B0JW babcock wilcox nuclear energymethodology development and application process (EMDAP) of RG...

B0JW babcock & wilcox nuclear energy

Safety Analysis Evaluation MethodologyRequirements for the B&W mPowerTM Reactor

Redacted VersionMPWR-TECR-005013(NP)

Revision 000November 2011

z" mPower TM

I a progressive energy solution

B&W mPowerTM Reactor ProgramBabcock & Wilcox Nuclear Energy, Inc.

109 Ramsey PlaceLynchburg, VA 24501

Open Items Safety Proprietary Files Attached Total No. PagesClassification

N N N N 153

© 2011 BABCOCK & WILCOX NUCLEAR ENERGY, INC. ALL RIGHTS RESERVED.This document is the property of Babcock & Wilcox Nuclear Energy, Inc. (B&W NE)

Babcock & Wilcox Nuclear Energy, Inc.a Babcock & Wilcox company

Document No. Title Rev. No.

MPWR-TECR- Redacted Version 000005013(N P)

SINIATrURES

Pm~red y:. Robed P. Martin, Ph.D. ?Meamos Develpmut Lead. -

Sfety Analyal Signdure O

PrprdJan $. Muraft.alcy ,A0 AI~L -Advioy Engies.

s ,•S•.yAnalyssS ure Sia..s DO

, By., Eric S• •i,.ms s/1m .. (litI

_________SaftyAnalysis signature Date

Approvd O. MidcaIT. Chuceleson ,(I19I

_______ Design Engierming Signawur Date

ApodO. Chet PosksPi .-Manager,

_____ _____Licening SgaueDt

©2011 Babcock & Wilcox Nuclear Energy, Inc. All rights reserved.

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RECORD OF REVISION

Revision No. Date Preparer Description of Changes

000 October 2011 R. P. Martin Initial Release

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4 F

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MPWR-TECR- Redacted Version 000005013(NP)

Executive Summary

The Babcock & Wilcox (B&W) mPowerTM reactor is an integral pressurized water reactor designwith the reactor, steam generator, and pressurizer all located in a single pressure vessel. Thedesign has no large cold or hot leg piping, which eliminates the potential for large break loss-of-coolant accidents. Most transients and accidents described in NUREG-0800 are either identicalor very similar to the commercial operating PWRs or to advanced PWRs with passive safetysystems. No new thermal-hydraulic or neutronics phenomena have been identified for thisdesign. Because the B&W mPower reactor employs conventional balance-of-plant systems,transients initiated by failures in these systems are very similar to other PWR designs.

This technical report describes the functional and infrastructural requirements and theassociated basis for performing design basis safety analysis supporting the development andcertification of the B&W mPower reactor. It has been prepared to facilitate the decision-makingprocess in the resolution of safety analysis issues addressing compliance with applicablefederal regulations and guidance. Nearly three dozen general requirements (with many moresubheading requirements) related to the design, development, deployment of safety analysismethods have been identified. Complete evaluation methodologies will be captured insupplements to this technical report. Refinement of requirements will continue as newinformation is gained through B&W's engineering and testing program.

The approach adopted for deriving these requirements examines the predominant regulatoryguides (RG) related to safety analysis: RG 1.203, RG 1.183, and RG 1.77. This technical reporthas been formatted in a manner recognizing that the first step in the preparation of anyevaluation methodology is to define requirements. Consistent with Element 1 of the evaluationmethodology development and application process (EMDAP) of RG 1.203, detail is providedherein on analysis purpose (Section 2), plant design (Section 3), transient classes and figure(s)-of-merit (Section 4), characterization of the modeling domain (Section 5) and on thedevelopment of phenomena identification and ranking tables (Section 6). The generaladequacy of an evaluation methodology is measured against the completeness of uncertaintytreatments addressing both analytical and process measures. This relates to EMDAP Elements2 - 4 and is discussed in Section 7. A compilation of evaluation methodobgy requirementsdrawn from the previous sections is given in Section 8.


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Babcock& Wilcox Nuclear Energy, Inc.a Babcock & Wilcox company



TABLE OF CONTENTS

1 . INTRODUCTION ........................................................................................................................... 102. SAFETY ANALYSIS W ORKFLOW REQUIREMENTS .................................................................. 12

2.1 Organizational Requirements and Defense-in-Depth Measures ........................................... 122.2 Quality Management ............................................................................................................ 172.3 Regulatory Requirements and Expectations ......................................................................... 18

3. B&W MPOW ER REACTOR AND SYSTEMS OVERVIEW ............................................................ 263.1 Reactor Coolant System ...................................................................................................... 263.2 Emergency Core Cooling System ......................................................................................... 283.3 Non-Safety Decay Heat Removal ......................................................................................... 313.4 Main Steam System ............................................................................................................. 323.5 Feedwater System ............................................................................................................... 353.6 Instrumentation and Control Systems ................................................................................... 37

4. IDENTIFICATION OF TRANSIENT CLASSES .............................................................................. 414.1 Description of Chapter 15 Events to be Analyzed for the B&W mPower Reactor .................. 474.2 Summary of Chapter 6 Containment Response Applicable to B&W m Power Reactor .......... 844.3 Combustible Gas Control in Containment ............................................................................. 92

5. CHARACTERIZED MODELING DOMAIN ...................................................................................... 955.1 Hierarchical Decomposition ................................................................................................. 955.2 Essential Features for Analytical Modeling ......................................................................... 1015.3 Computer Code Selection for B&W mPower Accident Analysis .......................................... 103

6. PHENOMENA IDENTIFICATION AND RANKING TABLE ........................................................... 1136.1 PIRT Objectives ................................................................................................................. 1136.2 Database Supporting PIRT Development ........................................................................... 1146.3 B&W mPower SBLOCA PIRT ............................................................................................ 1236.4 Figures of Merit .................................................................................................................. 1256.5 Identification of Plausible Phenomena ................................................................................ 1266.6 Phenomena Relative Importance and State of Knowledge ................................................. 1276.7 PIRT Results and Conclusions ........................................................................................... 128

7. MANAGING UNCERTAINTY IN EMDAP ..................................................................................... 1337.1 Managing Compliance ....................................................................................................... 1337.2 Managing Knowledge and Expertise .................................................................................. 1337.3 Managing Data Applicability ............................................................................................... 1347.4 Managing Code/Model Development .................................................................................. 1357.5 Managing Code Verification and Validation ........................................................................ 1357.6 Managing Uncertainty Quantification and Convolution ........................................................ 1367.7 Managing Human Reliability and Consequence of Failure .................................................. 138

8. REQUIREMENTS SUMMARY .................................................................................................... 1399. CONCLUSIONS ........................................................................................................................... 14710. REFERENCES ............................................................................................................................ 148

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List of Figures

Figure 1. Plant E ngineering D isciplines ............................................................................................... 15Figure 2. E M D A P Flow D iagram ............................................................................................................. 22Figure 3. RPV Upper and Lower Regions .......................................................................................... 28Figure 4. S im plified EC C S diagram .................................................................................................... 29Figure 5. Main Steam and Turbine Generator Systems ..................................................................... 34Figure 6. Feedw ater S ystem ................................................................................................................... 36Figure 7. Hierarchical RCS Response Model for a Generic B&W mPower Plant Event ....................... 97Figure 8. B&W mPower Plant Containment Heat and Mass Transfer Mechanisms ............................. 99Figure 9. Hierarchical Containment Response Model for a Generic B&W mPower Plant Event ............. 100Figure 10. Analysis Code Suite Supporting the B&W mPower Reactor ................................................. 103Figure 11. Contrast Between Deterministic and Probabilistic Convolution of Uncertainty ....................... 137

List of Tables

Table 1. General RCS Specifications ................................................................................................. 27T a ble 2 . R P S sensor inputs ..................................................................................................................... 39Table 3. Applicability of NUREG-0800 Transients to B&W mPower Reactor ....................................... 42Table 4. Transients that are Specific to B&W mPower Reactor .......................................................... 45Table 5. B&W mPower Applicability to SRP 6.2.1 Expectations ............................................................ 86Table 6. Small Break LOCA PIRT for the B&W 177 Design .................................................................. 116Table 7. Small Break LOCA PIRT for the MASLWR Design .................................................................. 117Table 8. PIRT for PWR Non-LOCA Transients ...................................................................................... 119Table 9. PIRT for Containment Conditions Following an LOCA .............................................................. 121Table 10. SSCs Important to Each Phase of the Design Basis Transient (Bolded Items Are Dominant) 124Table 11. Figures of Merit for B&W mPower Reactor LOCA Design Basis PIRT ................................... 126Table 12. Relative Importance Ranking Scale ....................................................................................... 127Table 13. State of Know ledge S cale ..................................................................................................... 127Table 14. G eneric R isk Perspective ...................................................................................................... 128Table 15. B&W mPower Reactor PIRT Risk Perspective Results .......................................................... 129Table 16. Summary of Organizational Requirements ............................................................................ 139Table 17. Summary of Regulatory Requirements for Safety Analysis .................................................... 140Table 18. Documentation Requirements based on RG 1.203 ............................................................... 143Table 19. Computing and Software Requirements ................................................................................ 145


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List of AcronymsACRS Advisory Committee on Reactor SafeguardsADV Automatic Depressurization Valves

ALARA As Low As Reasonably AchievableAOO Anticipated Operational OccurrenceASME American Society of Mechanical Engineers

ATWS Anticipated Transient Without ScramBEPU Best-Estimate-Plus-Uncertainty

BOP Balance of PlantBPR Burnable Poison RodBTP Branch Technical PositionB&W Babcock and WilcoxB&W NE Babcock and Wilcox Nuclear EnergyBWR Boiling Water Reactor

CFR Code of Federal RegulationsCGCS Combustible Gas Control SystemCHF Critical Heat Flux

COL Combined LicenseCRA Control Rod AssembliesCSAU Code Scaling Applicability and UncertaintyCVCS Chemical and Volume Control SystemDCD Design Control DocumentDNBR Departure from Nucleate Boiling RatioDOE Department of EnergyEBT Emergency Boron Tank

EC Emergency Core Cooling CondenserECCS Emergency Core Cooling SystemEMDAP Evaluation Methodology Development and Assessment Process

EPRI Electric Power Research InstituteEQ Equipment QualificationESFAS Engineered Safeguards Features Actuation System

FOM Figure-of-Merit

FSAR Final Safety Analysis ReportFWS Feedwater System

FWP Feedwater Pumps

FWH Feedwater Heater

GDC General Design Criteria

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IAC Interim Acceptance CriteriaI&C Instrumentation and Control

INL Idaho National LaboratoryIST Integrated Systems Test (B&W mPower-specific)

IT Information TechnologyLCO Limiting Condition for OperationLOCA Loss of Coolant Accident

LOEL Loss of External LoadLOFT Loss-of-Fluid Test

LWR Light Water ReactorMCR Main Control RoommPCS mPower Plant Control SystemMSIV Main Steam Isolation Valve

MSS Main Steam SystemMS/R Moisture Separator/Reheaters

MSRV Main Steam Relief ValveMSSV Main Steam Safety ValveNEA Nuclear Energy Agency (part of the OECD)

NPP Nuclear Power PlantOECD Organization for Economic Cooperation and DevelopmentU.S. NRC Nuclear Regulatory Commission

PCL Plant Control Layer

PIRT Phenomena Identification and Ranking TablePML Plant Monitoring Layer

PORV Pilot-Operated Relief Valve

PPL Plant Protection LayerPRA Probabilistic Risk AssessmentPSAR Preliminary Safety Analysis ReportPWR Pressurized Water Reactor

QA Quality AssuranceRCIPS Reactor Coolant Inventory and Purification SystemRIA Reactivity Insertion Accidents

RCS Reactor Coolant SystemRG Regulatory Guide

RPV Reactor Pressure VesselRPS Reactor Protection System


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RWST

SAFDL

SBLOCA

SECY

SGTRSLB

SOK

SRDCS

SRP

SSC

TBVTCV

TEDE

TMITSV

UHS

URD

Refueling Water Storage Tank

Specified Acceptable Fuel Design Limit

Small Break Loss of Coolant Accident

Office of the Secretary of the Commission

Steam Generator Tube Rupture

Steamline Break

State of Knowledge

Safety-Related Distributed Control System

Standard Review Plan

Structures, Systems, and Components

Turbine Bypass ValveTurbine Control Valve

Total Effective Dose Equivalent

Three Mile Island

Turbine Stop Valve

Ultimate Heat Sink

Utility Requirements Document





1. INTRODUCTION

The evaluation of nuclear power plant (NPP) safety issues requires analyses of the plant'sresponses to possible equipment failures or malfunctions. Such analyses inform the licensingprocess by validating the design envelope defined by the limiting conditions for operation (LCO),limiting safety system settings, and design specifications for safety-related structures, systems,and components (SSCs) employed to protect public health and safety. The purpose of thisreport is to present the principal functional and infrastructural requirements for the transient andaccident evaluation methodologies employed in the preparation of reactor coolant system (RCS)and containment response analysis supporting the licensing activities for the Babcock & Wilcox(B&W) mPowerTM reactor.

Like a conventional light water reactor (LWR), the B&W mPower reactor provides several leaktight barriers against the release of radioactivity to the environment and includes a reliableshutdown capability in response to design basis events. The SSCs that are important to safetyhave been engineered to withstand the anticipated loads from dynamic effects associated withnormal plant operation (including maintenance and testing), transients and accidents. Per theU.S. Code of Federal Regulations (CFR) Title 10 Part 50 (Reference 1) (refer, in particular, to10 CFR 50.2 in defining "Safety-related structures"), safety analyses are performed to ensurethat the plant's SSCs relied upon to remain functional during and following a design basis eventensure:

" the integrity of the reactor coolant pressure boundary

" the plant's capability to shut down the reactor and maintain it in a safe shutdowncondition

" the plant's capability to prevent or mitigate the consequences of accidents which couldresult in potential offsite exposures comparable to the applicable guideline exposures setforth in 10 CFR 50.34(a)(1) or 10 CFR 100.11, as applicable

For the certification of NPP designs, U.S. Nuclear Regulatory Commission (U.S. NRC)regulations require that the safety characteristics of these designs be demonstrated throughapproved evaluation methodologies, that is, a calculational framework for evaluating thebehavior of the reactor system following possible equipment failures or malfunctions. Evaluationmethodologies include one or more computer programs and all other information necessary tocharacterize a particular event scenario, such as mathematical models, assumptions embeddedin computer models and correlations, and procedures for treating the program input and outputinformation. The evaluation methodologies and associated computer codes are to be qualifiedagainst the expected system behavior as observed from applicable experimental programs.

Resolving technical issues through modeling and simulation is an essential component ofmodern engineering. Engineers have a relatively large suite of suitable analytical computercodes available to support complex design projects such as the B&W mPower reactor.Engineering, however, relies on more than simply "running a computer code." To draw credibleconclusions from analysis relies on a strict adherence to a verifiable process for which legitimatedesign inputs and assumptions are adopted that address the analytical uncertainties related to


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compliance, expertise, data applicability, code and model development, code verification andvalidation, uncertainty quantification, and human reliability and the consequence of failure.Technical, organizational, and interpersonal interfaces further complicate the migration of datafrom a design specification to a meaningful metric for decision-makers.

The objective of this report is to highlight the functional and infrastructural requirements anddescribe the implementation plan for the development, application, and maintenance of thesafety analysis methods supporting the B&W mPower reactor. It addresses requirementsassociated with the resolution of technical issues through analysis in compliance with applicableFederal regulations and guidance.

This technical report has been formatted in a manner recognizing that the first element in thepreparation of any evaluation methodology is to define requirements. Emulating Element 1 ofthe Evaluation Methodology Development and Assessment Process (EMDAP) of RegulatoryGuide 1.203 (Reference 2), detail is provided herein on analysis purpose (Section 2), plantdesign (Section 3), transient classes and figure(s)-of-merit (Section 4), characterization of themodeling domain (Section 5), and on the development of phenomena identification and rankingtables (PIRT) (Section 6). The general adequacy of an evaluation methodology is measuredagainst the completeness of uncertainty treatments addressing both analytical and processmeasures. This relates to EMDAP Elements 2 - 4 and is discussed in Section 7. Section 8includes a compilation of evaluation methodology requirements drawn from the previoussections.





2. SAFETY ANALYSIS WORKFLOW REQUIREMENTS

This section establishes the purpose of safety analysis through the identification oforganizational, quality program, and regulatory requirements. These requirements address abroad range of both functional and infrastructural characteristics. Functional requirements arethose that address core project goals, while infrastructural requirements describe thoseperformance and system characteristics essential to the actual process. Early identification ofthese requirements feeds directly into all downstream activities.

Fundamentally, safety analysis demonstrates and validates the B&W mPower reactor safetymission against the breadth of applicable regulations and regulatory guidance. The first step inthis process is to identify those requirements essential to this goal. The identification ofrequirements begins with characterizing the organizational role of safety analysis through thetechnical and administrative tools, processes, and interfaces that facilitate project success.

Regulations and the companion regulatory guidance ultimately establish the engineering andanalytical framework for safety analysis. The regulatory process provides a top-downexpression of rules and expectations, open to interpretation, but in practice defined byprecedents. The initial regulatory interface begins with the quality program, a corporateinterpretation of the regulatory language appearing in 10 CFR 50 Appendix B. Only after aquality program has been established can the activities commonly recognized as safety analysisbegin. Regulations most notably establish acceptance criteria; but, they also include guidancedirected at the mechanics of actually performing analysis.

2.1 Orcqanizational Requirements and Defense-in-Depth Measures

Safety analysis is a unique technical competency supporting the engineering projectorganization. Foremost, safety analysis relies on computation to translate the design conceptinto safety measures. The identification of organizational requirements begins by clarifying theinterfaces between technology and the project organization. These are dependent upon thekind of technology provided by safety analysis to the project. Detailing this role and howinformation is received, used, maintained, and delivered by staff both internal and external tothe project further characterizes the organizational requirements unique to safety analysis.

Like similar NPP design and deployment projects, the B&W mPower reactor engineering projectis structured around the defense-in-depth philosophy adopted in the nuclear industry. Therequirements established for advanced LWRs are very stringent with regard to the considerationof defense-in-depth and the radiological impact on the public. To demonstrate the safety of theplant, the following basic objectives should be fulfilled (Reference 3):

i) Prevention of abnormal operation and failures;

ii) Control of abnormal operation and detection of failures;

iii) Control of accidents within the design basis;


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iv) Control of severe plant conditions, including the prevention of accident progressionand mitigation of the consequence of severe accidents; and

v) Mitigation of radiological consequences of significant releases of radioactivematerials.

The safety goal of small, modular reactors like the B&W mPower reactor is to enhance thesesafety objectives by:

* Exhaustive review of initiating events used to confirm the adequacy of the safetyprovisions

* Minimizing or, if possible, eliminating the occurrence of complex phenomena and "cliff-edge" effects during normal operation, anticipated operational occurrences (AOOs), andaccidents

* Providing long grace periods when operator action is necessary

* Simplifying the whole operations architecture by

o The application of "As Low As Reasonably Achievable" (ALARA) principles forthe protection of workers against the radiation exposure in particular whenimplementing the necessary corrective actions in accidental conditions,

o Minimizing and mitigating of hazards other than radiological ones (e.g., chemicalhazards),

o Minimizing the production of wastes and effluents and developingaccommodation for their lifecycle, and

o Preventing by design possible types of malevolence and proliferation, andminimizing their potential consequences.

2.1.1 Engineering Interfaces with Safety Analysis

Figure 1 presents an illustration of the engineering project, emphasizing the coordination ofsafety analysis (highlighted in yellow) with the various design disciplines arranged (and shaded)in relation to their defense-in-depth role:

* Reactor coolant system and safety analysis - characterizes the RCS pressure boundary,a physical fission product barrier, and defines the design envelope for thermal-hydraulicconditions during normal operation and design basis events.

* Fuel and core design - characterizes the configuration of fuel and the material andstructural requirements for the fuel rod cladding, a physical fission product barrier.


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MPWR-TECR- Redacted Version 000

" Fuel burnup and core performance - characterizes the design envelope for fuelmaterials and reactor physics during normal operation and design basis events.

* Instrumentation and controls - characterizes the protection and safeguards systems,defined by several monitored plant process inputs, translated as binary signals thatresult in the actuation of systems serving the critical safety functions of reactivity control,RCS pressure and inventory control, RCS heat removal, containment environment andisolation, containment integrity, and radiation and radioactive effluent control.

* Containment performance - characterizes the design envelope for the containment, aphysical fission product barrier.

* Structural and loss-of-coolant accident (LOCA) loads - characterizes the designenvelope of structural forces on the containment and critical load-bearing and safetyequipment.

* Radiological, severe accidents, and probabilistic risk assessment (PRA) - characterizesthe design envelope of the radiological source term and the risk and consequencesrelated to the release and transport of radionuclides from the fuel into the environment.

" Technical specifications and equipment qualification/survivability - captures the designenvelope requirements for SSCs supporting critical safety functions from otherengineering disciplines to support administrative coordination and controls, includingprocurement.

* Accident management, plant simulator, and training - characterizes programs, plans,training, exercises, and resources necessary to prepare personnel to operate, maintainand effectively react to off-normal situations, including those arising from terrorism ornatural events. It includes emergency preparedness, whose objective is to ensure thatNPP operators can implement measures to protect public health and safety in the eventof a radiological emergency.

* Auxiliary systems - characterizes new and spent fuel handling and storage, watersystems, process auxiliaries, air handling systems, fire protection and other systems thatsupport the overall mission of plant operation.

Safety analysis must readily apply and supply data from different and diverse project teams withsimilar computational requirements without constraints to their physical location. As illustrated inFigure 1, analysis related to RCS performance effectively interfaces with all aspects of NPPdesign, including fuel, core, and containment performance, instrumentation and controls (I&C),equipment qualification (EQ), severe accidents, structural loads, dose assessments,radionuclide transport, emergency response, simulator design, and training programs.


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TechnicalSpecifications

I PRA (PSA) J Performance IAnalysis

AuxiliarySystems)

Figure 1. Plant Engineering Disciplines

2.1.2 Safety Analysis Organizational Considerations

The traditional mission of NPP safety analysis is to confirm the adequacy of a design undernormal operation, AOOs, and accidents - while supporting the overall design control processvia interfaces with other engineering disciplines. Safety analysis for the B&W mPower reactor isno different. Its key business processes rely extensively on a specialized and demanding use ofscientific computing power.

The volume of analysis and industry quality standards demand a unique information lifecycleand architecture beginning from design inputs to technical reports available to the end-userdecision-makers. As such, safety analysis employs data stores with varied forms subject to aregulated quality control program. These data stores can be quite large and are often written ina machine-dependent format. Technical and administrative decision-makers will utilize safetyanalysis information in both raw and editorialized formats, prepared for various audiences. Assuch, the ability to reduce, filter, and migrate data between formats while abiding by a qualitycontrol program is necessary to support end product objectives.

The nature of phenomena examined by safety analysts is diverse and subject to uncertaintiesinherent in the modeling process. Consequently, analysis relies on a hybrid of physics- and


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evidence-based computer codes for simulation of NPP processes, supplemented by varioussupport tools providing task automation, pre-/post- processing, and data transfer/migrationinterfaces to peripheral activities. All such simulation tools are considered "open" to continuousdevelopment subject to a qualification process that emphasizes verification and validation.

While the simulation tools are a central component to safety analysis, it is only one piece of ananalytical evaluation methodology that must address the broader uncertainties associated withthe application of analysis for resolving safety issues (see Section 7). The integrity ofproduction safety analysis results must be traceable from documented statements of resultsback to the original design inputs through unique references and file checksum properties. Allfiles must permanently reside in a data storage management system satisfying quality controlregulations. Calculation and evaluation methodobgy qualification results must also be readilyaccessible and reproducible on demand.

In general, modern computing environments have evolved towards a relatively small set ofaccepted standards. For the development of NPP safety analysis from qualified evaluationmethodologies, the technology architecture required by engineers begins with a highperformance computing system, supported by a stable operating system, such as UNIX. Mostcomputers can operate under UNIX or a variant such as LINUX; however, variability in computerplatform architecture results in simulation variability. To maintain fidelity with the qualificationprogram for computer codes used by safety analysis, a single platform must be selected beforeanalysis advances to production.

