Thermal effects of residual power on plasma facing components of net

Fusion Engineering and Design 6 (1988) 269-279 North-Holland, Amsterdam

269

THERMAL EFFECTS OF RESIDUAL POWER ON PLASMA FACING COMPONENTS OF NET Vito RENDA,

Antonio SORIA and Loris PAPA

CEC-JRC Ispra Estab., 21020 lspra (VA), Italy Submitted 23 June 1988; accepted 27 June 1988 Handling Editor: G. Casini

Residual power is of concern for the safety of all kinds of nuclear reactors and the environmental consequences of an accident caused by it. In the case of the NET (Next European Torus) fusion reactor, residual power is due to decay heat of activated structural materials. Preliminary temperature transients after plasma shut-down have been assessed assuming one year of nominal running and 100% of load factor. Loss of flow accidents (LOFA) were studied and the results show that no major hazard will arise from a LOFA on a single cirruit, while reactor melting could happen in the case of a LOFA in the whole reactor. Additional calculations show that o n e of the main cooling circuits remaining operational could be enough to avoid major hazards, so that thermohydraulic analysis is underway to arrange the circuit of the inboard first walls for natural cooling so as to guarantee that the reactor is inherently safe.

1. Introduction

2. Effects of the radioactive products

N E T (Next E u r o p e a n Torus) is a design for a fusion reactor of technological interest which is sponsored by the Commission of the European Communities. Its characteristics are similar to those of a possible First Commercial T o k a m a k Reactor ( F C T R ) so that the safety studies developed for N E T can also play a very i m p o r t a n t role in the future of fusion technology. JRCIspra is at present engaged in studies for an inherently safe reactor. Those studies are strictly d e p e n d e n t on plasma parameters, geometric configuration a n d choice of materials, so that reference is made to N E T with the aim also of finding a design strategy for a n inherently safe FCTR. One of the most i m p o r t a n t source of concern is due to the residual power stored in the reactor which could lead to the melting of the structures and the consequent release of radioactive products to the environment. The studies underway are last to design the reactor in such a way to avoid hazards by means of only inherent passive systems. The studies presented in the note are related to the residual power effects, after a reactor shut-down, in cases of Loss Of Flow Accident (LOFA). The aim of the work is the analysis of possible hazards and the identification of possible passive systems which could be used to keep the reactor safe.

The most serious concern related to safety a n d environmental aspects of nuclear plants is the hazard associated with the radioactive inventory. F r o m the safety point of view it is advisable to separate the effects of the radioactive products into the following groups of problems: - decay heat, - dismantling a n d repair, - decommissioning, - waste m a n a g e m e n t . - Hazards due to decay heat can rise in the case of loss of refrigeration, even assuming that a reactor shut-down is ensured by the control system. - Dismantling and repair is m u c h easier if the dose equivalent rate produced by the c o m p o n e n t is low. - F r o m the point of view of p l a n t decommissioning the most desirable conclusion is the achievement of a n allowed unrestricted site release. If the induced radioactivity of the p l a n t materials decays sufficiently, the decommissioned c o m p o n e n t s can be handled and reworked into new c o m p o n e n t s for new reactors. In general it can be useful to classify as short term effects those related to decay heat a n d dismantling a n d

0 9 2 0 - 3 7 9 6 / 8 8 / $ 0 3 . 5 0 © E l s e v i e r S c i e n c e P u b l i s h e r s B.V. (North-Holland Physics Publishing Division)

270

V. Renda, A. Soria, L. Papa / Thermal effects of residual power

repair, and as long term effects those related to decommissioning and waste management. Short term effects depend essentially on activated elements of short life, while long term effects depend on activated elements of long life. It is obvious that the elementary material composition plays a basic role in defining low activation materials; it is very difficult to optimize the composition to minimize both short and long term effects.

