Int. Conf. Accel. -driven Transmu. Tech. Appl. (Las Vegas, July 25-29, 1994)

Plutonium(TRU) Transmutation and 233U Production by Single-Fluid Type Accelerator Molten-Salt Breeder (AMSB)


Kazuo Furukawa*, Yoshio Kato*1, and Sergey E.Chigrinov*2
*Inst. of R & D, Tokai University, Hiratsuka, Kanagawa 259-12, Japan
*1 Dept. Chem. of Fuels Research, Japan Atom. Ene. Res. Inst. , Tokai, Ibaraki 319-11, Japan
*2 Radiation Phys. & Chem. Problems Inst., Academy of Science, Sosny, Minsk, Belarus


Abstract: For a near term, practical & industrial disposition of Pu(TRU) by an accelerator facility, considerations beyond high levels of physical soundness and safety should be required: (1) only a few R&D; items including radiation damage, heat removal and material compatibility; (2) reductions in operation/maintenance/processing works; (3) reduced production of radioactivity; (4) economical energy production. This will be achieved by the new modification of Th-fertilized Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy foundation for next century might be laid.



In addressing the Pu disposition issue, our final target should be:
(1) near complete elimination of the world's Pu stocks (cf. US-NAS report[1]), and achieving the following:
(2) economical utilization of the valuable energy potentiality of Pu (and TRU), and
(3) establishment of a practical, global, nuclear-energy industry in the next century.
For the complete disposition of weapons-grade Pu or reactor-grade Pu, the application of Th-233U fuel recycle system will be the best. However, its practical/economical realization would depend on fluid-fuel concepts due to the high gamma activity of 233U fuel, which is extremely effective for nuclear-proliferation and the prohibition of use by terrorists.
The only one, Molten Fluorides Fuel, among several fluid-fuel concepts, whose practicality has been verified by the long effort of Oak Ridge National Laboratory during 1947-80[2]. They aimed to establish a Fission Breeding power station (MSBR). However, it has several serious difficulties in design which effect its global applicability for the future. Therefore, our group is proposing more practical approach, a Th-breeding synergetic system: THORIMS-NES (Thorium Molten-Salt Nuclear Energy Synergetics") [3], which is composed of: [I] simple fission power stations [Molten-Salt Reactor (MSR) : FUJI-series] [4], which is a self-fueling reactor without continuous chemical processing or core-graphite exchange, [II] fissile-fuel producers by spallation/fission reactions of 1 GeV-proton [Accelerator Molten-Salt Breeder (AMSB) : ASO-series] [5], and [III] dry process plants.



The complete Pu disposition by this THORIMS-NES might be established in the following three plans:
(1) D-plan: Pu (and Trans-Uranium elements [TRU]) separation via Dry-process from the spent solid-fuels. No longer depending on the present Purex, the new Dry-process by Molten-Fluoride technology should be developed so as to improve economy and diversion resistance, and which will supply the fluoride salts for F- and A-plans. The technological basis has been examined by French, Russian, etc.

(2) F-plan: Pu-burning and 233U-production by MSR [FUJI-Pu]. The most practical, safe technology for effective Pu-incineration will be an application of MSR fertilized by Th[6]. A simple system, FUJI-Pu, has been studied, verifying the rapid and continuous Pu-burning by the stepwise isolation of 233U, which is produced for the next Th-U cycle operation[7],[8]. (3) A-plan: Pu-burning and 233U-production by AMSB-Pu [ASO*-Pu]. In parallel with the F-plan, spallation/fission reactions in AMSB will be applicable for the same purpose. A new design named AMSB-Pu or ASO-Pu has been developed with a goal of efficient production of 233U; 2-3 times higher than the Pu-incinerator (F-plan), and some excess electricity production[9], [8].

In the final stage, we would have an orthodox THORIMS-NES fully dependent upon the pure Th-233U fuel cycle. This system promises a global Nuclear Energy Industry with the following characteristics: (a) extreme safety: no severe accidents; (b) anti nuclear-proliferation and terrorism; (c) less Radio-wastes due to nearly no TRU produced and few maintenance/operational works (low-level waste production); (d) very few nuclear materials transportation; (e) small R&D; cost due to few, new items: (f) high economical potential; and (g) real establishment of an ideal Breeding Fuel Cycle, resulting a small need of Th of only about 2 M tons for about 1000 TWe-Year production in the next century [cf. the past nuclear production of 2 TWe-Year][8].


