Large spreading of core melt for melt retention—stabilization

ELSEVIER

Nuclear Engineering and Design 157 (1995) 447 454

Nuclear Engineering and Design

Large spreading of core melt for melt retention-stabilization H.A. Weissh/iupl, D. Bittermann Siemens K W U NA-M, PO Box 32 20, 91050 Erlangen, German),

Abstract

For future nuclear power plants possible core melt scenarios have to be taken into account in the plant layout. Different solutions for core melt retention are proposed in the literature. The core melt retention devices should as far as possible be simple in construction, to minimize the deepness of sophistication of the physical problems connected with it and to minimize unforeseen technological problems. Furthermore it should be in good compliance with normal operational needs. One very promising solution seems to be a large enough spreading of the melt with water cooling from above. This is the favoured solution for the development of the French-German European pressurized water reactor. The technical concept and the problems to be expected in the course of an anticipated core melt accident (to be overcome in an actual design) are shown. The needs for analytical and experimental support in the early conceptual phase and later on in the design stage of plant construction are discussed. Typical analytical and first experimental results for the large core melt spreading are presented.

1. Introduction

For future nuclear power plants (NPPs) core melt scenarios have to be taken into account in the design. In the case of a reactor pressure vessel (RPV) melt through, measures have to be foreseen to stabilize the core melt within the containment in order to maintain its functions. Different concepts for core melt retention are proposed: melt retention in the RPV itself due to outside cooling, and covering a broad range of compact crucible solutions with one- or two-sided cooling; surface increasing solutions, where the melt is distributed within or over special arrangements; and solutions where water is penetrating the core melt in an adequate manner.

The core melt retention device should as far as possible be simple in construction, to minimize the broadness and deepness of the physical and technological problems connected with it. Furthermore it should be in good compliance with normal operational needs and should not run into unforeseen constructional problems during a detailed design phase. One very promising solution seems to be a large enough spreading of the melt with water cooling from above. This is the favoured concept for the development of the F r e n c h - G e r m a n European pressurized water reactor (EPR) (Bonhomme, 1993; Weisshfiupl, 1994). In designing such a solution one has to look not only at the melt retention and stabilization capability itself but also at the

0029-5493/95/$09.50 © 1995 Elsevier Science S.A. All rights reserved S S D I 0029-5493(95)01001-7

H.A. Weisshiiupl, D. Bittermann / Nuclear Engineering and Design 157 (1995) 447 454

448

boundary conditions arising from the course of the accident. Research and development (R&D) work is already needed for the early conceptual phase. Ongoing R & D has to show the compliance of the chosen design.

2.

Technical

concept

interaction - - passive means for heat removal from the melt use of passive heat capacity of the containment structures to ensure an extensive delay time until containment heat removal system operation is required. Additional features to cope with potential phenomena during the course of the accident are -the reactor pit lower part is designed in order to

The basic concept for the core melt retention and stabilization proposed for the EPR is the spreading of the melt on a large area outside the reactor pit (Fig. 1). The spreading compartment characteristics are as follows: dedicated spreading area of about 150 m 2 bottom and lateral structures of the compartment have protection layers designed for thermal and mechanical loads the spreading area is dry during normal operation the spreading c o m p a r t m e n t is connected with the in-containment refuelling water storage tank (IRWST) with pipes for water flooding after spreading; these pipes are closed during normal operation and accident conditions by plugs which will be melted by the corium after spreading between the spreading compartment and the above containment compartment, an open flow area exists for steam escape (the steam condenses in the containment and the condensate flows back to the IRWST). The overall goal for the chosen concept is to stabilize the melt by water cooling within the containment in order to avoid -basemat melt through -intolerable thermal loads on or damage of structures with the potential consequence of loss of containment function - - a long-term source for fission products Additional goals are no melt interaction with the basemat concrete to avoid - - additional hydrogen production - - a d d i t i o n a l pressure increase caused by production of non-condensable gases - - no large energy release caused by fuel coolant

