Corrosion and Hydrogen Pickup – Vol. I

Corrosion and hydrogen pickup (HPU) mechanisms of Zr alloys remain a top priority of the nuclear industry. Commercial Zr alloys have today adequate in-reactor corrosion properties. However, hydrogen in fuel components limits the fuel performance today during normal operation and accident conditions as well as during transport of spent fuel.

DOWNLOAD SAMPLE  | ADDITIONAL INFORMATION

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Hydride Reorientation

Hydride orientation has an important effect on fracture toughness of hydride-containing zirconium alloys because hydrides form as approximately linear arrays of platelet-shaped microscopic precipitates with habits on or near the basal planes of the α–Zr matrix in which they form.

This Stand Alone Report (SAR) addresses a key aspect of the issues raised in the foregoing by providing a comprehensive, self-contained and up-to-date review and analyses of the results of studies carried out on the conditions governing hydride orientation in zirconium alloy pressure and fuel cladding tubes used in nuclear reactors. The report combines a detailed theoretical and experimental overview of this subject with the author’s own analyses of these results. These analyses make use of theoretical advances documented in the author’s 2012 book dealing with the effects of hydrogen and hydrides on the integrity of zirconium alloy components. In the author’s 2012 book, emphasis is placed on delayed hydride cracking, which is a localised failure mechanism.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

Covering the Zr Related Results Published During 2016-2017

(ZIRAT22/IZNA17 AR)

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.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

Dimensional Instability and Irradiation Growth of Zirconium Alloys

(ZIRAT22/IZNA17 STR)

In this STR review, aimed specifically at irradiation growth, addressed are conditions of direct interest to LWRs and CANDUs, including information that has mechanistic implications. Irradiation creep was covered earlier by ZIRAT14 Special Topic Report: In-reactor Creep of Zirconium Alloys, authored by Ron Adamson, Friedrich Garzarolli and Charles Patterson, 2009.

The STR addresses all data deemed relevant to understanding irradiation growth, including a broad review of growth mechanisms, and a summary of practical effects of growth on component performance.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

Performance Evaluation of New Advanced Zr Alloys for BWRs and PWRs/VVERs Vol. I

(ZIRAT22/IZNA17)

To meet the current situation with more aggressive reactor environments (higher burnups, changing water chemistries and loading patterns), and resolving fuel performance issues such as BWR channel bowing and PWR assembly bowing, a large number of zirconium alloys have been and are being developed. The main driver for the initial material development in Pressurised Water Reactors (PWRs) has been to reduce corrosion rates and Hydrogen Pick-Up Fractions (HPUFs), which have occasionally limited the maximum discharged burnup.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

Performance Evaluation of New Advanced Zr Alloys for BWRs and PWRs/VVERs Vol. II

(ZIRAT22/IZNA17)

To meet the current situation with more aggressive reactor environments (higher burnups, changing water chemistries and loading patterns), and resolving fuel performance issues such as BWR channel bowing and PWR assembly bowing, a large number of zirconium alloys have been and are being developed. The main driver for the initial material development in Pressurised Water Reactors (PWRs) has been to reduce corrosion rates and Hydrogen Pick-Up Fractions (HPUFs), which have occasionally limited the maximum discharged burnup.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET

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

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.

DOWNLOAD SAMPLE  | DOWNLOAD PRODUCT SHEET