MIPR燃料硝酸铀酰溶液中痕量氟、氯离子的测定

采用水蒸气蒸馏法分离出硝酸铀酰溶液中的痕量F-、Cl-,并用离子色谱对F-、Cl-进行测定,对F-、Cl-分离条件和测定条件进行了讨论。该法的线性范围为0.04~0.3 mg/L,线性相关系数大于0.999,相对标准偏差小于10%,样品的加标回收率为94%~97%。     

Scale-Up Preparation of Uranium Amalgam from Uranyl Ion Using Two-Compartment Electrolyzer

A solid uranium amalgam containing as high as 1.7 g U/ml Hg was prepared electrolytically using a two-compartment electrolyzer separated with a cation exchange membrane at a kilogram scale. The design and operation characteristics of the electrolyzer is described. The results indicate that ca. 170 g of uranium ion in an aqueous solution could be reduced to metallic state by forming amalgam within 4h with a current efficiency of 30% and uranium recovery of more than 80%     

Reduced Partition Function Ratio of Uranyl Ion and Uranium Isotope Effect

Density of 30v/o tri-n-butyl phosphate-n-dodecane solution loaded with uranyl nitrate, nitric acid and water was measured. An empirical density equation was derived from regression analysis of the density data. The equation represents the density values well in a wide range of composition and temperature.     

Precipitation Reaction of Uranyl Ion in Molten Potassium Thiocyanate

The qualitative and the quantitative analyses of the reaction products obtained by heating uranyl solution of molten potassium thiocyanate were carried out. It was found that in the presence of water in the melt, uranyl ion is converted into UO₂ accompanied with the evolution of CO₂ and the formation of free sulfur. The molar ratios on these products were almost equal to each other.

By detecting ammonia in reaction products, it was concluded that the precipitation reaction of uranyl ion in the melt is expressed as follows:

UO₂ ⁺⁺ + SCN^? + 2H₂O = UO₂ + CO₂ +S+ NH₄ ⁺.     

密度-电导-温度法测量铀-酸体系中铀和硝酸

提出了密度-电导-温度法测量铀-酸溶液中铀和硝酸含量。建立铀-酸溶液的密度、电导率和铀质量浓度、硝酸浓度以及温度之间的数学模型,该模型为一非线性二元二次方程组。只要测出铀-酸溶液的温度、密度和电导率,代入上述方程组,即可由计算机计算出铀-酸体系中铀和硝酸浓度。结果表明:当铀-酸体系中铀质量浓度范围在50~200g/L、硝酸浓度范围在1~4mol/L时,该法对铀和硝酸浓度的测量相对误差均在±5%之间。     

Reactivity Effect Measurement of Neutron Interaction between Two Slab Cores Containing 10% Enriched Uranyl Nitrate Solution without Neutron Isolator

The reactivity effect of neutron interaction between two identical units containing low enriched (10% ²³⁵U enrichment) uranyl nitrate solution without neutron isolator was measured in the STACY. The unit has 350 mm of thickness and 690 mm of width and distance between those two units was adjustable from 0 to 1,450 mm. Condition of the solution was about 290gU/l in uranium concentration, about 0.8 N in free nitric acidity, 24–27°C in temperature and about 1.4g/cm³ in solution density. The reactivity effect was estimated from variation of critical solution level from 495 to 763 mm depending on the core distance. The reactivity effect was also evaluated by the solid angle method and a computational method using the continuous energy Monte Carlo code MCNP-4C and the nuclear data library JENDL 3.2. Those estimations were compared and the distance to isolate neutron interaction was overestimated by the solid angle method. The computational method considering neutron reflection from surrounding structures reproduced well the isolation distance given by the experiment.     

Removal of Liquid and Solid Impurities from Uranyl Nitrate Hexahydrate Crystalline Particles in Crystal Purification Process

The purification behavior of uranyl nitrate hexahydrate (UNH) was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide (MOX) fuel in batch experiments. The UNH crystal recovered from the MOX fuel dissolver solution containing simulated fission products (FPs) was purified by a sweating and melt filtration process. Although the decontamination factors (DFs) of Pu, Cs, and Ba did not change in the sweating process, that of Eu increased with increases in temperature and time. These results indicate that liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs, and Ba were minimally affected in the batch experiments. On the other hand, the DF of Ba increased with 0.45 and 5.0 μ filters in the melt filtration process. Since Pu and Cs formed as Cs₂Pu(NO₃)₆ in the course of U crystallization and was accompanied with the UNH crystal, these behaviors were similar to each other. Although the DFs of Pu and Cs did not change with the 5.0 μ filter, it increased approximately twofold with the 0.45 μ filter. The particle size of Cs₂Pu(NO₃)₆ is relatively small and might pass through the 5.0 μ filter in the melt filtration process. The liquid impurities as Eu remained in the molten UNH crystal with some filters.     

