- Email: [email protected]
Neutronics study and analysis of a multifunctional blanket
Fusion Engineering and Design 41 (1998) 589 – 595
Neutronics study and analysis of a multifunctional blanket Bingjia Xiao *, Lijian Qiu Institute of Plasma Physics, Academia Sinica, P.O. Box 1126, 230031 Hefei Anhui, People’s Republic of China
Abstract The analysis codes of neutronics, thermal-hydraulics and radioactivity and after-heat calculation are developed and briefly introduced. A multi-function blanket is studied. The analysis shows that the blanket can transmute almost all the HLW loaded, breed large amounts of fissile material and, meanwhile, output a large amount of thermal power during a 30-year operation time. The blanket with its driver is verified as a radioactivity clean nuclear power system. © 1998 Elsevier Science S.A. All rights reserved.
1. Introduction The fusion power for commercial use needs a long time to develop. It is important to find near term applications of controlled fusion in which the parameter requirements are far less stringent, to promote the progress of the fusion power. Fusion breeder and high level waste (HLW) transmutor are candidate selections for fusion’s earlier applications. In 1990, Rubbia [1] proposed a concept that an accelerator driven subcritical system can be a radioactivity clean nuclear power system (RCNPS). This concept has attracted the attention of many scientists. By this concept, the main problems accompanying fission reactors, such as the shortage of fissile material, the high level waste and the risk of nuclear diffusion, can be effectively solved. Can the hybrid reactor be designed as a RCNPS? Many works [2 – 4] show the possibilities of hy-
* Corresponding author. E-mail: [email protected]
brid reactors to breed fissile material and to transmute HLW. Principally, the RCNPS concept can be realised in the hybrid reactor. Hybrid reactors designed as a breeder and HLW transmutor can serve as a new kind of RCNPS. It can put forward the fusion application and fusion development, and put forward the fission development as well. This paper gives the design and analysis of a blanket that can breed tritium for self-sustaining, breed fissile material and transmute both fission products and actinides. Finally, the RCNPS concept as a hybrid reactor is verified.
2. Analysis codes and their development
2.1. Re6isions to BISON 1.5 BISON 1.5 is a code for 1-D neutron transport and burn-up calculation [5]. A study shows that in a hybrid reactor, neutronics simulation results have little difference by BISON 1.5 (1-D)
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and by MCNP (3-D) [6]. BISON 1.5 code can be used in the blanket neutronics analysis. It has some limitation to be applied into the transmutation calculation such as: (1) the neutron data library does not contain the nuclides needed; (2) it does not consider the resonance self-shielding effect; and (3) the burn-up calculation has difficulty being applied in the complicated process for transmutation blanket may meet. So we made revisions to BISON 1.5. The new data library, a 46-group neutron plus 21-group photon group parameter library, which contains all the nuclides and all main cross sections that may be met in the transmutation issue, was created from ENDF-B/ VI by the codes NJOY and TRANSIX. For Burnup computations, we propose the Burlirch–Stoer numerical method [7] to replace the BATEMAN method [5] so that the complicated burn-up process can be solved and the burn-up data library does not need any modifications. A calculation example shows the correctness of this revisions [10]. The original burn-up chain is extended to consider all the main process as may occur in the HLW transmutation and fuel breeding issue, and described completely in a new burn-up library. We also revised BISON 1.5 to consider the resonance self-shielding effects by the BONDARENKO method [8]. Other minor revisions, such as the user interfaces and interface to the new library and thermal-hydraulics simulation, were also made.
2.2. Thermal-hydraulics and neutronics coupled calculation The thermal-hydraulics is an important aspect of the blanket. We developed a code, THPBHR, to analyze the 2-D thermal-hydraulics for the blanket with either pebble bed or plate-like fuel elements with various kinds of cooling scenarios [9,10]. In the case of a soft neutron spectrum and critical factor near to the critical point (keff = 1), the neutronics and thermal-hydraulics coupled effect is important. So we combined THPBHR with revised BISON 1.5 to form a thermal-hydraulics and neutronics coupled analysis code BITH [10].
Fig. 1. The neutron transport, burn-up, BHP and thermal-hydraulics calculation schemes.
