Text

LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, PM-0519-65372, Revision 0
	ML19134A074
Person / Time
Site:	NuScale
Issue date:	05/09/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of New Reactors
References
	LO-0519- 65373
	Download: ML19134A074 (78)
	v • d • e

L0-0519-65373 May 09, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Presentation Materials Entitled "ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation," PM-0519-65372, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee meeting on May 14 and 15, 2019.

The materials support NuScale's presentation of Chapter 19, "Probabilistic Risk Assessment and Severe Accident Evaluation," of the NuScale Design Certification Application. is the nonproprietary presentation entitled "ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation," PM-0519-65372, Revision 0.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions, please contact Rebecca Norris at 541-452-7539 or at rnorris@nuscalepower.com.

Since~~_____-

/~-~~

Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Robert Taylor, NRC, OWFN-7H4 Michael Snodderly, NRC, TWFN-2E26 Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Rani Franovich, NRC, OWFN-8H12 : "ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation," PM-0519-65372, Revision O NuScale Power, LLC 1100 NE Circle Blvd., Suite200 Corvallis, Oregon 97330 Office 541.360-0500 Fax541.207.3928 www.nuscalepower.com "ACRS Subcommittee Presentation: NuScale FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation," PM-0519-65372, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com L0-0519-65373

PM-0519-65372 Revision: O NuScale Nonproprietary ACRS Subcommittee Presentation:

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NuScale Nonproprietary Chapter 19 Section

. 19.0 19.1 19.2 19.3 19.4 19.5 2

PM-0519-65372 Revision: O Title Probabilistic Risk Assessment and Severe Accident Evaluation Probabilistic Risk Assessment Severe Accident Evaluation Regulatory Treatment of Nonsafety Systems Comment

Overview Level 1,2

Thermal hydraulic &.

phenomenological

analyses No RTNSS SSCs Strategies and Guidance to Address Loss of Add d.

resse 1n L~rge Areas of the Plant due to Explosions and Cha ter 20

. Fires p

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NuScale Nonproprietary ACRS Subcommittee Presentation:

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NuScale Nonproprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris Supervisor, Licensing 4

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NuScale Nonproprietary Section 19.0: Probabilistic Risk Assessment and Severe Accident Evaluation

Developed in accordance with applicable regulations, regulatory guidance, and industry standards

Performed for a single module

Considered all modes of operation for both internal and external initiating events

Provides risk insights including those related to risk-significant systems, components, human actions, relevant programs (e.g., RTNSS, SAMOA), and multiple module risk

PRA demonstrates that the NuScale design exceeds NRC safety goals with significant margin 5

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NuScale Nonproprietary Section 19.1: Probabilistic Risk Assessment

Objective: to assess risks associated with all modes and all hazards for a single NuScale Power Module (NPM)

Level-1 (CDF) and Level-2 (LRF)

- Full power, internal events (FP-1 E)

- Low power and shutdown (LPSD)

Include crane failure

- Internal fire

- Internal flood

- External flood

- High winds

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NuScale Nonproprietary PRA Quality Process

NuScale PRA quality procedure

- Follows guidance provided in NRC Regulatory Guide 1.17 4

NuScale PRA follows guidance provided by

- ASME/ANS PRA standard

- NRC Regulatory Guide 1.200 and Interim Staff Guidance 028

Each PRA notebook reviewed for conformance with PRA standard

- Self-assessment documented by notebook authors

- Self-assessment independently reviewed/verified by outside consultants

PRA reviewed by outside, independent expert panel 7

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NuScale Nonproprietary PRA Expert Peer Review Group

Separate and independent from PRA standard self-assessment reviewers

Expert review group members:

- George Apostolakis (chairman)

- Mark Cunningham

- Rick Grantom

- Dave Moore

- Per Peterson 8

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NuScale Nonproprietary Expert Panel Findings

Review group authored a final report

- No major concerns or objections 9

PM-0519-65372 Revision: O

Minor points that were raised include

>> NuScale multi-module risk approach represents an important "first step" in advancing the state-of-the-art

>> There are more detailed and sophisticated HRA methods available compared to what was done in the NuScale PRA

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NuScale Nonproprietary Independent Self Assessment

External review of the NuScale PRA self-assessment against the high level and supporting requirements of the ASME PRA Standard

In general, there was agreement, and in fact, in some cases, a higher capability category than identified was believed to be met. However, there were also some instances of a lack of concurrence, and possible enhancements were provided

NuScale was able to incorporate those recommendations into the design certification PRA 10 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Initiating Event Analysis

Multiple sources of input used to identify potential initiating events (I Es)

- NuScale design-specific master logic diagram

- NuScale design-specific simplified system-level failure modes and effects analysis (FMEA)

- Traditional lists of PRA initiating events

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NuScale Nonproprietary Full Power Internal Initiating Events

eves LOeA (injection line) inside containment vessel (eNV)

eves LOeA (injection line) outside eNV

eves LOeA (discharge line) outside eNV

Spurious opening of Eecs valve

Loss of DC power

Loss of offsite power

Steam generator tube failure

LOCA (other) inside CNV

Secondary-side line break (i.e., feedwater or main steam)

