TW4-TRP-002 D5
FUS-TN-SA-SE-R-129
ENTE PER LE NUOVE TECNOLOGIE, L'ENERGIA E L'AMBIENTE
Associazione ENEA-EURATOM sulla Fusione
FUSION DIVISION
NUCLEAR FUSION TECHNOLOGIES
Safety assessment of the
Power Plant Conceptual Study Model AB
R. Caporali(1)
G. Caruso(2)
L. Di Pace(3)
ENEA/TW4-TRP-002 Deliverable 5
EFDA Task TW4-TRP-002
Deliverable 5
(1)
(2)
(3)
ENEA Consultant
University “La Sapienza” of Rome
Thermonuclear Fusion Division
Via E. Fermi 45, I-00044, Frascati (Rome), Italy
E-mail:di_pace@frascati.enea.it, r_caporali@tin.it, gianfranco.caruso@uniroma1.it
Rev. 0 June 2005
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SAFETY ASSESSMENT OF THE POWER PLANT
CONCEPTUAL STUDY MODEL AB
EFDA Technology Workprogramme 2003
EFDA Task TW4-TRP 002
R. Caporali (ENEA consultant, Rome, Italy)
G. Caruso (University La Sapienza of Rome, Italy
L. Di Pace (ENEA FUS TN, C.R. Frascati, Italy)
This report deals with the safety assessment of the PPCS Model AB
carried out in the framework of EFDA workprogramme 2003.
The work is devoted to assess the safety and environmental point of
view the new conceptual design for the PPCS study, the HeliumCooled Lithium Lead (HCLL) Reactor.
The identification of the main accidental sequences has been carried
out by defining a set of postulated initiating events (PIEs) pointing
out representative accident scenarios for deterministic assessment.
The deterministic analyses of some accidental sequences has been
then performed by CONSEN5.1 code, with the aim to check the
secondary confinement design, (e.g. selected volume, design
pressure, size of the rupture disk), to quantify the environmental
source terms and to provide feedbacks to the design.
This report fulfils the EU task EFDA TW4 TRP 002 Deliverable 5.
Issued by
Reviewed by
Approved by
L. Di Pace
M. T. Porfiri
A. Pizzuto
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M. Samuelli, Fusion Division (ENEA, FUS Frascati, Italy)
Technologies (ENEA, FUS-TEC Frascati, Italy)
A. Pizzuto
T. Pinna
M. T. Porfiri
R. Andreani, D. Maisonnier, P. Sardain (EFDA, Garching, Germany)
D. Ward, R. Pampin (UKAEA, Culham, UK)
A. Li Puma, L. Giancarli (CEA-Saclay, France)
L. Buhler (FZK, Karlsruhe, Germany)
B. Branas (CIEMAT, Madrid, Spain)
E. Bogush, A. McCallum, A Orden Martinez, A. Paule, D. Puente
(EFET)
G. Cambi (Bologna University, Bologna , Italy)
ENEA FUS-TEC Secretarial Staff
Authors:
R. Caporali (ENEA consultant)
G. Caruso (University La Sapienza of Rome, Italy)
L. Di Pace (ENEA, FUS, Frascati, Italy)
FUS-TEC Archive
1. giustificare tutto il testo, fare attenzione ad escudere le celle delle tabelle nell’appendice
dell’FFMEA
2. I margini di pagina destro e sinistro sono troppo ridotti, meglio ampliarli.
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Table of Contents
Executive Summary .............................................................................................................................. 5
List of Figures ....................................................................................................................................... 6
List of Tables ......................................................................................................................................... 7
1. Introduction................................................................................................................................... 9
2. Model AB Design Data for Accident Analyses ......................................................................... 10
2.1.
Tokamak Basic Machine Configuration ................................................................................... 10
2.2.
FW/shielding blanket and divertor ........................................................................................... 10
2.3.
Vacuum vessel .......................................................................................................................... 10
2.4.
Cooling loops ............................................................................................................................ 10
2.5.
Other data of interest ................................................................................................................ 11
2.5.1.
Ventilation features for the containment .............................................................................. 11
2.5.2.
Tritium, dusts and other source terms................................................................................... 11
2.5.3.
Pb-17Li characteristics ......................................................................................................... 11
2.5.4.
Data needed to evaluate the status of faulted and un-faulted cooling loops ......................... 11
2.5.5.
Expansion volume ................................................................................................................ 11
3. Selection of accident sequences by FFMEA ............................................................................. 12
3.1.
Methodology ............................................................................................................................. 12
3.2.
LOCA accidents........................................................................................................................ 12
3.2.1.
In-VV LOCA ........................................................................................................................ 12
3.2.2.
Interface LOCA between FW and Breeding Blanket ........................................................... 13
3.2.3.
Ex-VV LOCA ....................................................................................................................... 14
3.2.3.1.
Primary loop LOCA ......................................................................................................... 14
3.2.3.2.
Secondary loop LOCA ..................................................................................................... 14
3.3.
Steam generator tube rupture .................................................................................................... 14
3.4.
LOFA accidents ........................................................................................................................ 15
3.5.
Loss of heat sink accidents ....................................................................................................... 15
3.5.1.
Turbine trip ........................................................................................................................... 15
3.5.2.
Loss of condenser ................................................................................................................. 15
3.5.3.
Load rejection ....................................................................................................................... 15
4. Accidents deterministic analysis by CONSEN 5.1 ................................................................... 19
4.1.
Brief description of CONSEN code (version 5.1) .................................................................... 19
4.2.
Accident sequences analyzed ................................................................................................... 20
4.3.
Plant nodalization ..................................................................................................................... 22
4.4.
Main Thermal-hydraulics Results............................................................................................. 29
5. Final remarks and conclusions .................................................................................................. 57
Appendix 1 MODEL AB DESIGN DATA FOR ACCIDENT ANALYSES .................................................. 59
Appendix 2 FAILURE MODE AND EFFECT ANLYSIS (FFMEA) ........................................................ 77
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List of Figures
Fig. 4.1 –
Fig. 4.2 –
Fig. 4.3 –
Fig. 4.4 –
Fig. 4.5 –
Fig. 4.6 –
Fig. 4.7 –
Fig. 4.8 –
Fig. 4.9 –
Fig. 4.10 –
Fig. 4.11 –
Fig. 4.12 –
Fig. 4.13 –
Fig. 4.14 –
Fig. 4.15 –
Fig. 4.16 –
Fig. 4.17 –
Fig. 4.18 –
Fig. 4.19 –
Fig. 4.20 Fig. 4.21 –
Fig. 4.22 –
Fig. 4.23 –
Fig. 4.24 –
Fig. 4.25 –
Fig. 4.26 –
Fig. 4.27 –
Fig. 4.28 –
Fig. 4.29 –
Fig. 4.30 –
Fig. 4.31 –
Fig. 4.32 –
Fig. 4.33 –
Fig. 4.34 –
Fig. 4.35 –
Fig. 4.36 –
Fig. 4.37 –
Fig. 4.38 –
Fig. 4.39 –
Scheme of a control volume in CONSEN
Dimensions of First Wall cooling channels
PPCS model AB nodalization for the LOFA + in vessel LOCA accident simulation
Case C1 – Pressure transient: Long term phase
Case C1 – Pressure transient: Short term phase
Case C1 – Temperature transient
Case C1 – Flow rates between volumes
Case C1 – Main structures temperature evolution
Case C1 – Tritium inventory in the PPCS volumes
Case C1 – Tritium released through different paths
Case C1 – Dust inventory in the PPCS volumes
Case C1 – Dust released through different paths
Case C2 – Pressure transient: Long term phase
Case C2 – Pressure transient: Short term phase
Case C2 – Temperature transient: Long term phase
Case C2 – Flow rates between volumes
Case C2 – Main structures temperature evolution
Case C2 – Tritium inventory in the PPCS volumes
Case C2 – Tritium released through different paths
Comparison of VV and TCHS pressures in Cases C1 and C2
Case C3: Comparison of the pressure transient between the design values and the following
values (EV 500,000 m3, area VV-EV 0.5 m2)
Case C3: Pressure transient (500,000 m3 EV)
Case C3: Helium flow rates (500,000 m3 EV)
Case C3: Tritium inventories in VV and EV (500,000 m3 EV)
Case C3: Tritium released to the external atmosphere (500,000 m3 EV)
PPCS model AB nodalization for the Ex-vessel LOCA accident simulation
Case C4 – Pressure transient: Long term phase
Case C4 – Temperature transient: Long term phase
Case C4 – Helium flow rates between volumes
Case C4 – Tritium inventory in the PPCS volumes
Case C4 – Tritium releases to the external atmosphere
Case C5 – Pressure transient: Comparison with design values
Case C5 – Pressure transient: Long term phase
Case C5 – Temperature transient: Long term phase
Case C5 – Helium flow rates between volumes
Case C5 – Tritium inventory in the PPCS volumes
Case C5 – Tritium releases to the external atmosphere
Case C6 – Flow rate between secondary and one BL-HTS loop
Case C6 – Pressure transient in the steam generator tube rupture accident
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List of Tables
Table 3.1 Table 4-1 –
Table 4-2 Table 4-3 Table 4-4 Table 4-5 Table 4-6 Table 4-7 Table 4-8 Table 4-9 Table 4-10 Table 4-11 -
Complete list of accident families (PIEs) for PPCS Model AB
Accident sequences analyzed
Leakages laws for VV and EV.
Tritium inventories.
Peak pressure and time in cases C1 and C2
Cases C1 and C2 -Tritium inventories in the volumes after 1 day from the start of the
accident.
Case C1 -Dust inventory in the volumes after 1 day from the start of the accident.
Case C3 -Tritium inventories in the volumes after 1 day from the start of the accident.
Pressure values in the expansion volumes (peak and final values after 1 day)
Case C5 - Tritium inventories in the volumes after 1 day from the start of the accident.
Summary of Tritium inventories and releases in the volumes after 1 day from the start of
the accident.
Summary of dust inventories and releases in the volumes after 1 day from the start of the
accident.
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Acronyms
ACP
ADS
AH
CDC
CODAC
CVCS
DEMO
DS
EV
FFMEA
FMS
FW-BK
FW/BL
HCLL
HT
HTS
HX
ID
ISS
LOCA
LOFA
LTS
PFC
PI
PIE
PPCS
QA
RD
RF
TCHS
UHS
VV
Activation Corrosion Products
Air Detritiation System
Additional Heating
Detritiation System
Expansion Volume
Functional Failure Mode and Effect Analysis
First Wall- Blanket
First Wall/Blanket
Helium Cooleed Lithium Lead
Heat transfer
Heat Transfer System
Heat Exchanger
Internal Diameter
Isotope Separation System
Loss of Coolant Accident
Loss of Flow Accident
Low Thermal Shield
Plasma Facing Components
Postulated Initiating Event
Power Plant Conceptual Study
Quality Assurance
Rupture Disk
Radio Frequency
Tokamak Cooling Helium System
Ultimate Heat Sink
Vacuum Vessel
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1. Introduction
The European Power Plant Conceptual Study (PPCS) was launched in January 2000 and completed at
the end of 2002. The conceptual design of 4 reactor models was performed, ranging from near term
plasma physics and materials (Models A and B) to an advanced model (Model D) based on the use the
most advanced thinkable plasma physics and technology.
Following a review of the European DEMO blanket development programme, it was decided to
consider the Helium-Cooled Lithium-Lead blanket (HCLL) as possible DEMO blanket concept.
A new PPCS reactor model (model AB) has therefore been studied in 2004 based on HCLL blanket and
He-cooled divertor.
The HCLL blanket is based on the use of EUROFER as structural material, of Pb-17Li (Li at 90% in
6
Li) as breeder, neutron multiplier and tritium carrier, and of helium as coolant with inlet/outlet
temperature of 300/500°C and 8 MPa pressure.
A helium cooled divertor capable of tolerating peak heat fluxes of at least 10 MW/m2 has been
considered as reference solution. The back-up solution is given by a water-cooled divertor using
EUROFER as structural material, a coolant pressure and outlet temperature respectively of 15.5 Mpa
and 325°C, and W-alloy monoblocks as armour.
This document summarizes the work carried out by ENEA to assess from the safety and environmental
point of view the Helium-Cooled Lithium Lead (HCLL) Reactor (PPCS Model AB).
The work has been divided into two main lines of activity; the former aimed to define a set of postulated
initiating events pointing out representative accident scenarios possible objects of deterministic
assessment, the latter devoted to the accident analyses of some of the accident sequences individuated.
The CONSEN code version 5.1 has been used to verify the containment response and assess the
environmental source terms released outside during the accident.
The probabilistic analysis for PPCS Model AB has being performed with the help of a Functional
Failure Mode and Effect Analysis (FFMEA), focused on the possible consequences of loss of
component or system functions. Other scope of the FFMEA application has been to ascertain if
previous analyses performed for the former PPCS models can be used as enveloping analyses applicable
also to Model AB.
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2. Model AB Design Data for Accident Analyses
In this section data relevant for performing accident analysis are presented. This list is not exhaustive of
the main design data, but it has only the purpose to present the data used in the accidental scenarios
analysed in the present work. For any other design data make reference to the dedicated documentation,
in particular refer to the relative design general document [1]. Related tables are shown in Appendix 1.
2.1. Tokamak Basic Machine Configuration
The general configuration of the machines, to be used in safety analysis, is shown in table A.1.1
Most of the data were taken from [2], while the plasma thermal energy (2.5 GJ) was given by reference
[3]. The plasma volume and surface was worked out by simple geometric considerations starting from
analogous values given in reference [4] for the PPCS Model A, quite similar in the overall dimensions
(major radius = 9.55 m).
The overall plasma thermal energy is assumed delivered in 1 second to one third of the FW surface in
case of a disruption.
2.2.
FW/shielding blanket and divertor
Relevant thermal and geometric data are given in tables A.1.2 and A.1.3. They are mainly taken from
[1].
The divertor design as written above is characterised by a reference design, helium-cooled divertor and a
back-up solution using water as coolant. In the tables cited above data are given for the divertor
reference design.
2.3. Vacuum vessel
Vacuum vessel features are reported in table A.1.4. EUROFER characteristics are reported in table
A.1.5.
2.4. Cooling loops
It has been assumed nine primary loops for the first wall/blanket [1] and three primary loops for the
divertor [§] (non c’è il riferimento 8 nelle referenze). Both use helium as coolant. (cancellata perché già
detto precedentemente).The secondary loop provides a superheated steam to be sent to turbines after
reheating the steam produced by the 9 steam generators (SGs) of the first wall/blanket primary heat
transfer system. The superheated steam outcomes from the three superheaters exchanging heat with the
helium coolant of the primary divertor heat transfer system (questa frase è un pò contorta e non si
capisce bene lo schema di scambio: è possibile fare uno schemino semplificato con le frecce?)
First wall/blanket cooling loops main features are reported in table A.1.6 as far as the primary and the
secondary loops are concerned.
Divertor cooling loops main features are reported in table A.1.7 for both primary and secondary loops.
Coolant inventories are accurately defined for in vessel components and related piping: the remaining of
the circuits has been extrapolated from previous projects design data (e.g.: Model B).
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2.5. Other data of interest
These are data generally referring to materials properties/parameters, containment and/or source term.
