Left: Scheme of Studsvik LOCA test facility Right: Microstructure of fragmented nuclear fuel, Studsvik

The Studsvik Cladding Integrity (SCIP) project investigates nuclear fuel behaviour and cladding failure that may arise during operation and Loss of Coolant Accident (LOCA) transients. More recently, investigations of fuel behaviour in storage after operation have been added to the project scope (in the fourth phase).

For the SCIP project, Studsvik Nuclear AB (Sweden) conducts advanced out-of-pile experiments on irradiated nuclear fuels and claddings in its laboratories. Most of the test rigs have been customised to simulate for example nuclear fuel accidents and storage conditions which, together with advanced microscopy and other characterisation technologies, provide the participating organisations from industry, regulation and research with important reference data for modelling fuel and clad behaviour.

The project utilises the large library of irradiated fuels at the Studsvik hot cells and involves integral heating and LOCA testing, small scale heating tests, mechanical testing of weakened and damaged rods, pre-test, in-situ and post-test examinations to obtain data on microstructure and microchemistry, fission gas release tests and analyses using different techniques.

Over the years, the SCIP project has fostered international co-operation and continues to establish a common understanding on key nuclear fuel safety issues between regulatory bodies, utilities, vendors and research organisations with an increasing membership.

SCIP phases

First phase (2004-2009)

In the first phase, detailed investigations of how cladding properties affect fuel rod failures that may arise during operation were performed. Irradiated PWR and BWR claddings ramp testing of 11 fuel rodlets, comprehensive non-destructive examination and destructive examinations, mechanical testing and extensive micro-structural characterisation were completed. Regarding pellet-clad interaction, a modified mandrel testing technique to study iodine induced stress corrosion cracking (ISCC) was developed and mechanical test techniques were enhanced to better simulate in-reactor behaviour. The project provided an enhanced understanding of the dominant failure mechanisms for LWR cladding during normal operation and anticipated transients focusing on high burn-up cladding (Studsvik Cladding Integrity Project (SCIP): Executive Summary).

At the end of the first SCIP phase, it was clear that fuel damage studies remained of high interest for reactor performance and safety. At the same time many fuel manufacturers were developing new advanced pellets, aiming to reduce the failure risk, which contributed to the high interest in these questions and a new SCIP phase on fuel failure mechanisms could be successfully launched with an increased number of members.

Data and reports from SCIP phase one may be obtained upon request at https://www.oecd-nea.org/tools/abstract/detail/csni2019.

Second phase (2009-2014)

A second phase (SCIP-2) of the project started in July 2009 and was completed in June 2014. It was built on the considerable knowledge generated in the previous SCIP programme phase. The goal of the second phase was to continue generating high quality experimental data to improve the understanding of the dominant failure mechanisms for water reactor fuels and to devise means for reducing fuel failures. The major focus had been on cladding failures that are caused by pellet-cladding mechanical interaction, especially stress corrosion and hydrogen-assisted fracture mechanisms, as well as on the propagation of cladding cracks.

In parallel with SCIP phase 2, Studsvik, on behalf of the U.S. Nuclear Regulatory Commission, carried out a LOCA test series and shared the results with SCIP members. The NRC LOCA test programme at Studsvik had the objective of investigating the strength and ductility of High Burn-up (HBU) rods after ballooning, rupture, oxidation and quench. The fine fuel fragmentation and dispersal observed in the first four Studsvik tests was unexpected (M. Flanagan, P. Askeljung and A. Puranen, “Post-Test Examination Results from Integral, High-Burnup, Fuelled LOCA Tests at Studsvik Nuclear Laboratory”, NUREG-2160, August 2013, Report on Fuel Fragmentation, Relocation and Dispersal (FFRD)) and it brought the subject to the fore when a new phase of the project was under discussion and the focus of the project moved to LOCA.

The scope and objectives of the phase 3 of the project were established following extensive discussions within the SCIP community, including a dedicated workshop in June 2013.

