Radiation Damage to Amine Extractants, (II)

Radiation damage to 5 v/0 alkyl amine/kerosene modified with lauryl alcohol was studied for changes brought thereby in their behavior in Th and Pa extraction by scrubbing with 3N nitric acid, and stripping with 0.01 N nitric acid at 25°C.

Eight amines were used; Tri-n-octyl amine, N-cyclohexyl dilauryl amine, N-benzyl dilauryl amine, N-cyclohexyl lauryl amine, Dilauryl amine, Amberlite LA-1, Amberlite LA-2 and Primene JM-T. These solvents were used virgin as well as irradiated by ⁶⁰Co γ-ray to various doses up to 10⁸R. No marked changes in extraction-scrubbing-stripping behavior were observed in all the amines tested, except N-cyclohexyl lauryl amine and Dilauryl amine, which tended to form either third phase or emulsion.     

ABSTRACTS OF JAPANESE ARTICLES

Radiation damage to 30v/0 TBP/kerosene was investigated to determine changes in the extraction behavior of uranium and thorium with the solvent irradiated up to 10⁸ R in nitrate system and similarly in the stripping behavior with 0.01 N nitric acid.

The effect on uranium and thorium was discernible only with irradiations reaching 10⁸ R.

Changes in the behavior of fission products present in company with uranium or thorium were found to be the similar to the case where they existed free of uranium or thorium. At irradiations below 10⁷ R, uranium and thorium manifested loading effect, but it disappeared when irradiation reached 10⁸R.

Calculation of one stage decontamination factors of fission products from uranium and thorium confirmed deterioration of the power of separation with irradiation, and the degree of deterioration was determined quantitatively. The decontamination factor for ⁹⁵Zr-⁹⁵Nb was decreased to 1/2,000 by irradiation to 10⁸ R, while the same dosage of irradiation had little effect for ¹⁰⁶Ru-¹⁰⁶Rh.

Physical properties such as density, viscosity, emulsification tendency and phase separation time were also observed to be effected by irradiation to 10⁸ R.     

Accurate Measurement of Isotope Ratio of Trace Plutonium in Uranium Matrix by Multi-collector Inductively Coupled Plasma Mass Spectrometry

An efficient separation procedure for trace plutonium from uranium matrix by TBP and aliphatic quaternary amine chromatographic extraction was developed. Various factors in the chromatographic extraction separation process were optimized experimentally.     

Isolation of Radioniobium from γ-lrradiated Molybdenum Oxide

A reverse-phase chromatographic isolation procedure for Nb from Mo and Tc is given. The separation is performed with TBP (tri-n-butyl phosphate) on celite column.

Molybdenum oxide was irradiated with 20MeV bremsstrahlung, which produces a mixture of ^93mMo, ⁹⁹Mo; ^91mTc; ^91mNb, ^95mNb, ⁹⁵Nb and ⁹⁶Nb. The irradiation was applied to a 6 N HC1 solution of the target. Separation was repeated twice, which resulted in isolation of the radioniobium with a decontamination factor of 10⁴ both for Mo and for Tc.

Decay analysis of the activity showed that radioniobium thus separated was composed of ^91mNb, ^95mNb, ⁹⁵Nb and ⁹⁶Nb. The γ-spectrum of ⁹⁶Nb was obtained by subtraction of the spectrum taken 119 hr after the end of irradiation from that of 74 hr.     

溶剂萃取法从辐照镎靶溶解液中提取分离钚的萃取价态选择Ⅰ:Pu(Ⅳ)-Np(Ⅳ)北大核心CSCD

采用磷酸三丁酯(TBP)溶剂萃取法对从辐照镎靶溶解液中提取分离钚的可行性进行了研究。从料液制备、流程设计两个方面研究了Pu(Ⅳ)-Np(Ⅳ)组合作为萃取价态组合的可能性。研究了1,1-二甲基肼(UDMH)还原-亚硝酸钠氧化两步法将镎、钚控制在Pu(Ⅳ)-Np(Ⅳ)的方法。结果表明,99.9%以上Pu(Ⅳ)-99.5%以上Np(Ⅳ)在4 h内能够保持稳定。基于此,设计了从辐照镎靶溶解液中提取分离钚的萃取流程,并用串级实验进行了验证:1A中镎的回收率为99.5%;1B中镎的反萃率为0.8%,钚的反萃率为99.9%;1C中镎的反萃率为99.5%。结果表明,采用Np(Ⅳ)-Pu(Ⅳ)的价态组合进料,基本可实现镎钚的分离,但料液中Np(Ⅳ)-Pu(Ⅳ)价态的长时间稳定性及TBP对Np(Ⅳ)萃取能力弱等问题将影响该工艺的实际应用。     

HNO3-H2C2O4体系中Np和Pu的分离研究

研究了30％TBP-煤油在不同的硝酸-草酸混合溶液中对Np,Pu各价态的萃取分配,在HNO     

Chemical Degradation of Kerosene Diluent with Nitric Acid

Chemical degradation of kerosene diluent with nitric, acid was studied to clarify its effect on uranium distribution ratio and fission product retention in the presence of degradation products. The decontamination of fission products retained in degraded kerosene was also investigated.

