OVERVIEW OF SAFETY OF EUROPEAN FUSION POWER PLANT DESIGNS

EFDA EUROPEAN FUSION DEVELOPMENT AGREEMENT

OVERVIEW OF SAFETY OF EUROPEAN FUSION POWER PLANT DESIGNS

Annual Meeting on

Nuclear Technology

May 10 - 12, 2005

Nuremberg

S.Ciattaglia, aL.Di Pace, W.Gulden, P.Sardain, bN.Taylor

EFDA Close Support Unit, Garching bei Muenchen, GermanyaEuratom/ENEA Fusion Association, Frascati, Rome, Italy

bEuratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK


Outline

•Introduction

•Fusion Reactor Design Overview

•Safety analysis and main results

• Overview

• Accident analysis

• Doses to the Public

• Occupational Safety

• Waste management

•Conclusions


Introduction

EFDA aim: to define and manage the EU R&D programme on nuclear fusion• JET• ITER• LT: Reactor study, IFMIF, ..

• From 1990 to 2000 preliminary studies on safety, environmental and economic potential of fusion power has demonstrated:

• the potential to provide energy with inherent safety and favourable environmental features,

• the cost of fusion electricity likely to be comparable with that from other sources of electricity generation

Further progress on experiments and R&D meanwhile:• substantial advances in the understanding of fusion plasma physics and in the

development of more favourable plasma operating regimes • progress in the development of materials and technology


Introduction (2)

• PPCS (Power Plant Conceptual Studies):• A comprehensive design study for commercial fusion power plants

performed from mid 2001 to mid 2004, to serve as a better guide for the further evolution of the fusion development programme

• Focussed on four (+1) power plant models, named PPCS A to PPCS D plus model AB, spanning a range from relatively near-term concepts, based on limited technology and plasma physics extrapolations, to a more advanced conception

• They differ from one another in their size, fusion power and materials compositions, and these differences lead to differences in economic performance and in the details of safety and environmental impacts

• The study was carried out with the help of a large number of experts from both the European fusion research community and its industrial partners


Fusion Reactor Design Overview

• Objectives• Demonstration of:

• Credibility of fusion power plant design• Safety and environmental advantages of fusions power• Economic viability of fusion power

• Set of requirements issued by industry and utilities• Safety• Operational aspects• Economic aspects

• Economic, safety and environmental analyses of these models were performed too


Schematic diagram of a tokamak fusion power plant

Breeding Blanket

Poloidal Field Coil

Toroidal Field Coil

Power Conversion System

Supply Electric Power to the Grid

Heating &Current drive

IsotopeSeparation

PumpingD+T+ashes

Breeding Blanket

Poloidal Field Coil

Toroidal Field Coil

Power Conversion System

Supply Electric Power to the Grid

Heating &Current drive

IsotopeSeparation

PumpingD+T+ashes

Vacuum

Vessel


General layout


Key parameters

• 1500 MWe

• Fusion power is determined by efficiency, energy multiplication and current drive power

• Given the fusion power, plasma size mainly driven by divertor considerations


PPCS main elements

• All the models PPCS A to D are based on the tokamak concept as the main line of fusion development proceeding through JET, the world’s largest and most advanced operating machine, that provides the basis for the plasma physics of ITER, under the final design phase

• Two main elements were focused in the design:

•Blanket:

• Takes the energy of the energetic neutrons produced by the fusion process (4/5th)

• Neutrons are absorbed by Li atoms to produce tritium (the fuel together with deuterium)

•Divertor

• for exhausting the fusion reaction products from the plasma chamber, mainly helium, and the associated heat power


Plants main features

Model A Model B Model C Model D

Fusion Power (GW) 5.0 3.6 3.4 2.5

Blanket Gain 1.18 1.39 1.17 1.17

Plant Efficiency 0.31 0.36 0.44 0.6

Bootstrap Fraction 0.45 0.43 0.63 0.76

Padd (MW) 246 270 112 71

H&CD Efficiency 0.6 0.6 0.7 0.7

DV Peak load (MW.m-2) 15 10 10 5

Average neutron wall load 2.2 2.0 2.2 2.4

Major Radius (m) 9.55 8.6 7.5 6.1

Structural material Eurofer Eurofer Eurofer SiC/SiC

Coolant Water Helium LiPb/Helium LiPb

Breeder LiPb Li4SiO4 LiPb LiPb

TBR 1.06 1.12 1.15 1.12

Structural material CuCrZr W alloy W alloy SiC/SiC

Armour material W alloy W alloy W alloy W alloy

Coolant Water Helium Helium LiPb

Conversion Cycle Rankine Rankine Brayton Brayton

Bla

nke

tD

V


PPCS A and PPCS B

• Limited extrapolations in plasma physics performance compared to the ITER design basis