Dissolution Behavior of Uranium Oxides with Supercritical CO2 Using HNO3-TBP Complex as a Reactant

Dissolution behavior of U₃O₈ and UO₂ using supercritical CO₂ medium containing HNO₃-TBP complex as a reactant was studied. The dissolution rate of the oxides increased with increasing the HNO₃/TBP ratio of the HNO₃-TBP complex and the concentration of the HNO₃-TBP complex in the supercritical CO₂ phase. A remarkable increase of the dissolution rate was observed in the dissolution of U₃O₈ when the HNO₃/TBP ratio of the reactant was higher than ca. 1, which indicates that the 2:1 complex, (HNO₃)₂TBP, plays a role in facilitating the dissolution of the oxides. Half-dissolution time (t_½ ) as an indication of the dissolution kinetic was determined from the relationship between the amount of uranium dissolved and the dissolution time (dissolution curve). A logarithmic value of a reciprocal of the t_½ was proportional to the logarithmic concentration of HNO₃, C_HNO3, in the supercritical CO₂. The slopes of the (l/t_½ ) vs. ln C_HNO3 plots for U₃O₈ and UO₂ were different from each other, indicating that the reaction mechanisms or the rate-determining steps for the dissolution of U₃O₈ and UO₂ are different. A principle of the dissolution of uranium oxides with the supercritical CO₂ medium is applicable to a method for the removal of uranium from solid matrices.     

超临界二氧化碳中TBP络合萃取硝酸钕

将超临界流体萃取技术应用于乏燃料后处理中,可简化后处理流程、减少二次废液的产生。本工作进行了含磷酸三丁酯(TBP)的超临界二氧化碳(SC-CO2)络合萃取硝酸钕的实验研究,考察了硝酸钕粒径、TBP流量、系统温度和压力对络合萃取过程的影响。实验结果表明,含TBP的SC-CO2可高效萃取硝酸钕,萃取率达97%以上,增大TBP流量可加快萃取过程,而粒径、温度和压力对萃取速率的影响则较小。由实验结果可推断,该络合萃取过程的动力学受络合反应控制,并用一动力学模型计算出表观反应速率常数。     

Removal of Radioactive Ions from Nuclear Waste Solutions by Electrodialysis

Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained ¹³⁷Cs⁺, ⁹⁰Sr²⁺, ¹⁰⁶Ru³⁺ or fission products (F.P.) as active ions and NaCl, Na₂SO₄ or Ca(NO₃)₂ as inactive coexisting salts. The decontamination factor of the active ions was in the order: ¹³⁷Cs⁺ (greater than 99%) ⁹⁰Sr²⁺F.P. ¹⁰⁶Ru³⁺. The dialysis time required to attain the saturation was the shortest for monovalent cations K⁺, Cs⁺ and Na⁺, intermediate for divalent cation Sr²⁺, and the longest for trivalent cation Ru³⁺. The ratio of the decontamination factor of an active ion ηA to the desalination factor of an inactive ion A₊ was nearly equal to unity for ²⁴Na, ⁴²K, ¹³⁷Cs and ^S),Sr. On the other hand, the apparent selective permeability of an active ion (A⁺) against Na⁺ ion, T_A+ was higher than unity for all the active ions tested, and was in the order of ¹³⁷>Cs⁹⁰Sr>⁴²>K²⁴Na, where T^A _Na+ is defined by the ratio of γA to γ_Na+ with γA being the ratio of dilution of A in the diluate and γ_Na+ being that of Na⁺ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentrations of the coexistent salts or on the concentration of the active ions.     

Solubility of Tributylphosphate in Plutonium Nitrate Solution and Highly Radioactive Liquid Waste Solution

Abstract

The solubility of tri-n-butylphosphate (TBP) in aqueous solutions of plutonium nitrate (PuN) and in highly radioactive liquid waste (HRLW) of PUREX nuclear fuel reprocessing was investigated. By an empirical formula the solubility of TBP in PuN solutions was described in the range of 0–0. 1 M Pu and 1–8M HNO₃ concentrations. The following items were elucidated:

(1) The logarithm of TBP solubility (S) in the solution of interest varies inversely in proportion to the concentration of Pu(IV) in the range of 0–0.1M PU(IV) at a constant concentration of HNO₃, indicating that Pu(IV) simply behaves as an electrolyte for the salting-out of TBP. Log S subsequently levels off with increasing Pu concentration, which would be due to a change in the principal dissolution form of TBP having an interaction with Pu (IV).

