conference on severe accident research … - validation... · [email protected] ... a large...

May16-18, 2017

Warsaw, Poland

8TH CONFERENCE

ON SEVERE ACCIDENT

RESEARCH

ERMSAR 2017

Validation Progress and

Exploratory Analyses of Three-

Dimensional Simulation for

BWR In-vessel Core

Degradation

Shigeki Shiba

Regulatory Standard and Research Department,

Secretariat of Nuclear Regulation Authority (S/NRA/R)

[email protected].

ERMSAR 2017, Warsaw, May 16-18, 2017

Contents

1.2. Objective & Targets

3. Model System

4. Validation Results

5. Exploratory Analyses

6. Summary

2


Backgrounds

Severe accidents at TEPCO’s Fukushima

Daiichi Nuclear Power Station Units 1–3

in March 11, 2011.

Taking the lessons learned from the

severe accidents into account, S/NRA/R

has started development of a detailed

model (called “Multifunction Model”) of

in-vessel core degradation for BWRs as a

fundamental safety research.

3


1. Backgrounds

2.3. Model System



6. Summary

4

Contents


Objective & Targets

To develop the Multifunction Model for BWR in-

vessel core degradation

Calculation Zones:

Calculation Stages:

Required Functions:

5

i. Core, ii. Region of core support plate, iii. Lower

and upper plenums.

i. Tight coupling calculation of neutronic, thermal-

hydraulic and fuel pin behavior,

ii. Direct modeling in three-dimensional geometry,

iii. Precise treatment of multiple eutectic reactions.

From a normal operational to severe accident

conditions.


1. Backgrounds

2. Objective & Targets

3.4. Validation Results


6. Summary

6

Contents


Model System

7

Three modules communicate each other through the

“control module.”

CONTROL

Module

Fuel Pin Behavior

Module, based on

FEMAXI-6

Normal operation

Transient

Fuel pin failure

behavior

Thermal-Hydraulic Module

Multi-phase, component & velocity fields

Vaporization/Condensation, Melting/ Freezing

Neutronic

Module, based on

SKETCH

Distribution of

neutron flux

Eigenvalue

Reactivity


1. Backgrounds


3. Model System

4.5. Exploratory Analyses

6. Summary

8

Contents


Validation with Existing Severe Accident Experiments

9

Severe accident tests related to BWR core degradation

are limited in the world.

Name Main Parameters Purpose

CORA

Facility： KfK

Period： 1987-1992

Burn-up： Un-irradiated

Number of fuel pins： 25-59

Fuel length： 1.0m

Control rod： Ag-In-Cd/B4C

Maximum temperature： 2300~2500K

To observe

initial core

degradation

process

XR-2

Facility： SNL

Period： 1993-1996

Burn-up： Un-irradiated

Number of fuel pins： 64

Fuel length： 1.0m

Control rod： B4C

Maximum temperature： ~2300K

To observe melt

flow near core

support plate

FARO/

KROTOS

Facility： JRC

Period： 1991-1999

For FARO

-High pressure condition.

-Average diameter of particle: 3.2－4.8mm.

For KROTOS

-Low pressure condition than that of the FARO.

-Average particle diameter： 0.18－2.5mm.

To observe

molten corium

interaction with

water pool


Absorber

blade

Validation with CORA Experiment: Calculation

Nodalization

10

CORA test facility

(Main components)

CORA bundle

arrangement

Axial cross-section

of calculation model

Horizontal cross-section of

calculation model

Absorber blade

Channel box

Heated rod

Zry shroud

Channel box

Unheated rod

Axial cross-section

Absorber blade

Zircaloy shroud

Insulation: ZrO2 fiber

SS-blade

Heated

Unheated

B4C powder absorber

Channel box

Horizontal cross-section


Validation with

CORA

Experiment:

Moving Image

of Simulation

11


(a) Relocation

of Zry-Inconel

Eutectics

(b) Relocation

of B4C-SS

Eutectics

(c) Relocation

of B4C-SS-Zr

Eutectics

(d) Molten Corium

on the Bottom of

Bundles

Unheated rod

Absorber blade (B4C-SS)

Channel box

Grid spacers

Heated rod

Horizontal cross-section of

experimental resultsCalculation results for axial relocation in CORA-18 experiment.

