MSR IRRADIATION PROGRAM AT NRG PETTEN MSR Workshop 2018, ORNL, US - PowerPoint PPT Presentation

MSR IRRADIATION PROGRAM AT NRG PETTEN MSR Workshop 2018, ORNL, US P.R. Hania 2018-10-04

2 NRG-2.4094/18.149980 - EU DuC=N MOLTEN SALT REACTOR CHALLENGES MSRs are complex, and difficulties are multidisciplinary ! Technological challenges need to be solved for a safe and economic MSR. ! Time-consuming and costly experiments are required, to tackle these ! challenges and provide a basis to license MSR designs With its experience and facilities available, NRG can provide a significant ! contribution to MSR research. In view of the large perspective of MSR technology for the (longer term) future, NRG has embarked on MSR R&D based on a government supported program.

NRG-2.4094/18.149980 - EU DuC=N 3 THE DUTCH MOLTEN SALT PROGRAM Molten Salt Technology fits well within the goals of the Dutch nuclear energy R&D program: Improve safety ! Reduce resource consumption / waste ! Contribute to CO 2 -free energy market ! Collaboration between NRG, JRC, TU Delft and CV Rez Complementary competences ! Objective: to contribute to molten salt technology development: 1. Obtain operational experience 2. Safety Confirm Fission Products (FP) stability in the salt and FP ! migration Investigate FP management methods ! 3. Material qualification: Material properties of irradiated containment materials ! In-pile corrosion / deposition of suitable alloys and SiC ! 4. Waste: Provide a waste route for spent molten salt fuel ! 5. Integral Demonstration: Feasibility of experimental Molten Salt loop for the HFR Petten !

4 NRG-2.4094/18.149980 - EU Du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

5 NRG-2.4094/18.149980 - EU DuC=N NRG MSR PROGRAM AT NRG 2!5 !"#$"%&' !"#$%&'() !"#$%&'(* !+,+ -./012+ 3"4'% 2!5(6778 !"#$%&'()*+,-# !"#$%&'()*+,-# 7'8)*/90)5# ?%/)+6# @%(%/*3# • @%.)24 • .'/012,'3&)0%# .'/016%0'/# 3,*8+50)*4#*:# A*,,*.)*4#0%.0. %6$,)00/%6%40 8).3*.'/#,*+0% )40%,'50)*4 • B':%09#'4'/9.%. )40%,'50)*4 ! ; <=! > • ?%/)+6#$+$$/)42# C*+0#*:#3)/%D Focus on irradiation technology ! Focus on generic topics (not specific for certain concepts) ! Ambitious program with limited funding, program open for ! partnering

6 NRG-2.4094/18.149980 - EU DuC=N LUMOS LOOP CONCEPT DESIGN In-pool loop positioned directly next to HFR core wall Main parameters: Actinide bearing FLIBE ! salt (20-25 L) Alloy N first ! containment Power: 125 kW ! Power density: 100- ! 150 W/cc Flow rate: ! 3 m/s ! ! T: ! 100 o C ! 5-6 operational years ! targeted

7 NRG-2.4094/18.149980 - EU DuC=N THE HIGH FLUX REACTOR (HFR)

8 NRG-2.4094/18.149980 - EU DuC=N CURRENT IRRADIATION ACTIVITIES 1. SALIENT-01: LiF-ThF 4 in graphite crucibles 2. SALIENT-03: LiF-ThF 4 -UF x -PuF 3 in Alloy N crucibles 3. ENICKMA: tensile and low-cycle fatigue samples of nickel based alloys

9 NRG-2.4094/18.149980 - EU DuC=N SALIENT-01 DESIGN ! Salt composition: 78LiF-22ThF 4 ! Nuclear-grade graphite ! Fuel power rises during irradiation due to production of U-233 ! Fixed crucible temperature (~600 o C)

10 NRG-2.4094/18.149980 - EU DuC=N SALIENT-01 ASSEMBLY Synthesis and crucible loading at JRC Karlsruhe Assembly of sample holder at NRG

11 NRG-2.4094/18.149980 - EU DuC=N SALIENT-01 STATUS Start of irradiation: ! !"#$%&'$()*+ #,-$%.&(/%$*0$ August 10, 2017 1.35 ! ! "#$%&'()*+!,! ! "#-&$.+ 1.30 9 out of 18 cycles completed ! according to specifications 1.25 273 Full Power Days ! Temperatures on target: ! 1.20 595 o C (L1, L4) ! 1.15 634 o C (L2, L3) ! 1.10 0 1 2 3 4 5 6 7 8 9 Experiment was moved to Cycle number ! !"#$"%&'(%") &'*+,,-*."/.*0.*1&)*1&$) lower-flux position (G7-> H4) 2#"&)(%"3 &'*4546" )'&%'7($8 after cycle 8

12 NRG-2.4094/18.149980 - EU DuC=N SALIENT-03: GOALS Investigate in-pile corrosion of Hastelloy N by fluoride fuel salt Determine whether corrosion is irradiation-enhanced ! Determine the influence of fission products and redox buffering on the ! corrosion rate Compare experimental mass transport in a non-isothermal salt column ! to CFD simulations Investigate fission product behavior Determine in-pile fission gas release ! Establish which fission products/species relocate to ‘cold spots’ during ! irradiation Determine post-irradiation fission product release temperatures ! (Knudsen Cell Effusion test at JRC Karlsruhe) Start of irradiation in 2019