2.3. Radioacti6ity, BHP and after-heat calculation The radioactivity, BHP and after-heat calculation can be performed by FDKR [11]. It was revised to be coupled with a revised BISON 1.5. After the burn-up calculation and neutron transport calculation are fulfilled at a certain time step, the results can be supplied into the FDKR to perform the radioactivity calculation. All the calculation processes can be seen in Fig. 1.
2.4. The neutron wall loading requirements for the transmutation of HLW It is proposed that Pu can serve as the neutron multiplier to reduce the requirements of the neutron wall loading to 1 MW m − 2 for the purpose of transmutation. If the requirements can be further lowered, fusion application for HLW transmutation can be more attractive. A blanket is designed to accomplish dual functions, namely, transmuting the high level nuclear waste (Np, Am, Cm and Pu isotopes produced in
Fig. 2. Blanket configuration and material composition for minor actinides transmutation.
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
591
Table 1 The HLW transmuted after 500 days with nuclear wall loading 1 and 0.5 MW m−2 241
243
244
2910c 3331d 13c 14d 11c 13.3d
360 404 9 9 7.8 9.12
−18a −31.3a −2a −4 −2.2a −3.9a
Am
Nuclides Transmuted (1.e24) Number of LWR? b (Normalized to 1 GW year) Transmutted fraction (%)
Am
Cm
237
238
239
2250 2594 11 12 10 11.7
−1107a −1343a −18a −20 −358a −444a
1470 1532 0.6 0.6 12.9 15.3
Np
Pu
Pu
240
241
242
270 321 0.3 0.3 5.1 6.22
273 301 0.8 0.8 16 17.5
−99a −129a −0.5a −0.6 −9.3a −12a
Pu
Pu
Pu
a
Negative values express the nuclide density increased. Normalizing method: N, nuclides transmuted in 1 GW year thermal power hybrid reactor blanket/nuclides produced in 1 GW year thermal power LWRs. c For 0.5 MW m−2 neutron wall loading. d For 1 MW m−2 neutron wall loading. b
Table 2 Transmutation efficiency (total fission reaction rate/(n, gamma) reaction rate) 241
243
Am
0 500 days a b
1.0a 1.0
244
Am
1.1b 1.1
1.4 1.4
237
Cm
1.4 1.5
3.3 3.4
238
Np
3.4 3.5
1.6 1.6
239
Pu
1.7 1.7
5.3 5.3
5.4 5.5
240
Pu
14. 14
241
Pu
14 15
5.0 5.1
242
Pu
5.2 5.3
12 11
Pu
11 12
4.6 4.7
4.7 4.8
For 0.5 MW m−2 neutron wall loading. For 1 MW m−2 neutron wall loading.
fission reactor) and breeding the tritium for the use in fusion reactor core. The 1-D layout and the material composition of the blanket are shown Fig. 2. The coolant is assumed as helium. The rest materials of all zones in Fig. 2, Table 4 are coolants (helium). Tables 1 – 3 give the results of the nuclides transmuted after 500 days burning. We can see that by raising the fuel loading density and hence, raising the keff and neutron flux, we can reach the transmutation object with lower neutron wall loading (0.5 MW m − 2) which may be achieved in the near future. One of the main figures of merit of transmutation is the transmutation efficiency, that is, the ration of the total fission reaction to the total absorption reaction rate. It is shown that the efficiency of this kind of blanket is so high that even if the neutron wall loading is lowered to 0.5 MW m − 2, the transmutation efficiency and the tritium breeding rate are high enough. The thermal power in the blanket is 4.9 (initial)–3.4 GW (after 500 days) and 9.7 – 3.4 GW with neutron wall loading 1 and 0.5 MW m − 2, respec-
tively. It is also shown that keff decreases with time that is beneficial to the safety of the blanket, and the tritium breeding rate (T) is so high that it can provide tritium for several fusion reactors of same fusion power. The heat generated in the blanket can be release off with the existing technology used in fast fission reactors.
