ML18310A343

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Part 02 - Final Safety Analysis Report (Rev. 2) - Part 02 - Tier 02 - Chapter 20 - Mitigation of Beyond-Design-Basis Events - Sections 20.01 - 20.04
ML18310A343
Person / Time
Site: NuScale
Issue date: 10/30/2018
From: Bergman T
NuScale
To:
Office of New Reactors
Cranston G
References
NUSCALESMRDC, NUSCALESMRDC.SUBMISSION.6, NUSCALEPART02.NP, NUSCALEPART02.NP.2
Download: ML18310A343 (22)
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NuScale Standard Plant Design Certification Application Chapter Twenty Mitigation of Beyond-Design-Basis Events PART 2 - TIER 2 Revision 2 October 2018

©2018, NuScale Power LLC. All Rights Reserved

COPYRIGHT NOTICE This document bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this document, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copies of the information contained in these reports needed for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of additional copies necessary to provide copies for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS CHAPTER 20 MITIGATION OF BEYOND-DESIGN-BASIS EVENTS . . . . . . . . . . . . . . . . . . . 20.1-1 20.1 Mitigating Strategies for Beyond-Design-Basis External Events . . . . . . . . . . . . . . . 20.1-1 20.1.1 Determining Applicable Extreme External Hazards . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-2 20.1.2 Extended Loss of AC Power and Loss of Ultimate Heat Sink Design Assessment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-3 20.1.3 Mitigating Strategies for an Extended Loss of AC Power Event. . . . . . . . . . . . . . . 20.1-5 20.1.4 Spent Fuel Pool and Reactor Pool Level Instrumentation . . . . . . . . . . . . . . . . . . . . 20.1-7 20.1.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-9 20.2 Loss of Large Areas of the Plant due to Explosions and Fires . . . . . . . . . . . . . . . . . . 20.2-1 20.2.1 Phase 1 - Enhanced Fire Fighting Capabilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.2-1 20.2.2 Phase 2 - Measures to Mitigate Damage to Fuel in the Spent Fuel Pool. . . . . . . 20.2-2 20.2.3 Phase 3 - Measures to Mitigate Damage to Fuel in the Reactor Vessel and to Minimize Radiological Release. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.2-2 20.2.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.2-2 20.3 Integration with Emergency Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.3-1 20.4 Enhanced Emergency Response Capabilities for Beyond-Design-Basis Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.4-1 20.4.1 Enhanced Emergency Plan Staffing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.4-1 20.4.2 Enhanced Emergency Plan Communications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.4-1 20.4.3 Enhanced Emergency Plan Training, Drills, and Exercises . . . . . . . . . . . . . . . . . . . . 20.4-1 20.4.4 Multi-Unit Multi-Source Dose Assessment Capability. . . . . . . . . . . . . . . . . . . . . . . . 20.4-2 Tier 2 i Revision 2

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 20.1-1: Core Cooling Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-11 Table 20.1-2: Containment Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-12 Table 20.1-3: Spent Fuel Pool Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20.1-13 Tier 2 ii Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events CHAPTER 20 MITIGATION OF BEYOND-DESIGN-BASIS EVENTS Following the earthquake and tsunami at the Fukushima Dai-ichi nuclear power plant, the NRC established a senior-level task force referred to as the Near-Term Task Force (NTTF). The NTTF conducted a systematic and methodical review of the NRC regulations and processes to determine if the agency should make safety improvements in light of the events in Japan. As a result of this review, the NRC issued SECY-11-0093 (Reference 20.1-1). SECY-11-0124 (Reference 20.1-2), and SECY-11-0137 (Reference 20.1-3) were issued to establish the NRC staff's recommendations and prioritization of the recommendations.

