高温气冷堆不确定性分析的新进展
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摘要
球床高温气冷堆由于采用流动球床堆芯和燃料多次通过的运行方式,不能直接套用轻水堆中一般采用的"系统分解,逐级传递"的分析思路,其不确定性的传播和分析具有特殊性。清华大学核能与新能源技术研究院基于高温气冷堆的设计分析经验,开展了高温堆的不确定性研究,并取得了一些进展。目前高温气冷堆已建立起完整的不确定性分析计算框架。在此框架内,基于VSOP程序,开发能反映球床高温气冷堆实际运行特点的不确定性分析程序VSOP-UAM,实现了核数据不确定性隐式效应和显式效应的完整分析。然后使用SCALE/TSUNAMI-3D和VSOP-UAM程序,建立燃料球、堆芯单元、初装堆芯和平衡堆芯的分析模型,量化了核数据的不确定性对各种模型关键参数的影响。此外,还量化了球流混流效应、燃料富集度、燃料孔隙率这些球床堆芯参数的不确定性对堆芯有效增殖因数k_(eff)和功率分布的影响。从计算结果可看出,高温气冷堆的不确定性分析显示出了有别于传统轻水堆的结果。
Considering the strong coupling effect due to the continuous flow of fuel pebbles in the core, the uncertainty analysis framework and tool of the light water reactor are not immedieatly applicable to the pebble bed high-temperature gas-cooled reactor(PB-HTR), and it requires special treatment due to the characteristics related to the pebble bed. Based on design and engineering experience of PB-HTR, Institute of Nuclear and New Energy Technology(INET) of Tsinghua University performed in-depth study on uncertainty analysis of HTR and has achieved some progresses in recent years. The framework of uncertainty analysis for PB-HTR has been established and the uncertainty analysis tool VSOP-UAM has been developed based on the HTR reactor physics code VSOP. VSOP-UAM has the capability of describing the realistic operation characteristics of the pebble bed core, and conducting the complete analysis of the implicit effect and the explicit effect of nuclear data uncertainty. SCALE/TSUNAMI-3 D and VSOP-UAM codes were used to model the fuel pebble, the core unit, the initial state and the equilbrium state of the core, and the propagations of the uncertainty of the nuclear data to the system parameters were quantified. Additionally, the effects on effective multiplication factor(k_(eff)) and the power distribution of the core from the mixture of pebble flow, the uncertainty of filling fraction of the pebble bed and the uranium loading in fuel pebble, were also quantified and assessed. Some valuable observations were made through the analysis, which are different from the light water reactor results.
Considering the strong coupling effect due to the continuous flow of fuel pebbles in the core, the uncertainty analysis framework and tool of the light water reactor are not immedieatly applicable to the pebble bed high-temperature gas-cooled reactor(PB-HTR), and it requires special treatment due to the characteristics related to the pebble bed. Based on design and engineering experience of PB-HTR, Institute of Nuclear and New Energy Technology(INET) of Tsinghua University performed in-depth study on uncertainty analysis of HTR and has achieved some progresses in recent years. The framework of uncertainty analysis for PB-HTR has been established and the uncertainty analysis tool VSOP-UAM has been developed based on the HTR reactor physics code VSOP. VSOP-UAM has the capability of describing the realistic operation characteristics of the pebble bed core, and conducting the complete analysis of the implicit effect and the explicit effect of nuclear data uncertainty. SCALE/TSUNAMI-3 D and VSOP-UAM codes were used to model the fuel pebble, the core unit, the initial state and the equilbrium state of the core, and the propagations of the uncertainty of the nuclear data to the system parameters were quantified. Additionally, the effects on effective multiplication factor(k_(eff)) and the power distribution of the core from the mixture of pebble flow, the uncertainty of filling fraction of the pebble bed and the uranium loading in fuel pebble, were also quantified and assessed. Some valuable observations were made through the analysis, which are different from the light water reactor results.