3. Multifunction transmutation blanket It is found that the tritium-breeding ratio in the scenario introduced last section for transmuting minority actinides is very large. If this ratio is reduced and the extra neutrons are lowered down to a certain extent after they pass across the tritium-breeding zone, we can use them to breed fissile material and transmute the fission products. Table 4 gives the blanket 1-D configuration and material compositions using Pu as neutron multiplier. In this scenario, we use natural uranium to breed 239Pu, both for self-sustaining of Pu and for providing fissile material for fission reactors.
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Table 3 keff, tritium breeding ratio and power density of the blanket Time (days)
keff T P (W cm−3) a b
0
100
200
300
400
500
0.89a/0.97b 8.3/29 180/348
0.88/0.95 7.6/18 162/211
0.87/0.94 7.1/14 148/170
0.86/0.93 6.6/13 138/149
0.85/0.92 6.3/12 130/135
0.85/0.92 6.1/11 123/124
For 0.5 MW m−2 neutron wall loading. For 1 MW m−2 neutron wall loading.
Table 4 The 1-D configuration and material composition of blanket with Pu as neutron multiplier Zone no.
Zone function
Width (cm)
Material
1 2 3 4 5 6 7 8 9 10
Fusion core First wall Minor actinide transmutation Wall Tritium breeding Wall Fissile material breeding Wall Fission product transmutation Reflector
100 0.5 40 0.5 5 0.5 10 0.5 20 40
Vacuum ss-316 (70%) ss-316 (5%), Np Am Cm (4.9%), Pu (1.6%), Zr (1%) ss-316 (80%) ss-316 (5%), natural Li2O (5%) ss-316 (80%) ss-316 (5%), natural U (55%) ss-316 (80%) ss-316 (5%), 135Cs (3%), 129I (3%), 99Tc (4%) C (75%), Be (15%)
The other fuel cycle candidate is Th–U. We also designed a blanket for this fuel cycle, that is using 233U as neutron multiplier and 232Th to
breed 233U. The blanket configuration for Th–U fuel cycle is almost the same as the one mentioned in Table 4 except that natural U (55%) is replaced by 232Th (55%) in zone 7 and Pu (1.6%) is replaced by 233U (1.4%). The reference fusion core parameters are from Ref. [12]. The calculation results are shown in Tables 3 and 4 and Fig. 3 From the calculation results, we can conclude that: 1. The peak power density of the blanket is about 190 W cm − 3 at initial and decreased with time (shown in Fig. 3). The power density Table 5 The HLW transmuted (kg) with multiplier after 500 day operation 241
243
1441 1296
181 165
Am
Fig. 3. Energy amplifying factor, peak power density and critical factor as the function of operation time with U – Pu cycle.
Pu U
233
Am
244
Cm
−51 −39
233
U or
237
235
1115 982
20 22
Np
239
Cs
Pu as neutron
129
99
73 81
325 360
I
Tc
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
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Table 6 The fissile material consumed and produced, and tritium production rate (T) Fissile material consumed (amount of nuclides) 239
Pu 9.21×1026 U 12.7×1026
233
Fissile material produced (amount of nuclides)
Tritium production rate (T)
69.4×1026 38.6×1026
1.5 1.4
Table 7 The HLW in the initial and after 30 year burning and the total
U produced (1024)
241
243
244
237
235
129
233
99
11 080 3
1959 23.8
353 25.7
9825 3.67
4886 57.6
13494 18.7
249 000
54 541 0.007
Am
Initial After 30 years operation
233
Am
Cm
Np
Cs
I
U
Tc
level is similar to a fast breeding reactor. It will not bring about many serious problems of thermal hydraulics. 2. The energy-amplifying factor is about 100 for both schemes. From the blanket, high energy multiplying can be attained, It can not only provide necessary plasma heating power on the condition that the fusion can not reach energy self-sustaining, but also act as an energy amplifier (EA) to output high thermal power. 3. The HLW can be effectively transmuted. After 500 days of operation, the blanket can burn 241 Am, 243Am, 237Np, 244Cm, 135Cs, 129I and 99Tc produced by 86, 61, 78, 2, 12 and 13 GWtY LWR, respectively, when Pu is neutron multiplier and
the blanket can burn 241Am, 243Am, 237Np, 244Cm, 135 Cs, 129I and 99Tc produced by 77, 55, 68, 2, 13 and 14 GWtY LWR, respectively when 233U is neutron multiplier Table 5. 4. The critical factor is 0.95 in both the schemes at initial and decreases with time. The blanket maintains deep subcritical status to ensure the critical safety. 5. The tritium can be self-sustained. 6. The blanket can not only provide the fuel for neutron multiplying in the blanket, but also can provide the fuel for fission reactors about 2770 Kg 239Pu with U-Pu cycle of 1011 Kg 233U with Th–U cycles.
