15 September Woo-Ho Lee, Seon Ho Song, Yong Lak Paek Nuclear Regulation Division. Korea Institute of Nuclear Safety (KINS)

Transcription

1 CNRA International Workshop on New Reactor Siting, Licensing and Construction Experience Licensing Experience of New Reactor (APR1400) in Korea 15 September 2010 Woo-Ho Lee, Seon Ho Song, Yong Lak Paek Nuclear Regulation Division Korea Institute of Nuclear Safety (KINS)

2 Table 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 - 2-

3 Introduction 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 - 3-

4 Introduction Location of NPPs in Korea KINS PWR Ulchin 1,2,3,4,5,6 Shin-Ulchin 1,2 Yonggwang 1,2,3,4,5,6 PWR PHWR & PWR PWR Wolsong 1,2,3,4 Shin-Wolsong 1,2 Kori 1,2,3,4 Shin-Kori 1,2 and 3,4 Slide - 4-

5 Introduction Current Status of New Reactor Shin-Kori Shin- Wolsong Shin-Kori Shin- Ulchin Plant Type MW #1 #2 #1 #2 #3 #4 #1 #2 PWR PWR PWR PWR PWR PWR PWR PWR 1,000 1,000 1,000 1,000 1,400 1,400 1,400 1,400 Commercial Operation Dec Dec Mar Jan Sep Sep Dec Dec Type OPR1000 OPR1000 OPR1000 OPR1000 APR1400 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) Slide - 5-

6 Introduction 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 - 6-

7 Introduction Shin-Kori Units 3 and 4 (APR 1400) May 2002: DC issued, April 2008: CP granted Structure inspection are undergoing Slide - 7-

8 Introduction 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 Scheduled to begin commercial operation in Dec and Dec. 2016, respectively. Slide - 8-

9 Introduction Development of Regulatory Requirements of APR1400 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 Legal System of Nuclear Safety Regulation KINS The Act provides developed the bases and a set of fundamental matters concerning the development and utilization of atomic regulatory requirements energy and safety regulations To achieve a higher level of safety for APR1400 The Decree provides particulars entrusted by the Act and necessary for the enforcement of the Act Based on the existing requirements and international safety standards The Regulation provides the technical standards and particulars entrusted by the Act and the Decree such as detailed procedures and format of documents The Notice provides detailed particulars legislated for the technical as standards a Ministerial and guidelines Ordinance on Technical Codes Standards for materials, for Nuclear design, test, and inspection of components and equipment Facilities in July 2001 Reg. Guide, Criteria, SRGs Slide - 9-

10 Licensing Process Structure of Licensing Process Slide - 10-

11 Licensing Process New Licensing Process Slide - 11-

12 Licensing Process Pre-application Safety Review (PSR) Prepare Preliminary Safety Information Documents (PSIDs) PSR Approach Safety Evaluation Results - Prepare PSIDs In the design development stage - Major information including in the PSIDs. Accident Calculation. Containment performance. Calculation of source-term. Probabilistic Safety Assessment. Safety test Program, etc. - 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 - Not official licensing review in accordance with formal licensing process - Reflected to the preparation of application Documents (SARs, SSARs) for formal licensing process Slide - 12-

13 Licensing Process Standard Design Approval (SDA) 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 - 13-

14 Licensing Process 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. Slide - 14-

15 Licensing Process 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. Slide - 15-

16 Licensing Process 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. Slide - 16-

17 Licensing Process Pre-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) Slide - 17-

18 Licensing Process Pre-Operational Inspection (POI) Inspection schedule for each stage (Shin-Kori unit 3) Structure Inspection Installation Inspection CFT Inspection CHT&HFT Inspection Start-up Test Inspection CP April 2008 First Con t R/V Install SIT/ILRT OL (Initial Fuel Loading) Remark : CFT (Cold Functional Test), CHT (Cold Hydrostatic Test), HFT (Hot Functional Test) Slide - 18-

19 Design 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, Safety Goal - CDF(/RY) - Cont. Failure Fre. (/RY) <10-5 <10-6 <10-5 < Design Life (yr) Containment Cylindrical Spherical Cylindrical Slide - 19-

20 Design 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) Thermal Margin(%) Operator Action (min) Radiation Source Term Realistic Realistic Deterministic 10.Hot Leg Temp. ( o F) Radiation Exposure Slide - 20-

21 Safety 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 - 21-

22 Safety 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 - 22-

23 Safety 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 : Acceptance criteria (1,204 ) The reflood model and code which were conservatively set in the current method Slide - 23-

24 Safety 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 - 24-

25 Safety 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 - 25-

26 Safety 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 - 26-

27 Safety 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 - 27-

28 Safety 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 - 28-

29 Conclusion 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. Slide - 29-

30 Thank you for your attention!