General reactor trip

Loss of support system (e.g., instrument air, AC power bus) 12 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Accident Sequence Analysis

Initiating events and subsequent plant responses evaluated

Key safety functions identified

- Fuel assembly heat removal, reactivity control, containment integrity

End states of the accident sequences defined

- Level-1: core damage frequency (CDF)

- Level-2: large release frequency (LRF)

Event trees constructed for each of the initiating events associated with system successes or failures to accomplish the applicable safety functions 13 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Success Criteria

The Level 1 PRA overall success criterion is the prevention of core damage, defined by maintaining a peak cladding temperature less than 2,200 degrees Fahrenheit

- This is demonstrated for a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months mission time

System success criteria is determined by the minimum system availability required to prevent core damage

The Level 1 success criteria evaluation is built upon a comprehensive simulation suite of more than 40 unique accident sequences

The Level 2 success criterion is large release defined as a source term resulting in acute whole body 200 rem dose to the maximally exposed individual stationary at the reactor site boundary for 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 14 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Success Criteria

PRA success criteria simulations use NuScale's safety-related NRELAPS code with an input model that starts with NuScale's safety-related input model

- The PRA simulations augment the safety-related input model with additional nonsafety-related models for beyond-design-basis phenomena 15 PM-0519-65372 Revision: O

Chemical and volume control system (CVCS) and containment flooding and drain system (CFDS) models

Multi-dimensional core thermal hydraulic and neutronic models are used to simulate complex beyond design basis transients such as ATWS Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Human Reliability Analysis

Human actions are not credited in the evaluation of design basis events

- Human actions only relevant to beyond design basis analyses

Human error probabilities for beyond design basis events based on methodologies provided in NUREG/CR-4772 and NUREG/CR-6883

- Latent human errors and recovery actions

As a modeling convenience, when quantifying the PRA model, the bounding human error probability of the complete set of post-initiator human failure events, is used for all independently modeled post-initiator human failure events

Risk significant human action candidates input to D-RAP

- Operator fails to initiate eFDS injection

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NuScale Nonproprietary Post-initiator HEPs in PRA Quantification

Post-initiator final human error probability (HEP) values range from 4E-3 to 2E-5

- Time available (based on bounding scenarios) for human actions range from 30 minutes to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

To simplify the quantification of the PRA model, bounding value of the set of HEPs used to quantify all post-initiator HEPs Event Description Value IBI HEP01 HEP02 HEP03 17 PM-0519-65372 Revision: O Human error probability for first HFE in cutset

4.0E-03 '.! 10;

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Human error probability for second HFE in cutset 1.SE-01 3

Human error probability for third HFE in cutset 0.5 i

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NuScale Nonproprietary NuScale PRA Human Errors Modeled (Pre-Initiator)

Name Description eFOS--HFE-0001A-UTM-N eFOS--HFE-0002A-UTM-N eveS--HFE-0001 A-UTM-N eveS--HFE-0002A-UTM-N EHVS--HFE-0001 A-UTM-N ELVS--HFE-0001 A-UTM-N ELVS--HFE-0002A-UTM-N MPS---HFE-0001 A-UTM-S Operator misaligns MOP 0001A eFOS train A manual valves during test and maintenance Operator misaligns MOP 0001 B eFOS train B manual valves during test and maintenance Operator misaligns MOP 0002A eves train A manual valves during test and maintenance Operator misaligns MOP 00028 eves train B manual valves during test and maintenance Operator misaligns eTG 0003X EHVS combustion turbine generator during test and maintenance Operator misaligns OGN 0001X ELVS standby diesel generator during test and maintenance Operator misaligns OGN 0002X ELVS standby diesel generator during test and maintenance Operator miscalibrates safety function modules during test and maintenance 18 1 ________________________________

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NuScale Nonproprietary NuScale PRA Human Errors Modeled {Post-Initiator)

Name Description Context CFDS--HFE-0001 C-FOP-N CVCS--H FE-0001 C-FOP-N CVCS--H FE-0002C-FOP-N ECCS--H FE-0001 C-FTO-N EHVS--H FE-0001 C-FTS-N ELVS--HFE-0001 C-FTS-N 19 PM-0519-653 72 Revision: 0 Operator fails to unisolate and initiate CFDS injection Operator fails to unisolate and initiate eves injection Operator fails to locally unisolate and initiate eves injection Operator fails to open ECCS valves Operator fails to start/load combustion turbine generator Operator fails to start/load backup diesel generator Copyright 2019 by NuScale Power, LLC.

