Annals of Nuclear Energy 73 (2014) 500–505

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Annals of Nuclear Energy

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Neutronic study of an innovative natural uranium–thorium basedfusion–fission hybrid energy system

http://dx.doi.org/10.1016/j.anucene.2014.07.0320306-4549/� 2014 Elsevier Ltd. All rights reserved.

⇑ Corresponding author.E-mail address: zhouzhw@mail.tsinghua.edu.cn (Z. Zhou).

S.C. Xiao a, J. Zhao a, Z. Zhou a,⇑, Y. Yang b

a Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, Chinab Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou, China

a r t i c l e i n f o a b s t r a c t

Article history:Received 29 May 2014Received in revised form 19 July 2014Accepted 19 July 2014Available online 10 August 2014

Keywords:FFHRU-modulesTh-modules233U breedingOptimization

An innovative design for a water cooled fusion–fission hybrid reactor (FFHR), aiming at efficiently utiliz-ing natural uranium and thorium resources, is presented. The major objective is to study the feasibility ofthis concept balanced with multi-purposes, including energy gain, tritium breeding and 233U breeding. Inorder to improve overall neutron economy of the system, the fission blanket is designed with two types ofmodules, i.e. the natural uranium modules (U-modules) and thorium modules (Th-modules), which arealternately arranged in the toroidal and poloidal directions of the blanket. This innovative design is basedon a simple intuition of neutron distribution: with the alternate geometrical arrangement, energy mul-tiplication by uranium fission, tritium breeding and 233U breeding are performed separately in differentsub-zones in the blanket. The uranium modules which has excellent neutron economy under the com-bined neutron spectrum, plays the dominant role in the energy production, neutron multiplication andtritium breeding. Excess neutrons produced by the uranium modules are then used to drive the thoriummodules (which have poor neutron economy) to breed 233U fuel. Therefore, it creates a new free dimen-sion to realize the blanket’s balanced design. The COUPLE code developed by INET of Tsinghua Universityis used to simulate the neutronic behavior in the blanket. The simulated results show that with the vol-umetric ratio of thorium modules about 0.4, the balanced design for multi purposes is achievable, withenergy multiplication M P 9, tritium breeding ratio TBR P 1.05, and at the end of the five years refuelingcycle, the 233U enrichment in thorium modules exceeding 1.0%. The neutronic analysis results also showthat the preliminary design of this innovative FFHR is of great potential to utilize the bred 233U effectivelyafter the initial fuel load of the first ten-year operation.

� 2014 Elsevier Ltd. All rights reserved.

1. Introduction blanket is filled with natural uranium, spent nuclear fuel or natural

Thorium is 3–4 times more abundant than uranium and iswidely distributed in nature as an easily exploitable resource inmany countries. During the pioneering years of nuclear energy,from the mid 1950s to mid 1970s, there was considerable interestworldwide to develop thorium fuels and fuel cycles in nuclearresearch and power reactors for conversion thorium to ‘fissile fuel’233U in order to supplement uranium reserves (IAEA, 2005). FFHRwas proposed to breed fissile fuel (Lidsky, 1975; Bethe, 1979; Leeand Moir, 1981; Kotschenreuther et al., 2012). A FFHR uses thefusion reaction T (D, n) 4He in the plasma confined in its tokamakto generate high-energy neutrons with 14.1 MeV. These fusionneutrons are employed to drive fission and tritium conversionreactions in the subcritical blanket surrounding the plasma. The

thorium for generating fission energy and multiplying neutrons,and lithium based breeders for producing tritium used to fuelfusion in tokamak.

However, for the FFHR fueled with only natural thorium, theenergy gain M is relatively small (less than 2.0) in the initial oper-ating stage, which would be challenging on the fusion capabilityrequirement for energy generation (Piera et al., 2010; Lafuenteand Piera, 2011). ITER is an experimental fusion reactor whichhas been investigated extensively for years. It has yet to achievethe performance required for power reactor operation. Early appli-cation of fusion energy may be realized under much lower plasmacondition than in a fusion-only power reactor. The blanket shouldhave higher energy multiplication factor M to accomplish this goal.The early design of a water cooled natural uranium–thorium fueledblanket concept (Xiao et al., 2012, 2013) could not reach the bal-anced operation with multi-purposes, i.e., high energy gain, tritiumself-sufficiency and good 233U breeding rate due to thermal neu-tron barrier effect of 232Th fuel arranged in the fission blanket.