minimize the fall height of a failed RPV bottom and the resulting forces and --- reduce the potential for water collection beneath the RPV bottom by constructional means to reduce the possibility and consequences of an ex-vessel steam explosion in case of a loss-of-coolant accident (LOCA) - - reactor pit and spreading compartment are connected via a melt discharge channel, which has a slope to the spreading area and is closed by a steel plate; this steel plate (possibly covered with refractory material) resists for a certain time melt through in order to accumulate the melt in the pit. Potential energetic fuel-coolant interactions during the spreading process itself and by flooding of the melt are avoided and respectively minimized by --- having an initially dry spreading compartment, where eventually only a very shallow water layer can form as a consequence of condensing steam in the case of a L O C A allowing only limited flooding rates in the order of about 50 to 100 kg s J. Owing to the outside arrangement of the spreading area a separation of short-term mechanical and thermal loads caused by the RPV failure and the long-term thermal loads caused by the spreaded melt is achieved. A spreading area of about 150 m 2 is selected to ensure on the one hand the complete spreading of the melt and on the other hand the coolability of the melt by water from above without having to take into account a partial quenching of the melt, i.e. heat transfer by conduction alone would be sufficient. The protection layers (e.g. zirconia bricks) have a high melting point and a low heat conductivity to protect the underlying concrete

H.A. Weisshdupl, D. Bittermann / Nuclear Engineering and Design 157 (1995) 447 454

,

PRT

Fig. 1. EPR layout for spreading and stabilization of core melt.

449

450

H.A. We&shgiupl, D. Bittermann / Nuclear Engineering and Design 157 (1995) 447- 454

Table 1 R&D work for the large area spreading of melt Phenomena

Priority

Boundary conditions (1) Melt down scenario (2) RPV failure mode

B B

Spreading behaviour (1) Dry spreading (2) Presence of some water (3) Flow through orifices (4) Melt through of steel plate

B A A A

Fuel coolant interaction (1) Jets into water (2) Spreading with (some) water present (3) Flooding of melt

Analytical (a) experimental (e)

Main aim of investigation

a

Composition of melt Pouring hi~tory into pit Temperature

a, e

Spreaded area Influence of layering Spreading time

a, e

Potential loadings Steam production Solidification behaviour (fragmentation)

e

Long term heating up of concrete (Partial) quenching

e

Thermal mechanical chemical behaviour Gassing of concrete Detachment of protection layer

a, e

Contribution to source term

B A B

Cooling of the melt (I) Water above

B

Structure behaviour (1) Protection layers (2) Concrete

A B

Fission products

B

A, high; B, medium. from m o l t e n core concrete i n t e r a c t i o n ( M C C I ) a n d melting.

3. Analytical and experimental needs T o s u p p o r t the large a r e a s p r e a d i n g concept investigations have to be p e r f o r m e d with respect to - - b o u n d a r y c o n d i t i o n s (Plank, 1994) (melt d o w n b e h a v i o u r , R P V failure m o d e ) - - s p r e a d i n g b e h a v i o u r ( s p r e a d i n g ability, steel plate melt t h r o u g h , flow t h r o u g h orifices) - - m e l t - c o o l a n t i n t e r a c t i o n (melt ejection after R P V failure, ingression o f melt in s p r e a d i n g area, flooding o f melt) - - c o o l i n g o f melt with w a t e r f r o m a b o v e - - structural b e h a v i o u r - - fission p r o d u c t b e h a v i o u r T h e different p h e n o m e n a to be addressed, the m a i n aims for investigations, the analytical a n d / o r e x p e r i m e n t a l a p p r o a c h , a n d the p r i o r i t y for invest i g a t i o n s are listed in T a b l e 1.