As a result of NRC's involvement with stakeholders and nuclear industry representatives, the following guidance documents were published:

  • NEI 12-01, "Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities" (Reference 20.1-4)
  • NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide" (Reference 20.1-6)
  • NEI 13-06, "Enhancements to Emergency Response Capabilities for Beyond Design Basis Events and Severe Accidents" (Reference 20.1-7)
  • NEI 14-01, "Emergency Response Procedures and Guidelines for Beyond Design Basis Events and Severe Accidents" (Reference 20.1-8)

Also as part of the NTTF recommendations, rulemaking is in progress (Reference 20.1-9). The proposed rule codifies the requirements in NRC orders for mitigating beyond-design-basis events and for having reliable SFP instrumentation for the licensed and operating fleet of nuclear power plants. In addition, the proposed rule establishes regulatory requirements for an integrated response capability and for enhanced on-site emergency response capabilities. This chapter describes the NuScale Power Plant response to the proposed rule.

Section 20.1 addresses strategies to mitigate BDBEEs, Section 20.2 addresses loss of large area of the plant due to fire or explosion, Section 20.3 addresses procedure integration, and Section 20.4 addresses emergency response.

20.1 Mitigating Strategies for Beyond-Design-Basis External Events This section discusses the mitigating strategies that address an extended loss of alternating current power (ELAP) and loss of normal access to the ultimate heat sink (LUHS) resulting from a BDBEE. JLD-ISG-2012-01, Revision 1 endorsed the approach of NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Revision 2 (Reference 20.1-6) as an acceptable method for COL applicants and licensees to define and deploy strategies to enhance their ability to cope with conditions resulting from BDBEEs. The NuScale design is inherently different from the assumptions or design features considered in the development of the strategies detailed in NEI 12-06. NuScale is providing an alternative method to NEI 12-06 for determining compliance with the proposed MBDBE rule in this section and TR-0816-50797.

Tier 2 20.1-1 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events 20.1.1 Determining Applicable Extreme External Hazards FLEX equipment credited in the mitigation strategies is required to be available following a BDBEE. The extreme external hazards required to be considered for a BDBEE are seismic, flooding, high winds (including applicable missiles), snow, ice, and extreme (cold and high) temperatures. Descriptions of the external hazards design criteria are provided in the following sections.

20.1.1.1 Seismic The seismic design criteria are identified in Section 3.7.1.

COL Item 20.1-1: A COL applicant that references the NuScale Power Plant design certification will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific seismic hazard.

20.1.1.2 External Flooding The external flood design criteria are identified in Section 3.4.2.

COL Item 20.1-2: A COL applicant that references the NuScale Power Plant design certification will determine if a flood hazard is applicable at the site location. If a flood hazard is applicable, then the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific flood (including wave action) hazard.

20.1.1.3 High Winds / Missile Protection The high winds (hurricane and tornado) and applicable missile design criteria are identified in Section 3.3 and Section 3.5.

COL Item 20.1-3: A COL applicant that references the NuScale Power Plant design certification will determine if high wind and applicable missile hazards are applicable at the site location. If high wind and applicable missile hazards are applicable, then the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific high wind and applicable missile hazards.

20.1.1.4 Snow, Ice, and Extreme Cold The snow and ice design criteria are identified in Section 3.8.4. The zero percent exceedance minimum outdoor design dry bulb temperature (i.e., extreme cold) is identified in Table 2.0-1.

COL Item 20.1-4: A COL applicant that references the NuScale Power Plant design certification will determine if snow, ice and extreme cold temperature hazards are applicable at the site location. If snow, ice and extreme cold hazards are applicable, the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific snow, ice or extreme cold temperature hazard.

Tier 2 20.1-2 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events 20.1.1.5 High Temperatures The zero percent exceedance maximum outdoor design dry bulb temperature (i.e. high temperature) is identified in Table 2.0-1.

COL Item 20.1-5: A COL applicant that references the NuScale Power Plant design certification will determine if extreme high temperature hazard is applicable at the site location. If extreme high temperature hazard is applicable, the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific extreme high temperature hazard.

20.1.2 Extended Loss of AC Power and Loss of Ultimate Heat Sink Design Assessment This section discusses the inherent coping capability of the NuScale Power Plant design to maintain the key safety functions following a loss of normal access to the ultimate heat sink (ELAP and an LUHS) event. The key safety functions are maintaining core cooling, containment, and spent fuel pool (SFP) cooling.

20.1.2.1 Definitions An ELAP event is defined as a loss of all alternating current (AC) electric power to the essential and nonessential switchgear buses except for those fed by qualified DC batteries through inverters.

