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ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0
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Issue date:	03/04/2020
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"ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident,"

PM-0320-69141, 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 Full Committee Presentation NuScale Topical Report Non-Loss-of-Coolant Accident

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1 March 5, 2020 PM-0320-69141

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Presenters Ben Bristol Supervisor, System Thermal Hydraulics Meghan McCloskey Thermal Hydraulic Analyst Matthew Presson Licensing Project Manager Paul lnfanger Licensing Specialist 2

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Outline

Scope of non-LOCA LTR

Non-LOCA events

- Events and acceptance criteria

- Interface to other methodologies

- Factors controlling margin to acceptance criteria

Development of non-LOCA EM

- PIRT and gap analysis

- Focus of NRELAPS validation for non-LOCA

General event analysis methodology

Specific event analysis 3

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Scope of Non-LOCA Topical Report In Scope Out of Scope

NRELAP5 system

SAFDLs evaluated in downstream subchannel transient analysis of non- analysis LOCAevents

Accident radiological dose

Interface to subchannel analysis and accident radiological

Control rod ejection analysis

LOCA and valve opening

Short-term transient events progression with DHRS

Peak containment cooling pressure/tern perature analysis

Long term transient progression with DHRS Riser uncovery Return to power 4

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Non-LOCAEM EM applicable to NuScale Power Module plant design Applicable initiating events:

Cooldown events

Reactivity events

- Decrease in FW temperature Uncontrolled bank withdrawal from subcritical

- Increase in FW flow

- Uncontrolled bank withdrawal at power

- Increase in steam flow Inadvertent opening of SG relief or safety valve - Control rod misoperation

- Steam piping failures (postulated accident)

Single rod withdrawal

- Loss of containment vacuum Containment flooding

Control rod drop

- Inadvertent decrease in RCS boron concentration

Heatup events

- Loss of external load Turbine trip

Inventory increase event

- Loss of condenser vacuum - eves malfunction

- Closure of MS~

- Loss of non-emergency AC power

Inventory decrease events

- Loss of normal FW flow Feedwater system pipe breaks (postulated accident) - Small line break outside containment (infrequent event)

- Inadvertent operation of DHRS

- Steam generator tube failure (postulated accident)

NuScale unique event 5

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Non-LOCA Event Acceptance Criteria AOO Infrequent Event Accident Description Analysis Acceptance Criteria Acceptance Criteria Acee ptance Criteria Reactor Coolant Non-LOCA System Pressure S 110% of Design S 120% of Design S 120% of Design NRELAP5 (Pdesign= 2100 psia)

Steam Generator Non-LOCA Pressure S 110% of Design S 120% of Design S 120% of Design NRELAP5 (Pdesign= 2100 psia)

Minimum If limit exceed, If limit exceed,

> Limit Subchannel Critical Heat Flux Ratio fuel assumed failed <1l fuel assumed failed <1l Maximum Fuel If limit exceed, If limit exceed,

< Limit Subchannel Centerline Temperature fuel assumed failed <1l fuel assumed failed <1l

< Limits < Limits < Limits Containment Containment Integrity (pressure, temperature) (pressure, temperature) (pressure, temperature) PIT analysis Escalation of an AOO to an accident (AOO)

If other or No No No acceptance Consequential loss of criteria are met system functionality (IE or accident)?

Normal or Normal Radiological Dose < Limit < Limit Accident Operations radiological (1) NuScale safety analysis methodologies developed to demonstrate fuel cladding integrity maintained.

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Evaluation Models - General Non-LOCA Approach r---------------. r--------------~

Plant design, :, I I II I

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I M &E releases from T/H Core design, NRELA PS I VIPRE-01 Fuel rod design, I ....

, 1 sys tem T/H I 1 I

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., subchannel response, other Plant initial conditions, respo nse analysis input I

SSC performance

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RCS pre ssure, Accident secon dary Fuel cladding radiological press ure, integrity analysis Safe sta bilized --

cond1flon Radiological Non-LOCA topical report Subchannel topical report TR-0516-49416-P TR-0915-17564-P-A dose

---------------~ acceptance criteria Accident source term topical report TR-0915-17565-P 7

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Non-LOCA Events -

Margin to Acceptance Criteria Design characteristics governing non-LOCA event transient response and margin to acceptance criteria

- MCHFR: Limited by combination of high power, high pressure, high temperature conditions occurring around time of reactor trip,

_for reactivity insertion events

- Primary pressure: Protected by RSV lift

- Secondary side pressure: Limited by primary side temperature conditions

- Radiological release: MPS designed to rapidly detect and isolate based on measured conditions

- Establishing a safe, stable condition: MPS designed to trip, actuate DHRS to protect adequate inventory in at least 1 steam generator 8

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Non-LOCA EM Development

Non-LOCA evaluation model developed to perform conservative analyses, following intent of the RG 1.203 EMDAP and applying a graded approach

Element 1 - Establish applicable transients and acceptance criteria, develop non-LOCA PIRT

Element 2, 3, 4

- Leverage NRELAPS development, NRELAPS assessments performed during LOCA evaluation model development.

