Fuel for sodium cooled fast reactors in Russia

Transcription

1 Fuel for sodium cooled fast reactors in Russia L. M. Zabudko, V.M. Poplavsky (1), I.A. Shkaboura, M.V. Skupov (2), V.A. Kisly, F.N. Kryukov (3), State Corporation for Atomic Energy ROSATOM (1)Institute for Physics and Power Engineering (IPPE), Obninsk, (2) High-Technological Institute of Inorganic Materials (VNIINM), Moscow (3) Research Institute of Atomic Reactors(RIAR), Dimitrovgrad FAIRFUELS Workshop, 9-10 February, 2011, Stockholm 1

2 Presentation Outline Background: SFR technology and fuel Advanced SFR cores (BN-800, BN-1200) Principal results of advanced fuels study Summary 2

3 Current Status of Sodium Cooled Fast Reactors Technology in Russia Russia (USSR): the best results of sodium cooled fast reactors technology assimilation ~ 120 reactor-years: ~ 40% of SFR-years world wide Experimental facilities Advanced design BR-10 (1959) BOR-60 (1969) BN-350 (1973) BN-600 (1980) BN-800 (under construction) BN-1800 (conceptual design) BN-1200 (under development) Different fuels (PuO 2, UC, UN, UPuN, UO 2, UPuO 2, alloyed and non-alloyed metallic fuels, inert-matrices fuels) have been irradiated and investigated 3

4 BR-10, BOR-60 Fuels BOR-60 standard fuel : MOX vibro (since 1980) Experimental fuel: Me alloyed, unalloyed Carbide UC, UPuC Nitride UN, UPuN Carbo-nitride UCN, Inert-matrices fuels, MOX pellet BR-10 standard fuel PuO2 UC UN 4

5 BN-600, BN-800 Fuel Standard fuel- MOX pellets I stage Max burn-up Max dose 10 at% 90 dpa BN-600 Standard fuel- UO 2 pellets Max burn-up 11.3at% Max dose 83 dpa (Some LFA 95 dpa) II stage Max burn-up 12.5 at% Max dose 110 dpa BN-800 Experimental MOX fuel (pellets and vibropack) 5

6 Топливо BN-1200 core is capable of using different fuels w/o changes of other reactor elements Homogeneous transmutation of MA MOX of increased density with MA Nitride with MA BN-1200 R&D Heterogeneous transmutation of MA in 42 special FAs (Inert matrices with MA) BN-1200 BN-1200(МА) Increase of burn-up (17at%-I stage, 20at%-II stage) using ODS cladding steels. FAIRFUELs Workshop, 9-10 February

7 Reactor tests to validate BN-1200 fuel performance 200 dpa up to 20 at% Post irradiation examinations Steel samples in BN-600 material assemblies (MA) Experimental fuel pins in BN-600 MA up to 15 at% Experimental fuel pins in BOR-60 FA Experimental fuel pins in BN-600 FA

8 MOX fuel MOX PELLETS Mechanical mixing of initial PuO 2 and UO 2 (MMO method) and mixed oxide co-precipitation (GRANAT method) have been developed in VNIINM, Moscow. VNIINM fabrication process of homogenous powder mixture is named as Vortex Milling Process (VR-process). Pu distribution uniformity in fuel pellet is comparable to chemically co-precipitated fuel. VR-process is universal for both oxides and nitrides of required quality. (Photos of the blend samples in characteristic X-radiation. The milling-blending time increases at left to right) 40 experimental FAs with pellet MOX irradiated in BN-600 up ~12at% MOX VIBRO The fabrication technique of vibropack MOX fuel has been developed in RIAR, Dimitrovgrad. Vibropacking fabrication and pyrochemical reprocessing permit to realize the remoteautomatic type of granulate and fuel pins fabrication. The required quality granulate is received by combined electro-crystallization of UO 2 -PuO 2 from molten alkali metal chlorides. 30 experimental FAs with vibro MOX irradiated in BN-600 up ~10.5 at%. 8

