Advanced-ORIENT Cycle Project: Summary of Phase I Fundamental Studies

Available online at www.sciencedirect.com

Procedia Chemistry 7 (2012) 222 – 230

ATALANTE 2012 International Conference on Nuclear Chemistry for Sustainable Fuel Cycles

Advanced-ORIENT Cycle Project: Summary of Phase I Fundamental Studies a b Shin-ichi Koyama *, Tatsuya Suzuki , Masaki Ozawab, Kiyoko Kurosawac ,a Reiko d e f a Fujita , Hitoshi Mimura , Ken Okada , Yasuji Morita ,Yasuhiko Fujii a

Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan b Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550, Japan c Kaken Co., Ltd., 1044 Horimachi, Mito, Ibaraki 310-0903, Japan d TOSHIBA Corporation Power Systems Company, 4-1 Ukishimacho, Kawasaki-ku, Kawasaki 210-0862, Japan e Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aoba-ku, Sendai 980-8579, Japan f National Institute of Advanced Industrial Science and Technology, 1-1-1 Higashi, Tsukuba, Ibaraki 305-8565, Japan

Abstract The Advanced Optimization by Recycling Instructive Elements (Adv.-ORIENT) Cycle strategy has been proposed as a basic concept for trinitarian research on the separation, transmutation, and utilization (S, T, & U) of nuclides and elements on the basis of the FBR fuel cycle. Working in this direction, validation of the principal separation methods and related safety research were performed from 2006 through 2011 as the first phase. The separation scheme was composed of four ion exchange (IXC) steps and one catalytic electrolytic extraction (CEE) step. The fundamental technological aspects are summarized as the Phase I program. © 2012 2012Elsevier The Authors. Published and/or by Elsevier B.V. under responsibility of the Chairman of the ATALANTE 2012 © B.V...Selection peer-review SelectionCommittee and/or peer-review under responsibility of the Chairman of the ATALANTE 2012 Program Program Keywords: Adv.-ORIENT Cycle, separation, rare metal fission product, actinide, lanthanide, utilization

* Corresponding author. Tel.: +81-29-267-4141; fax: +81-29-267-7130 E-mail address: [email protected]

1876-6196 © 2012 Elsevier B.V...Selection and/or peer-review under responsibility of the Chairman of the ATALANTE 2012 Program Committee doi:10.1016/j.proche.2012.10.037

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1. Introduction The PUREX process is now the main reprocessing method for conventional light water reactor systems.[1] The separation of fission products (FPs) and minor actinides (MAs) from the generated high-level liquid waste (HLLW) should be of interest as a means of more effectively utilizing resources for future advanced nuclear recycle systems as well as reducing the environmental impact. To achieve the ultimate minimization of ecological risks and utilization of new elements/nuclides obtained from nuclear fuel recycling, a new fuel cycle paradigm should be designed. The Advanced Optimization by Recycling Instructive Elements (Adv.-ORIENT) Cycle project concept has been proposed as an integrated technology (Fig. 1) [2].

Cs,Se,Sn Isotope Separation

Elution

䛆IXC(II)䛇

135Cs

(Transmutation)

137Cs-Zeolite

Application for Heat and/or Radiation Source

Mo

RMFP Recovery by CEE

Multi-functional Reprocessing

䛆IXC(III)䛇

Separation

Application for 99mTc Source (Near Surface and/or Deep Repository)

Ln 90Sr

-Zeolite

Application for Heat (Sr battery䠅 and/or Radiation Source

Ru,Rh,Pd,Tc,etc

Pd

Pd, Ru, Rh, Tc, Sb, etc

MA in An Separation U, Pu, Np 䛆IXC(IV)䛇 /Purification Am, Cm Process Am Am, Cm by IXC Mutual Sr

Other FPs

Fast Breeder Reactor

Reuse of Filter

RMFP Separation by IXC Filter

137Cs

ElectroCatalyst

䛆CEE䛇

Tc Fuel Fab.

