CNRA International Workshop on CNRA International Workshop on “New Reactor Siting, Licensing and Construction Experience”“New Reactor Siting, Licensing and Construction Experience”

Licensing Experience of New Reactor (APR1400) in KoreaLicensing Experience of New Reactor (APR1400) in Korea

Woo-Ho Lee, Seon Ho Song, Yong Lak Paek

Nuclear Regulation Division

Korea Institute of Nuclear Safety (KINS)

15 September 2010

Slide - 2-

Table of ContentsTable of Contents

Introduction

Licensing Process • Pre-application Safety Review • Standard Design Approval • Construction Permit (CP) for Nuclear Installation • Operating License (OL) for Nuclear Installation • Pre-Operational Inspection

Design Features of APR1400

Safety Review

• Standard Design Approval

• Shin-Kori unit 3&4

Conclusion

Slide - 3-

IntroductionIntroduction

Current status of NPPs in Korea

Korea has achieved a remarkable growth in nuclear power since commercial operation of Kori Unit 1 in 1978

Korea has now 20 operating NPPs : 16 PWRs and 4 PHWRs

Total licensed output of 20 units is 17,716MW

- generation capacity : 32.9%

- actual production : over 40 %

6 PWRs are now under construction and 2 additional units are planned

4 PWRs (1,000MW) and 2 PWRs (1,400MW) are under const.

2 PWRs (1,400MW) are under docket review for CP

Total units in operation by 2016 : 28 units

Slide - 4-

IntroductionIntroduction

Location of NPPs in Korea

Ulchin 1,2,3,4,5,6

Shin-Ulchin 1,2

Wolsong 1,2,3,4

Shin-Wolsong 1,2

Kori 1,2,3,4

Shin-Kori 1,2 and 3,4

Yonggwang

1,2,3,4,5,6

KINS

PWR PWR

PHWR & PWR

PWR

Slide - 5-

IntroductionIntroduction

Plant Type MW Commercial

Operation

Commercial

Operation Type

Shin-Kori #1 #1

#2

PWR PWR

PWR

1,000 1,000

1,000

Dec. 2010 Dec. 2010

Dec. 2011

OPR1000

OPR1000

OPR1000

Shin-

Wolsong

#1

#2

PWR

PWR

1,000

1,000

Mar. 2012

Jan. 2013

Shin-Kori #3 #3

#4

PWR PWR

PWR

1,400 1,400

1,400

Sep. 2013 Sep. 2013

Sep. 2014

OPR1000

OPR1000

APR1400 APR1400

APR1400

Shin-Shin-

Ulchin

#1 #1

#2

PWR PWR

PWR

1,400 1,400

1,400

Dec. 2015 Dec. 2015

Dec. 2016

APR1400 APR1400

APR1400

: Under Construction, : Under Docket Review, : Under Construction & DR for OL

OPR 1000 : Improved KSNP (Korea Standard Nuclear Power Plant, 1000MW PWR)

APR 1400 : Advanced Power Reactor (1400MW PWR)

Current Status of New Reactor

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IntroductionIntroduction

Background & Current status of APR1400

• Launch of a 10-year National Project : 1992

To develop technologies involved in the design of an advanced

reactor

Application for Pre-application Safety Review (PSR) : 2000

Application for Standard Design Approval (SDA) : 2001

• The SDA was issued in 2002 after two years of safety review

• Shin-Kori units 3&4(APR1400)

KHNP applied a Construction Permit *(CP) in 2003 , CP was

granted in 2008

Structure & Installation inspection are undergoing

• Shin-Uljin units 1&2 (APR1400)

KHNP applied a CP in October 2008, Review for CP are undergoing

Slide - 7-

IntroductionIntroduction

Shin-Kori Units 3 and 4 (APR 1400)

• May 2002: DC issued , April 2008: CP granted

• Structure inspection are undergoing

Slide - 8-

IntroductionIntroduction

Shin-Uljin units 1&2

• PWR-type reactor (1,400 MWe APR)

Almost same design as Shin-Kori units 3&4

• Docket Review & Safety Review for CP are undergoing since Oct. 2008

• Scheduled to begin commercial operation in Dec. 2015 and Dec. 2016, respectively.

Slide - 9-

IntroductionIntroduction

Development of Regulatory Requirements of APR1400

KINS developed a set of

regulatory requirements

• To achieve a higher level

of safety for APR1400

• Based on the existing

requirements and

international safety

standards

legislated as a Ministerial

Ordinance on Technical

Standards for Nuclear

Facilities in July 2001

Reg. Guide, Criteria, SRGs

Industrial Codes and Standards

(ASME, IEEE, ACI, KEPIC, etc.)

