ML18059A782

ANP-10338NP-A, Rev 0, Area - Arcadia Rod Ejection Accident.

ML18059A782
Person / Time
Site:	PROJ0728
Issue date:	12/31/2017
From:	Framatome
To:	Office of Nuclear Reactor Regulation
References
CAC MF7009, EPID L-2015-TOP-004, NRC:18:005 ANP-10338NP-A, Rev 0
Download: ML18059A782 (361)

Text

framatome AREA TM -ARCADIA Rod Ejection Accident Topical Report December 2017 Framatome Inc. (c) 2018 Framatome Inc. ANP-10338NP-A Revision 0 Copyright

© 2018 Framatome Inc. All Rights Reserved ANP-10338NP-A Revision 0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555~0001 Mr. Gary Peters, Director Licensing and Regulatory Affairs AREVA Inc. 3315 Old Forest Road Lynchburg, VA 24501 December 20, 2017 NRC-IC-17-051 T4.12.l Rec'd 1/02/18 HHE

SUBJECT:

FINAL SAFETY EVALUATION FOR AREVA INC. TOPICAL REPORT ANP-10338P, REVISION 0, "AREAŽ-ARCADIA ROD EJECTION ACCIDENT" (CAC NO. MF7009/EPID:

L-20-15~TOP-004)

Dear Mr. Peters:

By letter dated October 9, 2015 (Agencywide Documents Acce*ss and Management System (ADAMS)Accession No. ML 15300A298), AREVA Inc. (AREVA) submitted Topical Report (TR) ANP-10338P, Revision 0, "AREAŽ -ARCADIA Rod Ejection Accident," to the U.S. Nuclear Regulatory Commission (NRG) staff for review and approval.

By letter dated October 31, 2017 (ADAMS Accession No. ML 17243A 195), an NRC draft safety evaluation (SE) regarding our approval ,of TR ANP-10338P, Revision 0, was provided for your review and comment By letter dated November 17, 2017 (ADAMS Accession No. ML 17325A999), AREVA provided comments on the draft SE. The NRC staff's disposition of the AREVA comments oh the draft SE are discussed in the attachment (ADAMS Accession No. ML 17334A780) to the final SE enclosed with this letter. The NRG staff has found that TR ANP-10338P, Revision 0, is acceptable for referencing in licensing applications for nuclear power plants to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our aqceptance of the TR. Our acceptance applies only to material provided in the subject TR. We do not intend t9 repeat our review of the acceptable material described in the TR. When the TR appears as a

• reference in licensing action requests, our review will ensure that the material presented applies to the specific plant involved.

Requests for licensing cictions that deviate from this TR wiil be subject to a plant-specific review in accordance with applicable review standards.

NOTICE: The enclosure transmitted herewith contains Proprietary Information.

Whe.n separated from the enclosure, this transmittal document is decontroUed.

___J

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G. Peters In accordance with the guidance provided on the NRC website, we request that AREVA publish approved versions of TR ANP-10338P, l}evision 0, within 3 months of receipt of this letter. The approved versions shall incorporate this letter and the enclosed final SE after the title page. Also, they must contain historical review information, including NRC requests for additional information and your responses.

The approved versions shal! include ~n "-A" (designating approved) following the TR identification symbol, As an alternative to including the RAls and RAI responses behind the title page, if changes to the TR were provided to the NRC staff to support the resolution of RAI responses, and if the NRC staff reviewed and approved those changes as described in the RAI responses, there are two ways that the accepted version can capture the RAls: 1. The RAls and RAI responses can be included as an Appendix to the accepted version. 2. The RAis and RAI responses can be captured in the form of a table (inserted after the final SE) which summarizes the changes as shown in the approved version of the TR. The table should reference the specific RAls and RAI responses which resulted in any changes, as shown in the accepted version of the TR. If future changes to the NRC's regulatory requirements affect the acceptability of this TR, AREVA will be expected to revise the TR appropriately or justify its continued applicability for subsequent referencing.

Licensees referencing this TR would be expected to justify its continued applicability or evaluate their plant using the revised TR. Project No. 728 Docket No. 99902041

Enclosure:

Final Safety Evaluation (Proprietary)

FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT ANP-10338P.

REVISION 0, "AR.EAŽ -ARCADIA ROD EJECTION ACCIDENT" AREVA INC. PROJECT NO. 728/DOCKET NO. 99902041

1.0 INTRODUCTION

AND BACKGROUND By letter dated October 9, 2015, AREVA Inc. (AREVA) submitted Topical Report (TR) ANP-10338P, Revision 0, "AREAŽ -ARCADIA Rod Ejection Accident", to the U.S. Nuclear Regulatory Commission (NRC) for review and approval (Reference 1 ). The TR is a methodology for the evaluation 9f a control rod ejection accident in a pressurized water reactor (PWR) for referencing in licensing actions. 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 attaches, This failure can result in a control rod being ejected from the core by reactor coolant system (RCS) pressure forces on the control rod drive mechanism.

This rapid ejection of a high neuron absorber c~uses a step incre?se in rf3activity in the core and consequently, a rapid increase iii the core power in the vicinity of the ejected rod. The REA is postulated to be a large power increase that may potentially challenge the integrity of the RCS via a spike in presSlure and degradation of core coolability.

The AREAŽ methodology couples a variety of AREVA computer codes and analysis methods; such as the ARCADIA code system (Reference

2) f9r the analysis of the three-dimensional (3-D) neutronic, thermal-hydraulic, fuel pin thermal mechanical behavior and S-RELAPS (Reference 3), a RCS response code. Through this coupling of codes, the AREAŽ methodology provides a capability to demonstrate a conservative representation of the reactor response during a REA that is in compliance wit.h

• regulatory criteria.

This demonstration is based on the com-putation of quantities such as energy deposition, fuel rim melt, fuel centerline melt, minimum departure from nucleate boiling ratio (DNBR), and RCS pressure.

2.0 REGULATORY EVALUATION

The AREAŽ methodology basically consists of coupling the NRG approved reactor analysis system code ARTl:.MISŽ to the NRC approved Code S-RELAP5 to account for the RCS response in the rod ejection transient.

ARCADIA is a code package that provides a converged code system for neutroriic and thermal-hydraulic core design and safety evaluation.

The main components Of the ARCADIA system are the spectrai/lattice code APOLL02-A and the .core simulator ARTEMISŽ.

The core simulator ARTEMISŽ is a 3-0 nodal multigroup reactor burnup code with pin power reconstruction for PWRs. In addition, a thermal-hydraulic progr?m COBRA~FLXŽ that is capable of performing 3-D steady-state, -transient full-core, and

• subc;hannel analyses has been integrated into ARTEMISŽ to have th<:3 capability to solve complex two-phase flow problems.

The fuel pin temperatures for use with COBRA-FLXŽ are calculated by the fuel rod model .(FRM). This model, within the ARTEMISŽ code system, solves both the static and time~dependent, one-dimensional thermal equ?tions for th.e fuel rods to compute the fuel temperature di.stribution and heat flux to the coolant. Enclosure In addition, the AREAŽ methodology introduces elements of the currently under NRC review fuel performance code GALILEOŽ (Reference 4), used to define fuel and clad thermal properties for the FRM in both the neutronics solution and the thermal-hydraulic solution in ARTEMISŽ, The introduction of elements of GALILEOŽ also allows for capabilities on the pin level as opposed to ARTEMISŽ's use of nodal average properties.

Thus, the focus of the review is the NRC un-reviewed elements of the AREAŽ methodology.

In particular, the stability of the coupling between ARTEMISŽ and S-RELAP5 and the effect of the introduction of GALILEOŽ models on the figures of merit associated with the acceptance criteria.

2.1 GALILEOŽ 2.1.1 Thermal-Mechanical Properties.

The TR as submitted to the NRC forreview and approval, was written under the assumption that GALILEOŽ is an NRC approved FRM. Given that currently GALILEOŽ is not approved, a request for additional information (RAI) question (RAl-2, Reference

5) was formulated to allow AR.EVA to justify reference to GALILEOŽ as a computational component within the FRM in ARTEMISŽ.

AREVA responded that the basic motivation is for AREAŽ to remain consistent with both ARCADIA and GALILEOŽ, so that on approval of GALILEOŽ no further justification is needed. In light of the modular structure of GALILEOŽ, it is clearly feasible to introduce individual independent modules with regard to the thermal-mechanical properties of the fuel pins and link GALILEOŽ computed parameter values to ARTEMISŽ.

AREVA states that the FRM thermal property default equations in ARTEMISŽ (thermal conc!uctivity, specific heat; and fuel pellet radial power profile fit function) are equivalent to the thermal property equations in GALILEOŽ.

Thus, ARCADIA and GALILEOŽ use consistent thermal-mechanical properties.

In its response, AREVA shows a table that points to the specific sections of the GALIL.,EOŽ theory manual that describes the mathematical model for the computation of each of the relevant GALILEOŽ thermal parc!meter models. Furthermore, GALILEOŽ/ARTEMISŽ transient comparisons of fuel and clad temperatures are given in Tables 5.2, 5.3, and 5.4 of the TR. For all cases, the_ trend differences are well behaved. In addition, ARTEMISŽ does not account for thermal expansion, which primarily affects the gap conductance model. Gap conductance is a complex function of clad and fuel surface temperature, gap size, roughness, contact pressure, and fission gas content. To capture the complex effects of burnup on gap conductance, as well as transient effects due to thermal expansion, GALILEOŽ is run statically at [ .], and the resultant gap conductance.

These calculations generate!

for each unique fuel composition (weight percent U235 and Gadolinia content) a table, which represents a response surface for the gap conductance as a function of the state vector components.

• 2.1.2 Concomitant Relevant New Capabilities AREVA's response to RAl-2 points out that the advantage of the GALILEO TM modules lies in that they introduce capabilities that reflect pin properties, as opposed to the nodal (i.e. nodal average fuel pin) properties c:1vailable from ARTEMISŽ.

These capabilities are instrumental in *determining the REA specific regulatory criteria estimates associated with:

P*ellet Cladding Mechanical Interaction Fuel pin pressure Fuel and rim melt temperature Fuel and clad thermal properties

-Gap conductance Furthermore, AREVA presents in response to RAl-2 a table that describes the key differences expected in using the ARTEMISŽ FRM modwle relative to the new GALILEOŽ capabilities for the calculation of the REA specific regulatory criteria.

This is illustrated in Figures 5-2 through 5-7 in the TR, with comparisons of the transient of fuel center line temperatures and transient maximum rim temperatures between the ARTEMISŽ models and the GALILEOŽ models in AREAŽ. These show that using GALILEOŽ generated tables, ARTEMISŽ is capable of modeling fuel temperature behavior with respect to time for pellet centerline and for rim and pellet surface. The differences in the figures are attributed, in AREVA's response to RAl-2, to the fact that the FRM calculations in ARTEMISŽ use static gap conductance, whereas GALH-EOTM uses dynamic gap conductance.

Moreover, these results show that the [. . . -J that is applied to the rin:i temperature for fuel containing Gadolinia.

Based on AREVA's RAI responses and references to GALILEOŽ models of clad corrosion, therm*a1 properties, fuel pin pressure calculation, and attendant performance verification and validation in the TR, the introduction of the proposed GALILEOŽ modules in AREAŽ is well tested and acceptable.

• 2.2 ARTEMISŽ -S-RELAP5 Coupling The submitted TR contained inadequate information with regard to type of coupling, data flow of system and core coupled state calculations, calculation time steps, and specific state variables passed between ARTEMISŽ and S-RELAP5_.

Further clarification was requested by the NRC in the form of RAl-1 (Reference 5). AREVA responded with detailed clarification.

AREVA submitted an Appendix D that describes*in detail the ARTEMISŽ and S-RELAP5 coupling.

This Appendix D, ARTEMISŽ and S-RELAP5 Coupling P~scription, will be added to the TR when the approved version is issue.d. The new appendix addresses in detail ARTEMISŽ Nodal Transient Calculation, Model Transient Calculation, and ARTEMISŽ Time Step Manager. The AREA tM computational methodology fundamentally consists of a parallel processing scheme that allows coupling the large NRC approved multi-physics codes ARTEMISŽ and S-RELAP5 to be run separately-and exchange data during the calculation.

The data exchange is performed through Message Passing Interface (MPI) environment.

This approach is consistent with the current state-of-the-art modelling strategies by industry.

In that, it allows the coupling of existing codes, Which are linked in a parallel processing scheme that allows the codes to be run separately and exchange data during the calculation vi.a MPI. The approach described in detail in Appendix D shows that the approach to coupling ARTEMISŽ and S-RELAP5 is consistent with current industry practice and is satisfactory.

2.3 Sensitivity

and Stability Assessment The AREAŽ methodology consists of a system of coupled codes: ARTEMISŽ, with GALILEOŽ modules used to generate property lookup tables for the FRM in ARTEMISŽ and S-RELAP5.

