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Presented in this paper is a methodology for combining a Bayesian statistical approach with Data Quality Objectives (a structured decision-making method) to provide increased levels of confidence in analytical data when approaching a waste boundary. Development of sampling and analysis plans for the characterisation of radioactive waste often use a simple, one pass statistical approach as underpinning for the sampling schedule. Using a Bayesian statistical approach introduces the concept of Prior information giving an adaptive sample strategy based on previous knowledge. This aligns more closely with the iterative approach demanded of the most commonly used structured decision-making tool in this area (Data Quality Objectives) and the potential to provide a more fully underpinned justification than the more traditional statistical approach. The approach described has been developed in a UK regulatory context but is translated to a waste stream from the Fukushima Daiichi Nuclear Power Station to demonstrate how the methodology can be applied in this context to support decision making regarding the ultimate disposal option for radioactive waste in a more global context.

A modeling calculation methodology for estimating the radionuclide composition in the wastes generated at the Fukushima Daiichi nuclear power station has been upgraded by introducing an approach us...

The Fukushima Daiichi Nuclear Power Station, which is owned by the Tokyo Electric Power Company, was damaged by the great earthquake and tsunami on March 11, 2011, and serious contamination due to radioactive nuclides occurred. To investigate the waste management methodologies, contaminated materials were radiochemically analyzed. This paper reviews the analytical data concerning actinide elements. Contaminated water has accumulated in the basement of the reactor and other buildings, and actinide nuclides have been detected in this water. Actinides first get dissolved into the water inside the primary containment vessel, and then their concentration in the water decreases to a certain level with further flow. The contaminated water is chemically decontaminated; however, the actinide concentration does not decrease with time. This suggests that the actinides are continuously being supplied by the damaged fuel via slow dissolution. The dissolved transuranic (TRU) nuclides are recovered in the precipitate via a chemical treatment and are mostly removed from the water. Pu, Am, and Cm were detected in the topsoil at the site and appear to originate from the damaged fuel, whereas the detected U originates from natural sources. TRU nuclides slowly move in soil to deeper layers. The contamination of the rubble is nonuniform, and actinides are detected as well as fission products. Inside the reactor building of unit #2, the TRU nuclide concentration is comparatively higher near the boundary of the primary containment vessel, which experienced a fault during the accident. As for the vegetation, TRU nuclides were only found in fallen leaves near the reactor buildings.

A calculation methodology for estimating the radionuclide composition in the wastes generated at the Fukushima Daiichi nuclear power station has been developed by constructing a skeleton overview o...

Radioactive waste management requires planned and systematic actions to provide confidence that the entire system, processes and final products will satisfy given requirements for quality. The characterisation process is dependent on setting clear characterisation objectives and gathering the right information to underpin the decisions that need to be taken to manage the waste safely. This paper reviews experience of characterisation of waste generated from past nuclear activities that were not conducted in compliance consistent with current criteria, or from unexpected situations that were not planned for. This experience shows that the development of a reliable and efficient characterisation and categorisation methodology is a common challenge for such wastes, referred to here as unconventional and legacy (UL) waste. Through the activites of the Nuclear Energy Agency Expert Group on the Characterisation of Unconventional and Legacy Waste, consideration has been given to widely used waste stream characterisation procedures and methods that were originally developed primarily for application in conventional decommissioning work. Although they provide a substantial basis for characterisation, there are various additional factors that commonly need to be taken into account in the case of UL waste. By analysing the challenges and lessons learned from a variety of case studies and other international experience, it has been possible to identify opportunities for adaptations and enhancements to these characterisation methologies, and these are set out and explained. The need for integration of waste characterisation with other aspects of strategic planning for UL waste management is discussed, including characterisation to address any non-radiological hazards. The analysed case studies have also highlighted the importance of developing a robust legislative and regulatory framework in parallel with an appropriate waste infrastructure to treat, store and dispose of UL waste. Finally, the basic features of a UL waste characterisation roadmap are presented, including the interactions within a wider UL waste management programme and key areas for further consideration and possible development. It is anticipated such work can be supported by continued international cooperation.