Isotope Separation

133Cs

LLFP Removal by Inorganic Ion Exchanger

Application for Catalysts for Hydrogen Generation and/or Fuel Cell

Advanced Optimization by Recycling Instructive Elements Cycle Spent fuel EX-Cycle 䛆IXC(I)䛇

Pd 107Pd

Application as longlived Tc battery Catalyst after Transmutation 99Tc(n,ȕ-)100Ru

Pu (Cm ĺ)

Cm

Strategic Materials

IN-Cycle

Suzuki/Ozawa

Fig. 1 Advanced-ORIENT cycle concept

The main goals of this project were the selective separation of our elements of interest, the effective transmutation of the minor actinides (Ans) and long-lived fission products (LLFPs), and the utilization of rare metals such as the light platinum group metals (PGMs: Ru, Rh, Pd) and Tc as well as middle-lived fission products (MLFPs) such as Cs and Sr. Because an ion chromatographic technique was used as the main separation process, it was difficult to design the process conditions for engineering scale applications. Therefore, validation of the principal separation method and related safety research were performed as the first phase of the project from 2006 through 2011. In this paper, the fundamental technological aspects, except for the actual transmutation research, are summarized as the Phase I program.

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2. Separation Process Chemistry Thirty-one elements were defined as Nuclear Rare Metals (NRMs) in this program. The defined NRMs are shown in Fig. 2. For this determination, the produced weight of each element in the spent fuel (>10g/tHM) was considered and the noble gases, halogens, Cd, Sn, Sb, Bk, and Cf were excluded [3, 4]. H Li

Fission Product

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Defined Nuclear Rare Metal

Actinide

Be 䖃䠖 Rare Metal, 䕔䠖 Atrifical Radionuclide 䚷䕿䠖 Stable, 䕕䠖 Short Lived (T1/2
B

C

N

O

F

Ne

Al

Si

P

S

Cl

Ar

Na

Mg

K

Ca

Sc

Ti

V

Cr

Mn

Fe

Co

Ni

Cu

Zn

Ga

Ge

As

Se

Br

Kr

Rb

Sr

Y

Zr

Nb

Mo

Tc

Ru

Rh

Pd

Ag

Cd

In

Sn

Sb

Te

I

Xe

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Pt

Au

Hg

Tl

Pb

Bi

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At

Rn

Eu

Gd

Tb

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Er

Tm

Yb

Lu

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Ra

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Ir

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La

Ce

Pr

Nd

Pm

Sm

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Fig. 2 Defined nuclear rare metals

2.1. Cs/Sr Separation in the IXC(I) step For the selective separation of Cs, Cs-selective silica gels adsorbents loaded with ammonium molybdophosphate (AMP-SG) or ammonium tungstophosphate (AWP-SG) were developed by successive impregnation/loading methods (H3Mo12O40P + 3 NH4NO3 ĺ (NH4)3PO4·12MoO3(AMP) + 3 HNO3). The uptake of Cs+ by AMP-SG in the presence of 2.5 mol·dmí3 (M) HNO3 reached equilibrium within 1 h, and a relatively large uptake above 99% was achieved. The AMP-SG exchangers exhibited relatively large Kd values (~103 cm3/g) for Cs+, even in the presence of 2–3 M HNO3 and 1 M NaNO3, which are close to the concentrations in HLLW. In addition, the selective uptake of Cs+ by AMP-SG was confirmed using simulated HLLW (the experimental fast reactor “Joyo” fuel was used), and an uptake of 96% was obtained (Table 1) [5]. Table 1 Uptake (%) of 137Cs from a simulated HLLW (Joyo)

Adsorbent samples AMP-SG AMP-CaALG

ref.[5]

AWP-CaALG

ref.[5]

