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ACRS Full Committee Presentation: NuScale Topical Report- Non-Loss-of-Coolant Accident, PM-0320-69141, Revision 0
	ML20069A154
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Site:	NuScale
Issue date:	03/04/2020
From:	; NuScale
To:	; Office of Nuclear Reactor Regulation
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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

PM-0320-69141 Revision: 0

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2 PM-0320-69141 Revision: 0 Presenters Ben Bristol Supervisor, System Thermal Hydraulics Meghan McCloskey Thermal Hydraulic Analyst Matthew Presson Licensing Project Manager Paul lnfanger Licensing Specialist Copyright 2020 by NuScale Power, LLC.

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Outline

Scope of non-LOCA L TR

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

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

Cooldown events Decrease in FW temperature Increase in FW flow Increase in steam flow Inadvertent opening of SG relief or safety valve Steam piping failures (postulated accident)

Loss of containment vacuum Containment flooding

Heatup events Loss of external load Turbine trip Loss of condenser vacuum Closure of MS~

Loss of non-emergency AC power Loss of normal FW flow Feedwater system pipe breaks (postulated accident)

Inadvertent operation of DHRS NuScale unique event

Reactivity events Uncontrolled bank withdrawal from subcritical Uncontrolled bank withdrawal at power Control rod misoperation Single rod withdrawal Control rod drop Inadvertent decrease in RCS boron concentration

Inventory increase event

- eves malfunction

Inventory decrease events Small line break outside containment (infrequent event)

Steam generator tube failure (postulated accident)

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Non-LOCA Event Acceptance Criteria Description AOO Infrequent Event Accident 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

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

< Limit If limit exceed, If limit exceed, Subchannel Centerline Temperature fuel assumed failed <1l fuel assumed failed <1l Containment Integrity

< Limits

< Limits Containment (pressure, temperature)

(pressure, temperature)

PIT analysis Escalation of an AOO to an accident (AOO)

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

Normal Normal or Radiological Dose Operations

< Limit

< Limit Accident radiological (1) NuScale safety analysis methodologies developed to demonstrate fuel cladding integrity maintained.

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Evaluation Models - General Non-LOCA Approach Plant design, Core design, Fuel rod design, Plant initial conditions, SSC performance 7

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Fuel cladding integrity Non-LOCA topical report Subchannel topical report TR-0516-49416-P TR-0915-17564-P-A

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Accident radiological analysis Radiological dose acceptance criteria Accident source term topical report TR-0915-17565-P Template#: 0000-21727-F01 R5

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 Increase in RCS inventory Decay heat removal system Chemical volume control system 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 w

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NRELAP5 Applicability for Non-LOCA After non-LOCA PIRT developed, gap analysis performed to determine how to address high-ranked phenomena:

Validation performed as part of NRELAP5 assessment for LOCA evaluation model

Additional validation or benchmark for non-LOCA

Conservative input

Subchannel analysis Key areas identified from gap analysis for short-term non-LOCA analysis:

DHRS modeling and heat transfer NRELAP5 validation against KAIST tests; NIST-1 SETs HP-03, HP-04 NPM sensitivity calculations

Steam generator modeling and heat transfer NRELAP5 validation against S1ET-TF1, S1ET-TF2 tests NPM sensitivity calculations

Reactivity event response NRELAP5 benchmark against RETRAN-30

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 NRELAP5 input (geometry, control and protection systems, etc)

2. Adapt NRELAP5 base model as necessary for specific event analysis and desired initial conditions

3. Perform steady state and transient analysis calcufations with NRELAP5

4. Evaluate results of transient analysis calculations:

Confirm margin to maximum RCS pressure acceptance criterion Confirm margin to maximum SG pressure acceptance criterion Confirm appropriate transient run tin:,~

execution to demonstrate safe, stab1l1zed condition achieved

5. Identify cases for subchannel analysis and extract boundary conditions (if applicable)

Conservative bias directions:

Maximum reactor power

Maximum core exit pressure

Maximum core inlet temperature

Minimum RCS flow rate NRELAPS CHF calculations for dummy hot rod may be used as a screening tool to assist analysts in determining limiting cases to be evaluated in downstream subchannel analysis

6. Identify cases for radiological analysis (if applicable)

Maximum mass release case Maixmum iodine spiking case

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Non-LOCA Methodology General Methodology (Section 7.1 }:

Steady-state conditions Treatment of plant controls Loss of power Single failure Bounding reactivity parameter input Biasing of other parameters:

initial conditions, valve characteristics, analytical limits and response times Operator action Event-specific Methodology (Section 7.2)

Description of event initiation and progression Acceptance criteria 'of interest' Limiting single failure, loss of power scenarios, or need for sensitivity calculations Initial condition biases and conservatisms, or need for sensitivity calculations Tabulated representative results of sensitivity calculations Example analysis results provided in Section 8

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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 Occurrences

CNV - Containment Vessel

CVCS-Chemical andVolumeControl System

DHRS - Decay Heat Removal System

ECCS-EmergencyCore Cooling System

EM - Evaluation Model

EMDAP-Evaluation Mode I Development and Assessment Process

FW - Feedwater

FWIV - Feedwater Isolation Valve

IET - Integral Effects Test

KAIST - Korea Advanced Institute of Science and Technology

LOCA - Loss of Coolant Accident

MCHFR-Minimum Critical Heat Flux Ratio

MPS - Module Protection System

MSIV - Main Steam Isolation Valve

MSS-Main Steam System

NIST NuScale lntegralSystemTest-1

NPM - NuScale Power Module

PIRT-Phenomena Identification and Ranking Table

RCS - Reactor Coolant System

RPV - Reactor Pressure Vessel

RSV - Reactor Safety Valve

RVV -ReactorVentValve

SET-Separate Effects Test

SG - Steam Generator

SSC - Structures, Systems, and Components 15~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~...._-~~~

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Portland Office 6650 SW RedWJod 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, MD 20852 301.770.0472 Richland Office 1933 Jadwn Ave., Suite 130 Richland, WA 99354 541. 360. 0500 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://vwvw.nuscalepower.com W Twtter: @NuScale_Pooor 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

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

Module protection system designed to automate event mitigation actuations (no operator actions) 18 PM-0320-69141 Revision : 0 Copyright 2020 by NuScale Power, LLC.

feedwater line steam generator

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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 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

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

Fail safe: ECCS valves open on loss of DC power re actor vent valve reactor recirculation valve 19 PM-0320-691 41 Revision: O Copyright 2020 by NuScale Power, LLC.

NOT10SCALE reactor vent valves reactor recirculation valve

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Decay Heat Removal System (DHRS)

Rem aves heat after loss of normal cooling Boiling/condensing loop

Two redundant trains Redundant actuation and isolation valves for each train Initiates on:

- Loss of power

- Loss of cooling 20 PM-0320-69141 Revision: 0 indication (ESFAS Signal)

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Deterministic Event Mitigation Module Protection Functions Reactivity Control

Reactor trip

CVCS/Demineralized Water Isolation RCS and Secondary Inventory Control

Containment Isolation

Secondary Isolation Heat Removal

DHRS Actuation

ECCS Actuation Subcooling

Reactor trip Event Mitigation Increase in heat removal transients

Reactor trip Secondary Isolation Decrease in heat removal transients

Reactor trip DHRS Actuation Reactivity and povver transients

Reactor trip Demineralized Water Isolation Increase in RCS inventory transients

Reactor trip eves Isolation Decrease in RCS inventory transients

Reactor trip CNV Isolation E CCS actuation Stability

Reactor trip

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Nu Scale Confidential, Proprietary Class 2 Loss of Power - Non-LOCA Event AC power available, DC power available AC power unavailable, DC power available AC power unavailable, DC power unavailable 23 PM-0220-68852 Revision: 0 MPS Stable DHRS Rx trip DHRS cooling established Riser uncovery may occur 24 hrs 72 hrs

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