The latest investigation and activities of Fukushima's ... · The latest investigation and activities of Fukushima's accident Naoto KASAHARA Professor, ... – Subcommittee for Probabilistic

The latest investigation and activities of Fukushima's accident

Naoto KASAHARA

Professor, University of Tokyo

Jointly sponsored by

ASINCO-9 April 18-20, 2012

The Splendor Kaohsiung, Taiwan, R.O.C.

1

Contents

1. Earthquake and Tsunami

2. Overview of Nuclear Power Station Accident

3. Investigations and Current Lessons learned

4. Recent Activities of the Japan Welding Engineering Society (JWES)

2

Date & Time: 14:46 March 11, 2011

Magnitude: Mw = 9.0 (Ranked 4th largest in the world)

Epicenter：～130 Km offshore

Source Depth: 24Km

(Ref.: Japan Meteorological Agency）

Summary of the Earthquake

Epicenter

Summary of Damage

（Earthquake + Tsunami)

As of January 13, 2012

(Ref. :The Fire Department Report)

Item Category counts

Persons Fatality 16,131

Missing 3,240

Injured 5,994

Houses Full Damaged 128,497

Half Damaged 240,090

Partial Damaged 677,502

Fire (broken out) 286

1. Earthquake and Tsunami

3

Tsunami Observations

Tsunami approaching to Fukushima Daiichi (1F)

Flooded area in Fukushima Daiichi (1F)

( : inundation area)

Breakwater (10m)

Unit 6

Unit 5

Unit 2

Unit 1 Unit 4

Unit 3

April 18, 2011

（Ref. :The 2011 Tohoku Earthquake Tsunami Joint Survey Group

(Height ＜４ｍ）

4

Power Station (BWR)

Unit MWe PCV Type

C.O Cond. at EQ

Scram Cold S/D

Max. Acc.

(gal) (1)

Design Acc.

(Ss) (gal)

Onagawa

1 524 Mark-I 1984 ○ ○ ○ 587 (1.11) 529

2 825 Mark-I M 1995 △ ○ ○ 607 (1.02) 594

3 825 Mark-I M 2002 ○ ○ ○ 573 (1.12) 512

Fukushima

Daiichi

(1F)

1 460 Mark-I 1971 ○ ○ No 447 489

2 784 Mark-I 1974 ○ ○ No 550 (1.26) 438

3 784 Mark-I 1976 ○ ○ No 507 (1.15) 441

4 784 Mark-I 1978 ■ - (NA) 319 445

5 784 Mark-I 1978 ■ - (NA) 548( 1.21) 452

6 1,100 Mark-II 1979 ■ - (NA) 444 448

Fukushima

Daini

(2F)

1 1,100 Mark-II 1982 ○ ○ ○ 254 434

2 1,100 Mark-II M 1984 ○ ○ ○ 243 428

3 1,100 Mark-II M 1985 ○ ○ ○ 277 428

4 1,100 Mark-II M 1987 ○ ○ ○ 210 415

2. Overview of Nuclear Power Station Accident (Units near the Epicenter in the East Coast)

+ Condition when EQ: ○：Operating, △：Startup, ■：Outage, + 980gal=1G

+ Scrammed/Cold Shutdown : ○：Success

+ (Note1) “Max. Acc.” shows max.(EW, NS) at the Base mat in Reactor Building.

6

Sea

B-５．Alternative sea water injection by fire engines

Stack

SGT

C-３．Ventilation of radioactive stream from SC to depressurize RCV

7

B-２．IC automatically started however not continues by DC power loss

Turbine Generator

Condenser

Feed

wat

er

line

Main steam line

B-１．Isolation valve closed after Reactor shutdown

Reactor Pressure Vessel

Suppression Chamber

Summary of accident sequence of Unit 1 of the Fukushima Dai-ichi

B-4 ．SHC and CS could not operate by loss of AC and Ultimate heat sink

Reactor Building

Shutdown cooling system

C-１．From RPV, enhanced temperature and pressurized stream injected into PCV and SC

C-２．Water level decrease→Core exposure →Hydrogen Generation from

Zirconium water reaction →Core melt→Melt through

A. All control rods automatically inserted and shutdown chain reaction

:Disable A. Phenomena related to Shutdown B. Phenomena related to cooling C. Phenomena related to confinement

C-４．Hydrogen exposure of RB

Isolation Condenser

High Pressure Coolant Injection

B-3 ．Disable by DC power loss

Safety Release Valve

Core spray system

Reactor Containment Vessel

Core Melting Simulation

Heat balance model

Dec

ayed

hea

t( M

W)

date

RPV

Melted Fuels Mar. 15 Mar. 11

Unit 1

Unit 2

Unit 3

0

5

30

25

20

10

15

Fuels

+ Water injection in Unit 1

stopped longer than Unit 2 and 3, then higher decayed heat was remained.