Used for LOCA-OC (2 IEs), SGTFs, and transients (1 IE)

Used for LOCA-IC (3 IEs), LOCA-OC (letdown) (1 IE), transients (1 IE) and secondary steam line break (1 IE) upon failure of ECCS, and SGTFs Local unisolation due to lack of control from a partial loss of DC power Backup action to MPS autofunction failure Backup local action to control room initiation failure during loss of offsite power Backup local action to control room initiation failure during loss of offsite ower N ~!J.?.s~.L.r Template #: 0000-21727-F01 RS

NuScale Nonproprietary Data Sources

Industry information (e.g., NUREG/CR-6928, LERs) where applicable

- Common cause failure (CCF) modeling based NUREG/CR-5497

Design-specific analyses

- Passive safety system reliability (i.e., ECCS, DHRS)

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NuScale Nonproprietary Quantification

Quantification of the PRA model was performed with the SAPHIRE code

- Including CCF models, failure data correlations and uncertainty analyses

Using the ASME/ANS PRA Standard convergence criterion, a truncation value of 1 E-15 per module year was used for the CDF 21 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Uncertainty Analyses

Addressed using both quantitative uncertainty analyses and sensitivity studies

- SAPHIRE PRA code has capability for propagating parametric uncertainties

- Sometimes augmented using sensitivity studies (e.g., SGTF)

- Thermal hydraulic analyses typically use bounding inputs

Uncertainty addressed in all modes and all hazards of single module PRA

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NuScale Nonproprietary Parametric Uncertainty

The data parameters include initiating event frequencies, component failure probabilities, CCF events and their alpha factors, and human error probabilities

- Initiating event frequencies that rely on generic industry data were assigned an expanded uncertainty distribution (i.e., log normal error factor= 10)

SAPHIRE has the built-in ability to perform an uncertainty analysis

- Includes correlating failure probabilities

After cutsets were generated in SAPHIRE, an uncertainty analyses was performed using the Latin Hypercube uncertainty sampling methodology.

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NuScale Nonproprietary Importance

Systems

- CNTS (containment isolation valves), ECCS, MPS, and UHS

Components ECCS RWs and RRVs DHRS actuation valves RSVs

- eves and CES containment isolation valves

- Combustion turbine generator

Other events and initiators (FV>20°/o)

RBC, LOCA inside CNV, LOCA outside CNV, LOOP, internal fires, internal flood

Human actions (FV>20°/o)

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NuScale Nonproprietary Sensitivity Studies Parameter I, Base Case I

HEP Parameter Change All HEPs set to FALSE CDF Result LRF Result 2.7E-10 2.0E-10 1.7E-11 1.0E-11

~-------

/ HEP CCF iCCF LOOP-IE All HEPs set to TRUE All CCFs set to FALSE All CCFs set to max value of 0.002 LOOP frequency set to 1 per year (base = 3.1 E-2 per year) 3.2E-8 5.4E-12 4.2E-6 2.2E-9 2.BE-9 1

1.2E-12 3.7E-8 1.?E-11 r--------- ------------ -*----**- ----------- ----

! LOCA-IC-IE LOCA inside CNV frequency increased 1 order of magnitude 3.4E-10 1.?E-11 SGTF-IE SGTF frequency increased to generic value 2.BE-10 2.2E-11

, ECCS &

ECCS and DHRS passive heat transfer

3.2E-10 1.?E-11 2.8E-10 1.?E-11

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NuScale Nonproprietary Level 2 Methodology

Analysis indicates that the only applicable containment vessel (CNV) failure mechanisms are containment bypass events and failure of containment isolation

No bridge trees or Level 1 plant damage state binning

- Level 2 event tree is directly linked to the Level 1 event trees Core Damage Sequences CD LEVEL2-Ef 26 PM-0519-65372 Revision: O 0

Core Damage Cutset Mapped to Containment Isolation - CIVs Release Size Close CD-T01 CNTS-T01 0

1 I

2 I

0 0

3 I

End State Comments (Phase - PH1)

(Phase - PH1)

CD Core Damage NR RC1 :CD with Isolation LR RC2:CD with Release w

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NuScale Nonproprietary External Hazards

External events are evaluated using Level 1 PRA model and the following methodologies

- Internal fire: NUREG/CR-6850

- Internal flood: Part 3 of ASME/ANS RA-Sa-2009

- External flood: Part 8 of ASME/ANS RA-Sa-2009

- High winds: Part 7 of ASME/ANS RA-Sa-2009

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NuScale Nonproprietary Seismic Risk Evaluation

NuScale performed a PRA-based seismic margin assessment (SMA)

Design-specific fragility calculations were performed for SSCs that contribute to the seismic margin

- Consulted with Simpson, Gumpertz & Heger, Rizzo Associates, and Stevenson and Associates

Generic capacities with design-specific response factors were used for other SSCs

DC/COL-ISG-020 seismic margin goal: high confidence of low probability of failure (HCLPF) value of 1.67 times the certified seismic design response spectra (CSDRS)

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NuScale Nonproprietary SMA Methodology

PRA-based SMA uses internal event logic, seismically-induced initiators, and maps seismic failures to random failures

HCLPF: high confidence (95o/o) of low probability (5%) of failure

- HCLPF can also be interpreted as a 1 °/o probability of failure at the mean (or best-estimate) confidence level (i.e., at the HCLPF PGA there is a 1 % probability of core damage)

- Evaluated at the sequence level using min-max criteria

Seismic margin determined by those seismic failures that would result in a conditional core damage probability of greater than 1 °/o

Structural fragilities evaluated for those SSCs that contact the module, are located above the module, or where collapse might damage the module (which is assumed to result in core damage) 29 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Seismic Risk Evaluation Seismic plant response

