Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR

Pellet-Cladding Mechanical Interaction (PCMI) failure is the one of failure mode which must be evaluated in the nuclear fuel safety. PCMI is caused by the mechanical load to cladding due to the fuel pellet expansion. Under the high fuel burnup condition, the fuel cladding may become degraded by embrittlement under the neutron-induced irradiation and hydrogen accumulation due to the waterside corrosion. In order to consider the further deterioration of the material with higher burnup, the evaluation using mechanical indicators, e.g. strain and stress, might be required. In this study, a transient model is developed, which can mechanistically evaluate the PCMI behavior, in particular, for fuel rods under the higher burnup condition. The model is incorporated in a fuel behavior analysis code and verified by benchmarks with other similar codes. The PCMI predictability of this code is validated using the experimental test data.

Fueling Machine Model for Simulation of Spent Fuel Behavior in Postulated Fuel Handling Accidents

On-power fueling is an important feature of the CANDU® reactor. Fueling is a routine operation with a large number of channel fueling visits made each year with the fueling machines acting as the key system. Hence, safety analysis must consider fueling machine events typically when the fueling machine is in transit toward the spent fuel port. This paper presents a model of fueling machine containing spent fuel with complex configuration and multiprocess mechanisms. Using an integral approach with fuel and fueling machine, this model tends to improve previous modeling method, which only takes account of a slice of fuel or fueling machine. This fueling machine model is developed for simulations of the fueling machine coolant thermal hydraulics behavior, the spent fuel behavior, and potential fission product release during postulated loss of heat removal accidents. An example of its application is also presented in this paper.

Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

DIONISIO is a computer code designed to simulate the behavior of one nuclear fuel rod during its permanence within the reactor. Starting from the power history and the external conditions to which the rod is subjected, the code predicts all the meaningful variables of the system. Its application range has been recently extended to include accidental conditions, in particular the so-called loss of coolant accidents (LOCA). In order to make realistic predictions, the conditions in the rod environment have been taken into account since they represent the boundary conditions with which the differential equations describing the fuel phenomena are solved. Without going into the details of the thermal-hydraulic modeling, which is the task of the specific codes, a simplified description of the conditions in the cooling channel during a LOCA event has been developed and incorporated as a subroutine of DIONISIO. This has led to an improvement of the fuel behavior simulation, which is evidenced by the considerable number of comparisons with experiments carried out, many of them reported in this paper. Moreover, this work describes a model of high temperature capture and release of hydrogen in the nuclear fuel cladding, in scenarios typical of LOCA events. The corresponding computational model is being separately tested and will be next included in the DIONISIO thermal-hydraulic module.