NEI 12-06 (Reference 20.1-6) defines an LUHS as the loss of motive force for UHS flow, i.e., service water or circulating water pumps, with no prospect for recovery. The MBDBEE proposed rule (Reference 20.1-9) further defines an LUHS for passive reactor designs, such as the NuScale design, as a loss of normal access to the normal heat sink.

The LUHS event assumes the water inventory in the UHS remains available following the event, and the piping connecting the UHS to plant systems, which are qualified to survive the applicable external hazards, remains intact.

NEI 12-06 defines the following three phases for developing FLEX strategies (Reference 20.1-6):

  • Phase 1 (initial): cope relying on plant equipment
  • Phase 2 (transition): augment or transition from plant equipment to on-site FLEX equipment and consumables to maintain or restore key functions
  • Phase 3 (final): obtain additional capability and redundancy from off-site equipment until power, water, and coolant injection systems are restored or commissioned 20.1.2.2 Applicable Systems, Structures, and Components The UHS is a large pool of water consisting of the combined water volumes of the reactor pool, refueling pool (RFP) and SFP, as described in Section 9.2.5. Each NuScale Power Module (NPM) is partially immersed in the UHS. Since the NPMs are partially immersed and in direct contact with the UHS, there is no need for UHS piping or motive Tier 2 20.1-3 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events force such as service water or circulating water pumps to provide cooling functions following BDBEE. Therefore, the loss of the UHS, as defined in NEI 12-06 (Reference 20.1-6), is not plausible. No other heat sink is credited for maintaining the key safety functions.

The NPM is a self-contained nuclear steam supply system composed of a reactor core, a pressurizer, and two steam generators integrated within the reactor pressure vessel (RPV) and housed in a compact steel containment vessel (CNV). The plant design relies on passive systems for core cooling during a loss of AC power or DC power.

The reactor coolant system (RCS) design does not require inventory makeup following an ELAP event, but instead relies on maintaining the coolant inventory contained in the RPV and CNV, as described below.

The containment isolation valves (CIVs), which isolate the CNV, fail-safe to their closed position using stored energy. A discussion of the design and function of the CIVs is described in Section 6.2.4.

The decay heat removal system (DHRS) actuation valves fail-safe to the open position, and the main steam isolation valves (MSIVs) and feedwater isolation valves (FWIVs) fail-safe to the closed position to place the DHRS passive condensers in service. These valves fail-safe using stored energy and do not require operator actions or electric power to perform this function. Once a DHRS passive condenser is in service, a closed natural circulation loop is established transferring core decay heat and sensible heat to the UHS. DHRS is further described in Section 5.4.3.

Emergency Core Cooling System (ECCS) valves are also designed fail-safe to their open position. The ECCS valves fail-safe using the differential pressure between the RPV and CNV and do not require operator actions or electric power to perform this function. The function of the ECCS is further described in Section 6.3. With the ECCS valves open and the CIVs closed, a closed natural circulation loop is established in which decay heat is transferred from the core to the UHS. The containment heat removal capability of the CNV is described in Section 6.2.2, and the natural circulation process after ECCS initiation is described in Section 6.3. Once natural circulation is established, core cooling and containment integrity are assured by maintaining sufficient inventory in the UHS. Sufficient inventory to maintain core cooling and containment integrity is available in the UHS for greater than 30 days (Reference 20.1-10) without the need for operator actions.

Instrumentation and its associated indications remain available as needed to confirm proper CIV positions and verify that the natural circulation passive cooling is established. Once the valve positions are verified to be in the proper position and key parameters indicate that passive natural circulation is occurring, the only key parameters that require monitoring for the duration of the event are the UHS and SFP level when below the weir.

The UHS and SFP are monitored using four reliable independent level instruments that are designed to the augmented quality requirements specified in NEI 12-02 (Reference 20.1-5). Two level instruments are positioned in the SFP area, one level instrument in the RFP area, and one level instrument is positioned in the reactor pool Tier 2 20.1-4 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events area. All four instruments monitor the UHS level until the UHS level decreases below the SFP weir when the RFP and reactor pool inventory is separate from the inventory in the SFP.

