Anatech - Linking Theory and Practice.

CSED correlations for Zry-2 and Zry-4 cladding at low temperatures have been developed using mechanical property test data of irradiated materials. Both of the Zry-2 and Zry-4 CSED models are functions of hydrogen concentration in the cladding since the embrittlment of the cladding tube is caused principally by the hydride precipitates; Models have been used mainly for determining fuel failures during RIA transient featuring multi-axial loading and high strain rates. Thermal and mechanical responses of RIA test rods, HBO, TK, and FK series conducted at NSRR, were evaluated using EPRI’s fuel performance code FALCON with an emphasis on
modeling the cladding deformation driven by UO2 pellet expansion during the power pulse. Application of the CSED models to RIA tests shows that the model can predict the failures reasonably well, with an error of failure enthalpy from –10 to 15 cal/g.

EPRI’s FALCON fuel performance code contains a fuel failure assessment capability that calculates a cumulative damage index, CDI, applicable to the PCI fuel failure mechanism. It is based on an adaptation of the cumulative damage concept coupled with a local cladding stress evaluation. The CDI is tailored specifically for the SCC failure mechanism and represents a figure of merit that essentially describes cladding vulnerability to failure as a function of history-dependent phenomenon such as power, temperature, stress, and fluence. This paper presents an evaluation of results from the analysis of PWR test rods using the cumulative damage method as compared to more traditional threshold stress methods for determination of PCI-induced fuel rod failure vulnerability.

The final version of acceptance criteria are being developed in the United States (US) for use in the safety analysis of the hot-zero power (HZP) and hot-full power (HFP) Reactivity Initiated Accidents (RIA) in Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Recently, the staff at the US Nuclear Regulatory Commission (NRC) issued Interim RIA acceptance criteria within Revision 3 of the Standard Review Plan (NUREG-0800) for use in new reactor certification and licensing [1]. These criteria attempt to account for exposure induced changes in fuel rod behavior at higher burnup. The interim RIA acceptance criteria have been developed using an empirical approach with limited consideration for the differences between RIA-simulation tests and LWR RIA events or UO2 and Mixed Plutonium-Uranium (MOX) fuel.

The Electric Power Research Institute (EPRI), under the auspices of the Fuel Reliability Program Working Group 2 – Fuel Regulatory Issues, has undertaken an effort to evaluate the impact of irradiation on MOX fuel rod behavior during an RIA event. RIA-simulation tests have shown that irradiated MOX fuel pellets at similar burnup can experience a larger expansion response compared to UO2 pellets due to the dispersed distribution of high burnup structure of plutonium agglomerates throughout the pellet [2]. This enhanced pellet expansion increases the clad loading at a given fuel enthalpy level in comparison to UO2 fuel.

This paper summarizes the development and verification of a MOX fuel pellet transient gaseous swelling model for use in the FALCON fuel performance code and the evaluation results of MOX fuel rods being tested at the Japan Atomic Energy Agency (JAEA) Nuclear Safety Research Reactor (NSRR).

Data on the factors that influence cladding embrittlement during a LOCA event are reviewed and the ramifications of these factors on the ECCS acceptance criteria and evaluation models used in LOCA safety analyses are provided. These factors include higher fuel burn-up at discharge, the prevalence of an inner surface oxide and fuel-clad bonding layers, and development of new or improved cladding materials. The ramifications of these factors on the
ECCS acceptance criteria and evaluation models can be substantial and will require thorough consideration before revised criteria and requirements are issued.

A recently developed constitutive model treats the cladding as a multi-material composite in which the alloy matrix and the hydride platelets are treated as (i) separate material phases with their own elastic-plastic and fracture properties and (ii) interacting at their interfaces with appropriate constraint conditions between them to ensure strain and stress compatibility. An essential feature of the model is a multi-phase damage formulation that models the complex interaction between the hydride phases and the alloy matrix and the coupled effect of radial and circumferential hydrides on cladding stress-strain response. This gives the model the capability of directly predicting cladding failure progression during loading events and, as such, provides a unique tool for constructing failure criteria analytically where none can be readily developed by conventional material testing. Implementation of the model in a fuel behavior code provides the capability to predict fuel rod failures under operational and transient conditions without having to rely on failure criteria. The model is of particular use for assessing the structural integrity of high-burnup spent fuel subjected to hypothetical transportation and handling accident conditions, for which the development of valid failure criteria would require major efforts.

The prevention of fuel failures due to pellet-cladding interaction (PCI) has received renewed attention over the past several years. The Electric Power Research Institute (EPRI) developed technical guidelines to assist utilities with the prevention of fuel failures due to a number of causes, including PCI. As part of the EPRI effort, many detailed fuel rod mechanical analyses were performed by ANATECH Corporation (ANATECH) using EPRI’s FALCON computer code. Duke Energy Carolinas, LLC (Duke), other utilities, and fuel vendors provided ANATECH with detailed core power and power distribution data from recent operating experience in order to (i) identify fuel rods that are potentially vulnerable to PCI and (ii) analyze cladding stress levels in those fuel rods using FALCON. The results of the fuel mechanical analyses were used to provide insights into PCI risk for specific fuel designs, power maneuvering rates, and core designs. This paper describes core power and power distribution calculations using SIMULATE-3 and fuel rod mechanical analyses using FALCON for two pressurized water reactors operated by Duke Energy: Oconee Unit 2 and McGuire Unit 1.