2.5.1. Ventilation features for the containment
Table A.1.8 reports the data for model AB.
2.5.2. Tritium, dusts and other source terms
The data are scaled from models A&B and from SEAFP. Data are reported in table A.1.9.
2.5.3. Pb-17Li characteristics
They are given for solid and liquid state, depending on the temperature, in table A.1.10.
2.5.4. Data needed to evaluate the status of faulted and un-faulted cooling loops
Table A.1.11 gives the release modality in correlation with vault isolation, table A.1.12 defines the
status of not affected cooling loops, table A.1.13 outlines the behaviour of the pressure control system.
Data relevant for heat transmission are given in tables A.1.14 and A.1.15, i.e. emissivity and view.
2.5.5. Expansion volume
In Model B, there was an external expansion volume of 68000 m3 and an internal volume (comprising
the free volume of TCHS, pipechases, north, and east vaults) available for expansion of 49500 m3, for a
total of 117500 m3. However in Model AB, due to the increase of the building dimensions, it is possible
to consider an internal volume of 117600 m3 available for expansion, comprising the free volume of
TCHS, pipechases, north, south,east, west vaults). So it is not necessary to have an additional external
expansion volume.
Table A.1.16 shows the volumes available for the expansion and the related parameters. Table A.1.17
shows the internal volume of different chambers for both models.
The Tokamak Cooling Helium System (TCHS) vault, in Model AB, is fourth times bigger than in
Model B. The West and South chambers are considered also as internal expansion volume of Model
AB. The sum of all new volumes cover the volume of the external expansion volume of Model B.
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3. Selection of accident sequences by FFMEA
3.1. Methodology
The analysis for PPCS Model AB has been performed on the basis of a Failure Mode and Effect
Analysis (FFMEA), focused on the possible consequences of loss of component or system functions: the
FMEA is reported in Appendix 2.
The objective of the analysis is to define a set of postulated initiating events pointing out representative
accident scenarios, possible objects of deterministic assessment: also, it has to be ascertained if previous
analyses performed for the other PPCS models can be used as enveloping analyses applicable also to
Model AB.
The PPCS model AB is similar to model C as far as primary cooling and breeder circuits are involved,
but also similar to model B in the primary to secondary Heat Transfer System (HTS) interfaces.
The FFMEA defines the accident initiators which can be grouped within an accident family: the
complete list of accident families (PIEs) is given in table 3.1.
As most important accidents, and also because of their dependency on the specific design, LOCA,
LOFA, Loss of heat sink accident are considered, characterized by the affected loops and location in
case of LOCA.
These accident are discussed in the following.
3.2. LOCA accidents
This kind of initiators have been considered in a broader sense, including also the breeder loops.
The classification has been made by the location of the initiator.
3.2.1. In-VV LOCA
Possible initiators are related to:
- FW surface towards the plasma (front wall);
- FW surface towards an adjacent module (side wall);
- He manifold within HTS;
- Li-Pb manifold within HTS.
The last initiator is not properly related to a loss of coolant, but it can be logically included being a
dispersion of a hot and pressurized fluid within the VV.
Consequences of the first 3 initiators are:

plasma poisoning and disruption, with a timing that could change a little depending on the location
of the breach but variations are not expected to invalidate the assumption of an immediate
disruption.

He circuit depressurization (1 out of 9), at a rate depending on the size of the breach in the cases of
FW structural failure: the He manifold rupture is assumed a circumferential, double ended one as
usual. The He manifold rupture maximises the rate of circuit depressurisation.

VV pressurization which could require the intervention of pressure relief devices and the opening of
an expansion volume in the worst cases.

possible consequential ruptures due to induced stress and vibration involving modules and HTS
fixing devices, manifold restraints and other structures around the breach zone. A failure of the
integrity of the separation with the Li-Pb circuit within the module could also be hypothesized.
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possible plugging of pressure relief openings/devices if provided in the bottom of the plasma
chamber in the cases where a Li-Pb leak in the VV is involved.
Consequences of a break of a Li-Pb manifold can be:

plasma poisoning and disruption;

possible plugging of pressure relief openings/devices if provided in the bottom of the plasma
chamber;

possible consequential rupture of nearby manifolds (He or Li-Pb) or HTS;

mild VV pressurization.
The main interest of this accident is focused on VV pressurization and on the verification of the
effectiveness of provided measures, as pressure relief devices and expansion volumes.
Li-Pb circuit structural failures towards the VV will add the heat content of the metal to the energy
already available for VV pressurization deriving from He pressure and temperature.
3.2.2. Interface LOCA between FW and Breeding Blanket
Possible initiators are the structural failure of Breeding Units cooling plates or stiffening plates,
including the first wall, towards the breeding units: also the He and Li-Pb circuits could be mixed up
because of a breach or leak within the back plates.
This event pressurizes the Breeding Unit and the consequent in VV LOCA is likely, specially in the case
of the structural failure weakening the FW structure, depending on the effectiveness of the intervention
of pressure relief devices on the liquid metal circuit towards an expansion tank. In this case no account
should be taken of a passive shutdown because there are no impurities entering the VV until the FW
integrity is lost.
Also, the liquid Li-Pb part of the module is pressurized from 1.5 MPa of the liquid metal loop to the 8
MPa of the He primary loop: module collapse should be considered in dependence on the module design
pressure in the liquid metal section.
Module structural collapse could have an impact on the effectiveness of pressure relief devices because
the VV could be pressurised in a very short time, thus overcoming the pressure relief devices capacity.
The pressurization rate in this case is the same as in the in VV LOCA due to He manifold failure: in
both cases, in fact, the in-VV flow rate is the critical one evaluated in the manifold section, which is the
maximum possible in flow.
The consequences are in any case similar to the ones related to the in-VV LOCA accident, i.e.:
- Liquid metal within the VV will add its heat content to the energy available for VV pressurization;
- Liquid metal draining to the bottom of the VV could plug pressure relief openings/devices if
provided there.
If the pressure is not released through an in-VV LOCA, being the Li-Pb circuit overpressure mitigated
through pressure relief devices towards a safety tank, helium could enter the tank and cause its structural
failure, with release of impurities contained in the coolant and in the breeder to the room. If there is the
possibility that He pressurizes the room, there could be outside release.
A surge in the pressure in a module could also stop the liquid metal flow, with a consequent
solidification in the part of the Li-Pb circuit outside the VV. A plugging of both inlet and outlet lines
could make the circuit without possibility of expansion in the part outside the VV. The He gas which
should be trapped in the circuit after causing the plugging of the inlet and outlet lines should be expelled
from the loop in case of liquid metal dilatation due to temperature increase. This could be very bad
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situation where structural failure is almost sure in dependence of a temperature increase causing a
further dilatation of the liquid metal.
The main focus of the analysis of this initiators is to verify the effectiveness of the structural design of
the FW and Breeding Blanket in worst accident conditions.
3.2.3. Ex-VV LOCA
This kind of initiator has been considered in a broader sense, taking into account also breaches of the
secondary loops.
3.2.3.1. Primary loop LOCA
Possible initiator can be any size of leak from the primary, starting from components seal leakages up to
a double ended guillotine break of the main pipe, excluding a SG tube rupture considered in a dedicated
paragraph. The guillotine break can be considered as representative of the worst conditions for the
primary loop depressurisation and for the strictest timing requirements for accident mitigation. A
breach assumed to occur within a primary loop will release cooling He to a tokamak surrounding room.
The room is pressurized with a rate depending on the breach size and the possibility of a structural
failure should be assessed.
A consequential in-VV LOCA can be assumed due to loss of coolant to the FW: since the He loop is
already depressurised the structural failure of the FW towards a breeding loop can be excluded, given
also that an in-VV LOCA will cause an immediate plasma shutdown. The possible release of VV
contained activated products and dust will depend on the differential pressure balance between the VV
and the affected room.
The analysis of this kind of accident should focus on the verification of the structural integrity and leak
tightness of the surrounding room under a worst case room pressurization and on the possibility of
mobilisation of the VV contained source term.
3.2.3.2. Secondary loop LOCA
Small size breaches can be considered relatively safe because of the time available to detect the fault
and to provide mitigations, first of all plasma shuts down. Instead a large breach can be considered as a
loss of heat removal from a primary loop, which could cause an in-VV LOCA because of both primary
circuit pressurization and temperature increase. A consequent breeder loop LOCA is also likely in this
situation, being the FW failure due to He loop overpressurisation: the consequences on modules
structural integrity could then be worse than in the case of a direct in-VV LOCA.
The aim of the analysis is at identifying worst case conditions of an interface LOCA.
3.3. Steam generator tube rupture
The characteristic of the design choice for Model AB steam generator is the higher pressure of the
secondary loop with respect to the primary, so that following this accident steam/water mixture will
enter the He side of the steam generator. He loop pressurisation will depend on the thermodynamic
behavior of the steam/water mixture which enters a loop with higher temperature, and consequent steam
expansion and loop pressurization, but with a pressure loss due to the steam/water mixture cooling down
due to the expansion through the breach.
If water is introduced in the loop due to this accident a primary loop He circulator is expected to stop
because of machine protection devices sensing liquid through the suction side.
In any case water entering also the VV after the primary loop seems to be a low probability outcome:
this could be a very challenging event if the He loop looses its integrity also towards the breeder loop,
thus causing liquid metal and water to come in contact with consequent possible H2 formation.
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The primary loop pressurization is the focus of the analysis, because in this case little time is needed to
achieve structural failure conditions, with primary pressure strictly linked to the pressure of the
secondary loop and likely higher then in the primary loop ex-VV LOCA case.
3.4. LOFA accidents
Such accidents are related to loss of flow in one, or more, primary cooling loops.
The most representative initiator is a stop of the circulator with no coast-down causing an almost
immediate loss of flow in one primary cooling loop; as an outcome FW failure should follow.
Consequences are qualitatively the same as in the in-VV LOCA case, the energy available for VV
pressurization will depend on the He circuit conditions. In fact the circuit pressure tends to increase
because of FW heating up, but also to decrease because of the stop of the circulator.
Loss of flow in a breeder loop seems to have lesser consequences since FW overheating will depend on
the reduced heat transfer from the liquid metal to the cooling plates: the timing for FW overheating
should be relatively relaxed.
3.5. Loss of heat sink accidents
Loss of heat removal by secondary side can be due to many initiators, some related to a single loop, as
valve wrong alignment, some related to all loops.
The latter initiators are more challenging because they could impair the effectiveness of the VV
overpressure limitation deriving from the subdivision of the primary HTS of the Blanket in 9 loops: in
fact all 9 loops could breach towards the plasma chamber thus allowing for the discharge of the overall
inventory of He coolant due to a multiple in VV LOCA. In such a case it is maximized the challenge to
the structure of the VV and then of the expansion volume, provided that pressure relief devices are
successful.
Initiators generating a generalised loss of heat sink are discussed in the following
3.5.1. Turbine trip
Such an initiator can be started by the malfunction of turbine related subsystems and auxiliaries: the trip
is required to protect the machine from damage.
Loss of heat removal through the turbine is lost. In this case it is generally provided a bypass to the
condenser which could allow for heat removal for a limited time, usually from 10 to 20 minutes
depending on the condenser size.
This provision coupled with an immediate plasma shutdown could avert damage to the plasma facing
components.
3.5.2. Loss of condenser
System malfunction and supporting systems unavailability, as loss of service water, could be the causes
of this initiator.
This event is more challenging of the one above because loss of heat removal can not be achieved
through a bypass to the condenser. A few minutes could be available from the start of the accident to the
total loss of heat removal. Immediate plasma shutdown could limit the temperature of the plasma facing
components within acceptable limits, but this is to be ascertained, taking also into account the decay
heat within the plasma chamber.
3.5.3. Load rejection
The external grid could be lost because of abnormal power distribution calling for the trip of the
connecting circuit breakers in order to protect the grid from damage. Also, there could also be loss of
the grid due to internal causes, as spurious trip of the plant circuit breakers.
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This initiators can be seen as a turbine trip event but for the fact that external power is not available to
plant auxiliaries: power from internal generators could be provided, with the turbine working in reduced
power condition for little time in order to allow for the alignment of internal power generators. It has to
be taken into account that the nuclear fusion can not work at reduced power, or at least not under a given
limit, which is the power needed to keep the plasma burning, so that the turbine can not work at reduced
power but for a limited time trusting on an increase of heat dispersion to the environment through the
plant auxiliaries (a partial bypass to the condenser allowing to partialise the inlet of steam to the turbine
could be an example).
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Table 3.1 - Complete list of accident families (PIEs) for PPCS Model AB (evidenziare i selezionati nella
tabella con il grassetto)
- PIEs ACO
FB1
FB2
FB3
FD1
FD2
FD3
FF1
FF2
FF3
FS1
FS2
FV
HS
HSD
HSF
HSR
HSV
LAI
LAO2
LAO3
LBB1
LBO3
LCO
LDI 2
LDI1
LDO2
LDO3
LFI1
LFI2
LFO3
LRO1
LSI1
LSI2
LSO2
LSO3
LVI1
LVO2
LVO3
M3
M4
M5
M7
TFE1
TFE2
TFE3
TFS1
TFS2
TISS1
TISS2
TISS3
TPI1
TPI2
TPL1
TPL2
TT
VVG
Description
Rupture of VV pump cryogenic circuit out VV
Partial flow blockage (one tube) in a Breeder blanket cooling loop
Loss of flow in a Breeder blanket cooling circuit because of pump seizure
Loss of all Breeder blanket cooling pumps
Partial flow blockage (one tube) in a Divertor cooling loop
Loss of flow in a Divertor cooling circuit because of compressor failure
Loss of all Divertor compressors (sono uno per ogni circuito, se non sbaglio)
Partial flow blockage in a FW-BK cooling loop (e.g. in a small channel)
Loss of flow in a FW-BK cooling loop because of pump seizure
Loss of flow due to FW-BK cooling pump failure (with coastdown)
Partial flow blockage in a LTS cooling loop (e.g. in a channel)
Loss of flow in a LTS cooling loop because of compressor seizure
Total loss of flow in a VV cooling loop
Generalized loss of Heat Sink
Loss of heat sink in DV cooling circuit
Loss of heat sink in cooling circuit of FW and blanket structures
Loss of heat sink in Additional Heating (RF) cooling circuit
Loss of heat sink in VV cooling circuit
Rupture of additional heating (RF) cooling circuit inside VV boundary
Rupture of tubes in the HX of additional heating (RF) cooling circuit, between primary and secondary loops
Rupture of additional heating (RF) cooling circuit outside cryostat
Rupture of Breeder blanket cooling circuit within breeder blanket box
Rupture of Breeder blanket cooling circuit outside cryostat
Rupture of CVCS cooling pipe outside cryostat
Rupture of all divertor coolant loops
Rupture of one divertor coolant tube
Rupture of tubes in the HX of DV loop, between primary and secondary cooling loops
Rupture of divertor cooling circuit outside cryostat
Rupture of one FW cooling channel inside VV
Rupture of more than one FW cooling channel inside VV
Rupture of FW-BK cooling circuit outside cryostat (into primary cooling system room)
Rupture of Pb-17Li purification system inside related room
Rupture of one Low Thermal Shield cooling pipe inside VV
Rupture of one Low Thermal Shield cooling manifold inside VV
Rupture of pipes in the HX of Low Thermal Shield loop, between primary and secondary cooling loops
Rupture of Low Thermal Shield cooling circuit outside cryostat (into primary cooling system room)
Leakage of one VV segm. to plasma chamber
Rupture of a tube in the HX of VV loop, between primary and secondary cooling loops
Rupture of one loop of VV cryostat circuit outside cryostat
Quench
Arc inside the coils
Arc at joints near cryostat walls
Arcs in current leads near cryostat walls
Break in fuel exhaust processing within secondary confinement
Double failure in fuel exhaust processing and its secondary confinement
Membrane breach in front-end permeator
Breach in a fuel storage tank
Double failure of storage tank and secondary confinement
ISS pipe breach within secondary containment
Simultaneous failure of ISS pipe and its secondary containment
Loss of cooling to Cryogenic Distillation column
Pellet Injector pipe break within secondary containment
Simultaneous failure of Pellet Injector process line and secondary containment
Break in tritium process line within secondary confinement
Double failure of tritium process line and secondary confinement
Limited overpower transient
Ingress of gas in the VV
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4. Accidents deterministic analysis by CONSEN 5.1
4.1. Brief description of CONSEN code (version 5.1)
The CONSEN code runs on a PC to simulate the temperature and pressure transients in the
interconnected volumes affected by the accident, taking into account the relative heat and mass
exchanges. The code solves the equations of mass and energy conservation, evaluates the
thermodynamic evolution of the fluids (water, helium, oxygen, nitrogen and non-condensable gases),
including change of phase, and treating thermodynamic conditions also below the triple point of water.
