Research on reactor component activation and dose calculation
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摘要: 放射性源项调查是反应堆退役制订方案、估算费用和进度以及作好辐射防护和应急准备的重要依据。反应堆构件在中子辐照过程中由于中子活化反应会产生大量的放射性核素，其产生的衰变光子是反应堆退役过程中工作人员面临辐射剂量的主要来源。采用蒙特卡罗粒子输运程序（cosRMC、MCNP）和活化计算程序（DEPTH、ALARA）相结合的方法计算反应堆构件在运行一定时间后产生的放射性核素核子数密度、活度和几个主要构件的辐射剂量率。对比通过两个不同活化程序计算得到的计算结果，相对偏差在可接受范围内，表明了cosRMC的活化计算和剂量率计算功能应用于反应堆退役分析的可靠性和准确性。Abstract:
Background The investigation of radioactive source terms serves as a critical basis for formulating reactor decommissioning plans, estimating costs and schedules, and ensuring adequate radiation protection and emergency preparedness.Purpose During neutron irradiation, reactor components undergo neutron activation reactions that generate significant quantities of radionuclides. The decay photons emitted by these nuclides constitute the primary source of radiation exposure for personnel during reactor decommissioning.Methods A combined approach using Monte Carlo particle transport programs (cosRMC, MCNP) and activation calculation programs (DEPTH, ALARA) was employed to calculate the nuclide atom density, activity, and radiation dose rates for key components after a specified operational period.Results Comparing results from the two activation programs shows relative deviations within acceptable limits,Conclusions demonstrating the reliability and accuracy of cosRMC’s activation calculations and dose rate assessment capabilities for reactor decommissioning analysis.-
Key words:
- reactor decommissioning /
- cosRMC /
- ALARA /
- activation calculation /
- dose calculation
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图 1 NUREG/CR-
6115 水平剖面图与垂直剖面图Figure 1. Horizontal cross-section and vertical cross-section of NUREG/CR-
6115 
图 2 堆芯围板、堆芯吊篮、热屏蔽层和压力容器中子能谱
Figure 2. Core baffle, core barrel, thermal shield, and RPV neutron spectrum
图 3 堆芯围板、堆芯吊篮、热屏蔽层和压力容器各能群光子的停堆剂量率相对误差
Figure 3. Core baffle, core barrel, thermal shield, and RPV relative error of the shutdown dose rate for each energy group of photons
表 1 NUREG/CR-
6115 模型结构参数Table 1. Structural parameters of NUREG/CR-
6115 structure inner radius/cm outer radius/cm height/cm materials upper reflector \ \ 32.865 306.7 ℃ water + steel lower reflector \ \ 13.97 280 ℃ water + steel core baffle \ \ 382.115 SS-304 stainless steel by-pass flow \ 190.185 382.115 293.3 ℃ boron water core barrel 190.185 193.995 382.115 SS-304 stainless steel inner inlet water gap 193.995 196.535 382.115 280 ℃ boron water thermal shield 196.535 200.345 382.115 SS-304 stainless steel outer inlet water gap 200.345 218.440 382.115 280 ℃ boron water RPV Liner 218.440 219.075 382.115 SS-304 stainless steel RPV 219.075 240.665 382.115 SA-302B stainless steel 注：“\”表示不适用。 
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表 2 重要核素产额
Table 2. Yield of important nuclides
structure nuclides atom density/cm−3 |relative error|/% ALARA DEPTH core baffle 60Co 2.0926E+16 2.1057E+16 0.6278 51Cr 8.6670E+17 8.6683E+17 0.0148 55Fe 6.3342E+18 6.3550E+18 0.3280 59Fe 3.0794E+16 3.0802E+16 0.0268 54Mn 2.4371E+17 2.4416E+17 0.1826 48Sc 2.0590E+09 2.0673E+09 0.4039 65Zn 2.5127E+09 2.5267E+09 0.5573 core barrel 60Co 9.9192E+14 9.9564E+14 0.3751 51Cr 8.2801E+16 8.2817E+16 0.0197 55Fe 5.9648E+17 5.9844E+17 0.3283 59Fe 2.6898E+15 2.6903E+15 0.0188 54Mn 2.1917E+16 2.1957E+16 0.1864 48Sc 2.3200E+07 2.3296E+07 0.4123 65Zn 1.8511E+06 1.8616E+06 0.5649 thermal shield 60Co 3.6517E+14 3.6648E+14 0.3583 51Cr 2.1479E+16 2.1483E+16 0.0193 55Fe 1.5670E+17 1.5722E+17 0.3310 59Fe 7.4995E+14 7.5010E+14 0.0203 54Mn 8.0167E+15 8.0314E+15 0.1831 48Sc 2.4734E+06 2.4843E+06 0.4421 65Zn 3.9769E+04 4.0424E+04 1.6461 RPV 51Cr 6.5003E+11 6.5016E+11 0.0201 55Fe 2.3698E+15 2.3776E+15 0.3272 54Mn 2.7002E+14 2.7051E+14 0.1818
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表 3 四个目标栅元的停堆剂量率
Table 3. Shutdown dose rate of the four target grid cells
structure dose rate/(Sv/hr) relative error/% −3σ/% 3σ/% ALARA DEPTH core baffle 6.1679E+00 6.2696E+00 − 1.6489 − 3.6347 3.6347 core barrel 3.9238E+00 3.8921E+00 0.8079 − 3.3881 3.3881 thermal shield 1.1709E+00 1.1652E+00 0.4868 − 3.3986 3.3986 RPV 1.6995E-01 1.7454E-01 − 2.7008 − 4.9815 4.9815
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