Melt Pool Behavior and Coolability in the Lower Head of a Light Water Reactor- Progress in WP5-2 of SARNET2

Weimin Ma

Division of Nuclear Power Safety

Royal Institute of Technology (KTH)

The 5th European Review Meeting on Severe Accident Research (ERMSAR-2012)Cologne, Germany, March 21-23, 2012

KTH: W.M. Ma, W. Villanueva, P. Kudinov; KIT: A. Miassoedov, X. Gaus-Liu RUB: M. Hoffmann; PSI: B. Jaeckel, J. Birchley; IVS: P. Matejovic, M. Barnak

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WP5-2 of SARNET2:- Remelting of debris, melt pool formation and coolability (MPF)

This topical area of research is concerned with corium behaviour and coolability in the lower head.

The objectives are:

• To create and enhance the database on the corium simulant behavior in the lower head.

• To develop and validate the models and computational codes for simulation of in-vessel melt pool and vessel behavior.

• To assess the influence of potential SAM measures (e.g. CRGT cooling and ex-vessel cooling) on in-vessel corium coolability and retention.

The Scope

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Dual approach: Experimental and analytical.

Experimental work • The LIVE program (KIT) is to provide further experiments with

emphasis on the pool build-up and behavior in a well-scaled lower head using corium simulant.

• Experiments at the CNU facility (CEA) dedicated to study two-phase flow (and CHF) around a heated hemispherical geometry of a PWR. The data will be unique for the external vessel cooling in conditions more representative for the reactor case in 3D geometry (1:10 angular sector).

Approach

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Analytical work

Approach (contd.)

•USTUTT (IKE) – Stuttgart University, Germany•INRNE – Institute for Nuclear Research and Nuclear Energy, Bulgaria•TUS – Technical University of Sofia, Bulgaria

•IVS – Inzinierska Vypoctova Spolocnost, Slovakia•PSI – Paul Scherrer Institut, Switzerland•RUB – Ruhr-Universität Bochum, Germany

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Paper 2.1: Progress in Research on Melt Pool Behavior and Coolability in the Lower Head of a Light Water Reactor.

• Some results of LIVE-2D tests

• Calculations of fuel degradation tests on the ACRR.

• Thermo-mechanical analysis of BWR vessel failure under CRGT cooling

• Thermal load analysis related to the IVR of VVER440

Paper 2.5: Analysis and Interpretation of the LIVE-L6 Experiment

• To investigate the thermal-hydraulic behavior of a stratified melt pool in the reactor vessel lower head.

Paper 2.13: The Experimental Results of LIVE-L8B – Debris Melting Process in a Simulated PWR Lower Head.

Contents

LIVE-2D Test Facility

LIVE-2D– Study of both transient and steady-state behavior of the 2D

molten pool

– 12 cm thick slice, Ø1 m, wall thickness 25 mm

– NaNO3-KNO3 melts (Tmax ~ 370 °C)

– Phase diagrams similar to corium

– Supporting tests to compare with SIMECO and LIVE-3D

Main investigated thermal-hydraulic parameters:– Melt temperature distribution

– Heat flux distribution through the wall

– Heat transfer from the upper layer to the vessel wall

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TestCooling conditions

Time

(hh:mm)

Power

(W)Remarks

LidVessel Wall

L00A1 Water Water

0-27:15 7316

Power is decreasing with each step

27:15-54:27

6275

54:27-96:15

4206

L01 Water Water

0-30:15 7311

Order of Power levels:High, low, high

30:15-70:50

3691

70:50-99:15

7316

L02 Water Water

0-28:45 3694

Order of Power levels:Low, high, low

28:45-71:15

7315

71:15-99:15

3706

L03_01 Insulated Water0-50:13 2145

Cooling right from the start50:13-95:13

1112

L03_A Insulated Water0-50:25 2150

Cooling of vessel wall only with delay after pouring in

50:25-95:25

1118

L04 Insulated Water 0-46:28 2155At beginning of test 50%

debris already in the vessel

LIVE-2D Tests

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Results of LIVE-2D (1)

Aver

age

upw

ard

Nus

selt

num

ber

Internal Rayleigh number

LIVE-2D

Comparison of the upward heat transfer in LIVE-2D tests with earlier experiments

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Results of LIVE-2D (2)

Comparison of the upward and downward heat transfer in LIVE-2D tests with SEMICO experiments

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Analysis of Tests on ACRR

ACCR: Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL).

