Estimation of the Dose Rate of Spent Fuel-Related Components of Lingao Nuclear Power Plant Using ORIGEN2 and MCNP5
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摘要
The components from spent fuel are discharged from the core and then stored in the framework of the spent fuel pool for cooling. However, it is of great significance to save the storage space of the spent fuel pool by shrinkage and shearing work to increase the spare fuel lattice number. In order to solve the problem of shrinkage and shearing work of spent fuel involving the problem of radioactive safety, the radioactive source item is calculated by ORIGEN2 program base on Unit 1 Ⅱ of Lingao Nuclear Power Plant(NPP), and the radiation dose rate of the related component shrinkage operation scene is simulated by the MCNP5 program. In addition, the effectiveness of shielding measures is discussed, and the maximum dose rate is within 0.35 μSv/h at the distance of 2.5 m from component center, and the maximum dose rate is almost 0 at the distance of 3.2 m from the component center. The intensity of the radiation dose produced by the related components is very low and can be neglected, which belong to the green area of NPP. The program calculation system from source term calculation to shielding calculation is established, and an engineering example is referenced, and its application and analysis are carried out. It provides a basis for radioactive safety analysis and evaluation for the shrinkage operation of spent fuel and makes the shrinkage technology of fuel-related components safer and more reliable.
The components from spent fuel are discharged from the core and then stored in the framework of the spent fuel pool for cooling. However, it is of great significance to save the storage space of the spent fuel pool by shrinkage and shearing work to increase the spare fuel lattice number. In order to solve the problem of shrinkage and shearing work of spent fuel involving the problem of radioactive safety, the radioactive source item is calculated by ORIGEN2 program base on Unit 1 Ⅱ of Lingao Nuclear Power Plant(NPP), and the radiation dose rate of the related component shrinkage operation scene is simulated by the MCNP5 program. In addition, the effectiveness of shielding measures is discussed, and the maximum dose rate is within 0.35 μSv/h at the distance of 2.5 m from component center, and the maximum dose rate is almost 0 at the distance of 3.2 m from the component center. The intensity of the radiation dose produced by the related components is very low and can be neglected, which belong to the green area of NPP. The program calculation system from source term calculation to shielding calculation is established, and an engineering example is referenced, and its application and analysis are carried out. It provides a basis for radioactive safety analysis and evaluation for the shrinkage operation of spent fuel and makes the shrinkage technology of fuel-related components safer and more reliable.
The components from spent fuel are discharged from the core and then stored in the framework of the spent fuel pool for cooling. However, it is of great significance to save the storage space of the spent fuel pool by shrinkage and shearing work to increase the spare fuel lattice number. In order to solve the problem of shrinkage and shearing work of spent fuel involving the problem of radioactive safety, the radioactive source item is calculated by ORIGEN2 program base on Unit 1 Ⅱ of Lingao Nuclear Power Plant(NPP), and the radiation dose rate of the related component shrinkage operation scene is simulated by the MCNP5 program. In addition, the effectiveness of shielding measures is discussed, and the maximum dose rate is within 0.35 μSv/h at the distance of 2.5 m from component center, and the maximum dose rate is almost 0 at the distance of 3.2 m from the component center. The intensity of the radiation dose produced by the related components is very low and can be neglected, which belong to the green area of NPP. The program calculation system from source term calculation to shielding calculation is established, and an engineering example is referenced, and its application and analysis are carried out. It provides a basis for radioactive safety analysis and evaluation for the shrinkage operation of spent fuel and makes the shrinkage technology of fuel-related components safer and more reliable.
引文
[1]Hu Y, Chen Y X, Yang S H, et al. Calculation and analysis of internal heat source for PWR related components based on Monte Carlo method[J]. Nuclear Science and Engineer-ing, 2013, 33(2):168-174(Ch).
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[9]Sayyed M I, Issa S A M, Buyukyildiz M, et al. Determination of nuclear radiation shielding properties of some tellurite glasses using MCNP5 code[J]. Radiation Physics and Chemistry, 2018, 150:1-8.
[10]Karavasilis E, Dimitriadis A, Gonis H, et al. Effective dosein percutaneous transhepatic biliary drainage examination using PCXMC2.0 and MCNP5 Monte Carlo codes[J]. Physica Medica, 2014, 30(4):432-436.
[11]Dong M G, Sayyed M I, Lakshminarayana G, et al. Investigation of gamma radiation shielding properties of lithium zinc bismuth borate glasses using XCOM program and MCNP5 code[J]. Journal of Non-Crystalline Solids, 2017,468:12-16.