S.C. Xiao et al. / Annals of Nuclear Energy 73 (2014) 500–505 501

A new type of hybrid blanket aiming at efficiently utilizing tho-rium resource while maintaining higher M and tritium self-suffi-ciency, proposed by this paper, adopts the seed–blanket designconcept of Wang (2003) for a PWR with thorium–uranium fuelcycle. This innovative design uses U-modules as the seed zones,Th-modules and tritium breeding modules as the fuel breedingblanket zones. U-modules and Th-modules are separated geomet-rically to reach an excellent neutron economy. The natural ura-nium fuel (U-modules) functions as energy generation andneutron multiplication source (Seed). Excess neutrons releasedby the U-modules are then used to convert natural thorium to233U fuel in Th-modules and to breed tritium from lithium-basedmaterial in tritium-modules. The blanket employs water as coolantand operates under the combined thermal and fast neutron spec-trum. Different fuel to water volumetric ratio and thorium fuelfraction are investigated to find optimized blanket designparameters.

The preliminary results indicate that it is rather promising todesign a high-performance water cooled fission blanket of FFHRfor electric power generation and 233U breeding by directly loadingnatural uranium and thorium if fusion neutron source from anITER-scale tokamak is achievable.

2. Simulation model and method

Focusing only on the neutronic performance of the blanket withsome details omitted, a simplified one dimensional ‘D-Shape’model of FFHR has been chosen for simulating the plasma-blanketzone of a typical tokomak. The water cooled natural uranium/tho-rium fueled FFHR blanket is depicted also in Fig. 1.

The major radius of the plasma R is 510 cm, and the minorradius a is 154.5 cm and b equals 286 cm. The elongation b/a is1.85 and the aspect ratio R/a is 3.30 (Zhou et al., 2011). In the frontfuel region, the thickness of each natural uranium or thorium fuelplate is 2 cm, and the H2O coolant/moderator thickness is 1 cm forthe U-modules and 0.5 cm for Th-modules. While in the tritiumbreeding region, the Li4SiO4 plate with 90% 6Li enrichment is4 cm each and the H2O coolant/moderator thickness is 4 cm. Thethickness of shielding reflector is set to be 45 cm. Five uranium fuelplates are inserted into the U-modules. And Th-modules areequipped with four thorium fuel plates. The arrangement ofU- and Th-modules in toroidal directions is shown in Fig. 2. In

Fig. 1. Computational ‘D-Shape’ model of water cooled natural uranium/thoriumfueled FFHR.

the preliminary design, only the alternate arrangement of U- andTh-modules in toroidal direction is considered.

A uniform volumetric circular cylindrical fusion neutron sourcein toroidal direction is placed at the center of plasma region, theradius range of which is from 415 cm to 635 cm, and the heightis 560 cm. The fusion neutron source is based on physics similarto or less demanding than that used for the ITER design, so theexisting R&D program supporting ITER will cover the requirementsin tokamak physics design of this innovative FFHR. In all calcula-tions, the total fission power of the blanket is set to 3000 MW,which could be realized by adjusting the fusion power during thefission fuel depletion process. The refueling cycle is 3650 days.The code system Couple2.0 (Zhou et al., 2011) developed by INET,Tsinghua University, which couples the codes MCNPX and Origin2.0, is used to simulate neutron transport in fission blanket and cal-culate the fissile material depletion and conversion.

3. Results and discussion

3.1. The essence of blanket neutronic design

The essential goal of optimal neutronic design of subcriticalFFHR blanket is to achieve balance of its multi-operating purposes,i.e., high energy multiplication M, tritium self-sufficiency and good233U breeding rate, under the constraint of utilizing only naturaluranium and thorium as initial fuel. Neutron economy is the direc-tion towards optimal design. The blanket’s neutron multiplicationeffect should be strong enough to ensure that available excess neu-trons could be used to induce fission reaction and breed tritiumand 233U. Furthermore, these neutrons should be distributed opti-mally among these neutron competing reactions.

