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ACRS Subcommittee Presentation: NuScale Topical Report - Non-Loss-of-Coolant Accident, PM-0220-68852, Revision 0
	ML20054A962
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Site:	NuScale
Issue date:	02/19/2020
From:	; NuScale
To:	; Office of Nuclear Reactor Regulation
References
	LO-0220-68854 PM-0220-68852, Rev 0
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NuScale Nonproprietary ACRS Subcommittee Presentation NuScale Topical Report Non-Loss-of-Coolant Accident February 19, 2020 PM-0220-68852

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

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Outline

Scop_e 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 niisoperation

- 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 MSIV

- 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 Acee ptance Criteria Acee ptance Criteria Acee ptance Criteria Reactor Coolant Non-LOCA System Pressure S 110% of Design S 120% of Design S 120% of Design NRELAPS (Pdesign= 2100 psia)

Steam Generator Non-LOCA Pressure S 110% of Design S 120%-of Design S 120% of Design

NRELAPS (Pdesign= 2100 psia)

Minimum If limit exceed, If limit exceed,

> Limit Subchannel Critical Heat Flux Ratio fuel assumed failed <1> fuel assumed failed <1>

Maximum Fuel If limit exceed, If limit exceed,

< Limit Subchannel Centerline Temperature fuel assumed failed <1> fuel assumed failed (1>

< Limits < Limits < Limits Containment Containment Integrity (pressure, temperature) (pressure, temperature) (pressure, temperature) P/T 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---------------, ~--------------~

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Plant design, I

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Core design, I

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I I VIPRE-01 from T/H I - I I_ response, other Fuel rod design, I ,- system T/H T .,. subchannel Plant initial conditions, response analysis input SSC performance

i I RCS pressure, Accident secondary Fuel cladding radiological pressure, integrity analysis Safe stabilized condition ,,

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 PIRT 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 1\/lain feedwater system Decreased heat removal Containment vessel 1\/lain 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

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 e

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NRELAPS 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-

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

- NPM sensitivity calculations

Validation performed as part of NRELAPS assessment for LOCA

Steam generator modeling and heat transfer evaluation model

- NRELAPS validation against S IET-TF 1, S IET-TF2 tests

Additional validation or benchmark for non-LOCA - NPM sensitivity calculations

Reactivity event response

Conservative input

- NRELAPS benchmark against RETRAN-3D

Subchannel analysis

NPM non-LOCA integral response

- NRELAPS validation against NIST-1 IETs NLT-2a, NLT-2b, NLT-15p2 Overall conclusion: NRELAP5 code, with NPM system model, is applicable for calculation of th e NPM non-LOCAsystem response 11 PM-0220-68852 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 - NRELAP5 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 ti~~

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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-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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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 pressurizer

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-0220-68852 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Acronyms

AOO-AnticipatedOperational

MPS - Module Protection System Occurrences

MSIV - Main Steam Isolation Valve

CNV - Containment Vessel

MSS - Main Steam System

CVCS-Chemical and Volume Control System

NIST NuScale lntegralSystemTest-1

DHRS - Decay Heat Removal System

NPM - NuScale Power Module

ECCS - Emergency Core Cooling System

PIRT- Phenomena Identification and Ranking Table

EM - Evaluation Model

RCS - Reactor Coolant System

EMDAP- Evaluation Model Development and Assessment Process

RPV - Reactor Pressure Vessel

FW- Feedwater

RSV - Reactor Safety Valve

FWIV - Feedwater Isolation Valve

RVV - Reactor Vent Valve

IET - Integral Effects Test

SET - Separate Effects Test

KAIST - Korea Advanced Institute of

SG - Steam Generator Science and Technology

SSC - Structures, Systems, and

LOCA - Loss of Coolant Accident Components

MCHFR- Minimum Critical Heat Flux Ratio 15:- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - -

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Portland Office Richland Office 6650 SW Redoood Lane, 1933 Jaclwn 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://I/INI/W. nuscalepowercom

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Additional Material for Public Presentation Previously presented background material 17 PM-0220-68852

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Power Module Overv:iew Integral Pressurized Water Reactor steam line

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-0220-68852

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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 reactor 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 recirculation

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

Fail safe: ECCS valves open on loss of DC power 19 PM-0220-68852

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Decay Heat Removal System {DHRS)

Removes heat after FWIVs loss of normal cooling

Boiling/condensing loop - - DI-R actuation valves

Two redundant trains re actor pool

Redundant actuation and isolation valves for each train DI-R 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 power transients Control

Reactor trip Demineralized Water Isolation

Containment Isolation Increase in RCS inventory transients

Secondary Isolation

Reactortrip eves Isolation Heat Removal Decrease in RCS inventory transients

DHRS Actuation

Reactor trip CNV Isolation

ECCS Actuation ECCS actuation Subcooling Stability

Reactor trip

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Loss of Power - Non-LOCA Event Riser MPS Stable DH RS uncovery Rx trip DHRS cooling establis hed may occur 24 hrs 72 h rs

Availability of AC power

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) )) AC , DC power available DHRS event initiation cooling affects MPS Rx trip Stable DH RS cool ing Riser uncovery may 24 hrs ECCS whether ECCS DHRS established valves occur valve 72 A

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DC power available Non-LOCA event DHRS cooling ECCS cooling actuate, and initiation what time they Rx trip DHRS Stable DHRS cooling establis hed ECCS valve opening open 24 72 onlAB hrs hrs AC power unavailable, DC power Non-LOCA unavailable DHRS ECCS event cooling cooling initiation 23 PM-0220-68852

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