conceptual design activities of fds series fusion power · pdf file─fds-st : a spherical...

ASIPP

Conceptual Design Activities of FDS Series Fusion Power Plants

FDS

Presented by Yican WU

Contributed by FDS Team

Institute of Plasma Physics, Chinese Academy of SciencesP.O. Box 1126, Hefei, Anhui, 230031, China;

E-mail: [email protected]

______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

ASIPP

• FDS Team Fusion Design StudyFusion Driven (Subcritical) System

• ASIPP Academia Sinica, Institute of Plasma Physics

• Wide collaboration with other institutions in China

FDS

For further details: Link to Website: http://www.fds.org.cn


ASIPPFDS

1. Fusion Reactor Designs

2. Blanket Concepts

3. TBMs (Test Blanket Modules), Testing Strategy and R&D• DFLL (Dual Function Lithium Lead) –TBMs for ITER/EAST

To validate and demonstrate the technology of He-cooled and He/LiPb dual-cooled liquid lithium lead breeder blanket.

4. Design and Analysis Tools Development• VisualBUS (MCAM, SNAM.HENDL) : Integrated Neutonics Analysis• TOPCODE(SYSCODE+RiskA) : Integrated System Analysis

Contents

: a spherical tokamak-based reactor, to exploit innovative conceptual path─ FDS-ST: a high temperature fusion reactor, for hydrogen generation─ FDS-III: a fusion power reactor, for advanced electricity generation─ FDS-II

: a fusion-driven sub-critical system, for early application of fusion e.g waste transmutation and fuel breeding etc.

─ FDS-I

: High Temperature Liquid Blanket (outlet temp. 900~1000°C)─ HTL: Dual-coolant(He/LiPb) Lithium-Lead Breeder Blanket (outlet temp.~700°C)─ DLL: Single-coolant (He) Lithium Lead Breeder Blanket (outlet temp.~450°C)─ SLL: Dual-coolant (He/LiPb) Waste Transmutation Blanket (outlet temp.<450°C)─ DWT


FDS Fusion Reactor Designs

: a spherical tokamak-based reactor, to exploit innovative conceptual path─ FDS-ST: a high temperature fusion reactor, for hydrogen generation─ FDS-III: a fusion power reactor, for advanced electricity generation─ FDS-II

: a fusion-driven sub-critical system, for early application of fusion e.g waste transmutation and fuel breeding etc.• Status of nuclear energy in China and in the world• Advantages of a subcritical system

─ FDS-I


ASIPPFDS

• Fission nuclear industry has been falling on hard times─ Nuclear radioactive waste disposal ?─ Nuclear fuel supply shortage ?─ Nuclear proliferation ?

• Fusion development still needs hard work to economical utilization although it has a good progress

─ Great advances in plasma physics and technological experiments, R&D activities on fusion energy Fusion neutron source application─ New concepts on economical application of fusion Further efforts to economical fusion energy

• From the energy supply point of view, we have a big gap between fission energy and fusion energy ! From the technology development point of view, we need to pass am intermediate step to pure fusion energy

─ There hardly is a possibility to make a commercial use of pure fusion energy before 2050. ─ But we could have a near-term application of fusion as a neutron source (FDS-I).

FDS-I Necessity: Nuclear Energy Status in the World

What about a Hybrid ?______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

Yican WU,, Plasma Science & Technology, Vol.3, No.6 (2001).

ASIPPFDS

FDS-I Necessity: Energy Demand Status in China

Predicted Situation at 2050 in China:

> Sum all over the world3601830High LevelSum in US, France and RF2401220Mid. Level

Double in France120610Low Level

Total Nucl. Power Capacity(Approximate Scale)

Nucl. Power CapacityFraction BFraction AScenario

•Fraction A: Fraction of nucl. electricity in total electricity capacity•Fraction B: Fraction of nucl. electricity in total primary energy capacity

Predicted Fraction and Capacity of Nuclear Energy Supply in 2050 in China

: ~ 1200-1500 GWeInstalled Capacity: ~ 5 billion tons of tCEEnergy demand: ~ 6000-12000 billion US$Total GDP: ~ 1.5 billionPopulation

Power supply shortage ?(in the present, nuclear electricity ~1%)

Nuclear electricity should make an important contribution.

