Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals
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Xie Linjun
Abstract
The friction force between the contact surfaces of a reactor internal hold-down spring (HDS) and core barrel flanges can directly influence the axial stiffness of an HDS. However, friction coefficient cannot be obtained through theoretical analysis. This study performs a mathematical deduction of the physical model of an HDS. Moreover, a mathematical model of axial load P, displacement δ, and flexibility coefficient is established, and a set of test apparatuses is designed to simulate the preloading process of the HDS. According to the experimental research and theoretical analysis, P-δ curves and the flexibility coefficient λ are obtained in the loading processes of the HDS. The friction coefficient f of the M1000 HDS is further calculated as 0.224. The displacement limit load value (4,638 kN) can be obtained through a displacement limit experiment. With the friction coefficient considered, the theoretical load is 4,271 kN, which is relatively close to the experimental result. Thus, the friction coefficient exerts an influence on the displacement limit load P. The friction coefficient should be considered in the design analysis for HDS.
Kurzfassung
Die Reibungskraft zwischen den Kontaktoberflächen von Niederhaltefedern (HDS) und Kernrohrflanschen kann die axiale Steifigkeit von HDS direkt beeinflussen. Dieser Beitrag stellt eine mathematische Deduktion des physikalischen Modells von HDS vor. Ein mathematisches Modell der Axialbelastung P, der Verschiebung δ und des Flexibilitätskoeffizienten wird aufgestellt und eine Reihe von Testapparaten entwickelt um den Vorspannungsprozess von HDS zu simulieren. Auf der Basis der experimentellen Ergebnisse und der theoretischen Analyse erhält man P-δ Kurven und den Flexibilitätskoeffizienten λ. Der Wert des Reibungskoeffizienten f der M1000 HDS errechnet sich zu 0.224. Der Wert der Verschiebungsbelastungsgrenze (4,638 kN) kann experimentell bestimmt werden. Mit dem betrachteten Reibungskoeffizient ist die theoretische Ladung 4,271 kN, was relativ nahe bei den experimentellen Ergebnissen liegt. Somit übt der Reibungskoeffizient einen Einfluss auf die Verschiebungsbelastungsgrenze P aus und sollte deshalb bei der Auslegung berücksichtigt werden.
References
1 ZhangZ.; XueG.: Nuclear Techniques36 (2013) 608–613Search in Google Scholar
2 FangJian, DuanYuangang, RanXiaobing, DaiChangnian: Nuclear Power Engineering35 (2014) 42–45Search in Google Scholar
3 SigristJ. F.; BrocD.; LaineC.: Nucl. Eng. Des.23 (2006) 2431–244310.1016/j.nucengdes.2006.03.001Search in Google Scholar
4 Min-TsungKaoa, Chung-YunWua, Ching-ChangChien, YibanXub, KunYuanb, MiloradDzodzob, MichaelConnerb, StevenBeltzb, SumitRayb, TeresaBissettb: Nucl. Eng. Des10 (2011) 4181–419310.1016/j.nucengdes.2011.08.007Search in Google Scholar
5 Ehrnstén, U.; PakarinenJ.; KarlsenW.; KeinänenH.: Eng Fail Anal.33 (2013) 55–6510.1016/j.engfailanal.2013.04.021Search in Google Scholar
6 ChoiY.; LimS.; KoB.-H.; ParkK.-S.; ParkN.-C.; ParkaY.-P.; JeongK.-H.; ParkJ.-S.: Nucl. Eng. Des.255 (2013) 202–21110.1016/j.nucengdes.2012.10.010Search in Google Scholar
7 FangJian, DuanYuangang, RanXiaobing, DaiChangnian: Nuclear Power Engineering35 (2014) 42–45Search in Google Scholar
8 RCC-M Edition 2007: Design and Construction Rules for Mechanical Components of PWR Nuclear Islands [S]. 2007Search in Google Scholar
9 XiaX., XiongL.; SunK.; YuZ. P.: International Journal of Automotive Technology, 17 (2016) 991–100210.1007/s12239-016-0097-7Search in Google Scholar
10 ASME-boiler and pressure vessel Code[S]. Section III, Division 1, Appendix III, 1989Search in Google Scholar
11 HuaL.; DengJ.; QianD.; LongH.: International Journal of Machine Tools & Manufacture110 (2016) 66–7910.1016/j.ijmachtools.2016.09.003Search in Google Scholar
12 YoungW. C.; BudynasR. G.: California Medicine68 (2002) 366–374Search in Google Scholar
13 ZhouY.; CaiZ. B., PengJ. F.: Applied Surface Science388 (2016) 40–4810.1016/j.apsusc.2016.04.174Search in Google Scholar
14 XueG.; ZhangM.; XieL.: The Design of Reactor Internals Hold-Down Spring [C], 2016Search in Google Scholar
15 LiY.: Advanced Materials Research512–515 (2012) 1957–1960Search in Google Scholar
16 UryukovB. A.; EvdokimenkoY. I.; Kisel', V. M.; TkachenkoG. V.: Journal of Engineering Physics and Thermophysics77 (2004) 559–56410.1023/B:JOEP.0000036502.40275.e0Search in Google Scholar
17 ASME Boiler and Pressure Vessel Code [S]. Section III, Division 1, Appendix II, 2004Search in Google Scholar
18 ASME Boiler and Pressure Vessel Code, Section II, Part D, 776–778 (2007)Search in Google Scholar
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- Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals
- Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Modelling Human Resource Requirements for the Nuclear Industry in Europe
- Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis
- The integrity of NSSS and containment during extended station blackout for Kuosheng BWR plant
- Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR
- Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5
- The application of knowledge management and TRIZ for solving the safe shutdown capability of fire alarms in nuclear power plants
- Dose assessment for emergency workers in early phase of Fukushima Daiichi nuclear power plant accident
- Anti-neutrino flux in a research reactor for non-proliferation application
- Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals
- Robust observer based control for axial offset in pressurized-water nuclear reactors based on the multipoint reactor model using Lyapunov approach
- Internal and external hazards inside the containment in case of an emergency situation
- Slab albedo for linearly and quadratically anisotropic scattering kernel with modified FN method
