IAEA International Conference on Topical Issues in … New core internals with many reactor control...

© 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved.

June 7, 2017

Development for

Nuclear Power Plant Safety

Overview of Technology Developments for

Continuous Improvements of Nuclear Safety

Tomofumi YAMAMOTO

Nuclear Energy Systems Division, Mitsubishi Heavy Industries, Ltd. Tokyo, Japan

IAEA International Conference on Topical Issues in

Nuclear Installation Safety: Safety Demonstration of Advanced Water Cooled Nuclear Power Plants (CN-251)

© 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 2

Contents

1. Introduction

2. Strengthening of Overall Nuclear Safety

3. Advances for Core Cooling Measure Using

SG Secondary-side Depressurization

4. Development of Seismic Isolation System

For Nuclear Facilities

5. Summary

1. Introduction


Current Status of NPP in Japan : PWR : BWR

Kashiwazaki Kariwa

Tomari

Higashidori

Onagawa

Fukushima Daiichi

Fukushima Daini

Tokai Daini

Hamaoka

Ikata

Sendai

Shika

Tsuruga

Mihama

Ohi

Shimane

Genkai

Takahama

1 2 3

1 2 3

1 2 3 4 5 6

1

Oma １

１

1 2 3 4 1 2 3

1

1 2

1 2 3 4 5 6 7

2

1 2 3

1 2 3

1 2 1 2 3 4

1 2 3 4

1 2 3 4

PWR BWR

Restarted 5 0

Approved by NRA 7 0

Under review by NRA 4 10

Preparing review 4 14

To be decommissioned 4 10

1 2 3 4 5


New Regulatory Requirements

Natural Phenomena

Fire

Reliability of Power Supply

Reliability of Other SSCs

Seismic / Tsunami Resistance

Intentional Aircraft Crash

Suppress Radioactive Materials Dispersion

Prevent CV Failure

Prevent Core Damage (multiple failure)

Internal Flooding

Natural Phenomena (volcano, tornado, forest fire)

Fire

Reliability of Power Supply

Reliability of Other SSCs

Seismic / Tsunami Resistance

Previous

SSCs: Structure, Systems and Components CV: Containment Vessel

Focus on prevention from

severe accidents

(i.e. Regulation based on the

design basis: Regulatory body

have confirmed that a single

failure would not lead to core

damage)

Reinforced

Reinforced

&

Newly

introduced

Newly

introduced

New

2. Strengthening of Overall Nuclear Safety


Developments based on the DiD

levels

of DiD Objective

Direction to strengthen

safety functions Technology developments

Level 1 Prevention of

abnormal

operation and

failures

Earthquake-resistance - Seismic isolation system

- Enhancement of seismic evaluation

method for steam generator

Level 2 Control of

abnormal

operation and

detection of

failures

Maintain subcriticality to

cold shutdown only with

control rods

- New core internals with many

reactor control clusters

- Enhancement of CFD analysis for

core internals

Level 3 Control of

accident within

the design basis

Diversity for reactor core

cooling

- Enhancement of core cooling

capability by steam generator

- Air cooling system/equipment

Level 4 Control of severe

plant conditions

Cooling of melting core - In-vessel retention for large reactor

According to the concept of Defence in Depth, the middle and long

term direction and the technology developments were surveyed.

3. Advances for Core Cooling Measure Using SG Secondary-side Depressurization


原子炉容器

制御棒駆動装置

タービン

復水タンク

大気

蒸気発生器

使用済燃料ピット

給水車消火栓

地面

主蒸気逃し弁

大気

～～～～

発電機

復水器

～～

～～タービン動補助給水ポンプ

加圧器

一次冷却材ポンプ

原子炉格納容器

Feed Water

Gravity-driven CRD

1. Shutdown

Large volume

containment confines

radio activity and

hydrogen under accident.

3. Confinement

4. Cooling (ＳＦＰ)

Water can be fed at the

ground level. 一次系自然循環

① Coolant is injected by turbine-driven pumps(Passive safety).

