Host phases for actinides in simulated metallic waste forms

Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists primarily of stainless steel cladding hulls containing undissolved metal fission products and a low concentration of actinide elements. This waste will be immobilized in a metallic waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). This paper presents transmission electron microscope, energy-dispersive X-ray spectroscopy, and electron diffraction observations of SS-15Zr alloys containing 2-11 wt% U, Np, or Pu. The major U- and Pu-bearing materials are Cr-Fe-Ni-Zr intermetallics with structures similar to that of the C15 polymorph of Fe₂Zr, significant variability in chemical compositions, and 0-20 at.% actinides. A U-bearing material similar to the C36 polymorph of Fe₂Zr had more restricted chemical variability and 0-5 at.% U. Uranium concentrations between 0 and 5 at.% were observed in materials with the Fe₂₃Zr₆ structure.     

Decontamination of Zircaloy cladding hulls from spent nuclear fuel

The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The ⁹⁰Sr and ¹³⁷Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 μm of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit.     

Electrolytic Reduction of Spent Light Water Reactor Fuel Bench-Scale Experiment Results

A series of experiments were performed to demonstrate the electrolytic reduction of spent light water reactor fuel at bench-scale in a hot cell at the Idaho National Laboratory Materials and Fuels Complex. The process involves the conversion of oxide fuel to metal by electrolytic means, which would then enable subsequent separation and recovery of actinides via existing electrometallurgical technologies, i.e., electrorefining. Four electrolytic reduction runs were performed at bench scale using ～500 ml of molten LiCl–1 wt% Li₂O electrolyte at 650°C. In each run, ～50 g of crushed spent oxide fuel was loaded into a permeable stainless steel basket and immersed into the electrolyte as the cathode. A spiral wound platinumwire was immersed into the electrolyte as the anode. When a controlled electric current was conducted through the anode and cathode, the oxide fuel was reduced to metal in the basket and oxygen gas was evolved at the anode. Salt samples were extracted before and after each electrolytic reduction run and analyzed for fuel and fission product constituents. The fuel baskets following each run were sectioned and the fuel was sampled, revealing an extent of uranium oxide reduction in excess of 98%.     

乏燃料棒中关键核素径向分布特点及对废包壳测量准确度影响

废包壳是水法乏燃料后处理工艺首端高放固态废物的主要来源，通常采用非破坏性测量方法进行整体测量并分析其中残余的U、Pu等感兴趣关键核素的量，传统方法中多引用组件的平均燃耗作为分析计算的输入参数。但根据反应堆运行经验，乏燃料组件和乏燃料棒中燃烧生成的核素及残余U的浓度呈非均匀空间分布状态，这一特性增大了废包壳非破坏性测量分析结果的不确定度。本文采用模拟计算的方法重建乏燃料棒中感兴趣关键核素的径向分布特征，数据表明废包壳中感兴趣核素的质量浓度比与采用燃料棒平均燃耗计算的结果相差可达100%，由此建立了采用非均匀分布特性修正废包壳中感兴趣核素浓度测量分析方法。     

Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.     

Effects of hydrogen content and temperature on fracture toughness of Zircaloy-4

The influence of hydrogen content and temperature on the fracture toughness of a Zircaloy-4 commercial alloy was studied in this work. Toughness was measured on CT specimens obtained from a rolled material. The analysis was performed in terms of J-integral resistance curves. The specimens were fatigue pre-cracked and hydrogen charged before testing them at different temperatures in the range of 293–473 K. A negative influence of the H content on material toughness was important even at very small concentrations, being partially restored when the test temperature increased. Except for some specimens with high H concentration tested at room temperature, the macroscopic fracture behaviour was ductile. The role of Zr-hydrides and Zr(Fe,Cr)₂ precipitates in the crack growth and the dependence with hydrogen content were analysed by observation of the fracture surfaces and determination of the Zr(Fe,Cr)₂ precipitates density on them.     

Neutronic analysis of silicon carbide cladding accident-tolerant fuel assemblies in pressurized water reactors

In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ~(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.     