9 Gas release, % Swelling, % Nitride BR-10 reactor: 2 loadings with UN (660 and 590 fuel pins). I core design burn-up 8at%. Good fuel performance II core design burn-up (8.8 at%). 24 cases of fuel failures, mainly, at burn-up > 8at%. The reason is fuel-cladding mechanical interaction (FCMI). PIEs of 6 FA : data on swelling and gas-release from nitrides in dependence on burn-up have been received. BOR-60 reactor: 1) tens pins with UN, max burn-up > 8 at%, 2) few pins with UPuN, max burn-up 4 at %, 8.95 at %. 3) 4 pins with UPuN with increased Pu content (45 % and 60 % ) within the frame of joint Russian-French BORA-BORA experiment. Maximum burn-up 12 at%. All pins are helium -bonded. Nitride swelling 14 UN, max fuel temperature 1675 К, density 93%, 12 UN, max fuel temperature 1460 К, density 95%, UN, max fuel temperature К, density 84-94%, Calculation 8 at max fuel temperature 1675 К -, at К -, at 1173 К BOR-60 standard fuel : MOX vibro (since 1980) 25 Experimental fuel: 20 Me alloyed, unalloyed 15 UN, max fuel temperature 1675 K, density 93% USA, UN, max fuel temperature 1460 K, density 95%, USA UN, max fuel temperature 1173 K, density 85 94% BR-10 UN, max fuel temperature 1953 K, density 87%, USA 10 Carbide UC, UPuC 5 Nitride UN, UPuN 0 Carbo-nitride UCN, Inert-matrices fuels, Burn-up, at % Burn-up,at %. 9

10 BORA-BORA experiment: fabrication, irradiation in BOR-60 and PIE of high Pu content fuels Start of fabrication- 1996, irradiation , completion of PIE fuel pins with UPu 0.45 O 2 pellets VNIINM, Moscow 4 fuel pins with UPu 0.45 O 2 vibropac fuel RIAR, Dimitrovgrad 2 fuel pins with UPu 0,45 N pellets VNIINM, Moscow 2 fuel pins with UPu 0.6 N pellets VNIINM, Moscow 2 fuel pins with PuN + ZrN pellets VNIINM, Moscow 2 fuel pins with PuO 2 + MgO pellets -IPPE, Obninsk 10

11 1 BOR-60 irradiation device with removable fuel pins head 2 - spacer grid 3 pins bundle 4 - wrapper 5 - throttle 6 - adapter 7 tail 4P (Pu,U)N (121,88) A P (Pu,Zr)N (38,24) A P (Pu,U)N (122,74) A1-227 A ПИ (Pu,Mg)O 2 (18,67) 9P (Pu,Zr)N (38,82) A P A1-228 (Pu,U)N A0-113 A1-229 (107,18) A1-223 A1-234 A P (Pu,U)N (107,18) 13ПИ (Pu,Mg)O2 (18,56) FA-263E pellet fuel pins (UN-PuN) pellet fuel pins (PuO 2 - MgO) pellet fuel pins (PuN - ZrN) regular pins of disassembly FA 11

12 UPuN in BORA-BORA: PIE results UPu 0.45 N core top (а), core middle (b), a) b) Fuel type UPu 0.6 N UPu 0.45 N Irradiation time, 900/514 a) 900/514 a) efpd Max burn-up, at. %. 12.1/7.0 a) 9.4/5.4 a) UPu 0.6 N core top (c), core middle (d) c) d) Max fuel temperature, 0 С, BOL/EOL Swelling rate, % / at% Fission gas release, % Max cladding corrosion depth, µm 1760/ /930 (0,48 0,68) ±0,04% (0,64 1,11) ±0,04% no 15 ( upper section) No evident structure changing besides small intragranular porosity and coagulation of along border grain pores that confirms fuel temperature calculation correctness. No nitride dissociation was detected on microstructure and on nitrogen content changing. a) 1-st irradiation stage / 2-nd stage 12

13 UPuN in BORA-BORA: PIE results The positive results of irradiation tests of highpurity mixed nitride at BOR-60 reactor up to 12.1at% at maximum linear rating of 54,5 kw/m may be explained by high initial homogeneity of Pu distribution, low oxygen and carbon contents (<0,15 w% and < 0,1 w%), uniform porosity distribution, combination of intra-grain and along grains borders pores. The performance of nitride fuel pin as well as oxide is limited by irradiation stability of cladding steel, and also by fuel-cladding mechanical interaction (BR-10 reactor experience). The positive BORA-BORA results confirm the possibility to provide at least 12at% of burn-up for He-bonded pins at initial fuel porosity increase. The BORA-BORA nitrides porosity is 15%. 13

14 Principal results of fuels study: BORA-BORA inert matrices fuels Transmutation and incineration are innovative options in management and disposal of fission products and actinides. In order to improve the efficiency of these processes, materials inert to neutron activation are of interest. Irradiated PuN-ZrN has 2-phase structure consisting of solid solutions based on Zr and Pu nitrides differing by its proportion and fission products content. Swelling rate <0,1%/1at.%. Gas release ~ 1%. Max cladding corrosion depth ~ 15 m. Irradiated PuO2 MgO has 2-phase structure consisting of Pu and Mg dioxides with fission products dissolved. Swelling rate ~ 0.5%/1at%. Gas release - 9%. Max cladding corrosion depth < 10 m. Fuel type PuO 2 +MgO PuZrN Pu content, % Density, %th Irradiation time, efpd Max burn-up, at. %. Max cladding temperature, 0 С Max linear rating, kw/m, BOL/EOL 900/514 a) 900/514 a) 19.0/11.1 a) 19.4/11.3 a) / /18.2 a) end of 1-st irradiation stage / end of 2-nd stage 14

