area™ - arcadia revision 0 rod ejectionareva inc. anp-10338np revision 0 area™ - arcadia® rod...

ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection

Accident

Topical Report

October 2015

AREVA Inc.

(c) 2015 AREVA Inc.

ANP-10338NP Revision 0

Copyright © 2015

AREVA Inc. All Rights Reserved

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page i

Nature of Changes

Item Section(s) or Page(s) Description and Justification

1 All Initial Issue

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page ii

Contents

Page

1.0 INTRODUCTION ............................................................................................... 1-1

1.1 Range of Applicability ............................................................................. 1-2

1.2 Topical Report Content ........................................................................... 1-2

2.0 APPLICABLE REA REGULATORY REQUIREMENTS ..................................... 2-1

2.1 Current Criteria ....................................................................................... 2-2

2.2 NRC Proposed Changes to Criteria ........................................................ 2-5

2.3 Future Criteria ......................................................................................... 2-7

2.4 Maximum RCS Pressure ........................................................................ 2-7

3.0 ROD EJECTION ACCIDENT SCENARIO IDENTIFICATION ............................ 3-1

3.1 Reactivity Insertion ................................................................................. 3-1 3.1.1 Prompt Critical .............................................................................. 3-2 3.1.2 Sub-Prompt Critical ...................................................................... 3-3

3.2 RCS Pressure ......................................................................................... 3-4

4.0 PHENOMENA IDENTIFICATION RANKING TABLE (PIRT) EVALUATION OF REA MODEL REQUIREMENTS .......................................... 4-1

4.1 Fuel Pin Integrity During a Prompt Power Pulse ..................................... 4-1

4.2 DNBR ...................................................................................................... 4-2

4.3 System Pressure .................................................................................... 4-2

4.4 Regulatory Criteria for an REA ............................................................... 4-2

5.0 ANALYTICAL MODELS .................................................................................... 5-1

5.1 GALILEO™ ............................................................................................. 5-3 5.1.1 Enthalpy Rise Limits ..................................................................... 5-4 5.1.2 Thermal Properties ....................................................................... 5-4 5.1.3 Fuel Pin Pressure ......................................................................... 5-5

5.2 ARCADIA® .............................................................................................. 5-5 5.2.1 ARCADIA® Validation ................................................................... 5-5 5.2.2 Verification of Gap Conductance and Thermal

Conductivity Models ..................................................................... 5-6

5.3 COBRA-FLX™ ........................................................................................ 5-8 5.3.1 COBRA-FLX™ Validation ............................................................ 5-9

5.4 RELAP5 Computer Code ........................................................................ 5-9 5.4.1 S-RELAP5 Code and Model ...................................................... 5-10

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page iii

5.4.2 RELAP5/MOD2-B&W Code and Model ..................................... 5-12

6.0 AREA™ METHODOLOGY DESCRIPTION ...................................................... 6-1

6.1 Applicable Regulatory Criteria ................................................................ 6-1

6.2 GALILEO™ ............................................................................................. 6-1 6.2.1 PCMI Failure Criteria for Clad ...................................................... 6-1 6.2.2 Fuel Pin Pressure ......................................................................... 6-2 6.2.3 Fuel and Rim Melt ........................................................................ 6-4 6.2.4 Fuel and Clad Thermal Properties ............................................... 6-4 6.2.5 Fuel Pellet to Clad Gap Conductance .......................................... 6-4

6.3 ARTEMIS™ Models for REA Event Analysis .......................................... 6-5

6.4 ARTEMIS™ (Steady State Nodal Solution) ............................................ 6-6

6.5 ARTEMIS™ (Transient Nodal Solution) .................................................. 6-8 6.5.1 Trip Function ................................................................................ 6-8 6.5.2 Enthalpy Rise ............................................................................. 6-11 6.5.3 Adjustment Factors .................................................................... 6-12

6.6 Transient COBRA-FLX™ Calculations ................................................. 6-13 6.6.1 Adjustment Factors .................................................................... 6-13 6.6.2 DNBR Critical Heat Flux Correlations ........................................ 6-15 6.6.3 Mixed Core Applications ............................................................ 6-15

6.7 RELAP5 ................................................................................................ 6-15 6.7.1 RCS Pressure Evaluations ......................................................... 6-16 6.7.2 Pressure for DNB Evaluations (Scenario 2 Section

3.2) ............................................................................................. 6-16

6.8 Data Processing ................................................................................... 6-18 6.8.1 PCMI Failure Criteria .................................................................. 6-18 6.8.2 Total Enthalpy for High Clad Temperature Failure

Criteria ....................................................................................... 6-19 6.8.3 Fuel Melt Failure Criteria ............................................................ 6-20 6.8.4 Coolability .................................................................................. 6-20

6.9 Fuel Failures ......................................................................................... 6-22

6.10 Radiological Consequences ................................................................. 6-22

6.11 Update Process .................................................................................... 6-23

6.12 Level of Significance ............................................................................. 6-26

6.13 Method Summary ................................................................................. 6-27

7.0 UNCERTAINTY AND BIASING METHODOLOGY ............................................ 7-1

7.1 Core Sensitivity Analysis ......................................................................... 7-1

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7.1.1 Sensitivity Evaluation Method ...................................................... 7-4 7.1.2 Onset of Trip ................................................................................ 7-6 7.1.3 Core Biasing Strategy .................................................................. 7-6 7.1.4 Core Biasing Values ..................................................................... 7-7

7.2 RELAP5 Biasing for Pressure Calculations .......................................... 7-19 7.2.1 RELAP5 Peak RCS Pressure Calculations ................................ 7-20 7.2.2 RELAP5 Core Pressure for MDNBR Calculations ...................... 7-22

8.0 AREA™ PLANT SPECFIC APPLICATION ....................................................... 8-1

8.1 Initial Application of AREA™ Methodology ............................................. 8-1

8.2 Cycle to Cycle Evaluation ....................................................................... 8-2

9.0 SAMPLE PROBLEMS ....................................................................................... 9-1

10.0 CONCLUSIONS .............................................................................................. 10-1

11.0 REFERENCES ................................................................................................ 11-1

APPENDIX A W 4-LOOP 193 FA PLANT WITH 17X17 FUEL LATTICE ..................... A-1

APPENDIX B B&W 177 FA PLANT WITH 15X15 FUEL LATTICE .............................. B-1

APPENDIX C CE 217 FA PLANT WITH 14X14 FUEL LATTICE ................................ C-1

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page v

List of Tables

Table 4-1 PIRT Plant Transient Analysis ..................................................................... 4-3

Table 4-2 PIRT Fuel Rod Transient Analysis for Fuel and Clad Temperatures ........... 4-3

Table 4-3 Parameters Directly Addressed by AREA™ Methodology .......................... 4-4

Table 4-4 System Parameters Considered for Pressure Analysis ............................... 4-4

Table 5-1 ARCADIA® Validation Test Matrix for AREA™ .......................................... 5-13

Table 5-2 GALILEO™/ ARTEMIS™ Transient Comparisons for UO2 Fuel ............... 5-14

Table 5-3 GALILEO™/ ARTEMIS™ Transient Comparisons for 2 wt% Gadolinia Fuel ....................................................................................................... 5-15

Table 5-4 GALILEO™/ ARTEMIS™ Transient Comparisons for 8 wt% Gadolinia Fuel ....................................................................................................... 5-16

Table 5-5 COBRA-FLX™ Validation Test Matrix for AREA™ ................................... 5-17

Table 7-1 Criteria Applicability to Initial Conditions for Sensitivity Calculations ......... 7-24

Table 7-2 Core Biasing Strategies for the Key Parameters ....................................... 7-25

Table 7-3 Parameters Considered For Biasing for RCS Pressure Scenarios ........... 7-26

Table 9-1 Core Biasing Parameters and Values ......................................................... 9-3

Table 9-2 Biasing Parameters and Values for Overpressure ...................................... 9-4

Table A-1 General Timing of the Event ..................................................................... A-11

Table A-2 W 4-Loop Limiting Results Summary for Burnup 1 ................................... A-12





Table A-7 W 4-Loop, Measure of Conservatism for Limiting Result Cases ............... A-17

Table A-8 Transient and Static Difference in Limiting Conditions .............................. A-18

Table A-9 W 4-Loop Plant Overpressure Input Summary ......................................... A-19

Table A-10 W 4-Loop Plant Overpressure Results Summary (High Pressurizer Pressure Trip Modeled) ........................................................................ A-20

Table A-11 W 4-Loop Plant Overpressure Results Summary ................................... A-21

Table A-12 W 4-Loop Plant Core Pressure for MDNBR Input Summary ................... A-22

Table B-1 General Timing of the Event ....................................................................... B-8

Table B-2 B&W Plant Limiting Results Summary for Burnup 1 ................................... B-9

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page vi

Table B-3 B&W Plant Limiting Results Summary for Burnup 2 ................................. B-10




Table B-7 Measure of Conservatism for Each of the Limiting Cases ........................ B-14

Table B-8 B&W plant Overpressure Input Summary ................................................. B-15

Table B-9 B&W plant Overpressure Results Summary (no high pressure trip modeled) ............................................................................................... B-16

Table B-10 B&W Plant Overpressure Results Summary ........................................... B-17

Table B-11 B&W Plant Core Pressure for MDNBR Input Summary .......................... B-18

Table C-1 CE Plant General Timing of the Event ....................................................... C-4

Table C-2 CE Plant Limiting Results Summary for Burnup 1 ..................................... C-5





Table C-7 CE Plant Measure of Level of Conservatism for Each Limiting Parameter ............................................................................................ C-10

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page vii

List of Figures

Figure 2-1 Corrosion Limit Based on Relative Oxide Thickness ................................. 2-9

Figure 2-2 Corrosion Limit Based on RXA Clad Type and Excess Hydrogen ........... 2-10

Figure 2-3 Corrosion Limit Based on SRA Clad Type and Excess Hydrogen ........... 2-11

Figure 3-1 Prompt Critical Power Excursion ................................................................ 3-6

Figure 3-2 Sub-Prompt Critical Power Excursion (Prompt-Jump) ............................... 3-6

Figure 5-1 Coupling of the Time Dependent Models ................................................. 5-18

Figure 5-2 UO2 HZP EOL Transient Fuel Centerline Temperature ........................... 5-19

Figure 5-3 UO2 HZP EOL Transient Maximum Rim Temperature ............................. 5-20

Figure 5-4 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Centerline Temperature ......................................................................................... 5-21

Figure 5-5 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Surface Temperature .... 5-22

Figure 5-6 8 wt% Gadolinia Fuel HZP EOL Transient Fuel Centerline Temperature ......................................................................................... 5-23

Figure 5-7 8 wt% Gadolinia Fuel HZP EOL Transient Maximum Rim Temperature ......................................................................................... 5-24

Figure 6-1 REA Analysis Code and Data Links ......................................................... 6-28

Figure 6-2 SCRAM Position versus Drop Time ......................................................... 6-29

Figure 6-3 Pulse Width Definition for Prompt versus Non-Prompt ............................ 6-30

Figure 6-4 DNBR for a Prompt Pulse at 20% Power ................................................. 6-31

Figure 7-1 Doppler Test Result Comparisons .......................................................... 7-27

Figure 8-1 Increased Biasing for Cycle Verification .................................................... 8-4

Figure A-1 W 4-Loop Enthalpy Rise Limits for M5® Fuel Based on Relative Oxide Thickness ................................................................................... A-23

Figure A-2 W 4-Loop Enthalpy Rise Limits for Zr4 Fuel Based on Relative Oxide Thickness .............................................................................................. A-24

Figure A-3 W 4-Loop Limiting Pressure Parameters for UO2 Fuel with M5® Clad ..... A-25

Figure A-4 W 4-Loop Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad ...................................................................................................... A-26

Figure A-5 W 4-Loop Limiting Pressure Parameters for UO2 Fuel with Zr4 Clad ...... A-27

Figure A-6 W 4-Loop Limiting Pressure Parameters for Gadolinia Fuel with Zr4 Clad ...................................................................................................... A-28

Figure A-7 W 4-Loop Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad ........ A-29

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page viii

Figure A-8 W 4-Loop Limiting FGR for UO2 and Gadolinia Fuel with Zr4 Clad ......... A-30

Figure A-9 W 4-Loop General Depressurization Curve ............................................. A-31

Figure A-10 Transient FQ, F∆H, and Core Power for Max Enthalpy Condition ........... A-32

Figure A-11 Transient Maximum Enthalpy for Max Enthalpy Condition .................... A-33

Figure A-12 Total Enthalpy Limit with Burnup for Max Enthalpy Condition ............... A-34

Figure A-13 Enthalpy Margin to Limit Scatter Plot for Max Enthalpy Condition ......... A-35

Figure A-14 Transient FQ, FΔH, and Core Power for Max Enthalpy Rise Condition ... A-36

Figure A-15 Transient Maximum Enthalpy Rise for Max Enthalpy Rise Condition .... A-37

Figure A-16 Transient Maximum Enthalpy for Max Enthalpy Rise Condition ............ A-38

Figure A-17 Maximum Enthalpy Rise and Limits by Clad Type for Max Enthalpy Rise Condition ...................................................................................... A-39

Figure A-18 Transient FQ, FΔH, and Core Power for Max Fuel Temperature Condition ............................................................................................... A-40

Figure A-19 Transient Fuel, Fuel Rim and Clad Temperature for Max Fuel Temperature Condition ......................................................................... A-41

Figure A-20 Maximum Fuel Temperature by Fuel Type – Margin to Limits for Max Fuel Temperature Condition ......................................................... A-42

Figure A-21 Maximum Fuel Rim Temperature by Fuel Type – Margin to Limits for Max Fuel Rim Temperature Condition ............................................. A-43

Figure A-22 Transient FQ, FΔH, and Core Power for MDNBR Condition .................... A-44

Figure A-23 Transient MDNBR for MDNBR Condition .............................................. A-45

Figure A-24 SAFDL to MDNBR Ratio by Fuel Type as a Function of Burnup for MDNBR Condition ................................................................................ A-46

Figure A-25 SAFDL to MDNBR Ratio by Fuel Type as a Function of Fuel Pin to Core Pressure Difference for MDNBR Condition .................................. A-47

Figure A-26 Case 2 Power Response for High Pressurizer Pressure Trip ................ A-48

Figure A-27 Case 2 Pressure Response for High Pressurizer Pressure Trip ............ A-49

Figure A-28 Case 4 Power Response for High Pressurizer Pressure Trip ................ A-50

Figure A-29 Case 4 Pressure Response for High Pressurizer Pressure Trip ............ A-51

Figure A-30 Core Pressure for MDNBR Response Comparison ............................... A-52

Figure B-1 Enthalpy Rise Limits for M5® Fuel Based on Excess Hydrogen .............. B-19

Figure B-2 Limiting Pressure Parameters for UO2 Fuel with M5® Clad ..................... B-20

Figure B-3 Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad ............. B-21

Figure B-4 Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad ........................ B-22

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page ix

Figure B-5 B&W General Depressurization Curve .................................................... B-23

Figure B-6 Reactor Power for Biased Case .............................................................. B-24

Figure B-7 Peak RCS Pressure for Biased Case ...................................................... B-25

Figure B-8 Reactor Power For Prompt Critical – No Trip .......................................... B-26

Figure B-9 Peak RCS Pressure Response for Prompt Critical Reactivity Addition – No Trip ............................................................................................... B-27

Figure B-10 Hot Leg Pressure Comparison .............................................................. B-28

Figure B-11 Verification of the General Depressurization Curve. .............................. B-29