3. Comparison between fission and fusion Residual power is one of the most i m p o r t a n t concerns in fission reactors where most of the radioactive elements are concentrated in fuel assembles and are due to fission products and actinides. In the case of a Pressurized W a t e r Reactor (PWR), if a loss of coolant flow occurs followed by an automatic shut-down, the reactor is shifted to a state of hot-standby; in that case the reactor is partially depressurized to decrease the stresses on the vessel wall. Water boiling must be avoided to guarantee that hot spots, cladding failures and fuel coolant interaction do not occur. The reactor needs an emergency cooling circuit and the design parameters must be chosen to ensure a temperature lower than the water saturation temperature ( - 1 8 0 ° C ) at the design pressure related to the hot stand-by state. In the case of a Liquid Metal Fast Breeder Reactor ( L M F B R ) , if the p u m p i n g system fails and an automatic shut down occurs, the residual power delivered by the fission products in the core gives rise to a temperature increase of the sodium b o t h in the core and in the region near the vessel. Two main problems can arise: - Sodium boiling in the hot channel, local stop down of the natural convection and failure of the pins with consequent s o d i u m - f u e l interaction. - Temperature increase of the vessel, which contains the sodium and the core, with consequent structural damage with inelastic strains in the vessel itself. Because of these two phenomena, an emergency cooling circuit must be foreseen for the reactor. The design parameters must be chosen b o t h to avoid sodium boiling in the core and to guarantee the structural reliability of the vessel. The comparison between fission and fusion leads to the following conclusions. Most of the radioactive elements of fusion reactors are concentrated in the structural material of the First Wall a n d the Blanket. Fusion neutrons, generated at an energy of 14 MeV, may induce m a n y threshold activation reactions, much

as (n, p), (n, d), (n, a), which are practically absent in a fission reactor. This means that they are more effective in inducing activation reactions. In a fusion reactor more than 30% of the power is released in the first wall and more than 90% in the first wall and blanket. A b o u t 50% of the radioactive inventory is concentrated in the first wall and more than 95% in the first wall and blanket. This explains why these c o m p o n e n t s are the main concern for the radioactive hazard. Tritium may also be considered as a p r o d u c t of neutron activation. The problems related to this radioactive isotope are however different from those of the other isotopes. It is the main c o m p o n e n t of the fuel cycle, it is extracted from the blanket and it will probably be separated from the coolant, stored a n d reinjected into the plasma. The presence of tritium in the first wall is controlled by the processes of atomic b o m b a r d m e n t of plasma particles during b u r n and of the outgassing during shut-down. Tritium does not give rise to residual power; so that it is not considered in the context of the present note. Quantitative comparisons can be found in the literature [1] for some specific reactors; but it should be stressed that the hazards related to decay heat after a reactor shut-down, the subject of this note, are strictly dependent on the reactor geometry. While in a fission reactor a high residual power density is stocked in a small volume, in a toroidal fusion reactor a low residual power density is stocked in a great volume. Hazards due to decay heat depend o n the possibility of heat transfer from the source of power to the environment. A comparison between fission a n d fusion is shown quantitatively in table 1 for two reactors of c o m p a r a b l e power.

Table 1 Comparison between Phenix and INTOR Phenix

INTOR

Total power (MW) 560 620 Active volume (m3) 1.2 a 150 b Max. power density ( M W / m 3) 650 20 Average power density (M'W/m 3) 470 4 Max. neutron flux ( n / m 2 s) 7.2× 1019 7.0× 10 TM Vol. integrated flux (nm/s) 6.3 × 1019 2.1 × 102o a Core volume. b Volume of the first wall and blanket.

V. Renda, A. Soria, L Papa / Therraal effects of residual power 4.

Reference

reactor

t~

,.[] Hazards due to decay-heat can be evaluated only for a specific reactor. Reference will be made to N E T (Next European Torus) whose data base for the calculations is known. Moreover N E T is designed as a technological experiment on the way to a power plant, so that the results of the study can guide the design of a future commercial reactor.