Decay heat time history

3. Investigations and Current Lessons learned

11

1) Computer Code :MAAP (Modular Accident Analysis Program)

2) Analytical Conditions:

+ Thermal balance (decayed heat vs. water injection, etc.). + Water level indication (incl. uncertainty).

+ Gas concentration in Container Vessel.

+ Radiation time history in RCW (Reactor Cooling Water system).

+ Temperature time histories at RPV and CV.

Core Melting Simulation (cont.)

Unit1

Unit 2&3

+ Unit1:All Fuels are estimated to melt and penetrate through the vessel, then be on the pedestal covered with water.

+ Unit 2 &3: Fuels are estimated damaged and melted, but almost fuels are sustained in RPV. Fuels are covered with water.

+ penetrated depth to concrete is estimated about 70cm.

App. 70cm depth

3) Analysis Result:

4.7Hrs 5.3Hrs 14.3Hrs 15Hrs

2 Cases are calculated because of water level uncertainty.

+ Water level is correct : Melted but almost fuels remain at the original position.

+ Water Level is not correct : Fuels melted down but almost remain inside the PRV.

Pedestal

PCV

App. 7.6m App. 2.6m

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Container Vessel inside Observation

PCV Inside was observed by Fiber Scope at Unit 2 .

+ Picture was not clear because of moisture.

+ Water rain was falling. (Dew Dropping or Leaking)

+ Radiation noise was detected.

+ PCV Inside surface was corroded.

+ Inside Temperature was measured as 45 ℃, consisted with 43 ℃ by the usually using thermometer.

Inside surface PCV

pipe

support

grating

Piping or conduit

Fiber Scope

13

Defense in depth : principle IAEA

14

file:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htmfile:///C:/Users/Naoto KASAHARA/DailyDATA/外部発表/24年度/ASINCO9/defdep3.htm

Current lessons learned from Fukushima accident

① Insufficient measures against severe accident which should be

presupposition Pride to disregard experiences of TMI and Chernobyl

Fruitless discussion for safety improvement between affirmative and

negative sides

Desire of absolute safety which is conflict with further measures

against accident

Inappropriate treatment of residual risk

→Shortage of synthesis capability to social technology

② Insufficient preparedness against tsunami and the total loss of

power Amount of efforts for prevention and insufficient preparedness for

mitigation and evacuation

Unbalance between stick for details and overlook of weal points of

systems

Lack of leaders who look over cross fields and organizations

→Shortage of imagination and comprehensive understanding 15

As an example of the application of the concept of risk, a clear paradigm shift in awareness from "design to prevent accidents from occurring" to "design and measures that assume that accidents may occur" will be required.

With the introduction of the risk concept, a high level of safety can be achieved by analyzing the influence of accidents on nuclear power plants, determining levels of importance and the order of precedence of improvements to be made.

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Deterministic

Probabilistic

"What Engineering should be after the Unprecedented Disaster“, http://www.t.u-tokyo.ac.jp/etpage/topics/pdf/vision_e.pdf

Safety and risk(cont.)

Within the scope of assumption

Beyond the scope of assumption

Safety and risk Difficulties of risk-communication

① Safety for engineers （Risk）＝(Damage)×(Probability) Efforts to deduce risk under allowable level

② Requirement from citizens Stable Security of their properties

③ Discrepancy Accumulation enlarge their anxiety even though safety society Stable and secured society weaken human sensitivities and resistance to

dangers

→Correct recognition of society and citizens

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18

［Before troubles］［After troubles］

Pay attention

Abilities of concentration are limited.