Gap analysis performed to evaluate how high ranked phenomena are addressed

Focused on differences in high ranked Pl RT phenomena between LOCAand non-LOCA

Additional NRELAP5 code validation performed focused on DHRS and integral non-LOCA response

- Suitably conservative initial and boundary conditions applied for non-LOCA analyses

- Sensitivity calculations used to demonstrate factors controlling margin to acceptance criteria 9

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Non-LOCA PIRT Development Event Types SSCs Considered in PIRT Increased heat removal Reactor coolant system Main feedwater system Decreased heat removal Containment vessel Main steam system Reactivity anomaly Decay heat removal system Chemical volume control system Increase in RCS inventory Reactor pool Containment evacuation Steam generator tube failure system Phase Identification RCS Response DHRS Operation

PIRT Figures of merit 1 pre-trip transient higher flow levels at full inactive CHFR power levels RCS pressure 2 post-trip transitional flow levels at startup CHFR transition transitioned power levels RCS, secondary, containment pressures 3 stable natural lower flow levels at decay fully effective CHFR circulation power levels RCS mixture level Subcriticality

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If DHRS actuated by protection system

Different non-LOCA events involve different plant systems and responses

PIRT developed considering all non-LOCA event types and important SSCs

Short-term response divided into 3 generic phases with associated FoM P~-0320-69141 Revision: o Copyright 2020 by NuScale Power, LLC.

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NRELAP5 Applicability for Non-LOCA After non-LOCA PIRT developed, Key areas identified from gap analysis for short-term non-LOCA analysis:

gap analysis performed to

DHRS modeling and heat transfer determine how to address high-

- NRELAP5 validation against KAIST tests; ranked phenomena: NIST-1 SETs HP-03, HP-04

- NPM sensitivity calculations

Validation performed as part of NRELAP5 assessment for LOCA

Steam generator modeling and heat transfer evaluation model

- NRELAP5 validation against

Additional validation or benchmark S1ET-TF1, S1ET-TF2 tests for non-LOCA - NPM sensitivity calculations

Reactivity event response

Conservative input

- NRELAP5 benchmark against RETRAN-30

Subchannel analysis

NPM non-LOCA integral response

- NRELAP5 validation against NIST-1 IETs NLT-2a, NLT-2b, NLT-15p2 Overall conclusion: NRELAP5 code, with NPM system model, is applicable for calculation of the NPM non-LOCA system response 11 PM-0320-69141 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Non-LOCA Analysis Process Topical report Section 4

1. Develop plant base model 5. Identify cases for subchannel NRELAP5 input (geometry, control analysis and extract boundary and protection systems, etc) conditions (if applicable)

2. Adapt NRELAP5 base model as - Conservative bias directions:

necessary for specific event

Maximum reactor power analysis and desired initial

Maximum core exit pressure conditions

Maximum core inlet temperature

3. Perform steady state and transient

Minimum RCS flow rate analysis calcufations with - NRELAPS CHF calculations for NRELAP5 dummy hot rod may be used as a screening tool to assist analysts in

4. Evaluate results of transient determining limiting cases to be analysis calculations: evaluated in downstream subchannel

- Confirm margin to maximum RCS analysis pressure acceptance criterion

6. Identify cases for radiological

- Confirm margin to maximum SG pressure analysis (if applicable) acceptance criterion

- Maximum mass release case

- Confirm appropriate transient run tin:,~

execution to demonstrate safe, stab1l1zed - Maixmum iodine spiking case condition achieved 12 PM-0320-69141

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Non-LOCA Methodology General Methodology Event-specific Methodology (Section 7 .1 }: (Section 7 .2)

Steady-state conditions

Description of event initiation and progression Treatment of plant controls

Acceptance criteria 'of interest' Loss of power

Limiting single failure, loss of Single failure power scenarios, or need for Bounding reactivity sensitivity calculations parameter input