Fig. 4. keff, power density as the functions of time.
Fig. 5. The BHP as the function of decaying time.
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B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
Fig. 6. The BHP of the blanket as the functions of operation times.
4. Hybrid reactor as a novel RCNPS concept The multi-functional blankets mentioned in last section show their functions of energy amplifying, HLW transmutation, tritium self-sustaining and fissile material breeding (Table 6). One of the key issues of RCNPS is the radioactivity of the system. The possible medium products from the transmutation process may bring about even more serious radioactivity. Although these vice-products do last a much shorter time than the nuclides transmuted, the radioactivity and biological hazard (BHP) of some of them, such as 238Pu, are much more serious than the original nuclides. The Th–U cycle for multi-functional blanket is picked out for the analysis. Considering the difficulties of partitioning, the HLW was remained in the blanket and no further HLW added once the certain amount of HLW are loaded into the blanket initially. In order to keep the neutron flux at a certain level, the neutron multiplier (233U) should be added into the blanket after a certain operation time which can be easily fulfilled by adding 233U pebbles and taking out the burned 233 U fuel pebbles. The operation time of the blanket is 30 years. Different from the scenario mentioned in last section, the fusion neutron wall loading is assumed 1 MW m − 2. In our calculation, the time
step for burn-up and transport calculation is selected as 1 year. At the end of each operation year, the neutron multiplier is supplemented and U-Th pebbles are unloaded and replaced by new 232 Th pebbles when the enrichment of 233U is \ 3% (Table 7 shows the total 233U produced). Figs. 4–6 and Table 7 give the calculation results. The critical factor remains in the range of 0.94–0.96 and thus the critical safety can be ensured. From Fig. 4, we can see that, the power density has large variations. If sustaining the power density to a certain level, the choice of the density of the neutron multiplier needs modification and the material configurations of the fissile breeding zone need special considerations at a certain operation period. The work shown here is just a primary conceptual study and must be improved in the future. Table 7 shows that the initially loaded HLW are almost transmuted completely. Fig. 5 gives the comparison of the BHP level of the blanket at different decaying time in the case of the natural decay of loaded HLW and 10-, 20- and 30-year operation, the BHP of the blanket decreased to about 4% of the initial level. The blanket together with the driver is verified as a new kind of RCNPS.
5. Summary The multi-functional blanket, whether using Pu of 233U as neutron multiplier, can breed fissile material, transmute HLW and meanwhile produce a high energy output. After 30 year burning, the BHP level of the blanket is much lower compared with the initial level. The RCNPS concept of hybrid reactor is primarily verified. Some serious issues include thermal-hydraulics and structural material selection and the radiation damage need to be studied.
Acknowledgements This work is sponsored by the national natural science fund and national high-technology program.
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
References [1] F. Carminati, C. Rubbia, et al., CERN/AT/93-47 (ET) (1993). [2] Institute of Plasma Physics (ASIPP) and Institute of Southwest Institute of Physics (SWIP), The final report of the detailed conceptual design of fusion experimental breeder, 1996. [3] L.J. Qiu, Y.C. Wu, Q. Xu, Q.Y. Huang, Transmutation of 90Sr using fusion-fission hybrid rectors, IAEA-CN-56/ G-24, Wurzburg, 30 Sept.–7 Oct., 1992. [4] L.J. Qiu, et al., A compact tokamak transmutation CN60/F-II-6, Seville, Spain, 26 Sept.–1 Oct., 1994. [5] CCC-464, Bison 1.5, A one dimensional discrete ordinate neutron transport and burn-up calculation code system, RSIC computer collection. [6] Y. Wu, Neutronics study and analysis of the helium gas cooled blanket of fusion experimental breeder, Chin. J.
.