Reference Yotaro Hatamura, Knowledge of danger, Natsumesha Co.Ltd. (In Japanese)

Troublesome requests from

outsides

Lack oy attention to the essential part due to the limitation of human concentration

Imagination and perspective

Pay attention

Cliff edge

Power Center 19

Metal Clad Switchgear August 25, 2011 Photographed by Tokyo Electric Power Company

Dam

age

００

The scope of assumption (Slide according to accidents）

Beyond evaluation

Severity of accident

Regulation SA countermeasure

Cliff edge

Allowable level

Countermeasure against cliff edge phenomena (probabilistic approach)

Reference: International Symposium on Nuclear Safety 2011, AESJ

The scope of assumption (Slide according to Scientific evidence）

Regulation

Cliff edge

Regulation



Mitigation of the worst phenomena (deterministic approach)

20

Combined deterministic and probabilistic approach

8 Research Committees of JWES • Atomic Energy Research Committee • Special Materials Welding Research Committee • Chemical Plant Welding Research Committee • Robotic Welding Research Committee • Surface Modification and Processing Research

Committee • Welding and Joining Processes Committee • Nondestructive Inspection Technology Application

Research Committee

4. Recent Activities of the Japan Welding Engineering Society (JWES)

21

Structure and Activities of Atomic Energy Research Committee

• Chairman : Dr. Ayao Tsuge (President of Shibaura Institute of Tech.) • Secretary : Professor Shinobu Yoshimura (University of Tokyo) • Planning Board of the Committee

• Research Subcommittees (mostly sponsored by industries)

– Subcommittee for Probabilistic Fracture Mechanics (PFM) – Subcommittee for Structural Problems in Nuclear Engineering (SPN-II) – Subcommittee for Giga-Cycle Fatigue 3 (GCF3) – Subcommittee for Low-cycle Fatigue (LCF) – Subcommittee for Multi-axial Fatigue III (MF-III) – Subcommittee for International Research Communication & Collaboration Chairman : Professor Naoto Kasahara (University of Tokyo) Secretary : Dr. Naoki Miura (Central Research Institute of Electric Power Industries) ※ Preparation of new subcommittee for structural integrity and safety issues after the Great East Japan

Earthquake

• National Conferences on Various Structural and Materials Topics (twice a year)

• Training Course for Nuclear Structural Engineers 22

1

The Outlook on Taiwan Nuclear Power Industries after Fukushima

Accident

The Outlook on Taiwan Nuclear Power Industries after Fukushima

Accident

Li H. [email protected]

Material and Chemical LaboratoriesIndustrial Technology Research Institute

Copyright 2010 ITRI 工業技術研究院 2

Outline

• New Power Energy Policy in Taiwan• AEC’s Request-Integrated Examination

on Nuclear Safety• TaiPower’s Responses • Strategic Plan for the Promotion of

Local Nuclear Manufacturers/ TNA


New Power Energy Policy in Taiwan2011.10.31

• Nuclear (5% @Y-2025)– N1(2016),N2,N3 decommission at licensed due date – Lungmen (2700MW) operation under safe conditions– New constructions hold

• Coal/oil fired (30% @Y-2025)• Gas fired (40% @Y-2025)

– >20Mton• Renewable (16~20% @Y-2025) (12,502MW @Y-2030)

– Wind – 4200MW @Y-2030– Solar – 3000MW @Y-2030– others


龍門

國聖金山

馬鞍山

Chinshan GE BWR-4

636 MWe x 2# 1 Dec. 1978# 2 July 1979

Kuosheng GE BWR-6

985 MWe x 2# 1 Dec. 1981#2 Mar. 1983

Lungmen GE ABWR 1350 MWe x2

Maanshan W PWR

951 MWe x 2# 1 July 1984#2 May 1985

Taiwan current nuclear power capacity- 5,144 MWe

金山

國聖

馬鞍山


AEC request-Integrated Examination on Nuclear Safety

• AEC requested TaiPower to verify NPPs capability to respond both the DBAs and Beyond DBAs

• The program Includes– Protection measures on reactor safety– Off-site radiation protection and emergency

response• Agenda for operating reactors

– Near-term Examination (before June 30, 2011)– Mid-term Examination (before Dec. 31, 2011)Ref: Wen-Chun Teng, The 26th Japanese-Sino Seminar on Nuclear Safety, Tokyo, Japan July 26-27, 2011