3.2. The rationale of seed–blanket concept in the blanket design

It is well known that 232Th has poor neutronic behavior for fis-sion. For the thorium fueled blanket, the general method to obtaingood system neutron balance is the adoption of the seed–blanketconcept, in which natural uranium modules (Seed) act as theenergy generation and neutron multiplication components, whilenatural thorium modules and tritium breeding modules (Blanket)as the blankets to breed fissile and fusion fuels. The basic logic ofthis concept is to use the excess neutrons generated in the naturaluranium fuel region to breed fissile fuel 233U in the thorium fuelregion, while maintaining high energy multiplication factor (M)and tritium self-sufficiency. The subtle idea of the seed–blanket

Fig. 2. Alternately arrangements of natural uranium and thorium modules in thetoroidal directions of the blanket (equatorial plane).

502 S.C. Xiao et al. / Annals of Nuclear Energy 73 (2014) 500–505

concept is to decouple the neutron generation modules from ‘pure’neutron absorbing modules. The performance of neutron genera-tion modules is mainly kept by itself. Therefore, the U-modules,Th-modules and tritium breeding modules are functionally decou-pled to obtain an excellent neutron economy.

3.3. Key parameters in the blanket optimal design

For the seed–blanket concept, two aspects to improve the sys-tem overall performance related to multiple parameters, i.e. M,TBR and 233U breeding capability should be taken into account,namely: (1) optimization of U-modules’ performance; (2) decou-pling the U-modules and Th-modules efficiently.

3.3.1. Optimization of U-modules performanceThe light water is used to moderate both the fusion and fission

neutrons. For a water cooled uranium system operating under thecombined thermal and fast neutron spectrum, the volumetric ratioof natural uranium fission fuel to coolant water V (U)/V (H2O),denoted as VR, is the key parameter influencing the blanket’s Mand TBR performance of the blanket with a given natural uraniuminventory. We could change VR in the front fuel region to adjust thesystems neutron spectrum. Fig. 3 gives the system parameters Mand TBR vs. VR of the water-cooled blanket. One could find thatTBR increases with VR increasing. The fraction of neutron that leaksinto the lithium zone increases with the size of front fuel zonedecreasing. The reason for M curve to form a peak is that withthe hardening of neutron spectrum (by reducing the coolant thick-ness), the neutron fraction of the resonance capture for breedingfissile fuel has exceeded the neutron fraction scattered to the cool-ant. In addition, more neutrons leak into the lithium region. There-fore, the neutrons available for fission reactions decrease.

The guiding principle in designing the U-modules is to obtain anoptimized U-modules’ M and TBR performance. Higher VR implyingstronger neutron penetrating capabilities and larger neutrons leak-age rates into tritium breeding region yields higher tritium breed-ing rate. VR value between 1.0 and 2.0 is an optimized operatingregion for the U-modules to reach balanced system performance.

3.3.2. Decoupling the U-modules and Th-modules efficientlyFor the combined thermal and fast neutron spectrum, thermal

neutrons play the central role in inducing fission reaction, 233Uconverting and tritium breeding reactions. However, thorium fuelplates will act as a thermal neutron barrier both for tritium breed-ing modules and U-modules if they were installed in the same fuel

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Fig. 3. TBR and M change with VR.

region. The central point of decoupling these two main kinds ofmodules efficiently is to place the pure neutron absorbers, i.e.Th-modules away from the neutron generators, i.e. U-modules.Energy multiplication by uranium fission, tritium breeding and233U converting are performed separately in different sub-zonesin the blanket. This could be realized through geometrical alter-nately arrangement of these two modules in the blanket’s toroidaland poloidal directions and choosing an appropriate volume ratioof thorium modules to U-modules. One should be noted that Fig2 is a simplified model, the geometric sized of U-Modules andTh-modules in the toroidal direction are much larger than the dif-fusion length of thermal neutrons. The space self-shielding effect ofthermal neutrons in the toroidal direction may be rather strong todecrease the 233U breeding rate with a given volume ratio of tho-rium to uranium modules. Further studies should be carried outto find out an optimized thickness of both U-modules and Th-modules.