• Blanket• based, respectively, on the “water-cooled lithium-lead” and the “helium-cooled

pebble bed” concepts, using of a low-activation martensitic steel (Eurofer) as structural material

• Divertor• water-cooled divertor (Model A) is an extrapolation of the ITER design and uses

the same materials• helium-cooled divertor (Model B), operating at much higher temperature,

requires the development of a tungsten alloy as structural material• Balance of plant

• model A based on PWR technology, which is fully qualified• model B relies on the technology of helium cooling, the industrial development of

which is starting now, in order to achieve a higher coolant temperature and a higher thermodynamic efficiency of the power conversion system


Eurofer as structural materialWater as coolant LiPb as breeder and neutron multiplier

OutboardModulea 20˚ sector

Lithium-lead

Double Walled Tube

First Wall Tube

Stiffener

Back plate

x

y

Cut view

Side view

manifold

Model A: LiPb Blanket


Divertorcassette

High temperature DvLow temperature Dv

Model A: water-cooled Divertor


Model B: He-cooled divertor

Divertor concept using helium as coolant and W as structural material Peak load of 10 MW/m2 necessity to optimise the heat exchange


PPCS Models C and D

Based on successively more advanced concepts in plasma configuration and in materials technology: the objective is to achieve even higher operating temperatures and efficiencies

• Blanket:• Model C: a “dual-coolant” blanket concept, (helium and lithium-lead

coolants with steel structures and silicon carbide insulators)• Model D: a “self-cooled” blanket concept (lithium-lead coolant with a

silicon carbide structure)• Divertor:

• Model C: the divertor is the same concept as for model B• Model D: the divertor is cooled with lithium-lead like the blanket (the

pumping power for the coolant is minimised and BoP simplified)


RAFM = Reduced Activation Ferritic Martensitic; ODS = Oxide Disperse Strengthened

Model C : DC Blanket scheme


PPCS Safety analysis

• Aim• Critical design review and relevant recommendations in order to• Demonstrate that no design-basis accident and no internally

generated accident will constitute a major hazard to the population outside

the plant perimeter, e.g. requiring evacuation•

• Technique adopted to select accident sequences• Functional Failure Mode and Effects Analysis methodology to find out

representative accident initiators• a plant functional breakdown for the main systems • a FFMEA for each lower level function

• Two design-basis accidents and two beyond design basis accidents chosen and

analysed in detailed for both Models A and B, as well as a bounding accident

scenario


Overview

• Fusion Reactors will produce and contain radioactive materials that require careful management both during the operation (avoiding release in normal and accident conditions) and after decommissioning

• Main radioactive mobilisable inventories• tritium in the in-vessel components and in the fuel cycle• activated materials (dust originating from plasma-PFC interaction and

corrosion products)

• Energies that can mobilise the above inventories during accident conditions• plasma• decay heat• electromagnetic• chemical• coolant/cryogenic


SOURCE TERMS

ASSESSMENT

Normal working conditions Occupational dose

IEAS

Thermodynamic transients Aerosols and H3 transport

Containments Release from the plant DCF

Overall Plant AnalysisFFMEA

Radioactive waste Operational&Decomm waste

Identification&classification

Management•On-site•Recycling•Final disosal

Effluents

PST

PST EST

DCF

man*Sv/y

dose/sequence to MEI

frequency*dose

Quantity and waste categories

mSv/y

Nuclear fusion reactor safety analysis approach


PPCS safety analysis has benefited by the main conclusions of ITER safety analysis, which are:

• A comprehensive analysis of off-normal events and failures and combination of failures postulated to

critically verify the design • Source Terms: 1 Kg of tritium, 100 Kg of Be-dust 100 Kg W-dust, 200 Kg of carbon dust, 10 Kg of