The source-term study conducted by Sandia National Laboratories nearly two decades
ago for the spent fuel inventory known at the time, which was in the low-to-medium burnup range
(∼ 35 GWd/MTU), showed that the effects of transportation accidents on spent fuel failures, and
consequential radioactivity release to the environment, were relatively benign. However, with
today’s discharged fuel burnups routinely greater than 45 GWd/MTU, potential hydride
reorientation during interim dry storage, and its effects on cladding properties, has become one of
the primary concerns for spent fuel transportation. Laboratory tests of un-irradiated cladding
specimens subjected to heat treatments promoting hydride dissolution followed by re-precipitation
in the radial direction have shown that relatively moderate concentrations (∼70 ppm) of radial
hydrides can significantly degrade cladding ductility, at least at room temperature. The absence of
specific data that are relevant to high-burnup spent fuel under dry storage conditions have led to
the conjecture, deduced from those tests, that massive cladding failures, possibly resulting in fuel
reconfiguration, can be expected during cask drop events. Such conclusions are not borne out by
the findings in this paper. The analysis results indicate that cladding failure is bi-modal: a state of
failure initiation at the cladding ID remaining as part-wall damage with less than 2% probability
of occurrence, and a through-wall failure at a probability of 1E-5. These results indicate that
spent fuel conditions that could promote the formation of radial hydrides during dry storage are
not sufficient to produce radial hydrides concentrations of significant levels to cause major threat
to spent fuel integrity. It is important to note in this regard that the through-wall cladding failure
probability of 1E-5 is of the same order of magnitude as calculated in the cited Sandia study for
low burnup fuel.

EPRI, under the auspices of the Fuel Reliability Program Working Group 2 – Fuel Regulatory Issues, has undertaken an effort to evaluate the impact of irradiation on fuel rod behavior during and following a Loss of Coolant Accident (LOCA) in support of extending the current fuel rod burnup limits [1]. This initiative includes supplying high burnup fuel rods for testing in the NRC LOCA program at Argonne National Laboratory (ANL), providing analytical support for the design and conduct of the ANL and Halden integral LOCA experiments and post-quench mechanical property tests, and performing an independent assessment and interpretation of the
experimental results to evaluate the overall behavior of high burnup fuel. The primary objective of this
activity is to identify the impact of burnup on the cladding embrittlement criteria defined in 10 CFR 50.46 part b(1) and b(2). The purpose of this paper is to provide a summary of recent evaluations performed to assess the impact of irradiation on the cladding residual ductility following high temperature oxidation and quench during a LOCA
event.

This paper provides a review and analytical assessment of several RIA-simulation experiments performed between 1996 and 2002 on test rods refabricated from high burnup commercial UO2 fuel rods. The evaluation focused on twenty-four (24) RIA-simulation tests, including eight (8) tests with cladding failure, that were performed in the CABRI (REP-Na) and NSRR test facilities on both PWR and BWR rod segments with burnup levels between 45 and 75 GWd/tU. Experimental data, including available PIE reports, were collected, and the experiments were analyzed using the FALCON fuel behavior code to gain insights into the thermal and mechanical state of the test rods during and following the power pulse. The key conclusions from this evaluation are: 1) pellet thermal expansion is the primary PCMI loading mechanism during the early phase of the pulse in LWR UO2 fuel up to a peak pellet burnup of 75 GWd/tU, 2) cladding integrity is controlled primarily by a combination of the hydride phase (hydrogen content, distribution, and orientation) and the cladding temperature, and 3) post-DNB cladding
temperature excursions following the power pulse can allow for cladding deformation beyond pellet thermal expansion.

The recent advent of several key behavioral issues in nuclear fuel performance, resulting from the trend toward increased fuel utilization, has created the need for enhanced steady state and transient fuel rod analysis capabilities. These issues, derived primarily from burnup-induced pellet changes and increased hydrogen uptake (as a byproduct of corrosion) occurring at higher burnups, impact steady state operation and licensing issues such as postulated reactivity initiated accidents (RIA), loss-of-coolant accidents (LOCA), and spent fuel storage and transportation. In response, EPRI has undertaken the development of FALCON, an integrated steady state and transient fuel performance analysis code applicable to the full range of fuel operating regimes. The FALCON development program has emphasized the use of advanced thermo-mechanics in a fully two-dimensional finite element continuum framework. Based on the architecture of the FREY transient analysis code and including salient models from the ESCORE steady state analysis code, FALCON provides enhanced steady state and transient numerics, updated material property and behavioral models, and an expanded and relevant light water reactor (LWR) benchmarking and verification database. This paper will present an overview of the FALCON steady state analysis development program with an emphasis on the results from the benchmarking, verification, and validation activities.

Revised acceptance criteria have been developed for use in the safety analysis of the hot-zero power (HZP) and hot-full power (HFP) Reactivity Initiated Accidents (RIA) in Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The primary RIA events considered are the postulated control rod ejection accident (REA) for PWRs and the postulated control rod drop accident (RDA) for BWRs. The revised RIA acceptance criteria have been developed as part of the on-going nuclear industry effort to extend fuel rod average burnup levels beyond the current limit of 62 GWd/MTU.

The revised acceptance criteria have been developed using an evaluation methodology that combines both experimental data and analytical calculations to establish the influence of burnup on transient fuel rod behavior during RIA events. Used in the evaluation were experimental data from RIA-simulation tests on fuel segments extracted from commercial UO2-Zircaloy cladding fuel rods irradiated to 64 GWd/MTU. These technical bases were then translated to the PWR REA application using the state-of-the-art fuel rod behavior analysis code FALCON as a means to establish the fuel rod failure threshold and core coolability limit as a function radial average fuel enthalpy and rod average burnup. The advantage of using a combined approach of analytical evaluations and experimental data to derive the revised acceptance criteria is that the methodology can be applied to other fuel rod designs and cladding materials to determine application specific criteria, if relaxation of the criteria is required.