ML042260358

Hearing 07/15/04 - Intervenor Exhibit C, NEA/NSC/DOC(2003)12, Status of NSC Activities in the Field of Fuel Behaviour - Identified 07/15/04

ML042260358
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Site:	Catawba
Issue date:	05/20/2003
From:	Organization for Economic Co-operation and Development (OECD), Nuclear Energy Agency
To:	Office of Nuclear Reactor Regulation
Byrdsong A T
References
50-413-0LA, 50-414-OLA, ASLBP 03-815-03-OLA, JT00144590, RAS 8321 NEA/NSC/DOC(2003)12
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-R45 er-~,c-For Official Use NEA/NSC/DOC(2003)12 Organisation de Cooperation et de Ddveloppement Economiques Organisation for Economic Co-operation and Development 20-May-2003 English - Or. English NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE 0

~. r DOCKETED USNRC 0 August 9, 2004 (11:45AM)

W0 OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF STATUS OF NSC ACTIVITIES IN THE FIELD OF FUEL BEHAVIOUR May 2003 0(

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r NEA/NSC/DOC(2003)12 Status of OECD/NEA NSC Activities in the Field of Fuel Behaviour May 2003 Content

1. Introduction

2. Important Issues for Fuel Performance Modelling

3. Identification of Available Experimental Data

4. IFPE database

5. International Workshops

6. Future Activities

7. References Annex 1: Mandate of NSC Expert Group on Scientific Issues in Fuel Behaviour Annex 2: Issues and Topics for Seminar on Pellet Clad Interaction

1. INTRODUCTION The OECDINEA Nuclear Science Committee (NSC) has long recognised the need for improved knowledge and international co-ordination of important scientific issues related to fuel behaviour.

Consequently, a special Task Force was set up by the NSC in late 1993.

The objectives of the Task Force, as endorsed by the NSC, were to identify areas of high priority to Member countries, which would benefit from international co-ordination and co-operation on studies of the basic underlying phenomena of fuel behaviour under normal operating conditions.

The Task Force was asked also to advise the NSC on developments needed regarding data, models and experiments to meet the requirements for better understanding fuel behaviour and for improved predictive models. Work is coordinated with the CSNI Expert Group on Fuel Safety Margins.

The main findings and conclusions of the Task Force were summarised in a report [1], reviewed in Section 2 below, which set the scene for a series of actions taken by the NEA over the succeeding years. The present report reviews these initiatives and activities, highlighting the achievements already made as well as indicating for the future the direction in which there is benefit to be gained by continuing and extending the current activities.

2. IMPORTANT ISSUES FOR FUEL PERFORMANCE MODELLING The first action of the Task Force was to identify topics and potential authors to review the then current state of understanding, and to identify priority areas where further work was required. This was accomplished-aid a report addressing the following topics was issued [1]:

Thermalperformance; the calculation of fuel temperatures, the effect of design parameters and the effect of irradiation

• Fission Gas Release; the release under different operating regimes and the effect of high burn-up 2

a NEA/NSC/DOC(2003)12

• FissionProduct Swelling; the distinction between inexorable solid fission product swelling which is predominantly a function of bum-up and fission product accumulation, and gaseous swelling which is also dependent on high power operation

• Stress Corrosion Cracking; the conditions of stress, time and fission product release on the propensity for clad failure from this mechanism

• Water Chemistry; current practice and monitoring, its effect along with Zr composition and final treatment on clad corrosion

• Hydrogen in Cladding-,its distribution and measurement

• FailedFuel; detection and modelling degradation processes

• Spentfuel; long term storage under 'wet' and 'dry' conditions

• QualityAssurance; as applied to materials, experimental tests and data production.

The report concluded that although fuel in nuclear reactors had proven to be highly reliable in its performance and safety, reactor operation was often supported by rather generous bounding operating conditions. In the future the requirements would be more onerous. The economic requirement to increase efficiency meant that fuel performance must be calculated accurately on a 'best estimate' basis. To do this requires good fuel performance computer codes validated by high quality data.

As a result of this report the Task Force arrived at the following recommendations for future NEA activities:

i. Countries and organisations should be encouraged to make special research efforts to reduce uncertainties in modelling specific aspects of fuel behaviour, namely

- thermal performance and calculating fuel temperatures

- fission gas release

- fission product swelling and creep of U0 2

- thermo-mechanical behaviour

- high bum-up fuel behaviour in transient conditions ii. As a first step, a review should be made of existing data iii. A public domain database of well-qualified experiments and data should be assembled which can be used for model development and code validation. This database should be organised and maintained by the NEA Data Bank iv. International topical meetings covering high priority issues should be organised by the NEA in co-ordination with the IAEA. Of particular interest were workshops on: thermal performance, fission gas release (FGR) and pellet-clad mechanical interaction (PCM1)

3. IDENTIFICATION OF AVAILABLE EXPERIMENTAL DATA The second report commissioned by the Task Force and issued by the NEA was the review of existing data that could be made available to fulfil the aim of improving code performance. The report [2] concentrated primarily on data produced in the Joint Programme carried out by the Halden Reactor Project, Norway.

Conditions in the Halden reactor are particularly well suited to studies of fuel performance. The boiling conditions ensure a constant coolant temperature, and hence a well-defined boundary condition from which to assess thermal performance from measurements of centreline fuel temperatures. Also, the low system pressure of 32 bar and low fast flux ensures negligible clad creep down, thus removing one parameter from through life assessments of fuel dimensions and temperature. However, when more prototypic conditions are required, dedicated in-pile loops are available to simulate the thermal hydraulic conditions of temperature and pressure as well as neutron flux spectrum for PWR, BWR and, most recently, Advanced Gas Cooled Reactors (CAGR) as operated in the UlK.

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NEA/NSC/DOC(2003)12 The report outlined some of the experiments that have been performed in the Halden reactor and the most important data that they have provided which are available for use in fuel performance studies. As to be expected in a Project that has many years of experience and data production, several reviews of individual topics have already been written. This present report made extensive use of these with the aim of demonstrating the extent of the information available collated in a form that is most useful for the development of fuel performance codes and their validation.