For long-term preservation of safety analysis competency and capability, the evaluationmethodology and computing infrastructure needs to be sufficiently flexible to allow forinnovations that improve safety analyst performance and/or end-product quality. Suchinnovations might include new computing hardware, a new computer operating system, or amodification to the evaluation methodology itself. Since plant design is part of an evaluationmethodology, modifications to the plant design or operating conditions also fall into thiscategory. Central to this requirement is the expectation of occasional requalification ofevaluation methodology elements. Minor changes can be addressed through qualitativedisposition or qualitative analysis. Major changes may require a full-scope qualificationprogram. As such, decisions regarding the computing environment should employ practicescompliant with recognized standards and broad acceptance to reduce an evaluationmethodology exposure to situations requiring full-scope requalification.

Administratively, scientific computing requires unique information technology (IT) skills, muchdifferent than those associated with maintaining normal office personal computers and relatedsoftware. Employing dedicated IT professionals in support of scientific computing provides thereliability necessary to ensure uninterrupted computation services.


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2.2 Quality Managemen

A central theme of safety analysis is the attention to quality and the a posteriori demonstrationthereof. Quality priorities and success criteria for a project should be established early so thatthe corresponding project management plan develops in a manner that integrates relatedobjectives. While corporate and regulatory policy must be followed, additional project-specificcontrol measures and/or unique policy interpretations may be necessary to clarify proceduraldetail or to facilitate audits. This can include unique manuals and guidelines that describe thepreparation of computer code input and how to process software output.

Quality management begins with recognizing the corporate and regulatory policies that must befollowed. Development, assessment, maintenance and application of an evaluationmethodology are activities related to the requirements of 10 CFR 50 Appendix B, "QualityAssurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants". Several sectionsof 10 CIFIR 50 Appendix B cite requirements relevant to the development and application ofevaluation methodologies. These include:

" Section III:

- Appendix B design control measures specifically apply to reactor physics, thermal,hydraulic and accident analyses. Design control measures for evaluationmethodologies have been further interpreted to include configuration controlpractices to protect computer code and evaluation methodology integrity and allowtraceability of both the code version and plant input models.

- The design control measures shall provide for verifying or checking the adequacy ofdesign, such as by the performance of design reviews, by the use of alternate orsimplified analytical methods, or by the performance of a suitable testing program.

- Design changes are subject to appropriate design control measures.

" Section IV and VI I: Contractor supplied products and services used to support safetyanalysis must also comply with Appendix B design control measures.

" Section V: Documented instructions, e.g., user guides, are required.

" Sections VI and XVII: The appropriate level of document control and quality assurance(QA) records is identified on documentation.

" Section XVI: Errors must be promptly identified and corrected, and corrective actionsmust be taken to preclude repetition. All this must be documented. In some cases 10CFR 21, addressing defects and non-compliances may apply.

The core set of requirements for QA of safety analysis software appears in American Society ofMechanical Engineers (ASME) NQA-1 -1994 (Reference 4). Subpart 2.7 provides requirementsfor the acquisition, development, operation, maintenance, and retirement of software.Implementation of these requirements follows a prescriptive set of instructions. Subpart 2.7notes that the "appropriate requirements of this Subpart shall be implemented through thepolicies, procedures, plans, specifications, or work practices, etc., that provide the framework forsoftware engineering activities". Thus, the safety analysis software owners/vendors should





follow a documented procedural basis to develop their respective software. The quality programat B&W Nuclear Energy (B&W NE), the organization within B&W responsible for engineering theB&W mPower reactor, cites these relevant regulatory requirements and expectations(Reference 5).

2.3 Regiulatory Requirements and Expectations

B&W NE's safety analysis evaluation methodologies will adhere to the applicable parts of theU.S. NRC regulations and regulatory guidelines, Branch Technical Positions (BTP), statementsfrom the Office of the Secretary of the Commission (SECY), etc. in developing safety analysiscontent for the B&W mPower reactor design control document (DCD). These B&W mPower-specific evaluation methodologies will employ several computer codes to predict relevant safetymeasures addressing criteria expressed in NUREG-0800, the Standard Review Plan (SRP),and Regulatory Guide 1.206 (References 6 and 7).

2.3.1 Governing Regulatory Documents

Federal regulations addressing the safety of NPPs draw upon several U.S. NRC policystatements, including:

* U.S. NRC Policy Statement, "Regulation of Advanced Nuclear Power Plants"(Reference 8)

" U.S. NRC Policy Statement, "Nuclear Power Plant Standardization" (Reference 9)

* U.S. NRC Policy Statement, "Safety Goals for the Operations of Nuclear Power Plants"(Reference 10)

These U.S. NRC Policy Statements set the regulatory objectives for new reactor designs.They build upon the principle that new NPP designs will provide enhanced safety marginsand utilize simplified, inherent, passive, or other innovative means to accomplish their safetyand security functions. Specifically, designs are expected to employ features to prevent lossof containment integrity and to maintain core cooling through reliable, long-term decay heatremoval systems.

The key U.S. NRC regulations related to the preparation and reporting of design basisanalyses are:

* 10 CFR 50.34(b), Final safety analysis report (refer also to 10 CFR 52.47)

* 10 CFR 50.36, Technical specifications

* 10 CFR 50.43, Additional standards and provisions affecting class 103 licenses andcertifications for commercial power

* 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-waternuclear power reactors

* 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants


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* 10 CFR 100, Reactor Site Criteria

These regulations implicitly define the scope of analytical evaluation methodologies in termsof the documents to be submitted to the U.S. NRC by applicants and licensees. Thisincludes the following requirements regarding applications for construction permits andlicenses to operate a facility:

" Engineering analysis reports per 10 CFR 50.34 evaluating the design and performanceof SSCs, and their adequacy to prevent and mitigate the consequences of accidents

" A description and analysis of the SSCs of the facility, with emphasis upon performancerequirements, the bases, with technical justification per 10 CFR 50.43 upon which suchrequirements have been established, and the evaluations required to show that safetyfunctions will be accomplished

* Analysis and evaluation of emergency core cooling system (ECCS) performancefollowing LOCAs prepared in accordance with the requirements of 10 CFR 50.46

" The technical specifications for the facility based on the safety analysis and prepared inaccordance with the requirements of 10 CFR 50.36

* For site evaluation a fission product release from the core should be assumed, theexpected demonstrable leak rate from the containment and the meteorologicalconditions pertinent to this site to derive an exclusion area, a low population zone andpopulation center distance per 10 CFR 100

The General Design Criteria (GDC) appearing in 10 CFR 50 Appendix A refers to specificNPP design requirements. For safety and containment response analysis, the morepertinent design criteria include:

* GDC 10 - Reactor design

* GDC 13 - Instrumentation and control

* GDC 15 - Reactor coolant system design

* GDC 16 - Containment design

* GDC 19 - Control room

" GDC 20 - Protection system functions

* GDC 26 - Reactivity control system redundancy and capability

* GDC 28 - Reactivity limits

* GDC 29 - Protection against anticipated operational occurrences

* GDC 34 - Residual heat removal

* GDC 35- Emergency core cooling

" GDC 38 - Containment heat removal


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" GDC 41 - Containment atmosphere cleanup

* GDC 44 - Cooling water

* GDC 50 - Containment design basis

The regulatory position on the analytical interpretation of regulations has evolved throughrevisions of the U.S. NRC's Standard Review Plan, beginning with the publication ofNUREG-75/087 (Reference 11) in 1975. Before that, individual review plans existed butthey were not published in a single authoritative document. NUREG-75/087 was revisedand reissued as NUREG-0800 in 1981. During the late 1980s and early 1990s, the ElectricPower Research Institute (EPRI) and the U.S. NRC pursued separate evaluations of plantrequirements for advanced light water reactors. EPRI released its Utility RequirementsDocument (URD) in 1992 (Reference 12) and the U.S. NRC updated the SRP in 1996. InJune 2007, the U.S. NRC issued Regulatory Guide 1.206, following a revision to the SRP inMarch 2007, to specifically address the issuance of combined construction and operationlicenses (COL) for NPPs.

The current status of open and resolved generic safety issues appears in NUREG-0933, "APrioritization of Generic Safety Issues" (Reference 13). Regulatory Guide (RG) 1.206advises that design applications should address how the design and analysis of applicableevents incorporate the resolution of medium- and high-priority GSIs identified in the versionof NUREG-0933 that is current 6 months before the application submittal date.

2.3.2 Additional Regulation and Guidance Documents

As a complement to the U.S. NRC's Standard Review Plan, several Regulatory Guides havebeen prepared to clarify the U.S. NRC's expectation for preliminary and final safety analysisreport (PSAR/FSAR) content. Collectively, the following documents express therequirement that evaluation methodologies are developed following a structured designcontrol process:

* RG 1.7, "Control of Combustible Gas Concentrations in Containment" (Reference 14)

* RG 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear PowerPlants (LWR Edition)" (Reference 15)

* RG 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident forPressurized Water Reactors" (Reference 16)

* RG 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors" (Reference 17)

* RG 1.203, "Transient and Accident Analysis Methods" (Reference 2)

* RG 1.216, "Containment Structural Integrity Evaluation for Internal Pressure Loadingsabove Design Basis Pressure" (Reference 18)

Among the RGs cited, RG 1.77 and RG 1.203 are notable in how ongoing research anddevelopment is influencing the development of evaluation methodologies and how they align


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with regulatory sections 10 CFR 50 Appendix K (deterministic LOCA analysis) and 10 CFR50.43 (verification and validation programs). The relevance of these documents plusadditional detail on regulations governing radiological analyses appears in the followingsubsections.

2.3.2.1 Regulatory Guideline 1.203

In 2005, the U.S. NRC published RG 1.203, describing the structured evaluation modeldevelopment and assessment process (EMDAP). With its introduction, greater responsibilityhas been placed on applicants for defining the technical basis of evaluation methodologies,rather than procedural compliance with the elements of 10 CFR 50 (e.g., Appendix K) or theSRP. The EMDAP is generally applicable to the development of analysis methods for thepurpose of evaluating safety issues related to NPP AQOs and accidents. EMDAP startsfrom the definition of the objectives, the functional requirements, and the identification ofimportant phenomena. Guided by these top-level priorities, code development andassessment follow, ultimately leading to the evaluation model adequacy decision. Figure 2presents a diagram of the EMDAP.

Cosmetically, an evaluation methodology may appear to simply provide a procedure for howto perform an analysis. The EMDAP considers that the basis for the procedural mechanicsaddresses the many facets of uncertainty management, including compliance, expertise,data applicability, code and model development, code verification and validation, uncertaintyquantification, and human reliability and the consequence of failure (see Section 7). EMDAPstarts from the definition of the objectives, the functional and infrastructural requirements,and the identification of important phenomena. Guided by these top-level priorities, codedevelopment and assessment follow. As shown in Figure 2, the EMDAP emphasizes fourelements leading towards an evaluation methodology that meets an appropriate standard ofadequacy.

RG 1.203 presents a top-down approach formalizing the development of evaluationmethodologies. It is notable that while EMDAP follows the principles common to CodeScaling Applicability and Uncertainty (CSAU) (Reference 19), EMDAP is applicable to bothdeterministic and statistically based evaluation methodologies. Like CSAU, a key step in theEMDAP is the PIRT, part of the EMDAP element on defining evaluation methodologyrequirements. EMDAP-based methodologies build upon PIRT conclusions to address themany uncertainties associated with a particular accident or transient scenario. While it wouldbe challenging to apply a comprehensive Best-Estimate-Plus-Uncertainty (BEPU) EMDAPlevel of analysis, individually, to each of the approximately 50 design-basis events appearingin SRP Chapter 15, the EMDAP for conservative and deterministic approaches is relaxed atthe expense of unrealized safety margins. In that setting, the PIRT along with a completecode assessment suite provide confirmation that an evaluation methodology and associatedanalytical computer codes are adequately qualified for their particular application.Reference 20 provides an example of this concept as applied to pressurized water reactor(PWR) non-LOCA transients.


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Element 1Establish requirements for evaluation model capability

1. Specify analysis purpose, transient class and powerplant class

2. Specify figures of merit3. Identify systems, components, phases, geometries,

fields, and processes that should be modeled4. Identify and rank phenomena and processes (PIRT)

q

Element 2Develop assessment base

5. Specify objectives for assessment base6. Perform scaling analysis and identify similarity

criteria7. Identify existing data and/or perform integral

(lET) and separate effect tests (SET) tocomplete database

8. Evaluate effects of lET distortion and SETscale up capability

9. Determine experimental uncertainties

Element 4Assess evaluation model adequacy

Closure relations (bottom-up)

13. Determine model pedigree and applicability tcsimulate physical processes

14. Prepare input and perform calculations to assassmodel fidelity and/or accuracy

15. Assess scalability of models

Integrated EM (top-down)

16. Determine capability of field equations and nunericsolutions to represent processes and phenomena

17. Determine applicability of EM to simulate systemcomponents

18. Prepare input and perform calculations to assmasssystem interactions and global capability

19. Assess scalability of integrated calculations and datafor distortions

EM Resolution

20. Determine EM bases and uncertainties

Element 3Develop evaluation model

10. Establish EM development plan11. Establish EM structure12. Develop or incorporate closure models

No Yes

Sappropriate mt P•make and ae'f n anla-

Return toelements •t event

iscorrections

Figure 2. EMDAP Flow Diagram

Regulatory Position 2 of RG 1.203 notes that QA requirements are an essential feature ofthe EMDAP. Implementation of the QA program for the development of an evaluationmethodology is similar to that expected in the development of a computer code (Reference21). As such, accompanying an evaluation methodology are procedures for design control,document control, software configuration control and testing and error identification andcorrective actions used in its development and maintenance as discussed in Section 2.2.

Ultimately, compliance with the quality assurance program is communicated throughdocumentation. The EMDAP explicitly specifies the following documents:

" Evaluation methodology requirements (e.g., regulations and regulatory guidancedocuments)

* Evaluation methodology instructions (e.g., guidelines for developing code input andsafety analysis methods)

©2011 Babcock & Wilcox Nuclear Energy, Inc. All rights reserved. of 153




" Code description manuals (e.g., code structure, user input requirements, models andcorrelations, verification review, and programmers guide)

• Scaling reports (e.g., scaling methodology, comparison of important parameter groups,

distortion analysis)

* Assessment reports (e.g., general test problems, PIRT-targeted)

* Uncertainty analysis reports (e.g., uncertainty ranges, statistical treatment, analysis ofresults)

2.3.2.2 Regulatory Guide 1.77

RG 1.77 provides guidance for evaluating control rod ejection accidents for PWRs. Whilethe rod ejection accidents are excluded in the B&W mPower reactor, RG 1.77 appliesgenerally to short term reactivity insertion accidents (RIA). RIAs may result in the suddeninsertion of reactivity into the core, leading to an increase in reactor power until the DopplerEffect (fuel temperature) compensates for the reactivity imbalance. While this happensquickly, a few tenths of a second, the energy generated can be very significant, especially ata specific location or hot spot. Once the prompt neutron population derived from the powerburst becomes negligible, the neutronic population is sustained at a relatively small quasi-steady level by delayed neutrons from fission fragments decay. During the accident, spatialeffects become important and both radial and axial power greatly redistribute. The transientis not completely adiabatic and heat transfer from rod to coolant is also spatially dependent.Depending on the location in the core (and the particular scenario), a full spectrum of heattransfer regimes from single-phase liquid forced convection to film boiling may be present inthe core.

Most current evaluation models are highly conservative and rely on a simple yet veryconservative hypothesis with respect to the thermal-hydraulics and neutronics performance(point or one-dimensional axial kinetics) (Reference 22). These models are very efficientand entirely adequate to verify acceptance criteria within current licensing framework asestablished based on test programs during the 1970s and 1980s. However, certain testsperformed at the Cabri reactor in 1993 with sodium coolant and at the Nuclear SafetyResearch Reactor in 1994 resulted in failures for irradiated fuel rods for energy depositionssignificantly lower than the accepted limit. Although neither of these experiments were fullyrepresentative to the thermal-hydraulic or neutronics conditions typical of a PWR, generalconcurrence have been reached to extend the experimental program in order to enlightenthe true failure threshold of irradiated fuel rods in representative light water reactorconditions.

As of 2011, RG 1.77 has not been updated to reflect the research since the Cabri tests.However, the SRP Section 4.2, Appendix B defines an Interim Acceptance Criteria (IAC).The IAC addresses multiple criteria, including a method for determining the number of fuelclad failures for radiological assessment; however, regarding the peak fuel enthalpy, thepeak radial average fuel enthalpy has been reduced from 280 cal/g to 230 cal/g.





2.3.2.3 10 CFR 50 Appendix K

Significant regulatory attention has been placed on small- and large-break LOCAs. Mostnotably is the specific discussion of LOCA in Appendix K of 10 CFR Part 50. Appendix Kdescribes a deterministic approach to resolving ECCS performance issues with safetyanalysis. While best-estimate evaluation methodologies are allowed by the U.S. NRC, suchmethods generally need greater resources to develop. For advanced thermal-hydraulicsystems, many of the Appendix K requirements can be administered through code input;however, ten areas have been identified for current generation LWRs (Reference 23) inwhich such best-estimate codes may need assessment and/or modification to demonstrateconformance with Appendix K requirements. These areas are: (1) fission product decay, (2)metal-water reactions, (3) critical flow discharge, (4) ECCS bypass, (5) critical heat flux(CHF)/departure from nucleate boiling (DNB), (6) post-CHF heat transfer during blowdown,(7) prevention from return to nucleate boiling and transition boiling heat transfer, (8) coreflow distribution during blowdown, (9) reflood rate for PWRs, and (10) refill and reflood heattransfer for PWRs. (Note: the B&W mPower reactor eliminates several of these require-ments through design features intended to prevent the onset of particular phenomena.)

2.3.2.4 10 CFR 50.43 Criteria

Demonstrating the performance of the B&W mPower reactor under accident and transientscenarios requires a thorough verification and validation program involving scaled testingdata and mathematical modeling. The unique characteristics of plants using simplified,inherent, passive, or other innovative means to accomplish their safety functions areexplicitly recognized in the regulations governing the evaluation of standard plant designs.Prepared testing and modeling data must satisfy the regulations in 10 CFR 50.43(e)(1) that:

* The performance of each safety feature of the design has been demonstrated throughanalysis, test programs, experience, or a combination thereof.

* Interdependent effects among the safety features of the design have been foundacceptable by analysis, appropriate test programs, experience, or a combination thereof.

* Sufficient data exist on the safety features of the design to assess the analytical toolsused for safety analysis over a range of normal operating conditions, transientconditions, and specified accident sequences, including equilibrium core conditions.

Babcock & Wilcox has constructed a B&W mPower Integrated Systems Test (IST) facility inBedford County, Virginia. The IST program serves to extend the existing PWR database toconfirm the B&W mPower reactor design methodology, with an emphasis on the passiveengineered safety systems. The function and adequacy of plant control systems,engineered safety features, and protection systems will be demonstrated. Further, testprogram data will improve the analytical methodology for the B&W mPower facility designand operation and confirm the performance of the passive safety features during design-basis events. Experiments at the IST will also provide insights into the requirements foremergency operating procedures and improve their guidance.


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Following the accident at Three Mile Island Unit 2, the U.S. NRC staff developed an actionplan to provide a comprehensive and integrated plan to improve safety at power reactors. Inclarifying the action plan, simulation of a suite of thermal-hydraulic tests at the U.S. NRCsponsored Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering was requiredby plant vendors using their computer codes and analysis methods. Those specificallyidentified in Reference 24 include LOFT tests L3-1, L3-2, and L3-6 along with Semiscale S-07-10B.

2.3.3 Radiological Consequence Requirements

As described in RG 1.183, an accident source term assumes a major accident involvingsignificant core damage and is typically postulated to occur in conjunction with a largeLOCA. Although the LOCA is typically the maximum credible accident, NRC staff experiencein reviewing license applications has indicated the need to consider other accidentsequences of lesser consequence but higher probability of occurrence. Accident sourceterm is characterized by radionuclide composition and magnitude, chemical and physicalform of the radionuclides, and the timing of the release of these radionuclides. Beyond theguidance for analyses addressing RG 1.183, several requirements are imposed on NPPsiting and operation to assure that the consequences of design-basis events remain wellbelow that which would be expected to result in adverse impact on public health and safety.These are summarized in NUREG-0800, Section 15.0.3, and include:

* Section 50.34(a)(1) of 10 CFR 50, "Contents of applications; technical information," as itrelates to the evaluation and analysis of the offsite radiological consequences ofaccidents with fission product release. Of particular note:

o An individual located at any point on the boundary of the exclusion area for any2-hour period following the onset of the postulated fission product release, wouldnot receive a radiation dose in excess of 25 rem total effective dose equivalent(TEDE), and

o An individual located at any point on the outer boundary of the LPZ, who isexposed to the radioactive cloud resulting from the postulated fission productrelease (during the entire period of its passage), would not receive a radiationdose in excess of 25 rem TEDE.

* GDC 19 of Appendix A to 10 CFR 50, "Control room," as it relates to maintaining thecontrol room in a safe condition under accident conditions by providing adequateprotection against radiation

* Section 100.21 of 10 CFR Part 100, "Non-seismic siting criteria," as it relates to theevaluation and analysis of the radiological consequences of accidents for the type offacility to be located at the site in support of evaluating the site atmospheric dispersioncharacteristics

* Paragraph IV.E.8 of Appendix E, to 10 CFR Part 50, "Emergency Planning andPreparedness for Production and Utilization Facilities," as it relates to adequateprovisions for an onsite technical support center (TSC) from which effective direction canbe given and effective control can be exercised during an emergency


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3. B&W mPOWER REACTOR AND SYSTEMS OVERVIEW

The B&W mPower reactor is an advanced light water PWR system that uses an integral reactordesign in which the reactor core, control rod drive mechanisms (CRDMs), steam generator, andpressurizer are contained within a single reactor pressure vessel (RPV). The system willprovide up to 500 megawatts of thermal power (MWt). The information contained within thisdesign overview section is preliminary and subject to change.

Features of the B&W mPower design include:

* rail shippable components

" forgings within U.S. manufacturing capabilities

" no safety-rated emergency (diesel) power required

" underground containment

* no primary coolant penetrations larger than[] [CCI per Affidavit 4(a)-(d)]

" conventional balance-of-plant systems and components

* water-cooled condensers

* four year fuel cycle (target)

* standard fuel (U0 2 enriched up to 5 wt% 235U)

" no boron in primary system coolant for normal reactivity control

* digital instrumentation and controls

* underground spent fuel storage

* 60-year plant design life

3.1 Reactor Coolant System

The RCS consists of the reactor core, CRDMs, reactor coolant pumps (RCPs), steamgenerator, and pressurizer. With the exception of the RCPs, the RCS is contained in the RPV.There is no RCS loop piping external to the RPV as in a conventional PWR.

] [CCI per Affidavit 4(a)-(d)]


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Table 1. General RCS Specifications

Operating power, MWt 500

Operating pressure, psia (MPa) 2050 (14.1)

Design pressure, psia (MPa) [ ] [CCI per Affidavit 4(a)-(d)]

Cold leg temperature, 'F (°C) [ ] [CCI per Affidavit 4(a)-(d)]

Hot leg temperature, °F (°C) 608 (320)

Total coolant flow @ full power, millions of pounds [ ] [CCI per Affidavit 4(a)-(d)]per hour (kilograms per second)

The RPV section division appears in Figure 3. [


RPV penetrations consist of [

I [CCI per Affidavit 4(a)-

(d)]


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UpperVessel -

LowerVessel


Primary Loop Secondary Loop

Figure 3. RPV Upper and Lower Regions

3.2 Emergency Core Cooling System

The ECCS is a safety system that provides four basic functions: (1) depressurization of theRCS, (2) core decay heat removal, (3) cooling water injection into the reactor core, and (4)emergency shutdown of the reactor via injection of soluble boron should control rods fail to fullyinsert upon a reactor trip.



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L

][CCI per Affidavit 4(a)-(d)]

Figure 4. Simplified ECCS diagram

3.2.1 Emergency Core Cooling Condenser

I


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[CCI per Affidavit 4(a)-(d)]

3.2.2 Automatic Depressurization Valves

The ADVs are part of the ECCS.


3.2.3 Low Pressure Injection from the Refueling Water Storage Tank

Cooling water flows into the reactor vessel from the RWSTs after a LOCA[


3.2.4 Ultimate Heat Sink

I






3.2.5 Containment Recirculation

After cooling water begins flowing from the RWST through the low pressure injection valves,water levels in the reactor core and in the containment surrounding the reactor vessel(reactor cavity) will reach equilibrium after some LOCAs.


3.2.6 Emergency Boron Injection

Two redundant and independent emergency boron tanks (EBTs) provide an additional anddiverse reactivity control measure to shut down the reactor.

