ML20054B634

Nuscale - ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology, PM-1019-67365, Revision 0

ML20054B634
Person / Time
Site:	NuScale
Issue date:	02/17/2020
From:	NuScale
To:	Office of Nuclear Reactor Regulation
References
LO-0220-68846 PM-1019-67365, Rev 0
Download: ML20054B634 (18)

Text

L0-0220-68846

Enclosure:

"ACRS Subcommittee Presentation: NuScale Topical Report - Rod Ejection Accident Methodology,"

PM-1019-67365, Revision O 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 Subcommittee Presentation NuScale Topical Report Rod Ejection Accident Methodology I

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I February 19, 2020 PM-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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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 NRELAPS System T/H Analysis Basis code TR-0516-49422-P

• NRELAP5 code developed from f----~  : LOCA EM

• open ing ,

Va lve RELAP5-3D I I

event I

I TR-0516-49084--P I I

• Modified to address NuScale- ~-------------------------* *--------------

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specific phenomena/systems ,------ -------,

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r--------~!.. i c~~:~;::nt ! ~ analysis  :

, Cont rol rod

• TR-05 16-49416-P

• LOCA Evaluation Model (EM) I

ejection I

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

I L-------------'

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I Non-LOCA EM I

• LOCAEM extended to derive TR-0716-50350-P I I EMs for other events as shown in ~------ t----

• this figure. FSAR Ch 5, RAI 9508

• LOCA EM assessment basis .*--------------.*

Extend ed  :

leveraged for non-LOCA. ~ DHRS  :

cooling  :

• Additional supporting EMs include

• I I

• Nuclear Analysis Codes -

TR-0716-50350-P-A ~-____...

I FSAR Ch 15

- Overcoolin g

. I

• Critical Heat Flux -

return to power .*:

I I TR-0116-21012-P-A --------------

TR -0916-5129~P I I

• Subchannel Analysis -  : Long term  :

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

ECCS

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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 Desig n 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 Fa ilure 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 validated cladding H2 model to convert local exposure to excess cladd ing hydrogen

- Oxidation criteria from NUREG-0800 Section 4 .2, Append ix B (M L07074000) is used

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

- NuScale MS cladding less susceptible than other zirc alloy-type clad used in the ind ustry NUREG-800, Sec. 4 .2, Fig. B-1 DG-1327, Fig. 2 200 175 17'i '

(0.04, 150) i (0,1§0) (75.150)

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150  ! 150 1j Cladding Failure 0:addina: Jal ure

';' 125 _;: 125 1 Ill ii: iJ...

NuScale '11 100 *

~ 100  !

/ criteria

• ii 5C w 75 (0.08, 75)

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f 75 Oaddin1 Intact 11 "i I  :;;

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u. 50 1 +- NuScale ... 50 (0.20, 0) i I max ra t*10 per 25

• 25 1 1 corrosion limit.

0 0 0 50 100 150 200 250 300 0 0.04 0.08 0.12 0.16 0.2 ExatuCladd I Hydrocen (wppmJ Ox ide/Wall Thickness

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Unique Event Method {Flowchart)

NRELAPS Dynamic Peak RCS Pressure System Response Below 120% Limit SIMULATE-3K VIPRE-01 MCHFR Above SIMULATES Steady Dynamic Core SubchannelCHF Correlation Design State Initialization Response Evaluation Limit Fuel Temperature Adiabatic Heatup

+ - - - - ~ and Enthalpy Below Fuel Response Regulatory Criteria 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 Tern perature Coefficient

- Moderator Temperature Coefficient 10 PM-1019-67365 Revision : 0 Copyright 2020 by NuScale Power, LLC.

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CHF Evaluation

• VIPRE-01: Model detailed thermal-hydraulics

• Evaluate critical heat flux (CHF) acceptance criteria

• Code shown to be appropriate in TR-0915-17564-A (Subchannel 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 performed with each application to holistically justify differences 11 PM-1019-67365 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

- Example values provided in closed session 12 PM-1 019-67365

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Dynamic System Response I

• NRELAP5: Evaluate system response for input to CHF Evaluation COHTROL ROD ,

DRIVE MECHo\NISM

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

PRESSURIZER

• Transient power from S1MULATE-3K R1SER utilized as input (PRIMRY R(1NJ STEAII GENERATOR

- No reactivity calculation performed in NRE LAPS (SWlHl!MY FLOW}

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

- System flow, pressure, and inlet tern perature CORE

- 'Screens' cases for potentia l to be limiting (PRIMRY R(1NJ

- Family of limiting cases evaluated with VIPRE-01 13 PM-1019-67365

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Dynamic System Response II

• NRELAP5: 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-1019-67365

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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-1019-67365 Revision: 0 Copyright 2020 by NuScale Power, LLC.

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Acronyms

• CHF - Critical Heat Flux

• GDC - General Design Criteria

• HZP-HotAero 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-1019-67365

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