From the results of infrared spectrometry, it was recognized that nitro compounds, carboxylic acids, nitrate esters and nitroso compounds were formed as the main degradation products. The uranium distribution ratio in the extraction with TBP/degraded kerosene decreased with increasing concentration of nitric acid above 5 N, with rising temperature, and with increasing concentration of olefinic and aromatic hydrocarbons in the kerosene. It is suggested that the decrease of uranium distribution ratio may be due to the formation of carboxylic acids caused by the kerosene degradation. Some fission products, in particular ⁹⁵Zr-⁹⁵Nb and ¹³¹I, were retained in the organic phase, but ¹⁰⁶Ru-¹⁰⁶Rh did not remain after scrubbing with 3 N HNO₃. Fission product retention caused a decrease of uranium distribution ratio. Usual acid-alkali washing could not remove ⁹⁵Zr-⁹⁵Nb and ¹³¹I into the aqueous phase, hut ⁹⁵Zr-⁹⁵Nb was eliminated by oxalic acid washing.     

A contribution on the mechanism of plutonium nitrate extraction by mono- and dibutyl phosphates

In a previous communication [1] we examined the effect of monobutyl phosphate (MBP) and dibutyl phosphate (DBP) on the extraction of plutonium nitrate by tributyl phosphate (TBP).In the present work the mechanism of plutonium extraction by MBP and DBP is elucidated. An investigation was made of the mechanism of plutonium nitrate extraction by MBP and DBP with an ionic strength equal to 6. It was shown that plutonium is extracted in the form of PuK₄, where K is [(C₄H₉)₂PO₄ ^– or [C₄H₉HPO₄]^–. It was calculated that the equilibrium constant for the reaction of plutonium nitrate with MBP equals (1.5 ± 0.25)·10³ and with DBP, (6.15 ± 0.85) ·10³.     

Fractional Extraction of Zirconium,Niobium and Neptunium with Tri-n-butyl Phosphate

Radioactive Zr, Nb and Np were separated from each other by a continuous fractional extraction method.

The extraction of Zr and Nb was carried out with 1 % TBP in chloroform from the 10N hydrochloric acid solution of irradiated U. Under these conditions, Zr, Nb and Np could be extracted continuously with the organic solution, while U and other F.P. remained in the aqueous solution. Finally, Zr in the organic solution was back-extracted with concentrated hydrochloric acid containing 1sim;2 drops of hydrofluoric acid.

The method is convenient for the separation of Zr, Nb and Np from the F.P.

The distribution ratios in the extraction system have also been obtained for a number of radioactive nuclides.     

Separation of Uranium from Fission Products

A simple procedure has been developed for separating U from F.P. and neutron-irradiated Th. The separation is performed with tri-n-butyl phosphate in a system of dodecane-mixture of sulfuric acid and aluminium nitrate.

Uranium dioxide was irradiated with 20 MeV bremsstrahlung, which produced both ²³⁷U and F.P. The target was dissolved in dilute nitric acid and U was extracted into the organic phase of the above mentioned system. Finally, U in the organic phase was back-extracted into an aqueous phase. The γ-ray spectrum and decay curve of the separated U fraction show no radioactive nuclides other than U isotopes and its decay products.

This method can also be applied to the preliminary separation of ²³⁸U from neutron- irradiated Th.

The distribution ratios (Kd) for U, Th and some other elements in the extraction system are also given.     

Application of Electrically Driven Liquid-Liquid Dispersion to Solvent Extraction Using Tri-n-Buthyl Phosphate

Abstract

An alternative current (A.C.) electric field was applied to disperse an aqueous phase of nitric acid solution in a solvent extraction system using tri-n-buthyl phosphate (TBP) diluted with n-dodecane. The diameter of the aqueous droplet dispersed through the A.C. electric field showed a log-normal distribution and an average diameter was 0.02 mm in 10% TBP. This average diameter is approximately 1/10 of the value obtained in the pulsed column which is currently used for nuclear fuel reprocessing. Through a batch-wise extraction experiment of U(VI), it was shown that extraction efficiency increases with frequency of the A.C. electric field.

Through another experiment in continuous flow system, it was observed that the entrainment fraction of the aqueous phase into the product flow of an organic phase increases slightly with flow rate ratio, and yet it was found that it can be controled to be less than 0.2%. The experimental results indicate that application of the A.C. electric field is more suitable for lower TBP concentrations because of the increase of the electric conductivity of organic phase.     