This coupled system performs the calculations that demonstrat_e compliance with regulatory criteria related to enthalpy, DNBR, fuel temperature, fuel pin pressure, transient fission gas release (FGR), and RCS pressure.

To this end, the AREAŽ methodology uses the 3-D nodal transient code ARTEMISŽ with [ ] in the core. It computes fuel internal pressure to evaluate coolability when fuel failures occur. Given its [ ] it also provides the number of effective fuel pin failures for dose evaluation.

The coupling to the reactor system code S-RELAP5 computes the maximum system pressure, and the core pressure used to determine DNB. The stability of analyses of the AREA TM methodology, in the co11text of a REA, is approached by AREVA via a sensitivity and biasing analysis of key parameters that affect the calculated values used to compare to the acceptance criteria.

These latter are: Maximum Fuel Temperature Maximum Rim Temperature Maximum Enthalpy Rise Maximum Total Enthalpy Minimum DNBR Maximum Energy to the Coolant during the Transient The initial choice of key parameters that affect these calculated values for. comparison to regulatory criteria are determined by a Phenomena Identification Ranking Table (PIRT) evaluation in relation to REA model requirements.

This evaluation results in the identification of parameters directly addressed by th.e AREAŽ methodology and system parameters considered for pressure analysis.

2.3.1 Core Sensitivity Analysis For the core sensitivity analysis adjustment factors on the following parameters are used to account for uncertainty and conservative allowances:

Fuel conductivity Fuel heat capacity Gap conductance Power peaking Core inlet flow Cross section changes to control rod Cross section changes due to fuel temperature variation Cross section changes due to moderator temperature variation These adjustments are applied to assess the sensitivity of the system response.

In light of this preliminary sensitivity analysis, these parameter adjustments are used to bias the limiting cases in order to generate conservative results that meet the acceptance criteria.

This results in a base case for the. transient with the following integral parameters biased by a representative uncertainty:

Increase in Ejected Rod Worth (ERW) Increase in Doppler temperature coefficient (OTC) (less negative)

Decrease in delayed neutron fraction i3 Increase in moderator temperature coefficient (MTG) (more positive)

Increase in fuel pin power peaking In addition to this base case, a nominal case is established where the above parameters are not biased. The difference of these two cases (base case -nominal case) gives the minimum amount of conservatism with regard to computed results in the methodology.

Furthermore, sensitivity analysis is performed using a one-at-a-time (OAT) design (Reference 6). That is, the biasing is removed individually for each parameter from the base case and transient calculations are performed for each. This results in three .computed response values for each dependent variable:

-A value with an individual bias removed (R-bias) A value from the base case (R-base) A value from the nominal case (R-nominal)

The importance of the sensitivity as a function of an individual parameter is defined by the following ratio, R-parameter

= [ ]/[ 1 This response measure is computed for six transient conditions, two extremes of cycle burn up (beginning of cycle and end of cycle), and three power levels (0, 20, and 100 percent).

These results are tabulated for the six computed values of dependent variables used to compare to the acceptance criteria and shown previously above. It is recognized that one limitation of OAT designs is that, from a 'statistical' point of view, they do not enable estimation of interactions among parameters over the bias ranges. The OAT design used by AREVA is local, in that the response to the bias .is dependent on the choice of the nominal values and linear over the range of the bias. AREVA implicitly addresses this issue through a thorough PIRT analysis, and by implementation of a thorough sensitivity evaluation method. The former adds significant historical engineering judgement; the latter a reasonable level of assurance, through a thorough screening process of the results of the six transients described previously.

This approach is consistent with current industry practice and is satisfactory.

2.3.2 System

Pressure Response Sensitivity Analysis The S-RELAP5 computer code is automatically coupled to ARTEMISŽ, as described in the additional Appendix D, and used for RCS pressure calculations. (The NRC approved RELAP5/MOD2-B&W

[Babcock & Wilcox Company] computer code is utilized in the evaluation of the REA event for a B&W plant. In addition, the RELAP5/MOD2-B&W code is manually coupled to ARTEMISŽ for the REA analysis.)

Two scenarios are evaluated.

In the first scenario, the maximum system pressure is determined.

In the second scenario, core pressure for determining DNB is evaluated.

The 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.

To this end, the system parameters (excluding the core parameters) considered for biasing are; Initial pressure Initial pressurizer level Initial power level Reactor trip system settings PSV relief capacity L __ The impact, the value, and their basis are assessed for each. This approach is consistent with current industry practice and is satisfactory.

3.0 AREA TM Plant Specific Application and $ample Problems The application of the AREAŽ methodology presented in this TR is demonstrated in plant specific applications, which capture the impact of cycle-to-cycle variations in an REA, for three different PWR plant types. These are: a Westinghouse 4-Loop plant with a 17X17 fuel lattice, a B&W 177 fuel assembly (FA) plant with a 15X15 fuel lattice, and a Combustion Engineering 217 FA plant with a 14X14 fuel lattice. These plant applications demonstrate the AREAŽ methodology.

The applications of the methodology consist of two phases: an initial application of the methodology and a follow-on application to capture the impact of cycle-by-cycle variations in an REA. The key elements of the initial phase are: Estc;1blish appropriate fuel limits in the context of regulatory requirements.

Verify that the biases are acceptc;1ble by performing selected sensitivity analyses as described on this TR. Determine any biases and penalties that are plant specific.

Run a matrix of cases as defined in this TR to establish the margin to the limiting conditions.

Run RELAP5 for the two pressure scenarios that define the margin to the high pressure limit and verify the DNBR calculations.

The initial phase is followed by a cycle-to-cycle evaluation.

The initial phase's function is to establish biases of key parameters and provide a basis for the expectation that they are bounding for future cycles. 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 phase analyses.

The limiting cases are analyzed for a time-in..:life and power level matrix. The key parameters considered are: ERW Delayed Neutron Fraction 13 -*ore MTG . Initial 3-0 power peaking factor (Fa) Initial 2-0 enthalpy rise peaking factor (Ft.H) Static post ejection Fa Static post ejection Ft.H Fuel failures .If key parameters, established in the initial application of the AREAŽ methodology, are exceeded, the methodology allows two approaches to address future cycles: 1. Complete reanalysis of the matrix of cases. 2.. Reanalysis of a portion of the matrix of cases is repeated for the co.ndition where a specific parameter is found to be outside of the initial application analysis range. In the plant~specific application in the three sample problems, no criteria are exceeded nor are failures predicted.

4.0 CONCLUSION

S The technical review of the AREAŽ methodology for the evaluation of a control rod ejection accident in a PWR, and the results of sample plant applications has been completed.

As evidence of the efficacy of the AREAŽ methodology, the documentation provided in the TR and the responses by AREVA to the RAls, demonstrate that the methodology provides a stable and conservative representatiori of the reactor response during an REA. Furthermore, the documentation demonstri;ltes thcit the computed estimates of figures of merit are able to demonstrate compliance with the appropriate regulatory criteria.

Thus, based on the technical evaluation above, the NRC staff finds the AREAŽ methodology for the evaluation of a control rod ejection accident in a PWR acceptable for licensing appiications subject to the limitations.

and conditions specified in Section 5.0 of this SE. 5.0 LIMITATIONS AND CON-DITIONS

1) The AREAŽ methodology is limited to the evaluation of a control rod ejection accident in a PWR. 2) The AREAŽ methodology consists of coupled AREVA codes and methods. The applicatior, of the ARENrvi methodology is limited to the conditions and limitations of the SES of the approved codes that the AREA TM methodology uses in its analysis of a c::ontrol rod ejection accident.

-3) The AREAŽ methodology is limited to only the GALILEOŽ derived thermal-mechanical properties of fuel pins. The use of another NRG approved code for the thermal-mechanical properties must be noted and the compLJted d.ifferences quantified and Justifiec;l.

6.0 REFERENCES

1.. ANP-10338P, "AREA TM -ARCADIA Rod Ejection Accident," October 2015 (Agency-wide Documents Access and Management System (ADAMS) Package Accession No. ML 15300A314).

2. ANP-10297P-A, Revision 0, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," Topical Report, February 2013 (AOAMS Package Accession No. ML 14195A145).

3.. Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRG), "Document to Support the NRG Review of EMF-2103P, Revision 3,'Realistic Large Break LOCA Methodology for Pressurized Water Reactors,"'

January 10, 2014 (Enclosed is a copy of EMF-2100(P), Revision 16, "S-RELAP5 Models and Correlations Code Manual," December 2011) (ADAMS Accession No. ML 14016A220). 4. ANP-1 Q323P, Revision O, "Fuel Rod Thermal-Mechanical Methodology for Boiling Water Reactors and Pressurized Water Reactors," Juiy 2013 (ADAMS Package Accession No. ML 13218A013).

5. ANP-10338Q2, Revision 0, "Response to Request for Additional lnformation-ANP-10338P," August 2017 (ADAMS Accession No. ML17229B156).

6. A. Saltelli, K. Chan and E. M. Scott, "Sensitlvity Analysis," .John Wiley & Sons, 2000 (New York).

Attachment:

Comment Resolution Table Principal Contributor:

Y. Orechwa, NRR/DSS/SNPB Date: December 20, 2017 RESOLUTION OF COMMENTS BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON DRAFT SAFETY EVALUATION FOR TOPICAL REPORT ANP-10338P, REVISION 0, "AREAŽ-ARCADIA ROD EJECTION ACCIDENT" AREVA, INC. PROJECT NO. 728/DOCKET NO. 99902041 This attachment provides the U.S. Nuclear Regulatory Commission (NRC) staff's review and disposition of the comments made by AREVA Inc. on the draft safety evaluation (SE) for Topical Report ANP-10338P, Revision 0, "AREAŽ -ARCADIA Rod Ejection Accident." Page Line Proposed Change/Comment NRC Resolution of Comments 1 33 Replace ARCADIA with ARTEMISŽ, this The NRC staff agrees with the is because ARCADIA is a code package editorial suggestion.

Change and the specific code is ARTEMISŽ implemented in final SE. 2 7 Replace ARCADIA with ARTEMISŽ The NRC staff agrees with the editorial suggestion.

Change implemented in final SE. 2 8 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 2 9 Replace AREA with AREAŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 2 10 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 2 11 Replace GALILEO with GALILEOŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 2 41 Replace the word "statistically" with The NRG staff agrees with the "statically"

• editorial suggestion.

Change implemented in final SE. 3 5 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 3 15 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. Attachment Page Line Proposed Change/Comment NRC Resolution of Comments 3 24 Replace the word "without" with "compared The NRG staff agrees with the to" editorial suggestion.

Change implemented in final SE. 3 24 Delete the word "tables" The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 3 32 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 3 36 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 4 3 Replace ARCADIA with ARTEMISŽ The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 5 38 Replace the word "liner" with "linear" The NRG staff agrees with the editorial suggestion.

Change implemented in final SE. 7 49 Delete the words "in place of those of the The NRG staff agrees with the NRC approved FRM in the ARCADIA editorial suggestion.

Change code." This is because the FRM properties implemented in final SE. in the ARCADIA code package are the same properties as in the GALILEOŽ code as described in the response to RAI 2-A.

October 9, 2015 NRC:15:038 U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 A AREVA Request for Review and Approval of ANP-10338P, "AREAŽ -ARCADIA Rod Ejection Accident" AREVA Inc. (AREVA) requests the NRCs review and approval of the topical report ANP-10338P, "AREAŽ -ARCADIA Rod Ejection Accident

for referencing in licensing actions. This repqrt presents a methodology for the evaluation of a control rod ejection accident in a Pressurized Water Reactor (PWR). The methodology is used to demonstrate compliance with the criteria specified in NUREG-0800, Section 4.2, Appendix B. NUREG-0800, Section 4.2, Appendix B contains the current Nuclear Regulatory Commission criteria for a control rod ejection accident.

The methodology is consistent with the guidance in Regulatory Guide 1.77, and NUREG-0800, Section 15.4.8. In support of the Office of Nuclear Reactor Regula.tion's prioritization efforts, the Topical Report Prioritization Scheme is included as an endosure with this letter. AREVA would appreciate the NRC approval of this topical report by.September 2017. AREVA considers some of the materiai contained in the enclosed *document to be proprietary.

As required by 10 CFR 2.390(b), an affidavit is enclosed to support the withholding of the information from public disclosure.

A proprietary version and a non-proprietary version of the report are enclosed.

There are no commitments within this letter or its enclosures.

If you have any questions related to this information, please contact Ms. Gayle F. Elliott, Product Licensing Manager, by telephone at (434) 832-3347, or by e-mail at Gayle.Elliott@areva.com.

ro s, Director Licensing

& Regulatory Affairs AREVA Inc. cc: J. G. Rowley Project 728 AREVA.INC.

