Fast Breeder Reactor (FBR)

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Fast Breeder Reactor (FBR) Fugen (JAEA) Tsuruga

MOCHIZUKI Nuclear Reactor Thermal Hydraulics

1 Fast Breeder Reactor (FBR) Fugen (JAEA) Tsuruga NPP (JAPC) Tsuruga Peninsula JAEA Tsuruga HQ INSS APWR Site Mihama NPP (KEPCO) Monju (JAEA) Research Center (JAEA) Professor H. MOCHIZUKI Nuclear Reactor Thermal Hydraulics Division Research Institute of Nuclear Engineering 1 University of Fukui

2 Outline Pu production and fuel cycle Characteristics of neutrons Comparison between LWR and FBR Physical properties of sodium FBRs in the world Components of FBR Monju 2

FBR and LWR University of Fukui FBR: Breed nuclear

239Pu 238U Liquid sodium Core is small. Temp.

Atmospheric pressure Large core Temp: ~300ºC

3 FBR and LWR University of Fukui FBR: Breed nuclear fuel ( 238 U 239 Pu) We can utilize nuclear fuel for a couple of thousand years. CR Fuel Blanket fuel LWR: 235 U of 0.7% in the uranium ore is consumed. Uranium 235 U will be consumed within 100 years. 239Pu 238U Liquid sodium Core is small. Temp. is more than 500ºC. Atmospheric pressure Large core Temp: ~300ºC High-pressure FBR LWR Pu(~20%)+ 238 U(~80%) : MOX Fuel Slightly enriched uranium ( 3~ 4% 235 U) Liquid sodium Coolant water Non Moderator water Fast neutron Fission Thermal neutron 3

Recovery U & Pu Fuel fabrication Recovery U & Pu Reprocessing Plant Fuel fabrication (U &

4 Fuel cycle FBR cycle Mine LWR cycle Uranium FBR U 3 O 8 FBR fuel Depleted uranium Depleted uranium Recovery U Enrichment Enriched U UF 6 Spent fuel Recovery U & Pu Recovery U & Pu Fuel fabrication Recovery U & Pu Reprocessing Plant Fuel fabrication (U & MOX fuels) Reprocessing Plant HLW Spent fuel LWR LWR Fuels HLW Underground storage Intermediate storage 4

5 Fission in LWR or Graphite reactor University of Fukui Moderator Water, Graphite moderation Control Fast neutron 10,000 km/s 1/30 of light speed 2.2 km/s 10,000 km/s 235 U has a large fission cross section to a thermal neutron. Some fast neutrons are absorbed in 238 U and produce 239 Pu. 5

6 Neutron irradiation chain of 238 U University of Fukui 238 U 239 U 240 U min 14.1 h Half-life period disintegration 239 Np Np 2.35 d min 62 min Fission cross section Pu 240 Pu 241 Pu 242 Pu 243 Pu (n, ) reaction 14.4 y 4.96 h Am Am Am

7 Microscopic cross section of 235 U & 238 U University of Fukui Easiness of fission From R A Knief, Nuclear Engineering 0.7% 99.3% Thermal neutron Fission cross section is large for thermal neutron. Fast neutron Fission cross section is small for fast neutron. 7

8 Cross section of neutron capture of 238 U University of Fukui 8

Number of neutrons produced by one fission University of Fukui Number of neutrons produced by a fission Thermal neutron 239 Pu 241 Pu When η is less than 2, breeding

9 Number of neutrons produced by one fission University of Fukui Number of neutrons produced by a fission Thermal neutron 239 Pu 241 Pu When η is less than 2, breeding is impossible. One must be used for a fission, and the other one should be used for breeding. Some of them escape from the reactor. Fast neutron Energy of neutron (ev) 9

10 Characteristics of FBR FBR can produce nuclear fuel more than consumption. Configuration of FBR is different from LWRs. Neutrons are not moderated in order to breed nuclear fuel efficiently. Enrichment of fuel is higher than that of LWR in order to raise fission probability. Fuels should be cooled by good heat transfer coolant (sodium) because power density is high. 10

thin due to low system pressure Mihama Unit-3

11 Comparison of reactor vessel between FBR and LWR University of Fukui Monju 238MWe Reactor vessel is thin due to low system pressure Mihama Unit-3 826MWe Reactor vessel is thick due to high system pressure 11

12 238 U+Pu(20%~30%) Driver fuel Comparison of fuel arrangement between FBR and LWR Blanket fuel of 238 U is placed in the peripheral, upper and lower region of the core in order to breed effectively. Example: 238 U(96%)+ 235 U(4%) Small core Short fuel Small diameter fuel Large core Long fuel Large diameter fuel Enrichment is high in the outer region of the core in order to flatten the power distribution. 12

