Out-of-pile modelling of nuclear fuel elements for MTR type reactors – Part 1
-
K. Farhadi
Abstract
In the first part of the present paper, for a 5 MW thermal pool-type research reactor, the fuel element is modelled for when undergoing both natural circulation of the coolant and forced convection of the coolant operational conditions. First, the required dimensionless groups were identified and then the pertinent similarity criteria were derived accordingly. The derived similitude laws were modified under the conditions of identical pressure, identical temperature difference and identical coolant and fuel cladding in the model and the prototype. These modifications were done for the system under both natural and forced convections. The effect of varying cladding materials under normal operating conditions of the research reactor were observed via coolant channel thickness. Also the effect of a wider coolant channel on the nature of the coolant fluid was observed. The results obtained indicate that it is not possible to conserve all the dimensionless groups between the model and the prototype and hence achieve an errorless outcome. Among all the liquids available, methanol is the only liquid which nearly satisfies the thermal-hydraulic similitude and must be used in place of ordinary coolant water. This in turn necessitates the coolant channel to be wider and as a consequence the traditional Aluminium cladding in research reactors should be replaced by Iron. The derived scale down criteria can be used for the design of fuel element for the out-of-pile testing.
Kurzfassung
Im ersten Teil des vorliegenden Beitrags wird für einen 5 MW-Pool-Typ-Forschungsreaktor das Brennelement modelliert für zwei Betriebsbedingungen, Naturumlauf des Kühlmittels und erzwungene Konvektion des Kühlmittels. Zuerst wurden die erforderlichen dimensionslosen Gruppen identifiziert und dann entsprechend die passenden Ähnlichkeitskriterien abgeleitet. Die abgeleiteten Ähnlichkeitsgesetze wurden modifiziert für die Bedingungen identischen Drucks, identischer Temperaturdifferenzen and identischer Kühlmittel und Brennelementhülle im Modell und im Prototyp. Diese Modifikationen wurden im System sowohl für Naturumlauf wie auch für erzwungene Konvektion durchgeführt. Der Einfluss unterschiedlicher Hüllmaterialien wurde unter normalen Betriebsbedingungen des Forschungsreaktors beobachtet über die Kühlkanalwanddicke. Auch der Effekt eines weiteren Kühlkanals wurde beobachtet über die Art der Kühlflüssigkeit. Die erhaltenen Ergebnisse zeigen, dass es nicht möglich ist all die dimensionslosen Gruppen zwischen Modell und Prototyp zu erhalten und so zu einem fehlerlosen Ergebnis zu kommen. Zwischen den verfügbaren Flüssigkeiten ist Methanol die Einzige, die nahezu die thermohydraulische Ähnlichkeit befriedigt und muss deshalb anstelle von normalem Kühlwasser verwendet werden. Dies erfordert im Gegenzug einen weiteren Kühlkanal und konsequenterweise sollte die Hülle aus Aluminium durch Eisen ersetzt werden. Die abgeleiteten Kriterien können für die Auslegung von Brennelementen zur Überprüfung außerhalb des Reaktors verwendet werden.
References
1 Hamilton, D. C.; Lynch, F. E.; Palmer, L. D.: The nature of the flow of ordinary fluids in a thermal convection harp. Report ORNL-1624, March, 195410.2172/4371580Suche in Google Scholar
2 Madejski, J.; Mikielewicz, J.: A new device for heat transfer equipment. Int. J. Heat Mass Transfer14 (1971) 354–363 DOI: http://dx.doi.org/10.1016/0017-9310(71)90155-4Suche in Google Scholar
3 Japikse, D.: Advances in thermosyphon technology, in T. F.Irvine, Jr., and J. P.Harnett (eds.) Advances in Heat Transfer Vol. 9, 1973, Academic Press, New York, pp. 1–1110.1016/S0065-2717(08)70061-3Suche in Google Scholar
4 Zvirin, Y.: A review of natural circulation loops in pressurized water reactors and other systems, Nucl. Eng. Des.67 (1981a) 203–225 DOI: http://dx.doi.org/10.1016/0029-5493(82)90142-XSuche in Google Scholar
5 Metrol, A.; Greif, R.: A review of natural circulation loops. In: S.Kakac; W.Aung; R.Viskanta (eds.), Natural Convection: Fundamentals and Applications, Hemisphere, 1985, New York, pp. 1033–1071Suche in Google Scholar
