US-ABWR design features and FLEX concept for extended loss ...

J-STAGE Advance Publication date: 28 February, 2017Paper No.15-00614

© 2017 The Japan Society of Mechanical Engineers[DOI: 10.1299/mej.15-00614]

Vol.4, No.2, 2017Bulletin of the JSME

Mechanical Engineering Journal

US-ABWR design features and FLEX concept for extended loss of AC power events

1. Introduction

The ABWR is a third generation, evolutionary boiling water reactor. Several ABWR units have been operated in

Japan with excellent performance for more than a decade. In the United States, the US-ABWR obtained design certification from the USNRC in 1997 (ML11126A100, 1997); this design is the reference for the STP3&4 Combined License Application (COLA) for the construction and operation of two ABWRs in Texas (Fig. 1). Nuclear Innovation North America (NINA) is the Licensee Applicant for this new build project, and Toshiba is the selected primary technology contractor.

The STP3&4 ABWR design has incorporated numerous design and technology enhancements for improved safety performance and meeting recent USNRC regulations. For example, the design satisfies the aircraft impact rule, 10CFR 50.150 (NRC Regulation), and has amended the certified design to reflect this advancement. Cassette type Emergency

Kenji ARAI*, Yuji YAMAMOTO*, Steve THOMAS**, Bill MOOKHOEK**,

Jim POWERS*** and Thomas F CARTER**** *Toshiba Corporation

8, Shinsugita-Cho, Isogo-Ku, Yokohama 235-8523, Japan

E-mail:[email protected]

**Nuclear Innovation North America (NINA)

***Toshiba America Nuclear Energy (TANE)

**** Westinghouse Electric Company

Abstract The US-Advanced Boiling Water Reactor (ABWR), certified by the USNRC, is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3&4) Combined License Application (COLA) and incorporates numerous design and technology enhancements for improved safety performance. Nuclear Innovation North America (NINA) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3&4 project has finished the USNRC technical review of the COLA and the final safety evaluation report (FSER) was issued by the USNRC in September 2015. Following the accident at the Fukushima Dai-ichi plant, the US-ABWR has been further reviewed for Beyond Design Basis Event (BDBE) safety using industry and regulatory guidance for USNRC Order EA-12-049 “Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (BDBEE)”. By virtue of the design approach, the US-ABWR is capable of providing an indefinite coping period for a station blackout. The use of installed systems with extended coping times is a significant advantage of the US-ABWR compared to most of the plants currently operating in the U.S. In addition, STP3&4 design incorporates enhancements consistent with the current US industry Diverse and Flexible Coping Strategies (FLEX) initiative. This paper summarizes the progress of the US-ABWR licensing and describes the technology and features of the US-ABWR design that contribute to safety post-Fukushima. Key words : US-ABWR, Combined license application, Beyond design basis event, Diverse and Flexible

Coping Strategies, Extended loss of AC power

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Received: 30 October 2015; Revised: 20 January 2017; Accepted: 19 February 2017

2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

Core Cooling System (ECCS) strainers of STP3&4 ABWRs meet the Regulatory Guide 1.82 Rev. 3 (NRC Regulatory Guide, 2003).

In response to the events at the Fukushima Dai-ichi plants, the USNRC issued Order EA-12-049 including associated Interim Staff Guidance (ML12229A174, 2012) which requires that US nuclear power plants provide an integrated plan for adding mitigation strategies that will increase the defense-in-depth for Beyond Design Basis External Events (BDBEE) scenarios to address an Extended Loss of AC power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) occurring simultaneously at all units on a site. Nuclear power plants in the United States have a coping capability for Station Blackout (SBO) conditions for a limited time period ranging from approximately two to sixteen hours per 10CFR 50.63 (NRC Regulation). An ELAP event is defined as a loss of all off-site and on-site AC power sources for a potentially indefinite time period.

NINA has submitted an integrated plan which is consistent with the current US industry Diverse and Flexible Coping Strategies (FLEX) initiative for STP 3&4 ABWRs to address the USNRC requirement (ML13128A140, 2013). The US-ABWR is capable of providing a significant coping period for an ELAP and consequent LUHS without core damage by using existing plant systems, and equipment enhancements to support the FLEX integrated plan would further align the plant design and licensing basis with the FLEX concept.