Questions: How to solve these problems:

─ Nuclear fuel supply ?

─ Radioactive waste disposal ?

─ Safety problem ?

What about a Hybrid ?


ASIPPFDS

FDS-I Advantages: Attractive Features of Subcritical systems

Improve energy balance Energy amplification in fission blanket to easy the requirements of the plasma core and the relevant materials.Qt= Qp (0.2+ 0.8Qb); )1( eff

eff

fusion

fissionb E

EQ

κκ

υ −=

Improve neutron balance Fusion Neutrons enabling excess neutrons available for

─ Breeding fissile fuel─ Transmuting LLMA and LLFP

ν : neutrons per fission (critical reactor)ν/keff : neutrons per fission (subcritical)

Improve SafetySub-criticality allowing no critical accident risk and larger design marginkeff << 1─Suitable fuel cycle not to breed pure fissile materials─ Reduce the number of fuel breeders

Benefit both fission and fusion Providing a test-bed for pure fusion reactors to encourage continued work and continued progress.Solving the problems of fission development key concern:─ Long-lived watses─ Fuel breeding

Example: Qp=1, ν =3, keff=0.9, Qb=43, Qt=34


Yican WU,, Plasma Science & Technology, Vol.3, No.6 (2001).

ASIPPFDS

FDS-I: the Fusion-Driven Subcritical System

FusionFusion--Driven Hybrid SystemDriven Hybrid System

─ a hybrid system for multi-applications

─ early application of Fusion Neutron Source

Fusion

Core

Fission

Blanket

Blanket functions:─ fuel breeding─ waste transmutation─ energy production─ other applications


Yican WU, Fusion Eng. Des. 63-64 (2002) 73-80.

FDS-I: Fusion-Driven Subcritical SystemBased on Easy Plasma Physics/Engineering Level

Core: fusion power=100~200MW, Neutron Wall Loading~0.5MW/m2

Blanket : He/LiPb Dual-cooled Waste Transmutation (DWT) blanket

FDSASIPP

─ Goal: Early application of fusion (waste transmutation/energy)

─ Functions:─ Inboard : tritium breeding─ Outboard : multifunctional (waste disposal, fuel breeding,energy generating, material testing etc.)

─ Feasible plasma technology:low plasma core parameters(long-pulse/steady-state)

─ Feasible material technology:low neutron wall loading(RAFM /316SS & He/LiPb) Reference 3-D model of FDS-I


Y.C. WU et.al, J. of Nuclear Materials, 307-311 (2002) 1629-1636.

FDSASIPP

Neutronics and Fuel Cycle Balance

Neutronics Principle of DWT Blankets

AC Zone(MA / Pu / U)

FPZone

(a)

(b)

(c)

FP zoneAC (MA / Pu / U) zoneplasma

(a) Transmute LLMA / Pu wasteHigh energy neutrons (n, fission)

(b) Breed fissile material (e.g. Pu239)Middle energy neutrons (n, γ)

(c) Transmute LLFPThermal neutrons (n, γ)

D + T

neutrons


Y.C. Wu et.al, Fusion Eng. Design, 63-64(2002)133-138.

FDSASIPP

Ref

eren

ce

3-D g

eom

etri

cal m

odel

Design of DWT-CPL/OPG/NPG BlanketA series of design scenarios, with emphasis on circulating particle or pebble bed fuel configurations considering geometry complexity of tokamak, frequency of fuel discharge and reload (including design of an emergency fuel discharge sub-system to improve the safety potential of the system), are being evaluated and optimized.

A design and its analysis on the He-gas and liquid LiPb DWT blanket with Carbide heavy nuclide Particle fuel in circulating Liquid LiPb coolant (DWT-CPL) has been studied for years.

