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

L0-0519- 65373 :

"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

NuScale Nonproprietary ACRS Subcommittee Presentation:

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NuScale Nonproprietary Chapter 19 Section Title Comment

. 19.0 Probabilistic Risk Assessment and Severe

Overview Accident Evaluation 19.1 Probabilistic Risk Assessment Level 1,2

Thermal hydraulic & .

19.2 Severe Accident Evaluation phenomenological *

analyses 19.3 Regulatory Treatment of Nonsafety Systems No RTNSS SSCs Strategies and Guidance to Address Loss of Add d.

. resse 1n 19.4 L~rge Areas of the Plant due to Explosions and Cha ter 20

. Fires

p Adequacy of Design Features and Functional 19.5 Capabilities Identified and Described for Overview Withstanding Aircraft Impacts 2

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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 1.174

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

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

>> The terms CDF and LRF are tied to current large reactors and use of these terms in the NuScale design may be misleading 9

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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

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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

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 15 PM-0519-65372 Revision: O 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 Human error probability for first HFE in cutset ;:*4.0E-03 '.! 10;

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HEP02 Human error probability for second HFE in cutset 1.SE-01 3 i

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

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

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

Name Description Context Operator fails to unisolate and Used for LOCA-OC (2 IEs), SGTFs, CFDS--HFE-0001 C-FOP-N initiate CFDS injection and transients (1 IE)

Used for LOCA-IC (3 IEs), LOCA-OC (letdown) (1 IE), transients (1 IE)

Operator fails to unisolate and CVCS--H FE-0001 C-FOP-N and secondary steam line break (1 initiate eves injection IE) upon failure of ECCS, and SGTFs Operator fails to locally Local unisolation due to lack of CVCS--H FE-0002C-FOP-N unisolate and initiate eves control from a partial loss of DC injection power Operator fails to open ECCS Backup action to MPS autofunction ECCS--H FE-0001 C-FTO-N valves failure Backup local action to control room Operator fails to start/load EHVS--H FE-0001 C-FTS-N initiation failure during loss of offsite combustion turbine generator power Backup local action to control room Operator fails to start/load ELVS--HFE-0001 C-FTS-N initiation failure during loss of offsite backup diesel generator ower 19 PM-0519-653 72 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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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 1E-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 Parameter Change CDF Result LRF Result I

, Base Case 2.7E-10 1.7E-11 I

HEP All HEPs set to FALSE 2.0E-10 1.0E-11

------- -- ~------- --- - - ------- - ---------------*------ -- - ----- - -- ------ --------------------- - -- ----------- ------

/ HEP All HEPs set to TRUE 3.2E-8 2.BE-9 1 CCF All CCFs set to FALSE 5.4E-12 1.2E-12 iCCF All CCFs set to max value of 0.002 4.2E-6 3.7E-8 LOOP-IE LOOP frequency set to 1 per year 2.2E-9 1.?E-11 (base = 3.1 E-2 per year) r--------- ------------ -*----**- ----------- ---- ------------- *----- - -- - - - --- --------- -- --- ---------------------- --- --- ---- ----

! LOCA-IC-IE LOCA inside CNV frequency increased 1 3.4E-10 1.?E-11 order of magnitude 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

DHRS PSSR failure increased 1 order of magnitude l&C sensors Failure probability of sensors was increased 2.8E-10 1.?E-11 an order of magnitude 25 PM-0519-65372

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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 Core Damage Cutset Mapped to Containment Isolation - CIVs # End State Comments Release Size Close (Phase - PH1) (Phase - PH1)

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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 SMA results

Induced initiator event trees

Plant level HCLPF: 0.88g

- Structural failures

Structural failures dominate

- LOCAs - Crane

- Loss of offsite power - Exterior walls

Seismic failure mapping - Bay walls

- No pre-screening (all PRA cutsets - Module supports included in the SMA)

At lower PGAs, LOOP combined

- Evaluated at 14 ground motion with random failures dominate levels ranging from 0.05g to 3.5g 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 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 Comment Failure to scram events (ATWS) do Core characteristics result in ATWS power levels that are not lead directly to core damage. 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.

Passive heat removal capability is RSV cycling transfers adequate RCS water to the CNV to allow sufficient to prevent core damage if heat transfer through the RPV to the CNV and ultimately to a reactor safety valve (RSV) cycles. reactor pool to remove decay heat.

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

Passive, fail-safe emergency core The ECCS consists of 5 valves that fail-safe on a loss of power cooling system (ECCS) functions to and provides a natural circulation path through the core and preserve RCS inventory, which is CNV, thus providing heat transfer to the reactor pool. The sufficient to allow core cooling closed-loop system does not need additional inventory.

without RCS makeup from external source.

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NuScale Nonproprietary Level 1 Key Insights (2 of 2)

Design Feature/Insight Comment Containment isolation preserves Containment isolation eliminates the potential for breaks outside RCS inventory for core cooling of containment to result in loss of RCS inventory. For breaks without external makeup. inside of containment, containment isolation is not necessary to support passive core cooling and heat removal.

Passive, fail-safe safety systems Safety-related mitigating systems are fail-safe on loss of power (ECCS, DHRS , RSVs) include and do not require supporting systems such as lube oil, air or redundancy and do not need support HVAC to function. No single failure results in a loss of system systems, including electric power or function.

operator actions.

There are no risk significant, post- No operator actions, including backup and recovery actions, are initiator human actions associated risk significant to the CDF because of passive system reliability with the full-power PRA. and fail-safe system design.

Risk significant structures, systems The module response to external events is comparable to the and components (SSCs) for external response to internal event due to the passive features of the events are largely the same as those design and independence from support systems such as power.

