MYRRHA FUEL TRANSIENT TESTS PROJECT AT THE TRIGA-ACPR REACTOR

Transcription

1 MYRRHA FUEL TRANSIENT TESTS PROJECT AT THE TRIGA-ACPR REACTOR CSABA ROTH, BRIAN BOER*, MIREA MLADIN, ADRIAN DATCU, GEORGIANA BUDRIMAN, CALIN TRUTA Institute for Nuclear Research Pitesti, Romania * SCK CEN, Mol, Belgium The 7th Annual International Conference on Sustainable Development through Nuclear Research and Education May 28-30, 2014, Piteşti, România

2 Outline MYRRHA & MAXSIMA Reactivity Insertion Accident scenario in MYRRHA Modelling and simulation of the test rig Thermo-mechanical behavior of the fuel segment Experimental devices conceptual design Conclusions

3 MYRRHA & MAXSIMA MYRRHA - Multi-purpose hybrid Research Reactor for High-tech Applications One of the three new research reactors forming the cornerstons of the European Research Area of Experimental Reactors (ERAER) flexible fast spectrum irradiation facility high power proton accelerator, coupled to a lead-bismuth cooled subcritical core by means of a lead-bismuth spallation target irradiation tool for science and research in reactor and fuel cycle technology, reactor physics, material and fuel science, neutron physics and life sciences innovative material and fuel technologies testing and qualification by irradiation in representative conditions in MYRRHA neutron irradiation services and beam time of charged particles (proton) for scientific/industrial irradiation programs

4 MYRRHA & MAXSIMA The goal of FP7 MAXSIMA is to contribute to the "safety in MYRRHA" assessment. Duration: Consortium: SCK CEN (coordinator), ENEA, KIT, ANSALDO, GRS, NRG, CRS4, KTH, HZDR, ICN, CHALMERS, NNC, CIEMAT Safety Analysis in support of MYRRHA Core component Safety studies Steam generator and cooling safety Fuel Safety Enhanced Safety by Design for HLM reactors Education and training

5 MYRRHA & MAXSIMA Fuel safety Transient testing of MYRRHA fuel for the determination of the pin failure threshold Test methodology development and simulation of transient fuel behavior. Design and fabrication of fuel elements and irradiation device.

6 MYRRHA & MAXSIMA TRIGA Annular Core Pulsed Reactor Various experiments can be performed in the 9 dia. large dry irradiation channel by exposure either to a neutron fluency of n/cm 2 in a single pulse or to a flux rate of n/cm 2 s in steady-state operation. The Reactor is suitable for transient tests on fuel specimens. The TRIGA-ACPR Reactor supported a large experimental program of transient tests on CANDU type fuel using a suitable experimental set-up.

7 Pu-U mixed oxide (MOX), 30 % Pu Pellets of 5.42 mm in diameter The density of the MOX is 95 % of the theoretical density at 10.5 g/cm 3 Cladding DIN austenitic stainless steel with an inner and outer diameter of 5.65 mm and 6.55 mm, respectively. LBE cooled, 270 C at the core inlet

8 Reactivity Insertion Accident scenario in MYRRHA Several transient scenarios are considered for the safety evaluation of MYRRHA fuel. The control rod ejection scenario is seen as one of the most severe in terms of overpower and possible resulting pellet-clad interaction. Transient tests are aimed to determine the pin failure threshold value of the energy deposition which could produce cladding failure. Power evolution

9 Reactivity Insertion Accident scenario in MYRRHA During the transient the gap closes rapidly with hard contact between pellet outer surface and the cladding inner surface. Large mechanical stresses and cladding plastic deformation are the result, which is known as Pellet-Cladding Mechanical Interaction (PCMI). A sensitivity study shows that a cladding plastic deformation of 0.5%, used in past fast reactor programs as the design limit, is only exceeded for energy depositions > 300 J/g together with a gap size < 20 micron.

10 Modelling and simulation of the test rig The preliminary assessment shows that the required energy deposition level can be achieved during the pulse by using additional neutron moderator materials to modify the neutron spectrum aimed to increase the fission power density in the test segments. Be blocks, available on site optimal solution The model created with MCNPX for ACPR core was used and modified to include the capsule assembly. The assembly is composed of five beryllium blocks, the central one containing the capsule MCNPX model of the core & irradiation device

11 Modelling and simulation of the test rig Reactor power evolution in pulses with one rod (1.63 $) and two rods (3.48 $) Average energy deposition inside the three test fuel pins as a function of enrichment.

