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NEA-1766 IFPE/KOLA-3-MIR-RAMP

IFPE/KOLA-3-MIR-RAMP, KOLA-3 MIR test temperature during ramp, FGR and pressure at EOL, Bu up to 55 MWd/kgUO2

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1. NAME OF EXPERIMENT:  IFPE/KOLA-3-MIR-RAMP.
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2. COMPUTERS
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Program name Package id Status Status date
IFPE/KOLA-3-MIR-RAMP NEA-1766/02 Arrived 12-OCT-2011

Machines used:

Package ID Orig. computer Test computer
NEA-1766/02 Many Computers
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3. DESCRIPTION

In 1996-97 a number of tests with the high burn up VVER-440 fuel under transient conditions were carried out in the MIR reactor (SSC RIAR). Fuel rods that had operated under normal operating conditions at the Kola NPP Unit 3 during 4- and 5 fuel cycles up to the maximum burnup of about 50 and 60 MWd/kgU were tested. The tests were carried out under single ramp conditions (RAMP experiment) and step by step power increase (FGR-1 and FGR-2 experiments). The objective of the experiments was to determine the power ramp influence on fuel rods state, to evaluate FGR induced-threshold linear power values, to study dependence between the linear power, temperature, structure and properties of the fuel.
  
Nine refabricated fuel rods were tested in the above mentioned tests. All of these refabricated fuel rods (RFRs) were cut from fuel rods of the FAs 198 and 222. The positions of the RFRs, the average linear heat rate of the FAs 198 and 222 fuel rods during base irradiation, the operating conditions of the FA-198 and FA-222 fuel rods, details of the design values of the FA's 198 and 222 fuel rods and the main operating conditions of the Kola 3 during 5-9 fuel cycles are provided.
  
RAMP test:
The test rig with RFRs was installed for testing in the research reactor MIR in February 1996. During the RAMP experiment 3 non-equipped RFRs were tested under power ramp conditions. The RAMP test was divided into 3 stages: stage 1 - irradiation at initial power level (duration - ~346 hours); stage 2 - power ramp from initial to maximal level (duration - ~23  min); stage 3 - hold stage after ramp (duration - ~107 hours).
  
FGR-1 test:
The FGR-1 experiment started in March 1996 and finished in April 1996. The test rig included three RFRs. Two were instrumented with pressure sensors while the third was not instrumented.
  
FGR-2 test:
The FGR-2 experiment was the second of the tests series directed to investigation of the thermal physical behaviour of VVER fuel at the different power levels. It was carried out during April-June of 1997. The test rig included three RFRs, two of which were instrumented with thermocouples, and the third was not instrumented.
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9. STATUS
Package ID Status date Status
NEA-1766/02 12-OCT-2011 Arrived restricted
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10. REFERENCES

- A. Smirnov, B. Kanashov, G. Lyadov et al.:
'Examination of Fission Gas Release and Fuel Structure of High Burnup WWER-440 Rods under Transient Conditions', Proceeding of the third international seminar
'WWER fuel performance, modeling and experimental support', Pamporovo, Bulgaria, 4-8 October 1999

- A. Smirnov, B. Kanashov, V Ovchinnikov et al.:
'Study of Behaviour of WWER-440 Fuel Rods of Higher Fuel Burnup under Transient Conditions', Report HPR-349/43, Enlarged HPG Meeting on High Burnup Fuel Performance, Safety and Reliability, OECD Halden Reactor Project, Norway, Lillehammer, 15-20 March,1998

- S. Lemehov, A. Smirnov, V. Tsibuya:
'KOLA-3 High burnup fuel validation tests FA-198 and FA-222', Enlarged HPG Meeting on High Burnup Fuel Performance, Safety and Reliability and Degradation of In-Core Materials and Water Chemistry Effects, OECD Halden Reactor Project, Norway, Loen, 19-24  May,1996

- Yu. Bibilashvili, A. Medvedev, G. Khvostov et al.:
'Fission Gas Products Behaviour Modelling in the START-3 Code for the WWER Fuel at High Burnup and Transient Conditions', Proceeding of the third international seminar 'WWER fuel performance, modeling and experimental support', Pamporovo, Bulgaria, 4-8 October 1999

- Smirnov, V. Smirnov, A. Petuhov et al.:
'The Peculiarities of the WWER-440 Fuel Behaviour at Higher Burnups', Proceeding of the second international seminar 'WWER reactor fuel performance, modeling and experimental support', Sandanski, Bulgaria, 21-25 October 1997

- Smirnov, B. Kanashov, V. Kuzmin et al.:
'Results of Post Irradiation Examination to Validate WWER-440 and WWER-1000 Fuel Efficiency at High Burnups', Proceeding of the third international seminar 'WWER fuel performance, modeling and experimental support', Varna, Bulgaria, 1-5 October 2001
NEA-1766/02, included references:
- The tests of VVER-440 refabrication rods in reactor MIR (working materials)
(2005) Prepared by A.A. Bochvar All-Russia Research Institute of Inorganic
Materials
- 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)
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12. PROGRAMMING LANGUAGE(S) USED

No item found

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15. NAME AND ESTABLISHMENT

Federal State Unitary Enterprise
A.A. Bochvar All-Russia Research Institute of Inorganic Materials
(VNIINM)
123060 Moscow, P.B. 369
Russian Federation
  
FSUE "SSCR RIAR"
Dimitrovgrad-10
433510 Dimitrovgrad, Ulyanovsk region
Russian Federation
  
IAEA
Wagramerstrasse 5
P.O.Box 100
A-1400 Vienna
Austria
  
Released within FUMEX-II exercise
  
Compilation: J.A. Turnbull
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16. MATERIAL AVAILABLE
NEA-1766/02
Summary of irradiation prepared by A. A. Bochvar All-Russia Research Institute
of Inorganic Materials (Word)
Pre characterization      WORD file summarizing pre-characterization
PIE results    Post Irradiation Examination after Ramping (WORD)
QA report for compiling dataset    (WORD)
Readme.doc    Readme file
FA198-xx   Irradiation histories for each mother rod (from FA-198, xx = 20, 76,
99)
FA222-xx   Irradiation histories for each mother rod (from FA-222, xx = 2, 3, 5,

6, 25, 46)
FGR1_x_B    Base irradiation history for re-fabricated rod in FGR1 ramp test, x
= 32, 41, 48
FGR2_x_B    Base irradiation history for re-fabricated rod in FGR2 ramp test, x
= 50, 51, 52
RAMP_x_B    Base irradiation history for re-fabricated rod in RAMP test, x = 33,

37, 38
FGR1_x_R    Ramp irradiation history for re-fabricated rod in FGR1 ramp test, x
= 32, 41, 48
FGR2_x_R    Ramp irradiation history for re-fabricated rod in FGR2 ramp test, x
= 50, 51, 52
RAMP_x_R    Ramp irradiation history for re-fabricated rod in RAMP test, x = 33,

37, 38
PGas_41.txt          in-pile pressure data for FGR1 RFR 41
PGas_48.txt          in-pile pressure data for FGR1 RFR 48
FGR1_Pr_cool.txt     coolant pressure during FGR1 tests
RFR_50_tc.txt        in-pile temperature data for FGR2 RFR 50
RFR_51_tc.txt        in-pile temperature data for FGR2 RFR 51
FGR2_Pr_cool.txt     coolant pressure during FGR2 tests
Unreviewed files: RFR_50_tc_avg.txt, RFR_51_tc_avg.txt, IFPE_KOLAMIR.doc
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: VVER-440, fission gas release, fuel behaviour, power ramp.