3.4. Simulated results: the effect of Th-modules’ toroidal volumefraction Tvf

In the preliminary design of the thorium-uranium fueled blan-ket, three VR values, i.e. 1.0, 1.5 and 2.0, of U-modules are selectedto study its influence on the system overall performance, while V(Th)/V(H2O) of Th-modules is set as 4.0. The M, TBR and 233Ubreeding rate vs. Tvf are given in Fig. 4.

From Fig. 4, it can be found with smaller VR, the system canachieve higher thorium fuel converting rate with the constraintcondition of tritium self sufficiency and M requirement. ParameterVR has small influence on the system’s 233U breeding rate, but playsan important role in determining the U-modules energy multipli-cation M and TBR performance. In general, the M and TBR require-ments are more restricted conditions for the blanket design. Thevolume fraction of Th-modules in the toroidal direction and VR

are determined by three factors. First, in order to meet the require-ments to the blanket with 3000 MW fission power and 500 MWITER fusion power upper limit, M should be larger than 6. Second,the TBR requirement of self-sufficiency should be guaranteed.Third, with the current blanket arrangement, the 233U breedingrate should be larger enough.

It has been found that M and TBR are inversely proportional toTvf, but M is more sensitive than TBR to Tvf and the 233U breedingrate is proportional to Tvf. Fig. 5 displays the relations of TBR andenergy fraction contributed by the Th-modules vs. Tvf, respectively.It shows that TBR fraction contributed by the Th-modules is almost

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Fig. 4. Blanket’s operating performance vs. VR and Tvf.

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Fig. 6. M vs. burn up with different Th-modules volume fraction.

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Fig. 7. TBR vs. burn up with different Th-modules volume fraction.

60

S.C. Xiao et al. / Annals of Nuclear Energy 73 (2014) 500–505 503

linearly proportional to Tvf, while the energy fraction increasesslowly with small Tvf and sharply with Tvf exceeding 0.9.

Fig. 4 also shows that with TBR reaches about 0.94 if all space inthe blanket is filled with thorium modules. The results indicatesthat the fusion neutron plays the most important role in tritiumbreeding due to its stronger penetrating power and intensity, whiletritium breeding due to the neutrons generated by the U-modulesplays the secondary important role. The energy gain factor M canbe ensured by proper design of inventory of U-modules. Watercooled U-modules is an excellent energy multiplier under the com-bined thermal and fast neutron spectrum.

Figs. 6 and 7 show the histories of M and TBR vs. fuel burn upunder 0.4, 0.5 and 0.6 volume fractions of Th-modules with VR as2.0. Furthermore, M and TBR are slightly increasing with burn upand both requirements to M and TBR could be met with Tvf 6 0.6.

The fractions of M and TBR contributed by the Th-modules aregiven in Fig. 8 and Fig. 9, respectively. The results show that bothfractions of M and TBR increase with burn up increasing. At theend of operating stage, the thorium-modules could generate about40% percent energy with 0.6 thorium fuel volume fraction. FromFig. 6, at the end of 10 years operating, the energy gain factor Mapproaches about 11. Fusion energy will then be 272.72 MW.Energy released by the Th-modules divided by fusion power givesa value about 4.4. It means that the blanket could be energy selfsustained using only thorium modules with additional operatingtime, the U-modules could be replaced by the natural Th-modulesat the end of operating stage.

The 233U breeding rate under different thorium fuel volumefractions is shown in Fig. 10. Higher thorium fuel volume fractiongives higher enrichment and larger total mass of 233U. At the endof 10 years operating stage, the blanket could generate about 4.5,3.4 and 2.5 tons 233U for the 0.6, 0.5 and 0.4 thorium volume frac-tions respectively. And the 232U inventories are 0.62, 0.43 and0.29 tons respectively. The initial 232Th inventory for 0.4, 0.5 and0.6 thorium volume fractions are 154.96, 194.18, 233.01 tons.The bred 233U fuels in the irradiated thorium fuel modules are thenunloaded and inserted into a new FFHR, functioning as energy mul-tiplier and neutron source to drive other natural thorium modules.The blanket of FFHR can then gradually complete the transitionfrom initial mode with the combination of U-modules and Th-modules to final mode with all Th-modules. One critical point isto choose an efficient thorium fuel to water ratio and to find a bal-anced system with optimal parameters of M and TBR, similar to thedesign of uranium–water system.

