Preliminary application of neutronics calculation in LFR reactor with metallic fuel using dragon code
-
摘要: 铅铋合金或铅冷却快堆（LFR）是具有良好应用前景的第四代先进核能系统之一。针对环形芯体金属燃料（UZr，UPuZr）LFR的燃料组件与堆芯，利用Dragon/Donjon程序开展中子学计算，获得了基于ENDF/B 8.0库的172群和295群多群中子数据库、输运方法（SP3）和扩散方法（MCFD）的结果及其与蒙卡程序RMC的偏差。采用SP3算法针对UZr燃料得到的keff偏差小于550×10−5；对于UPuZr燃料采用MCFD算法得到的keff偏差小于−700×10−5。控制棒组件价值的偏差小于7.6%；172群和295群库的结果基本无差异。应用SP3算法的燃料组件功率偏差小于±6.0%；SP3算法的偏差小于MCFD的。结果证明，Dragon/Donjon程序在金属燃料铅铋快堆物理分析中具有可行性。
-
关键词:
- 铅铋快堆 /
- 金属燃料 /
- 反应堆物理 /
- Dragon/Donjon程序
Abstract: Lead-bismuth/lead cooled fast reactor (LFR) is one of the fourth-generation advanced nuclear energy systems with good application prospects. Aiming at the application of two metallic fuels (UZr, UPuZr) with annular slug in an LFR fuel assembly and a typical LFR core, the Dragon/Donjon code was used to perform neutronics calculations. The results of 172-group and 295-group neutron databases based on the ENDF/B 8.0 library and the transport method (SP3) and the diffusion method (MCFD) were obtained and compared with the results using RMC code. The keff deviations using SP3 algorithm for UZr fuel are less than 550×10−5, and for UPuZr fuel, the keff deviations obtained by MCFD algorithm are less than −700×10−5. The maximum deviation of control rod worth is 7.6%, and the results using 172-grouplib and 295-grouplib are basically the same. By using the SP3 algorithm, the fuel assembly power deviations are less than ±6.0%, and the deviation with the SP3 algorithm is less than that with MCFD algorithm. The results preliminarily prove the feasibility of the Dragon/Donjon code for reactor physics analysis of LFR with metallic fuel.-
Key words:
- lead bismuth cooled fast reactor /
- metallic fuel /
- reactor physics /
- Dragon/Donjon code
-
图 3 用于群常数制作的组件计算模型
Figure 3. Calculation models for assembly group constant computation
图 4 Dragon/Donjon程序中的LFR堆芯计算模型示意图
Figure 4. Calculation model of an LFR core with Dragon/Donjon code
FA: fuel pellet of fuel assembly; GC: gas chamber of fuel assembly; AR: axial reflector of fuel assembly; CR: B4C pellet section of control rod assembly; LBE; assembly section with empty HEXCAN and coolant. I/O: inner/outer fuel zone; U/L: upper/lower part of core.
图 5 Dragon/RMC程序得到的UZr燃料组件中子能谱
Figure 5. Neutron spectrum of UZr fuel in a fuel assembly using Dragon or RMC code
表 1 燃料组件、径向反射层组件和控制棒组件的设计参数(293 K)
Table 1. Designed dimensions of fuel, radial reflector and control rod assemblies (293 K)
item number of
rods in
assemblyHexcan outer
flat-to-flat
size/mmHexcan wall
thickness/mmpin
pitch/mmpin cladding
inner
diameter/mmpin cladding
outer
diameter/mmpin cladding
thickness/mmfuel assembly 271 173 4.5 9.80 7.37 8.500 0.565 radial reflector assembly 91 173 4.5 16.90 — 15.838 — CR assembly 31 173 4.5 22.77 18.00 19.000 0.500 item pin cladding
outer diameter,
including wrap
wires/mmfuel
pellet
outer
diameter/mmfuel pellet
inner hole
diameter/mmabsorber
pellet
diameter/mmrod body
outer diameter
of CR
assembly/mmrod body
wall thickness
of CR
assembly/mmassembly
(section)
axial
length/mfuel assembly 8.584 7.37 4.944 — — — upper reflector section: 0.5
gas chamber section: 1.1
fuel pellet section: 0.9
lower reflector section: 0.5radial reflector assembly — — — — — — 3.0 CR assembly — — — 17 149 2 absorber pellet section:1.0
lower reflector section: 0.5
LBE section: 1.5
下载: 导出CSV
表 2 并群后的24群能群结构
Table 2. 24 group energy structure for condensation
group
numberupper energy limit
of 172-group lib/eVupper energy limit