271

}

) ®

4.1. Geometry

The general architecture of N E T is shown in an artistic view in fig. 1. The toroidal chamber, working as vacuum vessel, is assembled in 16 sectors embracing an angle of 22.5 o each placed side by side in the toroidal direction. The horizontal and vertical sections of the reactor are shown in fig. 2. The reactor consists of the following in vacuum components: Inboard First Wall (IFW) facing the inner part of the plasma and including the two divertor regions, is made up of 48 panels, assembled (fig. 3) in groups of three per sector. As shown in the cross section of fig. 4, the panel consists of stainless steel cooling tubes embedded in a high conductivity copper matrix, protected by a graphite armour and stiffened by an AISI-316L structure. This structure, called the inboard shielding blanket (ISB), must be characterized by shielding properties so that a water circuit is planned; the system is arranged in such a way that this circuit could be a

NET NEXT EUROPEAN TORUS

A) VERTICALSECTION O) HORIZONTALSECTION Fig. 2. NET-DN vertical and horizontal sections.

Fig. 3. First wall/shielding blanket panels.

shielding blanket containing lithium salts dissolved in water. A detail of the first wall panel with U-cooling tubes is shown in fig. 5. Outboard First Wall (OFW) consisting of 48 boxes containing the blanket (fig. 6) with a general configuration shown in fig. 7. It is closed by an outboard back plate (OBP). The box envelope is made of two stainless steel plates including square cooling tubes as shown in the detail of fig. 8. The plasma side is protected by a graphite layer.

PLASMA

k " , ~ 3 .',Z~Y\\\\\\K~\"/

K~,\\\\\\\\\\\\~K-~ Fig. 1. Net general configuration.

Fig. 4. Panel cross section.

V. Renda, A. Soria, L Papa / Thermal effects of residual power

272

ARMOUR PLASMA ~" (GRAPHITEI %-//////.-/,~.-

.- ~

20 coolant-inletside H20 coolant-outlets4de

--~-

\\

,

F~EoTpSEIN) ~I~'I'I~!~i~,BAC

K PLATE

Fig. 5. Detail of the first wall.

t

®®$®®®®lJ ®51

/

liquid breeder Fig. 9. Breeding module cross section.

t P A MA t Fig. 6. Outboard movable part (cross section).

Fig. 7. First wall general configuration.

-///

Fig. 10. Vacuum vessel detailed design.

//, y ~ x 7

plates as shown in fig. 10. The multilayer structure contains water for reasons of neutronic shielding. 4.2. Model

|t.

o,o

tl

t

Fig. 8. First wall detail.

Breeding Blanket (BB) consisting of stainless steel tubes bent in the poloidal direction and enveloped by the O F W boxes. Each box contains 25 breeding modules organized in 5 rows as shown in fig. 6. Each module, whose cross section is shown in fig. 9, contains the liquid eutectic Pb-17Li breeding material and the cooling circuit. Vacuum vessel, consisting of the inboard (IVV) and outboard ( O W ) parts, enveloping all the plasma facing components and assembled in stainless steel -

-

The thermal transients on the reactor components were assessed by mean of the program T R E C G A [2] developed, for this specific problem, at the JRC-Ispra. It is a monodimensional analysis of coaxial cylindrical slabs taking into account radiation between faced and surfaces. An artistic view of the model is shown in fig. 11 and the details are shown in fig. 12. It was divided in three blocks: the movable inboard part of the reactor (IFW-ISB), - the movable outboard part of the reactor ( O F W / OBP-BB), - the semipermanent vacuum vessel (IVV/OVV). The figure also shows the numbering of the slabs (from 1 to 75), these materials (AISI-316L SS, Cu, -

1I. Renda, A. Soria, L Papa / Thermal effects of residual power

273

Table 2 Main component dimensions Components

Zone

Radius (m) Inner

Outer

IFW

Graphite Cu-heat-sink

3.647 3.590

3.657 3.635

ISB

Whole

3.42

3.590

OFW

Graphite AISI 316 box

6.538 6.560

6.558 6.577

OBP

Back plate

7.340

7.490

BB 1

Whole

6.660

6.744

BB 2

Whole

6.790

6.874

BB 3

Whole

6.920

7.000

BB 4

Whole

7.050

7.134

BB 5

Whole

7.180

7.264

IVV

Whole

2.250

3.250

OVV

Whole

7.680

8.270

Fig. 11. Artistic view of the model.

Pb-17Li, Graphite and H 2 0 ) and the five planned independent circuits (from C1 to C5). The main dimensions of the components are described in table 2 in terms of their radial position from the reactor axis.