引文
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[3] BRATTON R N,AVRAMOVA M,IVANOV K.OECD/NEA benchmark for uncertainty analysis in modeling (UAM) for LWRs:Summary and discussion of neutronics cases (phase Ⅰ)[J].Nuclear Engineering and Technology,2014,46(3):313-342.
[4] REITSMA F,STRYDOM G,TYOBEKA B,et al.The IAEA coordinated research program on HTGR reactor physics,thermal-hydraulics and depletion uncertainty analysis:Description of the benchmark test cases and phases[C]//Proceedings of HTR.Tokyo,Japan:[s.n.],2012.
[5] IVANOV K,AVRAMOVA M,ROYER E.Overview and status of the OECD/NEA LWR-UAM benchmark activities after the UAM-10 workshop[C]//WPRS Workshops 2017.Erlangen,Germany:[s.n.],2017.
[6] HOU J,BLYTH T,PORTER N.Benchmark for uncertainty analysis in modeling (UAM) for design,operation and safety analysis of LWR,volume Ⅱ:Specification and support data for the core cases (phase Ⅱ)[R].France:OECD/NEA,2017.
[7] HOU J,AVRAMOVA M,IVANOV K.Benchmark for uncertainty analysis in modeling (UAM) for design,operation and safety analysis of LWR,volume Ⅲ:Specification and support data for the system cases (phase Ⅲ)[R].France:OECD/NEA,2017.
[8] HAO C,LI F,ZHANG H.The challenges on uncertainty analysis for pebble-bed HTGR[C]//Proceeding of the International Topical Meeting on Advances in Reactor Physics (PHYSOR).Knoxville,Tennessee,USA:[s.n.],2012.
[9] RüTTEN H J,HAAS K A,BROCKMANN H,et al.V.S.O.P.(99/05) computer code system for reactor physics and fuel cycle simulation,Jül-4189[R].Germany:Forschungszentrum Jülich GmbH,2005.
[10] CHERNIKOV A,LYUTIKOV R,KURBAKOV S,et al.Behavior of HTGR coated fuel particles in high-temperature tests[J].Energy,1991,16(1-2):295-308.
[11] SCALE:A comprehensive modeling and simulation suite for nuclear safety analysis and design,ONRL/TM-2005/39[R].USA:Radiations Safety Information Computational Center,Oak Ridge National Laboratory,2011.
[12] WANG L,GUO J,LI F,el al.Direct evaluation of nuclear data uncertainty propagation in pebble-bed HTR core[C]//Proceedings of PHYSOR.USA:[s.n.],2016.
[13] KODELI I,SARTORI E.Neutron cross-section covariance data in multigroup form and procedure for interpolation to users’ group structures for uncertainty analysis applications[C]//Proceedings of PHYSOR.France:[s.n.],1990.
[14] WANG L,GUO J,LI F,et al.Effect of double heterogeneity treatment on neutronics modeling of HTGR unit cell[C]//Proceedings of HTR.USA:[s.n.],2016.
[15] 王黎东,郭炯,李富.核数据对球床式高温气冷堆燃料元件中子学特性的影响[J].原子能科学技术,2017,51(9):1 591-1 598.WANG Lidong,GUO Jiong,LI Fu.Effect of nuclear data on fuel element neutronic characteristics of pebble-bed high temperature gas-cooled reactor[J].Atomic Energy Science and Technology,2017,51(9):1 591-1 598(in Chinese).
[16] WANG L,GUO J,LI F.Difference of graphite capture cross sections in ENDF/B libraries[C]//Proceedings of ICONE.Japan:[s.n.],2015.
[17] HAO C,LI F,GUO J,et al.Quantitative analysis of uncertainty from pebble flow in HTR[J].Nuclear Engineering and Design,2015,295:338-345.
[18] HAO C,LI F,GUO J,et al.Uncertainty and sensitivity analysis of filling fraction of pebble bed in pebble bed HTR[J].Nuclear Engineering and Design,2015,292:123-132.
[19] KADAK A C,BAZANT M Z.Pebble flow experiments for pebble bed reactors[C]//Proceedings of ICONE HTR2004.Beijing,China:[s.n.],2004.