Characteristic Behavior of Actinides During Electrochemical Reprocessing of Nitride Fuel in Chloride Melts

The results of experiments on anodic dissolution of (U, Pu)N pellets in (LiCl-KCl)_eut-UCl₃-PuCl₃ melt and precipitation of uranium and plutonium on a solid cathode in the presence of a small amount of neptunium and americium are presented. For electrochemical cell voltage 2.1 V, the ratio of uranium and plutonium in the cathode deposit obtained corresponds to the ratio in the electrolyte. Thus the requirement for nonpartitioning of actinides is achieved. Under these conditions, americium and neptunium precipitate simultaneously on the cathode.The conditions for complete precipitation of uranium and plutonium are determined experimentally. A residual concentration of these elements of 0.43 and 0.05 mass %, respectively, in the electrolyte is achieved.__________Translated from Atomnaya Energiya, Vol. 98, No. 3, pp. 201–206, March, 2005.     

Fabrication of uranium-plutonium oxide microspheres by the hydrolysis process

The application of the Hydrolysis Process to the fabrication of (U, Pu)O₂ microspheres is described. A uranium feed solution of high uranium concentration is prepared from concentrated uranyl nitrate-urea solution by adding solid hexa-methylentetramine. The uranium feed solution is mixed with a concentrated acid-free plutonium nitrate solution and processed into high dense (U, Pu)O₂ microspheres with Pu/(U + Pu) atomic ratios between 0.1 and 0.2. The Hydrolysis Process is applied also for pure substoichiometric plutonium oxide microspheres.     

1BP中铀钚价态及其含量的分析方法

建立了1BP工艺点铀钚价态及其含量的分析方法。通过研究不同价态铀钚的可见吸收光谱，采用多点斜率法拟合了不同价态铀钚在414、480、600、659 nm波长下的摩尔吸光系数ε。利用摩尔吸光系数ε结合多元线性回归法（MLR），建立了1BP中Pu(Ⅲ)、U(Ⅳ)、U(Ⅵ)及Pu(Ⅳ)分析的数学模型。该方法测量了1BP模拟样品：在工艺正常情况下，Pu(Ⅲ)的质量浓度范围为0.50～8.00 g/L，测量精密度优于3.0%(n=6)，回收率为94.5%～103.9%；U(Ⅳ)的质量浓度范围为0.45～38.15 g/L，测量精密度优于3.0%（n=6），回收率为95.3%～104.7%；U(Ⅵ)的质量浓度范围为0.45～38.59 g/L，测量精密度优于3.0%（n=6），回收率为96.5%～103.0%；Pu总量的回收率为87.2%～100.6%。方法简单快速，精密度高，属于无损分析。     

1BP中铀钚价态及其含量的分析方法

建立了1BP工艺点铀钚价态及其含量的分析方法。通过研究不同价态铀钚的可见吸收光谱，采用多点斜率法拟合了不同价态铀钚在414、480、600、659 nm波长下的摩尔吸光系数ε。利用摩尔吸光系数ε结合多元线性回归法（MLR），建立了1BP中Pu(Ⅲ)、U(Ⅳ)、U(Ⅵ)及Pu(Ⅳ)分析的数学模型。该方法测量了1BP模拟样品：在工艺正常情况下，Pu(Ⅲ)的质量浓度范围为0.50～8.00 g/L，测量精密度优于3.0%(n=6)，回收率为94.5%～103.9%；U(Ⅳ)的质量浓度范围为0.45～38.15 g/L，测量精密度优于3.0%（n=6），回收率为95.3%～104.7%；U(Ⅵ)的质量浓度范围为0.45～38.59 g/L，测量精密度优于3.0%（n=6），回收率为96.5%～103.0%；Pu总量的回收率为87.2%～100.6%。方法简单快速，精密度高，属于无损分析。     