12 Modelling and simulation of the test rig Time evolution of central fuel (node1), outer face of cladding (node7) and middle of LBE layer (node8) temperatures for 11% wt. U-235

13 Thermo-mechanical behavior of the fuel segment Based on the neutronic evaluation a thermo-mechanical simulation of the fuel test segments during the transient was made with TRANSURANUS code. The fuel test segments will use UO2 instead of MOX during the first phase of the project (methodology validation). Deposited energy Maximum enthalpy Maximum fuel temperature Maximum cladding temperature Maximum equivalent stress Effective plastic strain Permanent hoop strain J/g J/g C C MPa 0.47% 0.40% J/g J/g C C MPa 0.91% 0.71% J/g J/g C C MPa 1.43% 1.04% J/g J/g C C MPa 1.99% 1.35% J/g J/g C C MPa 2.58% 1.66% J/g J/g C C MPa 3.16% 1.97% J/g J/g C C MPa 3.72% 2.31% J/g J/g C C MPa 4.26% 2.65% above ~300 J/g energy deposition the cladding (effective) plastic strain limit of 0.5 % is exceeded

14 Thermo-mechanical behavior of the fuel segment Test matrix Expected failure result Expected cladding deformation (equivalent strain) Standard cladding + extra cold worked cladding Failure Failure Failure more unlikely Possible likely 0.5 % > 1.0% > 2.5 % Energy deposition 300 J/g 500 J/g 720 J/g Fuel enthalpy increase 212 J/g 338 J/g 526 J/g Required fuel enrichment 5 % 10 % 20 % Next to the standard cladding material (DIN ) an extra cold worked version of the cladding will be used for some of the test segments to mimic loss of cladding mechanical ductility by irradiation induced hardening

15 Experimental devices conceptual design The test fuel: Active length: 15 cm Outer radius: cm, 0.45 mm thick Pellet radius: cm Cladding: stainless steel Ti-15/15 (DIN ).

16 Experimental devices conceptual design Irradiation device ACPR experimental dry channel

17 Experimental devices conceptual design Assembly of the experimental device as formed by its components placed in the ACPR core dry experimental channel

18 Experimental devices conceptual design Upper plug Internal capsule - single use device which contains three test fuel segments fixed on grids, fully inserted in Lead-Bismuth Upper grid Test fuel segments Lower grid Lower plug

19 Experimental devices conceptual design Positioning of the test fuel on the maximum of the axial flux distribution ( 20 cm) 1,2 1 Thermal flux (E<10 kev) measured Fast flux (E>10 kev) measured Thermal flux (E<10 kev) computed Fast flux (E>10 kev) computed Thermal/Fast flux (relativ units) 0,8 0,6 0,4 0, Distance (cm)

20 Experimental devices conceptual design Melting tank TC1 LBE filling installation gas source P1 Volume tube TC2 TC3 P2 TC4 pump All vessels and tubes will be heated and insulated. During the filling, the LBE will be heated up to C; the rest of the installation (tubes, valves) and the irradiation capsule will be heated up to C. A sufficient number of TC s for temperature monitoring. Standard vacuum pump to remove the air from the system (down to 10-2 mbar) before filling with LBE. Capsule The 7th Annual International Conference TC5 on Sustainable Development through Nuclear Research and Education,

21 Experimental devices conceptual design Internal capsule disassembling Internal capsule Heater & insulator Recovered LBE

22 Experimental devices conceptual design Instrumentation Irradiation capsule instrumentation 30 cm length Co SPND flux measurement Ni-Cr thermo-couple LBE temp. meas. Experiment instrumentation - to measure the cladding temperature during the pulse for all three test segments. TC fixed in the central position of the test segments, fastened in a small tub segment brazed on the test segment surface. The internal pressure of test segments will be not measured and recorded - no financially viable option available Fast data acquisition system

23 Experimental devices conceptual design Brazing parameters optimization Brazing test specimen Further investigation will be done on the brazing procedure effect on the DIN14970 cladding material, including alternative technical solutions

24 CONCLUSIONS The preliminary design of the fuel pin segments and irradiation capsule for transient testing of MYRRHA fuel in the ACPR TRIGA reactor has been established. The transient tests aim at the determination of the MYRRHA fast reactor fuel pin failure threshold during RIA type transients. Extensive modeling with neutronics, thermal and fuel performance codes was used to determine the design requirements of the fuel segments and capsule. Enrichment levels of the UO 2 fuel of the segments were fixed to 5 %, 10 % and 20 % enrichment in U-235 to achieve cladding plastic deformation due to PCMI of 0.5% or more. The capsule design satisfies the neutronic requirements and allows performing the transient tests in a safe manner.