Comparison of fuel assembly Fuel element Plug Handling head Spacer pad CR cluster Cladding Spacer pad Wrapper tube Plenum spring Upper nozzle Upper blanket pellet Pellet Lower blanket pellet Spacing

13 Comparison of fuel assembly Fuel element Plug Handling head Spacer pad CR cluster Cladding Spacer pad Wrapper tube Plenum spring Upper nozzle Upper blanket pellet Pellet Lower blanket pellet Spacing wire Fuel element Spacing wire Spacer pad Lower nozzle Fuel pin CR guide tube Fuel pin Spacer Cladding Pellet Fuel pin Plug Plug Length:2.8m Feul elements:162 Blanket pellets are inserted in the cladding. Entrance nozzle Length:4.2m Length:approx. 4m Effective length:approx. 3.6m Weight:670kg (17 17) 13

14 Main steam B University of Fukui Containment Vessel Control valve MSIV Comparison of turbine systems 13MPa 483 MSIV バイパス弁高圧タービン FBR 0.8MPa Low P turbine A G Generator D Condenser Sea water Condensate pump MPa at 30 PWR Containment vessel 6.4MPa 278 Feed-water pump High pressure feed-water Turbine rotor heaters Cross-over pipe Casing Control valve Main steam Extraction Casing Low pressure feed-water heaters Generator Efficiency: 40% 15.7MPa Feed water heaters Efficiency: 32% Vacuum 25 14

15 T ~300K 0 University of Fukui B Liquid B A S A Critical point MPa C 328 C Mixture D S C Super heated steam D E Saturation curve S C 487 S Rankin Cycle Higher energy - Receive Q 1 - Release Q 2 - Power output L 1 - Pump work L 2 T 0 Liq. B A S A Critical point K 22.1MPa B 280 C Mixture S C D Super heated S. E S Q Q L F i i L C D 2 i B i L1 L Q 1 A Area( BB' C' CS Area( ADS A) B) ic id ib ia Q1 Q2 2 Q1 Q Q 1 2 C S A C S A Area( ABB' C' CDA) Area( BB' C' CS S B) C A Q Q L i L C' i D' 2 i B i L1 L2 L Q 1 A Area( BB' C' S Area( AD' S ' S ' S A) B) ic ' id' ib ia Q1 Q2 Q1 Q Q 1 2 C C A A Area( ABB' C' D' A) Area( BB' C' S ' S B) C A 15

16 FBR & LWR FBR LWR Neutron life time 10-7 seconds 10-5 seconds Ratio of delayed neutron 0.34~0.37% 0.55~0.7% Mean free path long short Coolant Sodium (low corrosive) Water (high corrosive) Fuel UO 2,, PuO 2 (Enrichment ~ 20%) UO 2 (Enrichment 3~ 4%) Cladding S.S. Zircaloy Dia. of fuel pellet 4~6mm Approx. 1cm Outlet coolant 500~ ~320 temperature Temperature difference 130~ (B)~35 (P) System pressure Atmospheric 7MPa(B) 15MPa(P) Power density Approx. 300kW/l Approx. 90kW/l Burn-up 100,000MWd/t 30,000~60,000MWd/t Efficiency Approx. 40% Approx. 32% 16

Characteristics of sodium Sodium : alkali metal

Boiling point : 881.5ºC at atmospheric pressure.

17 Characteristics of sodium Sodium : alkali metal which is soft and has metallic color. Weight of sodium : 0.97 times of water at 20 ºC. Melting point : 98ºC(97.8 ºC ). Boiling point : 881.5ºC at atmospheric pressure. Sodium is lighter than water. Cut by a knife easily Liquid sodium 17

2Na + 2H 2 O 2NaOH+ H 2 (Hydrogen)+Heat Therefore, sodium must be separated from water.