A robust makeup line with an external connection point for providing inventory to the SFP is available to support SFP makeup following a BDBEE. The makeup to the SFP will overflow into the UHS when level reaches the weir. The makeup line is sized to provide at least 100 gpm of gravity-fed makeup inventory to the SFP. The 100 gpm makeup is greater than the UHS boil off rate. The SFP makeup line is discussed in Section 9.2.5.

During hot shutdown, safe shutdown, transition, and refueling conditions the NPM remains partially immersed in the UHS. The DHRS and ECCS are available or are in service to support core cooling.

Prior to entering the transition mode, the CNV is flooded and the ECCS valves are opened. The NPM remains partially immersed in the UHS while being transported to the RFP. In the RFP the upper module, lower module, and RPV are disassembled while immersed in the UHS.

The reactor core is submerged in water either via the CNV being flooded or directly in the UHS, while in refueling, through the duration of the event. Therefore, passive core cooling remains available in shutdown and refueling conditions.

The SFP will not begin to boil sooner than five days after the start of the ELAP event.

The UHS is sized such that sufficient inventory is available to provide spent fuel cooling for greater than 30 days (Section 9.2.5).

The two level instruments located in the SFP meet the guidance in accordance with NEI 12-02 (Reference 20.1-5). See Section 20.1.4 for more detail on the SFP level instruments.

20.1.3 Mitigating Strategies for an Extended Loss of AC Power Event A summary of the three phases as defined above for NuScale's FLEX strategies is provided in the following sections. The key parameters that are monitored for the FLEX strategies are summarized in Table 20.1-1, Table 20.1-2 and Table 20.1-3.

20.1.3.1 Phase 1 As described in Section 20.1.2, the key safety functions are maintained for greater than 30 days with installed plant equipment. No operator actions or supplemental equipment are necessary to perform these functions.

Core Cooling The core cooling function is automatically established and passively maintained by safety-related equipment, as follows:

Tier 2 20.1-5 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events

  • The containment isolation valves close to maintain RCS inventory.

Maintain Containment The containment function is automatically established and passively maintained by safety-related equipment as follows:

  • The containment isolation valves close to establish containment of the RCS.
  • Containment temperature and pressure control are provided by partial immersion of the CNV in the UHS.

Spent Fuel Pool Cooling The spent fuel pool cooling function is maintained by submergence of the spent fuel in the UHS.

20.1.3.2 Phase 2 A FLEX strategy for a transition phase is not needed for the NuScale Power Plant design.

The coping time utilizing installed plant equipment is greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The initial phase is of sufficient duration to transition directly to the final phase.

20.1.3.3 Phase 3 The baseline coping capability utilizing installed plant equipment is greater than 30 days, and therefore, immediate actions after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are not necessary. Sufficient time is available for detailed planning and procurement of offsite equipment, if necessary to maintain the key safety functions. Due to this extended baseline coping capability, the Phase 3 FLEX strategy is to monitor UHS pool level, using the level instruments described in Section 20.1.4, and add inventory to the UHS via the SFP assured makeup line if necessary.

COL Item 20.1-6: Not used.

COL Item 20.1-7: Not used.

Tier 2 20.1-6 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events 20.1.4 Spent Fuel Pool and Reactor Pool Level Instrumentation 20.1.4.1 Design Bases The design of the four (4) UHS level instruments meet the guidance of NEI 12-02. The design basis functions of the pool level instrumentation are to provide plant personnel with a reliable wide-range water level indication of the UHS level until the UHS level decreases below the SFP weir when the RFP and reactor pool inventory is separate from the inventory in the SFP and reactor pool relative to the following water levels:

  • Level 1 - level that is adequate to support operation of the normal pool cooling systems (See Table 9.2.5-1, Relevant Ultimate Heat Sink Parameters, Minimum level for SFPCS and RPCS suction penetrations),
  • Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the operating deck (See Table 9.2.5.1, Relevant Ultimate Heat Sink Parameters, Minimum level to support radiation shielding),
  • Level 3 - level where stored fuel remains covered and actions to implement make-up water should no longer be deferred (See Table 9.2.5-1, Relevant Ultimate Heat Sink Parameters, Top of spent fuel rack).