International audience; Following the NPP accident, several hundred tons of heat-generating corium and fuel debris have been cooled permanently by millions of m3 of flowing. Knowledge on the interaction with water is crucial for any decommissioning planning.Starting from knowledge on the evolutions of the accident in the three reactor cores and associated fuel debris formations and some additional isotopic and physio-chemical information on debris fragments collected in Fukushima soils, we review the temporal evolution of the chemistry and leached radionuclide contents of the cooling water, comparing measured concentration ratios of the actinides and fission products in the water to reported results of laboratory leaching studies with either spent nuclear fuel or simulated fuel debris.As for spent fuel leaching, the fractions of inventories of 134,137Cs in the cooling water are orders of magnitude larger than that of the actinides. After more than 10 years of fuel debris/water contact, 137Cs release rates have decreased by about a factor of 100. The total release of actinides from the fuel debris is orders of magnitude lower than that of 134,137Cs or of 90Sr. This high stability makes direct disposal of fuel debris in suitable containers after decommissioning a viable option.

Quantification of the radioactive waste inventory remaining inside the reactors at Fukushima Daiichi NPS is necessary to effectively plan their recovery, treatment, and disposal. Analysis of radionuclide concentrations and secondary wastes in the contaminated water treatment system can provide a means to estimate the radioactive waste inventory, which is not possible by more direct methods due to problems of accessibility and high levels of radiation. A predictive model has therefore been developed to estimate the radioactive waste inventory from the radionuclide concentrations and throughputs in the contaminated water. Model fitting has enabled the estimation of the key parameters, such as the initial radionuclide concentration C0, the continuous release rate F, and inventory of source of continuous release IS0. An estimated one-third of the total 137Cs inventory has already made its way into the water treatment system as secondary wastes, whereas half still remains inside the damaged reactors as...

Nd, which simulates minor actinides (MAs), was used for investigating residual minor actinides produced during the extraction chromatography separation of spent fuel from fast neutron reactors. A cross-sectional distribution of Nd in a minute globular adsorbent having diameter less than 50 μm was obtained using particle induced X-ray emission-computed tomography with a 3-MeV proton microbeam. The measurement area was 150 × 150 μm 2 corresponding to 128 × 128 imaging pixels in projection images with 9° resolution, image reconstruction was carried out by a modified ML-EM (maximum likelihood expectation maximization) method. As a result, the cross-sectional distribution of Nd in the adsorbent was successfully obtained, and it was first revealed that Nd existed both in the central region and on the outer surface even after an elution. This implies that the internal structure of the adsorbent must be modified for improving of the recovery of MAs.

The adsorption and desorption of cesium (Cs) on clays of contaminated soil in a rhizosphere zone can be greatly affected by various biogeochemical processes, the timespans of which are usually months to years. Herein, we present several representative scenarios of the binding of Cs on diverse sites of vermiculitized biotite by controlled Cs adsorption to particles of different sizes. We investigated whether and how the fixed Cs in the different scenarios is desorbed by ambient and hydrothermal treatments with several low-molecular-weight organic acids (LMWOAs). The results showed that the sorbed Cs was discriminatively retained in the un-collapsed, partially collapsed, and thoroughly collapsed structures of vermiculites. The desorption of the sorbed Cs by hydrothermal LMWOAs extractions was easily realized in the un-collapsed structure, but was limited or minimal in the partially collapsed and thoroughly collapsed structures. The Cs desorption varied in accord with the LMWOA species applied and increased with the acid concentration, temperature, and number of treating cycles. The analysis of Cs-desorbed specimens confirmed their partial destruction and interlayer expansion, suggesting that the underlying mechanism of Cs removal by LMWOAs involves not only acid dissolution and complexation but also the accelerated weathering of clays within a short time under hydrothermal conditions. Our findings contribute novel insights into the mobility, bioavailability, and fate of Cs in contaminated soils and its removal from these soils for environmental restorations.

The effective and efficient removal of radioactive Cs from contaminated soil is highly urgent for the nuclear post-accident remediation. In present study, we achieved rapid Cs desorption from both a typical micaceous clay (i.e., vermiculitized biotite, VB) and actually contaminated soil by high-speed ion exchange through temperature-controlled continuous leaching with Mg-solutions in a column reactor. Cs-sorbed VB was firstly employed as a soil surrogate to explore the macro-Cs desorption process and micro-mechanism in detail. Results showed that VB sandwiched the adsorbed Cs to its interlayers within collapsed structure (10.7 A) and prevent Cs release even by abundant extraction with H2O at 250 °C or Mg2+ at 25 °C. However, Mg2+-extracted Cs desorption boosted significantly with elevating temperatures and 100 % of sorbed-Cs was removed from Cs-VB leached above 150 °C. Further structural and composition analysis of the leached specimen ensured that solvated Mg2+ preferentially entered into Cs+-collapsed interlayers at 150 °C than K+-interlayers above 200 °C, leading to prior complete Cs removal over K from VB at lower temperatures. By contrast, the Cs-contaminated soil reduced by ∼39 % but ∼82 % of its initial radioactivity after equally leaching with same volumes of Mg2+-solution at 150 and 200 °C, respectively. These temperature-controlled Cs desorption validated that radioactive Cs in actual soil indeed be tightly trapped by micaceous clays nearly in the Cs-K co-collapsed interlayers, to which its extraction by other cations can conditionally occur above enough high leaching temperatures. These superior features would inspire new insights for the design of novel practical technologies for treatment and decontamination of the nuclear post-accident soils.