18 Reaction of sodium with water and air University of Fukui When sodium reacts with water, hydrogen gas and NaOH are produced. 2Na + 2H 2 O 2NaOH+ H 2 (Hydrogen)+Heat Therefore, sodium must be separated from water. Leak speed of sodium is slow due to atmospheric system pressure. However, sodium reacts with air as shown in photos, and produces a lot of white alkali aerosol. 18

Blow-down of high-pressure and high-temperature

of hightemperature and high-pressure light water

heat-up fuel Even a small amount of steam leak, a

This is a different point compared to a leak of

19 Blow-down of high-pressure and high-temperature coolant University of Fukui Blow-down: Discharge of hightemperature and high-pressure light water ECC water injection in order to cool down the heat-up fuel Even a small amount of steam leak, a jet is very dangerous. This is a different point compared to a leak of sodium. 9 Aug 2004 at Mihama NPP, 5 people were killed and 6 were seriously injured. 19

20 Sodium leak from secondary system on 8 Dec University of Fukui Leak location IHX Reactor Evaporator Super heater Debris of sodium from a broken thermocouple well Air cooler Pump for secondary system 20

21 Reason of superiority of Na among liquid metal Na K NaK Li Pb Bi Pb/Bi Hg University of Fukui (70/30) (eutectic) Melting point( ) Boiling point( ) Vapor pressure(600 )(mmhg) (atm) Neutron absorption cross section Thermal neutron (barns) Fast neutron (100eV)(mb) 1.1 5(400eV) Half-life period 15hr. 12.5hr. 0.8sec. 3.3hr. 5days 5.5min. Thermal conductivity (600 )(cal/cm/sec/ ) Specific heat(600 )(cal/g/ ) Density (600 )(g/cm 3 ) Heat transport capability (4in.φPipe)(C.H.V./ft 2 sec.) Pumping forth (1/ρ 2 c 3 ) Price ( /ft 3 )

22 Sodium cooled fast reactors University of Fukui Loop type FBR Monju Tank type FBR Phénix Secondary sodium Main motor Primary & Pony motor circulating Primary pump sodium Air cooler (AC) Secondary circulating pump Super heater (SH) Evaporator (EV) Turbine Generator Condenser Sea water cooler Core IntermediateHeat Exchanger (IHX) Primary Heat Transport System (PHTS) Secondary Heat Transport System (SHTS) Feed water pump Turbine system 22

23 FBR around the world University of Fukui DFR, PFR (Therso) Super Phénix BN-600, 800(Belouarsk) (Crays-Malville) BOR-60(Dimitrovgrad) BN-350(Aktau) Phénix (Marcoule) CEFR(Beijing) Monju (Tsuruga) Joyo (O-arai) FFTF(Hanford) EBR-Ⅱ(Idaho falles) FBTR, PFBR (Kalpakkam) :Closed :in operation (including outage) 23

24 First power plant in the world (USA) University of Fukui EBR-I (Experimental Breeder Reactor) Coolant was NaK. Four light bulbs were lit up on 20 December

25 Donreay Fast Reactor, PFR (UK) University of Fukui Rating: 60 MWt/15MWe Coolant: Na-K Loops: 24 Prototype Fast Reactor 25

26 Phénix (France) Oldest power reactor in France. Rating: 565MWt/255MWe Coolant: Na Loops: 3 26

27 Tank-type FBR(Phénix) 27

28 Super-Phénix (France) Electrical power: 1200MWe Demonstration plant in France Super-Phénix was sacrificed by Prime minister Jopspin on 2 February 1998 in order to have a coraboration between Socialist party and green party. 28

29 Super-Phénix Snapshots in side the plant were allowed after the workshop at SPX. This might be the first and the last chance. 29

30 FBTR (India) 40 MWt /13.2 MWe Fuel Mark I ( 25 Nos.) 70% PuC + 30% UC Mark II ( 13 Nos.) 55% PuC + 45% UC PFBR test SA 29% PuO % UO 2 From IAEA-TECDOC

PFBR (India) 01 Main Vessel 02 Core Support Structure 03 Core Catcher 04 Grid Plate 05 Core 06 Inner Vessel 07 Roof Slab 08 Large Rotating Plug 09 Small

31 PFBR (India) 01 Main Vessel 02 Core Support Structure 03 Core Catcher 04 Grid Plate 05 Core 06 Inner Vessel 07 Roof Slab 08 Large Rotating Plug 09 Small Rotating Plug 10 Control Plug 11 CSRDM / DSRDM 12 Transfer Arm 13 Intermediate Heat Exchanger 14 Primary Sodium Pump 15 Safety Vessel 16 Reactor Vault 31

32 PFBR in

33 BN600 (Russia) Beloyarsk NPP is close to Ekaterinburg. Pressure tube type reactor Site for BN800 BN600 33

34 BN600 Modular type steam generator Irradiation of vibro-pack fuel (Dismantled War head was used.) From IAEA-TECDOC

35 BN600 From IAEA-TECDOC

36 BN800 Construction site 36

37 CEFR(1/3) 37

38 CEFR(2/3) 38

39 CEFR(3/3) 39

40 Core of Monju 40

41 Fuel subassembly University of Fukui Length:2.8m Length:4.2m Handling head: for the gripper of refueling machine Spacer pad: keep clearance to the adjacent wrapper tube Wrapper tube: keep flow rate and protect fuel subassembly Entrance nozzle: adjusting flow rate. Many orifices to prevent blockage 41