During refueling, the NPM is disassembled in the RFP area to allow transferring of new and spent fuel to and from the reactor core. When an NPM is disassembled, the water level in the RFP area will be monitored to ensure the fuel in the reactor is covered during an ELAP event.

The UHS level instruments are designed to withstand external hazards; such as seismic, flooding, high winds (including applicable missiles) extreme temperatures, and snow and ice; without loss of capability to perform their monitoring function.

The UHS instruments are designed to withstand the effects of and to be compatible with the environmental conditions associated with the expected conditions in the Reactor Building during normal operations and an ELAP event.

The UHS level instruments and their power supplies are physically and electrically separated and independent.

20.1.4.2 Description Two (2) wide-range water level instruments are provided for both the SFP and the reactor pool, for a total of four (4) instruments. The wide-range instruments encompass the elevations from the top of the fuel storage racks to near the operating deck. The pool water level instrumentation is consistent with the guidance of NEI 12-02, Revision 1 (Reference 20.1-5).

Instruments Two (2) permanent wide-range instruments monitor the level of the SFP. Two permanent wide-range instruments are also installed to monitor the UHS water level, one in the RP area and one in the RFP area. The instruments transmit signals to the Tier 2 20.1-7 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events main control room. All four of these instruments are capable of monitoring Levels 1 and

2. The two (2) SFP level instruments are capable of monitoring Level 3, when the RP water level is below the weir wall elevation. The two (2) level instruments in the RP and RFP areas are capable of monitoring the level of the water above the fuel in the reactor core when the NPM is disassembled in the RFP during refueling.

Arrangement The two (2) SFP level instruments are arranged in opposite corners of the SFP. The two (2) UHS level instruments are located in the one (1) in the RFP area and the other instrument is located in the RP area. The separation of these instruments is adequate to provide protection from missiles that may be generated within the reactor building from affecting the instruments simultaneously. Protection from external missiles is provided by the Reactor Building structure.

The instrument cables are also separated to provide protection against a single missile damaging both trains.

Mounting The four (4) UHS instruments are seismically mounted such that the instruments will maintain their design configuration during and following an SSE (Seismic Category I).

Qualification The four (4) UHS instruments and associated cabling are environmentally qualified to operate following a BDBEE in the following environmental conditions:

  • SSE seismic event (Seismic Category I)
  • Concentrated borated water environment,
  • Temperature of approximately 212 degrees Fahrenheit and 100 percent relative humidity,
  • Boiling water or steam environment,
  • Radiological conditions existing from a normal refueling with a freshly discharged fuel batch that remains covered with SFP water (Level 3).

Independence The four (4) UHS level instruments are both physically and electrically independent.

Power Supplies The power to the four (4) UHS level instruments is supplied by the highly reliable DC power system (EDSS) with interface through the plant protection system (PPS). Power to the redundant level instruments is from separate bus sources such that the loss of one supply will not result in a loss of power supply function to both divisions of UHS level instrumentation. Additionally, a replaceable battery that is isolated from faults on the normal power supply provides an alternate source of power independent from the Tier 2 20.1-8 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events plant AC and DC power systems. Batteries are designed for easy replacement to power UHS monitoring level instruments indefinitely.

Accuracy The instrument channels are designed to maintain the minimum accuracy following a power interruption or change in power source without recalibration.

Testing The permanently installed UHS level instruments are designed such that testing and calibration can be performed in-situ.

Display The four (4) UHS level instruments transmit signals to the main control room and the remote shutdown panel, and are immediately available to the operators following an event. The instrument signals also initiate high or low level alarms, both locally and in the main control room.

Programs COL Item 20.1-8: A COL applicant that references the NuScale Power Plant design certification will develop procedures, training, and a qualification program for operations, maintenance, testing, and calibration of ultimate heat sink level instrumentation to ensure the level instruments will be available when needed and personnel are knowledgeable in interpreting the information as addressed in Nuclear Energy Institute (NEI) 12-02.