[CCI per

Affidavit 4(a)-(d)]

3.3 Non-Safety Decay Heat Removal

For nominal shutdown conditions, the feedwater system provides decay heat removal to themain condenser via the steam turbine bypass subsystem of the steam system. Should thissteam turbine bypass subsystem be unavailable, shutdown decay heat can be removed by theplant auxiliary condenser, in conjunction with the RCIPS.

3.3.1 Auxiliary Condenser

The auxiliary condenser system is a non-safety-related system that supports full decay heatremoval capability after shutdown. It directly interfaces with the main steam and feedwatersystems [


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3.3.2 Reactor Coolant Inventory and Purification System

The RCIPS is a non-safety, reactor support system that maintains and controls the qualityand distribution of the primary coolant inventory.


3.4 Main Steam System

The main steam system (MSS) includes all piping from the two OTSG outlet nozzles to acommon header in containment through a piping and turbine (high- and low-pressure) systemoutside of containment ending at the steam turbine stop valve (TSV), all take-off piping to theturbine bypass valves (TBVs), main steam safety valves (MSSVs), moisture separator/reheaters(MS/R), drains, valves, and instrumentation. The general configuration appears in Figure 5.The primary function of the main steam system is for transporting the steam generated in theOTSG to the steam turbine generators or the turbine bypass system over the entire range ofmain steam system operation, from warm-up to maximum operating conditions. The MSS alsosupplies steam to the MS/R and the gland seal system for the turbine. In the event that theturbines are unavailable, the MSS bypasses the high pressure turbine and transports the steamto the condenser via the TBVs.

[


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3.4.1 Steam Turbine Generator

The steam turbine generator consists of a high pressure turbine stage and a low pressureturbine stage. The high pressure turbine receives steam from the steam generator via theMSS piping. The low pressure turbine receives steam via a MS/R and low pressurecrossover piping. The high pressure turbine includes a TSV installed on each inlet of thehigh pressure turbine and a turbine control valve (TCV) for fine adjustment of inlet pressureand, thus, plant load. The TSV closes on several set points determined by the turbinemanufacturer. The high pressure turbine contains two extraction steams (MS/R reheaterand feedwater heater (FWH) #4 source steam), while the low pressure turbine containsthree extraction streams (source steam for FWH #1, FWH #2, and deaerator). The turbinedesign will employ equipment protection features such as limiting turbine overspeed duringloss-of-external load.



L


Figure 5. Main Steam and Turbine Generator Systems

3.4.2 Turbine Bypass Valves

The two TBVs accept main steam via a nozzle and reduce the pressure. Prior to exiting theTBVs, feedwater mixes with the steam in order to desuperheat the steam before enteringthe condenser. The TBVs are located on branch piping just upstream of the TSV.


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3.4.3 Main Steam Isolation Valve

The MSIV consists of one fast-acting valve actuated by either automatic or manual controlaction. The location of the MSIV is downstream of the containment penetration,[

] [CCI per Affidavit 4(a)-(d)] The MSIV isolates the main steaminside containment in response to low steam pressure, low RCS pressure, or highcontainment pressure setpoints. The MSIV may also be manually actuated from the maincontrol room (MCR).

3.4.4 Main Steam Safety Valves

The MSSVs prevent the MSS overpressure. The size of the MSSVs is sufficient to relievethe maximum main steam flow rate with additional margin. They discharge to vent stacksthat vent steam away from structures while minimizing backpressure, and preventingbackflow. The MSSVs are located in the reactor service building on branch piping from themain steam header, outside containment, and upstream of the MSIV. To help reduce MSSVopenings and issues arising from reseating of MSSVs after opening, a pilot-operated reliefvalve (PORV) or similar pressure relief function with upstream isolation will be employed.

3.5 Feedwater System

The feedwater system (FWS) provides heated deaerated water to the steam generator via theelectric motor-driven feedwater pumps (FWPs) and FWHs, which increase overall pressure andtemperature of the feedwater in the system to satisfy the requirements of the steam generator.The general configuration appears in Figure 6.

The FWS consists of the following:

[



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I'


Figure 6. Feedwater System


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3.6 Instrumentation and Control Systems

The instrumentation and control (I&C) systems strategy varies widely with different plantdesigns; however, the primary functions are similar for conventional PWRs. While thecharacterization of these systems for the B&W mPower reactor is ongoing and subject tochange, the principal functions of I&C systems are well established. These are (1) to control thenormal operation of the facility within design limits, (2) to provide information and alarms in thecontrol room to monitor the operation and status of the facility and permit corrective actions tobe taken for off-normal plant conditions, (3) to establish that the facility is operating withindesign and license limits, (4) to permit or support the correct operation of engineered safetyfeatures, and (5) to monitor and record important parameters during and following accidents,enabling necessary operator actions.

The B&W mPower plant control system (mPCS) is a digital system that provides protection,control, monitoring and alarming functions.

[CCI per

Affidavit 4(a)-(d)]

3.6.1 Reactor Protection and Engineered Safeguards Features Actuation Systems

The RPS and ESFAS are safety-related systems comprised of [ ] [CCI per Affidavit4(a)-(d)] independent, redundant divisions and channels of instrumentation that providemonitoring, control, warning, and trip functions for the protection of reactor fuel and thereactor coolant pressure boundary. The generation of trip commands depends on thecoincidence of instrument channel logic conditions that indicate an unsafe condition. TheRPS and ESFAS-related equipment includes:

* sensors

* wiring

* inter-connecting devices

* input processing modules

" logic processing modules

* output logic processing modules

* contactors

* relays

* load drivers



" displays

* communication modules

* other equipment required to execute the required safety functions of the system

I



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The RPS will utilize several process inputs to initiate safety functions. Table 2 presents theRPS sensor inputs being considered for the B&W mPower reactor.

Table 2. RIPS sensor inputs

Protected Function Sensed Parameter

i

] (CCI per Affidavit4(a)-(d)]

The instrumentation monitored by the RPS will likely include:

[




3.6.2 Other Control Systems

The following list identifies additional instrumentation and control systems that may appearas part of the power generation systems.

I



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4. IDENTIFICATION OF TRANSIENT CLASSES

The evaluation of the safety of a NPP includes analyses of the plant's responses to possibleequipment failures or malfunctions. Those design-basis events specifically address thoseexpected to occur over the lifetime of a plant plus a few exceptional events that are notexpected to occur over a plant's lifetime, but are examined to demonstrate design robustness.The U.S. NRC recognizes two event classifications determined on the basis of frequency is asfollows (Reference 2):

* AQOs: Events expected to occur on a frequency of one or more times during the lifetimeof the plant

* Accidents: Events not expected to occur that are evaluated to demonstrate designadequacy

Such safety analyses provide a significant contribution to the selection of LCO measures,limiting safety system settings, and design specifications for components and systems from thestandpoint of public health and safety. The following tables provide a summary of the accidentsto be analyzed for the B&W mPower reactor, as informed by NUREG-0800. Most transients andaccidents described in NUREG-0800 are either identical or very similar to the commercialoperating PWRs or to advanced PWRs with passive safety systems. Table 2 lists the suite ofsafety analyses that must be addressed to comply with 10 CFR 50.34. It includes a discussionof their applicability to the B&W mPower reactor. Table 4 identifies transients that are unique tothe B&W mPower reactor.


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Table 3. Applicability of NUREG-0800 Transients to B&W mPower Reactor

SRP Title B&W mPower ApplicabilitySection _ _ _ _,

15.1.1 - Decrease in Feedwater Temperature,15.1.4 Increase in Feedwater Flow, Increase

in Steam Flow, and InadvertentOpening of a Steam Generator Relief

I or Safety Valve15.1.5 Steam System Piping Failures Inside

and Outside of Containment (PWR)

15.1.5.A Radiological Consequences of MainSteam Line Failures OutsideContainment of a PWR

15.2.1 - Loss of External Load; Turbine Trip;15.2.4 Loss of Condenser Vacuum; and

Closure of Main Steam Isolation Valve15.2.5 Steam Pressure Regulator Failure

(Closed)

15.2.6 Loss of Nonemergency AC Power tothe Station Auxiliaries

15.2.7 Loss of Normal Feedwater Flow

15.2.8 Feedwater System Pipe Breaks Insideand Outside Containment (PWR)

15.3.1 - Loss of Forced Reactor Coolant Flow15.3.2 Including Trip of Pump Motor and Flow

Controller Malfunctions15.3.3 - Reactor Coolant Pump Rotor Seizure /15.3.4 Reactor Coolant Pump Shaft Break

15.4.1 Uncontrolled Control Rod AssemblyWithdrawal from a Subcritical or LowPower Startup Condition

15.4.2 Uncontrolled Control Rod AssemblyWithdrawal at Power

15.4.3 Control Rod Misoperation (SystemMalfunction or Operator Error)

(Table 3 continued on next page) [CCI per Affidavit 4(a)-(d)]


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SRP Title B&W mPower ApplicabilitySection

15.4.4 - Startup of an Inactive Loop or15.4.5 Recirculation Loop at an Incorrect

Temperature / Flow ControllerMalfunction Causing an Increase inBoiling Water Reactor (BWR) CoreFlow Rate

15.4.6 Inadvertent Decrease in BoronConcentration in the Reactor CoolantSystem (PWR)

15.4.7 Inadvertent Loading and Operation ofa Fuel Assembly in an ImproperPosition

15.4.8 Spectrum of Rod Ejection Accidents(PWR)

15.4.8.A Radiological Consequences of aControl Rod Ejection Accident (PWR)

15.4.9 Spectrum of Rod Drop Accidents(BWR)

15.4.9.A Radiological Consequences of ControlRod Drop Accident (BWR)

15.5.1 Inadvertent Operation of ECCS

15.5.2 Chemical and Volume Control System(CVCS) Malfunction that IncreasesReactor Coolant Inventory

15.6.1 Inadvertent Opening of a PWRPressurizer Pressure Relief Valve or aBWR Pressure Relief Valve

(Table 3 continued on next page)

[CCI per Affidavit 4(a)-(d)]@2011 Babcock & Wilcox Nuclear Energy, Inc. All rights reserved.

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SRP Title B&W mPower ApplicabilitySection

15.6.2 Radiological Consequences of theFailure of Small Lines CarryingPrimary Coolant Outside Containment

15.6.3 Radiological Consequences of SteamGenerator Tube Failure

15.6.4 Radiological Consequences of MainSteam Line Failure OutsideContainment (BWR)

15.6.5 Loss-of-Coolant Accidents ResultingFrom Spectrum of Postulated PipingBreaks Within the Reactor CoolantPressure Boundary

15.6.5.A Radiological Consequences of aDesign Basis Loss-of-Coolant AccidentIncluding Containment LeakageContribution

15.6.5.B Radiological Consequences of aDesign Basis Loss-of-CoolantAccident: Leakage From EngineeredSafety Feature Components OutsideContainment

15.6.5.D Radiological Consequences of aDesign Basis Loss-of-CoolantAccident: Leakage From Main SteamIsolation Valve Leakage ControlSystem (BWR)

15.7.3 Postulated Radioactive Releases Dueto Liquid-Containing Tank Failures

15.7.4 Radiological Consequences of FuelHandling Accidents

15.7.5 Spent Fuel Cask Drop Accidents

15.8 Anticipated Transients Without Scram

15.9 Boiling Water Reactor Stability



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Table 4. Transients that are Specific to B&W mPower Reactor

I

-I- I

+ 4

](Table 4 continued on next page) [CCI per Affidavit 4(a)-(d)]



E

+ 4



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4.1 Description of Chapter 15 Events to be Analyzed for the B&W mPower Reactor

The discussion herein emphasizes the range of applications for which the development of a full-scope thermal-hydraulic system code and related evaluation methodologies should address.The detail describing the individual events is a qualitative assessment based on conventionalPWR experience while acknowledging certain design elements unique to the B&W mPowerreactor. The progression of many events is a strong function of the instrumentation and controllogic, which is subject to change through design optimization. In addition, several non-safetysystems that would otherwise be expected to assist in event mitigation are not credited in safetyanalysis. These include, but are not limited to, turbine bypass, auxiliary condenser decay heatremoval, RCIPS decay heat removal, and pressurizer spray.

4.1.1 Increase in Heat Removal by the Secondary System

A number of transients can result in an inadvertent and uncontrolled increase in secondarysystem heat removal. Due to the decrease in reactor core inlet temperature and thepresence of a negative moderator coefficient in the core, these transients can result in anincrease in reactor power or a decrease in the shutdown margin. Any unplanned powerlevel increases may result in fuel damage or excessive reactor system pressure.

Because the B&W mPower reactor employs a BOP similar to conventional PWR designs,the failure modes in the secondary systems are expected to be similar to these plants andno new failures and phenomena are expected to occur. The following transients identified inthe SRP for PWRs are applicable to the B&W mPower plant.

* Decrease in feedwater temperature

" Increase in feedwater flow

" Increased steam flow

* Inadvertent opening of a steam generator relief or safety valve

* Steam system piping failures inside and outside of containment

4.1.1.1 Feedwater System Malfunctions Causing a Decrease in Feedwater Temperature

A decrease in feedwater temperature is an AOO expected to occur with moderatefrequency. In this scenario feedwater temperature decreases because of a failure of thefeedwater heater system (FWH), such as by an inadvertent opening of a feedwater heaterbypass valve.


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] [CCI per Affidavit4(a)-(d)]

SRP Acceptance Criteria

I.1


The analysis will be performed to confirm that the following applicable SRP AcceptanceCriteria for Section 15.1.1 events have been satisfied:

1. Pressure in the RCS and MSS should be maintained below 110 percent of thedesign values.

2. Fuel cladding integrity is maintained by keeping the minimum DNBR above the 95percent probability/95 percent confidence DNBR limit.

3. An incident of moderate frequency should not generate a more serious plantcondition without other faults occurring independently.


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4. The requirements stated in RG 1.105, "Setpoints for Safety-Related Instrumentation,"(Reference 26), are used with regard to their impact on the plant response to thetype of AQOs addressed in this SRP section.

5. The most limiting plant systems single failure, as defined in the "Definitions andExplanations" of Appendix A to 10 CFR 50, must be assumed in the analysis andshould follow the guidance stated in RG 1.53, "Application of the Single-FailureCriterion to Safety Systems," (Reference 27).

6. Parameter values in the analytical model should be suitably conservative.

a. Initial power level is rated output (licensed core thermal power) for the number ofloops initially assumed operating plus an allowance of two percent to account forpower measurement uncertainty unless a lower number can be justified throughthe measurement uncertainty methodology and evaluation or the uncertainty isaccounted for otherwise.

b. Conservative scram characteristics are assumed (e.g., maximum time delay withthe most reactive rod held out of the core).

c. The core burnup is selected to yield the most limiting combination of MTC, voidcoefficient, Doppler coefficient, axial power profile, and radial power distribution.

d. Mitigating systems should be assumed as actuated in the analyses at setpointswith allowance for instrument uncertainty in accordance with RG 1.105 and asdetermined by the organization responsible for instrumentation and controls,

4.1.1.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow

A failure or misoperation of the feedwater control system is an AOO, expected to occur withmoderate frequency. In this scenario the delivery of feedwater increases because of afailure related to the feedwater flow control system or the valve that responds to thefeedwater flow control system. [


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Analyses will examine the RCS response to a failurecontrol system. [

(d)] See SRP Acceptance Criteria in Section 4.1.1.1.

or misoperation of the feedwater

] [CCI per Affidavit 4(a)-

4.1.1.3 Failures That Result in Increased Steam Flow

The initiating event is a sudden increase in steam flow due to an inadvertent opening of theTCV or the turbine bypass valves (TBVs). This transient is an AOO, expected to occur withmoderate frequency. [



I


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I [CCI per Affidavit 4(a)-(d)] See SRP Acceptance Criteria in Section4.1.1.1.

4.1.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve

This transient is very similar to the inadvertent opening of the TCV or TBVs. This transient isan AOO, expected to occur with moderate frequency. [



I

] [CCI perAffidavit 4(a)-(d)] See SRP Acceptance Criteria in Section 4.1.1.1.

4.1.1.5 Steam System Piping Failures Inside and Outside of Containment

The steam system piping failure or steam line break (SLB) transient is a postulated accidentnot expected to occur during the life of the NPP. Analysis of the transient following a SLB issensitive to the fluid discharge rate at the break so that a range of break sizes must beconsidered both inside and outside containment to determine the acceptability of the systemresponse. The course that the transient takes and its ultimate effects also depend on theassumed initial power level and mode of operation (e.g., hot shutdown; full power).

The accident initiator is a break in the main steam line located either inside or outside thecontainment. In-containment breaks will be contained by closures of the MSIV andfeedwater control valve. As such, steam released from these breaks will continue until thefeedwater inventory in the steam generator and feedwater lines is exhausted.

[






The applicable SRP acceptance criteria for Section 15.1.5 events are as follows:

1. Pressure in the RCS and MSS should be maintained below acceptable design limits,considering potential brittle as well as ductile failures.


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2. The potential for core damage is evaluated on the basis that it is acceptable if theminimum DNBR remains above the 95/95 DNBR limit for PWRs based onacceptable correlations. If the DNBR falls below these values, fuel failure (rodperforation) must be assumed for the rods that do not meet this criterion unless it canbe shown, based on an acceptable fuel damage model (which includes the potentialadverse effects of hydraulic instabilities), that fewer failures occur. Fuel damage thatis calculated to occur must be of sufficiently limited extent that the core remains inplace and intact with no loss of core cooling capability.

3. The radiological consequences analysis meets the requirements summarized in SRP15.0.3 (see also Section 2.3.3).

4. The auxiliary feedwater system or other means of decay heat removal must besafety-related and, when required, automatically initiated.

5. Tripping of the RCPs should be consistent with the resolution to Three Mile Island(TMI) Action Plan item II.K.3.5 (identified in the SRP).

Breaks located inside the containment must be investigated for both RCS performance andfor peak containment pressure and temperature (refer to SRP Section 6.2.1.4).

4.1.1.6



See SRP Acceptance Criteria in Section 4.1.1.1.

4.1.1.7



.J [CCI per Affidavit 4(a)-(d)]



4.1.2 Decrease in Heat Removal by the Secondary System

A number of transients can result in an unplanned decrease in secondary system heatremoval. Standard Review Plan (SRP) Sections 15.2.1 - 15.2.8 provide the guidance andbasis for the development and performance of these analyses. Due to the decrease in heatremoval, transients discussed in this section can result in RCS and steam systempressurization and DNBR violations.

The following transients identified in the SRP for PWRs are applicable to the B&W mPowerplant.

* Loss of external load

* Turbine trip

* Loss of condenser vacuum

" Inadvertent closure of the MSIV

* Steam pressure regulator failure

* Loss of non-emergency AC power to the station auxiliaries

* Loss of normal feedwater flow

* Feedwater piping breaks inside and outside of containment

4.1.2.1 Loss of External Load

The loss of external load (LOEL) transient is typically a transmission line disturbance thatresults in the plant separation from the grid, causing a full or partial load rejection. Turbineinlet pressure increases with the LOEL. As such, the TCV partially closes, in an attempt tomaintain the nominal state. In doing so, it serves to reduce the generated megawatts to thehouse load level. Main steam pressure increases and the control-grade TBVs open at their


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setpoint to release excess steam directly to the condenser. This transient is an AOO,expected to occur with moderate frequency.



The applicable SRP acceptance criteria for Section 15.2.1 events are as follows:

1. The basic objectives of the review of this transient scenario are met as follows:


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a. To identify which moderate-frequency event that results in an unplanneddecrease in secondary system heat removal is the most limiting, in particular asto primary pressure, secondary pressure, and long-term decay heat removal.

b. To verify whether the predicted plant response for the most limiting eventsatisfies the specific criteria for fuel damage and system pressure.

c. To verify whether the plant PS setpoints assumed in the transients analyses areselected with adequate allowance for measurement inaccuracies as delineated inRG 1.105.

d. To verify whether the event evaluation considers single failures, operator errors,and performance of non-safety-related systems consistent with the RG 1.206regulatory guidelines.

2. With the ANS standards as guidance, specific criteria meet the relevant requirementsof GDCs 10, 13, 15, 17, and 26 for events of moderate frequency, as follows:

a. Pressure in the RCS and MSS should be maintained below 110 percent of thedesign values.

b. Fuel-cladding integrity is maintained by keeping the minimum DNBR above the95/95 DNBR limit for PWRs based on acceptable correlations and by satisfactionof any other specified acceptable fuel design limit (SAFDL) applicable to theparticular reactor design.

c. An incident of moderate frequency should not generate an aggravated plantcondition without other faults occurring independently.

d. The requirements in RG 1.105 are used for their impact on the plant response tothe type of AQOs addressed in this SRP section.

e. The most limiting plant system single failure, as defined in "Definitions andExplanations," 10 CFR 50, Appendix A, must be assumed in the analysisaccording to the guidance of RG 1.53 and GDC 17.

f. Performance of non-safety-related systems during transients and accidents andsingle failures of active and passive systems must be evaluated and verifiedaccording to the guidance of SECY 77-439, SECY 94-084, and RG 1.206.

3. The applicant should analyze events using an acceptable analytical model. Any otheranalytical method proposed by the applicant is evaluated by the staff foracceptability. For new generic methods, the reviewer requests an evaluation by theappropriate organization for reactor systems. The values of the parameters in theanalytical model should be suitably conservative. The following values areacceptable, and are used as described below:


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a.

b.

C.

d.


4.1.2.2 Turbine Trip

The turbine trip transient begins with the closure of the TSV. This transient is an AOO,expected to occur with moderate frequency. The steam flow is terminated, causing thesteam pressure to increase rapidly to the setpoint of the TBVs. [


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I

] [CCI per Affidavit 4(a)-(d)] See SRP Acceptance Criteria inSection 4.1.2.1.

4.1.2.3 Loss of Condenser Vacuum

Loss of condenser vacuum results in an automatic turbine trip;[





4.1.2.4 Inadvertent Closure of Main Steam Isolation Valve

The inadvertent closure of the main steam isolation valve results in isolation of steam flowfrom the steam generator. [








4.1.2.5 Steam Pressure Regulator Failure

The steam pressure regulator controls feedwater flow.

] [CCI per Affidavit 4(a)-(d)] This transient is an AOO, expected to occur with

moderate frequency.



4.1.2.6 Loss of Non-Emergency AC Power to the Station Auxiliaries

The loss of non-emergency AC power to the station auxiliaries is essentially a loss of offsitepower with a concurrent turbine trip. This transient is an AOO, expected to occur withmoderate frequency.



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4.1.2.7 Loss of Normal Feedwater Flow

Loss of main feedwater flow can be caused by a failure within the feedwater control system,a valve failure or failure of the main FWP. This transient is an AOO, expected to occur withmoderate frequency.

] [CCI perAffidavit 4(a)-(d)]



4.1.2.8 Feedwater Piping Breaks Inside and Outside of Containment

The steam and water release from a postulated feedwater line break results in a loss ofsecondary coolant which may result in a reactor system cool-down (by excessive energydischarge through the break) or a reactor system heat-up (from the loss of reactor systemheat sink). This is a postulated accident that is not expected to occur during the life of theNPP. [


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The applicable SRP acceptance criteria for this event are as follows:

1. Requirements for maintenance of adequate decay heat removal by the auxiliaryfeedwater system (AWFS) are in 10 CFR 50.34(f)(1)(ii), (TMI issue I1.E.1.1)and 10CFR 50.34(f)(2)(xii), (TMI issue II.E.1.2). Requirements for RCP operation are in 10CFR 50.34(f)(1)(iii), (TMI issue 2.K.2).

2. Pressure in the RCS and MSS should be maintained below 110 percent of thedesign pressures in the ASME Boiler and Pressure Vessel Code, Section III as citedin Reference 2 for low-probability events and below 120 percent for very low-probability events like double-ended guillotine breaks.

3. The potential for core damage is evaluated for an acceptable minimum DNBRremaining above the 95/95 DNBR limit for PWRs based on acceptable correlations. Ifthe DNBR falls below these values, fuel failure (rod perforation) must be assumed forrods not meeting this criterion unless, from an acceptable fuel damage model (referto SRP Section 4.2) including the potential adverse effects of hydraulic instabilities,fewer failures can be shown to occur. Any fuel damage calculated to occur must beof sufficiently limited extent that the core remains in place and intact with no loss ofcore cooling capability.

4. Calculated doses at the site boundary from any activity release must be a smallfraction of the 10 CFR 100 guidelines.

5. The auxiliary feedwater system (or other means of decay heat removal) must besafety-related and, when required, automatically initiated.

6. Certain assumptions should be in the analysis of important parameters that describeinitial plant conditions and postulated system failures.





a. The power level assumed and number of loops operating at the initiation of thetransient should correspond to the operating condition which maximizes accidentconsequences.

b. The assumptions as to whether offsite power is lost and the time of loss shouldbe conservative.

c. The effects (such as pipe whip, jet impingement, reaction forces, temperature,and humidity) of the postulated feedwater line breaks on other systems should beconsidered consistently with the intent of BTP 3-3 and BTP 3-4.

d. The worst single active component failure should be assumed to occur in thesystems required to control the transient. For new applications, LOOP should notbe considered a single failure; FWLBs should be analyzed with and withoutLOOP, as in assumption B, in combination with a single, active failure.

e. The maximum rod worth should be assumed to be held in the fully withdrawnposition per GDC 25. An appropriate rod reactivity worth versus rod positioncurve should be assumed.

f. The core burnup (time in core life) should be selected to yield the most limitingcombination of moderator temperature coefficient, void coefficient, Dopplercoefficient, axial power profile, and radial power distribution.

g. The initial core flow assumed for the analysis of the feedwater line ruptureaccident should be chosen conservatively.

h. During the initial 10 minutes of the transient, if credit for operator action isrequired (i.e., RCP trip), an assessment for the limiting consequence mustaccount for operator delay and error.

4.1.3 Decrease in Reactor Coolant System Flow

The primary safety concern for the loss of forced reactor coolant flow is a loss of heattransfer from the core, causing increased fuel temperature and eventual fuel damage. Thefollowing transients identified in the SRP for PWRs are applicable to the B&W mPower plant.

* Partial loss of forced reactor coolant flow from pump motor trip and flow controllermalfunction

* Complete loss of forced reactor coolant flow from pump motor trip and flow controllermalfunction

* Reactor coolant pump seizure (locked rotor)

* Reactor coolant pump shaft break


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4.1.3.1 Partial Loss of Forced Reactor Coolant Flow

A partial loss of forced reactor coolant flow results from a motor trip caused by electricalanomalies (e.g., overcurrent, phase imbalance), a mechanical failure of a RCP, or a fault inthe electrical power supply. As a consequence, the failed RCPs coast down and total flowto the core decreases. This event is an AOO, expected to occur with moderate frequency.

] [CcIper Affidavit 4(a)-(d)]


The SRP acceptance criteria for this event are as follows:

1. Pressure in the RCS and MSS should be maintained below 110 percent of thedesign values. The event does not challenge the design limit.

2. Fuel-cladding integrity is maintained by keeping the minimum DNBR above the 95percent probability/95 percent confidence DNBR limit.


4. The requirements stated in RG 1.105 are evaluated with regard to their impact on theplant response to the type of AQOs addressed in this SRP section.


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5. Onsite and offsite electric power systems must be maintained so safety-relatedSSCs function during normal operation and AQOs.

6. The most limiting plant systems single failure, as defined in the "Definitions andExplanations" of Appendix A to 10 CFR 50, must be assumed in the analysis andshould follow the guidance stated in RG 1.53.

7. The performance of non-safety-related systems during transients and accidents andof single failures of active and passive systems (especially the performance of checkvalves in passive systems), must be evaluated and verified by the guidance of SECY77-439 as cited in Reference 2, SECY 94-084 as cited in Reference 2, and RG1.206.

8. The applicant's analysis of the most limiting AQOs should use an acceptable model.Unapproved analytical methods proposed by the applicant are evaluated by the stafffor acceptability.

9. Parameter values in the analytical model should be suitably conservative.

a. Initial power level is rated output (licensed core thermal power) for the number ofloops initially assumed operating plus an allowance of two percent to account forpower measurement uncertainty unless a lower number can be justified throughthe measurement uncertainty methodology and evaluation or the uncertainty isaccounted for otherwise (refer to SRP 4.4, Reference 2). The number of loopsoperating at the initiation of the event should correspond to the operatingcondition which maximizes the consequences of the event.


c. The core burnup is selected to yield the most limiting combination of moderatortemperature coefficient, void coefficient, Doppler coefficient, axial power profile,and radial power distribution.

d. Mitigating systems should be assumed as actuated in the analyses at setpointswith allowance for instrument uncertainty in accordance with RG 1.105 and asdetermined by the organization responsible for instrumentation and controls.

4.1.3.2 Complete Loss of Forced Reactor Coolant Flow from Pump Motor Trip and FlowController Malfunction

A complete loss of forced reactor coolant flow results from a loss of electrical power to allRCP motors simultaneously. This event is an AOO, expected to occur infrequently during aplant lifetime. [


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] [CMI perAffidavit 4(a)-(d)]


As with the partial loss of forced coolant flow, [



4.1.3.3 Reactor Coolant Pump Seizure (Locked Rotor)

This transient involves a sudden seizure of one RCP rotor, and is very similar to the partialloss of reactor coolant flow involving one pump trip. The RCP locked rotor transient is apostulated accident. [

I [CCI per Affidavit 4(a)-(d)]



4.1.3.4 Reactor Coolant Pump Shaft Break

This transient involves a sudden break of one RCP shaft, and is very similar to the RCPseizure event. The RCP shaft break transient is categorized as a postulated accident. As


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described previously,[



The design of the B&W mPower reactor, with 8 RCPs serving the single coolant loop, makesthis event equivalent to the events described in Sections 4.1.3.1 and 4.1.3.2. As such, theSRP Acceptance Criteria in Section 4.1.3.1 apply.

4.1.4 Reactivity and Power Distribution Anomalies

The following transients involving reactivity and power distribution anomalies are identified inthe SRP for PWRs.

* Uncontrolled control rod assembly withdrawal

* Control rod misoperation

* Startup of inactive reactor coolant pump at incorrect temperature

" A malfunction or failure of the flow controller in a boiling water reactor loop that results inan increased reactor coolant flow rate

* Chemical and volume control system malfunction that results in a decrease in the boronconcentration in the reactor coolant

• Inadvertent loading and operation of a fuel assembly in an improper position

* Spectrum of rod cluster control assembly ejection accidents

* Spectrum of rod drop accidents in a BWR

4.1.4.1 Uncontrolled Control Rod Assembly Withdrawal

An uncontrolled withdrawal of a control rod bank is an AOO for the B&W mPower reactor.[


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1. The requirements of GDC 10, 20, and 25 concerning themet for this event when:

SAFDL are assumed to be

a. The thermal margin limits as specified in SRP Section 4.4 are met.

b. Fuel centerline temperatures for (PWRs) as specified in SRP Section 4.2 do notexceed the melting point.

4.1.4.2 Control Rod Misoperation

Control rod misoperation can encompass several different initiating events including:insertion of a single control rod assembly at power, axial misalignment of a control rodassembly with respect to its bank, and withdrawal of a single control rod assembly at power.Single control rod assembly insertion or withdrawal and axial misalignment (up to andincluding an assembly staying in the core on withdrawal of the bank) are considered AOOs.

For other PWR designs, withdrawal of a single control rod assembly is a postulated accidentrather than an AOO. In these designs, no single mechanical or electrical failure can result inwithdrawal of a single control rod assembly, and multiple operator errors are required tomanually achieve such a withdrawal. [


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1. The requirements of GDC 10, 17, 20, and 25 concerning the SAFDL are assumed tobe met for this event when:

a. The thermal margin limits as specified in SRP Section 4.4 are met.

b. Fuel centerline temperatures (for PWRs) as specified in SRP Section 4.2 do not

exceed the melting point.

4.1.4.3 Startup of Inactive Reactor Coolant Pump at Incorrect Temperature

[



1. The thermal margin limits as specified in SRP Section 4.4, subsection 11.1, are met.

2. Fuel centerline temperatures (for PWRs) as specified in SRP Section 4.2, subsectionII.A.2(a) and (b),do not exceed the melting point.


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4.1.4.4 A Malfunction or Failure of the Flow Controller in a Boiling Water Reactor Loop

that Results in an Increased Reactor Coolant Flow Rate

This accident is BWR-specific and is not applicable to the B&W mPower reactor.

4.1.4.5 Chemical and Volume Control System Malfunction that Results in a Decrease inthe Boron Concentration in the Reactor Coolant

The B&W mPower reactor does not use boron in the RCS for reactivity control. Therefore,this transient is not applicable to the B&W mPower reactor.

4.1.4.6 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position

Misloading of a fuel assembly is highly unlikely because of the multiple levels ofadministrative controls that will be established and maintained when a core is loaded.Similar care is taken with installation of incore reactor instrumentation (installed in thecorners of selected elements). Moreover, misloaded assemblies or instrumentation wouldbe detected during low power and/or power ascension testing. Startup testing would alsodetect the unlikely presence of an out-of-specification fuel element (i.e., enrichment orimproperly configured burnable poison rods).



The primary safeguards against fuel-loading errors are procedures and design features tominimize the likelihood of the event. Additional safeguards include incore instrumentationsystems which would detect errors. The following acceptance criteria cover the event ofoperation with misloaded fuel caused by loading errors:

1. To meet the requirements of GDC 13, plant operating procedures should include aprovision requiring that reactor instrumentation be used to search for potential fuel-loadingerrors after fueling operations.

2. In the event the error is not detectable by the instrumentation system and fuel rod failurelimits could be exceeded during normal operation, the offsite consequences should be asmall fraction of the 10 CFR Part 100 criteria. A small fraction is interpreted to be less than10% of the 10 CFR Part 100 reference values. For the purpose of this review, theradiological consequences of any fuel-loading error should include consideration of thecontainment, confinement, and filtering systems. The applicant's source terms and


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methodologies with respect to gap release fractions, iodine chemical form, and fissionproduct release timing should reflect NRC-approved source terms and methodologies.

Analyses will consider the possibility that an error occurs and goes undetected. AdditionalSRP acceptance criteria will be derived from a review of applicable challenges arising fromcoincident AQOs with misloaded fuel.

4.1.4.7 Spectrum of Rod Cluster Control Assembly Ejection Accidents

] [CCM per Affidavit 4(a)-(d)]

4.1.4.8 Spectrum of Rod Drop Accidents in a BWR


4.1.5 Increase in Reactor Coolant Inventory

The concern with an uncontrolled increase in reactor coolant inventory is overpressurizationof the RCS and flow of colder water into the core, causing an increase in power due to thenegative moderator coefficient. Also, a solid pressurizer may lead to liquid flow through thepressurizer relief valves. If the valves are not designed for liquid release, they might fail toclose causing loss of reactor coolant inventory.

The SRP identifies events with the potential to cause an unplanned increase in reactorcoolant inventory. The transients applicable to PWRs are (note that the RCIPS serves a rolesimilar to the CVCS in SRP terminology):

* Inadvertent operation of the ECCS that increases RCS inventory

* RCIPS malfunction that increases RCS inventory

* Inadvertent operation of the normal decay heat removal via RCIPS that increases RCSinventory

4.1.5.1 Inadvertent Operation of the ECCS that Increases RCS Inventory

[



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4.1.5.2 RCIPS Malfunction that Increases RCS Inventory

This transient is an AOO, expected to occur with moderate frequency. It can occur when theindicated pressurizer level signal in the level controller fails low. [




2. Fuel-cladding integrity is maintained by keeping the minimum DNBR above the 95percent probability/95 percent confidence DNBR limit.


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4. The values of parameters used in the analytical model should be suitablyconservative. The following values are considered acceptable for use in the model:

a. The initial power level is taken as the licensed core thermal power for the numberof loops initially assumed to be operating plus an allowance of two percent toaccount for power measurement uncertainties, unless a lower power level can bejustified by the applicant. The number of loops operating at the initiation of theevent should correspond to the operating condition which maximizes theconsequences of the event.


c. The core burnup is selected to yield the most limiting combination of moderatortemperature coefficient, void coefficient, Doppler coefficient, axial power profile,and radial power distribution.

4.1.5.3





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4.1.6 Decrease in Reactor Coolant Inventory

A number of transients can result in an unplanned decrease in RCS inventory. Due to theRCS depressurization and the subsequent decrease in subcooling margin, a decrease inreactor coolant inventory can result in an unacceptable DNBR. If the rate of inventorydecrease exceeds the makeup capability of the RCIPS, the core will uncover and fueldamage could occur. The primary safety concern is loss of cooling of the reactor coreresulting in fuel damage and a release of radioactivity to the environment.

The following transients involving a decrease in RCS inventory are identified in the SRP.

" Inadvertent opening of a PWR pressurizer pressure relief valve

* Failure of small lines carrying primary coolant outside containment

* Steam generator tube failure

* Spectrum of boiling water reactor steam system piping failures outside of containment

* Loss of coolant accidents resulting from a spectrum of postulated piping breaks

The following transients are specific to the B&W mPower reactor and have the potential toresult in a decrease in RCS inventory.

[


All of these events fall into the general category of LOCAs. The traditional analyticallydetermined LOCA figures-of-merit are described in 10 CFR 50.46 as peak clad temperature,maximum local oxidation, and core wide hydrogen production.



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4.1.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve

An inadvertent opening of a pressurizer relief valve causes release from the pressurizersteam space into the containment. This transient is an AOO. [



The applicable SRP acceptance criteria are as follows:


2. Fuel cladding integrity is maintained by keeping the core covered as inferred bypreventing the hottest rod from experiencing DNB.

3. An AOO should not develop into a more serious plant condition without other faults

occurring independently.

4.1.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment

Another event that could result in a decrease in reactor coolant inventory (thus leading to atemperature increase in the RCS) is the failure of small lines carrying primary coolantoutside of the containment vessel. This event postulates the failure of small lines outside ofcontainment that are connected to the primary coolant pressure boundary. Such linesinclude instrument, sample, radwaste, or RCIPS lines. This transient is an AOO.


The radiological consequences analysis meets the requirements summarized in SRP 15.0.3(see also Section 2.3.3). See SRP Acceptance Criteria in Section 4.1.6.1.


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4.1.6.3 Steam Generator Tube Rupture

The steam generator tube rupture (SGTR) accident is defined as a breach of the barrierbetween the RCS and the secondary side, with a double-ended break of one tube being thelimiting case. It is a postulated accident, because of its low probability combined withpotentially significant radiological consequences. The first concern of this accident is apotential steam generator overfill due to the reactor coolant leaking through the failed tube,and a subsequent loss of steam superheat and potential moisture or bulk liquid carryover tothe steam lines. The second concern is a radiological release to the environment via thefailed tube and the secondary side of the plant. If the size of the tube rupture too small toresult in a measurable impact on the plant thermal-hydraulic conditions, the event shouldstill be detectable by a monitored increase in radiation activity on the secondary.

A failed steam generator tube causes a decrease in RCS inventory. [



The radiological consequences analysis meets the requirements summarized in SRP 15.0.3(see also Section 2.3.3).

4.1.6.4 Spectrum of Boiling Water Reactor Steam System Piping Failures Outside ofContainment


4.1.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated PipingBreaks

A loss of coolant accident (LOCA) is a postulated accident that results from a loss of reactorcoolant at a rate in excess of the capability of the RCS makeup system, from breaks in theRCS pressure boundary up to and including a break equivalent in size to the double-endedrupture of the largest pipe of the RCS. This accident will not likely occur during the life of theNPP.

The initiating event is a pipe break in the RCS that causes a loss of reactor coolant at a ratein excess of RCS makeup system capabilities. [





The applicable SRP acceptance criteria are:

1. An evaluation of ECCS performance has been performed by the applicant inaccordance with an EM that satisfies the requirements of 10 CFR 50.46. RG 1.157and Section I of Appendix K to 10 CFR 50 provide guidance on acceptable EMs. Forthe full spectrum of reactor coolant pipe breaks, and taking into considerationrequirements for RCP operation during a small break LOCA, the results of theevaluation must show that the specific requirements of the acceptance criteria forECCS are satisfied. This also includes analyses of a spectrum of LOCAs and smallbreak loss of coolant accidents (SBLOCAs) to verify that boric acid precipitation isprecluded for all break sizes and locations.


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2. The analyses should be performed in accordance with 10 CFR 50.46, includingmethods referred to in 10 CFR 50.46(a)(1 ) or (2). The analyses must demonstratesufficient redundancy in components and features, and suitable interconnections,leak detection, isolation, and containment capabilities such that the safety functionscould be accomplished assuming a single failure in conjunction with the availability ofonsite power (assuming offsite electric power is not available, with onsite electricpower available; or assuming onsite electric power is not available with offsiteelectric power available). Additionally the LOCA methodology used and the LOCAanalyses should be shown to apply to the individual plant by satisfying 10 CFR50.46(c)(2), and the analysis results should meet the performance criteria in 10 CFR50.46(b).

a. The calculated maximum fuel element cladding temperature does not exceed2200°F.

b. The calculated total local oxidation of the cladding does not exceed 17 percent ofthe total cladding thickness before oxidation. Total local oxidation includes pre-accident oxidation plus the oxidation that occurs during the course of theaccident.

c. The calculated total amount of hydrogen generated from the chemical reaction ofthe cladding with water or steam does not exceed one percent of the hypotheticalamount that would be generated if all of the metal in the cladding cylinderssurrounding the fuel, excluding the cladding surrounding the plenum volume,were to react.

d. Calculated changes in core geometry are such that the core remains amenableto cooling.

e. After any calculated successful initial operation of the ECCS, the calculated coretemperature is maintained at an acceptably low value and decay heat is removedfor the extended period of time required by the long-lived radioactivity.

These criteria will be addressed via a surrogate figure-of-merit demonstrating thatfuel cladding integrity is maintained, such as by keeping the core covered as inferredby preventing the hottest rod from experiencing DNB (see Note below).

3. The radiological consequences of the most severe LOCA are within the guidelines ofRG 1.183 and 10 CFR 100. For applications under 10 CFR 52, reviewers should useSRP Section 15.0.3 (see also Section 2.3.3).

4. The TMI Action Plan requirements of I1.E.2.3, I1.K.2.8, I1.K.3.5, I1.K.3.25, I1.K.3.30,II.K.3.31, II.K.3.40 have been met.


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Note: Per RG 1.203, Section 1.1.2, a surrogate figure-of-merit can be used when providedwith appropriate justification.

4.1.6.6




4.1.6.7


SRP AcceDtance Criteria


4.1.6.8


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4.1.6.9

] [CCM per Affidavit 4(a)-(d)] This event is an AOO and is unique to the

B&W mPower reactor. It requires a radiological release analysis.



4.1.7 Radioactive Release from a Subsystem or Component

A number of transients can result in a radioactive release from a subsystem or component.The discussion provided in this section is generalized to the extent practical since themajority of the transients will follow the same methodology. All radiological transients willfollow the guidelines presented in SRP 15.0.3.

The analysis approach of the following group of events is discussed in this section.

* Release of radioactivity to the environment due to liquid tank failure

* Fuel handling accident

* Spent fuel cask drop accident


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Generic Method of Evaluation


This accident and certain others can result in the release of primary coolant to theenvironment. Fission products can be present in the primary coolant due to leaking fuelrods (i.e., rods that are leaking prior to the transient). In addition, the cladding on previouslynon-leaking fuel rods can become damaged during certain non-LOCA events involving fuelin the reactor core. This breached cladding can release fission products in the gap betweenthe fuel pellet and the cladding of the fuel rod. This fuel rod gap inventory can also betransferred to the primary coolant, and ultimately released to the environment.

Per RG 1.183, the inventory of fission products in the reactor core and available for releaseto the containment will be based on the maximum full power operation of the core, taking

into account fuel enrichment, fuel burn up, and rated thermal power. In addition, the periodradionuclides to reach equilibrium or to reach maximum values, of irradiation will be ofsufficient duration to allow the activity of dose-significant

The offsite atmospheric dispersion factors (x/Q values) will be taken directly from the EPRIURD. These factors are conservative in that they are part of the envelope of designparameters given in the URD to allow siting at most sites available in the U.S. and areconsistent with guidance in RG 1.145 (Reference 28).

4.1.7.1 Release of Radioactivity to the Environment Due to Liquid Tank Failure

Liquid waste from the facility includes equipment and floor drains, chemical drains (exceptstrong acids), and detergent drains. These liquids are collected separately and stored intanks in the Radwaste Building. The RCS fluid is not directed to the wastewater system.

The SRP recognizes that a wastewater system designed to RG 1.143 would be unlikely tofail catastrophically (i.e., a failure involving the near total loss of the systems inventory ofradioactive material). However, the malfunction of a tank or its components, a valvemisalignment, tank overflow, or an operator error appear more likely and are assumed to bethe types of failures warranting an evaluation of their consequences. Although no specificmodes of failure have been designated as being representative, the U.S. NRC considersthat for the safety evaluation of the wastewater system, the type of malfunction analyzedshould be limited to the postulated failure of a tank or pipe rupture located outside of


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containment. The evaluation would consider the impact of the failure on the nearest potablewater supply and the use of water for direct human consumption, or indirectly throughanimals (livestock watering), crops (agricultural irrigation), and food processing (water as aningredient). Small cracks and consequent low-level releases are bounded by this analysisand should be contained without any significant release. This evaluation is part of the U.S.NRC review in SRP Section 11.2 and the information supporting that evaluation will beincluded in that section of the DCD.


SRP acceptance criteria are based on meeting the relevant requirements of the followingregulations:

1. General Design Criterion 60 as it relates to the radioactive waste managementsystems being designed to control releases of radioactive materials to theenvironment.

2. 10 CFR Part 20 as it relates to radioactivity in effluents to unrestricted areas. Tanksand associated components containing radioactive liquids outside containment areacceptable if failure does not result in radionuclide concentrations in excess of thelimits in 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potablewater supply, in an unrestricted area, or if special design features are provided tomitigate the effects of postulated failures for systems not meeting these limits.

4.1.7.2 Fuel Handling Accident

A fuel handling accident can be postulated to occur either inside the containment (onto thereactor core) or in the fuel storage area (in the spent fuel pool). The fuel handling accidentis defined as the dropping of a spent fuel assembly, as a result of a failure of the fuelassembly lifting mechanism, such that every rod in the dropped assembly has its claddingbreached so that the activity in the fuel/cladding gap is released.


The radiological consequences analysis meets the requirements summarized in SRP 15.0.3(see also Section 2.3.3). Additional event-specific SRP acceptance criteria are based onrequirements of 10 CFR Part 100 with respect to the calculated radiological consequencesof a fuel handling accident and General Design Criterion 61 with respect to appropriatecontainment, confinement, and filtering systems. Specific criteria necessary to meet therequirements are:

1. The radioactivity control features of the fuel storage and handling systems insidecontainment and in the fuel building are acceptable if they meet the requirements ofGeneral Design Criterion 61, "Fuel Storage and Handling and Radioactivity Control,"with respect to appropriate containment, confinement and filtering systems.


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2. The model for calculating the whole-body and thyroid doses is acceptable if itincorporates the appropriate conservative assumptions in Regulatory Guide 1.25,"Assumptions Used for Evaluating the Potential Radiological Consequences of aFuel Handling Accident in the Fuel Handling and Storage Facility for Boiling andPressurized Water Reactors" (Reference 29), with the exception of the guidelines forthe atmospheric dispersion factors (X/Q values). The acceptability of the X/Q valuesis determined under SRP Section 2.3.4.

3. An engineered safety features grade atmosphere clean-up system is required for thespent fuel storage area to reduce the potential radiological consequences.

4. The containment design is acceptable with respect to a postulated fuel handlingaccident if it possesses the capability for prompt radiation detection by use ofredundant radiation monitors and automatic isolation if fuel handling operationsinside containment occur when the containment is open to the environment (i.e., witha containment purge exhaust system). An acceptable alternative approach iscontainment venting through an engineered safety features atmosphere cleanupsystem or containment isolation during fuel handling operations.

4.1.7.3 Spent Fuel Cask Drop Accident

This accident description addresses a spent fuel cask falling or tipping into the spent fuel pool,thereby resulting in damaged fuel and a release of fission products. When necessary, a utilityoperator of a B&W m Power reactor will transfer spent fuel casks in and out of the

fuel handling building with the use of a crane. [



The radiological consequences analysis meets the requirements summarized in SRP 15.0.3(see also Section 2.3.3). Additional SRP acceptance criteria are based on requirements of10 CFR Part 100 with respect to the calculated radiological consequences of a fuel handlingaccident and General Design Criterion 61 with respect to appropriate containment,confinement, and filtering systems. Specific criteria necessary to meet the requirements are:

1. The radioactivity control features of the fuel storage and handling systems insidecontainment and in the fuel building are acceptable if they meet the requirements ofGeneral Design Criterion 61, "Fuel Storage and Handling and Radioactivity Control,"with respect to appropriate containment, confinement and filtering systems.