This paper summarizes the progress of the US-ABWR licensing and describes the technology and features of the US-ABWR design that contribute to safety post-Fukushima. The evaluation of the coping capability for ELAP is included as well.

2. US-ABWR design and STP3&4 combined license application

The ABWR is a proven safe design with over 40 reactor years of safe reliable operation at three sites in Japan. There are currently four operational ABWRs at these sites: Kashiwasaki-Kariwa Units 6&7, Hamaoka Unit 5 and Shika Unit 2; these plants entered Commercial Operation in 1996, 1997, 2005 and 2006 respectively, and as such reflect much of the current industry philosophy for nuclear safety, technology advancements and design robustness, having evolved substantially from the earliest BWR designs that began at the Fukushima Daiichi site in Japan. The selection of the ABWR as the technology for the STP3&4 project was based to a major degree on the proven track record of the Japanese ABWRs, including nuclear safety, design, construction and operational performance.

2.1 Engineered safety features

The US-ABWR design fully addresses NRC requirements for design basis events, and incorporates three

redundant and independent divisions of ECCS as depicted in Fig. 2. Each division has one high pressure (High Pressure Core Flooder System or Reactor Core Isolation Cooling System) and one low pressure (Low Pressure Flooder System) coolant makeup system with the dedicated emergency diesel generator.

Fig. 1 STP3&4 site near Bay City, Texas

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

In addition to the conventional divisional ECCS systems, the US-ABWR incorporates a number of less conventional features that substantially contribute to plant safety, and provide capability to cope with beyond design basis conditions; some examples:

− Combustion Turbine Generator (CTG) provided to cope with SBO. − Alternate Feedwater Injection System (AFI) capable of injecting water directly to reactor. − AC Independent Water Addition (ACIWA) mode of the Fire Protection System for core cooling, spent fuel

pool cooling, or containment spray actuation using either the diesel driven fire pump, portable pump or fire truck.

These features were incorporated into the design basis for the plant as the US design prior to the event at Fukushima Dai-ichi and provide diversity that proved to be effective when evaluating mitigation of a potential BDBE.

Fig. 2 US-ABWR ECCS Configuration

2.2 STP3&4 combined license application The STP3&4 COLA was submitted to the USNRC in September 2007 by NINA for the construction and operation

of two ABWRs in Texas. The USNRC has completed its technical review for the COLA and the FSER was issued in 2015. The STP3&4 COLA is the reference-COLA (R-COLA) for the ABWR standard design which references the Design Control Document (DCD) Rev.4 certified in 1997 (ML11126A100, 1997) and incorporates design advances since that time to stay abreast of evolving industry standards for nuclear safety and reliability; some examples include:

Aircraft Impact Assessment (AIA) required by 10CFR50.150 was completed for the STP3&4 site, and the design incorporates appropriate structural and fire barrier enhancements. In addition, a new Alternate Feedwater Injection (AFI) System was included for diversity.

ECCS Strainers for the plant have been designed, including the downstream effects resolution. Cassette type ECCS strainers of STP3&4 ABWRs meet the Regulatory Guide 1.82 Rev. 3, long term cooling requirement.

Seismic Analysis and Design meets the latest standards of the industry, including consideration of advanced soils-structure interaction analysis and the Central and Eastern United States-Seismic Source Characterization (CEUS-SSC).

I&C Systems Design has advanced since the time of DCD certification in 1997, and the STP3&4 project incorporates the best of the Toshiba and Westinghouse platforms which are integrated to provide Operators with state-of-the-art Human-Machine Interface.

RCIC

LPFLHPCF

EDG EDG

EDG

HPCFLPFL

LPFL

Main Steam Line

High Pressure Core Flooder (HPCF) Pump

Emergency Diesel Generator (EDG)

External Power Source

Reactor Coolant Isolation Cooling (RCIC) Pump

Feed Water Line

Low Pressure Core Flooder (LPFL) Pump

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

The majority of these features have been reviewed by the USNRC Staff and the USNRC Advisory Committee on Reactor Safeguards (ACRS). The STP3&4 project has finished the USNRC technical review of the COLA and the final safety evaluation report (FSER) was issued by the USNRC in September 2015 (ML 15232A128, 2015). The USNRC issued the combined licenses for STP3&4 construction and operation in February 2016 (ML16056A428, 2016, ML16033A047, 2016).

