Development of the Methodologies for Evaluating Severe Accident

Management

Lim Joong Taek

Department of human resource development team

December 13, 2016

IAEA Technical Meeting on the Verification and Validation of Severe Accident Management Guidelines

December 12-14, 2016 IAEA Headquarters, Vienna, Austria

Contents

Introduction

Theoretical Background

Methodology

Preliminary Results

Concluding Remarks & Future Work

2

Introduction

Nuclear Power Generation in Korea

25 Power Reactors Operation • 21 PWRs and 4 PHWRs

• Installed capacity of 23.1 GWe

• 30 ~ 35% share of electricity supply

3 PWRs(APR1400) under Construction

4 PWRs(APR1400) under Planning

The Most Economical, Reliable & Semi-domestic Electricity Source in Korea

Operation Construction Planning

Hanbit 6 units

Hanul 6 units

Shin-Hanul 4 units

Wolsong 4 units

Shin-Wolsung 2 units

Kori 4 units

Shin-Kori 6 units

4

Domestic severe accident legislation

Amendment of Nuclear safety act • Purpose

Severe accident management performed by administrative order with weak legal basis

Clarification of responsibility for accident management and regulatory requirements in Nuclear Safety Act

Even in case of a severe accident, to minimize the release of radioactive materials and to restore the NPP to a safe condition through accident management program

To introduce accident management program as a licensing document

• Application New NPP : AMP is required to obtain operating license

NPP in operation : AMP is required to submit within 3 years(June 23, 2019)

5

Purpose

The new nuclear safety act

Purpose • Evaluate the effectiveness of

the severe accident management strategies using probabilistic manner

• Combine PSA, AMP, and severe accident research

6

Accident Management

Program

Severe Accident Research

Integrated Methodology

Probabilistic Safety

Analysis

Theoretical Background

Integrated ROAAM • ROAAM + level 2 PSA

• PSA is a tool for determining relevant sequences that must be mitigated in the SAM approach

• ROAAM increases the credibility and transparency of the level 2 PSA study providing a framework for modeling and quantifying complex physical phenomena.

• Application Westinghouse AP600 : passive LWR design

SAMEM • Performed the development and

demonstration of integrated models for the evaluation of severe accident management strategies using PSA and ROAAM

8

Background

Probabilistic framework for containment failure due to hydrogen combustion

- Application of the risk oriented accident analysis methodology (ROAAM)to severe accident management in the AP600 advanced light water reactor : J.H.Scobel, et al(1998) - The development and demonstration of integrated models for the evaluation of sever accident management strategies-SAMEM : M.L. Ang(2001)

Methodology

10

Level 1 PSA Level 2 PSA

Fission Product

Noble Gas

I

Cs

Te

Sb

Sr

Ru

Ba

La

Ce

Source Term Category

STC1 (SGTR)

STC2 (ISLOCA)

STC3 (NOTISOCS)

STC4 (NOTISONOCS)

STC5 (CFBRB)

STC6 (MELTSTOP)

STC7 (NOCF)

STC8 (BMT)

STC9 (ECF1)

STC10 (ECF2)

STC11 (ECF3)

STC12 (ECF4)

STC13 (LCF1)

STC14 (LCF2)

STC15 (LCF3)

STC16 (LCF4)

STC17 (LCF5)

Containment Failure Mode

BYPASS

NOTISO

NOCF

BMT

ECF

LCF

Severe Accident

Phenomena

Molten Core- Concrete

Interaction

In-Vessel Steam Explosion

Ex-Vessel Steam Explosion

Hydrogen Generation &

Reaction

High Pressure Melt Ejection

Direct Containment

Heating

Containment Failure Mitigation

CIS Containment Isolation before Core Damage

SDR Sudden Depressurization

for mitigation

RACV Power Recovery before Reactor Vessel Failure

RACC Power Recovery before

Containment Failure

INJ Coolant Injection to Reactor Vessel

CFS Reactor Cavity Flooding

HMSI Hydrogen Control System

Operation

CHR Containment Heat Removal

Initiating Event

LOCA

Large LOCA

Medium LOCA

Small LOCA

SGTR

ISLOCA

RVR

T r an s i en t

LSSB-IN/OUT

LOFW

LOCV

LOIA

PLOCCW

TLOCCW

LOKV(4.16kV)