At 254mm elevation.

At 545mm elevation.

At 1158mm elevation.

12

Validation with CORA Experiment: Axial- & Lateral-Directional Material Relocation According to simulation results, the Multifunction model could simulate the axial and

lateral relocation.

At 1158 mm elevation

At 545 mm elevation

At 254 mm elevation

Grid spacers

Heated rod

Absorber blade

(B4C-SS)

Calculation results for axial relocation in CORA-18 experiment. Horizontal cross-section

of experimental results


Calculation

target Fuel rods

X

Y

Z

1 2 3 4 5 6

1

2

3

4

5

6

Horizontal view of test

section.

Horizontal cross-

section of calculation model.

Control blade

Z

Y

Axial cross-section of calculation model.

XR - 2 test facility.

13

Validation with XR-2 Experiment: Calculation

Nodalization

Catcher

box

Velocity limiter

Inlet orifice

Nosepiece

Control

blade gap

Core support

plate

Y-Z cross

section


Location

Volume of relocated materials (liter)

Experimental result Calculation result

Material found below the core

support plate 7.06 6.08

Inlet nozzle 0.96 0.29

On velocity limiter 3.2 3.13

Catcher box 2.9 2.66

Material above the core support

plate 1.7 2.15

Above core support plate 0.62 0.52

In nosepieces 0.77 1.24

Control blade gap 0.31 0.40

Total

8.76 8.23

After melt relocation, a large amount of particle Zr piled up on the bottom

catcher.

Calculated results reproduced to the experimental results, in spite of the very

complicated structure.

14

Validation with XR-2 Experiment: Calculation Results


15

Validation with FARO L-19 and KROTOS K-37: Molten Corium Breakup with Water Particle diameters were simulated well by the calculations in the whole range of

mass fraction.

Some differences would imply that the sensitivity study of adjustment parameters

necessary to decide the particle diameter is required.


1. Backgrounds


3. Model System


5.6. Summary

16

Contents


Exploratory Analyses: Calculation Conditions

and Initial Parameters

17

Calculation Conditions:

Initial Parameters:

i. Geometry: ¼ Sector-Core.

ii. Simulation time: 250s.

To quantify sensitivities of initial parameters for the BWR core

degradation phenomena, the following parameters were

selected.

i. Initial Water Level (BAF*/TAF)

ii. Decay Heat (10% of thermal output*/Way-Wigner equation)

iii. Radial Power shape (Cosine*/ Homogenous)

iv. Oxidation of fuel cladding (Depleted*/Fresh fuel loading)

*Reference case


18

XZ cross-section

2127.5 mm 2352.5 mm

Control rod

drive housing

Dryer

Separator

Shroud head

Top of active fuel

Bottom of active

fuel Control rod guide

tube

Upper head

Bottom of RPV

18,8

60

mm

(3

0 c

ells

)

3 cells

6 cells

8 cells

4 cells

1 cells

1 cells

2 cells

1 cells

4 cells

Upper head

Dryer

Separator

Shroud head

Top of active fuel

Bottom of active fuel

Control rod guide tube

Control rod drive housing

Bottom of

RPV

XY cross-section

18

86

0 m

m

i. Water level: BAF

ii. Simulation time: 250s

iii.Decay heat level: 10%

iv. Pressure: 7 MPa

Exploratory Analyses: Simulation Model of ¼

Sector-Core Geometry

Horizontal cross-section :

7x7 cells


19

Exploratory

Analyses: Moving

Image of

Simulation


20

Exploratory Analyses: Results for Middle-

Power-Level BWR In-Vessel Core Degradation

(a) Distribution of

material volume fraction

(b) Pressure

distribution

(c) Structural

temperature

distribution

(d) Molten corium

temperature

distribution

7.25 MPa

7.20 MPa

7.15 MPa

7.10 MPa

7.05 MPa

7.00 MPa

3300 K

2700 K

2100 K

1500 K

900 K

300 K

3300 K

2700 K

2100 K

1500 K

900 K

300 K


Exploratory Analyses: Calculation Results from

Sensitivity Analyses

21

Variables Reference

case

Initial

water level

(TAF)