(2) The variation in S in PuN solutions (0–0.1M PU) with nitric acid concentration shows almost the same tendency as that in HNO₃ solution.

(3) A dependency of S on fission product metal ions in HNO₃ for HRLW similar to that for PuN was observed.

(4) The logarithm of the ratio of TBP solubility in water to that in solution of interest was nearly proportional to l/T for HRLW solution or for low concentration of PuN solution. That deviates from the linear relation at high temperature when the concentration of PuN is increased, which can be explained by the change in ionic form of Pu.     

Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors

This paper is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. To minimize plutonium proliferation concern the adoption of long-life core with no fuel radiochemical treatment on site is suggested. Current investigation relies upon light water reactor technology and plutonium-free fresh fuel. Erbium doped to uranium oxide (enrichment 19.8%) fuel is selected as the reference. Such a high enrichment is selected in attempt to approach the longest irradiation time in one batch mode. In addition to that, uranium enriched up to 20% does not consider as a nuclear material for direct use in weapon manufacture. A sequence of two irradiation cycles for the same fuel rods in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140GWd/tHM without compromising safety characteristics. Being as large as 8% in the final isotopic vector, fraction of ²³⁸Pu serves as an inherent protective measure against plutonium proliferation.     

Uranium Extraction from Sea Water with Composite Adsorbents, (II)

A composite hydrous oxide, prepared from the mixed solution of titanium tetrachloride and ferrous chloride by addition of sodium hydroxide solultion, was investigated because of its rather high uranium adsorption capacity and its magnetic property. Results obtained may lead to easier handling of adsorbent species in the extraction of uranium from sea water.

The uranium adsorption capacity of the composite hydrous oxide was measured using sea water to which a small amount of uranyl chloride was added. The initial uranium concentration was 10.1 μg/l. Physical and chemical properties, such as specific surface area, mean pore radius and amount of surface OH groups, were also measured. The composite hydrous oxide was found to be composed mainly of relatively small particles of anatase and large particles of magnetite. Uranium adsorption capacity reaches its maximum when the precipitation temperatures are 50–70°C. The capacity of the composite hydrous oxide was found to be closely related to the mean pore size and the amount of surface OH groups.     

Estimation of Electrical Resistivity Increase due to Carbon Deficients in Uranium Monocarbides

Pyrolysis of spent ion exchange resins is used to reduce radioactive waste volume and to make the final waste form more stable. The weight loss of cation exchange resin after pyrolysis is only 50w/o, while that of anion exchange resin is 90w/o. Fundamental experiments were performed to investigate the reason for the small weight loss of the former.

The cation resin consists of base polymer and functional sulfonic acid groups. Chemical analyses of the pyrolysis products showed that 65% of the functional groups decomposed at about 300°C and generated SO₂ gas. However, only a small amount of the base polymer was pyrolyzed even at 600°C and the weight loss was only 50w/o. The IR and XPS studies on the residue showed that 35% of the functional sulfonic acid groups was converted to sulfonyl and sulfur bridges between the base polymers during pyrolysis. These bridges made the base polymers thermally stable. Therefore, the small weight loss of the cation resin was attributed to formation of bridges, which originated from the functional groups.     

Carbothermic Synthesis of High Purity Plutonium Nitride from Plutonium Oxide

Synthetic conditions of PuN from carbothermic reduction of PuO₂ has been studied in a mixed 8%H₂+92%N₂ stream at a temperature range of 1,270–1,680°C. In the course of both reactions of the carbothermic synthesis of PuN from PuO₂ and the hydrogenation of C, the vaporization loss of Pu was observed. It increased with temperature in the temperature range of 1,350–1,450°C, and reached to a constant value 1.3% of total Pu in the temperature range of 1,450–1,680°C, at which PuN was synthesized at a reaction rate of high enough. The minimum mixing ratio (C/PuO₂, mole ratio) for the formation of high purity PuN depends on temperature. The value is 2.15 for 1,620°C and 2.35 for 1,680°C. The oxygen and carbon impurities in the PuN obtained were found to be 0.095–0.028 and 0.17–0.012w/0, respectively.     

