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NEA-1599 IFPE/FUMEX-I.

IFPE/FUMEX-I, Data from OECD Halden Reactor Project for FUMEX-1 (Fuel Modelling at Extended Burnup)

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1. NAME OF EXPERIMENT:  IFPE/FUMEX-I.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
IFPE/FUMEX-I NEA-1599/01 Arrived 30-SEP-2003

Machines used:

Package ID Orig. computer Test computer
NEA-1599/01 Many Computers
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3. DESCRIPTION

This set contains the experimental data from the OECD Halden Reactor Project made available for the FUMEX-I exercise (Fuel Modelling at Extended Burnup) carried out under the auspices of the IAEA.
  
The six cases are:
  
FUMEX 1:This data set represents the irradiation of production line PWR type fuel underbenign conditions. Temperatures remained low but increased slightly with burnup.
  
FUMEX 2:This was a small diameter rod designed to achieve rapid accumulation of burnup. Temperatures were estimated to remain low. The internal pin pressure was measured in-pile and an assessment of FGR was also provided by PIE.
  
FUMEX 3: This case consisted of 3 short rods equipped with centreline thermocouples each with a different gap and fill gas composition. After steady state irradiation to ~30 MW.d/kg UO2, they were given a severe increase in power (power ramp).
  
FUMEX 4: Two rods filled with 3 bar He and 1 bar He/Xe mixture were irradiated to ~33 MW.d/kg UO2. Both rods experienced a period of increased power part way through the irradiation.
  
FUMEX 5: The test case comprised a single rod base irradiated at low power to 16 MW.d/kg UO2 with a power ramp and a hold period at the end of life. The main purpose of this case was to assess pellet clad mechanical interaction (PCMI) and fission gas release (FGR) under ramp conditions.
  
FUMEX 6: Two rods were base irradiated at low power. The rods were refabricated to include pressure transducers. Rod internal pressure was monitored during power ramps, one fast, one slow.
   
Irradiation conditions in the Halden Reactor:
  
The Halden Reactor is a heavy water moderated and cooled boiling water reactor. The nominal operation conditions are 240 deg.C coolant temperature and a corresponding saturation pressure of 34 bar. These conditions imply decreased uncertainties for some effects with an influence on experimental results and data evaluation, namely:
- cladding creep-down is very small,
- cladding oxidation can be practically neglected,
- boiling conditions can be assumed for the entire length of the rod and consequently the clad surface temperature is known with high accuracy.
  
The reactor is operated with three major shut-downs per year for loading/unloading of driver fuel and experiments.
  
In-core instrumentation and experimental techniques:
  
In-core instrumentation combined with suitable experimental techniques and test designs is the key to meaningful results for model development and validation. While PIE ascertains the state existing at the end of irradiation, in-pile instrumentation gives information on how phenomena developed during in-core service. The ten instrumented rods of the FUMEX cases thus provide a good basis for studying key parameters of fuel modelling.
  
A general overview of instrumentation and experimental techniques applied in the Halden Project fuel testing programmes can be found in Refs [19-21]. Those employed for producing FUMEX data are repeated below.
  
Instrumentation:
  
Fuel centre temperatures were measured with refractory metal thermocouples, the rod internal pressure was determined with bellows pressure transducers, and elongation sensors were used to obtain the length increase of the cladding. In FUMEX 4, the rods contained all three types of instrument, providing comprehensive information on the state of the fuel.
  
The diameter gauge is a two or three point contact feeler that can be moved along the length of a fuel rod during operation. Diameter changes can be detected with a micrometer resolution as demonstrated in FUMEX 5.
  
Re-instrumentation of irradiated fuel rods:
  
Re-instrumentation of irradiated fuel rods is a method of shortening experiment execution times and costs. Since the instrumentation is exposed to irradiation for a shorter duration, the failure probability is decreased. FUMEX 6 is an example of re-instrumentation with a pressure transducer. The data obtained from this experiment gave a good indication of on-set and kinetics of fission gas release in response to a power increase.
   
Design features for simulation of burnup effects:
  
Increasing burnup in general incurs increasing uncertainties in data interpretation, e.g. fuel temperature changes can be effected by a combination of causes like fission gas release, changes in gap size and conductivity degradation. Test designs with controlled and known influential parameters therefore facilitate the assessment of separate effects.
   
FUMEX 3 and FUMEX 4 are examples where fission gas release was simulated by addition of xenon to the fill gas. Xenon fill gas in combination with a small as-fabricated diametral gap (50-100 micron) simulates a high burnup situation with a large amount of released fission gas and a closed gap due to fuel swelling and clad creep-down. Gap conductance models can be validated with measured data from tests with such a design without the uncertainties caused by high burnup.
  
Error estimation of power and temperature data:
  
At the first start-up of an experiment, assembly power is determined calorimetrically, resulting in a relation between total power and average neutron flux measured by neutron detectors. The calibration error is about 3% which has to be combined with a similar uncertainty for the distribution of total power to individual rods and local positions. The evaluation of start-up data of a large number of comparable HBWR experiments has indeed shown that the observed spread agrees with these considerations and is about 4%. During the course of irradiation, small local changes of the neutron flux distribution, which cannot be resolved with the arrangement of neutron detectors, will add another uncertainty. The combination of all sources leads to an error estimate for power data of about 5 % for the time after the initial calibration. It is customary, however, to assign this uncertainty to temperature rather than power. In a representative selection of the FUMEX evaluation sheets, a 5% error on temperature above coolant temperature is indicated with a corresponding error bar.
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9. STATUS
Package ID Status date Status
NEA-1599/01 30-SEP-2003 Arrived at NEADB
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10. REFERENCES
NEA-1599/01, included references:
- Dr. W. Wiesenack:
Further Specifications of FUMEX input data
OECD Halden Reactor Project
Notes to recipients of FUMEX cases dated Sept.2, 1993 and Jan.12, 1994
- Fuel Modelling at Extended Burnup
Report of the Co-ordinated Research Programme on Fuel Modelling
at Extended Burnup - FUMEX (1993-1996)
IAEA-TECDOC-998 (January 1998)
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12. PROGRAMMING LANGUAGE(S) USED

No item found

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15. ESTABLISHMENT OF AUTHORS

Data released by:
   OECD Halden Reactor Project
   Institutt for Energiteknikk
   P.O. Box 173
   N-1751 HALDEN
   Norway

Compilation: Rodney J. White
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16. MATERIAL AVAILABLE
NEA-1599/01
/original/FUMEX_I/data  Six sub-directories with raw fumex histories
/original/FUMEX_I/doc  *.pdf files of fumex project
FUMEX/doc  Master QA documents for new history generation and
   Graphical history files
FUMEX/data  Fortran programs for history generation (*.f90 files)
Raw fumex histories - *.dat files - plus supplementary information - *.h1 files
New power histories - *.o1 files
Simplified power histories for graphical checks - *.o2 files
1st files
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: burnup, cladding, creep, fission gas release, fuel behaviour, post-irradiation examination, power ramp, pwr reactors.