• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.179-188
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• 2007

In this work $^{14}C\;and\;^3H$ distribution characteristics of spent resins from nuclear power plants(NPPs), pressurized water reactors(PWRs), was investigated. It was found that the recovery percent of $^{14}C$ by the wet oxidation-acid stripping was $81%{\sim}100%$ for the added activity range of $^{14}C,\;0.72\;Bq{\sim}460\;Bq$, and it was not affected by the kinds of stripping acids, 3N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$. And the recovery percent of $^3H$ by distillation using the same apparatus was $81%{\sim}101%$ for the added activity range of $^3H,\;0.60\;Bq{\sim}435\;Bq$. Among the tested stripping acids, 3\;N-HCl, $3\;N-HNO_3\;and\;3\;N-H_2SO_4$, only the trapped $^3H$ solution by distillation in $3\;N-H_2SO_4$ was compatible with the 3H scintillator, Ultimagold XR. Neither of the $^{14}C\;and\;^3H$ trapping solutions from the spent ion exchange resin samples by the wet oxidation-3 $N-H_2SO_4$ stripping contained gamma nuclides. However, some gamma nuclides, $^{60}Co,\;^{134}Cs,\;^{137}Cs\;and\;^{54}Mn$, were found in the trapped $^3H$ solutions of the spent resins by the wet oxidation-3 N-HCl stripping. It was the same for the $^3H$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). Meanwhile only two nuclides, $^{134}Cs,\;and\;^{134}Cs$, were found in the $^{14}C$ trapping solutions of the spent resins by Sample Oxidizer(PACKARD MODEL 307). It was found that most of the $^{14}C$ in the spent resins existed as inorganic carbon form, more than about 70% of the total $^{14}C$ content. Among the analyzed 30 spent ion exchange resin samples, the average concentration of $^{14}C$ and $^3C$ for the high radioactive samples, 8 samples, was $19000\;Bq/g{\pm}41000\;Bq/g,\;670\;Bq/g{\pm}460\;Bq/g$ and that for the low radioactive samples, 22 samples, was $4.2\;Bq/g{\pm}4.3\;Bq/g,\;6.0\;Bq/g{\pm}5.3\;Bq/g$, respectively. And the average $^{14}C/^3H$ ratio for the high radioactive samples, was higher, 28, than that of low radioactive samples, 0.70. Some linear relationship trend was found between the activity concentrations of $^{14}C\;and\;^3H$.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• Resources Recycling
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• v.7 no.2
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• pp.9-15
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• 1998

Recovery af molybdenum in spent desulfuriring catalyst of petrochemical industries was studied from MfGnatc solulion which is a resultant of firstly remvercd vanadium by wet processes. In order to separate and recover molybdenum from upper mentioned rafinatz solution containing several mctal ions, such as molybdenum (1,100 ppm), vanadium (150 ppm), aluminium (19 ppm), and nickel (33 ppm), either adsorption technique by chelate resin or solvent extr~ction by tertiary amine as extractant was applied. In case of adsorption method, palyamine type chelate resin showed the highest selectivily far molybdenum ion up lo 60 ddm' of ancentration aftcr eluting with 3.0 rnolld~n' of NH,OH. On the othcr hand. molybdenum ion wa cffectlvely cxtractcd in Ule whole ranges of equilibrilrm pR by solvent extraction method with 10 ~01%-alamine 336 which was pretreated with 2N-HCI

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.49-56
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• 2005

Many different kinds of radwastes are discharged from the nuclear power plants, and $^{129}$I is included in these radwastes. Recovery test of $^{129}$I was evaluated for different radwastes(dry active waste, sludge, spent resin and simulated evaporator bottom). Recovery of $^{129}$I for dry active waste by acid leaching with $1.8\%$ NaClO was $74.3\%$$(RSD,\;2.2\%)$ and l291 for spent rein by alkali fusion method with KOH as a flux agent was $87.7\%$$(RSD,\;0.9\%$), respectively. iodide in simulated evaporator bottom containing a high concentration of borate was adsorbed with anion exchange resin at pH 7 phosphate buffer solution. Recovery of $^{129}$I for anion exchange resin was $92.5\%$ and not affected up to 1,200 $\mu$g/mL $H_3BO)3$(as a Boron). Recovery of $^{129}$I for the spent resin from nuclear power plant was $87.2\%$ $(RSD,\;1.2\%)$.