Near-term Actions for Operating NPPs1. Re-examination of Capability for Loss

of All AC Power (SBO)2. Re-evaluate Flooding and Tsunami

Protection3. Ensure Integrity and Cooling of Spent

Fuel Pool4. Assess Heat Removal and Ultimate

Heat Sink5. EOPs re-examination and re-training


Near-term Actions for Operating NPPs6. The Ultimate Response Guidelines

(procedure to abandon the reactor)7. Support between different units8. Considerations for Compound Accidents9. Mitigation Beyond DBA Events10. Preparedness and backup equipment11. Manpower, Organization, Safety

Culture


Comparison between AEC and International Authorities/ Organizations

Item AEC NRC NEI WENRA WANO NISA

(1) Re-examination of Capability for Loss of All AC Power (SBO)

V V V V V

(2) Re-evaluate Flooding and Tsunami Protection

V V V V V

(3) Ensure Integrity and Cooling of Spent Fuel Pool

V V V

(4) Assess Heat Removal and Ultimate Heat Sink

V V V

(5) EOPs re-examination and re-training V V V V

(6) The procedure to abandon the reactor V

(7) Support between different units V V

(8) Considerations for Compound Accidents V V V V V

(9) Mitigation Beyond DBA Events V V V V V

(10) Preparedness and backup equipment V V V V

(11) Manpower, Organization, Safety Culture V V V


Mid-term Actions for Operating NPPs• Periodic Integrated Safety Assessment

– Routine Periodic Assessment for every 10 Years

– To Include the Coping Capability of Fukushima Accident

– To advance the schedule by 2 year for Maanshan NPP


TaiPower Responses after Fukushima Accident• Comprehensive Safety Assessments to

Operating NPPs• Comprehensive Safety Assessments to

Under Construction NPP• Newly Authored Ultimate Response

Guidelines• Imposing European Version of Stress

Test on Taipower• Planning of Decommission of Taiwan NPPs

Ref: Hwai- Chiung, Hsu, The 26th Japanese-Sino Seminar on Nuclear Safety, Tokyo, Japan July 26-27, 2011


Comprehensive Safety Assessments to Operating NPPs• For Design Basis Accident (DBA）, the Emergence Response Procedure

and Scheduled Exercises have been established.• For Beyond Design Basis Accident (BDBA), proactive plans and

countermeasures must be prepared.• TaiPower performed the Comprehensive Safety Assessments with a

humble attitude, and take enhancement actions as below:– � 11 items for Near- term (with 63 detailed items)– � 1 item for Mid-term (with 4 detailed items)– � 3 task forces

• 4 Major Undertakings：– Earthquake Protection– Tsunami Protection– Assess the rescue capabilities

• � Backup and Rescue Power Supply• � Backup and Rescue Water Supply (heat sink)• � Rescue of Spent Fuel Pool• � Integrated Management of Rescue Resources

– Newly authored Ultimate Response Guidelines


Decommission of Taiwan NPPs• A decommissioning plan shall be submitted by the

licensee (Taipower) to the Atomic Energy Council (AEC) for review and approval three years prior to the scheduled permanent cessation of operation of nuclear reactor facilities.

• A TaiPower task force is developing and preparing the decommissioning plan and environmental impact assessment report, which are scheduled to be submitted to AEC & EPA before the end of 2015.

• The estimated total backend cost is NTD 335 billion (~US $10.5 billion), for the existing 6 operating units, based on 2008 currency value, in which NTD 67.5 billion (~US $2.2 billion) is allocated for decommissioning.


Taiwan Nuclear Grade Industry Association, TNA • Objective

– To create a platform for Taiwan local industry to exchanges nuclear related information of technology and business

– To upgrade local industry technology and facilitate easier entry of local nuclear products into the global market

• Characteristics– Majority members are from local industry – Non-profit organization– Government registered organization


機械材料類(Materials and Mechanical)PIPE, PUMPS-ALL TYPES, VALVES-ALL TYPES, FITTINGS & FLANGES, TUBING, HEAT EXCHANGERS, TANKS, …

工程服務類(Engineering service)DESIGN, TESTING (COMM. GRADE PROD. ANALYS), INSPECTION, QA/QC, IEEE QUALIFICAT ON TESTING, REPAIR SERVICES, NDE, …

儀控電氣類(Instrumentation and control)WIRE &CABLE, TRANSFORMERS, INSTRUMENT PANELS, MOTORS-ALL TYPES, POWER SUPPLIES, RELAYS, …

輻射防護類(Radiation protection)RADIATION MONITORING, NUCLEAR WASTE DISPOSAL, NUCLEAR INSTRUMENTATION, DOSIMETER CALIBRATION, FUEL RACKS,HYDROGEN MONITORING SYSTEMS, …