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Fig. 5. TBR and Energy fraction contributed by the Th-modules.

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Fig. 8. Evolution history of TBR fraction contributed by the Th-modules.

The isotope burn-up history of U-modules is given in Fig. 11, itcould be found that total fissile fuel inventory of 235U plus 239Pu inthe U-modules is increasing with burn up due to the 238U to 239Pu

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Fig. 9. Evolution history of energy fraction contributed by the Th-modules.

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Fig. 10. 233U enrichment vs. burn up with different Th-modules volume fractions.

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Fig. 11. Fissile isotope inventory evolution history of U-modules.

504 S.C. Xiao et al. / Annals of Nuclear Energy 73 (2014) 500–505

neutron capture conversion reaction. At the end of 10 years opera-tion, the U-modules could generate about 4.75 tons 239Pu with 0.4volume fraction of thorium modules. Besides generating energy,the uranium modules is also an excellent fissile breeding modules.The bred Th-modules could be removed and new natural thoriumfuel modules are inserted. The irradiated U-modules will still func-tion as the energy multiplier and neutron source to drive the Th-modules.

There was general consensus that a hybrid capable of producinga certain amount of electric power would be noticeably moreexpensive than an LWR producing the same amount of power(DOE report, 2009). Deonigi and Engel (1976) estimated that thecapital cost of hybrid reactor is 1.8–2.7 times the cost of LWRs.However, the large uncertainty is difficult to evaluate dependingon construction experience and safety regulatory control of eachcountry. For example, in the 13th International Nuclear IndustryExhibition recently held in China, Westinghouse’s vice presidentTimothy Collier said that the cost of AP1000 (the state-of the artthird generation PWR developed by Westinghouse) could decreaseto about $3600/kWe after commercial batch construction. And thecost of AP1000 currently under construction in Vogtle of USreaches about $6360/kWe (News: AP1000 total unit cost: $3600 /KWe, 2014). Professor Qiu Lijian from AIPP of Chinese Academy

of Science, in Chapter 15 of his book Fusion Energy and Its Applica-tions (Qiu, 2008), evaluated the economy of AREIS series FusionReactor as a reference with the cost between $3670 and 4400/kWe.

These cost targets appear to leave sufficient room for the fusion-fission system to realistically operate when considering the risingprice of natural uranium fuel. A conclusion of the hybrid reactor’seconomic potential was made in reference (Richardson and Cohen,1987): Fusion Hybrid reactors could become economically viable,especially as a source of fissile fuel for light-water reactors, if theprice of uranium oxide becomes high enough; however this pricecan be estimated only roughly at present and may lie between$100 and $330 per pound.

4. Conclusions and remarks

The major results of this study are summarized as follows:

(1) An innovative water cooled natural uranium–thorium fueledblanket using the seed–blanket concept with multi-pur-poses, i.e. high M, tritium self sufficiency and good 233Ubreeding rate, is realized. The energy gain factor M is majorlycontributed by U-modules. Excess neutrons generated bythe U-modules are then used to drive the Th-modules andtritium breeding materials to breed 233U and tritium.

(2) The uranium to water volumetric ratio and the volume frac-tion of thorium-modules in the toroidal direction are thetwo kernel factors determining the overall performance ofFFHR blanket. VR value between 1.0 and 2.0 is an optimizedoperating region for the U-modules to reach balanced sys-tem performance. With smaller V (U)/V (H2O), the systemcould operate with higher volume fraction of thorium mod-ules in the blanket subject to the constraints of tritium selfsufficiency and reasonable value of M.

(3) The energy fraction contributed by Th-modules increaseswith burn up increasing. At the end of 10 years refuelingcycle, it can reach about 40% with Tvf = 60% and VR = 2.0.And the 233U enrichment and 233U fuel inventory can reachabout 1.9% and 4.5 tons respectively. The irradiated thoriumfuel modules could then be unloaded and inserted into anew FFHR, functioning as energy multiplier and neutronsource to drive other natural thorium modules. The blanketof FFHR can then gradually complete the transition from ini-tial mode with the combination of U-modules and Th-mod-ules to final mode with all Th-modules.

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Acknowledgment

This work is supported by the Chinese magnetic confinednuclear fusion energy research with grant number 2010GB111006.

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