Other concepts such as the DWT blanket with Oxide heavy nuclide pebble bed fuel in circulating helium-Gas (DWT-OPG) and with Nitride heavy nuclide Particle fuel in circulating He-Gas (DWT-NPG) are also being investigated.

DWT-CPL: The AC appears in the form of the TRISO(TRi-ISOtropic)-like carbide particles coated with SiC suspending in the LiPb slurry. The circulating fuel form has the advantages of good compatibility with complex geometry, easy control of fuel cycle and fast response to emergency fuel removal etc.


Y.C. WU et.al, ISFNT-7, Tokyo, Japan (2005)

FDSASIPP

Design Parameters of DWT-CPL Blanket

Reference Design Parameters of DWT-CPL with fertile-free fuelReference parametersConstraint and objectiveItems

AC (Np, Am, Cm, U, Pu): carbide(coated with C and SiC)

FP (Tc, I, Cs): Tc / NaI / CsClTechnology and engineering feasibilityFuel

24 (MA)24 (Pu)

7/9/13 (Cs / I / Tc)To maximize

Transmutedwaste (UPRW/y)

238 (MA)197 (Pu)200 (FP)

≤300 (the LLMA and LLFP compositions of existing and available spent fuels )

Initial MA&FP Inventory (UPWR*)

3.4≥1.1(tritium sustainability limit)TBR

100≤100 (cooling capability limit)Pdmax (MW·m-3)

0.93≤0.95 (safety margin limit)Keff

* UPWR represents the equivalent amount of identified x-isotopes from the spent fuel at the burnup of 33,000MWD/MTU from a standard 3000MWt PWR in a full power year.


FDS-II: Fusion Power ReactorBased on Advanced Plasma Physics/Engineer LevelCore: fusion power = 2~3 GW, Neutron Wall Loading ~ 3MW/m2

Blanket: high thermal power density and high thermal efficiency(1) He/LiPb Dual-cooled Lithium Lead (DLL) Blanket(2) He-cooled Quasi-Static Lithium Lead (SLL) Blanket

FDSASIPP

─ Goal: Highly efficient application of fusion energy

─ Advanced plasma technology:steady-state

─ Advanced material technology:(RAFM/ODS & He/LiPb)

─ Blanket scheme options:Single/dual-coolants

1234 5

6789

100

Reference 3-D model of FDS-II______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

WU Yican et. al.,, Chinese J. of Nucl. Sci. & Eng., Vol.25, No.1 (2005)

SLL: He-cooled Quasi-Static Lithium Lead BlanketSingle Coolant: He-gas (R-T + P-directions) T-Breeder: Quasi-Static LiPb: (slowly flowing in P-direction,

outlet temp.~450°C)Coating: to protect the steel structure and to reduce T-

permeation and MHD effects.DLL: He/LiPb Dual-cooled Lithium Lead Blanket• Coolant 1: He-gas (R-T + P-directions) • Coolant 2 & T-Breeder: LiPb (flowing in P-direction,

(outlet temp.~700°C)• Thermal and electric insulators: to avoid RAFM working

at high temp. 700C

FDSASIPP

Design of DLL/SLL Blanket for FDS-II

To avoid the critical issues such as MHD effects and FCI, relevant to DLL, SLL blanket is designated to use quasi-static LiPb flow instead of quick moving LiPb in DLL

─ “Multi-large-modules” blanket─ Liquid breeder blanket system as a primary option─ The RAFM steel steel as structural material.─ He gas cooling structure (FW and SPs )


Design of SLL/DLL Blanket Module

FDSASIPP

Reference design parameters of the DLL/SLL blankets

1.2TBR

~199/78/12Ave velocity of LiPb in breeder zone /mm·s-1

/450480/700In/Out temp. /℃Breeder material LiPb

88Pressure /MPa

~120/80115/40FW/SP velocity /m·s-1

300/450300/450In/Out temp. /℃

Coolant He

~5.7/155.7/15.3Deposition nuclear heat (FW/breeder zone) /MW

3.543.54Neutron wall load /MW·m-2

0.70.7Heat flux /MW·m-2

Heat source

Coating:e.g. Al2O3

FCI：e.g .SiCf/SiC,Coating: e.g.Al2O3

Functional material(s)

CLAM steelCLAM steelStructural material(s)

SLL-DEMODLL-DEMOBlanket


FDSASIPP

FDS-III: a High Temp. Fusion Reactor for H-Generation(as the First Generation Fusion Power Plant ?)