Figure C-1 Enthalpy Rise Limits for M5® Fuel Based on Relative Oxide Thickness ............................................................................................. C-11

Figure C-2 Limiting Pressure Parameters for UO2 Fuel with M5® Clad .................... C-12

Figure C-3 Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad ............. C-13

Figure C-4 Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad ....................... C-14

Figure C-5 CE Plant General Depressurization Curve ............................................. C-15

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page x

Nomenclature

(If applicable) Acronym Definition

AFD Axial Flux Difference

AREA™ ARCADIA® Rod Ejection Accident

AO Axial Offset

ASI Axial Shape Index

ASME American Society of Mechanical Engineers

B&W Babcock & Wilcox

BOC Beginning of Cycle

BOL Beginning of Life

BWR Boiling Water Reactor

CE Combustion Engineering

CEA Control Element Assembly

CFR Code of Federal Regulations

CHF Critical Heat Flux

COLR Core Operating Limits Report

CPR Critical Power Ratio

CRDM Control Rod Drive Mechanism

DPC Doppler Power Coefficient

DTC Doppler Temperature Coefficient

EFPD Effective Full Power Days

EOC End of Cycle

EOL End of Life

ERW Ejected Rod Worth

FGR Fission Gas Release

FP Full Power

GDC General Design Criteria

GWd/MTU Gigawatt days per Metric Tonne Uranium

HFP Hot Full Power

HZP Hot Zero Power

LOCA Loss of Coolant Accident

M(DNB)R Minimum (Departure from Nucleate Boiling) Ratio

MOL Middle of Life

MPI Message Passing Interface

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page xi

Acronym Definition

MTC Moderator Temperature Coefficient

MWd/mtU Megawatt days per Metric Tonne Uranium

NRC Nuclear Regulatory Commission

PC Power Coefficient

PCMI Pellet Cladding Mechanical Interaction

PDIL Power Dependent Insertion Limit

PIRT Phenomena Identification Ranking Table

PSV Pressure Safety Valve

PWR Pressurized Water Reactor

RCCA Rod Control Cluster Assembly

RCS Reactor Coolant System

REA Rod Ejection Accident

RG Regulatory Guide

RIA Reactivity Initiated Accident

RPS Reactor Protection System

RXA Recrystallized Annealed

SAFDL Specified Acceptable Fuel Design Limit

SRA Stress Relief Annealed

SRP Standard Review Plan

UO2 Uranium Dioxide

W Westinghouse

wt% Weight Percent

Zr4 Zircaloy 4 Alloy

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page xii

ABSTRACT

This report presents the AREA™ methodology for the evaluation of a control rod

ejection accident in a PWR. The methodology is used to demonstrate compliance with

the acceptance criteria specified in NUREG-0800, Section 4.2, Appendix B which

contains the current Nuclear Regulatory Commission criteria for a control rod ejection

accident. The AREA™ methodology is flexible and is capable of demonstrating

compliance with potential revisions to the rod ejection accident criteria. The

methodology is consistent with the guidance in Regulatory Guide 1.77 and NUREG-

0800, Section 15.4.8.

The methodology makes use of a variety of AREVA codes and methods. The

ARCADIA® code system is used to analyze the three dimensional neutronics and

thermal-hydraulics behavior during the transient. The code GALILEOTM provides the

thermal-mechanical properties of the fuel pins. The code S-RELAP5 is used to model

the reactor coolant system response for Westinghouse and Combustion Engineering

plants and the code RELAP5/MOD2-B&W is used for Babcock & Wilcox plants.

The methodology is applicable to PWRs for which the codes and methods are

applicable. These include all currently operating Westinghouse, Combustion

Engineering and Babcock & Wilcox plants.

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page 1-1

1.0 INTRODUCTION

A Rod Ejection Accident (REA) is initiated by the failure of the housing in the upper

head of the reactor vessel where the Control Rod Drive Mechanism (CRDM) attaches.

This failure allows a control rod to be ejected from the core by Reactor Coolant System

(RCS) pressure forces on that control rod or drive mechanism. This rapid ejection

causes a step increase in reactivity in the core increasing core power and peaking

around the location of the ejected rod.

The REA is a postulated accident in which large power increases can occur. Large

power increases potentially challenge RCS integrity from a spike in pressure and core

coolability. The purpose of this topical report is to define a methodology to demonstrate

that in the event of an REA, the appropriate criteria for RCS pressure, core coolability,

and consequences from failed fuel are met.

Since the REA is classified as a design basis accident, the Specified Acceptable Fuel

Design Limits (SAFDLs) are allowed to be exceeded. This methodology estimates the

consequences of an REA and compares the results to criteria that address fuel failure,

coolability, and RCS integrity.

The ARCADIA® Rod Ejection Accident (AREA™) methodology provides a conservative

representation of the reactor response during an REA and demonstrates compliance

with the appropriate criteria. Energy deposition, fuel rim melt, fuel centerline melt,

Minimum Departure from Nucleate Boiling Ratio (MDNBR), and RCS pressure are

considered in the evaluation of the REA. The methodology includes the use of a nodal

3-D kinetics solution with open channel thermal-hydraulics and fuel temperature

feedback and a detailed model that includes an open channel thermal-hydraulic model

with a fuel rod thermal model. These models provide localized neutronic and thermal

conditions to demonstrate compliance with the REA criteria that would be the same as

or similar to those presented in Reference 1.


1.1 Range of Applicability

The AREA™ methodology is applicable to all operating Pressurized Water Reactors

(PWRs) that can be modeled with ARCADIA® and RELAP5 {includes S-RELAP5

(Reference 8) and RELAP5/MOD2-B&W (Reference 9)}. The ARCADIA® Code System

(References 11 and 12) is capable of modeling a variety of PWR reactor types and

sizes over a large range of enrichments with a variety of burnable absorber and control

rod absorber types. The capabilities of the code have been shown to be valid from cold

to hot conditions which cover modes 1 through 6 of plant operation. Since COBRA-

FLX™ is part of the ARCADIA® Code System, it is also validated for modeling of a

variety of plant types with varying core sizes and assembly lattice designs. RELAP5

has been used and approved for Westinghouse, Combustion Engineering (CE), and

Babcock & Wilcox (B&W) plants.

1.2 Topical Report Content

The following discussion provides a general structure for the remaining content of this

topical report. Section 2.0 provides the applicable regulatory guidance for the REA.

Section 3.0 provides a description of the accident scenarios. Section 4.0 contains a

discussion of the important phenomena for an REA. The analytical models are

described in Section 5.0. Section 6.0 contains the AREA™ methodology descriptions.

The sensitivity evaluation, results, and biasing are described in Section 7.0. Application

of the AREA™ methodology is discussed in Section 8.0. An overview of the sample

problems is contained in Section 9.0. The conclusion is summarized in Section 10.0.

References are listed in Section 11.0. Results of the sample problems are provided in

Appendices A, B, and C.


2.0 APPLICABLE REA REGULATORY REQUIREMENTS

This AREA™ methodology is designed to be consistent with the regulatory guidance for

a Reactivity Initiated Accident (RIA). There are two RIAs explicitly addressed within the

regulatory guidance; the REA for PWRs and a control rod drop accident for Boiling

Water Reactors (BWRs). The regulatory criteria which must be met are specified in

10CFR50 Appendix A. The General Design Criteria (GDC) define the criteria for all

aspects of a nuclear plant design to ensure safe operation. Not all GDCs apply to the

REA.

10CFR50 Appendix A requirements apply to all power reactors. Those specific to the

REA event are GDC 13 and GDC 28.

GDC 13: Instrumentation and control. Instrumentation shall be provided to monitor

variables and systems over their anticipated ranges for normal operation, for

anticipated operational occurrences, and for accident conditions as

appropriate to assure adequate safety, including those variables and systems

that can affect the fission process, the integrity of the reactor core, the reactor

coolant pressure boundary, and the containment and its associated systems.

Appropriate controls shall be provided to maintain these variables and

systems within prescribed operating ranges.

GDC 13 provides for the use of prescribed instrumentation and plant design features to

be used to terminate the REA event.

GDC 28: Reactivity limits. The reactivity control systems shall be designed with

appropriate limits on the potential amount and rate of reactivity increase to

assure that the effects of postulated reactivity accidents can neither (1) result

in damage to the reactor coolant pressure boundary greater than limited local

yielding nor (2) sufficiently disturb the core, its support structures or other

reactor pressure vessel internals to impair significantly the capability to cool

the core. These postulated reactivity accidents shall include consideration of


rod ejection (unless prevented by positive means), rod dropout, steam line

rupture, changes in reactor coolant temperature and pressure, and cold water

addition.

In addition to the 10CFR50 Appendix A requirements, the other regulatory requirements

pertaining to the REA event are 10CFR100.11 and 10CFR50.67. Both of these

requirements refer to radiological consequences of an REA event. These two

requirements are not directly addressed by the AREA™ methodology. Rather, they are

indirectly addressed by showing that the number of potential fuel failures and enhanced

release related to an REA is such that the event is not limiting with regards to dose

consequences.

In addition, this report is structured to be consistent with the guidance in the Standard

Review Plan (SRP), NUREG-0800 (Reference 3) Section 15.0.2, Revision 0

(Methodology Guidance) and Section 15.4.8, Revision 3 (Control Rod Ejection

Guidance).

Additional criteria are established to help mitigate the consequences of an REA to

ensure that the regulatory requirements stated in GDC 13 and GDC 28 are met. An

interim set of RIA criteria are defined in Reference 1. An NRC position memorandum

outlining proposed draft criteria has been issued (Reference 2). Clearly, if the draft

criteria are approved, the criteria in NUREG-0800 will change after the submittal of this

topical report. In addition, international RIA tests are planned which indicates that the

bases for the regulatory criteria in both References 1 and 2 may evolve. Therefore, the

AREA™ methodology includes the selection of the appropriate criteria upon application.

2.1 Current Criteria

Excerpts from Reference 1 are shown below.

B. FUEL CLADDING FAILURE CRITERIA


The total number of fuel rods that must be considered in the radiological assessment is

equal to the sum of all of the fuel rods failing each of the criteria below. Applicants do

not need to double count fuel rods that are predicted to fail more than one of the criteria.

1. The high cladding temperature failure criteria for zero power conditions is a peak

radial average fuel enthalpy greater than 170 cal/g for fuel rods with an internal

rod pressure at or below system pressure, and 150 cal/g for fuel rods with an

internal rod pressure exceeding system pressure. For intermediate (greater than

5% rated thermal power) and full power conditions, fuel cladding failure is

presumed if local heat flux exceeds thermal design limits (e.g. DNBR and CPR).

2. The PCMI failure criteria is a change in radial average fuel enthalpy greater than

the corrosion-dependent limit depicted in Figure B-1 (PWR) {Figure 2-1 in this

topical report} and Figure B-2 (BWR).

Fuel cladding failure may occur almost instantaneously during the prompt fuel enthalpy

rise (due to PCMI) or may occur as total fuel enthalpy (prompt + delayed), heat flux, and

cladding temperature increase. For the purpose of calculating fuel enthalpy for

assessing PCMI failures, the prompt fuel enthalpy rise is defined as the radial average

fuel enthalpy rise at the time corresponding to one pulse width after the peak of the

prompt pulse. For assessing high cladding temperature failures, the total radial average

fuel enthalpy (prompt + delayed) should be used.

C. CORE COOLABILITY CRITERIA

Fuel rod thermal-mechanical calculations, employed to demonstrate compliance with

criteria #1 and #2 below, must be based upon design-specific information accounting for

manufacturing tolerances and modeling uncertainties using NRC approved methods

including burnup enhanced effects on pellet power distribution, fuel thermal conductivity,

and fuel melting temperature.

1. Peak radial average fuel enthalpy must remain below 230 cal/g.


2. Peak fuel temperature must remain below incipient fuel melting conditions.

3. Mechanical energy generated as a result of (1) non-molten fuel-to-coolant

interaction and (2) fuel rod burst must be addressed with respect to reactor

pressure boundary, reactor internals, and fuel assembly structural integrity.

4. No loss of coolable geometry due to (1) fuel pellet and cladding fragmentation

and dispersal and (2) fuel rod ballooning.

D. FISSION PRODUCT INVENTORY

The total fission-product gap fraction available for release following any RIA would

include the steady-state gap inventory (present prior to the event) plus any fission gas

released during the event. The steady-state gap inventory would be consistent with the

Non-LOCA gap fractions cited in RG 1.183 (Table 3) and RG 1.195 (Table 2) and would

be dependent on operating power history. Whereas fission gas release (into the rod

plenum) during normal operation is governed by diffusion, pellet fracturing and grain

boundary separation are the primary mechanisms for fission gas release during the

transient.

Based upon measured fission gas release from several RIA test programs, the staff

developed the following correlation between gas release and maximum fuel enthalpy

increase:

Transient FGR = [(0.2286*∆H) – 7.1419]

Where:

FGR = Fission gas release, % (must be ≥ 0)

∆H = Increase in fuel enthalpy, ∆cal/g


The transient release from each axial node which experiences the power pulse may be

calculated separately and combined to yield the total transient FGR for a particular fuel

rod. The combined steady-state gap inventory and transient FGR from every fuel rod

predicted to experience cladding failure (all failure mechanisms) should be used in the

dose assessment. Additional guidance is available within RG 1.183 and 1.195.

2.2 NRC Proposed Changes to Criteria

Excerpts from Reference 2 are used to show the proposed changes to the criteria.

(Paraphrasing is used.)

B. FUEL CLADDING FAILURE CRITERIA

1. For zero power conditions, the high temperature cladding failure threshold is

expressed in the following relationship, as shown in Figure 3.2.1-5.

o Cladding differential pressure < 1.0 MPa,

Peak radial average fuel enthalpy = 170 cal/g

o Cladding differential pressure > 1.0 MPa, < 4.5 Mpa

Peak radial average fuel enthalpy = 170 – ((∆P – 1.0)*20) cal/g

o Cladding differential pressure > 4.5 MPa,

Peak radial average fuel enthalpy = 100 cal/g

Predicted cladding differential pressure must consider the impact of transient

FGR on internal gas pressure. An acceptable means of determining the amount

of transient FGR is described in Section 3.5 of this report. … (DNBR remains the

same) …


2. The PWR PCMI failure criteria is a change in radial average fuel enthalpy greater

than the corrosion-dependent limit depicted in Figure 3.2.2-21 {Figure 2-2 in this

TR} and Figure 3.2.2-22 {Figure 2-3 in this TR for Fully Recrystallized Annealed

(RXA) clad and Stress Relief Annealed (SRA) clad, respectively. }

C. CORE COOLABILITY CRITERIA

2. A limited amount of fuel melting is acceptable provided it is restricted to (1) fuel

centerline region and (2) less than 10% of any pellet volume. For the outer 90%

of the pellet volume, peak fuel temperature must remain below incipient fuel

melting conditions. Fuel temperature predictions must be based upon design-

specific information accounting for manufacturing tolerances and modeling

uncertainties using NRC approved methods including burnup-enhanced effects

on pellet radial power distribution, fuel thermal conductivity, and fuel melting

temperature.

… However, until regulatory guidance exists to address items #3 and #4 above,

applicants need only demonstrate compliance to coolability criteria #1 and #2.

D. FISSION PRODUCT INVENTORY

The revised transient FGR correlations are listed below. The total fission product

inventory is equal to the steady state gap inventory plus the transient FGR derived with

these correlations.