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2. The model for calculating the whole-body and thyroid doses is acceptable if itincorporates the appropriate conservative assumptions in Regulatory Guide 1.25,"Assumptions Used for Evaluating the Potential Radiological Consequences of aFuel Handling Accident in the Fuel Handling and Storage Facility for Boiling andPressurized Water Reactors." with the exception of the guidelines for theatmospheric dispersion factors (X/Q values). The acceptability of the X/Q values isdetermined under SRP Section 2.3.4.

3. An engineered safety features grade atmosphere clean-up system is required for thespent fuel storage area to reduce the potential radiological consequences.

4. The plant design with regard to spent fuel cask drop accidents is acceptable withoutcalculation of radiological consequences if potential cask drop distances are lessthan 30 feet and appropriate impact limiting devices are employed during caskmovements.

4.1.7.4 Anticipated Transients Without Scram

10 CFR 50.62 defines an anticipated transient without scram (ATWS) as an AOO with asubsequent failure of the RPS to trip the reactor. Since the plant design must satisfy boththe single-failure criterion and GDC 26 (see Section 3.2.6), multiple failures or commonmode failures would have to be present for ATWS to occur. Therefore, an ATWS event isneither an AOO nor a design basis accident. Nonetheless, the diverse scram signals of theRPS must be demonstrated for a suite of AOOs, including loss of feedwater, turbine trip,loss of balance-of-plant (several initiated events possible), and loss-of-offsite-power. Thefailure of the reactor to shut down during certain transients can lead to unacceptable RCSpressures, fuel conditions, and/or containment conditions. The ATWS rule (10 CFR 50.62)requires that certain light water-cooled plants have prescribed systems and equipment thathave been determined to reduce the risks attributable to ATWS events, for each of theNSSS vendor's designs, to an acceptably low level. The rule also requires applicants todemonstrate the adequacy of their plant's prescribed systems and equipment. Forevolutionary plant design such as the B&W mPower reactor, the U.S. NRC developedadditional requirements criteria (i.e., to provide a diverse scram system or to demonstratethat the consequences of ATWS events are acceptable).


10 CFR 50.62(c)(1) requires that "each pressurized water reactor must have equipment fromsensor output to final actuation device, that is diverse from the reactor trip system, toautomatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine tripunder conditions indicative of an ATWS. This equipment must be designed to perform itsfunction in a reliable manner and be independent (from sensor output to the final actuationdevice) from the existing reactor trip system."


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As discussed in Section 3.6.2, the B&W mPower reactor has incorporated several l&Csystems other than the RPS to accommodate the 10 CFR 50.62(c)(1) requirement. SRPacceptance criteria for established for existing PWR designs will be examined forapplicability to the B&W mPower reactor. These include:

A. Provide measures to automatically initiate the auxiliary (or emergency) feedwatersystem and a turbine trip under conditions indicative of an ATWS. This equipmentshall be independent and diverse from the reactor trip system from sensor output tothe final actuation device.

B. Combustion Engineering or Babcock and Wilcox reactors applicants shall haveprovision for a scram system that is independent and diverse from the reactor tripsystem, from sensor output to the points of interruption of power to the control rods.

C. These system and equipment shall be demonstrated to provide reasonableassurance that unacceptable plant conditions do not occur in the event of ananticipated transients.

D. The reactor coolant system (RCS) pressure shall not exceed ASME Service Level Climits (approximately 22 MPa or 3200 psig) containment safety parameters (e.g.,temperature or pressure) should not exceed design limits

4.2 Summary of Chapter 6 Containment Response Applicable to B&W mPower Reactor

The containment encloses the reactor system and is the final barrier against the release ofsignificant amounts of radioactive fission products in the event of an accident. The containmentstructure must be capable of withstanding, without loss of function, the pressure andtemperature conditions resulting from postulated loss of coolant, steam line break, or feedwaterline break accidents. The containment structure must also maintain functional integrity in thelong term following a postulated accident. Specifically, it must remain a low leakage barrieragainst the release of fission products.

The following transient analyses calculations will be performed for the B&W mPower reactor toconfirm that the containment design basis withstands pressure and temperature requirements.

* Containment peak pressure and temperature following secondary system pipe ruptures

* Containment peak pressure and temperature following a LOCA

The containment design basis includes the effects of stored energy in the RCS, decay heatenergy, and energy from other sources such as the secondary system, and, if applicable, metal-water reactions including the recombination of hydrogen and oxygen.


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4.2.1 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures

A break of a main steam or main feedwater line inside the containment causes a rapidpressurization and a temperature increase that could challenge the containment designpressure and the equipment qualification (EQ) limits.


4.2.2 LOCA Mass and Energy Release Calculation

Following a break of the largest attached pipe to the reactor vessel

[CCI perAffidavit 4(a)-(d)]

4.2.3 SRP Acceptance Criteria

Table 5 reflects the applicability of SRP Acceptance Criteria appearing in SRP Section6.2.1, Containment Functional Design, sections to the B&W mPower reactor. Combustiblegas control is addressed in Section 4.3.


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Table 5. B&W mPower Applicability to SRP 6.2.1 Expectations

SRP 6.2.1.1.APWR Dry Containments, Acceptance Criteria B&W mPower Applicability

6.1.a GDC 16 and 50- The peak calculatedcontainment pressure following a loss-ofcoolant accident, or a steam or feedwaterline break, should be less than thecontainment design pressure.*

6.1 .b GDC 38 - The containment pressureshould be reduced to less than 50% ofpeak calculated pressure for the designbasis loss of coolant accident with 24hours after the postulated accident

6.1 .c GDC 38 - The containment pressure forsub-atmospheric containments should bereduced to below atmospheric pressurewithin one hour after the postulatedaccident, and the sub-atmosphericcondition maintained for at least 30 days.

6.1 .d GDC 38 and 50 - For containmentresponse to the loss-of-coolant accident,the analysis should be based on theassumption of loss of off-site power andthe most severe single failure in theemergency power system (e.g., a dieselgenerator failure) the containment heatremoval systems (e.g., a fan, pump, orvalve failure), or the core cooling systems(e.g., a pump or valve failure). Theselection made should result in the highestcalculated containment pressure.

6.1 .e GDC 38 and 50 - For containmentresponse to secondary system piperuptures, the analysis should be based onthe most severe single failure in thecontainment heat removal systems, e.g., afan, pump, or valve failure. The analysisshould also be based on a spectrum ofpipe break sizes and reactor power levels.

* For plants at the construction permit stage of review, the containment design pressure should

provide at least a 10% margin above the accepted peak calculated containment pressure.




SRP 6.2.1.3 M&E for LOCASources of Energy (1111.13.11) B&W rnPower Applicability

(10 CFR Part 50 Appendix K, I.A)

1 Reactor Power - The reactor should be assumed tohave operated continuously at least 1.02 times thelicensed power level; however, a lower core powerlevel - no less than licensed power - could bejustified.

2 Stored Energy in the Core - The steady-statetemperature distribution and stored energy in the fuelshall be calculated for the burn-up that yields thehighest calculated stored energy.

3 Fission Heat - Fission heat shall be calculated usingreactivity and reactor kinetics. Shutdown reactivityresulting from temperatures and voids shall be giventheir minimum plausible values, including allowancefor uncertainties.

4 Decay of Actinides - The heat from the radioactivedecay of actinides, including neptunium andplutonium generated during operation, as well asisotopes of uranium, shall be calculated inaccordance with fuel cycle calculations and knownradioactive properties.

5 Fission Product Decay - The heat generation ratesfrom radioactive decay of fission products shall beassumed to be equal to 1.2 times the values forinfinite operating time in the 1971 ANS Standard.The fraction of the locally generated gamma energythat is deposited in the fuel (including the cladding)may be different from 1.0; the value used shall bejustified by a suitable calculation.

6 Metal - Water Reaction Rate - The rate of energyrelease, hydrogen generation, and cladding oxidationfrom the metal/water reaction shall be calculatedusing the Baker-Just equation. The reaction shall beassumed not to be steam limited.

7 Stored Energy in the Reactor Coolant system metal- Heat transfer from piping, vessel walls, and non-fuel internal hardware shall be taken into account.

8 Stored Energy in the Secondary System - Heattransfer between the primary and secondary systemsin the steam generator shall be taken into account.

9 Fuel Clad Swelling and Rupture - The prediction offuel clad swelling and rupture should not beconsidered.

i

ý (Table 5 continued on next page) [CCI per Affidavit 4(a)-(d)]





SRP 6.2.1.3 M&E for LOCABreak Size and Location (ll.B.2)

Containment design basis calculations should beperformed for a spectrum of possible pipe breakssizes and locations to assure that the worst casehas been identified.

SRP 6.2.1.3 M&E for LOCABlowdown Calculations (ll.B.3.b)

The initial mass of water in the reactor coolantsystem should be based on the reactor coolantsystem volume calculated for the temperature andpressure conditions existing at 102% of full power.

2 Mass release rates should be calculated using amodel that has been demonstrated to beconservative by comparison to experimental data.

3 Calculations of heat transfer from surfacesexposed to the primary coolant should be based onnucleate boiling heat transfer. For surfacesexposed to steam, heat transfer calculations shouldbe based on forced convection.

4 Calculations of heat transfer from the secondarycoolant to the steam generator tubes for PWRsshould be based on natural convection heattransfer for tube surfaces immersed in water andcondensing heat transfer for the tube surfacesexposed to steam.

SRP 6.2.1.3 M&E for LOCAPWR Core Reflood B&W mPower ApplicabilityCalculations (ll.B.3.c)

Following initial blowdown of the RCS, the waterremaining in the RV should be assumed to besaturated.

2 Justification should be provided for the refill period.An acceptable approach is to assume a water levelat the bottom of the active core at the EOB so thereis no refill time.






SRP 6.2.1.3 M&E for LOCAPWR Core Reflood B&W mPower ApplicabilityCalculations (ll.B.3.c)

3 Calculations of the core flooding rate should bebased on the ECCS operating condition thatmaximizes the containment pressure either duringthe core reflood phase or the post-reflood phase.

4 Calculations of liquid entrainment should be basedon PWR FLECHT experiments.

5 Liquid entrainment should be assumed to continueuntil the water level in the core is 2ft from the top ofthe core.

6 The assumption of steam quenching should bejustified by comparison with applicableexperimental data. Liquid entrainment calculationsshould consider the effect on the carryout ratefraction of the increase core inlet watertemperature caused by steam quenching assumedto occur from mixing with the ECCS water.

7 For Cold Leg Breaks only.

Steam leaving the steam generators should beassumed to be superheated to the temperature ofthe secondary coolant.

SRP 6.2.1.3 M&E for LOCASRP .2.13 MEfor OCAB&W mPower ApplicabilityPWR Post-Reflood Calculations (ll.B.3.d)

1 All remaining stored energy in the primary andsecondary systems should be removed during thepost-reflood phase.

2 Steam quenching should be justified bycomparison with applicable experimental data.

3 The results of post-reflood analytical models shouldbe compared to applicable experimental data.



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SRP 6.2.1.3 M&E for LOCA mPower ApplicabilityPWR Decay Heat Phase Calculations (ll.B.3.e) -- _ m_ owerApplicability

1 The dissipation of the core decay heat shouldbe considered during this phase of theaccident. The fission product decay energymodel is acceptable if it is equal to or moreconservative than the decay energy modelgiven in Branch Technical Position ASB 9-2 inSRP §9.2.5.

2 Steam from decay heat boiling in the core should beassumed to flow to the containment by the pathwhich produces the minimum amount of mixing withECCS injection water.

SRP 6.2.1.4, M&E for Secondary System Pipe mPower ApplicabilityRuptures, 11.1, Sources of Energy

The stored energy in the affected steam generatormetal, including the vessel tubing, feedwater line,and steam line.

2 The stored energy in the water contained within theaffected steam generator.

3 The stored energy in the feedwater transferred tothe affected steam generator prior to closure of theisolation valve in the FW line.

4 The stored energy in the steam from the unaffectedsteam generator(s) prior to closure of the isolationvalves in the steam generator crossover lines.

5 The energy transferred from the primary coolant tothe water in the affected steam generator duringblowdown.

6 The SLB should be analyzed for a spectrum of pipesizes and various plant conditions from hot standbyto 102% of full power. Only the 102% powercondition need be analyzed provided the applicantcan demonstrate that the feedwater flows and fluidinventory are greatest at full power.



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SRP 6.2.1.4 M&E for Secondary System PipeRuptures, 11.2, M&E Release Rate Calculations B&W mPower Applicability

1 Mass release rates should be calculated using theMoody model for saturated conditions, or a modelthat is demonstrated to be equally conservative.

2 Calculations of heat transfer to the water in theaffected steam generator should be based onnucleate boiling heat transfer.

3 Calculations of mass release should consider thewater in the affected steam generator and FW line,the FW transferred to the affected steam generatorprior to the closure of the isolation valves in the FWlines, the steam in the affected steam generator,and the steam coming from the unaffected steamgenerator(s) as the secondary system is beingdepressurized prior to the closure of the isolationvalves in the steam generator crossover lines.

4 If liquid entrainment is assumed in the SLB,experimental data should support the predictions ofthe liquid entrainment model.

5 The effect on the entrained liquid of steamseparators located upstream from the break shouldbe taken into account.

6 A spectrum of the steamline breaks should beanalyzed beginning with the double-ended breakand decreasing in area until it has beendemonstrated that the maximum release rate hasbeen considered.

7 A single active failure in the steam or feedwater lineisolation provisions or feedwater pump, such thatthe containment peak pressure and temperature aremaximized, should be assume to occur in steamand feedwater line break analyses. For theassumed failure of a safety grade steam orfeedwater line isolation valve, operation of non-safety grade equipment may be relied upon as abackup to the safety grade equipment.

9 Operator action to terminate auxiliary feedwater flowmust be justified.



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4.3 Combustible Gas Control in Containment

Combustible gases may be generated following an accident as a result of the release ofdissolved hydrogen in primary system fluid, radiolysis of reactor coolant water, corrosion ofaluminum and zinc in the containment. Further, per the 2003 revision of 10 CFR 50.44, theevaluation of combustible gas must also consider sources resulting from the consequence ofsevere accident phenomena such as high temperature reactions between the fuel cladding andcoolant and molten core-concrete interaction. The ignition of the combustible sources couldbreach the containment or damage essential systems and components needed to mitigate anaccident.

The B&W mPower reactor will employ the necessary features to prevent the accumulation of

combustible gas mixtures.



Specific SRP acceptance criteria acceptable to meet the relevant requirements of the U.S.NRC's regulations are all applicable to the B&W mPower reactor. These are as follows:

1. In meeting the requirements of 10 CFR Part 50, § 50.44, and GDC 41 to providesystems to control the concentration of hydrogen in the containment atmosphere,materials within the containment that would yield hydrogen gas due to corrosion fromthe emergency cooling or containment spray solutions should be identified, and theiruse should be limited as much as practicable.

2. In meeting the requirements of 10 CFR Part 50, § 50.44, and GDC 41 to providesystems to control the concentration of hydrogen or oxygen in the containmentatmosphere, the applicant should demonstrate by analysis, for non-inertedcontainments, that the design can safely accommodate hydrogen generated by anequivalent of a 100 percent fuel clad-coolant reaction, while limiting containmenthydrogen concentration, with the hydrogen uniformly distributed, to less than 10percent (by volume), and while maintaining containment structural integrity.

3. In meeting the requirements of 10 CFR Part 50, § 50.44(c)(3), regarding equipmentsurvivability, equipment necessary for achieving and maintaining safe shutdown ofthe plant and maintaining containment structural integrity should perform its safetyfunction during and after being exposed to the environmental conditions attendantwith the release of hydrogen generated by the equivalent of a 100 percent fuel clad-coolant reaction, including the environmental conditions created by activation of thecombustible gas control system.

4. In meeting the requirements of 10 CFR Part 50, § 50.44, to provide the capability forensuring a mixed atmosphere in the containment during design bases and the more


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likely beyond-design-bases accidents, and of GDC 41 to provide systems asnecessary to ensure that containment integrity is maintained, this capability may beprovided by an active, passive, or combination system. Active systems may consistof a fan, a fan cooler, or containment spray. For passive or combination systems thatuse convective mixing to mix the combustible gases, the containment internalstructures should have design features which promote the free circulation of theatmosphere. For all containment types, an analysis of the effectiveness of themethod used for providing a mixed atmosphere should be provided. This analysis isacceptable if it shows that combustible gases will not accumulate within acompartment or cubicle to form a combustible or detonable mixture that could causeloss of containment integrity.

5. Atmosphere mixing systems prevent local accumulation of combustible or detonablegases which could threaten containment integrity or equipment operating in a localcompartment. Active systems installed to mitigate this threat should be reliable,redundant, single-failure proof, able to be tested and inspected, and remain operablewith a loss of onsite or offsite power.

6. In meeting the requirements of 10 CFR Part 50, § 50.44, and GDC 41 regarding thefunctional capability of the combustible gas control systems to ensure thatcontainment integrity is maintained, the design should meet the provisions of RG 1.7,Section C.1, discussing equipment survivability.

7. To satisfy the design requirements of GDC 41:

a. Performance tests should be performed on system components, such ashydrogen igniters and combustible gas monitors. The tests should support theanalyses of the functional capability of the equipment.

b. Combustible gas control system designs should include instrumentation neededto monitor system or component performance under normal and accidentconditions. The instrumentation should be capable of determining that a systemis performing its intended function, or that a system train or component ismalfunctioning and should be isolated. The instrumentation should have readoutand alarm capability in the control room. The containment hydrogen and oxygenmonitors should meet the provisions of RG 1.7, Section C.2.

8. To satisfy the inspection and test requirements of GDC 41, 42, and 43, combustiblegas control systems should be designed with provisions for periodic in-serviceinspection, operability testing, and leak rate testing of the systems or components.

9. In meeting the requirements of 10 CFR Part 50, § 50.44(c)(5), regarding containmentstructural integrity, an analysis must demonstrate containment structural integrity,using an analytical technique that is accepted by the U.S. NRC staff and includingsufficient supporting justification to show that the technique describes the


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containment response to the structural loads involved. The analysis must address anaccident that releases hydrogen generated from 100 percent fuel clad-coolantreaction accompanied by combustible gas burning. Systems necessary to ensurecontainment integrity must also demonstrate the capability to perform their functionsunder these conditions. One acceptable analytical technique is a demonstration thatspecific criteria of the ASME Boiler and Pressure Vessel Code, described in RG 1.7,Section C.5, are met.

10. In meeting the requirements of 10 CFR Part 50, § 50.44(c), and GDC 41 for thedesign and functional capability of the combustible gas control systems, preliminarysystem designs and statements of intent in the safety analysis report are acceptableat the construction permit stage of review if the guidelines of RG 1.7 are endorsed.


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5. CHARACTERIZED MODELING DOMAIN

A considerable amount of experience related to design basis LWR accident analysis exists fromsmall- and large-scale experimental programs, theoretical evaluations, and computer codedevelopment. With this knowledge base the modeling domain requirements can becharacterized with experiential detail. A preferred approach (Reference 30) involves organizingthis information through a hierarchical or "top-down" decomposition of the design and eventdescriptions. With the top-down approach, the evaluation of the modeling domain considers theperspective of the variables affecting the principal figures-of-merit.

The initial motivation for this exercise is to inform the phenomena identification and rankingtable review process (PIRT, see Section 6). At a high level this characterization can be used todetermine essential features of the analytical tools that will be incorporated within the evaluationmethodology. Further, it can be used as an initial screening of competing candidate analyticaltools.

5.1 Hierarchical Decomposition

The "top-down" approach for characterizing the modeling domain focuses on the capabilitiesand performance requirements of the analysis tools to be employed, specifically, three issuesrelated to evaluation methodology applicability: model, nodalization, and scale:

* Model applicability: The analysis tools must be capable of modeling the key phenomena

in the system components and subsystems.

* Nodalization applicability: The resolution of an analysis tool must align with the design.

* Scale applicability: Model and nodalization applicability are insensitive to the problemscale.

In preparation of the PIRT exercise, a list of SSC are identified by the several phenomena thatcould theoretical appear during one or more events. A common approach to developing that listbegins with a hierarchical decomposition from a "system level", which may actually encompassa single component or multiple systems. Following from the system level, subsystems,modules, constituents, phases, geometry, fields, and transport processes defines the genericevent hierarchy. This process is illustrated in the next two subsections for a generic eventimpacting (1) the RCS and (2) the containment.

5.1.1 Reactor Coolant System Response


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To inform the PIRT process, and thus begin the process of identifying evaluation modelcharacteristics, it is useful to present the key phenomena and processes within ahierarchical description.

Figure 7 provides a representative hierarchical model addressing the RCS response to ageneric event in a B&W mPower plant. Core cooling is the focus which begins withidentifying the primary and secondary coolant systems as subsystems to the "B&W mPowerSystem". The major components, pressurizer, core, OTSG, feedwater, and steamline, areidentified as modules. The constituent materials of each module are water, fuel, and steel.The fundamental phases are liquid, vapor, and solid. Each phase can exist in severalgeometric forms, such as ambient single-phase or stratified two-phase fluid or tube/rodstructures. Associated with each geometrical form are fields that describe the distribution ofmass, momentum, and energy.



Figure 7. Hierarchical RCS Response Model for a Generic B&W mPower Plant Event


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5.1.2 Containment Response

Containment response to a breach in the RCS may be characterized by two distinct phases.The first one (blowdown) is an injection of hot steam corresponding to RCSdepressurization. The high rate of steam injection during this period increases containmenttemperature and pressure and quickly disperses a steam and air mixture throughout thecontainment. With cold walls and other structure surfaces, condensation is the mostimportant heat transfer mechanisms. Natural convection and conduction-limited heattransfer to the UHS distinguishes the second phase. Figure 8 provides a simple illustrationof the expected heat and mass transfer mechanisms defining the B&W mPowercontainment response dynamics.


Figure 9 provides a representative model addressing the containment response to a genericevent in a dry containment relying on passive heat removal (Reference 31), like the B&WmPower reactor. For in-containment atmospheric mixing, individual enclosures, andinterconnecting channels between enclosures can be grouped as subsystems. The largeenclosures are divided into pool, air space, and structure modules. The constituent



materials of each module are water, non-condensable gases, steel, and concrete. Thefundamental phases are liquid, vapor, and solid. Each phase can exist in several geometricforms, such as droplets, films, bubbles, jets, etc. Associated with each geometrical form arefields that describe the distribution of mass, momentum, and energy.

E

I[CCI per Affidavit 4(a)-(d)]

Figure 8. B&W mPower Plant Containment Heat and Mass Transfer Mechanisms


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Hierachy

*From Peterson, P. F., "Scaling and Analysis of Mixing in Large Stratified Volumes,"Int. J. Heat Mass Transfer, 37(Suppl. 1), 97-106, 1994.(Aerosol contribution neglected as a simplification)

Figure 9. Hierarchical Containment Response Model for a Generic B&W mPower PlantEvent


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5.2 Essential Features for Analytical Modeling

While the completeness of any analysis tool depends on the specific plant and scenario ofinterest, some general code features characterize the current state-of-the-art in large RCS andcontainment analysis codes. The essential code features supporting LWR evaluationmethodologies represent a selection of notable code capabilities inherent in existing U.S. NRC-approved evaluation methodologies currently supporting the operating fleet of NPPs in the U.S.

(1) Multi-Dimensional Capability. Resolution of core flow distribution is an explicit requirementin 10 CFR 50 Appendix K. While "pseudo" multi-dimensional modeling, incorporating a cross-flow momentum approximation, have appeared in previous evaluation models, these evaluationmethodologies have all since been superseded by full multi-dimensional hydrodynamic models.Full multi-dimensional hydrodynamic field equations, accommodating Cartesian and cylindricalcoordinate systems, are considered essential for simulating the asymmetric flow conditions inthe reactor vessel that are characteristic in many reactor accident and transient scenarios. Withfew exceptions, coolant system performance codes supporting the U.S. installed base of NPPsincorporate multi-dimensional hydrodynamic models.

(2) Energy Equations. The energy equation formulation appearing in all U.S. NRC and U.S.Department of Energy (DOE) sponsored thermal-hydraulic codes are approximated usinginternal energy rather than total energy. This formulation can produce energy error when asizeable pressure gradient exists between two adjacent control volumes. This deficiency is adirect consequence of ignoring specific energy terms that are difficult to approximatenumerically. This deficiency has been recognized by the safety analyst community and manycodes have either retrofitted a special user option for convecting the total energy across aparticular user-specified junction or completely reformulated the energy equation to a totalenergy equation. Analyses involving containment modeling are more numerically appropriatewith either approach.

