NEA-1310: IFPE/SOFIT, WWER-440 Fuel Thermal Performance and Fission Gas Release

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NEA-1310 IFPE/SOFIT. (Abstract last modified 23-DEC-2002)

IFPE/SOFIT, WWER-440 Fuel Thermal Performance and Fission Gas Release

Abstract Table of Contents:

1. NAME OF EXPERIMENT - IFPE/SOFIT.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

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Explanation of the status codes.

Program-name Package-ID Status IFPE/SOFIT-1.1 NEA-1310/01 Arrived IFPE/SOFIT-1.3 NEA-1310/02 Obsolete IFPE/SOFIT-1.3 REV.3 NEA-1310/03 Arrived

Machines used: Package-ID Orig.Computer Test Computer NEA-1310/01 Many Computers NEA-1310/03 Many Computers

3. DESCRIPTION -
The SOFIT program is described as a sub-task under the Finnish- Russian co-operation on VVER fuel research and consists of a series of irradiation tests in the MR reactor at the Kurchatov Institute, Moscow. The contracting parties are Imatran Voima Oy (IVO) and the Russian National Research Centre Kurchatov Institute (IRTM), (Reference 1). The program is divided into three distinct phases each addressing specific objectives:
    SOFIT 1   Parametric fuel rod irradiations with basic
              steady state power histories up to moderate
              levels of burn-up as dictated by instrumentation
              endurance.
    SOFIT 2   Parametric studies based on irradiation of
              instrumented high burn-up rods.
    SOFIT 3   Irradiation testing under transient conditions.
The results of 2 assemblies of SOFIT 1 are at present available where the main objective was to obtain well qualified data on VVER-440 fuel for verifying and improving codes. In each assembly, rods of different design were irradiated with in-pile instrumentation to measure fuel centreline temperature, fuel stack and cladding elongation. PIE has been performed to obtain data on microstructural changes and measurement of fission gas release (FGR).
The first series of irradiations were completed by May '92 and some destructive PIE has been performed.

NEA-1310/01:

DESCRIPTION -
Precharacterization and irradiation histories for SOFIT 1.1 rods 1-6, 7 and 12.
In-pile temperatures for rods 1-6 and PIE fgr from rods 7 and 12.

NEA-1310/03:

DESCRIPTION OF PROGRAM OR FUNCTION -
Precharacterization and irradiation histories for SOFIT 1.3 rods 1, 3, 4 and 5.
In-pile fuel and clad extension measurements for rods 1 and 3.
In-pile temperature data for rods 4 and 5.

4. METHOD OF SOLUTION -
- IRRADIATION CONDITIONS IN THE MR REACTOR
The MR reactor is a pool type research reactor with a total power of 50 MW. The reactor contains several pressurized loops which can be connected to one or more in-pile test channels. These channels are located between blocks of beryllium which act as moderators. The reactor is operated in 30 to 40 day cycles of near constant power. At the time of the SOFIT tests, the axial form factor was 1.37 and the in-channel radial form factor was 1.2.
Rod power determination was based on the calorimetric bundle power measurements and calculations. The axial and radial power distributions were calculated using a model based on experimental data from the MR reactor. Axial power profiles were refined using signals from the 5 axially located neutron detectors by fitting a 5th order polynomial to the measurements to obtain a smooth power profile. The uncertainty in the axial power profiles is estimated at +/-5%. The in-bundle radial power profile was calculated using signals from 3 neutron detectors located on the same plane as the core central elevation. The detectors measured some azimuthal variation in power during irradiation and the uncertainty of the radial profile of +/-5% combined to provide an estimated total uncertainty of +/-10% in local linear power.
- EXPERIMENTAL DETAILS
SOFIT 1.1 - 1.4 comprised 4 assemblies each of 18 rods arranged in hexagonal geometry around a central non fuelled tube used for instrumentation including neutron detectors and coolant temperature thermocouples. The fuel-to-clad gaps ranged between 140 and 290 microns, the fill gas, both helium and xenon were used, varied in pressure between .1 to 2 MPa. The UO2 density varied between 10.4 and 10.75 g/cc and the fabricated grain size was around 5 microns. The fuel active length was 1000 mm within a total rod length of 1200 mm. In most cases 6 of the 12 outer rods were instrumented with thermocouple hot junctions located between 300 and 500 mm above the lower end of the fuel stack and near the maximum power position.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME -

7. UNUSUAL FEATURES OF THE PROGRAM -

8. RELATED AND AUXILIARY PROGRAMS -

9. STATUS
NEA-1310/01: 12-NOV-1997 Arrived at NEADB
NEA-1310/02: 23-DEC-2002 Obsolete
NEA-1310/03: 23-DEC-2002 Arrived at NEADB

10. REFERENCES  -
        SOFIT: A Joint Experimental Programme Between the USSR and
        Finland on VVER Fuel Performance, V Yakovlev, A Moshajev,
        P Strizhov, J Johansson, P Liuhto, P Hyvarinen and R
        Terasvirta, Paper presented at IAEA International Symposium
        on the Utilization of Multi-purpose Research Reactors and
        related International Co-operation, Grenoble, France, 19-23
        October, 1987.
NEA-1310/01:
- P. Losonen:
OECD/NEA Data Bank, SOFIT-DATA, Note (28/11/96)
- P. Losonen:
Verification of Transuranus Against Temperature Data from WWER type Test Fuel
Rods from Sofit Experiments Paper to be presented in IAEA/OECD Data Base
Training meeting in Halden, Norway, 25-27 September 1996
- V. Yakovlev, R. Strijov, V. Murashov, J. Johansson, R.P. Terasvirta, O.
Tiihonen and K. Ranta-Puska:
Research carried our on WWER-440 Type Fuel Rods in the MR Reactor
IAEA-SM-288/64 Reprint from Improvements in Water Reactor Fuel Technology and
Utilization
- V. Yakovlev, R. Strijov, V. Murashov, A. Senkin, R.P. Terasvirta, P. Liuhto,
J. Moisio, O. Tiihonen, S. Kelppe and K. Ranta-Puska:
Qualification and Interpretation of MR Test Reactor Irradiation Data on
VVER-440 Type Fuel Rods for Fuel Thermal Model Validation IEA-TC-659/1.4
- A.V. Smirnov et al.:
WWER-1000 and WWER-440 Fuel Operation Experience American Nuclear Society, Int.
Topical Meeting on LWR Fuel Performance Florida, USA, April 16-19, 1994
- Yu. Bibilashvili et al.:
Toward High Burnup in Russian WWER Reactors and Status of Water Reactor Fuel
Technology American Nuclear Society, Int. Topical Meeting on LWR Fuel
Performance Florida, USA, April 16-19, 1994
- D. Elenkov and K. Lassmann:
The Development of the Transuranus-WWER Version
- Solonin M. et al.:
WWER Fuel Performance and Material Development for Extended Burnup in Russia,
Proceedings of the Second International Seminar, WWER Reactor Fuel Performance,
Modelling and Experimental Support, 21-25April 1997, Sandanski, Bulgaria (Ibid.
[4] pp. 48-57)
NEA-1310/03:
- P. Losonen:
  OECD/NEA Data Bank, SOFIT-DATA
  Note (28/11/96)
- P. Losonen:
  Verification of Transuranus Against Temperature Data from
  WWER type Test Fuel Rods from Sofit Experiments
  Paper to be presented in IAEA/OECD Data Base Training meeting
  in Halden, Norway, 25-27 September 1996
- V. Yakovlev, R. Strijov, V. Murashov, J. Johansson, R.P. Terasvirta
O. Tiihonen and K. Ranta-Puska:
Research carried our on WWER-440 Type Fuel Rods in the MR Reactor
IAEA-SM-288/64
Reprint from Improvements in Water Reactor Fuel Technology and
Utilization
- V. Yakovlev, R. Strijov, V. Murashov, A. Senkin, R.P. Terasvirta
P. Liuhto, J. Moisio, O. Tiihonen, S. Kelppe and K. Ranta-Puska:
Qualification and Interpretation of MR Test Reactor Irradiation Data on
VVER-440 Type Fuel Rods for Fuel Thermal Model Validation
IEA-TC-659/1.4
- A.V. Smirnov et al.:
WWER-1000 and WWER-440 Fuel Operation Experience
American Nuclear Society, Int. Topical Meeting on LWR Fuel Performance
Florida, USA, April 16-19, 1994
- Yu. Bibilashvili et al.:
Toward High Burnup in Russian WWER Reactors and Status of Water
Reactor Fuel Technology
American Nuclear Society, Int. Topical Meeting on LWR Fuel Performance
Florida, USA, April 16-19, 1994
- D. Elenkov and K. Lassmann:
The Development of the Transuranus-WWER Version