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

Active systems providing makeup Inventory addition is possible by the active systems chemical inventory to the RPV are not risk and volume control system (CVCS) and containment flooding significant. 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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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

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 U0 2 relocates at first observed relocation time from database

Assume debris is molten, pure U0 2 composed of no filler materials (e.g., steel, zirconium)

No credit for water in lower plenum at time of relocation 47 PM-0519-65372 Revision: O 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

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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 U0 2 - 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

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 Comment

Containment Isolation The primary purpose of CNTS is to If coolant remains primarily within the RPV, then the core retain primary coolant inventory is covered. If the core is not covered in the RPV then within the CNV. With primary sufficient primary coolant is in the CNV to submerge the coolant inventory maintained in the outside of the lower RPV and establish conductive heat RPV or CNV, cooling of core debris removal from the core debris to the coolant in the CNV is ensured.

- - - - *-- ... - through th~ _RflY '-'YaJJ. --* .. -- . *--- ____ ----------*-*-

CNTS terminates releases through . Containment penetrations through which releases are

'. penetrations leading outside *. assumed to occur that dominate risk include those that

containment. ! 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 Comment

C - I

Passive Heat Removal I

- --* -- -- - - __ J The RPV has no insulating material Retaining primary coolant in the containment results in collection and passive heat removal capability of sufficient RCS water in the CNV to allow heat transfer through from the RPV to the CNV is RPV to CNV and ultimately UHS to remove heat generated in sufficient to prevent core debris from the fuel regardless of its location.

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

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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 Addition of water to the containment from external overpressure failure cannot lead to sources (CFDS) results in submergence of the reactor vessel overpressurization of containment and establishes passive heat removal through the containment (i.e., loss of decay heat removal wall to the reactor pool. Even if containment flooding is not through the steam generators plus successful, the RPV failure mode is such that containment failure of the RSVs to open). ultimate capacity would not be exceeded.

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

In-vessel steam explosions are not Core support failure is expected before the fuel has a chance to likely due to core support design and become molten. With the core uncovered there is little water in volume of lower vessel head. the bottom of the RPV with which core debris can interact.

HPME cannot occur 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 Submergence of the lower RPV establishes passive heat occur with a submerged RPV. removal and prevents core debris from exiting the RPV. No ex-vessel challenges occur if the core remains within the vessel.

Overpressure of containment due to There is no concrete in the containment with which the core non-condensable gas generation is debris could interact and generate non-condensable gases.

not applicable to the NuScale design.

Basemat penetration is not There is no basemat making up the containment boundary. This applicable to the NuScale design. 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 mitigating systems are fail-safe on loss of power safety-related system functions (i.e ., and do not require supporting systems such as lube oil ,

containment isolation) important to instrument air, or HVAC to function.

the Level 2 PRA.

With one exception , there are no risk Operator actions, including backup and recovery actions, are significant, post-accident human not significant to the Level 2 analysis because of passive actions associated with the full- system reliability and fail-safe system design. The operator power internal events Level 2 PRA. action to align CFDS during a eves break outside containment The exception is alignment of CFDS meets the risk significance thresholds because of a during accident sequences in which mathematical limitation of the calculation of the Fussell-Vesely isolation of a broken eves line measure of importance 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 The module response to external events is comparable to the events are largely the same as those response to internal event due to the passive features of the found risk significant for internal design which are not affected by the external events and plant events 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

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:

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-05 19-65372 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

SFP integrity maintained for all strikes 70 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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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 system

CCF common cause failure

CSDRS certified seismic design response

CD core damage spectra

CDF core damage frequency

CTG combustion turbine generator

CES containment evacuation system

DIC demand/capacity

CFDS containment flooding and drain

DGN diesel generator system

DHRS decay heat removal system

CFR Code of Federal Regulations

ECCS emergency core cooling system

CHF critical heat flux.

EHVS 13.8 kV and switchyard system

CIV containment isolation valve

ELVS low voltage AC electrical

CNV containment vessel distribution system

CNTS containment system

ESF earthquake scale factor

COL combined license 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

NPM NuScale Power Module conditioning

NR no release

IE initiating event

NRC Nuclear Regulatory Commission

IVR in-vessel retention

PGA peak ground acceleration

LOCA loss of coolant accident

PRA probabilistic risk assessment

LOOP loss of offsite power 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

SAPHIRE Systems Analysis Programs for Hands-on Integrated Reliability

RCS reactor coolant system Evaluations

RG Regulatory Guide

SGTF steam generator tube failure

RPV reactor pressure vessel

SMA seismic margin assessment

RSV reactor safety valve

SOARCA State-of-the-Art Reactor

RTNSS regulatory treatment of nonsafety Consequence Analysis systems

SSC structures, systems, and

RVV reactor vent valve components

RWB Radioactive Waste Building

SFP spent fuel pool

RXB Reactor Building

SRP Standard Review Plan

SAMOA severe accident mitigation design* TGB Turbine Generator Building alternative

UHS ultimate heat sink 75 PM-0519-65372 Revision: O Copyright 2019 by NuScale Power, LLC.

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NuScale Nonproprietary Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541. 360. 0500 971.371.1592 Arlington Office Corvallis Office 2300 Clarendon Blvd., Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541. 360. 0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave., Suite 205 London SW1 E 5BH Rockville, MO 20852 United Kingdom 301. 770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive,.

Suite 230 Charlotte, NC 28217 980. 349. 4804 http://www.nuscalepower.com W Twitter: @NuScale_Power NUSCALE' Power for all humankind

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