N

Xa

b

a

KTH2

B

1

mttbtctcar[o[a

ltsat

0d

Fusion Engineering and Design 85 (2010) 2227–2231

Contents lists available at ScienceDirect

Fusion Engineering and Design

journa l homepage: www.e lsev ier .com/ locate / fusengdes

eutronic calculations of a thorium-based fusion–fission hybrid reactor blanket

.B. Maa,∗, Y.X. Chena, Y. Wanga, P.Z. Zhanga, B. Caoa, D.G. Lua, H.P. Chengb

North China Electric Power University, School of Nuclear Science and Engineering, Beijing 102206, ChinaChina Nuclear Power Engineering Co., Ltd., Beijing 100037, China

r t i c l e i n f o

eywords:horiumybrid reactor

33Pa effectlanket

a b s t r a c t

Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-lifenuclear waste, and having inherent safety. Considering the fast development of nuclear power industryand the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuelproduction is proposed. The blanket using ThN or ThO2 dispersed in graphite or BeO is investigatedunder a first neutron wall loading of 0.57 MW/cm2 as ITER. The design helium cooled pebble bed designis employed according to the experience in the fusion reactor field. Preliminary neutronic calculations areperformed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo

transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulativefissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritiumbreeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficientin the tritium required for the (DT) fusion driver in those modes during the operation period. The blanketenergy multiplication factor M, varies between 13.66 and 15.85 depending on the fuel types at the end

add 233 233

of the operation period. In

. Introduction

World thorium reserves are estimated to be about three timesore than natural uranium reserves. Early work has investigated

he possibility of production of 233U in the fusion–fission reac-or [1–4]. Considering the abundance of thorium, a thorium-basedreeding blanket concept of fuel production is proposed. The neu-ron wall loading is set to be 0.57 MW/cm2 as ITER [5]. Heliumooled pebble bed design is used according to the experience inhe fusion reactor field. 233Pa is an important nuclide in the Th–Uonversion chain with long half-life (27.4 days) and large neutronbsorption cross section, which influences the Th–U conversionatio and the operation of Light Water Reactor with Th–U fuel6–9]. Therefore, 233Pa influences the Th–U conversion ratio and theperation of hybrid reactor, which is not considered in early work10–13]. In this paper, the effect of 233Pa on the 233U productionnd keff is discussed.

The neutronics calculations have been performed using an ITER-ike blanket geometric model as depicted in Fig. 1. Table 1 shows

he material composition in different zones of the blanket. The fis-ion zone containing spherical thorium dispersed fuels is insertedt the immediate neighborhood of the first zone in order to achievehe maximum production of 233U after 960 days. Fissionable tho-

∗ Corresponding author. Tel.: +86 10 51963823.E-mail address: maxb917@163.com (X.B. Ma).

920-3796/$ – see front matter. Crown Copyright © 2010 Published by Elsevier B.V. All rioi:10.1016/j.fusengdes.2010.08.044

ition, the effect of Pa on the U production and keff are also discussed.Crown Copyright © 2010 Published by Elsevier B.V. All rights reserved.

rium fuels, namely ThO2 or ThN having spherical geometry, weredispersed in a graphite or BeO matrix with a volume fraction of 50%as shown in Fig. 2 [13]. In this concept, a volume neutron sourcesimulates the fusion plasma chamber. The latter is surrounded bya first wall made of ferritic steel, type F82H [14].

2. Calculation methods

The one-dimensional transport and burnup code BISONC [15]is used for calculating the key parameter of the blanket, such askeff, the production of 233U, the cumulative fissile fuel enrichmentand power density of the blanket etc. The transport library wereobtained from JENDL-3 [16], which has 42 neutron energy groupstructure and P-5 Legendre scattering order. Since JENDL-3 hasno photon production files, in the current library only the neu-tron cross sections are present. However, the gamma heating isincluded in the kerma factors, under the assumption of local depo-sition of the photon energy. The burnup library contains the burnupchain of actinides followed by the cross sections for the reactions ofimportant actinides, in the number of 22, occurring in the burnup

chain. At this point, importance is given to not rapidly decayingactinides; those which are present in the transport library are con-sidered important. The tritium breeding ratio (TBR) is calculatedby the Monte Carlo transport code MCNP [17] with the point-wisecross section library FENDL/2 [18].

ghts reserved.

2228 X.B. Ma et al. / Fusion Engineering and Design 85 (2010) 2227–2231

Fig. 1. One-dimensional modeling of the hybrid blanket (LZ = lithium zone, FZ = fission zone, FW = first wall).

Table 1Material and volume fraction of fusion–fission reactor.

