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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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"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 L0-0320-69151

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Opening Remarks - NuScale T/H Methods System T/H Analysis Basis NRELAP5 code developed from RELAP5-3D Modified to address NuScale-specific phenomena/systems LOCA Evaluation Model (EM) developed following RG 1.203 EMDAP LOCA EM extended to derive EMs for other events as shown in this figure.

LOCA EM assessment basis leveraged for non-LOCA.

Additional supporting EMs include Nuclear Analysis Codes -

TR-0716-50350-P-A Critical Heat Flux -

TR-0116-21012-P-A Subchannel Analysis -

TR-0915-17564-P-A NRELAP5 code TR-0516-49422-P I

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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 Dominant Physics Timing Spatially Peak power Integrated Energy Postulated Cause Acceptance Criteria 5

PM-0320-69146 Revision : 0 Rod Ejection Nuclear milli-sec Local

Other Events Thermal-Hydraulics secto hr Global

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Unique Event Acceptance Criteria Criteria Description Topical 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:

200 175 Closed session presents example results, showing large margins for enthalpy rise A technical 'formality' inhibits complete adoption at this time. NuScale does not currently have a validated 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 Lical/gm)

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Unique Event Method (Flowchart)

SIMULATES Steady State Initialization 8

PM-0320-69146 Revision: 0 S1MULATE-3K Dynamic Core

Response

MCHFR Above 0 Correlation Design Limit

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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 Revision : 0 Copyright 2020 by NuScale Power, LLC.

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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

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

Transient power from S1MULATE-3K utilized as input

- No reactivity calculation performed in NRELAPS

Provides system thermal-hydraulic conditions to subchannel (CHF) evaluation

- System flow, pressure, and inlet temperature

- 'Screens' cases for potential to be limiting

- Family of limiting cases evaluated with VIPRE-01 CON1IOI. ROD DRIVE MECHANISM MAit STEAM RISER (PRIMARY R.OW)

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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 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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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-0320-69146 Revision : 0 Copyright 2020 by NuScale Power, LLC.

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Portland Office 6650 SW Redoood 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 Jadvtin Ave., Suite 130 Richland, WA 99354 541. 360. 0500 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980. 349. 4804 http:l!WNW. nuscalepower.com

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