Peak Pellet BU < 50 GWd/MTU: Transient FGR (%) = [(0.26 * ∆H) - 13]

Peak Pellet BU > 50 GWd/MTU: Transient FGR (%) = [(0.26 * ∆H) - 5]

Where:

FGR = Fission gas release, % (must be > 0)

∆H = Fuel enthalpy increase (∆cal/g)


These transient FGR correlations supersede the correlation derived in Reference 4 and

presented in DG-1199.

2.3 Future Criteria

The AREA™ methodology is flexible and capable of demonstrating compliance with

potential revisions to the REA criteria in Reference 1. The codes that support the

AREA™ methodology (ARTEMIS™, GALILEO™, and RELAP5) are capable of

performing calculations that demonstrate compliance with various formulations of

criteria related to enthalpy, Departure from Nucleate Boiling Ratio (DNBR), fuel

temperature, fuel pin pressure, transient Fission Gas Release (FGR), and RCS

pressure.

2.4 Maximum RCS Pressure

The REA overpressure acceptance criteria are taken from NUREG-0800 SRP Section

15.4.8, Revision 3 (Reference 3). These acceptance criteria specify that the peak RCS

pressure does not result in stresses that exceed the "Service Level C" limits as defined

in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel

Code. Consistent with the ASME requirement, maintaining the RCS pressure below

120% of the system design pressure is used to demonstrate compliance with the

requirement. The pressure limit for the REA is established in the plant licensing bases.

The AREA™ methodology supports either limit.

To show compliance with the ASME requirements, Regulatory Guide (RG) 1.77

(Reference 17) is used. The RG 1.77 guidance are:

• Calculations based on conventional heat transfer from the fuel

• A conservative metal-water reaction threshold

• Prompt heat generation in the coolant to determine heat flux variation and volume

surge

• Volume surge used in the pressure transient calculation


• Account for fluid transport in the RCS

• Heat transfer to the steam generators

• Credit action of the pressurizer relief and safety valves

• No credit for pressure reduction caused by the failure of a CRDM pressure housing


Figure 2-1 Corrosion Limit Based on Relative Oxide Thickness

*This figure is extracted from Reference 1.


Figure 2-2 Corrosion Limit Based on RXA Clad Type and Excess Hydrogen

*This figure is extracted from Figure 3.2.2-21 in Reference 2.

**The failure value of 95 cal/g at 130 wppm is used and the plotted value is ignored for the sample problems.

**


Figure 2-3 Corrosion Limit Based on SRA Clad Type and Excess Hydrogen

*This figure is extracted from Figure 3.2.2-22 in Reference 2.


3.0 ROD EJECTION ACCIDENT SCENARIO IDENTIFICATION

The REA is postulated to occur from a mechanical failure of the CRDM pressure

housing resulting in a fast ejection of a Rod Control Cluster Assembly (RCCA) or a

Control Element Assembly (CEA) along with the drive shaft. The consequence of this

mechanical failure is a rapid reactivity insertion resulting in a rapid increase in power

and an adverse core power distribution. The rapid power increase in conjunction with a

skewed power distribution can challenge thermal and mechanical limits of the fuel and

system. Fuel failures can challenge radiological release limits for the plant and

catastrophic failures can challenge system integrity (GDC 28).

3.1 Reactivity Insertion

An REA can be considered a fixed reactivity insertion event. The REA can be

categorized by two distinct types.

• Reactivity (ρ) inserted is greater than the effective delayed neutron fraction (β) or

prompt critical

• Reactivity inserted less than the β or sub-prompt critical.

Technical Specification limits for PWRs define the allowed control rod positions with

respect to power level which are referred to in this method as Power Dependent

Insertion Limits (PDILs). The PDILs allow more than one bank of control rods to be

inserted at low powers and typically only one bank partially inserted at full power. In

general, a core containing more inserted control rod banks and/or more deeply inserted

positions, results in higher ejected rod worths. Hence, the highest ejected rod worths

occur at low powers and can result in prompt critical power excursions. At high powers

the ejected rod worths are lower and result in sub-prompt critical power excursions.

The prompt critical and sub-prompt critical power excursions are quite different and are

explained using simple analytic expressions from point kinetics. Sections 3.1.1 and

3.1.2 discuss the characteristics of prompt critical and sub-prompt critical power

excursions, respectively.


3.1.1 Prompt Critical

For ρ>β, the core is prompt critical and a simple relationship between energy deposited

from a power pulse and other core parameters (Reference 19) is shown below:

=2 ∗ − ) ∗

where:

- Energy deposited

ρ - step reactivity insertion (ejected rod worth)

β - effective delayed neutron fraction

- heat capacity of the fuel

- Doppler Temperature Coefficient (DTC)

For this condition, the power increase is fast (peak powers can be reached in terms of

milliseconds) where only terminates the prompt power excursion (see Figure 3-1).

After the maximum power is achieved, core power decreases at a rate similar to the

increase and continues to decrease to a much lower power level where it remains

relatively constant (referred to as "the residual power level") until a reactor trip occurs.

Since heatup of the fuel is very fast, fuel/clad thermal-mechanical processes are very

complex. Because of this complexity, limits are based upon tests that measure thermal

energy of the fuel during the event (enthalpy based limits). The Nuclear Regulatory

Commission (NRC) has correlated the results from these tests to establish the enthalpy

limits. From the above equation for prompt critical excursions, the key parameters for

REA that affect the energy deposition are the ejected rod worth (ρ), the effective

delayed neutron fraction (β), the heat capacity ( ), and the DTC ( ).


For Hot Zero Power (HZP) conditions, if the initial flux is very low, the control rod can be

fully ejected before any fuel temperature feedback occurs; hence, the highest peak

power is reached when starting at very low powers because no feedback occurs during

the reactivity insertion.

3.1.2 Sub-Prompt Critical

For ρ<β, the core power excursion is limited by delayed neutrons. The multiplication of

the prompt neutron production reaches a peak that is represented by the simple

analytical prompt-jump expression shown below.

Pj/Po = β / (β – ρ)

where:

Pj - prompt-jump power

Po - initial power

β - effective delayed neutron fraction

ρ - step reactivity insertion

Following the prompt-jump, the power tends to approach the power level (referred to as

"the residual power level") where the feedback from the Moderator Temperature

Coefficient (MTC) and DTC is balanced with ρ. This progression with time is highly

dependent upon the rate of delayed neutron buildup and the feedback response to the

heatup (see Figure 3-2). The thermal conditions that can occur after this prompt-jump

result in higher fuel temperatures and higher heat fluxes that can result in fuel failures.

For a sub-prompt critical rod ejection from HZP, the prompt-jump occurs from the initial

power and the core power escalates over a period of many seconds to minutes. For a

prompt-jump REA, the results are more limiting when initiated from a higher power for

the same ejected rod worth. For REAs occurring at power as ρ approaches and

exceeds β there is a smooth transition as the prompt rise turns into a pulse.


3.2 RCS Pressure

The control rod ejection accident is postulated in RG 1.77 (Reference 17) to occur due

to “a mechanical failure of the CRDM housing such that the RCS pressure would eject

the control rod and drive shaft to the fully withdrawn position”. Historically, the REA

methodologies have evaluated two scenarios. The first scenario is for the determination

of the maximum RCS pressure. The second scenario is for the determination of fuel

failure due to DNB or other causes.

RG 1.77 specifies a number of elements of an REA methodology. Two of these

elements are important for the REA scenarios with respect to the reactor coolant

pressure as stated in RG 1.77.

• “No credit should be taken for the possible pressure reduction caused by the

assumed failure of the control rod pressure housing.”

• “It should be assumed that clad failure occurs if the heat flux equals or exceeds the

value corresponding to the onset of the transition from nucleate boiling (DNB), or for

other appropriate causes.”

The postulated failure of the CRDM housing can result in a breach of the reactor coolant

vessel ranging in size from zero (ejected control rod plugs the hole) to something less

than the size of the hole associated with the mechanical failure. The postulated

mechanical failure for the REA leads to a large uncertainty in the amount of coolant that

would be lost and the rate of that loss of coolant.

Historically, the first scenario assumes that there is no coolant leakage from the

mechanical failure. The scenario is evaluated to calculate the energy deposited in the

coolant from the accident to determine the maximum RCS pressure. The assumption of

zero coolant loss is conservative from a maximum RCS pressure perspective. The

AREA™ methodology uses this first scenario to evaluate the maximum RCS pressure.

The maximum RCS pressure determination is addressed further in Section 6.7.1.


Historically, a second scenario assumes that the reactor RCS pressure is held constant

at the initial value for the accident. This scenario is necessary because the first

scenario results in an increase in the RCS pressure which improves Departure from

Nucleate Boiling (DNB.) An additional conservatism (relative to the assumption of

constant initial pressure) that is applied to this second scenario in the AREA™

methodology is addressed in Section 6.7.2.


Figure 3-1 Prompt Critical Power Excursion

Figure 3-2 Sub-Prompt Critical Power Excursion (Prompt-Jump)


4.0 PHENOMENA IDENTIFICATION RANKING TABLE (PIRT) EVALUATION OF REA MODEL REQUIREMENTS

This section addresses parameters to be modeled and/or considered in the AREA™

methodology. The three aspects of the REA that address relevant regulatory guidance

are:

• Integrity of the fuel pin during the prompt power pulse

• Potential failures due to overheating after the power excursion (DNBR)

• Integrity of the RCS due to potential over pressurization

4.1 Fuel Pin Integrity During a Prompt Power Pulse

Fuel pin integrity during a prompt power pulse has been characterized in Reference 6

and divided into two parts, the system transient and the fuel rod transient. A list of the

phenomena, their “importance ratio” and “knowledge ratio” is presented in Table 4-1 for

the plant transient analysis.

A similar list is presented in Table 4-2 for fuel and clad temperatures. [

] Therefore, these items are not included in

Table 4-2.

Reference 6 states that the phenomena with importance ratios above 75 are important

and those with knowledge ratios above 75 are well known. It also states that

parameters near the threshold should be considered.


4.2 DNBR

Additional parameters are added to address the impact on DNBR since the scope of

Reference 6 was primarily concerned with PCMI type failures and not DNBR. Each of

the parameters listed in Table 4-3 are addressed with respect to the requirements to

bound, apply uncertainty, or demonstrate a negligible consequence.

4.3 System Pressure

The RCS pressure response can be affected by the system parameters in Table 4-4. In

addition, RCS pressure can be affected by the core parameters presented in

Table 4-3.

4.4 Regulatory Criteria for an REA

The importance of each parameter is tested or evaluated in the AREA™ methodology

relative to its effect on fuel temperature, fuel rim temperature, enthalpy rise, total

enthalpy, MDNBR, and/or RCS pressure. Section 2.0 presents the acceptance criteria

for the REA. Section 7.0 provides a discussion on the parameter investigations.


Table 4-1 PIRT Plant Transient Analysis

Subcategory Phenomenon IR* KR** Calculation of power history during pulse (includes pulse width)

Ejected control rod worth 100 100 Rate of reactivity insertion 61 88 Moderator feedback 38 93 Fuel temperature feedback 100 96 β 95 96 Reactor trip reactivity 0 96 Fuel cycle design 92 100

Calculation of rod fuel enthalpy increase during pulse (includes clad temperature)

Heat resistances in high burnup fuel, gap, and clad (including oxide layer)

58 67

Transient clad-to-coolant heat transfer coefficient

56 64

Heat capacities of fuel and clad 94 90 Fractional energy deposition in pellet 4 93 Pellet radial power distribution 63 88 Rod peaking factors 97 100

Notes: * Importance ratio IR>75 important **Knowledge ratio KR<75 not completely understood

Table 4-2 PIRT Fuel Rod Transient Analysis for Fuel and Clad Temperatures

Subcategory Phenomenon IR* KR** Initial conditions Pellet and clad dimensions 91 96

Burnup distribution 55 89 Clad oxidation 46 73 Power distribution 100 89 Coolant conditions 93 96 Transient power specification 100 94

Fuel and clad temperature changes

Heat resistances in fuel, gap, and clad 75 77 Transient clad-to-coolant heat transfer coefficient (oxidized clad)

50 58

Heat capacities of fuel and clad 88 93 Coolant conditions 85 88

Notes: * Importance ratio IR>75 important **Knowledge ratio KR<75 not completely understood


Table 4-3 Parameters Directly Addressed by AREA™ Methodology

Neutronic Thermal (Neutronic and Detailed Model)

Other

Ejected rod worth Fuel conductivity Computational accuracies* β Gap conductance Manufacturing tolerances* Moderator feedback Clad conductivity, oxide Fuel temperature feedback Heat capacity of fuel Rate of reactivity insertion Heat capacity of clad Neutron velocities* Direct energy deposition

in coolant

Reactor trip reactivity Pellet radial power profile

Ejected rod location* RCS pressure* Excore flux* RCS temperature* RCS flow Peaking * Parameters added for DNBR considerations and completeness

Table 4-4 System Parameters Considered for Pressure Analysis

Overpressure Pressure Decrease for DNBR

Initial RCS pressure Initial pressure Initial pressurizer level Initial pressurizer level Initial RCS temperature Initial RCS temperature Trip setpoints RCS breach area Pressurizer safety valve settings and uncertainties

Secondary heat removal settings

Non-safety systems


5.0 ANALYTICAL MODELS

The AREA™ methodology is capable of evaluating an REA to demonstrate compliance

with the acceptance criteria discussed in Section 2.0. The methodology requires the

following analytical models:

• GALILEO™ (Reference 7) {COPERNIC (Reference 16) can also be used if the

outlined validations are performed}

• ARTEMIS™ ( References 11 and 12), a coupled 3-D kinetics solution with

neutronics, fuel rod thermal model, and 3-D thermal hydraulic model

• COBRA-FLX™ (Reference 13) as the 3-D thermal hydraulic model implemented in

Reference 12

• S-RELAP5 (Reference 8) for Westinghouse and CE plants or RELAP5/MOD2-B&W

(Reference 9) for B&W plants

Figure 5-1 shows the coupling of the time dependent models. The fuel performance

code is the source of thermal properties of the fuel, clad, and gap for the time

dependent models which is why it is not shown in Figure 5-1. The ARTEMIS™ nodal

and detailed model are approved in Reference 11. The interface with RELAP5 is

introduced in this topical report. As shown in Figure 5-1 three distinct models can be

used together with information exchange between the models where appropriate. A

description of these models follows.

• [

]


- [

]


• [

]

5.1 GALILEO™

GALILEO™ is the fuel performance code that provides the following information

pertinent to the AREA™ methodology:

• Enthalpy rise criteria functionalized by clad corrosion is converted to enthalpy rise

limits versus burnup

• Fuel thermal properties with burnup dependencies for the time dependent solutions

of temperature

• Fuel pin internal pressure to determine fuel enthalpy limits for high clad temperature

failure criteria

COPERNIC can also provide this information. For the AREA™ topical report, whenever

GALILEO™ is used, COPERNIC can also be used with differences noted.


5.1.1 Enthalpy Rise Limits

The enthalpy rise criteria in Section 2.3 are based on clad corrosion in terms of either

relative oxide thickness or excess hydrogen content. The corrosion model in

GALILEO™ for oxide thickness or hydrogen uptake is used to maximize the corrosion

obtained at a given burnup to obtain an enthalpy rise limit with burnup.

5.1.2 Thermal Properties

GALILEO™ is used to define fuel and clad thermal properties for the fuel rod model

used by both the neutronics solution and the thermal-hydraulic solution in ARTEMIS™.

This fuel rod model is described in Section 5 of Reference 12. These properties include

fuel and clad thermal conductivity which includes clad oxide formation, heat capacity for

the fuel pellet and clad, radial power distribution in the fuel pellet, porosity of the fuel,

and gap conductance. Fuel burnup affects fuel conductivity, pellet radial power profile,

and clad oxide thickness. Either thermal property equations are used directly or input

as polynomial equations in the ARTEMIS™ fuel rod model.