DF-4 was an in-pile experiment to investigate early-phase degradation of BWR fuel canisters and control blades.

MP-2 test performed was to investigate late-phase melt procession in LWRs.

DF-4 test was simulated by different versions of MELCOR, and the results for the degradation were essentially similar, but MELCOR 1.8.5 has kinetic models for eutectic interactions.

MP-2 test was simulated by the ATHLET-CD code with and without late phase module MEWA, and the results show that the code with the MEWA module has better prediction.

Control Rod Guide Tube (CRGT) cooling is being considered as a potenetial SAM measure for Nordic BWRs.

Given a thermal and mechanical load on the lower head, what is the mode of failure of the vessel wall?

Is in-vessel retention possible by means of combined CRGT cooling and ex-vessel cooling?

Analysis of Vessel Failure Modes

Temperature

H = 0.7 m DB

H = 1.9 m DB

t = 4.44 h

t = 3.47 h

Different amounts of melt relocated to the lower-head have been considered, corresponding to 0.7m, 1.1 m, 1.5m and 1.9 m deep pool.

PECM model was used to predict heat transfer and phase changes in the melt.

Snapshots of Temperature

ANSYS 13 specifics• Full transient analyses• Strong structural-thermal coupling• Plane223: 2D 8-node coupled-field element• 850 quadrilateral elements and 2901 nodes

Creep Analysis in ANSYS

0.7 m deep melt pool

The maximum displacement is located at the bottom of the RPV

t = 5.0 h

‘Ballooning’ Mode of Vessel Failure

t = 4.94 h

1.1 m deep melt poo1

1.5 m deep melt pool

‘Localized’ Mode of Vessel Failure

1.9 m deep melt poo1

t = 3.89 h t = 3.61 h

• Creep intensifies in one location (localized creep).• Displacement of the lower part of the vessel is relatively uniform.

CHF limit

• External cooling reduces the heat fluxes on the outer wall

H = 1.9 m

• Simulations taking into account the external vessel cooling on the melt heat transfer.

• Cooling of the vessel from outside effectively reduces temperature on the inner side of the vessel wall.

Effect of External Cooling

Melt Pool Convection in VVER-440

In-vessel retention of corium in the lower head with external cooling is considered for VVER-440.

The molten pool in the lower head was simulated by the stand-alone DIVA module of ASTEC V1.3 rev.2 code

Boundary condition – constant heat transfer coefficient and coolant temperature – was applied on the outer surface in order to simulate the external reactor vessel cooling.

Assumptions of base case:100% of UO2, 50 tons of stainless steel (SS) and 19 tons of zirconium were relocated into the lower head.

Sensitivity Study (1)

Effect of molten pool configuration on the thermal load.

Effect of stainless steel mass on the thermal load.

Sensitivity Study (2)

Effect of decay heat distribution between oxidic and metallic layers on the thermal load.

Sensitivity Study (3)

Concluding Remarks

Progress on melt pool and lower head behaviour related to WP5-2 of SARNET2 was reported.

The LIVE-2D experiment shows a good agreement for downward heat transfer with SIMECO, but upward heat transfer is considerably higher than in the SIMECO.

The MELCOR analysis of the ACRR DF-4 test indicates only minor oxidation of B4C because of the lack of any treatment in MELCOR for oxidation of the molten mixture.

The ACRR MP-2 test was better simulated by the ATHLET-CD code with the MEWA module.

The coupled thermo-mechanical creep analysis of a BWR lower head reveals that if only CRGT cooling is activated, the cases with 0.7 m and 1.1 m melt pool depths result in a ballooning type of vessel failure while the cases with 1.5 m and 1.9 m melt pool depths result in a localized creep. If external vessel cooling plus CRGT cooling is implemented, in-vessel retention of the melt is possible.

Heat transfer of molten pool in the lower head of a VVER-440 reactor was studied by the ASTEC V1.3 code. Sensitivity study shows for various molten pool configuration, stainless steel mass in molten pool and decay heat distribution, the heat flux is well below the cooling limit (CHF).