥浆的流动度达0.19m 以上;近似绝热养护后的中放蒸残液和偏铝酸钠废液固化体的抗压强度分别大于7.8和10MPa,固化体性能良好,近似绝热养护28天,两种固化体42天时~(134)Cs 和~(85)Sr 的浸出率均低于1.0×10~(-2)cm/d。     

模拟放射性含氟废液水泥固化配方的研究

本文以沸石、硅灰、石英砂为添加剂,按照质量比m（沸石）∶m(硅灰)∶m（石英砂）∶m（水泥）=1∶1∶3∶10配方对模拟放射性含氟废液进行水泥固化。由配方得到的水泥浆流动度和初、终凝时间满足桶内固化要求。测定了水泥固化体28 d的抗压强度、抗浸泡性和抗冻融性实验后的强度损失,进行了抗冲击性能测试和模拟核素浸出实验。结果表明,该配方可有效地固化模拟放射性含氟废液,固化体28 d抗压强度、各项实验强度损失和模拟核素浸出率均满足GB 14569.1-2011的要求。水泥固化体的F^-浸出率很低,XRD显示F^-以CaF₂形式存在。废液中F^-质量分数控制在1%较为合适,此时水泥固化体终凝时间为14 h,F^-的42 d浸出率为2.54×10^-3 cm/d。     

模拟放射性含硼废液的水泥固化研究

为了比较硫铝酸盐水泥和普通硅酸盐水泥含硼废液的固化,为配方优化提供依据,研究采用两种配方对模拟放射性含硼废液进行水泥固化。测定了固化体28d抗压强度、抗浸泡性、抗冻融性和耐γ辐照试验后的强度损失,进行了模拟核素浸出试验,并对固化体水化产物进行XRD分析。结果表明,两种配方可有效固化模拟含硼废液,固化体28d抗压强度、各项试验强度损失和模拟核素浸出率均满足GB14569.1—93的要求,试验所用的硫铝酸盐水泥配方对Cs＋的滞留能力优于普通硅酸盐水泥配方,固化体中的硼以B(OH)4－形式固溶在钙矾石中。     

含~(60)Co和~(152)Eu放射性废液的水泥固化配方研究

针对含60 Co 3 8× 10 5Bq/L、152 Eu 6 67× 10 5Bq/L、总放射性活度为 2× 10 7Bq的放射性废液进行了水泥固化配方及工艺试验研究。结果表明 :水泥浆流动度和初凝时间随水灰比增大而增大 ,而固化体的抗压强度则随其增大而降低。优选配方的水泥固化体各种性能均满足中低放废液固化体性能要求 :水泥浆流动度≥ 130mm ;水泥固化体 2 8d抗压强度 >7MPa ;4 2d浸出率60 Co为1 84× 10 - 4cm/d、152 Eu为 2 76× 10 - 5cm/d(剂灰比 0 15 ) ,60 Co为 5 4 7× 10 - 4cm/d、152 Eu为1 5 5× 10 - 4cm/d(无添加剂 ) ;总 β的累积浸出分数 (4 2d)分别为 1 7× 10 - 2 cm(剂灰比 0 15 )和3 5× 10 - 2 cm(无添加剂 )     

对模拟TBP有机试剂固化处理的初步研究

研究了废有机溶剂TBP/煤油碱性水解工艺的可行性与相关条件,分析其水解的主要产物,并对水解溶液进行水泥固化处理。经过气相色谱分析,当温度为95～105℃时,TBP在12.5mol/L NaOH溶液中水解率接近100%,滴定分析表明主要水解产物为HDBP,并有少量的H2MBP。对水灰比为0.4和0.6的水泥固化体保养28天后进行抗压强度测试,其中水灰比为0.4的水泥固化体抗压强度可以达到12MPa。     

碱矿渣水泥固化模拟高放废液的研究

研究了用碱矿渣水泥固化模拟高放废液。结果表明,碱矿渣水泥在抗压强度、孔结构、Cs离子浸出性能及热稳定性等方面均优于硅酸盐水泥和高铝水泥。以碱矿渣水泥为基体,掺入适量沸石和硅灰,采用特殊工艺,在废物包容量小于25％时。固化体抗压强度65-100MPa,孔隙率小于10％,核素Cs和Sr离子浸出率仅为10~(-5)和10~(-6)g·cm~(-2)·d~(-1)的水平,与现有玻璃固化体性能相近。另外还探讨了核素离子在碱矿渣水泥固化体中的固化机理。     

模拟高放废物玻璃固化体冲击试验研究

本研究采用普通固体废物块冲击试验装置,对两种组分模拟高放废物玻璃固化体进行了冲击试验,研究了冲击能量、固化体组分对碎粒粒径分布的影响,估算了单位撞击能量所增加的表面积值,以建立玻璃固化体抗冲击性能的测试试验方法。     

Solidification of spent radioactive organic solvent by sulfoaluminate and Portland cements

