Challenges in Two-phase Flow and Reactor

Safety Research after Fukushima Accident

Mamoru Ishii

Walter H. Zinn Distinguished Professor

School of Nuclear Engineering

Purdue University

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Outline

• Status of Two-phase Flow Modeling and Reactor Safety

Analysis

Design Basis Accident (DBA)

Beyond DBA

Severe Accident

• Lesson Learned from Fukushima I Accident – Reactor Safety

• Future Direction of Reactor Safety Research

• Direction of Two-phase Flow Research – Challenges

• Direction of Two-phase Flow Research – Need

• Cause of Inaccuracy of Current Two-phase Flow Model

• One Approach to Improve Models Significantly

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Status of Two-phase Flow Modeling and Reactor Safety Analysis

Design Basis Accident

• For Current LWR (PWR & BWR)

Large Break LOCA: Predictive Capability Well Developed

Phenomena Understanding

Data Base and Scale-Up Capability

Safety Analysis code (Best Estimate)

Small Break LOCA: Some Challenges

Prolonged Transients

Natural Circulation Dominated Flow

Water Level Switches in Coolant Flow

Variety of Scenarios and Boundary Conditions

Difficult to diagnose Plant Conditions

Much Smaller Body of Data Basis

However, Large Safety Margin and Time to Manage

Other Accidents (LOHS & LOF) are covered

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Status of Two-phase Flow Modeling and Reactor Safety Analysis

Beyond Design Basis Accident

Multiple Failure Scenarios

(DBA + Failure of Safety Systems)

As Extrapolation of DBS: Relatively Well Covered

Prolonged Station Blackout (SBO)

Common Mode Failure of Active Safety Systems

Phenomena much Different from DBS

Importance of Operator Actions

Multiple Options Possible

Coupling of Reactor & Containment

Loss of Ultimate Heat Sink (Fukushima Type)

Similar to SBO

Not Extensively Studied and Accident Phenomenology Not Well

Understood

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Status of Two-phase Flow Modeling and Reactor Safety Analysis

Severe Accident

Beyond Design Basis Accidents has Higher Probability of

Severe Accident Consequence

Mitigation of Severe Accidents is Key for Reactor Safety

rather than Severe Accident Analysis itself

(Fukushima I Accident Lesson)

Once Reactor is Uncovered and Uncoolable, Severe

Accident Phenomena follows

(Relatively well understood)

Several Computer Codes exist based on Phenomenological

Models

(This may be sufficient)

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Lesson Learned from Fukushima I Accident

Reactor Safety

(1) Implementation of True Defense in Depth Accident Mitigation Strategy is

necessary

Defense against Earthquake and Tsunami was quite insufficient in terms of

Heat Sink Protection

Emergency AC Power Protection

Avoidance of Common Mode Failures of Systems

However there were for Severe Accident Mitigation

Retrofit Reactor Coolant Injection System

Retrofit Automatic Containment Depressurization System

Severe Accident Mitigation Strategy was not known or Personnel were not

trained,

No Effective Defense in Depth Approach against Prolonged Station Blackout &

Loss of Heat Sink due to

Defective Safety Regulation (Slow to change)

Inaction at TEPCO even High Risk from Tsunami became known

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Lesson Learned from Fukushima I Accident

Reactor Safety

(2) Development of Severe Accident Mitigation Strategy & Training

Understanding of Accident Phenomenology

Focus of Mitigation Strategy

Reactor Cold Shut-down

Reactor Depressurization

Sufficient Core Coolant Inventory

Containment Pressure Regulation and Venting

(3) Improvement of Emergency Reactor Instrumentation Systems

Development of Autonomous Critical Reactor Diagnostic System for Beyond

DBS

Independent Power

Autonomous and Possibly Wireless

Critical Instruments (Protected)

Reactor Water Inventory

Reactor and Containment Pressure & Temperature

Heat Sink Status

Key Valve Configuration (Safety and Non-Safety)

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Future Direction of Reactor Safety Research

(1) Much More Emphasis on Beyond Design Basis Accident

Multiple Failure Scenarios

Prolonged Station Black-out

Loss of Ultimate Heat Sink

Objective of Research (1)

Find out Best Accident Mitigation Strategy (Use of Defense in Depth 2nd & 3rd Options )

Find out Limit Envelop of Protection based on Existing Systems (Hours and Days)

Margin of Errors and Options

Objective of Research (2)

At the Limit Envelop: What Additional Systems?