3. US-ABWR capability for beyond design basis events

A consolidated listing of design features that prevent or mitigate Design Basis Accidents (DBA) and/or a Beyond

Design Basis Event (BDBE) is shown in Table 1.

Table 1 US-ABWR DBA/BDBE prevention & mitigation features

The US-ABWR design incorporates B.5.b equipment as an extensive damage mitigation feature to address the

requirements from Order EA-02-026, Section B.5.b (Interim Compensatory Measures Order, 2002) and 10 CFR 50.54(hh)(2) (NRC Regulation) which were imposed after the events of September 11, 2001 by the USNRC. These features were incorporated in the US-ABWR design prior to the accident at Fukushima Dai-ichi, and reveal that the US-ABWR has a considerable level of inherent safety and diversity to cope with BDBEs. The safety analyses presented in the USNRC certified DCD include many beyond design basis severe accident scenarios that demonstrate safety using the capability of the above listed systems and components.

3.1 US-ABWR core cooling diversity

Key to coping with a BDBE is the ability to inject water into the core in a timely and effective manner. The

US-ABWR features several diverse means of accomplishing core cooling, should the conventional AC-powered ECCS be unavailable. First there is the Reactor Core Isolation Cooling (RCIC) system, which can provide core cooling for approximately 8 hours during Station Blackout (SBO) conditions with availability of DC Control Power as described in the USNRC certified ABWR Design Control Document (DCD), and substantially longer given battery load management and manual operation during an ELAP. The STP3&4 ABWR incorporates a new RCIC compact turbine water lubricated (TWL) pump, which features single shaft mounted turbine and pump impellers enclosed in a monoblock casing capable of operating with suppression pool temperature up to 394 K. This design is free of rotating shaft seals, which simplifies the design by eliminating need for a barometric condenser, vacuum pump and associated valves to control seal leakage; and, the lube oil system and coolers are also no longer required. The result is a robust and reliable RCIC system that can operate for long periods independent of Direct Current (DC) battery power when

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

using manual control of the steam admission valve. Recently the manufacturer of the turbine water lubricated (TWL) pump performed a submergence test with two 8 hour runs to demonstrate the robust nature of this component; videos of the test are available on the manufacturer’s website (www.spx.com/en/clydeunion-pumps/).

A second alternative for core cooling is the AFI system, which with availability of the Combustion Turbine Generator (CTG) for power, can provide water to the reactor through the Feedwater system. This diverse and independent system was provided to cope with scenarios related to a hypothetical aircraft impact, but is available in any situation where it would be necessary or expedient to provide water to the core from outside the reactor building and turbine building. A third alternative is the AC Independent Water Addition (ACIWA) mode (Fig. 3) of the Fire Protection System (FPS) using the diesel driven fire pump, a portable pump or a fire truck. This mode of operation credits seismically designed portions of the Fire Protection System as a means to cool the core, using water from the Fire Water Storage Tanks (FWSTs). Because the ACIWA connects to the Residual Heat Removal (RHR) system, it is capable of supporting not only core cooling, but also Containment Spray and Spent Fuel Pool (SFP) makeup functions.

Fig. 3 AC independent water addition mode

3.2 US-ABWR containment overpressure protection One of the challenges to the station staff at Fukushima Dai-ichi during the course of the event was the inability to

effectively vent the containment in a controlled manner as pressure, and hydrogen content, increased over time. Although the plants included containment vents, they proved difficult to operate in the conditions during the accident. The US-ABWR design incorporates a Containment Overpressure Protection System (COPS) that is passively actuated using rupture disks (Fig. 4) and vents the containment to the plant stack.

The inner rupture disk in this arrangement has a design burst pressure set at approximately twice the containment design pressure (0.72 MPa) based on containment structural capability criteria. The outer rupture disk is set to a nominal value and serves as an isolation boundary to maintain the COPS piping system inert during standby conditions. Air operated valves are provided which are normally open – fail open to allow containment isolation following system actuation. The containment venting system at the Fukushima Dai-ichi nuclear power plants had normally closed isolation valve and therefore needed operator action to achieve the containment vent under the SBO condition.