LODC(125V)

LOOP

SBO

GTRN

ATWS

Core Damage Prevention

EVENT Tree

Core Damage Frequency

Containment Failure Probability

Plant Damage State Frequency

Fission Product Release Frequency

Korean Level 2 PSA Process

Database the level 2 PSA information

Transfer to database • Accident progress

Core Damage Prevention function

Severe accident mitigation function

• PDS group of Accident Sequences

• Screening PDS ET Containment failure sequence

Frequency : Below 10-14 /RY

• Success/failure of the functions Success : (AAC)

Failure : AAC

11

Plant Damage State Event Tree

12

Database the level 2 PSA information

Source term release fraction • The mass of the initial fission

product in a reactor core and the mass fraction of the fission product released into the environment

• Each PDS has its own 17 STC quantification values and the probabilities by containment failure modes

• 10 fission products Noble gas, I, Cs, Te, Sb, Sr, Ru,

Ba, La, Ce

Source Term Category Release fraction

13

Initiating Event Core Damage

Prevention Functions Severe Accident

Mitigation Function Source Term

Database the level 2 PSA information

Preliminary application

APR1400 • Thermal power : 4,000 MWth

• Hot-leg temperature

: 615 ℉(323.9 ℃)

• Severe accident respond facilities Cavity flooding system

Passive auto-catalytic recombiner

ECSBS

Selection of initiating event • Based on CDF contribution

14

Initial Event

Co

re D

amag

e Fr

equ

en

cy SBO

SLOCA

Risk information to display

Severe Accident Prevention / Mitigation • Core Damage Frequency

• Conditional Containment Failure Probability

Containment failure / Source term release • Containment Failure Frequency

• Source Term Release Fraction

15

𝐶𝐹𝐹 = 𝐹𝑅𝐸𝑄 × 𝐶𝐹𝑃

𝐶𝐶𝐹𝑃 =𝐶𝐹𝐹

𝐶𝐷𝐹=

𝐹𝑅𝐸𝑄

𝐶𝐷𝐹 × 𝐶𝐹𝑃

* FREQ : Frequency of PDS ET end state

Preliminary Results

Database of the risk information

17

STATION

BLACKOU

T

AAC

UNAVAIL

ABLE

DELIVER

AUX.

FEEDWATER

(TD PUMPS)

RCP

SEAL

FAILURE

RECOVER

AC

POWER

(IN

1HRS)

DEL.

AFW

AND

REM.

STEAM

(MSSV)

DEL.

AFW

AND

REM.

STEAM

(ADV)

RECOVER

AC

POWER

LATE

REFUEL

AAC

STORAGE

TANK

SHUT

DOWN

COOLING

MAINTAIN

SECONDARY

HEAT

REMOVAL

RECOVER

AC

POWER

AND

REFUELING

AAC

SAFETY

DEPRESSU

RIZATION

SAFETY

INJECTION

FOR

FEED

(2/4)

CONTAIN

MENT

HEAT

REMOVAL

(COOL

IRWST)

CONTAIN

MENT

HEAT

REMOVAL

(CS

SPRAY)

CONTAIN

MENT

ISOLATIO

N

SYSTEM

RAPID

DEPRESSU

RIZATION

RECOVER

AC

POWER

PRIOR

TO

VESSEL

RECOVER

AC

POWER

PRIOR

TO

CONTAIN

MENT

IN-VESSEL

INJECTIO

N

CAVITY

FLOODIN

G

SYSTEM

HYDROGEN

CONTROL

SYSTEM

IGNITORS

CONTAIN

MENT

HEAT

REMOVA

L

FREQ

(Accident

Sequence

Frequency

)

CLASS

(PDS No.)I Cs

83 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - (MSHR) RACL-2 - - - - 3.39E-10 (CIS) - (RACV) - (INJ) (CFS) (HMSI) (CHR1) 1.31E-10 113 3.57E-02 1.21E-11 2.01E-03 9.68E-04