Atomic

Energy Act

Enforcement

Decree of the Act

(Presidential Decree)

Enforcement

Regulations of the Act

(Ministerial Ordinances)

Notice of the Minister of

Education, Science and Technology

The Act provides the bases and

fundamental matters concerning the

development and utilization of atomic

energy and safety regulations

The Decree provides particulars

entrusted by the Act and necessary for

the enforcement of the Act

The Notice provides detailed

particulars for the technical standards

and guidelines

Codes and Standards for materials,

design, test, and inspection of

components and equipment

The Regulation provides the technical

standards and particulars entrusted by the Act

and the Decree such as detailed procedures and

format of documents

Industrial Codes and Standards

(ASME, IEEE, ACI, KEPIC, etc.)

Atomic

Energy Act

Enforcement

Decree of the Act

(Presidential Decree)

Enforcement

Regulations of the Act

(Ministerial Ordinances)

Notice of the Minister of

Education, Science and Technology

The Act provides the bases and

fundamental matters concerning the

development and utilization of atomic

energy and safety regulations

The Act provides the bases and

fundamental matters concerning the

development and utilization of atomic

energy and safety regulations

The Decree provides particulars

entrusted by the Act and necessary for

the enforcement of the Act

The Decree provides particulars

entrusted by the Act and necessary for

the enforcement of the Act

The Notice provides detailed

particulars for the technical standards

and guidelines

The Notice provides detailed

particulars for the technical standards

and guidelines

Codes and Standards for materials,

design, test, and inspection of

components and equipment

Codes and Standards for materials,

design, test, and inspection of

components and equipment

The Regulation provides the technical

standards and particulars entrusted by the Act

and the Decree such as detailed procedures and

format of documents

The Regulation provides the technical

standards and particulars entrusted by the Act

and the Decree such as detailed procedures and

format of documents

Legal System of Nuclear Safety RegulationLegal System of Nuclear Safety Regulation

Slide - 10-

Licensing ProcessLicensing Process

Structure of Licensing Process

Slide - 11-

Licensing ProcessLicensing Process

New Licensing Process

Slide - 12-

PrePre--application Safety Review (PSR)application Safety Review (PSR)

Licensing Process Licensing Process

Prepare Preliminary Safety

Information Documents (PSIDs)PSR Approach Safety Evaluation Results

- Prepare PSIDs In the design

development stage - The applicable current regulations,

and codes & standards are applied

first

- When necessary, enquired

regulatory requirements for the new

design features will be developed as

appropriate

- Major information including in the

PSIDs

. Accident Calculation

. Containment performance

. Calculation of source-term

. Probabilistic Safety Assessment

. Safety test Program, etc.

- Not official licensing review

in accordance with formal

licensing process

- Reflected to the preparation of

application Documents (SARs,

SSARs) for formal licensing

process

Slide - 13-

Standard Design Approval (SDA)Standard Design Approval (SDA)

Licensing Process Licensing Process

Purpose

• Encourage the use of standard plant designs to enhance plant

safety, to improve the efficiency, and to reduce the complexity of

the regulatory process

• Improve the efficiency of licensing review by early resolution of

safety issues for the timely construction of nuclear power plants

Documents submitted for SDA

• Standard Design Safety Analysis Report (SSAR)

• Description on the technical capability for the design of the reactor

• Preparation plan for the Emergency Operating Procedure (EOP)

Slide - 14-

Construction Permit (CP)Construction Permit (CP)

Purpose

to ensure that the technical standards for the location, structure,

facility, and performance of NPP are met

Documents submitted for CP

Radiation Environmental Report (RER)

Preliminary Safety Analysis Report (PSAR)

Quality Assurance Program (QAP) for construction

Description on the Technical Capability (DTC) for nuclear plant

installation, etc.

Licensing Process Licensing Process

Slide - 15-

Construction Permit (CP)Construction Permit (CP)

Processing Period of CP

New reactor is 24 months

15 months for the reactors that have similar type and size in design to the previously licensed ones,

Review method

KINS’ Safety Review Guides (SRGs)

Provide not only the regulatory requirements but also review procedures to keep the consistence in the review results

Safety review

The principle and concept of reactor facility design

The implementation of the regulatory criteria

The evaluation of the environmental effects resulting from the construction, etc.