42 Shielding plug Hole for upper core structure Hole for refueling machine 42

43 UIS and refueling machine 43

44 Arrangement of initial core Driver fuel subassemblies (fuels for driving the core) Blanket fuel subassemblies (fuels for breeding) Inner core Outer core Blanket PuO 2 20% PuO 2 25% 238 U C F B Neutron source Control rod B 4 C 44

45 Flow distribution mechanism Monju Wrapper tube Core support plate Phénix Slit Connection pipe Orifices of entrance nozzsle High pressure plenum Entrance nozzle High pressure plenum Low pressure plenum Low pressure plenum 45

A part of a fuel structure fell down at the inlet of the fuel assembly and

46 Fuel subassembly of Fermi reactor University of Fukui A fuel melt accident occurred in October A part of a fuel structure fell down at the inlet of the fuel assembly and blocked the entrance. Orifices were provided in order to prevent total blockage. This is a lesson learned from the accident. 46

47 Piping of primary heat transport system University of Fukui Piping is winding in order to absorb stress due to elongation caused by high temperature coolant flow. 47

Intermediate heat exchanger Inlet of secondary side Outlet of secondary side Primary Secondary Flowrate 5120t/h 3730t/h Inlet temp. 529 325 Outlet temp.

48 Intermediate heat exchanger Inlet of secondary side Outlet of secondary side Primary Secondary Flowrate 5120t/h 3730t/h Inlet temp Outlet temp Configuration of heat transfer tube Number of HT tubes Rating OD 21.7mm, thickness:1.2mm, length:6.07m MW Height 12.1m Approx.12m Approx. 6m Inlet of primary side Outlet of primary side 48

Monju Seban-Shimazaki [1] Seban & Shimazaki (Nu=5+0.025Pe 0.

49 Heat transfer in IHX 100 [3] Nu [1] Nu(1) 50MWSG Nu(1) Joyo Nu(1) Monju Nu(2) 50MWSG Nu(2) Joyo Nu(2) Monju Seban-Shimazaki [1] Seban & Shimazaki (Nu= Pe 0.8 ) [2] Martinelli & Lyon (Nu= Pe 0.8 ) [3] Lubarsky & Kaufman (Nu=0.625Pe 0.4 ) Pe 49

50 Inlet windows University of Fukui Inlet window Inlet of primary sodium Shell Outlet of primary sodium CFD analysis of IHX Rectifying mechanism 小孔 Nozzle Small holes more than are meshed at present. Plate #1 #2 #3 #4 #5 #6 #7 Bank of heat transfer tubes Rectifying plate Small holes Inlet window Plates (7) Outlet window 50

51 3 4 University of Fukui 2 5 Analysis using 60 sector model 1 6 Inlet window Heat transfer tubes Plate #1 #2 #3 Secondary outlet Primary inlet 48% flowrate #4 #5 Velocity 0.65 Temp. 486 Outlet window Flowrate ratio Primary outlet M. Takano and H. Mochizuki, CFD Flow Pattern Analysis on Primary-side of IHX for Fast Reactors, ICONE-19, 43412, Chiba, Japan (2011). #6 #7 Secondary inlet 0 m/s

52 Internals of IHX 52

53 Primary main pump Intake pressure should be positive in order to prevent cavitation. Liquid surface is formed inside the casing of the pump. Height of the pump is restricted by this surface height. 53

54 Primary main pump Impeller blade 54

55 Electoro-magnetic pump Fabrication of a large EM pump is difficult. If one can fabricate a large EM pump, the location problem will be solved. Maintenance becomes easy. 55

56 Steam generator 140 heat transfer tubes are coiled. 56

57 Super heater used at 50MW SG facility University of Fukui Sodium flow Inner duct Helically coiled heat transfer tube Water flow 57

58 Approx. 30m University of Fukui Air cooler for auxiliary system Design: 15MW, Real rating :20MW Outlet damper (controlled with inlet vanes) Heat transfer tubes with fins ~4.5m ~5.3m ~6.5m Inlet damper Inlet vanes Blower 58

59 H.T. C. of Finned Heat Transfer Tube University of Fukui Nu hd k a 1000 e MWSG Monju Joyo Data by Jameson Turbulent +10% -10% Carbon S. Copper S. S. Nu/Pr 1/3 10 Laminar Nu/Pr 1/3 =0.1370Re % -20% Nu/Pr 1/3 = Re Re 59