4. Analytical and experimental results To investigate the feasibility o f the large a r e a s p r e a d i n g m e a n s l o o k i n g at different key p h e n o m ena d u r i n g an a s s u m e d accident progression. In the first stage o f investigation, slightly conservative, simplified analytical tools have been used for a few representative scenarios with a b r o a d p a r a m e t e r variation. In the second stage, m o r e s o p h i s t i c a t e d c o m p u t e r p r o g r a m s covering a variety o f possible accident p r o g r e s s i o n s (subsequently c o m b i n e d in a limited n u m b e r o f accident sequences), s u p p o r t e d by selected s e p a r a t e effect experiments, are used to d e m o n s t r a t e the adeq u a c y o f the chosen design. In a third stage, a d d i t i o n a l analytical investigations a n d a b r o a d e r e x p e r i m e n t a l verification c o n c e r n i n g a d d r e s s e d p h e n o m e n a shall p r o v e that the goals envisaged will be sufficiently fulfilled. In the following some selected e x a m p l e s o f analytical a n d e x p e r i m e n t a l results following the accident progression are presented.

H.A. Weisshgiupl, D. Bittermann /Nuclear Engineering and Design 157 (1995) 447 454

350

2O/o]

451

19,50 Yol

300 250 145

200

[ ] met. m a s s 85

75

150

IIIII1[1111 IIIIIlillll

125

50

i1,l l3lol l il !!!!!!!!!!!H

7O

100

[hi o x i d . m a s s

105

85

Ill

0 CM1

CM2

i

CM3

CM4

Coremelt

CM5

CM6

CM7

Group

Fig. 2. Melt propagation scenarios: core melt groups immediately after PRV failure (percentages are estimated).

Core degradation leading to slumping has been analyzed using the MAAP 3.0 code and partly SCDAP/RELAP. The time period from slumping to RPV failure has been addressed by special investigations. Fig. 2 shows the result of a variety of calculations performed with respect to the amount of core material, molten at the time of RPV failure. One can see that in most cases about 155 to 200 tons of core melt are to be expected, whereas only in a small number of cases less and even more (additional mass of core internals, shrouding etc.) has to be reckoned with. The relative probabilities depicted in Fig. 2 are based on probabilistic safety analyses of a typical 1 3 0 0 M W P W R plant (Konvoi-type), supplemented by accident progression event trees for the different phases of core degradation. Calculations for the spreading of the melt have been performed with the code CORFLOW (see Fig. 3). Flooding of the melt is achieved via pipes connected to the IRWST. Flooding progression for different bounding conditions (full quenching, i.e. assuming a perfect heat transfer from melt to water, to film boiling heat transfer) can be seen

from Fig. 4. With a flooding rate of 50 kg s ~ the time to cover the spreaded melt with a 10 cm water layer ranges between 2500 s and about 700 s. Long term temperature evolution is shown in Fig. 5 for the assumption that not even partial quenching will occur but only film boiling. The m a x i m u m interface temperature for the protection layer is 1900 °C, the m a x i m u m temperature of the basemat concrete is 900 °C. The dashed line at 350 °C indicates the region where the load bearing capability of the concrete is significantly decreased. Didurit is a high temperature resistant concrete with a melting point of about 1800 °C. For lay-out considerations the region of lessened load bearing capability and the release of gases (vapour) from the concrete have to be taken into account. Separate effect tests have been performed with an arrangement as depicted in Fig. 6. Test series with different surface conditions (concrete to a ceramic protection layer) and different amounts of water present (dry to 40 cm) using 30-50 kg of a l u m i n a - i r o n thermite of 2200°C have shown that even under water a sufficient spreading of melt is achieved, whereas care has to be taken

452

H.A. Weisshiiupl, D. Bittermann ~Nuclear Engineering and Design 157 (1995) 447-454

..."" j.. ....,-

......" """...

....

f...~

..j"

..,,

-....

....

.,...

..

"......

'....,.

.s'"

'"..,,.,... 1.

x3[ m ]

+_0. 0.

0.25

10.25

10.

