NuScale ISI, IST, RVI, and Appendix JProgram Development

Randall Morrill, PE

Reactor Module Design Engineer

July 24, 2014

NuScale Nonproprietary

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PurposeThe purpose of this presentation is to provide anintroduction to NuScale's approach to addressing ISI andIST in relation to its unique design features.

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Overview" Address and outline steps for NuScale SSC and its FOAK

design in the areas of

-ISI

-IST

- RVI inspection

- Appendix J leak test considerations

* Provide a brief description of additional activities plannedfor these programs to support DCA application and theCOL application.

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Agenda" Applicable codes, standards, and regulations

" Examples of requirements in relation to the NuScaledesign

" Overall NuScale ISI, IST, RVI inspection, and Appendix Jleak testing development

* Inspection and test element evaluation and alternativeflow charts

* Summary

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Abbreviations

CNV

COL

cvcsDMA

ECCS

FMECA

FOAK

ISI

IST

NDE

NPS

PDI

PWR

RCPB

containment vessel

combined operating license

chemical and volume control system

degradation mechanism assessment

emergency core cooling system

failure modes, effects, and criticality analysis

first-of-a-kind

inservice inspection

inservice testing

nondestructive examination

nominal pipe size

performance demonstration initiative

pressurized-water reactor

reactor coolant pressure boundary

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AbbreviationsDefinition

RCS

RPV

RVI

RXM

SSC

UT

reactor coolant system

reactor pressure vessel

reactor vessel internals

reactor module

structures, systems, and components

ultrasonic testing

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NuScale Power Module Components

CIVs

CNV

RXM piping

RW-

RPV

Integrated PZRbaffle plate andsteam plenum

SG annularspace

RRV

RVI lower riserassembly

RVI core supportassembly

CIV = containment isolation valve

CNV = containment vessel

PZR = pressurizer

RPV = reactor pressure vessel

RRV = reactor recirculation valve

RVI = reactor vessel internals

RVV = reactor vent valve

RXM = reactor module

SG = steam generator

}}3(a)

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Considered Scope* 10 CFR 50.55a, "Codes and Standards"

* 10 CFR 50 Appendix J, "Primary Reactor Containment LeakageTesting for Water-Cooled Power Reactors"

* ASME Section Xl Division 1, "Rules for Inservice Inspection ofNuclear Power Plant Components"

* ASME O&M Code, "Operation and Maintenance of Nuclear PowerPlants"

* EPRI MRP-227-A, "Materials Reliability Program: Pressurized WaterReactor Internals Inspection and Evaluation Guidelines"

* NRC RIS 2012-08, Regulatory Issue Summary, "Developing InserviceTesting and Inservice Inspection Programs under 10 CFR Part 52,"Revision 1

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Current Codes, Standards, and Regulations

" Current codes, standards, and regulations may not bewell suited for application to NuScale's integratedpressurized water reactor design.

* ISI, IST, RVI inspections and Appendix J leak testingrequirements were developed for current light waterreactor designs and associated SSC.

" Application of these current requirements would

- be non-conservative in some instances

- be overly burdensome in other instances without an increase in

overall reactor and SSC safety or reliability

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Example: Dual Clad Vessel* Containment vessel and reactor pressure vessel are clad

on both sides, since all surfaces contact borated water.

- current ASME Xl MandatoryAppendix VIII Supplements do notaccount for this two-sided cladgeometry

- technology likely exists, but UT PDIdemonstrations will be necessary

- surface exams would interrogateonly nonstructural clad surfaceversus load carrying membrane

- Unique identification of eachindividual weld and inspectionelement }}3(b)

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Example: Visual, VT-2 Exam

* ASME Xl VT-2 examinations ofreactor coolant pressureboundary following refueling andperiodic system leakage tests

- RCPB internal to CNV

* access for performing direct visualVT-2 during startup is limited

}}3(a)

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Example: ASME Xl Containment Requirements

Current ASME Xl IWE containment requirements consistonly of visual examinations looking primarily for wallthinning and corrosion.

