dr. m. brovchenko & prof. e. merle- lucotte lpsc/in2p3/cnrs - grenoble inp, france

Dr. M. Brovchenko & Prof. E. Merle-LucotteLPSC/IN2P3/CNRS - Grenoble INP, France

INTRODUCTION TO THE PHYSICSOF THE MSFR

EVOL Winter School 04/11/2013 in Orsay, France

Outline

1.Breeding of the GEN IV systems

2.Historical overview of MSRs

3.MSFR concept

4.Validation of neutronic characteristics

5.MSFR kinetic and load following

1.Breeding of the GEN IV systems

1. Breeding of the GEN IV systems

Introduction to the Physics of the MSFR 1/55

Nuclear Waste

Light Water Reactor (LWR)

Transuranics

Enriched Uranium

Fission Products

Fissile NucleiNeutron

CaptureNeutron

Actinides:

Current Reactors



Sustainability:- Minimize the nuclear waste- Extending the nuclear fuel supply by breeding

Nuclear Waste

LWRTransuranics

= FuelEnriched Uranium

Fission Products

Fertile material: Thorium or U238

GEN IV

Fission Products

Breeds:Fissile

U233 or Pu = Fuel

Closed Cycle

Since 2002 : Molten Salt Reactor = one of the 6 reactor concepts selected by the Generation IV International Forum. Challenging technology goals are: Safe and Reliable Systems Sustainable Nuclear Energy Competitive Nuclear Energy Proliferation Resistance and Physical Protection



FBR-Na

MSFR

LWR

thermal fast

U233 fission

Am241 fission

Am241 capture

Am241 not fissile

Neutron population

Probability to do the nuclear reaction

Example of Am241

Am241 fissile

Fast neutron spectrumallows to burn the

transuranics


How to make Transuranics fissionable ?


2.Historical overview of MSRs

1964 – 1969: Molten Salt Reactor Experiment (MSRE) Experimental Reactor Power: 7.4 MWth Temperature: 650°C U enriched 30% (1966 - 1968) 233U (1968 – 1969) - 239Pu (1969) No Thorium inside

· 1970 - 1976: Molten Salt Breeder Reactor (MSBR) Never built

Power: 2500 MWth Thermal neutron spectrum

· 1954 : Aircraft Reactor Experiment (ARE) Operated during 1000 hours

Power = 2.5 MWth

2. Historical overview of MSRs

Historical studies of MSR : Oak Ridge Nat. Lab. - USA


Which constraints for a liquid fuel?• Melting temperature not too high• High boiling temperature• Low vapor pressure• Good thermal and hydraulic properties (fuel = coolant)• Stability under irradiation• Good solubility of fissile and fertile matters• No production of radio-isotopes hardly manageable• Solutions to reprocess/control the fuel salt

Lithium fluorides fulfill all constraints

Molten Salt Reactors

Advantages of a Liquid Fuel Homogeneity of the fuel (no loading plan)

Heat is produced directly in the heat transfer fluid

Possibility to reconfigure passively the geometry of the fuel: - One configuration optimize the electricity production managing the criticality

- An other configuration allow a long term storage with a passive cooling system

Possibility to reprocess the fuel without stopping the reactor: - Better management of the fission products that damage the neutronic and physicochemical characteristics

- No reactivity reserve


Liquid fuelled-reactors: why “molten salt reactors” ?


Which constraints for a liquid fuel?• Melting temperature not too high• High boiling temperature• Low vapor pressure• Good thermal and hydraulic properties (fuel = coolant)• Stability under irradiation• Good solubility of fissile and fertile matters• No production of radio-isotopes hardly manageable• Solutions to reprocess/control the fuel salt

Lithium fluorides fulfill all constraints

Neutronic cross-sections of fluorine versus neutron

economy in the fuel cycleThorium /233U Fuel Cycle

Molten Salt Reactors

F[n,n’]Na[n,γ]


Liquid fuelled-reactors: why “molten salt reactors” ?