4.3. Material data Five different materials were taken into account in the calculations. They are: - AISI-316L stainless steel, main structures, - copper, heat sink of IFW, - graphite, armour for both IFW and OFW, - Pb-17Li, breeding material, water, coolant and neutronic shielding. The material properties are summarized in table 3. -

4. 4. Steady state temperatures The steady state is characterized by the temperatures of the components. U p till now the steady state thermal

MODEL OUTBOARD

INBOARD

2tl

PLASMA 16

19

BLANKET DETAIL

LEGENDA

i!ilii ~/-~ ,~.~,-~!!,.~, g

...H20 [ ~ ...COPPER ...GRAPHITE ...LIPB I

Fig. 12. Detailed model.

V. Renda, A. Soria, L Papa / Thermal effects of residual power

274 Table 3 Main material properties Material

AISI-316L

Copper

100-300 -

305

1380 -

350 -

7920-7825 512- 540 15- 18 30

8850 420 385 30

2100 1950 30 100

9430 90 16 /

Temperature ( o C) Pressure (MPa) Mass. density ( k g / m 3) Specific heat ( J / k g ° K ) Conductivity ( W / m OK) Emissivity (%)

Graphite

Pb-17Li

Water 100-300 0.5 15 960- 730 4200-5470 300 a -

" Equivalent conductivity simulating convective heat exchange.

Table 4 Main component temperatures

0.g I (FLUENCE

Components

Zone

Temperature ( ° C)

IFW

Graphite Copper

1380 305

ISB

AISI-316 SS

OFW

Graphite AISI-316SS-BOX

i 0r~ 06 ~

COPpeR0rw~

as , 04

325

AISI 316 SS (OFW)

03

1360 310

OBP

Backplate

300

BB

Row 1 Pbl7Li

350

1 MWA/M2)

o.2~ . ~ . . . k . o ,~,k" ",__'~_~ oo I~

~

~ ......

,~

- ......... 2b

PB,rL, (~STROW) / ./. . . . . . . . . . . . . . . . . 3b 4b

TIME (HOURS)

Fig. 13. Residual power transients. W

Inboard and outboard

100

e q u i l i b r i u m h a s n e v e r b e e n assessed, so t h a t t h e s t e a d y state temperatures have been assumed starting from design consideration, power distribution and expected fluxes b e t w e e n c o m p o n e n t parts. E v e n if arbitrarily d e f i n e d t h e t e m p e r a t u r e s c a n n o t be far f r o m reality a n d their i n f l u e n c e o n t h e t r a n s i e n t c a n n o t be m e a n i n g f u l . T h e v a l u e s o f t h e c o m p o n e n t t e m p e r a t u r e s u s e d in t h e c a l c u l a t i o n s a r e s h o w n in table 4.

4.5. Residual power T h e residual p o w e r u s e d in t h e c a l c u l a t i o n s [3] w a s evaluated assuming the hypothesis of reactor operations listed o n table 5. T h r e e m a i n h e a t g e n e r a t i o n f u n c t i o n s were a s s u m e d ,

Table 5 Data for residual power evaluating Neutron wall loading ( M W / m 2) Years of operation (a) Availabifity (%) Integrated wall loading (MW a / m 2)

1 1 100 1

w h i c h were related compo-nents: - A I S I - 3 1 6 SS of - copper of - Pb-17Li of

to

the

following

materials

IFW and OFW, IFW, t h e first row o f BB.

Table 6 Component attenuation factors Component

AISI-316L

Copper

Pb-17Li

IFW

0.60

1.0

-

OFW BOX

Plasma side Back plate

1.00 0.04

BB 1

0.43

1.00

BB 2

0.28

0.70

BB 3

0.14

0.54

BB 4

0.09

0.40

BB 5

0.05

0.30

W

Inboard Outboard

0.03 0.01

and

V. Renda, A. Soria, L, Papa / Thermaleffects of residualpower These residual power transients are shown in fig. 13 for the three materials. For all other components the heat generation is also known from monodimensional neutronic calculation. In the note these powers are indicated as fractions of the main curves for the attenuation factors listed in table 6. N o heat generation was assumed in graphite and water.