Augmenting the COPS pressure relief function is the ACIWA mode of the Fire Protection System, which can be aligned through the RHR system to provide water to the wetwell or drywell sprays. The US-ABWR includes additional severe accident mitigation features to protect containment integrity, including fusible-link flooder valves in the under-vessel region, which will open on increased temperature to flood the area below the reactor vessel with Suppression Pool water to cool core material should an ex-vessel event occur.

Standpipe

LPFL PumpDiesel DrivenFire Pump(Installed)

OutdoorFire TruckConnection

ReactorBuildingWall

DrywellSpray

Drywell

MO MO

MO

ReactorVessel

WetwellSpray

Supply Tank

MO

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

Fig. 4 Containment overpressure protection system

3.3 US-ABWR SFP cooling during ELAP

The US-ABWR incorporates a CTG that can power the Fuel Pool Cooling and Cleanup System (FPC) during a SBO. In the event of a BDBE that renders the CTG unavailable, the time to boil considering maximum abnormal spent fuel heat load is approximately 12 hours. Maintaining the SFP full of water at all times during the ELAP event is not required; instead the requirement is to maintain adequate level to protect the stored spent fuel and limit exposure to personnel onsite and offsite. Once boiling starts at 12 hours, it would take another 64 hours (76 hours total from the event start) for the level to drop to ten feet above the top of the fuel, giving the operators sufficient time to provide pool makeup. SFP level instrumentation with remote readout is provided for operators to monitor SFP level. There are redundant seismically designed standpipes at opposite sides of the reactor building to connect an external pump or fire truck to provide makeup and cooling to the SFP through spray headers in the pool. Additionally, the ACIWA mode of the FPS can provide makeup water to the pool through the RHR system or external standpipes.

4. The STP3&4 FLEX integrated plan

NINA, the licensee for STP3&4, has submitted a FLEX integrated plan to the USNRC consistent with other US licensees to demonstrate compliance with Order EA-12-049, and the commitments associated with post-Fukushima mitigating strategies will be included in a new Appendix 1E to the COLA. The USNRC has completed their review for the FLEX integrated plan in 2014 and confirmed that the FLEX integrated plan satisfies the Order EA-12-049.

The STP3&4 FLEX integrated plan is based on an evaluation of the plant capability for extended coping conducted by NINA and Toshiba, considering equipment availability during an ELAP, battery load management, water inventory for cooling and makeup to the core and SFP, and importantly, the robustness of equipment credited for coping. The CTG was not credited in this evaluation for ELAP coping.

4.1 STP3&4 ELAP coping time line

The systems and time line used for coping with an ELAP are described below. As an overview, the coping plan consists of running RCIC until Suppression Pool temperature reaches the limit of the pump capability at 394 K, at which point suction is aligned to the Condensate Storage Tank (CST) and the operators begin reactor cooldown and depressurization. At approximately 20 hours into the event, containment pressure will reach the COPS activation set point, and depressurization will occur. When the CST is nearing the end of its inventory at approximately 36 hours, the reactor is fully depressurized using a Safety Relief Valve (SRV), and core injection will continue using the ACIWA mode of FPS. The plant then continues coping until additional support resources arrive from the Regional Response Center. The time line steps are as follows:

− Apply load management to extend DC power for a coping duration up to 72 hours.

Innerrupture disk

Outerrupture disk

System valves(Air operated valves)

Drywell

Wetwell

Plant Stack

PP

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Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

− Run RCIC in manual control to provide core cooling. After enough CST water has been injected to raise the Suppression Pool level to the high-high alarm set point, RCIC suction will shift to the Suppression Pool. RCIC suction will remain aligned to the Suppression Pool until the pool temperature approaches 394 K, which is the qualification limit of the RCIC pump. Operators will then switch suction back to the CST at approximately 10 hours.

− Reactor cooldown and depressurization to approximately 2.5 MPa is initiated. − Suppression Pool temperature and pressure continue to increase until the COPS rupture disk opens at a

nominal pressure of 0.72 MPa; this will occur at approximately 20 hours. COPS will then vent the containment from the wetwell via the plant stack and provide containment cooling.