Molten Salt Reactor (MSR)Renewal of the concept

• Thorims-NES5 then FUJI-AMSB in Japan since the 80’sReactor of very low specific power fed with 233U produced in sub-critical reactors

• Resumption of the MSBR’s studies by CEA and EDF since the 90’s

• TIER-1 project of C. Bowman in the 90’sPu burner (LWR’s spent fuel assemblies dissolved in liquid fuel) in sub-critical reactors

• TASSE (CEA) project in the 90’sPlutonium burner (liquid fuel) in sub-critical reactors

• AMSTER (EDF) project in the 90’sPlutonium burner then breeder reactor in Thorium cycle

• REBUS (EDF), MOSART (Kurchakov Institute), SPHINX (Czech Republic)Projects of actinide burners

• MOST Network 2001-2004European network having assessed the studies, the experiments and the state of knowledge

concerning molten salt reactors

2. Historical overview of MSRsMolten Salt Reactor (MSR)

Renewal of the concept


• Participation to the project TIER I of C. Bowman (1998)

• Re-evaluation of the MSBR from 1999 to 2002Use of a probabilistic neutronic code (MCNP) Development of an in-house evolution code for materials (REM)Coupling of the neutronic code with the evolution code

• From the Thorium Molten Salt Reactor to the Molten Salt Fast ReactorBreeder in the Thorium fuel cycle and Actinide Burner ReactorDevelop to solve the problems of the MSBR project

– Bad (null to positive) feedback coefficients– Positive void coefficient– Unrealistic reprocessing– Problems specific to the graphite moderator

- Lifespan- Reprocessing and storage- Fire risk


Thermal vs Fast Spectrum Renewal of the concept at CNRS


Coupling of the in-house code REM for materials

evolution with the probabilistic code MCNP for

neutronic calculations


Tools for the Simulation of Reactor Evolution


'( ) ( ) ( ) ( ) ( ) ( )j i

ij j j i j j i i i i

j

Nt N t b N t b t N t N t

t

production from nucleus j disappearancesum over all nuclei j

production by nuclear reaction production by

radioactive decay

disappearance by radioactive decay

disappearance by nuclear reaction

Molten Salt Reactors: addition of a feeding term, equal to the number of nuclei added per time unit for each element (flow)

Reprocessing: new terms. .

.

1extr extri i i reprocess

i

NT

with

Efficiency linked to the nucleus extraction probability


Tools for the Simulation of Reactor EvolutionIntegration Module: Bateman Equation for nucleus i


TMSR general parameters:- total power: 2500 MWth (1000 MWe)- salt composition:78% LiF – 22% (HN)F4 (21.4% ThF4 – 0.6% UF4)- mean temperature: 630 °C (900 K)

TMSR geometrical parameters:- core radius: 1.6 m- core shape: cylindrical (H=D)- salt volume: 20 m3

- fertile blanket: Thorium- hexagon size (moderator): 15 cm- channel radius (fuel salt): varying

Molten Salt Reactor operated in the Thorium Fuel Cycle (TMSR)


Historical MSR Studies at CNRS Systematic studies (L. Mathieu)


- core volume adjusted to keep the same salt volume

r = 4 cm

r = 8.5 cm

single channel

3 different moderation ratios:

thermal fast


Variation of the channel radius (moderation ratio)


1graphiteL

no graphite

moderator+

Influence studied through 4 core characteristics:

- Total feedback coefficient- Breeding ratio- Neutron flux / Graphite lifespan- Fissile inventory


Influence of the channel radius on the core behavior


Three types of configuration:

- thermal (r = 3-6 cm)- epithermal (r = 6-10 cm)

- fast (r > 10 cm)




Thermal spectrum configurations

- low 233U initial inventory- quite long graphite life-span

- iso-breeder- positive feedback coefficient

Epithermal spectrum configurations

- quite low 233U initial inventory- very short graphite life-span

- quite negative feedback coefficient- iso-breeder

Fast spectrum configurations (no moderator)