331'5 Old Forest Road, LynchbiJrg.VA.24591 Tel.: 434 832 3000 -www.areva.com Document C_ontrol Desk October 9, 2015

Enclosures:

NRC:15:038 Page 2 1. Proprietary version of ANP-10338, "AREAŽ -ARCADIA Rod Ejection Accident," October 2015. 2. Non-Proprietary version .of ANP-10338, "AREAŽ~ ARCADIA Rod Ejection Accident," October 1015. 3. Topical Report Prioritizc;1tion

$theme 4. Notarized Affidavit AFFIDAVIT COMMONWEAL TH OF VIRGINIA ) ) ss. CITY OF LYNCHBURG ) 1. My name is Gayle Elliott. I am Manager, Product Licensing, for AREVA Inc. (AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.

I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in the following document:

",ANP-10338P, 'AREAŽ -ARCADIA Rod Ejection Accident Topical Report'," referred to herein as "Document," Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type custoinarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contc:1ined in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Comlilis.sion in confidence With the request that the information contained in this Document be Withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied qy AREVA to determine whether information should be classified as proprietary: (a) The information reveals qetails of AREVA's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The 'information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA. (d) The information reveals certain distinguishing -aspects of a process, methodology, or component, the exclusive use of which provides a competltive advantage for AREVA in product optimization or marketabili~y. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA. The information in this Document is considered proprietary for the reasons set forth iri paragraphs 6(c) through 6(e) above. 7. In accordance with AREVA's policies governing

{he protection and control of information, prnprietary information contc!ined in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured flle Cir area and distributed oh a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowled9e, information, and belief. $UBSCR1BED before me this _qth. day of Qc%6:}rv( I 2015. Sherry L. McFade.n NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMIS.SION EXPIRES: 10/31/18 Re~;i. # 7079129 SHERRY(. MCFADEN Notary Public oommon'wealth of Virginia '7079129 My commission Expires Oct 31, 2018.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Gary Peters, Director Licensing and Regulatory Affairs AREVA Inc. 3315 Old Forest Road Lynchburg, VA 24501 June 1, 2016

SUBJECT:

ACCEPTANCE FOR REVIEW OF AREVA INC. TOPICAL REPORT ANP-10338P, "AREAŽ -ARCADIA ROD EJECTION ACCIDENT" (TAC NO. MF7009)

Dear Mr. Peters V\ rv-\t..* it,

• o\O Y(,.U!Alf'-A ltln'IIC, By letter dated October 9, 2015 (Agencywide Documents Access and Management System Accession Number ML 15300A298), AREVA Inc. (AREVA) submitted for U.S. Nuclear Regulatory Commission (NRC) staff review Topical Report (TR) ANP-10338P, "AREAŽ -ARCADIA Rod Ejection Accident." The NRC staff has performed an acceptance review of TR ANP-10338P.

We have found that the material presented is sufficient to begin our review. The NRC staff expects to issue its request for additional information by October 14, 2016 and issue its draft safety evaluation (SE) by April 7, 2017. This schedule information takes in consideration the NRC's current review priorities and available technical resources and may be subject to change. If modifications to these dates are deemed necessary, we will provide appropriate updates to this information.

The NRC staff estimates that the review will require approximately 680 staff hours including project management and contractor time. The review schedule milestones and estimated review costs were discussed and agreed upon in a telephone conference between AREVA Product Licensing Manager, Jerald Holm, and the NRC staff on May 23, 2016. Section 170.21 of Title 10 of the Code of Federal Regulations requires that TRs are subject to fees based on the full cost of the review. You did not request a fee waiver; therefore, NRC staff hours will be billed accordingly.

As with all topical reports, the SE will be reviewed by the NRC's Office of the General Counsel (OGC) to determine whether it falls within the scope of the Congressional Review Act (CRA). During the course of this review, OGC considers whether any endorsement or acceptance of a TR by the NRC amounts to a rule as defined in the CRA. If this initial review concludes that the SE, with its accompanying TR, may be a rule, the NRG will forward the package to the Office of Management and Budget (0MB) for further review and consideration.

Any review by 0MB would impact the schedule for the issuance of the final SE.

G.Peters If you have questions regarding this matter, please contact Jonathan G. Rowley at (301) 415-4053.

Project No. 728 Sincerely, £=.Chief Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation UNITED STATES NRt-Y:C-11-002..

f<ec-'4 , J~o (11~ NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Gary Peters, Director Licensing and Regulatory Affairs AREVA Inc. 3315 Old Forest Road Lynchburg, VA 24501 January 23, 2017

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RE: AREVA INC. TOPICAL REPORT ANP-10338P, "AREA-ARCADIA ROD EJECTION ACCIDENT" (CAC NO. MF7009)

Dear Mr. Peters:

By letter dated October 9, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15300A298), AREVA INC. (AREVA) submitted for U.S. Nuclear Regulatory Commission (NRC) staff review and approval Topical Report (TR) ANP-10338P, "AREA-ARCADIA Rod Ejection Accident." Upon review of the information provided, the NRC staff has determined that additional information is needed to complete the review. On December 12, 2016, Jerald Holm, AREVA Product Licensing Manager, and I agreed that the NRC staff will receive the response to the enclosed request for additional information (RAls) questions by March 31, 2017. If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-4053.

Project No. 728

Enclosure:

RAI Questions Sincerely, /**, I *' ( l ( r /J . *. ~--i//-' ~,, ,, ?4;-i, 1 JtL /11,,.,J iJ6nathan G. Rowley, Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION RELATED TO TOPICAL REPORT ANP-10338 "AREA -ARCADIA ROD EJECTION ACCIDENT" AREVA INC. (CAC NO. MF7009) The ARCADIA Rod Ejection Accident (AREA) analytic methodology consists of a sequence of multi-physics coupled codes. The transient be.havior of the computed figures of merit are dependent on the evolution of the state vector that defines, in a consistent manner, the evolution of the values of the state vector components.

These allow the computation of integral values of the figures of merit for comparison to the defined regulatory limits. A. Define the time-dependent state vector (i.e., give the vector components) that is computed by AREA for the calculation of the time behavior of the figures of merit that demonstrate that the regulatory acceptance criteria are met. B. Elaborate Figure 5-1 so that the flow of the coupling within ARTEMIS and of ARTEMIS/S-RELAP5 is shown clearly as either internal, external, or parallel with regard to the information in the state vector as it flows through the coupled codes. C. Elaborate in detail the coupling scheme shown in Figure 6-1 by following the components of the state vector over one time step. 1. Identify the couplings and show whether they are simultaneous or staggered.

In particular, is the ARTEMIS/GALILEO one-way or two-way? 2. How do you assure that the state vector components are converged within a time step? That is, it appears that you are applying an Operator Splitting methodology.

In that case, the action of the governing equations on the variables is decomposed into a separate, uncoupled physical description for each part, leading to an inconsistent treatment of the nonlinear terms. 3. Identify the automatic couplings and the manual couplings in the context of one time step. Give the rational for this distinction and the rules for the application.

D. Define a reference case to demonstrate the results of an application of an AREA analysis.

Enclosure 1. Outline the algorithm for the computation of the converged steady state (i.e., initial condition) for a rod ejection analysis of the reference case. Outline the algorithm for the computation of the converged steady state (i.e., initial condition) for a rod ejection analysis of the reference case. 2. Outline the algorithm for th.e computation of the converged time behavior of the state vector for a rod ejection analysis of the reference case (i.e., of each component of the state vector).

• March 31, 2017 NRC:17:015 U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 A AREVA Response to Request for Additional Information Regarding ANP-10338P Revision O "AREAŽ -ARCADIA Rod Ejection Accident" Ref. 1: Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), "Request for Review and Approval of ANP-10338P, 'AREAŽ -ARCADIA Rod Ejection Accident'," NRC:15:038, October 9, 2015. Ref. 2: Letter, Jonathan G. Rowley (NRC) to Gary Peters (AREVA Int_.), "Request for Additional Information Re: AREVA Inc. Topical Report ANP-10338P, 'AREA-ARCADIA Rod Ejection Accidt:!rit' (CAC No. MF700Q)," January 23, 2017. AREVA Inc. (AREVA) requested the NRC's review and approval of the topical report ANP-10338P, "ARENM -ARCADIA Rod Ejection Accident" in Reference

1. The NRC provided a Request for Additional Information (RAI) in Reference

2. The response to this RAI is endosed with this letter. AREVA considers some of the material contained in the enclosed to be proprietary.

As required by 10 CFR 2.390(b), an affidavit is enclosed to support the withholding of the information from public disclosure.

Proprietary and non-proprietary versions of the RAI response are provided.

There are no commitments within this letter or its enclosures.

If you have any questions related to this information, please contact Ms. Gayle F. Elliott (Product Licensing Manager) by telephone at (434) 832-3347, or by e-mail at Gayle.Elliott@areva.com.

Sincerely, Gary Peters, Director Licensing

& Regulatory Affairs AREVA Inc. cc: J. G. Rowley Project 728 AREVA INC. 3315 Old Forest Road, Lynchburg, VA 24501 Tel.: 434 832 3000 -www.areva.com Document Control Desk March 31, 2017

Enclosures:

1. Proprietary copy of ANP-10338 QlP, Revision 0, "Response to Request for Additional Information -ANP-10338P" NRC:17:015 Page 2 2. Non-Proprietary copy of ANP-10338 QlNP, Revision 0, "Response to Request for Additional Information -ANP-10338P" 3. Notarized Affidavit AFFIDAVIT COMMONWEAL TH OF VIRGINIA ) ) ss. CITY OF LYNCHBURG ) 1. My name is Morris Byram. I am Manager, Product Licensing, for AREVA Inc. (AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.

I am familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in ANP-10338 Q1P, Revision 0, entitled "Response to Request for Additional Information -ANP-10338P," and referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that 9ther companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

' requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA. The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d), and 6(e) above. 7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis. I ____ ---------

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this-"'--7)-~t.b

__ dayof I 2017. Sherry L. McFaden NOTARY PUBLIC, COMMONWEAL TH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg.# 7079129 SHERRY (, MCFADeN Noiiry Publlc ConlmonWHlttl of Virginia . 7079129 My Comml11lon E>1plres Oct 31, 2018 A AREVA Response to Request for Additional Information -ANP-10338P Topical Report March 2017 AREVA Inc. (c) 2017 AREVA Inc. ANP-10338 Q1NP Revision O Copyright

© 2017 AREVA Inc. All Rights Reserved ANP-10338 Q1NP Revision 0 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Item 1 Section(s) or Page(s) All Nature of Changes Description and Justification Initial Issue ANP-10338 Q1NP Revision 0 Page i AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Contents Page 1.0 RAI 1 ....................................................................................................................

1

2.0 REFERENCES

...................................................................................................

16 3.0 TOPICAL REPORT MARKUP PAGES ...............................................................

17 Page ii AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report List of Tables ANP-10338 Q1 NP Revision 0 Page iii Table 1-1 Maximum Fuel Values from Case 4, Detailed Model. .................................

10 Table 1-2 Rate of Change for S-RELAP Parameters

.................................................

10 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report List of Figures Figure 1-1 GALILEOŽ Process for Thermal Properties

..............................................

11 Figure 1-2 Model Validation

.........................................................................................

12 Figure 1-3 ARTEMISŽ Core and Cycle Design Calculations

......................................

13 Figure 1-4 Coupled Calculations with ARTEMISŽ and S-RELAP5 .............................

14 Figure 1-5 Core Fa during Time of Peak Power ..........................................................

15 Page iv AREVA Inc. Response to Request for Additional Information

-ANP-10338P Topical Report Acronym BOC DNBR EOC NRC RAI Nomenclature Definition Beginning of Cycle Departure from Nucleate Boiling Ratio End of Cycle United States Nuclear Regulatory Commission Request for Additional Information ANP-10338 Q1NP Revision 0 Pagev AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Introduction The United States Nuclear Regulatory Commission (NRC) provided a request for additional information (RAI) regarding the topical report ANP-10338 (Reference

1) in Reference

2. One question was received from the NRC. Page vi AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report 1.0 RAI 1 Question:

The ARCADIA Rod Ejection Accident (AREA) analytic methodology consists of a sequence of multi-physics coupled codes. The transient behavior of the computed figures of merit are dependent on the evolution of the state vector that defines, in a consistent manner, the evolution of the values of the state vector components.

These allow the computation of integral values of the figures of merit for comparison to the defined regulatory limits. A. Define the time-dependent state vector (i.e., give the vector components) that is computed by AREA for the calculation of the time behavior of the figures of merit that demonstrate that the regulatory acceptance criteria are met. B. Elaborate Figure 5-1 so that the flow of the coupling within ARTEMIS and of ARTEMIS/S-RELAP5 is shown clearly as either internal, external, or parallel with regard to the information in the state vector as it flows through the coupled codes. C. Elaborate in detail the coupling scheme shown in Figure 6-1 by following the components of the state vector over one time step. 1) Identify the couplings and show whether they are simultaneous or staggered.

