Modeling of PARAMETER-SF4 experiment with SOCRAT/V1 code · 6 20 years Project (since 1999). Versions V1, V3 One-through coupled modeling of SA at LWR reactors Validated on experimental

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Modeling of PARAMETER-SF4 experiment

with SOCRAT/V1 code

Vienna, Austria, 2-5 September 2019

K.S. Dolganov, A.E. Kiselev, D.Yu. Tomashchik, T.A. Yudina

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The experiment SF4 was conducted with the aim to study the

model of VVER-1000 FA under simulated conditions of a

severe accident including the stages of air ingress and

further quenching by the bottom flooding with water, namely:

❑study of the behavior of FA structural components (fuel

pellets and claddings, spacer grids);

❑study of the oxidation degree of FA structural components;

❑study of interaction and structural-phase changes in the

materials of FA model (fuel pellets and claddings);

❑study of the hydrogen release.

Aim of PARAMETER-SF4 experiment

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PARAMETER test facility

1 – power supply; 2 – coolant outlet; 3 – condensate

drain; 4 – vessel flange; 5 – fuel rod; 6 – grid; 7,15 –

shroud; 8,16 – thermal insulation; 9 – bellows; 10 –

bottom heater; 11 – quench water; 12 – air ingress; 13,24

– cooling water; 14 – top cooling; 17 – argon inlet; 18 –

steel tube; 19 – body; 20 – tantalum heater; 21 – fuel

pellet; 22 – cladding; 23 – vessel; 25 – unheated rod; 26

– steam inlet; 28 – peripheral rods.

“Protocol of PARAMETER-SF4 Experiment Results”, ISTC

Project#3690 “Study of fuel assemblies under severe accident top

quenching conditions in the PARAMETER-SF test series”, FSUE

SRI SIA “LUCH”, 2009.

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Experimental Scenario

-preparation and heating up stage (0 – 8000 s) –

stabilization of coolant parameters: argon flow

rate ( 2 g/s, inlet temperature 530°C), steam

flow rate ( 3.5 g/s, inlet temperature 280°C),

pressure (~ 0.48 MPa at 1475 mm);

-pre-oxidation stage (8000 – 13886 s) – FA

holdup at temperature of 1200оС in the hottest

zone (1250 -1300 mm) for ∼ 6000 s. Electrical

power was 7500 to 8500 W

-cool down stage (13886 – 16035 s) – maximal

FA temperature at the end of the stage 600оС.

Electrical power was 4000 W;

-air ingress stage (16035 – 17500 s) with

subsequent (from 16355 s to 17434 s) assembly

heating-up to FA maximum temperature of ∼1750оС in the hottest zone;

- bottom flooding stage (~17500 – 17900 s) –

flooding of the bundle from bottom.

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Test facility modeling

The main areas of SOCRAT/V1

application include the estimates of

hydrogen source to the containment,

and source of corium (mass, energy

and composition) from reactor

pressure vessel after its melt-

through.

SOCRAT/V1 has a modular

structure. Each module contains the

realistic models of a separate set of

physical processes, and interaction

between different processes is

assured by coupling of modules

through the common interface

standards. SOCRAT/V1 contains the

data base of thermo-physical

properties of different materials.

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❑ 20 years Project (since 1999). Versions V1, V3

❑ One-through coupled modeling of SA at LWR reactors

❑ Validated on experimental data

System Of Codes for Realistic AssessmenT of severe accidents

SOCRAT description (1)

The code is essentially developed to model VVER NPPs but it

was also applied to simulate the severe accidents at integral-

type light water reactors, BWRs (Fukushima Daiichi) and

PWR reactors (TMI-2).

Bolshov, L.A., Dolganov, K.S., Kiselev A.E., Strizhov V.F., Results of SOCRAT code

development, validation and applications for NPP safety assessment under severe

accidents, Nuclear Engineering and Design, Volume 341, 2019, Pages 326-345.

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* Reprinted from Nuclear Engineering and Design 341 (2019) 326–345

Copyright Elsevier 2019

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Primary and secondary thermal

hydraulics (RATEG module)

1D approach:Phase mass 2

Phase enthalpy 2

Momentum 2

Noncondensable mass N_gas

2D approach:Momentum +2

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( ) ( ),// rrrftC j

Oj −=

( )0=

r

vr j

jjj

j

O CvFTtrf +=),,(

( ) rCTDF j

j

Oj /−=

1−jFjF jC

j

ju

1−jC

1−j

1−ju

+ BC:

UO2 α-Zr(O)+(U,Zr) (U,Zr) α-Zr(O) β-Zr α-Zr(O) ZrO2

UO2 α-Zr(O)+(U,Zr) (U,Zr) α-Zr(O) α-Zr(O) ZrO2

UO2 α-Zr(O)+(U,Zr) (U,Zr) ZrO2

UO2 α-Zr(O)+(U,Zr) (U,Zr)O2

UO2 (U,Zr)O2

Core degradation (SVECHA module)

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T > 2300 K & e < 0.2 mm or T > 2500 K

U-Zr-O: 2250…2850 K

( ) v c

dVU F F Vg

dt = − − +

( )d

V h Udt

= −

( )( )s s

dV h C T T Q M W

dt + − = − − +

Free candling or

candling in gaps

Core degradation (SVECHA module)

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SOCRAT Validation extent (1)

❑PBF SFD 1-1, 1-4, ST

❑QUENCH (-3, 6, 10, 12, 13)

❑PARAMETER-SF (1, 2, 3, 4)

❑CORA (-7, 13, 15, 17, 31)

❑CORA (-W1, W2)

❑PHEBUS B9+

❑PHEBUS FPT1

❑FARO, MAGICO, QUEOS

❑RASPLAW-AW

❑FZK (Leistikow & Schanz, Hofmann)

❑ORNL (Pawel et al.)