137

Cs-HLLW/ȝg 6.39 6.39 6.39

137

Cs-filtrate/ȝg 0.26 0.20 0.04

Uptake/% 95.9 96.9 99.4

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Sr-selective adsorbents composed of microcapsules (MCs) containing crown ether compounds were developed using sol-gel methods. The crown ether compound decyl-18-crown-6 (D18C6), with a selective uptake ability for Sr, was used as the active component, and calcium alginate (CaALG) microcapsules containing D18C6 were developed for the efficient separation of Sr2+ from a concentrated HNO3 solution. Hybrid MCs (CLD-MC) enclosing D18C6, sodium lauryl benzenesulfonate (Na-LBS), and 1-decanol had relatively large Kd values above 80 cm3/g, even in the presence of 3 M HNO3. Using a CLD-MC column, the breakpoint and breakthrough capacity were estimated to be 99 cm3 and 1.6 × 10í2 mmol/g, respectively. 㻝㻜㻜 㻥㻜

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Fig. 4 EDS of the surface of the deposit RMFP from the simulated HLLW of HCl media

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2.2. Rare Metal Fission Product (RMFP) Separation and Utilization of a CEE step Catalytic electrolytic extraction (CEE) [6] was performed on simulated HLLW in HCl and HNO3 media. The reduction ratios for Pd, Rh, Re (simulator of Tc), Ru, Se, and Ag were more than 70% for CEE of the simulated HLLW in HCl media (Fig. 3), whereas the reduction ratios for Mo, Zr, and Nd were less than 10%. In the case of the simulated HLLW in HNO3 media, although the reduction ratio for Pd was the highest at 85%, those for Ru, Rh, and Re were only approximately 20% and those for Zr, Fe, and Nd were less than 10%, as was observed for the HLLW in the HCl media. Figure 4 shows the elemental map of the surface of the deposits obtained from the simulated HLLW in HCl media using energy dispersive spectroscopy (EDS) analysis. Pd, Rh, Ru, Mo, and Re were detected on the surface, whereas Te, Fe, Nd, and O were not detected. The form of the RMFP deposits was almost metallic and not oxidic. The shape of the deposits from the simulated HLLW in HCl media was dendritic. In contrast to previous experiments [7], the shape of deposits from the simulated HLLW in HNO3 media was not dendritic, but rather uniformly spherical. Therefore, the deposit behavior of the RMFPs in different HLLW media must be investigated in detail. 2.3. An and RMFP Separation in the IXC(II), (III), and (IV) steps RMFPs were adsorbed using a gelated tertiary pyridine-type resin (TPR) [8] and a dilute HCl solution in the ion exchange step [IXC(II) step]. It was confirmed that the PGMs including Ru, Rh, and Pd were separated from the actual spent fuel. The detailed distribution of the elements in the spent fuel was also investigated in the IXC(II) process using a simulated spent fuel solution containing U but no MAs.

Concentration / arb. units

Zr, Sr, Cs, RE, 1

Mo

0.5

Se Te

0 0

50 Effluent volume /ml

100

Fig. 5 Main elements eluted in the IXC(II) process. (HCl concentration/0.5 M. Column volume/8 mL)

In this experiment, Re was used as a substitute for Tc. It was confirmed that the TPR is effective for recovering or removing PGMs, Tc(Re), Ag, Cd, and Sb from the spent fuel. In other words, these elements, when present in a dilute HCl solution, are strongly adsorbed onto the TPR. However, approximately 10% elution of Ru and Rh in the early phase was observed, whereas the remaining approximately 90% of Ru and Rh were not observed in the eluted solution. The main elements eluted in the IXC(II) process are shown in Fig. 5. All alkali metal elements, all alkaline earth elements, and all rare earth elements as well as Zr were not adsorbed onto the