In particular, is the ARTEMIS/GALILEO one-way or two-way? 2) How do you assure that the state vector components are converged within a time step? That is, it appears that you are applying an Operator Splitting methodology.

In that case, the action of the governing equations on the variables is decomposed into a separate, uncoupled physical description for each part, leading to an inconsistent treatment of the nonlinear terms. 3) Identify the automatic couplings and the manual couplings in the context of one time step. Give the rational for this distinction and the rules for the application.

Page 1 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report D. Define a reference case to demonstrate the results of an application of an AREA analysis.

1) Outline the algorithm for the computation of the converged steady state (i.e., initial condition) for a rod ejection analysis of the reference case. Outline the algorithm for the computation of the converged steady state (i.e., initial condition) for a rod ejection analysis of the reference case. 2) Outline the algorithm for the computation of the converged time behavior of the state vector for a rod ejection analysis of the reference case. (i.e., of each component of the state vector). Response 1.A: . Appendix D has been developed to describe the coupling between ARTEMISŽ and RELAP5. Appendix D is shown in the markup section of this document.

A markup of page 5-1 of the topical report is provided to reference Appendix Din the ANP-10338P (Reference

1) topical report. These markup pages will be included in the approved version of the topical report when it is issued. The time-dependent state variables are provided in the section "System State Variables" of Appendix D. Response 1.8: Appendix D, Figure D-1 provides additional details of the coupling between the codes of the coupled system; S-RELAP5 (Reference 5), ARTEMISŽ (Reference 3 and 6) nodal and ARTEMISŽ detailed model [ ] The equations solved by each code are provided in the section "System Equations" of Appendix D. The internal solution of the ARTEMIS TM nodal calculation is provided in Appendix D.1. The internal solution of the ARTEMISŽ detailed model is provided in Appendix D.2. The parallel progression of the codes and the external exchange of data are described in Appendix D.7. Paqe2 AREVA Inc. ANP-10338 Q1 NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Response 1.C.1: Figure 6-1 of the topical report depicts the work flow process used for the AREA TM (Reference

1) methodology analysis.

The "Fuel and Clad Thermal Properties" and "Gap Conductances" in Figure 1-1 are described in Sections 6.2.4 and 6.2.5 of the topical report. During the initial part of the analysis, the GALILEO TM (Reference

4) code is run to [ ] The "Model Validation" step in the AREA TM analysis provides the assurance that the fuel rod module of ARTEMISŽ agrees with GALILEOŽ (or alternative approved code). The output from this step is a document that compares results from ARTEMISŽ and GALILEOŽ.

This part of the process is shown in Figure 1-2 and described in Section 5.2.2. There is no data provided from this step to the coupled calculations with ARTEMISŽ and S-RELAP5.

The "ARTEMISŽ Core and Cycle Design Calculations" step consists of developing the model for the plant and cycle of interest.

The model is based on the actual or planned cycle design. The output from this step is an ARTEMIS TM restart file at the various burnup points during the cycle. The "ARTEMISŽ (Static) Core Neutronics Analysis" step reads the restart file generated from the cycle design, updates the state parameters to the desired conditions of the core and writes a new restart file. Multiple times during the cycle may be handled. The output from this step is an ARTEMISŽ restart file at the desired initial conditions for the transient.

This step is illustrated in Figure 1-3. Page 3 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report The next step in the work flow is the coupled calculations with ARTEMISŽ and RELAP5 as depicted in Figure 1-4. The external inputs to this process are the data tables from GALILEOŽ, and the ARTEMISŽ restart file at the desired initial conditions.

If desired, penalization of the system pressure can be applied for DNBR calculations in the detailed model. These penalty factors are contained in the data file "General DNBR Pressure" that is generated by S-RELAP5 as described in Section 5-4. The data file "Input to RELAP5" is not used for coupled calculations.

It is used for the manual coupling process with standalone ARTEMISŽ and standalone RELAP5 or S-RELAP5 calculations.

The coupled calculation is described in detail in Appendix D. The main output values are the maximum system pressure, minimum DNBR, maximum fuel temperature and maximum rim temperature.

The remaining parts of Figure 6-1 are evaluation steps concerning the results from the coupled transient calculations.

Response 1.C.2: The multi-physics calculation is using Operator Splitting methodology.

This is a typical approach used in the industry (Reference 7). Each physics code has an internal time step control. Section D.5 provides the description of the S-RELAPS internal time step control. Section D.4 provides the description of the ARTEMISŽ internal time step control. Section D.7 provides the description of the data controller that provides the time step control between S-RELAP5 and ARTEMISŽ.

The minimum and maximum time step sizes over the various time spans are provided as inputs to S-RELAP5 and ARTEMISŽ based on experience with the type of transient.

The internal time step control of S-RELAP5 will stop the code if it determines that the time step size should be reduced to a value smaller than the minimum. The internal time step control of ARTEMISŽ will continue with the minimum time step size even when the internal time step control attempts to further reduce the value. The method used to insure that the selected ARTEMIS TM minimum time steps are sufficiently small is to repeat a transient with a further reduction in the minimum time step size. If the figures of merit are not Page 4 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report significantly affected, the original minimum time step size is confirmed to be appropriate for this type of transient.

This time step evaluation was performed for Case 4 that is described in Section A.4 of the topical report. It is the reference case for this RAI and is described in the response to RAI 1.0.1. [ ] The maximum time step size of S-RELAP5 is also important to whether the coupled solution is converged.

S-RELAP5 starts the problem with the maximum allowed time step. If successful, it continues to use the maximum time step. [ ] Page 5 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Core power and inlet temperature are the key coupling variables between ARETMISŽ and S-RELAP5 for the rod ejection transient.

These values are not distinguishably different between the calculations with the original and the reduced time step sizes. Figure 1-5 shows the core Fa during the part of the transient with the highest power. The values from the original (Base) calculation and the small time step (Small) are plotted together.

The differences are insignificant.

Response 1.C.3: For the AREA TM methodology analysis, the manual couplings to the system core coupled calculations are the data files from GALILEOŽ and the ARTEMISŽ restart file for the initial conditions of the transient.

All of these are provided at the beginning of the transient calculation.

The system core coupled calculations are automatically coupled and there are no further manual steps to advance from time step from tN to tN+l* [ ] Page 6 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Manual coupling between RELAP5 and ARTEMISŽ is demonstrated in Appendix B of the topical report. The severe power response from a rod ejection accident is essentially over in less than a half a second. Feedback from the system is minimal, since the loop time is roughly 10 seconds. The power response after the power pulse is nearly constant so that. the feedback from the system to the core is slow. Hence, the system and core response are separable and comparisons are made between the end results of the manual coupling rather than a per time step basis. In the sample problem, no feedback from the system to the neutronics code is credited so that this check was not necessary.

The distinction is that automatic coupling is used for ARTEMISŽ and S-RELAP5 calculations, while manual coupling is used for ARTEMISŽ and RELAP5 (used for B&W plants) calculations.

The distinction is due to having an automated coupling process with S-RELAP5 and not having an automated coupling process with RELAP5. Response 1 D: Case 4 in Appendix A from the topical report was selected as the reference case. This case represents a limiting condition for system pressure.

BOC conditions are used to minimize the neutronic feedback and the maximum ejected rod worth for the entire cycle in terms of dollars is used to preclude running a case at EOC. The high flux trip is disabled to extend the time at power which yields a maximum pressure response.

This case has not been evaluated with the detailed model in the topical report. The detailed model results are run for Case 4 and are shown in Table 1-1 for maximum fuel temperature, maximum fuel rim temperature and maximum fuel enthalpy rise. This transient would not evaluate DNBR since it's a prompt critical condition.

The maximum pressure for this case is also provided in the table. Note that the enthalpy rise for this case [ ] is slightly larger than maximum enthalpy rise [ ] of any of the cases run in Section A.3 of the topical report. Therefore, Case 4 is a more severe case with respect to the fuel thermal response, in addition to being the limiting condition for system pressure.

Page 7 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Response 1.D.1: The S-RELAP5 code was run in standalone mode with its steady-state controllers for a long time period for the reference case. Major parameters were checked to insure that a viable steady state condition had been reached. A null transient was then run from the end of the steady-state calculation and continuing with the steady-state controllers turned off. The null transient further confirmed that steady state conditions had been achieved.

The system core coupled calculation was then started from the S-RELAP5 steady-state conditions.

Table 1-2 provides the rate of change of several parameters from 4500 seconds and 5000 seconds of the S-RELAP5 steady state initialization.

After the S-RELAP5 steady state model is established, the boundary conditions (exit pressure, inlet temperature and core mass flux) are used in a standalone ARTEMISŽ steady state calculation.

Then, the coupled calculation (External Controller, S-RELAP5, ARTEMISŽ)

is run at the steady state conditions using a null transient to confirm that steady state conditions have been established for the coupled system. Response 1.D.2: The algorithm for the advancement of-a coupled S-RELAP5 and ARTEMISŽ transient calculation is provided in Appendix D.7. Each code provides convergence criteria for the internal solution of its equations.

For example, the ARTEMISŽ calculation has convergence criteria for the neutronics solution, thermal-hydraulics solution and thermal model of the fuel rod module. Page 8 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report The original calculation of this reference case did not include the ARTEMISŽ detailed model since minimum DNBR and individual fuel rod temperatures were not desired. This reference case has been repeated including the ARTEMISŽ detailed model as an example of a system fully core coupled calculation.

The coupled system uses an Operator Splitting approach as outlined in Section 0.7. The convergence of the system is assured by using appropriate

• time step sizes for the type of transient and by a time step reduction test as outlined in the response to RAI 1.C.2. This reference case was used as the sample problem for this time step size confirmation test. Page 9 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Table 1-1 Maximum Fuel Values from Case 4, Detailed Model Table 1-2 Rate of Change for S-RELAP Parameters ANP-10338 01 NP Revision 0 Page 10 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Figure 1-1 GALILEO TM Process for Thermal Properties GALILEOŽ Gap Conductances Fuel and Clad Thermal Properties ANP-10338 Q1 NP Revision 0 Page 11 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Figure 1-2 Model Validation GALILEOŽ ANP-10338 Q1 NP Revision 0 Page 12 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Figure 1-3 ARTEMISŽ Core and Cycle Design Calculations ANP-10338 Q1NP Revision 0 Page 13 AREVA Inc. Response to Request for Additional Info rm ation -ANP-10338P Topical Report Figure 1-4 Coupled Calculations with ARTEMISŽ and S-RELAP5 ANP-1033 8 Q1 NP Revision 0 Page 14 AREVA Inc. Response to Request for Additional Info r mation -ANP-10338P Topical Report Figure 1-5 Core Fa during Time of Peak Power Core FQ 10 9 8 7 a 6 ... 5 ... 0 4 V 3 , ----......

/ "-/ ,/ _/" , 2 1 0 0.05 0.07 0.09 0.1 1 0.13 0.15 Time (s) -Base -Small ANP-10338 Q1NP Revision 0 Page 15 AREVA Inc. ANP-10338 Q1NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report

2.0 REFERENCES

1. ANP-10338P Revision 0, "AREAŽ-ARCADIA Rod Ejection Accident Topical Report," October 2015. 2. Letter, Jonathan G. Rowley (NRC) to Gary Peters (AREVA Inc.), 'Request for Additional Information Re: AREVA Inc. Topical Report ANP-10338P, "AREA -ARCADIA Rod Ejection Accident"', (Accession No. ML 163548596), January 23, 2017. 3. ANP-10297P-A Revision 0, 'The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," February 2013. 4. ANP-10323P Revision 0, "Fuel Rod Thermal-Mechanical Methodology for Boiling Water Reactors and Pressurized Water Reactors," July 2013. 5. Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), 'Document to Support the NRC review of EMF-2103P, Revision 3, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors,"'

NRC:14:001, January 10, 2014, contains a copy of EMF-2100(P), Revision 16, "S-RELAP5 Models and Correlations Code Manual," (Accession No. ML 14016A220), Dec 2011. 6. ANP-10297P-A Revision 0, Supplement 1, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," June 2015. 7. Zertak, 0. et al, "Review of Multi-Physics Temporal Coupling Methods for Analysis of Nuclear Reactors," 0. Zerkak, T. Kozlowski and I. Gajev, Ann. Nucl. Energy, 2015. Page 16 AREVA Inc. ANP-10338 Q1 NP Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report 3.0 TOPICAL REPORT MARKUP PAGES Markup pages to ANP-10338 are provided reflecting the changes discussed in this document.

The markups are for: Page 5-1 Appendix D, which will be added to topical report when approved version is issued. Page 17 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 5.0 ANALYTICAL MODELS Proprietary ANP-10338P Revision O. Page 5-1 The AREA TM 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:

• GALILEOŽ (Reference

7) {COPERNIC (Reference

16) can also be used if the outlined validations are performed}

• ARTEMISŽ ( References 11 and 12), a coupled 3-D kinetics solution with neutronics, fuel rod thermal model, and 3-D thermal hydraulic model

• COBRA-FLXŽ (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 ARTEMISŽ 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 the ARTEMISŽ and S-RELAP5 coupling is provided in Appendix D. A description of these models follows. ti [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary APPENDIX D-ARTEMISŽ AND S-RELAP5 COUPLING DESCRIPTION ANP-10338P Revision 0 Page D-1 ARTEMISŽ (Reference 01 and 04) and S-RELAP5 (Reference

03) can be run as a coupled system to solve time dependent multi-physics problems.

