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atw Vol. 63 (2018) | Issue 1 ı January FUEL 42 underlines that further tests on MOX fuel would improve knowledge about its sensibility as regard to fuel melting, especially for high burn-up and high plutonium levels characteristics of MOX fuel loaded in EDF’s reactors. 4 Taking into account rod failures and assembly bow 4.1 Impact of fuel assembly bowing on safety demonstration Fuel assemblies distort in-core and the gap between fuel assemblies can vary away from the design value. Using several ex-core fuel assembly bow measurements (from different reactors and cycles), EDF has developed a mechanical model to estimate interassembly gap size distributions in cores. IRSN assessed assumptions and considered that the predictive model is consistent with the current state- ofthe-art. In addition to slower drop times of RCCA due to friction in guide tubes, fuel assembly distortion potentially leads to neutronic, thermohydraulic and mechanical impacts on safety demonstration: • the presence of larger inter-assembly gaps causes a local variation in the fuel-to-moderator ratio and hence the local neutron moderation. Basically, as fuel assemblies move apart, the concentration of thermal neutrons in the gap increases and so does the power in peripheral pins. EDF has developed a new methodology for quantifying and taking into account this effect in the safety demonstration. IRSN estimates this methodology satisfactory; • for the same reasons, the Critical Heat Flux Ratio (CHFR) decreases at periphery, but also the hydraulic diameter of the corresponding flow channel increases, so does the CHFR. Because of these antagonistic effects, the flow channel in which the minimum CHFR is reached could become a peripheral one (instead of a channel within the fuel assembly in the current safety demonstration). In such flow channel, grid straps do not have mixing vanes, which significantly reduces CHFR. For EDF, the minimum CHFR remains in the center of the fuel assembly. However, to evaluate the global effect, EDF realized sensitivity studies notably using unappropriate CHF correlations. Because of the large number of justifications still to be provided, IRSN can’t conclude on EDF evaluation; • the presence of smaller water gaps, and particularly the existence of contact between grids, is likely to increase the maximum impact forces on fuel assembly grids under seismic and LOCA loads. EDF not yet assessed the effect of variable inter-assembly gaps, repre sentative of the in-reactor situation, on the assembly grids buckling risk. In addition, for IRSN, the validation of EDF’s model to calculate the impact force on grids during accidentel conditions needs to be completed, particularly because it doesn’t include a comparison with sufficiently representative tests results. Thus, the safety demonstration will be updated. 4.2 Leaking fuel rods during normal operating conditions The behavior of defective fuel rods, especially under REA, is an important aspect of safe reactor operation, since some EDF’s reactors (7 out of the 58 operating reactors currently) contain a very small percentage of leaking fuel rods (only 0,11% leaking fuel assemblies). This issue has been assessed for several years. The complexity of the physical phenomena to be taken into account and the lack of available experimental data on waterlogged fuel rods under this transient explain the difficulty to conclude on the potential unwanted effects: surrounding fuel rods failures due to mechanical and thermal effects or even potential vessel damage [23, 24]. IRSN considered that EDF’s demonstration takes into account satisfactorily the state-of-the-art. Finally, the large pressure pulse does not lead to additional fuel rods failures nor to vessel damage. However, for IRSN, EDF should still justify that the models used for assessing thermal interaction and its consequences are appropriate. Considering other PCC-2 and PCC-4 transients, IRSN estimates that it is likely that in many cases, application of stress would lead to the fuel rods failure in the secondary defect area and to fuel dispersal in the primary coolant. However, these phenomena are unlikely to affect the core coolability or to have any significant impact on the the radiological doses calculations, except for steam