Physics Design Studies for Indian MSRs

Transcription

1 Physics Design Studies for Indian MSRs D.K. Dwivedi, A.K. Srivastava, Indrajeet Singh, Anurag Gupta and Umasankari Kannan Reactor Physics Design Division, Bhabha Atomic Research Center, Mumbai , India Technical Meeting on the Status of Molten Salt Reactor Technology 31 October to 3 November, 2016

2 Outline of the presentation History of MSRs Molten Salt Reactor concepts worldwide Current Interests Why resurge in interest? MSR among 6 reactors in GIF Indian interest better utilization of thorium in 3 rd gen reactors Development and Studies done at BARC Major R&D work initiated Physics design studies Code validation IMSBR -Loop type and pool type design Simplified model of Refueling, effect of Protactinium removal, Axial Precursor distribution Zero Power dynamics Summary

3 Experimental reactors with Molten Salt Fuel Aircraft Reactor Experiment : 1954 (100 hrs), US Power : 2.5 MW th Molten fuel : NaF-ZrF 4 -UF 4 ( mol %) Enrichment : 93.4% in 235 U Peak temperature : 860 o C Fuel Temperature Coefficient : -9.8 pcm/k Ref.: E.S. Bettis et al. NSE, 2, (1957) Reactor core top view Molten Salt Reactor Experiment : , US Power : 8 MW th Molten fuel : 7 LiF-BeF 2 -ZrF 4 -UF 4 ( mole %) (also used 233 U and 239 Pu in later part of experiment) Secondary coolant: LiF-BeF 2 (66 34 mole %) Enrichment : 33% in 235 U Mean temperature : 650 o C Ref.:Paul N. Haubenreich et al. NAT, 8, (1970) Schematic of MSRE plant

4 Operating Parameters Thermal/electrical power Values Power density 87.4 Wcm -3 Fluoride salt comp (mol%) 2250 MWth/1000 MW(e) (HN)F %- 7 LiF72%- BeF 2 16% Salt volume inside/outside m 3 /48.7 m 3 Fuel &graphite temperature 635 C Breeding ratio 1.05 Fuel processing scheme Fuel inventory Doubling time On-line continuous processing 1500 kg 19 years vessel height/ diameter (m) 4.6 / 4.3 Core height/diameter (m) 3.8 / 3.05 Vessel design pressure Fuel velocity Ref.: E.S. Bettis et al. NAT, 8, (1970) The Molten Salt Breeder Reactor (MSBR) 0.5 MPa 4.6 m/s Fig.: Schematic of MSBR

5 MSR concepts worldwide Family Concepts Spectrum Fuel Cycle MSR- (Breeder/ Nearbreeder) MSR- Burner Power (MW th ) Comments MSBR T 233U-Th 2250 BR ~ 1.05; FRC > 0 (Slightly +ve) AMSTER-B T 233U-Th 2250 BR > 0.95 REBUS* F U-Pu 3700 BR ~ 1.03; FRC < 0 FUJI T 233U-Th 450 BR ~ 0.97; FRC < 0 TMSR T, E, F 233U-Th 2500 BR > 1 & FRC < 0 in both T, F MSFR # F 233U-Th 3000 BR~ 1.12 & FRC < 0 AMSTER-I T U-Pu-MA 2250 SPHINX F Pu-MA 1208 MOSART F Pu-MA 2400 FRC ~ -3.9 pcm/k China launched 2 MW th research Thorium Molten-Salt fuelled and cooled Reactor (TMSR) in 2011 * Chloride based fuel salts # MSFR derived from non-moderated TMSR

6 Work on MSBR at BARC was carried out in collaboration with ORNL in 1970s Preparation of pure ThF 4, LiF salts Solubility of PuF 3 in LiF- BeF 2 -ThF 4 salt Thermodynami cs of U-Bi alloys and vapour pressure measurements Work was suspended when the ORNL programme was shutdown 6

7 Current Interests in MSRs Thorium Fuel Cycle : 3 rd Stage Indian Program It is possible to have MSR breeder with Th- 233 U fuel cycle in Thermal, Epithermal and Fast spectrums Better utilization of Thorium in Indian third stage of Nuclear programme. 234 Pa 6.75 h n, γ (40 b) 232 Th 1.41x10 10 y 233 Th 22.3 m 233 Pa 27 d n, γ (7.4 barn) β - n, γ (1500 b) β Th 24.1 d 233 U 1.41x10 5 y Renewed interest in MSR due to inclusion among 6 promising reactor concepts by GIF Capable of meeting diverse needs such as breeding, burning actinides etc. Operation at higher temperature makes it suitable for hydrogen production Offers many advantages over other conventional reactor Technology improvements worldwide Fig.: Thorium conversion chain to 233 U