❑NIIAR

❑AEKI

❑REBEKA

❑QUENCH-LOCA (-0)

❑VIAM, VNIINM, OKB GP, NRC KI

❑AECL (Hayward), NIIAR

❑Kim-Olander

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SOCRAT Validation extent (2)

❑ PARAMETER-SF (ISTC) Core degradation with HA reflood /2005-2010/

❑ OSU MASLWR (IAEA–OSU) Long term cooling* /2010-2012/

❑ BSAF (OECD NEA) IC (unit 1) as BC** /2012-2014/

❑ ATLAS (OECD NEA) PAFS* simulation /2016/

❑ FUMAC HALDEN (IAEA) Spray (HA) cooling and FP transport /2017/

❑ PKL4 (OECD NEA) Passive heat removal by SG /2018/

❑ PERSEO Passive heat removal system* /2018/

* Special system connected to primary or secondary side

** No design data

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SOCRAT Application (1)

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SOCRAT vs ORIGEN: < 20 % in FP

Except for Kr85M, Pm148, Pm148M, Cs135: 20…35 %

Difference in actinides: < 20 %

Except for Am241, Cm242, Cm244: 25…45 %

Cross-validation of FP inventory in Fukushima-1 U1 core

SOCRAT Application (2)

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Reference data for ST during first 34 hours (from earthquake): solution of

inverse problem for atmospheric transport with account for weather

JAEA, GRS, IRSN: reconstructed source term based on field measurements

(dose rates, volumetric and surface activity, activities ratios)

Nuclide Activity, Bq Relative difference, %SOCRAT/V3 WSPEEDI GRS/

WSPEEDISOCRAT/V3 vsGRS/WSPEEDI

SOCRAT/V3 vs WSPEEDI

WSPEEDI vsGRS/WSPEEDI

Xe-135 1.07E+17 – 1.57E+17 32 – –Cs-137 7.87E+14 1.40E+15 7.29E+14 8 44 48I-131 8.00E+15 1.40E+16 1.18E+16 -32 -43 16

135Хе 131I

SOCRAT Application (3) : ST from Fukushima-1 U1*

* http://www.gidropress.podolsk.ru/files/proceedings/mntk2019/autorun/article059-en.htm

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SOCRAT Application (4) : PERSEO facility

PERSEO is a full scale facility

aimed at studying a new passive

decay heat removal system

operating in natural circulation.

It was built at SIET laboratories in

Piacenza (Italy) modifying the

existing PANTHERS IC-PCC

facility.

* The authors express their gratitude to ENEA for distributing the

PERSEO facility and Test 7 description and the Test 7 experimental data

along the OECD/NEA/CSNI/WGAMA activity on the “Status report on

thermal-hydraulic passive systems design and safety assessment”.

*

*

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Initial and Boundary condition for SF4 test

Total electrical power history and coolant pressure in test bundle

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Bundle thermal response before flooding stage (1)

Rod temperature at elevations 200, 300 and 400 mm

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Rod temperature at elevations 600 and 900 mm

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Bundle degradation and hydrogen production (1)

Oxygen concentration at bundle outlet Temperature profile in the second row rods

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Rod temperature at elevations 200, 300 and 400 mm

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Total hydrogen production

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Post-test bundle examination

Post-test bundle examination demonstrates that at the

elevation 130 mm the assembly does not have any visible

damages (all fuel rod claddings and periphery rods kept

their integrity). The oxide scale on fuel rod claddings is thin

and well attached to metal; the measured thickness of

zirconium oxide scale varies within 4 – 9 μm. SOCRAT/V1

code predicts the thickness of zirconia to be 6 – 7 μm in

heated rods at this elevation.

Ignatiev, D., 2009. Post-test examination of the PARAMETER-SF4

fuel assembly. In: Proceedings of 15th International QUENCH

Workshop, 3–5 November 2009, Karlsruhe, Germany. ISBN 978-3-

923704-71-2.

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Conclusion

The air ingress test PARAMETER-SF4 was performed in

July 2009 with the aim to investigate the behavior of a

VVER fuel assembly during water flooding after air ingress

phase. The scenario and conditions of the experiment are

prototypical for a severe accident with fuel rod uncover,

degradation and reflooding in a spent fuel pool. SF4

experiment has been simulated with the severe accident

code SOCRAT/V1.

The results of the simulations show a reasonable

agreement of calculations with measurements of

temperature histories, outlet oxygen concentration and

hydrogen production. Insignificant differences were found in

melt progression and maximal temperatures in the bundle.

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Thank you for [email protected]

Documents

Modeling of PARAMETER-SF4 experiment with SOCRAT/V1 code · 6 20 years Project (since 1999). Versions V1, V3 One-through coupled modeling of SA at LWR reactors Validated on experimental