− RCIC continues to provide core cooling until the CST approaches the end of its usable volume. Operators will then align ACIWA to commence injection to the reactor through RHR, and the Reactor Pressure Vessel (RPV) will be depressurized using an SRV. This transition to ACIWA is predicted to occur at about 36 hours.

− ACIWA has water inventory in the FWSTs for continued coping, and is also capable of connecting to additional reserves in the UHS Basin if required.

− The Regional Response Center commitment is 32 hours for backup resources to arrive at the site, which is well within the coping capability of the plant, and will enable operators to continue to safe shutdown.

The plant response to an ELAP was evaluated by using the Modular Accident Analysis Program (MAAP4, 1994-2005) code and the result is shown in Section 5 of this paper.

4.2 STP3&4 design enhancements to support FLEX

There are several design enhancements that are being incorporated into the STP3&4 design to support the FLEX integrated plan. These enhancements relate to the robustness of water and fuel supplies, and providing manual local controls and monitoring. Some specific enhancements include:

− Robust seismic design for the CST, FWST and Diesel Fire Pump (DFP) Fuel Oil Tank. − Ensure structures, systems and components used for FLEX mitigation, such as the ACIWA pump, are protected

from flooding and hurricane missiles.. − Increase the size of the DFP Fuel Oil Tank for 36 hours of service. − Provide a permanent piping connection from the ACIWA system to the Ultimate Heat Sink Basin for extended

water inventory. − Provide accessible local manual controls for RCIC. − Provide Class 1E power to the plant stack radiation monitor.

4.3 Experiencing ELAP on the TANE ABWR simulator

In order to develop a sense of the US-ABWR response to an ELAP, and in particular the core cooling function of RCIC and response of the plant as experienced by control room operators responsible to implement the FLEX integrated plan, NINA exercised the US-ABWR Demonstration Simulator at the Toshiba America Nuclear Energy (TANE) office in Charlotte, North Carolina to replicate this type of event (Fig. 5). The Simulator is benchmarked and tuned to the Japanese ABWR, and as such it is representative of the STP3&4 design, making it useful for such exercises.

The plant response to ELAP in the Simulator was found to be controlled and protracted, with RCIC slowly cycling to maintain RPV inventory between Level 2 and Level 8 while the Suppression Pool gradually heated up and Containment pressure rose similar to the analyses supporting the FLEX integrated plan. Time frames available for operators to implement the FLEX integrated plan appeared to be realistic based on this exercise, and the licensee was able to gain a further appreciation for the tolerance of the US-ABWR design to such a challenging BDBE.

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

Fig. 5 NINA STP3&4 licensee management exercise the ABWR demonstration simulator at TANE, Charlotte

5. Assessment of US-ABWR capabilities for ELAP

The coping capability of US-ABWR against an ELAP event has been evaluated by conducting a MAAP analysis (Toshiba, 2013) using Version 4.0.7. The primary system was nodalized with the active core modeled with 5 nodes radially and 24 nodes axially in the active fuel zone. The containment model was nodalized as shown in Fig. 6.

Fig. 6 MAAP model of the US-ABWR reactor containment In the MAAP analysis, the CTG is assumed to be failed, and RCIC System and ACIWA (firewater) are available

for core cooling. Inputs for major analytical conditions are listed in Table 2. The sequence of events for the ELAP scenario is summarized in Table 3. Operations staff will conduct a deep load

shed of the Engineered Safety Feature (ESF) DC batteries to extend battery life for instrumentation to monitor key parameters and support RCIC operation during ELAP events. This deep load shed, in conjunction with a cross-tie of the four divisional batteries, will provide more than 72 hours of Class 1E 125V DC power for the STP3&4 ABWRs. In the analysis, therefore, RCIC operation is calculated to continue for about 40 hours until the CST water for RCIC water injection exhausts, based on the assumed initial water volume available.