其他類(Others)PLATING, PAINT, O-RINGS & GASKETS, …

Nuclear-grade Component IndustryNuclear Industry

Bringing existing industry into nuclear power market.+ Nu Dedication + Tech Upgrade (Higher Reliability)(excluding heavy forgings, eg. RPV, SG, Main TB and Generator)

cf. Aero Industry

Bringing existing industry into aero- market+Aero Certification +Tech Upgrade (Higher Reliability)

(Engine exclusive)


Events & Milestone of TNA 2009~2010• Held National Energy Meeting. ,2009/04/15-16

• ITRI held the workshop of “promoting Taiwan Nuclear Grade Industry Association” -Taiwan Vice President Vincent Siew gave the welcome speech. ,2009/11/3

• Submitted application document to Ministry of the Interior and apply for registered party. ,2009/12/25

• TNA held the first planning meeting. ,2010/03• TNA was formally organized. ,2010/06/14• TNA visited AEC. ,2010/07• TNA visited NE-NTHU. ,2010/07• TNA visited Straits Exchange Foundation, Chairman

Chiang. ,2010/08• TNA visited MOEA. ,2010/09• HUST in Vietnam visited ITRI and TNA ,2010/11• Sino-Japanese Nuclear Workshop on Nuclear Safety, Nuclear

Development, and Piping Design ,2010/11~12• NSC & IDB granted the budget for the development of Taiwan

Nuclear Grade Industry ,2010/12


Events & Milestone of TNA 2011• TNA visited State-Owned Enterprise Commission(國營

會) ,2011/2/21• Technical Workshop on NPP system and component design ,2011/3/8 • Industry Workshop on the promotion of local nuclear component

manufacturing ,2011/4/22• TNA visited Straits Exchange Foundation, Chairman Chiang

2ndly ,2011/4/26. • Workshop on the localization of power equipment ,2011/7/4 • TNA Annual Meeting and Business Opportunity Forum for Power

Equipment ,2011/7/15• TNA visited major power equipment companies in Japan, including

J-Power, JINED, MHI to establish business opportunities exchange platform ,2011/7/24

• Beijing Nuclear Power Equipment Exhibition and Visit ,2011/9/19• Taiwan Power Forum – Global supply chain ,2011/11/2 • China Fujian Nuclear Power Office visited TNA ,2011/11/28


TNA on cross-strait activities • 參加中國核工業國際展及拜會有關單位（2010年3月）(11th Nuclear

Industry China Exhibition, China Nuclear Power Technology Research Institute )

• 拜會中廣核核電設備國產化聯合研發中心（2010年4月） (China GuangDong Nuclear)

• 協辦中國核動力院張副院長專家團訪台『2010兩岸核能電廠技術交流會』（2010年8月）(Nuclear Power Institute of China)

• 兩岸金屬材料創值論壇大陸專家團拜會工研院及TNA（2010年9月）(Cross-strait Metal Forum)

• “兩岸精密機械產業合作及交流會議”大會, 南京, 2010/10/12 ~ 13(Cross-strait Mechanical Industry Meeting)

• Beijing Nuclear Power Equipment Exhibition and Visit (2011/9/19)

• China Fujian Nuclear Power Office visited TNA (2011/11/28)


2009/11/3, ITRI held the forum of “promoting

Taiwan Nuclear Grade Industry Association”

Taiwan Vice President Vincent Siew gave the opening speech.


2011/04 Visit Japan

J-Power (Fossil) MHI (Fossil)

JINED (Nuclear)JAIF (Nuclear)TiTech-APS (CSP)


2011/07 Beijing Nuclear Exhibition

拜訪大陸核能相關機構，建立聯絡管道--福建發改委核電辦、中國核學會及核工業集團公司、中國核能建設集團公司、能源局科技裝備司、國家核電技術公司、中國電力投資集團公司、核安全局核安全管理司、中國核電工程有限公司(核二院) 、中國核行業協會

核安全局首次表明台灣廠商適用HAF-601(民用核安全设备设计制造安装和无损检验监督管理规定，非604境外單位)中國核能建設集團與俊鼎機械、南寧工程展開接觸，工研院將協助產品設計技術