- very negative feedback coefficients

- very good breeding ratio

- no problem of graphite life-span

- large 233U initial inventory




3.MSFR Concept

Thermal power 3000 MWth

Input/output operating temp. 650-750 °C

Melting Temp. 565 °C

Fuel Salt Volume 18 m3

Fuel circulation time 3.9 s

233U initial inventory 3400 kg

233U production 95 kg/year

Fuel Molten salt LiF-ThF4-233UF4 with 77.5 % LiF initial composition (mol%) or LiF-ThF4-(Transuranics)F3

3. MSFR Concept

Molten Salt Fast Reactor

Why this configuration?Introduction to the Physics of the MSFR 17/55

Physical Separation (in the core) Gas Reprocessing Unit through bubbling extraction Extract Kr, Xe, He and particles in suspension

Chemical Separation (by batch) Pyrochemical Reprocessing Unit Located on-site, but outside the reactor vessel

Fission Products Extraction: Motivations Control physicochemical properties of the salt (control deposit, erosion and corrosion phenomena's) Keep good neutronic properties

Gas injection

Reprocessing by batch of 10-40 l per day

Gas extraction

3. MSFR Concept

Fuel reprocessing


1/ Salt Control + Fluorination to extract U, Np, Pu + few FPs - Expected efficiency of 99% for U/Np and 90% for Pu – Extracted elements re-injected in core

2/ Reductive extraction to remove actinides (except Th) from the salt – MA re-injected by anodic oxidation in the salt at the core entrance

3/ Second reductive extraction to remove all the elements other than the solvent - lanthanides transferred to a chloride salt before being precipitated

1/ 2/ 3/

3. MSFR Concept

On-site Chemical reprocessing unit


To remove all insoluble fission products (mostly noble metals) and rare gases, helium bubbles are voluntary injected in the flowing liquid salt (bottom of the core) → Separation salt / bubbles → Treatment on liquid metal and then cryogenic separation (out of core)

3. MSFR Concept

Online noble gazes bubbling in core


Element Absorption (per fission neutron)

Heavy Nuclei 0.9

Alkalines < 10-4

Metals 0.0014

Lanthanides 0.006

Total FPs 0.0075

Fast neutron spectrum Þ very low capture cross-sections

low impact of the FP extraction on neutronics

Þ Parallel studies of chemical and neutronic issues possible

Batch reprocessing: On-line (bubbling) reprocessing:

3. MSFR Concept

Influence of fuel reprocessing on the neutronics


Optimization studies: Initial fissile matter (233U, Pu), salt composition, fissile

inventory, reprocessing, waste management, deployment capacities, heat exchanges, structural materials, design..

Generation IV reactors: fuel reprocessing mandatoryNeutronic core of the MSFR associated to reprocessing units (on-site)

What is a MSFR ?

Molten Salt Reactor (molten salt = liquid fuel also used as coolant)

Based on the Thorium fuel cycle

With no solid (i.e. moderator) matter in the core Fast neutron

spectrum

3. MSFR Concept

MSFR: result of neutronic optimization studies


Reactor Design and Fissile Inventory Optimization = Specific Power Optimization2 parameters:

3 limiting factors:

• The produced power• The fuel salt volume and the core geometry

• The capacities of the heat exchangers in terms of heat extraction and the associated pressure drops (pumps) large fuel salt volume and small specific power• The neutronic irradiation damages to the structural materials which modify their physicochemical properties. Three effects: displacements per atom, production of Helium gas, transmutation of Tungsten in Osmium large fuel salt volume and small specific power• The neutronic characteristics of the reactor in terms of burning efficiencies small fuel salt volume and large specific power and of deployment capacities, i.e. breeding ratio (= 233U production) versus fissile inventory optimum near 15m3 and 400W/cm3

Liquid fuel and no solid matter inside the core possibility to reach specific power much higher than in a solid fuel

