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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-400085, India [email protected]
Technical Meeting on the Status of Molten Salt Reactor Technology 31 October to 3 November, 2016
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
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 3rd 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
Outline of the presentation
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Experimental reactors with Molten Salt Fuel Aircraft Reactor Experiment : 1954 (100 hrs), US
•Power : 2.5 MWth • Molten fuel : NaF-ZrF4-UF4 (53-41-6 mol %) • Enrichment : 93.4% in 235U • Peak temperature : 860 oC • Fuel Temperature Coefficient : -9.8 pcm/K
Ref.: E.S. Bettis et al. NSE, 2, 804-825 (1957)
Molten Salt Reactor Experiment : 1965-1969 , US •Power : 8 MWth • Molten fuel : 7LiF-BeF2-ZrF4-UF4 (65-29.1-5-0.9 mole %) (also used 233U and 239Pu in later part of experiment) • Secondary coolant: LiF-BeF2 (66–34 mole %) • Enrichment : 33% in 235U • Mean temperature : 650 oC
Ref.:Paul N. Haubenreich et al. NAT, 8, (1970)
Reactor core top view
Schematic of MSRE plant
Fig.: Schematic of MSBR
Operating Parameters Values Thermal/electrical power 2250 MWth/1000 MW(e) Power density 87.4 Wcm-3 Fluoride salt comp (mol%) (HN)F412.3%-7LiF72%-
BeF216% Salt volume inside/outside 25.73 m3/48.7 m3 Fuel &graphite temperature 635°C Breeding ratio 1.05 Fuel processing scheme On-line continuous
processing Fuel inventory 1500 kg Doubling time 19 years vessel height/ diameter (m) 4.6 / 4.3 Core height/diameter (m) 3.8 / 3.05 Vessel design pressure 0.5 MPa Fuel velocity 4.6 m/s
The Molten Salt Breeder Reactor (MSBR)
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244 Ref.: E.S. Bettis et al. NAT, 8, (1970)
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Family Concepts Spectrum Fuel Cycle
Power (MWth)
Comments
MSR- (Breeder/ Near-breeder)
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
MSR-Burner
AMSTER-I T U-Pu-MA 2250
SPHINX F Pu-MA 1208
MOSART F Pu-MA 2400 FRC ~ -3.9 pcm/k
MSR concepts worldwide
* Chloride based fuel salts # MSFR derived from non-moderated TMSR
• China launched 2 MWth research Thorium Molten-Salt fuelled and cooled Reactor (TMSR) in 2011
Work on MSBR at BARC was carried out in collaboration with ORNL in 1970s
6
Work was suspended when the ORNL programme was shutdown
• Preparation of pure ThF4, LiF salts
• Solubility of PuF3 in LiF-BeF2-ThF4 salt
• Thermodynamics of U-Bi alloys and vapour pressure measurements
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Current Interests in MSRs
• 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 world-wide
Thorium Fuel Cycle : 3rd Stage Indian Program
• It is possible to have MSR breeder with Th-233U fuel cycle in Thermal, Epithermal and Fast spectrums
• Better utilization of Thorium in Indian third stage of Nuclear programme.