Fig. 3. Simulation of melt spreading (oxidic melt, isothermal boundary conditions).

with respect to energetic fuel-coolant interactions especially if water is trapped below the melt (initially wet concrete).

5. Conclusions

For the EPR, core melt retention stabilization is proposed to be achieved by a large spreading of

melt on a dedicated spreading area of about 150 m 2, with high temperature resistant protection layers and flooding of the melt after spreading by passive means via connection pipes to the IRWST. The feasibility and adequacy of the chosen design were first investigated with simplified tools assuming enveloping boundary conditions. In a second stage, sophisticated analytical tools were

H.A. Weisshiiupl, D. Bittermann / Nuclear Engineering and Design 157 (1995) 447-454

453

4000 ~

3500

parameter: ~

3000

~

heat f ux to water

initial h e a t c a p a c i t y = 1 6 0 0 M W s / m 2

\

~

to°tsMq;7:h

\

w'o2

2500

2000 15oo 1000 500 0 20

I

I

I

I

I

I

I

30

40

50

60

70

80

90

100

flow rate(kg/s) F i g . 4. Flooding time to cover the s p ~ a d e d melt with a 10cm layer of water.

used covering a variety of possible accident progressions to give, for example, the relative fractions and overall amounts of oxidic and metallic melts that would be released from a failed RPV. To look at the feasibility of the proposed core melt retention means to start with the melt down

behaviour and to end with long-term temperature evolutions in the spreading area. Still more analytical and especially experimental efforts have to be undertaken to show the achievement of the envisaged goals for accident control (Kuczera, 1994).

2400 -

:;

•

TIME1: 9 ooo (s) ( 0.1 DAYS) TIME2: 1000C~(s) ( 1.2DAYS) TIME3: 250000(s} ( 2.9 DAYS) x TIME5: 750000(s) [ 87 DAYS) o TIME6: 10C~ 0CO{s) ( 11:6DAYS)

o

2000

BASE ~AT (C:ONCR "TE)

_

1600

DID

Z'

/

i

,2oo

AREA =150 m2 LAYEREDCOREMELT METALLIC CM ABOVE OXlDIC CM BELOW 0.0~5m/0.1~lm ,00

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

0 -2.8

350-

-

~ -2.4

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

~ -2.0

.

~

J -1.6

-1.2

-

J

<

-(~.8

-~ -0.4

.

0.0

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

.

PROTECTION LAYERS ZrO2 ABOVE DIDURIT 135 BELOW

.

0.4

HEIGHT (m)

Fig. 5. Temperature profiles--base case.

08

112

COREMELT-WATER HTC=800 W/m2 K

454

H.A. Weisshiiupl, D. Bittermann Nuclear Engineering and Design 157 (1995) 447-454 References steel plate (lOmm+Smm)

catcher

therm~e ~,

,

isolation

l

concrete

moulding sand

romp

ceramic tiles

\

/ \ / \

¢,

V

/

Fig. 6. Test facility 3-2.

steel crucible

N. Bonhomme and U. Krugmann, Safety and containment of the European pressurized water reactor (EPR), 5th Int. Seminar on Containment of Nuclear Reactors, Karlsruhe, Germany, August 1993. B. Kuczera, W. Eglin and H. Weissh~upl, Cooperative R&D work on LWR servere accident phenomena, Proc. Int. Topical Meet. on Advanced Reactor Safety (ARS '94), Pittsburgh, PA, April 1994. H. Plank and H. Weissh/iupl, Ermittlung der Randbedingungen ffir eine Kernschmelzrfickhalteeinrichtung, EnergieTechnik Umwelt/Festschrift, Ruhr-Universit~t Bochum, September 1994, pp. 299-303. H.A. Weissh/iupl, Preventive and mitigative measures for the European pressurized water reactor (EPR) for severe accidents with core melt down, Jahrestagung Kerntechnik '94, Stuttgart, Germany, May 1994.