85 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - (MSHR) RACL-2 - - - - 3.39E-10 (CIS) - (RACV) - (INJ) (CFS) HMSI (CHR1) 3.54E-11 113 9.61E-03 3.26E-12 2.01E-03 9.68E-04

134 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - MSHR - - - - - 6.38E-10 (CIS) SDR (RACV) - (INJ) (CFS) (HMSI) (CHR1) 6.27E-11 113 9.04E-03 5.77E-12 2.01E-03 9.68E-04

136 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - MSHR - - - - - 6.38E-10 (CIS) SDR (RACV) - (INJ) (CFS) HMSI (CHR1) 1.40E-11 113 2.01E-03 1.28E-12 2.01E-03 9.68E-04

261 SBO AAC (AFW-T) (RSF) - - - RACL-1 - - - - - - - - 6.23E-07 (CIS) - (RACV) - (INJ) - (HMSI) (CHR1) 3.16E-07 113 4.67E-02 2.91E-08 2.01E-03 9.68E-04

263 SBO AAC (AFW-T) (RSF) - - - RACL-1 - - - - - - - - 6.23E-07 (CIS) - (RACV) - (INJ) - HMSI (CHR1) 1.15E-07 113 1.70E-02 1.06E-08 2.01E-03 9.68E-04

418 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR (RACV) - (INJ) (CFS) (HMSI) (CHR1) 5.66E-10 113 7.57E-03 5.21E-11 2.01E-03 9.68E-04

420 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR (RACV) - (INJ) (CFS) HMSI (CHR1) 1.46E-10 113 1.95E-03 1.34E-11 2.01E-03 9.68E-04

434 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR RACV (RACC) (INJ) (CFS) (HMSI) (CHR1) 2.77E-10 113 3.70E-03 2.54E-11 2.01E-03 9.68E-04

436 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR RACV (RACC) (INJ) (CFS) HMSI (CHR1) 6.21E-11 113 8.31E-04 5.72E-12 2.01E-03 9.68E-04

CCFP

(=CFF/CDF)

CFF

(=FREQ*CFP)

Source TermContainment Failure MitigationCore Damage Prevention

No.

CDF

(Level1

Accident

Sequence

CDF)

CDF

Frequency of PDS ET end state

PDS Group

CCFP

CFF

Distribution chart

18

1.00E-12

1.00E-11

1.00E-10

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00

Co

re D

amag

e Fr

equ

ency

Conditional Containment Failrue Probability

SLOCA Accident Sequence

CDF Vs CCFP • Core Damage Frequency

• Conditional Containment Failure Probability

PDS 18

PDS 19

PDS 40PDS 41

PDS 76

PDS 77

PDS 92

PDS 93

PDS 4

PDS 18

PDS 19

PDS 40

PDS 41PDS 76

PDS 77

PDS 92

PDS 93

PDS 4

1.00E-12

1.00E-11

1.00E-10

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00

Co

re D

am

ag

e F

req

ue

ncy

Conditional Containment Failrue Probability

CHR1 Failure in SLOCA

19

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E-16 1.00E-15 1.00E-14 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07

방

출

분

율

(

S

T)

원자로건물손상빈도 (CFF)Containment Failure Frequency

Sou

rce

Term

Rel

ease

Fra

ctio

n

1.00E-12

1.00E-11

1.00E-10

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-050

0.001

0.002

0.003

0.004

0.005

0.006

0.007

Co

re D

am

ag

e F

req

ue

ncy

So

urc

e T

erm

Re

lea

se F

ract

ion

Containment Failure Probability

CCFP-CDF-ST

Source term release fraction • CFF Vs RF (2D)

• CCFP Vs CDF Vs RF (3D)

Distribution chart

Concluding Remarks & Future Work

Insight from this works

Effective evaluation for SAMP • Legislation

• Verify the impact of mitigation success and failure

Prioritization of severe accident management strategies • Selection of the most important mitigation functions

• Application Improvement of the equipment

Modification of Technical specification

Personnel training

21

Further Work

Concept for evaluating severe accident

Methodology for DET/CET branch probability calculation

Uncertainty analysis of CET using computational code

Level 2 PSA re-quantification • The new index for measuring the effectiveness of AMP

22

Thank you!!

23