232Th 1.41x1010 y
233Th 22.3 m
233Pa 27 d
233U 1.41x105 y
234Pa 6.75 h
n, γ (7.4 barn)
β- n , γ (1500 b)
β-
234Th 24.1 d
n , γ (40 b)
Fig.: Thorium conversion chain to 233U
MSBRs are an attractive option for the third stage of the Indian nuclear power programme
8
Indicative case study and is neither a statement of targets of the country nor any commitment of installation
Fig.: Evolution of installed capacities of various reactors in the Indian context, as calculated by TEPS, considering nuclear material supply limitations
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
• 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
Major R&D initiated at BARC
9
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Facility for handling of active salts
10
Active salt preparation and purification facility Inert gas gloveboxes with closed loop purification and recirculation system LiF-ThF4 prepared and purified using HF and H2 in this facility
Facility for handling inactive salts
11
Facility for electrochemical studies and component development in in-active salts Inert gas gloveboxes with closed loop recirculation and purification system Development of electrochemical tools for monitoring of salt condition
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Thermal hydraulic and corrosion testing facility using active salts
12
MAFL: Natural circulation loop for thermal hydraulic studies using LiF-ThF4 MAF-Corr: Static corrosion studies using LiF-ThF4
Thermal hydraulic and corrosion testing facility using inactive salts
13
MELT TANK
COOLER
EXPANSION TANK
HEATER
FILTER
SAFETY TANK
CONTROL VALVE
Molten Salt Natural Circulation Loop: to study natural circulation behaviour of active salts
Molten Salt Corrosion Test Facility: Corrosion studies using in-active salts
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Facilities for thermo-physical evaluation of salts
14
100 200 300 400 500 600 700-120
-90
-60
-30
0
30
Heat
Flow
(/m
W)
Sample temperature (/°C)
550C
Fuel Salt
-10
-5
0
5
10
Mas
s Ch
ange
(/m
g)
432C
TG
DTA
Drop calorimeter DTA plot of eutectic composition
High temperature viscometer
Differential Scanning Calorimeter
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Physics studies carried out at BARC
Validation of Tools : • MSFR (French) • MSRE (ORNL, US)
Indian concepts:
IMSBR-Loop Type IMSBR-Pool Type
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Molten Salt Fast Reactor : French Design
• Core Dimension (m): 2.2 X 2.2
• Power : 3000 MWth/ 1300 MWe
• Power density : 330 Wth/ cm3
• Breeding Ratio : 1.12
• Initial fissile inventory : 3.26 T/GWe
• Mean Fuel salt temperature : 750 oC
Ref.: D. Heuer et.al. ANE64 (2014) 421–429’
Fuel Salt : LiF (77.5%)-ThF4 (20%)-233UF4 (2.5%) Blanket Salt : LiF (77.5%)-ThF4 (22.5%)
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
MSFR found CRITICAL for given composition Breeding Ratio : 1.17 Initial fissile inventory : 3.23 T/GWe
MSFR core simulation : Results
Spectrum from published in paper ‘D. Heuer et.al. ANE64 (2014) 421–429’
Fig.: Spectrum predicted in our simulation
10-1 100 101 102 103 104 105 106 1071E-10
1E-9
1E-8
1E-7
1E-6
1E-5
Flux
(E*dφ/
dE)
Energy (eV)
Flux
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
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
Fig.: Radial and axial distribution of two group fluxes with (ARCH + DRAGON)
Analysis of MSRE with indigenous code ARCH
Ref.: Haunbenreich, P.N. et al., MSRE Design and Operations Report-III, ORNL,1964
No beryllium To avoid chemical toxicity Simplified design of experimental facilities (no need to protect against beryllium in