The numerical model is based on the homogeneous equilibrium of all the fluids, improved with
modelling capability of the ice formation on cold structures and of the evaporation phenomenon at the
liquid-gas interface. The model is based on zero-dimensional (lumped parameter) mass and energy
balances. The junction between volumes can be modelled in different ways in order to simulate rupture
panels, simple connections, breaks and relief or control valves. The model can also simulate solid
structures of different material (carbon and stainless steel, concrete, copper, beryllium, tungsten,
insulants, superconductors and user-defined materials), and shape, with internal energy generation and
several boundary conditions. The thermal field is solved using the finite difference implicit method
(Fourier equation in 1-D and in 2-D only for a target) by jet impingement model. The structures can be
completely immersed inside the volume or can act as boundaries between adjacent volumes. Exchange
of energy can be modelled between different fluids, between the structures and the fluids and between
volumes.
The code allows for heat transfer mechanism such as nucleate and film boiling, critical heat flux
evaluation, evaporation at gas-liquid interface, condensation, natural convection, ice formation and
thermal conduction inside the structures. Other heat transfer mechanism and values can be defined by
the user as input data. The CONSEN code can also simulate chemical reactions between Be, W, C and
steam or air, taking into account the influence of pressure and temperature on the reaction rate and the
evolution of the produced gases (Hydrogen, Carbon dioxide, Carbon monoxide). Another feature is the
capability to model the jet impingement heat transfer.
The critical flow model (for high and low
subcooling conditions) is included into the code.
The CONSEN code was used for several
accident analysis in the fusion field in the past [5,
6]. A new version of the CONSEN code was
used for this analysis. The version 5.1 has been
developed to allow the model of several
compartments connected following a free
topology defined by the user, so that all the
compartments can be connected, through
junctions, to all the other volumes. This
overcomes the limitation of the previous versions
of the code in which a maximum of six volumes
could be connected only to a "central"
compartment.
In this version user also imposes volumetric or
mass flow rates depending on the pressure
difference between volumes, to simulate
typically the leakages from compartments.
Fig. 4.1 – Scheme of a control volume in CONSEN
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A simple model to evaluate tritium migration between the compartments has been implemented too in
CONSEN 5.1. The tritium flow between two connected volumes is evaluated multiplying the tritium
volumetric concentration in the compartment (assumed uniform in all the fluid phases) and the
volumetric flow rate evaluated by the existing models in CONSEN. In this new model, actually under
testing, the tritium capture in the solid phase if ice is formed in a volume is considered only for the
volumetric fraction in the water. Also filtering capability of tritium can be simulated in the junctions
between volumes.
4.2. Accident sequences analyzed
Six different accident sequences have been analyzed using the CONSEN 5.1 computer code.
They have been defined by FFMEA in the previous section.
In the following table the accidents are summarized. An identification number for each calculation has
been assigned.
Accident
LOFA + in-vessel LOCA
Case ID
C1, C2
Generalised loss of heat
sink
C3
Ex-vessel LOCA
C4
Interface LOCA between
FW and Breeding Blanket
C5
Steam generator tube
rupture
C6
Description
In-VV LOCA due to a break of 5 (C1) or 10 (C2) FW
cooling channels when FW temperature reaches 1073 K
after the LOFA
As in the previous cases but affecting all the 9 loops of the
cooling helium. The rupture of 5 channels has been
considered.
Double guillotine break of a main piping inside the TCHS
vault (58,000 m3) and rupture disk intervention at 0.14
MPa towards the other vaults (59,600 m3). The aim is to
define the rupture disc area between TCHS and the
expansion vaults and to verify the volumes available to
limit the pressure ≤ 0.16 MPa in the expansion volume
itself
In-VV LOCA. Double guillotine break of an helium
manifold of a module (i = 220 mm). The aim is to define
the rupture disc area between VV and the expansion
volume to limit the pressure inside VV ≤ 0.2 MPa
Preliminarily analysis of a steam generator tube rupture (10
tubes affected,i = 20 mm) to verify the pressurization of
one helium loop.
Table 4-1 – Accident sequences analyzed
4.3. Cases C1 and C2
4.3.1 Accident sequences description for the cases C1 and C2
The first postulated accident is a pump trip in one of the FW/BL HTS leading to a loss of flow in one of
the cooling loops, without pump coast-down. The Fast Plasma Shutdown System (FPSS) does not
intervene. The Plasma Facing components increase their surface temperatures until 1,073 K are reached
and a break in the in-vessel FW/BL cooling channels happens, with a reference cross section
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corresponding to 5 (Case C1) or 10 (Case C2) cooling channels, each with average dimensions 0.014 x
0.0156 m (Fig. 4.2).
All the control systems (relief valves and pressurizer) in the cooling loops will be excluded at the
beginning of the accident. A plasma disruption occurs, having an energy deposition of 2.5 GJ hitting an
area of 1/3 of the FW zone for 1.0 s. The FW/BL surface of the failed loop is first cooled down with
residual coolant present before the complete drainage.
When the VV total pressure reaches 0.11 MPa a rupture disk (assumed of 0.2 m2) opens toward the
TCHS vault, having a free volume of 58000 m3 (initial conditions of the air internal atmosphere 30.0 °C
and 0.1 MPa). A rupture disk connecting the TCHS vault to the other vaults (volume available for
expansion 59600 m3) intervenes at the set point pressure of 0.14 MPa, with a flow area of 2 m2 (mi
sembra che la sezione di 2m2 si sia dimostrata sufficiente e quindi il caso di 5 m2 non sia stato
considerato. Corretto?), to limit the pressure inside the containment to values lower than 0.16 MPa. (see
Table A.1.16).
25
14
4
d
p
21.6
w
15.6
Fig. 4.2 – Dimensions of the First Wall cooling channels
Leakages from the VV and TCHS (and for other expansion volumes considered) building towards the
external environment has been initially considered with a daily leak rate of 5% and 10% respectively.
VV
Design pressure
PD (MPa)
Leak rate
(% volume/day)
0.2
5%
(at design pressure)
0.16
10 %
(at design pressure)
Scale rules
leakage [m3/s]
Scales with square root of differential pressure
leakage
0.05  Volume P  P0
24  3600
PD  P0
Scales with square root of differential pressure
TCHS
P current pressure
leakage
0.1  Volume P  P0
24  3600
PD  P0
P0 atmospheric pressure
Table 4-2: Leakages laws for VV and TCHS vault.
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In table 4-3 the tritium inventories assumed in the calculations are reported.
Source terms
Tritium in VV
Tritium in coolant
Model B
1 kg
1.0 g (per loop)
Notes
Initially present in VV
Released into VV
Table 4-3: Tritium inventories.
A transient of 24 hours has to be analysed, taking into account the initial nuclear heating and the decay
heat in the in-vessel structures.
The final goal is to demonstrate that the overpressure in the VV is safely mitigated under the design
pressure of 0.2 MPa and the secondary containment (THCS vault) overpressure is also safely mitigated
under the design pressure of 0.16 MPa.
4.3.2 Plant nodalization for the sequences C1 and C2
A PPCS Model AB nodalisation has been developed for the CONSEN calculations. In Fig. 4.3 the
scheme of the nodalization for this accident sequence is presented.
The model is composed by :

6 control volumes (1 -> Blanket PHTS, 2 -> VV, 3 -> TCHS, 4 -> DS1, 5 -> VV external zone, 6
-> TCHS external zone)

1 fixed condition volume ((7-F) -> DS2)

6 structures (blanket, 3 zones of first wall, divertor, concrete walls)

6 junctions [2 normal junctions (J-1, J-2), 2 imposed mass flow rate junctions (J-4, J-5), 2 pdependent volumetric flow rate junction (J-3, J-6)

(modificare la figura mettendo al posto delle due labels Detrit. Sys un DS1 ed un DS2)
PPCS MODEL AB
LOFA + IN-VESSEL LOCA
6
ATMOSPHERE
CONSEN 5.1 MODEL
J-6
J-4
4
2
J-2
VV
TCHS
3
J-3
Detrit. Sys
5
DF=99.9%
J-5
Detrit. Sys
J-1
7 -F
5
ATMOSPHERE
R
1
1
Blanket
PHTS
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Fig. 4.3 – PPCS model AB nodalization for the LOFA + in vessel LOCA accident simulation
The presence of the Detritiation System (DS1) (once through) for the TCHS vault atmosphere was
considered with a retention efficiency equal to 99.9% for tritium. This DS1 connects the TCHS
atmosphere with the external environment (Volume 4) through the junction J-4, with a constant mass
flow-rate (3.0 kg/s). A further back-environment control node (DS2), simulating the external clean
atmosphere (Vol. 7-F), has been added into the PPCS nodalisation in order to simulate the 3.0 kg/s clean
air ingress (from the junction J-5) into the TCHS vault atmosphere due to the presence of the DS2.
No dust simulation has been performed in the calculations, due to the lack of a specific model in
CONSEN. A simple treatment of dust migration could be also performed, but leading to conservative
results, considering a perfect mixing of dust in the volume atmosphere without retention or deposition
processes, as gravitational settling, thermophoresis, diffusional deposition, centrifugal or turbulent
deposition. This evaluation has been performed only in case C1, to provide some preliminary data. In
this case an inventory of 10 kg of dust (100% resuspended) has been considered into the VV and a
scrubber with a filtration efficiency (for dusts only) of 0.9 has been included in the CONSEN model at
the junction J-2 (linking VV to the TCHS vault).
In the following, all the main input data describing the nodalization are reported. (centrare tutte le
tabelle e numerarle per distinguerle e citarle)
VOLUMES and STRUCTURES
Tab. 4.3.2-1 VOLUME 1: Blanket PHTS broken single loop
Fluid:
Volume:
Mass:
Tritium content:
Temperature:
Pressure:
Structures:
Connected to:
helium
358.30 m3
2022.46 kg
1g
673.15 K
8 MPa
1/9 blanket
Volume 2 (Vacuum Vessel)
Tab. 4.3.2-2 - STRUCTURES in Volume 1
Structure 1.1: 1/9 Blanket
HT Surface:
1161 m2
Initial temperature:
680 K
Shape:
plate, 0.138 m equivalent thickness
Cooling:
He in Volume 1:
HT coefficient during
LOFA:
Time (s)
h (W/m2K)
0.
35000.
1.
35000.
16.
20000.
31.
5000.
1000.
5.
100000000.
0.
Mass:
1042890 kg
Composition:
100% CS (che cosa è CS? Metterlo negli
acronimi)
Volume:
160 m3
Internal power:
1.46493E+06 W/m3 nominal
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(decay heat)
Time (s)
0.
1.
10.
20.
30.
60.
600.
1800.
3600.
21600.
43200.
64800.
86400.
q"' (W/m3)
4.67E+04
2.49E+04
8.68E+03
8.51E+03
8.42E+03
8.31E+03
7.53E+03
6.95E+03
6.34E+03
3.51E+03
2.63E+03
2.34E+03
2.18E+03
VOLUME 2: VV (Vacuum Vessel)
Fluid:
air
Volume:
5596 m3
Tritium content:
1000 g
Temperature:
423 K
Pressure:
4 10-5 Pa
Structures:
1/9 FW, 1/3 FW, 5/9 FW, Divertor
Connected to:
Volume 1 (PHTS), Volume 3 (Expansion
Volume), Volume 5 (atmosphere)
STRUCTURES in Volume 2
Structure 2.1 + 2.2 + 2.3 : (1/9 + 1/3 + 5/9 FW)
Surface exposed to plasma:
1667 m2
HT Surface:
3978 m2 (442 + 1326 +2210 m2)
Initial temperature:
707-641 K (significato delle due temperature?)
Shape:
plate, 0.007 m equivalent thickness
Cooling:
He in Volume 1:
HT coeff. (1/3 FW):
Time (s)
h (W/m2K)
0.
35000.
1.
35000.
16.
20000.
31.
5000.
1000.
5.
100000000.
0.
First layer:
W
Mass:
54500 kg (6055.56 + 18166.66+30277.78 kg)
Composition:
100% W
Volume:
2.83 m3 (0.3144 + 0.94333+1.57222 m3)
Internal power:
75.854E+06 W/m3 nominal
plasma disruption
2.5 GJ in 1 second only on 1/3 of FW
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(decay heat)
time
0.1
1.
10.
20.
30.
60.
600.
1800.
3600.
21600.
43200.
64800.
86400.
Second layer:
Mass:
Composition:
Volume:
Internal power:
(decay heat)
time
0.1
1.
10.
20.
30.
60.
600.
1800.
3600.
21600.
43200.
64800.
86400.
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q"' (W/m3)
1174683.11
1158460.247
1077672.403
1048481.484
1038304.064
1026493.463
985897.2792
977433.6749
968177.4558
891081.1661
815178.6219
752392.5795
699440.
Heat sink
Structure 2.4: Divertor
Mass:
Composition:
Volume:
Shape:
Surface:
Cooling:
HT coefficient :
Initial temperature:
Internal power:
197000 kg (21888,89 + 65666.67 + 109444.444 kg)
100% CS
25 m3 ( 2.777 + 8.333 + 13.889 m3)
75.854E+06 W/m3 nominal
q"' (W/m3)
337385.6073
337121.5315
335838.8655
334533.1875
333487.8533
330390.59
302581.4519
276611.0307
246443.0911
91061.13728
46851.9709
37760.57649
35698.25847
168000 kg
100% W
8.75 m3
plate, 0.02 m thick
437.5 m2
He in Volume 1:
35000 W/m2K
1165-935 K
29771428. W/m3 nominal
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(decay heat)
Time (s)
0.1
1.