• Analytical Science and Technology
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• v.11 no.6
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• pp.421-428
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• 1998

A method has been studied to separate Zr from various fission products in PWR spent nuclear fuels. A solution containing metal ions in place of radioactive fission products was prepared. The Zr was separated with 5 M HCl followed by eluting metal ions such as Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag and Cd with 12 M HCl on Dowex $1{\times}8$, anion exchange resin. The recovery of Zr was more than 95%. The purification of Zr was carried out on anion exchange resin, Dowex $1{\times}8$, in 5 M HCl in order to remove Mo causing isobaric effect during mass spectrometry. The method was applied to separate Zr from a spent PWR fuel. From mass spectrometric measurement, the purified Zr portion was not showed the isobars from other elements such as Mo and Sr.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.41-47
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• 2005

Preparation of ferrocyanide-anion exchange resin and adsorption test of the prepared resin on the Cs$^{+}$$ion were performed. Adsorption capability of the prepared resin on the Cs$^{+}$ion in the simulated citric acid based soil decontamination waste solution was 4 times greater than that of the commercial cation exchange resin. Adsorption equilibrium of the prepared resin on the Cs$^{+}$ion reached within 360 minutes. Adsorption capability on the Cs$^{+}$ion became to decrease above the necessary Co$^{2+}$ion concentration in the experimental range. Recycling test of the spent ion exchange resin by the successive application of hydrogen peroxide and hydrazine was also performed. It was found that desorption of Cs$^{+}$ion from the resin occurred to satisfy the electroneutrality condition without any degradation of the resin.

• Resources Recycling
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• v.31 no.3
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• pp.73-80
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• 2022

Reduction smelting of spent lithium-ion batteries at high temperature produces metallic alloys. Following solvent extraction of the leaching solutions of these metallic alloys with either sulfuric or hydrochloric acid, the raffinate is found to contain Ni(II), Co(II), Mn(II), and Si(IV). In this study, two cationic exchange resins (Diphonix and P204) were employed to investigate the loading behavior of these ions from synthetic sulfate and chloride solutions. Experimental results showed that Ni(II), Co(II), and Mn(II) could be selectively loaded onto the Diphonix resin from a sulfate solution of pH 3.0. With a chloride solution of pH 6.0, Mn(II) was selectively loaded onto the P204 resin, leaving Ni(II) and Si(IV) in the effluent. Elution experiments with H₂SO₄ and/or HCl resulted in the complete recovery of metal ions from the loaded resin.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.147-157
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• 2004

The production of $^{14}C$ occurs in the Moderator(MOD), Primary Heat Transport System (PHTS), Annulus Gas System(AGS) and Fuel in the CANDU reactor. Among the four systems, The MOD system is the largest contributor to $^{14}C$ production(approximately 94.8%). $^{14}C$ is distributed of $^{14}CO_2$, $H_2^{14}CO_3$, $H^{14}{CO_3}^-$ and $^{14}{CO_3}^{2-}$ species as a function of the pH of water. Of these species, $H_2^{14}CO_3$ and $H^{14}{CO_3}^-$ form are predominant because the pH of MOD system is ＞ 5. In this paper, adsorption-desorption characteristics of bicarbonate ion (${HCO_3}^-$) by IRN 150 resin was investigated. ${HCO_3}^-$ ion existed in neutral condition(app. pH 7)was reacted with ion exchange resin (IRN-150) and saturated with it. Then $NaNO_3$ and $Na_3PO_4$ solutions selected as extraction materials were used to make an investigation into feasibility of ${HCO_3}^-$ extraction from resin saturated with ${HCO_3}^-$. Desorption of $CO^{2+}$ and $Cs^+$ ion by $Na^+$ ion was not occurred, and desorption of ${HCO_3}^-$ ion by ${NO_3}^-$ and ${PO_4}^{3-}$ was occurred slowly. Also, the status of ion exchange which is used in Wolsong NPPs and generation of spent resin yearly were surveyed.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.191-195
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• 2003

Zinc acetate has been proposed and used to evaluate ionic zinc as a means to reduce reactor radiation buildup at several nuclear plants. Thermodynamic analysis of the aqueous zinc system using reliable data shows that the stability of the hydrolyzed zinc species increases with pH and temperature. Adsorption kinetics and isotherm studies were carried out to investigate the mixed resin performance of the zinc adsorption. The equilibrium isotherms of the zinc adsorption onto nuclear grade resin indicate that the data correlate well with the Langmuir model and that the adsorption is physical in nature. The maximum capacity according to the Langrnuir model is about 0.6meq/g for an initial zinc concentration of 100ppm at $50^{\circ}C$. The use of natural zinc could result in the generation of a $^{65}Zn$ activity with about $500{\mu}Ci/mL$ of resin after 12 months of operation.