experimental facilities) • Avoid BeF2 (i.e. Avoid LiF-BeF2), 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
Salient design guidelines for arriving at conceptual design of IMSBR
19
Indian Molten Salt Breeder Reactor
IMSBR – Loop type: Major design parameters
20
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-ThF4-UF4
6 Blanket salt LiF-ThF4
7 Secondary salt LiF-NaF-ZrF4
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 233U)
Design of a low power demonstration IMSBR is in progress with temperature limited to 700 °C, and Ni-Mo-Cr-Ti based alloy
Blanket salts LiF-ThF4 (22.4 mole % of ThF4)[22.6-77.4% by wt] (Current reference)
• Liquidus: 568 °C LiF-NaF-ThF4 (30-57-13) [10.8 – 33.4 – 55.8 by wt]
• Liquidus: 505 °C LiF-CaF2-ThF4 (70-8-22)[19.7-6.8-73.5 by wt]
• Liquidus: 510 °C NaF-CaF2-ThF4 (Eutectic composition not known)
Fuel salt Same as blanket salt, but with UF4 dissolved as required
Coolant salts LiF-NaF-ZrF4 (26-37-37) [8.0-18.5-73.5 by wt](Current reference)
• Liquidus: 436 °C LiF-NaF-ZrF4 (42-29-29) [15.2 – 17.0 - 67.8 by wt]
Selected salts
21
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
IMSBR: design parameters and results Size: 2 X 2.05 (m) Power: 850 MWe Fuel Salt: LiF(77.6%)- ThF4(19.7%)- 233UF4(2.7%) Blanket Salt: LiF (77.6%)-ThF4 (22.4%) Core Average Temperature: 750 oC K-eff: 1.0447 ± 0.0004 Enrichment: ~ 12% ( 233U in heavy metal) Initial Fissile Inventory in core: 1.9 T Initial Conversion Ratio: 1.078
Physics design simulation of IMSBR - Loop Type
2 m
2.05 m
Longitudinal view Schematic cross-sectional view of core
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Height Core radius Partition (Ni-W-Cr)
Blanket thickness
Shell thickness (Ni-W-Cr)
200 cm 100 cm 2.5 cm 50 cm 2.5 cm
Case - 1 Composition (LiF-ThF4-UF4) 77.6% -19.7% -2.7%
He % (0.5%)
K-eff 1.0447 ICR_fuel 0.989 ICR_blanket 0.089 ICR 1.078 Inventory – 233U 1.9 T Enrichment (233U in HM) ~ 12 %
Case - 2 Composition (LiF-ThF4-UF4) 77.6% -19.9% -2.5%
He % (0.5%)
K-eff 1.0006 ICR_fuel 0.979 ICR_blanket 0.086 ICR 1.065 Inventory – 233U 1.76 T Enrichment (233U in HM) ~ 11.1%
IMSBR - Loop Type (continue..)
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
10-2 10-1 100 101 102 103 104 105 10610-10
10-9
10-8
10-7
10-6
10-5
10-2 10-1 100 101 102 103 104 105 10610-10
10-9
10-8
10-7
10-6
10-5
Spectrum in core region Spectrum in blanket region
Norm
alis
ed F
lux
Neutron Energy (eV)
Fig.: Comparison of spectrum in core region and blanket region of revised IMSBR core
IMSBR - Loop Type (continue..)
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Fig.: Schematic of pool type IMSBR with natural circulation ( Ref.: A. Borgohain et al. ThEC15, 2015)
Indian Molten Salt Breeder Reactor: IMSBR - Pool Type
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Analyses and Results: In presence of B4C lining Composition: LiF-ThF4-UF4
He (%mol)
Keff ICR ( fuel )
ICR (blanket )
ICR (total )
77.6% -19.9% -2.5% 0 0.9839 0.966 0.136 1.102 77.6% -19.7% -2.7% 0 1.0266 0.976 0.140 1.116 77.6% -19.9% -2.5% 0.5 0.9867 0.957 0.131 1.088
IMSBR : Pool Type (continue..)
B4C lining replaced by blanket salt Composition: LiF-ThF4-UF4
He (% mol)
Keff ICR (fuel)
ICR (blanket)
ICR (total)
77.6% -19.9% -2.5% 0 0.9852 0.962 0.150 1.112 77.6% -19.7% -2.7% 0 1.0288 0.972 0.153 1.125 77.6% -19.9% -2.5% 0.5 0.9872 0.953 0.142 1.095
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
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-232
77.6% -19.9% -2.5% 4.1 T 5.03 T 32.66 T 77.6% -19.7% -2.7% 4.44 T 5.44 T 32.28 T
IMSBR : Pool Type (continue..)