of 295-group lib/eV1 19.6400×106 19.6400×106 2 10.0000×106 10.0000×106 3 6.0653×106 6.0653×106 4 3.0119×106 3.3287×106 5 2.0190×106 1.9014×106 6 1.2246×106 1.2870×106 7 8.2085×105 8.6001×105 8 4.9787×105 4.9400×105 9 3.0197×105 3.2065×105 10 1.2277×105 1.6506×105 11 1.1109×105 1.1562×105 12 6.7379×104 6.7379×104 13 3.6979×104 3.6979×104 14 1.6616×104 2.2699×104 15 1.5034×104 1.4900×104 16 9.1188×103 9.1188×103 17 5.0045×103 5.0045×103 18 3.3546×103 3.4811×103 19 1.5073×103 1.8118×103 20 1.2341×103 1.1347×103 21 6.7729×102 6.7729×102 22 3.7170×102 4.1909×102 23 3.0432×102 2.9592×102 24 1.3674×102 1.4666×102 lower energy limit/eV 1.0000×10−5
下载: 导出CSV
表 3 铅铋堆燃料组件中子无限增殖因子的计算结果
Table 3. Calculation results of kinf in an LFR fuel assembly
solution kinf difference with RMC/10−5 UZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyUZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyRMC code 1.33254 1.35197 1.23743 — — — Dragon code (172-group lib) 1.31722 1.33137 1.22903 −873 −1144 −552 Dragon code (295-group lib) 1.31272 1.32382 1.22524 −1133 −1573 −804
下载: 导出CSV
表 4 铅铋堆燃料组件冷却剂密度效应的计算结果
Table 4. Calculation results of coolant density effect in an LFR fuel assembly
solution reactivity change of coolant density effect, ΔρLBE/10−5 relative difference/% UZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyUZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyRMC code 265 383 214 — — — Dragon code (172-group lib) 247 374 224 −6.8 −2.3 4.7 Dragon code (295-group lib) 248 375 223 −6.4 −2.1 4.2
下载: 导出CSV
表 5 铅铋堆燃料组件燃料多普勒系数的计算结果
Table 5. Calculation results of fuel Doppler coefficient in an LFR fuel assembly
solution fuel Doppler coefficient, kD/10−5 relative difference/% UZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyUZr fuel
assemblyUPuZr fuel
assemblyUO2 fuel
assemblyRMC code −297 −380 −743 — — — Dragon code (172-group lib) −280 −359 −715 −5.7 −5.5 −3.8 Dragon code (295-group lib) −300 −378 −725 1.0 −0.5 −2.4
下载: 导出CSV
表 6 Dragon/Donjon程序得到的铅铋堆keff及其计算偏差(293 K)
Table 6. keff
and its discrepancy of a LFR core using Dragon/Donjon code (293 K) solution keff difference with RMC/10−5 LFR core with
UZr fuelLFR core with
UPuZr fuelLFR core with
UZr fuelLFR core with
UPuZr fuelRMC code 1.02934 1.03132 — — 172-group lib MCFD 1.04436 1.03005 1398 −120 SP3 1.03517 1.01923 548 −1151 295-group lib MCFD 1.04139 1.02432 1124 −663 SP3 1.03102 1.01247 159 −1805
下载: 导出CSV
表 7 Dragon/Donjon程序得到的铅铋堆控制棒组件价值及其计算偏差
Table 7. Discrepancy of control rod worth of an LFR core using Dragon/Donjon code
solution CR worth/10−5 relative difference/% LFR core with
UZr fuelLFR core with
UPuZr fuelLFR core with
UZr fuelLFR core with
UPuZr fuelRMC code −10570 −10022 — — 172-group lib MCFD −10459 −10250 −1.1 2.3 SP3 −11056 −10787 4.6 7.6 295-group lib MCFD −10405 −10235 −1.6 2.1 SP3 −10985 −10743 3.9 7.2
下载: 导出CSV
表 8 Dragon/Donjon程序得到的铅铋堆燃料组件功率的计算偏差(D/R-1)
Table 8. Discrepancy of fuel assembly power of an LFR core using Dragon/Donjon code (D/R-1)
XS library method relative difference of LFR core with UZr fuel/% relative difference of LFR core with UPuZr fuel/% Max Min RMS Max Min RMS CR-P and CR-S
withdrawn172-group lib MCFD 1.6 −4.4 1.7 1.7 −4.8 1.9 SP3 5.5 −1.5 1.8 3.9 −1.3 1.4 295-group lib MCFD 3.0 −8.9 3.7 2.7 −8.3 3.4 SP3 1.4 −3.9 1.4 1.3 −3.7 1.3 CR-P at critical
location;
CR-S withdrawn172-group lib MCFD 3.2 −5.4 2.3 3.1 −5.9 2.4 SP3 5.1 −1.7 1.7 3.9 −1.9 1.3 295-group lib MCFD 4.7 −9.9 4.3 4.3 −9.2 4.0 SP3 2.2 −4.7 1.9 2.1 −4.9 1.8
下载: 导出CSV
-