工研院協助推動『台灣核能級產業發展協會』(TNA)於2010/6/14成立工研院、益鼎工程、俊鼎機械、亞炬企業、國森企業、常熟華新特殊鋼、榮剛材料、南寧工程、中鋼機械、台灣端板、福臨參展東方電機洽詢榮剛合金鋼材料，工研院將協助資質審訂問題中國核能建設集團公司詢問REDEX防火產品相關資訊，對產品效能很感興趣

『台灣核能級產業發展協會』首次參展『台灣核能級產業發展協會』首次參展

建立兩岸核能級產業交流平台建立兩岸核能級產業交流平台


舉辦「全球電能設備商機交流研討暨座談會」

合作備忘錄簽訂儀式齋藤社長積極促成「台灣核能級產業發展協會」與「日本技術者聯盟」，針對台灣產業技術需求以及商機媒合為前提，將於明年(101年)二月在台灣舉行台日產業交流暨座談會，透過此次合作備忘錄的簽訂，台灣核能級產業發展協會與ANNEX RESEARCH INC. 將在商機媒合、技術引進輸出、人才交流等方面繼續保持合作，作為協會廠商與日本產業界之間的橋樑。

註：MOU簽訂儀式，由齊藤衛社長(右2)、許文都理事長(右3)為簽證人；周能傳副局長(左2)、劉仲明副院長(右1)、吳再益院長(左1)為見證人。

2011/11/02 Global Power Equipment


Major Missions of TNA• To coordinate the consortium of local nuclear industry• To act as the advisory committee of Taiwan nuclear

industry policy• To become the exchange center for of nuclear

industry related information• To form the technology center to develop and upgrade

local manufacturing technology• To initiate the cross-strait bridge to engage in the

Chinese mainland nuclear market• To be the contact window between foreign nuclear

Rx/ or nuclear component vendors• To operate a holding company of marketable nuclear

grade components manufactured locally


Taiwan’s Strategic Plan for the Promotion of Nuclear-grade Industry

宜有產業政策宣示建立技術研發策略法令政策制訂配合提供基金以及商機啟動兩岸搭橋計畫成立科專及輔導專案

策略聯盟產業成形提出產業發展建言兩岸核電溝通平台產業技術資訊交流形成產業之供應鏈合法立案法人組織

產業轉型永續發展大供應商合作夥伴生產核能品保產品進入全球核電市場

結合產學研之能量輔導產業提升技術特殊關鍵技術突破建立核能品保制度積極推動國際合作

Action Items

TNA Government

法人(ITRI)Industry

Development Policies

1.In Taiwan：MOEA/Taipower’s policies to assure market opportunities for domestic manufacturers, and assist them to obtain nuclear-grade qualifications.

2.With China：Promoting cross-strait cooperation：Mutual recognition of Regulations/Standards/Licenses, and assisting a smoother entry into Mainland China’s markets.

3.To the World：Partnership with major nuclear A/E and vendors, entering global supply chain system


廠商

•台灣

•大陸

•全球

In Taiwan - Organizing the nuclear certification platform with (in) TNA

國內測試、驗證能力盤點與整合

核電關鍵組件規範及標準建立及

研擬

國內核電級組件盤點

核電級產業推動及技術研發策略

規劃

核電級認證平台及驗證技術(Nuclear-grade certification and accreditation techniques)•核能品保及檢證評估技術•環境模擬測試技術•設備組件老化模擬測試技術•耐震驗證分析技術•輔導廠商建立核能品保…

核電級產品設計、製造、檢驗及分析技術(Nuclear-grade products design, manufacturing, testing and analysis techniques)•核電級產品可靠度設計技術•(Design for Higher Reliability)•金屬成型及接合技術•特殊合金及焊材製造技術•檢測及驗證技術…

TNA整合國內測試能量

•工研院材料、機械、電子測試驗證

•核研所檢/驗證中心

•台灣電子檢驗中心

•地震驗證實驗室•金屬中心閥製品檢測驗證實驗室


两岸交互認證

大陸核電法規及工業規範資料收集，並納入國內法規中，使本土核電製造標準，與大陸核安全局取得相互認證。

協助國內廠商經原委會認證，即可取得大陸核電製造輸入核准證（HAF601,HAF604），而使產品能直接進入大陸核電市場。

With China – Start Cross-strait Bridge Project


To the World – Global Supplier Chain• Nuclear Vendors

• Seller’s Market, Look for global supplier chain.