While the AREAŽ methodology allows a manual coupling between the two codes and also allows ARTEMISŽ to be run standalone, the focus of this appendix is on the coupled system method of solution.

Figure D-1 provides a schematic of the major modules and the data coupling between them. The S-RELAP5 code calculates the system response.

The "ARTEMISŽ nodal calculation" calculates the neutronic response for the core. The "ARTEMISŽ Detailed Calculation" calculates the DNBR and fuel rod response [ ] COBRA-FLXŽ (Reference.02) is the name for the core thermal-hydraulics module within ARTEMISŽ that is used for both the nodal simulator and the detailed model (calculation).

The interface between S-RELAP5 and ARTEMISŽ is controlled externally using the "External Controller".

This controller provides the overall time step control for the problem and translation of data between the S-RELAP5 and ARTEMISŽ geometric models. The coupled calculation with S-RELAP5 and ARTEMISŽ includes two simultaneous ARTEMISŽ calculations; one based on a core nodal model and the other based on a detailed core model. Each model is run with its own ARTEMISŽ executable.

The nodal model couples the neutronic, thermal-hydraulics, fuel rod module and dehomogenization modules. The detailed model is [ ] with only the thermal-hydraulic module and the fuel rod module . . The following sections describe the data sharing of the nodal solution, the detailed model solution, the time step management, parallelization, and the coupling between ARTEMISŽ and S-RELAP5.

The code GALILEOŽ (Reference

05) is not coupled with ARTEMISŽ and S-RELAP5 in the time dependent solution.

GALILEOŽ is only used to generate input data for the ARTIEMISŽ fuel rod module and detailed model. As noted in Section 5.0, COPERNIC (Reference

06) could be used in its place. System Equations The equations for the modules in ARTEMISŽ are provided in References 01 and 02. The equations for S-RELAP5 are provided in Reference

03.

AREVA Inc. Proprietary ANP-10338P Revision 0 AREAŽ -ARCADIA Rod Ejection Accident Topical Report Page D-2 a. Neutronics module -See pages 3-1 to 3-11 of Reference 01 for the steady state and transient equations.

b. Thermal-hydraulics module -See pages 2-35 to 2-40 of Reference 02 for the basic equations.

c. Fuel rod module -See pages 5-1 to 5-3 of Reference 01 for the heat transfer equation and its formulation.

d. Oehomogenization module-See pages 3-15 to 3-19 of Reference 01 for the equations.

e. S-RELAP5 -See pages 2-1 to 2-3 of Reference 03 for the two-fluid field equations and page 2-12 for the noncondensable gas and boron concentration in the liquid field equations.

f. The algorithm for coupling between ARTEMISŽ and S-RELAP5 is described in Appendix 0.7. System State Variables The coupled multi-physics system involves many time-dependent state variables.

These variables are used both directly and to derive additional data elements.

The combined set of time-dependent state variables for S-RELAP5 model is denoted as SR(uR, tR) where tR is the time of the S-RELAP5 solution and uR is the spatial location in the S-RELAP5 model. The set SR(uR, tR) consists of the variables listed above in item e.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary ANP-10338P Revision 0 Page D-3 The combined set of time-dependent state variables for the ARTEMISŽ nodal model is denoted as SA(uA, tA) where tA is the time of the ARTEMISŽ nodal solution and uA is the spatial location in the ARTEMISŽ nodal model. The set SA(uA, tA) consists of the variables listed above in items a, b, c and d. The combined set of time-dependent state variables for the ARTEMISŽ detailed model is denoted as SD(uD, tD) where tD is the time of the ARTEMISŽ detailed solution and uD is the spatial location in the ARTEMISŽ detailed model. The set SD(uD, tD) consists of the variables listed above in items b and c. D.1 ARTEMIS rM Nodal Transient Calculation The ARTEMISŽ nodal transient calculation begins with a steady state solution at the defined core conditions.

Then, the transient proceeds from one time step to the next as described in this section. The time steps are controlled by a time step manager as described in Appendix D.4. For the following description, it is assumed that the current transient solution has been completed for time tA. The time step A=O refers to the initial steady state solution.

The transient determines the solution at time tA+i using the following steps.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary The process is repeated for each time step until the end of the transient is reached. D.2 Detailed Model Transient Calculation ANP-10338P Revision 0 Page D-4 The calculations and data flow for the ARTEMISŽ thermal-hydraulic module and the fuel rod module of the detailed model are described below. The steps are the same for the detailed model and the nodal model calculations.

Note that the time steps of the detailed model are obtained from the nodal model, thus, tD = tA AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary ANP-10338P Revision 0 Page D-5

[ AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.3 S-RELAPS Calculation Proprietary ANP-10338P Revision 0 Page 0-6 J The S-RELAP5 equations and state variables were previously provided in the sections "System Equations" and "System State Variables" earlier in this Appendix.

An overview of the transient solution process is provided in Section 1.2.4 of Reference

03. D.4 ARTEMISŽ Time Step Manager The ARTEMISŽ time step control method (see Section 3.1 O of Reference

04) is based on [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.5 S-RELAP5 Time Step Manager Proprietary ANP-10338P Revision 0 Page D-7 The S-RELAP5 code contains an internal time step manager that provides a variety of checks on solution applicability to control the time step size. [ ] D.6 Parallel Application The solution of the ARTEMISŽ nodal model takes considerably less time than the solution of the detailed model. ARTEMISŽ has been developed for use with multiple threads (parallel calculation).

Typically, more threads are assigned to the detailed model calculation to allow it to complete over the same time frame as the nodal calculation.

This allows for faster completion of the coupled calculation.

D.7 S-RELAP5 and ARTEMISŽ Coupling The data exchange between S-RELAP5 and ARTEMISŽ is made through the External Controller.

This controller performs geometric translation of data and time step control between S-RELAP5 and ARTEMISŽ as needed. The time progression is shown in Figure D-2 and data flow is illustrated in Figure D-3. The automatic control of the calculation steps and their timing are described below.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary The process (steps 1 through 4) is repeated until the end of the transient is reached. ANP-10338P Revision 0 Page D-8 AREVA Inc. Proprietary AREAŽ -ARCADIA Rod Ejection Accident Topical Report D.8 Variable Definitions Variable C h p p Q'_coolant Q'_fuel Description precursor concentration enthalpy inlet flow pressure fuel rod power power deposited in the coolant power deposited in the fuel Q'_fuel surface heat flux transferred from the clad to the coolant SA ARTEMISŽ nodal model time-dependent state variables Sn ARTEMISŽ detailed model time-dependent state variables SR S-RELAP5 time-dependent state variables tA time in ARTEMISŽ nodal model tn time in ARTEMISŽ detailed model tR time in S-RELAP5 model Tc moderator temperature T clad clad temperature Terr effective fuel temperature Tr fuel temperature Tin inlet temperature T wan wall temperature u . internal energy uA spatial location in ARTEMISŽ nodal model un spatial location in ARTEMISŽ detailed model uR spatial location in S-RELAP5 model U axial velocity Ur liquid specific internal energy U 9 gas specific internal energy v 1 liquid velocity v 9 gas velocity Xn noncondensable quality ANP-10338P Revision 0 Page D-9 AREVA Inc. Proprietary AREA TM -ARCADIA Rod Ejection Accident Topical Report Og void fraction p density Ps boron density Pm moderator density L macroscopic cross section cp neutron flux cp surface heat flux ANP-10338P Revision 0 Page D-10 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.9 References Proprietary ANP-10338P Revision 0 Page D-11 D1 ANP-10297P-A Revision 0, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," February 2013. D2 ANP-10311 P-A Revision 0, "COBRA-FLX:

A Core Thermal-Hydraulic Analysis Code Topical Report," January 2013. D3 Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), "Document to Support the NRC review of EMF-2103P, Revision 3, 'Realistic Large Break LOCA Methodology for Pressurized Water Reactors'," NRC:14:001, January 10, 2014, contains a copy of EMF-2100(P), Revision 16, "S-RELAP5 Models and Correlations Code Manual," (Accession No. ML 14016A220), December 2011. D4 ANP-10297P-A Revision 0, Supplement 1, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," June 2015. D5 ANP-10323P Revision 0, "Fuel Rod Thermal-Mechanical Methodology for Boling Water Reactors and Pressurized Water Reactors," July 2013 D6 BAW-10231 P-A Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004. J AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary Figure D-1 Data Flow of System and Core-Coupled Calculations ANP-10338P Revision 0 Page D-12 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary Figure D-2 Overview of System-Core Coupled Calculation Time Steps ANP-10338P Revision 0 Page D-13 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Proprietary Figure D-3 Variables passed between S-RELAP5 and ARTEMISŽ ANP-10338P Revision 0 Page D-14 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Gary Peters, .Director Licensing and Regulatory Affairs AREVA Inc. 3315 Old Forest Road Lynchburg, VA 24501 July 31, 2017 NRC-IC-17-025 T4.12.1 Rec'd 8/11/17 HHE

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RE: AREVA INC. TOPICAL REPORT ANP-10338P, "AREA-ARCADIA ROD EJECTION ACCIDENT' (CAC NO. MF7009)

Dear Mr. Peters:

By letter dated October 9, 2015 (Agencywide Documents Access a.nd Management System (ADAMS) Accession No. ML 15300A298), AREVA INC. (AREVA) submitted for U.S. Nuclear Regulatory Commission (NRC) staff review and approval Topical Report (TR) ANP-10338P, "AREA -ARCADIA Rod Ejection Accident." Upon review of the information provided, the. NRC _ staff has determined that additional information is needed to complete the review; On July 14, 2017, Jerald Holm, AREVA Product Licensing Manager, and I agreed that the NRC staff will receive the response to the enclosed request for additional information (RAls) questions by August 15, 2017. If you have any questions regarding the enclosed RAI questions, please contact me at 3_01-415-4053.

Project No. 728

Enclosure:

RAI Questions

/.,.....-7 I <_,_,..,*'< , ./ Sincer~ly

-_ -------~---_,,. , Y1 -~ It/ Jonathan G. Rowley, Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

---

RAl2: REQUEST FOR ADDITIONAL INFORMATION RELATED TO TOPICAL REPORT ANP-10338P "AREA -ARCADIA ROD EJECTION ACCIDENT" AREVA INC. (CAC NO. MF7009) The methodology for the ARCADIA Rod Ejection Accident (AREAŽ) analysis code (ANP-10338P, Revision 0) is predicated on the U.S. Nuclear *Regulatory Commission (NRC) approved methodology of the ARCADIA code system (ANP-10297P and ANP-10297P-A, Revision 0, Supplement 1 ). However, in lieu of the fuel rod model (FRM) of the approved ARCADIA code system, AREA TM applies the NRC uni;lpproved fuel performance code GALILEO (ANP-10323P) to provide the thermal-mechanical properties of the fuel pins for analyses performed with AREA TM. In line with the Update Process set forth in Section 6.1 of ANP-10338P, Revision 0, for including codes without current NRC approval, this RAI requests the following information.

A Justify the motivation for using GALILEO derived thermal-mechanical properties of fuel pins rather than those of the NRC approved FRM in ARCADI.A.

B. List the capabilities that are introduced by the GALILEO models to ARCADIA that either do not exist or introduce better estimates with less uncertainty of the figures of merit for meeting regulatory reactivity-insertion accident (RIA) acceptance criteria.

• C. The FRM in the ARCADIA code solves the heat transfer equation given by Equation 5-1 in ANP-10297P, Revision 0. Is the FRM thermal .solver identical to the one in GALILEO? D. The thermal equation solver given by Equation 5-1, in order to compute the time-dependent spatial distribution of the local temperature, requires values for the spatially dependent local heat generation source term, local thermal conductivity, local specific heat, and local density. The description for the computation of the thermal models for these physical properties is given, in the case of GALILEO, iri Chapter 5 of GALILEO Fuel Rod Performance Code Theory Manual FS1-0004682, Revision 2.0. Comparable detailed descriptions do not exist in the ARCADIA topical report. Thus, in the context of the supplementary summary of the interfaces with GALILEO in the AREA TM methodology, submitted by AREVA to facilitate a teleconference with NRC on May 31, 2017 (shown below), the following information is requested.

