Overview of Mitsubishi Advanced PWR

IAEA INPRO 7th Dialogue Forum in Vienna

November 19, 2013 Sumio FUJII Acting General Manager, Nuclear Systems Engineering Department

• The data and information in the presentation materials are proprietary of Mitsubishi Heavy Industries.

• Utilization of the data and information is limited to the purpose of INPRO 7th Dialogue Forum.

• The data and information concerning advanced PWR in the presentation materials are never permanent. They may be changed in the process of nuclear power plant construction for individual site.

• Mitsubishi Heavy Industries has the copyright of all figures and photos in the presentation materials.

1. Development History of Mitsubishi PWR 2. Advanced PWR for Global Deployment 3. Safety of Mitsubishi Advanced PWR 4. State-of-Art Technology for Digital Instrumentation & Control System of Mitsubishi Advanced PWR 5. Safety Enhancements for Beyond Design Basis Accident of Mitsubishi Advanced PWR 6. Construction Management of Mitsubishi Advanced PWR 7. Operation and Maintenance Management of Mitsubishi Advanced PWR 8. Conclusion

2Loop （（（300-600MWe ））） 3loop （（（900-1000 MWe ））） 4Loop （（（1200-1500MWe ））） Unit C/O Outpt Unit C/O Output Unit C/O Output Ohi1 1979.03 1175 MkWe Mihama1 1970.11 340 MkWe Takahama1 1974.11 826 MkWe Ohi2 1979.12 1175 MkWe Mihama2 1972.07 500 MkWe Takahama2 1975.11. 826 MkWe Turuga2 1987.02 1160 MkWe Genkai1 1975.10 559 MkWe 1976.12 Mihama3 826 MkWe Ohi3 1991.12 1180 MkWe Ikata1 1977.09 566 MkWe Sendai1 1984.07 890 MkWe Ohi4 1993.02 1180 MkWe Genkai2 1981.03 559 MkWe Takahama3 1985.01 870 MkWe Genkai3 1994.03 1180 MkWe Ikata2 1982.03 566 MkWe Takahama4 1985.06 870 MkWe Genkai4 1997.07 1180 MkWe Tomari1 1989.06 579 MkWe ～ Senfai2 1985.11 890 MkWe Tsuruga3 201X 1538 MkWe Tomari2 1991.04 579 MkWe Tsuruga4 201X～ 1538 MkWe Ikata3 1994.12 890 MkWe

Tomari3 2009.12 912 MkWe

Tsuruga 3 & 4 are Advanced PWRs.

• The development of the Advanced PWR (APWR) was started in late 1980s as a joint cooperative development project by five Japanese PWR owner utilities and Mitsubishi Heavy Industries (MHI), financially supported by the Ministry of International Trade and Industry (Currently the Ministry of Economy, Trade, and Industry) as a part of the Phase III Improvement and Standardization Program of Japanese PWR. • The design of the APWR was based on MHI’s conventional 4-loop plant technologies, on which MHI has accumulated significant operating experiences, and was scaled up to achieve higher electrical output. • In addition to adopting those proven technologies after the first step development, further modifications were also made on the prototype APWR design to improve economy, safety, reliability, operability, and maintainability by incorporating advanced technologies. • The first APWR plant (Japanese APWR) is Tsuruga-3 producing 1538 MWe operated by the Japan Atomic Power Company. The Tsuruga-3 is now under safety review for construction permission, which got suspended due to the Fukushima Daiichi accident.

70’s 80’s 90’s 2000’s 2010’s 2020’s

Development & Improvement of PWR Technology APWR Tsuruga 3, 4 licensing process

Enhanced up to APWR

US-APWR US Utilities US NRC Licensing Comanche Peak 3, 4

European Utilities EU-APWR

European and ATMEA1 Global Utilities AREVA by ATMEA (Joint Venture)

• By referring the J-APWR, MHI has designed the US-APWR which meets regulatory requirements in United States of America as well as Utilities Requirements Document (URD). The turbine-generator system and all electric equipment of the US-APWR are designed for 60 Hz electric grid. • Basic designs of the US-APWR except for the fuel length are same as those of the J-APWR whose design has completed. The US-APWR has been developed as a larger-output version of the J-APWR, aiming at higher electrical outputs and improved economics, by modifying some design features mainly in the secondary side without increasing core thermal output. • The fuel length was changed to 4.2 meters in place of 3.66 meter, and the electric output was increased to about 1700 MWt. • EU-APWR is a sister plant of the US-APWR and is aiming to satisfy European Utilities Requirements (EUR). Nevertheless the turbine- generator system and all of electric equipment are designed as 50 Hz, designs of core and other major equipment on the reactor coolant system are same as those of the US-APWR except the reactor coolant pumps.