A promising method for H- production is that nucl. power would be used as a provider of• electricity in the electrolysis process or • high-temperature heat in the thermochemical cycles technology, which needs the high temp. range above 900 oC to achieve high efficiency of H-production.

To achieve a high temperature above 900 oC, one of the most challenging issues is the structural material under irradiation. As a result, the development of high temp. fusion reactor is limited by the current status of material technology

An optimized blanket design with innovative idea is considered to obtain high temp. heat based on the relatively mature and most promising RAFM steel (allowed temp. up to 550oC) as structural material and SiCf/SiC composite material or other high temp. materials (allowed temp. up to 900~1000℃) as flow channel electrical and thermal insulator in a dual-cooled liquid lithium-lead blanket

DTL: High Temperature Liquid Lithium Lead______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

FDSASIPP

FDS-III: HTL Blanket for H-Generation(High Temperature Liquid Lithium Lead):

─ Goal:Production of hydrogen

─ Material technology:• RAFM steel as structural material• SiCf/SiC composite or

other refractory materials as FCIs─ Blanket scheme feature:• Multilayer FCIs in LiPb channel:

1) increasing LiPb temperature above 900 oC 2) reducing interface temperature RAFM steel

/LiPb below 500oC3) temperature gradient across FCIs

─ Hydrogen production technology:• Thermochemical I-S cyclesschematic drawing of multilayer FCIs


H. CHEN et.al, Chinese J. Nucl. Sci. & Eng. (2005)

ASIPP

FDS-ST: Spherical Tokamak-Based Reactor

FDS

Based on Spherical Tokamak --- Neutron SourceCore: fusion power=100~200MW, Neutron Wall Loading =0.5~1MW/m2

Blanket: optional (DWT); CCP: innovative concepts of Center Conductor Posts

Outboard can be designated to be a subcritical system with a high multiplication of energy, which can compensate the large fraction of recirculating power in a ST and mitigating the requirement of the neutron wall loading, leading to a strong irradiation on FW, to achieve the highly economical operation.

The plasma βT in a ST can be high enough such that resistive TF can be small to reduce Joule losses in TF coils made of normal conductor. This eliminates the need for a thick inboard shield for cryogenic toroidal-field coil so that fusion devices with smaller major radius are possible.

Advantages:- Improved performance of tokamak plasma- Low neutron wall load of 0.5~1MW/m2- High multiplication of energy


L.J. QIU et.al, Nuclear Fusion, Vol.40, No.3y (2000)

ASIPPInnovative Concepts of Center Conductor Post

FDS

Great Challenges of CCP:- Compact inboard space- Serious radiation effects- Large Joule losses

Requirements:- Protect CCP against radiation- Prolong lifetime of CCP- Reduce Joule losses- Must be replaceable, reliable, and

maintainable

Innovative Concepts of CCP:• Liquid Li self-cooled (Li-SC)

avoid radiation issues, enhance TBR • Water-cooled Copper (water-Cu)

easy assembly and replacement • Water-cooled solid Li (LM-Li)

recovery of CCP, enhance TBR• Liquid metal-blanketed Copper (LM-Cu)

LiPb as shielding for CCP , enhance TBR______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

Y. WU et.al, Fusion Technology, Vol.35, No.1, 1999.Y. KE et al, Plasma Science and Technology, No.3, 2002.Y. WU et.al, Fusion Eng. and Design 51-52, 2000.J. YU et.al, Vol 307-311, 2002Y. KE et al, Nuclear Techn. (in Chinese)，Vol. 26, 2002.

