Covering the Zr Related Results Published During 2015-2016

(ZIRAT21/IZNA16)

The overall objective of the ZIRAT Programme is to enable the
nuclear utilities and laboratories to:

• Gain increased understanding of material behaviour related to
  successful core options for the back end of the fuel cycle.

The objective is met through review and evaluation of the most recent data on zirconium alloys, identification of the most important new information, and discussion of its significance in relation to fuel performance now and in the future. Included in the review are topics on materials research and development, fabrication, component design and in-reactor performance.

The evaluations are based on the large amount of non-proprietary data presented at technical meetings, published in the literature and provided through discussions with zirconium materials manufacturers.

The open literature information will be collected throughout the year and the data most important to the utilities and laboratories is selected for the Annual Report. The large collective experience gained by the reviewers in past and current projects is an important factor in making the evaluation, hence ensuring that the presented compiled information is put in perspective, and that the most important information is emphasized. The data will be useful to
utilities and laboratories to assist them in evaluating:

• New and potential fuel performance problems and performance limits
• The effect of new data on current fuel design bases
• Qualifications desirable for new design features
• The effect of modified or new fabrication processes on properties
• Potential use of new Quality Control (QC) methods
• QC requirements for new materials features
• Qualification needs for new alloys
• Lessons learned from fuel performance regarding design bases, fabrication process control, QC and reactor operation

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

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.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

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.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET