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Enclosuacrs Presentation: Chapter 4, Reactor Overview, PM-0419-65096, Revision 0
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"ACRS Presentation: Chapter 4, Reactor OveNiew," PM-0419-65096, 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

NuScale Nonproprietary ACRS Presentation:

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Presentation Team Larry Linik Fuels Engineer Allyson Callaway Supervisor, Nuclear Analysis Ken Rooks Safety Analysis Engineer Matthew Presson Licensing Engineer 2

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Chapter 4: Reactor 4.1 Summary Description 4.2 Fuel System Design r----- -------- ----- --*-*;**------*-------* -***----*----- ------ - -- . - --**-~'"*----

4.3 :; Nuclear Design 4.4 Thermal and Hydraulic Design

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4.1 - Summary Description

160 MW Thermal Integral Natural Circulation PWR*

37 NuFuelHTP2' Fuel Assemblies

16 Hybrid AIC/84 C Control Rods NuScale Reactor Design Parameters Key Reactor Parameter Value Core thermal output (MWt) 160 System pressure (psia) 1850 Inlet temperature (°F) 497 Core average temperature (°F) 543 Average temperature rise in core (°F) 100 Best estimate flow (lb/hr) 4.66E+06 Core bypass flow (%)(best estimate) 7.3 Average linear power density (kw/ft) 2.5 Peak linear power for normal operating conditions (kw/ft) 5 Normal operation peak heat flux (Btu/hr-ft2) 170,088 Total heat flux hot channel factor, FQ

2 Heat transfer area on fuel surface (ft2) 6275.6 Normal operation core average heat flux (Btu/hr-ft2) 85,044 Core flow area (ft2) 9.79 Core average coolant velocity (ft/sec) 2.7 4

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4.2 - Summary Description

NuScale design based on Framatome's proven US 17x17 PWR Technology

Over 1500 17x17 HTP fuel assemblies with maximum burnup of 54 GWd/mTU

NuScale design features:

- Zircaloy-4 HTP' upper and intermediate spacer grids

- lnconel 718 HMP' lower spacer grid

- Coarse-mesh filter plate on bottom nozzle

- Zircaloy-4 MONOBLOC' guide tubes Quick-disconnect top nozzle

- Alloy M5 fuel rod cladding 5

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4.2 - Fuel Assembly Design

17x17 HTP' Spacer Grid Design

- Zircaloy-4 strip

- Proven Grid-to-Rod-Fretting (GTRF) resistance across many US PWRs

- Multiple (8) line contacts on each fuel rod

- Flow channels to promote flow mixing 6

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4.2 - Fuel Assembly Design

17x17 HMP' Spacer Grid Design

- Alloy 718 strip

- Similar construction to HTP spacer grids

- Straight channels

Coarse Mesh Bottom Nozzle 304 ss Frame Alloy 286 Filter plate (for debris capture)

Use in several US 17x17 PWR plants with no debris failures 7

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4.2 - Design Basis TR-0116-20825-P-A, Rev. 1, Applicability of AREVA Fuel Methodology for the NuScale Design SRP Criteria Review Summary SRP 4.2 Acceptance Analysis Framatome Topical Report Criteria Shipping And Handling Stress Analysis 1.A.i Fuel Assembly/Component Stress Analysis 1.A.i FIV Assessment 1.A.iii EM F-92-116(P)(A)

Axial Growth (Rod and Assembly) 1.A.v Fuel Lift Analysis 1.A.vii Internal Hydriding 1.B.i Clad Stress Analysis 1.A.i Fuel Rod Buckling Analysis 1.A.i BAW-10227P-A Clad Fatigue Analysis 1.A.ii Clad Corrosion Analysis

1.A.iv Fuel Rod Internal Pressure 1.A.vi BAW-10231 P-A Fuel Centerline Melt Analysis 1.B.iv Transient Clad Strain Analysis 1.B.vi BAW-10084P-A Clad Creep Collapse Analysis 1.B.ii BAW-10227P-A Rod Bow Evaluation 1.A.v XN-75-32(P)(A) 8 PM-0419-65096 Revision: 0 Copyright 2019 by NuScale Power, LLC.

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4.2 - Design Basis TR-0716-50351-P, Rev. 0, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces SRP Criteria Review Summary SRP 4.2 Acceptance Analysis Framatome Topical Report Criteria LOCA/Seismic Stress Analysis Appendix A ANP-10337P-A 9

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4.2 - Fuel Testing

BOL/EOL testing to characte_rize the mechanical response of the fuel assembly for Seismic/LOCA calculations

- Axial Stiffness Testing

- Lateral Pluck Testing

- Lateral Stiffness Testing

- Forced Vibration testing

- Vertical Drop testing

Life and Wear testing

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4.2 - Fuel Testing

Hydraulic Flow Te.sting to determine pressure drop and lift characteristics

- Flow Lift Testing

- Pressure Drop Coefficient Testing

Mechanical testing of bottom nozzle

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4.2 - Control Rod Design

CRA design based on Framatome's proven US 17x17 PWR Technology CoupUn. to CROM Hybrid design - B4 C and AIC absorbers - SprlnC Retainer Bolt 24 control rods with Stainless Steel cladding - Spider Body One-piece cast stainless steel spider Standard 17x17 rod configuration Flex Joint Flex joint formed by the combination of the pin, nut, upper end plug and spider boss Parameter Value CRA total weight (lb) 43 CRA total height (inch) 94 .37 Control rod length - short/medium/long (inch) 87 .065 I 87.425 I 87 .875 Control rod outer diameter (inch) 0.381 Control rod inner diameter (inch) 0.344 Control rod bottom end plug length (inch) 1.913 B4C outer diameter (inch) 0.333 B4C stack length (inch) 62.0 Ag-In-Cd outer diameter (inch) 0.336 Ag-In-Cd stack length (inch) 12.0 Height of CRA spider assembly (inch) 10.387 CRA shaft outer diameter (inch) 1.804 12 PM-0419-65096

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4.2 - Control Rod Design

CRA Analyses Cladding stress and strain Cladding creep Spider stress due to shipping and SCRAM Absorber melt Rod internal pressure Rod and spider spring loading due to SCRAM and absorber growth Component Material Spider 304L stainless steel Rod end plugs 308L stainless steel Cladding 304L stainless steel Solid spacer, lock pin, nuts, tension bolt 304L stainless steel Spring retainer 17-4 PH stainless steel Spider spring Alloy 718 Control rod plenum spring 302 stainless steel Absorber materials 80% Ag - 15% In - 5% Cd and B4C Stack support Alloy X750 13 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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4.2 - Control Testing

CRA Drop Alignment Test

- Limiting drop

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4.2 - Conclusion

Proven Framatome 17X17 components

NRC Accepted Framatome codes and methods

Standard Framatome fuel design, analysis, prototype fabrication, and testing

Standard CRA design, analysis, and testing

Both fuel and CRA designs are essentially reduced height versions of current designs 15 PM-0419-65096 Revis ion: 0 Copyright 2019 by NuScale Power, LLC.

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4.3 - Design Basis and Methods

Core design conforms to NUREG-0800, SRP 4.3 guidance

Latest version of Studvik's Core Management Suite (CMS5)

- CASM05 and SIMULATES steady-state neutronics software

- CMS5 analytical methods for neutronic analysis are approved for use

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4.3 - Nuclear Design

Fuel Cycle Design 2-year, 3-batch cycle Out-to-in fuel shuffle Gadolinia burnable absorber

Equilibrium Core Reference design and analysis in DCA Equilibrium core design is representative ; used for demonstration of methods Limits placed on th is core design are applied to the design of all cycles Initial , transition , and equilibrium cycles must meet analyzed limits A-01 : Batch A Type 1, 4 .05 wt% 235 U A-02 : Batch A Type 2, 4 .55 wt% 235 U, with Gadolinia B-01: Batch B Type 1, 4.05 wt% 235 U B-02 : Batch B Type 2, 4.55 wt% 235 U, with Gadolinia C-01 : Batch c Type 1, 4.05 wt% 235 U C-02: Batch C Type 2, 4.55 wt% 235 U, with Gadolinia C-03: Batch C Type 3, 2.60 wt% 235 U A - Twice burned , B - Once burned, C - Fresh 17 PM-0419-65096

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4.3 - Core Design Limits

Cycles are designed to meet constraints and requirements

- Energy output and burnup

- Enrichment limits, zoning, and gadolinia loading

Core design limits are verified for each cycle to confirm the safety analysis bases and ensure specified acceptable fuel design limits (SAFDLs) are not exceeded

- Moderator temperature and Doppler reactivity coefficients

- Kinetics parameters

- Critical and refueling boron concentration

- Axial and radial peaking

- Shutdown margin and long term shutdown capability

- Event-specific limits (i.e. power peaking)

Core design limits are set to ensure that sufficiently conservative values are analyzed 18 PM-0419-65096

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4.3 - Core Power Distributions

Power distributions are protected by monitoring to ensure that limits are not exceeded during normal operation

Control rod assemblies (CRAs) are arranged into regulating and shutdown banks

The power dependent insertion limits (POils) and axial offset (AO) window ensure axial and radial peaking are within design limits

The NuScale power module is stable with respect to axial and radial xenon imbalances due to small core size, H/D ratio 225 I I I I I I I Axial Offset Window l

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4.3 - Shutdown Margin

Shutdown Margin (SOM)

- The instantaneous amount of reactivity by which the reactor is subcritical, or would be subcritical from its present condition , assuming all CRAs are fully inserted with the worst CRA assumed stuck out of the core.

- Power defect, temperature defect to hot zero power, margin for uncertainties, WRSO, no boration

Long Term Shutdown (LTSD)

- The instantaneous amount of reactivity by which the reactor is subcritical , or would be subcritical from its present condition , assuming all CRAs are fully inserted and the RCS is cooled to equilibrium conditions.

- Power defect, temperature defect to pool temperature, margin for uncertainties, all rods in , no boration

Distinct definitions for SOM and LTSD establish the design basis and satisfy GDC 26 and PDC 27

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4.4 - Thermal and Hydraulic Design Parameter Basis Lim it / Protection Critical heat flux 95/95 criteria that hot rod SL2.1.1.1 avoids boiling transition Fuel temperature Fuel centerline temperature SL2 .1.1.2 remains below melting limit Reactor core coolant Primary system flow remains TS 3.4.1 flow within ranges assumed in the safety analysis Hydrodynamic Normal operation and AOOs Module protection stability do not lead to instability system analytical limits prevent loss of subcooling and instability Acceptance criteria ensure GDC 10 and 12 compliance 21 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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4.4 - NuScale Design Fuel and Core Conditions Parameter NuScale Design Avg. Fuel Rod Linear Heat Generation Rate 2.5 kW/ft Peak Fuel Temperature (cycle beginning) 1620 °F Core Inlet Mass Flux 0.4 Mlb/hr-ft2 Core Inlet Subcooling 135 °F Core Exit Subcooling 25 °F Core Exit Void Fraction 0.0 Hot Channel Exit Thermodynamic Quality 0.05 Nominal core average exit conditions similar to traditional PWRs 22 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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4.4 - Evaluation Methods & Implementation

CHF and fuel temperature

Methodology topical reports:

TR-0915-17564-P-A TR-0116-21012-P-A TR-0716-50350 [in review]

Subchannel Analysis NuScale Power Critical Heat Rod Ejection Accident Methodology Flux Correlations Methodology

VIPRE-01 used to calculate reactor core flow and enthalpy distribution

- Assess thermal margin to CHF for normal operations and most DBEs to support FSAR Chapter 15

Peak linear heat rate using maximum local peaking to assess power margin to fuel melt

RCS flow driven by density gradient and system flow resistances

Application: FSAR Chapter 15 except 15.6, 15.9

Hydrodynamic stability Methodology Software Implementation TR-0516-49417 {in review] PIM 15.9 23 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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4.4 - CHF Limit and Margin 'Stack-up' Normal Operation > 3.0 I\

Ranges MCHFR Margin for transients (Applied Methodology Input Biases & Uncertainties)

CHF Analysis Limit 1.284

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Margin for CHF penalties (rod bow, F5) 95/95 CHF Safety Limit 1.21 l t

CHF correlation uncertainties and biases CHF Failure 1.00 Robust treatment of analytical and design uncertainties 24 PM-0419-65096

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4.4 - Pressure vs. Temperature Operation M _ a,_

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Module protection system (Ch. 7, red )

Technical specification LCOs (Ch . 16, blue)

CHF and fuel melting SAFDLs precluded for normal operation & DBEs 25 PM-0419-65096

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4.5.1 - Control Rod Drive System Materials 0

All pressure boundary materials are CROM SIJ'PORT FRAME designed in accordance with ASME Code

Pressure boundary materials

- Austenitic stainless steel materials as addressed in Tier 2, Section 5.2.3 and RG 1.44 Revision 1,

- TYPICAL CONTROL ROO ORNE (CRO) SHAFT PRESSURIZER with corresponding weld materials

Non-pressure boundary materials

- Austenitic & martensitic stainless steels, nickel-

TYPICALCRO SliAFT SUPPORT base materials

- Cobalt-based materials are used in a very small

( portion where alternate material will not perform TYPICAL CONTROL ROil ASSEM81.Y satisfactorily.

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4.5.2 - Reactor Internals and Core Support Structure Materials

Materials are selected based on compatibility with their environment

Components are considered for peak neutron fluence and evaluated using Electric Power Research Institute (EPRI) materials reliability program criteria

Parts exposed to reactor coolant are made of corrosion resistant material

- Made mostly of austenitic stainless steel, some nickel-base materials and limited cobalt base materials Components and materials are consistent with those for existing, proven designs 27 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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4.6.1 Description of the Control Rod Drive System Control Rod Drive Mechanisms (CROM)

Components and materials are consistent MAST ASSEMBLY with those for existing PWR magnetic jack CROM design.

SENSOR COIL

Additional remote disconnect mechanism HOUSING (OUTSIDE THE PRESSURE (ROM) coil and latch are provided on top SENSOR COILS BOUNDRY) of the typical configuration of three coils.

Components external to primary pressure REMOTE DISCONNECT boundary are designed for operation in an COIL LATCH HOUSING DRIVE COILS evacuated containment, but not required LIFT COIL to operate during ECCS blowdown .

LATCH MECHANISM ,

MOVEABLE COIL

Components internal to primary pressure STATIONARY COIL DRIVE COIL HOUSING boundary (drive shaft & latch mechanism)

(OUTSIDE THE PRESSURE are exposed to steam & non-condensable BOUNDRY) gases (N 2 , H2 ) on top of the pressurizer.

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4.6.1 Description of the Control Rod Drive System

Components external to primary pressure boundary are non-safety

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II related (B2) and support SCRAM function by de-energizing from RTB.

The CROM electrical coils are exposed to a high-temperature vacuum, and ELECTROMAGNETIC DRIVE COIL require a cooling water system. The cooling coils envelope the electrical coils and are protected against impact COOLING COILS by an outer coil housing and mast assembly (pipe).

Disassembly of electrical components OUTERDRNE COIL HOUSING from the pressure boundary by upward retraction (slip-fit) after disconnection of ISOMETRIC VIEW WITH OUTER DRIVE cables and cooling water hoses.

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4.6.1 Description of the Control Rod Drive System

Initially, the Disconnect Rod is withdrawn and shaft fingers are in the retracted position allowing insertion of the shaft fingers into CRA Hub CONTROL ROD

Once shaft fingers are inserted in the CRA hub, the DRIVE SHAFT Disconnect Rod is released and forces shaft fingers to expand outward nd engage CRA Hub for normal DISCONNECT ROD operation FINGERS

To release, the Disconnect Rod is withdrawn allowing the PLUG CONTROL ROD shaft fingers to return to the ASSEMBLY HUB retracted position

The drive shaft can now be withdrawn and sepa rated from the CRA CONNECTED AND DISCONNECTED AND FINAL DISCONNECTED LOCKED POSITION RELEASED POSITION RESTING POSITION 30 PM-041 9-65096

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Chapter 4 - COL Items COL Item 4.2-1

A COL applicant that references the NuScale Power Plant design certification and wishes to utilize non-baseload operations will provide justification for the fuel performance codes and methods corresponding to the desired operation.

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Acronyms

AO - Axial Offset

ECCS - Emergency Core Cooling System

AOO - Anticipated Operational Occurrences

EOL - End of Life

ASME - American Society of

EPRI - Electric Power Research Mechanical Engineers Institute

BOL - Beginning of Life

FIV - Flow Induced Vibration

CHF - Critical Heat Flux

FSAR - Final Safety Analysis Report

COL - Combined License

H/D - Height over Diameter

CRA - Control Rod Assembly

HMP - High Mechanical Performance (Spacer Grid)

CRDM - Control Rod Drive Mechanism

HTP - High Thermal Performance (Spacer Grid)

DBE - Design Basis Event

LOCA - Loss of Coolant Accident

DCA - Design Certification Application

LTSD - Long Term Shutdown 32 PM-0419-65096

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Acronyms

MCHFR - Minimum Critical Heat Flux Ratio

MPS - Module Protection System

PDIL - Power Dependant Insertion Limit

PWR - Pressurized Water Reactor

RCS - Reactor Coolant System

RG - Regulatory Guide

RTB - Reactor Trip Breaker

SAFDL - Specified Acceptable Fuel Design Limit

SDM - Shutdown Margin

SRP - Standard Review Plan

TS - Technical Specifications 33 PM-0419-65096 Revision : 0 Copyright 2019 by NuScale Power, LLC .

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Portland Office Richland Office 6650 SW Redwood Lane, 1933 Jadwin Ave. , Suite 130 Suite 210 Richland, WA 99354 Portland, OR 97224 541 .360.0500 971 .371 .1592 Arlington Office Corvallis Office 2300 Clarendon Blvd. , Suite 1110 1100 NE Circle Blvd., Suite 200 Arlington, VA 22201 Corvallis, OR 97330 541 .360.0500 London Office 1st Floor Portland House Rockville Office Bressenden Place 11333 Woodglen Ave. , Suite 205 London SW1 E 5BH Rockville, MO 20852 United Kingdom 301 . 770.0472 +44 (0) 2079 321700 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 http://www. nuscalepower com

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