The data can be divided into a number of categories depending on the use to be made of them. In this respect, the data are not of universal interest. The simplest division is into the three broad categories:

• Data useful for model development and validation, e.g.: radial flux depression, fuel creep, fuel densification and swelling and clad creep down. These are the 'unsung needs' of the code developer, not often apparent in code predictions but essential to obtain a good description if reliable predictions are to be obtained. Very often these data require special measuring techniques and non-prototypic rig designs.

• Data of direct relevance to licensing requirements, e.g.: fuel temperatures, stored heat, fission product release and rod internal pressure, waterside corrosion. Such data are of particular use for validation purposes.

• Data for fuel development and optimisation, e.g.: performance of fuel variants and new cladding materials, effect of changes in fuel design. These data are of most interest to fuel vendors in developing and supporting new products.

Within these categories, the report discussed data under the following phenomenological headings:

- Radial Flux Depression

- Thermal Performance

- Fuel Densification and Swelling

- U0 2 Grain Growth

- Fission Product Release

- Clad Properties

- PCMI - Pellet Clad Mechanical Interaction

- Integral Behaviour

- High Bum-up Effects.

The report reviewed experiments that addressed 'single effects', i.e., where single parameters were isolated for study, (fuel centre temperature as a function of fuel pellet to clad gap), and experiments which addressed 'integral behaviour', where several effects were studied simultaneously (fuel temperature, fuel and clad axial extension and rod internal pressure as a function of power and bum-up).

In addition to the Halden Project data, the report briefly reviewed data from certain Studsvik ramp tests and from the three Riso fission gas release projects. These had already been shown to be compatible with Halden experiments, particularly in the areas of PCI/SCC failure and PIE of ramp tested fuel.

Despite the comprehensive nature of the data available, as outlined in the report, it was clear that there were a number of areas where further experiments were necessary. With the move to higher discharge bum-up, it is necessary to extend the data to cover the extremes of bum-up and power expected in commercial reactors.

This implied a progressive extension of well qualified data in excess of 70 MWd/kgUO 2. New effects were being discovered at these levels of bum-up and further data were required on such topics as: 'the rim effect, further definition of the effect of bum-up on fuel thermal conductivity, clad corrosion and hydriding and clad mechanical properties. Although there were data on temperature changes during reactor scrams, for transient 4

NEA/NSC/DOC(2003)12 analysis, there was a need for data on the evolution of temperatures during rapid power changes. There were (and still are) few data on this last topic and there is clearly a benefit to be claimed in reactivity faults for example, to exploit the lag between power increase and consequent increase in temperature.

The report supported the Task Force initiative to assemble an International Fuel Performance Experiment (IFPE) database, and recommended the inclusions of experimental data identified in the report from all three international programmes.

4. IFPE DATABASE From the outset it was recognised that the database should apply to all commercially operated thermal reactor systems and that the data should be both prototypic, originating from power reactor irradiations with pre- and post-irradiation characterisation, and test reactor experiments with in-pile instrumentation and PIE exploring normal and off-normal behaviour. It was recognised that experiments have been performed where the data remain of commercial interest, for example, much of the details of modem MOX performance remains proprietary to the manufacturers, and it was not the intention to compromise such arrangements. However, Zircaloy clad UO0 pellet fuel can be largely regarded as a 'standard product' and as such, release of what was previously proprietary data can only benefit the nuclear community at large.

A particular aspect of the compilation was the inclusion of data generated within internationally sponsored research programmes whose confidentiality agreements had expired. Such data, although available in principle, have not been widely used. The inclusion of such data is of particular importance where the originating organisation has changed its terms of reference. For example, the Riso laboratories in Denmark no longer perform nuclear research and find difficulty in resourcing the supply of information for their three fission gas release projects. In such cases, there exists the danger of losing access to the data altogether.

4.1 Extent of parameters included The database is restricted to thermal reactor fuel performance, principally with standard product Zircaloy clad U02 fuel, although the addition of advanced products with fuel and clad variants is not ruled out. Recent additions to the Database include MOX fuel, (U,Gd)0 2 and defect fuel. Data encompass normal and off-normal behaviour but not accident condition entailing melting of fuel and clad, resulting in loss of geometry.

Of particular interest to fuel modellers are data on: fuel temperatures, fission gas release (FGR), clad deformation (e.g. creep-down, ridging) and mechanical interactions. In addition to direct measurement of these properties, every effort is made to include PIE information on the distribution of: grain size, porosity, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements on caesium, xenon, other fission product and actinides.

Emphasis has been placed on including well-qualified data that illustrate specific aspects of fuel performance.

For example, cases are included which specifically address the effect of gap size and release of fission gas on fuel-to-clad heat transfer. Also in the context of thermal performance, the effect of bum-up on UO2 thermal conductivity has been addressed. This is illustrated by cases where fuel temperatures have been measured throughout prolonged irradiation and at high bum-up where sections of fuel have been refabricated with newly inserted thermocouples. Regarding fission gas release, data are included for normal operations and for cases of power ramping at different levels of burn-up for fuel supplied by several different fuel vendors. In the case of power ramps, the data include cases where in-pile pressure measurements show the kinetics of release and the effect of slow axial gas transport due to closed fuel-to-clad gaps. Supplementing these in-pile studies, there are datasets of Out-of-Pile annealing studies measuring FGR under well defined conditions of temperature and time.

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NEA/NSC/DOC(2003)12 To date datasets about 446 rods/samples from various sources encompassing PWR, BWR, PHWR and WWER reactor systems have been included:

Halden irradiated IFA-432 5 rods Halden irradiated IFA-429 7 rods Halden irradiated IFA-562.1 12 rods Halden irradiated IFA-533.2 I rod Halden irradiated IFA-535.5 &.6 4 rods The Third Ris0 Fission Gas Release Project 16 rods The Ris0 Transient Fission Gas Release Project 15 rods The SOFIT WWER fuel Irradiation Programme 12 rods The High Bum-up Effects Programme 81 rods WWER rods from Kola-3 32 rods Rods from the TRIBULATION programme 19 rods Studsvik INTER-RAMP BWR Project 20 rods Studsvik OVER-RAMP PWR Project 39 rods Studsvik SUPER-RAMP PWR Sub-Programme 28 rods Studsvik SUPER-RAMP BWR Sub-Programme 16 rods Studsvik DEMO-RAMP I - BWR 5 rods Studsvik DEMO-RAMP II - BWR 8 rods Studsvik TRANS-RAMP-I BWR 5 rods Studsvik TRANS-RAMP-I1 PWR 6 rods Studsvik TRANS-RAMP-IV BWR 7 rods REGATE FGR and cladding diameter during and after a transient 1 rod at Bu - 47 MWd/kg CEA/EDF/FRAMATOME Contact I & 2 3 rods AEAT-IMC NFB 8 and 34 22 samples CEA/EDF/FRAMATOME PWR and OSIRIS ramped fuel rods 4 rods CENG defect fuel experiments 8 rods CANDU elements irradiated in NRU 36 rods Siemens PWR rods irradiated in GINNA 17 rodlets CEA failed PWR rods irradiated in SILOE: EDITH-MOX 01 I rod CNEA six power ramp irradiations with (PHWR) MOX fuels 5 rods BN GAIN (U,Gd)0 2 fuel 4 rods INR Pitesti - RO-89 and RO-51 CANDU fuel type 2 rods HRP IFA-597.3 rods 7, 8 and 9 (cladding degradation, FCT, FGR 3 rods at Bu o 60 MWd/kgUO2 HRP IFA-534.14 rods 18 and 19 ( EOL FGR and pressure, grains 2 rods size of 22 and 8.5 micrometers and Bu >52 MWd/kgUO 2 Total 446 cases 4.2 Format of files The criterion adopted for the file format was one primarily of simplicity. It was considered that users should be able to read the files independent of commercial software. It was recognised that the majority of 6

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NEA/NSCIDOC(2003)12 codes were written in FORTRAN and therefore all files are in simple ASCII format for easy interrogation by file editors. Even text files are of this format despite the enforced limitations imposed by this approach.

The adopted ASCII format does not preclude reformatting into a commercial database system sometime in the future if so desired. All databases have the following common elements:

• Summary file. This is a text file which describes the purpose of the experiment or test matrix and the scope of the data obtained.

• Indexfile. This is also a text file and lists all file titles and gives a brief summary of their contents.

• Pre-characterisation.This includes information on the fuel pellets and cladding used, their manufacturing route, dimensions and chemical composition including impurities. For the fuel, this is augmented by details of enrichments, porosity distribution, re-sintering test data and microstructure.

For the cladding, additional information includes mechanical properties, corrosion characteristics and texture as and when available. Details of the fuel rod geometry include: relevant dimensions, fuel column length, weight, fill gas composition and pressure. Details of reactor irradiation conditions.

• Irradiationhistories. All histories are in condensed form with care to ensure that all important features are preserved. Where there was a significant axial power profile the history is provided in up to 12 axial zones. For each time step, the data are provided as: time, time increment over which power was constant, clad waterside temperature and local heat rating for the prescribed number of axial zones. Information is provided to calculate the fast neutron flux and its spatial variation if the information is available.

• In-pile data. When applicable, separate files tabulate data from in-pile instrumentation as a function of variables such as: time, bum-up or power. For example, IFA432 fuel rods were equipped with centreline thermocouples and cladding elongation detectors. Files were created tabulating temperatures at constant powers of 20 and 30 kW/m as a function of local bum-up throughout life. At several times during the irradiation, at approximately 5 MWd/kg UO2 intervals, temperature is tabulated against local power during slow ramps. In this case, the data are reproduced directly from the original files where signals were logged every 15 minutes. A similar procedure was adopted for clad elongation measurements; during short periods of variable power, 200-5000 hours duration, elongation is tabulated against rod average power showing cases where there was pellet-clad mechanical interaction.

• PIE data. Where such examinations were made, data are recorded either in tabular or text form.

Dimensional data include axial and diametral dimensions before and after irradiation, and post irradiation ridge heights where available. Data on fission gas release include rod averaged values obtained by puncturing and mass spectrometry, local values from whole pellet dissolution and across pellet spatial distributions as measured by gamma scanning, EPMA and XRF. Before and after porosity and grain size distributions are given as is the radial position for the onset of grain boundary porosity when such measurements have been made from metallographic examination.

All data files are held centrally on the OECD/NEA Data Bank computer system from which all files are dispatched. This single source for distribution is necessary for Quality Assurance purposes, particularly for the tracking and release of upgraded or corrected files. With the OECD's experience of data bank management, this arrangement ensures long-term availability of the service.

Often PIE data are only available in graphical form and photomicrographs which are difficult to preserve in the current ASCII file format. For this reason, the possibility of scanning figures and photographs for storage and retrieval using the medium of the CD ROM was investigated. As a result, all reports available 7

NEAINSC/DOC(2003)12 for compiling the datasets to date have been scanned and copied onto a single CD-ROM. The files are all in PDF format and can easily be read and printed using a wide range of software.

5. INTERNATIONAL WORKSHOPS 5.1 Thermal Performance The initiative to hold international workshops commenced with Thermal Performance of High Bum-up LWR Fuel held 3-6 March 1998 at CEA Cadarache. The workshop was co-organised by the OECD NEA in co-operation with the IAEA and hosted by the Departement d'dtudes des Combustibles (DEC). The meeting was attended by 66 experts from 19 countries, and four international organisations.

The first two sessions were devoted to Fuel Thermal Conductivity Data and Thermal Conductivity Modelling. This was followed by a first panel discussion, chaired by Hubert Bairiot and Louis-Christian Bernard addressing Solving and Emerging Thermal Conductivity Questions. The third session covered Fuel Clad-Gap Evolution Modelling followed by a second panel discussing Gap Evolution and Heat Transfer Questions, chaired by Gary Gates and Marc Lippens. The two final sessions covered Experimental Databasesand Advances in Code Development on ThermalAspects.

A final paneL chaired by J.A. Tumbull and Daniel Baron, reviewed the conclusions of the two panels on thermal conductivity and gap conductance and summarised - after discussion among all participants - the key conclusions of the seminar. These conclusions are appended herewith:

Fuel thermalconductivity Fuelthermal conductivity correlations The initial thermal conductivity correlations are well established for almost all fuel oxides loaded in commercial reactors. The influence of parameters such as temperature, stochiometry, plutonium content and gadolinium content are relatively well known. Some models are able to account for all these parameters simultaneously.