Enclosure i. Indicate for the six usages in the context of the GALILEO models given in Chapter 5 of the GALILEO Theory Manual the differences, if any, with the models in FRM of ARCADIA for the lookup tables for the requisite spatially dependent physical constants or the information for the post processing of the ARTEMISŽ results. ii. Is the radially dependent source computed with the same codes and fitting procedures in FRM and GALILEO? iii. Furthermore, indicate which differences delineated above are the main contributors to the ARTEMISŽ -GALILEO differences shown in Figures 5-2 through 5-7 in the AREA TM topical report under review. interfaces with GALILEO The different uses of the fuel rod code in AREA Trvi to determine fuel pin performance relative to the RIA criteria are highlighted below by showing the use of GALILEOŽ in the sample problems.

There are six different types of usage of the data as listed below. Note that any of this information can be derived from any NRG approved fuel rod code and that information from GALILEOŽ for ARE;'.\ TM would only be used aft~r it~ approV?!I by the NRG.

• 1. The fuel rod thermal properties are equations included in the ARTEMISŽ Fuel Rod Module and are listed below. The ARTEMISŽ default equations are equivalent to the thermal property equations in GALILEOŽ.

a. For uranium oxide (U02), U02-Gd20a

[gadolinium oxide], zircaloy-4 (Zr4), and M5 clad i. Thermal conductivity ii. $pacific heat b. Fuel pellet radial power profile fit function 2. Data generated by GALILEOŽ computer ruris that are used as input (lookup tables) to ARTEMISŽ.

a. Gap Conductance

b. Porosity (used to calculate fuel thermal conductivity)

3. Data generated by GALILEOŽ are compared to ARTEMISŽ results to verify the adequacy of the gap conductance lookup table model in ARTEMISŽ.

a. Fuel centerline, average, and surface temperatures

b. Clad internal and average temperatures L_ 4. The GALILEOŽ equations that are used to post process the ARTEMISŽ results. a. Fuel melt temperature equation and its uncertainty for U02 and U02-Gd203

5. Data generated by GALILEOŽ that are used to post process the ARTEMISŽ results for comparison to limiting conditions for RIA criteria.

a. Maximum rim burnup versus average pellet burnup for rim melt temperature

b. Clad corrosion versus burnup to convert to enthalpy rise failure limits c. Max internal pressure and back fill pressure versus burnup for pressure related . requirements

d. Fission gas release versus burnup 6. The following data from GALILEOTM models are used for AREATM sensitivity studies to establish the need for biasing. a. Uncertainties for thermal conductivity and specific heat b. Oxide results c. Fuel expansion August 15, 2017 NRC:17:036 U.S. Nuciear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 A . AREVA Response to Second Request for Additional Information Regarding ANP-10338P, Revision 0, "AREA 1 M -ARCAD1A Rod Ejection Accident

Ref. 1: Letter, Pedro Salas (AREVA Inc.) to Document C::9ntrol Desk (NRC}, "Request for Review and Approval ofANP-10338P, 'AREAŽ -ARCADIA Rod Ejec:tion Accident'," NRC:15:038, October 9, 2015. Ref. 2: Letter, Jonathan G. Rowley (NRC} to Gary Peters {AREVA Inc.), "Request for Additional Information Re: AREVA Inc. Topical Repo.rtANP-10338P, 'AREAŽ-.ARCADIA Rod Ejection Accident' (CAC No. MF7009,)," July 31, 2017. AREVA Inc. (AREVA) requested the NRC's review and approval of the topical report ANP-10338P, "AREAŽ -ARCADIA Rod Ejection Accident" in Reference

1. The NRC provided a Request forAddltional Information (RA!) in Reference

2. The response to this RA! is enc:losed with this letter. There are no commitments within this fetter or its enclosl)res.

lfyou have any questions related to this information, please contact Ms. (;ayfe F. Elliott (Deputy Director, Licensing and Regulatory Affairs) by tele*phone at (434) 832-3347, or by e-mail at Gayle.Elfiott@areva.com.

_,Lr/ A/0 / Gar eters, Director 7 Licensing

& Regulatory Affai AREVA Inc. cc: J. G. Rowley Project728

Enclosures:

1. Non-Proprietary copy of ANP-10338Q2, Revision O, "Response to Reqt1est for Additional Information -ANP-10338P" AREVA INC. 3315 Old Forest Road, Lynchburg, VA 24501 Tel.: 434 832 3000

• www:areva.com A AREVA Response to Request for Additional Information -ANP-10338P Topical Report August 2017 AREVA Inc. (c) 2017 AREVA Inc. AN P-10338Q2 Revision 0 Copyright

© 2017 AREVA Inc. All Rights Reserved ANP-10338Q2 Revision 0 AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Item 1 Section(s) or Page(s) All Nature of Changes Description and Justification Initial Issue AN P-1033802 Revision 0 Page i AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Contents ANP-10338Q2 Revision 0 Page ii Page 1.0 RAI 2 ....................................................................................................................

1

2.0 REFERENCES

.....................................................................................................

9 AREVA Inc. ANP-1033802 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report List of Tables Table 1-1 ARCADIA and GALILEOŽ Thermal Properties

..........................................

6 Table 1-2 Important Differences between ARCADIA FRM and GALILEOŽ Models ..........................................................................................................

6 Table 1-3 Expected Result Differences Between ARCADIA FRM and GALILEO TM **************************************************************************************************

7 Page iii AREVA Inc. Response to Request for Additional Information -ANP-10338P Topical Report Acronym *FRM MDNBR RAI Nomenclature Definition Fuel Rod Model Minimum Departure from Nucleate Boiling Ratio Request for Additional Information AN P-10338Q2 Revision 0 Page iv AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Introduction The United States Nuclear Regulatory Commission (NRC) provided a request for ?dditional information (RAI) regarding the topical report ANP-10338 (Reference

1) in Reference

2. One question was received from the NRC. Pagev AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P

• Topical Report 1.0 RAI 2 Question:

The methodology for the ARCADIA Rod Ejection Accident (AREAŽ) analysis code (ANP-10338P, Revision 0) is predicated on the NRC approved methodology of the ARCADIA code system (ANP-10297P and ANP-10297P-A, Revision 0, Supplement 1). However, in lieu of the fuel rod model (FRM) of the approved ARCADIA code system, AREA TM applies the NRC unapproved fuel performance code GALILEO (ANP-10323P) to provide the thermal-mechanical properties of the fuel pins for analyses performed with AREAŽ. In line with the Update Process set forth in Section 6.1 of ANP-10338P, Revision 0, for including codes without current NRC approval, this RAI requests the following information.

A. Justify the motivation for using GALILEO derived thermal-mechanical properties of fuel pins rather than those of the NRC approved FRM in ARCADIA.

B. List the capabilities that are introduced by the GALILEO models to ARCADIA that either do not exist or introduce better estimates with less uncertainty of the figures of merit for meeting regulatory reactivity-insertion accident (RIA) acceptance criteria.

C. The FRM in the ARCADIA code solves the heat transfer equation given by Equation 5-1 in ANP-10297P, Revision 0. Is the FRM thermal solver identical to the one in GALILEO? D. The thermal equation solver given by Equation 5-1, in order to compute the time-dependent spatial distribution of the local temperature, requires values for the spatially dependent local heat generation source term, local thermal conductivity, local specific heat, and local density. The description for the computation of the thermal models for these physical properties is given, in the case of GALILEO, in Chapter 5 of GALILEO Fuel Rod Performance Code Theory Manual FS1-0004682, Revision 2.0. Comparable detailed descriptions do not exist in the ARCADIA topical report. Page 1 AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Thus, in the context of the supplementary summary of the interfaces with GALILEO in the AREA TM methodology, submitted by AREVA to facilitate a teleconference with NRG on May 31, ?017 (shown below), the following information is requested.

i. Indicate for the six usages in the context of the GALILEO models given in Chapter 5 of the GALILEO Theory Manual the differences, if any, with the models in FRM of ARCADIA for the lookup tables for the requisite spatially dependent physical constants or the information for the post processing of the ARTEMISŽ results. ii. Is the radially dependent source computed with the same codes and fitting procedures in FRM and GALILEO? iii. Furthermore, indicate which differences delineated above are the main contributors to the ARTEMISŽ-GALILEO differences shown in Figures 5-2 through 5-7 in the AREAŽ topical report under review. Interfaces with GALILEO The different uses of the fuel rod code in AREA TM to determine fuel pin performance relative to the RIA criteria are highlighted below by showing the use of GALILEOŽ in the sample problems.

There are six different types of usage of the data as listed below. Note that any of this information can be derived from any NRG approved fuel rod code and that information from GALILEOŽ for AREA TM would only be used after its approval by the NRG. 1. The fuel rod thermal properties are equations included in the ARTEMISŽ Fuel Rod Module and are listed below. The ARTEMISŽ default equations are equivalent to the thermal property equations in GALILEOŽ.

a. For uranium oxide (U02), U02-Gd203

[gadolinium oxide], zircaloy-4 (Zr4), and M5 clad i. Thermal conductivity ii. Specific heat Page 2 AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report b. Fuel pellet radial power profile fit function 2. Data generated by GALILEOŽ computer runs that are used as input (lookup tables) to ARTEMISŽ.

a. Gap Conductance

b. Porosity (used to calculate fuel thermal conductivity)

3. Data generated by GALILEOŽ are compared to ARTEMISŽ results to verify the adequacy of the gap conductance lookup table model in ARTEMISŽ.

a. Fuel centerline, average, and surface temperatures

b. Clad internal and average temperatures

4. The GALILEOŽ equations that are used to post process the ARTEMISŽ results. a. Fuel melt temperature equation and its uncertainty for U02 and U02-Gd203

5. Data generated by GALILEOŽ that are used to post process the ARTEMISŽ results for comparison to lim'iting conditions for RIA criteria.

a. Maximum rim burnu'p versus average pellet burnup for rim melt temperature

b. Clad corrosion versus burnup to convert to enthalpy rise failure limits c. Max internal pressure and back fill pressure versus burnup for pressure related requirements

d. Fission gas release versus burnup 6. The following data from GALILEOŽ models are used for AREA TM sensitivity studies to establish the need for biasing. a. Uncertainties for thermal conductivity and specific heat b. Oxide results c. Fuel expansion Page 3 AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Response 2.A: The thermal properties in the approved Fuel Rod Model in ARCADIA are not changed for AREAŽ. The FRM model in ARCADIA obtained its thermal properties from a development version of GALILEOŽ.

Hence, ARCADIA and GALILEOŽ have the same thermal properties which are listed in Table 1-1. Therefore, the motivation for AREA TM is to remain consistent with both ARCADIA and GALILEOŽ and no further justification will be needed for this when GALILEOŽ is approved.

If another fuel performance code is used, justification would be developed for (1) using its thermal properties in ARCADIA or (2) a justification would be developed for not changing the thermal properties in ARCADIA (that is using the GALILEOŽ properties that are in ARCADIA currently).

It is anticipated that the use of the second option will be feasible.

Response 2.8: There are no changes to the ARCADIA model due to the use of GALILEOŽ in AREA TM. In order to show compliance with the RIA criteria for control rod ejection accidents, AREA TM requires a model to calculate the thermal conditions of specific fuel pins and ARCADIA has only a nodal model (4 radial nodes per assembly).

For example, the thermal conductivity for a fuel pin containing various amounts of Gadolinia is different than U0 2 which affects the calculation of the fuel temperature.

The capabilities of GALILEOŽ that are utilized in AREA TM to calculate the pin specific parameters that are not available from ARCADIA are:

• Calculate clad corrosion to determine enthalpy rise limits ** Pin internal pressures for fuel cladding failure criteria due to high temperature

• Pin internal pressures used to address potential coolability concerns

• Fuel melt temperature limit

• Rim burnup as a function of pellet burnup used to preclude rim melting Page 4 AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report

• Estimated uncertainties of the fuel properties for sensitivity analysis

• Pellet specific gap conductance and porosity tables are used in the detailed model to calculate specific pin properties (i.e. Gadolinia pins) Response 2.C: The FRM thermal solver is not identical to the solver in GALILEOŽ.

The FRM thermal solver is described in Section 5 of Reference

3. The key differences between the GALILEOŽ and ARCADIA FRM are shown in Table 1-2. Response 2.D: Table 1-3 describes the differences expected from using the ARCADIA FRM model for each of the listed applications for AREA TM. As noted in the response to RAl.2A, the FRM in ARTEMISŽ uses the same radial power profiles as GALILEOŽ.