Induced initiator event trees Structural failures LOCAs Loss of offsite power

Seismic failure mapping No pre-screening (all PRA cutsets included in the SMA)

Evaluated at 14 ground motion levels ranging from 0.05g to 3.5g SMA results

Plant level HCLPF: 0.88g

Structural failures dominate Crane Exterior walls Bay walls Module supports

At lower PGAs, LOOP combined with random failures dominate results

Negligible seismic risk from low power and shutdown states 30 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Select NuScale SMA Structures RXB wall 31 PM-0519-65372 Revision: O Bioshield Bay Walls Refueling Machine

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NuScale Nonproprietary Fragility Calculation Parameters

Design calculations for demand/capacity (D/C) ratio inputs

- Uses bounding, conservative values

- For fragility purposes, design calculates are adjusted to median-centered values, uncertainties quantified

Structural response factor variables

- Ground motion response

- Damping

- Modeling

- Mode combination

- Time history simulation

- Foundation-structure interaction

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NuScale Nonproprietary Fragility Calculation Parameters

Capacity variables

- Strength

- Ductility

Earthquake scale factor (ESF)

- Used in wall calculations, where capacity changes with demand

- Ratio by which the seismic demand must increase for overall demand to equal capacity

Static demand + ESF

seismic demand = static capacity +/- ESF

dynamic capacity (sign is dependent on load in compression/ tension)

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NuScale Nonproprietary Low Power and Shutdown

Potential initiating events are those considered for full power and those unique to LPSD

- Reduced inventory (drain down) events not applicable

No reduced inventory operations in the NuScale design

Evaluated external events shown to be not important

Dropped module event most significant CDF contributor

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NuScale Nonproprietary Dropped Module Evaluation

Drop probability developed based on conceptual reactor building crane design

Core damage conservatively assumed for dropped module

- For a horizontal module the core partially uncovers

- Containment assumed to fail in a manner that prevents pool water incursion but allows radionuclide release

Maximum radiological release much less than large release due to pool scrubbing effect

Up to two operating modules theoretically could be struck by free-falling module, potentially inducing LOCA or transient in struck module 35 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Postulated Dropped Module Impacts

Potential damage to the decay heat removal system (DHRS) because the heat exchangers are located external to the containment and face central pool channel

- Likelihood is an insignificant contributor to the modeled frequency of secondary side line break initiating event

Potential damage to the chemical and volume control system (CVCS) piping where the piping penetrates the bay wall as a result of movement of the struck module

- Likelihood is an insignificant contributor to the modeled frequency of the eves pipe break outside containment initiating event

Considering the probability of a load drop, the contribution of a potential module drop to the initiating event frequencies of an operating module is judged to be negligible both in absolute terms and in comparison to the frequency of a randomly occurring initiating events 36 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Multiple Module Evaluation

Each NPM comprises a separate, independent RPV and CNV, and is serviced by separate, independent safety systems

Systematic evaluation performed per SRP 19.0

Single module PRA with bounding multi-module adjustment factors (MMAF) applied to each and every basic and initiating event

- MMAF value of 1.0 for SSCs shared amongst multiple modules and plant wide initiating events (e.g., LOOP)

- MMAF values from 0.1 to 0.3 for SSCs with potential coupling mechanisms between modules (e.g., potential for common cause failures)

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NuScale Nonproprietary Level 1 Insights

NuScale design exceeds NRC core damage frequency safety goal with significant margin

- Full power internal event CDF 3.0E-10/mcyr

External initiator CDFs: 1.0E-09 to 6.1 E-11/mcyr

- LPSD CDF dominated by module drop event: 8.8E-08/mcyr

- Focused PRA CDF (no credit for nonsafety-related systems): 3.1 E-06/mcyr

Approximately equivalent to a long-term station blackout with no recovery of ac power

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NuScale Nonproprietary Level 2 Insights

NuScale design exceeds NRC large release frequency safety goal with significant margin

-Full power internal event LRF 2.3E-11/mcyr

External initiator LRFs: 4.3E-11 to <1 E-15/mcyr

- Module drop event does not result in large release

- Focused PRA LRF (no credit for nonsafety-related systems): 1.6E-07/mcyr

Approximately equivalent to a long-term station blackout with no recovery of ac power

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NuScale Nonproprietary Level 1 Key Insights (1 of 2)

Design Feature/Insight Failure to scram events (ATWS) do not lead directly to core damage.

Passive heat removal capability is sufficient to prevent core damage if a reactor safety valve (RSV) cycles.

Post-accident heat removal through steam generators or decay heat removal system (DHRS) is unnecessary if RSVs cycle.

Passive, fail-safe emergency core cooling system (ECCS) functions to preserve RCS inventory, which is sufficient to allow core cooling without RCS makeup from external source.

40 PM-0519-65372 Revision: O Comment Core characteristics result in ATWS power levels that are comparable to decay heat levels. Heat transfer from the containment vessel (CNV) to reactor pool is adequate to prevent core damage and most ATWS sequences require approximately the same system success criteria as non-ATWS events.

RSV cycling transfers adequate RCS water to the CNV to allow heat transfer through the RPV to the CNV and ultimately to reactor pool to remove decay heat.