11. MACHINE REQUIREMENTS -

12. PROGRAMMING LANGUAGE(S) USED -
NEA-1310/01:
NEA-1310/03:

13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED -

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. ESTABLISHMENT OF AUTHORS -
             Imatran Voima Oy (IVO)
             Vantaa
             Finland

             Kurchatov Institute (IRTM)
             Kurchatov Square
             123182 MOSCOW
             Russian Federation
             Compiled by: J.A. Turnbull, U.K.

16. MATERIAL AVAILABLE -
NEA-1310/01:
README.SOF Information file
SUMMARY.SOF SOFIT programme summary
S1-1R1_6.PC Precharacterization data rod1-6
S1-1R7.PC Precharacterization data for rod7
S1-1R12.PC Precharacterization data rod 12
S1-1R1.IRR 10 zone irradiation history rod1
S1-1R2.IRR 10 zone irradiation hist. rod 2
S1-1R4.IRR 10 zone irradiation history rod4
S1-1R5.IRR 10 zone irradiation history rod5
S1-1R6.IRR 10 zone irradiation history rod6
S1-1R7.IRR 10 zone irradiation history rod7
S1-1R12.IRR 10 zone irradiation hist. rod12
S1-1R1.BOL 10 zone histories, BOL rod 1
S1-1R2.BOL Detailed 10 zone hist. BOL rod 2
S1-1R3.BOL Detailed 10 zone hist. BOL rod 3
S1-1R4.BOL Detailed 10 zone hist. BOL rod 4
S1-1R5.BOL Detailed 10 zone hist. BOL rod 5
S1-1R6.BOL Detailed 10 zone hist. BOL rod 6
S1-1R1.TF Fuel temperatures for rod 1
S1-1R2.TF Fuel temperatures for rod 2
S1-1R4.TF Fuel temperatures for rod 4
S1-1R5.TF Fuel temperatures for rod 5
S1-1R6.TF Fuel temperatures for rod 6
S1-1R1.TFB Fuel temperatures BOL rod 1
S1-1R2.TFB Fuel temperatures BOL rod 2
S1-1R4.TFB Fuel temperatures BOL rod 4
S1-1R5.TFB Fuel temperatures BOL rod 5
S1-1R6.TFB Fuel temperatures BOL rod 6
S1-1R3L.IRR Irr.hist.lower thermocoup.rod 3
S1-1R3U.IRR Irr.hist.upper thermocoup.rod 3
S1-1R3L.TF Fuel temp.lower thermocoup.rod 3
S1-1R3U.TF Fuel temp.upper thermocoup.rod 3
S1-1R3L.TFB Fuel temp.low. thermoc.BOL rod3
S1-1R3U.TFB Fuel temp.upp. thermoc.BOL rod3
NEA-1310/03:
README.3  Readme file
SUMMARY.SOF SOFIT programme summary
S1-3R.PC Prechara. data for SOFIT 1.3 rods 1,3,4 & 5
S1-3Rx.IRR Complete 10 zone irr.hist. rod 1,3, 4 and 5
S1-3R1.EX Fuel stack & clad elongation rod 1
S1-3R3.EX Fuel stack & clad elongation rod 3
S1-3R4.TF Fuel temperatures rod 4
S1-3R5.TF Fuel temperatures rod 5
Review of SOFIT Experiments.doc

17. CATEGORIES -

- Y. Integral Experiments Data, Databases, Benchmarks

Keywords: EXPERIMENTAL DATA, FISSION PRODUCTS, FUEL ELEMENTS, FUEL PELLETS, FUEL RODS, FUEL-CLADDING INTERACTION, FUEL-COOLANT INTERACTIONS, WWER REACTORS, XENON