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

Thorium (Th) oxide fuel offers a significant advantage over traditional low-enriched uranium and mixed uranium/plutonium oxide (MOX) fuel irradiated in a Light Water Reactor. The benefits of using thorium include the following: 1) unlike depleted uranium, thorium does not produce plutonium, 2) thorium is a more stable fuel material chemically than LEU and may withstand higher burnups, 3) the materials attractiveness of plutonium in Th/Pu fuel at high burnups is lower than in MOX at currently achievable burnups, and 4) thorium is three to four times more abundant than uranium. This paper quantifies the irradiation of thorium fuel in existing Light Water Reactors in terms of: 1) the percentage of plutonium destroyed, 2) reactivity safety parameters, and 3) material attractiveness of the final uranium and plutonium products. The Monte Carlo codes MCNP/X and the linkage code Monteburns were used for the calculations in this document, which is one of the first applications of full core Monte Carlo burnup calculations. Results of reactivity safety parameters are compared to deterministic solutions that are more traditionally used for full core computations.Thorium is fertile and leads to production of the fissile isotope ²³³U, but it must be mixed with enriched uranium or reactor-/weapons-grade plutonium initially to provide power until enough ²³³U builds in. One proposed fuel type, a thorium-plutonium mixture, is advantageous because it would destroy a significant fraction of existing plutonium while avoiding the creation of new plutonium. ²³³U has a lower delayed neutron fraction than ²³⁵U and acts kinetically similar to ²³⁹Pu built in from ²³⁸U. However, as with MOX fuel, some design changes may be required for our current LWR fleet to burn more than one-third a core of Th/Pu fuel and satisfy reactivity safety limits. The calculations performed in this research show that thorium/plutonium fuel can destroy up to 70% of the original plutonium per pass at 47 GWd/MTU, whereas only about 30% can be destroyed using MOX. Additionally, the materials attractiveness of the final plutonium product of irradiated plutonium/thorium fuel is significantly reduced if high burnups (∼94 GWD/MTU) of the fuel can be attained.     

Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor

A fuel irradiation program is being conducted using the experimental fast reactor ‘Joyo’. Two short-term irradiation tests in the program were completed in 2006 using a uranium and plutonium mixed oxide fuel which contains minor actinides (MA-MOX fuel). The objective of the tests is the investigation of early thermal behavior of MA-MOX fuel such as fuel restructuring and redistribution of minor actinides. Three fuel pins which contained MA-MOX: 2% neptunium and 2% americium doped uranium plutonium mixed oxide (Am,Pu,Np,U)O_2−x fuel were supplied for testing. The first test was conducted with high-linear heating rate of approximately 430 W cm⁻¹ for only 10 min. After the first test, one fuel pin was removed for examinations. Then the second test was conducted with the remaining two pins at nearly the same linear power for 24 h. In these tests, two oxygen-to-metal molar ratios were used for fuel pellets as a test parameter. Non-destructive and destructive post-irradiation examinations results are discussed with early on the behavior of the fuel during irradiation.     

钚的溶剂萃取三循环3A槽的工艺过程实验研究

利用核临界安全的混合澄清槽以铀代钚对钚的溶剂萃取三循环的３Ａ槽的工艺过程进行了实验研究。实验过程中，两相总液面高度小于临界安全高度极限值。经９级萃取和８级洗涤使铀（钚）与裂变产物进一步分离，得到了钠、酸的各级分布，铀（钚）收率可达９９．９％。萃取段槽的级效率为９５％。     

Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2®) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 °C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO₂ matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO₂ grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH)_4(am) phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Waste disposal aspects of the long term cooling characteristics of irradiated nuclear fuels

A study has been made of the long term cooling characteristics of nuclear fuels irradiated in commercial reactor designs of interest within the U.K. In the case of thermal reactors, Magnox, AGR, SGHWR, LWR and HTR systems fuelled with either natural or enriched uranium are considered, together with a fast reactor fuelled with plutonium derived either from the Magnox or the AGR programmes. Alternative uses for Magnox plutonium are considered by simulating a plutonium fuelled HTR thermal system and the development of a $\text{Th}{}^{233}\text{U}$ fuel cycle has been anticipated for both a fast reactor and an HTR.For each system the activities as a function of cooling time are considered on the assumption of U/Pu recovery from the fuel during reprocessing within a year of discharge from the reactor and for the alternative case of no U/Pu extraction. The reprocessing waste products associated with the various fuel cycles have then been compared both on the basis of decay heating and radiological hazard per GW(e) yr. Finally, recycling of transplutonium elements is also considered with a view to reducing the long term heating commitment from the higher actinides.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

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.     