8 MSBRs are an attractive option for the third stage of the Indian nuclear power programme Fig.: Evolution of installed capacities of various reactors in the Indian context, as calculated by TEPS, considering nuclear material supply limitations Indicative case study and is neither a statement of targets of the country nor any commitment of installation 8

9 Major R&D initiated at BARC Code development for coupled thermal hydraulic physics analysis Chemistry related areas, salt purification, fission product solubility, electrochemistry Materials related development, corrosion studies and qualification of material as per design code Component and instrumentation development and qualification Reprocessing studies 9

10 Facility for handling of active salts Active salt preparation and purification facility Inert gas gloveboxes with closed loop purification and recirculation system LiF-ThF 4 prepared and purified using HF and H 2 in this facility 10

11 Facility for handling inactive salts Facility for electrochemical studies and component development in inactive salts Inert gas gloveboxes with closed loop recirculation and purification system Development of electrochemical tools for monitoring of salt condition 11

12 Thermal hydraulic and corrosion testing facility using active salts MAFL: Natural circulation loop for thermal hydraulic studies using LiF-ThF4 MAF-Corr: Static corrosion studies using LiF-ThF4 12

13 Thermal hydraulic and corrosion testing facility using inactive salts SAFETY TANK EXPANSION TANK FILTER COOLER HEATER CONTROL VALVE Molten Salt Corrosion Test Facility: Corrosion studies using in-active salts MELT TANK Molten Salt Natural Circulation Loop: to study natural circulation behaviour of active salts 13

14 Facilities for thermo-physical evaluation of salts DTA 432C 550C HeatFlow (/mw) TG Fuel Salt Mass Change (/mg) Drop calorimeter Sample temperature (/ C) DTA plot of eutectic composition High temperature viscometer Differential Scanning Calorimeter 14

15 Physics studies carried out at BARC Validation of Tools : MSFR (French) MSRE (ORNL, US) Indian concepts: IMSBR-Loop Type IMSBR-Pool Type

16 Molten Salt Fast Reactor : French Design Fuel Salt : LiF (77.5%)-ThF 4 (20%)- 233 UF 4 (2.5%) Blanket Salt : LiF (77.5%)-ThF 4 (22.5%) Core Dimension (m): 2.2 X 2.2 Power : 3000 MW th / 1300 MW e Power density : 330 W th / cm 3 Breeding Ratio : 1.12 Initial fissile inventory : 3.26 T/GW e Mean Fuel salt temperature : 750 o C Ref.: D. Heuer et.al. ANE64 (2014)

17 MSFR core simulation : Results MSFR found CRITICAL for given composition Breeding Ratio : 1.17 Initial fissile inventory : 3.23 T/GW e 1E-5 Flux 1E-6 Flux(E*dφ/dE) 1E-7 1E-8 1E-9 Spectrum from published in paper D. Heuer et.al. ANE64 (2014) E Fig.: Spectrum predicted in our simulation Energy (ev)

18 Analysis of MSRE with indigenous code ARCH Fig.: Radial and axial distribution of two group fluxes at core mid-plane and 8.4 in. from core centre line respectively as calculated by ORNL Ref.: Haunbenreich, P.N. et al., MSRE Design and Operations Report-III, ORNL,1964 Fig.: Radial and axial distribution of two group fluxes with (ARCH + DRAGON)

19 Indian Molten Salt Breeder Reactor Salient design guidelines for arriving at conceptual design of IMSBR No beryllium To avoid chemical toxicity Simplified design of experimental facilities (no need to protect against beryllium in experimental facilities) Avoid BeF 2 (i.e. Avoid LiF-BeF 2 ), which was used in earlier ORNL development Aim to optimise fissile material inventory Minimise waste generation Avoid /reprocess graphite (which was used in earlier reactors as moderator material) Replaceability & inspectability of in-core components Enhanced inherent safety Large scale deployment in third stage (locate near population centres) Initial design for large power Demonstration facility at lower power to demonstrate all systems Conceptual Design Report issued for 850 MWe 19