(3) Numerical Solution of Hydrodynamic Field Equations. While once representing thestate-of-the-art, the earliest generation of sparse matrix solvers, those originally appearing inlarge thermal-hydraulic systems codes, are comparatively inefficient to those benefiting from thegrowing experience in numerical methods. Along with advances in numerical modeling, modernsolver algorithms have been designed to optimize changes in computer architecture (e.g.,pipelines, vector hardware, shared-memory parallel architecture, etc.), further diminishing thevalue of legacy matrix solvers. Modern solvers make multi-dimensional hydrodynamic modelingpractical while providing significant speed up for one-dimensional models. Significant speed upis also possible through greater implicitness in the numerical model formulation (although,sometimes at the cost of accuracy).

(4) Hydrodynamic Constitutive Models. Evaluation methodology emphasis on phenomenaidentification, ranking, and a subsequent "bottom-up" approach to analysis has focusedattention on the completeness of physical models and correlations and the treatment of modelparameters in the execution of analysis. While new physics can be added directly to a systemscode, the complexity of mature thermal-hydraulic codes has made this a particularly challengingactivity. As such, interest in unified, multi-physics solution through the coupling of


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complementary physics codes has become more common, employing either the Parallel VirtualMachine (Reference 32) or Message Passing Interface (Reference 33) subroutine libraries.

(5) Heat Transfer Uncertainty Model. Heat transfer represents the largest source ofuncertainty in NPP safety analysis. Large thermal-hydraulic system codes model thepredominant heat transfer regimes based on fluid and wall conditions, including specialtreatment for core reflooding scenarios. Uncertainty treatments often require either a hardwiredbias or user-specified bias through code input. The latter is preferred.

(6) Fuel Performance Model. Since fuel design has a strong bearing on both NPP economicsand safety, safety analyses are expected to explicitly evaluate the impact of unique fuel productcharacteristics with the initial loading of fuel and subsequently with each follow-on reload. Eventscenarios are evaluated at various times in the operating life of the fuel product. In general, thecoolant system performance codes supporting the U.S. installed base of NPPs intrinsicallymodel fuel behavior both using time-in-life (or cycle) initial conditions and propagating theunique fuel condition through the conditions specific to a given design basis event.

(7) Containment Model. With the increasing complexity required of thermal-hydraulic codes,the challenge of maintaining a suite of computer codes can be significant and costly. As such,coolant system performance codes currently supporting the U.S. installed base of NPPsincorporate a containment response model either through an intrinsic physics package orthrough a loosely coupled framework. This tool is then used for LOCAs to support both ECCSperformance analyses (SRP 15.6) and containment response (SRP 6.2).

(8) Multi-dimensional reactor kinetics. General Design Criterion 28 on reactivity limitsrequires that "the reactivity control systems shall be designed with appropriate limits on thepotential amount and rate of reactivity increase". GDC 28 is clarified in Regulatory Guide 1.177and SRP 15.4.8 with specific limits on fuel enthalpy rise, RCS pressure rise, and fuel rodradionuclide inventory. As a consequence of the demand for longer fuel cycles (i.e., higherburnup), research reported by the U.S. NRC in 1998 (References 34 - 37) has suggested thatcurrent regulatory limits are not satisfactory and that the best solution is to have nuclear fuelsuppliers move to multi-dimensional reactor kinetics models. SRP Section 4.2, Appendix Bexpresses an IAC impacting both RIA and radiological assessment methodologies. As aconsequence nuclear fuel suppliers supporting the U.S. installed base of NPPs have developednew evaluation models using multi-dimensional reactor kinetics in either a fully- or loosely-coupled manner.

(9) FORTRAN 90. Like computer codes, compilers evolve with the changing computinglandscape. The original thermal-hydraulic system codes were developed using FORTRAN 66and have migrated over time to FORTRAN 77. Changes in computing architecture andprogramming trends (e.g., object-orient programming) inspired new FORTRAN releases in1990, 1995, and 2000. FORTRAN 90 represented a major restructuring of the FORTRANlanguage. As such, support for FORTRAN 77 compilers is vanishing. Both the U.S. NRC'sTRACE and U.S. DOE's RELAP5-3D codes have moved to FORTRAN 90 and the U.S. plantand fuel suppliers have initiated projects to begin moving to FORTRAN 90.


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5.3 Computer Code Selection for B&W mPower Accident Analysis

Because full-scope analysis of a NPP relies upon the examination of many diverse and complexphenomena that behave in widely varying time domains, multiple computer codes are employedfor safety analysis that focus on an optimal balance of physics and specific system behaviors.While the more widely used RCS performance analysis codes evolved to emulate much of theexpected and more important thermal-hydraulic phenomena for RCS performance analysis (i.e.,multi-phase hydraulics, heat conduction, control systems, noncondensable gases, etc.), reactorphysics, fuel performance, and containment response codes are often integrated in evaluationmethodologies to adequately resolve event-specific uncertainties when related phenomena areidentified as being important.

Many computer codes are available from various industry and government organizations. Thefollowing subsections provide a brief description of the full suite of computer codes that will beused to support the B&W mPower reactor transient and accident analyses. Figure 10 presentsa general flow diagram of the safety analysis code suite adopted for the B&W mPower reactor.RELAP5-3D is the principal analytical tool, either receiving or providing information forcalculating acceptance criteria measures.

Figure 10. Analysis Code Suite Supporting the B&W mPower Reactor





5.3.1 Reactor Coolant System Performance - RELAP5-3D

RELAP5-3D broadly accommodates the essential features for analytical modeling appearingin Section 5.2. For this reason B&W NE plans to use RELAP5-3D (Reference 38) for thesimulation of all B&W mPower reactor postulated AQOs and accidents of Chapter 15 and toprovide mass and energy releases for Chapter 6 containment response analyses. Themodel will include all major systems and components that are required to adequatelysimulate the plant response.

Several derivative versions of RELAP5 are available (e.g., Reference 39). All versionsemploy a non-homogeneous and non-equilibrium model for the two-phase system, solvedby a fast, partially implicit numerical scheme. Generic models facilitate the simulation ofgeneral thermal-hydraulic systems. These include pumps, valves, and pipes, heat releasingor absorbing structures, reactor point kinetics, electric heaters, jet pumps, turbines,separators, accumulators, and control system logic elements. Further, code developmenthas benefited from extensive application and validation against a broad database ofexperimental programs.

The U.S. DOE is the principal sponsor of RELAP5-3D. Beyond multi-dimensionalhydrodynamics, RELAP5-3D features multidimensional heat transfer capability (radiation,convection, and conduction), three-dimensional reactor kinetics, plus other generic andspecial component modeling for describing the behavior of complex fluid-filled systemsunder single- and two-phase conditions with accommodation for non-condensable gasesand boron.

5.3.2 Fuel Thermal-Hydraulics- VIPRE-01

For B&W mPower reactor safety analysis, VIPRE-01 (Reference 41) will be used to quantifythe design basis margins to the SAFDL related to DNBR, fuel temperature, and core flowand enthalpy distribution. VIPRE-01 is a subchannel thermal-hydraulic computer codedeveloped by Battelle Pacific Northwest Laboratories under the sponsorship of EPRI. Giventhe geometry of the reactor core and coolant channel, and the boundary conditions orforcing functions, VIPRE-01 simulates the physical processes and phenomena whilecalculating the measures relating to safety criteria. VIPRE-01 developed as an outgrowth ofthe COBRA series of codes (Reference 40), combining features of several of the differentCOBRA variations into a single computer code.

As a subchannel thermal-hydraulics code, VIPRE-01 is a quasi-1 D code that models flow inparallel channels with cross flow and turbulent mixing between the channels. IndividualVIPRE-01 channels can be discrete fuel element subchannels, internal flow tubes, orlumped channels consisting of several channels modeled together. A typical VIPRE-01model could include several different types of channels. The general flow model assumesincompressible and homogeneous fluid conditions, but it also includes special processmodels and correlations to account for subcooled boiling and two-phase flow with vapor slip.In each axial segment of the model, VIPRE-01 solves the conservation equations for mass,momentum (axial and lateral), and energy. VIPRE-01 also includes conduction models for


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the prediction of fuel pin centerline temperatures. A proven gap closure model is alsoavailable to allow internal calculation of fuel/clad gap conductivity.

Input to the VIPRE-01 for subchannel analysis consists of assembly geometry and hydraulicdescriptive information, including information regarding which assemblies are adjacent toeach other. A RELAP5-3D calculation provides the fuel assembly thermal-hydrauliccondition, including, for example, coarse core distributions, power, temperatures, andpressures.

5.3.3 Fuel and Containment Analysis Codes

For design basis safety analysis, fuel mechanical performance and containment responsecapability are effectively boundary conditions. While high fidelity modeling is becomingmore common, particularly in applications requiring improved design and safety margins,more simplistic models, incorporating conservatism, are still prevalent in modern evaluationmethodologies. When simple models are found to be adequate, fuel and containmentmodels may be represented by static boundary conditions, that is, with code input derivedindependently from the RCS analysis. The preferable approach will be dependent on theparticular application requirements.

Regarding fuel mechanical performance, the transient fuel behavior codes include models,correlations and properties for various phenomena of interest during the transients andaccidents such as stored energy, cladding plastic behavior, phase changes, and largecladding deformations (ballooning), fission gas release and subsequent fuel pinpressurization. Modern transient codes for fuel behavior, such as EPRI's FALCON(Reference 42) and the U.S. NRC-sponsored FRAPCON and FRAPTRAN (References 43and 44), are highly reputable for analyzing fuel response during reactivity induced accidents(RIAs) and LOCAs.

The containment thermal-hydraulic codes used for design basis safety analysis have beentraditionally developed as the lumped parameter codes that model thermal-hydraulicbehavior of the containment compartments and heat structures. Containment analysis codesgenerally solve for the transport of steam and liquid appearing in the containment following abreach of the RCS and heat transfer by condensation and conduction through structure.The application of containment codes for design basis analysis address:

* maximum pressure and temperature analysis during the LOCA and steam line breakaccidents,

* minimum back pressure analysis during a LOCA as a boundary condition for ECCSperformance analysis (i.e., core reflood), and

* analysis of differential pressures on the containment internal structures during LOCAand steam line break as a basis for estimating loads on containment internals.

The U.S. NRC (and its predecessor, the Atomic Energy Commission) has sponsored severalcontainment analysis codes including CONTEMPT (Reference 45) and MELCOR


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(Reference 48). These are all considered lump parameter codes. The limitations of thelumped parameter codes have initiated many attempts to develop a more detailed codesystem for the fluid flows inside the containment. The improvement in the computer capacityand speed has made it feasible to apply meso-scale computational fluid dynamicsmethodology to the containment calculations, which has been successfully realized with theEPRI-sponsored GOTHIC code that has both lump- and distributed-parameter modelingcapability.

5.3.3.1 FALCON - Fuel Mechanical

FALCON is a two-dimensional transient fuel behavior program developed to analyze theresponse of LWR fuel rods during RIA, LOCA and other transient conditions FALCONutilizes a coupled thermal and mechanical finite element methodology to represent thetransient behavior of the fuel column, cladding, and gap. It's fuel pellet mechanicalconstitutive model includes pellet cracking, creep, plasticity, and thermal expansion. Thethermal and mechanical properties of the fuel and cladding also consider the effects ofburnup and fast neutron fluence.

FALCON is the primary fuel design tool for B&W mPower reactor cores consisting ofgadolinium-enriched U0 2 fuels. It can be run in a best-estimate or bounding manner.FALCON is used either to directly evaluate compliance or to provide initialization input toother computer codes for the following figures-of-merit (some are required only as part ofanalyses prepared for SRP Chapter 4):

" DNBR

" fuel rod internal pressure

* fuel melt temperature

" cladding strain

* cladding creep collapse

* cladding peak oxide thickness

The U.S. NRC's FRAPCON and FRAPTRAN code will be used in a self-audit capacity toinform the development and assessment of evaluation methodology adequacy.

5.3.3.2 GOTHIC - Containment

GOTHIC (Reference 46), provided by Numerical Applications underthe sponsorship ofEPRI, is a general purpose thermal-hydraulics software package for design, licensing, safetyand operating analysis of NPP containments and other buildings. Applications of GOTHICinclude evaluation of containment and containment sub-compartment response to the fullspectrum of high energy line breaks. Applications may include, but are not limited to,pressure and temperature determination, equipment qualification (EQ) profiles andinadvertent system initiation, and degradation or failure of engineered safety features.


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GOTHIC has been developed and maintained in a rigorous QA program that has receivedregular audits for compliance with 10 CFR 50, Appendix B and 10 CFR 21.

GOTHIC is a state-of-the-art program that solves the conservation equations for mass,momentum and energy for multi-component, multi-phase flow. Conservation equations aresolved for up to three primary fields:

* steam/gas mixture

* continuous liquid

* liquid droplet

And up to two secondary fields:

* mist

* liquid components

Detailed description of the code can be found in Reference 46.

B&W NE intends to use GOTHIC for all design basis containment analyses. This includespeak containment analysis following a design basis LOCA and steamline breaks (SLB) toverify the acceptance of peak pressures. Another application will be for peak temperaturecalculations inside the containment following the design basis LOCA and SLB. The codemay also be used to perform design calculations and sensitivity studies for the ECCS,RWST, containment cooling, ECCS condensers, and the UHS performance, design andlicensing.

B&W NE will demonstrate that GOTHIC is capable of the simulation of similar transients andaccidents for the B&W mPower reactor. The physical phenomena expected to occur in thecontainment, the containment sub-compartments, and in the auxiliary and turbine buildingsare the same as for the currently operating plants or newly approved passive PWRsdesigns. While the GOTHIC code is commonly used as a containment analysis tool, itemploys features to extends its capability to support RCS performance simulation, such asmulti-dimensional two-phase flow modeling, control and trip systems, valves, materialproperties modeling, etc. The fast running and integral (primary and containment) featuresof GOTHIC will also be used for the complementary sensitivity and uncertainty analysisexpected in an EMDAP-based evaluation methodology.

5.3.3.3 MAAP5 - Combustible Gas

MAAP5 is an integral systems analysis computer code for assessing a wide range of integralplant evaluations including severe accidents, PRA, and design basis events. Fauske andAssociates, Inc. develops and maintains MAAP5 under the sponsorship of the EPRI(Reference 47). MAAP5 includes models required for analysis of a broad spectrum ofthermal-hydraulic and severe accident phenomena that might occur within the primarysystem and containment. It has the capability to simulate the progression of an accident





from the initiating event through the onset of core damage and containment failure. MAAP5builds on the substantial modeling enhancements over MAAP4 that include PWR RCSmodel improvement, containment model improvements which can perform design basisaccidents analyses with either a single node or multi-node containment model, coupled dosecalculation, 1-D neutronic and a point kinetic models, and many other improvements.

As with previous MAAP versions, MAAP5 includes models for the generation, transport,removal (i.e., igniters or PARs) and burning of combustible gases. Per 10 CFR 50.44,ignition prevention of a combustible gas must consider the release of hydrogen generatedfrom 100 percent oxidation of fuel cladding surrounding the active fuel, as might occur in asevere accident. For beyond design basis accidents, simulation of the realistic rateprocesses from the "more likely" event scenarios is acceptable per RG 1.216.

MAAP5 will be used to simulate severe accident sequences in the B&W mPower reactorthat result in combustion. The principal criteria of interest is the 10 CFR 50.44 requirementthat power plant designs be capable of mitigating the combustible gas threat from a Zr-waterreaction of 100 percent of the cladding surrounding the active fuel.

5.3.4 Radiological Response Codes

5.3.4.1 MELCOR

MELCOR (Reference 48) is a fully integrated, engineering-level computer code that modelsthe progression of LWR design basis and severe accidents in NPPs. MELCOR treats abroad spectrum of thermal-hydraulic and severe accident phenomena in both boiling andpressurized water reactors. These include: thermal-hydraulic response in the RCS;containment and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product releaseand transport behavior. For the B&W mPower reactor applications, MELCOR will be appliedin the estimation of accident source terms and radiological transport.

5.3.4.2 LOCADOSE

LOCADOSE is a multi-region radiation transport and dose calculation program developed byBechtel Corporation (Reference 48). The program calculates activities, activity releases,dose rates, and doses as a function of time for each region, including a special control roomregion. Dose rates and doses are also calculated at offsite locations. An unlimited numberof time steps are available to model time-dependent parameters. Among the inputparameters that the user can specify are the isotopes to consider, flow rates betweenregions, spray and deposition removal terms, filter efficiencies, partition factors, and thedose conversion factors. The program includes an extensive library of isotopic data.LOCADOSE can be used to model any accident scenario and calculate the radiologicalconsequences. It automatically calculates the maximum two-hour dose at the exclusionarea boundary.


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5.3.4.3 PWR-GALE

The PWR-GALE code is a computerized mathematical model for calculating the releases ofradioactive material in gaseous and liquid effluents from PWRs (Reference 50). Thecalculations are based on data generated from operating reactors, field and laboratory tests,and plant specific design considerations incorporated to reduce the quantity of radioactivematerials that may be released to the environment during normal operation, including AQOs.RCS and secondary system source terms represent radioactive liquid and gaseous materialthat may be transported or released to the environment by radioactive waste systems.These liquid and gaseous effluents are determined using PWR-GALE.

5.3.4.4 ARCON96

ARCON96 is a model for calculating relative concentrations (or atmospheric dispersionfactors, x/Q) in the vicinity of buildings (Reference 51). It is applied in the assessment ofNPP control room habitability under certain accident conditions. The calculation considersthe hourly averaged meteorological data and recently developed methods for estimatingdispersion in the vicinity of buildings to calculate x/Q values at control room air intakes thatwould be exceeded no more than 5% of the time. Dispersion factors can be determined foraveraging periods ranging from 1 hour to 30 days duration to support control roomhabitability dose calculations to ensure compliance with GDC 19. The code implements astraight-line Gaussian dispersion model with dispersion coefficients that are modified toaccount for low wind meander and building wake effects.

5.3.4.5 ORIGEN-S

ORIGEN-S computes time-dependent concentrations and radiation source terms of a largenumber of isotopes, which are simultaneously generated or depleted through neutronictransmutation, fission, and radioactive decay (Reference 52). Provisions are available tosimulate input feed rates, and physical or chemical removal rates from a system. Thecalculations may pertain to fuel irradiation within a nuclear reactor, or the storage,management, transportation, or subsequent chemical processing of spent fuel elements.Applications of ORIGEN-S include nuclear reactor and processing plant design studies,design studies for spent fuel transportation and storage, burnup credit evaluations, decayheat and radiation safety analyses, and environmental assessments. In addition, the codemodels the integration of actinide or fission product decay energies and radiation sourcesover any decay interval.

ORIGEN-S is the depletion and decay module in the SCALE code system. It may be calledfrom a control module, or it may be run as a stand-alone program. The primary objective inthe design of ORIGEN-S is to make it possible for the depletion calculations to utilize multi-energy-group cross sections processed from any standard ENDF/B formatted nuclear datalibrary. This function has been implemented through the execution of physics codes withineither the SCALE system or the AMPX system, both developed and maintained at OakRidge National Laboratory (ORNL). These codes compute problem-dependent neutron-spectrum-weighted cross sections that are representative of conditions within any given


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reactor fuel assembly, and convert the data into a cross-section library format that can beused by ORIGEN-S. Time-dependent cross-section libraries may be produced that reflectfuel composition variations during irradiation.

5.3.5 Reactor Physics Codes

Reactor physics phenomena can have a strong influence on total fission power and powerdistribution, stored energy, and decay heat. The preparation of design basis safety analysesrely on several neutronics parameters, including delayed neutron fraction, prompt neutronlifetime ratio, U-238 capture-to-fission ratio, and moderator density and fuel temperaturereactivity. The reactor physics response is most pronounced for the class of reactivityaccidents that include control rod withdrawal, control rod ejection (not applicable for theB&W mPower reactor, see Section 4.1.4.1), loss of primary flow, RCP startup, and ATWS.

The traditional role of reactor physics for design basis safety analysis has been limited tovery simplistic models, such as using a point-kinetics model, with parameters set at eithercycle or lifetime limiting values that maximize core region power. As discussed in Section2.3.2.2, there has been a considerable amount of regulatory debate that has resulted instricter acceptance criteria. In an effort to remove conservatisms associated with the point-kinetics model and subsequently offer opportunities to improve the efficiency and economicsof nuclear power, there has been a gradual shift in the nuclear industry towards high fidelity,multidimensional reactor physics simulations in nuclear reactor safety analysis. This hasbeen made possible through considerable improvements in numerical methods.

As with fuel and containment performance analysis, reactor physics response can beperformed standalone. Nonetheless, considerable investment has been made to couplemulti-dimensional reactor physics codes with advanced thermal-hydraulic system codes.While NESTLE (Reference 54) is fully integrated within the RELAP5-3D code, SIMULATE-3K, a product from Studsvik Scandpower, has been coupled with several derivative versionsof RELAP5 (References 55 and 56) and will be used in analysis of the B&W mPowerreactor. B&W NE submitted to the U.S. NRC in August 2010, a Topical Report: CoreNuclear Design Codes and Methods Qualification (Reference 57) that provided a descriptionof the plans to use both SIMULATE-3 and CASMO-5 (Reference 58, justification for the useof these codes, and also provided benchmarking information for the latter code.

5.3.5.1 CASMO-5

CASMO-5 is a multi-group two-dimensional transport theory code for burnup calculations forLWR assemblies or pin cells). It models a geometry consisting of cylindrical fuel rods ofvarying composition in a square pitch array with allowance for fuel rods loaded withgadolinium, erbium, integral fuel burnable absorber pellets, burnable absorber rods, clustercontrol rods, incore instrument channels, water gaps, and cruciform control rods in theregions separating fuel assemblies. Reflector/baffle analysis can also be performed.CASMO-5 incorporates the direct microscopic depletion of burnable absorbers such asgadolinia and erbium into the main calculation and the two-dimensional transport calculationconsiders a fully heterogeneous model.


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The CASMO-5 output generates edits for the eigenvalue, the power distribution, reactionrates and few-group parameters for use in core calculations. The output also contains fluxdiscontinuity factors for assembly interfaces and reflector regions. These discontinuityfactors can be used by SIMULATE-3 in two or multi-group diffusion theory in order topreserve net currents calculated by the CASMO-5 multigroup transport solution.

CASMO-5 has been rigorously benchmarked by the code vendor (Studsvik Scandpower)against measured critical experiments, post-irradiation benchmarks, and continuous-energyMonte Carlo calculations. These tests have demonstrated excellent agreement with nosignificant bias versus the number of gadolinia pins, number of Ag-In-Cd rods, boronconcentration, geometry, or presence of reflector/baffle. The overall accuracy of CASMO-5and its associated ENDF/B-VII neutron data library have been repeatedly validated,ensuring reliably accurate results regardless of core type, fuel type, or operating strategy.

5.3.5.2 SIMULATE-3

SIMULATE-3 is a three-dimensional, two-group, steady-state reactor core simulator thatperforms incore fuel management studies, core design calculations, and calculation ofsafety parameters (Reference 55). SIMULATE-3 employs an advanced nodal expansionmethod to solve the two-group neutron diffusion theory representation of the reactor corewithout requiring normalization to fine-mesh calculations or to measured data. SIMULATE-3provides for thermal-hydraulic feedback, modeling of equilibrium or time-dependent xenon andsamarium, and isotopic depletion. In addition, it allows for the generation of pin-by-pin powerdistributions using a pin power reconstruction technique.

The SIMULATE-3 solution methodology involves subdividing the spatial domain of thereactor into a set of rectangular parallelepiped nodes. The three-dimensional diffusionequation is integrated over the volume of each node to obtain the neutron balance equation.

The thermal-hydraulic model used in SIMULATE-3 for PWRs is a simple heat balance modelthat assumes: (1) known coolant inlet flow and temperature distributions; (2) coolant flow isin parallel channels, with no cross flow, and subcooled core exit water condition; (3)deposition of the power produced by fuel rods within a node is in the coolant associated withthat node; and (4) a negligible pressure drop across the core, and (5) all water propertiesevaluated at a single pressure. These assumptions imply that the coolant enthalpydistribution can be calculated by a simple heat balance of the enthalpy at the inlet of a node,the heat generated within the node, and the enthalpy at the outlet of a node.

