1. OVERVIEW OF SAFETY OF EUROPEAN FUSION POWER PLANT DESIGNS
Annual Meeting on
Nuclear Technology
May , 2005
Nuremberg
S.Ciattaglia, aL.Di Pace, W.Gulden, P.Sardain, bN.Taylor
EFDA Close Support Unit, Garching bei Muenchen, Germany
aEuratom/ENEA Fusion Association, Frascati, Rome, Italy
bEuratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK

2. Outline Introduction Fusion Reactor Design Overview
Safety analysis and main results
Overview
Accident analysis
Doses to the Public
Occupational Safety
Waste management
Conclusions

3. 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
potential of becoming a clean, zero-CO2 emission and inexhaustible energy source

4. 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

5. 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

6. Schematic diagram of a tokamak fusion power plant
Vacuum
Vessel

7. General layout

8. 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

9. 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
which is currently being characterised in the European fusion programme,

10. Plants main features Blanket DV 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
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
0.7
DV Peak load (MW.m-2) 15 10 5
Average neutron wall load
2.2
2.0
2.4
Major Radius (m)
9.55
8.6
7.5
6.1
Structural material
Eurofer
SiC/SiC
Coolant
Water
Helium
LiPb/Helium
LiPb
Breeder
Li4SiO4
TBR
1.06
1.12
1.15
CuCrZr
W alloy
Armour material
Conversion Cycle
Rankine
Brayton
Blanket DV

11. 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

12. Model A: LiPb Blanket Eurofer as structural material Water as coolant
LiPb as breeder and neutron multiplier
Cut view
Side view
manifold
Outboard
Module
a 20˚ sector

13. Model A: water-cooled Divertor
Low temperature Dv
High temperature Dv

14. 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

15. 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)

16. Model C : DC Blanket scheme
RAFM = Reduced Activation Ferritic Martensitic; ODS = Oxide Disperse Strengthened

17. 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

18. 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

19. Nuclear fusion reactor safety analysis approach
SOURCE
TERMS
ASSESSMENT
Normal working
conditions
Occupational dose IE AS
Thermodynamic transients
Aerosols and H3 transport
Containments
Release from the plant
DCF
Overall Plant Analysis
FFMEA
Radioactive waste
Operational&Decomm waste
Identification&classification
Management
On-site
Recycling
Final disosal
Effluents
PST
EST
man*Sv/y
dose/sequence
to MEI
frequency*dose
Quantity and
waste
categories
mSv/y

20. Overview
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
PPCS safety analysis has a reference the main conclusions of ITER safety analysis
Substantial safety and environmental advantages
Low levels of fuel inventory in the burning chamber
Relatively low levels of after-burn power density and long-term activation
These advantages depend upon the details of design and materials selection, so full safety and environmental analyses are being performed for each model
Results available for models A and B.
PIEs grouped for typology. Considering radioactive inventory mobilization and possible environmental release, four main accidents initiators were pointed out for deterministic assessment:
Ex-Vacuum Vessel loss of coolant
In-Vacuum Vessel loss of coolant due to ex-VV LOCA
Loss of Flow inducing an in-VV LOCA
Loss of Heat Sink.

21. Accident analysis 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
Tens of GJ in the magnets

22. Decay Heat Models A & B

23. Plant Model B Plant Model A DT
Bounding Temperature Accident Analysis (PFC highest temperature reached by 5 days for Mod A, by days for Mod B)
Plant Model A
Plant Model B DT

24. Activation of tokamak structures and components
Dominant radioisotopes
H3, Be10, Ni 63, C14, Co60, Nb94, Ag108m,
Specific activity of the mid-plane outboard
first wall in four Plant Models
Some dominant radioisotopes
H3, Be10, Ni 63, C14, Co60, Nb94, Ag108m,

25. Accident analysis
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

26. BDBA - Model B Loss of flow in one primary cooling loop with consequential in-vessel LOCA
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
EST for 24-h time interval

27. 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
UFOTRI and COSYMA computer programs were applied to assess the consequences of accidental tritium and activation product releases, making reference to German regulations and to a standard set of weather data representing a German site. Here, the dose criterion is defined as committed effective dose equivalent to the first 7 days exposure to the most exposed individual (MEI). This includes the exposure pathways, external irradiation from the passing cloud, the first week of external irradiation from the ground, the internal exposure from inhalation, plus skin absorption, and the internal exposure from inhalation and skin absorption from the reemitted radionuclides during the first week. Dose conversion factors according to ICRP-60 were applied in the public dose calculations. It was assumed that the release takes place over a 24-hour period, which is a conservative assumption, as the characteristic release time ranges typically from 1-7 days. Another important parameter is the release height, which was set to ground level as this results in the highest dose in the vicinity of the plant. Table 2 reports the maximum dose value to MEI at the plant fence.

28. DBA for Model A: ex-vessel LOCA
7-day ACP release <1 mg
7-day T release <3 mg
MELCOR results DT
ACP ST
Ex-base: m3 suppression tank (2000 m3 wet pool)
Drain tank: 5000 m3
Rupture disk between drain tank and vault=2m3
Pipe area between vault and suppression tank (equivalent rupture disk area): 40 m3
Exp p5:
m3 ST (1/10 wet pool)
80 m2
TCWS T
Pressure in TCWS vault, ST and DT

29. Doses to the Public
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 environment
Conservatively calculated worst case doses to the MEI
MODEL A: 1.2 mSv
MODEL B: 18.1 mSv
DBA and BDBA for Models A ad B
Worst 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
Model C and Model D worst case doses lower than those for Models A and B, based on the results of the activation analysis and temperature transient calculations, and other features of the designs

30. Occupational Radiation Exposure
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

31. 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 years on the plant allows reduction of radioactivity level waste masses

32. Waste Management: evaluation and categorisation
Wastes from model B
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

33. Waste Management: masses

34. Waste Management 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
ITER Waste
According to ITER decommissioning plan
the total wastes of ITER wastes are about t
the waste amount after 100 yr from shutdown would be about t

35. 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
Plasma performance only marginally better than the design basis of ITER is sufficient for economic viability of fusion reactors
Conceptual design of a helium-cooled divertor capable of tolerating a peak heat load of 10 MW/m2
Definition of a maintenance concept capable of delivering high availability (75%)
A first commercial fusion power plant - accessible by a “fast track” route of fusion development - will be economically acceptable, with major safety and environmental advantages

36. Conclusions: issues and relevant R&D
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
development of divertor systems, capable of combining high heat flux tolerance and high temperature operation with sufficient lifetime in power plant conditions
PFCs erosion/deposition and transport in SOL
Tritium retention and distribution in the tokamak
Detritiation techniques

37. 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