Gap conductance is a complex function of gap and surface temperatures, gap size

(i.e., creep and thermal expansion), contact pressure, and fission gas content. To

capture these effects in downstream codes using a [ ]

the gap conductance (Section 5.3 of Reference 12,) is [

]


5.1.3 Fuel Pin Pressure

The internal pressure of the fuel pin is needed to determine the high clad temperature

failure criteria or to resolve potential ballooning coolability issues with fuel pins

exceeding the critical heat flux.

5.2 ARCADIA®

The ARCADIA® code system is a neutronics, fuel thermal and thermal-hydraulic code

that performs core design and safety evaluations. It has 3-D neutronics static and

transient solvers with time dependent fuel and coolant models. It is used as the core

transient model for AREA™. It is capable of calculating all neutronics and thermal

effects discussed in Section 4.0 that are needed to demonstrate compliance with the

criteria listed in Section 2.0.

5.2.1 ARCADIA® Validation

Validation of ARCADIA® is provided in References 11 and 12. Table 5-1 contains the

neutronics and fuel temperature validation matrix of ARCADIA® specific to AREA™.

The thermal-hydraulic model in ARCADIA® is COBRA-FLX™ (Reference 13) as

described in Section 5.3.


5.2.2 Verification of Gap Conductance and Thermal Conductivity Models

Comparisons between GALILEO™ and the ARTEMIS™ fuel thermal model are

performed to verify the use of the [ ] described in Section

6.2.5. Representative rod ejection transients starting at HZP and Hot Full Power (HFP)

conditions are used for the verification. This comparison highlights any significant

differences between the ARTEMIS™ fuel thermal model and a more detailed treatment

of the fuel rod thermal properties in GALILEO™. [

]

[

]


Summary statistics are shown in Table 5-2, Table 5-3, and Table 5-4 for UO2 fuel, 2

wt% gadolinia fuel, and 8 wt% gadolinia fuel, respectively. The average, minimum, and

maximum ratios of ARTEMIS™ to GALILEO™ results are shown for each transient

simulation for the centerline, the average and surface temperatures of the fuel pellet

along with the internal surface and average temperatures of the clad. The standard

deviation in units of percent is also shown.

For all cases, the trends of the differences are well behaved and the differences in

maximum fuel centerline temperature are [ ] having

the largest differences. Some of the larger differences are examined in more detail. For

UO2 fuel at EOL and HZP both the centerline and the surface fuel temperatures have

minimum ratios of [ ] respectively. For the HZP transient simulation,

the rim temperature is the peak fuel temperature during the power pulse. For this

reason, the maximum rim temperature is of more interest than the surface temperature

since the surface is cooler than the pellet just inside the surface. The centerline and

maximum rim temperature plots are shown in Figure 5-2 and Figure 5-3, respectively.

The behavior is well captured by ARTEMIS™ using [ ]

For 2 wt% gadolinia fuel the HFP MOL case is examined for centerline and surface

temperatures as shown in Figure 5-4 and Figure 5-5. For this HFP transient the surface

and centerline temperatures are examined since they provide the temperature extremes

for the pellet. Most of these transient differences are the same as the steady state

temperature differences and simply propagate the difference through the transient.


For 8 wt% gadolinia fuel at HZP EOL conditions, the centerline and rim temperature are

shown in Figure 5-6 and Figure 5-7, respectively. [ ]

ARTEMIS™ is capable of modeling the fuel temperature behavior with respect to time

for the pellet centerline, the rim and the pellet surface. The peak centerline temperature

is predicted to be within [

] The rim effect is quite complex during a prompt critical REA where heat

flow occurs in both directions, toward the centerline and toward the clad. The results

above show that the maximum rim temperature is [

]

5.3 COBRA-FLX™

The COBRA-FLX™ core thermal-hydraulic code is AREVA’s latest development for

performing nuclear core thermal-hydraulic simulations. COBRA-FLX™ is the thermal-

hydraulic code module used in the core simulator ARTEMIS™. COBRA-FLX™ is

incorporated into the ARTEMIS™ code in its entirety. Within ARCADIA®,

COBRA-FLX™ can be used as part of ARTEMIS™ or stand-alone. The AREA™

methodology uses COBRA-FLX™ through ARTEMIS™. COBRA-FLX™ is used for

both the nodal simulator and the detailed model. The detailed model is a [

]


5.3.1 COBRA-FLX™ Validation

Validation of COBRA-FLX™ is provided in Reference 13. COBRA-FLX™ is an integral

part of the ARTEMIS™ moderator feedback solution in References 11 and 12. Table

5-5 contains the COBRA-FLX™ validation matrix specific to AREA™.

5.4 RELAP5 Computer Code

The NRC approved S-RELAP5 (Reference 8) is used in the automatically coupled

analysis for Westinghouse and CE plants and the NRC approved RELAP5/MOD2-B&W

(References 9 and 15) is used for the manually coupled calculation for B&W plants.

These codes generically referred to as RELAP5, have been previously approved for use

in REA analysis.

The purpose of the RELAP5 computer code for AREA™ is twofold: 1) to calculate the

pressure response during an REA based on taking no credit for the possible pressure

reduction caused by the assumed failure of the CRDM pressure housing, and 2) to

provide a pressure boundary condition to the core transient model for the DNBR

calculation. The RELAP5 computer code models the primary and secondary systems

that determine the change in RCS pressure, inlet temperature, and/or flow during an

REA.

Separate RELAP5 analyses are performed to determine the maximum pressure

scenario and DNBR scenario RCS responses. The biasing and uncertainties from the

sensitivity studies (Section 7.0) that maximize the energy deposited in the coolant are

used to generate the forcing function for input into RELAP5 maximum pressure

calculations. Sensitivities using RELAP5 are also performed to determine conservative

system biases and settings for maximum pressure calculations and core pressure

calculations for MDNBR.


5.4.1 S-RELAP5 Code and Model

The system code used for automatically coupled REA analyses is the S-RELAP5

computer code. S-RELAP5 is a general purpose thermal-hydraulic, best estimate,

system computer code that is used for a variety of safety-related and non-safety related

transient calculations. The code modeling capabilities include the simulation of large

and small break loss-of-coolant accidents (LOCAs), as well as operational transients

such as anticipated transient without SCRAM, loss-of-offsite power, loss of feedwater,

loss of flow, and REA.

The S-RELAP5 model used for the automatically coupled REA RCS pressure analysis

is generally consistent in modeling approach and level-of-detail as the models in a

previously approved AREVA methodology (Reference 8). Most of the aspects of the

model are unchanged compared to the previously approved method; however, some

modifications are made as discussed below:

Kinetics Modeling: The S-RELAP5 REA model for the RCS pressure analysis uses a

3-D automatically coupled core nodal model versus a point kinetics model in the

previously approved method. Time-dependent data are transferred from S-RELAP5 to

ARTEMIS™. ARTEMIS™ calculates the 3-D core power response to an ejected rod

and data are transferred to S-RELAP5 which determines the system thermal-hydraulic

response. S-RELAP5 and ARTEMIS™ are coupled via a Message Passing Interface

(MPI).


Sectorized Reactor Vessel: The S-RELAP5 model used by the AREA™ methodology

contains a sectorized reactor vessel model. For example, in the model for a 4-Loop

plant, [

] Each sector consists of [

] A sectorized core [ ] was

reviewed and approved by the NRC in Reference 18 (Section 6.2). The Reference 18

topical report is for use with the RELAP5 code version RELAP5/MOD2-B&W

(Reference 9). Similar to the modeling in Reference 18 (Section 6.2), the sectors are

[

]

Reactor Vessel Upper Head: The reactor vessel upper head contains an increased

number of nodes relative to previous S-RELAP5 models. This increase in the number

of nodes is consistent with the approved modeling in Reference 8. [

]

Mixing Junctions: Mixing junctions are included at the [

] This modeling approach is consistent with that approved by the NRC in

Reference 18 (Section 6.1). The Reference 8 (Section 6.0) S-RELAP5 model [

]


Steam Generator: The number of tube-side (primary side) and shell-side (boiler region)

nodes in the steam generator is increased relative to previous S-RELAP5 models.

[

] This increase in the number of nodes is consistent with the

approved modeling of Reference 8 (Section 3.0). [

]

5.4.2 RELAP5/MOD2-B&W Code and Model

The NRC approved RELAP5/MOD2-B&W (Reference 9) computer code is utilized in the

evaluation of the REA event for a B&W plant. RELAP5/MOD2-B&W is a general

purpose thermal-hydraulic, best estimate system computer code that is used for a

variety of safety-related and non-safety related transient calculations. The code

modeling capabilities include the simulation of large and small break LOCAs, as well as

operational transients such as anticipated transient without SCRAM, loss-of-offsite

power, loss of feedwater, loss of flow, and REA.

The system model utilized in the performance of the REA manually coupled analysis is

developed in compliance with NRC approved BAW-10193 (Reference 15) topical report.

The system model utilized in the REA system analysis includes detailed nodalization of

the reactor vessel, primary system piping, pressurizer, steam generators, and

secondary piping up to turbine entrance. The only modification to the system model is

the removal of the core reactivity components which are replaced with heat structures

that use the power and heat flux response tables that are created from the time

dependent axial power and heat flux shapes generated by ARTEMIS™ (Section 5.2).


Table 5-1 ARCADIA® Validation Test Matrix for AREA™

Parameter Benchmark Comparison Reference

and Section

Accuracy

Ejected rod worth Reference 11 Section 6.2

ITC Reference 11 Section 9.3

DTC Reference 11 Section 6.3

Trip worth Reference 11 Section 9.2

β Section 7.1.4.2 of this topical report

Power peaking Reference 11 Section 10.5 in Table 10-3

Core power versus time for fast reactivity insertion –NEACRP rod ejection

Reference 11 Section 7.1

Core power versus time for fast reactivity insertion –SPERT comparisons


Static fuel temperatures, transient fuel temperatures, and heat fluxes


Excore power versus time for dropped rod transients



Table 5-2 GALILEO™/ ARTEMIS™ Transient Comparisons for UO2 Fuel


Table 5-3 GALILEO™/ ARTEMIS™ Transient Comparisons for 2 wt% Gadolinia Fuel


Table 5-4 GALILEO™/ ARTEMIS™ Transient Comparisons for 8 wt% Gadolinia Fuel


Table 5-5 COBRA-FLX™ Validation Test Matrix for AREA™

Parameter Benchmark Comparison Reference and Section

Accuracy

Conservation of mass and energy


Fluid flow solution Reference 13 Section 5.2

Validity of steady state CHF correlations for transients

Reference 13 Section 5.3.2

Supports approved CHF correlations

Reference 13 Appendix C


Figure 5-1 Coupling of the Time Dependent Models


Figure 5-2 UO2 HZP EOL Transient Fuel Centerline Temperature


Figure 5-3 UO2 HZP EOL Transient Maximum Rim Temperature


Figure 5-4 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Centerline Temperature


Figure 5-5 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Surface Temperature


Figure 5-6 8 wt% Gadolinia Fuel HZP EOL Transient Fuel Centerline Temperature


Figure 5-7 8 wt% Gadolinia Fuel HZP EOL Transient Maximum Rim Temperature


6.0 AREA™ METHODOLOGY DESCRIPTION

This section provides an overview of the AREA™ methodology that is used to

demonstrate compliance with the regulatory guidance addressed in Section 2.0. The

AREA™ calculational process is illustrated in Figure 6-1. The process for each of the

codes in this figure is described relative to its function for the methodology.

6.1 Applicable Regulatory Criteria

As defined in Section 2.0 there are specific regulatory criteria that must be considered

when evaluating the potential consequences of an REA. These criteria are established

at the onset of an REA analysis with AREA™ as they define the limits of the analysis.

6.2 GALILEO™

As shown in Figure 6-1, GALILEO™ is used to generate or provide the basis for the

following:

• Pellet Cladding Mechanical Interaction (PCMI) failure criteria for the clad

• Fuel pin pressure

• Fuel and rim melt temperatures

• Fuel and clad thermal properties

• Gap conductance

The processes for generating the above information are described in the following

sub-sections.

6.2.1 PCMI Failure Criteria for Clad

GALILEO™ is used to convert the failure criteria from corrosion to burnup. It uses a clad

corrosion model and a process to generate acceptable corrosion (oxide or hydrogen

uptake) for fuel pin designs. For the AREA™ methodology, two options are available to

calculate fuel clad corrosion (or internal pin pressure).


1. [

]

The regulatory guidance that provides the PCMI failure limits as a function of corrosion

is used for the appropriate fuel clad type. At each burnup, the clad corrosion from

GALILEO™ in conjunction with the regulatory corrosion based failure criteria is used to

determine the failure limit, typically given in terms of enthalpy rise (∆cal/g). This

provides the failure limit for each clad type deployed in the core.

6.2.2 Fuel Pin Pressure

As described in Section 2.2, there are high clad temperature failure criteria due to

overheating that is expressed as a function of internal fuel pin pressure. A fuel pin

internal pressure calculation is needed to support these criteria. The internal pin

pressure is also used to address potential coolability criteria. Internal fuel pin pressure

needs to account for the heatup of the fuel pin and the amount of transient FGR during

an REA. The process used to calculate the internal pressure is as follows:


A conservative fuel pin pressure versus burnup relationship is generated for the

AREA™ methodology. Either option from Section 6.2.1 is used to obtain limiting

pin pressure information versus burnup. To simulate overheated conditions [

]


6.2.3 Fuel and Rim Melt

Fuel melt temperature for UO2 and gadolinia fuel versus burnup, rim melt temperature

(same as fuel melt at a higher localized burnup), uncertainty of the melt temperature,

and the predicted fuel temperature uncertainty are obtained from GALILEO™. Criteria

regarding centerline melt are established at the time of the plant specific application.

Predictions of fuel melt are conservative when ignoring heat of fusion, convection, and

conduction of melted fuel. If fuel melt is allowed by the regulatory criteria, the melted

volume is used for dose term evaluations and comparison to coolability criteria. The

maximum rim temperature is calculated to ensure no melt in the rim occurs in order to

maintain coolability (see Section 5.2.2).

6.2.4 Fuel and Clad Thermal Properties

Thermal conductivity and heat capacity of the fuel and clad are obtained from

GALILEO™.

If COPERNIC is used, there are two alternatives. [

]

6.2.5 Fuel Pellet to Clad Gap Conductance

[

]


Gap conductance is a complex function of clad and fuel surface temperatures, gap size

(i.e., creep and thermal expansion), roughness, contact pressure, and fission gas

content. [

]

The verification of the gap conductance model is shown in Section 5.2.2.

6.3 ARTEMIS™ Models for REA Event Analysis

Once the GALILEO™ data have been generated, the next phase includes several

ARTEMIS™ models to set the boundary conditions and to perform the REA simulations.

These models include:

• A cycle model (typically developed during the cycle design) is required. This model

uses a [ ] consistent with application of

ARTEMIS™ presented in References 11 and 12.

• Static ARTEMIS™ calculations are used to establish the initial conditions for the

REA event analysis.

• An ARTEMIS™ transient calculation is performed for the specific REA calculations

based on initial conditions defined by the previous steps. This step includes the

setup of all the time-dependent information.


• An ARTEMIS™ [

]

Fuel temperatures are calculated using the ARTEMIS™ fuel rod thermal model while

DNBR calculations are performed using the thermal-hydraulic model COBRA-FLX™.