20 IMSBR Loop type: Major design parameters Attributes Parameter 1 Power 850 MWe 2 Thermal efficiency 45% 3 Active core diameter/height 2m / 2.05m 4 Core inlet/outlet 700 / 800 C 5 Fuel salt LiF-ThF 4 -UF 4 6 Blanket salt LiF-ThF 4 7 Secondary salt LiF-NaF-ZrF 4 8 Flow rate (primary) 10.9 t/s 9 Velocity (core) 0.85 m/s 10 Fuel salt inventory (total) 41 t (2.7 t of 233 U) Design of a low power demonstration IMSBR is in progress with temperature limited to 700 C, and Ni-Mo-Cr-Ti based alloy 20

21 Selected salts Blanket salts LiF-ThF 4 (22.4 mole % of ThF 4 )[ % by wt] (Current reference) Liquidus: 568 C LiF-NaF-ThF 4 ( ) [ by wt] Liquidus: 505 C LiF-CaF 2 -ThF 4 ( )[ by wt] Liquidus: 510 C NaF-CaF 2 -ThF 4 (Eutectic composition not known) Fuel salt Same as blanket salt, but with UF 4 dissolved as required Coolant salts LiF-NaF-ZrF 4 ( ) [ by wt](current reference) Liquidus: 436 C LiF-NaF-ZrF 4 ( ) [ by wt] 21

22 Physics design simulation of IMSBR - Loop Type IMSBR: design parameters and results Size: 2 X 2.05 (m) Power: 850 MW e Fuel Salt: LiF(77.6%)- ThF 4 (19.7%)- 233 UF 4 (2.7%) Blanket Salt: LiF (77.6%)-ThF 4 (22.4%) Core Average Temperature: 750 o C K-eff: ± Enrichment: ~ 12% ( 233 U in heavy metal) Initial Fissile Inventory in core: 1.9 T Initial Conversion Ratio: m 2 m Schematic cross-sectional view of core Longitudinal view

23 Height IMSBR - Loop Type (continue..) Core radius Partition (Ni-W-Cr) Blanket thickness 200 cm 100 cm 2.5 cm 50 cm 2.5 cm Shell thickness (Ni-W-Cr) Case - 1 Composition (LiF-ThF4-UF4) 77.6% -19.7% -2.7% He % (0.5%) K-eff ICR_fuel ICR_blanket ICR Inventory 233U 1.9 T Enrichment ( 233 U in HM) ~ 12 % Case - 2 Composition (LiF-ThF4-UF4) 77.6% -19.9% -2.5% He % (0.5%) K-eff ICR_fuel ICR_blanket ICR Inventory 233U 1.76 T Enrichment ( 233 U in HM) ~ 11.1%

24 IMSBR - Loop Type (continue..) Spectrum in blanket region Spectrum in core region Normalised Flux Neutron Energy (ev) Fig.: Comparison of spectrum in core region and blanket region of revised IMSBR core

25 Indian Molten Salt Breeder Reactor: IMSBR - Pool Type Fig.: Schematic of pool type IMSBR with natural circulation ( Ref.: A. Borgohain et al. ThEC15, 2015)

26 Analyses and Results: IMSBR : Pool Type (continue..) In presence of B4C lining Composition: LiF-ThF4-UF4 He (%mol) K eff ICR ( fuel ) ICR (blanket ) ICR (total ) 77.6% -19.9% -2.5% % -19.7% -2.7% % -19.9% -2.5% B4C lining replaced by blanket salt Composition: LiF-ThF4-UF4 He (% mol) K eff ICR (fuel) ICR (blanket) ICR (total) 77.6% -19.9% -2.5% % -19.7% -2.7% % -19.9% -2.5%

27 IMSBR : Pool Type (continue..) Fuel Inventory & neutron spectrum Total molten salt fuel : 63 T (4.3 g/cc) (excluding fuel in IHX region) Compo.: LiF-ThF4-UF4 U-233(excluding IHX) U-233 (including IHX) Th % -19.9% -2.5% 4.1 T 5.03 T T 77.6% -19.7% -2.7% 4.44 T 5.44 T T IMSBR pool type : Natural Circulation case Core Blanket Normalized flux Neutron Energy (ev) Fig.: Comparison of normalized flux in core and blanket