5. I n d e p e n d e n t

cooling circuits

Up to now there is no design stating the number and the independence of the circuits. A very attractive feature could be in the meantime to combine design with safety and maintenance. Cooling circuits can be chosen based on the following guidelines. - minimization of the independent circuits, - optimisation of the heat transfer between facing components due to radiation effect, - compliance with maintenance. Following these guidelines, five independent cooling circuits were chosen. - Two circuits for the inboard movable part of the reactor and related to the IFW and to the ISB. Two independent circuits are needed because the IFW contains the divestor target with special local thermal problems and because the ISB contains highly tritiated water. - Two circuits for the outboard movable part of the reactor and related to the O F W and the BB. The need for two circuits is obvious because many types of BB could take place in N E T for experimental reasons while the 48 boxes are connected to one circuit only. - One circuit for the semipermanent vacuum vessel, both inboard and outboard. With this choice, components linked to independent cooling circuits radially face each other, offering the maximum surface for radiation. When maintenance is being performed, only one circuit will be disconnected at a time. In that case the uncooled component can radiate on the facing cooled ones so that, in principle, no duplication of the circuit is necessary simplifying considerably remote-handled operations. The independent circuits are shown in the detailed model of fig. 12.

275

6. A c c i d e n t s c e n a r i o

The global model set-up and the choice of the independent cooling circuit allow a rational description of the accident scenarios, so that the results presented can be considered a considerable improvement of the preliminary ones of ref. [4]. At present only Loss Of Flow Accidents ( L O F A ) will be studied. The thermal transients from the steady state were evaluated assuming adiabatic boundary conditions, as the reactor is separated from the cryostat by means of insulating material. The following accident scenarios were studied. - L O F A on a single independent circuit and automatic plasma shut-down due to the control command system. Five cases were assessed corresponding to the five independent circuits. The temperatures of the working circuits were assumed constant as the heat transferred to the circuits is negligeable compared to the nominal heat normally removed by the circuits. The assumption is pessimistic; in fact a temperature decrease could be envisaged and programmed. - L O F A in the whole reactor due to a loss of electric power supply and inherent plasma shut-down due to loss of magnetic field. A third real but low probability scenario can be envisaged corresponding to a L O F A on a single circuit and non-intervention of the control command system. In that case a delayed plasma shut-down must be envisaged due to plasma impurities coming from a partial vaporization of the first wall. This scenario will be studied in the future when the conditions for this form of inherent plasma shut-down will be assessed.

7. R e s u l t s

In the following only the most meaningful results will be presented.

7.1. Loss of flow in the inboard first wall (IFW) Fig.14A shows the temperature distribution in space for different times of the transient; a temperature rise to 370 ° C in about 6 min has been assessed, see fig. 14B, in the water of the faulty circuit.

7.2. Loss of flow in the inboard shielding blanket (ISB) As shown in fig. 15A and fig. 15B the transient evaluation is similar to the previous one. The maximum

V. Renda, A. Soria, L Papa / Thermal effects of residual power

276

1500-

G

1500

Oh

1300-

130o1

1100-

1100

"

900-

~ ~

900

700-_

1/2 h

500-

2h

300100-

0 h

1 h

~; uJ

700

~-

5ooi

_j

1

h

2h .....

j

300

L ~_

3 12h

lOOi 2 POSITION (METERS)

POSITION (METERS)

A) SPATIAL DISTRIBUTON

A) SPATIAL DISTRIBUTION

800 •

380370-

23-24

360350-

600.

340 • t.u I.-

500.

330320 •

400-

310 •

300',, O0

8 ,

,

,

0 1 02

03

,

i

,

,

0 4 05 06 07 TIME (HOURS)

,

,

=

08

09

10

300- ~ a - - - - ~ ' ~ 010 011 012 013

B) TIME TRANSIENT

014 015 016 0.'7 TIME (HOURS)

018

019

B) TIME TRANSIENT

Fig. 14. LOFA in the IFW.

1500-

.

Fig. 16. LOFA in the O F W / O B P .