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO3 concentrations. The flowsheet with a supply of 0.15 mol/dm3 HNO3 solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U ...

As a part of developing extraction chromatography technology for minor actinides (MA(III); Am and Cm) recovery from spent fast reactor fuels, improvement on the TODGA/SiO2-P adsorbent to enhance its desorption efficiency was carried out. Batchwise adsorption/elution experiments showed that optimizations in amount of the extractant impregnated in the support of the SiO2-P which is the porous silica coated with polymer and degree of the cross linkage of polymer succeeded in finding the optimum values. Inactive column separation experiments with the simulated high level liquid waste and the optimized adsorbent revealed that decontamination factors of fission products can also be improved as well as the recovery yields.

(2011). Separation of Trivalent Minor Actinides from Fission Products Using Single R-BTP Column Extraction Chromatography. Journal of Nuclear Science and Technology: Vol. 48, No. 6, pp. 855-858.

The extraction chromatography separation experiments with the engineering scale column were carried out on a simulated waste solution. Ten times repeated separation experiments with two steps flow sheet using the CMPO/SiO2-P and the HDEHP/SiO2-P adsorbents provided durability of the adsorbents. The chromatograms obtained from the CMPO/SiO2-P column showed reproduced profiles through the multi-cycle operations. On the other hand, the time of development and peak shapes of the representative elements for HDEHP/SiO2-P column became slower and broader with the number of repeating the procedure due to leaching of HDEHP from the adsorbents.

In the case where uranium recovered by an advanced aqueous reprocessing is utilized in light water reactors (LWRs) with the thermal neutron spectrum, the effects of the decontamination factor (DF) of the reprocessing on core neutronic characteristics were examined. The amounts of transuranium (TRU) elements and fission products (FPs) contained in the recovered uranium depend on the DF of reprocessing, and also 236U is generated by neutron capture of 235U during a reactor operation. These all act as poisons in the fuel. Therefore, in this paper, the additional 235U enrichment necessary to compensate for the produced TRU elements, FPs, and 236U was evaluated for three representative DF values: 102, 103, and infinity. The low value of 102 corresponds to the advanced aqueous reprocessing investigated here. An APWR core with a discharge burnup of 49 GWd/t when the initial 235U enrichment is 4.6% was considered as the reference core. Uranium of the reference core was enriched from the natural one. It was calcul...

1,3‐[(2,4‐Diethyl‐heptylethoxy)oxy]‐2,4‐crown‐6‐calix[4]arene (Calix[4]arene‐R14) is a supramolecular compound exhibiting Cs ion recognition in high level liquid waste (HLLW). To separate effectively Cs(I) from HLLW, a novel silica‐based Calix[4]arene‐R14 polymeric material (Calix[4]arene‐R14/SiO2‐P) was prepared. It was done through impregnating a mixture of Calix[4]arene‐R14 and tri‐n‐butyl phosphate (TBP), a molecular modifier, into the pores of macroporous SiO2‐P particles utilizing a vacuum sucking technique. The sorption of some typical fission product (FP) elements Na(I), K(I), Cs(I), Rb(I), Sr(II), and La(III) towards Calix[4]arene‐R14/SiO2‐P was investigated by examining the influence of contact time and the HNO3 concentration. It was found that with an increase in the HNO3 concentration from 1.0 M to 4.0, the sorption of Cs(I) towards Calix[4]arene‐R14/SiO2‐P increased quickly and then decreased with further increase of HNO3 concentration to 5.0 M. At the optimum HNO3 concentration of 4...

Using the advanced aqueous reprocessing system named NEXT process, actinides recovery was attempted by both a simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery from the raffinate. In U, Pu and Np co-recovery experiments a single cycle flow sheet was used under high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution aided Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction accomplished Np extraction by TBP with U and Pu. Most of Np could be recovered into the product solution. In the SETFICS process, a TRUEX solvent of 0.2 mol/dm3 CMPO and 1.4 mol/dm3 TBP in n-dodecane was employed instead of 0.2 mol/dm3 CMPO and 1.0 mol/dm3 TBP in n-dodecane in order to increase the loading of metals. Instead of sodium nitrate, hydroxylamine nitrate was applied to this experimental flow sheet in accordance with ...