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TPR. Mo, Te, and Se were slightly adsorbed on the TPR. However, nearly the entire Se content may be removed from the feed solution in the IXC(II) step. It should be noted that a method for removing PGMs from the TPR, their separation from Tc using thiourea, and a removal method for Tc using a highly concentrated HCl solution (6–9 M) have already been established [9]. Because U(IV), U(VI), Pu(IV), and Np(IV) are strongly adsorbed by the anion-exchange resin, highly concentrated 12 M HCl was used for elution of the Ln(III) compounds. The eluent was then changed to 6M HCl for An(III) separation. By changing the eluent further to 2M HCl, the Pu/Np group with a small amount of U was eluted from the resin [IXC(III) step]. The eluent was finally changed to 0.1 M HCl for recovery of the U residue [IXC(III) step]. A mixed solution of methanol (MeOH) and conc. HNO3 was then used as the first eluent for the mutual separation of Am and Cm in the IXC(IV) step. The decontamination factor for Am from Cm exceeded 103. The detailed procedures and results are cited in references 10 and 11. 3. Utilization of the Nuclear Rare Metals 3.1. Utilization of the separated Cs/Sr Zeolites that adsorb Cs and Sr can be converted to stable ceramic solid forms. For example, Cs forms of zeolite A and mordenite are readily converted to CsAlSi2O6 and CsAlSi5O12, respectively, by heat treatment. These solid forms have the potential for use as heat and radiation sources; the central temperatures of the abovedescribed Cs-solid forms with diameter 10 cm were estimated to be 776 qC and 442 qC, respectively. These solid forms have potential application in thermoelectric power generation through the attachment of thermoelectric materials (thermoelectric conversion elements). Thermoelectric materials should have a high figure of merit (Z) with high thermoelectric power and electric conductivity and low thermal conductivity. 㻝㻞㻜

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Fig. 6 Relationship between the cathode current corresponding to hydrogen evolution at í1.25 V vs. Ag/AgCl and the initial hydrogen evolution potential (ĭHinit) of RMFPs from simulated HLLW in 1 M NaOH

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3.2. Utilization of RMFP-deposited electrodes The RMFP-deposited electrodes obtained from CEE step were investigated as possible hydrogen evolutional electrodes. The cathode currents corresponding to hydrogen evolution at the given polarized potential (ĭ = í1.25 V vs. Ag/AgCl) as a function of the initial hydrogen evolution potential (ĭHinit) are shown in Fig. 6. The initial hydrogen evolution potential ĭHinit was obtained by the extrapolation of the cathodic IE curves. The RMFPdeposited electrodes from the simulated HLLW belong to the first quadrant and appear to have a higher catalytic reactivity than the smooth Pt electrode. Therefore, the catalytic reactivity of the RMFP deposits for hydrogen evolution must be further explored. 3.3. Creation of RMFPs as a strategic metal Partitioning and transmutation (P&T) have been researched during the past two decades as a way to decrease the long-lived radioactivity enclosed in HLLW. Iodine-129 and 99Tc are two nuclides among several LLFPs currently under investigation, and theoretical and experimental studies of the P&T for 99Tc have been conducted. These nuclear transmutations are based on the (n,Ȗ) reaction followed by ȕí decay to promote the change of “Z” to the right “Z+1” elements in the periodic table. The idea to create new rare metals is based on the utilization of this reaction at the LWR/FBR by optimizing the nuclear conditions, i.e., neutron energy versus neutron absorption cross section. A A+1 ĺ Z+1FPA+1 ZFP (n,Ȗ ) Z FP In the case of the transmutation of radioactive Ru, reloaded and irradiated RMFP Ru at an FBR (neutron flux: 2.27 × 1015[n/cm2/s], four cycles, 800 days/cycle [3, 4]) resulted in the creation of Rh and Pd with transmutation ratios of 3.5% and 3.7%, respectively. With such transmutations, the time for the radioactivity of these created elements to decrease to their exemption levels can be significantly reduced. 4. Safety Research 4.1. Corrosion Evaluation of Metals in Hydrochloric Acid Media To evaluate the applicable construction metals with corrosion resistance to HCl media, four primary candidate metals—Ta, Zr, Nb, and Hastelloy-B (28% Mo–Ni)—were selected along with SUS316L as a reference. Using 20 mm × 50 mm × 2 mmt samples, immersion, gas exposure, and electrochemical tests were performed in pure HCl and HCl-type simulated HLLW solutions. One immersion test was performed at room temperature for a maximum of 300 days and a second immersion and exposure test was conducted at 90 qC for 14 days. Measurement of corrosion rate, corrosion form, and dissolved elements resulting from corrosion as well as the observation of the surfaces and cross-sections of the specimens were conducted. Based on all results, it was concluded that Ta exhibits complete anticorrosive properties with a corrosion rate of less than 0.05 mm/y under all conditions. On the other hand, the corrosion rate of Nb, Zr, and Hastelloy-B exceeded 0.1 mm/y without pitting corrosion. In addition, the anticorrosive properties of Hastelloy-B at RT and Ta at 90 °C were confirmed in highly concentrated HCl media in order to determine their engineering feasibility. Ta is the most promising metal with perfect anticorrosive properties under all HCl conditions. However, Ta is a very expensive metal, and it is not realistic to fabricate all materials of a practical plant using Ta. For a practical study of the Adv.-ORIENT cycle project, it is important to consider economic efficiency when selecting the plant materials for the fuel cycle. The corrosion resistance should also be considered under the process conditions including radiation, solvent conditions, and temperature. It is also important to further investigate and practically evaluate other alloys and lining materials.