Decontamination of Radioactive Contaminants from Iron Pipes using Reactive Microemulsion of Organic Acid in Supercritical Carbon Dioxide

The ferrite fixed on the iron pipes was decontaminated by a reactive microemulsion in supercritical carbon dioxide (SC-CO₂). The specimens were prepared by treating the iron pipes with steam at 1,273 K for 2 min. The specimen was not dissolved in 3 mol·dm^?3 HNO₃ because its surface was covered with ferrite, while the original iron pipe was easily dissolved. This difference was used for determination of the fraction of ferrite. The fraction of ferrite covering the iron pipes was 1.5±0.3 wt%. A microemulsion containing organic acid was prepared using a fluorinated reagent, pentade-cafluorooctanoic acid (PFOA), and a non-fluorinated surfactant, polyoxyethylene (2) nonylphenyl ether (NP-2) and citric acid. In the former system, PFOA acted as a surfactant as well as an acid. By observation of the phase equilibrium, the microemulsion was found to be stabilized when the molecular ratio of water to surfactant, the w value, was 5.0 for the PFOA+H₂O+SC-CO₂ system, and 8.7 for the NP-2+citric acid+SC-CO₂ system at 25 MPa, and 323 or 353 K. Although the removal fractions of the ferrite were 0 and 1% for the PFOA and NP-2 system, respectively, at 25 MPa, 323 K, they increased to 92 and 56% at 25 MPa, 353 K.     

Fundamental Study on Separation of Uranium from Other Metals by Solvent Extraction with Tributylphosphate or Tributylphosphine Oxide in Molten Naphthalene

The extraction behaviors of U and other metals has been studied as functions of the concentration of nitric acid or hydrochloric acid in aqueous media and as functions of the amount of naphthalene as diluents, when tributylphosphine oxide (TBPO) or tributylphosphate (TBP) was used as a extractant. Uranium could be quickly extracted over 95% from 25 cm³ of the 1 mol·dm^?3 nitric acid solution with 150 mg of TBPO and 350 mg of naphthalene at 80°C. On the other hand, about 80% of U could be extracted under better conditions, when 25 cm³ of 4 mol·dm^?3 nitric acid solution and 1cm³ of TBP and 3 g of naphthalene were used. In either case, the extraction phases could be easily separated as a solid by cooling to room temparature. Uranium could be separated from alkaline earth metals, transition metals and rare earth metals, and U was efficiently concentrated by extraction with molten naphthalene, compared with usual solvent extraction which used toluene, etc.     

Numerical Model of Radionuclide Migration in Geologic Media

A three-dimensional radionuclide migration model has been developed by using of the direct-simulation method. The phenomena taken into account are radioactive decay, convection and dispersion in the ground water and sorption and desorption in the geologic media. Decay chain is represented by particle's character change using decay probability. Smoothing method is proposed to make an even distribution of particles and to obtain the exact radioactive inventory. Numerical calculations of ²⁴⁵Cm decay chain and ²³⁴U decay chain in the single layered geologic media have been carried out, and reasonable results have been obtained in comparison with INTRACOIN study. Nuclide migration of ²³⁷Np decay chain in the three-layered geologic media were examined and shown graphically.     

Development of Transuranium Element Recovery from High-Level Radioactive Liquid Waste

In the present study, the jet breakup characteristics of molten material is experimentally investigated in nonboiling condition using Wood's metal to isolate the key features of jet breakup phenomenon from the conjugated nature of melt breakup and steam generation. The experimental apparatus consists mainly of melt generating furnace and melt crucible equipped with variable nozzle diameter, a rectangular water tank of 350×350×800mm equipped with temperature controlling heater and thermocouples. The jet diameters were 10 and 20 mm and the jet velocity was varied by pressurizing the melt container. Wood's metal of 70°C melting temperature was used. Visualization of jet breakup provided characteristics of jet breakup in water. The range of jet velocity was 2.2–5.5 m/s. The debris were collected and sieved and it was shown that the debris sizes of 1.0–2.8 mm had the largest mass fraction, up to 50%. In the present experimental conditions, the Kelvin-Helmholtz instability is considered the most probable cause of jet breakup.     

Distribution of Plutonium Isotopes in Cooling Water from a PWR

Abstract

An investigation of plutonium isotopes in the primary cooling system of the Ringhals unit 2 (PWR) during normal operation has shown average concentrations of 0.03 Bq·l ^?1 of ²²⁸Pu and 0.02 Bq·l ^?1of ^{239 240}Pu. A major fraction of plutonium is associated with particles in contrast to dissolved plutonium in ionic form. The observed concentrations of plutonium isotopes in cooling water are characterized by log-normal distributions. The overall efficiency of the ion- exchange cleaning system with respect to plutonium in the primary circuit has been estimated as approximately 60%. During normal operation the ion-exchange cleaning system annually collects about 3 MBq of ²³⁸Pu and ²³⁹MBq of ^{239 +240}Pu. The cleaning system efficiently removes plutonium particles with sizes in the range 0.05 to 0.15 um from the primary cooling water. Particles with sizes outside this range are removed less efficiently.