Preparation and characterization of multiphase ceramic designer waste forms

Abstract

The long-term performance, or resistance to elemental release, is the defining characteristic of a nuclear waste form. In the case of multiphase ceramic waste forms, correlating the long-term performance of multiphase ceramic waste forms in the environment to accelerated chemical durability testing in the laboratory is non-trivial owing to their complex microstructures. The fabrication method, which in turn affects the microstructure, is further compounding when comparing multiphase ceramic waste forms. In this work, we propose a “designer waste form” prepared via spark plasma sintering to limit interaction between phases and grain growth during consolidation, leading to monolithic high-density waste forms, which can be used as reference materials for comparing the chemical durability of multiphase waste forms. Designer waste forms containing varying amounts of hollandite in the presence of zirconolite and pyrochlore in a fixed ratio were synthesized. The product consistency test (PCT) and vapor hydration test (VHT) were used to assess the leaching behavior. Samples were unaffected by the VHT after 1500 h. As measured by the PCT, the fractional Cs release decreased as the amount of hollandite increased. Elemental release from the zirconolite and pyrochlore phases did not appear to significantly contribute to the elemental release from the hollandite phase in the designer waste forms.

Matrices for radioactive waste immobilization: a review

“…The chemical durability of the matrix defines the long-term structural integrity and the elemental release from the matrix (Ojovan and Lee, 2014;Clark et al, 2021). The standard protocols are employed to study the leaching behavior of the immobilizing matrix.…”

Exploring the Thermodynamic Behavior and the Effect of Temperature and Electron Radiation on the Structure of Fluorapatite Compounds

“…However, ceramics and glass ceramics (GC) can be used as alternatives to borosilicate matrix due to higher thermodynamic stability [8–11] . Various ceramic matrices like monazite, zircon, thorite, britholite, apatite, pyrochlore, perovskite or zirconolite have gained considerable research interest due to their higher thermodynamic stabilities and structural flexibility to accommodate different oxidation states of metal ions [12,13] . Apatites are one of the prominent matrices based on phosphates and halides with the general formula M 10 (PO 4 ) 6 X 2 , where M= Ca 2+ , Sr 2+ , Ba 2+ , X= OH − , Cl − , F − which can tolerate non‐stoichiometry in their cation and anion sub‐lattices without collapsing the crystal structure [11,14] .…”

“…[8][9][10][11] Various ceramic matrices like monazite, zircon, thorite, britholite, apatite, pyrochlore, perovskite or zirconolite have gained considerable research interest due to their higher thermodynamic stabilities and structural flexibility to accommodate different oxidation states of metal ions. [12,13] Apatites are one of the prominent matrices based on phosphates and halides with the general formula M 10 (PO 4 ) 6 X 2 , where M = Ca 2 + , Sr 2 + , Ba 2 + , X = OH À , Cl À , F À which can tolerate non-stoichiometry in their cation and anion sub-lattices without collapsing the crystal structure. [11,14] However, for any matrix proposed for waste immobilization, its thermal stability should be ascertained to mitigate the effect of generated heat by radioactive decay of self-heated immobilized fission products as well as due to rise in temperature due to geothermal effect externally.…”

Structural Evolution and Chemical Durability of Thorium-incorporated Nd2Zr2O7 Pyrochlore at A and B Sites

“…Key words: pyrochlore; thorium; structural evolution; chemical durability; A-site substitution; B-site substitution 高放废物(HLW)中长寿命放射性锕系核素的安 全处理处置已成为制约核工业可持续发展主要障碍 之一。 高放废物安全固化要求基材具备优异的物理、 化学、抗辐照稳定性, 其中人造岩石(Synroc)固化技 术直接将核素固化到矿相晶格位置, 是高放废物固 化处理的理想方案 [1][2][3] 。Wang 等 [4] 研究发现类萤石 结构 A 2 B 2 O 7 烧绿石陶瓷可以耐受质量分数 10%的 239 Pu 近 3000 万年辐照剂量仍能保持较好的结构稳 定性, 因此烧绿石成为锕系核素固化的潜在候选基 材 [5][6][7] 。A 2 B 2 O 7 烧绿石结构(Fd-3m)是一种缺陷的萤 石结构(AX 2 ), A、B 位为金属阳离子(过渡金属或稀 土离子), 其中 A 位占据 16c 晶格位(0, 0, 0), 与氧离 子形成八配位立方六面体结构; B 位占据 16d 晶格 位(0.5, 0.5, 0.5), 与氧离子形成六配位八面体结构。 氧离子占位为 48f(x, 0.125, 0.125)和 8a(0.125, 0.125, 0.125), 8b(0.375, 0.375, 0.375)为空位; 48f 氧离子占 位 x 具有可调性, 介于 0.3125~0.3750 之间 [8] 。 近 500 种烧绿石结构化合物被成功合成, 其中烧绿石结构 A、B 位均能实现放射性核素晶格固化, 包容+2~+5 价核素, 从而实现核素高包容量 [9] 。Belin 等 [10] 在锆 基 烧 绿 石 A 位 通 过 Am 的 晶 格 全 替 代 形 成 了 Am 2 Zr 2 O 7 固化体; Kulkarni 等 [11] 以 La 2 Zr 2 O 7 作为基 材实现了 Pu 在烧绿石 A 位晶格固化; Mandal 等 [12] 利用 Gd 2 Zr 2 O 7 作为基材实现了较小包容量 Th 在烧 绿石 A 位的替代。Kutty 等 [13] 在 Gd 2 Zr 2 O 7 的 A、B 位开展了 U 的固化研究, 烧绿石 A 位对 U 的固溶量 为 10%(原子分数); B 位固溶量较低, U 掺杂导致烧 绿石结构快速转变为萤石结构。Tang 等 [14] [17] , 能更好地包容锕系核素。锕系核 素 Th 作为重要核燃料, 随着先进重水堆(AHWR) [18] 及第四代反应堆钍基熔盐堆(TMSR) [19] [21] 。采用英国 Renishaw 公司的 [22] , 531)), 未观测到其他物相结 构衍射峰, 说明样品形成了单一烧绿石结构固化 体。烧绿石结构 A 和 B 位阳离子半径比(r A /r B )为 1.46~1.78, 超过该比例范围, 烧绿石结构会转变为 萤石结构 [9] 。烧绿石结构中八配位和六配位 Th 4+ 离 子半径分别为 0.105 和 0.094 nm, 八配位 Nd 3+ 和六 配位 Zr 4+ 离子半径分别为 0.111、0.072 nm。与 Nd 2 Zr 2 O 7 的 r A /r B 值相比(~1.54) [15] ,…”