5.3.6 Software Control

As described in Section 2.2, B&W NE's Quality Assurance Program follows the core set ofquality assurance requirements for the development and application of safety analysissoftware contained in Subpart 2.7 of ASME NQA-1 -1994 (Reference 4). This standardclarifies the regulatory expectation for codes developed for safety analysis, includingverification and validation, the computer system on which codes may be installed and run,and the system used to control user access. Use of this document ensures that each code


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MPWR-TECR- Redacted Version005013(N P)

is updated and verified on approved computer systems, that only approved users haveaccess to the codes, and that all code users comply with established software controlmeasures established. B&W NE is committed to following this guidance in its computerbased design analysis activities for the B&W mPower reactor project.





6. PHENOMENA IDENTIFICATION AND RANKING TABLE

Phenomena identification and ranking table (PIRT) is a tool to provide guidance throughout theevaluation methodology development process. In particular, it is used to determine therequirements for physical model development, scalability, validation, and sensitivities anduncertainty studies. In particular, a PIRT provides the basis for:

* determining the adequacy of analysis tools (i.e., does the code properly model theimportant phenomena)

* establishing the assessment matrix (i.e., identifying test data that contain the appropriatephenomena during each accident phase)

* identifying phenomenological parameters that can be quantified for evaluating andpropagating uncertainties.

A detailed description of PIRT techniques is described in Reference 59.

PIRT development is a formal process that facilitates the structured collection anddocumentation of informed (expert) judgment with respect to phenomena and their importanceto a particular activity or event. The process focuses on a particular transient event (accident orotherwise), breaking the event into phases, establishing the figures-of-merit for each phase,identifying the phenomena for each system and component involved in the transient, andranking the importance of the phenomena relative to the figures-of-merit in each phase. Thestate of knowledge is also assessed for each phenomenon identified and an overall riskassessment made based on the importance and the state of knowledge.

For the B&W mPower reactor design and its counterpart IST facility, the primary issue is toassure that a sufficient experimental and analytical database exists to support the design andlicensing process. That is, the experimental and analytical database development must beplanned to address design behavioral questions in a hierarchical sequence in which the plantresponses that are postulated to be of the highest safety significance are explored first. In thiscontext, "safety significance" denotes the combination of how influential a behavior may be on asuccessful mitigation of an accident scenario and how well that behavior may be predicted viaexperimental data and/or analytical modeling. From this perspective, it naturally follows thatbehaviors of "highest" safety significance are those that have the most influence on the plantresponse and are the least well understood and/or predicted with the current state ofknowledge. The range of decreasing safety significance determinations progresses to the "mostinfluential behaviors that are moderately well-predicted", "moderately influential behaviors thatare least well-predicted", "moderately influential behaviors that are moderately well-predicted",etc.

6.1 PIRT Obiectives

For technical issues to be addressed supporting evaluation methodology development, theB&W mPower reactor PIRTs have two primary and one secondary objective.


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* To provide expert elicitation of the relative importance of phenomena occurring from theaccident or transient scenarios appearing in NUREG-0800 Chapter 15 and Section 6.2.(primary objective)

* To identify the "state of knowledge" relative to phenomena in the context of adequate,marginal and inadequate experimental databases and analytical tools with which toconduct safety analyses. (primary objective)

* To provide the framework whereby the phenomena relative importance and state ofknowledge (SOK) assessments can be used to establish "risk perspectives" that willguide continued experimental database and analysis tool development in a cost effectivemanner. (secondary objective)

In this context, risk perspectives relate to the uncertainty in understanding the combined relativeimportance and SOK of a phenomenon. For example, a dominant phenomenon during aspecific postulated accident/transient scenario that has a large or undetermined SOK poses thehighest level of risk in understanding the safety significance of that phenomenon. Further, aphenomenon having a high relative importance and moderate SOK poses the next relative levelof risk, and so on through decreasing levels of risk).

6.2 Database Supportingq PIRT Development

Several PIRTs examining conventional PWRs and design basis events have been performedpreviously for both LOCAs and Non-LOCAs (for example, see References 19, 20, 60 - 65). Assuch, there is merit to reviewing open literature to determine if an applicable PIRT already existsfor similar designs and transient conditions. A well-informed PIRT is based on a combination ofphenomenological observations at both reduced- and full-scales and the performance ofphysical models in applicable processes studies. Beyond the regulatory framework (Section2.3) and the B&W mPower reactor-specific information (Section 3), the database supportingB&W NE's PIRT activities originate from previous applications of the PIRT process, publishedanalytical characterizations of phenomena, and the large body of relevant integral- andseparate-effects test program data.

6.2.1 Open Literature SBLOCA PIRT Summary

There are two SBLOCA PIRT studies, appearing in the open literature, having featuressimilar to the B&W mPower reactor. One of these studies is for a B&W lowered-loop 177assembly PWR rated at 2568 MWt (Reference 63). The other study is for the Multi-Application Small Light Water Reactor (MASLWR), the design concept that has sinceevolved into the NuScale design (Reference 64).

Shortly following the introduction of the CSAU methodology in 1989, which presented aPIRT for a large-break LOCA in a Westinghouse PWR, the U.S. NRC commissioned asimilar exercise to demonstrate the PIRT process for an SBLOCA in a B&W lowered loopPWR design. Like the B&W mPower reactor, B&W's original fleet of PWRs includes OTSGsand facilitates natural circulation. In addition, for pipe SBLOCA break sizes below a certainthreshold the core will not uncover. As such, the investigators of that study used liquid level


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as the surrogate figure-of-merit to peak clad temperature. Table 5 presents the results fromthe B&W 177 PWR SBLOCA PIRT.

Like the B&W mPower reactor, the MASLWR employs an integral vessel and is capable ofpassively rejecting core and containment heat to an UHS. Rather than employing activepumps, it is fully reliant on natural convection and is rated at 150 MWt. Table 7 presents theresults from the MASLWR SBLOCA PIRT.

A list of highly important phenomena common in both studies is as follows:

* break flow

* natural circulation (mass flow, fluid resistance)

* decay heat

* RCP performance

* steam generator heat transfer

* RCS depressurization



Table 6. Small Break LOCA PIRT for the B&W 177 Design

Component Process Phenomena Rank

Break Break flow 9

Decay heat 6Forced convection 2

Vessel Upper head mixing 1Phase 1: Flashing 2Blowdown Wall stored heat 1

Reactor coolant pumps Pump performance 4High pressure coolant ECCS flow 4injection

Pressurizer Level 2Wall stored energy 1

Natural circulation 9

Vessel Decay heat 8Vent valve 6

2-phase liquid level 2

Break Break flow 7

Phase separation in U-bend 7Phase 2: U-bend U-bend voiding 5Natural Cold leg void fraction 4circulation High pressure coolant ECCS flow 6

injectionUpper downcomer Condensation 5

Natural circulation 9Primary heat transfer 1

Steam generator Steam generator heat transfer 1

Secondary conditions 1

Reactor coolant pumps Pump performance 9High pressure coolant ECCS flow 7injection

Phase_3:_LossBreak Break flow 6Phase 3: Loss Bek_______________________of natural Decay heat 6circulation Vessel Internal vessel circulation 4

2-phase liquid level 3Cold leg ECCS mix and spill 2Steam generator Primary heat transfer 1U-bend U-bend draining 1High pressure coolant ECCS flow 9injectionBreak Break flow 9

Phase 4: Steam generator Steam generator heat transfer 8Boiler- Decay heat 6

Core 2-phase liquid level 3Internal vessel circulation 1Core heat transfer 1



Table 7. Small Break LOCA PIRT for the MASLWR Design

System Period Component Period ProcesslPhenomenon Period1,2.3 1 2 3 1 2 3

Vent Valves HI H H Valves H H H Mass Flow (Choked/Nonchoked) H H HPiping M M H Line Flow Resistance M M H

ADS I H H Valves I H H Mass Flow (Choked/Nonchoked) I H HPiping I M H Line Flow Resistance I M HSparger I H M Condensation I H M

Energy Release I H MMass Release I H M

Passive Cont. __ _ _ _ _ _ _

Cooling S P P H External Cont. Cooling Pool I I H Natural Convection Heat Transfer I I HThermal Stratification I I L

Containment Shell P P H InternalPressure L M HBuoyancy Driven Flow P P PHeat Transfer L L HWall - condensaton rate L L HNoncondensible Gas Mass Fraction L L H

WallThermal Capacitance L L HThermal Conduction L L H

Passive Safety _ _ _ _ _ _ _ _ _

Recirculation L L H Sump L L H ADS Heat-up of Sump I L MCondensation (Surface of Pool) L I IThermal Stratification L L LRecirculation (Flow Resistance) I I HResupply from Containment L L H

Primary Coolant 71 7 _

System H H H Hot Leg Riser H H Flashing M H LFlow Resistance (wall/control rod tubes) M M MRiser Inventory/Circulation/Level H H H

SG Tube Annulus H H H SG Tube Condensation H H HFlow

Entrainment/De-entrainment L L ICCF I I IFlow Resistance H L LMultidimensional Flow L L L

Level L L HFlashing H H LStored Energy Release- Hot wall effect M M H

ReactorSystem H IH HI Vessel -Control Rods vuH L L Reactivity Change H L LVessel - Core Subchannels H H H Flw

Interfacial Drag L H LMass Flow H H HFlow Resistance H H H

Two-Phase Mixture Level H H HFlashing M H L

Vessel - Downcomer H H I H FlowEntrainment/De-entrainment L L ICCF L L LFlow Resistance H M HMultidimensional Flow L L L

Level L L HFlashing M H L

I Stored Energy Release- Hot wall effect M M HNOTES

Phase I - Blowdown H -Significantly Impacts PCT I - Inactive during the transient PhasePhase 2 - ADS Operation M - Moderately Impacts PCT P - Plausible

Phase 3 - Long Term Cooling L -Little Impact on PCT



Table 7. Small Break LOCA PIRT for the MASLWR Design (continued)System Period Component Period Process/Phenomenon Period__

1 2 31 2 31 2 3

Reactor System(continued) H Vessel - Fuel Rods H H H Fuel Heat Transfer

Conduction H H HGap Conductance H H H

Stored Energy Release H H LCladding Convective Heat Transfer

Subcooled Liquid H L LSubcooled Boiling H L LNucleate Boiling L L HCHF by DNB L L HFilm Boiling L L HForced Convection to vapor L L H

ReactivityVoid H I IModerator temperature H I IFuel Temperature (Doppler) H I I

Decay Heat H H HVessel - Guide Tubes L L I L Film Draining L L I

Stored Energy Release L L LVessel - Lower Plenum H H I H Flow transient

Flow Resistance H L HFlashing M H L

Stored Energy Release M M HVessel - Structures L LI L Stored Energy Release L L LVessel - Upper Head H H L Flow Transient

Entrainment/De-entrainment L L LStored Energy Release H H LFlashing M H L

SteamGenerator/HeatExchanger L L H Tubes L L H Heat Transfer with FW Available H H H

NOTESPhase I - Slowdown H - Significantly Impacts PCT I - Inactive during the transient PhasePhase 2 -ADS Operation M - Moderately Impacts PCT P - PlausiblePhase 3 -Long Term Cooling L - Little Impact on PCT

6.2.2 Open Literature Non-LOCA PIRT Summary

Reference 20, prepared by staff at Siemens Power Corporation, describes the developmentof a PIRT for conventional Westinghouse and Combustion Engineering PWRs. Since Non-LOCA transients are typically less challenging and less complex events than the LOCA, thePIRT team chose to simplify the activity by not dividing the event domains into characteristictime periods. Table 8 summarizes their results.


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Table 8. PIRT for PWR Non-LOCA Transients

Cooldown Heatup

SRP Section C ,- 1 ,- N N , N Co Cowi L. t. U. U5 t6 , N • N •

Criteria Challen edCore-DNBR H H H H H M M M H

Rod-FCM H IPrimary Pressure H H H H H HSteam GeneratorPressure H H M

Enthalpy Deposition

Key PhenomenaFuel/Rod HT H H H H H M M M H M

Kinetics Feedback M M M M H M M M M M

Decay Heat H H M

Pressurizer M M M M M H H H H H H

Flow Coastdown M M M

Boron Tracking MSteam GeneratorH H HSta~nrtr H H H H H H H H H H H HPri/Sec HTSteam GeneratorHSemGnrtrH M H H H H M M MSecondaryCritical Flow H H

Containment Pressure M

Natural Circulation H H H

Core-Loop Mixing M MCooldown Heatu _

SRP Section I T7 00 -- o - 7 ,f-

Criteria Challen ed

Core-DNBR H H H M H H H H H

Rod-FCM H H

Primary Pressure M M M H HSteam GeneratorPressure M

Enthalpy Deposition I I H



Table 8. PIRT for PWR Non-LOCA Transients

Key Phenomena

Fuel/Rod HT H H H H H H H H H

Kinetics Feedback H H H H H H H H

Decay Heat M

Pressurizer M M M M M M M H H H H

Flow Coastdown H H H M

Boron TrackingSteam Generator M M MPri/Sec HTSteam Generator MSecondary MCritical Flow H

Containment Pressure

Natural Circulation M

Core-Loop Mixing

6.2.3 Open Literature Containment Pressure PIRT Summary

Reference 65 captures the current state-of-the-art understanding of containmentphenomena, including the identification and ranking of important containment responsephenomena in a large, dry PWR containment. The Organization for Economic Cooperationand Development / Nuclear Energy Agency (OECD/NEA) experts prepared a PIRT thatconsidered a LOCA through the "core damage phase of a severe accident." In addition,they considered no delineation of LOCA phases and addressed three figures-of-merit:pressure, local temperature, and steam-air-hydrogen composition.

The OECD/NEA PIRT, expressing phenomena ranking for total pressure only, appears inTable 9 with some modification to reflect the B&W mPower-specific application. Suchmodification includes acknowledgement that this design does not employ containmentsprays or fans and that the UHS resides as a pool on top of the containment dome. Amongthe phenomena listed, only two are identified as being highly important for containmentpressure. These are free convection (condensation and evaporation driven) and structureconduction. As a result, they have little impact on peak pressure, which is a strong functionof containment structure surface area and total volume. Atmospheric buoyancy andstratification phenomena are given low-medium and medium rankings for intra-compartmentmixing and inter-compartment transport, respectively. Liquid advection is the only otherphenomenon receiving a ranking other than low.



Table 9. PIRT for Containment Conditions Following an LOCA

Component Process Phenomena RankPressure

Multi-component gas compression/expansion M

Pressurization Aerosol mass and energy exchange L

Depressurization Spray mass and energy exchange MVolume displacement/pool filling or draining LAtmosphere cooling by fan-cooler MJet-plume gas interaction/entrainment L(localized)

Atmosphere Mixing Buoyancy/stratification (regional) L-MAtmosphere Mixing Buoyancy/wall interaction (regional) L(intracompartment) Diffusion (turbulent) L

Spray dynamics N/AFan dynamics N/ABuoyancy/stratification M

Transport Form and friction losses L(intercompartment) Aerosol coupling L

Liquid water carryover LOne-dimensional transient conduction HInterior Two- or three-dimensional transient conduction LMass Transfer Outgassing (concrete) LSpray/aerosol deposition or impingement LFree convection LForced/mixed convection L

Sensible Heat Transfer Radiation (structure to atmosphere) LStructure: Radiation (structure to structure) LSurface Liquid film resistance L(solid and Liquid film advection Lfilm) Latent Heat and Mass Free convection H

Transfer (condensation/ Forced/mixed convection Levaporation)

Transport (film flow) Liquid film advection L-MInterfacial shear (film/gas interaction) L

Mixing Buoyancy/stratification LBubble dynamics L

Pool (UHS): Transport Filling and draining L

Interior Displacement (pressure driven) LConvection (flooded structures) L

Heat Transfer Boiling LSteam condensation (bubbles) LFree convection L

Sensible Heat Transfer Forced/mixed convection LPool (UHS): Aerosol/spray deposition LSurface Latent Heat and Mass Free convection H

Transfer (condensation/ Forced/mixed convection Levaporation)


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The OECD PIRT is applicable to the B&W mPower reactor with few exceptions.


6.2.4 General Observations from PIRT literature review

While the rankings demonstrate the uniqueness of the many events previously evaluated,they also show that certain phenomena are more common than others. The followingidentifies several phenomena that are expected to be of general importance for the B&WmPower reactor transient and accident analysis.






6.3 B&W mPower SBLOCA PIRT

During calendar year 2010, B&W NE hosted a team of subject matter experts through severalmeetings to develop a PIRT for B&W mPower reactor SBLOCA calculations. The participantsemployed a ranking procedure similar to past PIRT exercises to identify the phenomena andprocesses important to evaluating the RCS response to an SBLOCA. The objective of PIRT is torank the importance by the relative figure-of-merit sensitivity to a particular phenomenon.

The principal objective in identifying a scenario for a PIRT team is to select one that exhibits asystem response that is the most challenging to the safety-related SSC responsible formitigating the adverse conditions realized from the event. At the time the B&W mPower reactorSBLOCA PIRT team convened, [

] [CCI per Affidavit 4(a)-(d)] In this PIRT, five distinct phasescharacterize the transient. The SSCs that are important in each phase are identified in Table10.


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Table 10. SSCs Important to Each Phase of the Design Basis Transient (Bolded Items Are

Dominant)

I [CCI per Affidavit 4(a)-(d)]

I


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6.4 Figqures of Merit

A figure-of-merit (FOM) is the measure against which the relative importance of each"phenomenon" is judged. Useful FOMs have the following distinct characteristics:

" directly related to the scenario selected

* directly related to the issue

* directly related and explicit to the phenomena identified

* easily comprehended

* measurable

The FOMs determined to be appropriate for the objectives of this PIRT are shown in Table 11.





Table 11. Figures of Merit for B&W mPower Reactor LOCA Design Basis PIRT

I 1


6.5 Identification of Plausible Phenomena

The objective of this step is to identify the complete spectrum of phenomena that may havesome influence on plant response to a postulated accident or off-normal scenario. Plausiblephenomena are freely defined and can relate to a physically realized FOM (e.g., peak cladtemperature, RCS pressure, RPV liquid level, etc.), a material characteristic (e.g., thermalconductivity) or "process" (e.g., heat transfer) as may best satisfy the objective of understandingthe relative safety importance (controlling nature) of phenomena. Beyond the individual.expertise of the each PIRT participant, a standard approach was used in this step.

" Review of the background information presented in Section 6.2

" Recognition of the phenomenological hierarchy in the context presented in Section 5.1

- List of potentially active systems,

- List of relevant components of each system, and

- Identification of processes and/or phenomena in each component

* Incorporation of the PIRT objectives given in Section 6.1, with the understanding thatthere is little need to proceed further down the hierarchy than the physical processesmodeled by safety analysis codes, and/or the physical processes present and known(measured) in experimental facilities

• Deferral of phenomena evaluation (ranking) until after the identification step to helpassure no plausible phenomenon is ignored because of a priori perceptions of aninsignificant relative importance


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" Allowed downstream iteration on the initial list during the subsequent relative importanceand state of knowledge (SOK) ranking

" Generated written definitions of each phenomenon

6.6 Phenomena Relative Importance and State of Knowledge

The importance rankings for this PIRT are relative to the time phases identified in Section 6.3and the FOMs identified in Section 6.4. The ranking scale used is defined in Table 12.

Table 12. Relative Importance Ranking Scale

Rank Definition Application Outcomes

High (H) Phenomena has controlling Experimental simulation and analytical modeling with aimpact on Figure of Merit high degree of accuracy is critical

Medium (M) Phenomena has moderate Experimental simulation and/or analytical modeling with aimpact on Figure of Merit moderate degree of accuracy is required

Low (L) Phenomena has low impact Modeling must be present only to preserve functionalon Figure of Merit dependencies

Insignificant Phenomena has no, or Modeling must be present only if functional dependencies(N) insignificant impact on Figure are required

of Merit

Functional Dependencies denote Phenomenon A is of low or insignificant importance; however,Phenomenon A must be characterized so that Phenomenon B (and/or C), of moderate or high importance,will be adequately represented.

The objective of the state of knowledge ranking is to help provide a risk-based perspective ofthe safety significance of the plausible phenomena. That is, the SOK rank combined with therelative importance rank of each plausible phenomenon allow categorization into levels of riskrelative to the safety significance regarding experimentation and analysis associated withlicensing and regulation activities. The SOK ranking scale used is defined in Table 13.

Table 13. State of Knowledge Scale

Rank Meaning

H Fully known, small uncertainty

M Known, moderate uncertainty

L Partially known or largely unknown, large uncertainty

N Not applicable

@2011 Babcock & Wilcox Nuclear Energy, Inc. AJI rights reserved.

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In the context of this PIRT, a risk perspective relates to the relative uncertainty in understandingthe combined relative importance and SOK of a phenomenon. For example, a phenomenonthat is controlling in the reactor response to a postulated accident scenario (that is, high relativeimportance) and has an undetermined SOK rank (that is, low SOK) poses the highest level ofrisk in understanding the safety significance of that phenomenon. Further, a phenomenonhaving a high relative importance and moderate SOK rank poses the next relative level of risk,and so on through decreasing levels of risk. The generic risk perspective applicable to thisPIRT is described in Table 14.

Table 14. Generic Risk Perspective

Risk Level Phenomena Phenomena Probable Response To Reduce RiskRelative State Of

Importance Knowledge

1 H L Additional research is required; Need for additional(Highest) analysis is certain; Need for development of new

experimental data is highly likely

2 H M Additional research is required; Need for additionalanalysis is highly likely; Need to search for additional

3 M L existing experimental data is highly likely

4 M M Additional research may be required; Resolution withadditional analysis and/or existing data highly likely5 L L

6 L M

Phenomena with H, M and L Relative Importance and with High SOK do not require additional research

6.7 PIRT Results and Conclusions

Based on both prior PIRTs and the B&W mPower reactor design basis LOCA PIRT, thethermal-hydraulic behavior of an NPP during an SBLOCA is complex; however, no newphenomena have been identified for the B&W mPower reactor. Several phenomena aredeemed important throughout the transient.

.] [CCI perAffidavit 4(a)-(d)]

6.7.1 Safety-related Risk Perspective

[

] [CCI per Affidavit 4(a)-(d)] Those results are summarized in Table 15 includingpotential research that may have promise in further reductions in the risk.


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Table 15. B&W mPower Reactor PIRT Risk Perspective Results

I1* I I

4 4- 4

+ +




I (Table 15 continued on next page)


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4 * 4

I[CCI per Affidavit 4(a)-(d)]

6.7.2 Components and Phenomena that Dominate Each Phase

The SSCs that dominate the various phases have been identified and delineated in Table10. The characteristics of those SSCs contributing to this dominance are discussed asfollows:


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6.7.3 Recommendations on B&W mPower Reactor Use of PIRT Results

Based on the state of the design the following conclusions can be made:

[

][CCIper Affidavit 4(a)-(d)]

6.7.4 Expected Additional B&W mPower PIRTs

A revision to this technical report will provide content on separate short-term and long-termNon-LOCA PIRTs. The short-term Non-LOCA events are to include reactivity anomalieswhile the long-term Non-LOCA events address the balance of Chapter 15 events.


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7. MANAGING UNCERTAINTY IN EMDAP

Associated with NPP accident and transients analysis is a degree of uncertainty that provides asignificant challenge to the evaluation and resolution of design performance and plant licensingissues. For production-grade industrial applications, technical issue resolution using analysisrequires an evaluation methodology framework that begins with a review of the engineeringaccomplishments that demonstrate proof-of-principle, such as the identification of relevantphenomena, the credited test programs, the evolution of analytical techniques, and relatedranges-of-applicability of the conclusions drawn from these activities. With this foundation, thedevelopment of modeling and simulation tools and subsequent analysis can begin in earnest.Issue goals are translated into analysis measures, uncertainties are characterized, andcalculations are designed and executed to demonstrate the completeness of the design in termsof the expected domain of possibilities.

Whereas Sections 2 - 6 of this technical report address the Requirements Element of EMDAP,the remaining EMDAP elements address the overall adequacy of an evaluation methodology byaddressing the broader issue of uncertainty management as applied to the resolution oftechnical issues through analysis. While an evaluation methodology may appear to simplyprovide a procedure for how to perform an analysis, the technical basis for the proceduralmechanics must address the many facets of uncertainty management, including compliance,expertise, data applicability, code and model development, code verification and validation,uncertainty quantification, and human reliability and the consequence of failure. Underscoringthe evaluation methodology framework is the documentation that communicates the technicalbases supporting how these uncertainty domains are satisfied.