The FRM thermal calculations in ARTEMISŽ using static gap conductance tables compared to those in GALILEOŽ which use dynamic gap conductance is the primary cause of the differences of the results in these figures. Page 5 AREVA Inc. AN P-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Table 1-1 ARCADIA and GALILEOŽ Thermal Properties Thermal Parameter ARCADIA (Reference

3) GALILEOŽ (Reference

4) U0 2 thermal conductivity Same source as GALILEOŽ Section 5.6.2.1 in Reference 4 U0 2 -Gd 2 0 3 thermal conductivity Same source as GALILEOŽ Section 5.6.2.4 in Reference 4 Zr4 thermal conductivity Same source as GALILEOŽ Section 9.5.2 in Reference 4 M5 thermal conductivity Same source as GALILEOŽ Section 9.5.3 in Reference 4 Fuel specific heat Same source as GALILEOŽ Section 9.6.4 in Reference 4 Zr4 specific heat Same source as GALILEOŽ Section 9.6.2 in Reference 4 M5 specific heat Same source as GALILEOŽ Section 9.6.3 in Reference 4 Fuel pellet radial power profile Same source as GALILEOŽ Section 5.2 in Reference 4 Table 1-2 Important Differences between ARCADIA FRM and GALILEOŽ Models Phenomena ARCADIA GALILEOŽ Gap A contact resistance formulation is Detailed thermal expansion, Conductance used for gap conductance that does mechanical deformation, and fission not require a gap thickness for its gas release models are used to solution and uses a fitting table for calculate the pellet-clad gap, contact the gap conductance.

pressure, and gap gas content to calculate the gap conductance.

Coolant The thermal hydraulic module is a A simple static energy balance using Properties time dependent, 1 D energy and a closed channel thermal model is momentum balance with cross flow used for the coolant thermal and calculates the thermal properties of the pin. (This coolant properties of the coolant for the model is not capable of calculating node and pin coolant channels.

prompt critical transients.)

See section 4.0 of Reference

3. Corrosion No corrosion is used in the thermal A detailed corrosion and crud model Model solution of the FRM for the sample is needed to determine the corrosion problems.

A sensitivity study on based RIA criteria.

clad conductivity (identified in Table 4-3 of Reference 1 is performed to assess no corrosion as a part of the Section 7 sensitivity evaluation in Reference

1. Page 6 AREVA Inc. AN P-10338Q2 .Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Page 7 Table 1-3 Expected Result Differences Between ARCADIA FRM and GALILEOŽ Usage Expected differences

1. Fuel Thermal Properties Since the same source for this information is used in the ARCADIA FRM and GALILEOTM, no significant differences are expected.

-2a. Gap Conductance Tables GALI LEO TM is used to calculate these tables. The expected temperature errors are estimated and assessed in Section . 5.2.2 of Reference

1. 2b. Porosity The porosity is a very slow changing correction with burnup. The fitting table versus burnup obtains the primary radial variation of the porosity with burnup. No significant differences are expected.

3. Benchmarks This use of GALILEOŽ assesses the differences primarily of 2a but would also include all solution differences.

The outer clad temperature is fixed so that the coolant model is not a source of error in these benchmarks. (Section 5.2.2 of Reference

4) Also, both GALILEOŽ and ARCADIA FRM use no corrosion for these benchmarks.

4. Melt temperature Melt temperatures are defined in GALILEOŽ as a function of burnup. Maximum and rim temperature predictions by the FRM are assessed in Section 5.2.2 of Reference

1. 5a. Rim Burnup Rim burnups are conservatively estimated with GALILEOŽ runs to determine the melt temperature of the rim. The temperature prediction of the FRM is compared to this limit. 5b. Clad corrosion GALILEOŽ is used to estimate the amount of corrosion that occurs with burnup. Sensitivity calculations to identify the key parameters in Chapter 7 of Reference 1 determined that ignoring clad corrosion had an insignificant impact on the dependent variables (MDNBR, maximum fuel temperature, , total enthalpy, rim temperature, enthalpy rise, and integrated power to the coolant).

5c. and d. Internal Pressure and GALILEOŽ is used to conservatively estimate these variables

' using a conservative pin power history as described in Section fission gas release 6.2.2 of Reference

1. These are long term steady state effects and there is no direct relationship to the FRM. _,

AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report Page 8 6. Sensitivity Studies These studies are performed using the ARCADIA nodal and detailed models to assess the importance of ignoring or biasing different parameters.

Estimated uncertainties of the thermal properties from GALILEO TM are used to validate the biasing model of AREA TM. GALILEOŽ does not have a gap conductance model uncertainty; rather it is included in the overall prediction of temperature uncertainty.

GALILEOŽ estimates the effect of manufacturing tolerances on the gap conductance through direct simulations.

The gap conductance model in the ARCADIA FRM is also tested with sensitivity studies using a multiplier of 2.0 and 0.5 on the calculated value and also investigates manufacturing tolerances.

L AREVA Inc. ANP-10338Q2 Revision 0 Response to Request for Additional Information -ANP-10338P Topical Report

2.0 REFERENCES

1. ANP-10338P Revision 0, "AREAŽ -ARCADIA Rod Ejection Accident," October 2015. 2. Letter, Jonathan G. Rowley (NRC) to Gary Peters (AREVA Inc.), "Request for Additional Information Re: AREVA Inc. Topical Report ANP-10338P, 'AREA-ARCADIA Rod Ejection Accident'," (CAC No. MF7009), July 31, 2017. 3. ANP-10297P-A, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results," February 2013. 4. FS1-0004682-2.0, "GALILEO Fuel Rod Performance Code Theory Manual," May 2013. Provided to the NRC in Reference

5. 5. Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), "Documents to Support the NRC Review of ANP-10337P, 'Fuel Rod Thermal Mechanical Methodology for Boiling Water Reactors and Pressurized Water Reactors'," NRC:13:073, September 17, 2013. Page 9 October 5, 2017 NRC:17:041 U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 A AREVA Additional Information Regarding ANP-10338P, Revision O, "AREAŽ -ARCAD1A 0 Rod Ejection Accident" Ref. 1: Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), "Request for Review and Approval of ANP-10338P, 'AREAŽ -ARCADIA 0 Rod Ejection Accident'," NRC:15:038, October 9, 2015. AREVA Inc. (AREVA) requested the NRC's review and approval of the topical report ANP-10338P, "AREAŽ -ARCADIA Rod Ejection Accident" in Reference

1. In September 2017, AREVA notified the NRC by telephone of an error affecting the topical report. The enclosures to this letter contain a description of the error, its impact, and correc:tions to the proprietary and non-proprietary versions of the topical report. These corrections will be included when the approved version of the topical report is published.

There are no regulatory commitments within this letter or its enclosures.

AREVA considers some of the information contained in the enclosed documents to be proprietary.

As required by 10 CFR 2.390(b) an affidavit is enclosed to support the withholding of the information from public disclosure.

Proprietary and non-proprietary versions of the report are enclosed.

If you have any questions related to this information, please contact Ms. Gayle Elliott, Product Licensing Manager, by telephone at (434) 832-3347, or by e-mail at Gayle.Elliott@areva.com.

cc: J. G. Rowley Project 728 AREVAJNC.

3315 Old Forest Road, Lynchburg, VA 24501 Tel.: 434 832 3000

• www.areva.com Document Control Desk October 5, 2017 Enclo.sures:

NRC:17:041 Page 2 1. ANP-10338Q3P, "Additional Information -ANP-10338P, 'AREAŽ--'ARCADIA~

Rod Ejection Atcident 111 2. ANP-10338Q3NP, "Additional information ANP-10338P, 'AREAŽ -ARCADIA 0 Rod Ejection Accident 111 3. Notc1rized Affidavit AFFIDAVIT COMMONWEAL TH OF VIRGINIA ) ) ss. CITY OF LYNCHBURG ) 1. My name is Nathan E. Hottle. I am Manager, Product Licensing, for AREVA Inc. (AREVA) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA to determine whether certain AREVA information is proprietary.

lam familiar with the policies established by AREVA to ensure the proper application of these criteria.

3. I am familiar with the AREVA information contained in the following document:

ANP-10338Q3P Revision 0, "Additional Information -ANP-10338P, AREAŽ -ARCADIA Rod Ejection Accident," referred to herein as "Document." Information contained in this Document has been classified by AREVA as proprietary in accordance with the policies established by AREVA Inc. for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 1 O CFR 2.390. The information for which withholding from disclosure is requested qualifies under 1 O CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA, would be helpful to competitors to AREVA, and would likely cause substantial harm to the competitive position of AREVA. The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d), and 6(e) above. 7. In accordance with AREVA's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this "4ih ---'---day of __ o_~~--'-=------'-----'

2017. Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg.# 7079129 SHERRY L,' MCFADEN Notary Public Commonwilltll or Virginia 7079129 My Comm1111on E11plr11 Oct 31, 2018 A AREVA Additional Information -ANP-10338NP AREA TM -ARCADIA Rod Ejection Accident Topical Report October 2017 AREVA Inc. (c) 2017 AREVA Inc. ANP-10338Q3NP Revision O*

Copyright

© 2017 AREVA Inc. All Rights Reserved ANP-10338Q3NP Revision 0

• -----------------------------------------

AREVA Inc. Additional Information -ANP-10338NP AREA TM -ARCADIA Rod Ejection Accident Topical Report

SUMMARY

ANP-10338Q3NP Revision O Page 1 In September, AREVA notified the NRG by telephone of an issue impacting Topical Report ANP-10338NP, "AREA -ARCADIA Rod Ejection Accident." The issue was an inconsistency between the definition of nominal given in the text and the initial conditions for the sample problem results. The I ] as defined on page 7-2. Additionally, for the B&W plant in the Appendix B sample problems, the nominal cases from low power were incorrectly set to full power inlet conditions.

A clarification was added to tables A-7, B-7, and C-7 to indicate other conservatisms that are utilized in the methodology but are not reflected in the values of the table. Updates are provided for all the changes noted above in the Topical Report. For the tables containing the "Measure of Conservatism for Limiting Result Cases" (Tables A-7, B-7, and C-7), the I ]. In general, there is a decrease in the measure of conservatism.

There were two changes for the B&W plant initial conditions and I ] on the limiting parameter is countered by the inlet temperature change. Therefore, the direction of the change to the nominal results depends upon which error condition dominates.

The nominal pressure calculations are not significantly impacted.

The limiting conditions of the methodology are not affected by these changes and the overall conclusions of the Topical Report remain valid.

AREVA Inc. Additional Information -ANP-10338NP AREA TM -ARCADIA Rod Ejection Accident Topical Report MARKUP PAGES

• Page 6-26

• Page A-8

• Page A-17

• Page A-20

• Page 8-4

• Page 8-14

• Page 8-16 .. Page 8-28

• Page C-10 ANP-10338Q3NP Revision O Page 2 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-26 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 TM 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 TM. If future regulatory commitments are made relative to the. approved codes supporting AREA TM, the changes affecting AREA TM 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: . [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PaqeA-8 The nominal and biased cases reach similar peak RCS pressures with the difference in peak pressures of less than [ ] l AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-17 Table A-7 W 4-Loop, Measure of Conservatism for Limiting Result Cases AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-20 Table A-10 W 4-Loop Plant Overpressure Results Summary (High Pressurizer Pressure Trip Modeled)

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report [ ] These two cases reach similar peak RCS pressures.

[ ] [ ] ANP-10338NP Revision 0 Page B-4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision O Page B-14 Table B-7 Measure of Conservatism for Each of the Limiting Cases AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-16 Table 8-9 B&W plant Overpressure Results Summary (no high pressure trip modeled)

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure B-10 Hot Leg Pressure Comparison ANP-10338NP Revision 0 Page B-28 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-10 Table C-7 CE Plant Measure of Level of Conservatism for Each Limiting Parameter A AREVA. AREA rM -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 ANP-10338NP Revision 0 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Item 1 Section(s) or Page(s) All Nature of Changes Description and Justification Initial Issue ANP-10338NP Revision 0 Page i AREVA Inc. ANP-10338NP Revision 0 AREA TM -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 GALI LEO TM *.************......**.*.............*.*.**.*.*.*.***.***.*.***.*..*..*..**.**.*..*.********.

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-FLXŽ

........................................................................................

5-8 5.3.1 COBRA-FLXŽ 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 TM -ARCADIA Rod Ejection Accident Topical Report Page iii 5.4.2 RELAP5/MOD2-B&W Code and Model .....................................

5-12 6.0 AREA TM METHODOLOGY DESCRIPTION

......................................................

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

6-1 6.2 GALILEO TM ..*......*....*..*.*..*.*.*.*.**.*..*.*.*..*.*.*.*....*...*.***.************..**.*.*..*.**.*.

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 ARTEMISŽ Models for REA EventAnalysis

..........................................

6-5 6.4 ARTEMISŽ (Steady State Nodal Solution)

............. , ..............................

6-6 6.5 ARTEMISŽ (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-FLXŽ 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 AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page iv 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 TM PLANT SPECFIC APPLICATION

.......................................................

8-1 8.1 Initial Application of AREA TM Methodology

.............................................

8-1 8.2 Cycle to Cycle Evaluation

.......................................................................

8-2 9.0 SAMPLE PROBLEMS .......................................................................................

9-1

10.0 CONCLUSION

S

..............................................................................................

10-1

11.0 REFERENCES

................................................................................................

11-1 APPENDIX A W4-LOOP 193 FA PLANT WITH 17X17 FUEL LATTICE .....................

