Enclosure: 'ACRS Presentation: Chapter 4, Reactor Overview ... · L0-0419-65169 Enclosure: "ACRS Presentation: Chapter 4, Reactor OveNiew," PM-0419-65096, Revision 0 NuScale Power,

L0-0419-65169

Enclosure:

"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

PM-0419-65096

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NuScale Nonproprietary

ACRS Presentation: NuScale Chapter 4, Reactor Overview

April 17, 2019

Copyright 2019 by NuScale Power, LLC. N ~!J,li.~~-~,§.

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Presentation Team

Larry Linik

Fuels Engineer

Allyson Callaway

Supervisor, Nuclear Analysis

Ken Rooks

Safety Analysis Engineer

Matthew Presson

Licensing Engineer

Copyright 2019 by NuScale Power, LLC. i-1 ~!:'.~.~~!:.~"

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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

• ·• • - • • • •• -• I •• "'" • • ·•• • • • •· ~ --

<,

4.5 , Reactor Materials

4.6 Functional Design of Control Rod Drive System

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4.1 - Summary Description

• 160 MW Thermal Integral Natural Circulation PWR·

• 37 NuFuelHTP2™ Fuel Assemblies

• 16 Hybrid AIC/84C Control Rods

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NuScale Reactor Design Parameters

Key Reactor Parameter

Core thermal output (MWt)

System pressure (psia)

Inlet temperature (°F)

Core average temperature (°F)

Average temperature rise in core (°F)

Best estimate flow (lb/hr)

Core bypass flow (%)(best estimate)

Average linear power density (kw/ft)

Peak linear power for normal operating conditions (kw/ft)

Normal operation peak heat flux (Btu/hr-ft2)

Total heat flux hot channel factor, FQ ·

Heat transfer area on fuel surface (ft2)

Normal operation core average heat flux (Btu/hr-ft2)

Core flow area (ft2)

Core average coolant velocity (ft/sec)

Copyright 2019 by NuScale Power, LLC.

Value

160

1850

497

543

100

4.66E+06

7.3

2.5

5

170,088

2

6275.6

85,044

9.79

2.7

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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

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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

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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

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Alloy 286 Filter plate (for debris capture) Use in several US 17x17 PWR plants with no debris failures


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4.2 - Design Basis

TR-0116-20825-P-A, Rev. 1, Applicability of AREVA Fuel Methodology for the NuScale Design

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SRP Criteria Review Summary Analysis

SRP 4.2 Acceptance Criteria

Shipping And Handling Stress Analysis 1.A.i

Fuel Assembly/Component Stress Analysis 1.A.i

FIV Assessment 1.A.iii

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

Clad Fatigue Analysis 1.A.ii

Clad Corrosion Analysis · 1.A.iv

Fuel Rod Internal Pressure 1.A.vi

Fuel Centerline Melt Analysis 1.B.iv

Transient Clad Strain Analysis 1.B.vi

Clad Creep Collapse Analysis 1.B.ii

Rod Bow Evaluation 1.A.v


Framatome Topical Report

EM F-92-116(P)(A)

BAW-10227P-A

BAW-10231 P-A

BAW-10084P-A BAW-10227P-A XN-75-32(P)(A)

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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 Analysis

SRP 4.2 Acceptance Framatome Topical Report

Criteria LOCA/Seismic Stress Analysis Appendix A ANP-10337P-A

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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 - 1,000-hour test to characterize the grid to rod fretting

performance of the fuel

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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

- Develop load/deflection data to determine load limits

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4.2 - Control Rod Design

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

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Hybrid design - B4C and AIC absorbers

24 control rods with Stainless Steel cladding

One-piece cast stainless steel spider

Standard 17x17 rod configuration

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


CoupUn. to CROM

- SprlnC Retainer Bolt

- Spider Body

Flex Joint

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4.2 - Control Rod Design

• CRA Analyses

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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

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4.2 - Control Testing

• CRA Drop Alignment Test - Limiting drop

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!'. 500 t------+-----'~ ----------------j, ~

f 400 t---------lr-----"t----------------1 .. C

rs

11me (s)


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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

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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

- Topical Report TR-0616-48793-P-A "Nuclear Analysis Codes and Methods

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Qualification"

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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

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Reference design and analysis in DCA

Equilibrium core design is representative; used for demonstration of methods Limits placed on this core design are applied to the design of all cycles

Initial , transition , and equilibrium cycles must meet analyzed limits

Copyright 2019 by NuScale Power, LLC .