Licensing Process Licensing Process

Slide - 16-

Operating License (OL)Operating License (OL)

Purpose to confirm whether the components, systems, structures are installed and

their performances are assured as designed. In addition, to perform safety review on the operating capability and accident

management

Licensing documents for OL

Final Safety Analysis Report (FSAR)

Quality Assurance Program (QAP) for operation

Technical Specifications (TS) for Operation

Radiation Environmental Report (RER)

Radiation Emergency Plan (REP)

Description on the Technical Capability (DTC) for the reactor operation

Description on nuclear fuel loading plan

Description of the technical background and verification method to be for the Emergency Operating Procedure, etc.

Licensing Process Licensing Process

Slide - 17-

PrePre--Operational Inspection (POI)Operational Inspection (POI)

Purpose to confirm that the structures, systems, components (SSCs) of plants

are manufactured, installed, and tested in compliance with the SAR and QAP

to ensure that the completed nuclear reactor can be operated as expected throughout the design life

Inspection Items

Structure Inspection (19 items)

Installation Inspection (52 items)

Cold Functional Test (CFT) Inspection (77 items)

Cold Hydro Test (CHT) & Hot Functional Test (HFT) Inspection (23 items)

Initial Fuel Loading & Start-up Test Inspection (33 items)

Licensing Process Licensing Process

Slide - 18-

PrePre--Operational Inspection (POI)Operational Inspection (POI)

Inspection schedule for each stage (Shin-Kori unit 3)

Licensing Process Licensing Process

2008 2009 2010 2011 2012 2013

Structure Inspection

Installation Inspection

CFT Inspection

CHT&HFT Inspection

Start-up Test Inspection

CP April 2008

R/V Install

OL (Initial Fuel Loading)

Remark : CFT (Cold Functional Test), CHT (Cold Hydrostatic Test), HFT (Hot Functional Test)

SIT/ILRT

First Con’t

Slide - 19-

Design Features of APR1400Design Features of APR1400

Design Features

based upon the design, construction and operation experiences of KSNP

Overall design concept

very similar to that of KSNP and System80+

Major design differences between APR1400, System80+ and KSNP

Design Features APR1400 System80+ KSNP

1. Capacity (Mwe) 1,400 1,300 1,000

2. Safety Goal - CDF(/RY) - Cont. Failure Fre. (/RY)

<10-5 <10-6

<10-5 <10-6

10-4 10-5

3. Design Life (yr) 60 60 40

4. Containment Cylindrical Spherical Cylindrical

Slide - 20-

Design Features of APR1400Design Features of APR1400

Design Major design differences between APR1400 and KSNP

Design Features APR1400 System80+ KSNP

5.ECCS - No. of Trains - Safety Injection - RWST location

4

DVI Inside Cont.

4

Cold Leg Injection Inside Cont.

2

Cold Leg Injection Outside Cont.

6.Seismic Design (g) 0.3 0.3 0.2

7.Thermal Margin(%) 10-15 15 8

8.Operator Action (min) 30 30 10

9.Radiation Source Term Realistic Realistic Deterministic

10.Hot Leg Temp. (oF) 615 615 621.2

11.Radiation Exposure 20 20 50

Slide - 21-

Safety Review of APR1400Safety Review of APR1400

Standard Design Approval (SDA)

Licensing issues during the safety review Thermal-hydraulic loads and pool temperature of In-containment Refueling

Water Storage Tank (IRWST)

Performance Evaluation of ECCS

Consideration of Environmental Effect in Fatigue Evaluations of ASME Code Class 1 Components

Soft Control Application for Digital I&C System

Human Factors Engineering for the Advanced Control Room

Alternative Accident Source Term

In-Vessel Core Debris Retention through External Reactor Vessel Cooling

Steam Generator Tube Integrity

Supplementary actions

Uncertainty analysis for the core cooling capability at late reflood phase

Verification of design suitability for soft control and safety console

Submission of design report concerning steam generator tube integrity

Slide - 22-

Safety Review of APR1400Safety Review of APR1400

Shin-Kori units 3&4

Schedule of Shin-Kori units 3&4

Design certification of APR1400 : May, 2002

Application for CP : October, 2003

Safety review for CP : October, 2003 ~ April, 2008

Issuance of CP : April, 2008

Safety review focused on

Implementation of the follow-up actions for SDA

Comparison with the previous units and the standard design of APR1400

Major design features of Shin-Kori units 3&4

Domestic and overseas experience in nuclear power plant operation

Other matters

Slide - 23-

Safety Review of APR1400Safety Review of APR1400

Shin-Kori units 3&4 Assessment of Implementation of the Follow-up

Actions for SDA Safety Injection System Performance Evaluation during Late Reflood

of a LBLOCA

Evaluation of Design Adequacy of the Soft controller, Safety Control Panel and Remote Shutdown Room (RSR)