28 Refuelling Studies A simple Model The effect of refueling of fresh fuel and removal of burned fuel has been taken into account by appropriately adjusting each nuclide number density by using the following relation: K-eff Refuelling at 5 & 7 FPD 12600_litres_7FPD 10000_litres_7FPD 5000_litres_7FPD 1000_litres_7FPD 500_litres_7FPD 200_litres_7FPD 80_litres_7FPD 40 litres_7fpd 5000 litres_5fpd Fresh Fuel Where,, and are number density of mixed, fresh and burned fuel of i th nuclide., and are volume of fresh, removed and total fuel Burnup (FPD) Fig.: Two step refueling & effect of removal of different amount of fuel Pa-233 removal at 800 FPDs The reactivity effects of addition/removal of individual nuclide e.g. Pa-233 can be estimated by suitably adjusting the number density of nuclide. K-inf Burnup (FPD) Fig.: Effect of Pa-233 removal at 800 FPD and 820 FPD

29 Time Independent Precursor Distribution in circulating fuel reactor Time dependent multi group neutron diffusion equation can be written for circulation fuel reactor as 1 vv gg φφ gg tt GG = DD gg φφ gg + χχ pp,gg 1 ββ ννσ ff φφ gg gg + χχ dd,gg,ii λλ ii CC ii + φφ gg Σ gg gg φφ gg Σ rr,gg ii gg=1 II ii=1 CC ii tt = UU CC ii λλ ii CC ii + ββ ii ννσ ff φφ gg gg GG gg=1 gg 1 gg =1 The time independent equation with fuel velocity U in one-group diffusion theory with one group delayed neutron is shown as following DD gg φφ 0 (zz + ννσ ff 1 ββ Σ aa zz ]φφ 0 zz + λλ CC 0 = 0 UU dddd 0 dddd = λλcc 0 zz + ββββσ ff φφ 0 zz CC 0 zz = ee λλλλ UU ββββσ ff UU 1 ee λλλλ 1 HH λλλλ ee UU 0 zz φφ 0 zz dddd + ee λλλλ UU 0 φφ 0 zz dddd

30 Delayed Neutron Precursor Distribution in circulating fuel reactor Normalized flux distribution along z-axis Precursor distribution along z-axis for several fuel velocities

31 Zero and One Dimensional Model Point kinetics equation in case of circulation fuel reactor can be written with modified precursor equation as follow: ddcc ii (tt) dddd dddd(tt) dddd = ρρ tt ββ Λ = ββ ii Λ nn tt λλ ii CC ii tt CC ii tt ττ cc II nn tt + λλ ii CC ii tt ii=1 + CC ii tt ττ ee ττ cc exp λλ ii ττ ee 0-D Model dddd(tt) dddd CC ii (zz, tt) tt = ρρ tt ββ Λ + UU CC ii(zz, tt) zz nn tt + λλ ii CC ii tt = λλ ii CC ii + ββ ii Λ II ii=1 nn(zz, tt) 1-D Model Boundary Conditionφφ 0 0 = φφ 0 HH = 0and CC 0 0 = CC 0 HH ee λλλλ LL.

32 Axial DNP distribution at t = 0 6 Group DNPs distribution with axial height for stationary fuel for U Group DNPs distribution with axial height (z) for recirculating fuel for U-235 at 0.24 cm/s

33 Effective delayed neutron calculation in circulating fuel reactor Beta effective calculation using analytical model in Molten Salt Fast Reactor (French) In CFR, β eff differs from physical delayed neutron fraction β 0 for two reasons 1 st difference in spectrum of delayed and prompt neutron 2 nd delayed neutron precursors are transported with fuel salt flow Analytical model has been adopted to compute βeff in Circulating Fuel Reactor.* The graph shown for U-235 fuel and nominal flow rate is 1m/s. * Ref.: Manuele Aufiero et al. ANE 65 (2014),

34 Summary IMSBR is an attractive option for the third stage of the Indian nuclear power programme Indian MSBR programme will aim for a design which provides safety features consistent with the need for large scale deployment in the third stage Developmental activities in many areas have been initiated at BARC Physics design feasibility study of IMBRs are being carried out The existing neutronics codes are being validated and efforts are being taken for in-house development of transient simulation tools for Circulating Fuel Reactors 34

35 Acknowledgment: I acknowledge the support of Shri Abhishek Basak, RED, BARC for providing information regarding Engineering portion of the presentation.

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