The basic idea of AMSB was invented on 1980, and is based on the single phase Molten-Salt target/blanket concept[10], which is extremely simple in configuration [cf. Fig. 1]. The design largely solves several serious problems related to (i) radiation damage, (ii) heat removal, (iii) spallation chemistry, & (iv) target shuffling. The biggest remaining problem will be the proton beam injection port engineering, which may be solved by a real beam test which increases intensity step by step[11,12].

The AMSB concept also has several modifications, such as (A) Standard AMSB [AMSBst], (B) High-gain AMSB [AMSBhg], (C) Super-gain AMSB [AMSBsg], (D) High-gain Pu-burn AMSB [AMSB Puhg], (E) Super-gain Pu-burn AMSB [AMSB-Pusg]. The words "standard", "high-gain" or "super-gain" means, respectively, no, medium or large amounts of fissile components within target salts, respectively, which would result few, medium or significant fission events, with the varying levels of fissile production and heat generation.

The relationship between two different, mole-fraction salt compositions with a varying fissile content and the resultant total thermal output in 1 GeV-300 mA proton facility is shown in Fig. 2. The operation of such facility's accelerator will require about 600 MWe, obtained from the facility's generated heat of 1400 MWth. The first stage, hg-type will be easier to develop, but the matured stage, sg-type, could be operated to supply an excess electricity of 1 GWe to the public, thereby improving economy.

For Pu disposition, (D) High-gain Pu-burn AMSB [AMSB-Puhg] and (E) Super-gain Pu-burn AMSB [AMSB-Pusg] are especially important. Their performance has been preliminarily examined in 1993[8], [3]. These conservative values are presented in Table 1 along with the performance of FUJI and FUJI-Pu.


Table 1. Preliminary Performance of FUJI [per 1 GWe] and AMSB[1 GeV-300 mA]

Pu-inv. 233U-inv. Pu burn/Y 233U prod/Y Elec.output
FUJI-Pu 3 t - 0.86 t 0.7 t 1 GWe
FUJI - 2 t self-sust. 1 GWe
AMSB-Puhg 4 t 3 t 0.35 t 0.7 t 0 -
AMSB-Pusg 5 t 5 t 0.6 t 1.0 t 0.5-1 GWe



The global nuclear energy demand during the next century could be huge, as predicted in Fig. 3 [3] [8], which is an extension the original work of Marchetti[13], because his projections are not even enough to solve the CO2 Greenhouse effect.

U-Pu cycle system could not produce even the amount shown by the dotted curve in Fig.3 (B), and is also unrealistic due to the huge Pu handling. However, to realize the 1,000 TWe-Year production in the next century, there many scenarios based on THORIMS-NES by using the above D-, F-, and A plans. Here, a simple example has been shown in Fig.4 and Table 2.

Possibly, the U-Pu cycle power stations system sizes might be up to 4 times larger than present, decreasing the dotted curve in Fig. 3 (B). Even this low will still produce more than 10,000 ton Pu till 2050, which will be separated by Purex or D-plan process,with accompanying TRU in all the more proliferation resistant forms, because storage is not a solution.

Pu (TRU) disposition could be started from 2010 by F-plan, and from 2020 by A-plan in parallel. The former activity will reach a 500 GWe maximum at about 2030, burning about 7,400 tons Pu (TRU) or more. The latter will require 400 facilities at the peak at ~2040, burning about 3,000 ton Pu(TRU) or more.

The duty of FUJI-Pu will be finished at ~2050. However, with no modifications, operation as Th 233U power stations, can continue until the end of the reactor life. The technological development of AMSB-Pu will be significant among 2020 and 2050. Initially, AMSB-Puhg will be deployed in the first stage and not produce any excess electricity. The next AMSB-Pusg will produce about 0.5 GWe/facility, gradually improving in performance till 1 GWe or 2 GWe/facility.

Generally, FUJI-Pu will be more economical than AMSB-Pu for Pu disposition. However, after the middle of 2040's decade, in which Pu would be almost eliminated, AMSB-Pu would be gradually replaced by AMSMsg of higher performance by operating near criticality. This is essential to establish the THORIMS-NES. At 5-10 years before the peak on the nuclear energy demand curve (Fig. 3), which would occur at 2055-65 in our tentative prediction. AMSB-Pusg would be dismantled, and the 233U fissile recovered, which is useful for fueling more FUJI power stations. Therefore, the main role of AMSB will be only last less than 40 years, although it could be useful for radio-waste incineration as a continuing minor work[5].