MOXfuels The non-homogeneity of MOX fuel can be accounted for, using mathematical homogenisation techniques.

However, even for high Pu content the fuel thermal conductivity is close to the uranium oxide fuel conductivity providing O0M close to 2.000. Participants seemed to agree on a degradation of 4 to 5% per 10% Pu. However, it was shown that a deviation from stochiometry has a stronger effect.

The burn-up effect The burn-up effect on fuel thermal conductivity has been assessed up to 80 MWd/KgU, thanks to in-pile centreline temperature measurements and out-of-pile thermal diffusivity measurements. There is some agreement between these two methods.

Cp variation with burn-up Analysing shut-down temperature records in the Halden reactor, a slight increase of the Cp with bum-up is likely to happen. Nevertheless the discussion outlined that the Cp variation with bum-up could be neglected. The experimental data expected in the near future will certainly allow checking this assumption.

Heat transferin the rim The rim thermal conductivity evolution is not yet clearly known. The thermal degradation due to the onset of numerous micrometer porosities could be balanced by the cleaning of the matrix subsequent to this porosity build-up. In order to investigate the net effect of the rim structure, samples irradiated to high burn-8

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NEA/NSC/DOC(2003)1 2 up and temperatures Tower than 800 K are required. Other questions are still open about the rim fission gas release and the rim volume variation during the porosity build -up.

Furtherneeds A lack of data is identified on the bum-up degradation at temperatures higher than 1800 K A decrease of the conductivity improvement due to the electronic heat transport is likely to happen with bum-up. This effect was already observed when increasing the gadolinium content.

Gap conductance Pelletfragmentationand relocation The stochastic nature of this phenomenon prohibits analytical modelling of gap conductance. Reliance must be placed on benchmarking models against in-pile data. There is no general consensus as to a definitive formulation for gap conductance. Fortunately a large database exists and providing this is employed, there are limited difficulties in formulating an adequate gap conductance model which correctly reflects the effects of gap dimension, fill gas composition and pressure.

Surface roughness In principle this should be an important contributor to gap conductance by its effect of inhibiting heat flow.

In practice, at both the beginning of line and at high burn-up, there appears to be a little effect on heat transfer between the fuel pellet and the cladding.

Formationof innerclad oxide layer At high bum-up, when the fuel and clad have been in intimate contact for some time, an inner oxide layer is formed, 6-10 microns thick. It has been shown that this has a complex structure, which can include ZrO-, uranium and caesium, depending on the heat rating and bum-up. It is supposed that its contribution to gap heat transfer is small and if anything, beneficial, as it would tend to eliminate effects due to surface roughness and misaligned pellet fragments.

Closedgap conductance There is evidence to suggest that under these conditions, heat transfer is good and substantially independent of the fill gas composition and pressure, since it corresponds to a solid bound between fuel and cladding. There is some debate as to whether or not the conductance depends on interfacial pressure.

5.2 Fission Gas Release The second workshop in the series was the Fission Gas Release Workshop, held at CEA Cadarache, 26-29 September 2000, organised and hosted under the same arrangement as before. The seminar was attended by about 100 participants from 24 countries, representing 46 organisations. The objective of this Seminar was to emphasise the more recent developments, from both the experimental and modelling points of view.

The areas covered by the seminar included: diffusion coefficient of rare gases including helium, properties of bubbles, the effect of irradiation on re-solution and diffusion, FGR as a function of operating conditions, modelling and code validation.

The meeting was divided into four sessions which occupied three days of presentations. The session titles were: Feedbackfrom Experience, Basic Mechanisms, Analytical Experiments, Industrial Modelling and Software Packages. Summaries of each session were prepared by session chairpersons and were appended to the final report. On the fourth and final morning, there was a session entitled: Panel Discussion:

Conclusions and Perspectives where each chairperson gave an oral presentation of their sessions which 9

NEA/NSC/DOC(2003)1 2 was discussed among the participants. In view of the 'maturity' of the subject, each chairperson was requested to address the three questions:

- What do we know, what is new?

- What don't we know, what are the uncertainties and what remains unclear?

- What initiatives should be implemented in the future?

From the first session, Feedbackfrom Experience, it was concluded that there was an enhancement of fission gas release at high bum-up, >50 MWd/kg, both in steady state operation and in transient overpower, but there was no accepted phenomenon to explain this. One result of this was that the well established Halden 1% FGR criterion relating fuel centreline temperature to bum-up overestimated the onset temperature at these levels of burn-up. That is, the temperature for the onset of release at high bum-up was lower than predicted by this criterion. It has been observed that high bum-up fuel contains a region of restructured fuel microstructure close to the pellet periphery, the so-called 'rim' structure. It was tempting to use this as a reason for the enhanced release, but a direct correlation could not be made. Indeed a measure of the ratio of released krypton and xenon isotopes implied that their origin was the hot central regions of the fuel and not the cold plutonium rich restructured region in the pellet rim. The implication of this observation was that the restructured rim served mainly to increase the thermal resistivity of the fuel, thus increasing fuel temperatures.

A discussion of this dilemma continued into the second session on Basic Mechanisms. Investigations showed that this rim structure consisted of a refinement of the original - 10 micron diameter grains into

-0.1 micron diameter grains and the introduction of a micron size population of bubbles which could account for a swelling up to -10 volume %. In addition, the fission gas concentration in the matrix fell to a low level. The impression of most workers was that the shortfall in fission gas resided in the bubbles and had not been released from the fuel; however the exact distribution of gas remained uncertain. The fission gas release from the rim was not more than 15 to 20 %, showing a high retention capacity of such a restructured material. The mechanism of rim formation was not clear, one suggestion was that the build-up of irradiation damage caused the grain refinement and that this was followed by the collection of gas into the porosity. Alternatively, it has been postulated that the porosity was the first to form and that the sub-micron grains were nucleated from the pore surfaces. It remained a matter of debate whether this rim restructuring was beneficial or detrimental.

Several presentations addressed FGR from the rim both in steady-state and transients such as RIA, concentrating on mechanisms and experimental observations. Measurement of krypton and xenon isotopic ratios in released gas was clearly a valuable technique to determine the relative contribution of the rim region to FGR, and application to RIA tests had proved informative. It was concluded that although we knew details of this new high bum-up structure at the micron level, it was clear that a goal for the future was to understand it at the nanometer or atomic level, and that there was scope for simulation studies using high energy accelerators to reproduce the restructuring without the constraints of time and radioactivity.