FDSASIPP

Main Plasma Core Parameters of FDS Reactors

ITER**EAST*FDS-STFDS-IIFDS-IParameters

0.270.20.20.540.1Average surface heat load (MW·m-2)

0.574.99E-41.02.720.5Average neutron wall load(MW·m-2) ≥10/5313Energy multiplication /Q73/198050Auxiliary power /Padd(MW)3/5.55.03.5Safety factor /q_95

5.34.02.55.96.1Toroidal field on axis (T)151.59.2156.3Plasma current (MA)

0.330.450.450.60.4Triangularity 1.701.82.51.91.78Plasma elongation3.14.21.434Aspect ratio20.461.021Minor radius(m)

6.21.951.464Major radius(m)5000.081002500150Fusion power (MW)

* : Phase-III; ** : D-T phase


FDSTest Blanket Modules• DFLL (Dual Function Lithium Lead) –TBMs for ITER/EAST

To validate and demonstrate the technology of He-cooled and

He/LiPb dual-cooled liquid lithium lead breeder blanket.


Common Features of Blanket Concepts

• Structural Material: RAFM (e.g. CLAM)/+ODS• Insulation/Anti-erosion/T-permeation barrier layer:

---Coating(Al-based) /insert(e.g.SiC)• Tritium Breeder/Neutron Multiplier: LiPb• Coolant: He or He/LiPb

FDSASIPP

TBM

(DWT, DLL, SLL, HTL)

To define TBM for Testing in ITER:Dual Function Lithium Lead liquid breeder TBM

626

1832

476______________________________________________________________________________________________________Y. Wu, presented at the First IAEA Technical Meeting on the First Generation of Fusion Power Plants, July 5-7, 2005, Vienna

DFLL-TBMs for ITER and EAST

FDSASIPP

LiPb Tank

TBM

Pump

Flow MeterGas injection

Gas injectionHe inlet

He outlet

Dump Tank Fill/Drain

Heating

HeHe Detritiation unit

He inlet

He outlet

LiPb/HeHeatExchanger

Quasi-StaticLiPb(SLL)

Dual-cooled(DLL)

DFLL(Dual-Functional Lithium Lead)-TBM system is designated to check and validate the technology of both SLL and DLL blankets, therefore two types of SLL-TBM and DLL-TBM are to be tested with as similar as possible basic structure and auxiliary system except for including FCIs and quicker flowing LiPb in DLL-TBM. The DFLL design allows the strategy of earlier testing of SLL-TBM, evolving to later testing of DLL-TBM after the issues on FCI and MHD effects etc. can be solved.

Main differences between DLL and SLL• The SLL/DLL-TBM structures are similar• No FCIs in the SLL-TBM• LiPb is only selected as breeder in the SLL-TBM• No heat exchange between LiPb and He at transport for SLL• Mass flow rate of He in SLL-TBM will increase about 20%


Yican WU, ITER TBWG-13 Meeting, Garching, July 7-9, 2004

Design Parameters of DLL/SLL-TBM

ASIPPFDS

VLiPb=~1 mm/s ; Tin/out = /450 ℃VLiPb=20/11/5 mm/s ; Tin/out = 480/700 ℃3 rows poloidal flowing channel;

Breeder/multiplier LiPb

3-stage collector; Thickness: 20/10/10/20 mmHe collectorVHe =47m/s; Tin/out He = 392/402 ℃Thickness: 32 mm; 8 parallel cooling channels; (8 x 16) mm2, pitch 17.5mm;

Covers

VHe =48m/s; Tin/out He = 392/414 ℃

Thickness: 10 mm (3/4/3); Cooling channels: (4 x 8 )mm2, pitch 11 mmStiffening plates“7” type tpSP; rpSP

Tin/out He = 300/392 ℃; VHe = 49 m/s

U-shape; Toroidal He cooling; 4 paths; Thickness: 30 mm (5/15/10)Cooling channels: (15 x 20 )mm2, pitch 25 mm; First Wall

He: Tin/out = 300/410 ℃; Pin = 8 MPa; Qtot =1.96kg/sHe: Tin/out = 300/410 ℃; Pin = 8 MPa; Qtot =1.66kg/sCoolant He