Phase diagram analysis of (U,Pu)O2−x sub-system

The high plutonium, hypo-stoichiometric fuel exists as two phase system at low temperatures. The partial phase diagram of (U,Pu)O_2−x with two coexisting cubic phases was extensively investigated in this work using theoretical models. The critical temperature of the miscibility gap varies with Pu/M and O/M of the system. Based on the similar miscibility gap behaviour observed in PuO_2−x system and the experimental data available on the phase boundaries of (U,Pu)O_2−x for various Pu/M, some semi-empirical relationships and solution models were developed. With the help of these relationships, ternary isothermal sections of the miscibility gap, O/M at different temperatures and the critical temperature of the miscibility gap of (U,Pu)_2−x for different Pu/M values were calculated. These calculated values were compared with the available literature data.     

Radial redistribution of actinides in irradiated FR-MOX fuels

The redistributions of neptunium, plutonium and americium during two kinds of short-term irradiation tests for 10 min and 24 h at high linear heating rate around 430 W cm⁻¹ were studied in the uranium and plutonium mixed oxide fuel containing Am and/or Np. It was found in the irradiation test for 24 h that the concentrations of Pu and Am increased toward the central void, but there was no change in the concentration of Np. The obtained experimental redistributions of Am and Pu were analyzed, based on both pore migration and thermal diffusion models. As a result, the calculated redistributions of Pu and Am showed good agreements with the experimentally obtained ones.     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

强碱性阴离子交换树脂吸附DPA-Pu(Ⅳ)/U(Ⅵ)配合物研究

2,6-吡啶二羧酸（DPA,以H₂C表示）是一种可用于乏燃料后处理Purex流程高保留钚废有机相中钚洗脱的洗脱剂。为将DPA洗脱液中的钚与铀分离并回收钚,本文通过静态吸附实验研究了DPA-Pu(Ⅳ)/U(Ⅵ)配合物在强碱性阴离子交换树脂DOWEX 1上的吸附性能,考察了DPA浓度、酸度、温度以及主要辐解产物对DOWEX 1吸附钚和铀的影响。培养了DPA与U(Ⅳ)/U(Ⅵ)配合物的单晶并测定了其结构,通过配合物晶体与吸附金属离子树脂光谱的对比确定了Pu(Ⅳ)（以U(Ⅳ)模拟代替）和U(Ⅵ)吸附在树脂上的配合物形态,通过变温吸附实验获得了相应吸附反应的热力学数据。吸附实验结果表明,DOWEX 1树脂能在低酸（0.1 mol/L HNO₃）条件下同时吸附钚和铀,在高酸（8 mol/L HNO₃）条件下只吸附钚不吸附铀。根据上述实验所得结果,提出低酸吸附铀/钚、高酸柱上转型除铀、低酸解吸回收钚的方案,并进行了实验验证。结果表明,采用所提出的回收钚的方案,钚的回收率达96%,对铀的去污因子约为2.8×10³。     

Integrated Experiments of Electrometallurgical Pyroprocessing Using Plutonium Oxide

Electrometallurgical pyroprocessing is a promising technology to realize actinide fuel cycle. Integrated experiments to demonstrate electrometallurgical pyroprocessing of PuO₂ in continuous operation were carried out. In each test, 10–20 g of PuO₂ was reacted with Li reductant to form metal product. The reduction products were charged in an anode basket of the electrorefiner with LiCl-KCl-UCl₃ electrolyte. Using the anode, deposition of uranium on the solid cathode was carried out when PuCl₃/UCl₃ concentration ratio was low. After the Pu/U ratio in the salt electrolyte was increased enough, Pu and U were recovered simultaneously on a liquid cadmium cathode. By heating up the deposits for distillation of the salt and the cadmium, U metal or Pu-U alloyed metal was obtained as residues in the crucible. It was the first result to demonstrate the recovery of metal actinides in the continuous operation of pyroprocessing of oxide fuels.     

First Czechoslovak atomic-electric station A-1 with heavy-water reactor KS-150 (construction,start-up,and initial experience)

Conclusions In energy reactors similar to the KS-150 but larger, high fuel-expenditure levels (native metallic uranium) are entirely attainable. Thanks to the high conversion ratio, they can simultaneously produce plutonium for fast reactors in quantities approximately double the unit yield of thermal reactors using enriched uranium.Another method for developing such reactors [1, 2] may be to use in them a ring with equilibrium concentration of Pu and²³⁸U with additional²³⁵U contained in native uranium or diffusion-plant waste. In this case, they would be able to burn significant quantities of²³⁸U and, using their plutonium many times, also produce transuranic elements.Translated from Atomnaya Énergiya, Vol. 36, No. 3, pp. 163–170, March, 1974.     

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.