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Nucl. Sci. Eng. 16 (2) 133. [7] J. Stoer, R. Burlirch, Introduction to Numerical Analysis, Ch 7, Springer – Verlag, New York, 1980. [8] I.I. Bondarenko, Group constants for nuclear reactor calculation. [9] B.J. Xiao, L.J. Qiu, Steady state thermal-hydraulic models of pebble bed blankets on hybrid reactor, Fus. Eng. Des. 27 (1995). [10] Bingjia Xiao, The neutronics and thermal-hydraulics of the hybrid reactor blanket and the study of a novel radioactivity clear nuclear power system. Ph.D dissertation, Institute of Plasma Physics, Academia Sinica, Jan., 1997. [11] K.M. Feng, Y.X. Yang, J.H. Huang, The radioactivity calculation of hybrid reactors, Chin. J. Nucl. Sci. Eng. 9 (1) (1989) 76. [12] L.J. Qiu, B.J. Xiao, Y.P. Chen, Q.Y. Huang, Z. Guo, et al., A low aspect ratio tokamak transmutation reactor, Fusion Eng. Des. 41 (1998) 437 – 442.
.
Neutronics study and analysis of a multifunctional blanket Bingjia Xiao *, Lijian Qiu Institute of Plasma Physics, Academia Sinica, P.O. Box 1126, 230031 Hefei Anhui, People’s Republic of China
Abstract The analysis codes of neutronics, thermal-hydraulics and radioactivity and after-heat calculation are developed and briefly introduced. A multi-function blanket is studied. The analysis shows that the blanket can transmute almost all the HLW loaded, breed large amounts of fissile material and, meanwhile, output a large amount of thermal power during a 30-year operation time. The blanket with its driver is verified as a radioactivity clean nuclear power system. © 1998 Elsevier Science S.A. All rights reserved.
1. Introduction The fusion power for commercial use needs a long time to develop. It is important to find near term applications of controlled fusion in which the parameter requirements are far less stringent, to promote the progress of the fusion power. Fusion breeder and high level waste (HLW) transmutor are candidate selections for fusion’s earlier applications. In 1990, Rubbia [1] proposed a concept that an accelerator driven subcritical system can be a radioactivity clean nuclear power system (RCNPS). This concept has attracted the attention of many scientists. By this concept, the main problems accompanying fission reactors, such as the shortage of fissile material, the high level waste and the risk of nuclear diffusion, can be effectively solved. Can the hybrid reactor be designed as a RCNPS? Many works [2 – 4] show the possibilities of hy-
* Corresponding author. E-mail: [email protected]
brid reactors to breed fissile material and to transmute HLW. Principally, the RCNPS concept can be realised in the hybrid reactor. Hybrid reactors designed as a breeder and HLW transmutor can serve as a new kind of RCNPS. It can put forward the fusion application and fusion development, and put forward the fission development as well. This paper gives the design and analysis of a blanket that can breed tritium for self-sustaining, breed fissile material and transmute both fission products and actinides. Finally, the RCNPS concept as a hybrid reactor is verified.
2. Analysis codes and their development
2.1. Re6isions to BISON 1.5 BISON 1.5 is a code for 1-D neutron transport and burn-up calculation [5]. A study shows that in a hybrid reactor, neutronics simulation results have little difference by BISON 1.5 (1-D)
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B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
and by MCNP (3-D) [6]. BISON 1.5 code can be used in the blanket neutronics analysis. It has some limitation to be applied into the transmutation calculation such as: (1) the neutron data library does not contain the nuclides needed; (2) it does not consider the resonance self-shielding effect; and (3) the burn-up calculation has difficulty being applied in the complicated process for transmutation blanket may meet. So we made revisions to BISON 1.5. The new data library, a 46-group neutron plus 21-group photon group parameter library, which contains all the nuclides and all main cross sections that may be met in the transmutation issue, was created from ENDF-B/ VI by the codes NJOY and TRANSIX. For Burnup computations, we propose the Burlirch–Stoer numerical method [7] to replace the BATEMAN method [5] so that the complicated burn-up process can be solved and the burn-up data library does not need any modifications. A calculation example shows the correctness of this revisions [10]. The original burn-up chain is extended to consider all the main process as may occur in the HLW transmutation and fuel breeding issue, and described completely in a new burn-up library. We also revised BISON 1.5 to consider the resonance self-shielding effects by the BONDARENKO method [8]. Other minor revisions, such as the user interfaces and interface to the new library and thermal-hydraulics simulation, were also made.