To separate the long-lived minor actinides (MA(III) = Am(III), Cm(III)) and some specific fission products (FP) such as Pd(II), Mo(VI), Cs(I) and Sr(II) from high-level liquid waste (HLLW), we have been studying a new partitioning process by extraction chromatography using several novel silica-based extraction resins. In this work, we examined the separation behavior of the elements contained in a simulated MA-effluent by the CMPO/SiO2 P packed column. In addition, as an attempt to further isolate Am(III) and Cm(III) from the heavy RE(III) such as Eu(III), Gd(III) and Dy(III) contained in the MA-effluent, we investigated the possibility of separation by using a silica-based cationic exchange resin. Furthermore, to isolate Sr(II) from the HLLW, adsorption and separation performances of Sr(II) and some other FP elements were studied by using a novel silica-based crown ether extraction resin, DtBuCH18C6/SiO2 P. The experimental results demonstrated that the elements in the simulated MA-effluent can be successfully separated to (1) Pd, (2) MA-hRE and (3) Zr–Mo, by CMPO/SiO2 P packed column using water and a dilute DTPA solution as eluents. Am(III) and Cm(III) are expected to be effectively separated from light RE(III) and Y(III) by the SiSCR cationic resin. However, more effective separation between Am(II), Cm(III) and heavy RE(III) such as Eu(III), Gd(III) and Dy(III) needs further approach. DtBuCH18C6/SiO2 P showed a highly selective adsorption for Sr(III) so that the Sr(II) could be completely separated from other FPs except a portion of Ba(II).

The resistant behavior of a novel silica-based octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphoshine oxide (CMPO) extraction resin (CMPO/SiO2-P) against HNO3, temperature and γ-radiation had been investigated. The adsorption properties of the treated CMPO/SiO2-P resin were evaluated by a 3 M HNO3 solution containing 0.01 M Nd(III). It was found that 3 M or 0.01 M HNO3 concentration did not decrease the stability of CMPO/SiO2-P at 25°C. The HNO3 concentration effect on the leakage of CMPO/SiO2-P resin at 80°C was obviously higher than that of at 25°C. The adsorbed amount of Nd(III) onto the treated CMPO/SiO2-P resin was determined in the range of 0.1769–0.1839 mmol/g, which basically approached to 0.1875 mmol/g, the adsorbed quantity of fresh CMPO/SiO2-P resin for Nd(III). The contents of P and C leaked from CMPO/SiO2-P resin obviously increased with an increase in the radiation dose (ID) from 1.0 to 4.1 MGy in terms of the linear equations: [P]=110.05ID+21.92 and [C]=7683.8ID−919.8. The adsorbed amount of Nd(III) onto the radiated CMPO/SiO2-P resin quickly decreased from 0.1499 to 0.0900 mmol/g according to the equation QNd(III)=−0.0222ID+0.1778, showing a large quantity of CMPO leaked from CMPO/SiO2-P resin.

The design study of an aqueous reprocessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified PUREX process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery process was proposed as the NEXT process. For the simplified PUREX process and the SETFICS/TRUEX process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 105 for the U, Pu and Np product was a reasonable target for the simplified PUREX process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 102, respectively. For economical competitiveness, an aqueous reprocessing plant with the capacity of 200 tHM/yr cleared the target cost, while the plant with the capacity of 50tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99wt%. Usin...

The extraction behavior of Am(VI) was studied for a nitric acid solution containing ammonium dihydrogenphosphate using tri-n-butylphosphate (TBP) and other extractants. TBP and HDEHP solutions diluted with n-dodecane do not reduce Am(VI) in the extraction procedure, whereas CMPO and TOA do. In TBP extraction, phosphate anion raises the separation factor of Am from Nd as high as 120 with 1 MTBP solvent from 3 M HNO3-1 M NH4H2PO4, although it suppresses the extraction of both. A distribution ratio value of 6.6 was obtained using undiluted TBP. Dissociated phosphate anion H2PO4- forms a complex that lowers the distribution ratio of Am even in highly concentrated nitrate solution at low acidity. Americium(VI) is extracted as AmO2(NO3)2.2 TBP. The stability constant, log β, for Am(VI)- H2PO4- complex was estimated as 1.2.

