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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Dry Storage Handbook

This handbook contain a technical assessment of the expected performance of spent nuclear fuel (SNF) during extended dry-storage time periods and the condition of such fuel at the end of dry storage.

The principal focus of the reviews is on SNF and the effects of dry storage rather than on dry-storage containers and the related storage facilities. The objective is to provide background information on the likely behavior of materials comprising water reactor fuel assemblies and on the performance of integral assemblies under conditions typical of dry storage for extended intervals of time.

In brief, the technical assessment supports a conclusion that, although technical issues have been postulated with regard to long-term storage, there are no high-risk concerns with the extension of dry storage to long times; with proper planning and implementation.

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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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Fuel Material Technology Report (FMTR) – Volume I

This first volume of the Fuel Material Technology Report covers short general outlines of the designs of PWR, BWR, CANDU, VVER and RBMK reactors but with focus on the corresponding fuel assembly and supporting structure designs together with the rational for the selection of the materials used in the different applications. This volume also gives an overview of the in-reactor fuel performance, of fuel performance codes and the manufacturing process of the fuel assembly.

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Fuel Design Review Handbook (FDRH)

This Fuel Design Review Handbook intends to provide a guide to the items that have the greatest influence on fuel performance and prioritize the audits that are recommended. A review of all aspects of the fuel design is not feasible or necessary within the time constraints of the utility and the vendor. The objective is to do the most effective audit in the shortest time period. The Handbook provides the “what, why and how” for the audits by describing the design criteria, their influence on performance and the approach to reviewing the associated design features for the three distinct technical areas of nuclear, thermalhydraulic, and mechanical/materials design, each written by experts in their field. A guide for design tool verification is included as well as a guide to auditing the vendor design QA system.

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