Study on Activation-Induced Dose Calculation of Reactor Structural Materials
-
摘要: 反应堆结构材料在堆芯中子辐照下由于中子活化反应而产生大量的放射性核素，其衰变光子是反应堆停堆检修、换料、退役过程中工作人员职业照射剂量的重要来源。本文基于严格两步法（R2S），研究了反应堆结构材料栅元活化计算方法，并基于蒙卡粒子输运程序（MCNP）与点活化计算程序（ORIGEN）建立了反应堆结构材料活化剂量计算软件（MOCA）。通过开发功能接口与数据接口程序实现输运程序与活化计算程序的自动耦合，进而实现“中子输运-活化分析-剂量计算”全自动耦合分析。利用M5包壳活化计算模型、不锈钢活化计算模型和NUREG/CR-6115压水堆模型对MOCA进行基准验证，证明了MOCA的正确性与可靠性。Abstract: Numerous radioactive nuclides are produced from the neutron activation reaction of reactor structural materials under the core neutron irradiation. The decay photon from activated reactor structural materials is an important source of occupational exposure dose for workers during reactor shutdown & maintenance, refueling or decommissioning. This paper studies the cell activation calculation method for reactor structural materials based on Rigorous two step (R2S) method, and develops a reactor structural material activation dose calculation program, MOCA, based on the Monte Carlo particle transport code MCNP and point activation calculation code ORIGEN. Then, this study develops the functional interface and data interface codes to realize the automatic coupling of the transport code and activation calculation code thus to realize fully automated “neutron transport-activation analysis - dose calculation” coupled analysis. Finally, the M5 fuel cladding activation calculation model, stainless steel activation calculation model and NUREG/CR-6115 PWR model are used to perform benchmark verification of the MOCA, which proves that the MOCA is adequate and reliable.
-
Key words:
- Structural material /
- R2S method /
- Activation calculation /
- Dose calculation
-
图 3 高度为−60 cm处的径向光子剂量率分布
Figure 3. Radial Activation-Induced Dose Rate Distribution at the Height of −60 cm
图 4 径向位置处于5 cm的轴向光子剂量率分布
Figure 4. Axial Activation-Induced Dose Rate Distribution at the Radial Position of 5 cm
图 5 高度为−60 cm处的径向光子剂量率分布
Figure 5. Radial Activation-Induced Dose Rate Distribution at the Height of −60 cm
图 6 径向位置处于5 cm的轴向光子剂量率分布
Figure 6. Axial Activation-Induced Dose Rate Distribution at the Radial Position of 5 cm
图 7 NUREG/CR-6115水平剖面（左）与垂直剖面（右）
Figure 7. Horizontal Profile (Left) and Vertical Profile (Right) of NUREG/CR-6115
图 8 高度为9 cm处的径向光子剂量率分布
Figure 8. Radial Photon Dose Rate Distribution at Height of 9 cm
图 9 径向位置处于5 cm的轴向光子剂量率分布
Figure 9. Axial Activation-Induced Dose Rate Distribution at the Radial Position of 5 cm
表 1 M5包壳模型计算结果
Table 1. Calculation Results of M5 Cladding Model
核素 衰变光子的
产生份额核子数 相对
误差/%SuperMC MOCA 59Ni 9.971630×10−1 2.371270×1020 2.366446×1020 −0.02 91Y 8.080200×10−5 4.050070×1010 4.094195×1010 1.09 93Zr 2.693620×10−3 2.974000×1010 3.017826×1010 1.47 94Nb 4.875270×10−5 3.048820×1012 3.283291×1012 7.69 
下载: 导出CSV
表 2 不锈钢活化计算模型计算结果
Table 2. Results of Stainless Steel Activation Calculation Model
核素 衰变光子的
产生份额核子数 相对
误差/%SuperMC MOCA 14C 6.459063×10−5 8.570000×1013 8.254000×1013 −3.71 32P 2.538009×10−6 2.308067×1010 2.490000×1010 7.87 55Fe 2.617085×10−5 1.666091×1013 1.590000×1013 −4.50 58Co 9.998066×10−1 4.516019×1013 4.230000×1013 −6.34 59Ni 9.723072×10-6 1.720055×1016 1.750000×1016 1.58
下载: 导出CSV
表 3 NUREG/CR-6115模型结构参数
Table 3. Parameters of NUREG/CR-6115 Structure