The steam generators and DHRS provide effective heat removal paths to prevent core damage, but are unnecessary if RSV cycling allows heat transfer to reactor pool. Passive, fail-safe DHRS provides a natural circulation closed loop system that does not require pumps, power, or additional water.

The ECCS consists of 5 valves that fail-safe on a loss of power and provides a natural circulation path through the core and CNV, thus providing heat transfer to the reactor pool. The closed-loop system does not need additional inventory.

Copyright 201 9 by NuScale Power, LLC.

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NuScale Nonproprietary Level 1 Key Insights (2 of 2)

Design Feature/Insight Containment isolation preserves RCS inventory for core cooling without external makeup.

Passive, fail-safe safety systems (ECCS, DHRS, RSVs) include redundancy and do not need support systems, including electric power or operator actions.

There are no risk significant, post-initiator human actions associated with the full-power PRA.

Risk significant structures, systems and components (SSCs) for external events are largely the same as those found risk significant for internal events.

Active systems providing makeup inventory to the RPV are not risk significant.

Comment Containment isolation eliminates the potential for breaks outside of containment to result in loss of RCS inventory. For breaks inside of containment, containment isolation is not necessary to support passive core cooling and heat removal.

Safety-related mitigating systems are fail-safe on loss of power and do not require supporting systems such as lube oil, air or HVAC to function. No single failure results in a loss of system function.

No operator actions, including backup and recovery actions, are risk significant to the CDF because of passive system reliability and fail-safe system design.

The module response to external events is comparable to the response to internal event due to the passive features of the design and independence from support systems such as power.

Additional systems and components have been identified as risk significant for external events due to a conservative evaluation.

Inventory addition is possible by the active systems chemical and volume control system (CVCS) and containment flooding and drain system (CFDS). Due to the reliability of the passive safety systems, the active systems providing this backup function were found not to be risk significant.

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NuScale Nonproprietary Section 19.1 COL Items Item Number Description COL Item 19.1-1 A COL applicant that references the NuScale Power Plant design certification will identify and describe the use of the probabilistic risk assessment in support of licensee programs being implemented during the COL application phase.

COL Item 19.1-2 A COL applicant that references the NuScale Power Plant design certification will identify and describe specific risk-informed applications being implemented during the COL application phase.

COL Item 19.1-3 A COL applicant that references the NuScale Power Plant design certification will specify and describe the use of the probabilistic risk assessment in support of licensee programs during the construction phase (from issuance of the COL up to initial fuel loading).

COL Item 19.1-4 A COL applicant that references the NuScale Power Plant design certification will specify and describe risk-informed applications during the construction phase (from issuance of the COL up to initial fuel loading).

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l NuScale Nonproprietary Section 19.1 COL Items Item Number Description COL Item 19.1-5 A COL applicant that references the NuScale Power Plant design certification will specify and describe the use of the probabilistic risk assessment in support of licensee programs during the operational phase (from initial fuel loading through commercial operation).

COL Item 19.1-6 A COL applicant that references the NuScale Power Plant design certification will specify and describe risk-informed applications during the operational phase (from initial fuel loading through commercial operation).

COL Item 19.1-7 A COL applicant that references the NuScale Power Plant design certification will evaluate site-specific external event hazards (e.g.,

liquefaction, slope failure), screen those for risk-significance, and evaluate the risk associated with external hazards that are not bounded by the design certification.

COL Item 19.1-8 A COL applicant that references the NuScale Power Plant design certification will confirm the validity of the "key assumptions" and data used in the design certification application probabilistic risk assessment (PRA) and modify, as necessary, for applicability to the as-built, as-operated PRA.

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NuScale Nonproprietary ACRS Subcommittee Presentation:

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NuScale. Non proprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris Supervisor, Licensing 45 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Section 19.2: Severe Accident Evaluation

Accident sequences resulting in core damage are evaluated in the Level 2 PRA for potential to challenge containment vessel (CNV) integrity and result in a large radionuclide release

- Large release defined as source term resulting in acute whole body 200 rem dose to the maximally exposed individual stationary at the reactor site boundary for 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months

- MACCS off-site consequence calculations demonstrate that sequences with intact CNV are not large release

- CNV bypass accidents counted as large release (simplification for convenience)

Potential challenges to CNV integrity identified from SRP, PRA standard, and NUREGs

There are no unique phenomenological challenges that are introduced by the NuScale design 46 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Use of MELCOR

Provides a best estimate evaluation of severe accident challenges to CNV

Informs conservative evaluations of severe accident challenges Provides a physical basis for parameters 47 PM-0519-65372 Revision: O

Timing of core damage, core relocation

Quantity of relocated material, composition of relocated material

System pressures, temperatures, quantity of hydrogen produced Evaluations use limiting values from database of simulations that each involve bounding/conservative simplifications

End of cycle decay heat load

DHRS not credited to slow down accident progression Evaluations also consider parameters that bound all results observed from database of simulations

100% of fuel U02 relocates at first observed relocation time from database

Assume debris is molten, pure U02 composed of no filler materials (e.g., steel, zirconium)

No credit for water in lower plenum at time of relocation Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary MELCOR Model Development