[1] Carmack J. Fuel development for advanced reactors[R]. Washington DC: Department of Energy, 2016. [2] Waltar A E, Todd D R, Tsvetkov P V. Fast spectrum reactors[M]. New York, USA: Springer, 2012. [3] Mum J W, Kin H J, Sweidan F B, et al. Metallic fuel performance evaluation for micro lead cooled fast reactor[C]//Proceedings of Transactions of the Korean Nuclear Society Virtual Spring Meeting. 2020. [4] 邹小亮, 柏云清, 王明煌, 等. 铅冷行波堆堆芯物理设计及中子学特性分析[J]. 核科学与工程, 2018, 38(4):590-597. (Zou Xiaoliang, Bai Yunqing, Wang Minghuang, et al. Core neutronics design and analysis for the Pb-cooled traveling wave reactor[J]. Nuclear Science and Engineering, 2018, 38(4): 590-597 doi: 10.3969/j.issn.0258-0918.2018.04.009 [5] Nuclear Energy Agency. Benchmark for neutronic analysis of sodium-cooled fast reactor cores with various fuel types and core sizes[R]. NEA/NSC/R(2015)9, 2016. [6] Hébert A. DRAGON5 and DONJON5, the contribution of École Polytechnique de Montréal to the SALOME platform[C]//Proceedings of the 3rd International Conference on Physics and Technology of Reactors and Applications. Tetouan, Morocco, 2014. [7] Sukarno D H. Neutronics analysis of PWR core using DRAGON, TRIVAC, and DONJON computer codes[C]//The 2nd International Conference on Energy Sciences (ICES). Bandung, Indonesia, 2018. [8] Al Zain J, El Hajjaji O, El Bardouni T, et al. Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code[J]. Journal of Radiation Research and Applied Sciences, 2018, 11(3): 255-261. doi: 10.1016/j.jrras.2018.04.002 [9] 赵文博, 谢金森, 谢芹, 等. DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性[J]. 核技术, 2017, 40:060602-862). (Zhao Wenbo, Xie Jinsen, Xie Qin, et al. Feasibility of DRAGON&DONJON code for MSR core burnup calculation[J]. Nuclear Techniques, 2017, 40: 060602-862) doi: 10.11889/j.0253-3219.2017.hjs.40.060602 [10] 贾国斌, 伍建辉, 陈金根, 等. 基于Dragon与Donjon程序的液态熔盐实验堆临界计算与分析[J]. 原子能科学技术, 2019, 53(5):853-862. (Jia Guobin, Wu Jianhui, Chen Jingen, et al. Critical calculation and analysis of liquid molten salt experimental reactor based on Dragon and Donjon codes[J]. Atomic Energy Science and Technology, 2019, 53(5): 853-862 [11] Choi H, Choi M, Hon R. Benchmarking DRAGON/PARCS against KRITZ and FFTF measurements[J]. Nuclear Technology, 2019, 205(3): 486-505. doi: 10.1080/00295450.2018.1495001 [12] Ponomarev A, Mikityuk K, Zhang Liang, et al. Superphénix benchmark Part I: results of static neutronics[J]. Journal of Nuclear Engineering and Radiation Science, 2022, 8: 011320. doi: 10.1115/1.4051449 [13] Wang Kan, Li Zeguang, She Ding, et al. RMC - a Monte Carlo code for reactor core analysis[J]. Annals of Nuclear Energy, 2015, 85: 121-129. doi: 10.1016/j.pnucene.2015.06.013 [14] Ponomarev A, Bednarova A, Mikityuk K. New sodium fast reactor neutronics benchmark[C]//Proceeding of PHYSOR 2018: Reactor Physics Paving the Way Towards More. Cancun, Mexico, 2018. [15] Hofman G L, Leibowitz L, Kramer J M, et al. Metallic fuels handbook[R]. Argonne: Argonne National Laboratory, 1985. [16] Nuclear Energy Agency. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies[M]. Nuclear Energy Agency, 2015. [17] Brown D A, Chadwick M B, Capote R, et al. ENDF/B-VIII. 0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data[J]. Nuclear Data Sheets, 2018, 148: 1-142. doi: 10.1016/j.nds.2018.02.001 [18] Faure B, Marleau G. Simulation of a sodium fast reactor: effect of B1 leakage models on group constant generation[J]. Annals of Nuclear Energy, 2017, 99: 484-494. doi: 10.1016/j.anucene.2016.10.002 [19] Rachamin R, Wemple C, Fridman E. Neutronic analysis of SFR core with HELIOS-2, serpent, and DYN3D codes[J]. Annals of Nuclear Energy, 2013, 55: 194-204. doi: 10.1016/j.anucene.2012.11.030 [20] Hebert A, Sekki D, Chambon R. A user guide for DRAGON version5[M]. Canada: École Polytechnique de Montréal, 2019. -
点击查看大图
计量
- 文章访问数: 928
- HTML全文浏览量: 312
- PDF下载量: 76
- 被引次数: 0