The solidification of simulated spent radioactive organic solvent, tri-butyl phosphate/kerosene, was investigated by emulsification–solidification method using sulfoaluminate cement (SAC) and Portland cement (PC). Zeolite, calcium hydroxide and MR-1 type emulsifier were mixed into the cement blends for improving the performance of solidified waste forms (SWF). The properties of SWF were evaluated in terms of mechanical strength, leachability and mineral phase analyses. The hydration products of SWF were characterized by X-ray diffraction (XRD). The experimental results showed that the 28 d compressive strengths of SAC solidified waste forms (SACF) and PC solidified waste forms (PCF) were 14.23 and 19.07 MPa, respectively. Leaching sequence of three radionuclides in two kinds of SWF is Cs⁺ > Sr²⁺ > Co²⁺. Compared with PCF, SACF had better performance in preventing nuclides Co²⁺ and Cs⁺ from leaching to the environment. The XRD patterns suggested that simulated spent radioactive organic solvent and emulsifier in SWF did not obviously change the hydration products of the two cements (SAC and PC).     

核电厂放射性废保温棉玻璃固化比较研究

为了验证所研发的等离子体系统对废物的适用性,利用等离子体炉熔融模拟核电厂废保温棉,得到了固化体。与实验室马弗炉制得的固化体相比,等离子体炉制得的固化体中同样无晶相结构,成分因炉膛耐火材料的熔蚀而出现差异,抗压强度则更优;二者的元素浸出实验结果相近;等离子体炉的出料实验证实,熔融体的高温黏度适合所选定的出料工艺。这些结果表明,利用所研发的等离子体系统可以得到性能与实验室相当的玻璃固化体,核电厂的废保温棉可以用于含硼浓缩液的玻璃固化。     

压水堆核电站产生的硼酸废液和浓缩废液的水泥固化研究

本文介绍了压水堆核电站产生的硼酸废液和浓缩废液水泥固化的实验室研究结果，硼酸废液在水泥中有良好的分散性和缓凝作用，水泥浆的泌水和终凝时间太长限制了水泥固化体中废液的包容量。     

含氚废水水泥固化的工艺及设备

本文介绍了本院水泥固化含氚废水的工艺概况和主要设备，并给出了含氚废水水泥浆和固化体性能及浸出实验结果。实验结果表明，沥青涂覆固化体表面和金属密封容器相结合处理含氚废水是实际可行的。     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste and tritium valorization

Nuclear waste management has to be taken into account for fusion machine using tritium as fuel. Soft housekeeping waste (e.g. gloves, tissues, protective clothes, etc.) is produced during the whole life as well as during the dismantling of the reactor and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms.In collaboration with ENEA, a lab-scaled facility has been built at CEA Cadarache for soft housekeeping waste detritiation and tritium valorization. The previously milled waste is placed in a reactor to be heated up to a temperature lower than the housekeeping melting point. A carrier gas is then injected in the detritiation reactor to remove tritium, thanks to the combined effects of temperature and carrier gas (type and feed flow). The tritiated gas exhausted from the detritiation reactor is then sent through a catalytic Pd–Ag membrane reactor (CMR) where tritium is recovered via isotopic exchange reaction and permeation phenomenon.Based on previous studies that have allowed defining the most efficient operating conditions for the detritiation process, this work presents the results obtained by the coupling of the detritiation facility with the CMR. Due to safety considerations, restrictions on the nature of the carrier gas were applied, rejecting air as the carrier gas even though air was the best candidate for the detritiation part of the process. The performance of the whole system was estimated by means of a parametric study on the influence of flow rates in the CMR and transmembrane pressure.     

高放废液合成岩石固化研究

建立了高效废液合成岩石固装置，确定了固化工艺和性能测试方法，制备的合成岩石固化体样吕测试样结果表明，采用的实验装置，工艺流程和测试方法可行，将Ｎａ０．５ＲＥＥ０．５ＴｉＯ３型钙钛矿和Ｎａ２Ａｌ２ＴｉＯ８Ｏ１６型黑钛铁钠矿作为包容钠的主要矿相，分别研制了国内生产高放废液的合成岩石基料配方，氧化钠的包容量可达５．７％。     

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.     

Studies on mechanical, thermal and morphological properties of irradiated recycled polyamide and waste rubber powder blends

The aim of this article was to show the effect of gamma irradiation on mechanical and thermal properties of recycled polyamide (rPA) copolymer blended with different content of waste rubber powder (WRP). In order to study the structural modifications of prepared blends have been subjected to irradiation doses up to 200 kGy were applied to all samples. Non-irradiated blends were used as control samples. Mechanical properties, namely, tensile strength (TS), elastic modulus, elongation at break and hardness have been followed up as a function of irradiation dose and degree of loading with rubber content. Furthermore, the influence of radiation dose in the thermal parameters, melting temperature, heat of fusion, ΔH_f of the recycled PA and its blend with waste rubber powder (WRP) was also investigated.     

应用焚烧技术处理核电厂放射性固体废物的技术经济分析

介绍了中国改进型三环路压水堆(CPR1000)放射性固体废物的来源和当前放射性固体废物处理系统,并以含4台CPR1000机组的厂址为例,对当前废物处理工艺和使用焚烧技术的处理工艺进行了比较分析。结果表明,焚烧技术在核电厂低、中水平放射性固体废物减容和废物处理经济性方面具有明显的优势。