10.
20.
30.
60.
600.
1800.
3600.
21600.
43200.
64800.
86400.
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q"' (W/m3)
946661.8498
935288.8648
875658.1685
854784.3604
847828.5126
840602.6752
813535.0863
805474.5398
796194.3523
717434.2632
639112.4614
574816.0741
520084.0059
VOLUME 3: TCHS - Expansion volume
Fluid:
air
Volume:
58000 m3
Temperature:
300 K
Pressure:
101320 Pa
Structures:
Walls
Connected to:
Volume 2 (VV), Volume 4 (Detritiation system - receiving), ,
Fixed condition volume 7 ( Detritiation system – clean),
Volume 6 (atmosphere)
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STRUCTURES in Volume 3
Structure 3.1: walls
Mass:
Composition:
Volume:
Shape:
Surface:
Initial temperature:
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11324544 g
100% CLS
5055.6
plate, 0.4 m thick
12639 m2
300 K
VOLUME 4: DS1 (Detritiation system – receiving volume)
Fluid:
air
Volume:
98000000 m3
Temperature:
300 K
Pressure:
101320 Pa
Structures:
NO
Connected to:
Volume 3 (TCHS)
VOLUME 5: VV external zone (atmosphere)
Fluid:
air
Volume:
98000000 m3
Temperature:
300 K
Pressure:
103500 Pa
Structures:
NO
Connected to:
Volume 2 (VV)
VOLUME 6: TCHS external zone (atmosphere)
Fluid:
air
Volume:
98000000 m3
Temperature:
300 K
Pressure:
103500 Pa
Structures:
NO
Connected to:
Volume 3 (TCHS)
JUNCTIONS BETWEEN VOLUMES
J-1: Connection between volumes 1 and 2 (PHTS - VV)
Area:
1.092E-3 m2 (2.184E-3 for C2)
Time:
Open when FW temperature is 1073 K
Check:
bidirectional, permanent
DP coeff.:
1.5
Flow:
calculated
0
P open:
J-2: Connection between volumes 2 and 3(VV - TCHS)
Area:
0.2 m2
Check:
bidirectional, permanent
Filter
efficiency = 0.9 only for dusts (Case C1)
DP coeff.:
1.5
Flow:
calculated
10000 Pa
P open:
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J-3: Connection between volumes 2 - 5 (VV – External Atmosphere)
Area:
N.A.
Time:
0
Check:
bidirectional, permanent
Flow:
table (5% vol/day)
Flow (m3/s)
p (Pa)
0.00E+00
0.
3.26E-05
10.
6.10E-05
35.
7.29E-05
50.
8.93E-05
75.
1.03E-04
100.
1.93E-04
350.
2.31E-04
500.
2.82E-04
750.
3.26E-04
1000.
4.61E-04
2000.
5.65E-04
3000.
6.52E-04
4000.
7.29E-04
5000.
8.31E-04
6500.
9.22E-04
8000.
1.03E-03
10000.
1.46E-03
20000.
1.79E-03
30000.
2.06E-03
40000.
2.31E-03
50000.
2.63E-03
65000.
2.92E-03
80000.
3.26E-03
100000.
3.42E-03
110000.
3.57E-03
120000.
3.72E-03
130000.
3.99E-03
150000.
4.31E-03
175000.
4.61E-03
200000.
5.15E-03
250000.
5.65E-03
300000.
J-4: Connection between volumes 3 - 4 (TCHS – DS1)
Time:
0
Check:
unidirectional from TCHS to DS1,
permanent
Filter
efficiency = 0.999 for dusts and tritium
Flow:
constant 3 kg/s
J-5: Connection between volumes 3 – (7-F) (TCHS – DS2)
Time:
0
Check:
unidirectional from DS2 to TCHS,
permanent
Flow:
constant 3 kg/s
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J-6: Connection between volumes 3 - 6 (TCHS – External Atmosphere)
Area:
N.A.
Time:
0
Check:
bidirectional, permanent
Flow:
table (10% vol/day)
Time (s)
Flow (m3/s)
0.
0.00E+00
10.
8.76E-04
35.
1.64E-03
50.
1.96E-03
75.
2.40E-03
100.
2.77E-03
350.
5.18E-03
500.
6.20E-03
750.
7.59E-03
1000.
8.76E-03
2000.
1.24E-02
3000.
1.52E-02
4000.
1.75E-02
5000.
1.96E-02
6500.
2.23E-02
8000.
2.48E-02
10000.
2.77E-02
20000.
3.92E-02
30000.
4.80E-02
40000.
5.54E-02
50000.
6.20E-02
65000.
7.07E-02
80000.
7.84E-02
100000.
8.76E-02
110000.
9.19E-02
120000.
9.60E-02
130000.
9.99E-02
150000.
1.07E-01
175000.
1.16E-01
200000.
1.24E-01
250000.
1.39E-01
300000.
1.52E-01
4.3.3 Main Thermal-hydraulics Results for sequences C1 and C2
The main results in the whole PPCS system (HTS, VV and TCHS) are shown from Fig. 4.4 through Fig.
4.12 for the reference case C1 and from Fig. 4.13 through Fig. 4.19 for the sensitivity case C2, where
the four characteristic phases of the accident are quite evident:
1. 1/9 FW heating before the break. The FW reaches 1073 K on the cooling helium surface at the
time 524.9 s after the start of the LOFA. (Fig. 4.8 and Fig. 4.17)
2. Helium pressurisation of the VV only, before the rupture disk openings at 0.11 MPa, obviously
dependent by the values of the VV free volume and the size of the break (Helium flow rate, see
Fig. 4.7 for the case C1 and 4.16 for the case C2).
3. Discharge into the TCHS vault, trough the RD, until pressures equilibrium between VV
and TCHS has reached. This value of the equilibrium pressure mainly depends by the system
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total free volume, but the helium flow-rate (break size) has some influence (comparing the
results in the C1 and C2 cases, see Fig. 4.20). (ricontrollare)
4. Long term phase, where the whole system has the same pressure but different temperatures,
with this pressure descending towards the atmospheric value, for the combined effect of the daily
leakages and of the heat transfer towards the TCHS structure.
In both cases C1 and C2 the heating phase of helium inside the cooling loop has the same evolution.
Helium reaches a temperature of 856 K and a pressure of 10.17 MPa at the time of the break ( 524.9 s).
At this time the plasma disruption (see the temperature peak of the FW affected in Fig. 4.8) and shut
down occurs (subsequent temperature decrease in the structures).
In Case C1 (5 FW channels break), rupture disk opens at 574 s (49 s after the break) and the maximum
pressure reached is 0.134 MPa at 1094 s (569 s after the break), and the TCHS vault pressure is the
same from this point (fig. 4.4 and 4.5).
The maximum helium flow rate from the break is 6 kg/s and pressure equilibrium practically occurs at
1850 s after the break. The maximum flow rate between the VV and the TCHS vault is 3.8 kg/s (Fig.
4.7)
Also for the VV and TCHS atmosphere temperature trends predicted by CONSEN (Fig. 4.6) three
different phenomenological phases are present:
1) the initial compression effect inside the VV atmosphere due to the helium blow-down, until the
RD opening, is shown;
2) the subsequent VV cooling due to expansion phase of the compressed helium into the TCHS
vault atmosphere, leading to at the TCHS vault temperature increase;
3) The long term thermal re-equilibrium with a decrease towards the final temperature levels,
deriving from a balance of the atmosphere internal energy, thermal capacities/losses, mass
exchanges and the Detritiation System cooling action.
In Case C2 (10 FW channels break), rupture disk opens at 547 s (22 s after the break) and the maximum
pressure reached is 0.141 MPa at 872 s (347 s after the break), and the TCHS vault pressure is the same
from this point figs. 4.13 and 4.14).
The maximum helium flow rate from the break is 12 kg/s and pressure equilibrium practically occurs at
880 s after the break. The maximum flow rate between the VV and the TCHS vault is 6.55 kg/s. (Fig.
4.16).
CASE
Max press. in the VV
[MPa]
Time from start of LOFA
[s]
Time from break
[s]
C1
0.134
574
49
C2
0.141
547
22
Table 4-4 – Peak pressure and time in cases C1 and C2
In table 4-5 the tritium inventories after 1 day from the start of the accident, in the different volumes, are
reported. The residual tritium content in the VV is 19 g in the case C1 and 21 g in the case C2; the
amount in the TCHS is negligible. The leakage of tritium from VV is 0.13-0.14 g respectively and from
the TCHS is about 2 g in both the cases. About 1 g of tritium is released by the detritiation system after
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filtration. Tritium inventories and release histories in the different volumes are reported in Figs. 4.9 and
4.10 for the case C1 and Figs. 4.18 and 4.19 for the case C2.
CASE 1
CASE 2
Tritium in
the VV
[g]
Tritium in
the TCHS
[g]
Tritium in
atmosph
from VV
[g]
Tritium in
atmosph from
TCHS
[g]
18.7
21.2
2.96E-09
2.88E-09
0.133
0.142
1.92
1.96
Tritium in
Total Tritium
atmosph from in atmosph
[g]
DS1
[g]
0.980
0.978
3.03
3.08
Tritium
captured by
the DS1
[g]
979.3
976.7
Table 4-5
Cases C1 and C2 - Tritium inventories in the volumes after 1 day from the start of the accident.
A simplified and preliminary evaluation of dust migration has been performed only in case C1. Dust
inventory and release after 1 day from the accident are reported in Table 4-6 and in Figs 4.11 –4.12. It
could be noted that the amount of dust in the TCHS vault and DS1 are the same as Tritium in Tab. 4-5
and that the amount related to VV are about ten times higher than the values for tritium. This is due to
the initial amount of dust that is ten times the initial tritium inventory in the VV. The filtration factor of
0.9 of the scrubber between the VV and the TCHS vault reduces the amount of dust transferred to the
TCHS vault by a factor of 10, reproducing the same values as tritium. Due to the same simplified model
for migration of tritium and dust in CONSEN, the same evolution will be calculated from the TCHS
vault release and DS1 filtration.
CASE 1
Dust in
the VV
Dust in the
TCHS
Dust in
atmosph
from VV
Dust in
atmosph
from TCHS
Dust in
atmosph
from DS
Total dust
in atmosph
Dust
captured
by the DS
[g]
[g]
[g]
[g]
[g]
[g]
[g]
Dust
captured
by the
scrubber
[g]
184.7
2.96E-09
1.32
1.92
0.980
4.22
979.3
8831.8
Table 4-6 – Case C1 - Dust inventory in the volumes after 1 day from the start of the accident.
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Base Case c1
1.60E+05
1.40E+05
Pressure [Pa]
1.20E+05
1.00E+05
8.00E+04
BL-PHTS
VV
6.00E+04
TCHS
4.00E+04
2.00E+04
0.00E+00
100
1000
10000
100000
Time [s]
Fig. 4.4 – Case C1 – Pressure transient: Long term phase
Base Case c1
1.00E+07
BL-PHTS
Pressure [Pa]
VV
TCHS
1.00E+06
1.00E+05
1.00E+04
0
600
1200
1800
Time [s]
Fig. 4.5 – Case C1 – Pressure transient: Short term phase
2400
3000
3600
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Base Case c1
1600
BL-PHTS
Temperature [K]
1400
VV
TCHS
1200
1000
800
600
400
200
0
1
10
100
1000
10000
100000
Time [s]
Fig. 4.6 – Case C1 – Temperature transient
Base Case c1
7
Flow rates [kg/s]
6
5
4
HTS-VV
3
VV-TCHS
2
1
0
100
1000
Time [s]
Fig. 4.7 – Case C1 – Flow rates between volumes
10000
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Base Case c1
1400
1/9 FW
Wall Temperature [K]
1200
DIV
3/9 FW
1000
800
600
400
200
0
1
10
100
1000
10000
100000
Time [s]
Fig. 4.8 – Case C1 – Main structures temperature evolution (meglio scala delle temperature da 600 a
1200)
Base Case c1
1.2E+03
VV
TCHS
Tritium [g]
1.0E+03
8.0E+02
6.0E+02
4.0E+02
2.0E+02
0.0E+00
100
1000
10000
Time [s]
100000
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Fig. 4.9 – Case C1 – Tritium inventory in the PPCS volumes
Base Case c1
VV-ATM
EV-DS-ATM
TCHS-ATM
2.5E+00
Tritium [g]
2.0E+00
1.5E+00
1.0E+00
5.0E-01
0.0E+00
100
1000
10000
100000
Time [s]
Fig. 4.10 – Case C1 – Tritium released through different paths
Base Case c1
1.2E+04
VV
TCHS
1.0E+04
Dust [g]
8.0E+03
6.0E+03
4.0E+03
2.0E+03
0.0E+00
100
1000
10000
Time [s]
100000
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Fig. 4.11 – Case C1 – Dust inventory in the volumes
Base Case c1
VV-ATM
EV-DS-ATM
TCHS-ATM
2.5E+00
Dust [g]
2.0E+00
1.5E+00
1.0E+00
5.0E-01
0.0E+00
100
1000
10000
100000
Time [s]
Fig. 4.12 – Case C1 – Dust released through different paths
Case c2
1.60E+05
1.40E+05
Pressure [Pa]
1.20E+05
1.00E+05
8.00E+04
BL-PHTS
VV
6.00E+04
TCHS
4.00E+04
2.00E+04
0.00E+00
100
1000
10000
Time [s]
100000
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Fig. 4.13 – Case C2 – Pressure transient: Long term phase
Case c2
1.00E+07
BL-PHTS
Pressure [Pa]
VV
TCHS
1.00E+06
1.00E+05
1.00E+04
0
600
1200
1800
2400
3000
3600
Time [s]
Fig. 4.14 – Case C2 – Pressure transient: Short term phase
Case c2
1600
BL-PHTS
Temperature [K]
1400
VV
TCHS
1200
1000
800
600
400
200
0
1
10
100
1000
Time [s]
10000
100000
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Fig. 4.15 – Case C2 – Temperature transient: Long term phase
Case c2
14
Flow rates [kg/s]
12
10
8
HTS-VV
6
VV-EV
4
2
0
100
1000
10000
100000
Time [s]
Fig. 4.16 – Case C2 – Helium flow rates between volumes
Case c2
1400
1/9 FW
Wall Temperature [K]
1200
DIV
3/9 FW
1000
800
600
400
200
0
1
10
100
1000
Time [s]
10000
100000
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Fig. 4.17 – Case C2 – Main structures temperature evolution Scala y da 600 a 1200 per leggere meglio)
Case c2
1.2E+03
VV
TCHS
Tritium [g]
1.0E+03
8.0E+02
6.0E+02
4.0E+02
2.0E+02
0.0E+00
100
1000
10000
100000
Time [s]
Fig. 4.18 – Case C2 – Tritium inventory in the PPCS volumes
Case c2
VV-ATM
TCHS-DS-ATM
TCHS-ATM
2.5E+00
Tritium [g]
2.0E+00
1.5E+00
1.0E+00
5.0E-01
0.0E+00
100
1000
10000
Time [s]
100000
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Fig. 4.19 – Case C2 – Tritium discharged through different paths
C1-C2 comparison
1.50E+05
VV-C1
TCHS-C1
VV-C2
TCHS-C2
Pressure [Pa]
1.40E+05
1.30E+05
1.20E+05
1.10E+05
1.00E+05
9.00E+04
0
600
1200
1800
2400
3000
3600
Time [s]
Fig. 4.20 – Comparison of VV and TCHS pressures in Cases C1 and C2
4.4. Case C3
4.4.1 Accident sequence description for the case C3
In the case C3 a generalised loss of heat sink has been postulated. As in cases C1 and C2, the break of
the BL/FW cooling loop occurs when the temperature in the FW cooled surface reaches 1073 K, but
affecting now all the 9 loops of the cooling helium. The rupture of 5 channels for each loop has been
considered (total break area 9.828 E-03 m2).
