Experimental study on in-vessel debris coolability in ALPHA program

Nuclear Engineering and Design

Volumn 187, Issue 2, 1999, Pages 241-254

  Maruyama, Yu ^a   Yamano, Norihiro ^a   Moriyama, Kiyofumi ^a   Sun Park, Hyun ^a   Kudo, Tamotsu ^a   Yang, Yanhua ^a   Sugimoto, Jun ^a  

^a Severe Accident Research Laboratory   (Japan)

Author keywords

[No Author keywords available]

Indexed keywords

ALUMINA; COOLING; HEAT FLUX; HEAT RESISTANCE; INTERFACES (MATERIALS); POROUS MATERIALS;

IN-VESSEL DEBRIS COOLABILITY; THERMITE REACTION;

NUCLEAR ENGINEERING;

EID: 0033077885     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/S0029-5493(98)00278-7     Document Type: Article

References (36)

1. Akers D.W., Schuets B.K. Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head. Nucl. Saf. 35(2):1994;288-300.
2. Asmolov, V., et al., 1998. Experimental Study of Heat Transfer in the Slotted Channels at ClE Facility. Proceedings OECD/NEA/CSNI Workshop on In-Vessel Core Debris Retention and Coolability, Garching, Germany, March 3-6.
3. Blose, R.E., et al., 1993. Core-Concrete Interactions with Overlying Water Pools, NUREG/CR-5907 SAND92-1563, November.
4. Blomquist R.A. et al. Viscosity of molten alumina. Am. Ceram. Soc. Bull. 57(5):1986;222.
5. Bradley, D.R., et al., 1993. CORCON-Mod3: An integrated Computer Models for Analysis of Molten Core-Concrete Interactions, NUREG/CR-5843 SAND92-0167, October.
6. Broughton J.M. et al. A Scenario of the Three Mile Island Unit 2 Accident. Nucl. Technol. 87(1):1989;34-53.
7. Cheung, F.B., et al., 1997. Critical Heat Flux (CHF) Phenomenon on a Downward Facing Curved Surface, NUREG/CR-6504 PSU/ME-97-7321, June.
8. Chu T.Y. et al. Ex-Vessel Boiling Experiments: Laboratory- and Reactor-Scale Testing of the Flooded Cavity Concept for In-Vessel Core Retention Part I: Observation of Quenching of Downward-Facing Surfaces. Nucl. Eng. Des. 169:1997;77-88.
9. Chu T.Y. et al. Ex-Vessel Boiling Experiments: Laboratory- and Reactor-Scale Testing of the Flooded Cavity Concept for In-Vessel Core Retention: Part II: Reactor-Scale Boiling Experiments of the Flooded Cavity Concept for In-Vessel Core Retention. Nucl. Eng. Des. 169:1997;89-99.
10. Diercks D.R., Korth G.E. Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head. Nucl. Saf. 35(2):1994;301-312.
11. Dinh T.N., Nourgaliev R.R. Turbulence Modelling for Large Volumetrically Heated Liquid Pools. Nucl. Eng. Des. 169:1997;59-76.
12. 2-Cladding Composite Body with Simultaneous Solidification and Melting. Nucl. Sci. Eng. 51:1973;84-87.
13. Fujita Y. et al. Nucleate Boiling Heat Transfer and Critical Heat Flux in Narrow Space between Rectangular Surfaces. Int. J. Heat Mass Transf. 31(2):1988;229-239.
14. Gitzen, W.H., (Ed.), 1970. Alumina As a Ceramic Material, The American Ceramic Society.
15. Henry R.E., Dube D.A. Water in the RPV: A Mechanism for Cooling Debris in the RPV Lower Head, Proceedings Specialist Meeting on Selected Containment Severe Accident Management Strategies, Stockholm, Sweden. June. 13-15:1994;93-104.
16. Henry R.E. A Scaling and Experimental Approach for Investigating In-Vessel Cooling, Proceedings. Twenty-Fourth Water Reactor Safety Information Meeting, 2, Bethesda, MD. October. 21-23:1996;141-159.