Microbial biofilm growth on irradiated, spent nuclear fuel cladding

A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 10³ Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.     

废锆包壳的α去污北大核心CSCD

开展了利用Ag(Ⅱ)间接氧化溶解废锆包壳的α去污研究。结果表明:非放锆包壳不溶于Ag(Ⅱ)-硝酸溶液中,可能是包壳表面生成氧化锆,抑制了包壳的溶解;建立了废锆包壳电化学溶解工艺,开展了废锆包壳α去污工艺验证,溶解后废锆包壳残留的α比活度为2.0×10^(5)Bq/kg;分析了废锆包壳α污染源项,确定了废锆包壳α比活度的主要贡献是^(241)Am和^(238)Pu。     

Reactions of Uranium-Plutonium Alloys with Iron

In metallic U-Pu-Zr fuel for fast reactors, metallurgical reactions occur between the fuel alloy and the stainless steel cladding, and a liquid phase may be formed in the reaction zone at a higher temperature. In order to clarify the condition for liquefaction at the fuel-cladding interface, the reactions of U-Pu alloys with Fe have been examined at 923 and 943 K. The test results confirmed that the liquid phase is not formed at 923 K in any region of the reaction zone when the maximum Pu content in the (U,Pu)₆Fe phase is less than the Pu solubility limit in this phase. Comparison of the present test results with the liquefaction data from the various tests on metallic fuel-cladding compatibility suggested that the liquefaction condition is independent of the Zr content in the fuel alloy and can be expressed as a function of the atom fraction ratio of Pu/(U+Pu) in the fuel alloy and the reaction temperature. At 923 K, liquefaction will occur when the Pu/(U+Pu) ratio is larger than 0.25.     

Formation of Pellet-Cladding Bonding Layer in High Burnup BWR Fuels

Formation process of the pellet-cladding bonding layer was studied by EPMA, XRD, and SEM/TEM for the oxide layer on a cladding inner surface and the bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27 and 42 GWd/t in BWRs. In the lower burnup specimens of 15 and 27GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and previously reported 49 GWd/t had a typical bonding layer. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the cladding was made up mainly of ZrO₂. The structure of this ZrO₂ consisted of cubic phase, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U, Zr)O₂ and amorphous phase. Even in the lower burnup specimens having no bonding layer, cubic ZrO₂ phase was identified in the cladding inner oxide layer. The formation process of the bonding layer were discussed in connection with phase transformation by irradiation damage of fission products and conditions for contact of pellet and cladding.     

Phase characteristics of a number of U-Pu-Am-Np-Zr metallic alloys for use as fast reactor fuels

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.     

N36锆合金包壳辐照生长经验模型研究

利用N36锆合金包壳燃料棒堆内辐照考验的部分池边检查数据，计算了4个典型辐照生长经验模型对N36锆合金包壳的适用参数。计算结果表明，在典型辐照生长经验模型中，双曲正切经验模型最适合描述N36锆合金包壳辐照生长行为。在双曲正切经验模型基础上，建立了N36锆合金包壳辐照生长最佳估算模型和包络模型。通过添加工程因子，建立了不同加工工艺的N36锆合金包壳辐照生长经验模型。利用池边检查剩余数据对N36锆合金包壳辐照生长经验模型进行了验证，模型与数据吻合较好。     

Bench Simulation of the Stages of an Anticipated Loss-of-Coolant Accident in a VVÉR Reactor System

A computational-experimental procedure for performing a bench simulation of the first and second stages of an accident are described. Experiments studying the behavior of fuel elements and model fuel assemblies under conditions simulating the first stage of an accident showed, for the first time, that the claddings of the most highly heat-stressed fuel elements bulge out and rupture. Tests performed on model fuel assemblies with 19 and 37 fuel elements with Zr + 1%Nb and É635 alloy cladding under the conditions characteristic for the second stage of an accident showed that the fuel-element cladding becomes deformed in the course of the tests. The claddings of the fuel-element simulators rupture at the heating stage in the temperature range 820–920°C. The blockages of the flow section of the model assemblies with Zr + 1%Nb and É635 alloy claddings were 42 and 70%, respectively.     