3. MSFR Concept

MSFR: Design and Fissile Inventory Optimization


Core:No inside structure,perhaps thermo-hydraulical injection element (plate or deflector) to be added

Outside structure: Upper and lower Reflectors, Fertile Blanket Wall

+ 16 external modules:

• Pipes (cold and hot region) • Bubble Separator• Pump • Heat Exchanger• Bubble Injection

Fuel Salt Loop = Includes all the systems in contact with the fuel salt during normal operation

3. MSFR Concept


Structural materials


Fuel salt volume / specific power t(100 dpa) t(100 ppm He) t(-1 at% of W)

12 m3 - 500 W/cm3 85 years 2.2 years 4.7 years

18 m3 – 330 W/cm3 133 years 3.2 years 7.3 years


3. MSFR ConceptMSFR: Design and Fissile Inventory Optimization

Structural material composition Ni based alloy:

Ni W Cr Mo Fe Ti C Mn Si Al B P S

79.432 9.976 8.014 0.736 0.632 0.295 0.294 0.257 0.252 0.052 0.033 0.023 0.004


Most irradiated area (central part of axial reflector – radius 20 cm/thickness 2 cm)



Fast neutrons (En > 100 keV) eject nucleus from structure material creating damages through atom

displacement (evaluated with cross section MT444)

DPADisplacement per atom


79.432 9.976 8.014 0.736 0.632 0.295 0.294 0.257 0.252 0.052 0.033 0.023 0.004



mainly on 59Ni, (58Ni) and 10B(Boron quantity may be re-ajusted)


He production


79.432 9.976 8.014 0.736 0.632 0.295 0.294 0.257 0.252 0.052 0.033 0.023 0.004




Transmutation Cycle of W in Re and Os (neutronic captures + decays)

Tungstene transmutation


79.432 9.976 8.014 0.736 0.632 0.295 0.294 0.257 0.252 0.052 0.033 0.023 0.004


Optimization = Medium Fuel Salt Volumes

Fuel salt volume / specific power t(100 dpa) t(100 ppm He) t(-1 at% of W)


18 m3 – 330 W/cm3 133 years 3.2 years 7.3 years




Reactor Design and Fissile Inventory Optimization = Specific Power Optimization2 parameters:

3 limiting factors:

• The produced power• The fuel salt volume and the core geometry

Liquid fuel and no solid matter inside the core possibility to reach specific power much higher than in a solid fuel

Reference MSFR configuration with 18m3 et 330 W/cm3 corresponding to an initial fissile inventory of 3.5 tons per GWe

• The capacities of the heat exchangers in terms of heat extraction and the associated pressure drops (pumps) large fuel salt volume and small specific power• The neutronic irradiation damages to the structural materials which modify their physicochemical properties. Three effects: displacements per atom, production of Helium gas, transmutation of Tungsten in Osmium large fuel salt volume and small specific power• The neutronic characteristics of the reactor in terms of burning efficiencies small fuel salt volume and large specific power and of deployment capacities, i.e. breeding ratio (= 233U production) versus fissile inventory optimum near 15m3 and 400W/cm3

3. MSFR Concept



Starting mode 233U [kg] TRU [kg] enrU + TRU [kg]

Th 232 25 553 20 396 10 135Pa 231 U 232 U 233 3 260 U 234 U 235 1 735U 236 U 238 11 758

Np 237 531 335Pu 238 229 144Pu 239 3 902 2 464Pu 240 1 835 1 159Pu 241 917 579Pu 242 577 364Am 241 291 184Am 243 164 104Cm 244 69 44Cm 245 6 4

Initial heavy nuclei inventories per Gwe:

3. MSFR ConceptMSFR starting modes


4.Validation of MSFR neutronic characteristics

EVOL objective: to propose a design of MSFR by 2012 given the best system configuration issued from physical, chemical and material studies