17. Hohorst, J.K., 1990. SCDAP/RELAP5/MOD2 Code Manual, Volume 4: MATPRO; A Library of Material Properties for Light-Water-Reactor Accident Analysis, NUREG/CR-5273 EGG2555, February.
18. Jeong, J.H., et al., 1998. Experimental Study on CHF in a Hemispherical Narrow Gap, Proceedings. OECD/NEA/CSNI Workshop on In-Vessel Core Debris Retention and Coolability, Garching, Germany, March 3-6.
19. Kohler, W., et al., 1998. Experiments on Heat Removal in a Gap between Debris Crust and RPV Wall, Proc. OECD/NEA/CSNI Workshop on In-Vessel Core Debris Retention and Coolability, Garching, Germany, March 3-6.
20. Kymäläinen O. et al. Heat Flux Distribution from a Volumetrically Heated Pool with High Rayleigh Number. Nucl. Eng. Des. 149:1994;401-408.
21. Kymäläinen O. et al. In-Vessel Retention of Corium at the Loviisa Plant. Nucl. Eng. Des. 169:1997;109-130.
22. Mishima K., Nishihara H. The Effect of Flow Direction and Magnitude on CHF for Low Pressure Water in Thin Rectangular Channels. Nucl. Eng. Des. 86:1985;165-181.
23. Maruyama, Y., et al., 1993. Evaluation of Specific Heat and Temperature of Thermite Melt Used in Melt Coolant Interaction Test of ALPHA Program, JAERI-M 93-096 (in Japanese), March.
24. Maruyama, Y., et al., 1996. In-Vessel Debris Coolability Experiments in ALPHA Program, Proceedings International Topical Meeting on Probabilistic Safety Assessment, vol. III, Park City, UT, September 29-October 3, pp. 1367-1374.
25. Maruyama Y. et al. Studies on In-Vessel Debris Coolability in ALPHA Program, Proceedings Twenty-Fourth Water Reactor Safety Information Meeting, 2, Bethesda, MD. October. 21-23:1996;161-172.
26. Rempe J. et al. Margin-to-Failure Calculations for the TMI-2 Vessel. Nucl. Saf. 35(2):1994;313-327.
27. Rempe, J.L., et al., 1994. Investigation of the Coolability of a continuous mass of Relocated Debris to a Water-Filled Lower Plenum, EGG-RAAM-11445, September.
28. Rubin A.M., Beckjord E. Three Mile Island-New Findings 15 Years After the Accident. Nucl. Saf. 35(2):1994;256-268.
29. Stickler, L.A., et al., 1994. Calculations to Estimate the Margin to Failure in the TMI-2 Vessel, NUREG/CR-6196, February.
30. Sudo Y. et al. Experimental Study of Differences in DNB Heat Flux between Upflow and Downflow in Vertical Rectangular Channel. J. Nucl. Sci. Technol. 22(8):1985;604-618.
31. Suh K.Y. et al. The SONATA-IV Project In Pursuit of In-Vessel Defense-in-Depth Against a Severe Accident, Proceedings. 11th KAIF/KNS Annual Conf., Seoul, Korea. April. 11-12:1996;841-853.
32. Theofanous T.G. et al. In-Vessel Coolability and Retention of a Core Melt. Nucl. Eng. Des. 169:1997;1-48.
33. Theofanous T.G. et al. The First Results from the ACOPO Experiment. Nucl. Eng. Des. 169:1997;49-57.
34. Theofanous T.G., Syri S. The Coolability Limits of a Reactor Pressure Vessel Lower Head. Nucl. Eng. Des. 169:1997;59-76.
35. Turland, B.D., Morgan, J., (Ed.), 1985. Compendium of Post Accident Heat Removal Models for Liquid Metal Cooled Fast Breeder Reactors, European Applied Research Report, Vol. 6, No. 5, April.
36. Wolf, J.R., et al., 1994. TMI-2 Vessel Investigation Project Integration Report, NUREG/CR-6197 TMI V(93)EG10 EGG-2734, March.