Investigation of Pressure-Tube and Calandria-Tube Deformation Following a Single Channel Blockage Event in ACR-700
Author(s)Gerardi, Craig Douglas; Buongiorno, Jacopo
Advanced Nuclear Power Technology Program (Massachusetts Institute of Technology)
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The ACR-700 is an advanced pressure-tube (PT) reactor being developed by Atomic Energy of Canada Limited (AECL). As in conventional CANDU reactors, the PTs are horizontal. Each PT is surrounded by a calandria tube (CT), and the gap in between is filled with carbon dioxide gas. The space between the CTs is filled with the heavy-water moderator. One postulated accident scenario for ACR-700 is the complete coolant flow blockage of a single PT. The flow is not monitored within each individual PT, thus during the early stages of this accident the reactor remains at full power and full pressure, resulting in rapid coolant boil-off and fuel overheating. Melting of the Zircaloy (Zry) components of the fuel bundle (cladding, end plates and end caps) can occur, with relocation of some molten material to the bottom of the PT. The hot spot caused by the molten Zry/PT interaction may cause PT/CT failure due to localized plastic strains. Failure of the PT/CT results in depressurization of the primary system, which triggers a reactor scram, after which the decay heat is removed via reflooding, thus PT/CT rupture effectively terminates the accident. Clearly, prediction of the time scale and conditions under which PT/CT failure occurs is of great importance for this accident. We analyzed the following key phenomena occurring after the blockage: (a) Coolant boil-off (b) Cladding heat-up and melting (c) Dripping of molten Zircaloy (Zry) from the fuel pin (d) Thermal interaction between the molten Zry and the PT (e) Localized bulging of the PT (f) Interaction of the bulged PT with the CT Simple one-dimensional models were adequate to describe (a), (b) and (c), while the three-dimensional nature of (d), (e) and (f) required the use of more sophisticated models including a finite-element description of the thermal transients within the PT and the CT, implemented with the code COSMOSM. The main findings of the study are as follows: (1) Most coolant boils off within 3 s of accident initiation. (2) Depending on the magnitude of radiation heat transfer between adjacent fuel pins, the cladding of the hot fuel pin in the blocked PT reaches the melting point of Zry in 7 to 10 s after accident initiation. (3) Inception of melting of the UO2 fuel pellets is not expected for at least another 7 s after 2Zry melting. (4) Several effects could theoretically prevent molten Zry dripping from the fuel pins, including Zry/UO2 interaction and Zry oxidation. However, it was concluded that because of the very high heat-up rate typical of the flow blockage accident sequence, holdup of molten Zry would not occur. Experimental verification of this conclusion is recommended. (5) Once the molten Zry relocates to the bottom of the PT, a hot spot is created that causes the PT to bulge out radially under the effect of the reactor pressure. The PT may come in contact with the CT, which heats up, bulges and eventually fails. The inception and speed of the PT/CT bulging and ultimately the likelihood of failure depend strongly on the postulated mass of molten Zry in contact with the PT, and on the value of the thermal resistance at the Zry/PT interface. It was found that a Zry mass =/< 10 g will not cause PT/CT failure regardless of the contact resistance effect. On the other hand, a mass of 100 g would be sufficient to cause PT/CT failure even in the presence of a thick 0.2 mm oxide layer at the interface. The characteristic time scales for this 100-g case are as follows: PT bulging starts within 3 s of Zry/PT contact - PT makes contact with the CT in another 2 s - CT bulging starts in less than 1 s - CT failure occurs within another 5 s. Thus, the duration of the PT/CT deformation transient is 11 s, which gives a total duration of the accident (from PT blockage to PT/CT failure) of 18 to 21 s.
Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Advanced Nuclear Power Program