Development for Nuclear Power Plant Safety

Transcription

1 IAEA International Conference on Topical Issues in Nuclear Installation Safety: Safety Demonstration of Advanced Water Cooled Nuclear Power Plants (CN-251) 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 June 7, 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 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 2

3 1. Introduction

4 Current Status of NPP in Japan Shika Tsuruga Mihama Ohi Takahama Shimane Genkai Kashiwazaki Kariwa Sendai 1 2 Ikata Tomari : PWR : BWR Hamaoka Oma 1 Higashidori Onagawa Fukushima Daiichi Fukushima Daini Tokai Daini MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved 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

5 New Regulatory Requirements Previous 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) Natural Phenomena Fire Reliability of Power Supply Reliability of Other SSCs Seismic / Tsunami Resistance New 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 Newly introduced Reinforced & Newly introduced Reinforced SSCs: Structure, Systems and Components CV: Containment Vessel 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 5

6 2. Strengthening of Overall Nuclear Safety

7 Developments based on the DiD According to the concept of Defence in Depth, the middle and long term direction and the technology developments were surveyed. levels of DiD Objective Level 1 Prevention of abnormal operation and failures Level 2 Control of abnormal operation and detection of failures Level 3 Control of accident within the design basis Level 4 Control of severe plant conditions Direction to strengthen safety functions Earthquake-resistance Maintain subcriticality to cold shutdown only with control rods Diversity for reactor core cooling Cooling of melting core Technology developments - Seismic isolation system - Enhancement of seismic evaluation method for steam generator - New core internals with many reactor control clusters - Enhancement of CFD analysis for core internals - Enhancement of core cooling capability by steam generator - Air cooling system/equipment - In-vessel retention for large reactor 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 7

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

9 ~ ~ Safety Measures under SBO/LUHS Natural-circulation core cooling can be achieved by SG secondary-side depressurization under SBO/Loss of UHS 3. Confinement Large volume containment confines radio activity and hydrogen under accident. 給水車 消火栓 Spent 使用済 Fuel Pit 燃料ピット 制御棒駆動装置 原子炉格納容器 Containment 1. Shutdown Gravity-driven CRD 加圧器 PRZ 蒸気発生器 SG 大気 MSRV 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. ~ 主蒸気逃し弁 Turbinedriven Pumps タービン動補助給水ポンプ 2. Cooling 大気 Turbine Generator タービン 発電機 地面 Ground 4. Cooling (SFP) Water can be fed at the ground level. 原子炉容器 PV Natural 原子炉容器一次系自然循環 circulation 一次冷却材 RCP ポンプ 3 Feed Water ~ Water 復水タンク tank 復水器 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 9

10 Primary Pressure Purpose Key lesson learned from Fukushima Dai-ichi Accident To secure core cooling measures 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 Accumulator Pressurizer Hot Leg Cold Leg Low Pressure Injection Pump Steam Generator Core Main Steam Relief Valve Small Break LOCA Cross-over Leg Small Break LOCA occur Reactor trip SI signal SG depressurization initiation Accumulator Injection Start Low-pressure injection start 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 10 Time 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).

11 Subjects Subjects of this study To show the validity of the core cooling with SG secondaryside 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 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 11

12 Demonstration Large Scale Test Facility (LSTF) LSTF is full-height and full-pressure (approximately 15MPa) thermalhydraulic simulator of typical four-loop PWR. Volume Height Nominal pressure Loop LSTF Scaled by 1/48 to the reference PWR corresponds to the reference PWR corresponds to the reference PWR Two loops (One broken/ One intact) 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 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 12

13 Depressurization Demonstration of the core cooling under SB-LOCA Start of SG secondaryside depressurization Primary and Secondary Pressure Actuating level of Accumulator Experiment the primary pressure decrease along with the SG secondaryside 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 thermalhydraulics are well simulated MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 13

14 Temperature (K) PCT Demonstration of the core cooling under SB-LOCA Activation timing of Accumulator 100 Peak Cladding Temperature (PCT) 200 M-RELAP5 Time (s) 300 Experiment M-RELAP5 Experiment 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 heatup 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 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 14

15 Typical Results - Core Cooling Measure 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 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 15

16 4. Development of Seismic Isolation System for Nuclear Facilities

17 Purpose 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. Reactor building Table 4.1 Development process PWR Plant Seismic isolator BWR Plant Fig.4.1 Base isolation concept for NPP 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 17

18 General Conditions Ground motion for this study Artificial wave enveloping at general Japanese NPP sites Maximum acceleration : 800 cm/sec 2, 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 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 18

19 Characteristic of Isolators (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. Breaking test equipment (Max horizontal load kn) Photo under breaking Fig. 4.2 Breaking tests of full-scale isolators Schematic diagram for breaking capacity 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 19

20 Evaluation of Building (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. 3D-FEM model MDOF stick model Response results Fig D-FEM model and MDOF stick model of base-isolated building 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 20

21 Evaluation of Crossover Piping (c) Verification tests of the crossover piping between base-isolated and non base-isolated buildings 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 Shaking test using 1/10 routing Routing design (PWR plant) Repeated test using 1/4 piping Fig. 4.4 Verification tests of crossover piping 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 21

22 Residual Risk Evaluation (d) Residual risk evaluation PRA method for fragilities of base-isolated buildings based on various failure modes Evaluation of the validity of these fragilities. Failure probability of seismic isolators Fragilities of base-isolated building Fig. 4.5 Fragility evaluation of a quake-absorbing building 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 22

23 Typical Results Seismic Isolation System 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 PWR Plant consist of the base-isolated reactor building and non-base isolated turbine building Countermeasure for earthquake beyond design basis 2017 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 23

24 5. Summary

25 Summary Based on the lessons learned from Fukushima Daiichi Accident, several technology developments to strengthen NPP safety have been completed in March, The results will be considered for continuous improvement of NPP safety MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. 25

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