7.1 Managing Compliance

The ultimate safety goal of any NPP is the protection of the public from uncontrolled release offission products through a breach in the containment following a severe accident. Regulationshave evolved to recognize the value of a defense-in-depth strategy that reduces the likelihood ofa catastrophic event through multiple layers of physical, design, and operational defenses.Before analysis can be performed, the constraints imposed by regulations must be consideredas demonstrated in Section 2; this is the task of managing compliance. Relatively few explicitrequirements exist for analysis supporting the licensing of a NPP; however, there is aconsiderable record of acceptable methods presented as regulatory guidance and precedent,including the U.S. NRC's Standard Review Plan (i.e., NUREG-0800).

7.2 Managing Knowledge and Expertise

Managing knowledge comprises the strategy used to translate the existing knowledge base ofinsights and experiences into a complete and self-consistent evaluation methodology. In rawform this knowledge exists in individuals and appears in reports and journals that preserve ourunderstanding of a technology. EMDAP can be viewed as a knowledge management system,repackaging the "Top-down/Bottom-Up" approach originally developed in CSAU methodology(Reference 19) for the specific purpose of reducing the breadth of understanding of NPPtransient and accident analysis into basic phenomena, captured into a computer code.





Fundamentally, thermal-hydraulic system codes like RELAP5 must model the principalphenomena associated with the particular plant and scenario of interest. A considerableamount of effort has been afforded for the identification and ranking of phenomena for commonplant transients and accidents, as demonstrated with the PIRT developed in Section 6 for theB&W mPower reactor. The PIRT is the principal tool for capturing expertise through the latterelements of EMDAP. It guides the development of analytical tools and evaluation methodologiesand the establishment of measures to demonstrate their adequacy. The PIRT is not an infalliblesource of information; rather, this is usually supplemented by scaling analysis and test programbenchmark studies. Knowledge completeness is always a concern and vigilant scrutiny isexpected. Good documentation, in the form of reports addressing code models andcorrelations, developmental assessment, programmers guide, and application guidelines, is themost effective means for communicating the level of quality achieved with a particularevaluation methodology.

7.3 Managqing Data Applicability

The fundamental technical basis supporting the use of analysis for issue resolution comes fromexperimentally derived data. Data serves to assure model completeness, demonstratedthrough code verification and validation (see Section 7.5) and uncertainty quantification (seeSection 7.6). Efficient engineering of an evaluation methodology depends on appropriatereduction of the breadth of topical research to the essential constituents (i.e., systems,components, phases, geometries, fields and processes). Specifically, all datasets are notequally important. The PIRT and scaling analysis are used in EMDAP to initially discern relativeimportance. There is arguably a threshold of importance for which the inclusion of an explicit ordetailed model is unnecessary; however, this is a subjective measure.

For completeness, datasets should exist that address the important phenomena over therequired range-of-applicability. It is insufficient to credit test program results that only partiallycover or are totally outside the domain of interest. Of course, the practical limitations ofinvestigation (e.g., sponsorship funding, scale, I&C systems) can always be made to appearinadequate and requiring further experimentation. Rather, the completeness must be shown asa composite of separate-effects tests (SETs) and integral-effects tests (lETs). Beyond therange-of-applicability concern, the density of available data must be sufficient to be useful;however, conservative or bounding analytical treatments can often be developed to account forsuch limitations.

Processes and phenomena appearing in test program data should be insensitivity to scaleeffects. Scaling analysis involves characterizing a process or phenomenon in terms of itsessential constituents (i.e., systems, components, etc.), independent of scale.Phenomenological impact on the principal figures-of-merit appear in terms of nondimensionalratios, for specific event phases, that relate back to the first-order phenomenologicaldependencies in field equation form. Review of both the static and dynamic response of theseparameters reveal the important phenomena as well as experiment and model distortion.Accommodation for notable scale distortion is incorporated into the evaluation methodologydevelopment process.


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7.4 Managing Code/Model Development

Addressing the completeness and range-of-applicability of individual models and correlationsrepresents a necessary, but not unique, code/model challenge to demonstrating the adequacyof an analytical tool. Code/model development challenges may begin with the form of thegoverning equations. For example, the common differential two-phase flow model that inspiredcodes like RELAP5 describes an ill-posed set of conservation equations. Further, simplificationsintroduce round-off error and lack of information can impact numerical convergence.

Beyond code structure, flexible user-defined input presents endless opportunities for mistakennodalization and/or user-options. The so-called User Effect reflects the consequence of a robustand flexible modeling interface popular within the modeling and simulation community(Reference 66). The most common strategy used to address the User Effect is to define a priorirestrictions on nodalization and the application of user options that are codified into automation.

7.5 Managing Code Verification and Validation

The credibility of an evaluation methodology relies on the associated verification and validation(V&V) through code assessments. Verification is the confirmation that documented statementsaccurately reflect the coding and evaluation methodology procedure mechanics, while validationis the act of demonstrating or testing. Verification typically takes the form of a line-by-line reviewof coding and supporting evaluation documentation, and confirmation of compliance with anapproved quality assurance plan. Standard problems, benchmarks with other codes, oranalytical exercises with known solutions (e.g., Method of Manufactured Solutions) are alsouseful complements to explicit code-to-data comparisons.

Computational tools are validated using an appropriate developmental assessment matrixconsisting of both separate and integral effects test program data that address the moreimportant phenomena influencing the ultimate figure(s)-of-merit. The assessment matrixsupports the evaluation methodology development in defining nodalization, quantifying code,accuracy, and demonstrating any code or model scaling effects. The principal objectives are todemonstrate sufficient accuracy in modeling dominant physical processes (determined from aPIRT), appropriate nodalization, independence d scale effects, and the relative insensitivity ofcompensating error. As with the other elements of an evaluation methodology, all V&V tasks arenot equally important and the PIRT results are used to guide such decisions.

Scaling analysis supports both verification and validation by providing independent insight intothe dominant physical processes, as indicated by the relative magnitudes of non-dimensionalcoefficients modifying terms in the field equations. There are two premises on which the scalingimpacts the assessment process. The first premise is that the tests are scalable to similarevents at full-scale and the second is that the code models themselves provide predictionscalability. For the first premise to be true, the selection of tests needs to be such that theimportant phenomena are captured by one or more appropriately scaled tests. For the secondpremise to be true, the phenomenological models in the computer codes should apply to boththe full-scale and the scaled test. Generally, this is done by selecting a number of assessmentsin facilities of different scale and demonstrating that results from those assessments along with


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similar studies in a full-scale plant model consistently predict phenomena appearing in theexperimental data.

RELAP5 has undergone extensive review and validation by the U.S. NRC, the U.S. DOE, andothers. The process includes developmental assessment by the code's developers,independent assessment by members of the U.S. NRC's cooperative research programs,review by the Advisory Committee on Reactor Safety (ACRS), and formal peer review by anindependent committee of experts. This process has been documented through the publicationof formal reports with the release of each version of the code.

RELAP5/MOD2, /MOD3, and -3D have also been widely used by regulatory and researchorganizations around the world to support international standard problem exercises andexperimental programs. As such, the impact of user experience, the ability of the codes topredict thermal-hydraulic phenomena, and the applicability of the codes to prototypic planttransient data have been well characterized. Although plant data for accident conditions arelimited, these codes have also been widely used to assess the performance of plants underdesign basis accident conditions.

7.6 Managqingq Uncertainty Quantification and Convolution

Analysis uncertainty in modeling and simulation has many sources. These include thoseassociated with approximated models that describe the underlying physics; those associatedwith the settings of parameters used in physical models (including, for example, equation-of-state); those associated with performing a simulation at a given spatial resolution; and thoseassociated with approximations in the numerical algorithms. Uncertainty quantification is theprocess of characterizing, estimating, propagating, and analyzing all kinds of uncertainty tosupport decision-making.

Since the 1988 U.S. regulatory rule change to 10 CFR 50.46, there has been notable industryinvestment in uncertainty quantification methods for nuclear plant thermal-hydraulic analysis.These methods center upon the evaluation of parametric and system response uncertainties,where system response uncertainties are derived from the convolution of one or moreparametric uncertainty models. In this context uncertainty quantification begins with a model orcorrelation, data, and a clear qualitative statement that captures the assumptions associatedwith measured values. The latter element is key in that observations of phenomena will ofteninherit a bias associated with data acquisition methods. For example, core heat flux datum is afunction of the movement of coolant, the fluid properties, the geometry of the core, fuel materialsand configuration, the possible stochastic nature of the underlying phenomena, the location anddensity of instrumentation, calibration of the instrumentation, and post-test data reduction. Aswith the type of data, such assumptions are not equally important. What is important is that theapplication of uncertainty models is consistent with the original assumptions.

Because of the investment involved in quantifying uncertainty, the outcomes from the previousevaluation methodology development tasks, in particular, expertise documented in a PIRT,identify a manageable set of uncertainty parameters. The quantification process requires thatdata from SETs be separated into control and validation sets. The control set is used to derive


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the uncertainty and, as the name implies, the validation set is used to validate the integrity of theuncertainty model. A general uncertainty model is characterized by a bias and a probabilitydensity function; however, the uncertainty model does not necessarily reflect a task unique toperforming a probabilistic analysis. A deterministic analysis can be viewed as one built fromuncertainty models consisting of parameter biases. While the bias may simply reflect a staticerror in a model parameter, it can also be used to define a conservative or bounding treatmentbased on a limited set of test data.

Figure 11 illustrates the difference in the convolution of uncertainty within deterministic andprobabilistic analysis. With deterministic treatment key parameters are conservatively bounded,effectively "stacked" upon other conservatisms in a single calculation. With probabilistictreatment key parameters are sampled over an uncertainty range, requiring several calculations.

a) Deterministic Analysis b) Probabilistic Analysis

Figure 11. Contrast Between Deterministic and Probabilistic Convolution of Uncertainty

A broad definition of uncertainty quantification includes risk quantification. While risk is generallyassociated with safety and health, it could also consider investment, security, human resourcesor even psychological risks such as political and institutional integrity - anything that, if lost,would impact project goals. Risk quantification requires a model that takes system response


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metrics and their uncertainties as an input, and produces risk metrics and their uncertainties asthe output. In all likelihood, the risk model itself may be uncertain, e.g., the health impact of agiven dose of radiation. This calls for additional risk model detail defined from the convolution ofthe probability distributions of system response metrics with the probability distributions of therisk metrics.

7.7 Managing Human Reliability and Consequence of Failure

In many instances the design basis performance of the components, processes, and systems ina NPP is demonstrated assuming no operator action. While for NPPs with active safetysystems, operator action may be credited in certain analyses (assuming a delay of 30 minutesor more, for example), for passive plant designs like the B&W mPower reactor, incorporation ofoperator action prior to 72 hours after event initiation is only performed if the action couldinadvertently reduce the margin-to-limit for a particular acceptance criteria figure-of-merit.

Similar to design basis safety analysis, there is a large scope of RCS performance analysis forthe purpose of evaluating the integrity of I&C and accident management that require that therole of the operator be integrated into the analyses. These analyses address two non-physicalplant defense-in-depth mechanisms (see Section 2.1) and the supporting analyses shouldexplicitly address the uncertainty associated with operator performance. Probabilisticmethodologies have not been used extensively in this area; however, the analytical tools thatare commonly used for this activity can be applied in a manner that accounts for thisuncertainty.

Ultimately, the possibility of multiple failures leading to an undesirable situation is intrinsic in anycomplex system. As such, the consequence of such failure should be evaluated. In a broadsense, beyond design basis accident analysis addresses the failure of the protection and safetysystems addressing design basis transients and accidents; although, ATWS could also fall intothis category. Further, accident management strategies are developed to extend the defense-in-depth concept to respond to the complete failure of the balance of a plant's defenses. PRA is acommon tool to quantify the frequency, consequence, and overall risk of failure across the full-scope of possible plant failure mechanisms.


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8. REQUIREMENTS SUMMARY

The overriding principle to performing design basis safety analysis, as stated in Section 1, isthat to draw credible conclusion from analysis, strict adherence to a verifiable process must befollowed. Legitimate design inputs and assumptions are adopted addressing the analyticaluncertainties related to compliance, expertise, data applicability, code and model development,code verification and validation, uncertainty quantification, and human reliability and theconsequence of failure. Table 17 through Table 19 summarizes of the safety analysisinfrastructural requirements inferred from the discussion appearing in the previous sections thatsupport the above principle. Table 16 summaries infrastructural requirements recognizingorganizational and division-of-labor logistics supporting the B&W mPower reactor and facilitydesign. Table 17 summarizes both functional and infrastructural requirements of notablegovernment regulations and guidance documents with specific relevance to safety analysis.Table 18 summaries the documents necessary to completely describe safety analysisevaluation methodologies per RG 1.203. Table 19 summaries a combination of functional andinfrastructural computing and software requirements.

Table 16. Summary of Organizational Requirements

ReauirementSafety analysis characterizes the RCS pressure boundary, a physical fissionproduct barrier, and defines the design envelope for thermal-hydraulicconditions during normal operation and design basis events

RCS performance metrics explicitly interface with unique project design teamsresponsible for fuel, core, and containment performance, instrumentation andcontrols, equipment qualification, severe accidents, structural loads, doseassessments, radionuclide transport, emergency response, simulator design,and training programs.

The complexity of engineering and software interfaces places a practical requirementthat task automation, pre-/post- processing tools, or data transfer/migration activitiesbe incorporated into an evaluation methodology networked to a large common fileserver for storing quality records

Plant design or process changes trigger reanalysis or disposition to determine itsimpact on plant conditions measured against acceptance criteria


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cument No. Title Rev. N

WR-TECR- Redacted Version 00015013(N P)

Table 17. Summary of Regulatory Requirements for Safety Analysis

# rRequirement1 10 CFR 50 Appendix B and ASME NQA-1

Design control measures explicitly apply to reactor physics, thermal, hydraulicand accident analysesDesign control measures include configuration control practices to protectcomputer code and evaluation methodology integrity and allow traceability ofboth the code version and plant input modelsDesign control measures shall provide for verifying or checking the adequacy ofdesign, such as by the performance of design reviews, by the use of alternateor simplified calculational methods, or by the performance of a suitable testingprogram

Design changes are subject to appropriate design control measures

Documented instructions, e.g., user guides, are requiredThe appropriate level of document control and quality assurance (QA) recordsis identified on documentation

Errors must be promptly identified and corrected, and corrective actions mustbe taken to preclude repetition. All of this must be documented

2 10 CFR 50.34, Content of Application, technical information, final safety analysisreport (as paraphrased in RG 1.203)

Safety analysis reports must analyze the design and performance of structures,systems, and components, and their adequacy to prevent accidents andmitigate the consequences of accidents

Analysis and evaluation of emergency core cooling system (ECCS) coolingperformance following a LOCA must be performed in accordance with therequirements of 10 CFR 50.46The technical specifications for the facility must be based on the safety analysisand prepared in accordance with the requirements of 10 CFR 50.36

3 10 CFR 50.43 Additional standards and provisions affecting class 103 licenses andcertifications for commercial power

The performance of all design safety features (i.e., SSC) has beendemonstrated through analysisInterdependent effects among design safety features have been foundacceptable by analysisSufficient data exist on the safety features of the design to assess the analyticaltools used for safety analysis over a range of normal operating conditions

4 10 CFR 50 Appendix A

GDC 10 - The reactor core and associated coolant, control, and protectionsystems shall be designed with appropriate margin to assure that specifiedacceptable fuel design limits are not exceeded during any condition of normaloperation, including the effects of anticipated operational occurrences

I (Table 17 continued on next page)


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# IRequirementGDC 13 - Instrumentation shall be provided to monitor variables and systemsover their anticipated ranges for normal operation, for anticipated operationaloccurrences, and for accident conditions as appropriate to assure adequatesafety

GDC 15 - The reactor coolant system and associated auxiliary, control, andprotection systems shall be designed with sufficient margin to assure that thedesign conditions of the reactor coolant pressure boundary are not exceeded

GDC 16 - Reactor containment and associated systems shall be provided toestablish an essentially leak tight barrier against the uncontrolled release ofradioactivity to the environment

GDC 19 - A control room shall be provided from which actions can be taken tooperate the nuclear power unit safely under normal conditions and to maintain itin a safe condition under accident conditions, including loss-of-coolantaccidents. Adequate radiation protection shall be provided to permit access andoccupancy of the control room under accident conditions

GDC 20 - The protection system shall be designed to initiate automatically theoperation of appropriate systems including the reactivity control systems, toassure that specified acceptable fuel design limits are not exceeded as a resultof anticipated operational occurrences

GDC 26 - Two independent reactivity control systems of different designprinciples shall be provided.

GDC 28 - The reactivity control systems shall be designed with appropriatelimits on the potential amount and rate of reactivity increase

GDC 29 - The protection and reactivity control systems shall be designed toassure an extremely high probability of accomplishing their safety functions inthe event of anticipated operational occurrences

GDC 34 - A system to remove residual heat shall be provided. The systemsafety function shall be to transfer fission product decay heat and other residualheat from the reactor core at a rate such that specified acceptable fuel designlimits and the design conditions of the reactor coolant pressure boundary arenot exceeded

GDC 35 - A system to provide abundant emergency core cooling shall beprovided. The system safety function shall be to transfer heat from the reactorcore following any loss of reactor coolant

GDC 38 - A system to remove heat from the reactor containment shall beprovided.

GDC 41 - Systems to control fission products, hydrogen, oxygen, and othersubstances shall be provided as necessary to reduce their concentrationfollowing an accident to assure that containment integrity is maintained

GDC 44 - A system to transfer heat from structures, systems, and componentsimportant to safety, to an UHS shall be provided. The system safety functionshall be to transfer the combined heat load of these structures, systems, andcomponents under normal operating and accident conditions




# IRequirementGDC 50 - The reactor containment structure, including access openings,penetrations, and the containment heat removal system shall be designed sothat the containment structure and its internal compartments can accommodate,without exceeding the design leakage rate and with sufficient margin, thecalculated pressure and temperature conditions resulting from any loss-of-coolant accident

5 The regulatory position on the analytical interpretation of the CFR requirements hasevolved through revisions of the U.S. NRC's Standard Review Plan and RegulatoryGuides

RG 1.7, Control of Combustible Gas Concentrations in Containment

RG 1.70, Standard Format and Content of Safety Analysis Reports for NuclearPower Plants

RG 1.77, Assumptions Used for Evaluating a Control Rod Ejection Accident forPressurized Water Reactors

RG 1.183, Alternative Source Term

RG 1.203, Transient and Accident Analysis Methods (evaluation methodologyrequirements)

RG 1.216, Containment Structural Integrity Evaluation for Internal PressureLoadings above Design Basis Pressure

A complete set of safety analyses must be prepared consistent with the reviewcriteria appearing in NUREG-0800

A notable Interim Acceptance Criteria is identified in SRP Section 4.2, AppendixB that supersedes statements appearing in RG 1.77 and RG 1.183.


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Table 18. Documentation Requirements based on RG 1.203

# Requirement1 Element 1: Establish requirements for evaluation mode capability

Specify analysis purpose, transient class and power plant class

Specify figures of merit

Identify systems, components, phases, geometries, fields and processes thatshould be modeled

Identify and rank phenomena and processes

2 Element 2: Develop assessment base

Specify objectives for assessment base

Perform scaling analysis and identify similarity criteria

Identify existing data and/or perform IETs and SETs to complete database

Evaluate effects of IET distortions and SET scale up capability

Determine experimental uncertainties

3 Element 3: Develop evaluation model

Establish EM developmental plan

Establish EM structure

Develop or incorporate closure models

4 Element 4: Assess evaluation model adequacy

Determine model pedigree and applicability to simulate physical processes

Prepare input and perform calculations to assess to assess model fidelity and/oraccuracy

Assess scalability of models

Determine capability of field equations and numeric solutions to representprocesses and phenomena

Determine applicability of evaluation methodology to simulate systemcomponents

Prepare input and perform calculations to assess system interactions and globalcapability

Assess scalability of integrated calculations and data for distortions

Determine EM bases and uncertainties

5 Required documentation

Internal quality records (e.g., drawings, system descriptions, analytical inputrequirements, calculations, analysis reports, etc.)

Evaluation methodology requirements (per RG 1.203, RG 1.77, and RG 1.183)

Evaluation methodology instructions (e.g., guidelines for developing code inputand safety analysis methods)

Code description manuals (e.g., code structure, user input requirements,

(Table 1 8 continued on next Paw) 1@2011 Babcock & Wilcox Nuclear Energy, Inc. All rights reserved.

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# IRequirementmodels and correlations, verification review, and programmers guide),

Scaling reports (e.g., scaling methodology, comparison of important parametergroups, distortion analysis)

Assessment reports (e.g., general test problems, PIRT-targeted)

Uncertainty analysis reports (e.g., uncertainty ranges, statistical treatment,analysis of results)


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Table 19. Computing and Software Requirements

# RequirementComputing Environment

1 Safety analysis requires a high performance computing environment, supported by astable operating system, such as UNIX

2 Large and diverse data, including code subroutines and executables, utility codesand scripts, parameter models of plants, test program facilities and related data,verification and validation suites, and documentation in various formats, is managedwithin a highly structured and transparent quality program

3 Safety analysis must readily deliver/receive data to/from other project teams withsimilar computational requirements without constraints to their physical location.

4 Safety analysis depends on the ability to reduce, filter, and migrate data betweenformats while abiding by a quality control program is required

5 Analysis and qualification results must reside in a quality records managementsystem, be readily accessible and reproducible

6 The computing environment is part of the evaluation methodology. A single platformmust be selected

Code Selection or Development

7 The codes must be evaluated as to whether they support the ability to accuratelyquantify the phenomena important to measuring the key figures-of-merit.

8 Essential features of RCS performance code

Multi-Dimensional Capability

Energy Equations

Numerical Solution of Hydrodynamic Field Equations

Hydrodynamic Constitutive Models

Heat Transfer Model

Fuel Performance Model

Containment Model

Multi-dimensional reactor kinetics.

FORTRAN 90

9 To credit analysis exemption from the Appendix K treatment of fluid-flowuncertainties at post-CHF conditions, the B&W mPower test program must provideample evidence to conclude that vessel liquid level never drops below the top of thecore under the most challenging conditions and accounting for any scale distortion.

Code and Methodology Qualification

10 Software and evaluation methodology must submit to a qualification program

11 Code adequacy is demonstrated emphasizing the PIRT assessment conclusions.The assessment is accomplished by comparing the event and important phenomena

(Table 19 continued on next Daael


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# Requirementidentified in the PIRT with the models and correlations documents for the selectedcodes

Field equations that address global processes

Closure (constitutive) equations which support the field equations bymodeling specific phenomena or processes

Code numerics that demonstrate that the code can efficiently and reliablyperform the necessary calculations

Structure and nodalization, which address the ability of the code to modelthe NPP geometry and components, and to provide an accurate predictionof the NPP response.

12 Changes to computing hardware, software, and methodologies elements require arequalification of the evaluation methodology

Testing & Debugging

13 The assessment matrix must include experiments that address the importantphenomena

14 Experiments should be selected that cover the range of each important phenomenonobserved in the NPP analyses

15 The plant model must be nodalized finely enough to represent both the importantphenomena and design characteristics of the NPP but coarsely enough to remaineconomical

16 The phenomenological models in the computer codes need to be demonstrated atboth reduced- and full-scale.

17 It should be demonstrated that the compensating errors will not produce erroneousresults for the selected scenario and NPP being analyzed.

18 Causes for code-to-data deviations must be explain in terms of the modeled physics(or lack thereof) and how important a particular phenomenon is to the overall results.


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9. CONCLUSIONS

The B&W mPower reactor is an integral PWR design with the reactor, steam generator andpressurizer all located in a single pressure vessel. The design has no large cold or hot legpiping, which eliminates the potential for large break LOCA accidents. Most transients andaccidents described in NUREG-0800 are either identical or very similar to the commercialoperating PWRs or to advanced PWRs with passive safety systems. No new thermal-hydraulicor neutronics phenomena have been identified for this design. Because the B&W mPowerreactor employs conventional balance-of-plant systems, all transients initiated by failures inthese systems are very similar to other PWR designs.

This requirements document has been prepared to describe the functional and infrastructuralrequirements and the associated basis for performing design basis safety analysis supportingthe development and certification of the B&W mPower reactor. Nearly three dozen generalrequirements (with many more subheading requirements) related to the design, development,and deployment of safety analysis methods have been identified. Complete evaluationmethodologies addressing the remaining elements in the EMDAP will be captured insupplements to this technical report. Refinement of requirements will continue as newinformation is revealed through plant engineering and the IST program.


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10. REFERENCES

The following references are cited within the text:

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