结构 内半径/cm 外半径/cm 高度/cm 材料 上反射层 — — 32.865 306.7℃水+钢 下反射层 — — 13.97 280℃水+钢 围板 — — 382.115 SS-304不锈钢 旁流 — 190.185 382.115 293.3℃含硼水 吊篮 190.185 193.995 382.115 SS-304不锈钢 内进水间隙 193.995 196.535 382.115 280℃含硼水 热屏蔽层 196.535 200.345 382.115 SS-304不锈钢 外进水间隙 200.345 218.440 382.115 280℃含硼水 压力容器内衬 218.440 219.075 382.115 SS-304不锈钢 压力容器 219.075 240.665 382.115 SA-302B不锈钢 “—”表示不适用
下载: 导出CSV
表 4 NUREG/CR-6115结构材料重要核素产额
Table 4. Key Nuclide Yield of Structural Materials of NUREG/CR-6115
结构 核素 衰变光子的
产生份额核子数 相对误差/% SuperMC MOCA 围板 51Cr 0.0512 9.04304×1021 9.01111×1021 −0.3530 54Mn 0.0812 1.65722×1023 1.61091×1023 −2.7948 55Fe 0.8296 4.57695×1024 4.79988×1024 4.8707 59Fe 0.0051 1.46866×1021 1.46521×1021 −0.2354 58Co 0.0159 1.23000×1022 1.16172×1022 −5.7550 60Co 0.0023 1.35802×1022 1.45523×1022 7.1582 59Ni 0.0001 2.05000×1025 1.89945×1025 −7.3124 63Ni 0.0145 3.39182×1024 3.18262×1024 −6.1678 吊篮 51Cr 0.0536 3.13175×1021 3.07964×1021 −1.66389 54Mn 0.0497 3.25893×1022 3.21736×1022 −1.27565 55Fe 0.8423 1.68271×1024 1.65714×1024 −1.51958 59Fe 0.0047 4.41057×1020 4.42317×1020 0.28571 58Co 0.0325 5.36206×1021 4.76203×1021 −7.46049 60Co 0.0004 1.56667×1021 1.62038×1021 3.42838 59Ni 0.0001 7.80399×1024 7.47500×1024 −4.21564 63Ni 0.0166 1.17560×1024 1.15718×1024 −1.56705 热屏蔽层 51Cr 0.0487 6.06371×1020 5.93619×1020 −2.10300 54Mn 0.0875 1.22718×1022 1.20177×1022 −2.07025 55Fe 0.7782 3.34461×1023 3.24861×1023 −2.87020 59Fe 0.0046 9.44776×1019 9.16060×1019 −3.03941 58Co 0.0652 2.30271×1021 2.15916×1021 −6.53987 60Co 0.0006 5.55609×1020 5.49024×1020 −1.18515 59Ni 0.0001 1.52538×1024 1.47439×1024 −3.34253 63Ni 0.0150 2.28406×1023 2.22262×1023 −2.68968 压力容器 51Cr 0.0000860 1.08765×1017 1.12238×1017 3.19297 54Mn 0.1718164 2.47656×1021 2.52859×1021 2.10120 55Fe 0.8229250 3.79205×1022 3.58747×1022 −5.39515 59Fe 0.0051717 9.89552×1018 1.07346×1019 7.46847 60Co 0.0000006 5.71326×1016 5.50578×1016 −3.63147 93Mo 0.0000002 1.08459×1019 1.14242×1019 5.33207
下载: 导出CSV
-
[1] 王小胡,胡一非,李江波,等. 退役反应堆放射性活化源项计算[J]. 原子能科学技术,2014, 48(5): 893-897. doi: 10.7538/yzk.2014.48.05.0893 [2] 李卫. 基于栅元计数的停堆剂量率计算程序研发及应用[D]. 合肥: 中国科学技术大学, 2018. [3] VALENZE D, IIDA H, PLENTEDA R, et al. Proposal of shutdown dose estimation method by Monte Carlo code[J]. Fusion Engineering and Design, 2001, 55(4): 411-418. doi: 10.1016/S0920-3796(01)00188-0 [4] CHEN Y X, WU Y C, FISCHER U. The rigorous-2-step calculation of shutdown dose rates for nuclear fusion devices[J]. Nuclear Techniques, 2003, 26(10): 763-766. [5] 郑剑,李斌,邹俊,等. 基于网格计数的停机剂量率计算方法研究[J]. 核科学与工程,2016, 36(6): 784-789. doi: 10.3969/j.issn.0258-0918.2016.06.010 [6] CROFF A G. A user’s manual for the ORIGEN2 computer code: ORNL/TM-7175[R]. Clinton Town, Tennessee, USA: Oak Ridge National Lab, 1980 [7] WU Y C, SONG J, ZHENG H Q, et al. CAD-based Monte Carlo program for integrated simulation of nuclear system SuperMC[J]. Annals of Nuclear Energy, 2015(82): 161-168. doi: 10.1016/j.anucene.2014.08.058 [8] WU Y. CAD-based interface programs for fusion neutron transport simulation[J]. Fusion Engineering and Design, 2009, 84(7-11): 1987-1992. doi: 10.1016/j.fusengdes.2008.12.041 [9] WU Y C. Multifunctional neutronics calculation methodology and program for nuclear design and radiation safety evaluation[J]. Fusion Science and Technology, 2018, 74(4): 321-329. doi: 10.1080/15361055.2018.1475162 [10] 杨寿海. 基于遗传算法的多目标智能辐射屏蔽方法研究[D]. 北京: 华北电力大学, 2012. -
计量
- 文章访问数: 193
- HTML全文浏览量: 283
- PDF下载量: 41
- 被引次数: 0