15 Inert matrices fuels: MATINE ISTC Project #2680 MATINE - Study of Minor Actinide Transmutation in Nitrides: modeling and measurements of out-of-pile properties. Objective - to identify the optimum fuel pin design that would perform well under irradiation in ADS fast neutron spectrum, fuel (Pu,Am,Cm,Zr)N (with ZrN=50-65at%, Pu/Am/Cm=40/50/10). Task 1: Compilation and analysis of literature data on nitride fuels and on MA fuels. The objective is to perform input data for calculation modeling of fuel performance. Task 2: Experimental study of out-of-pile PuZrN properties: high temperature stability, thermal conductivity, high temperature creep, linear expansion. Task 3: Modeling of (Pu,Am,Cm, Zr)N behavior under irradiation up to 35at% in ADS with Pb-Bi coolant (pellet for helium, sodium and lead-bismuth bonded fuel pin, vibro for helium-bonded pins only). Task 4: Technical economical assessments of fabrication feasibility of (Pu,Am,Cm, Zr)N with (60 10) at% ZrN and with 10 at% Cm at RIAR site. Participants IPPE (Obninsk), VNIINM (Moscow), RIAR (Dimitrovgrad). Collaborators CEA (France), KTH (Sweden). 15

16 Inert matrices fuels: out-of-pile properties MATINE, Task 2 (Pu,Zr)N samples with ZrN=60mol% of 85%th, 93%th density done by direct nitridation of metals. Powder mixing by VR process, solid solution has received. Study of (PuZr)N properties : High-temperature creep by uni-axial compression under very-high-purity argon. At steady-state phase the creep rate temperature dependence is satisfactorily described by Arrhenius equation. Creep rate increases linearly with load increasing. Thermal conductivity from 500 C to 1600 C - by laser flash method under vacuum.. High-temperature stability: heating and cooling rate - 600ºС/hour, holding time at maximum temperature - 20 minutes. For gas environment tests the heating started in vacuum up to 1100ºС under fill gas pressure - 5 MPa. Each test of (Pu, Zr)N samples - with standard ZrN samples. Thermal expansion of ZrN and (Pu,Zr)N samples in dilatometric system at temperatures from 20ºС to 1400ºС in argon + hydrogen. For ZrN good agreement with literature data. 16

17 MATINE Fabrication Process Finalizing MATINE Fabrication Process Finalizing Pellet Density, g/cm 3 Fabrication Process homogeneous blending of initial nitrate powders with electromagnetic vortex blender (VR-process), pelletezing, sintering. Process Finalizing ZrN as a dummy. Sintering: up to 1400 о С vacuum, then up to 1800 о С argon, heating and cooling 10 о С/min. 7,0 6,5 6,0 ZrN PuN Heat Treatment of Initial ZrN Components Mixing Mixing of PuZrN with Zinc Stearate Pellet Pressing 5,5 Pellet Sintering 5,0 Pellet Quality Control 4, Sintering Time, hours 17

18 Creep rate, 1/h Creep rate, 1/h MATINE Creep rate of PuZrN samples Samples investigated at the stress of 2 kg/mm 2 Sample Density: 93% of TD Samples Density: 85% of TD Inverse temperature, 1000/K Sample 1 Sample 2 Sample 3 Inverse temperature, 1000/K Approximation Averaged values 18

19 Thermal conductivity, W/м К Теплопроводность, Вт/м К MATINE: (PuZr)N and ZrN thermal conductivity, 100% density Temperature, С Температура, С - (Pu0,4Zr0,6)N, (наст. работа); - (U0,8 Pu0,2)N, (работа [8]; - ZrN, (работа [9]); - (Pu0,4Zr0,6)N; - ZrN. - UN, (работа [8]); - PuN, (работа [10]); - (Pu0,36 Zr0,64)N, (работа [11]; - ZrN, (наст. работа) (Pu0,25 Zr0,75)N, (работа [4]). Certain discrepancies with respective foreign data can be explained by different physicalchemical and structural samples properties due to the difference of fabrication processes 19