13 NRG-2.4094/18.149980 - EU DuC=N SALIENT-03 DESIGN Changes with respect to SALIENT-01: Channel of Gas: He+Ne mix 3 rd Containment REFA-170 capsule Heaters to avoid radiolysis during ! Gas: He+Ne mix external 2 nd Containment Sample holder HFR downtime internal Addition of Pu for fission power at ! Gas:He Sample 1 st Containment holder start of irradiation (sealed) Electrodes Addition of U for ‘salt buffering’ Pressure ! transducer (UF 4 /UF 3 ) Graphite Welded Alloy N capsules: ! Heaters shroud Corrosion test ! Gas: He inside the Pressure measurement ! capsules Inclusion of 3 inert electrodes ! Nuclear fuel Molten Salt Large measurable temperature ! gradients : Transport phenomena !

14 NRG-2.4094/18.149980 - EU DuC=N ENICKMA IRRADIATION Irradiation of Alloy N based ! material specimens for post- irradiation mechanical testing: Tensile testing ! Low Cycle Fatigue ! Small Punch testing ! Microstructure analysis ! Irradiation parameters: ! "#$%&'()*$+',(-(.// Temperature: 650 and 750 o C ! Up to 1E21 n/cm 2 thermal, 3E21 ! n/cm 2 fast (up to 50 appm helium, >1 dpa expected) Oven anneal test at same ! temperatures for comparison

15 NRG-2.4094/18.149980 - EU DuC=N OTHER ACTIVITIES 1. Gamma irradiation near RT 2. Waste treatment 3. Lab-scale helium bubbling

16 NRG-2.4094/18.149980 - EU DuC=N SAGA: SALT RADIOLYSIS TEST HFR Spent fuel is used as the gamma source ! ~50 o C base irradiation (solid salt samples) ! Monitoring of pressure, dose and temperature ! 5 salts investigated, Salt samples provided by CV Rez ! Start Q4 2018 !

17 NRG-2.4094/18.149980 - EU DuC=N WASTE PROJECT Commitment to convert the waste produced by the MSR ! irradiations to a chemical form to be transported to Dutch center for interim storage (COVRA) Actinide- and fission product bearing fluoride salt samples are ! not an acceptable waste form by COVRA Fuel waste (containing actinides and fission products) can be ! accepted only in chemically stable forms Irradiated fuel salts release the corrosive fluorine gas following ! radiolysis at near room temperature, reduced salt is itself corrosive conversion to well-known stable matrices

18 NRG-2.4094/18.149980 - EU DuC=N LAB-SCALE HELIUM BUBBLING !"#$% &%'($)*%&"+($, ! "#$%&'(#$&)*+%, ! -.%+)(/$ 0%'1+# ! 2/'+.+)'3 !"#$%&'()*+&,+%"(()&-./&0"(1#2

19 NRG-2.4094/18.149980 - EU DuC=N SUMMARY NRG develops irradiation capabilities, provides information to ! mitigate risks and increases knowledge on Molten Salt Reactor Technology NRG seeks to collaborate with (support) MSR developers to ! accelerate their path towards MSR technology

20 NRG-2.4094/18.149980 - EU DuC=N ACKNOWLEDGEMENTS D. A. Boomstra O. Benes E. Capelli J. Uhlir P. Soucek M. Hoogmoed J.L. Kloosterman M. Marecek M. Naji K. Kottrup A.L. Smith R.J.M. Konings H. Uitslag-Doolaard D. Bykov E. d’Agata A.J. de Koning E. de Visser-Tynova J.D. Bruin !""#$%&'()&)#%*+,-%,"%./0"1)'2%'2#%2',+"2'&%)31"0,%'/,-"0+4',+"2%*-)2% )31"0,)#%50"6%.7%8/9 :./0"1)'2%8/'&;/$)%9"#+5+<',+"2=%2",%)>/'&%,"%?@A%'0)% $/(B)<,%"6%,-)%.7%'2#%6'C%()%$/(B)<,%,"%2',+"2'&%)31"0,%'/,-"0+4',+"2%*-)2% )31"0,)#%,"%'2",-)0%.7%<"/2,0C%'$%*)&&D%.E)2%*+,-"/,%'2%.7%8/9F%"0%*+,-%.7% 8/9 ?@AF%'/,-"0+4',+"2%6'C%()%0)>/+0)#%#/)%,"%,-)%5+2'&%#)$,+2',+"2%'2#% 1/01"$)%5"0%*-+<-%,-)%G""#$%'0)%,"%()%/$)#D%@"%0+G-,$%6'C%()%#)0+E)#%50"6% ,-)%$1)<+5+)#%.7%8/9 "0%'($)2<)%"5%'2%.7%8/9D

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22 NRG-2.4094/18.149980 - EU DuC=N !"##"$ $%&'"()*+$,)(-#)%&#%)"

23 NRG-2.4094/18.149980 - EU DuC=N THE HIGH FLUX REACTOR (HFR)

24 NRG-2.4094/18.149980 - EU DuC=N THE HIGH FLUX REACTOR (HFR) High flux ! 45 MW thermal power ! Stable and constant ! flux profile in each irradiation position Main applications ! Isotope ! production Nuclear energy ! irradiation services R&D ! 31 operation days per ! irradiation cycle, 9 cycles a year