A course of design for the US-APWR and EU-APWR is to; ‹To achieve the best performance in safety, reliability, operability, and maintainability by incorporating advanced technologies. ‹To provide the world’s largest NPP for large electric grid utilities aiming at economical production costs. ‹ To contribute toward safe, stable and flexible power production through satisfying American “Utilities Requirements Document ” and “ European Utilities Requirements ”.

V Full 4-train safety systems with b est-mix of passive

SH SH and active systems, which RV allows o n-power SH ACC ACC SH maintenance

RWSP

V Full digital I&C technology enabling one-man operation

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 10 Major Features on economy V Large reactor producing thermal output of 4,466MWt V 14-ft fuels creating additional thermal margin and flexible core operation V 14-ft fuels making 24-month operation without deterioration in fuel economy V Enhanced SG heat transfer performance by enlarged heat transfer area with triangular lattice arrangement of SG tubes V High-performance s team-water separators generating high quality steam in SG V High performance LP-turbine having last stage blades of 70 inches length V Secondary system enabling high thermal efficiency over 36%

Reactor Vessel Core design

Control rod drive mechanism

Outlet nozzle

Inlet nozzle

Fuel assembly

Reactor Fuel and Rod cluster control assemblies vessel A, B, C, D : Control group bank SA, SB, SC, SD : Shutdown group bank

Best estimates Design limits of initial core Active Core Active core equivalent diameters (cold) 3.89 m Active fuel height (cold) 4.2 m Hydrogen/Uranium atomic ratio (cold) 5.57 Core average linear power density 15.2 kW/m Maximum linear power density 39.5 kW/m 31.2 kW/m

N Nuclear enthalpy rise hot channel factor F ∆H 1.73 1.50

Delayed neutron fraction βeff (%) 0.44 to 0.75 0.50 to 0.69 Prompt neutron lifetime, l* (µsec) 8 to 20 14.0 to 15.3 Reactivity coefficient

Doppler power coefficient (pcm/%power) BOC -12.4 to -7.4 EOC -12.1 to -7.6 Moderator temperature coefficient (pcm/ ℃℃℃) -71.1 to -1.4 Moderator density coefficient (pcm/g/cm 3) < 0.51x10 5 < 0.32x10 5 Boron coefficient (pcm/ppm) - -9.3 to -8.0 Neutron multiplication factor

Maximum core reactivity keff (BOC,cold,noXe) - 1.223

Maximum fresh fuel assembly k∞ - 1.456 Boron concentration (ppm) Refueling boron concentration > 4000 -

Cold shutdown, BOC, noXe, ARI, keff <0.95 - 1850

Cold shutdown, BOC, noXe, ARO keff =0.99 - 1796

Hot shutdown, BOC, noXe, ARO keff =0.99 - 1706

Hot zero power, BOC, noXe, ARO keff =1.0 - 1579

Hot full power, BOC, noXe, ARO keff =1.0 - 1444 Hot zero power, BOC, equilibrium Xe, ARO - 1086 Shutdown margin BOC 2.94 > 1.6 EOC 2.15

Design Parameters Design Values Coolant condition Primary coolant system pressure (MPa(absolute)) 15.51 Thermal design flow rate (ton/hr) 76,300 Effectiveflowrateforcorecooling(ton/hr) 69,500 Reactor vessel inlet temperature (℃) 288.1 Average rise temperature in reactor vessel (℃) 36.9 Heat transfer at normal condition Fraction of heat generated in fuel (%) 97.4 Core average linear heat rate (kW/m) 15.2

Maximum local peak linear heat rate at FQ=2.6(kW/m) 39.5 Power density (kW/l) 89.2 Specific power (kW/kg uranium) 32.0 Minimum DNBR by WRB-2 correlation At nominal condition; for typical hot channel 2.05 for thimble hot channel 1.98 During AOO; for typical hot channel > 1.35 for thimble hot channel > 1.33 Maximum fuel centerline temperature during AOO (℃℃℃) < 2548

Conventional 4-loop PWR US-APWR

RWSP SprayHeader SprayHeader Spray Header Header Spray Spray Header Header Spray Spray

‹ Four independent safety trains • Four-train mechanical safety systems • Four-train electrical safety systems • No inter-connecting piping between trains ‹ Direct vessel injection (DVI) • The cooling water from the safety injection pumps is directly injected into the reactor vessel. ‹ Refueling water storage pit (RWST) inside the containment • Operation of changing the suction from the RWST to the containment recirculation sump is eliminated. ‹ Advanced accumulator (ACC) • An advanced accumulator is connected to each cold leg to refill the reactor vessel lower plenum and down-comer immediately after a LOCA. ‹ No low-head safety injection system • Installation of ACCs enabled us to eliminate low-head safety injection systems, which resulted in reduction of the number of active components.