- NuScale containment is designed for high pressure accidentconditions and contains full penetration pressure boundary welds

{ 1}}3(a)

- during operation, the containment is under a vacuum

- NuScale containment is immersed in borated water in the pool

- NuScale containment is physically moved to accommodaterefueling and generally accessible

}}3(a)

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Example: Small Diameter Piping

* Current ASME XI requirements for Class 1 small diameterpiping system welds (e.g., <_ 4" NPS) are exempt fromexamination.- basis for exemption is CVCS and ECCS makeup capacity

NuScale has RCS makeup capability through the CVCS, but the ECCS

is based on volume control and does not require makeup

}}3(a) and (b)

- integrity of these small lines less important to overall safety of the

design

- exemption basis in current code will be evaluated with respect to

the NuScale design

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Overall NuScale ISI, IST, RVI Inspection, and Appendix JLeak Testing Development Plans

113(a)

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Degradation Mechanisms Assessment

A DMA will be performed

" to identify degradation mechanisms present in theNuScale design (material, stress, environment, etc.)

" for each SSC to provide insight into which inspectiontechnique(s) (e.g., visual, surface, or volumetric) areappropriate to assure safety and reliability

" to determine SSC failure likelihood

" to identify any unique DM that results from materials orconditions not found in existing transferable and relevantoperating experience

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Alternative Inspection and Testing Methods

Each SSC and its associated elements will be evaluated to determineaccessibility, inspectable geometry, exposure, etc., to determine ifinspections are feasible with inspection methods prescribed bycurrent codes and standards.

" For SSC, whose inspection geometries, access issues, ALARA, orsafety concerns for performing current inspection and testingrequirements, alternate inspection or monitoring techniques will bedeveloped.

" In some cases, inspection or testing techniques will need to bemodified or alternate techniques developed for the NuScale design.

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Alternative Inspection and Testing Methods

" Estimates for radiological exposures resulting from the conduct ofinspections and tests will be compared to 10 CFR 20 limits forrequired inspections, tests, and monitoring or alternative methods.

" General worker safety conditions such as confined spaces or heatstress will be evaluated.

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ISI Performance Demonstration Initiative Actions

" Any unqualified geometries requiring ultrasonic inspectionwill be qualified in accordance with ASME BPVC SectionXl, Mandatory Appendix VIII.

" Prior to submittal of the first COL application, NuScale willengage with the vendors and EPRI NDE Center toprovide samples for qualification of any inspectiongeometries outside of current PDI configurations.

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Inservice TestingOM Code - Operation and Maintenance of Nuclear Power Plants, currentrequirements

" ISTA- General Requirements

" ISTB - Inservice Testing of Pumps-Pre 2000 Plants

" ISTC - Inservice Testing of Valves

* ISTD - Inservice Examination and Testing of Dynamic Restraints(Snubbers)

" ISTE - Risk Informed Inservice Testing of Components

" ISTF - Inservice Testing of Pumps-Post 2000 Plants

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Inservice Testing" NuScale RXM-specific SSC relevant to OM Code

- multiple containment isolation valves

- two reactor vent valves and two RPV safety valves

- two reactor recirculation valves and {{ }}3(b)

" Other NuScale RXM-specific SSC relevant to OM Code

- no RCS or ECCS pumps for the RCPB*}}3(a)

• Alternate period and testing frequencies will be developed consistentwith OM IST protocol

}}3(a) and (b)

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Reactor Vessel Internal Inspections

EPRI MRP 227 is currently the PWR standard forreactor vessel internals inspections and augmentsASME Section XI, Div 1

- addresses OEM RVI geometries (i.e., B&W,Westinghouse, and CE internals-specificconfigurations) with some similarities to NuScale

}}3(a)

}}3(b)

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Inspection and Test Element Evaluation Flow Chart

}}3(a)

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Inspection and Test Element Alternatives Flow Chart

}}3(a)

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Summary for DCA Information" Where existing ASME and industry standards can be meaningfully employed for the

NuScale FOAK design, those codes and standards, as endorsed or modified by the

NRC under 10 CFR 50.55a, will be utilized.

" Where existing codes and industry standards are not appropriate to address the FOAK

design features, NuScale will provide proposed alternatives that

- provide an acceptable level of quality and safety

- demonstrate that compliance with the specified current requirements would result in hardship orunusual difficulty without a compensating increase in the level of quality and safety

* Along with any alternatives proposed at the time of the DCA, NuScale will provide theNRC staff with sufficient information to ensure that an adequate foundation for theplant-specific operational programs can be developed by COL applicants. This may

include technically supported alternatives provided under the provisions of 10 CFR

50.55a (3)(ii)

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