10-2 10-1 100 101 102 103 104 105 10610-11
10-10
10-9
10-8
10-7
10-6
10-5
10-2 10-1 100 101 102 103 104 105 10610-11
10-10
10-9
10-8
10-7
10-6
10-5
Blanket
Norm
alize
d flu
x
Neutron Energy (eV)
Core
IMSBR pool type : Natural Circulation case
Fig.: Comparison of normalized flux in core and blanket
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Refuelling Studies – A simple Model
0 5 10 15 201.000
1.002
1.004
1.006
0 5 10 15 201.000
1.002
1.004
1.006
0 10 201.000
1.002
1.004
1.006
0 5 10 15 201.000
1.002
1.004
1.006
0 10 201.000
1.002
1.004
1.006
0 10 201.000
1.002
1.004
1.006
0 5 10 15 201.000
1.002
1.004
1.006
0 10 201.000
1.002
1.004
1.006
0 5 10 15 201.000
1.002
1.004
1.006
0 5 10 15 201.000
1.001
1.002
1.003
1.004
1.005
1.006
1.007
Fresh Fuel
K-ef
f
Burnup (FPD)
5000 litres_5FPD 40 litres_7FPD 80_litres_7FPD200_litres_7FPD 500_litres_7FPD 1000_litres_7FPD 5000_litres_7FPD 10000_litres_7FPD 12600_litres_7FPD
Refuelling at 5 & 7 FPD
0 200 400 600 8001.00
1.05
1.10
1.15
1.20
K-in
f
Burnup (FPD)
Pa-233 removal at 800 FPDs
Fig.: Two step refueling & effect of removal of different amount of fuel
Fig.: Effect of Pa-233 removal at 800 FPD and 820 FPD
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:
Where, , and are number density of mixed, fresh and burned fuel of ith nuclide. , and are volume of fresh, removed and total fuel.
The reactivity effects of addition/removal of individual nuclide e.g. Pa-233 can be estimated by suitably adjusting the number density of nuclide.
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Time Independent Precursor Distribution in circulating fuel reactor
1𝑣𝑣𝑔𝑔
𝜕𝜕𝜑𝜑𝑔𝑔𝜕𝜕𝑡𝑡
= 𝛻𝛻 ∙ 𝐷𝐷𝑔𝑔𝛻𝛻𝜑𝜑𝑔𝑔 + 𝜒𝜒𝑝𝑝,𝑔𝑔 ∙ 1 − 𝛽𝛽 ∙ � 𝜈𝜈Σ𝑓𝑓 𝑔𝑔
𝐺𝐺
𝑔𝑔=1
∙ 𝜑𝜑𝑔𝑔 + �𝜒𝜒𝑑𝑑,𝑔𝑔,𝑖𝑖 ∙ 𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖
𝐼𝐼
𝑖𝑖=1
+ � 𝜑𝜑𝑔𝑔′ ∙ Σ𝑔𝑔′→𝑔𝑔𝑖𝑖
𝑔𝑔−1
𝑔𝑔′=1
− 𝜑𝜑𝑔𝑔 ∙ Σ𝑟𝑟,𝑔𝑔
𝜕𝜕𝐶𝐶𝑖𝑖𝜕𝜕𝑡𝑡
= − 𝛁𝛁 𝑼𝑼 ∙ 𝑪𝑪𝒊𝒊 − 𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖 + 𝛽𝛽𝑖𝑖 ∙ � 𝜈𝜈Σ𝑓𝑓 𝑔𝑔
𝐺𝐺
𝑔𝑔=1
∙ 𝜑𝜑𝑔𝑔
• The time independent equation with fuel velocity U in one-group diffusion theory with one group delayed neutron is shown as following
𝛻𝛻 ∙ 𝐷𝐷𝑔𝑔𝛻𝛻𝜑𝜑0(𝑧𝑧� + �𝜈𝜈Σ𝑓𝑓 1 − 𝛽𝛽 − Σ𝑎𝑎 𝑧𝑧 ]𝜑𝜑0 𝑧𝑧 + 𝜆𝜆 𝐶𝐶0 = 0
𝑈𝑈 𝑑𝑑𝐶𝐶0𝑑𝑑𝑧𝑧
= −𝜆𝜆𝐶𝐶0 𝑧𝑧 + 𝛽𝛽𝜈𝜈Σ𝑓𝑓𝜑𝜑0 𝑧𝑧