Material and volume fraction Zone Thickness (cm)

Center zone Void 1 375Plasma Void 12 400

Inner blanketBeryllium layer Be: 100% 11 0.2First wall F82H: 70%; He-gas: 30% 10 3Structure wall F82H: 70%; He-gas: 30% 2, 4, 6, 8 1, 1, 1, 1Lithium zone Li2O: 40%; Be: 20%; He-gas: 30% 3, 5, 7, 9 10.8, 6.0, 9.0, 12.0

Outer blanketBeryllium layer Be: 100% 13 0.2First wall F82H: 70%; He-gas: 30% 14 3

1.7%;

3

3

cdTegttd

Structure wall F82H: 70%; He-gas: 30%Fission zone ThO2 or ThN: 31.7%; C or BeO: 3SiC: 10.6%; He-gas: 26% 15Lithium zone Li2O: 40%; Be: 20%; He-gas: 40%

. Numerical results

.1. TBR

TBR calculation have been performed for the following fourases: ThO2 dispersed in graphite, ThO2 dispersed in BeO, ThNispersed in graphite, ThN dispersed in BeO. Table 2 showsBR for all the investigated cases at the beginning and at the

nd of the operation period of 960 days. In all the investi-ated modes, the blanket is self-sustaining with respect to theritium breeding ratio (TBR > 1.05). Therefore, the blanket inhese modes can provide enough tritium for their own fusionriver. The tritium production in lithium (mainly 6Li isotope)

Fig. 2. Dispersed spherical fuel.

16, 18, 20, 22 1, 1, 1, 1

8.017, 19, 21 12.8, 9.0, 8.0

increases almost linearly, caused by an increase of the neutronpopulation due to the accumulation of fissile fuel in the blan-ket.

Beryllium is an excellent neutron multiplier, which can multi-ply neutron with the Be (n, 2n)2� reaction. TBR is calculated as afunction of the beryllium fraction in the case of ThO2 dispersed ingraphite, as shown in Fig. 3, while the enrichment of 6Li is assumedto be 90%, and the total volume fraction of beryllium plus Li2Ois fixed as 60%. The neutron flux of 19th zone is shown in Fig. 4.As shown in Fig. 3, TBR increases with the increasing of berylliumfraction. It is caused by the increasing of beryllium fraction thatleads to the neutron flux increasing in the low energy as depictedin Fig. 4. TBR begins to decrease when the beryllium fraction isabout more than 50%, because the Li2O fraction is too small thatthe neutron flux increasing cannot compensate the atomic den-sity decreasing of 6Li and 7Li. The differences at lower energy areinduced by effects of beryllium multiplication and slowing downof neutrons. In addition, a trap at the neutron spectrum can be seenbecause of the neutron absorption peak of 6Li at the energy range0.1–1 MeV.

3.2. keff and energy multiplication M

The total energy generation in the blanket can be expressedwith the help of the energy multiplication factor M, which can be

Table 2TBR in investigated blanket.

Time (day) ThO2 + C ThO2 + Beo ThN + C ThN + BeO

0 1.050 ± 0.003 1.084 ± 0.003 1.059 ± 0.003 1.078 ± 0.003960 1.201 ± 0.003 1.215 ± 0.003 1.209 ± 0.003 1.224 ± 0.003

X.B. Ma et al. / Fusion Engineering and Design 85 (2010) 2227–2231 2229

605040302010

0.7

0.8

0.9

1.0

1.1

1.2

TB

R

Be fraction (%)

c

M

MotwiwTBtgawBrbi

10008006004002000

0.0

0.1

0.2

0.3

0.4

0.5

0.6

k eff

(ThO2+C)

(ThO2+BeO)

(ThN+C) (ThN+BeO)

ThN dispersed in BeO showed the best performance when takinginto account fissile fuel breeding because ThN has a higher tho-

Fig. 3. TBR versus beryllium volume fraction.

alculated as below [12]:

= Blanket energy release in MeV14.1

+ 1 (1)

Figs. 5 and 6 show keff results and energy multiplication factorrespectively. All cases meet the requirements of subcriticality

ver the operation period, and there seems to be a same trend onhe keff and energy multiplication factor M. M increased graduallyith increasing operation period for all the fuel types. This trend

s caused by the behavior of the effective multiplication factor keff,hich is fairly affected by the fuel composition and arrangement.

he energy multiplication values blanket using fuels dispersed ineO are a little more than that fuels dispersed in graphite becausehe fission rates in the blanket with fuels dispersed in BeO werereater than those dispersed in graphite. The M values were 13.70nd 15.85 for the blankets using ThO2 dispersed in BeO or graphite,hich were 13.66 and 15.79, respectively, as using ThN dispersed ineO or graphite at the end of the operation period. It is essential toise keff in order to improve the reactor performance of the system,ut the power density increases in this case. It is thus needed to

nvestigate the cooling capability in detail.