6.4 ARTEMIS™ (Steady State Nodal Solution)

The ARTEMIS™ steady state analysis defines a matrix of cases at various power levels

and core burnups to define the initial boundary conditions for the REA transient

simulation. The matrix consists of [

] This matrix of cases is selected based on the plant being

modeled. The end points of burnup Beginning of Cycle (BOC) and End of Cycle (EOC)

and power level (HZP and HFP) and [

] The selection of the [

] The following trends are examined for this behavior to

select the intermediate powers and burnups:

• [

]


A core design for the plant of interest is used for plant specific application of this

method. The core design contains the core loading and depletion history of the cycle.

[

]


6.5 ARTEMIS™ (Transient Nodal Solution)

ARTEMIS™ performs 3-D neutronics kinetics simulations with a time dependent, fuel

thermal model and a nodal based COBRA-FLX™ thermal-hydraulic model. The [

] The ejected rod is simulated in the nodal

method by removing a fully inserted control rod in 0.1 seconds. Partially inserted rods

are removed by a time corresponding to the fraction of initial insertion multiplied by 0.1

seconds. The transient is modeled for [

]

Some of the features utilized in the ARTEMIS™ transient calculation are discussed in

the subsections below.

6.5.1 Trip Function

PWRs typically have a high flux trip function using excore detector signals. ARTEMIS™

has the following models to implement a trip function:

• An excore core detector model

• Signal processing

• Control rod drop (SCRAM) function


Excore detectors are located in four nearly symmetric locations around the core, which

causes the excore signal response to differ from the core average value when an

asymmetric rod is ejected. The excore signals generated by the transient simulation of

the REA are compared to the trip setpoint. Once the trip setpoint is reached (three of

the four signals exceed the trip setpoint emulating 2/4 logic with the highest signal

failed), a time delay is employed before the control rods are dropped. Rod position with

time in ARTEMIS™ is defined by the plant licensing basis for the control rod drop

position versus time (SCRAM curve). Physical models for the excore signals and the

dropping of the control rods are discussed in the following subsections.

6.5.1.1 Excore Detector Model

Reactor Protection Systems (RPSs) typically sense and respond to power range excore

detector signals. These signals measure fast flux exiting the reactor core and provide an

indication of the actual incore reactor conditions. The incore assembly powers are

multiplied by excore weighting factors to translate the incore conditions to excore

signals. [

]

6.5.1.2 Signal Processing

ARTEMIS™ simulates the instrumentation and processing that determines a reactor trip

based on excore flux signals. [

] When the trip criteria are reached, the time to start

control rod drop is set based on an input delay time between the trip sensed and start of

physical drop.


6.5.1.3 Control Rod Drop (SCRAM)

Control rod drop can be initiated by a high flux trip signal or other system trip functions

such as high pressure. [

]

Westinghouse and CE plants restrict the flow of water around the control rod as it drops

by reducing the diameter of the guide tube called the dashpot. B&W plants have a

similar mechanism in the CRDM for the lead screw. In this context, a deceleration of the

control rod drop is caused by these devices and is referred to as the dashpot region.

[

]


[

]

6.5.2 Enthalpy Rise

In order to calculate fuel enthalpy rise to assess PCMI failures, the prompt fuel enthalpy

rise is defined as the radial average fuel enthalpy increase from initial conditions to the

time corresponding to one pulse width after the peak of the prompt pulse (Reference 1).


For power excursions where the ejected rod worth (ρ) is less than β, the power rise is

much smaller and the characterization of the power rise and decline is no longer a

prompt pulse. The enthalpy rises during a prompt power excursion and a non-prompt

power excursion are shown in Figure 6-3. The prompt enthalpy rise is clearly seen in

the top figure. However, the bottom figure does not have a prompt rise in enthalpy even

though it has a pulse width of approximately 300 milliseconds. For the AREA™

methods [

]

6.5.3 Adjustment Factors

ARTEMIS™ adjustment factors are used to account for uncertainty and conservative

allowances. These adjustment factors are used in the AREA™ methodology on the

following parameters:

• [

]


[

] These adjustments are applied to examine sensitivities. Following the

sensitivity analysis, these parameter adjustments are used to bias the limiting cases as

described in Section 7.0.

6.6 Transient COBRA-FLX™ Calculations

In addition to being the nodal simulator, ARTEMIS™ is also the driver for the

COBRA-FLX™ solution. The COBRA-FLX™ solution is directly coupled with the

time-dependent fuel thermal model in ARTEMIS™. [

]

6.6.1 Adjustment Factors

In ARTEMIS™, there are adjustment factors that can be used to account for uncertainty

and conservative allowances for the detailed model calculations that are applied to the

fuel thermal model and/or COBRA-FLX™. These adjustment factors are multipliers or

adders to the following parameters:


• [

]

[

] These adjustments are applied to

examine sensitivities. Following the sensitivity analysis, these parameter adjustments

are used to bias the limiting cases as defined in Section 7.0.


6.6.2 DNBR Critical Heat Flux Correlations

The AREA™ method for DNBR calculations uses approved CHF correlations which are

used in COBRA-FLX™ detailed model calculations. The regulatory guidance from

Reference 1 and Reference 2 states that the DNBR SAFDL should be used as a failure

criterion for powers greater than 5%. [

]

6.6.3 Mixed Core Applications

For a mixed core configuration, COBRA-FLX™ can be used to [

] hydraulic resistances (pressure loss coefficients) and

other hydraulic and physical characteristics or an NRC approved mixed core

methodology can be used.

6.7 RELAP5

The RELAP5 computer code is used for RCS pressure calculations. There are two

scenarios described in Section 3.2 for the REA that are evaluated. The first scenario is

related to the determination of the maximum system pressure and the second scenario

is related to the core pressure used for the determination of DNB.


6.7.1 RCS Pressure Evaluations

The REA scenario for over pressurization is maximized when no credit is taken for the

possible pressure reduction caused by the assumed failure of the CRDM housing. The

REA event can be terminated by one of two types of reactor trip functions: (1) high

neutron flux/high core power or (2) high RCS pressure/high pressurizer pressure.

Which reactor trip occurs is dependent on the magnitude of the reactivity insertion.

A large prompt reactivity insertion results in a high neutron flux/high core power reactor

trip within one second of event initiation. Although the peak neutron power for this

scenario is extremely high, the heating of the coolant is delayed because the energy is

initially deposited into the fuel and then must be conducted to the coolant. Hence, a

power excursion that does not trip on high flux continues to deposit energy into the RCS

that can result in higher pressures. The AREA™ methodology addresses the transition

between the high flux trip and no trip scenarios for REAs.

If the core power excursion is not matched by a similar secondary heat removal over

time, a reduction in steam generator inventory can occur. If the event extends long

enough, the loss of secondary inventory can lead to a reduction in the steam generator

heat removal and cause a more rapid pressure increase.

6.7.2 Pressure for DNB Evaluations (Scenario 2 Section 3.2)

As an additional conservatism for the DNBR analysis, a more conservative value is

used for the core pressure (relative to the assumption of constant initial pressure) than

has historically been used for this scenario. The core pressure used for the evaluation

of DNBR (and other fuel criteria) in the AREA™ methodology is [

]


1. [

]

The focus for the rod ejection event is the short term potential for high energy deposition

in the fuel and then in the coolant that could challenge the coolability criterion and the

system pressure criterion. This is the GDC 28 requirement. Thus, the focus of the

AREA™ methodology for demonstrating compliance with the fuel failure criteria is

placed on the assessment of the consequences to the fuel that are a direct result of the

rapid energy insertion that follows the control rod ejection.

[

]


6.8 Data Processing

Results from the transient nodal and detailed model solutions are processed to provide

tables and figures for the AREA™ methodology. The processing performed relative to

regulatory limits and criteria are discussed below.

6.8.1 PCMI Failure Criteria

The PCMI failure limits from Section 6.2.1 for all applicable clad types are input as a

function of fuel pin burnup. [ ] the difference

between the calculated enthalpy rise (see Section 6.5.2) and the clad limit for that fuel

pin is calculated and displayed as a function of burnup. The differences are analyzed to

determine if failure occurs. The maximum difference is recorded for the core (a negative

difference is less than the limit and yields no failures in the core). If a positive difference

is reached for a fuel pin, then it is counted as failed and coolability issues may need to

be addressed relative to the regulatory criteria defined for the AREA™ application.


6.8.2 Total Enthalpy for High Clad Temperature Failure Criteria

Different limits for these criteria are specified in References 1 and 2. For Reference 1

the failure limit is 170 cal/g when the internal fuel pin pressure is less than core

pressure and 150 cal/g for fuel pin internal pressures above core pressure.

Reference 2 states that this enthalpy limit is functionalized with rod internal pressure.

When the high clad temperature failure is a function of internal fuel pin pressure, the

AREA™ methodology uses the [

] Total enthalpy calculations are performed for all

cases. For prompt critical power excursions, the differences between the total enthalpy

and the fuel high clad temperature failure limit are analyzed to determine if failure

occurs. The maximum difference is recorded for the core (a negative difference is less

than the limit and yields no failures in the core). If a positive difference is reached [

] The regulatory criteria for total enthalpy, for

high clad temperature failures, and any coolability issues relative to this type of failure

are defined in the plant specific application of the AREA™ methodology.


6.8.3 Fuel Melt Failure Criteria

The fuel melt temperature is a function of burnup and fuel pellet material (i.e., in

GALILEO™, burnup and gadolinia content reduces the melt temperature). For [

] the melt temperature for the pellet and rim is calculated. The

melt temperature function from GALILEO™ is [

] The difference between the maximum temperature of the pellet and the melt

temperature is analyzed to determine if the melt temperature is reached. The maximum

difference is recorded for the core (a negative difference is less than the limit and yields

no failures in the core). If a positive difference is reached [ ] it is either

unacceptable if not allowed or it is counted as failed and coolability issues may need to

be addressed relative to the regulatory criteria used for the method. If the melt criteria

have a volume or location requirement, it is checked for acceptability. [

]

The plant specific application defines the applicable regulatory criteria for fuel melt that

are used by the AREA™ methodology.

6.8.4 Coolability

The coolability criteria from References 1 and 2 are summarized as follows:

1. A peak radial average fuel enthalpy cannot be exceeded.

2. No fuel rim melt is allowed and centerline melt is either precluded or <10% is

allowed.


3. Mechanical energy generated as a result of (1) non-molten fuel-to-coolant interaction

and (2) fuel rod burst must be addressed with respect to reactor pressure boundary,

reactor internals, and fuel assembly structural integrity.

4. No loss of coolable geometry due to (1) fuel pellet and clad fragmentation and

dispersal and (2) fuel rod ballooning is allowed.

With these coolability criteria, the following AREA™ processes are presented:

1. Maximum enthalpy is calculated and can be shown to meet the stated criterion.

2. The rim temperature is precluded from exceeding the fuel melt temperature and the

amount of fuel near the centerline that is to be precluded from melting can be

demonstrated.

3. Failures that occur during the power pulse could lead to significant energy deposition

to the coolant because [

]

These failures do not pose a coolability concern relative to coolability criterion and

are not precluded. DNBR and fuel melt failures (if allowed) are included in the fuel

pin failure census related to dose calculations.


4. Exceeding CHF causes the fuel to overheat. [

]

6.9 Fuel Failures

The AREA™ methodology uses [

] the failure criteria defined. The number

of failures is used in determining dose consequences.

6.10 Radiological Consequences

The AREA™ methodology only addresses the source term for the number of fuel pins

failed during an REA. The design basis dose evaluation is plant specific and is not

defined here. Consideration is also given to the fission-product gap inventory for an

REA which is defined in the interim acceptance criteria (Reference 1) and in

Reference 5. The amount that the radiological source terms increase due to REAs is

defined by the regulations and is specified in a plant specific application of the AREA™

methodology.


[

]

6.11 Update Process

There are many situations that might require an update to processes, codes, or

libraries. These include but are not restricted to:

• Improved computer models (first principle or empirical models)

• Data processing is [

]

• Incorporate an improvement in the input or output data structure (these types of


changes have no impact on code numerics and do not require NRC review)

• Improvements, updates, or use of new data libraries (e.g., gap conductance models)

• Updates to codes used by the AREA™ methodology

For all codes supporting AREA™, test cases are included in the code test suite

qualification with respect to its application to AREA™. For those codes that have an

NRC approved update process, those processes are followed to support the AREA™

methodology.

Codes supporting AREA™ without an NRC approved update process use the following

update process when methodology updates are necessary:

• Documentation justifying the required modifications

• Execution of test cases including regression testing

• Updated documentation for theory and users manuals

• Validation testing to show continued applicability to AREA™

• Generation of a summary report documenting the code updates impacting AREA™

along with the validation testing results to be provided to the NRC.

Code updates are allowed for any code supporting AREA™. AREVA maintains a

quality program (including software quality) that is compliant with 10CFR50 Appendix B

requirements. This quality program assures updates are made within the bounds of

NRC licensing requirements for safety evaluations.


Updates are defined as changes in the method that improve the man/machine interface

through better input and output processing and checking, enhance the computational

performance, improve numerical robustness, accelerate convergence, etc. AREVA will

perform such updates as necessary to maintain modern, flexible software that is easy to

use and computationally efficient. Modifications or updates that have a significant

impact as described in Section 6.12 of this topical report will not be implemented unless

they are submitted to the NRC for review and approval. Updates that do not have a

significant impact as defined in Section 6.12 will be summarized in a letter to the NRC

for information. Examples include:

• Source Coding and Structure: Changes in source coding and code structure that

improve the readability and maintainability of the computer codes supporting

AREA™.

• Numerical Methods and Software Architecture: Changes in the numerical

methods may be made to improve computational efficiency and numerical accuracy.

Examples include: improvements to code convergence and numerical algorithms,

improvements to the temporal coupling, implicit coupling, and

parallelization/vectorization of the solution and coupling.

• Computational Platform and Compilers: Movement to newer computational

platforms and compilers may be made as new platforms and compliers become

available.

• Updating Physical Models and Correlations: Updates and improvements in

physical models and correlations may be made as new data or expanded

assessments become available. These updates and improvements are a necessary

element of maintaining a modern and accurate methodology; one that remains state

of the art.


Flexibility to perform discretionary updates is important to maintaining modern and

robust computer codes. For instance, making updates and improvements to physical

models and correlations (that have no more than a small impact on the results) is a

necessary element to expand the robustness of the application. This flexibility provides

AREVA the ability to maintain the AREA™ methodology so that it keeps pace with

subsequent updates and improvements from new data or expanded assessments and

to keep pace with potential changes in regulatory guidance.

It is foreseen that NRC approval may be granted for updates to approved codes and/or

correlations that revise or extend a code’s capabilities for use with AREA™. If future

regulatory commitments are made relative to the approved codes supporting AREA™,

the changes affecting AREA™ will be incorporated without further NRC notification or

request for renewal/approval.

6.12 Level of Significance

The following definition is used to classify a significant update as it affects the results to

the dependent variables listed in Section 7.1.1, when determining the impact of updates

to computer codes, correlations or data libraries:

• [

]


These conditions are consistent with the biasing method described in Section 7.0.

6.13 Method Summary

The AREA™ methodology (Figure 6-1) provides a generic approach to analyze an REA.

The methodology provides the flexibility to perform the REA analysis based on criteria

specified for enthalpy rise, total enthalpy, DNBR, fuel rim temperature, fuel centerline

temperature, and RCS pressure. AREA™ uses the 3-D ARTEMIS™ nodal transient

code with [

] Capability of analyzing fuel pin internal pressure has been

incorporated to evaluate coolability issues when fuel failures occur. The AREA™

methodology also provides the [ ] for dose

evaluations. The methodology also evaluates the RCS pressure criterion for an REA.