Oh

1500 •

1300~

13oo1

1100 ~

11001

900~

0 h

8001 1/2 h

w

700 ~

7001 1

500 ~ 300 ~

100~

h

...... 2.~ ..... ~.~-=-,,-~T~-

_/

3

1/2

k-

5001

•. ' ~ - - . - - ' - - ' ~ "

- --

h

1 h .....

k_

300 •

12h

100. POSITION (METERS)

J

2/h

3-

12h

POSITION (METERS)

A) SPATIAL DISTRIBUTION

A) SPATIAL DISTRIBUTION 700 •

o35°t!'

600" G

33oti 9

500 •

320 400" 31oll

~

- -

f~'--a-:..-a..-_..-_..:_..__.__..T_..

0.0

0.1

0.2

03

0 4 0 5 0.6 0 7 TIME (HOURS)

B) TIME TRANSIENT

Fig. 15. LOFA in the ISB.

_

08

09

'

10

300-

2'0

~o

TIME (HOURS) B) TIME TRANSIENT

Fig. 17. LOFA in the BB.

110

V. Renda, A. Soria, L Papa / Thermal effects of residual power water temperature in the faulty circuit, 325 ° C , is reached in about 1 min. 7.3. Loss of flow in the outboard first wall ( O F W / OBP)

G

:°°t

---"1

277

L

2,*.

48. L

The evaluation is given in fig. 16A and fig. 16B; the water temperature in the first wall reaches 7 3 0 ° C in about 25 min. The same temperature is reached by the stainless steel. '

5

6

9

POSITION (METERS)

7.4. Loss of flow in the breeding blanket (BB)

A) SPATIAL DISTRIBUTION

The space configuration shown in fig. 17A stresses the differences between this and the other cases. The most loaded blanket elements are those of the second row where a temperature of 650 ° C is reached in nine hours as shown in fig. 17B.

13°°t 11001 v tu

7.5. Loss of flow in the vacuum vessel ( I V V / OVV)

9oo~. 7007 5001

3001

As shown in fig. 18A and fig. 18B the transient is slow on the vacuum vessel.

25 - 29

t~ ;/

~....,.~,.~"~.~'~

12-is 68 ~ -

j

. . . .

.............

. .....

_

_ I

.."- ' " ' " ~ 3 ~ . , ~ " h , . / ' ' / -

1001

1'0 2'0 3'0 .b

sb

6'0 7'0 ~o 9'0 1;o

TIME (HOURS) B) TIME TRANSIENT

Fig. 19. LOFA in the whole reactor. 1500

0h

-

1300-

7.6. Loss of flow in the whole reactor

1100. 900"

w v-

The evolution of the phenomenon is shown in fig. 19A where the spatial distribution is plotted for different times. As shown in fig. 19B, after a certain time the maximum temperature is reached in the first row of the breeding blanket where about 9 0 0 ° C have been assessed at about one day. The behaviour of the inboard part of the reactor is similar, fig. 19B, but temperatures of the order of 720 ° C are reached in one day. Because of the adiabatic conditions imposed at the reactor boundary, the temperatures rise in the long term and the structures could melt after about 10 days.

700500-

112 h 1 h

300 ,j2

-

3_12 ,

100POSITION (METERS) A) SPATIAL DISTRIBUTION 400-

6 - 70

~

~--

300 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 -3-4

200.

8. Analysis of the results

100.~ o

"

73

2'0

lb

TIME (HOURS) B) TIME TRANSIENT

Fig. 18. LOFA in the I W / O W .

-

75

~o

In the analysis, the only natural cooling mechanism taken into account is radiation from faulty components to normally cooled ones. To this pessimistic hypothesis excluding every self cooling capability of the faulty

