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ACRS Full Committee Presentation: NuScale Topical Report-Rod Ejection Accident Methodology, PM-0320-69146, Revision 0
	ML20069A163
Person / Time
Site:	NuScale
Issue date:	03/05/2020
From:	; NuScale
To:	; Office of Nuclear Reactor Regulation
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ML20069A577	List: ML20069A157; ML20069A163;
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	L0-0320-69151 PM-0320-69146, Rev 0
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"ACRS Full Committee Presentation: NuScale Topical Report- Rod Ejection Accident Methodology," PM-0320-69146, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., SJite 200 Corvallis, 0-egon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Nonproprietary ACRS Full Committee Presentation N uScale Topical Report Rod Ejection Accident Methodology March 5, 2020 PM-0320-69146

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Presenters Kenny Anderson Nuclear Fuels Analyst Matthew Presson Licensing Project Manager 2

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Opening Remarks - NuScale T/H Methods NRELAP5 System T/H Analysis Basis code TR-0516-49422-P I

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NRELAP5 code developed from RELAP5-3D 1---;.i: :

LOCA EM

Valve opening event I I I I TR-0516-49084-P Modified to address NuScale- *-------------------------* ,--------------

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Containment :

specific phenomena/systems :----- ;-------: 1 - - - - - - - -~, respo nse :

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,I Control rod *I TR-0516-49416-P

LOCA Evaluation Model (EM) : ejection : I I-----------*** I I

I developed following RG 1.203 EMDAP : (T/H response)

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I Non-LOCA I

L--------------' i' EM

t_ ______ r____ _*;

LOCA EM extended to derive TR-0716-50350-P EMs for other events as shown in this figure. FSARCh 5, RAI 9508 LOCA EM assessment basis --------------

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Extended :

leveraged for non-LOCA. ~ DHRS :

cooli ng :

Additional supporting EMs include

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FSARCh 15

- Nuclear Analysis Codes - ----*-----*---

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, power :

Critical Heat Flux - I I

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TR-0116-21012-P-A ******-***--**

TR-0916-51299-P I

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Subchannel Analysis - : j Long tefm :

TR-0915-17564-P-A ' - - - - - - - - - - - . .;,iI cooling with :

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Agenda

Event Overview

Acceptance Criteria

PCMI Criteria - DG-1327

Method Flowchart

Steady State Initialization

Event Evaluations

Summary 4

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Overview

NuScale seeks approval of methodology for modeling rod ejection accident (REA) events

Bounding reactivity initiated accident (RIA) from General Design Criteria (GDC) 28

REA is unique in comparison to other Ch. 15 events Description Rod Ejection Other Events Dominant Physics Nuclear Thermal-Hydraulics Timing milli-sec secto hr Spatially Local Global Peak power -5x Full Power -1 .2x Full Power Integrated Energy Low Lowto High Failure of ASME Class 1 Postulated Cause Single Equipment Failure Pressure Boundary Acceptance Criteria Specialized Generic 5

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Unique Event Acceptance Criteria Topical Criteria Description Unique?

Section Maximum reactor coolant system pressure 5.3 No Hot zero power (HZP) fuel cladding failure 5.5.2 Yes FGR effect on cladding differential pressure N/A Yes Critical heat flux (CHF) fuel cladding failure 5.4.1 No Cladding oxidation-based PCMI failure 5.5.3 Yes Cladding excess hydrogen-based PCMI failure N/A Yes Incipient fuel melting cladding failure 5.5.1 No Peak radial average fuel enthalpy for core cooling 5.5.2 Yes Fuel melting for core cooling 5.5.1 No Fission product inventory (failed fuel census) 5.6 Yes

Submitted NuScale design and method inherently precludes fuel failure, thus no accident radiological consequences are evaluated.

PCMI: Pellet-Clad Mechanical Interaction 6

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Revised PCMI Criteria

In general, the NuScale REA methodology has adopted the limiting criteria of the 'Clifford Letter'(ML14188C423), now included in draft guide DG-1327 (ML16124A200). In spirit, NuScale is prepared forthis regulatory change:

- Closed session presents example results , showing large margins for enthalpy rise

- A techn ical 'formality' inhibits complete adoption at this time. NuScale does not currently have a val idated cladding H2 model to convert local exposure to excess cladding hydrogen

- Oxidation criteria from NUREG-0800 Section 4.2 , Appendix B (ML07074000) is used

- To simplify method , no exposure is credited (Limit: 75 Li cal/gm)

- NuScale MS cladding less susceptible than other zirc alloy-type clad used in the industry 200 200 - - - - - - - - - - - - - - - - - ,

175 (0.04, 150) ~ (0, lSO) (75. 150)

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150 }1~ - - ~ ~ ~ -

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"; 125 Cladding Failure t 125 Oadding Failure II)

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75 Cladding Intact ci

<< 50 LI. 50 i (0.20, 0) l.