• Recommendations for the design of the core and fuel heat exchangers • Definition of a safety approach dedicated to liquid-fuel reactors - Transposition of the

defence in depth principle - Development of dedicated tools for transient simulations of molten salt reactors

• Determination of the salt composition - Determination of Pu solubility in LiF-ThF4 • Control of salt potential by introducing Th metal - Evaluation of the reprocessing

efficiency (based on experimental data) – FFFER project • Recommendations for the composition of structural materials around the core

WP2: Pre-Conceptual Design and Safety WP3: Fuel Chemistry and Reprocessing WP4: Structural Materials

+ Coupling to the ROSATOM project MARS (Minor Actinides Recycling in Molten Salt)

European ‘EVOL’ (Evaluation and Viability Of Liquid fuel fast reactor systems) Project (7th PCRD) - EURATOM/ROSATOM cooperation

Neutronic Benchmark

European participant :CNRS, HZDR, KIT, POLIMI, POLITO, TU Delft + KI

4. Validation of neutronic characteristics

EVOL project


Simplified Geometry of MSFR for MCNP calculations

4. Validation of neutronic characteristicsNeutronic Benchmark definition

233U-started MSFR TRU-started MSFRTh 233U Th Actinide

38 281 kg

19.985 %mol

4 838 kg

2.515 %mol

30 619 kg

16.068 %mol

Pu 11 079 kg5.628 %mol

Np 789 kg0.405 %mol

Am 677 kg0.341 %mol

Cm 116 kg0.058 %mol

Initial composition


On-line bubbling extraction:Removal period T1/2 30 seconds Extracted elements Z 1, 2, 7, 8, 10, 18, 36, 41, 42, 43, 44, 45, 46, 47, 51, 52, 54 and 86.

Pyrochemical reprocessing:Rate 40 l/dayExtracted elements Z 30, 31, 32, 33, 34, 35, 37, 38, 39, 40, 48, 49, 50, 53, 55, 56, 57, 58,

59, 60, 61, 62, 63, 64, 65, 66, 67, 68, 69, 70.

4. Validation of neutronic characteristicsNeutronic Benchmark definition Thermal power (MWth) 3000

Electric power (MWe) 1500

Fuel Molten salt LiF-ThF4-233UF4 initial composition (mol%) with 77.5 % LiF

or LiF-ThF4-(Pu-MA)F3

Fertile Blanket Molten salt LiF-ThF4 initial composition (mol%) (77.5%-22.5%)

Melting point (°C) 565

Input/output operating temp. (°C) 625-775

Fuel Salt Volume (m3) 18 9 out of the

core 9 in the core

Blanket Salt Volume (m3) 7.3

Total fuel salt cycle in the system 3.9 s


CNRS/IN2P3/LPSC (LPSC)

4. Validation of neutronic characteristicsTools used for the Neutronic Benchmark

- Probabilistic code MCNP with a home-made materials evolution code REM;

- Extraction of nucleus i with a specific removal constants λchem; - Fissile and fertile composition adjusted to control the reactivity.

Helmholtz-Zentrum Dresden-Rossendorf (HZDR)

- HELIOS 1.10 code system with internal 47 energy group library;- Pre- and post-processor for reprocessing and re-fuelling adapted

to MSFR calculation

Kurchatov Institute (KI) - MCNP-4B code coupled with ORIGEN2.1 code, solves depletion equations;

- Extraction of FP and fissile and fertile adjustment simulated.


Politecnico di Torino (POLITO)

Stochastic calculations:- SERPENT calculation of the core without burn-up;- 3-group energy cross section used for DYNAMOSS. Deterministic calculations:- Diffusion model in cylindrical r-z geometry- For circulating fuel system

Technical University of Delft (TU-Delft)

- HEAT an in-house developed CFD-program; - DALTON an in-house developed diffusion code; - SCALE used to cross sections calculations.

