physics design studies for indian msrs - … design studies for indian msrs d.k. dwivedi, a.k....

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


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


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)


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


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


• 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


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


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


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


Physics studies carried out at BARC

Validation of Tools : • MSFR (French) • MSRE (ORNL, US)

Indian concepts:

IMSBR-Loop Type IMSBR-Pool Type


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%)


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


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


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


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..)


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..)


Fig.: Schematic of pool type IMSBR with natural circulation ( Ref.: A. Borgohain et al. ThEC15, 2015)

Indian Molten Salt Breeder Reactor: IMSBR - Pool Type


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


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


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.


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


Normalized flux distribution along z-axis

Precursor distribution along z-axis for several fuel velocities

Delayed Neutron Precursor Distribution in circulating fuel reactor


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:


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


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.


• 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


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

physics design studies for indian msrs - … design studies for indian msrs d.k. dwivedi, a.k....

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