A major contribution to the release process is known to be single gas atom diffusion and quite satisfactory models could be constructed using this, along with irradiation resolution from grain boundaries. Many models employed the diffusion coefficient formulated by Turmbull, White and Wise as described in the IAEA Preston Meeting in 1988. Here the authors gave a two-term diffusion coefficient with a third low temperature term to be applied close to surfaces and used for short lived radioactive species. There was much discussion on this topic in the meeting and it transpires that White has an alternative description using only the two high temperature terms with a 'fractal' treatment of the surface to account for the difference in kinetics between long and short lived species. Subsequent to the meeting, this has now been published in the external literature.

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NEA/NSC/DOC(2003)12 An extensive study of retained caesium and xenon using Electron Probe Micro Analysis (EPMA) carried out at ITU showed that systematically, caesium diffusion was some 3 times slower than for the rare gases.

This was an important result since in many accident calculations it is assumed that their diffusivities are comparable; the assumption therefore is reassuringly pessimistic.

In addition we know from electron microscopy that intragranular bubbles are formed within grains and intergranular bubbles and tunnels along grain boundaries. Theoretical studies of krypton atoms in the U02 matrix concluded that both krypton and xenon were insoluble thus casting doubt on the existence of

'thermal resolution' as a means of moderating absorption of single gas atoms into intragranular bubbles. In the absence of intragranular bubble mobility, such a mechanism was used in several models in order to predict substantial release in transients. It was agreed that observations support only slow or limited random migration of small bubbles, but a 'directed' movement towards grain boundaries due to a vacancy concentration gradient was an interesting proposal, (Evans, UK). If this mechanism had a significant contribution, the resulting microstructure would be quite distinctive and amenable to future testing. Topics which required further attention was the re-solution of gas atoms from grain boundary bubbles, particularly for large grain fuel where the grain size was greater than the fission fragment range and the response of grain boundary porosity in fast transients.

It was clear that there was much more information available on these features than currently accessible through the open literature. There was general support for a more open distribution of information and all participants were urged to publish as much information as possible in the interest of a general improvement in models and their application to safe reactor operation. It was noted that although the level of fundamental research in many establishments had been reduced, there were still original and comprehensive works going on in ITU, for example, improving the investigation techniques (microhardness tests, lattice parameter measurements). Also, there was a refreshing intake of young engineers and scientists within CEA Cadarache. The participants looked forward to publication of new and original work from these quarters.

As a digression from fission gas, it was clear that helium generation in long term storage of high bum-up fuel can pose a significant problem. In particular for high burn-up MOX. Further data were required on low temperature helium diffusion coefficients.

The session entitled Analytical Experiments started with two presentations describing a novel method of determining the disposition of gas between the matrix and the grain boundaries. Such studies complement those performed with X-Ray Fluorescence (XRF) and EPMA and provide vital information for modellers.

The techniques still need refining, particularly regarding the effect of small intragranular bubbles on the results, but promise an evaluation of the grain boundary capacity for gas prior to and during interlinkage.

As mentioned previously, Halden has obtained data from high bum-up fuel which suggests that the temperature for the onset of FGR is lower than previously expected. It was clear that further data were required to substantiate this observation. In addition, there was a need to obtain further data on MOX fuel to compare it with U02 fuel performance under identical conditions.

Enhanced release at high bum-up can cause the rod internal pressure to exceed the coolant pressure. The effect of this requires attention. So far, experimental data suggests that the gap does not reopen by clad creep-out and positive feedback does not occur. However, the data are sparse and further experiments are necessary.

Both mechanistic and empirical approaches to FGR modelling were presented in the session: Industrial Modelling and Software Packages. It was clear that empirical modelling can be very valuable, but is limited in applicability and is essentially only valid within the confines of parameters and irradiation conditions covered in the database on which it is developed. Modellers were also cautioned about 11

NEA/NSC/DOC(2003)12 employing multiple mechanistic models and expecting to obtain good predictions by using appropriate fitting parameters. There is a case for independent assessment of mechanistic models and their supporting data before consideration for inclusion within fuel performance codes. The requirement for further data comparing U0 2 and MOX behaviour was re-affirmed in a paper by Struzik of CEA which showed that there was a significant difference in behaviour in ramp test above 30 MWd/kg. Below 30 MWd/kg the differences in behaviour could be explained in terms of the radial power profile but at higher bum-up FGR and swelling in MOX was greater than expected.

In conclusion there was unanimous agreement that the Workshop had been a success with an excellent range of presentations which initiated extensive and in-depth discussion. It was clear that the topic of fission gas release modelling was in a mature state with agreement about most of the important mechanisms contributing to the phenomena. High bum-up and MOX behaviour are the current challenges where there is an incomplete understanding, but the meeting produced some good suggestions as to where future work could be focussed. Participants were urged to publish their work in the open literature so that all could benefit towards the goal of safe operation of commercial reactors.

5.3 Pellet-Clad Interaction The third and final workshop is to be on Pellet Clad Mechanical Interaction as this was deemed the least advanced area of modelling. The date for this meeting is scheduled for 8-11 March 2004, to be held at CEA Cadarache. Issues and topics are described in Annex 2.

References [3,4] concern the proceedings of these two seminars.

6. FUTURE ACTIVITIES 6.1 The IFPE Database Now that the IFPE Database is firmly established, there is a continuing need for its support and improvement. Indeed, on the assumption that the Database will continue to be maintained by the NEA, several organisations are using the database as the sole source for their code development and validation.

Further data will be added to the Database in the future including the following for which agreement for release has already been obtained:

• BR-3 High Bumup Fuel Rod Hot Cell Program (DOE/ET 34073-1, Vol. 1 & 2)

• RIS0-I experiment

• IFE/OECDAHRP FUMEX 1-6

• Zaporoshye VVER1000 fuel behaviour data (4-8 cycles, Bu >)50 MWd/kgUO2

• IMC (UK) swelling data from ramping CAGR U0 2 fuel

• HATAC FGR and cladding diameter during and after a transient at Bu - 49 MWd/kg

• B&W Extended-Burnup Gadolinia Lead Test Assembly Experiments

• HRP He/Ar/Xe gas flow, Nb doped fuel, IFA-504

• VNIINM ramp data from WWER440 up to 4 cycles and WWER-1000 up to 3 cycles

• IFA-508 and IFA-515 conducted by JAERI at HRP - PCMI behaviour data on different cladding thickness by means of diameter rig.