4.4.2 Plant nodalization for the sequence C3
The plant nodalization used is the same of the cases C1 and C2 (è corretto questo? Altrimenti va
specificato)
4.4.3 Main Thermal-hydraulics Results for sequence C3
A first calculation considering a volume available for expansion of 58,000 m3 and a vent area between
VV and TCHS of 0.2 m2 (design values) were preliminary considered. In this case (Fig. 4.21) a peak of
0.41 MPa in the VV and 0.391 MPa in the TCHS vault, at 802 s and 994 s respectively after the start of
the LOFA (278 s and 470 s after the break) were reached. To limit the pressure peaks to the maximum
allowable values (0.2 MPa for VV and 0.16 MPa for the expansion volumes), different calculations
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were performed and the following values were obtained: an expansion volume of 500,000 m3 and a vent
area between VV and EV of 0.5 m2 have been imposed. The main results are reported in Figs. 4.21
through 4.25.
The heating phase after the LOFA looks like the cases C1 and C2. The peak in the VV is 0.205 MPa,
still exceeding the design pressure. A value lower than 0.2 MPa could be reached with a larger volume
of expansion.
In table 4-7 the tritium inventories after 1 day from the start of the accident, in the different volumes, are
reported. As shown in Figs. 4.24 and 4.25, tritium is transferred completely in the EV after the break, so
that the main release to the atmosphere is from that volume. (perché non ne rimane niente nel VV?)
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CASE C3
Tritium in
the VV
Tritium in
the TCHS
[g]
[g]
0.1948
48.59
Tritium in
atm. from
VV
[g]
Tritium in
atm. from
TCHS
[g]
0.0096
16.77
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Total
Tritium in
atmosph
from DS
[g]
0.9433
PAGE
42 of 87
Total
Tritium in
atmosph
[g]
17.7
Tritium
captured
by the DS
[g]
942.23
Table 4-7 – Tritium inventories in the volumes after 1 day from the start of the accident (generalised
loss of heat sink)
Case C3
5.0E+05
VV (DESIGN)
Pressure [Pa]
4.5E+05
EV (DESIGN)
4.0E+05
VV
3.5E+05
EV
3.0E+05
2.5E+05
2.0E+05
1.5E+05
1.0E+05
0
600
1200
1800
2400
3000
3600
Time [s]
Fig. 4.21 – Case C3: Comparison of the pressure transient between the design values (58,000 m3, RD
area VV-EV 0.2 m2) and the following values (EV 500,000 m3, RD area VV-EV 0.5 m2)
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Case c3
2.50E+05
BL-PHTS
Pressure [Pa]
2.00E+05
VV
EV
1.50E+05
1.00E+05
5.00E+04
0.00E+00
100
1000
10000
100000
Time [s]
Fig. 4.22 – Case C3: Pressure transient (500,000 m3 EV)
Case c3
60
Flow rates [kg/s]
50
40
30
HTS-VV
VV-EV
20
10
0
100
1000
10000
Time [s]
Fig. 4.23 – Case C3: Helium flow rates (500,000 m3 EV)
100000
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Case c3
1.2E+03
VV
EV
Tritium [g]
1.0E+03
8.0E+02
6.0E+02
4.0E+02
2.0E+02
0.0E+00
100
10000
1000
100000
Time [s]
Fig. 4.24 – Case C3: Tritium inventories in VV and EV (500,000 m3 EV)
VV-ATM
Case c3
EV-DS-ATM
EV-ATM
1.8E+01
1.6E+01
Tritium [g]
1.4E+01
1.2E+01
1.0E+01
8.0E+00
6.0E+00
4.0E+00
2.0E+00
0.0E+00
100
1000
10000
100000
Time [s]
Fig. 4.25 – Case C3: Tritium released to the external atmosphere (500,000 m3 EV)
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4.5. Case C4
4.5.1 Accident sequences description for the case C4
The case C4 concerns a double guillotine break of a main piping inside the TCHS vault (58,000 m 3) and
rupture disk intervention at 0.14 MPa towards the other vaults (59,600 m3). The aim was to define the
rupture disc area between TCHS and the expansion vaults and to verify the volumes available to limit
the pressure ≤ 0.16 MPa. The break involve a main collector of 1.25 m ID, with a break area of 2 x
1.22718 m2 = 2.45437 m2. The break occurs at time 0 and the Fast Plasma Shutdown System does
intervene in 3 seconds. A high energy deposition on the PFC’s walls occurs due to the disruption: 2.5
MJ/m2 hitting an area of 1/3 of FW zone for 1 second. After this time, only decay heat in the structures
is considered. Helium from the pipe is discharged directly into the TCHS vault, and the rupture disk
with the other vaults intervenes at 0.14 MPa.
4.5.2 Plant nodalization for the sequence C4
In this case the VV is not involved and the nodalization is shown in Fig. 4.26.
4.5.3 Main Thermal-hydraulics Results for sequence C4
A rupture disk area of 2 m2 has been verified between the TCHS vault and the EV and this values
appears adequate to limit the pressure in the vaults (pressure peak 0.155 MPa after 1 s in the TCHS
vault). Due to the very quick pressurization of the TCHS vault, the rupture disk opens immediately after
the break, and an area of 2 m2 at least is needed to limit the pressure increase.
Max. pressure in TCHS
vault
[MPa]
Final pressure in TCHS
vault
[MPa]
Max pressure in the EV
[MPA]
Final pressure in the
EV
[MPa}
0.155
0.1067
0.128
0.1067
CASE C4
Table 4-8 – Pressure values in the expansion volumes (peak and final values after 1 day)
In Figs. 4.27 – 4.31 the main results are reported. Tritium releases is due to the small amount contained
into the affected loop only. Figs 4.30 and 4.31 show the amount of tritium during the accident.
PPCS MODEL AB
EX-VESSEL LOCA
5
CONSEN 5.1 MODEL
ATMOSPHERE
J-4
Detrit. Sys
J-3
3
4
J-2
EV
J-5
J-1
TCHS
DF=99.9
%
1
J-6
6
5
Detrit. Sys
7 -F
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Fig. 4.26 – PPCS model AB nodalization for the ex-vessel LOCA accident simulation (nello schema del
modello fare le stessse modifiche che vengono richieste nella figura dei casi C1 e C2, mettendo DS1 e
Ds2 al posto di Detrit. Sys.)
Case c4
2.40E+05
BL-PHTS
Pressure [Pa]
2.20E+05
TCHS
EV
2.00E+05
1.80E+05
1.60E+05
1.40E+05
1.20E+05
1.00E+05
0.1
1
10
100
Time [s]
Fig. 4.27 – Case C4 – Pressure transient: Long term phase
1000
10000
100000
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Case c4
2500
BL-PHTS
TCHS
EV
Temperature [K]
2000
1500
1000
500
0
1
10
100
1000
10000
100000
Time [s]
Fig. 4.28 – Case C4 – Temperature transient: Long term phase (l’aumento di temperatura nel BL-PHTS
non mi sembra giustificato da niente. Il plasma shut down ha spento dopo 3 s)
Case c4
Flow rates [kg/s]
800
700
HTS-TCHS
600
TCHS- EV
500
400
300
200
100
0
0.1
1
10
Time [s]
Fig. 4.29 – Case C4 – Helium flow rates between volumes
100
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Case c4
1.2E+00
TCHS
EV
1.0E+00
Tritium [g]
BL-PHTS
8.0E-01
6.0E-01
4.0E-01
2.0E-01
0.0E+00
1
10
100
1000
10000
100000
Time [s]
Fig. 4.30 – Case C4 – Tritium inventory in the PPCS volumes
TCHS-ATM
Case c4
TCHS-DS-ATM
2nd EV-ATM
6.0E-03
Tritium [g]
5.0E-03
4.0E-03
3.0E-03
2.0E-03
1.0E-03
0.0E+00
1
10
100
1000
Time [s]
Fig. 4.31 – Case C4 – Tritium releases
10000
100000
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4.6. Case C5
4.6.1 Accident sequences description for the case C5
In case C5 another in-vessel LOCA has been analyzed. It involves a double guillotine break of the
helium manifold of a module (i = 220 mm). The aim was to define the rupture disk area between VV
and the TCHS vault to limit the pressure inside VV ≤ 0.2 MPa.
4.6.2 Plant nodalization for the sequence C4
The CONSEN model is similar to the cases C1, C2 and C3, with a larger break area: 7.6 E-02 m2.
4.6.3 Main Thermal-hydraulics Results for sequence C4
The break occurs at time 0, and a preliminary calculation with an area of 0.2 m2 shows that in this case a
pressure peak of 0.474 MPa should occur in the VV after 8.5 s from the break. The TCHS pressure
would be leant towards 0.146 MPa after 60 s, in equlibrium conditions.
To limit the VV pressure, a rupture disk area of 1.8 m2 is needed. The peak reaches 0.197 MPa after 3
s in the VV. The equilibrium pressure in the TCHS valt (Fig. 4.32) reached at 25 s is of 0.148 Mpa; the
increment, with respect to the previous value of 0.146 Mpa, is limited..
In table 4-9 the tritium inventories after 1 day from the start of the accident, in the different volumes, are
reported. As shown in Figs. 4.36 and 4.37, an amount of 56 g of tritium remains into the VV, but the
main release to the atmosphere is from the TCHS vault (1.85 g from leakages and about 1 g through the
DS).
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Tritium in
the VV
Tritium in
the TCHS
[g]
[g]
Tritium in
atm. from
VV
[g]
Tritium in
atm. from
TCHS
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Total Tritium
in atmosph
from DS
[g]
Total Tritium
in atmosph
[g]
Tritium
captured by
the DS
[g]
0.9426
3.19
941.6
[g]
CASE C5
56.13
0.02729
0.3966
1.846
Table 4-9 – Tritium inventories in the volumes after 1 day from the start of the accident.
Case C5
5.0E+05
VV (DESIGN)
Pressure [Pa]
4.5E+05
TCHS (DESIGN)
4.0E+05
VV
3.5E+05
TCHS
3.0E+05
2.5E+05
2.0E+05
1.5E+05
1.0E+05
0
10
20
30
40
Time [s]
Fig. 4.32 – Case C5 – Pressure transient: Comparison with design values
50
60
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Case c5
2.50E+05
BL-PHTS
Pressure [Pa]
2.00E+05
VV
TCHS
1.50E+05
1.00E+05
5.00E+04
0.00E+00
1
10
100
1000
10000
100000
Time [s]
Fig. 4.33 – Case C5 – Pressure transient: Long term phase
Case c5
1200
BL-PHTS
VV
TCHS
Temperature [K]
1000
800
600
400
200
0
1
10
100
1000
10000
100000
Time [s]
Fig. 4.34 – Case C5 – Temperature transient: Long term phase (la derivata della curva del BL-PHTS è
sospetta)
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Case c5
350
Flow rates [kg/s]
300
250
200
HTS-VV
150
VV-TCHS
100
50
0
1
10
100
Time [s]
Fig. 4.35 – Case C5 – Helium flow rates between volumes
Case c5
1.2E+03
VV
TCHS
Tritium [g]
1.0E+03
8.0E+02
6.0E+02
4.0E+02
2.0E+02
0.0E+00
1
10
100
1000
Time [s]
10000
100000
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Fig. 4.36 – Case C5 – Tritium inventory in the PPCS volumes (con la scala logaritmica sulle ordinate si
vedono meglio i risultati)
VV-ATM
Case c5
TCHS-DS-ATM
TCHS-ATM
2.0E+00
1.8E+00
1.6E+00
Tritium [g]
1.4E+00
1.2E+00
1.0E+00
8.0E-01
6.0E-01
4.0E-01
2.0E-01
0.0E+00
1
10
100
1000
10000
100000
Time [s]
Fig. 4.37 – Case C5 – Tritium releases to the external atmosphere
4.7. Case C6
4.7.1 Accident sequences description for the case C6
A preliminary analysis of a steam generator (SG) tube rupture (10 tubes i = 20 mm affected) to verify
the pressurization of one helium loop. The break area is 3.1416 E-03 m2.
4.7.2 Plant nodalization for the case C6
A simple model, including only one loop of the BL/FW Heat transfer system with the connected blanket
mass (similar to the input used in the previous cases) where only the decay heat is considered and the
full secondary system, has been used. The secondary system has been simulated as a volume of 7600
m3, filled with about 3.8 E+06 kg of water at 322 °C, saturated at 11.61 MPa (3.64% average steam
quality in the volume).
4.7.3 Main Thermal-hydraulics Results for the case C6
Results are shown in Fig. 4.38 and 4.39.
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In 200 s, the pressure equilibrium between the volumes is reached at 11.57 Mpa (non si verifca la
rottura del primario, con una tale pressione?). Flow rates of water into the BL-HTS loop varies from 42
kg/s to 0. Temperatures are not significantly affected by the rupture.
Case c6
50
45
Sec-HTS
Flow rates [kg/s]
40
35
30
25
20
15
10
5
0
0
200
400
600
800
1000
1200
Time [s]
Fig. 4.38 – Case C6 – Water flow rate through a break section between secondary and one BL-HTS loop
(come mai non c’è asintoto?)
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Case c6
1.40E+07
Pressure [Pa]
1.20E+07
1.00E+07
8.00E+06
BL-HTS
Secondary
6.00E+06
4.00E+06
2.00E+06
0.00E+00
0
200
400
600
800
1000
1200
Time [s]
Fig. 4.39 – Case C6 – Pressure transient in the steam generator tube rupture accident (meglio scala delle
ordinate da 8e6 a 1.2e7)
Summary of Thermal-hydraulic results
As a summary of the calculation performed the following considerations can be done:
Cases C1 and C2
Design values for the TCHS vault volume (58,000 m3) and vent area between VV and TCHS (0.2 m2)
are sufficient to face at the LOFA + in-vessel LOCA accident involving the rupture of 5 or 10 FW
cooling channels.
Case C3
A generalised loss of heat sink (LOFA + in-vessel LOCA affecting all the 9 loops) could lead to
elevated pressures if design values of the previous parameters are used. A very large expansion volume
of 500,000 m3 and a vent area of 0.5 m2 are needed to limit the pressure.
Case C4
In the case of a double guillotine break of a main piping inside the TCHS vault, a rupture disk area of 2
m2 between this volume and the other vaults has been verified and this values appears adequate to limit
the pressure in the vaults (pressure peak 0.155 MPa after 1 s in the TCHS vault).
Case C5
If a double guillotine break of an helium manifold of a module (i = 220 mm) considered as an invessel LOCA, a rupture disk area of 1.8 m2 between the VV and the TCHS is needed to limit the VV
pressure within the design value.