A-01 : Batch A Type 1, 4 .05 wt% 235U

A-02: Batch A Type 2, 4 .55 wt% 235U, with Gadolinia B-01: Batch B Type 1, 4.05 wt% 235U B-02: Batch B Type 2, 4.55 wt% 235U, with Gadolinia C-01 : Batch c Type 1, 4.05 wt% 235U C-02: Batch C Type 2, 4.55 wt% 235U, with Gadolinia C-03: Batch C Type 3, 2.60 wt% 235U

A - Twice burned, B - Once burned, C - Fresh

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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

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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

j 200 .s ~ 175 N

c" 150 I -o 125 .r;

I j 100 Ill Cl.

~ 75 .!!!. 5 50 :;::; 'in ~ 25

0

I

1 -I

r-::-I I I I

I

f

I I I I I I I Axial Offset Window Power Dependent lnsertio n Limits

. AL[ l !- - -- ·-1-L ------- __ i ___ --

NOT ALLOWED --Group 1J - Group 2

(-040, 25) (0.25, 25)

-

r 1 r 1 0 r r , 0 10 20 30 40 50 60 70 80 90 100 -0.5 -0.4 -0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 0.4 0.5

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Power Level (% rated) Axial Offset Fraction ~--- -- -- --- -- --- -- ---------~

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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 - Acceptability of design basis is demonstrated in DCA Chapter 15 analyses

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4.4 - Thermal and Hydraulic Design

Parameter

Critical heat flux

Fuel temperature

Reactor core coolant flow

Hydrodynamic stability

Basis

95/95 criteria that hot rod avoids boiling transition

Fuel centerline temperature remains below melting limit

Primary system flow remains within ranges assumed in the safety analysis

Normal operation and AOOs do not lead to instability

Lim it / Protection

SL2.1.1.1

SL2.1.1.2

TS 3.4.1

Module protection system analytical limits prevent loss of subcooling and instability

Acceptance criteria ensure GDC 10 and 12 compliance

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4.4 - NuScale Design Fuel and Core Conditions

Parameter

Avg. Fuel Rod Linear Heat Generation Rate

Peak Fuel Temperature (cycle beginning)

Core Inlet Mass Flux

Core Inlet Subcooling

Core Exit Subcooling

Core Exit Void Fraction

Hot Channel Exit Thermodynamic Quality

NuScale Design

2.5 kW/ft

1620 °F

0.4 Mlb/hr-ft2

135 °F

25 °F

0.0

0.05

Nominal core average exit conditions similar to traditional PWRs

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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 Subchannel Analysis

Methodology NuScale Power Critical Heat

Flux Correlations

TR-0716-50350 [in review] Rod Ejection Accident

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

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Methodology Software

TR-0516-49417 {in review] PIM


Implementation

15.9

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4.4 - CHF Limit and Margin 'Stack-up'

Normal Operation Ranges MCHFR

> 3.0

(Applied Methodology Input Biases & Uncertainties)

CHF Analysis Limit 1.284

I\

Margin for transients

' I

1' Margin for CHF penalties (rod bow, F5)

95/95 CHF Safety Limit 1.21

CHF Failure 1.00

l t

CHF correlation uncertainties and biases

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4.4 - Pressure vs. Temperature Operation M_ a,_ 2100

High Pressure Analytical, Limit (2000 psia ) 2000 · • .... · · - ----------------------------------·

-n:, "iii a. -(IJ

""

1900

:::J 1800 11'1 11'1 (IJ

"" Q.

"" (IJ

-~ 1700 :::J 11'1 11'1 (IJ

"" Q.

1600

~ .3 ~ (]J a. E (]J

I-CU u

·.;:; ·c u E ::i E C

~

H\gh Pressure Operating Limit (1920 psia)

; ~ I [ U:-

Norllflal Operating Pressure (1850!psia) ~ I ! ~ er - - - - - !T1 - • -t ~ - - - - •

! <( I ~

i T Cold I- I T Avg j := T Hot

Low Nonnal Pcessuce (1780 psia) 1 L - -1 · 1w '"""" Analytical

·~-- L1m1t (1720 psi aJ

' ' i Low Low Pressure Analytica \ Limit (1600 psia) Subcooled Margin 5°F

rn C <I:

~ ::i .., ~ ···· -··or · ·· a. E (]J

I-

Saturation Curve

1500 L ----- --i- -- ___ -+-- ______ ---t-------~~-----------i

400 450 500 550 600 650

Reactor Coolant System Temperature (°F)

• Module protection system (Ch. 7, red ) • Technical specification LCOs (Ch . 16, blue)