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4.2. Thermal Stability of the Tertiary Pyridine Resin (TPR) With the application of an exotic ion exchange resin such as TPR in the reprocessing steps, evaluation of its thermo- and radiochemical stability was important. Therefore, a miniature scale experiment was conducted, and it was found that no hazardous reaction occurred in the TPR/MeOH/HNO3 system up to 35°C. However, above 150°C, an exothermic reaction of the NO3-type TPR or the TPR/HNO3 system was observed [12]. The thermochemical properties were evaluated in reference to tri-butyl phosphate (TBP), a fully mature technology used for conventional reprocessing. A gram-scale heating experiment was conducted to evaluate the exothermic behavior of a TBP/HNO3 system compared with that of a TPR-HNO3 system. In the case of the TBP-HNO3 system, the organic phase was a liquid and convection did not exist in the vessel. No thermal accumulation was observed. In the TPR-HNO3 system, a runaway reaction did not occur up to approximately 200°C whereas the solvent remained. However, after drying out of the solvent, a reaction between TPR and HNO3 did proceed [13]. According to precise calculations based on this experiment, TPR will be ignited at 109°C after 5 h of holding time. From these results, it was concluded that TPR should not be stored dry in the presence of HNO3 in order to prevent fires and explosions. 5. Conclusions and perspectives The fundamental aspects of lab-scale experiments have been obtained during the Phase I program of the Adv.-ORIENT Cycle project. Based on the results of these experiments, it can be concluded that the concept of the Adv.-ORIENT Cycle is established as an assembly of each integrated technology. Therefore, its feasibility as a process should be evaluated and the requirements for safe operation should be solved with engineering scale designs. As for the process flow, an HCl solution in the IXC(II), IXC(III), and CEE steps should be used to increase the recovery rates of the target elements. In a conventional reprocessing process, HNO3 is used for fuel dissolution; therefore, a conversion process between HCl and HNO3 will be needed if HCl is chosen as the solvent. To address this issue, preliminary tests are now underway to convert between HNO3 and HCl via electrodialysis using a porous ceramic membrane tube. In addition, it is favorable to use the same solvent throughout all separation processes. Therefore, fuel dissolution in HCl media should be considered. The corrosion properties of structural materials in HCl solution were evaluated for some candidate metals. A cost evaluation is needed to make economically sound decisions in the next phase. Finally, the use of HCl media should be fixed on the basis of the results of all research and development activities during the Phase I program of the Adv.-ORIENT Cycle project. The policies of the phase II program will depend on the summary of the phase I program. There will be two possibilities. One is the continuation of basic research to evaluate the most favorable processes on the basis of lab-scale experiments. The other is to shift to larger scale experiments and consider engineering design aspects. In all cases, a sequential demonstration process using actual spent fuel as hot experiments should be performed. And also the cost estimation should be needed to assess its feasibility of this project and the benefit compared with conventional reprocessing process in the next phase II program. References [1] G. Choppin, J.Rydverg, J. O. Liljenzin, “RADIOCHEMISTRY and NUCLEAR CHEMISTRY”, 2nd Edition of Nuclear Chemistry, Theory and Applications, ISBN 0-7506-2300-4. [2] M. Ozawa, S. Koyama, T. Suzuki, R. Fujita, H. Mimura, Y. Fujii, “Advanced Orient Cycle, Toward Realizing Intensified Transmutation and Utilization of Radioactive Wastes”, Proc. of the Global 2007, Boise, USA, Sep., (2007)