Figure 6-1 REA Analysis Code and Data Links


Figure 6-2 SCRAM Position versus Drop Time


Figure 6-3 Pulse Width Definition for Prompt versus Non-Prompt


Figure 6-4 DNBR for a Prompt Pulse at 20% Power


7.0 UNCERTAINTY AND BIASING METHODOLOGY

This section describes the process used to define the key parameters that are biased in

order to generate conservative results that meet the criteria as outlined in Section 2.0.

The parameters listed in Table 4-3 are evaluated to determine the appropriate biasing

strategy for the AREA™ methodology. The evaluation of these parameters determines

which parameters are tested with a sensitivity analysis. These parameters are tested

using a sensitivity analysis process as described in Section 7.1. The evaluation of the

sensitivity results is described in Section 7.1.1 and the onset of trip is discussed in

Section 7.1.2. The method to identify which parameters are to be biased and the final

core biasing strategy is described in Section 7.1.3. The magnitude of the bias for the

parameters is defined in Section 7.1.4. Sections 7.2.1 and 7.2.2 describe the biasing

for the overpressure analysis and the minimum pressure for DNBR analysis,

respectively.

7.1 Core Sensitivity Analysis

The sensitivity analysis is performed on the parameters identified by the evaluation

using the methodology described in Section 6.0. The transient calculations in

Sections 6.5 and 6.6 are used to generate the sensitivities. The base case is defined

with the following parameters biased by a representative uncertainty:

• Increase in Ejected Rod Worth (ERW)

• Increase in DTC (less negative)

• Decrease in β

• Increase in MTC (more positive)

• Increase in fuel pin power peaking (detailed model calculation only)

[

]


This initial biasing is necessary to obtain sensitivities that are representative of limiting

results. A nominal case is also run where the biasing of the parameters listed above is

not included. [

] The difference between the results of the base case

(with biasing) and the nominal case establishes the minimum amount of conservatism

inherent to the methodology and provides a means to determine the importance of the

sensitivities.

Sensitivity calculations are performed [

]

For each of the parameters listed above (already biased), [

]

Results from the sensitivity transient cases are tabulated for the six dependent

variables:

• Maximum fuel temperature

• Maximum rim temperature

• Maximum enthalpy rise

• Maximum total enthalpy

• MDNBR

• Maximum energy to the coolant during the transient


Each of these defined dependent variables is used to meet a regulatory requirement.

For each sensitivity case, [

]

[

]


7.1.1 Sensitivity Evaluation Method

Categorization of the parameters relative to their variability and their impact on the

dependent variable results determines the manner in which the parameter is treated.

[

]


[

]


7.1.2 Onset of Trip

It is recognized that if a trip does not occur with all other conditions being the same,

then the results from the no trip case can be the same or more severe than the results

for the trip case. Since the condition of trip or no trip is dependent upon the proximity of

the transient response to the trip condition [

]

7.1.3 Core Biasing Strategy

The biasing of the key parameters for the six dependent variables [

]


7.1.4 Core Biasing Values

Table 7-2 identifies the parameters to be biased and the direction of each bias. This

section describes the definition, physical significance (impact), value, and bases for

each parameter.

7.1.4.1 Ejected Rod Worth

Definition: The ERW is the reactivity worth of an individual RCCA or CEA that is

removed from the core without thermal feedback.

Impact: ERW is the driving force of an REA. Once ejected, the core power increases

and the local power increases around the location of the ejected RCCA or CEA.

Value: The uncertainty is defined by Section 6.2 of Reference 11 [ ] The

possible initial position of an ejected rod is defined by control rod positions allowed by

the PDIL specified in the COLR.

Basis: [

]

7.1.4.2 Effective Delayed Neutron Fraction (β)

Definition: β is the effective fraction of total neutrons produced by fission that are

delayed (emitted by decay by excited isotopes). Effective refers to the relative “worth”

of a delayed neutron relative to the entire fission neutron energy spectrum.


Impact: REAs with reactivity insertions less than β rely on delayed neutron production

to maintain power escalation and have doubling times of seconds or longer. For ERWs

greater than β, the doubling time can be 1000 times smaller. Hence, a lower β produces

a higher power pulse as described in Section 3.1.1.

Value: The AREA™ methodology uses a [ ]

Basis: [

]

7.1.4.3 Doppler Temperature Coefficient (DTC)

Definition: The DTC is the reactivity change per fuel temperature change with all other

conditions held constant.

Impact: The DTC is the major feedback mechanism to mitigate prompt critical

transients.

Value: The value used is [ ] which is a reduction in the magnitude of the DTC

since it is a negative quantity.


Basis: [

]

7.1.4.4 Moderator Temperature Coefficient (MTC)

Definition: The MTC is the reactivity change per unit change in moderator temperature

with all other conditions held constant.


Impact: The MTC can be a major feedback mechanism in mitigating REAs.

Value: The MTC is a delayed but important feedback and should be increased by

[

]

Basis: [

]

7.1.4.5 Peaking Uncertainty

Definition: Peaking in this context refers to the relative power distribution effects due

to uncertainties. Peaking uncertainties are typically defined as 2-D (F∆H) and 3-D (FQ)

uncertainties.

Impact: Higher peaking directly affects all the local thermal results. Biases for

uncertainties are conservatively applied in the detailed model calculations. Inclusion of

peaking uncertainties in the nodal model would increase the temperatures and reduce

the transient response and are conservatively ignored. In addition, if voiding occurs

around individual fuel pins during the transient, the powers in these fuel pins would be

reduced and are conservatively ignored.


Value: [

] The licensing bases of a plant

may also include other peaking penalties/uncertainties in addition to those from

Reference 11 that are applicable to the REA event. [

]

Basis: [

]

7.1.4.6 Initial Condition Peaking

Definition: Peaking in this context refers to the peaking that can exist at different initial

conditions.

Impact: The initial AO can skew the power to the top or bottom of the core. Higher

peaking directly affects all the local thermal results and can change the ERW. These

initial conditions affect both the nodal and detailed model calculations.

Value: Initial conditions are set to reflect the limiting conditions of AO defined by the

Technical Specifications or COLR.


Basis: [

]

7.1.4.7 Core Power

Definition: Core power is the rate of energy produced by the core. The actual core

power is parameterized by examining several power levels. This sensitivity is for the

amount that the power could be lower or higher than the indicated power by the thermal

power heat balance uncertainty.

Impact: It primarily affects the thermal analysis and provides some benefit in the nodal

model.

Value: The heat balance uncertainty is well defined and available in the Technical

Specifications.

Basis: [

]

7.1.4.8 Gap Conductance

Definition: Gap conductance is the amount of heat flow across the gap between the

fuel pellet and clad per degree of temperature difference across the gap.


Impact: Gap conductance is an integral part of the fuel pellet thermal solution which

affects fuel temperature by allowing or restricting heat flow out of a pellet during a fast

heatup like an REA. When considering maximum fuel temperature, a lower gap

conductance is conservative. When considering DNBR and peak RCS pressure, a high

gap conductance is conservative.

Value: The base gap conductance values and sensitivity values are generated with

GALILEO™ (Reference 7). A range of values are used based on sensitivity

calculations.

Basis: [

]


7.1.4.9 Fuel Conductivity

Definition: Fuel conductivity is the rate of heat transfer through the fuel per degree

change in temperature per unit distance.

Impact: Fuel conductivity is an integral part of the fuel pellet thermal solution which can

affect fuel temperature by allowing or restricting heat transfer through the fuel pellet

during a fast heatup such as an REA. Low fuel conductivity increases fuel temperature

and high thermal conductivity increases the transient heat flux (lowering DNBR). DNBR

is unaffected by fuel thermal conductivity and gap conductance at steady state

conditions.

Value: Fuel thermal conductivity values used are obtained from GALILEO™ which

defines a thermal fuel conductivity uncertainty of [ ] (Reference 7 page 5-76).

Basis: [

]

7.1.4.10 Fuel Heat Capacity

Definition: Fuel heat capacity is the heat increase per unit volume or mass per degree

change in temperature.

Impact: Since an REA is an energy insertion event, heat capacity could be an

important parameter. In steady state conditions or slow transients, heat capacity is not

a key parameter and can be ignored.


Value: The uncertainty on heat capacity from GALILEO™ [

]

Basis: [

]

7.1.4.11 Burnup

Definition: Burnup is a measure of the depletion of the fuel.

Impact: There are burnup dependent limits that are important to this methodology.

Most burnup dependent phenomena vary slowly as a function of burnup with the

exception of gap conductance. Clad creeps down and upon contact with the pellet, gap

conductance rises abruptly as the gap closes. After gap closure, gap conductance

slowly varies with burnup.


Value: [

]

Basis: [

]

7.1.4.12 Critical Heat Flux (CHF) Correlation

Definition: CHF is a phenomenon that occurs when a fuel pin is surrounded by a vapor

layer and the heat transfer coefficient decreases with increasing clad temperature. This

state is the DNB condition. Empirical correlations are developed at steady state

conditions to determine when this phenomenon may occur.

Impact: If the heat flux increases beyond this critical condition, a sharp increase in the

clad and fuel temperature can occur. At this point, fuel pin failure is assumed to occur

which is consistent with the criteria in Section 2.3.

Value: Approved CHF correlations are used with their respective correlation limit.

Basis: [

]

7.1.4.13 Core Flow

Definition: Core flow is the amount of coolant moving through the active core.


Impact: Core flow is a key parameter in the DNBR calculations. Flow is also a key

parameter to heat removal from the core. It can affect the nodal model through the

coolant effects on MTC. For overpressure calculations a maximum flow results in a

slightly higher predicted power which leads to more energy transferred to the coolant

resulting in increased core pressure.

Value: Minimum and maximum core flow is available from the plant licensing

documentation. The flow uncertainty also is considered in the biasing of the models.

Basis: [

]

7.1.4.14 Core Inlet Temperature

Definition: Temperature of the coolant entering the active core. Coolant temperatures

in the core increase from inlet temperature based on core heatup from the power

produced and heat removal from the fuel to the coolant by the flow.

Impact: Core inlet temperature is a key parameter in MDNBR calculations and in

determining the thermal properties of the coolant. Also, the coolant density from

temperature differences can affect the reactivity of the nodal model (by MTC effects).

For overpressure calculations a minimum temperature may result in more energy

transferred to the coolant resulting in an increased pressure.

Value: Temperature deadband and uncertainty are available from the plant licensing

documentation.


Basis: [

]

7.1.4.15 Core Pressure

Definition: Pressure is the force per unit area of the active core that keeps the coolant

from boiling at highly elevated temperatures. The minimum pressure for the core is at

the top of the active fuel.

Impact: Core pressure is an input to DNBR correlations. It affects the coolant density

which can have a secondary effect on the MTC. It also affects the differential pressure

between the fuel rod and system. This decrease in core pressure also affects the high

clad temperature failure criteria and the evaluation of ballooning failures relative to DNB

propagation.

Value: RCS pressure deadband and uncertainty are available from the plant licensing

documentation. The two are combined to give a low value for the detailed model

calculations. The initial value is dependent upon the allowed operational range before

systems are activated to correct (deadband) and the measurement uncertainty on

pressure.


Basis: [

]

7.2 RELAP5 Biasing for Pressure Calculations

There are two scenarios for RELAP5 Calculations. The following sections describe the

biasing for the peak RCS pressure calculations and the pressure for DNBR calculations.


7.2.1 RELAP5 Peak RCS Pressure Calculations

Biasing of parameters for RCS peak pressure are intended to maximize the energy

added to the RCS while minimizing the ability of the secondary systems and pressure

relief components to mitigate the RCS pressure response. Items addressed in this

section are the assessment of system parameters which are in addition to the neutronic

behaviors discussed in Section 7.1.4. The list of parameters considered for biasing,

including core parameter biasing is provided in Table 7-3.

7.2.1.1 Initial RCS Pressure

Definition: Initial RCS pressure is the pressure used at the start of the event. The

RCS pressure is not the same throughout the system and is dependent upon the

elevation and location in the coolant flow path. Typically, initial pressure at a key

location in the system or at a sensor is used as the reference pressure point. Maximum

pressure in the RCS is usually located at the lowest elevation of the system near the

bottom of the reactor downcomer or vessel.

Impact: A higher initial pressure has the least margin to the regulatory pressure limit.

In general, the proximity of the initial pressure to the system setpoints for trip and safety

valves may affect the transient response of the pressure.

Value: The value is dependent upon the allowed operational range before a system is

activated to correct (deadband) and the pressure measurement and signal processing

uncertainty.

The value can be higher or lower than nominal and the deadband and uncertainty may

have different effects if non safety controls are used. The pressure deadband and

uncertainty are available from plant licensing documentation.

Basis: [

]


7.2.1.2 Pressurizer Safety Valve Capacity/Setpoints/Tolerance

Definition: Overpressure protection of the RCS is provided by pressurizer safety

valves. These ASME required valves act to relieve pressurizer steam and thereby limit

RCS pressure.

Impact: Higher lift setpoints maximize the RCS pressure response by allowing the

RCS to reach higher pressures before the safety valves open. Lower relief capacity

minimizes the pressure relief. Higher lift tolerance delays valve opening and maximizes

RCS pressure response.

Value: The capacity, setpoints, and tolerances are obtained from existing licensing

basis documents.

Basis: [

]

7.2.1.3 Reactor Protection System (RPS) Setpoints

Definition: The RPS ensures reactor trip and control rod insertion when events exceed

the specified setpoints.

Impact: A reactor trip significantly reduces neutron power and terminates the energy

addition into the RCS.

Value: Key trip functions for the REA are high neutron power and high RCS pressure,

or pressurizer pressure. The setpoints and uncertainties for these trip functions are

obtained from existing licensing basis documents.

Basis: [

]


7.2.1.4 Non-Safety Systems

Definition: Non-safety systems include controls and systems/components that provide

normal controls and limits that keep the plant operation within acceptable bounds for

normal operation. Non-safety systems include the normal control system, rod controls,

pressure controls, inventory control and secondary plant controls.

Impact: In general, non-safety systems act to counter/reduce any off-normal operation

or event. However, since controls may affect timing of reactor trip, the system

behaviors must be reviewed to ensure the delay in reactor trip actuation does not make

the consequences of the event worse.

Value: The philosophy for simulation of non-safety control systems is to either model a

control system to perform its normal control function, or to assume the control function

is set to its state at the beginning of an event. Non-deterministic failures of the non-

safety systems are not considered. Nominal control points and operational

characteristics are obtained from plant specific documentation.

Basis: [

]

7.2.2 RELAP5 Core Pressure for MDNBR Calculations

The pressure calculation supporting MDNBR analysis is biased independently of the

overpressure analysis as the intent is to conservatively model the pressure during the

event.

The discussion below indicates the conditions that must be treated differently than the

overpressure cases.

7.2.2.1 Initial RCS Pressure

Definition: Initial RCS pressure is the pressure used at the start of the event.


Impact: A lower initial pressure leads to lower pressure conditions when the breach in

the upper head associated with the REA is modeled and can lead to lower calculated

DNBRs.

Value: The value should be biased to the lowest allowable operating conditions. The

variations are dependent upon the allowed operational range before systems are

activated to correct (deadband) and the measurement uncertainty on pressure. The

RCS pressure deadband and uncertainty are obtained from the plant licensing

documentation.

Basis: [

]

7.2.2.2 Breach Size

Definition: This is the largest area around the CRDM that could remain open after a

CRDM is removed.

Impact: [

]

Value: The area for the breach is available in plant licensing documentation and

drawings.