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Case C6
The pressure equilibrium between secondary and the BL-HTS loop is reached at 11.57 MPa in 200s.
A tritium migration evaluation has been also calculated in the sequences considered and the table 4-10
summarizes the results obtained, including the fraction of the initial inventory.
Initial
Tritium
Tritium
in the
inventory
VV
[g]
[g]
CASE C1
1001
CASE C2
1001
CASE C3
1009
CASE C4
1
CASE C5
1001
Tritium in
the TCHS
[g]
Tritium in
atmosph
from VV
[g]
Tritium in
atmosph
from TCHS
[g]
Total
Tritium in
atmosph
[g]
Tritium
captured
by the DS
[g]
1.92
0.192%
1.96
0.196%
16.77
1.66%
1.41E-0.3
Total
Tritium in
atm. from
DS
[g]
0.980
0.098%
0.978
0.098%
0.9433
0.093%
8.17E-04
18.7
2.96E-09
1.87%
0.000%
21.2
2.88E-09
2.12%
0.000%
0.1948
48.59
0.019%
4.82%
0.1744
9.1E-05
(EV)
17.440% 0.009%
56.13
0.02729
5.61%
0.003%
0.133
0.013%
0.142
0.014%
0.0096
0.001%
5.01E-03
(EV)
0.501%
0.3966
0.040%
3.03
0.303%
3.08
0.308%
17.7
1.75%
7.25E-03
979.3
97.832%
976.7
97.57%
942. 23
93.39%
0.816
0.141%
1.846
0.184%
0.082%
0.9426
0.094%
0.725%
3.19
0.319%
81.6%
941.6
94.07%
Table 4-10 – Summary of Tritium inventories and releases in the volumes after 1 day from the start of
the accident. (aumentare la altezza delle celle, altrimenti non si legge niente. Aggiungere un grafico a
barre per una migliore comprensione)
A preliminary and conservative evaluation of dust migration in the volumes and its release into the
atmosphere has been performed with a specific calculation in Case C1, considering 10 kg of dust
completely resuspended. With simple considerations based on the similar model used for tritium
migration, not considering the possible deposition or removal processes, the following table 4-11
summarizes the results for dust also for the cases C2, C3 and C5.
CASE C1
CASE C2
CASE C3
CASE C5
Dust in Dust in the
the VV
TCHS
[g]
[g]
Dust in
atmosph
from VV
[g]
Dust in atm.
from TCHS
[g]
Dust in
atmosph
from DS
[g]
184.7
1.85%
212
2.12%
1.95
0.02%
561.3
5.61%
1.32
0.013%
1.42
0.014%
0.096
0.001%
3.97
0.040%
1.92
0.019%
1.96
0.020%
16.77
0.168%
1.846
0.018%
0.980
0.010%
0.978
0.010%
0.9433
0.009%
0.9426
0.009%
2.96E-09
0.0%
2.88E-09
0.0%
48.59
0.486%
0.02729
0.0%
Total dust
Dust
in atmosph captured
[g]
by the DS
[g]
4.22
0.042%
4.358
0.044%
17.8
0.178%
6.76
0.068%
979.3
9.79%
976.7
9.77%
942.3
9.42%
941.6
9.42%
Dust
captured
by the
scrubber
[g]
8831.78
88.32%
8806.94
88.07%
8989.36
89.89%
8490.31
84.9%
Table 4-11 – Summary of dust inventories and releases in the volumes after 1 day from the start of the
accident. (vedere commento precedente)
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5. Final remarks and conclusions
The FFMEA analysis has been carried out focused on the possible consequences of loss of component
or system functions. The objective of the analysis has been defined a set of postulated initiating events
pointing out representative accident scenarios for the deterministic assessment: also, it has to be
ascertained if previous analyses performed for the other PPCS models can be used as enveloping
analyses applicable also to Model AB.
Some of the representative sequences pointed out by FFMEA have been analysed in deterministic way
by using the CONSEN code to evaluate the appropriateness of the design choices made for the
containments system and to define some additional design data like size of the rupture disk.
The other aim of the CONSEN analysis has been to determine in conservative way the environmental
source terms released as consequence of the accident sequences analysed showing a limited T and dust
release to the environment. The adoption of a scrubber and of a DS has been very effective in reducing
the dust and tritium migration to the environment.
The related dose to the population will be calculated using the results of analogous calculations carried
out for Model B (see below TO BE DONE).
It would be important to point out a possible problem relative to the design choice of adopting an
internal expansion volume (EV). It has been ascertained that in case of in-VV LOCA with pressure
relief of VV to secondary containment, contamination of operating areas with T and dust (e.g. TCHS)
would occur, but the quantities can be negligible if detritiation system will intervene..
A generalised loss of heat sink (LOFA + in-VV LOCA affecting all the 9 FW/BK loops) could lead to
elevated pressures to require a very large expansion volume (500,000 m3) to accommodate the
pressurization. It would be necessary to consider the possibility to have an external EV, to mitigate the
consequences.
Problem with the interconnection of the secondary heat transfer system through condenser and turbines,
with the possibility to involve all the primary loops in case of loss of heat sink could occur.
There is the need to deepen the accident analyses for events involving SGs and BUs (SG tube rupture
and interface LOCA between FW and Breeding Blanket).
TO BE COMPLETED
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References
[1]
A. Li Puma, L. Giancarli, “Helium-Cooled Lithium-Lead Fusion Power Plant (PPCS model AB)
Design and integration of in-vessel components and associated systems”, RAPPORT DM2S
SERMA/LCA/RT/04-3543/A, March 2005
[2]
D. J. Ward, “Modified Parameters for PPCS Plant Model AB with Modified Radial Build”,
PPCS/UKAEA/HCLL4, November 2004
[3]
D. J. Ward, private communication by e-mail dated 11th May 2005
[4]
L. Di Pace et al. , “Accident Description for Power Plant Conceptual Study”, PPCS/ENEA/TW1TRP-PPCS4/3, Rev. 2, October 2002
[5]
G. Caruso, M.T. Porfiri “Magnet induced confinement bypass accident simulation using
CONSEN 5 computer code” - FUS-TN-SA-SE-R-066 – March 2003
[6]
G. Caruso, M.T. Porfiri “CONSEN validation against EVITA. Complementary cryogenic tests
(2003): Pre- and Post-test calculations” - FUS-TN-SA-SE-R-089 – October 2003
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Appendix 1
MODEL AB DESIGN DATA FOR ACCIDENT ANALYSES
(metterei tutte le referenze usate in questa appendice nelle referenze generali perchè vengono usate referenze citate nella pagina di una tabella in
un’altra pagina contenente un’altra tabella e si fa fatica a ritrovarle. Unificandole nelle referenze generali la ricerca è più rapida)
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Table A.1.1 - Tokamak parameters
Major radius, R
Minor radius, a
Plasma current, Ip
Additional power
Fusion power
Toroidal field on axis
Plasma volume
Plasma surface
plasma thermal energy
Operational plasma cycle
Unit
m
m
MA
MW
MW
T
m3
m2
GJ
days
Nominal
Max
9.56
3.18
30
257
4290
6.7
3476(°)
1756 (°)
3.5 [3]
Steady state
Note: data taken from [2] with the exception of plasma volume and plasma surface and plasma thermal energy
(°) values calculated starting from related values given for PPCS Model A [4]
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Table A.1.2- FW/BLK and Divertor Parameters
FW/BL
Unit
Total thermal power
MW
Heat flux on FW/BLK, av./max.
MW/m2
Average neutron wall loading,
MW/m2
nominal/max.
First wall surface area inboard and
m2
outboard
First wall grid internal channel surface W/m2K°
heat transfer coefficient
Number of blanket modules
Non necessario, cancellare
Divertor
4219
0.50/1.84/2.58
1583 (°)
Total thermal power
Heat flux on divertor av./max.
Average neutron wall loading,
nominal/max.
First wall surface area
Unit
MW
MW/m2
MW/m2
926
-/10
-/-
m2
494 (°)
6150 [*]
180
Number of cassettes
(°) values calculated starting from related values given for PPCS Model A [4]
[*] private communication A. Li Puma
72
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Table A.1.3 - In-vessel components, dimensions and surfaces
Blanket Modules
Module 1
Module 2 (inboard equatorial)
Module 3
Module 4
Module 5 (outboard equatorial)
Module 6
Poloidal length [mm]
Toroidal length [mm]
Radial lenght [mm]
4750
4657
3874
5217
4460
4879
1998
2265
1903
1812
1812
1812
742
742
890
1331
1109
1109
FW Module surface
[m2]
9.5
10.5
7.4
9.5
8.1
8.8
(°) values calculated starting from dimensions taken from figures 1, 2, 12 and 15, of reference [1], adjusted to take into account the new values of R
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Table A.1.4 - Vacuum Vessel features
VV
Units
Free volume (includes the port extension volume)
Inner surface
m3
Inner surface port extension (equatorial port) (°°)
m2
9.6
Thickness
Mass of vessel
Mass of shielding
Total mass
Gap between VV and FW/BL
Walls temperature
Operating temperature (atmosphere)
Design pressure
Thermophysical properties of walls:
1. thermal conductivity  (20 °C < T < 1000 °C)
2. specific heat (20 °C < T < 1200 °C)
3. density (20 °C < T < 1200 °C)
m
t
t
t
m
°C
°C
MPa
0.47 (IB) / 1.02 (OB)
11553 [*]
9570 [*]
21123 [*]
N.A.
200 (°)
200 (°)
0.2
Stainless steel 316L [†], [T] in °C
 = 189.9 -0.2694·T + 2.5429E-4·T2 -1.0104E7·T3
Cp = 1741.8 + 3.3358·T -3.1125E-3·T2 + 1.2748E6·T3
 = 1823 - 6.933E-2·T -1.5139T-5·T2
m2
W/K·m
J/kg·K
kg/m3
5596 (°)
1667 (FW) (°)
521 (DV) (°)
(°) assumed for accident analyses
(°°) original equatorial port dimensions 4350 mm (pol), 2000 mm (tor.) [1], updated according to the increased machine dimensions (from R = 9.1 to 9.56 m). New calculated
equatorial port dimensions are: 4565 mm (pol.), 2100 (tor.)
[*]
[†]
R. Pampin, “Neutron transport and activation calculations for PPCS plant model AB“,UKAEA/TW4-TRP-002 Deliverable 2e January 2005
ITER Safety Analysis Data List, Vers. 4.0.1, IDoMS G 81 RI 10 03-08-08 W 0.1 (2003)
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Table A.1.5 - EUROFER thermophysical properties
Temperature
[°C]
Specific Heat
[J/kg·K] [‡]
Thermal
Conductivity
[W/m·K] [‡]
Density
[kg/m3] [‡]
20
448.85
25.9
7730
50
462.76
100
484.11
27.0
7710
150
503.92
200
523.04
28.1
7680
250
542.34
300
562.69
28.8
7650
350
584.94
400
609.96
29.2
7610
450
638.61
500
671.75
29.0
7580
550
710.25
600
754.96
28.5
7610
Melting point = 1400 – 1415 °C
FW thermal crisis = 800 °C at internal pressure of 8 MPa and with 2 mm W thickness
[‡]
Thermal Diffusion
Coefficient [cm2/s]
0.0885
0.0865
0.0822
0.0785
0.0725
0.0656
0.0575
P. Norajitra et al, “Conceptual Design of the Dual-Coolant Blanket within the Framework of the EU Power Plant Conceptual Study (TW2-TRP-PPCS12)” Final Report,
FZKA 6780, 2003
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Table A.1.6 - FW/BLK Cooling Loop
Primary side
Type of coolant
Number of loops
Coolant pressure (inlet blanket)
Total pressure drop
Coolant temperature (inlet blanket)
Coolant temperature (outlet blanket)
Total temperature increase
Heat load/loop
Mass flow rate
Internal diameter main pipes (cold leg/hot leg)
Internal diameter collectors to blanket modules (inbord/outboard)
Coolant speed in main pipe (hot leg/cold leg)
Coolant speed in collectors to blanket modules (min/max)
Units
MPa
MPa
°C
°C
°C
MW
kg/s/loop
m
m
m/s
m/s
In-vessel coolant hold-up components
m3
Collector pipes and main pipes coolant hold-up
m3
Heat Exchanger (primary side)
m3
Total hold-up
m3
(*) assuming 5.67 kg/m3 as average density
Based on a total thermal power to be removed of 4219 MW [1]
Helium
9 (1 loop per 40 ° sec tor)
8.25
0.35
300
500
200
469
452.2
1.05 / 1.25
0.22 / 0.25
84 / 86
50 / 125
289
1636 kg (*)
1921
10883 kg (*)
330
1871 kg (*)
2540
14390 kg (*)
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Table A.1.6 - FW/BLK Cooling Loop (cont.)
Secondary side [§], [**]
Number of loops
Type of coolant
Coolant pressure (inlet water)
Coolant pressure from steam generator (steam)
Pressure drop
Coolant inlet temperature (water)
Coolant outlet temperature from steam generator (steam)
Total temperature increase
Mass flow rate
Feedwater pipe diameter
Steam pipes to superheater (No. 3) diameter
Coolant speed in pipes (water / steam)
Coolant hold-up [††]
Steam Generator [data marked with (*) have been estimated]
Mass flow rate
tmean = (ta - tb)/ln(ta/tb) ta = THe in – Tsteam out = 500 – 444.8 = 55.2; tb = THe out – Twater in = 300 - 250 = 50.0
Heat transfer coefficient
Pressure drop (primary side)
Active height of tubes bundle
Pipe cross section (water /steam) [**] (penso sia il richiamo al riferimento sottostante. Corretto?)
Heat transfer area [**]
Primary side coolant hold-up [**]
Secondary side coolant hold-up [**]
[§]
[**]
[††]
MPa
MPa
MPa
°C
°C
°C
kg/s
m
m
m/s
m3
kg/s
°C
W/m2K
MPa
m
m2
m2
m3
m3
9
Water/steam
11.61
8.86
2.75
250
444.8
194.8
2068
0.508
0.923
1.4 / 34.8
7600
229.8
52.6
3708 (calculated with
tmean
2.75
12.2
0.12
3283
36.7
11.3
A Paule, presentation at the Final meeting of Task TW4-TRP-002, “PPCS Model AB, Primary & Secondary Heat Transport Systems, 3D Drawings, EFDA ORDER nº
93/851 UK”, IBERTEF A.I.E., Garching (D), May 12 th , 2005
D. Puente, presentation at the Final meeting of Task TW4-TRP-002, “B.O.P. specification, Task Order Number 93/851 JK”, IBERTEF A.I.E. & SENER, Ingeniería y
Sistemas, Garching (D), May 12th, 2005
A. Orden Martínez, private communication by e-mail dated 19th May 2005
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Empty weight [**]
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ton
~ 143
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Table A.1.6 - FW/BLK Cooling Loop (cont.)
Primary loop pump/circulator
Motor power
He coolant inventory
(i dati contenuti nelle le righe eliminate non servono per le analisi)
MW
m3
31
to-day technological limit of
motor : ~ 20 MW)
N.A.
