UK ABWR - NEW NPP DESIGN FOR UK

Transcription

1 IAEA PLiM2017 [IAEA-CN ] UK ABWR - NEW NPP DESIGN FOR UK 25-October-2017 Kazuhiro YOSHIKAWA Masaaki HAYASHI Yasuhiro MABUCHI Daisuke TANIGUCHI Hitachi-GE Nuclear Energy, Ltd.

2 Contents 1. Introduction of UK ABWR 2. Basic Strategy for Safety Measures 3. Overview of Safety Enhancement 4. Structural Integrity of the UK ABWR 5. Summary 2

3 1.1 Overview of UK ABWR The UK ABWR is a nuclear reactor design proposed for development and construction in the UK. Large scale, highly efficient plant. Highly economical reactor core Reactor coolant recirculation system driven by internal pumps Advanced control rod drive mechanism Overall digital control and instrumentation Reinforced concrete containment vessel These features constitute a highly-functional, enhancedsafety nuclear reactor system, with a compact, easy-tooperate, and efficient turbine of excellent performance. 3

4 1.1 Overview of UK ABWR (cont.) Thermal / Plant Output Reactor rated pressure Reactor Core Fuel Assemblies / Control Rods Recirculation System Primary Containment Vessel (PCV) ECCS / PCV cooling System 3,926MWt / 1,350 MWe class 7.07MPa 872 bundles / 205 rods Internal pump system Reinforced concrete with built-in liner 3 divisions 4

5 1.2 Generic Design Assessment (GDA) For new nuclear build in the UK, the first stage is a rigorous assessment of the reactor design by UK regulators, known as GDA. Hitachi-GE applied GDA in GDA assessment process for the UK ABWR is now in Step4andDesign Acceptance Confirmation (DAC) is expectedbyendof2017. Next step is site-specific assessment. 5

6 Contents 1. Introduction of UK ABWR 2. Basic Strategy for Safety Measures 3. Overview of Safety Enhancement 4. Structural Integrity of the UK ABWR 5. Summary 6

7 2. Basic Strategy for Safety Measures It is very important to enhance safety feature of ABWR design with consideration of lessons learnt from the accident at Fukushima Daiichi NPS. They indicate the need to consider the potential for sitewide damage caused by Beyond Design Basis (BDB) external hazards and event management under a severe plant condition. Hitachi summarized seven lessons learnt from the Fukushima accident. 7

8 2.1. Lesson 1 Lesson 1: Arrangement, and backup and recovery of important equipment The BDB tsunami flooded the first floor of the Reactor Building(R/B) through air inlets, outlets and through doors broken by the impact force of water. Emergency AC, DC power supply system, and switchboard lost their function Rearrangementof important equipment and facilities on a higher place and/or waterproofing are effective. Backup provisions and recovery of important equipment are also effective Following Backup provisions outside of the R/B on elevated location or a waterproof backup building power supply trucks, switchboards, fire trucks Emergency connection for the mobile equipment on the R/B Heavy machinery available for establishing access routes. 8

9 2.2. Lesson 2 Lesson 2: Configuration and deployment of isolation valves The Isolation Condenser (IC) works during isolated condition of the RPV. For a loss of its function, high pressure coolant injection systems and/or low pressure coolant injection systems with RPV depressurization are installed. In unit 1 of Fukushima Daiichi NPS, isolation valves of the IC seems to be closed due to loss of instrumentation DC power, because DC power for the instrumentation is lost prior to AC power, which is the power source of closing valves, in large tsunami. Measures for opening closed isolation valves, which can be operated outside the containment or remotely, should be added to similar system in ABWR design. 9

10 2.3. Lesson 3 Lesson 3: Provision of backup DC power supply Loss of DC power caused a following situation during the Fukushima accident: Difficulties to understand the plant status due to loss of all instrumentation functions, which caused misunderstanding of the plant status and failure of appropriate recovery actions, Loss of RPV depressurization function with SRV, which needs to open valves of accumulator with DC power, and Loss of control signal for reactor coolant injection system. it is important to provide portable DC power and DC power backup suppliesfor these important functions 10

11 2.4. Lesson 4 Lesson 4: Instrument reliability and credibility Instrumentations of the reactor water level, pressure and temperature, which are important for accident management, had a problem in reliability and credibility with a progression of the accident. Credibility of the instrumentations lost due to exceeding range of design condition. Reliability of instrumentations during a severe accident is important. These instrumentations should have a measure of confirming credibility by others. Accident management procedures and practices for loss of all instrumentation should be prepared. For SFP, temperature and level indication should be added, considering above (i.e. severe accident condition). 11