4. Validation of neutronic characteristicsTools used for the Neutronic Benchmark

Politecnico di Milano (POLIMI)

ERANOS-based EQL3D procedure and extension for the MSFR:- ERANOS 2.2-N code system for core calculation (33-group energy);- FP extraction and re-fueling adjustment simulated.SERPENT-2 extension for on-line fuel reprocessing:- SERPENT is a three-dimensional Monte Carlo code;- developed extension of SERPENT-2 code takes into account online

fuel reprocessing and features a reactivity control algorithm.


Same data basis JEFF-3.1 for the calculations of different partners (probabilistic tool)

Differences due to different energy binning: POLITO much smaller

Small deviations due to different cross section reconstruction methods

4. Validation of neutronic characteristicsNeutronic energy spectrum

Good agreement between the POLITO, LPSC and POLIMI calculations



Cros

s Se

ction

[bar

n]

Neutron energy [MeV]

Neutron energy [MeV]

Neu

tron

Flu

x

Why this shape?



Multi-group spectra of 233U-started MSFR or steady state composition


Sensibility to different nuclear data bases (LPSC and POLIMI tools)



Different data bases used

One of the reasons of the observed differences



Sensibility to composition of the fuel salt (LPSC-MCNP tool with JEFF-3.1)



0.05 0.5 5 50-5

-4.5

-4

-3.5

-3

-2.5

-2

-1.5

-1

-0.5

0233U-started MSFR POLIMI Density

POLIMI Doppler

KI Density

KI Doppler

LPSC Density

Linear (LPSC Density)

LPSC Doppler

Logarithmic (LPSC Doppler)

POLITO Density

POLITO Doppler

Density TU Delft

Doppler TU DelftOperation time [years]

Feed

back

coe

ffici

ent [

pcm

/K]

4. Validation of neutronic characteristicsThermal feedback coefficient

Initial composition Different data bases

2 calculations performed:𝑑𝑘𝑑𝑇

=𝑑𝑘𝑑𝑇 𝐷𝑒𝑛𝑠𝑖𝑡 𝑦

+𝑑𝑘𝑑𝑇 𝐷𝑜𝑝𝑝𝑙𝑒𝑟

Thermal expansion Cross sections


4. Validation of neutronic characteristicsThermal feedback coefficient

0.05 0.5 5 50-4.0

-3.5

-3.0

-2.5

-2.0

-1.5

-1.0

-0.5

0.0TRU-started MSFR

POLIMI DensityPOLIMI DopplerKI DensityKI DopplerLPSC DensityLinear (LPSC Density)LPSC DopplerLogarithmic (LPSC Doppler)

Reactor operation time [years]

Ther

mal

feed

back

coe

ffici

ent [

pcm

/K]

Different data bases

Overall good agreement: both contributions are negativ for all compositions


4. Validation of neutronic characteristicsEvolution calculation

Different tools, same data basis : ENDF-B6



Same tool, different data bases (SERPENT - POLIMI)



BG =Extra Produced Fissile Material

Operation TimeBreeding gain :

Higher sensibility on the data basis choice than on the tool


4. Validation of neutronic characteristics

Neutron capture cross section of 233U

= main reason of the observed differences


Nuclear data bases

5.MSFR kinetics and load following

o Model hypotheses :• Homogeneous reactor• One energy group• Φ(r,t) = Φ(r) ˑ φ(t)• No precursor transport

𝜌 (𝑡 )= 𝑑𝑘𝑑𝑇 (𝑇 (𝑡 )−𝑇 0 )+ 𝐼 (𝑡)

𝜕𝑃𝜕𝑡

=𝜌 (𝑡 )− 𝛽𝑒𝑓𝑓

Λ𝑃 (𝑡 )+ 𝐴∑

𝑖

𝜆𝑖𝐶𝑖(𝑡)

𝜕𝑇𝜕𝑡

=𝑃 (𝑡 )−𝑃𝑟𝑒𝑓𝑟 (𝑡)