Thermal-hydraulics modeling developed from NRELAPS model Matching elevations, volumes, flow areas, frictional losses, heat structure material, surface area, thickness, heated diameters, etc

Benchmarking of steady-state operation and transients demonstrate reasonable to excellent agreement with NRELAPS Goal is to approximately match NRELAP5 accident simulation to the point of core damage and then extend simulation into severe accident space

Severe accident modeling based on appropriate and accurate modeling of NPM design characteristics Decay power curve, core component masses and locations, radionuclide inventory, core flow geometry

Incorporates modeling best practices from MELCOR code development staff and industry leading subject matter experts State-of-the-Art Reactor Consequence Analyses (SOAR CA) reports MELCOR guides, manuals, assessments 48 PM-0519-65372 Revision: o Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary In-Vessel Retention (IVR)

Conservative analysis demonstrates that RPV lower head integrity is maintained if core debris relocates to lower plenum

Maximum heat flux remains below critical heat flux (CHF) on exterior surface

- Heat generation rate based on conservative assumptions/inputs (e.g., 100°/o core U02 - no upward radiation heat losses)

- Assumed CHF threshold conservatively does not credit high absolute pressure and large subcooling in CNV

With effective external vessel cooling, the lower head remains intact and the severe accident progression is stabilized in RPV 49 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Severe Accident Phenomena CNV integrity not challenged by severe accident phenomena 50 PM-0519-65372 Revision: O

Hydrogen combustion not challenging due to limited oxygen concentration

In-vessel fuel-coolant interactions (FCI) (i.e., steam explosion) are not sufficiently energetic to induce alpha mode failure due to factors including:

- Small core size, low debris temperatures, small drop height, shallow pool, relatively high system pressure

Containment overpressure does not occur

- High pressure steel CNV designed for most limiting LOCA blowdown which exceeds maximum severe accident pressures

- Submergence of CNV in UHS provides highly effective pressure suppression

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NuScale Nonproprietary Consideration of Uncertainty

If IVR in RPV fails

- High pressure melt ejection (leading to direct containment heating) does not occur because there is no driving pressure differential

- Energetic ex-vessel FCI not likely for similar reasons as in-vessel FCI

- Debris relocated to CNV would be retained by CNV lower head

Effective external cooling of CNV by reactor pool

If lower CNV fails

- Pool scrubbing minimizes release

If upper CNV fails

- Instantaneous release of entire airborne radionuclide inventory in module at time of postulated CNV failure would not constitute a large release 51 PM-0519-65372 Revision: a Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Level 2 Insights

Core damage events are stabilized within the RPV

Severe accident phenomena do not challenge CNV integrity

Large release does not occur even if RPV and CNV are postulated to fail

The large release frequency is dominated by containment bypass events 52 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Level 2 Key Insights (1 of 5)

Design Feature/Insight

Containment Isolation The primary purpose of CNTS is to retain primary coolant inventory within the CNV. With primary coolant inventory maintained in the RPV or CNV, cooling of core debris is ensured.

CNTS terminates releases through

'. penetrations leading outside

containment.

53 PM-0519-65372 Revision: O Comment If coolant remains primarily within the RPV, then the core is covered. If the core is not covered in the RPV then sufficient primary coolant is in the CNV to submerge the outside of the lower RPV and establish conductive heat removal from the core debris to the coolant in the CNV through th~ _RflY '-'YaJJ.

. Containment penetrations through which releases are

. assumed to occur that dominate risk include those that

! bypass containment such as CVCS (injection and

discharge) and paths through the steam generator tubes (main steam and feedwater piping). Isolation of normally

'open valves in these penetrations prevents releases from

bypassing containment.

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NuScale Nonproprietary Level 2 Key Insights (2 of 5)

Design Feature/Insight

Passive Heat Removal The RPV has no insulating material and passive heat removal capability from the RPV to the CNV is sufficient to prevent core debris from j)_~rietrati11~ _thE:}_ r_~~ct_or_~~~s~I_.

The CNV is uninsulated and passive

heat removal capability from the

, CNV to the UHS is sufficient to

prevent the containment from

pressurizing and or core debris from

. penetrating the containment 54 PM-0519-65372 Revision: O Comment C

I I

- - __ J Retaining primary coolant in the containment results in collection of sufficient RCS water in the CNV to allow heat transfer through RPV to CNV and ultimately UHS to remove heat generated in the fuel regardless of its location.

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NuScale Nonproprietary Level 2 Key Insights (3 of 5)

Design Feature/Insight Comment Severe Accident Containment Challenges (1 of 2)

Primary coolant system overpressure failure cannot lead to overpressurization of containment (i.e., loss of decay heat removal through the steam generators plus failure of the RSVs to open).

Hydrogen combustion is not likely as the containment is normally evacuated.

In-vessel steam explosions are not likely due to core support design and volume of lower vessel head.

HPME cannot occur 55 PM-0519-65372 Revision: O Addition of water to the containment from external sources (CFDS) results in submergence of the reactor vessel and establishes passive heat removal through the containment wall to the reactor pool. Even if containment flooding is not successful, the RPV failure mode is such that containment ultimate capacity would not be exceeded.