Table 2. Complete Pu Elimination by THORIMS-NES and
Replacement of U-Pu Reactors by Th-U Reactors [cf. Fig. 4]

Total U-Pu Cumulative D-Plan F-Plan A-Plan
Capacity Stations Pu Pu supply FUJI-Pu, FUJI AMSB-Pu
(GWe) (GWe) (ton) (t/10Y) (GWe) (fac. /10Y)
2000 300 300 1,200
2010 550 550 2,500 start start
2020 1,000 750 4,400 990 200, 50 start
2030 1,850 800 7,000 4,170 500, 550 hg: 100
2040 3,460 550 9,300 5,920 300, 2,210 sg: 400
2050 6,800 200 10,600 0 0, 6,200 sg: 400*

Th-U Pu 233U
Stations burn produc.
(GWe) (t/10Y) (t/10Y)
2000 [ /10Y : cumulative values in
2010 10 years till the date]
2020 250 390 310
2030 1,050 2,870 2,450 [* : replacing by improved
2040 2,910 4,320 4,480 AMSBsg gradually]
2050 6,600 3,090 7,380*


The technological basis of the F-plan has been prepared by the excellent efforts of 0RNL during the 1947-76 period. Therefore, the commercialization of FUJI in smaller size (100-300 MWe) could be performed in at least 15 years. Such commercial nuclear power stations should be simpler in configuration/operation/maintenance, and the power size will be flexible; unlike the Fission or Spallation Breeding power stations [MSBR or AMSB].

Based on the initial developmental effort of MSR basic technology, A-plan developing AMSB-Pu could proceeded after about 10 years from FUJI's deployment. The most important items for R&D; of AMSB-Pu would be the following:

(A) 1 GeV, 100-300 mA proton Linear Accelerator: In the target/blanket salt system a little lower voltage, 1 GeV, will be convenient due to deeper beam penetration. which is effective for heat removal, although it should be optimized in final design.

As a long term program, more economical non-monochromatic beam accelerator development should be encouraged as more of an industrial machine than a research Linac.

(B) Injection port engineering: Is the most serious, unresolved item. However, the vapor of salt might be mostly condensable on duct wall (cf. Fig. 1). which might allow applying several add-on techniques such as electrostatic collection. Gaseous species in molten salt should be carried away to be separated in outside of reactor core.

(C) Accuracy of neutronic calculations: The target/blanket salt system of AMSB contains several kind of nuclei including light ones. Thus, neutronic calculations are not easy and of low accuracy in predicting reaction products yield and heat generation rates. Furukawa and Kato unsuccessfully tried to obtain an experimental analysis of a large target salt block using the help of SIN (now PSI) group in 1981, but now planning it under the cooperation with Russian group. It will also be valuable for the development of the spallation theorem in general.

(D) Reactor chemical aspects: Several chemical issues relating to "spallation chemistry" has been successfully examined[11, 12, 5] based on the MSR chemistry developed by 0RNL. The chemical processing procedures of salt will be more flexible (less stressing) in our substantially subcritical system. The transmutation of hazardous radioisotopes could be accomplished by a minimum separation work of simply circulating them through the target/fuel salt cycle in THORIMS-NES.

(E) System engineering design optimization: The size of target/blanket system is important regarding [a] radiation damage of reflector graphite and reactor vessel wall, [b] inventory of fissile, fertile and reaction products for producing optimal reactor performance, and finally [c] total economy.

The optimization of the salt composition and the operation conditions of this facility could be improved, over time, depending on the development progress of several technological items.
The time schedule of this developmental program has been shown in Table 3.


For the practical industrialization of Pu (TRU) disposition, the entire system should be constructed as a positive endeavor seeking not only the minimization of the negative steps in separation, target fabrication, transportation, dismantling, and the R&D; work necessary for Pu disposition, but also the transformation required to create a new rational nuclear energy system without Pu(TRU).

This paper has briefly proposed one of the most promising approaches, THORIMS-NES, and included a scenario of realistic elimination of all Pu produced by U-fueled solid reactors in the past, present and near future. This might be an useful example to verify the feasibility of complete Pu elimination and potential huge nuclear energy production, and suggest criteria to judge the facility performance necessary for that purpose.