278

K Renda, A. Soria, L Papa / Thermal effects ofresidualpower

circuit, a pessimistic value of the radiation coefficient must also be added. In general the time-scale of the p h e n o m e n o n is of the order of minutes when the plasma faced components are concerned, while it can be of many hours in the case of the breeding blanket and vacuum vessel; the time scale becomes of many days for a L O F A in the whole reactor. The heat stored in the graphite on the inboard and outboard first wall is the most i m p o r t a n t cause of the fast transients on the walls themselves. In the case of a L O F A in a simple independent circuit, the temperatures always remain far below the melting point of the structural materials, both stainless steel and copper, while in each case water boiling in the faulty water channel c a n n o t be avoided. In the case of loss of flow in the whole reactor, the temperatures rise in all components; the transient is very slow and the calculations show that all the residual power c a n n o t be absorbed by the thermal capacity of the reactor and the melting of structures must be expected after many days. In reality the situation is not so pessimistic because the energy released in the first few minutes can be removed by the flow due to water inertia in the circuits and, in the long term, by a thermosyphon effect provided that an appropriate design is carried out. In principle in all cases, but in the total LOFA, just the inertia of the water and of the p u m p i n g system would probably be enough to avoid water boiling. The results also allow some preliminary considerations about the importance of the residual power on maintenance to be formulated. The cases of L O F A in the I F W and the case of L O F A in the ISB show that the temperature in the faulty channel remain low thanks to conduction to the working one; so that it is possible to disconnect one of the two circuits and connect it to the maintenance machine without problems. The situation is more complicated for the O F W for which only radiation can be used to remove the heat coming from the graphite coverage. Taking into account the water inertia, the delay in the maintenance operations after shut-down and the possibility of decreasing the temperature in the BB and particularly in the IFW, in principle the O F W can also be maintained. More.specific calculations are necessary, but there is a high probability that duplication of the circuits is not necessary for reasons of maintenance. This result, if confirmed in the future, greatly simplifies the design and the remote handling of plasma facing components.

9. Conclusions As part of N E T safety analysis, a preliminary inherently safe oriented design of the reactor is underway at JRC-Ispra. The aim would be to integrate safety targets, process and maintenance strategies in the main conceptual design. This approach is at present applied to the problem of the removal of decay heat from the plasma facing c o m p o n e n t s in cases of loss of flow accidents. A m i n i m u m n u m b e r of i n d e p e n d e n t cooling circuits has been chosen in such a way as to optimize the heat transfer by radiation taking account of the structure of the reactor (movable and p e r m a n e n t parts), the quality of the water (highly or lowly tritiated) and the possibility of cooling the c o m p o n e n t u n d e r m a i n t e n a n c e only by radiation or conduction on the nearest working circuit. The first step of the study, presented in the note, is the analysis of the thermal transient from the steady state after a shut-down in the reactor. The cases analysed up to now are related to a L O F A on one single independent circuit and plasma shut-down due to the intervention of the control c o m m a n d system, and to a L O F A in the whole reactor, followed by an inherent plasma shut-down, due to loss of power supply. No emergency power supply (diesels) has been envisaged. Preliminary pessimistic calculations have shown that reactor melting could occur after m a n y days (about 10), only in the case of total L O F A while in the other cases problems could arise because of water boiling and structural deformation of the plasma facing components. In these cases no major safety problems should occur, but the maintainability and operability of the reactor could not be ensured. The results show that just a little i m p r o v e m e n t in the heat removal could allow the classification of those accidents as upset conditions for which no specific intervention on the plant is mandatory. The improvement in the cooling capability could easily be obtained by designing the circuits to allow a thermosyphon effect. A preliminary engineering design of the I F W cooling circuit is underway at JRC-Ispra and the preliminary results seem to indicate that water boiling and structural deformations can be avoided by a simple arrangement of the circuit for natural cooling.

V. Renda, A. Soria, L. Papa / Thermal effects of residual power

References [1] C. Ponti and S. Stramaccia, Problems raised by the neutron activation products in a fusion reactor, IAEA - Environmental and Safety Aspects of Fusion Technical Committee, 17-21 October 1983. [2] A. Soria, Program TRECGA: A radiation heat transrnis-

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sion modal model to analyze thermal transients, CECJRC-Ispra T.N. No. 1.87.142. [3] C. Ponti, Activation calculations for NET-DN, CEC-JRCIspra T.N. No. 1.87.116 (October 1987). [4] V. Renda and L. Papa, Thermal effects due to residual power on the reactor internals, CEC-JRC-Ispra T.N. No. 1.87.50 (April 1987).