• CEA failed PWR rods irradiated in SILOE; EDITH-3 and EDITH- MOX 02

• Qinshan Phase I fuel in-pile experiment A meeting of the Expert Group on IFPE is scheduled for September 2003 at Halden. A further meeting is scheduled in connection with workshop on pellet-clad interaction, in March 2004.

12

NEA/NSC/DOC(2003)12 Based on the success of the FUMEX-1 Code Comparison Exercise, the IAEA have agreed to launch an initiate project, FUMEX-2. The global outcome of the FUMEX 1 programme was very encouraging, with a high degree of participation from member countries. AlU agree that it was a worthwhile exercise, and that the cases chosen were stringent tests of models and codes performances. The exercise was useful in demonstrating the strong points of the codes as well as highlighting deficiencies where improvements were necessary. As a consequence, most of the codes underwent some development during this programme. It was also apparent that many of the codes have been developed on a limited database and that the FUMEX I cases provided a valuable addition. However developments pointed out that limited knowledge have been gained in the extended bum-up range (above 50 MWd/kg). Bum-up extension corresponds to a general trend in the field of fuel management, and reliable prediction of fuel behaviour at high bum-up constitutes a basic demand for safe and economic operation of nuclear fuel.

This was the basis for launching FUMEX-2 which started in December 2002 and should be completed in 2006. Central to the project are selected cases in the IFPE Database. This will help promote the use of the Database and at the same time, will generate feedback and peer review from users, thus assisting the Quality Assurance of the datasets within the Database. The cases chosen for FUMEX-2 are tabulated below, with cases already in the Database shown in bold type.

It is important therefore that the NEA provide additional support to maintaining and improving the Database over this period in addition to the introduction of further data as they become available.

At the joint International Fuel Performance Experiments Database (IFPE) and Fuel Modelling Exercise (FUMEX-IH) meeting organised by the LAEA and OECD/NEA, held at IAEA Headquarters from 16-19 December 2002 enhanced co-operation was established in this field. The advantages of this co-operation are:

• Exposure of code developers to a wide ranging database

• Assistance in qualification of the IFPE database, correction of errors and detection of missing data, brought about by use of the database for comparison of predictions with data by a large number of workers.

The discussion, conclusions and recommendations regarding the IFPE Database can be summarized as follows:

• The interest of the users, not only at this meeting, showed that a continued need exists for the service and function provided by the IFPE

• Concerns expressed as to the proper functioning of the IFPE were related to:

- Funding, resources to maintain/extend/qualify the data base

- Procedures for feedback and notification of updates

• Proposals for improvements included:

- Extension of the description chapters to include More information on uncertainties if available Models (as such, or correlations for older materials); note however that this suggestion was not unanamously agreed Data assumed by users to fill gaps in information provided User experience in general Alternative (un)condensed histories

• IFPE should concentrate on fuels having rod geometry, but inclusion.of non-standard fuel variants with this geometry are welcome (e.g. MOX, disprosium doped fuel etc.).

Separate effects experiments not closely linked to an irradiation history will be considered on a case by case basis 13

NEA/NSC/DOC(2003) 12 No Case identification Measurements made for comparison

1. Halden IFA 534.14, rod 18 EOL FOR and pressure, grain size 22 pin, Bu 52 MWd/kgUO 2

2. Halden IFA 534.14, rod 19 EOL FGR and pressure, grain size 8.5 gm, Bu 52 MWdlkgUO 2

3. Halden IFA 5973, rod 7 Cladding elongation, at Bu = 60 MWd/kgUO 2

4. Halden IFA 597.3, rod 8 FCT, FGR at Bu s 60 MWd/kgUO2

5. Halden IFA 507, TF3 Transient temperature during power increase

6. Halden IFA 507, TF5 Transient temperature during power increase

7. REGATE FGR and clad diameter during/ after transient at Bu m47 MWdlkg 8 HATAC FGR and clad diameter during and after a transient at Bu w 49 MWd/kg

9. Kola-3, rod 7 from FA222 FGR, pressure and creepdown at Bu - 55 MWdk9gUO 2

10. Kola-3, rod 52 from FA222 FGR, pressure and creepdown at Bu - 46 MWd/kgUO 2

11. Kola-3, rod 86 from FA222 FGR. .u pressure and creepdown at 44 MWd/cgUO 2

12. Kola-3, rod 120 from FA222 FGR, pressure and creepdown at Bu 50 MWd/kgUO2

13. Riso-3 AN2 Radial distribution of fission products and FGR-EOL, Bu - 37 MWd/kgU0 2

14. Riso-3 AN3 FGR and pressure-EOL, FCT, Bu %37 MWd/kgUO 2

15. RIso-3 AN4 FGR and pressure-EOL, FCT, Bu X 37 1IWd/kgUO2

16. HBEP, rod BK363 FGR-EOL, Bu - 67 MWd/kgU02

17. HBEP, rod BK365 Fission products and PU distribution, FGR-EOL, Bu - 69 MWd/lkgUO2

18. HBEP, rod BK370 Fission products and Pu distribution, FGR-EOL, Bu - 51 MWd/kgUO 2

19. TRIBULATION, rod BNI/3 Pressure, FGR, cladding creepdown, Bu - 52 MWd/kgUO 2

20. TRIBULATION, rod BN1/4 Pressure, FGR, cladding creepdown, Bu = 51 MWd/kgUO 2

21. TRIBULATION, rod Pressure, FGR, cladding crecpdown, Bu - 51 MWd/kgUO 2 BN3/115

22. EDF/CEAIFRA, rod H09 Fission products and Pu distribution, FGR-EOL, Bu - 46 MWd/kgUO 2

23. Kola-3 + MIR test Temperature during ramp, FGR-EOL, Bu - 55 MWd/kgUO 2 24 Kola-3 + MIR test Pressure-EOL, Bu - 55 MWd/kgUO2