𝐶𝐶0 𝑧𝑧 = 𝑒𝑒 −𝜆𝜆𝜆𝜆𝑈𝑈
𝛽𝛽𝜈𝜈Σ𝑓𝑓𝑈𝑈
1𝑒𝑒𝜆𝜆𝜆𝜆 − 1
� 𝑒𝑒 𝜆𝜆𝜆𝜆′𝑈𝑈
𝐻𝐻
0𝜑𝜑0 𝑧𝑧′ 𝑑𝑑𝑧𝑧′ + � 𝑒𝑒
𝜆𝜆𝜆𝜆′𝑈𝑈
𝜆𝜆
0𝜑𝜑0 𝑧𝑧′ 𝑑𝑑𝑧𝑧′
• Time dependent multi group neutron diffusion equation can be written for circulation fuel reactor as
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Normalized flux distribution along z-axis
Precursor distribution along z-axis for several fuel velocities
Delayed Neutron Precursor Distribution in circulating fuel reactor
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Zero and One Dimensional Model
𝑑𝑑𝑑𝑑(𝑡𝑡)𝑑𝑑𝑡𝑡
=𝜌𝜌 𝑡𝑡 − 𝛽𝛽
Λ 𝑑𝑑 𝑡𝑡 + �𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖 𝑡𝑡
𝐼𝐼
𝑖𝑖=1
𝑑𝑑𝐶𝐶𝑖𝑖(𝑡𝑡)𝑑𝑑𝑡𝑡
= 𝛽𝛽𝑖𝑖Λ
𝑑𝑑 𝑡𝑡 − 𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖 𝑡𝑡 − 𝐶𝐶𝑖𝑖 𝑡𝑡𝜏𝜏𝑐𝑐
+ 𝐶𝐶𝑖𝑖 𝑡𝑡 − 𝜏𝜏𝑒𝑒
𝜏𝜏𝑐𝑐exp −𝜆𝜆𝑖𝑖𝜏𝜏𝑒𝑒
𝜕𝜕𝐶𝐶𝑖𝑖(𝑧𝑧, 𝑡𝑡)𝜕𝜕𝑡𝑡
+ 𝑈𝑈𝜕𝜕𝐶𝐶𝑖𝑖(𝑧𝑧, 𝑡𝑡)𝜕𝜕𝑧𝑧
= − 𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖 + 𝛽𝛽𝑖𝑖Λ
𝑑𝑑(𝑧𝑧, 𝑡𝑡)
𝑑𝑑𝑑𝑑(𝑡𝑡)𝑑𝑑𝑡𝑡
=𝜌𝜌 𝑡𝑡 − 𝛽𝛽
Λ 𝑑𝑑 𝑡𝑡 + �𝜆𝜆𝑖𝑖 ∙ 𝐶𝐶𝑖𝑖 𝑡𝑡
𝐼𝐼
𝑖𝑖=1
0-D Model
1-D Model
Boundary Condition𝜑𝜑0 0 = 𝜑𝜑0 𝐻𝐻 = 0and 𝐶𝐶0 0 = 𝐶𝐶0 𝐻𝐻 𝑒𝑒−𝜆𝜆𝜆𝜆𝐿𝐿.
• Point kinetics equation in case of circulation fuel reactor can be written with modified precursor equation as follow:
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
6 Group DNPs distribution with axial height for stationary fuel for U-235
6 Group DNPs distribution with axial height (z) for recirculating fuel for U-235 at 0.24 cm/s
Axial DNP distribution at t = 0
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Effective delayed neutron calculation in circulating fuel reactor
Beta effective calculation using analytical model in Molten Salt Fast Reactor (French) • In CFR, βeff differs from physical delayed neutron fraction β0 for two reasons 1st – difference in spectrum of delayed and prompt neutron 2nd– 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), 78-90.
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
• 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
Summary
34
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244
Acknowledgment: I acknowledge the support of Shri Abhishek Basak, RED, BARC for providing information regarding Engineering portion of the presentation.
D.K. Dwivedi, IAEA Technical Meeting, 622-I3-TM-52244