A numerical simulation code for the transuranium extraction (TRUEX) process was developed to determine the optimum operational conditions for the separation and recovery of transuranium (TRU) elements, such as americium. Equilibrium constants for the americium extraction by octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were determined by the least-squares method using experimental data. The activity coefficients for nitric acid and water were calculated with equations derived from experimental data. The calculated concentration profiles of americium, by using the best available equations, did not agree well with the experimental data in the extraction section. The stage efficiency was added in this code as a parameter to correct the differences between the calculated and the experimental profiles. The calculated concentration profiles using the modified code were in good agreement with the experimental data when 0.9 was used for the stage efficiency.

The mutual separation behavior of rare earth elements was studied in the system based on the neutral extractant CMPO (n-octyl(phenyl)-N,N-diisobutylcarbamoyl methylphosphine oxide) and aminopolyacetic acid DTPA (diethylenetriamine pentaacetic acid). The extractability of lanthanides decreased monotonously with increasing of atomic number in the system of 0.2 M CMPO-1.0 M TBP-n-dodecane (TRUEX solvent) and 0.05 M DTPA-NaNCO3 or hydroxylamine nitrate (HAN) solution. The separation factor of La/Lu was 6,600 using 0.05 M DTPA-3M NaNO3 solution (initial pH=2.0). The mutual separation of lanthanides in the CMPO/DTPA system was greatly improved when compared with the general TRUEX system. The separation factor was not affected by the initial pH of the solution (1.8–2.2), the salting out reagent (NaNO3 and HAN), or its concentration (2, 3 M). The experimentally obtained separation factors of lanthanide elements were reproduced by a simple model based on the distribution ratios in the TRUEX extraction system and t...

The SETFICS process, a variation of TRUEX process, was developed for the recovery of Am and Cm from acidic waste solution and the separation of actinides (III) and light lanthanides. The process uses the general TRUEX solvent as the extracting reagent and a DTPA-sodium nitrate solution for selective stripping of actinides(III). The basic flow sheet is composed of four steps: extraction-scrubbing; acid stripping; actinide(III) stripping; and lanthanide stripping. To demonstrate the usefulness of the SETFICS process, a counter current experiment was conducted using a real TRUEX product solution. Americium and curium were successfully recovered with a solution of 0.05 M DTPA-4 M NaNO3 (pH 2.0). Although the actinide(III) product solution contained Sm and Eu, the decontamination factor of 144Ce/241Am was 72, and most of the light lanthanides, specifically La, Ce, Pr, Nd, were removed. At least 80 % of the lanthanides were separated from the Am and Cm end products. In order to minimize the acidity in ...

With the aim of establishing an advanced fuel cycle enabling us to utilize all actinides in the spent fuels, separation of minor actinides in high-level liquid wastes (HLLW) has been studied using a CMPO/TBP mixture extractant. A novel flowsheet is being developed, by counter-current experiments using real HLLW from PUREX experiments, characterized by employing acidification for enhancing extraction of actinides with increasing RuNO 3+ decontamination, and a few mild salt-free reagents for fractional stripping and solvent clean-up steps. The flowsheet can give α- and salt-free HLLW providing a D F Am ≥10 3 . In the counter-current experiments for SETFICS a flowsheet has been developed to separate An(III) and Ln using DTPA, Am and Cm was recovered without light lanthanide, and the applicability of hydroxylamine nitrate as a salting-out reagent was confirmed.

For trivalent actinide and lanthanide separation, the solvent extraction system which employs neutral bifunctional extractant n-octyl(phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and aminopolyacetic acid was investigated. The system is based on CMPO-tri-n-butylphosphate (TBP) mixed solvent and diethylenetriaminepentaacetic acid (DTPA)-sodium nitrate solution. The separation is achieved by complexing trivalent actinide preferentially with DTPA in aqueous phase and extracting lanthanide into organic phase. Mutual separation between light lanthanide, Am and Cm was mainly governed by complex formation with DTPA. The acidity is most affecting parameter for partitioning and its increase results in decrease of distribution ratio. For selective stripping of trivalent actinide from 0.2 M CMPO—1.0 M TBP in n-dodecane at room temperature, 0.05 M DTPA and 2–4 M sodium nitrate solution (pH 2) is useful. From the observed distribution behavior, a basic flow sheet which is composed of four steps was prop...