12 2.5. Lesson 5 Lesson 5: Provision of a flexible coolant injection For a BDB flooding, not only protection measures such as waterproofing and rearrangement of important equipment but also provisions for flexible coolant injection into RPV and containment are vital. Water sources should also be flexible. Considering an ultimate situation in which the loss of all permanentlyinstalled coolant injection systems occurs, the coolant injection system should be diversified by mobile equipment. Flexible and simple event management procedures with these mobile equipment and practices are vital. 12

13 2.5. Lesson 6 Lesson 6: Accessibility, operability and assurance of effectiveness To resolve the problems of accessibility, operability and assurance of effectiveness of accident management equipment, the following countermeasures are important: a heavy machinery for development of clear area a remote manual valve to reduce radiation exposure, a separate installation of several connections on the R/B for alternative coolant injection systems with mobile equipment, an isolation valve to prepare appropriate line and to avoid bypass, an alternative coolant injection line into SFP, and a bypass line of the rupture disk or removal of it. 13

14 2.7. Lesson 7 Lesson 7: Provision of alternative means for protecting the Containment Vessel A major reason of a land contamination by the Fukushima accident is assumed to be leakage from the containment gasket, because of the over temperature of containment following a delay of core debris cooling. Therefore, countermeasure for over temperature of containment is important. The most important approach for the BDB external hazards is to provide diversified measures of coolant injection into RPV because the BWR uses direct-cycle operation at low pressure, and it is easy to inject water directly into the reactor. 14

15 2.7. Lesson 7 (continued) The following basic strategy for safety measures should be applied: 1. Important safety equipment to be protected from design loads caused by external events. (e.g. embankments, waterproof door, and relocation of equipment etc.) 2. Portable equipment to be used in the event of failure regarding protection for safety equipment, and the preparation of flexible responses. 3. Accident management procedures to be as simple as possible to ensure coordination to proceed smoothly in the event of an external event leading to wide damage across the entire site and a need for off-site assistance. 15

16 Contents 1. Introduction of UK ABWR 2. Basic Strategy for Safety Measures 3. Overview of Safety Enhancement 4. Structural Integrity of the UK ABWR 5. Summary 16

17 3. Overview of Safety Enhancement Overview of the countermeasures of UK ABWR design: A) Accident management should be simple. That is, direct coolant injection into the RPV can be achieved with mobile equipment. This simple strategy shall also be an effective way of cooperating with off-site support during times of confusion. B) Connections into R/B should be provided. Connection points should be diversified. C) Alternative ways to achieve coolant injection into the containment and containment head should be provided. D) A backup building with an alternative power supply and coolant injection function should be installed. This facility is located separately from the R/B, which is kept sealed off during normal operation. It should be useful for functions such as providing a frontline base during emergencies or a storage facility for mobile equipment. 17

18 3.1 Safety Enhancement A) B) SFP Cooling (Example) Alternative coolant injection into SFP Power Supply (Example) Portable DC power and DC power backup supplies for initial response Alternative AC power supply Containment Cooling (Example) Alternative coolant injection into the containment Coolant injection into the reactor well C) A) C) Core Cooling (includes long term heat removal) (Example) Alternative coolant injection into RPV Alternative long-tem heat removal Alternative RCW D) (Example) Backup Building Protection against specific hazards Coolant injection, power supplies, and mobile equipment 18

19 3.2 Backup Building Design Concept Reactor building Backup building DC power supply for SRV Alternative AC power supply (Air-cooled DG) 高台 Mobile equipment Connections for mobile equipment SFP Core cooling Coolant injection into SFP Containment head cooling PCV Containment cooling RPV W/W Alternative Coolant injection system S/P Water source :Countermeasures 19

20 Contents 1. Introduction of UK ABWR 2. Basic Strategy for Safety Measures 3. Overview of Safety Enhancement 4. Structural Integrity of the UK ABWR 5. Summary 20

21 4.1 Structural Integrity Classification Structural Integrity Classification : It is required for the metal components and structures required high levels structural reliability that it sets higher classification than the ordinary classification, which sub-divided Class 1 into two further classes (VHI and HI); Structural Integrity Classes Failure Modes and Effects Criticality Analysis (FMECA): To identify the higher classification, general factors were considered to record in the FMECA. 1) Component Identification, Weld Location / Region FMECA (Sample) 2) Postulated Failure Mode 3) Direct Consequences of Failure 4) Protection 5) Indirect Consequences of Failure 6) Component Classification Class VHI (Very High Integrity) HI (High Integrity) SC 1 (Standard Class 1) SC 2 & 3 (Standard Class 2 & 3) Consequences of failure Severe core damage and large off-site release of radiation. Severe core damage. Containment protects against large off-site release. Limited release of radioactive material. Localised damage to fuel. Minor off-site release. Significant release within nuclear island. No core damage. Fault within capability of protective systems. Contamination within nuclear island. 21