𝐶𝑝𝑑

𝜕𝐶𝑖

𝜕𝑡=𝛽𝑖 ,𝑒𝑓𝑓

Λ ∙𝐴𝑃 (𝑡 )− 𝜆𝑖𝐶𝑖(𝑡)Group i :

with = - 5 pcm/K , I(t) : 500 pcm in 0,001s

keff ~ kinf

Fast spectrum

βeff

5. MSFR kinetics and load following

Reactivity insertion with point kinetic model

Nor

mal

ized

fissi

on p

ower

Time after transient begin [s]

Tem

pera

ture

[C]

Reac

tivity

[pcm

]




o Linear reactivity insertion of 500 pcm (ρ > β ≈ 170 pcm)

o In 0,001 s and 0,01 s :

• Prompt criticality,

• Fission power x 1000

with = - 5 pcm/K



Nor

mal

ized

fissi

on p

ower

Reac

tivity

[pcm

]


Tem

pera

ture

[C]

Reac

tivity

[pcm

]




with = - 5 pcm/K o Linear reactivity insertion of 500 pcm (ρ > β ≈ 170 pcm)

o In 0,001 s and 0,01 s :

• Prompt criticality,

• Fission power x 1000

o In 1 s and 10 s :

• Fission power,

• No temperature peak

o Reactor stabilization at T > T0



No dynamic precursor following nor fuel salt temperature in and out flow:

neglected at short and long term. For t ~4s (circulation time) :

These effects have to be studied


Nor

mal

ized

fissi

on p

ower

Tem

pera

ture

[C]

Reac

tivity

[pcm

]





Point kinetic model with zones

Temperature distribution

Extension of the point kinetic model :- To follow the delayed neutron precursor- To model the heat exchanger outside the core

Two meshes are defined : fix and moving

Fuel salt temperature for each cell:

Reactivity:

Fission power: CoreCore

CoreDelayed neutron precursor abondance for each cell:


Ferti

le b

lank

et

Neu

tron

ic p

rote

ction

Heat exchanger

Pump R

Z

To go even further:

With code COUPLEo Tool based on neutronics and thermo

hydraulics

o Developed at KIT (Karlsruhe)

o 2D axial symmetry

o Neutronics: • Diffusion with 2 energy groups

o Thermo hydraulics : • Mass conservation• Navier –Stokes• Energy conservation

o 112/130 cells R/Z

o Pump model fixing the fluid velocity

o Heat exchanger model as negative heat source


Coupling of neutronic and thermalhydrulic

More about thermalhydraulics and the coupling this afternoon



Comparison of 3 kinetic tools

Point kineticPoint kinetic with zonesCOUPLE

Fast load following transients:

•Heat extraction drop to 50%

•Heat extraction increase to 150%

With a heat exchanger model (COUPLE + Point kinetic with zones): transient starts ~ 1 s later

Point kinetic with zones : recirculation clearly observed

COUPLE: not that clearly cause circulation form time dependent


Nor

mal

ized

fissi

on p

ower

Point kinetic Point kinetic with zones

Tem

pera

ture

[K]


Reac

tivity

[pcm

]




Slow load following transients

Point kinetic with zonesCOUPLE

Slow load following transients:•Heat extraction decrease to 50%, 25%, 4% (decay heat)

Very good load following capabilities

BUT precise themalhydraulics

simulation needed!


Nor

mal

ized

fissi

on p

ower Point kinetic with zones

Extracted heat Fission power:

Load following :

Time after transient begin [s]Te

mpe

ratu

re [K

]


- MSFR Concept

- Neutronic characteristics

- Basis in reactor kinetics

Summary

More about

- MSFR thermal-hydraulics

- Coupled neutronics and thermal hydraulics

- Fuel cycle aspects and transmutation

dr. m. brovchenko & prof. e. merle- lucotte lpsc/in2p3/cnrs - grenoble inp, france

Documents

reactor concepts

nuclear fuel supply

aircraft reactor experiment

hydraulic properties

fuel saltlithium fluorides

historical overview

msfr concept4

msfr kinetic