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 20 Advanced Accumulator having two flow rates • A short while high flow rate of cooling water is required just at early stage of LOCA, a large volume of water is injected to fill up the down-comer. After that, the flow rate is reduced to the amount necessary for decay heat removal. Advanced accumulator system has a longer supplying duration that can cover the time expected to the low pressure injection system. • This design makes it possible to eliminate the LPIS, which leads to the reduction of active components.

Nitrogen Nitrogen Conventional Design

Injection Water Main stand pipe Injection Water

Side inlet

Flow Damper Flow Damper

Side inlet

Advanced Accumulator

High Flow Rate Mode Low Flow Rate Mode

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 22 Full Digital Instrumentation & Control System • Fully digitalized technology is applied to the instrumentation and control system (I&C system) for both safety and non-safety functions. • The reactor protection system and actuation system of engineered safeguard features (ESF) have 4-time redundancy in the I&C system. • Conventional operating and monitoring devices such as switches and indicators have been eliminated, except devices for diverse actuation system (DAS).

System Features ・Fully computerized Main control board ・Visual Display Units (VDU) for safety and non-safety channels ・Limited number of conventional switches and indicators ・Fully digitalized with Mitsubishi digital controllers Safety I&C ・Four-train redundancy in reactor trip system, ESF actuation system and safety logic system for component control ・Fully digitalized with Mitsubishi digital controllers Non-safety I&C ・Duplex digital architecture for each control and process monitoring sub-system ・Fully multiplexed, including class 1E signals Data communication ・Multi-drop data bus and serial data link ・Fiber optics communication networks for noise immunity and required isolation

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 24 Human Engineering & Control Room Design • Plant parameters and associated operating switches are displayed on same screens, and touch screen operations are applied. Thus, the operator’s work load is reduced and the reliability of operation is increased. • A large display panel is installed to display major parameters for normal and abnormal situations. Thus, the current status of the entire plant can be understood by all crews and their communication is improved. • Thus, the digital system provides significant benefits to the safety of nuclear power through reduction in operation works and maintenance work loads, which reduces the potential for human error.

• Diverse Actuation System (DAS) is installed to Diverse Human-System- Interface (HSI) Panel provide back-up actuations for safety and non- Switches safety components as countermeasures against Indicators common cause failure (CCF) in the software of Alarms the digital I&C system.

• The DAS was designed for beyond design basis Measured parameter CCF incidents. The DAS consists of diverse signals devices from the digital safety system so that a CCF in the digital system doesn’t impair the DAS Automatic actuation functions. system • The DAS initiates safety functions independent from the output of the digital safety system. Manual actuation is provided for all functions. Each Automatic actuation is also provided for functions Remote if the time for manual operator action is I/O Diverse inadequate. Trip Components such as pumps, valves

• Countermeasures against Airplane Crash • The US-APWR is designed against the airplane crash (APC) by taking into account the potential effects of the impact of a large commercial airplane for the building strengths and plant layout. • The countermeasures for the APC in the US-APWR design: - Physical / functional separation and segmentation in the layout design for safety related structures, systems and components (SSCs) - Reinforcement of structural design of the buildings, which contain safety related SSCs • It was confirmed by the assessment that the inherent robustness of the US-APWR design would be maintained for the followings; - The reactor core remains cooled or the containment remains intact - Spent fuel pit integrity is maintained

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 28 Countermeasures against Severe Accidents • Prevention of severe accidents 1. Reduction of potential precursor of design basis accidents and severe accidents. • Reduction of latent LOCA precursors. E.g.; - Adoption of top-mounted in-core instrumentation system (ICIS) to eliminate penetrations at the reactor vessel bottom. - Reduction of reactor vessel welding seams and improvement of welding methodologies - Improvement of piping bypassing the containment to a higher rating to reduce probability of interface LOCA 2. Enhanced reliability in safety functions • Installation of four-train structure for safety systems • Installation of refueling water storage pit (RWSP) in the containment • Installation of advanced accumulators to reduce the number of active components • Countermeasures against Anticipated Transient Without Scram (ATWS). E.g.; Installation of diverse actuation system to cope with common cause failure in the reactor protection system

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 29 Countermeasures against Severe Accidents • Countermeasures against station blackout (SBO) E.g.; Installation of four-train emergency gas turbine generators (GTGs) and additional alternate power source by two-train GTGs. • Countermeasures against fire E.g.; Clear physical separation between the redundant trains of safety systems • Countermeasures against the intersystem LOCA E.g.; Prevention of over-pressurization of the residual heat removal system • Frequency of severe accidents (PRA result for core damage frequency) • The result of calculation for US-APWR core damage frequency (CDF) meets the NRC goal and the Utility Requirements Document (URD) goal.