Basis: [

]


Table 7-1 Criteria Applicability to Initial Conditions for Sensitivity Calculations


Table 7-2 Core Biasing Strategies for the Key Parameters


Table 7-3 Parameters Considered For Biasing for RCS Pressure Scenarios


Figure 7-1 Doppler Test Result Comparisons


8.0 AREA™ PLANT SPECFIC APPLICATION

This section describes the intended use of the AREA™ methodology for plant specific

applications. As defined in this section, there is the initial use of the AREA™

methodology for the plant analysis and the follow-on applications that captures the

impact of cycle-to-cycle variations on an REA.

8.1 Initial Application of AREA™ Methodology

The initial application of the AREA™ methodology consists of the following:

1. Applicable regulatory requirements establish appropriate fuel limits. These fuel

limits are used as the bases for the AREA™ analyses performed.

2. Define the fuel performance code (GALILEO™ or COPERNIC) to be used for the

analyses. This topical report defines two fuel performance codes that could be

used; COPERNIC or GALILEO™. If COPERNIC is used, then the thermal

properties, biasing values, and gap conductance values are determined or verified

with respect to the requirements of Section 5.2.2. Parameters based on GALILEO™

are defined and provided in this topical report.

3. The AREA™ methodology defines the use of S-RELAP5 for Westinghouse and CE

plants or RELAP5-B&W for B&W plants.

4. Verify that the biases presented in this topical report remain acceptable by running

selected parameter sensitivities. If COPERNIC is used, all uncertainties defined in

step 2 for COPERNIC replace the biases presented for GALILEO™.

5. Determine any biases and penalties used that are for the plant specific analyses.

6. Run the matrix of cases as defined in Section 6.4 with biasing strategy defined in

Table 7-2 to define margin to the limiting conditions.

7. Run RELAP5 for both pressure scenarios. Define margin to the high pressure limit.

Verify DNBR calculations.


[

]

If the peak values with minimal biasing are already higher than the high burnup criteria,

another approach is required. This condition is described in Section 6.4. The plant

specific application defines the approach used for this condition.

8.2 Cycle to Cycle Evaluation

The first application of the AREA™ methodology biases key parameters so that it

provides a basis for the initial cycle that is expected to be bounding for future cycles.

The application of the methodology summarizes the key parameters for each of the

limiting cases in the time-in-life and power level matrix (Section 6.4) analyzed. Steady

state calculations are performed to verify that the key parameters for a follow-on cycle

remain within the range of these key parameters from the initial application. These key

parameters are:

• ERW

• β


• DTC

• MTC

• Initial FQ (at power cases only)

• Initial FΔH (at power cases only)

• Static post ejection FQ

• Static post ejection FΔH

• Fuel failures (see Section A.3.8)

If the key parameters established in the initial application of AREA™ are not exceeded

in future cycles, then no additional analysis is required. In the event that any of the key

parameters are not bounded, there are two approaches (listed below) that are available

to address future cycles.

1. Complete reanalysis of the matrix of cases is performed. This approach is selected

when a new baseline matrix of cases is needed. This option is typically employed

for major fuel design changes that are outside the scope of the original analysis.

2. Reanalysis of a portion of the matrix of cases is repeated for the condition where a

specific parameter is found to be outside of the initial application analysis range.

This option is typically employed for minor fuel design changes that are challenging

isolated conditions of the original analysis.


Figure 8-1 Increased Biasing for Cycle Verification


9.0 SAMPLE PROBLEMS

The AREA™ sample problems contain the detailed results of this REA methodology for

three plant types:

• Westinghouse 4-Loop 193 FA plant with 17x17 fuel lattice containing 1 cell water

holes and control rod pins.

• B&W 177 FA plant with 15x15 fuel lattice containing 1 cell water holes and

control rod pins.

• CE 217 FA plant with 14x14 fuel lattice containing 4 cell water holes and control

rod pins.

The sample problems encompass the general application to three different fuel and

plant types that cover the range of the fuel pin sizes, control rod pin sizes, and absorber

types for the current PWRs. Sample pressure calculations are provided for the

Westinghouse 4-Loop (with recirculating steam generators) and the B&W plant (with

once through steam generators). A sample pressure calculation is not provided for the

CE sample problem as it is adequately illustrated by the sample problem for the

Westinghouse plant.

For each plant type, the PCMI limits for the REA criteria are defined. The high clad

temperature failure criteria in Reference 2 are used for all the sample problems. The

transient FGR equation in Reference 2 is also used. No fuel or rim melt is encountered

so the fuel melt criteria from Reference 1 and 2 are met. [

]


The core biasing strategy in Table 9-1 is the same for all three sample problems. Core

summary tables of the limiting results are provided for the matrix of conditions. For the

first sample problem, detailed figures are provided for the transient that produced the

limiting values relative to each of the limiting criteria {fuel melt, rim melt, enthalpy rise,

total enthalpy for high clad temperature failure criteria (for prompt critical cases), and

MDNBR (for non-prompt critical cases)}.

[

] Table 9-2 provides the system parameter biasing, in

addition to the biases of Table 9-1, for the overpressure assessment. The overpressure

sample problems are presented for the first two plants representing a recirculating

steam generator system and a once through steam generator, respectively.


Table 9-1 Core Biasing Parameters and Values


Table 9-2 Biasing Parameters and Values for Overpressure


10.0 CONCLUSIONS

The AREA™ methodology provides a conservative representation of the reactor

response during an REA to demonstrate compliance with the appropriate criteria.

Energy deposition, fuel rim melt, fuel centerline melt, MDNBR and RCS pressure are

considered in the evaluation of an REA. The methodology includes the use of a nodal

3-D kinetics solution with open channel thermal-hydraulic and fuel temperature

feedback and a [

] These models provide localized

neutronic and thermal conditions to demonstrate compliance with the REA criteria that

are the same as or similar to the criteria in Reference 1 or Reference 2.

The AREA™ methodology is applied to three different PWR plant types that result in

very similar conclusions. The AREA™ methodology demonstrates the level of

conservatism applied to the analyses and compares the results to the criteria outlined in

both References 1 and 2. The sample problems are based upon the conservatisms

specified in Section 7.0 and illustrate the methodology. No criteria are exceeded nor

are failures predicted in these sample problems (Section 6.0). Section 8.0 provides an

overview of the AREA™ methodology as it applies to specific applications.


11.0 REFERENCES

1. NUREG-0800, Chapter 4, “Standard Review Plan for the Review of Safety

Analysis Reports for Nuclear Power Plants: LWR Edition — Reactor,”

March 2007.

2. Memorandum from Paul M. Clifford (NRC) to Timothy J. McGinty (NRC),

“Technical and Regulatory Basis for the Reactivity-Initiated Accident

Acceptance Criteria and Guidance, Revision 1,” ML14188C423, March 16,

2015.

3. NUREG-0800, Chapter 15, “Standard Review Plan for the Review of

Safety Analysis Reports for Nuclear Power Plants: LWR Edition —

Transient and Accident Analysis,” March 2007.

4. Regulatory Guide 1.183, “Alternative Radiological Source Terms for

Evaluating Design Basis Accidents at Nuclear Power Reactors,” July

2000.

5. Memorandum from Anthony J. Mendiola (NRC) to Travis L. Tate (NRC),

“Technical Basis for Revised Regulatory Guide 1.183 (DG-1199) Fission

Product Fuel-to-Cladding Gap Inventory,” ML111890397, July 26, 2011.

6. NUREG/CR-6742 LA-UR-99-6810, “Phenomenon Identification and

Ranking Tables (PIRTs) for Rod Ejection Accidents in Pressurized Water

Reactors Containing High Burnup Fuel,” Los Alamos National Laboratory,

September 2001.

7. ANP-10323P, “Fuel Rod Thermal-Mechanical Methodology for Boiling

Water Reactors and Pressurized Water Reactors,” July 2013.

8. EMF-2310PA Revision 1, “SRP Chapter 15 Non-LOCA Methodology for

Pressurized Water Reactors,” May 2004.


9. BAW-10164P-A, Revision 6, “RELAP5/MOD2-B&W – An Advanced

Computer Program for Light Water Reactor LOCA and Non-LOCA

Transient Analysis,” June 2007.

10. ANSI/ANS-19.6.1-2011, “Reload Startup Physics Tests For Pressurized

Water Reactors,” January 2011.

11. ANP-10297P-A, Revision 0, Supplement 1, “The ARCADIA® Reactor

Analysis System for PWRs Methodology Description and Benchmarking

Results Topical Report,” June 2015.

12. ANP-10297P-A, Revision 0, “The ARCADIA® Reactor Analysis System for

PWRs Methodology Description and Benchmarking Results Topical

Report,” February 2013.

13. ANP-10311P-A, Revision 0, “COBRA-FLX™: A Core Thermal-Hydraulic

Analysis Code,” January 2013.

14. BAW-10120PA, “Calculation of Core Physics Calculations with

Measurements,” July 1979.

15. BAW-10193P-A, Revision 0, “RELAP5/MOD2-B&W for Safety Analysis of

B&W Designed Pressurized Water Reactors,” January 2000.

16. BAW-10231P-A, Revision 1, “COPERNIC Fuel Rod Design Computer

Code,” January 2004.

17. Regulatory Guide 1.77, “Assumptions Used for Evaluating a Control Rod

Ejection Accident for Pressurized Water Reactors,” May 1974.

18. BAW-10169P-A, “RSG Plant Safety Analysis – B&W Safety Analysis

Methodology for Recirculating Steam Generator Plants,” October 1989.

19. “Dynamics of Nuclear Reactors,” David L. Hetrick, La Grange Park, IL:

American Nuclear Society, 1993, p. 166.

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page A-1

APPENDIX A W 4-LOOP 193 FA PLANT WITH 17X17 FUEL LATTICE

This sample problem is for a Westinghouse 4-Loop plant. GALILEO™ (Reference 7) is

used as the fuel performance code and S-RELAP5 (Reference 8) is used as the system

thermal-hydraulic code in the coupled calculation with ARTEMIS™ for the maximum

RCS pressure evaluation (Reference 11 and Reference 12). The biases used for this

application are as stated in Table 9-1. [

]

A.1 REA Limits Generated by GALILEO™

This core for this plant contains both M5® and Zr4 clad. [

] The enthalpy rise limits are based

upon the relative oxide thickness criteria from Reference 1 and are shown in Figure A-1

and Figure A-2 for M5® and Zr4, respectively. The limiting fuel pin pressures versus

burnup for M5® clad fuel are shown in Figure A-3 and Figure A-4 for UO2 and gadolinia

fuel, respectively. The limiting fuel pin pressure versus burnup curves described in

Section 6.2.2 are generated for Zr4 high clad temperature failure criteria shown in

Figure A-5 and Figure A-6 for UO2 and gadolinia fuel, respectively. Figure A-7 and

Figure A-8 contain the fission gas release for M5® and Zr4 fuel, respectively.


A.2 Boundary Conditions

For this sample problem, [

] The high flux trip of 118% of rated power and the high pressurizer pressure

trip at [ ] are used as noted. The general depressurization curve

supporting the MDNBR analysis is shown in Figure A-9. These simulations have an

assumed pressure decrease with time that is confirmed with S-RELAP5 (calculation in

Section A.5.)

A.3 Fuel Integrity Sample Problem Summaries

[

]

The general timing of these events is shown in Table A-1. The most limiting results

[ ] are displayed in Table A-2 through Table A-6

[ ] No failures are found against the specified criteria for the

applicable conditions. More detail is provided for the case with the least margin to the

limit for each of the criteria {total enthalpy (high clad temperature failure criteria),

enthalpy rise, fuel melt, rim melt, and MDNBR}. The overpressure biased case is

addressed later in Section A.4.


A.3.1 Minimum Margin to Total Enthalpy Limits (Failure Criterion for High Clad Temperature Failure Criteria for Prompt Critical Excursions)

The minimum margin to the enthalpy failure limit for high clad temperature failure criteria

is [ ] which occurs for the [ ] REA transient at EOC.

This [ ] is prompt critical [ ] The results for core power,

F∆H, and FQ are shown in Figure A-10. The maximum cal/g in the core with time is in

Figure A-11. The total enthalpy limit calculated based on the equation from Reference

2 is shown in Figure A-12. The cal/g margin for [ ] is shown in the

[ ] in Figure A-13. As shown in Figure A-13, the [

] The sharp rise in the loss of margin reflects the sharp drop in the

limit seen in Figure A-12. [

]

A.3.2 Minimum Margin to Enthalpy Rise Limits (PCMI Failure Limit)

The minimum margin to the limit for enthalpy rise is [ ] which occurs for

the [ ] (EOC). No failures are seen for

either M5® or Zr4 clad. The results for core power, F∆H, and FQ are shown in Figure

A-14. The maximum Δcal/g with time is in Figure A-15. The enthalpy rise in Figure

A-15 is terminated one pulse width after the peak. The progression of the enthalpy rise

with time can be inferred from the total enthalpy versus time displayed in Figure A-16.

The Δcal/g results and limits for M5® and Zr4 clad types are shown in the [

] in Figure A-17 for [ ] versus burnup. As expected, the

Zr4 clad has the least margin at high burnups but remains more than [ ]

below the limit. For this core, the enthalpy rise values for the fuel with M5® clad are

more than [ ] below the PCMI failure limit at any burnup.


[

]

A.3.3 Minimum Margin to Fuel Melt Limits

Fuel melt is potentially a coolability issue and a failure criterion. The minimum margin to

the limit for fuel melt is [ ] which occurs for the [

] No fuel melt occurs which meets the fuel melt criteria in

Section 2.0. The results for core power, F∆H and FQ are shown in Figure A-18. The

fuel, fuel rim, and clad temperatures with time are shown in Figure A-19. The difference

for [ ] between its fuel temperature and melt limit is shown in

the [ ] in Figure A-20. In general, the margin to the melt limit increases

with burnup indicating that the melt temperature limit is decreasing with burnup much

slower than the peaking is decreasing with burnup.

A.3.4 Minimum Margin to Rim Melt Limits

Melting of the fuel rim is a coolability issue. The minimum margin to the limit for fuel rim

melt is [ ] which occurs for the [

] No rim melt occurs so that the fuel rim melt criterion in Section 2.0 is met.

The results for core power, F∆H and FQ are shown in Figure A-18. The fuel, fuel rim,

and clad temperatures with time are shown in Figure A-19. The minimum difference

[ ] between the fuel rim temperature and its limit is shown in the

[ ] in Figure A-21.


A.3.5 Minimum Margin to MDNBR SAFDL or Maximum DNBR Failures

Exceeding the MDNBR SAFDL is a failure limit for non-prompt critical power excursions.

The minimum margin to the limit (SAFDL/MDNBR – 1) is [ ] which occurs for

the [ ] The results for core power, F∆H

and FQ are shown in Figure A-22. The MDNBR versus time is shown in Figure A-23.

The SAFDL is divided by the MDNBR for the [

] The limit becomes [ ] and all the fuel pins above are

assumed failed and those below [ ] are not. SAFDL/DNBRs for [ ]

are shown in the [ ] in Figure A-24 with burnup. There is significant

DNBR margin shown in this plot and there are no failures to report. Figure A-25 is the

plot of SAFDL/MDNBR versus differential fuel pin to core pressure. If there are fuel

failures, this curve shows if the pin pressure for any failed fuel pin is higher than core

pressure. Without a higher internal fuel pin pressure than the core pressure, no

coolability or DNB propagation issues due to fuel pin ballooning exist. Hence, this

condition meets coolability Criterion 4 in Section 2.0.