BWRPWR PWR PWR BWR BWR

Approved Vendor List

WH-AP1000; AREVA-EPR

SourcingPre-

selectionPre-

qualification Qualification

12~20億USD/year34 Rxs

69 Rxs


TNA Members 50 members (up to 2011.12)

Eng. Service I & C

Key MaterialsMechanical

益鼎、俊鼎、台灣汽電、詹記、南寧、銘榮元、振鍵

東元電機、欣歐、華敏光纖寅輝台塑、公元資訊

中鋼機械、泰豐、三太造機、萬機鋼鐵、力鋼、亞炬、鑫綠泰、捷流閥業、柏卡、晶騰、良聯、誠兌、祥景

中鋼、華新麗華、榮剛、精剛梧濟、中鋼焊材、安利金屬

R & D工研院、核研所、清大原科院台綜院、金屬中心、電檢中心

Others永記、柏林、仁望、TUV 、歐雅所羅門、騰馳

Information Technology

Power Plant

3%Machinery

36%

Engineering Consulting

9%

Software/Information

9%

Academia

26%

Instrument/Control

17%

Power Plant

Machinery

Instrument/Control

Engineering Consulting

Software/Information

Academia


Summary (1/2)

• Taiwan new power energy policy re-adjust the share of nuclear power energy. The license renewal of NPPs and all the new constructions will be suspended.

• ROC-AEC and TaiPower perform the Near-term Examination and Mid-term Examination to enhance the nuclear safety of Taiwan NPPs.


Summary (2/2)

• TNA was organized and aims to assist local industry to participate in the global nuclear market.

• TNA is looking for the cooperation with global supply chain, and initiating the cross-strait nuclear project.

Vendor

ITRI/ INER/ TRI/ NTHU

Taiwan Local

Industry

TNA

T S I M I N G T S E N GS T E V E N S O N & A S S O C I A T E S

A S I N C O 9 , K A O H S I U N G , T A I W A NA P R I L 1 9 , 2 0 1 2

U.S. Seismic Margin Assessment Methodologies

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1

Seismic Re-Evaluation in the U.S.

Systematic Evaluation Program (SEP), late 1970’s to early 1980’s for 11 oldest plants

1980 USNRC I&E Bulletin 80-11 evaluate masonry block walls

Unresolved Safety Issue (USI) A-46, late 1980’s, verify seismic adequacy of mechanical and electrical equipment in many older plants

Independent Plant Examination of External Events (IPEEE), early 1990’s

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2

Seismic Re-Evaluation in the U.S. - Continued

Generic Issue GI-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern U.S. on Existing Plants, USNRC Generic Letter 2011-XX: Seismic Risk Evaluation for Operating Reactors

Near-Term Task Force review, Request for Information pursuant to 10CFR50.54(f) regarding to recommendations 2.1, 2.3, and 9.3 of the insights from the Fukushima Dai-Ichi accident

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3

Purpose of Seismic Re-Evaluation

Identifying plant-specific vulnerabilities to severe accidents beyond design basis considerations Develop an appreciation of severe accident behavior Understand the most likely severe accident sequences that

could occur under full power conditions Gain qualitative understanding of the overall likelihood of core

damage and fission product releases

If necessary, to reduce the overall likelihood of core damage and radioactive material releases by modifying hardware and procedures that would help prevent or mitigate severe accidents

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4

Acceptable Methodologies

Seismic Probabilistic Risk Assessment (PRA) NUREG-1407, Section 3.1 ASME/ANS RA-Sa-2009

Seismic Margin Assessment (SMA) USNRC Seismic Margin Method

NUREG-1407, Section 3.2.4 NUREG/CR-4334

EPRI Seismic Margin Method NUREG-1407, Section 3.2.5 EPRI NP-6041-SL ASME/ANS RA-Sa-2009

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5

Comparison of Methodologies - Approach

Seismic PRA NRC SMA Method EPRI SMA Method

Probabilistic Semi-probabilistic Partially probabilistic

Should be used for NTTF

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6

Comparison of Methodologies – Scope of Review


Event trees and fault trees are usually developed from the event/fault trees developed for the internal event analyses. Structures and elements where failure could impact and fail safety-related elements are added to the trees

For PWR, the safety functions of reactor criticality and early emergency core cooling are considered. For BWR, the safety functions of reactor sub-criticality, emergency core cooling, and residual heat removal are considered. In addition, a small break LOCA is postulated to occur, soil failure modes, potential for earthquake-induced flooding, and earthquake-induced fires, as well as non-seismic failure and human actions are considered.