Safety Evaluation The four (4) UHS level instruments are designed to withstand and be protected from natural phenomena such as earthquakes, tornados, hurricanes, floods, tsunamis and seiches without loss of function.

These instruments are also designed to accommodate or be protected from the effects of the postulated environmental conditions, including missiles, pipe whipping, and jet impingement.

The instruments and associated cabling is protected by both physical and electrical separation such that a failure in one channel will leave the other channel functional.

20.1.5 References 20.1-1 U.S. Nuclear Regulatory Commission, "Near-Term Report and Recommendations for Agency Actions Following the Events in Japan,"

Commission Paper SECY-11-0093, July 12, 2011.

Tier 2 20.1-9 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events 20.1-2 U.S. Nuclear Regulatory Commission, "Recommended Actions to be Taken Without Delay from the Near-Term Task Force Report," Commission Paper SECY-11-0124, September 9, 2011.

20.1-3 U.S. Nuclear Regulatory Commission, "Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned," Commission Paper SECY-11-0137, October 3, 2011.

20.1-4 Nuclear Energy Institute, "Guideline for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities," NEI 12-01, Rev. 0, May 2012.

20.1-5 Nuclear Energy Institute, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation," NEI 12-02 Rev. 1, August 2012.

20.1-6 Nuclear Energy Institute, "Diverse and Flexible Coping Strategies (FLEX)

Implementation Guide," NEI 12-06, Rev. 2, December 2015.

20.1-7 Nuclear Energy Institute, "Enhancements to Emergency Response Capabilities for Beyond Design Basis Events and Severe Accidents," NEI 13-06, Rev. 0, September 2014.

20.1-8 Nuclear Energy Institute, "Emergency Response Procedures and Guidelines for Beyond Design Basis Events and Severe Accidents," NEI 14-01, Rev. 0, September 2014.

20.1-9 U.S. Nuclear Regulatory Commission, "Mitigation of Beyond-Design-Basis Events," Federal Register, Vol. 80, No. 219, November 13, 2015, pp.

70610-701647.

20.1-10 NuScale Power LLC, "Mitigation Strategies for Extended Loss of AC Power Event," TR-0816-50797, Rev. 0, November 2016.

Tier 2 20.1-10 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events Table 20.1-1: Core Cooling Parameters Function Parameters for Assuring Function is Parameters for Assuring Function Established is Maintained RCS inventory control RPV water level None(1)

Containment isolation valve positions Reactivity control Neutron flux None(1)

Core inlet temperature Core exit temperature Reactor trip breaker status CVCS containment isolation valve positions DHRS decay heat removal Core exit temperature Spent fuel pool level(2)

DHRS valve positions MSIV positions MSIV bypass isolation valve positions FWIV positions Spent fuel pool level(2)

ECCS decay heat removal ECCS valve positions Spent fuel pool level(2)

Containment water level RPV water level Core exit temperature Spent fuel pool level(2)

(1) By design, once these functions are established, they are maintained indefinitely.

(2) Spent fuel pool level provides indication of UHS level when UHS level is above the SFP weir.

Tier 2 20.1-11 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events Table 20.1-2: Containment Parameters Function Parameters for Assuring Function is Parameters for Assuring Function Established is Maintained Containment isolation Containment isolation valve positions None(1)

Containment heat removal Wide range containment pressure Spent fuel pool level(2)

Spent fuel pool level(2)

(1)

By design, once these functions are established, they are maintained indefinitely.

(2)

Spent fuel pool level provides indication of UHS level when UHS level is above the SFP weir.

Tier 2 20.1-12 Revision 2

NuScale Final Safety Analysis Report Mitigating Strategies for Beyond-Design-Basis External Events Table 20.1-3: Spent Fuel Pool Parameters Function Parameter for Assuring Function is Parameter for Assuring Function is Established Maintained Spent Fuel Pool Cooling Spent fuel pool level Spent fuel pool level Tier 2 20.1-13 Revision 2

NuScale Final Safety Analysis Report Loss of Large Areas of the Plant due to Explosions and Fires 20.2 Loss of Large Areas of the Plant due to Explosions and Fires NRC regulation 10 CFR 50.54(hh)(2)1 requires licensees to develop and implement guidance and strategies to maintain or restore core cooling, containment, and spent fuel cooling under the circumstance associated with the loss of large areas of the plant due to explosion or fire (LOLA). The strategies that are required to be addressed are: (i) fire fighting; (ii) operations to mitigate fuel damage; and (iii) actions to minimize the radiological release.