Requirement NRC URD US-APWR

CDF 1×10 -4/RY 1×10 -5/RY 3.0 ×10 -6/RY

• Mitigation of severe accidents (To keep the integrity of the containment vessel) • For debris dispersion, reinforcement of depressurization function of the primary system and an improvement of RV cavity geometry are considered. • For quasi-static over pressurization, an usual containment vessel air recirculation system and an alternative containment spray supplied from the fire service water system can be used. These systems can be used to reduce the pressure if the containment spray system is not available. • As countermeasures to avoid containment vessel damage due to hydrogen combustion, a hydrogen control system (igniters) is installed to control the hydrogen concentration.

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 32 Construction Management of Advanced PWR (1) Total Engineering Capability with “Single Responsibility” in the full areas; V Conceptual/Basic/Detailed Engineering V Manufacturing of components V Erections and installations of structures, systems and components V Support of human resource development for operation & maintenance V Technical support after start of commercial operation

¢ To keep quality and schedule, Material procurement MHI’ Data management system and management integrates;

V Design Integration V Procurement CAM* V Manufacturing & Inspection

V Construction Welding inspections

On-site installation inspections

Construction process management

•CAM : Computer-Aided Manufacturing

Major Components (Reactor Vessel, Steam Generator, Reactor Coolant Pump, Reactor Internals, Control Rod Drive Mechanism, Pressurizer, Turbine, etc.) are manufactured in our hands. Works and machines have been updated/enlarged and are prepared for the global deployment.

(Photo) A reactor vessel is manufactured by a super- large combined machine tool named “Super Miller”.

Reactor vessel is now under processing in upright installation position with high- accuracy & high-quality .

Reduction of on-site work volume and construction period

40m-dia. upper containment VRational designs like; • Internal structures using steel plate reinforced concrete (SC) (left) • Large modular (prefabricated) block construction (Right)

VTools for efficient construction like Super-large-capacity cranes Brilliant Successes which enables on-site welding and (1st Concrete to Fuel Loading) formation of containment vessel . 2 loop : 34.5 months 3 loop : 37.5 months VAbility for comprehensive coordination 4 loop : 40.0 months of civil & installation work

RV inspection unit A-UT machine RV head nozzle inspection unit

Censer

Phased array type probe for ECT probe for J-type welds surface Inspections can be done for J-groove welds 4 days by utilizing two units SG tubes inspection unit simultaneously. Pressurizer (for 4 loops)

・・・Application of world’s first submerged automated guided vehicle with robot RV ・・・Using a manipulator with SG 7 shafts One unit （（（more than 3,000 tubes ））） can be inspected for around 4 days.

UT / ECT for BMI nozzle inspection unit Piping inspection unit ：：： MHI point focus probe Automated UT unit for piping ECT (for detection 0.5mm) Phased array censer •Lightweight and compact unit TOFD-UT ( for sizing ) •Point focus phased array enables high accuracy flaw Achievement of high accuracy sizing. sizing with point focus (Approx. ±±±3mm) capability

© 2013 MITSUBISHI HEAVY INDUSTRIES, LTD. All Rights Reserved. ８AS-EXP-20130003 38 Conclusion MHI developed Advanced PWRs based on accumulation of half-century experiences in construction and operation of conventional PWR power plants. Proven and advanced technologies have been introduced in APWR to improve economy, safety, reliability, operability and maintainability. Advanced accumulator was employed to reduce the number of active components, and low pressure injection system was eliminated. US- and EU-APWR were developed to let them meet regulator’s and user’s requirements including flexible operation. The fuel length was changed to 4.2 meters. ECCS and supporting systems have 4 divisions in redundancy. Due to enhancements in safety systems, the total of the CDF is estimated 3.0x10 -6 /RY in a PRA calculation. MHI’s manufacturing and construction management is integrated to one data base system to accomplish them with high quality, high reliability and strict keeping of schedule.

Design Control Documents of the US-APWR are found on; http://www.nrc.gov/reactors/new-reactors/design-cert/apwr.html