6 Greif, R.: Natural circulation loops. J. Heat Transfer110 (1988) 1243–1258 DOI: http://dx.doi.org/10.1115/1.3250624Suche in Google Scholar
7 Nahavandi, A. N.; Castellana, F. S.; Moradkhanian, E. N.: Scaling laws for modelling nuclear reactor systems. Nuclear Science Engineering72 (1979) 75–8310.13182/NSE79-A19310Suche in Google Scholar
8 Zuber, N.: Problems in modelling of small break LOCA, Rep. NUREG-0724, October 198010.1615/ICHMT.1982.AdvCourHeatTransfNucReactSaf.20Suche in Google Scholar
9 Karwat, H.: Principal characteristics of experimental simulators suitable for SBLOCA events of LWRs and scaling principles adopted in their design. Proc. Spec. Meet. on Small Break LOCA Analyses in LWRs, Pisa, Italy, 23–27 June 1985, Vol. 1, pp. 399–433Suche in Google Scholar
10 Heisler, M. P.: Development of scaling requirements for natural convection liquid-metal fast breeder reactor shutdown heat removal facilities. Nuclear Science Engineering80 (1982) 347–35910.13182/NSE82-A19819Suche in Google Scholar
11 Ishii, M.; Kataoka, I.: Scaling laws for thermal-hydraulic systems under single-phase and two-phase natural circulation. Nuclear Engineering and Design81 (1984) 411–425 DOI: http://dx.doi.org/10.1016/0029-5493(84)90287-5Suche in Google Scholar
12 Vijayan, P. K.; Mehta, S. K.; Date, A. W.: On the steady state performance of natural circulation loops. Int. J. Heat Mass Transfer34 (1991) 2219–2230 DOI: http://dx.doi.org/10.1016/0017-9310(91)90048-JSuche in Google Scholar
13 Vijayan, P. K.; Nayak, A. K.; Pilkhwal, D. S.; Saha, D.; Venkat Raj, V.: Effect of loop diameter on the stability of single-phase natural circulation in rectangular loops. Proc. 5th Int. Topical Meet. on Reactor Thermalhydraulics, NURETH-5, Salt Lake City, UT, 21–24 September 1992, Vol. 1, pp. 261–267Suche in Google Scholar
14 Rohsenow, W. M.; Choi, H.: Heat, Mass and Momentum Transfer, Prentice-Hall, Inc., Englewood Cliffs, New Jersey (1961).Suche in Google Scholar
15 Farhadi, K., Bousbia-Salah, A.; D'Auria, F.: A model for the analysis of pump start-up transients in Tehran Research Reactor. Progress in Nuclear Energy49 (2007) 499–510 DOI: http://dx.doi.org/10.1016/j.pnucene.2007.07.006Suche in Google Scholar
16 Kocamustafaogullari, G.; Ishii, M.: Scaling of Two-Phase Flow Transient Using Reduced Pressure System and Simulated Fluid, Nuclear Engineering Design104 (1987) 121–132 DOI: http://dx.doi.org/10.1016/0029-5493(87)90293-7Suche in Google Scholar
© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of processes in RBMK-1500 fuel rods during the operation, short and intermediate term storage
- Can mechanical stresses noticeably influence the diffusion of hydrogen in zircaloy?
- Out-of-pile modelling of nuclear fuel elements for MTR type reactors – Part 1
- Experimental study of PHWR debris bed under boil-off condition
- Effectiveness of radial flow on rewetting of AHWR fuel cluster
- Radiotracers in performance evaluation of nuclear grade resins Amberlite IRN-78 and Purolite NRW-8000
- The production of 238–242Pu(n,γ)239–243Pu fissionable fluids in a fusion-fission hybrid reactor
- Theoretical study of deuteron induced reactions on 6,7Li, 9Be and 19F targets
- Upgrading of neutron radiography/tomography facility at research reactor
- Assessment of the radiological health damage costs of the Yeniköy and Kemerköy lignite-fired power plants in Muğla
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Analysis of processes in RBMK-1500 fuel rods during the operation, short and intermediate term storage
- Can mechanical stresses noticeably influence the diffusion of hydrogen in zircaloy?
- Out-of-pile modelling of nuclear fuel elements for MTR type reactors – Part 1
- Experimental study of PHWR debris bed under boil-off condition
- Effectiveness of radial flow on rewetting of AHWR fuel cluster
- Radiotracers in performance evaluation of nuclear grade resins Amberlite IRN-78 and Purolite NRW-8000
- The production of 238–242Pu(n,γ)239–243Pu fissionable fluids in a fusion-fission hybrid reactor
- Theoretical study of deuteron induced reactions on 6,7Li, 9Be and 19F targets
- Upgrading of neutron radiography/tomography facility at research reactor
- Assessment of the radiological health damage costs of the Yeniköy and Kemerköy lignite-fired power plants in Muğla