Technical Report TR-0816-50796 (Reference 20.2-1) documents an assessment evaluating the NuScale Power Plant response to a LOLA event using the guidance in Nuclear Energy Institute (NEI) 06-12 (Reference 20.2-2). The report defines LOLA criteria and identifies the design features that meet those criteria and expected combined license (COL) applicant requirements.

The analysis was performed using the three phases recommended in NEI 06-12 (Reference 20.2-2): Phase 1 - Enhanced Fire Fighting Capabilities; Phase 2 - Measures to Mitigate Damage to Fuel in the Spent Fuel Pool; and Phase 3 - Measures to Mitigate Damage to Fuel in the Reactor Vessel and to Minimize Radiological Release.

This section describes the results of the assessment with no Security Related Information.

20.2.1 Phase 1 - Enhanced Fire Fighting Capabilities The firefighting response to a LOLA event includes the operational aspects of responding to explosions or fire including items such as prearranging for the involvement of outside organizations, planning and preparation activities (e.g., pre-positioning equipment, personnel, and materials to be used for mitigating the event), and developing procedures and training for the event.

The fire protection system includes an underground yard fire main loop. Hydrants are provided on the yard fire main loop in accordance with the National Fire Protection Association at intervals up to 250 feet and located on each side of the Reactor Building. The lateral to each hydrant is controlled by an isolation valve. There are several connections in the yard main that can support supplying the yard main using a portable diesel-driven pump and several valves that can isolate damaged section(s) when required. The fire protection system is described in more detail in Section 9.5.1.

COL Item 20.2-1: A COL applicant that references the NuScale Power Plant design certification will develop enhanced firefighting capabilities by implementing the guidance in NRC guidance document "Developing Mitigating Strategies/Guidance for Nuclear Power Plants to Respond to Loss of Large Areas of the Plant in Accordance with B.5.b of the February 25, 2002, Order" dated February 25, 2005 (Reference 20.2-3).

The enhanced firefighting capabilities should address the expectation elements listed in Section 4.1.3 of the Technical Report TR-0816-50796 (Reference 20.2-1).

1. Regulation 10 CFR 50.54(hh)(2) may be moved to new regulation 10 CFR 50.155, presently in rulemaking process reference Federal Register notice Vol. 80, No. 219 pages 70610-70647 Tier 2 20.2-1 Revision 2

NuScale Final Safety Analysis Report Loss of Large Areas of the Plant due to Explosions and Fires 20.2.2 Phase 2 - Measures to Mitigate Damage to Fuel in the Spent Fuel Pool Additional spent fuel cooling strategies are not required, in accordance with the guidance in NEI 06-12 (Reference 20.2-2).

20.2.3 Phase 3 - Measures to Mitigate Damage to Fuel in the Reactor Vessel and to Minimize Radiological Release The generic pressurized water reactor key safety functions identified in NEI 06-12 (Reference 20.2-2) were developed based on a traditional pressurized water reactor plant design. These key safety functions are applicable to the NuScale Power Plant. The key safety functions that must be evaluated for a LOLA event are:

  • RCS heat removal
  • containment isolation
  • containment integrity
  • release mitigation A primary and alternate means for RCS inventory control, RCS heat removal, containment isolation, and containment integrity are maintained for the NuScale Power Plant with installed plant capabilities. An additional means or strategy of minimizing a potential radiological release from a LOLA event is needed to augment the NuScale Power Plant installed plant capabilities.

COL Item 20.2-2: A COL applicant that references the NuScale Power Plant design certification will provide a means for water spray scrubbing using fog nozzles and the availability of water sources, and address runoff water containment issues (sandbags, portable dikes, etc.) as an attenuation measure for mitigating radiation releases outside containment.