1. Retrofit Passive Safety System: Coolant Injection System, Containment Cooling and

Depressurization System

2. Off-site Capability Establishment (When and How)

New Reactor Design Challenges

Generation III+ , Small Modular Reactors , Other Type of New Reactors

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Direction of Two-phase Flow Research

Challenges

Accurate Prediction of Beyond DBS

Multiple Failure of Safety Systems

Prolonged Station Black-Out

Loss of Ultimate Heat Sink

Variety of Reactor Designs; Generation III, Generation III+, SMR

Two-phase Flow Phenomena dominated by Natural Circulation

Large Number of Two-phase Flow Phenomena and Conditions

Multiple Passes and Scenarios and Significant Effect of Operator Actions

Limited Resources for Research

Conventional Research Approach: Inefficient & Expensive

(Each Scenario, Design Test Code Tuning)

Need More General & Reliable Approach applicable to Wide Ranges of Conditions

Adhoc Semi-empirical Model should be changed to Mechanistic Model

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Direction of Two-phase Flow Research

Need

Accurate Two-phase Flow Models applicable to Variety of Phenomena,

Conditions and Geometry particularly under Natural Circulation

Conditions without Extensive Separate and Integral Tests and Code

Benchmark (Beyond DBS, SBO, LUHS, etc)

Accurate Model for Coupled Behaviors of Reactor System and

Containment

Model capable of Simulating Non-Safety Grade System Interactions

Shortcomings of Current Models and Safety Analysis Codes

Transient Specific

Geometry Specific

Model and Code tuned by adjusting them relative to Integral and Separate

Effect Tests

Models and Safety Codes work not because mechanistically correct

Extrapolation to Beyond DBS, SBO, LUHS or New Reactor Design may

not be accurate

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Cause of Inaccuracy of Current Two-phase Flow Models

Use of Very Inaccurate Flow Regime Map

Use of Regime Dependent Constitutive Relations or Correlations

Two-phase Flow Regimes highly dependent on Geometry and Boundary

Conditions

Existence of Global (Common 1-D) Flow Structures and Local Flow Structures

(One Model cannot fit both)

Almost All Complicated Two-phase Flow Phenomena dominated by Interface

Transfer of

Momentum

Mass and Energy

Interface Transfer dominated by Interface Structure

Most Two-phase Flow Phenomena can be modeled, if Interface Structures can

be predicted (Bottle Neck of Modeling Capability)

Currently Reactor specific, Transient specific, Geometry specific Adhoc Models

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

One Approach to Improve Models Significantly

(Introduction of Interfacial Area Transport Equation)

Introduction of Interfacial Area and its Transport Equations in Two-Fluid Model

Use of Two-group Interfacial Area Transport Equations

Interface Geometrical Variables

Two-phase flow Interfacial Structure Characterized by 4 Geometric Variables,

α1, α2, a1 and a2

Changes of Interfacial Area Modeled dynamically by Interfacial Area Transport

Equation

Interfacial Transfer of Momentum and Energy modeled separately for Group I

and II

Much more Mechanistic than Current Approach

Interface Area 1 1

Mixture Volume Interface Length Scalei

i

aL

Group II Volumetric Fraction α2

Interfacial Area Concentration a2

Group I Volumetric Fraction α1

Interfacial Area Concentration a1

Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University

Summary

Status of Two-phase Flow Modeling and Reactor Safety Research

Lesson Learned from Fukushima I Accident

Future Direction of Reactor Safety Research

Direction of Two-phase Flow Research

Cause of Inaccuracy of Current Models

One Approach to Improve Models Significantly