22 4.2 Structural Integrity Classification Structural Integrity Classification of UK ABWR FMECA identified SI classification for each UK standard class 1 components being involved in the function of either radioactive containment or core support. Where necessary, classification was established on a region by region basis. Reactor Pressure Vessel Core Support Structures Main Steam Line Component Weld Location / Region Classification RPV Major Boundary Small diameter Nozzles and Penetrations, Support Skirt, Stabilizers, Brackets Top Guide / Core Plate / Shroud Support / Core Shroud / CR Guide Tube / Fuel Support Connecting welds inside containment vessel including Inboard MSIVs and MS RCCV Penetrations Connecting welds outside containment vessel including Outboard MSIVs Very High Integrity Standard Class 1 Standard Class 1 Very High Integrity Standard Class 1 Steam Relief Valve Bonnet bolt Standard Class 1 Feed Water Line Stand by Liquid Control System High Pressure Core Flooder System Residual Heat Removal System Reactor Core Isolation Cooling Reactor Water Clean-up System UK ABWR Structural Integrity Classification Each connecting weld / Valves / RCCV Penetrations Standard Class 1 Each connecting weld / Valves / RCCV Penetrations Head Vent and Spray Nozzle (boundary portion) Standard Class 1 FMCRD (boundary portion) Spool Piece Standard Class 1 22

23 4.3 Avoidance of Fracture Defect Tolerance Assessment Objectives Defect Tolerance Assessment The demonstration of defect tolerance provides an important contribution to any claim of high structural reliability. Defect tolerance can be mostly, ensured by qualified inspection supported by the defect tolerance assessment (DTA). R6 Procedure In UK ABWR, DTA was conducted to apply R6 procedure which is the elastic plastic fracture mechanics assessment based on two parameters method. Elastic Plastic Fracture Mechanics HGNE Assessment COMMERCIAL based on 2 Parameters Method 23

24 4.4 Avoidance of Fracture DTA Procedure and Qualified NDE DTA Procedure There are some processes to determine Qualified Examination Defect Size (QEDS) which the criteria are used for Qualified NDE. - DTA Stage to develop SLLDS - QEDS Stage to discuss QEDS for the inspection qualification. - Reconsideration process before QEDS will be finally determined. Qualified NDE Qualified Non-Destructive Examination was Reconsideration of conservativeness considered to assure high level integrity of highly classified component. - End-of-manufacture NDE (Ultrasonic Testing) - European Network for Inspection & Qualification (ENIQ) Methodology - Applied for Highest classified component (VHI) DTA Stage QEDS Stage Yes Design Conditions Stress Calculation R6 Calculation SLLDS Choice of Applicable Inspection Method Internal Discussion or Expert Panel Re-calculation required? No Final QEDS Inspection Qualification General DTA Procedure 24

25 4.5 UK ABWR RPV Design RPV Design of UK ABWR has been developed: - UK regulatory requirement and therefore - Designed and Manufactured in accordance with ASME - Additional beyond Code basis requirement such as: Qualified End of Manufacturing Inspection (UT) DTA to support the criteria of UT Additional Fracture Toughness Testing to support DTA calculation using R6 Code Environmental Fatigue Assessment, which the additional analysis is non-mandatory requirement of ASME Archive materials Hydrogen control and additional chemical analysis points for the VHI large parent forging material as the measure for OPEXs (Both Hydrogen Flaking and Carbon Segregation). No any design changes were additionally raised from the original Japanese design which was design and manufactured in accordance with Japanese code (JSME). 25

26 5. Summary UK ABWR design, which is proposed for development and construction in the UK, has enhanced safety feature considering lessons learnt by Fukushima Daiichi NPS accident. UK ABWR design also has structural integrity feature for 60 years plant life time which includes design of important component, of which replacement will be difficult during the plant life time. General Design Assessment of UK ABWR design is expected to complete by end of The site licensing of WylfaNewydd nuclear power station will follow based on the accepted UK ABWR design. 26

27 UK ABWR -NEW NPP DESIGN FOR UK Kazuhiro YOSHIKAWA Masaaki HAYASHI Yasuhiro MABUCHI Daisuke TANIGUCHI Hitachi-GE Nuclear Energy, Ltd. 27