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[3] S. Koyama, T. Suzuki, M. Ozawa, “From Waste to Resource, Nuclear Rare Metals as a Dream of Modern Alchemists,” Energy Conversion and Management, 51, issue 9, pp.1799-1805, (2009). [4] K. Sato, I. Amano, et al., “Feasibility Study on Commercialization of Fast Breeder Reactor Cycle Systems Interim Report of Phase II Technical Study Report for Nuclear Fuel Cycle Systems-”, JNC TN9400 2004-03, June, (2004) (in Japanese) [5] H. Mimura, Y. Wu, W. Yufei, Y. Niibori, I. Yamagishi, M. Ozawa, T. Ohnishi and S. Koyama a, “Selective Separation and Recovery of Cesium by Ammonium Tungstophosphate-Alginate Microcapsules”, Journal of Nuclear Engineering and Design, 241, 12, pp.4750-4757, (2011) [6] M. Ozawa, M. Ishida, Y. Sano, “Strategic Separation of Technetium and Rare Metal Fission Products in Spent Nuclear Fuel: Solvent Extraction Behavior and Partitioning by Catalytic Electrolytic Extraction”, Radiochemistry, 45, 4, pp.225-232, (2003) [7] M. Ozawa, T. Suzuki, S. Koyama, I. Yamagishi, R. Fujita, K. Okada, K. Tatenuma, H. Mimura and Y. Fujii, “Adv.-ORIENT Cycle, Its Scientific Progress and the Engineering Feasibility” , Proc. of GLOBAL 2009, Paris, France, Sep., (2009) [8] M. Nogami, Y. Fujii, T. Sugo, “Radiation resistance type anion exchange resins for spent fuel treatment”, J. Radioanal. Nucl. Chem., 203, 109, (1996) [9] T. Suzuki, Y.Fujii , Y. Wu, H. Mimura, S. Koyama, and M. Ozawa, “Adsorption behavior of VII group elements on tertiary pyridine resin in hydrochloric acid solution”, J. Radioalal. Nucl. Chem., 282, p.641, (2009) [10] S. Koyama, M. Ozawa, T. Suzuki and Y. Fujii, “Development of Multi-functional Reprocessing process Based on Ion-exchange Method by Using Tertiary Pyridine-type Resin,” J. of Nuclear Science and Technology, 43, 6, pp.681–689, (2006) [11] T. Suzuki, Y. Fujii, S. Koyama, M. Ozawa, “Nuclide separation from spent nuclear fuels by using tertiary pyridine resin,” Progress in Nuclear Energy, 50, pp.456-461, (2008) [12] Y. Sato, K. Okada, M. Akiyoshi, T. Matsunaga, S. Koyama, T. Suzuki, M. Ozawa, “Thermochemical safety evaluation of tertiary pyridine resin for the application to multi-functional reprocessing process -Adv.-ORIENT cycle development-”, Progress in Nuclear Energy, 53, Issue 7, pp.988-993, (2011) [13] S. Koyama, T. Suzuki, H. Mimura, R. Fujita, K. Kurosawa, K. Okada, M. Ozawa,” Current status and future plans of Advanced ORIENT Cycle strategy”, Progress in Nuclear Energy, 53, Issue 7, pp.980-987, (2011)