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[DOI: 10.1299/mej.15-00614]

Table 2 Major analytical conditions Table 3 ELAP sequence of events

Input Value

Initial Reactor Power 3926 MWt

Initial Reactor Pressure 7.17 MPa

D/W Airspace Volume 7350 m3

W/W Airspace Volume 5960 m3

S/P Water Volume 3580 m3

CST Water Volume (assumption)

950 m3

Inner Rupture Disk (COPS) Open Pressure

0.72 MPa

Decay Heat Model ANSI/ANS 5.1-1979

Figs. 6-8 illustrate the reactor pressure, reactor water level and containment pressure responses to the ELAP. It is

assumed that the plant is operating at steady state full power when a BDBEE occurs that results in an ELAP and consequently, a LUHS for the reactor. The plant response is automatic while Operators diagnose and classify the event.

Following the reactor scram, the Main Steam Isolation Valves (MSIVs) close, the reactor pressure (Fig. 7) reaches the SRV opening pressure, and the reactor steam flows to the suppression pool through the relief valves. RCIC starts to function when the reactor water level reaches a low reactor water level (Level 2), and keeps injecting until the reactor water level reaches a high reactor water level (Level 8), at which point RCIC injection is terminated.

When RCIC injects coolant at Level 2, reactor pressure reduces which results in SRV closure. Once RCIC injection is terminated at Level 8, reactor pressure slowly rises to the SRV setpoint, and steam discharge to the suppression pool lowers reactor level back to Level 2. RCIC operation continues manually (following ESF battery load shedding) between Level 2 and Level 8 (Fig. 8) before the CST available water inventory exhausts at around 40 hours. The time duration of each RCIC cycle is initially greater than an hour, and extends as the scenario continues to greater than 4 hours per cycle. Operators are thereby afforded reasonable time to implement the steps of the FLEX integrated plan.

Fig. 7 Reactor pressure response to ELAP

Event Time

SBO, MSIV Closure 0.0 sec

Reactor Scram 1.0 sec

RCIC Injection Start 58 sec

COPS Rupture Disk Open 22 hr

RCIC Trip 40 hr

Manual SRV Open 40 hr

ACIWA Injection Start 40 hr

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Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

Prior to the CST available water inventory becoming depleted, the Operator aligns the ACIWA system to prepare for injection to the RPV. The Operator depressurizes the reactor vessel by opening one safety valve at 40 hours to initiate the ACIWA injection. When the reactor pressure decreases below the shutoff head of the firewater addition system, ACIWA initiates water injection to the reactor vessel through the RHR piping. The Operator maintains the reactor water level within the specified range by controlling the water flow. The ACIWA water injection quickly recovers the reactor water level and prevents core damage.

Fig. 8 Reactor water level response to ELAP The steam flow through the SRVs causes the continuous temperature rise in the suppression pool which results in

the pressurization of the wetwell (Fig. 9). The vacuum breakers, which allow flow from the wetwell to the drywell when the drywell is at a negative pressure with respect to the wetwell, open and the steam-nitrogen mixture flow to the drywell causes the drywell pressure to rise accordingly. The RCIC flow from the CST condenses the steam in the reactor vessel and, with the effect of the vacuum breaker flow, causes temporary pressure decreases in the wetwell between 10 and 20 hours. At around 22 hours, the wetwell pressure reaches 0.72 MPa and the COPS rupture disk opens to discharge the wetwell steam-nitrogen to the atmosphere. The COPS flow with the core cooling by RCIC or ACIWA is sufficient to keep the containment pressure below the allowable limit, and compensates for the LUHS. There is no fission products release predicted because RCIC or ACIWA prevents core damage.

Fig. 9 Containment pressure response to ELAP

6. Conclusions This paper provides insights into the US-ABWR design for safety and robustness to deal with a BDBE such as the

ELAP that occurred at Fukushima Dai-ichi. Because the US-ABWR is an evolutionary design which has been proven

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in Japan and advanced through Americanization to comply with up to date US codes, standards and regulatory requirements, the design incorporated much of the industry philosophy on severe accident mitigation prior to the event at Fukushima Dai-ichi. Evaluation of plant capabilities for coping following an ELAP was completed using a MAAP Version 4.0.7 model of the Reactor/Containment. The analyses confirmed that the basic US-ABWR design is capable of coping with an ELAP without suffering core damage well beyond the timeframe where backup resources would be available from the Regional Response Center. Design enhancements to support the FLEX integrated plan were identified and are not extensive, but will provide additional robustness for associated System, Structure or Components consistent with current regulatory guidance.