Third phase (2014-2019)

Based on the outcome of the 2013 workshop and the needs and interests of the SCIP community, the third phase (SCIP-3) of the project was devoted to:

• Determine and quantify parameters affecting fuel fragmentation and dispersal in loss-of-coolant accident (LOCA) – The mechanism behind the fine fragmentation and fuel dispersal observed in LOCA tests with very high burn-up fuel were studied to determine and quantify the parametric thresholds of fuel fragmentation and dispersal. The investigation included the effects of parameters, such as fuel burn-up and microstructure, cladding strain, temperature, internal gas pressure and gas flow at the time of rupture and the importance of transient fission gas release.
• Analyse the consequences of off-normal temperature transients – Transients leading to cladding overheating even without large cladding strain or failure still affects cladding material properties. Depending on the severity of the transient fuel, it may be affected to an extent that might challenge continued operation or normal waste handling, transport and storage. The objective of this part was to define criteria and assess overheated fuel rods for classifying them as undamaged or damaged. Damaged fuel rods would require non-standard handling, transport and storage procedures which might have a large economic impact.
• Observe effects of axial load on cladding failure – Large axial dimensional changes are induced in fuel rods by the temperature occurring during a LOCA. Restricting axial contraction during quenching, e.g. by mechanical interaction with spacer grids, might impact the extent of cladding damage and fragmentation. The objective here was to determine the impact of axial load on cladding performance under LOCA conditions.
• Study the effect of power ramp rates on PCI failure risk – SCIP-3 studied the mitigating effect of low ramp rates on PCI fuel failure risk and created the basis for eliminating unnecessary conservatism during plant operation.
• Support model development and verification – SCIP-3 contributed to the mechanistic understanding of fuel fragmentation and dispersal phenomena and thus supported model development. Modelling was also used to support the set-up of experimental parameters. Modelling workshops were an integral part of the SCIP-3 programme.
• Knowledge transfer: Knowledge transfer was integrated into the programme, using renowned experts. Expert contributions were an integral part of the workshops and programme review group meetings.

There was also a smaller part of the project related to pellet cladding interaction (PCI) failures. Modelling was also an essential and integral part of the project.

Report on Fuel Fragmentation, Relocation and Dispersal (FFRD)

Pellet-Clad Interaction (PCI) in Water-Cooled Reactors Workshop on pellet-clad interactions (PCI) in water-cooled reactors.

Fourth phase (2019-2024)

The fourth phase, SCIP-4, was completed in 2024, and extended the experimental and modeling work related to Pellet Clad Interaction (PCI) and to LOCA issues and investigated fuel and cladding performance related to interim storage. For PCI, microstructural characterisations and modeling activities were conducted. For LOCA, investigations of fission gas release in transients and of fuel fragmentation, relocation and dispersal, including microstructural characterisations of fuel fragmentation have been performed. Spent fuel pool LOCA studies were also conducted. For back-end studies, creep and hydride reorientation phenomena were investigated. Spent fuel rod transport and handling accident studies were also conducted.  

SCIP-4 generated high-quality experimental data to improve the understanding of the dominant failure mechanisms for water reactor fuels and devise means for reducing fuel failures, achieved results of general applicability (i.e. not restricted to a particular fuel design, fabrication specification or operating condition) and experimental efficiency through the judicious use of a combination of experimental and theoretical techniques and approaches.

Related external links

https://www.studsvik.com/scip-project/what-is-scip/

Flanagan, P. Askeljung and A. Puranen, “Post-Test Examination Results from Integral, High-Burnup, Fuelled LOCA Tests at Studsvik Nuclear Laboratory”, NUREG-2160, August 2013

Fifth phase (2024-2029)

The fifth phase of the project (2024-2029), which started in July 2024, continues to address the back end and LOCA, extending investigations conducted in the previous phases. A new task is dedicated to accident-tolerant fuels (ATFs) investigations. This new project phase is supported by 44 organisations from 15 countries. The first task of the project includes experimental investigations of fuel and cladding performance issues related to interim storage: creep of irradiated cladding and fuel under dry storage conditions with a focus on high burn-up fuels, hydride precipitation and reorientation in irradiated cladding, cladding annealing in long-term dry storage, and treatment of failed fuel. The second task investigates phenomena in LOCA transients: FFRD and transient fission gas release and axial gas communication. Regarding FFRD, work will aim at extending the database for standard fuels and fuels which have not yet been investigated (e.g. doped fuel, Gd fuel, MOX fuel). Detailed microstructure analyses of transient tested standard and non-standard fuels will be performed. The third task addresses ATFs with tests on mechanical properties of coating and cladding in irradiated coated ATF cladding and with LOCA tests to investigate ballooning and burst of ATF cladding types, in comparison to standard Zr-based alloy cladding tubes. A fourth task includes analytical and modelling activities (pre- and post-test calculations, benchmarking).

The first two meetings of the projects were held in July and December 2024 and discussed material selection, qualification and methods for the testing for the various project tasks. Regarding analytical activities, modelling cases have been defined. Actual testing is planned to start in 2025.