20.2.4 References 20.2-1 NuScale Power, LLC, "Loss of Large Areas Due to Explosions and Fires Assessment," TR-0816-50796 (Security Related Information).

20.2-2 Nuclear Energy Institute, "B.5.b Phase 2 & 3 Submittal Guideline," NEI 06-12, Rev. 3, September 2009.

20.2-3 U.S. Nuclear Regulatory Commission, Developing Mitigating Strategies/

Guidance for Nuclear Power Plants to Respond to Loss of Large Areas of the Plant in Accordance with B.5.b of the February 25, 2002, Order, February 25, 2005.

Tier 2 20.2-2 Revision 2

NuScale Final Safety Analysis Report Integration with Emergency Procedures 20.3 Integration with Emergency Procedures COL Item 20.3-1: A COL applicant that references the NuScale Power Plant design certification will ensure that the severe accident management guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines are integrated with the emergency operating procedures consistent with Recommendation 8.1 of SECY-11-0093, "Near-Term Report and Recommendations for Agency Actions Following the Events in Japan."

Tier 2 20.3-1 Revision 2

Enhanced Emergency Response Capabilities for Beyond-Design-Basis NuScale Final Safety Analysis Report Events 20.4 Enhanced Emergency Response Capabilities for Beyond-Design-Basis Events This section describes the provisions implemented to enhance the emergency response capabilities as they relate to emergency response staffing, communication, training, drills, exercises, and multi-source dose assessment capabilities for beyond-design-basis events.

A description of the emergency plan is provided in Section 13.3.

The beyond-design-basis event (BDBE) emergency response enhancements of Section 20.4.1, Section 20.4.2, and Section 20.4.3 are not required to be part of the plant's emergency plan, and therefore do not require the change control provisions of 10 CFR 50.54(q).

20.4.1 Enhanced Emergency Plan Staffing COL Item 20.4-1: A COL applicant that references the NuScale Power Plant design certification will perform an analysis that demonstrates the emergency response organization staff has the ability to implement the strategies of the emergency operating procedures, severe accident mitigation guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines. The analysis will be performed with the offsite response organization access to onsite being impeded. The event shall be a loss of all onsite and offsite alternating current power and loss of normal access to the ultimate heat sink.

COL Item 20.4-2: A COL applicant that references the NuScale Power Plant design certification will develop a supporting emergency response organization structure with defined roles and responsibilities to implement the strategies of the emergency operating procedures, severe accident mitigation guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines.

20.4.2 Enhanced Emergency Plan Communications The installed plant communication capabilities are described in Section 9.5.2.

COL Item 20.4-3: A COL applicant that references the NuScale Power Plant design certification will develop and describe at least one onsite and one offsite communications system capable of remaining functional during an extended loss of alternating current power including the effects of the loss of the local communications infrastructure.

20.4.3 Enhanced Emergency Plan Training, Drills, and Exercises COL Item 20.4-4: A COL applicant that references the NuScale Power Plant design certification will develop, implement, and maintain the training and qualification of personnel that perform activities in accordance with diverse and flexible coping strategies support guidelines (FSGs), severe accident mitigation guidelines, and extensive damage mitigation guidelines. The training and qualification on these activities will be developed using the systems approach to training as defined in 10 CFR 55.4 except for elements already covered under other NRC regulations.

Tier 2 20.4-1 Revision 2

Enhanced Emergency Response Capabilities for Beyond-Design-Basis NuScale Final Safety Analysis Report Events COL Item 20.4-5: A COL applicant that references the NuScale Power Plant design certification will develop drills or exercises that demonstrate the ability to transition to one or more of the strategies and guidelines of the emergency operating procedures, diverse and flexible coping strategies support guidelines (FSGs), extensive damage mitigation guidelines, and severe accident mitigation guidelines using only the station communication equipment designed to be available following an extended loss of alternating current including effects of the loss of the local communications infrastructure.

20.4.4 Multi-Unit Multi-Source Dose Assessment Capability COL Item 20.4-6: A COL applicant that references the NuScale Power Plant design certification will develop and describe the means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials to the environment including releases from reactor core and spent fuel pool sources.

Tier 2 20.4-2 Revision 2

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