Pulsed Neutron Experiment in Fast Natural Uranium Systems

Separation of fission products was carried out by solvent extraction using tri-n-butyl phosphate (TBP). ^{144, 144}Ce, ⁹¹Y and ⁹⁵Zr and ⁹⁵Nb were extracted with TBP from a freshly prepared nitric acid-potassium bromate solution. These nuclides in TBP were successively back-extracted with various aqueous solutions: ^{144, 144}Ce with hydrogen peroxide and nitric acid solution, ⁹¹Y with hydrochloric acid solution, and ⁹⁵Zr-⁹⁵Nb with oxalic acid solution. The other nuclides were not extracted by the TBP and remained in the nitric acid and potassium bromate solutions.     

Activation Analysis of Gadolinium Impurity in High Purity Europium Oxide

Abstract

An analytical method for determining Gd impurity in high purity Eu₂O₃ is proposed, which makes use of neutron activation and cation-exchange separation to examine its suitability as target material for the production of ^{152 m, 152, 154}Eu.

Long-term irradiation of an Eu sample resulted in ¹⁵³Gd activity amounting to 1.8 times that of the same nuclide produced from an equal quantity of Gd by (n, γ) reaction. This experimental value is quite consistent with that obtained by calculation under the assumption that the ¹⁶³Gd results from nuclear reaction on ¹⁵¹Eu induced by secondary neutrons.

For the accurate determination of Gd, Gd impurity was separated from the Eu sample prior to neutron irradiation in order to reduce the self-shielding effect in the Eu sample. Separation by cation-exchange with α-hydroxyisobutyrate (0.33 M, pH 3.77) made it possible to reduce the content of Eu in the Gd fraction below 7×10^?3%. This sufficed to assure that the ¹⁵⁹Gd content in ^152mEu was smaller than 1×10^?3%. For the determination of Gd content below 10⁴ ppm, however, should necessitate further purification of the irradiated Gd fraction.     

Photochemically-Induced Valency Adjustment of Plutonium and Neptunium in Nitric Acid Solution Using Mercury Lamp

A photochemically-induced valency adjustment method has been studied to remove Np from the mixed nitric acid solutions of Pu and Np in connection with the Purex reprocessing. The valencies of Pu and Np ions were adjusted to be Pu(HI) and Np(V) under the initial conditions and their concentrations were 1x10^?4 and 1x10^?3 mol·dm^?3, respectively. The experiments were carried out under the various conditions changing the irradiation intensities of the Hg lamp in the various concentrations of HNO₃. It was found that the rates of the redox reactions of the Pu ions were significantly affected by the irradiated light as well as the acid strength. Under the irradiation of the 0.015 W Hg lamp in 3 M HNO₃ solution containing a tenfold excess of a hydroxylamine and hydrazine, more than 95% Pu(ID) was oxidized rapidly to Pu(IV) within 10 min irradiation and it remained at the same valency even after the continuous further irradiation.

On the other hand, the irradiation did not change the valency of Np(V) under the conditions studied. These valency conditions, i.e. Pu(IV) and Np(V), are appropriate for separating Np from Pu by the solvent extraction with TBP-n-dodecane.

The present results lead to the conclusion that the photochemical method has a high potential for removing Np from the mixed solution of Pu and Np. The photochemical redox reaction mechanisms of Pu and Np in the nitric acid solution were discussed from the stand-points of the thermodynamic and kinetic considerations related to the variation in their standard electrode potentials of the photo-excited ion species by the light irradiation.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

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.     

Extraction of Some Elements by Mixture of DIDPA-TBP and Its Application to Actinoid Partitioning Process

Effect of TBP on the extraction of Pu(IV), Zr(IV), Nb(V), Nd(III) and Am (III) was studied with diisodecylphosphoric acid(DIDPA) as an extractant. Hydrolyzable elements contained in the high-level liquid waste of fuel reprocessing,i.e. Pu(IV), Zr(IV) and Nb(V), could be extracted with mixtures of DIDPA and TBP of different compositions. Addition of TBP to DIDPA causes an increase and a decrease of respectively distribution rate and ratio of Zr(IV). These elements extracted were completely stripped with the aid of oxalic acid. An effect of TBP on separation of transplutonides (III) from lanthanoids(III) in the DIDPA and DTPA extraction system was also studied.

Based on the results, a process flow sheet utilizing the extractant of DIDPA and TBP mixture was contrived for partitioning actinoids in the high-level liquid waste.     

Separation of Curium from Neutron Irradiated Am-241

The conditions under which curium can be separated from irradiated ²⁴¹Am target were elucidated. The isolation process consists of three steps: In the first step, Am(III) is oxidized to pentavalent state in a dilute nitric acid solution, and then plutonium and curium are extracted from the irradiated target by solvent extraction with HDEHP. Curium in the organic phase is back-extracted with 1 N nitric acid, and thereafter plutonium with a reducing solution containing ferrous sulfamate. The curium is finally purified by cation exchange, using α-hydroxy isobutyrate as the eluting solution. About 0.4 μg of ²⁴²Cm and 4×10^?3 μg of ²⁴³Cm were found in the curium fraction, which had been separated from 1 mg of irradiated ²⁴¹Am sample.     

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.     

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.