Review includes electrical, mechanical, and NSSS equipment, piping, tanks, heat exchangers, cable trays and conduit raceways, containment, and structures. In addition, leakage equivalent to a small break LOCA is postulated to occur in one success path. Soil failure mode, potential for earthquake-induced floodingand earthquake-induced fires, as well as non-seismic failures and human actions are considered.

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7

Comparison of Methodologies – Seismic Input


Site-specific hazard curves for peak ground acceleration and response spectra should be used.

A response spectrum shape is specified, for example, NUREG-CR-0098 median shape, anchored to level significantly higher than the plant design basis earthquake, may be used. Development of new structure response spectra, including effects of soil-structure interaction, is encouraged.

Same as the NRC SMA method.

NTTF requires enveloping site-specific GMRS and SSE

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8

Comparison of Methodologies – Selection of Equipment


Elements whose failure could lead to core damage (i.e., Level 1 PRA) are considered initially. Fault trees are “pruned” based on systems and fragility considerations.

Elements whose failure could lead to core damage are considered initially.Fault trees are “pruned” based on systems and fragility considerations.

Two separate and independent shutdown success paths are selected. One path postulates leakage to a small break LOCA.

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9

Comparison of Methodologies – Screening Requirements


Screening based on system and fragility considerations.

In general, equipment functionality is investigated based on seismic experience or test data. Equipment anchorage is analyzed for each component. Caveats and guidance are provided in the criteria screening tables in NUREG/CR-4334 and EPRI NP-6041 for three ranges of seismic input.

In general, equipment functionality is investigated based on seismic experience or test data. Equipment anchorage is analyzed for each component. Caveats and guidance are provided in the criteria screening tables in EPRI NP-6041 for three ranges of seismic input.

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10

Comparison of Methodologies – Required Experience and Training


The seismic PRA should be performed by experienced systems and seismic capability engineers who can perform seismic fragility analysis.

The seismic margin assessment should be performed by trained, experienced seismic capability and systems engineers. Seismic capability engineers must be capable of performing fragility analysis if this method is used.

The seismic margin assessment should be performed by trained, experienced seismic capability and systems engineers.

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11

Comparison of Methodologies – Walkdown Procedure

Seismic PRA NRC SMA Method EPRI SMA MethodPrincipal elements of the walkdown are (1) seismic capacity versus seismic demand, (2) caveats based on earthquake experience and generic testing data bases, (3) anchorage adequacy, and (4) seismic spatial interaction with nearby equipment, systems, and structures. Walkdown procedures should follow the requirements contained in EPRI NP-6041.

Principal elements of the walkdown are (1) seismic capacity versus seismic demand, (2) caveats based on earthquake experience and generic testing data bases, (3) anchorage adequacy, and (4) seismic spatial interaction with nearby equipment, systems, and structures. Elements not screened out are identified as outliers for further review. Walkdown procedures should follow the requirements contained in EPRI NP-6041.

Same as the NRC SMA method.

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12

Comparison of Methodologies – Evaluation of Component Capacity


For elements not screened out during walkdown, calculate fragility parameter values, that is, median capacities and logarithmic standard deviations.

The capacity of components that were not screened out can be calculated using the fragility analysis (FA) or the conservative deterministic failure margin (CDFM) method.

The capacity of components that were not screened out can be calculated using the conservative deterministic failure margin (CDFM) method.

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13

Containment Performance

SMA requires an evaluation of containment performance

Primary purpose of the containment performance evaluation Identify sequences and vulnerabilities that involve

containment, containment function, and containment systems seismic failure modes or timing

Containment Penetrations Rigorous fragility analysis needed for RLE greater than 0.3 g Evaluate of backup air system of the equipment hatch and

personnel lock that employ inflatable seals Evaluate cooling loss for penetrations need cooling

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14

Containment Performance - Continued

Valves Valves involved in the containment isolation system are

expected to be seismically rugged Walkdown to ensure there are no spatial interactions Seismic failure of actuation and control systems should be

examined Valves relying on backup air should be included

Heat Removal/Pressure Suppression Systems Components of the containment heat removal and pressure

suppression functional systems should be examined

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15

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