2.2. Thermal-hydraulics and neutronics coupled calculation The thermal-hydraulics is an important aspect of the blanket. We developed a code, THPBHR, to analyze the 2-D thermal-hydraulics for the blanket with either pebble bed or plate-like fuel elements with various kinds of cooling scenarios [9,10]. In the case of a soft neutron spectrum and critical factor near to the critical point (keff = 1), the neutronics and thermal-hydraulics coupled effect is important. So we combined THPBHR with revised BISON 1.5 to form a thermal-hydraulics and neutronics coupled analysis code BITH [10].
Fig. 1. The neutron transport, burn-up, BHP and thermal-hydraulics calculation schemes.
2.3. Radioacti6ity, BHP and after-heat calculation The radioactivity, BHP and after-heat calculation can be performed by FDKR [11]. It was revised to be coupled with a revised BISON 1.5. After the burn-up calculation and neutron transport calculation are fulfilled at a certain time step, the results can be supplied into the FDKR to perform the radioactivity calculation. All the calculation processes can be seen in Fig. 1.
2.4. The neutron wall loading requirements for the transmutation of HLW It is proposed that Pu can serve as the neutron multiplier to reduce the requirements of the neutron wall loading to 1 MW m − 2 for the purpose of transmutation. If the requirements can be further lowered, fusion application for HLW transmutation can be more attractive. A blanket is designed to accomplish dual functions, namely, transmuting the high level nuclear waste (Np, Am, Cm and Pu isotopes produced in
Fig. 2. Blanket configuration and material composition for minor actinides transmutation.
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
591
Table 1 The HLW transmuted after 500 days with nuclear wall loading 1 and 0.5 MW m−2 241
243
244
2910c 3331d 13c 14d 11c 13.3d
360 404 9 9 7.8 9.12
−18a −31.3a −2a −4 −2.2a −3.9a
Am
Nuclides Transmuted (1.e24) Number of LWR? b (Normalized to 1 GW year) Transmutted fraction (%)
Am
Cm
237
238
239
2250 2594 11 12 10 11.7
−1107a −1343a −18a −20 −358a −444a
1470 1532 0.6 0.6 12.9 15.3
Np
Pu
Pu
240
241
242
270 321 0.3 0.3 5.1 6.22
273 301 0.8 0.8 16 17.5
−99a −129a −0.5a −0.6 −9.3a −12a
Pu
Pu
Pu
a
Negative values express the nuclide density increased. Normalizing method: N, nuclides transmuted in 1 GW year thermal power hybrid reactor blanket/nuclides produced in 1 GW year thermal power LWRs. c For 0.5 MW m−2 neutron wall loading. d For 1 MW m−2 neutron wall loading. b
Table 2 Transmutation efficiency (total fission reaction rate/(n, gamma) reaction rate) 241
243
Am
0 500 days a b
1.0a 1.0
244
Am
1.1b 1.1
1.4 1.4
237
Cm
1.4 1.5
3.3 3.4
238
Np
3.4 3.5
1.6 1.6
239
Pu
1.7 1.7
5.3 5.3
5.4 5.5
240
Pu
14. 14
241
Pu
14 15
5.0 5.1
242
Pu
5.2 5.3
12 11
Pu
11 12
4.6 4.7
4.7 4.8
For 0.5 MW m−2 neutron wall loading. For 1 MW m−2 neutron wall loading.