25 25 0

0 0 so 100 150 200 250 0 0.04 0.08 0.12 0.16 0.2 txcau a..ddlnt Hydn>sen (wppm)

Oxide/Wall Thickness

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Unique Event Method (Flowchart)

NRELAPS Dynamic System Response SIMULATES Steady State Initialization S1MULATE-3K Dynamic Core

Response

VIPRE-01 SubchannelCHF Evaluation MCHFR Above Correlation Design Limit 0

Adiabatic Heatup Fuel Response 8

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Steady-State Initialization

SIMULATES: Setup the core response analysis

Code shown to be appropriate in TR-0616-48793-A (Nuclear Analysis Codes and Methods Qualification)

Determination of the worst rod stuck out (WRSO)

- Assumption bounds potential for ejected assembly to damage adjacent control rod assembly

- Due to rapid nature of the event, location does not significantly affect the results in NuScale application 9

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Dynamic Core Response

S1MULATE-3K: Model transient core response

Benchmarked to SPERT-111 experiment and NEACRP computational benchmark

- Benchmarks demonstrate the combined transient neutronic, thermal-hydraulic, and fuel pin modeling capabilities

- S1MULATE-3K results generally in excellent agreement with the results from the two benchmark problems

Uncertainties applied for each simulation:

- Delayed Neutron Fraction

- Ejected Rod Worth

- Doppler Temperature Coefficient

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CHF Evaluation

VIPRE-01: Model detailed thermal-hydraulics

Evaluate critical heat flux (CHF) acceptance c ~---------____,

Code shown to be appropriate in TR-0915-17.

Analysis Methodology)

Unique event differences in method:

- Smaller axial nodalization (smaller time steps)

- Radial power distribution (case-specific)

- Axial power distribution (peak assembly)

- Convergence parameters

Additional parametric sensitivity cases pe application to holistically justify difference 11 PM-0320-691 46

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Adiabatic Fuel Heatup

Hand-Calculation: Model fuel response

Total energy (from S1MULATE-3K) during the transient is integrated

Conservative as no energy is allowed to leave the fuel rod

Energy is then converted into either a temperature or enthalpy increase

Fuel rod geometry, heat capacity, and power peaking factors taken into account

Calculated values compared to NRC developed acceptance criteria

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Dynamic System Response I

NRELAPS: Evaluate system response for input to CHF Evaluation CON1IOI. ROD ,

DRIVE MECHANISM

Code shown to be appropriate in TR-0516-49416 (Non-LOCA Methodologies)

Transient power from S1MULATE-3K MAit STEAM RISER utilized as input (PRIMARY R.OW)

STEAIJ GENERATOR

- No reactivity calculation performed in NRELAPS (SECON DARY FlOWI

Provides system thermal-hydraulic CONTAINMENT 1/ESSB.

conditions to subchannel (CHF) evaluation

- System flow, pressure, and inlet temperature REAClOR PRES$URE vessa.

CORE

- 'Screens' cases for potential to be limiting (PRIIARY FLOW)

- Family of limiting cases evaluated with VIPRE-01 13 PM-0320-69146

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Dynamic System Response II

NRELAPS: Evaluate system response for pressurization

Limiting scenario: Low ejected worth that raises the power quickly to just below both the high power and high power rate trip 'setpoints'

Point-kinetics model used based on bounding static worth

Peak system pressure calculated compared to acceptance criteria

Example results to be presented in closed session 14 PM-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.

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Summary

A conservative analysis method for the unique rod ejection accident

Topical report provides details and justification for:

- Software tools and acceptance criteria used

- Applicability of the method and tools

- Appropriate treatment of uncertainties

Results from application of the method provide input to FSAR Chapter 15 15 PM-0320-69146

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Acronyms

CHF - Critical Heat Flux

GDC-General Design Criteria

HZP-HotZero Power

MCHFR- Minimum Critical Heat Flux Ratio

NEACRP- Nuclear Energy Agency Committee on Reactor Physics

PCMI - Pellet Clad Mechanical Interaction

REA - Rod Ejection Accident

RIA- Reactivity Initiated Accident

WRSO - Worst Rod Stuck Out 16 PM-0320-69 146

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