Nomenclature

ABWR - Advanced Boiling Water Reactor ACIWA - AC Independent Water Addition System ACRS - Advisory Committee on Reactor Safeguards AFI - Alternate Feedwater Injection System AIA - Aircraft Impact Assessment BDBE - Beyond Design Basis Event BDBEE -Beyond Design Basis External Events CEUS -The Central and Eastern United States COLA - Combined License Application COPS - Containment Overpressure Protection System CST - Condensate Storage Tank CTG - Combustion Turbine Generator DBA - Design Basis Accident DC - Direct Current DCD - Design Control Document DFP - Diesel Fire Pump ECCS - Emergency Core Cooling System ELAP - Extended Loss of AC Power ESF - Engineered Safety Feature FLEX - Diverse and Flexible Coping Strategies (FLEX) FPC - Fuel Pool Cooling and Cleanup System FPS - Fire Protection System FSER - Final Safety Evaluation Report FWST - Fire Water Storage Tank LUHS - Loss of Ultimate Heat Sink MAAP - Modular Accident Analysis Program MSIV - Main Steam Isolation Valve NEI - Nuclear Energy Institute NINA - Nuclear Innovation North America RCIC - Reactor Core Isolation Cooling System R-COLA - Reference Combined License Application RHR - Residual Heat Removal System RPV - Reactor Pressure Vessel SBO - Station Blackout SFP - Spent Fuel Pool SRV - Safety Relief Valve SSC - Seismic Source Characterization TWL - Turbine Water Lubricated Pump by SPX Clyde Union USNRC - United States Nuclear Regulatory Commission

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2© 2017 The Japan Society of Mechanical Engineers

Arai, Yamamoto, Thomas, Mookhoek, Powers and Carter, Mechanical Engineering Journal, Vol.4, No.2 (2017)

[DOI: 10.1299/mej.15-00614]

Product names mentioned herein may be trademarks of their respective companies.

References American National Standards Institute (ANSI)/American Nuclear Society (ANS)-5.1-1979, Decay Heat Power in Light

Water Reactors, Revised American National Standard (1979). Interim Compensatory Measures (ICM) Order, EA-02-026, NRC (2002). MAAP4 Modular Accident Analysis Program for LWR Power Plants User’s Manual. EPRI, Palo Alto, CA (1994–

2005). NRC ADAMS Document ML11126A100, ABWR Design Control Document, Rev. 4, GE Nuclear Energy (1997). NRC ADAMS Document ML12229A174, Interim Staff Guidance (ISG) JD-ISG-2012-01, Compliance with Order

EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation of Beyond Design Basis External Events (2012).

NRC ADAMS Document ML12242A378, NEI 12-06, Rev. 0, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide (2012).

NRC ADAMS Document ML13128A140, NINA Letter to NRC, U7-C-NINA-NRC-130031, Response to Request for Additional Information (includes FLEX Integrated Plan, Rev. 0 dated May 2, 2013) (2013).

NRC ADAMS Document ML15232A128, FSER for South Texas Project, Unit 3 and 4 (Public) (2015). NRC ADAMS Document ML16033A047, Combined License South Texas Project Unit 3 Nuclear Innovation North

America LLC STP Nuclear Operating Company NINA Texas 4 LLC City of San Antonio, Texas, Acting by and through the City Public Service Board Docket No. 52-013 (2016).

NRC ADAMS Document ML16056A428, Combined License South Texas Project Unit 3 Nuclear Innovation North America LLC STP Nuclear Operating Company NINA Texas 3 LLC City of San Antonio, Texas, Acting by and through the City Public Service Board Docket No. 52-012 (2016).

NRC Regulation, 10CFR 50.150, Aircraft Impact Assessment, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0150.html.

NRC Regulation, 10CFR 50.54, Conditions of Licenses, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0054.html.

NRC Regulation, 10CFR 50.63, Loss of All Alternating Current Power, https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0063.html.

NRC Regulatory Guide 1.82, Revision 3, Water Sources For Long-Term Recirculation Cooling Following A Loss-of-Coolant Accident, U.S. Nuclear Regulatory Commission (2003).

Toshiba Design Report NSO-2013-000311/PSNN-2013-0513, Rev. 1, STP3&4 MAAP Analysis for SBO Sequence (2013).

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