Initial condition biases and Biasing of other parameters: conservatisms, or need for initial conditions, valve sensitivity calculations characteristics, analytical

Tabulated representative results limits and response times of sensitivity calculations Operator action Example analysis results provided in Section 8 13 PM-0320-69141

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Conclusions

Non-LOCA system transient evaluation model developed following a graded approach in accordance with guidance provided in RG 1.203

Applies to NPM-type plant design natural circulation water reactor with helical coil SG and integral pressunzer

NRELAP5 used to simulate the system thermal-hydraulic response

Demonstrate primary and secondary pressure acceptance criteria are met

Demonstrate safe, stabilized condition achieved

System transient results provide boundary conditions to downstream subchannel and radiological analyses 14 PM-0320-69141 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Acronyms

AOO - Anticipated Ope rational

MPS - Module Protection System Occurrences

MSIV - Main Steam Isolation Valve

CNV - Containment Vessel

MSS-Main Steam System

CVCS-Chemical andVolumeControl System

NIST NuScale lntegralSystemTest-1

DHRS - Decay Heat Removal System

NPM - NuScale Power Module

ECCS-EmergencyCore Cooling System

PIRT- Phenomena Identification and Ranking Table

EM - Evaluation Model

RCS - Reactor Coolant System

EMDAP- Evaluation Mode I Development and Assessment Process

RPV - Reactor Pressure Vessel

RSV - Reactor Safety Valve

FW - Feedwater

FWIV - Feedwater Isolation Valve

RVV -ReactorVentValve

IET - Integral Effects Test

SET-Separate Effects Test

KAIST - Korea Advanced Institute of

SG - Steam Generator Science and Technology

SSC - Structures, Systems, and Components

LOCA - Loss of Coolant Accident

MCHFR- Minimum Critical Heat Flux Ratio 15~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~...._-~~~

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Portland Office Richland Office 6650 SW RedWJod Lane, 1933 Jadwn Ave., Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541. 360. 0500 971.371.1592 Charlotte Office Corvallis Office 2815 Coliseum Centre Drive, 1100 NE Circle Blvd., Suite 200 Suite 230 Corvallis, OR 97330 Charlotte, NC 28217 541.360. 0500 980.349.4804 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301.770.0472 http://vwvw.nuscalepower.com W Twtter: @NuScale_Pooor NUSCALE' Power for all humankind 16 PM-0320-69141 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Power Module Overview Integral Pressurized Water Reactor

Core, steam generator and pressurizer in one vessel feedwater line

Integrated reactor design, no large-break loss-of-coolant accidents

Reactor coolant system operated in single phase (liquid) density driven flow

Safety decay heat removal systems are passive and fail safe steam generator

Module protection system designed to automate event mitigation actuations (no operator actions) 18 PM-0320-69141

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ECCS Emergency Core Cooling System

ECCS valves open to a boiling/condensing circulation flow path to transfer decay and residual heat to reactor pool

- Liquid from containment vessel enters RCS through reactor recirculation re actor vent valve reactor vent valves valves

- Vapor vented from RCS to containment vessel through reactor vent valves

- Steam condenses on inside surface of containment vessel

- Heat transfer through vessel walls to the reactor pool reactor recirculation reactor recircu lation

Actuation Signals: High CNV valve valve level, 24hr loss of AC power

Fail safe: ECCS valves open on loss of DC power NOT10SCALE 19 PM-0320-691 41

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Decay Heat Removal System (DHRS)

Rem aves heat after FWIVs loss of normal cooling

Boiling/condensing loop

Two redundant trains reactor pool

Redundant actuation and isolation valves for each train C*fl pass ive condenser

Initiates on:

- Loss of power

- Loss of cooling indication (ESFAS Signal)

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Deterministic Event Mitigation Module Protection Functions Event Mitigation Reactivity Control Increase in heat removal transients

Reactor trip

Reactor trip Secondary Isolation

CVCS/Demineralized Water Decrease in heat removal transients Isolation

Reactor trip DHRS Actuation RCS and Secondary Inventory Reactivity and povver transients Control

Reactor trip Demineralized Water Isolation

Containment Isolation Increase in RCS inventory transients

Secondary Isolation

Reactor trip eves Isolation Heat Removal Decrease in RCS inventory transients

DHRS Actuation

Reactor trip CNV Isolation

ECCS Actuation E CCS actuation Subcooling Stability

Reactor trip

Reactor trip 21 PM-0320-69141

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ML20069A154

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