fission reactor) and breeding the tritium for the use in fusion reactor core. The 1-D layout and the material composition of the blanket are shown Fig. 2. The coolant is assumed as helium. The rest materials of all zones in Fig. 2, Table 4 are coolants (helium). Tables 1 – 3 give the results of the nuclides transmuted after 500 days burning. We can see that by raising the fuel loading density and hence, raising the keff and neutron flux, we can reach the transmutation object with lower neutron wall loading (0.5 MW m − 2) which may be achieved in the near future. One of the main figures of merit of transmutation is the transmutation efficiency, that is, the ration of the total fission reaction to the total absorption reaction rate. It is shown that the efficiency of this kind of blanket is so high that even if the neutron wall loading is lowered to 0.5 MW m − 2, the transmutation efficiency and the tritium breeding rate are high enough. The thermal power in the blanket is 4.9 (initial)–3.4 GW (after 500 days) and 9.7 – 3.4 GW with neutron wall loading 1 and 0.5 MW m − 2, respec-
tively. It is also shown that keff decreases with time that is beneficial to the safety of the blanket, and the tritium breeding rate (T) is so high that it can provide tritium for several fusion reactors of same fusion power. The heat generated in the blanket can be release off with the existing technology used in fast fission reactors.
3. Multifunction transmutation blanket It is found that the tritium-breeding ratio in the scenario introduced last section for transmuting minority actinides is very large. If this ratio is reduced and the extra neutrons are lowered down to a certain extent after they pass across the tritium-breeding zone, we can use them to breed fissile material and transmute the fission products. Table 4 gives the blanket 1-D configuration and material compositions using Pu as neutron multiplier. In this scenario, we use natural uranium to breed 239Pu, both for self-sustaining of Pu and for providing fissile material for fission reactors.
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
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Table 3 keff, tritium breeding ratio and power density of the blanket Time (days)
keff T P (W cm−3) a b
0
100
200
300
400
500
0.89a/0.97b 8.3/29 180/348
0.88/0.95 7.6/18 162/211
0.87/0.94 7.1/14 148/170
0.86/0.93 6.6/13 138/149
0.85/0.92 6.3/12 130/135
0.85/0.92 6.1/11 123/124
For 0.5 MW m−2 neutron wall loading. For 1 MW m−2 neutron wall loading.
Table 4 The 1-D configuration and material composition of blanket with Pu as neutron multiplier Zone no.
Zone function
Width (cm)
Material
1 2 3 4 5 6 7 8 9 10
Fusion core First wall Minor actinide transmutation Wall Tritium breeding Wall Fissile material breeding Wall Fission product transmutation Reflector
100 0.5 40 0.5 5 0.5 10 0.5 20 40
Vacuum ss-316 (70%) ss-316 (5%), Np Am Cm (4.9%), Pu (1.6%), Zr (1%) ss-316 (80%) ss-316 (5%), natural Li2O (5%) ss-316 (80%) ss-316 (5%), natural U (55%) ss-316 (80%) ss-316 (5%), 135Cs (3%), 129I (3%), 99Tc (4%) C (75%), Be (15%)
The other fuel cycle candidate is Th–U. We also designed a blanket for this fuel cycle, that is using 233U as neutron multiplier and 232Th to
breed 233U. The blanket configuration for Th–U fuel cycle is almost the same as the one mentioned in Table 4 except that natural U (55%) is replaced by 232Th (55%) in zone 7 and Pu (1.6%) is replaced by 233U (1.4%). The reference fusion core parameters are from Ref. [12]. The calculation results are shown in Tables 3 and 4 and Fig. 3 From the calculation results, we can conclude that: 1. The peak power density of the blanket is about 190 W cm − 3 at initial and decreased with time (shown in Fig. 3). The power density Table 5 The HLW transmuted (kg) with multiplier after 500 day operation 241
243
1441 1296
181 165
Am
Fig. 3. Energy amplifying factor, peak power density and critical factor as the function of operation time with U – Pu cycle.