ACP/loop• ORE design target: < 0.5 person Sv/y• Energies evaluation: fusion power, plasma, magnetic, decay heat, chemical, coolant, cryogenic • Protection/Mitigation systems definition (VV suppression tank, plasma shutdown, HVAC systems

and capability of dust and tritium filtering)• Low decay heat at shutdown (Tmax of PFC = 360 ºC after 9hr in case of LOCA in-vessel)• Radioactive releases for all accident events below the project release guidelines (relevant doses ~

average annual natural background) • DBA and BDBAs (e.g. all cooling systems not operating or common cause failure damaging both

vacuum vessel and cryostat) result into: • no need for evacuation, (<50 mSv)• Tmax of PFC ~ 650 ºC

Overview


• PPCS source terms (model A and B)• 1 kg of T in PFC plus a few kg in the Fuel Cycle, • 10 kg-dust in plasma chamber on hot surfaces • 50 kg/loop of ACP in Model A

• Energies• Decay heat: 66 MW at 1 min for Model A; 39 MW at 1 min for Model B• Plasma magnetic energy 3.1 GJ (model A), 1.8 GJ (model B)• Plasma thermal energy 4.3 GJ (model A), 2.5 GJ (model B)• ~50 GJ in the coolant loops (Model B)• Several 10 GJ in the coils

Accident analysis


Decay Heat Models A & B


Bounding Temperature Accident Analysis (PFC highest temperature reached by 5 days for Mod A, by 50-60 days

for Mod B)

Plant Model A

Plant Model B

T

T


Specific activity of the mid-plane outboard

first wall in four Plant Models

Some dominant radioisotopesH3, Be10, Ni 63, C14, Co60, Nb94, Ag108m,

Activation of tokamak structures and components


• Bounding accident sequences: complete unmitigated loss of cooling; no safety systems intervention

• Temperature distribution (ºC) in Plant Model B (a) and in Plant Model D (b) (100 days after onset of bounding accident scenario)

• Maximum temperatures never approach structural degradation for all models

Accident analysis

(a) (b)


• Parametric analyses on the building leakage rate from Expansion Volume;

• Possibility to operate an Emergency Detritiation System to reduce environmental releases

ECART results

Pressure inside VV, EV and PS

BDBA - Model B Loss of flow in one primary cooling loop with consequential in-vessel

LOCA

Referencecase

Case 1 Case 2 Case 3 Case 4 Case 10

EV lkg. =75% noEDS

EV lkg. =1% no EDS

EV lkg. =75% EDS3.0 kg/s

EV lkg. =10%, EDS3.0 kg/s

EV lkg. =1%, EDS3.0 kg/s,

EV lkg. = 1%,EDS 3.0 kg/s,plus scrubber

Tritium [g] 52.8 3.5 30.5 8.1 1.9 2.0

W dust [g] 102.0 4.6 62.0 15.0 3.3 0.2

Steel dust [g] 323.0 14.5 196.4 47.4 10.5 0.6

EST for 24-h time interval


Doses to the Public

• Environmental source terms• activated dust/ corrosion products• tritium

• During normal operation• negligible release (doses to MEI < 1% of the natural background),• ALARA principle is applied for public and workers

• No emission of any of the greenhouse gases

• Conservatively assumed

• a mobilisation fraction of 100 % for the dust at the beginning of the accident sequence

• 90% as HTO for T

• worst atmosphere conditions

• UFOTRI and COSYMA computer programs

• reference to German regulations and to a standard set of weather (for a German site)

• dose conversion factors according to ICRP-60

• the release takes place over a 24-hour period

• the release height set to ground level


Pressure in TCWS vault, ST and DT

TCWSST

DT

7-day ACP release <1 mg

7-day T release <3 mg

ACP

T

MELCOR results

DBA for Model A: ex-vessel LOCA


Bounding accident sequences for Models A and B: complete unmitigated loss of cooling; no safety systems intervention, mobilisation, transport within the plant, release and transport to the environmentConservatively calculated worst case doses to the MEI

MODEL A: 1.2 mSv MODEL B: 18.1 mSv

DBA and BDBA for Models A ad BWorst case dose values (mSv) for the 7-day dose to MEI at 1000 m distance

(24-h release, 95% fractile)

* tritium as HTO

Model C and Model D worst case doses lower than those for Models A and B

Doses to the Public

PlantModel

Ex-VV LOCA(mSv)

Ex-VV LOCA + in-VV LOCA(mSv)

LOFA + in-VV LOCA(mSv)

A 1.7 1E-3 0.1 6 N.A.