10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 101 102

1014

1015

1016

1017

1018

Neu

tron

flux

(cm

-2.s

-1.M

eV-1

)

Energy (MeV)

Be:50% Be:30% Be:10%

Fig. 4. Effect of beryllium fraction to the neutron flux of 19th zone.

Operation time (day)

Fig. 5. keff results for different fuel types.

3.3. 233Pa effect to the 233U production

The fertile isotope 232Th converts to a high quality fissile isotope233U by capturing a neutron as given below:

23290 Th + 1

0n → 23390 Th → 233

91 Pa + 0−1e → 233

92 U + 20−1e

It will be interesting to evaluate the fuel regeneration ability ofthe hybrid reactor in the form of figure-of-merit (FOM) which canbe defined as the ratio of net 233U production per year to the fissionpower output of the hybrid blanket. The mathematical formulationof FOM in net 233U (g MWth−1 a−1) can be described as follows:

FOM = net 233U production per yearPth

where Pth is the thermal power production in the hybrid blanket.FOM, the cumulative fissile fuel 233U production and 233Pa pro-

duction varying with the operation period are depicted in Figs. 7–9.As expected, fissile fuel production increased almost linearly withincreasing operation period. Among the investigated fuel types,

rium density than ThO2 and BeO enhances the neutron economy ofthe blanket. FOM increases rapidly with operation period increas-

10008006004002000

4

6

8

10

12

14

16 (ThO

2+C)

(ThO2+BeO)

(ThN+C) (ThN+BeO)

Ene

rgy

mut

iplic

atio

n (M

)

Operation time (day)

Fig. 6. Energy multiplication M for different fuel types.

2230 X.B. Ma et al. / Fusion Engineering and Design 85 (2010) 2227–2231

10008006004002000200

400

600

800

1000

1200

1400

1600 (ThO

2+C)

(ThO2

+BeO)

(ThN+C)

(ThN+BeO)

FO

M

Operation time (day)

Fig. 7. FOM results for different fuel types.

0

1000

2000

3000

4000

5000

6000

7000 (ThO

2+C)

(ThO2+BeO)

(ThN+C)

(ThN+BeO)

233U

pro

duct

ion

(Kg)

ifapt

0

2

4

6

8

10

(ThO2+C)

(ThO2+BeO)

(ThN+C)

(ThN+BeO)

Fue

l enr

ichm

ent (

%)

fuel enrichment and fissile isotope U production for all the fuel

10008006004002000

Operation time (day)

Fig. 8. 233U production for different fuel types.

ng before about 100 days, but FOM begins to decrease for all the

uel types. Because 233Pa has a long-half life (27 days), and notll production of 233Pa can be converted to 233U at the beginningeriod as depicted in Fig. 9. At the same time, a rapid accumula-ion of the fissile fuel component 233U causes a faster increase of

10008006004002000100

150

200

250

300

350

400

450 (ThO2+C)

(ThO2+BeO)

(ThN+C)

(ThN+BeO)

233

Pa

prod

uctio

n (K

g)

Operation time (day)

Fig. 9. 233Pa production for different fuel types.

10008006004002000

Operation time (day)

Fig. 10. Fuel enrichment versus operation period for different fuel types.

the fission power production in the hybrid blanket and burns thefissile fuel more efficiently in the blanket, in situ. It is generallyshown that the FOM values are higher in the ThN fuels dispersedin BeO or graphite than ThO2 fuels dispersed in BeO or graphiteafter about 100 days. The effect of 233Pa to keff and 233Pa produc-tion quickly reaches a basic balance with depletion at the beginningoperation.

3.4. Generated fuel enrichment and power density

Figs. 10 and 11 show the cumulative fissile fuel enrichment val-ues and maximum power density as a function of the operationperiod in the investigated modes. It is noted that the cumulativefuel enrichment increases almost linearly with increasing opera-tion period. The cumulative fuel enrichment values were 8.56% and7.78% for the blankets using ThO2 dispersed in BeO or graphite, andwhich were 7.57% and 6.88%, respectively, as using ThN dispersedin BeO or graphite at the end of the operation period. Power densityis also increasing with the operation period because of increasing

233

types. The maximum power density values blanket using fuels dis-persed in BeO are a little higher than that fuels dispersed in graphitebecause the fission rates in the blanket with fuels dispersed in BeOwere greater than those dispersed in graphite. The behaviors of

100080060040020000

50

100

150

200

250

300

350

(ThO2+C)

(ThO2+BeO)

(ThN+C)

(ThN+BeO)

Max

imu

m p

ower

de

nsity

(W

/cm

3 )

Operation time (day)

Fig. 11. Maximum power density for different fuel types.

ing an

tw

4

wbwdibtfp1oibaF

A

FeR

R

[

[

[

[

[

[

X.B. Ma et al. / Fusion Engineer

hose fuels in the LWR and power flattering for the hybrid blanketill be investigated in the future work.