The primary systems, consisting of fission MSRs (FUJI and superFUJI - commercial power stations) and proton beam AMSB need to be developed. However, their R&D; will be not difficult nor costly, and their further improvements will be hopeful by few efforts, promising the world-wide application to solve the energy, environment and North-South problems opening a new nuclear era.

ACKNOWLEDGEMENTS: The authors wish to express their sincere thanks to Dr. Y. Nakahara, JAERI, and Dr. M. 0dera on their cooperation for this study, and also to Mr. F.Atchison, PSI, Drs. H. Takahashi and P. Grand. BNL, Dr. C. Marchetti. IIASA, Drs. I. Chuvillo and O.V. Kiselev, ITEP, and many Japanese and foreign friends on their encouragement and help in developing this work.



[1] Comm. Int.Security & Arms Control, Nat. Acad. Sci., "Management & Disposition of Excess Weapons", Washington: Nat. Acad. press, 1994, pp. 2-3.
[2] Rosenthal, M. W., Haubenreich, P. N., Briggs, R. B. :ORNL-4812 (1972) ; Engel, J. R., Grimes, W. R. , Bauman, H. F., McCoy, H. E., Dearing, J. F. , Rhoades, W. E. :ORNL/TM-7207 (1980).
[3] Furukawa, K., Lecocq, A., Kato, Y. & Mitachi, K. : J. Nucl.Sci. Tech., 27,1157 (1990).
[4] Furukawa,K., Minami,K., Oosawa,T., Ohta, M., Nakamura, N. Mitachi, K., Kato, Y. Emerg., Nucl. Energy System, p. 235,World Sci. (1987): Furukawa,K., Mitachi,K., Kato,Y. : NucI. Engineering & Design, 136, 157 (1992).
(5] Furukawa, K., Lecocq, A., Kato, Y. and Mitachi, K. : LA-12205-C, pp. 686-697 (1991); Furukawa, K. :Atomkernenergie/Kerntech., 44, 42-45 (1984).
[6] Gat, U., Engel, J. R., & Dodds, H. L., Nuci. Tech., 100, 390-394 (1992).
[7] Mitachi,K.,Furukawa,K., Murayama, M. and Suzuki, T. : Nucl. charac. of a small M. S. power reac. fueled with Pu", in Emerging Nuci. Energy Systems, World Sci. 1994, pp. 326-331.
(8] Furukawa, K., Chigrinov, S. E. , Kato, Y. , & Mitachi, K : "Accelerator Molten-Salt Breeding Power Reactor useful for Pu-burning and 233U-production", in Emerging Nuci. Energy Systems, World Sci., 1994, pp. 429-433.
[9] Chigrinov, S., Kievitskaya, A., Petlitski, V., Rutkovskaya, K. and Furukawa, K. : "Calcu. method of energy systems based on high energy particle and nuclei accelerators", in Emerging NucI. Energy Systems, World Sci. , 1994, pp. 434-438 : Kato, Y., Furukawa, K., Mitachi, K., and Chigrinov, S. E. : "Fuel trajectory in Accel. M. S. Breeding Power Reactor system including Pu burning", in Emerging Nuci. Energy Systems, World Sci., 1994, pp. 439-443.
[10] Furukawa, K., Tsukada, K. and Nakahara, Y. : Proc. 4th ICANS, 1980, pp. 349-354; J. Nuci. Sci. Tech., 18, 79 (1981) ; JAERl-M83-050 (1983)
[11] Furukawa,K., Kato,Y., 0homichi,T. & Ohno, H. : Thorium Fuel Reactors, "Proc. Japan- U.S. Semi. Th Fuel Reactors", Nara, 1982, Atomic Ene. Soc. of Japan, 1985, pp. 271-280.
[12] Furukawa, K. et al., First Int. Sympo. on Molten Salt Chemistry & Technology (April, 1983, Kyoto) Proceedings, 1983, J-303 pp. 405-408, J-304 pp. 409-413, K-210 pp. 497-499.
[13] Marchetti, C. & Nakicenovic, N. :RR-79-13, IIASA, (1987); Marchetti, C. :Nucl, Sci. Eng., 90, 521 (1985).


Go back to Bruce's Eclectic (main) Page