② SG secondary-side is cooled by the steam discharge via. MSRV.

③ Core cooling is achieved by the

primary-side natural-circulation.

2. Cooling

原子炉容器

①

②

③

Natural-circulation core cooling can be achieved by SG

secondary-side depressurization under SBO/Loss of UHS

Containment

PRZ

PV RCP

Natural circulation

SG

MSRV

Turbine- driven Pumps

Water tank

Turbine Generator

Spent Fuel Pit

Ground

Safety Measures under SBO/LUHS


Hot Leg

Cold Leg

Pressurizer

Accumulator

Low Pressure Injection Pump

Steam Generator

Small Break LOCA

Cross-over Leg

Core

Main Steam Relief Valve

Key lesson learned from Fukushima Dai-ichi Accident

To secure core cooling measures

Purpose

Purpose of this study

Development of SG cooling system and procedure as an

additional safety measure in order to secure the diversity of

core cooling measures.

An example of core cooling process by SG under SB-LOCA

Time

Prim

ary

Pre

ssure

Accumulator Injection Start

Small Break LOCA occur

SI signal

SG depressurization initiation

Reactor trip

Low-pressure injection start

Early activation of SG

main steam relief valve

(MSRV) is considered

after SI signal.

The accumulator and

low-pressure injection

system are started

earlier to fulfill the core

cooling requirement

(PCT:1473K).


Subjects of this study

• To show the validity of the core cooling with SG secondary-

side depressurization

• To develop an analytical method which can apply to actual

reactors

Actions to resolve the subjects

• To perform experiments using an appropriate test facility

which fulfills the requirements for scalability to actual

reactors

- Large Scale Test Facility (LSTF) at JAEA is used as

the test facility

• To validate an analytical method using the database

- M-RELAP5 is used as the analytical method

Subjects


Large Scale Test Facility (LSTF) • LSTF is full-height and full-pressure (approximately 15MPa) thermal-

hydraulic simulator of typical four-loop PWR.

Test parameters to be examined • Pipe break size（2in, 4in, 6in, 8in＊, 10in） • Loop-unbalance cooling

• Low pressure/Low power natural circulation

• Non-condensable gas in Accumulator

• SG secondary-side depressurization timing

＊Major results will be presented.

Demonstration

LSTF

Volume Scaled by 1/48 to the reference PWR

Height corresponds to the reference PWR

Nominal

pressure corresponds to the reference PWR

Loop Two loops (One broken/ One intact)


Demonstration of the core cooling under SB-LOCA

Experiment

the primary pressure decrease

along with the SG secondary-

side depressurization.

The depressurization

successfully achieves the

activation of accumulator and

low-pressure injection system.

M-RELAP5

Good predictions are obtained,

which mean that break flow

rate and the SG primary and

secondary-side thermal-

hydraulics are well simulated.

Depressurization

Primary and Secondary Pressure

Start of SG secondary-

side depressurization

Actuating level

of Accumulator


Peak Cladding Temperature (PCT)

Experiment

After the Accumulator

activation the cladding

temperature turns down and

the PCT fulfills the safety

requirement (1473K). M-RELAP5

The start point of core heat-

up is earlier and the PCT is

higher than the experiment.

M-RELAP5 gives a

conservative prediction for

the PCT and can be used as

a safety evaluation code for

the actual reactor.

PCT

Demonstration of the core cooling under SB-LOCA

1000

900

800

700

600

500

400

Tem

per

atu

re (

K)

5004003002001000

Time (s)

Experiment

M-RELAP5

Experiment

M-RELAP5

Activation timing

of Accumulator


Performance of SG Secondary-side Depressurization The core cooling safety measure works well under SBLOCA even if no

high-pressure injection system is activated.

Although this presentation shows only one typical case, the

depressurization system also worked well for other several parameters

(Break size, Loop-unbalance cooling, Non-condensable gas in

Accumulator, SG secondary-side depressurization timing).