There is very little oxygen available (oxygen generated from radiolysis is only a long-term issue) and containment is steam inerted under severe accident conditions. In addition, conservative AICC analyses predict containment pressures that do not exceed the design pressure.

Core support failure is expected before the fuel has a chance to become molten. With the core uncovered there is little water in the bottom of the RPV with which core debris can interact.

Submergence of the lower RPV With passive heat removal establishes passive heat from the reactor to removal and prevents core containment established, the debris from exiting the RPV. No reactor is depressurized ex-vessel challenges occur if the even if core debris is core remains within the vessel.

postulated to exit the vessel.

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NuScale Nonproprietary Level 2 Key Insights (4 of 5)

Design Feature/Insight Comment Severe Accident Containment Challenges (2 of 2)

Ex-vessel steam explosion does not occur with a submerged RPV.

Overpressure of containment due to non-condensable gas generation is not applicable to the NuScale design.

Basemat penetration is not applicable to the NuScale design.

56 PM-051 9-65372 Revision: O Submergence of the lower RPV establishes passive heat removal and prevents core debris from exiting the RPV. No ex-vessel challenges occur if the core remains within the vessel.

There is no concrete in the containment with which the core debris could interact and generate non-condensable gases.

There is no basemat making up the containment boundary. This issue is addressed as a part of considering protection against contact of core debris with the containment wall.

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NuScale Nonproprietary Level 2 Key Insights (5 of 5)

Design Feature/Insight Comment Support Systems, Human Action, External Events Support systems are not needed for safety-related system functions (i.e.,

containment isolation) important to the Level 2 PRA.

With one exception, there are no risk significant, post-accident human actions associated with the full-power internal events Level 2 PRA.

The exception is alignment of CFDS during accident sequences in which isolation of a broken eves line outside containment fails, ECCS is successful but coolant inventory in containment needs replenishment in order to maintain natural circulation between CNV and the RPV.

Risk significant SSC for external events are largely the same as those found risk significant for internal events Safety-related mitigating systems are fail-safe on loss of power and do not require supporting systems such as lube oil, instrument air, or HVAC to function.

Operator actions, including backup and recovery actions, are not significant to the Level 2 analysis because of passive system reliability and fail-safe system design. The operator action to align CFDS during a eves break outside containment meets the risk significance thresholds because of a mathematical limitation of the calculation of the Fussell-Vesely measure of importance The module response to external events is comparable to the response to internal event due to the passive features of the design which are not affected by the external events and plant systems that are protected against external event challenges.

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NuScale Nonproprietary Section 19.2 COL Items Item Number Description

COL Item 19.2-1 : A COL applica~t--th~t r~f~r~~~~~-the NuScale Power Plant design I

. certification will develop severe accident management guidelines and

, other administrative controls to define the response to beyond-

. design-basis events.

COL Item 19.2-2 A COL applicant that references the NuScale Power Plant design certification will use the site-specific probabilistic risk assessment to evaluate and identify improvements in the reliability of core and containment heat removal systems as specified by 10 CFR 50.34(f)(1 )(i).

COL Item 19.2-3 58 PM-0519-65372 Revision: O

A COL applicant that references the NuScale Power Plant design

certification will evaluate severe accident mitigation design

. alternatives screened as "not required for design certification application."

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NuScale Nonproprietary ACRS Subcommittee Presentation:

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NuScale Nonproprietary Presentation Team Sarah Bristol Supervisor, Probabilistic Risk Assessment Etienne Mullin Probabilistic Risk Analyst

-Bill Galyean Probabilistic Risk Assessment Consultant Rebecca Norris.

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NuScale Nonproprietary Section 19.3

There are no RTNSS SSCs in the NuScale design

- None of the five RTNSS criteria were met by any NuScale SSC

RTNSS is also discussed in FSAR 17.4.3.3 61 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Section 19.3 COL Item Item Number Description

. COL Item 19.3-1 A COL applicant that references the NuScale Power Plant design certification will identify site-specific regulatory treatment of nonsafety systems (RTNSS) structures, systems, :

and components and applicable RTNSS process controls.

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NuScale Nonproprietary ACRS Subcommittee Presentation:

PM-0519-65372 Revision: O NuScale FSAR Chapter 19.5 Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts

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NuScale Nonproprietary Presentation Team Amber Berger Civil/Structural Engineer Rebecca Norris Supervisor, Licensing Marty Bryan Licensing Project Manager 64 PM-0519-65372 Revision: o Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Introduction and Background

Plant design for potential effects of beyond design basis large commercial aircraft impact [10 CFR 50.150(a)]

- The reactor core remains cooled, or the containment remains intact

- Spent fuel cooling or spent fuel pool integrity is maintained

Design-specific impact assessment per RG 1.217, which endorses NEI 07-13

NEI 07-13 methods followed with no exceptions

Aircraft impact informed the plant design 65 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Assessment Scope

Reactor Building assessed for effects in three areas for postulated aircraft impact

- Physical damage

- Shock damage from shock-induced vibration on structures, systems, and components

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NuScale Nonproprietary Assessment Methodology

NEI0?-13

Reactor Building is structure of concern

- NuScale Power Modules

- Ultimate heat sink

- Spent fuel pool

Impact locations

- Screening by NEI 07-13

- Radioactive Waste Building (RWB) is "intervening structure" to mitigate physical damage to RXB, conservatively do not credit RWB in shock assessment

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NuScale Nonproprietary NuScale Site Plan NuScale DCA Tier 2 Figure 1.2-1 Conceptual Site Layout 68 PM-0519-65372 Revision: 0 Copyright 201 9 by NuScale Power, LLC.