20 MATINE: (PuZr)N high temperature stability High-temperature stability study revealed differences in (PuZr)N behavior at 2200 C C in various environments. In vacuum fuel stability was the lowest. Presumably, Pu evaporated through matrix open porosity. In argon and nitrogen the samples revealed high stability; however separation of phases with high plutonium concentration was observed. Initial sample In vacuum In argon In nitrogen Zr Pu Zr Pu Zr Pu % мас % мас % мас мкм мкм мкм 1.Distribution of Zr (blue)and Pu (red ) from pellet edge. 2. Pu distribution in samples after testing in nitrogen at 2200 o C, in vacuum, argon at 2300 o C 20

21 MATINE - Coefficient of Thermal Expansion (CTE) of ZrN and (Pu,Zr)N Samples Sample Composition Physical СTE ( ) at 600 о С, 1/К Physical СTE ( ) at 1200 о С, 1/К Technical CTE ( ) at о С, 1/К ZrN (Pu,Zr)N dl/lo *10-3 Temp./ C T. Alpha/(1/K) 400.2, : E Physical Alpha: C,7.180E-06 1/K Physical Alpha: C,6.986E-06 1/K (Pu,Zr)N Physical Alpha: C,10.135E-06 1/K Temp./ C T. Alpha/(1/K) 400.2, : E-06 2 ZrN Physical Alpha: C,7.530E-06 1/K Temperature / C 21

22 MA fuels Russian-French AMBOINE experiment (AMericium in BOr-60: INcineration Experiment): possibility of Am recycling by pyrochemical methods has been studied. The program included fabrication of BOR-60 type fuel pin with vipac (UAm)O 2 in the core and with vipac UAmO 2 +MgO in axial blankets. Investigations on Am/REE (rare earth elements) and MgO separation by selective precipitation in molten salts have been carried out also. DOVITA program: irradiation of UNpO 2 vi-pack fuel to 20at% has been done at BOR-60. No principal difference is seen comparing MOX or uranium oxide. (UNp)O 2 after BOR-60 irradiation up to ~20 at % ISTC MATINE Project: the possibility has been considered on fabrication of (Pu,Am,Cm,Zr)N with up to 10mol% Cm on RIAR site. Electrolytic refining in molten chlorides on liquid metal cathode was proposed as main flow sheet. The technicaleconomical estimations have shown the technical feasibility of the offered processes at RIAR site. 22

23 Fuel codes development The principle code for the performance evaluation of cylindrical fuel pins is KONDOR code made in IPPE, Obninsk. On the base of this code the KORAT code in VNIINM, Moscow, for MOX pellet fuel pins and the VIKOND code in RIAR, Dimitrovgrad, for vibro MOX fuel pin. All these codes realize the calculation of thermo-mechanical characteristics of one separate axial section of a fuel pin. With the purpose of computer resources consumption minimization and calculation efficiency increase DRAKON-3D code has been developed in IPPE. KONDOR code modules are the key ones in DRAKON code structure. DRAKON code may be used for temperature and stress-strain state calculations of cylindrical fuel pins with MOX pellet and different types of dense fuels both in steady conditions and transients. It considers Nz ~ pin axial sections. For MATINE (PuAmCmZr)N fuel DRAKON code modernization: calculation scheme improving (automatic transition from w/o contact loading scheme to a contact one), development of temperatures calculation module, module of fuel swelling, module of He and fission gas release. 23

24 Stress, kg/mm2 Fuel codes development: DRAKON code For performance evaluation of fuel pins under conditions of so called rigid loading scheme typical for dense fuels the correct data on fuel swelling and creep are of principal importance as well as cladding deformation capability. BORA-BORA results of four UPuN pins PIE data are important for the verification and modification of some code modules (temperature, fuel swelling and gas release) z = 1 z = 10 z = 25 z = 40 z = Time, h Hoop stress per time in different axial section (z) of UPu 0.6 N pin 24

25 Conclusion Different fuels: PuO 2, UO 2 (pellet, vibro), UPuO 2 (pellet, vibro), UC, UN, UPuC, UPuN, oxide, nitride and carbide inert-matrices fuels, alloyed and non-alloyed metallic fuels have been studied in Russian SFRs (BN reactors). Recently the experiments with UPuN and MgO, ZrN based fuels have been completed in BOR-60. In order to meet the requirements to nuclear installations of 4-th generation: sustainability, proliferation resistance, waste management, safety, economics, the R&D on commercial BN-K reactor are currently under way in Russia. As reference fuel MOX fuel is considered, as more long-term option nitride is assumed to be. Other types of dense fuels are planned to be studied as well. 25

26 Thank you for your attention! 26