A.3.6 Conservatism of Biasing Method

Based on the results in these tables, an assessment of the limiting case for each of the

limiting criteria is presented and summarized in Table A-7. For each of the limiting

criteria, the power level, cycle burnup, [

] are provided. There

is ample conservatism for each limiting criterion.

In addition, the [ ] has the highest energy deposited in

the coolant for the over-pressure analysis.


A.3.7 Sensitivity to High Flux Trip

The matrix of analyzed cases includes REAs that do not trip on high flux. Thus, the

analysis inherently contains results without a trip. This provides assurance that if the no

trip condition is limiting, then the summary table of results includes the effect of an REA

with no trip. For example, the case at [ ] actuates the

high flux trip and the lowest responding detector is only [ ] higher than the trip

setpoint. The trip is deactivated and the case is re-examined. The resultant delta

MDNBR is only [ ] While the fuel temperature [ ]

for the [ ] this condition of no trip remains less than the maximum

fuel temperature of the remaining transients. Hence, no change is needed to Table A-7.

A.3.8 Static Cases

[

]


A.4 Peak RCS Pressure Assessment

The peak RCS pressure analysis is performed with S-RELAP5 automatically coupled

with ARTEMIS™. The limiting pressure criterion is 120% of the design pressure

(2485 psig) which yields a pressure limit of 2982 psig. Since the maximum integrated

power to the coolant from the cases in the previous section occurs at [ ]

the sample problems for overpressure are calculated at a [ ]

Two cases are presented: (Case 1) at nominal conditions and (Case 2) with biasing

applied. The conditions for each case are summarized in Table A-9. [

] The results are listed in Table A-10.


The nominal and biased cases reach similar peak RCS pressures with the difference in

peak pressures of less than [

]

[

]


These results demonstrate that even with a prompt critical excursion of [ ] the

RCS pressure limits are far from being challenged. In fact, Case 4 demonstrates that if

the high pressurizer pressure trip function is employed, the RCS pressure would not

reach the PSV lift setpoints. Even with the conservatisms used, the peak RCS pressure

[ ] remains well below the acceptance criterion limit for this plant (2982

psig).

The sample problems demonstrate that a cycle specific evaluation of REA conditions

with biased cases does not challenge the reactor coolant pressure boundary limit. The

results of Case 2 and Case 4 also show that the peak RCS pressure is relatively

insensitive to whether the rod ejection [

]

A.5 Core Pressure for MDNBR Evaluation

[

]


Figure A-9 is an approximated curve from a HFP break without a power increase. To

test the validity of this curve for other power levels, [

]

A.6 Sample Summary

For this sample problem, the REA results meet all the acceptance criteria and no fuel

failures are calculated. The biasing strategy provides significant conservatism to the

best estimate calculations. No coolability concerns exist since there are no total

enthalpies above 230 cal/g, no fuel melt failures, no enthalpy rise failures, no high clad

temperature failures, and no DNBR failures. If DNBR failures occur, examination of fuel

pin pressure above core pressure for the DNBR failures can address those failures for

coolability and propagation.

System overpressure results demonstrate that reactivity insertions of less than

[ ] are not challenging the pressure limits. Significantly higher reactivity

insertions greater than [ ] are needed to challenge the pressure integrity of

the RCS.


Table A-1 General Timing of the Event

Event Timing

Time to eject rod 0.1 second * fraction of insertion

Trip signal reached If trip occurs, the time is provided with the power plot

Time to peak core neutron power Included with power plot

Time to max enthalpy rise With power plot or 1 second past the time of peak core neutron power if not prompt critical

Rods begin to drop Total delay time (1 second after trip actuation)

Rods to full insertion Total drop time (3.68 seconds)

Simulation ended for the event [

]


Table A-2 W 4-Loop Limiting Results Summary for Burnup 1


Table A-7 W 4-Loop, Measure of Conservatism for Limiting Result Cases


Table A-8 Transient and Static Difference in Limiting Conditions


Table A-9 W 4-Loop Plant Overpressure Input Summary


Table A-10 W 4-Loop Plant Overpressure Results Summary (High Pressurizer Pressure Trip Modeled)


Table A-11 W 4-Loop Plant Overpressure Results Summary


Table A-12 W 4-Loop Plant Core Pressure for MDNBR Input Summary


Figure A-1 W 4-Loop Enthalpy Rise Limits for M5® Fuel Based on Relative Oxide Thickness


Figure A-2 W 4-Loop Enthalpy Rise Limits for Zr4 Fuel Based on Relative Oxide Thickness


Figure A-3 W 4-Loop Limiting Pressure Parameters for UO2 Fuel with M5® Clad


Figure A-4 W 4-Loop Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad


Figure A-5 W 4-Loop Limiting Pressure Parameters for UO2 Fuel with Zr4 Clad


Figure A-6 W 4-Loop Limiting Pressure Parameters for Gadolinia Fuel with Zr4 Clad


Figure A-7 W 4-Loop Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad


Figure A-8 W 4-Loop Limiting FGR for UO2 and Gadolinia Fuel with Zr4 Clad


Figure A-9 W 4-Loop General Depressurization Curve


Figure A-10 Transient FQ, F∆H, and Core Power for Max Enthalpy Condition


Figure A-11 Transient Maximum Enthalpy for Max Enthalpy Condition


Figure A-12 Total Enthalpy Limit with Burnup for Max Enthalpy Condition


Figure A-13 Enthalpy Margin to Limit Scatter Plot for Max Enthalpy Condition


Figure A-14 Transient FQ, FΔH, and Core Power for Max Enthalpy Rise Condition


Figure A-15 Transient Maximum Enthalpy Rise for Max Enthalpy Rise Condition


Figure A-16 Transient Maximum Enthalpy for Max Enthalpy Rise Condition


Figure A-17 Maximum Enthalpy Rise and Limits by Clad Type for Max Enthalpy Rise Condition


Figure A-18 Transient FQ, FΔH, and Core Power for Max Fuel Temperature Condition


Figure A-19 Transient Fuel, Fuel Rim and Clad Temperature for Max Fuel Temperature Condition


Figure A-20 Maximum Fuel Temperature by Fuel Type – Margin to Limits for Max Fuel Temperature Condition


Figure A-21 Maximum Fuel Rim Temperature by Fuel Type – Margin to Limits for Max Fuel Rim Temperature Condition


Figure A-22 Transient FQ, FΔH, and Core Power for MDNBR Condition


Figure A-23 Transient MDNBR for MDNBR Condition


Figure A-24 SAFDL to MDNBR Ratio by Fuel Type as a Function of Burnup for MDNBR Condition


Figure A-25 SAFDL to MDNBR Ratio by Fuel Type as a Function of Fuel Pin to Core Pressure Difference for MDNBR Condition


Figure A-26 Case 2 Power Response for High Pressurizer Pressure Trip


Figure A-27 Case 2 Pressure Response for High Pressurizer Pressure Trip


Figure A-28 Case 4 Power Response for High Pressurizer Pressure Trip


Figure A-29 Case 4 Pressure Response for High Pressurizer Pressure Trip


Figure A-30 Core Pressure for MDNBR Response Comparison

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page B-1

APPENDIX B B&W 177 FA PLANT WITH 15X15 FUEL LATTICE

This sample problem is for a B&W 177 fuel assembly plant. GALILEO™ (Reference 7)

is used as the fuel performance code and RELAP5/MOD2-B&W (Reference 9) is used

as the system thermal-hydraulic code for pressure calculations. The RELAP5 interface

with ARTEMIS™ (Reference 11 and Reference 12) is manually coupled for the

maximum RCS pressure calculation. For the manual coupling ARTEMIS™ provides a

forcing function to RELAP5/MOD2-B&W for the maximum RCS pressure analysis. This

provides sufficient conservatism so that no feedback is required from the system code

to the neutronics code. The biases used for this application are as stated in Table 9-1.

[ ]

B.1 REA Limits Generated by GALILEO™

This plant is assumed to have only M5® clad fuel. The PCMI limit for excess hydrogen

is used for this sample problem and is calculated using GALILEO™. [

]


The limiting fuel pin pressure versus burnup that is used for M5® high clad temperature

failure criteria is shown in Figure B-2 and Figure B-3 for UO2 and gadolinia fuel,

respectively. Figure B-4 contains the fission gas release for M5® clad fuel.

B.2 BOUNDARY CONDITIONS

For this sample problem, the power levels of [

] are selected. The high flux trip of 112% of rated power and the high pressure

trip at 2446 psia are used as noted. The general depressurization curve for the breach

is given in Figure B-5. These simulations use this assumed pressure decrease with

time and it is confirmed with RELAP5/MOD2-B&W (Section B.6).

B.3 Fuel Integrity Sample Problem Summaries

[

] as specified in Sections 7.1.3 and 7.1.4.6.

• [

]


The general timing of these events is shown in Table B-1. The most limiting results for

[ ] are displayed in Table B-2 through Table B-6 for

each sampled burnup. [

] No

failures are found against the specified criteria for the applicable conditions. The

overpressure biased case is addressed later in Section B.6.

B.4 CONSERVATISM OF BIASING METHOD

Based on the results an assessment of the limiting case for each of the limiting criteria

is presented and summarized in Table B-7. [

] There is ample

conservatism for each limiting criterion.

In addition, the [ ] has the highest energy deposited in

the coolant for the overpressure analysis.

B.5 Peak RCS Pressure Assessment

The maximum integrated power to the coolant from the cases in Section B.4 occurs at

[ ] used for the overpressure calculations. The

two required cases are presented: at nominal (Case 1) and with biased conditions

(Case 2). The conditions for each case are summarized in Table B-8. Power and peak

RCS pressure plots for Case 2 are provided in Figure B-6 and Figure B-7, respectively.


[

]

These two cases reach similar peak RCS pressures. [

]

[

]


[

] Even with the conservatisms used, the peak RCS pressure [

] remains well below 120% overpressure limit value of 3014.7 psia for this plant.

[

] Since the high pressure reactor trip case would prevent

the RCS from pressurizing to the PSV setpoints, there is a large margin to the 120%

overpressure limit of 3014.7 psia.

The case results also indicate that the timing of the peak pressure, without reactor trip,

is a strong function of the integrated energy added to the RCS. [

]

These results demonstrate that even with a prompt critical excursion of [ ] the

RCS pressure limit is far from being challenged. Case 5 demonstrates that if the high

pressure reactor trip function is employed, the RCS pressure would not reach the PSV

setpoints.


The sample problems demonstrate that this cycle-specific evaluation of REA conditions

with biased cases does not challenge the pressure safety limit. [

] the PSV capacity is shown to be capable of

relieving the initial RCS pressure excursion.

[

]

B.6 Core Pressure for MDNBR Evaluation

[

]


The MDNBR case in Section B.3 occurs at HFP. [

]

B.7 Sample Summary



best estimate calculations. No coolability concerns exists since there are no total




coolability and DNB propagation.

RCS overpressure results demonstrate that reactivity insertions of less than [ ]

are not challenging the pressure limits. Significantly higher reactivity insertions greater

than [ ] are needed to challenge the pressure integrity of the RCS.


Table B-1 General Timing of the Event

Event Timing


Trip signal reached If trip occurs, it is noted on the summary tables and it occurs prior to reaching the peak power.

Time to peak core neutron power Within 0.25 seconds after the rod is ejected

Time to max enthalpy rise One pulse width after peak power or 1 second past the time of peak core neutron power if not prompt critical




]


Table B-2 B&W Plant Limiting Results Summary for Burnup 1


Table B-7 Measure of Conservatism for Each of the Limiting Cases


Table B-8 B&W plant Overpressure Input Summary


Table B-9 B&W plant Overpressure Results Summary (no high pressure trip modeled)


Table B-10 B&W Plant Overpressure Results Summary


Table B-11 B&W Plant Core Pressure for MDNBR Input Summary


Figure B-1 Enthalpy Rise Limits for M5® Fuel Based on Excess Hydrogen


Figure B-2 Limiting Pressure Parameters for UO2 Fuel with M5® Clad


Figure B-3 Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad


Figure B-4 Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad


Figure B-5 B&W General Depressurization Curve


Figure B-6 Reactor Power for Biased Case


Figure B-7 Peak RCS Pressure for Biased Case


Figure B-8 Reactor Power For Prompt Critical – No Trip


Figure B-9 Peak RCS Pressure Response for Prompt Critical Reactivity Addition – No Trip


Figure B-10 Hot Leg Pressure Comparison


Figure B-11 Verification of the General Depressurization Curve.

AREVA Inc. ANP-10338NP Revision 0 AREA™ - ARCADIA® Rod Ejection Accident Topical Report Page C-1

APPENDIX C CE 217 FA PLANT WITH 14X14 FUEL LATTICE

This sample problem is for a CE 217 fuel assembly plant. GALILEO™ (Reference 7) is

used as the fuel performance code. The previous two samples problems adequately

show the AREA™ methodology for pressure calculations. No RELAP5 calculations are

performed for this sample problem. The biases used for this application are as stated in

Table 9-1. [ ]

C.1 REA Limits Generated by GALILEOTM

This plant is assumed to have only M5® clad fuel. [

] The enthalpy rise limits are based

upon relative oxide thickness from Reference 1 and are shown in Figure C-1. The

limiting fuel pin pressure versus burnup curves described in Section 6.2.2 are generated

for M5® clad fuel and are shown in Figure C-2 and Figure C-3 for UO2 and gadolinia

fuel, respectively. Figure C-4 contain the fission gas release for M5® clad fuel.

C.2 Boundary Conditions

For this sample problem, [

] are selected. CE has a variable high flux trip and the respective trips used in

these simulations [

]


C.3 Fuel Integrity Sample Problem Summaries

[

] as specified in Sections 7.1.3 and 7.1.4.6. The overpressure biased

case is not addressed in this sample problem.

• [

]

The general timing of these events is shown in Table C-1. The most limiting results for

these cases at each power level are displayed in Table C-2 through Table C-6 [

] It should be noted that none of the HZP cases reported in the

tables are prompt critical. No failures are found against the specified criteria for the

applicable conditions.

All the HZP cases in this sample problem are biased but the rod worths and β are not

artificially raised to be prompt critical. As seen in many of the HZP cases from Table

C-2 through Table C-6, the reactor period is so long that the core power did not achieve

a power level above [ ] At EOC for HZP the ejected rod worth is slightly less

than prompt critical and after [ ] the power is [ ] These results

are clearly non limiting and [

]


C.4 Conservatism of Biasing Method

Based on the results in these tables, an assessment of the limiting case for each of the

limiting criteria is presented and summarized in Table C-7. For each of the limiting

criteria, [ ] the limiting value, the nominal value, and

the estimated level of conservatism (limiting value – nominal value) are provided. There

is ample conservatism for each limiting criterion.

C.5 Sample Summary



best estimate calculations. No coolability concerns exists since there are no total




coolability and propagation.


Table C-1 CE Plant General Timing of the Event

Event Timing


Trip signal reached No trips occurred

Time to peak core neutron power Within 0.1 seconds after the rod is ejected

Time to max enthalpy rise 1 second past the time of peak core neutron power




]


Table C-2 CE Plant Limiting Results Summary for Burnup 1


Table C-7 CE Plant Measure of Level of Conservatism for Each Limiting Parameter


Figure C-1 Enthalpy Rise Limits for M5® Fuel Based on Relative Oxide Thickness


Figure C-2 Limiting Pressure Parameters for UO2 Fuel with M5® Clad


Figure C-3 Limiting Pressure Parameters for Gadolinia Fuel with M5® Clad


Figure C-4 Limiting FGR for UO2 and Gadolinia Fuel with M5® Clad


Figure C-5 CE Plant General Depressurization Curve

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