25. RIA to be specified (real data or simplified case)

26. LOCA to be specified (real data or simplified case) 27 Simplified case (I) Temperature vs Bu for onset of FGR (2a) FGR for constant 15 kW/rn to 100 MWd/kgU (2b) FGR for 20 kW/m at BOL decreasing linearly to 10 kW/m at 100 MWd/kgU 7c) FGR for idealized history supplied by BNFL (2d) FGR for idealized history supplied by FANP (3a) FGR for CANDU idealized history

__ (3b) FGR for CANDU idealized history

• Note: high priority cases are in bold 6.2 Other Activities Based on the survey of member countries and their safety related R & D activities, the NEA formed a good idea of priorities for current and future work. They are active therefore in supporting these activities particularly related to high bum-up studies, the introduction of MOX into commercial reactors and the issues arising in the use of weapons grade fuel, W-MOX in order to reduce the inventories held by both the US and Russia.

Two workshops were held on "Advanced Reactors with Innovative Fuels (ARWIF)" in 1999 and 2001.

Fuel behaviour issues were also discussed. For details see Ref. [6,7]. The next ARWIF workshop will be held in September 2004 at ORNL, Oak Ridge, TN, USA.

14 A/

. I NEA/NSC/DOC(2003)12 The commercial nuclear industry is currently at a very critical stage. The concern over Global Warming has focussed attention on fossil fuel usage and its emissions. As a result, many countries are extending the life of their current installed plant as well as showing a renewed interest in building new nuclear generators. It is clear, however, that this is unlikely to be in the form of current designs. There are several new designs, some are extensions of the current Generation II LWR, but some are more innovative and as such are classified under the generic Generation IV designs. Common features of these new designs include passive safety, high thermodynamic efficiency and proliferation resistant fuel. Before entering full scale commercial operation, these will require extensive qualification both in their design and in the materials used in their construction and operation. It is a goal for the NEA to assist in the quest for improved nuclear design and installation by supporting R & D into these new concept and new fuel types and cladding alloys.

7. REFERENCES

[1] Scientific Issues in Fuel Behaviour, a report by an NEA Nuclear Safety Committee Task Force, OECD Document, January 1995

[2] Review of Nuclear Fuel Experimental Data, fuel behaviour data available from IFE-OECD Halden Project for development and validation of computer codes, J. A. Turnbull, OECD Document, January 1995

[3] Thermal Performance of High Bum-Up LWR Fuel Seminar Proceedings, Cadarache, France, 3-6 March 1998, 389 pages, ISBN: 92-64-16957-1, OECD 1999

[4] Fission Gas Behaviour in Water Reactor Fuels Seminar Proceedings, Cadarache, France, 26-29 September 2000, 564, pages, ISBN: 92-64-19715-X, OECD 2001

[5] Fuel Safety Research in NEA Member States, a compilation of member contributions, February 2002

[6] Advanced Reactors with Innovative Fuels: (ARWIF'98) Workshop Proceedings, Villigen, Switzerland, 21-23 October 1998, OECD 1999, ISBN: 92-64-17386-2

[7] Advanced Reactors with Innovative Fuels (ARWIF-2001), Workshop Proceedings, Chester, United Kingdom, 22-24 October 2001, OECD Paris 2002, ISBN 92-64-19847-4 Text prepared by: J.A. Tumnbull Reviewed by: E. Sartori 9 May 2003 15

NEA/NSC/DOC(2003)1 2 Annex 1 OECD/NEA Nuclear Science Committee (NSC)

Expert Group on Scientific Issues of Fuel Behaviour (TFSFB)

SCOPE and OBJECTIVES

• Scope The Expert Group deals with the status and trends of scientific issues of fuel behaviour.

• Objectives

- Compile high quality experiments for the International Fuel Performance Experiments (IFPE) database. Emphasis is given to high bum-up in water reactors under normal operating conditions.

Priority goes to completed international programmes, the data of which would otherwise be lost, but also to data released from national programmes.

- Co-ordinate the qualification of these data through review and by organising user group meetings. Take initiatives so that it may be adopted as an international standard.

- Co-ordinate computer code validation and benchmark studies.

• Work Plan

- Organise co-ordination meetings combined with IFPE database user group meetings every 12-18 months.

- The IFPE database will be consolidated during the next two years through updating and addition of new experiments.

- Seminars and Experts' meetings will be held as needed to address high priority issues for validating the modelling of phenomena of particular concern (thermal performance, fission gas behaviour, pellet-clad mechanical interaction, etc.)

• Co-operation This activity will be carried out in co-ordination with the IAEA, the NEAICSNI Special Expert Group on Fuel Safety Margins and the OECD/IFE/ Halden Reactor Project.

16

. J 1- I NEA/NSC/DOC(2003)12 Annex 2 Issues and Topics for the Seminar on Pellet Clad Interaction The aim of this third seminar is to draw up a comprehensive picture of our current understanding of pellet clad interaction and its impact on the fuel rod, under the widest possible conditions.

Pellet-cladding mechanical interaction (PCMI)

I Two contrary mechanisms lead to a mechanical interaction between pellet and cladding, namely, cladding creepdown and fuel swelling. Once the fuel-clad gap has closed, after 2-3 cycles, the compressive stress experienced by the cladding due to the system pressure is reversed to a tensile stress induced by continued fuel swelling. From this point, the fuel swelling is the main factor contributing to radial and axial constraints on the cladding. Enhanced clad strain is observed in the region of the pellet ends, where ridging is observed caused by the 'wheatsheaf' shape adopted by the underlying cracked pellet due to the influence of the large radial temperature gradient. However, clad strain is reduced in the case of a compliant and dished fuel pellet. In these cases, clad deformation is reduced as fuel swelling and thermal expansion is accommodated by dish filling. This requires remnant volume to accommodate the expanding pellet and a relatively rapid irradiation creep rate of the fuel material. At high bum-ups, it is possible that fuel swelling could exhaust the remnant volume offered by pellet dishing. In this event, and in the absence of other effects, it may be expected that clad deformation would again accelerate, thus increasing the propensity of clad failure.