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Table A.1.7 - Divertor Cooling Loop
Primary side
Type of coolant
Number of loops
Coolant pressure (inlet blanket)
Total pressure drop
Coolant temperature (inlet blanket)
Coolant temperature (outlet blanket)
Total temperature increase
Heat load/loop
Mass flow rate
Internal diameter main pipes (cold leg/hot leg)
Internal diameter collectors to blanket modules
Coolant speed in main pipe (hot leg/cold leg)
Coolant speed in collectors to blanket modules (min/max)
MPa
MPa
°C
°C
°C
MW
kg/s/loop
m
m
m/s
m/s
In-vessel coolant hold-up components including collectors
m3
Main pipes coolant hold-up
m3
Superheaters (primary side)
m3
Total hold-up
m3
(*) assuming 5.28 kg/m3 as average density
Units
Based on a total thermal power to be removed of 4219 MW [1]
Helium
3 (non necessario)
10
0.44
540
717
177
309
336.7
1.05 / 1.15
N.A
77 / 77
N.A.
97.3
513.5 kg (*)
586.8
3050.5 kg (*)
86.9
458.8 kg (*)
771
4022.8 kg (*)
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Table A.1.7 - Divertor Cooling Loop (cont.)
Secondary side same as the one for the FW/BLK Cooling
Number of loops
Type of coolant
Coolant pressure (steam from SG)
Coolant pressure (superheated steam)
Total pressure drop
Coolant inlet temperature from steam generator (steam)
Coolant outlet temperature from superheater (superheated steam)
Total temperature increase
Mass flow rate
Steam pipes (No. 3) diameter
Superheated steam pipes (No. 6) diameter
Coolant speed in pipes (steam /superheated steam)
Coolant hold-up
Superheater
Mass flow rate
tmean = (ta - tb)/ln(ta/tb)
ta = THe in – Tsteam out = 717 – 642.5 = 74.5; tb = THe out – Twater in = 642.5 – 444.8 =
95.2
Heat transfer coefficient
Pressure drop (primary side)
Heat transfer area
Primary side coolant hold-up
Secondary side coolant hold-up
MPa
MPa
MPa
°C
°C
°C
kg/s
m
m
m/s
m3
9
Water/steam
8.86
8.60
0.26
444.8
642.5
197.7
2068
0.923
0.681
34.8 / 44.8
7600
kg/s
229.8
°C
84.4
W/m2K
MPa
m2
m3
m3
N.A.
0.26
N.A.
N.A.
N.A.
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Table A.1.7 - Divertor Cooling Loop (cont.)
Primary loop pump/circulator
Motor power
MW
30
to-day technological limit of
motor : ~ 20 MW)
N.A.
m3
He coolant inventory
Table A.1.8 - Ventilation features of the containments
Containment
Vacuum Vessel (VV)
Design
pressure
(MPa)
0.2
THCS vault
0.16
Internal Expansion
Volume (EV)
0.16
(*) Legenda:
P0 = atmospheric pressure (Pa)
P = current pressure (Pa)
Pd = design pressure (Pa)
Leak rate
(% of the volume/day)
Scale rules
Leakage [m3/s]
1
(at design pressure)
1
(at design pressure)
1
(at design pressure)
Scales with square root of pressure differential (*)
Leakage (B) = [0.01 * Volume * SQRT[ (P-P0) / (Pd-P0) ] / (24*3600)
Scales with square root of pressure differential
Leakage (B) = [0.01 * Volume * SQRT[(P-P0) / (Pd-P0) ] / (24*3600)
Scales with square root of pressure differential Leakage = [0.01 *
Volume * SQRT[(P-P0) / (Pd-P0)] / (24*3600)
24 = hours per day
3600 = seconds per hour
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Table A.1.9 - Tritium, dusts, Activation Corrosion Products (ACPs)
Source terms
Tritium in VV
Dust
Tritium in coolant
Sputtering products
Model AB
1 kg
10 kg
(W-dust)
[3 E-3 g/m3* He loop Inventory (m3)] 0.85 g
0g
This table is built on the basis of SEAFP values for tritium in VV and amount of dusts.
1. For the sputtering products (Model B) the estimation performed in the UKAEA report (July 94), with pipe materials made of 89% of iron,
the sputtering resulting was 1.e-12g/m3, and a total amount of 2.47e-10 g. The SEAFP model 1 (helium cooled) had a fusion power of 3308
MW (against 3410 MW for PPCS Model C perché si fa il confronto con la potenza del modello C e non AB?), and a neutron wall load of 2.1
MW/m2 (against 2.2 MW/m2 for PPCS). The two most important terms for sputtering are quite similar. Due to that the final assumption
may be 0 g.
2. The tritium in the cooling loop in the SEAFP model 1 (helium cooled) was 7 g in total for all the 8 cooling loops. In PPCS Model B the He
cooling loops are 8 and the helium inventory per loop is 391 m3 against the 325 m3 of SEAFP. Average T concentration in SEAFP Model 1
= 7/8 /325 = 2.7 mg/m3. You can assume the same concentration or to round the figure to 3 mg/m3. The average loop inventory for Model
AB is 2540/9 =282.2 m3. Hence the estimated T inventory for Model AB cooling loop is  1 g/loop. Same assumption might be valid for
divertor loops (0. 8 g-T/loop). (il passaggio intermedio per modello B non mi sembra serva a molto. Lo eviterei.
3. Dust is only tungsten, as it is used as blanket armour material (2 mm thick) and for the divertor tile ( 5 mm minimum thickness)
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Table A.1.10 - Lithium lead properties
Property
Law
Unit
- density in solid state:
10.6(1.0 - 12.210-5T)
g cm-3
- density in liquid state
10.45(1.0 - 16.110-5 T)
g cm-3
- melting temperature (at atmospheric pressure)
508
K
- latent heat of fusion:
33.9
KJ/Kg
- specific heat in solid state:
-0.02417 + 3.92710-4 T + 4986.7 T-2 (*)
KJ/Kg K
- specific heat in liquid state (<T800 K):
0.195 - 9.11610-6T (*)
KJ/Kg K
- thermal conductivity in solid state:
17.7 + 2.9410-4 T (*)
W/m K
- thermal conductivity in liquid state:
1.95 +19.610-3 T (*)
W/m K
- dynamic viscosity:
0.187 exp (11640/RT) (*)
mPa s
- electrical resistivity in liquid state:
1x10-8 + 0.0428 x 10-8T (*)
m
(*) T in K
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Table A.1.11 - Vault isolation
Before vault isolation
Isolation
After vault isolations
The effluents are assumed to be released from the In case of helium ingress into the vaults, they A cooler of 10 MW keeps the temperature and the
pressure in the vault under control
plant exhaust. The effective vent area is 0.001 m2 will isolate within 30 s of receiving the signal
(10 cm2), the minimum duct length from the
vault to the exhaust is 20 m
Table A.1.12 - Conditions in not affected loops after the initiating event
Model AB
Pump or circulator
Temperature in coolant
circulator
coolant at value of 403 °C (constant).
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Table A.1.13 - Pressure control system
Pressure control
Model AB
When the flow rate is 80% of the nominal one, the pressure
control system stops the gas injection.
Table A.1.14 - Emissivity
Component
Emissivity
Temperature range:300 < T < 3500 K
FW surface
(W)
= -0.0434 + 1.8524e-4·T + 1.954e-8·T2 [-]
VV towards the shield
(neglecting the LT shield in model B)
Stainless steel 316L
0.17
[†]
Table A.1.15 - View factors
Number of sectors
View factor
Adjacent sectors to the affected one
2
0.70
Non adjacent sectors to the affected one
6
0.2
T in K [†]
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Table A.1.16 - Expansion volume related parameters
VV design pressure
2nd containment design pressure (TCHS, corretto?)
Disk rupture opening set point pressure from VV to
Expansion Volume (EV)
Disk rupture opening set point pressure from TCHS
vault to EV
Total volume available for expansion
Area of the disk rupture
from VV to Expansion Volume
Area of the disk rupture
from TCHS vault to Expansion Volume
0.2MPa
0.16 MPa
0.10 MPa
0.14 MPa
117600 m3
0.2 m2
5 m2
Table A.1.17 - Comparison Expansion Volume
Model B vs. Model AB
VAULT
TCHS
Pipechases
North
South
East
West
Total
External Expansion Volume
Total Volume available for
expansion
MODEL B
Volume (m3)
14900
15500
6600
12500
49500
68000
MODEL AB
Volume (m3)
58000
12000
8600
7500
14000
17500
117600
-
117500
117600
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Appendix 2
FUNCTIONAL FAILURE MODE AND EFFECT ANLYSIS (FFMEA)
Metter intestazione tabella su tutte le pagine. Il contenuto delle celle va allineato a sinistra per leggerlo meglio. Correggere il file excel in modo da
avere anche il file di orgine adeguato a questa versione)
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Function
Failure Mode
P1 VV and in vessel components conditioning
P1.1 Provide Vacuum
P1.1.1 Vacuum Boundary
Loss of vacuum boundary
Causes
Loss of VV leak tightness for
penetration break or loads impact
or untimely venting of plasma
chamber
Break of Cryopump panel in VV
P1.1.2 Active Vacuum Pumping Failure in active vacuum Loss of integrity outside VV in
pumping system
pump cryogenic circuit
LOFA in pump cryogenic circuit;
Surface fouling in cryopumps
channels
Control system failure
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Corr./Prev. Act. on
Consequence
Ingress of air and/or inert gas into VV;
Plasma disruption;
Risk of PFCs failure;
Loss of Pb-17Li in VV
PIEs
VVG
H2 production for the reaction of Li with air moisture: risk
of H2 explosion
Keep low the air moisture in
rooms surrounding the VV
Ingress of cryogenic fluid into VV;
VV contained radioac. products and T to cryopump circuit;
VV pressurisation
T release through containment wall leaks and stack
effluents
Isolation of faulty circuit;
VV pressure relief to expansion
volume
Air treatment
VV pressurisation;
Plasma shutdown;
Release of T contained in cryogenic circuit in cryo-system
room
VV pressurisation;
Plasma shutdown;
Cryogenic circuit pressurization
VV pressurisation;
Plasma shutdown
Air treatment
ACO
P1.2 Provide Conditioning of invessel components
P1.2.1 Cleaning discharge
Loss of cleaning discharge
Electrical discontinuity in PFC
None (Delayed impurity release to the plasma)
N/S
P1.2.2 Glow-discharge
Electrodes failure
None safety relevant consequences
N/S
FW localised damage
N/S
Loss of glow discharge
Glow discharge excessive Control system failure
power
P1.2.3 Pre-heating of in-vessel Loss of heating capability
component
Electrical discontinuity in PFC;
Loss of power supply;
Control system failure
None safety relevant consequences;
Delay in starting operations
N/S
P1.2.4 Pre-heating of Pb-17Li
Heater failure;
Loss of power supply;
Control system failure
None safety relevant consequences;
Delay in starting operations
N/S
Loss of heating capability
Comment
Vacuum boundary in
fuel
systems
and
diagnostics have to be
considered too
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P2 Confine/Ignite/Burn Plasma
P2.1 Provide Magnetic Confinement
P2.1.1 Provide Magnetic Field
P2.1.1.1
Keep Geometry of Loss of coil integrity for Material defects;
Coils and Supports
break or short circuit (inside Unexpected electromagnetic
a coil or at current leads)
load;
External event
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Asymmetrical loads;
Overvoltages;
Short circuit;
Arcing
Energy Dumping
Release of activated material (aerosols generated by the
arcing or by the mobilization of material trapped in
cryosurfaces) to the building through wall leaks and circuit
breaches
Air treatment
M4;
M7;
M5
P2.1.1.2 Provide Power Supply Loss of coil power supply
to Coils
and control
Equipment failure;
Loss of power supply
Plasma shutdown;
None safety relevant consequences
N/S
P2.1.1.3
Provide control of Loss of control capability
plasma shape and position
Equipment failure;
Control system failure
Plasma shutdown;
None safety relevant consequences
N/S
Failure in magnetic field
configuration
Equipment failure;
Coil CODAC failure;
Coil failure;
Loss of power supply
None safety relevant consequences
N/S
Failure in fuel injection
Equipment failure in PI system;
Control of PI system failure
None safety relevant consequences
N/S
Loss of power supply to
additional heating devices
Equipment failure;
Control of PI system failure;
Loss of power supply
None safety relevant consequences
N/S
P2.2 Ignite/Burn DeuteriumTritium Plasma
P2.2.1 Provide Plasma Ignition
The effectiven. of the
ED depends on the
power supply and ED
design, and on the kind
of failure
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P2.2.2 Provide Burn Control
P2.2.2.1 Control Fuelling
P2.2.2.1.1
Control Fuel Wrong composition: less
Composition
fuel content than needed
P2.2.2.1.2 Inject Fuel
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Equipment failure;
Control of FMS failure
None safety relevant consequences
Wrong composition: more
fuel content than needed
Equipment failure;
Control of FMS failure
Limited overpower transient (PFC in-vessel LOCA could
be induced);
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Isolation of fuel system
Inadvertent pellet injection
Control of PI system failure;
CODAC system failure
Possible structural failures on PFC;
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Redundant control systems; LFI1
Pellet
injection
shutdown
(manual, automatic)
Faulty fuelling supply: less
pellet injection than needed
Equipment failure in FMS or PI
system (e.g. FMS process
boundary failure)
Release of T (and D, H) from the faulty equipment to
secondary containment;
Loss of VV vacuum boundary integrity in case of loss of PI
vacuum integrity and PI not isolated (see function "P1.1.1
Vacuum Boundary - Loss of vacuum boundary")
T
monitoring; TPI1;
Isolation
of
PI; TPI2
Inert
gas in PI room;
Air treatment
Control of FMS or PI system
failure;
CODAC system failure
None safety relevant consequences
Faulty fuelling supply: more Equipment failure in FMS or PI
pellet injection than needed system;
Control of FMS or PI system
failure;
CODAC system failure
P2.2.2.2 Provide Control of Failure on control plasma Equipment failure (i.e.:
Plasma Impurity
impurity
Cryopumps);
Loss of plasma configuration
Limited overpower transient (PFC in-vessel LOCA could
be induced);
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Plasma shutdown;
PFC damages
N/S
TT
Plasma diagnostic (Temperature, TT
density)
;
Emergency Plasma Shutdown
N/S
Possible
loss
of
primary vacuum in
case of PI isolation
failure
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P3 Remove Heat from Plasma during Normal Operation
P3.1 Convert Neutron load and Reduction in converting
radiation load into Thermal capability
Energy
Increase of converting
capability
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Incorrect Li-Pb flow because of
wrong design or faulty system
behaviour (e.g.: LOFA,
unexpected turbulence)
Increase of radiation to the magnets;
General increase of radiation to the out of vessel
components;
Long term damage of coils thermal insulation
Design
in
Design margins
QA; M3
Control system failure
Local abnormal thermal load on PFCs;
Possible PFC local melting and coolant leak inside plasma
chamber;
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Design
in
Design margins
QA; TT
P3.2 Remove Thermal Energy
P3.2.1 Provide FW Cooling
P3.2.1.1 Keep FW Cool.Loop Loss of loop integrity in VV
Integrity
Component break;
LOCA in-vessel from primary cool.loop;
Impact of heavy loads (missiles); Plasma disruption
Abnormal operating conditions
(more heat load than designed)
Design margins;
Design against missile
generation;
Monitor plasma facing
components degradation
Primary loop He discharged in the VV;
VV pressurisation
VV pressure relief to expansion
volume
Release of radioactive products and T from VV to vault
after pressure overcoming in VV with respect to external
pressure;
Release of T to the environment through wall leaks and
ADS (unfiltered materials)
Air treatment
HX tube rupture due to primary cooling circuit sudden
depressurization increasing dynamic loads;
Steam from secondary loop enters primary loop;
He enters VV;
VV over pressurisation;
VV boundary failure, e.g.: penetration seal break towards
cryostat or room surrounding cryostat;
Release of radioactive products and T from VV to
secondary containment;
Release of radioactive products and T from VV to building
through wall leaks and to environment through ADS
(unfiltered materials)
Isolation of secondary loop;
VV pressure relief to expansion
volume
LFI1;
LFI2
Depending on the
design an initial
leakage could lead to
a catastrophic rupture
Adequate routing to
expansion volume to
be provided
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Associazione ENEA-EURATOM sulla Fusione
P3.2.1.1 Keep FW Cool.Loop Loss of loop integrity in Component break;
Integr. (cntd)
cooling room
Impact of heavy loads (missiles);
Abnormal operating conditions
(i.e.: vibration)
FUS-TN
SA-SE-R-129
EMISSION
DATE
24-06-2005
REV. 0
PAGE
82 of 87
LOCA in cooling room
Design against missile
generation
Release of T contained in VV coolant, into cooling
system room;
Release of T contained in coolant to the environment
through wall leaks and ADS (unfiltered materials)
Air treatment
Cooling room pressurization;
Cooling room inertization because of He ingress;
FW in VV LOCA due to overheating;
Ingress of air and/or inert gas into VV;
VV pressurisation due to heat load and release of T and
activated materials to the cooling room
Isolation of faulty circuit;
Cooling room pressure relief
Release of radioactive products and T from VV to
building through wall leaks and to environment through
ADS (unfiltered materials)
Air treatment
LFO3
P3.2.1.2
Provide FW Cool. Loss of coolant flow
Loop Flow
Circulator failure
PFC structures heat up and primary cooling loop
pressurisation;
LOCA inside VV;
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Redundant circulator;
Preventive maintenance;
Flow detection/control;
Pressure detection / control;
Emergency Plasma Shutdown
FF1;
FF2;
FF3
Design
robustness
will depend on the
time the structures are
able to sustain the
temperature increase
compared with the
time of fault detection
and
plasma
shutdown. In case of
low margins there
could
be
a
generalised structural
failure
P3.2.1.3 Provide FW Heat Sink
Turbine trip;
Pump failure;
Loss of power supply;
Untimely valve closure in sec.