1.12MW1.15MWTotal deposited power

Pol. 1832 mm × Tor. 626 mm × Rad 476 mm (without external headers)Gap TBM/Frame = 20 mmTBM dimensions

China Low Activation Martensitic steel (RAFM Steel)Structural material

Ave.0.3MW/m2, Max. 0.5 MW/m2

0.78 MW/m2Heat FluxNeutron Wall Load

SLL-TBMDLL-TBM


Forced Convection LiPb Experimental Loop

blanket

glove box

coldtrap

heater

MHD test

N S

EM--pump

Dump tank

heater

Ar tank

corrosiontest

Tensile corrosiontest

Expansion tank

specimen handlingmechanisms

Devices for tensileload test

heat exchanger

TL

Gas valves

Gas valves

Gas valves

Gas valves

Gas valves

Liquid Metal Valve

Liquid Metal ValveLiquidMetal Valve

LiquidMetal Valve

Gas trap

Gas trap

Gas trap

flow meter

flow meter

flow meter

flow meter

EM--pump

Vacuum Pump

LiPb/He system for TBM in ITER

ASIPPFDS

LiPb Tank

TBM

Pump


Gas injectionHe inlet

He outlet


Heating

HeHe Detritiation unit

He inlet

He outlet


Quasi-StaticLiPb(SLL)

Dual-cooled(DLL)

Development Roadmap/Plan of TBMs

Thermal convection loopForced convection loop

LiPb/He system for TBM in EAST

LiPb Tank

TBM

Cold Trap

Detritiation unit

Pump


Gas injection

Flow Meter

He inletHe inlet

He outlet He outlet


He Purge

Magnetic Trap

Plugging meter

CT Heat Exchanger

C Tdraining


Heating MT draining

electricalresistivity meter

Ar tank

Gas Valve

Vacuum Pump

Cold leg Hot leg

Water Jackets

Heated leg

Specimen

Expansion tank

Measurement &control System

R&D on materials development (RAFM, Coating and FCI, MHD) and fabrication technology of TBM,diagnostic and measurement, out-of-pile test of mockup etc.

1:3 mockup, concerning mainly EM and thermo-mechanics effects, partially neutronicseffects

confirming EM/ Thermo-mechanics test of EAST, neutronics, tritium and integration test


Main parameters comparison of ITER and EAST

EAST superconducting tokamak in comparison with ITER

A comprehensive testing platform for blanket technology

a fully superconducting tokamak device, is expected to start its operation in 2005/2006

400s (burn ) ,1800s

(repetition)

100-200 s (flat-top) 1800 sec

(repetition

~1000 secPulse length

2200mm(H)×1748mm(W)970mm(H)×528mm(W)Dimension of port0.270.110.200.160.08Avg.HF (MW/m2)

1.77E+2064E+1522 E+159E+15Neutron Rate (n/s)5.34.03.53.5BT(T)123252010Ptotal(MW)1.773.563.82.42bN

0.672.072.871.84bp(%)0.350.450.40.3D1.71.81.81.7k20.460.460.46a(m)

6.21.951.951.95R(m)151.51.01.0Ip(MA)

1.0140.60.60.3ne(1020m-3)

D-T phaseD-D phaseH-H phasePhase IIIPhase IIPhase IITEREAST(D-D)

ASIPPFDS


Y. WU et. al, Plasma Sci. and techn., 2005; Also ITER DD (2005).

Test in EAST (D-D operation) before ITER

1. EAST-SLL-TBM─ Test of Effects of EM and MHD on TBMs and on Plasma─ Test of structural material with coating anti-corrosion─ Test of FW anti-heat flux and capacity of removing heat

2. EAST-DLL-TBM─ Validation of endurance heat and anti-corrosion for FCI ─ Testing EM and MHD ─ Measurement of neutrons and validation of instruments

EM-TBM (1:3 mock-up):

EM and Thermo-Mechanics Testing, partly Neutronics Testing

ASIPPFDS


H-H phaseElectroMagnetic-TBM (SLL-TBM):will start to be tested on the H-H phase to confirm the EAST testing results for a period.