CHF and fuel melting SAFDLs precluded for normal operation & DBEs 25

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4.5.1 - Control Rod Drive System Materials

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CROM SIJ'PORT FRAME

- TYPICAL CONTROL ROO ORNE (CRO) SHAFT

PRESSURIZER

----- TYPICAL CRO SliAFT SUPPORT

TYPICAL CONTROL ROil ASSEM81. Y GUIDE TUIIE

TYPICAL F\.El A.SSEMBL Y

• All pressure boundary materials are 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, with corresponding weld materials

• Non-pressure boundary materials

- Austenitic & martensitic stainless steels, nickelbase materials

- Cobalt-based materials are used in a very small portion where alternate material will not perform satisfactorily.

Components and materials are consistent with those for existing, proven designs

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4.5.2 - Reactor Internals and Core Support Structure Materials

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• 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

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4.6.1 Description of the Control Rod Drive System

MAST ASSEMBLY

SENSOR COILS

REMOTE DISCONNECT COIL

LIFT COIL

MOVEABLE COIL

STATIONARY COIL

Control Rod Drive Mechanisms (CROM)

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

SENSOR COIL • HOUSING (OUTSIDE THE PRESSURE BOUNDRY)

Additional remote disconnect mechanism (ROM) coil and latch are provided on top of the typical configuration of three coils.

LATCH HOUSING

- DRIVE COILS

• Components external to primary pressure boundary are designed for operation in an evacuated containment, but not required to operate during ECCS blowdown.

LATCH MECHANISM , •

• Components internal to primary pressure DRIVE COIL HOUSING (OUTSIDE THE PRESSURE BOUNDRY)

boundary (drive shaft & latch mechanism) are exposed to steam & non-condensable gases (N2, H2) on top of the pressurizer.

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' ,.,., .

• r .. ., •

ISOMETRIC VIEW DRIVE COIL ASSEMBLY

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ELECTROMAGNETIC DRIVE COIL

COOLING COILS

OUTERDRNE COIL HOUSING

II

ISOMETRIC VIEW WITH OUTER DRIVE COIL HOUSINGS NOT SHOWN

• Components external to primary pressure boundary are non-safety related (B2) and support SCRAM function by de-energizing from RTB.

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

• Disassembly of electrical components from the pressure boundary by upward retraction (slip-fit) after disconnection of cables and cooling water hoses.

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CONNECTED AND LOCKED POSITION

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CONTROL ROD DRIVE SHAFT

DISCONNECT ROD

FINGERS

PLUG

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

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

DISCONNECTED AND RELEASED POSITION

CONTROL ROD ASSEMBLY HUB

• To release, the Disconnect Rod is withdrawn allowing the shaft fingers to return to the retracted position

• The drive shaft can now be withdrawn and separated from the CRA

FINAL DISCONNECTED RESTING POSITION

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COL Item 4.2-1

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Chapter 4 - COL Items

• 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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• AO - Axial Offset

• AOO - Anticipated Operational Occurrences

• ASME - American Society of Mechanical Engineers

• BOL - Beginning of Life

• CHF - Critical Heat Flux

• COL - Combined License

• CRA - Control Rod Assembly

• CRDM - Control Rod Drive Mechanism

• DBE - Design Basis Event

• DCA - Design Certification Application

Acronyms

• ECCS - Emergency Core Cooling System

• EOL - End of Life

• EPRI - Electric Power Research Institute

• FIV - Flow Induced Vibration

• FSAR - Final Safety Analysis Report

• H/D - Height over Diameter

• HMP - High Mechanical Performance (Spacer Grid)

• HTP - High Thermal Performance (Spacer Grid)

• LOCA - Loss of Coolant Accident

• LTSD - Long Term Shutdown

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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

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Portland Office 6650 SW Redwood 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, MO 20852 301 . 770.0472

Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980. 349. 4804

Richland Office 1933 Jadwin Ave. , Suite 130 Richland, WA 99354 541 . 360. 0500

Arlington Office 2300 Clarendon Blvd. , Suite 1110 Arlington, VA 22201

London Office 1st Floor Portland House Bressenden Place London SW1 E 5BH United Kingdom +44 (0) 2079 321700

http://www. nuscalepower com '!I Twitter: @NuScale_Power

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Documents

Enclosure: 'ACRS Presentation: Chapter 4, Reactor Overview ... · L0-0419-65169 Enclosure: "ACRS Presentation: Chapter 4, Reactor OveNiew," PM-0419-65096, Revision 0 NuScale Power,