Impact for Safety Advances The results of this project give the technical evidence that the AM

measure can be activated without any concerns for several

uncertainties like break size, loop-unbalance cooling, non-condensable

gas in accumulator, SG secondary-side depressurization timing.

This contributes to enhance the reliability of the AM measure and also

be useful to refine the time-margin for operator action in future.

Typical Results - Core Cooling Measure

4. Development of Seismic Isolation System for Nuclear Facilities


Table 4.1 Development process

Purpose of this study • To secure the integrity of reactor buildings against huge earthquakes in

the future

• To realize the standard design not to depend on site conditions.

Subjects of this study • Obtaining highly aseismic performance by installing base-isolation

• Grasping the ultimate strength of isolator based on the full-scale

breaking tests

• Establishing the evaluation of “a residual risk” for phenomena

exceeding the design conditions.

Fig.4.1 Base isolation concept for NPP

PWR Plant BWR Plant Seismic isolator

Reactor building

Purpose


Ground motion for this study • Artificial wave enveloping at general Japanese NPP sites

Maximum acceleration : 800 cm/sec2 , Maximum velocity : 200 cm/s

Isolator • Lead rubber bearing (LRB)

Diameter : 1600mm (One of the largest scale manufactured in Japan)

Target spectrum of ground motion Seismic isolator

General Conditions


(a) Characteristic tests for isolators • Static breaking tests using the 1600mm-dia. LRBs

• Static hardening tests using the 1200mm-dia. LRBs

• Dynamic horizontal and vertical simultaneous loading tests using the

250mm-dia. LRBs

Design restore model of isolator combined with

horizontal and vertical characteristic

Schematic diagram for breaking capacity, etc.

Fig. 4.2 Breaking tests of full-scale isolators

Breaking test equipment (Max horizontal load 25.1×103kN)

Photo under breaking Schematic diagram for breaking capacity

Characteristic of Isolators


(b) Seismic evaluation for base-isolated building • Determine actual isolator specifications under the ground motion

• Evaluate seismic integrity based on the MDOF stick model and 3D-FEM

Evaluation method for base-isolated building considering horizontal and

vertical motions of isolator

Optimized design for isolated pedestal, etc.

Fig. 4.3 3D-FEM model and MDOF stick model of base-isolated building

3D-FEM model Response results MDOF stick model

Evaluation of Building


• Routing design for seismic relative

displacement between these buildings

• Verification of integrity of crossover piping

- shaking tests using 1/10 scale

- static-loading repeated tests using 1/4

scale piping

Establish crossover piping design method

Fig. 4.4 Verification tests of crossover piping

Routing design (PWR plant)

Shaking test using 1/10 routing

Repeated test using 1/4 piping

Evaluation of Crossover Piping (c) Verification tests of the crossover piping between

base-isolated and non base-isolated buildings


(d) Residual risk evaluation • PRA method for fragilities of base-isolated buildings based on

various failure modes

• Evaluation of the validity of these fragilities.

Fig. 4.5 Fragility evaluation of a quake-absorbing building

Failure probability of seismic isolators Fragilities of base-isolated building

Residual Risk Evaluation


Results • Expanding flexibility of the aseismic design of NPP facilities based

on the evaluation method of seismic isolation system

Further works • Improvement of highly damping isolator for further huge-earthquakes

• Examination of fail-safe devices against the earthquake beyond the

design basis

Countermeasure for earthquake

beyond design basis

Typical Results – Seismic Isolation System

PWR Plant consist of the base-isolated

reactor building and non-base isolated

turbine building

5. Summary


Summary

• Based on the lessons learned from Fukushima

Daiichi Accident, several technology developments

to strengthen NPP safety have been completed in

March, 2017.

• The results will be considered for continuous

improvement of NPP safety.

IAEA International Conference on Topical Issues in … New core internals with many reactor control...

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