. Conclusion

Aiming at early realization of a fusion–fission hybrid reactorith proven technology, an one-dimensional neutron analysis and

urnup calculations were performed. Four types of thorium fuelere loaded as nuclear fuel in the blanket of the reactor. The resultsemonstrate that the cumulative fissile fuel enrichment values var-

ed from 6.88% to 8.56% depending on the fuel types. The tritiumreeding ratio values are greater than 1.05 for all investigated fuelypes and the hybrid reactor is self-sufficient in the tritium requiredor the (DT) fusion driver in those modes during the operationeriod. The blanket energy multiplication factor M, varies between3.66 and 15.85 depending on the fuel types at the end of theperation period. The effect of 233Pa on the 233U production is alsonvestigated. It is noted that 233Pa is important to the FOM at theeginning operation period, but when the 233Pa production reachesbasic balance with depletion, it becomes less important to the

OM.

cknowledgements

The work was supported by the National Natural Scienceoundation of China (Nos. 10705011, 10875042), the Beijing Sci-nce New Star Plan Project (No. 2007B058) and the Fundamentalesearch Funds for the Central Universities (09MG11).

eferences

[1] Y.X. Chen, Y.C. Wu, Conceptual study on high performance dual-cooled blanketin a spherical tokamak fusion-driven transmuter, Chinese Journal of NuclearScience and Engineering 19 (3) (2004) 215–220 (in Chinese).

[[

[

d Design 85 (2010) 2227–2231 2231

[2] Y.C. Wu, S.L. Zheng, W.H. Wang, D.L. Chu, Q.Y. Huang, Y. Ke, et al., Concep-tual study on the fusion-driven sub-critical system, Chinese Journal of NuclearScience and Engineering 24 (1) (2004) 72–80 (in Chinese).

[3] M.G. Deng, K.M. Feng, Comparison study of transmutation neutronics calcula-tion for a spherical tokamak, Nuclear Fusion and Plasma Physics 23 (1) (2003)7–12 (in Chinese).

[4] B.Q. Deng, J.C. Yan, K.M. Feng, Reactor core design of spherical tokamak trans-mutation of nuclear wastes, Chinese Journal of Nuclear Science and Engineering22 (1) (2002) 79–84 (in Chinese).

[5] ITER-FEAT—Outline Design Report by the ITER Director, Tokyo, ITER Meeting,1, 2000, p. 6.

[6] IA EA, Thorium Fuel Cycle Potential Benefits and Challenges, IAEA 2TEC-DOC21450, IAEA, Vienna, 2005.

[7] Z.X. Wu, X.Q. Jing, Fuel management of HTR210, Journal of Tsinghua University(Science & Technology) 41 (4/5) (2001) 120–123 (in Chinese).

[8] Y.W. Yang, X.Q. Jing, Y.X. Gu, et al., Physics of Pu burning in modular hightemperature gas cooled reactor, Journal of Tsinghua University (Science & Tech-nology) 42 (5) (2002) 652–654 (in Chinese).

[9] G.L. Yu, K. Wang, Characteristics of Pa233 in the Th–U fuel fast spectra sys-tem, Journal of Tsinghua University (Science & Technology) 48 (3.) (2008) (inChinese).

10] H. Yaplcl, M. Bayrak, Neutronic analysis of denaturing plutonium in a thoriumfusion breeder and power flattening, Energy Conversion and Management 46(2005) 1209–1228.

11] S. Sahin, et al., Power flattering and minor actinide burning in a tho-rium fusion breeder, Energy Conversion and Management 43 (2002) 799–815.

12] H. Yaplcl, et al., Investigation of neutronic potential of a moderated (D-T) fusiondriven hybrid reactor fueled with thorium to breed fissile fuel for LWRS, Ener-gyConversion and Management 41 (2000) 435–447.

13] M. Ubeyli, Neutronic analysis of ARIES-RS fusion reactor fueled with thorium,Energy Conversion and Management 47 (2006) 322–330.

14] H.M. Sahin, Monte Carlo calculation of radiation damage in first wall ofan experiment hybrid reactor, Annals of Nuclear Energy 34 (2007) 861–870.

15] Jerzy Cetnar, Piotr Gronek, BISON-C Upgraded One-dimensional Transport andBurnup Calculation Code for Unix System, Krakow, 1997.

16] K. Shibata, et al., IENDL-3 JAERI (1990) 1319.17] J.F. Briesmeister (Ed.), MCNP/4C, A General Monte Carlo N-Particle Transport

Code, 1997.18] H. Wienke, M. Herman, FENDL/MG-2.0 and FENDL/MC-2.0, The Processed

Cross-section Libraries for Neutron–Photon Transport Calculations, Version 1,March 1997, IAEA Vienna, Report IAEA-NDS-176, Rev. 1, October 1998.