中国核动力院U-Mo合金燃料研究现状及进展

目前,U-Mo合金燃料是研究试验堆新一代燃料的研究重点.文章介绍U-Mo合金燃料在中国核动力研究设计院(NPIC)的研究现状和进展.NPIC于2006年正式开始研制U-Mo合金弥散燃料元件,几年间开展的研究工作主要有:U-Mo合金熔炼,γ相U-Mo合金粉末制备,(U-Mo)-(Al-Si)弥散燃料板制造工艺研究,U-Mo合金与基体材料、包壳材料和阻挡材料诸如Al、Nb、Zr、Mg等的相容性研究,Si添加到Al基体中对U-Mo/Al反应的影响以及U-Mo合金燃料成分分析及无损检测方法研究等.目前,NPIC已制备出基本满足要求的(U-Mo)-Al弥散燃料板,并计划于2010年前掌握满足技术要求的改进型(U-Mo)-Al弥散燃料板的制造技术.     

Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Thermochemical modelling of electrotransport of uranium and plutonium in an electrorefiner

Electrotransport behaviour of U and Pu in a molten salt electrorefining cell has been numerically simulated with an improved thermochemical model. Depending upon saturated or unsaturated states of the liquid Cd electrodes with respect to U or Pu or with both U and Pu, 16 conditions of electrorefiner cell operation have been categorised and electrotransport simulated for all the realistic conditions. Algebraic equations for determining the compositions of the salt phase and the two electrodes under each condition of electrotransport are derived. Fractional mass transport coefficients and relative fractional mass transport coefficients are derived for each condition to illustrate the electrotransport behaviour. Comparison is made between modeling with concentration dependent and concentration independent activity coefficients for U and Pu in liquid Cd. The electrotransport to a solid cathode and anodic dissolution have also been simulated. Application of the model to reprocessing of spent metallic fuel is discussed with respect to U recovery, Pu enrichment and reconstitution of the spent fuel with desired fuel composition.     

Injection casting of U–Zr and U–Zr–RE fuel slugs and their characterization

U–Zr fuel slugs containing rare-earth elements can be difficult to cast because of the high reactivity of rare-earth elements. In this study, U–Zr and U–Zr–RE (RE: a rare-earth alloy comprising 53% Nd, 25% Ce, 16% Pr, and 6% La by weight) fuel slugs were prepared by injection casting, and their characteristics were evaluated. The as-cast fuel slugs were fabricated to the full length of the mold, and they showed no thin sections or cracks. Compared to the theoretical density, the measured density of the U–Zr and U–Zr–RE fuel slug was lower and higher, respectively. Chemical analysis revealed that the Zr and RE compositions in the U–10Zr and U–10Zr–3RE fuel slugs matched the target composition within 1.0 wt%. However, the RE composition in the U–10Zr–7RE fuel slug differed from the target composition by over 4 wt%. The melting crucible was further deteriorated and the casting yield was lower for the casting of a high rare-earth bearing fuel slug.     

影响CaCl2-NaCl熔盐电脱氧法制备金属Zr的工艺条件

将氧化物转化为金属是熔盐电解精炼干法后处理氧化物乏燃料流程的关键步骤之一。在等摩尔CaCl₂-NaCl混合熔盐体系中,以石墨棒为阳极,采用高温烧结后的ZrO₂模拟UO₂开展了电脱氧制备金属Zr的FFC剑桥工艺条件优化。研究了工艺条件（槽电压、电解时间、烧结温度和电解温度等）对电脱氧制备Zr的影响。采用场发射扫描电子显微镜(SEM)和X射线衍射(XRD)分别分析了电解前后ZrO₂阴极的微观结构和物相组成。优化后的工艺条件为：电压3.4 V、电解时间12 h、烧结温度900 ℃和电解温度722 ℃。同时,研究结果表明, ZrO₂电脱氧还原为Zr时,存在中间产物CaZrO₃和ZrO。