D-D phaseNeutronics-TBM (SLL-TBM):Measurement of neutron fluence and spectra. Validation of neutronics codes and nuclear data.

D-T phaseTritium-TBM during low duty D-T (DLL-TBM):─ Measure of production rate and inventory of tritium ─ Measure of tritium permeation inventory, ─ validation of anti-permeation technology. ─ Measure of stress and temperature field distributing Integral-TBM during high duty D-T (DLL-TBM):─ Demonstration of DEMO blanket integrated performance ( EM, thermal, structure, neutronics, removal heat capacity of cooling system.) ─ Test of MHD integrated effects

Test in ITER

ASIPPFDS


Schedule of DFLL-TBM developing & testingin EAST and ITER

ASIPPFDS


FDSDesign Optimization and Tools

• VisualBUS (MCAM, SNAM.HENDL) :

Integrated Multi-functional Neutonics Analysis

• TOPCODE(SYSCODE+RiskA) :

Integrated Safety, Economics and System Optimization Analysis

• CFD(Computational Fluid Dynamics) + MHD Simulation

• Material Irradiation Simulation

• Virtual Assembly


ASIPPFDS

An integrated neutronic analysis code platform has been developed by integrating and improving existing codes

(Neutronics and Nuclear Data)Design Optimization and Tools Development

VisualBUS – Functions & Features

─ Transport Calculation: 1D/2D/3D-SN & MC methods─ Burn-up Calculation: Bateman/ Runge-Kutta methods─ Activation and BHP Calculation: Bateman method─ Parameters Optimization: GA / ANN algorithms

─ Auto-Modeling & Visualization: MCAM, SNAM─ User Interface: Graphical User Interface─ Distributed Calculation and Network-based Operation

Hybrid Evaluated Nuclear Data Library (Ver.-1)Evaluated Data for 264 isotopes selected from various national libraries:─ Light and Middle Elements : from the IAEA/FENDL2.0─ Heavy/FP Isotopes : ENDF/B-VI4 sub-libraries are nearly ready:─ 2 Transport sub-libraries

MG:Group-wise, 175N/42G; MC:point-wise, ACE-formatted)

─ 1 Burnup sub-library (BU)─ 1 Response Function sub-library (RF)

HENDL

MCAM (MCNP Automatic Modeling)─ An interface code for modeling and visualization for MCNP

SNAM (SN Automatic Modeling)─ An interface code for modeling and visualization for SN codes

1. CAD MCNP file Conversion2. MCNP file CAD Reverse Conversion3. MCNP Attributes Editing

─ Cell/Surface Properties─ Material Info─ Importance Info

4. 3D Model Creation

MCAM: An Interface Program between MCNP and CAD Softwares

ASIPPFDS


• ITER Benchmark Model

ASIPPFDS

Successfully converted the ITER benchmark model into MCNP input

Application Examples


Application Examples

• Converted ITER Brand Model

ASIPPFDS

For the first time, got the full 3D visualized MCNP model of ITER

Manifolds is symmetrical

Manifolds are not symmetrical

Old blanket around equatorial ports

New blanket

Discrepancies between CAD and MCNP models have been found ?

Outlet


SNAM: An Interface Program between SN codes and CAD Softwares

ASIPPFDS

1. CAD SN codes (VisualBUS, DOORS etc.)2. SN code input 3D Drawings3. SN code output 3D Visualization4. Model Attributes Edit

─ Surface Properties─ Material Info

5. 3D Model Creation


ASIPPFDS

─ MHD effects : FLUENT+User-rountine─ CFD simulation : FLUENT, CFX─ Stress Analysis : ANSYS

(Thermal-hydraulics/Thermo-machanics)Design Optimization and Tools Development

MHD Simulation and Evalaution

LiPb inlet

LiPb outlet

Symmetry face

3D MHD simulation3D ModelCAD Model


Fusion Material Database LibraryUnder development at ASIPPFunction of FMDL:– Collecting various data of fusion materials – Contributing to R&D on fusion materials– comparisons

Main contents:– Compositions (designed/measured)– Properties

(physical/mechanical/radioactive/nuclear/chemical/manufacturing/economic)

– Applications (if yes, references/comparisons)

ION损失给反冲核

红: ION损失给靶电子蓝: RECOIL损失给靶电子

离子传给各种靶原子核的能量(主要是反冲核),加上离子传给靶电子的能量,之和就是离子在靶中的能量损失.