Scheme for Suppressing Wear Damage of the Steam Generator Tube and Flow Induced Vibration Assessment

Safety Injection System Performance Evaluation during Late Reflood of a LBLOCA

The maximum clad temperature in a LBLOCA : 987.2℃ Acceptance criteria (1,204℃)

The reflood model and code which were conservatively set in the

current method

Slide - 24-

Safety Review of APR1400Safety Review of APR1400

Assessment of Implementation of the Follow-up Actions for SDA

Evaluation of Design Adequacy of the Soft controller, Safety

Control Panel and Remote Shutdown Room (RSR)

Design class categorization and the detailed design plan

Human factors engineering program implementation plan

Physical and electrical isolation of RSR from MCR and ICCR (Instrumentation and Component Control Room)

Scheme for Suppressing Wear Damage of the Steam Generator Tube

and Flow Induced Vibration Assessment

Material, capacity, number of the steam generator tube and the design of the top steam generator tube support structure were changed

Wear resistance of the steam generator tube : 62% improved

Hydraulic expansion method : Contact strength & Remaining stress

Heat treated Alloy 690 : No flaws caused by corrosion in the steam generator tube expansion transition areas

Standardization of expansion pressure

Slide - 25-

Safety Review of APR1400Safety Review of APR1400

Shin-Kori units 3&4 Licensing issues for major design features

Evaluation of the Hydrogen Control System Design Evaluation of the Design of the Reactor Cavity Flooding System Fatigue Design of the Safety Class 1 System Considering Environmental

Impact Design Excluding Operating Basis Earthquake Loads in Seismic Design Adequacy of the Reactor Coolant System Overpressure Protection Facilities Evaluation of the Design of the Human-System Interface of the Main

Control Room The Integrity of the Structure of Engineering Safety Feature Actuating

System Adequacy of the Integrated Design of the Soft Controller and Engineering

Safety Feature-Component Control System Analysis of Human Reliability for Establishing the Human Error Mechanism Application of Human Factors Engineering Activity to Local Control Panel

Design Redundancy Design of the Digital Protection Relay System Design Change of the Off-Site (Preferred) Power Supply System Classification of Underwater Intake and Discharge Facilities

Slide - 26-

Safety Review of APR1400Safety Review of APR1400

Review of Major Design Features

Evaluation of the Design of the Reactor Cavity Flooding

System(CFS)

The basic design of the reactor cavity

flooding system

The flooding time of the reactor

cavity

CFS series 1 : 50 minutes

CFS series 2 : 25 minutes

The schematic diagram of the reactor cavity flooding system of Shin-Kori Units 3 & 4

Slide - 27-

Safety Review of APR1400Safety Review of APR1400

Shin-Kori units 3&4

Evaluation of Reflection of Experience in NPP operation

Review of Ductility Dip Cracking (DDC) Resistance of Alloy 690

Design Change Reflecting Experience of Swedish Forsmark-1 Accident

Review of Ductility Dip Cracking (DDC) Resistance of Alloy 690

Ductility Dip Cracking (DDC) between the welded layers or in the root of

Alloy 690.

Inconel 52M filler material (AWS Classification, ERNiCrFe-7A) for the

buttering areas of Alloy 690

Non-destructive tests (PT, UT and RT) for all Alloy 690 welds

A test for qualifying the ductility dip cracking resistance of Alloy 690

Slide - 28-

Safety Review of APR1400Safety Review of APR1400

Shin-Kori units 3&4

Review of Other Matters operation

Survey of Faults in the Site

Design Standard of Gas Effluent Sampling Facilities

Inclusion of the Screw Fixture Control Program in the Preliminary Safety Analysis Report

Delta-Ferrite Content Requirement for Austenite Series Stainless Steel

Slide - 29-

ConclusionConclusion

PSR and SDA

Improved the efficiency of licensing review by early resolution of safety issues for the timely construction of new reactor (APR 1400)

APR 1400 (Shin-Kori units 3&4)

The location, structure and equipment of the nuclear reactor and related facilities for Shin-Kori units 3&4 satisfied the current safety requirements

APR 1400 can protect the public health and the environment from the impact of the radioactive materials generated from the construction of the facilities

Relatively recent experience accumulated in Korea can be utilized

effectively to facilitate mutual cooperation in the area of new reactor regulation.

KINS has a willingness to cooperate actively with other regulatory authority if the KINS have abilities to give a hand.

Thank you for your attention !Thank you for your attention !