Hot Cell Post-Irradiation Examinations Volume II

(ZIRAT21/IZNA16)

Maintaining and improving reliability of fuel and structural components requires an understanding of their behaviour in reactor and the mechanisms that have been observed to cause failures. A key factor in improving reliability is the identification of the cause or causes of failure. Such information, in turn, requires the examination and analysis of irradiated fuel (including bundle hardware) and structural components at reactor sites (poolside examinations), in hot cells and in related laboratories. Thus, to make progress toward ultra-high reliability fuel and to reduce the potential for fuel failure, it is imperative to examine both failed and non-failed (reference) fuel.

Post-irradiation examinations (PIE) provide fuel vendors and nuclear utilities with data on how newly developed or established materials withstand normal operating conditions in new environments. Post-irradiation examinations are largely carried out at a Hot Cell Laboratory where irradiated fuel rods and other hardware can be received, handled, examined, and tested. The investigation results provide information for fuel and component improvement and, thereby, can potentially enhance operating efficiency and reliability.

• Section 1 provide an overview about the status of post-irradiation examination (PIE) and inspection techniques for nuclear fuel and other zirconium alloy components used in CANDU reactors and their applications for analysis of materials behaviour in a CANDU reactor core.
• Section 2 discusses these techniques along with real world examples of in-reactor microstructural changes and impact on material behaviour.
• Section 3 provides information on PIE capabilities of some of the major hot cell facilities. This information will be useful for utility engineers when they need to have PIE performed on failed nuclear fuel or other components.

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Nuclear Fuel Behaviour Under RIA Conditions

(ZIRAT21/IZNA16)

The design basis RIA in a PWR is the Control Rod Ejection (CRE), while in a BWR, it is the Control Rod Drop Accident (CRDA). The CRE is based on the assumption of a mechanical failure of the control rod drive mechanism located on the reactor vessel top, followed by the ejection of the mechanism and the control rod by the internal reactor pressure. The resulting, significant power surge is limited partly by Doppler feedback and finally terminated by the reactor trip. The BWR CRDA is assumed to occur if a control rod is detached from its drive mechanism in the core bottom, stays stuck while inserted in the core, then if loosened, drops out of the core by gravity, without involvement of a change in reactor pressure as in the CRE. Partly as a result of these differences, the BWR power pulses are slower than for a PWR. The pulse widths for PWRs are in the range of 10–30 ms and for BWRs in the range of 20–60 ms.
The reactivity transient during an RIA results in a rapid increase in fuel rod power leading to a nearly adiabatic heating of the fuel pellets.
The RIA-simulation experiments conducted in the 1960’s and 1970’s using zero or low burnup test rods showed that cladding failure occurred primarily by either:

• Post-Departure from Nucleate Boiling (DNB)
• Cladding contact with molten fuel

This Special Topic Report (STR) will give insight and understanding of the parameters impacting the fuel RIA performance and reviews the applicability of the data to high burnup fuel cladding. The STR also provides the latest RIA regulatory acceptance criteria.

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Hot Cell Post-Irradiation Examination Techniques for Light Water Reactor Fuels

(ZIRAT19/IZNA14 STR)
Hot Cell Post-Irradiation Examination Techniques for Light Water Reactor Fuels The growing operational demands on nuclear fuel, such as longer fuel cycles, higher burnups, and use of transient regimes, call for more robust fuel designs and more radiation resistant materials. Implementation of new materials and fuel designs that are able to meet these more challenging conditions requires adequate operational feedback and practical verification of models for prediction of fuel behavior. Post-irradiation examinations (PIE) provide fuel vendors and nuclear utilities with data on how newly developed or established materials withstand normal operating conditions in new environments. Post-irradiation examinations are largely carried out at a Hot Cell Laboratory where irradiated fuel rods and other bundle hardware can be received, handled, examined, and tested. The investigation results provide information for fuel improvement and, thereby, can potentially enhance operating efficiency and reliability. The objectives of the hot cell examination of failed and sound sibling rods are to; characterize the fuel rod conditions associated with failure, identify the fuel failure mechanism, and provide insight into the root cause of the failures. The objectives of the hot cell examination of the fuel bundle hardware vary with the component being examined. The hot cell examinations/testing include a number of tasks selected to address these objectives using available hot cell capabilities. This Special Topic Report provides an overview about the status of post-irradiation examination (PIE) and inspection techniques for nuclear fuel and their applications for analysis of material degradation during fuel operation in a reactor core. Emphasis is given to advanced non-destructive and destructive PIE techniques applied to LWR fuel rods and bundle hardware. The objective of this STR is to provide this knowledge.

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Fuel Fabrication Process Handbook Rev 1 (FFPH)

The objective of the Fuel Fabrication Process Handbook is to provide guidance for a cost effective audit which uses audit time on areas which are most likely to affect the performance of the PWR/VVER and BWR fuel. The FFPH focuses on a “Process Audit” procedure, the audit of the fabrication process parameters for making high quality fuel. The FFPH provides the “what, why and how” to look at in an audit by:

• Listing the generic fabrication process steps for all components and their assembly (what to look for).
• Identifying important audit points and the attendant potential effect of deviations on performance (why to look).
• Assess the fabrication and QC process control at critical points (how to look).

This Handbook is an updated and expanded version of the previous FFPH Handbook published in 2005. More than 35 organisations worldwide bought this Handbook. The expansion constitutes two sections on Statistical Quality Control and Software Quality Assurance. Statistical QC is a vital part of process control, the establishment of sampling plans and the qualification of inspection methods. The QA of software is important for auditing the software for the expanding automation of fabrication methods.

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Control Assembly Technology Report (FMTR Volume III)

The Report on Control Assembly Technology which constitutes Volume III of the series of Fuel Material Technology Reports (FMTRs) will be available during the Spring of 2014. It describes the designs, manufacturing, performance and issues related to BWR/PWR/VVER/CANDU Control Assemblies with Ag-In-Cd (AIC), B4C, Hf absorber materials and stainless steel structural materials.

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PWR Zr Alloy Cladding Water Side Corrosion (PZAC)

The Report discusses the different parameters impacting PWR fuel corrosion and provides a computer code which allows an equivalent comparison of new alloys irradiated in different reactors at different conditions. The computer code may also assist in identifying the mechanism why the corrosion rate starts to accelerate under certain conditions.

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Management of BWR Control Rod (MBCR)

The objective of this report is to provide the customers with information which will be a useful tool for management of control rods.

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Fuel Material Technology Report (FMTR) – Volume IV

The primary objective of this volume of the Fuel Material Technology Report, FMTR Vol. IV, is to provide guidance in improving fuel reliability. To reach this objective, various Poolside and Hot Cell Examinations techniques may be used. A good knowledge of the pros- and cons- of the different techniques can guide the utility or fuel vendor to select the most cost efficient techniques for their specific goals. A second objective of this Report is to document this knowledge in a form which can be updated as new information, and methods become available.

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Fuel Material Technology Report (FMTR) – Volume II

In this second volume of the Fuel Material Technology Report, the following topics for different water cooled type reactors are covered:

• Effect of radiation on the various fuel assembly materials as well as the interaction between materials and cooling water chemistry in this radioactive environment.
• Water chemistries in different reactor types
• Fuel design criteria and operating limits
• Fuel performance during Loss Of Coolant Accidents and Reactivity Initiated Accidents
• Fuel performance during intermediate storage
• Trends and potential burnup limitations

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