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New nuclear fuel designs need to be verified with respect to relevant performance and safety aspects, notably resistance to corrosion and resistance to pellet-clad mechanical interaction (PCMI) under normal operating conditions and in transients. Assessments should also cover conditions such as those prevailing in fuel handling and storage.
In order to achieve this goal, a fundamental understanding of the dominant cladding failure mechanisms is needed. This project addresses this need. The overriding objective of this project is to obtain an improved fundamental understanding of the dominant failure mechanisms for light water reactor (LWR) fuel cladding under pellet-clad mechanical interaction (PCMI) loading that can arise during normal operation or anticipated transients. The focus will be on the following failure mechanisms, which will be studied within the corresponding project tasks:
PCMI driving force;
pellet-clad interaction (PCI): stress corrosion cracking (SCC) initiated at the cladding inner surface under the combined effect of the mechanical loading and chemical environment caused by an increase in the fuel pellet temperature following a power increase;
hydride embrittlement: time-independent fracture of existing hydrides;
delayed hydride cracking (DHC): time-dependent crack initiation and propagation through fracture of hydrides that can form ahead of the crack tip.
For each of these failure mechanisms, the project objectives can be broken down into the following types of investigation:
quantifying the key parameters and their influences through separate effects studies;
identifying the key physical processes and phenomena that are involved in each failure mechanism and theoretical modelling of the fracture mechanisms;
developing and refining testing techniques for the experimental verification of the fuel behaviour with respect to these failure mechanisms.
The project will seek to sustain the existing information legacy and facilitate the transfer of knowledge by means of seminars and workshops. The project also has the following general objectives:
to improve the general understanding of cladding integrity at high burn-up;
to study both boiling water reactor and pressurised water reactor/VVER fuel cladding integrity;
to complement two large international projects (CABRI and ALPS), which focus on fuel behaviour in design basis accidents (RIA), where some of the mechanisms are similar to those that may occur during normal operational transients or anticipated transients;
to achieve results of general applicability (i.e. not restricted to a particular fuel design, fabrication specification or operating condition). The results can consequently be used in solving a wider spectrum of problems and can be applied to different cases;
to achieve experimental efficiency through the judicious use of a combination of experimental and theoretical techniques and approaches.
Although the primary concern of this project is the integrity of LWR cladding during in-reactor service, a number of closely related areas may also be addressed, which can be relevant to water reactors in general. As already mentioned, the cladding behaviour of discharged fuel during handling, transportation and storage shares some features in common with the cladding's integrity whilst in the reactor.
Nuclear safety organisations, research laboratories and industry are being contacted in order to establish the technical and financial basis for this project. From the Swedish side, the project is supported by the utilities, the Swedish Nuclear Inspectorate and Westinghouse Atom.
The project has so far focused on the execution of several power ramps and in defining a hot cell programme with focus on the various failure mechanisms, which will be studied within the corresponding project tasks. These are as follows:
Pellet-clad interaction (PCI): stress corrosion cracking (SCC) initiated at the cladding inner surface under the combined effect of the mechanical loading and chemical environment caused by an increase in the fuel pellet temperature following a power increase.
Hydride embrittlement: time independent fracture of existing hydrides.
Delayed hydride cracking (DHC): time dependent crack initiation and propagation through fracture of hydrides that can form ahead of the crack tip.
Recently, the US NRC decided to carry out a test programme on the behaviour of high burn-up fuel in Loss-Of-Coolant Accident (LOCA) conditions in the Studsvik hot cells and in conjunction with the SCIP programme. Studsvik is a well-established international nuclear research centre located in Sweden. It performs, amongst other tasks, the examination and testing of nuclear fuel and specialises in studies of cladding behaviour under a variety of test conditions. Studsvik aims to become a centre of excellence for investigations of irradiated fuel cladding materials and maintains close co-operation with the Swedish Royal Institute of Technology and Malmo University on cladding integrity investigations. This research is being carried out under a Swedish-sponsored programme similar to the present project. As recommended by the CSNI, comprehensive industry participation was sought in the project establishment phase.
Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, United Kingdom, United States.
Project period: Current mandate: July 2004 to June 2009.
The distribution of this package is restricted and subject to prior approval.
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A-M. Alvarez-Holston, G. Lysell, V. Grigoriev (2007):
Studies of Hydrogen Assisted Failures Initiating at the Cladding Outer Surface of High Burnup Fuel Using a Modified Ring Tensile Technique, Proceedings of the 2007 International LWR Fuel Performance Meeting, San Francisco, California, USA, 30 September-3 October 2007.
Keywords: LWR reactors, cladding, hydriding, pellet clad interaction (pci), pellet clad mechanical interaction.