Pu U
233
Am
244
Cm
−51 −39
233
U or
237
235
1115 982
20 22
Np
239
Cs
Pu as neutron
129
99
73 81
325 360
I
Tc
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
593
Table 6 The fissile material consumed and produced, and tritium production rate (T) Fissile material consumed (amount of nuclides) 239
Pu 9.21×1026 U 12.7×1026
233
Fissile material produced (amount of nuclides)
Tritium production rate (T)
69.4×1026 38.6×1026
1.5 1.4
Table 7 The HLW in the initial and after 30 year burning and the total
U produced (1024)
241
243
244
237
235
129
233
99
11 080 3
1959 23.8
353 25.7
9825 3.67
4886 57.6
13494 18.7
249 000
54 541 0.007
Am
Initial After 30 years operation
233
Am
Cm
Np
Cs
I
U
Tc
level is similar to a fast breeding reactor. It will not bring about many serious problems of thermal hydraulics. 2. The energy-amplifying factor is about 100 for both schemes. From the blanket, high energy multiplying can be attained, It can not only provide necessary plasma heating power on the condition that the fusion can not reach energy self-sustaining, but also act as an energy amplifier (EA) to output high thermal power. 3. The HLW can be effectively transmuted. After 500 days of operation, the blanket can burn 241 Am, 243Am, 237Np, 244Cm, 135Cs, 129I and 99Tc produced by 86, 61, 78, 2, 12 and 13 GWtY LWR, respectively, when Pu is neutron multiplier and
the blanket can burn 241Am, 243Am, 237Np, 244Cm, 135 Cs, 129I and 99Tc produced by 77, 55, 68, 2, 13 and 14 GWtY LWR, respectively when 233U is neutron multiplier Table 5. 4. The critical factor is 0.95 in both the schemes at initial and decreases with time. The blanket maintains deep subcritical status to ensure the critical safety. 5. The tritium can be self-sustained. 6. The blanket can not only provide the fuel for neutron multiplying in the blanket, but also can provide the fuel for fission reactors about 2770 Kg 239Pu with U-Pu cycle of 1011 Kg 233U with Th–U cycles.
Fig. 4. keff, power density as the functions of time.
Fig. 5. The BHP as the function of decaying time.
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B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
Fig. 6. The BHP of the blanket as the functions of operation times.
4. Hybrid reactor as a novel RCNPS concept The multi-functional blankets mentioned in last section show their functions of energy amplifying, HLW transmutation, tritium self-sustaining and fissile material breeding (Table 6). One of the key issues of RCNPS is the radioactivity of the system. The possible medium products from the transmutation process may bring about even more serious radioactivity. Although these vice-products do last a much shorter time than the nuclides transmuted, the radioactivity and biological hazard (BHP) of some of them, such as 238Pu, are much more serious than the original nuclides. The Th–U cycle for multi-functional blanket is picked out for the analysis. Considering the difficulties of partitioning, the HLW was remained in the blanket and no further HLW added once the certain amount of HLW are loaded into the blanket initially. In order to keep the neutron flux at a certain level, the neutron multiplier (233U) should be added into the blanket after a certain operation time which can be easily fulfilled by adding 233U pebbles and taking out the burned 233 U fuel pebbles. The operation time of the blanket is 30 years. Different from the scenario mentioned in last section, the fusion neutron wall loading is assumed 1 MW m − 2. In our calculation, the time
step for burn-up and transport calculation is selected as 1 year. At the end of each operation year, the neutron multiplier is supplemented and U-Th pebbles are unloaded and replaced by new 232 Th pebbles when the enrichment of 233U is \ 3% (Table 7 shows the total 233U produced). Figs. 4–6 and Table 7 give the calculation results. The critical factor remains in the range of 0.94–0.96 and thus the critical safety can be ensured. From Fig. 4, we can see that, the power density has large variations. If sustaining the power density to a certain level, the choice of the density of the neutron multiplier needs modification and the material configurations of the fissile breeding zone need special considerations at a certain operation period. The work shown here is just a primary conceptual study and must be improved in the future. Table 7 shows that the initially loaded HLW are almost transmuted completely. Fig. 5 gives the comparison of the BHP level of the blanket at different decaying time in the case of the natural decay of loaded HLW and 10-, 20- and 30-year operation, the BHP of the blanket decreased to about 4% of the initial level. The blanket together with the driver is verified as a new kind of RCNPS.
5. Summary The multi-functional blanket, whether using Pu of 233U as neutron multiplier, can breed fissile material, transmute HLW and meanwhile produce a high energy output. After 30 year burning, the BHP level of the blanket is much lower compared with the initial level. The RCNPS concept of hybrid reactor is primarily verified. Some serious issues include thermal-hydraulics and structural material selection and the radiation damage need to be studied.
Acknowledgements This work is sponsored by the national natural science fund and national high-technology program.
B. Xiao, L. Qiu / Fusion Engineering and Design 41 (1998) 589–595
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