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NuScale Nonproprietary Assessment Results

Physical damage

- Local assessment per NEI formulas for perforation and scabbing

- Global response performed using detailed finite element models and NRC specified force-time history

- RXB external walls prevent physical damage from entering RXB

- No internal missiles for secondary impact

- No impact on containment boundary

- Spent fuel pool protected inside RXB below grade

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NuScale Nonproprietary Assessment Results (cont'd)

Shock damage

-Aircraft impact causes short duration, high acceleration, high frequency vibration

- Core cooling

At-power and shutdown scenarios considered

No active equipment required for success

Adequate heat removal is shown for all strikes

-Spent fuel 70 PM-0519-65372 Revision: O

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NuScale Nonproprietary Assessment Results (cont'd)

Fire damage

- Design and location of 3-hr fire barriers and 3-hr, 5-psid fire barriers prevent propagation of fire into RXB

- Design and location of 5-psid, fast-acting blast dampers at RXB HVAC key design feature

- Concrete shrouds protect exterior wall pipe and HVAC penetrations from physical damage and prevent fire propagation into the RXB

- Fire that enters through external personnel doors at grade level does not propagate beyond stairwells

All required operator actions occur prior to impact 71 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Assessment Conclusions

Design and functional capabilities provide adequate protection of public health and safety

NuScale plant meets 10 CFR 50.150 regulation

- Maintain containment integrity AND core cooling capability (only required to meet one)

- Maintain SFP integrity

For most postulated aircraft impact strikes, spent fuel pool cooling maint_ained, meeting all four CFR requirements 72 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Acronyms (1 of 3)

ATWS anticipated transient without scram* CRB Control Building

BOB beyond design basis

eves chemical and volume control

CCF common cause failure

CD core damage

CDF core damage frequency

CES containment evacuation system

CFDS containment flooding and drain system

CFR Code of Federal Regulations

CHF critical heat flux.

CIV containment isolation valve

CNV containment vessel

CNTS containment system

COL combined license system

CSDRS certified seismic design response spectra

CTG combustion turbine generator

DIC demand/capacity

DGN diesel generator

DHRS decay heat removal system

ECCS emergency core cooling system

EHVS 13.8 kV and switchyard system

ELVS low voltage AC electrical distribution system

ESF earthquake scale factor 73 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Acronyms (2 of 3)

FCI fuel-coolant interaction

LPSD low power and shutdown

FMEA failure modes and effects analysis

LR large release

FP-IE full power, internal event

LRF large release frequency

FSAR Final Safety Analysis Report

mcyr module critical year

HCLPF high confidence of low probability

MOP motor driven pump of failure

MMAF multi-module adjustment factor

HEP human error probability

MPS module protection system

HPME high pressure melt ejection

NEI Nuclear Energy Institute

HVAC heating ventilation and air conditioning

NPM NuScale Power Module

IE initiating event

NR no release

IVR in-vessel retention

NRC Nuclear Regulatory Commission

LOCA loss of coolant accident

PGA peak ground acceleration

LOOP loss of offsite power

PRA probabilistic risk assessment 74 PM-0519-65372 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Acronyms (3 of 3)

RBC reactor building crane

RCS reactor coolant system

RG Regulatory Guide

RPV reactor pressure vessel

RSV reactor safety valve

RTNSS regulatory treatment of nonsafety systems

RVV reactor vent valve

RWB Radioactive Waste Building

RXB Reactor Building

SAPHIRE Systems Analysis Programs for Hands-on Integrated Reliability Evaluations

SGTF steam generator tube failure

SMA seismic margin assessment

SOARCA State-of-the-Art Reactor Consequence Analysis

SSC structures, systems, and components

SFP spent fuel pool

SRP Standard Review Plan

SAMOA severe accident mitigation design* TGB Turbine Generator Building alternative 75 PM-0519-65372 Revision: O

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Portland Office 6650 SW Redwood Lane, Suite 210 Portland, OR 97224 971.371.1592 Corvallis Office 1100 NE Circle Blvd., Suite 200 Corvallis, OR 97330 541. 360. 0500 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MO 20852 301. 770.0472 Charlotte Office 2815 Coliseum Centre Drive,.

Suite 230 Charlotte, NC 28217 980. 349. 4804 NuScale Nonproprietary Richland Office 1933 Jadwin Ave., Suite 130 Richland, WA 99354 541. 360. 0500 Arlington Office 2300 Clarendon Blvd., Suite 1110 Arlington, VA 22201 London Office 1st Floor Portland House Bressenden Place London SW1 E 5BH United Kingdom

+44 (0) 2079 321700 http://www.nuscalepower.com W Twitter: @NuScale_Power

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