[12]Liem P H, Sembiring T M. Development of new ORIGEN2data library sets for research reactors with light water cooled oxide and silicide LEU(20 w/o)fuels based on JENDL-3.3nuclear data[J]. Nuclear Engineering and Design. 2013, 262:52-62.
[13]Jia Z Y. Calculation and Analysis of Source Terms[D]. Beijing:China Institute of Atomic Energy, 2004(Ch).
[14]Su Z, Zou S L, Yu T, et al. Calculation and analysis of the source term of PWR spent fuel[J]. Journal of University of South China(Natural Science Edition), 2011, 25(4):9-12(Ch).
[15]Su Y Y. Shielding Calculation and Method Evaluation of Commonly Used Gamma Radiation Sources[D]. Changchun:Jilin University,2010(Ch).
[16]Li G F. MC Study and Safety Analysis of Spatial Dose Rate Distribution for Spent Fuel Pools in Nuclear Power Stations[D]. Sichuan:Chengdu University of Technology, 2012(Ch).
[17]Snoj L, Kavcic A, Zerovnik G, et al. Calculation of kinetic parameters for mixed TRIGA cores with Monte Carlo[J].Annals of Nuclear Energy, 2010, 37(2):223-229.
[2]Makmal T, Aviv O, Gilad E. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements[J]. Nuclear Instruments and Methods in Physics Research A, 2016, 834:175-182.
[3]Hocine B, Tahar Z,Mohamed B. Lumped pseudo fission products during burnup step in MCNP5-ORIGEN coupling system[J]. Progress in Nuclear Energy, 2016, 88:277-284.
[4]Foudil Z, Mohamed B, Tahar Z. Estimating of core inventory, source term and doses results for the NUR research reactor under a hypothetical severe accident[J]. Progress in Nuclear Energy, 2017, 100:365-372.
[5]Khurram M, Cao X R, Raheel A, et al. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents[J].Nuclear Engineering and Design, 2013, 262:435-451.
[6]Abrefah R G, Birikorang S A, Nyarko B J B, et al. Design of serpentine cask for Ghana research reactor-1 spent nuclear fuel[J]. Progress in Nuclear Energy, 2014, 77:84-91.
[7]Shiba T, Maeda S, Sagara H, et al. A simple method to create gamma-ray-source spectrum for passive gamma technique[J]. Energy Procedia, 2017, 131:250-257.
[8]Liu K, Wu H C, Cao L Z, et al. A code development for LLFP transmutation analysis based on the whole pin-wise calculation in PWRs[J]. Nuclear Engineering and Design,2013,256:56-66.
[9]Sayyed M I, Issa S A M, Buyukyildiz M, et al. Determination of nuclear radiation shielding properties of some tellurite glasses using MCNP5 code[J]. Radiation Physics and Chemistry, 2018, 150:1-8.
[10]Karavasilis E, Dimitriadis A, Gonis H, et al. Effective dosein percutaneous transhepatic biliary drainage examination using PCXMC2.0 and MCNP5 Monte Carlo codes[J]. Physica Medica, 2014, 30(4):432-436.
[11]Dong M G, Sayyed M I, Lakshminarayana G, et al. Investigation of gamma radiation shielding properties of lithium zinc bismuth borate glasses using XCOM program and MCNP5 code[J]. Journal of Non-Crystalline Solids, 2017,468:12-16.
[12]Liem P H, Sembiring T M. Development of new ORIGEN2data library sets for research reactors with light water cooled oxide and silicide LEU(20 w/o)fuels based on JENDL-3.3nuclear data[J]. Nuclear Engineering and Design. 2013, 262:52-62.
[13]Jia Z Y. Calculation and Analysis of Source Terms[D]. Beijing:China Institute of Atomic Energy, 2004(Ch).
[14]Su Z, Zou S L, Yu T, et al. Calculation and analysis of the source term of PWR spent fuel[J]. Journal of University of South China(Natural Science Edition), 2011, 25(4):9-12(Ch).
[15]Su Y Y. Shielding Calculation and Method Evaluation of Commonly Used Gamma Radiation Sources[D]. Changchun:Jilin University,2010(Ch).
[16]Li G F. MC Study and Safety Analysis of Spatial Dose Rate Distribution for Spent Fuel Pools in Nuclear Power Stations[D]. Sichuan:Chengdu University of Technology, 2012(Ch).
[17]Snoj L, Kavcic A, Zerovnik G, et al. Calculation of kinetic parameters for mixed TRIGA cores with Monte Carlo[J].Annals of Nuclear Energy, 2010, 37(2):223-229.