Pellet-cladding chemical interaction (PCCI)

After the fuel-clad gap has been closed for a period of time, it has been observed that a chemical bond forms between the internally oxidised cladding and the fuel. The composition can be basically described as a mixed Zr-U (Pu)-oxide, containing up to 34 w/o Cs according to EMPA results. The beneficial aspect of this is that the thermal conductivity of the fuel-clad interface is improved over that of the original gas gap between the two components. This, in turn, leads to stabilised fuel temperatures and a reduction in thermal feedback effects caused by fission gas release. The implication to PCMI depends largely on the mechanical properties of this bonding layer. For strong mechanical bonding, a detrimental aspect is that it can lead to strain concentration in the cladding ligaments situated immediately adjacent to pellet cracks. In the event of pellet expansion in an over-power transient, strain is concentrated in these ligaments and premature failure can occur by ductility exhaustion, particularly in the vicinity of pellet ends. For weak bonding, and soft mechanical properties, the bonding layer can act as a lubricant, thus minimising stress and strain concentrations in the cladding. Thus a full understanding of this bonding layer is necessary to predict the likely outcome of fuel swelling and pellet expansion in transients. It is possible that the properties of the bonding layer are determined by its composition and structure, and this in turn could be a function of the irradiation history by way of temperature and fission product release and migration.

High Burn-up Fuel Restructuring At a pellet average burn-up 260 MWd/kg, the local bum-up within -100 gm of the pellet surface is increased by a factor 2 or 3 depending on the flux spectrum of the reactor. At the same time and in the same location, ceramography of discharged fuel has shown the presence of a large population of pores some 0.5 - I plm diameter and a refinement of the original U0 2 grain size from typically 10 pm to 0.1 - 0.3 17

.. U NEA/NSC/DOC(2003)12 im diameter. This small grain sized re-crystallized structure is best observed by Scanning Electron Microscopy of a fractured specimen where the surface morphology takes on a distinctive 'cauliflower' like appearance. Because of its location at the pellet surface, this has become known as the 'rim effect'.

However, this is mis-named and should be referred to as a the High Burm-up Structure or HBS.

Observations have shown that this 'rim' has mechanical properties different from those of the original pellet. In particular, it appears to be more resistant to crack propagation and deforms more easily under applied stress. Like the bonding layer descriped earlier, the properties of this rim will have an important influence on PCMI1 at high bum-up. With the reported properties, in this case, it should provide an alleviation of PCMI under normal and conditions of operational transients. However, there is a suspicion that its properties may change during rapid transients leading to enhanced PCMI thus contributing to clad failure and fuel dispersion during RLAs. This issue is the subject of substantial R&D programmes throughout the world using pulse reactors to generate the fast transients proposed as the event initiators.

Stress Corrosion Cracking Initial research identified iodine as the most probable active species for stress corrosion cracking (SCC). Subsequent observations indicated that this type of failure was generic to zirconium alloy clad fuel that was subjected to significant power increases. Today, it is grouped with other types of pellet-clad-interaction (PCI) failure mechanisms observed to occur in zirconium alloy fuel cladding.

Power changes are inevitable during the operation of nuclear power plants; hence, the possibility of a large number of fuel failures ensured a large allocation of resources to better understand and solve this problem. Cures were attempted based on the characteristics learned about the SCC process in zirconium alloys. Three type of solutions were widely applied: internal coatings (graphite or siloxane), barrier cladding (pure zirconium) and, equally important, a restriction on the reactor power changes and refuelling schedules. The incorporation of these solutions greatly reduced the frequency of SCC failures. Today, the number of SCC failures is well below the number of failures due to coolant debris.

In the event of a primary failure, and ingress of coolant into the fuel rod, secondary failures may occur due to excessive local clad hydriding and additional clad stress due to fuel oxidation. A scenario has been envisaged where the formation of a secondary failure results in a progressive deterioration of the cladding, leading to extended axial cracks along the length of the fuel rod. Such cracks have been observed in operating commercial BWRs.

Current Manufacturing Remedies to avoid failure by PCMI Cladding without the liner is manufactured to achieve an optimised combination of grain texture and grain size, which improves resistance to pellet-to-clad interaction (PCI). Cladding with pure zirconium on the inner surface provides additional protection against PCI. Iron-enhancement of the liner precludes the risk of excessive activity release in the unlikely event of cladding tube failure. Special low-temperature annealing processes used during manufacturing increase the resistance to both nodular and uniform corrosion.

Rods with niobia-doped or chromia-doped fuel pellets and pellets doped with different alumnino-silicates, the so called 'soft pellets', have been manufactured and supplied as possible PCI remedy fuel. It can withstand considerably larger power ramps without failure than is the case for standard undoped fuel.

Enhanced creep of the doped fuel into large voids associated with controlled fabrication porosity, as well as into cracks and pellet dimples (dishes), was confirmed to be the mechanism responsible for the improvement. These references show that the presence of additives to standard fuel can affect the swelling and creep properties; thus to proceed with modelling of these and other variants, e.g. gadolinia-doped fuel 18

NEA/NSC/DOC(2003)12 and MOX fuel, requires further dedicated experiments and modifications to the established principles of fuel property models.

Scope and structure of the seminar:

1. Opening Session and Industrial Goals

- Operating constraints in relation to PCI management

- Industrial approach to preclude PCI hazards

- Objectives of current R&D efforts

2. Fuel material behaviour under PCI situation

- Mechanical properties, cracking of standard and remedy fuel, e.g., improved microstructure, doped material, MOX etc.

- Properties of fuel-clad bonding layers and the HBS structure in high bum-up pellet rims

- Contribution to PCI of the effects of fission gases (swelling, release...)

- Experimental basis in support of fuel behaviour assessment

3. Cladding mechanical behaviour relevant to PCI

- Standard cladding behaviour activated in PCI situations

- Behaviour of advanced cladding alloys with respect to PCI

- Dependence of cladding behaviour on hydriding

- Stress-corrosion cracking: experimental facts and modelling

- Experimental basis in support of clad behaviour assessment

4. In-pile Rod behaviour

- PCI tests under operational and design basis transients

- Simulation on re-irradiated fuel in MTRs

5. Modelling of the mechanical interaction between pellet and cladding

- Hypothesis of numerical simulations

- Code performance with regard to experimental data and prospect for improvement

- Needs for extension of the data bases 19