60 Inter-subassembly Heat Transfer Model University of Fukui Sodium flow in subassembly t δ g Inter-wrapper flow T k-1 Center 1 U Nu k liq hd k e liq g h g 5 t k s Pe 0. 8 T k T k T k+1 k: thermal conductivity h: heat transfer coefficient k+1 th layer k th layer k-1 th layer 60

61 Heat Transfer to the Concerned Channel University of Fukui Q j m 6 i 1 N n m,i l z j U j m T j n,i T j m T n,i j T m j Layer k+1 T o,i j j: concerned axial mesh m: concerned channel group l: width of hexagonal wrapper tube z j : length of mesh j N m n,i : number of subassemblies of channel group n facing to the face i of channel m Layer k 61

62 Exit temperature of 3 rd Layer Subassembly University of Fukui Temperature ( deg-c ( ) ) SSC-L Measured NETFLOW++ with ISHT model NETFLOW++ without ISHT model Time (sec) 62

63 Downsizing of FBR Volume of reactor building = 810,000 m 3 Prototype FBR Monju Thermal output 714 MWt Electrical output 280 MWe Gen-IV FBR (JSFR) Thermal output 3570 MWt Electrical output 1500 MWe 63

64 Combined IHX and pump Pump support Motor primary in primary out secondary out Bellows NSSS Pump can be withdrawn. DHX IHX heat transfer tubes Pump secondary in IHX support 64

65 IAEA benchmark problems #1) Natural circulation test at Phenix and #3) LOF&ULOF tests at EBR-II University of Fukui IHX Inlet temp. Mochizuki Lab. was only organization in Japan and only university in the world who participated in the calculation. IHX Outlet temp. Our result is at the midst of the calculation curves Normalized flowrate (-) Measured Calculated Flow rate change Time (sec) H. Mochizuki, N. Kikuchi and S. Li, Calculation of Natural Convection Test at Phénix using the NETFLOW++ Code, Proceeding of International Congress on Advances in Nuclear Power Plant (ICAPP 2012), Chicago, Paper 12104, USA, (2012-June). The benchmark input deck was created on the basis of plant configuration and fuel patterns of EBR-II. Now this analysis a kind of blind test. However left result was presented in an international conference. 65

IAEA benchmark problem #2 about thermal stratification in upper plenum of Monju University of Fukui TC plug Upper instrumental structure Inner

66 IAEA benchmark problem #2 about thermal stratification in upper plenum of Monju University of Fukui TC plug Upper instrumental structure Inner barrel Flow holes Reactor vessel Fuel handling machine Reactor outlet nozzle Upper core structure Core SA Flow guide tube Surface of upper support plate 66

67 Overall view of the CAD model University of Fukui 4.38 m 5.30 m Upper flow-holes (24) Lower flow-holes (48) 67

68 CAD model around UCS Model-1 UIS Model-2 Model-3 Control rod guide tube Honeycomb structure Flow guide tube 68

69 Mesh around a flow hole 69

70 Configuration of the reactor core at the test University of Fukui Inner core 108 Outer core 90 Blanket 172 Neutron shield 324 Neutron source 2 Control rod (Coarse) 10 Control rod (Fine) 3 Control rod (Back-up) 6 IAEA Classification Ch. 1-5 Ch. 6-8 Ch.9-11 Ch.12,13 and FC Ch. N Ch. C Ch. F Ch. B 70

71 Contour maps from 0 sec to 20 min. 0 sec 60 sec 120 sec 300 sec 600 sec 1200 sec [m/s] [ºC] 71

(Measured) (SKE) (SKE) (SKE) (SKE) (SKE) (RKE) (RKE) (RKE) (RKE) (RKE) 0 Elevation from the

72 Results with boundary conditions calculated by 1D code and with rounded edge University of Fukui Time: 0 sec s (Measured) 120 s (Measured) 300 s (Measured) 600 s (Measured) 1200 s (Measured) (SKE) (SKE) (SKE) (SKE) (SKE) (RKE) (RKE) (RKE) (RKE) (RKE) 0 Elevation from the surface (mm) Temperature ( o C) 72

Safety Practices in Chemical and Nuclear Industries

Lecture 10 Safety Practices in Chemical and Nuclear Industries Fast Breeder Reactor (FBR) & Sodium-Water Reaction (SWR) Dr. Raghuram Chetty Department of Chemical Engineering Indian Institute of Technology