Decay

Heat

(Way-Wigner

equation)

Radial

Power

shape

(Homogenous)

Oxidation

of fuel

cladding

(Fresh fuel

loading)

Timing of fuel

failure [s] 131 255* 586* 243 102

Fuel failure mode Pellet

melting

Pellet

melting

Pellet

melting

Pellet

melting

Cladding

melting

Hydrogen

production [kg] 13.7 2.3 1.6 16.7 12.6

Weight of corium

on the bottom of

lower head [kg]

7,300 760 0 7,200 13,200

* The calculations were continued up to the fuel failures to evaluate the fuel failure

mode.


22

Sensitivity Analyses of Radial Power Shape: Temperature

Distributions & Core Degradation Appearances

The reference (a) :

Radially cosine-

power-shape.

The other case (b) :

Simplified

homogenous radial

power shape.

The time to fuel

failure of the

reference (a) was

fast due to the

higher radial

peaking factor.

3300 K

2700 K

2100 K

1500 K

900 K

300 K

3300 K

2700 K

2100 K

1500 K

900 K

300 K

Cosine-power-shape

Simplified homogenous power shape


1. Backgrounds


3. Model System



6.

23

Contents


Summary

24

The Multifunction Model was validated through the analyses

of the existing severe accident experiments, CORA-18, XR-2,

FARO L-19 and KROTOS K-37.

Exploratory analyses regarding initial conditions of BWR In-

vessel core degradation were executed and initial

parameters such as water level etc. were significantly

sensitive for BWR core degradation progression.

As further assignments of the model, we found that

adjustment factors in FARO L-19 and KROTOS K-37 must

be adjusted to enhance more accuracy of the particle

diameter calculations,

and thermal-Hydraulic Module in the model must be verified

due to no crust formation expected around the core support.


Annex

25


26

Item Content Value

Thermal output - ≈1500 MWth

RPV dimensions Height

Diameter

19 m

5 m

Core dimensions Height

Diameter

4 m

3.5 m

Fuel Assembly

Fuel lattice

Average enrichment

Average burn-up

9×9

3.7 wt%

45 GWd/t

Fuel pin

Fuel stack length

Plenum gas pressure

Gap width

Cladding material

Cladding outer diameter

Cladding thickness

Oxide layer thickness at

45GWd/t

3.71 m

1.0 MPa

0.0002 m

Zircaloy-2

0.011 m

0.0007 m

10 μm

Fuel pellet

Material

Diameter

Density

UO2

0.0094 m

97 %T.D.

Initial weight of core

components

Fuel: Uranium/Zircaloy

Control rod: B4C/Steel

18525 kg/3997 kg

220 kg/2396 kg

Exploratory Analyses: Simulation Model of ¼

Sector-Core Geometry


27

Sensitivity Analyses for Initial Water Levels

Timing of fuel failure were considerably sensitive to parameters.

Especially, cooling and decay heat conditions must be considered

in order to mitigate core degradation.


28

Sensitivity analyses for the BWR In-Vessel

Core Degradation: Fuel failure mode

Almost all of sensitivity parameters was not sensitive to fuel

failure mode.

Oxidation of fuel cladding must be considered.


29

Sensitivity analyses for the BWR In-Vessel

Core Degradation: Hydrogen production

In short simulation time, amount of Hydrogen production was

relatively small in comparison with results of conventional severe

accident codes.

Hydrogen production is slightly sensitive to these parameters.


30

Sensitivity analyses for the BWR In-Vessel Core

Degradation: Weight of corium on the bottom of lower

head In short simulation time, amount of corium on the bottom of

lower head was strongly sensitive to parameters

Especially, initial water level etc. leaded to mitigation of core

degradation.

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