B N.A. N.A. 0.4 2 (*)


• A minimisation of ORE was proposed as an important requirement for a fusion power plant

• Defined an annual collective dose target of 0.7 pers-Sv/y as design target

• The main sources are ACPs for Model A (water cooled) and tritium for all Models

• Preliminary results• further optimization is necessary for Model A

• 180 person·mSv per year is the target for Fuel Cycle System. Three fuel cycle systems – the fuelling, vacuum pumping and blanket tritium recovery systems – were highlighted as needing more attention

• development of cryogenic pumps with a larger pumping capacity is recommended

Occupational Radiation Exposure


Waste Management

The fusion radioactive wastes are characterised by low heat generation density and low radiotoxicity. Therefore recycling could be a viable option

Storing the fusion radioactive materials for 50-100 years on the plant allows reduction of radioactivity level waste masses

Table 2 – Classification of fusion radioactive waste

Activated material classifications Contact dose rate

after 50 y (mSvh-1

)

Decay heat per

unit volume after

50 y (Wm -3)

Clearance index

after 50 y [5]

PDW, Permanent Disposal Waste (Not

recyclable)

>20 >10 >1

CRM, Complex Recycle Material 2-20 1-10 >1

SRM, Simple Recycle Material <2 <1 >1

NAW, Non Active Waste (to be cleared) <1


For ALL the Models:

Activation falls rapidly: by a factor 10,000 after a hundred years

Significant contribution to SRM and CRM from operational wastes

Potentiality to have no waste for permanent repository disposal

Also tritiated + activated wastes

Wastes from model B

Waste Management: evaluation and categorisation


NAWSMR

CMRPDW

D

C

BA

0

10,000

20,000

30,000

40,000

50,000

60,000

70,000

80,000

90,000

Tonnes

Radioactive material category

Power Plant model

Figure 2 - Masses of the material after 100 years decay

D

C

B

A

Waste Management: masses


• If no recycling is planned

• the amount of waste to be disposed after 100 years, is equal to the CRM+SRM amounts

•

• Suitability and capability analyses of the final waste repositories in a few EU countries to store the PPCS wastes were performed (Konrad and Gorleben in Germany, SFR and SFL 3-5 in Sweden, CSA in France, El Cabril and DGR in Spain)

• With reference to Model B and German regulations, the fusion reactor waste can be all disposed in Konrad. For a few ones, detritiation is necessary to meet the relevant limits for storage

Waste Management


Conclusions

• The four PPCS conceptual design for commercial fusion power plants differ in their

dimensions, gross power and power density

• All models meet the overall objectives of the PPCS from design, safety,

economics point of view

• Comprehensive safety analysis of PPCS has showed

• “No evacuation” criteria is met with margin also in case of beyond design

basis accidents driven by internal events

• Intrinsic-passive safety features of nuclear fusion plants has been confirmed

also from the bounding accident sequence analyses

• Model B, BDBA LOFA + in-vessel LOCA provides the largest environmental

source terms

• ORE needs attention

• Wastes amount are significant but they are characterised by low decay heat and

low radiotoxicity. There is the potentiality to have no need of permanent waste

disposal after 100 years from shutdown if recycling is applied


Most of the open issues are relevant to the life time of first wall components, in particular divertor and blanket

Physics

• advanced plasma scenarios (improved confinement), in particular• good confinement regime with divertor tolerant ELMs

• regimes with large fraction of plasma current driven not inductively

• control of plasma transients: ELMs, VDEs and disruptions

• SOL (phenomena, transport), particle exhaust and control

Materials&components• optimisation of low activation martensitic steels (Eurofer)

• use of ODS (temperature, welding)

• development of more advanced materials envisaged in PPCS (e.g. W and SiC/SiC as

structural material )

• He cooled divertor

• development and test of blanket and divertor systems

• development and qualification of RH for maintenance, recovery action, test&inspection

Conclusions: issues and relevant R&D


Safety• Control of dust and tritium in the VV (source terms)• Lack of operating experience

• Reliability of prototypes • ORE minimisation

• Waste management• Quantity of operational waste, Tritiated + waste disposal

• Detritiation• Recycling

Answers expected from• ITER operation• EFDA R&D technology programme • DEMO power plant study (launched recently)

PPCS results have ulteriorly demonstrated the potentiality of Nuclear Fusion Reactors as viable and safety source of energy, pointing out the main lines of R&D necessary

Conclusions: issues and relevant R&D

OVERVIEW OF SAFETY OF EUROPEAN FUSION POWER PLANT DESIGNS

Documents

tokamak fusion power

commercial fusion power

power plant models

overview of safety

details of safety

inherent safety

reactor study

environmental analyses