loop;
LOCA in secondary cooling loop;
Loss of UHS
General heat up of PFC structure;
Primary cooling loop pressurisation;
Possible primary loop LOCA in-VV;
Consequences as for "P3.2.1.1 Keep FW Cool.Loop
Integrity - Loss of loop integrity in VV" could follow
Emergency Plasma Shutdown; HSF;
Emergency Power supply (e.g. HS
diesel generators)
Turbine trip or load
rejection is a very
frequent event the
reactor has to deal
with.
Plasma
shutdown has to be
effective in order to
avoid general PFCs
structural failure
Loss of heat sink
DOCUMENT
Associazione ENEA-EURATOM sulla Fusione
P3.2.2 Provide BB Cooling
P3.2.2.1
Keep BB Loop Loss of loop integrity in VV
Integrity
P3.2.2.2 Provide BB Loop Flow
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DATE
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He ingress within Pb-17Li circuit: circuit pressurization;
Partial loss of flow in the breedeer module;
FW in VV LOCA due to overheating
Flow detection/control;
Pressure detection / control;
Emergency Plasma Shutdown
LBB1
Emergency shutdown
could avoid FW
LOCA;
Breeder circuit is
assumed not designed
to secondary He
pressure
Loss of loop integrity in As for P.3.2.1.1 function
cooling room
Liquid metal loss in system room;
Tritium contained in the breeder loop released in the room;
Release of T to the environment through wall leaks and
ADS (unfiltered materials);
FW in VV LOCA due to overheating
Flow detection/control;
Pressure detection / control;
Emergency Plasma Shutdown;
Air treatment
LBO3
FW
consequential
LOCA should be a
very late event
Los of liquid metal breeder Loss of power supply;
flow
Pipe/channel plug;
Pump failure
Breeder circuit failure towards FW;
Structural failure of FW channels towards VV
Flow detection/control;
Pressure detection / control;
Emergency Plasma Shutdown;
VV pressure relief to expansion
volume
FB1;
FB2;
FB3
Design
robustness
will depend on the
time the structures are
able to sustain the
temperature increase
compared with the
time of fault detection
and
plasma
shutdown. In case of
low margins there
could
be
a
generalised structural
failure
LOCA inside VV;
Plasma disruption
Design margins
LSI1;
LSI2
LTS is water cooled,
only
activate
corrosion
products
and tritium for release
Loss of loop integrity in Component break;
heat exchanger
Abnormal operating conditions
(i.e.: vibration)
Release of ACPs to secondary circuit
Secondary loop HX provide
additional barrier
LSO2
Loss of loop integrity in Component break;
cooling room
Impact of heavy loads (missiles);
Abnormal operating conditions
(i.e.: vibration)
Release in cooling room of T and activated products
contained in coolant;
Release of T and activated products contained in coolant to
the environment through wall leaks and ADS (unfiltered
materials)
Air treatment;
Drainage
LSO3
P3.2.3 Provide Low Temperature Shield Cooling
P3.2.3.1 Keep LTS Cool.Loop Loss of loop integrity in VV
Integrity
Abnormal operating conditions
(i.e.: vibration);
Abnormal operating conditions
(more heat load than designed);
Alteration of structural materials
properties
FUS-TN
SA-SE-R-129
Component break;
Abnormal operating conditions
(more heat load than designed)
DOCUMENT
Associazione ENEA-EURATOM sulla Fusione
FUS-TN
SA-SE-R-129
EMISSION
DATE
24-06-2005
REV. 0
PAGE
84 of 87
P3.2.3.2 Provide LTS Cool. Loss of coolant flow
Loop Flow
Pump failure;
Loss of power supply;
Pipe/channel plug
Slow LTS circuit overheating
Emergency Plasma Shutdown
FS1;
FS2
P3.2.3.3 Provide LTS Heat Sink
Loss of circulating water
Slow LTS circuit overheating
Emergency Plasma Shutdown
HSF;
HS
Loss of heat sink
P3.2.4 Provide Divertor Cooling
P3.2.4.1
Keep
Divert. Loss of loop integrity in As for P.3.2.1.1 function
Cool.Loop Integrity
VV
As for "P3.2.1.1 Keep FW Cool.Loop Integrity - Loss of
loop integrity in VV"
LDI1;
LDI 2
Loss of loop integrity in As for P.3.2.1.1 function
heat exchanger
As for "P3.2.1.1 Keep FW Cool.Loop Integrity - Loss of
loop integrity in heat exchanger"
LDO2
Loss of loop integrity in
cooling room
As for P.3.2.1.1 function
As for "P3.2.1.1 Keep FW Cool.Loop Integrity - Loss of
loop integrity in cooling room"
LDO3
P3.2.4.2 Provide Divertor Cool. Loss of coolant flow
Loop Flow (bulk)
As for P.3.2.1.2 function
As for "P3.2.1.2 Provide FW Cool. Loop Flow - Loss of
coolant flow"
FD1;
FD2;
FD3
P3.2.4.4 Provide Divertor Heat Loss of heat sink
Sink
As for P.3.2.1.3 function
As for "P3.2.1.3 Provide FW Heat Sink - Loss of heat
sink"
HSD;
HS
Component break;
Abnormal operating conditions
(i.e.: vibration)
LOCA inside VV;
Plasma disruption
Design margins;
LVI1
VV pressure relief to expansion
volume
Loss of loop integrity in
heat exchanger
Component break;
Abnormal operating conditions
(i.e.: vibration)
Release of ACPs and T to secondary circuit
Secondary loop HX provide
additional barrier
LVO2
Loss of loop integrity in
cooling room
Component break;
Impact of heavy loads (missiles);
Abnormal operating conditions
(i.e.: vibration)
Release in cooling room of T and activated products
contained in coolant;
Release of T and activated products contained in coolant to
the environment through wall leaks and ADS (unfiltered
materials)
Air treatment;
Drainage
LVO3
Pump failure;
Loss of power supply;
Pipe/channel plug
VV structure heat-up;
Quench due to VV overheating
Emergency Plasma Shutdown
FV
P3.2.5 Provide VV Cooling
P3.2.5.1
Keep VV Coolant Loss of loop integrity in
Loop Integr.
VV
P3.2.5.2 Provide VV Coolant Loss of coolant flow
Loop Flow
Steam pressurisation
protection
is
oversized,
being
designed for He
LOCA
The heat up transient
should be mild, given
that the VV structures
do not face the
plasma
DOCUMENT
Associazione ENEA-EURATOM sulla Fusione
P3.2.5.3 Provide VV Heat Sink
Loss of heat sink
Loss of UHS
FUS-TN
SA-SE-R-129
EMISSION
DATE
24-06-2005
REV. 0
PAGE
85 of 87
VV structure heat-up;
Quench due to VV overheating
Emergency Plasma Shutdown
LOCA inside VV;
Plasma disruption
Design margins;
LAI
Design against missile
generation;
VV pressure relief to expansion
volume
Loss of loop integrity in Component break;
heat exchanger
Abnormal operating conditions
(i.e.: vibration)
Release of ACPs and T to secondary circuit
Secondary loop HX provide
additional barrier
LAO2
Loss of loop integrity in Component break;
cooling room
Impact of heavy loads (missiles);
Abnormal operating conditions
(i.e.: vibration)
Release in cooling room of T and activated products
contained in coolant;
Release of T and activated products contained in coolant to
the environment through wall leaks and ADS (unfiltered
materials)
Air treatment;
Drainage
LAO3
P3.2.6 Provide Addition.Heating syst.Cool.
P3.2.6.1 Keep AH syst. Coolant Loss of loop integrity in Component break;
Loop Integr.
VV
Impact of heavy loads (missiles);
Abnormal operating conditions
(more heat load than designed)
HSV;
HS
P3.2.6.2
Provide AH syst. Loss of coolant flow
Coolant Loop Flow
Pump failure;
Loss of power supply;
Pipe/channel plug
Local heat-up in antenna;
General heat-up of antenna;
Thermal stress on structures;
Primary cooling loop pressurisation;
LOCA inside VV
Emergency Plasma Shutdown
LAI;
VVG
P3.2.6.3 Provide AH syst. Heat Loss of heat sink
Sink
Loss of UHS
General heat-up of antenna;
Thermal stress on structures;
Primary cooling loop pressurisation;
LOCA inside VV
Emergency Plasma Shutdown
HSR;
HS
LOCA in cooling room;
Release of T and ACPs’contained in coolant to the
environment through wall leaks and ADS (unfiltered
materials);
Cooling room pressurization
Isolation of faulty circuit;
Air treatment;
Cooling room pressure relief
LCO
P3.3
Provide
Coolant Loss of loop integrity in Component break;
purification,
inventory
and cooling room
Impact of heavy loads (missiles);
chemistry
Abnormal operating conditions
(i.e.: vibration)
No
information
exist on cooling
loop
for
AH
systems. Here it has
been assumed they
are water cooled
DOCUMENT
Associazione ENEA-EURATOM sulla Fusione
FUS-TN
SA-SE-R-129
EMISSION
DATE
24-06-2005
REV. 0
PAGE
86 of 87
P4 Provide Fuel Cycle Functions
P4.1 Produce and extract tritium
P4.1.1 Produce tritium in the Loss of tritium production Neutron
flux
reduction; None safety relevant consequences
Pb-17Li
in breeder blanket
Breeding material degradation;
Loss of control and adjustment
of 6Li-concentration
P4.1.3
Keep
Pb-17Li Loss of integrity in system Wearing due to corrosion, Liquid metal loss in system room;
purification system integrity
room
vibration and pressure transient; Release of T and activated products contained in primary
Impact of heavy loads (missiles) coolant to system room;
Relase of gaseous T and He purge gas into system room;
Release of T and activated products contained in coolant to
the environment through wall leaks and ADS (unfiltered
materials)
H2 production for the reaction of Li with water contained
in air: risk of H2 explosion
P4.2 Fuel recovery
Loss of fuel recycle from Cryopumps
failure; None in the short term;
plasma exhausts
Baking pump train failures; Possibility of tritium release in the cryopump room in the
Regulation or isolation valve long term
failure
Loss of plasma exhausts Equipment
loop integrity
Process barrier leakage
failure; T or tritiated compound release from process barrier;
Possible air/H isotope mixture;
Loss of primary vacuum boundary integrity
N/S
Radioactivity monitoring;
Room leaktightness;
Isolation of faulty circuit;
Air treatment
Keep low the air moisture in
rooms hosting the Pb-17Li
circuit
Room atmosphere inertizationtreatment
Secondary containment
provision;
Secondary containment
atmosphere control;
Plasma chamber isolation from
fuel cycle systems
LRO1
N/S
TFE1;
TFE2;
TFE3
Loss of flow in breeding Circulator
failure; Increase of tritium permeation
blanket purge line
Loss of pump. power supply;
Valve closure or stuck close;
Pipe/channel plug
N/S
Loss of integrity in breeding Equipment
blank. purge line
Process barrier leakage
TPL1;
TPL2
failure; T or tritiated compound release from process barrier;
Possible air/H isotope mixture
DOCUMENT
Associazione ENEA-EURATOM sulla Fusione
P4.3 Process fuel
Loss to purify fuel
Cryotrapping failure;
Membrane failure;
Loss of fuel flow
FUS-TN
SA-SE-R-129
EMISSION
DATE
24-06-2005
REV. 0
PAGE
87 of 87
None safety relevant consequences
N/S
Loss to separate hydrogen Equipment failure in CDC;
isotopes
Loss of cryogenic system
capability;
Pipe/channel plug
Release of H isotopes in cold box ;
H isotopes expansion in expansion tanks
Loss to store fuel
T or tritiated compound release from process barrier;
Possible air/H isotope mixture;
Loss of primary vacuum boundary integrity
TFS1;
TFS2
T or tritiated compound release from process barrier;
Possible air/H isotope mixture;
Loss of primary vacuum boundary integrity
Avoid missiles generation;
TPL1;
Avoid failure propagation
TPL2
providing break segregation
(e.g.: via buffer tank operating in
push-pull configurat.);
Always double process barrier;
Maintain vacuum into the inter space between double process
barrier (T monit.)
Loss of loop
(catastrophic
noncatastrophic)
Equipment failure
integrity Equipment failure;
/ Impact of heavy loads (missiles);
Process barrier leakage;
Process barrier pressurisation;
Spurious safety relief valve
opening
T monitoring;
Isolation of faulty circuit;
Provide a good layout of HVAC
TISS1;
TISS2;
TISS3