Highly sensitive neutron activation analysis of uranium and thorium in high quality silica and aluminium has been investigated using the Japan Materials Testing Reactor (JMTR), having a thermal neutron flux higher than 1014 n/cm2/s. In order to determine ultra-low contents of uranium and thorium,239Np and233Pa as activation products were separated by using anion exchange and LaF3 coprecipitation methods. As a result, a number of interfering radioactive isotopes containing double neutron capture product such as183Ta were removed completely from the isolated239Np and233Pa fraction and the detection limits for uranium and thorium were found to be 2·10−12 g and 4·10−13 g, respectively.

Neutron activation analysis for thorium and uranium in high purity silica and aluminum was investigated. The activation products 233Pa and 239Np for the determination of thorium and uranium were separated from majority of interfering radioactivity by a method combined with anion exchange and LaF3 coprecipitation. It was found that the analysis with detection limit of several ppt for both uranium and thorium was possible when a high flux reactor was used. The method is useful for the determination of uranium and thorium in semiconductor grade silica and aluminum.

The leakage behaviour of a neutral chelating agent, i.e. octyl(phenyl)- N,N-di-isobutylcarbamoylmethylphosphine oxide (CMPO), from its macroporous silica-based (CMPO/SiO2-P) extraction resin (a novel polymeric adsorption material developed in the MAREC process) and the recovery of CMPO by some selected polymer- and silica-based porous adsorbents were investigated by column operation in 0.01 M HNO3at 25°C or 50°C. The concentration of CMPO in the effluent leaking from CMPO/SiO2-P was determined as ca. 48 ppm at 25°C and ca. 37 ppm at 50°C. The corresponding elution volumes and the amounts of CMPO leaking were 8700.9 BV/g and 239.3 mg/g at 25°C and 11 842.9 BV/g and 359.4 mg/g at 50°C, respectively. The adsorption experiment showed that a polymer-based porous material, SEPABEADS® SP-825, was capable of recovering CMPO effectively from 0.01 M HNO3. Its adsorption ability towards CMPO was considerably greater than those of macroporous silica-based SiO2-P particles, polymer-based Amberlite™ XAD-7 and activated carbon. The bed volume and the amount of CMPO adsorbed by SEPABEADS® SP-825 at the breakthrough point at 25°C were 13 549.2 BV/g and 498.7 mg/g, respectively, thereby demonstrating that SEPABEADS® SP-825 was promising for the recovery of CMPO from Am(III)- and Cm(III)-containing solutions in the MAREC process.

A conceptual design study for some alternatives as well as the advanced aqueous reprocessing system has been progressed as a feasibility study on commercialized fast reactor cycle systems. Among these alternatives, the Super critical fluid Dir ect Ex traction (Super-DIREX) process is expected to minimize the reprocessing cost because the heavy metals (U, Pu, Np, Am and Cm) are directly extracted from the spent fuel powder in a column covering the dissolution, clarification and extraction process of the advanced aqueous process. The conceptual design for the reprocessing plant using Super-DIREX process was conducted considering the flowsheet, mass balance, specification and number of the components and layout of the components. From the result of the evaluation for the reprocessing cost of the plant with the capacity of 50tHM/y, it is found out that about 30% of the reprocessing cost is reduced compared with the advanced aqueous process (the NEXT process). In addition, using the Super-DIREX process, about 30% of wastes generated from the plant are reduced.Copyright © 2004 by ASME

In the extraction chromatography system, accumulation of hydrogen gas in the chromatography column is suspected to lead to fire or explosion. In order to prevent the hazardous accidents, it is necessary to evaluate behaviors of gas radiolytically generated inside the column. In this study, behaviors of gas inside the extraction chromatography column were investigated through experiments and Computation Fluid Dynamics (CFD) simulation. N2 gas once accumulated as bubbles in the packed bed was hardly discharged by the flow of mobile phase. However, the CFD simulation and X-ray imaging on γ -ray irradiated column revealed that during operation the hydrogen gas generated in the column was dissolved into the mobile phase without accumulation and discharged.

Concerning decontamination of a radioactive liquid waste which includes seawater and nuclides from irradiated fuels and activated materials, the in-situ generation of metal hexacyanoferrates(II) by adding potassium hexacyanoferrate(II) and codecontamination of 134,137Cs and some activation products were investigated. Transition metals arising from seawater in the waste solution precipitates in the preference order of Zn>Ni≥Co>Mn according to their solubility. The precipitate adsorbs 134,137Cs, and decontamination will be attained by the following sedimentation with a polymer and filtration, as an example. Decontamination factor of activated products, 60Co and 54Mn, is dependent on concentration of hexacyanoferrate(II) in the solution.