Simulation of Ion Transport（使用SRIM程序等）

离子迳迹

Activation Analyses of Nuclear Materials(使用FISPACT程序等）

• Dose rate, DPA etc.

• Impurities Effect

• Transmutation

图（5.2f）t=1.7皮秒

图（5.2a）t=0.2皮秒

图（5.2b）t=0.5皮秒图（5.2d）t=1.1皮秒

Cu铜辐照损伤演化过程的仿真

MD Simulation of Neutron Irradiation（使用MDCASK程序等)

ASIPPFDS

Design Optimization and Tools Development( Material, started )


ASIPPFDS

─ Deterministic Transient AnalysisNTC2D: Neutronics-Thermalhydraulics Coupling calculation code for transient analysis

─ Probabilistic Reliability and Risk AnalysisRISKA :the Risk Analysis Code System developed as an advanced general-service tool for PSA, including fault tree and event tree analysis, important analysis, common cause failure analysis, human failure analysis, uncertainty and sensitivity analysis, etc. Development of the fusion-oriented version of RISKA is planned

─ Material Activation and Potential Environmental Impact

VisualBUS/ACT module

(Safety and Environmental Impact )Design Optimization and Tools Development

10-3 10-2 10-1 100 101 102 10310-1210-1110-1010-910-810-710-610-510-410-310-210-1100101102103104105

? ?

??

?(S

v/h)

? ? ? ? /a

? ? ? W? ? ? Cu? ? ? RAFM12? RAFM13? LiPb13? SiC19? LiPb19? SiC20? RAFM

hands-on

SRMCRM


ASIPPFDS

(System Optimization and Economics )Design Optimization and Tools Development

The Economics Difference between a subcritical system and a pure fusion systemmainly from the extra benefit (BOE: Benefit of Electricity) of a multifunctional blanket which can transmute high level wastes and produce net fissile fuel and tritium if needed. The net cost of electricity (COE’) can be represented by

COE’ = COE –BOE = (TCC* FCR0 + COP– BOP)/ (H0faPE)SYSCODE: a System Analysis Code integrating the details of physical, engineering and financial models developed for the Cost-and-Benefit calculation of fusion and fusion-fission hybrid systems with a function of multi-objective parameters optimization based on the Generic Algorithms and sensitivity/uncertainty analysis.TOPCODE: a more integrated code system is planned to achieve comprehensive Risk-Cost-Benefit analysis on fusion and related systems, including analysis of different scenarios considering carbon dioxide emission restrictions, price of fuel for fission and fusion, waste disposal, public attitude to risk and acceptance of fusion. It may be a coupling tool of SYSCODE and RISKA for the Risk-Cost-Benefit analysis.


ASIPPFDS

─ Precise real time collision detection.─ Interactive and automatic path planning in assembly─ Synchronous multi-screen displaying and data field visualization.─ Entity modeling and enduing physical properties for virtual objects

( Virtual Assembly )Design Optimization and Tools Development


ASIPPFDS

Summary

1. Very strong activities on Fusion-related system designs, with emphasizing blanket design optimization by FDS Team, are underway in China

2. Design of four concepts of Fusion Power Plants, four concepts of blankets with LiPb tritium breeder, and the related evaluation of safety, economy and environmental impact are being performed in parallel. The studies are still evolving.

3. Basic design and analysis tools are being developing by making full use of recent computer technology progress including CG, IC and Network technology.


Thanks !

conceptual design activities of fds series fusion power · pdf file─fds-st : a spherical...

Documents