generator tube rupture accidents. Indeed, these transients are characterized by a break in the second barrier, containment bypass and the possibility that some contaminated reactor coolant will be released into the environment. EDF will study the potential consequences of this scenario. References 1. A.M. Garde et al., Hydrogen Pick-Up Fraction for ZIRLO Cladding Corrosion and Resulting Impact on the Cladding Integrity, Proceedings of Top Fuel 2009 Paris, France, September 6-10 (2009) 2. S. K. Yagnik, R-C Kuo, Y.R. Rashid et al., Effect of hydrides on the mechanical properties of Zircaloy-4, Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September (2004) 3. R.L. Yang, R.O. Montgomery, N. Waeckel, EPRI TR #1002865, Topical report on reactivity-initiated accident: bases for RIA fuel and core coolability criteria (2002) 4. T. Sugiyama, High burnup fuel behavior under high temperature RIA conditions, FSRM 2010, Tokai, Japan, May (2010) 5. B. Julien et al., Performance of advanced fuel product under PCI conditions, Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22 (2004) 6. P.E. Macdonald, W.J. Quapp et al., Response of unirratiated and irratiated PWR fuel rods tested under Powercooling mismatch conditions, Nuclear Safety, vol.19, n°4, (1978) 7. F. M. Haggag, Zircaloy-cladding embrittlement criteria : comparison of in-pile and out-of-pile results, NUREG/ CR-2757 (1982) 8. T. Fuketa, Transient response of LWR fuels (RIA), Compr. Nucl. Mater. 579-593 (2012) 9. W.G. Lussie, The response of mixed oxide fuel rods to power bursts, IN-ITR-114, Idaho Nuclear Corporation (1970) 10. W.G. Lussie, The response of UO2 fuel rods to power bursts, IN-ITR-112, Idaho Nuclear Corporation (1970) 11. M.D. Freshley, Behavior of discret plutonium dioxide particles in mixedoxide fuel during rapid power transient, Nuclear technology, Vol.15 (1972) 12. N. Waeckel, Fuel Assembly distortion in EDF NPPs, Oral communication on OECD WGFS, Paris (2014) 13. C. Durand, Fuel bowing performances, EDF oral communication at OECD NEA Workshop Advanced fuel modelling for safety and performance enhancement (2017) 14. Report OECD NEA/CSNI/R(2014)10, Leaking Fuel Impacts and Practices (2014) 15. Y. KIM, S. KIM, Kinetic studies on massive hydriding of commercial zirconium alloy tubing, Journal of nuclear materials, 270, pp. 147-153 (1999) Fuel Review of Fuel Safety Criteria in France ı Sandrine Boutin, Stephanie Graff, Aude Foucher-Taisne and Olivier Dubois
Kommunikation und Training für Kerntechnik Wissen wirksam weitergeben Seminar: Erfolgreicher Wissenstransfer in der Kerntechnik – Methoden und praktische Anwendung Ein erfolgreicher Wissenstransfer ist die kontinuierliche Grundlage für jeden Kompetenzerhalt, besonders in Phasen von Umstrukturierungen oder umfänglicherem Personalwechsel in einer Organisation. In diesem Praxisseminar werden Ihnen die Maßnahmen und Anwendungsmöglichkeiten für den erfolgreichen Wissenstransfer vermittelt. Seminarinhalte ı ı ı ı ı Grundlagen zum Wissenstransfer Häufig genutzte Methoden in der deutschen kerntechnischen Branche Erfolgskritisches Wissen identifizieren – Methoden im Überblick Chancen und Grenzen ausgewählter Wissenstransfermethoden Zusammenhang zwischen Unternehmenskultur und Wissenstransfer Zielgruppe ı ı ı Fach- und Führungskräfte Projektverantwortliche Mitarbeiterinnen und Mitarbeiter der kerntechnischen Branche in Deutschland Maximale Teilnehmerzahl 12 Personen Referentinnen Frau Dr. Christien Zedler | Managing Director des IAOP Frau Dr. Tanja-Vera Herking | Senior Consultant des IAOP Wir freuen uns auf Ihre Teilnahme! Bei Fragen zur Anmeldung rufen Sie uns bitte an oder senden uns eine E-Mail. Freie Plätze – jetzt anmelden! Termin 2 Tage 21. bis 22. März 2018 Tag 1: 10:00 bis 18:30 Uhr Tag 2: 09:00 bis 16:00 Uhr Veranstaltungsort Geschäftsstelle der INFORUM Robert-Koch-Platz 4 10115 Berlin In Kooperation mit dem IAOP, Institut für Arbeitspsychologie, Organisation und Prozessgestaltung Teilnahmegebühr 898,– € ı zzgl. 19 % USt. Im Preis inbegriffen sind: ı Seminarunterlagen ı Teilnahmebescheinigung ı Pausenverpflegung inkl. Mittagessen Kontakt INFORUM Verlags- und Verwaltungsgesellschaft mbH Robert-Koch-Platz 4 10115 Berlin Petra Dinter-Tumtzak Fon +49 30 498555-30 Fax +49 30 498555-18 seminare@kernenergie.de
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