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

8-1 APPENDIX C CE 217 FA PLANT WITH 14X14 FUEL LATTICE ................................

C-1 APPENDIX D ARTEMISŽ AND S-RELAP5 COUPLING DESCRIPTION

..................

D-1 AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Pagev 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 TM Methodology

..........................

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

4-4 Table 5-1 ARCADIA Validation Test Matrix for AREAŽ ..........................................

5-13 Table 5-2 GALILEOŽ/

ARTEMISŽ Transient Comparisons for U02 Fuel ...............

5-14 Table 5-3 GALILEOŽ/

ARTEMISŽ Transient Comparisons for 2 wt% Gadolinia Fuel .......................................................................................................

5-15 Table 5-4 GALILEOŽ/

ARTEMISŽ Transient Comparisons for 8 wt% Gadolinia Fuel .......................................................................................................

5-16 Table 5-5 COBRA-FLXŽ Validation Test Matrix for AREAŽ ...................................

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-3 W 4-Loop Limiting Results Summary for Burnup 2 ...................................

A-13 Table A-4 W 4-Loop Limiting Results Summary for Burnup 3 ...................................

A-14 Table A-5 W 4-Loop Limiting Results Summary for Burnup 4 ...................................

A-15 Table A-6 W 4-Loop Limiting Results Summary for Burnup 5 ...................................

A-16 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 .......................................................................

8-8 Table B-2 B&W Plant Limiting Results Summary for Burn up 1 ...................................

B-9 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page vi Table B-3 B&W Plant Limiting Results Summary for Burnup 2 .................................

B-10 Table B-4 B&W Plant Limiting Results Summary for Burnup 3 .................................

B-11 Table B-5 B&W Plant Limiting Results Summary for Burnup 4 .................................

B-12 Table B-6 B&W Plant Limiting Results Summary for Burnup 5 .................................

B-13 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-3 CE Plant Limiting Results Summary for Burnup 2 .....................................

C-6 Table C-4 CE Plant Limiting Results Summary for Burnup 3 .....................................

C-7 Table C-5 CE Plant Limiting Results Summary for Burnup 4 .....................................

C-8 Table C-6 CE Plant Limiting Results Summary for Burnup 5 .....................................

C-9 Table C-7 CE Plant Measure of Level of Conservatism for Each Limiting Parameter

............................................................................................

C-10 AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report List of Figures Page vii 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 U0 2 HZP EOL Transient Fuel Centerline Temperature

...........................

5-19 Figure 5-3 U02 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 MS 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 U02 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 U02 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 U02 and Gadolinia Fuel with M5 Clad ..... .' .. A-29 AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page viii Figure A-8 W 4-Loop Limiting FGR for U02 and Gadolinia Fuel with Zr4 Clad .........

A-30 Figure A-9 W 4-Loop General Depressurization Curve .............................................

A-31 Figure A-10 Transient Fa, Ft.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 Fa, F LlH, 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 Fa, F LlH, 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 Fa, F LlH, 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 U02 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 U0 2 and Gadolinia Fuel with M5 Clad ........................

B-22 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 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 ......................................................

8-25 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 U02 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 U02 and Gadolinia Fuel with M5 Clad .......................

C-14 Figure C-5 CE Plant General Depressurization Curve .............................................

C-15 Figure D-1 Data Flow of System and Core-Coupled Calculations

............................

D-12 Figure D-2 Overview of System-Core Coupled Calculation Time Steps ...................

D-13 Figure D-3 Variables passed between S-RELAP5 and ARTEMISŽ ........................

D-14 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report Acronym AFD AREA TM AO ASI ASME B&W BOC BOL BWR CE CEA CFR CHF COLR CPR CROM DPC OTC EFPD EOC EOL ERW FGR FP GDC GWd/MTU HFP HZP LOCA M(DNB)R MOL MPI Definition Nomenclature (If applicable)

Axial Flux Difference ARCADIA Rod Ejection Accident Axial Offset Axial Shape Index American Society of Mechanical Engineers Babcock & Wilcox Beginning of Cycle Beginning of Life Boiling Water Reactor Combustion Engineering Control Element Assembly Code of Federal Regulations Critical Heat Flux Core Operating Limits Report Critical Power Ratio Control Rod Drive Mechanism Doppler Power Coefficient Doppler Temperature Coefficient Effective Full Power Days End of Cycle End of Life Ejected Rod Worth Fission Gas Release Full Power General Design Criteria Gigawatt days per Metric Tonne Uranium Hot Full Power Hot Zero Power Loss of Coolant Accident Minimum (Departure from Nucleate Boiling) Ratio Middle of Life Message Passing Interface ANP-10338NP Revision 0 Page x AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Acronym MTC MWd/mtU NRC PC PCMI POil PIRT PSV PWR RCCA RCS REA RG RIA RPS RXA SAFDL SRA SRP U02 w wt% Zr4 Definition Moderator Temperature Coefficient Megawatt days per Metric Tonne Uranium Nuclear Regulatory Commission Power Coefficient Pellet Cladding Mechanical Interaction Power Dependent Insertion Limit Phenomena Identification Ranking Table Pressure Safety Valve Pressurized Water Reactor Rod Control Cluster Assembly Reactor Coolant System Rod Ejection Accident Regulatory Guide Reactivity Initiated Accident Reactor Protection System Recrystallized Annealed Specified Acceptable Fuel Design Limit Stress Relief Annealed Standard Review Plan Uranium Dioxide Westinghouse Weight Percent Zircaloy 4 Alloy ANP-10338NP Revision 0 Page xi AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report ABSTRACT Page xii This report presents the AREA TM 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 TM 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 GALILEOŽ 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. AREA TM -ARCADIA Rod Ejection Accident Topical Report 1 :o INTRODUCTION ANP-10338NP Revision O Page 1-1 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 (CROM) 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 (AREAŽ) 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 1. 1 Range of Applicability ANP-10338NP Revision 0 Page 1-2 The AREA TM 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 FLXŽ 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 TM methodology descriptions.

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

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 2.0 APPLICABLE REA REGULATORY REQUIREMENTS ANP-10338NP Revision 0 Page 2-1 This AREA TM 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 1 OCFR50 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. 1 OCFR50 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 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 2-2 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 1 OCFR50 Appendix A requirements, the other regulatory requirements pertaining to the REA event are 10CFR100.11 and 1 OCFR50.67.

Both of these requirements refer to radiological consequences of an REA event. These two requirements are not directly addressed by the AREA TM 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 AREAŽ 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 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 2-3 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 ca/lg for fuel rods with an internal rod pressure at or below system pressure, and 150 ca/lg 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. DN8R and CPR). 2. The PCM/ failure criteria is a change in radial average fuel enthalpy greater than the corrosion-dependent limit depicted in Figure 8-1 (PWR) {Figure 2-1 in this topical report} and Figure 8-2 (8WR). Fuel cladding failure may occur almost instantaneously during the prompt fuel enthalpy rise (due to PCM/) or may occur as total fuel enthalpy (prompt+ delayed), heat flux, and cladding temperature increase.

For the purpose of calculating fuel enthalpy for assessing PCM/ 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 COOLA8/LITY 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 NRG 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 ca/lg.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 2-4 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 Joss of coo/able 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*L1H)

-7.1419] Where: FGR = Fission gas release, % (must be ;:;: OJ L1H = Increase in fuel enthalpy, L1callg AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 2-5 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 ca/lg o Cladding differential pressure>

1.0 MPa, < 4.5 Mpa Peak radial average fuel enthalpy = 170 -((!JP -1. OJ *20) ca/lg o Cladding differential pressure > 4. 5 MPa, Peak radial average fuel enthalpy = 100 ca/lg Predicted cladding differential pressure must consider the impact of transient FGR on internal gas pressure.

An acceptable means <:Jf determining the amount of transient FGR is described in Section 3.5 of this report.* ... (DNBR remains the same) ...

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page 2-6 2. The PWR PCM/ 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 specific information accounting for manufacturing tolerances and modeling uncertainties using NRG 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 coo/ability 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

• f),,H) -13] Peak Pellet BU> 50 GWd/MTU: Transient FGR (%) = [(0.26

• f),,H) -5] Where: FGR = Fission gas release, % (must be > OJ f),,H = Fuel enthalpy increase (f),,ca//g)

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 2-7 These transient FGR correlations supersede the correlation derived in Reference 4 and presented in DG-1199. 2.3 Future Criteria The AREA TM methodology is flexible and capable of demonstrating compliance with potential revisions to the REA criteria in Reference

1. The codes that support the AREAŽ methodology (ARTEMISŽ, GALILEOŽ, and RELAPS) 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 TM 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 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report

• Account for fluid transport in the RCS

• Heat transfer to the steam generators

• Credit action of the pressurizer relief and safety valves ANP-10338NP Revision 0 Page 2-8

• No credit for pressure reduction caused by the failure of a CROM pressure housing AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 2-9 Figure 2-1 Corrosion Limit Based on Relative Oxide Thickness FIGURE B-1: PWR PCMI Fuel Cladding Failure Criteria 175 (0.04, 150) -150+-----... iti u -125 51 ii! 100 iti :S 75 ci ::, IL 50 25 Cladding Failure (0.08, 75) (0.20, 0) 0-----------~----~----~-------1 0 0.04 0.08 0.12 0.16 0.2 Oxide/Wall Thickness

• This figure is extracted from Reference

1.

AREVA I nc. AREA T M -A R C ADIA Rod E j ection Ac c i d ent Topica l Report ANP-10338NP Revision 0 Page 2-10 Figure 2-2 Corrosion Limit Based on RXA Clad Type and Excess Hydrogen 200 175 --*--(0, 150) fll 150 QI "' ii: > E-125 fll .c .... C LU ] 100 ... QI i ... 75 cu :> <t "iii '5 fll 50 a: .:.: fll QI Q. 25 0 0 50 Revised PCMI Cladd i ng Failure Threshold RXA Cladding at PWR Operating Conditions I +-I -r (75 , 150) i -, c ,aaa mg li a1 u r e -i--I 1 (130, 95) *'i I C l adding Intact--:+ 1 0 0 150 200 Excess Cladding Hydrogen (wppm) *T his fig ure i s ex t racte d from Fi g u re 3.2.2-21 in R efe r e nc e 2. ---------(300, 70) ---. 250 300 **Th e fa i l ure v a l ue o f 9 5 c a l/g a t 130 wppm i s u se d an d t h e p l o tte d valu e i s i g nor e d for the sam p le probl e m s.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Top i cal Report ANP-10338NP Revision 0 Page 2-11 Figure 2-3 Corrosion Limit Based on SRA Clad Type and Excess Hydrogen 200 175 ni 150 5: ii: 125 ni .c ... C UI ] 100 u.. cu 75 ni 'ii m a: 50 ii cu Q. 25 0 (0, 150) 0 I ' ' I I I I I Revised PCMI Cladding Failure Threshold SRA Cladding at PWR Operating Conditions I I I I I I I I (165, 1sd) I C l ac dingfa i lur k I " I I ' (l\Hfail=

I 424 -53.Slf [Hex]) I :~ t ladding In t act I I I I I I I I I I I I : I 100 200 300 40 0 50 0 60 0 Excess Cladding Hydrogen (wppm) *Th is f igur e i s e xtracte d fr om Fi gure 3.2.2-22 i n R eferen c e 2. -----700 800 ----____ _______j AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 3.0 ROD EJECTION ACCIDENT SCENARIO IDENTIFICATION ANP-10338NP Revision 0 Page 3-1 The REA is postulated to occur from a mechanical failure of the CROM 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 (p) inserted is greater than the effective delayed neutron fraction (~) or prompt critical ,o Reactivity inserted less than the or sub-prompt critical..

Technical Specification limits for PWRs define the allowed contro'I rod positions with respect to power level which are referred to in this method as Power Dependent Insertion Limits (POils). The POils 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 3.1.1 Prompt Critical ANP-10338NP Revision O Page 3-2 For p>~, 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: where: 2 * (p -P)

• Cp av Ed -Energy deposited p -step reactivity insertion (ejected rod worth) -effective delayed neutron fraction Cv -heat capacity of the fuel av -Doppler Temperature Coefficient (OTC) For this condition, the power increase is fast (peak powers can be reached in terms of milliseconds) where only av 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 (p), the effective delayed neutron fraction (~), the heat capacity (Cp), and the OTC (av)-

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 3-3 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 p<~, 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. P/P o = / (~ -p) where: Pi -prompt-jump power Po -initial power -effective delayed neutron fraction p -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 OTC is balanced with p. 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 heatfluxes 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 asp approaches and exceeds there is a smooth transition as the prompt rise turns into a pulse.

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report 3.2 RCS Pressure Page 3-4 The control rod ejection accident is postulated in RG 1.77 (Reference

17) to occur due to "a mechanical failure of the CROM 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 ONB 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 CROM housing can result in a breach of the reactor coolant v~ssel 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 TM methodology uses this first scenario to evaluate the maximum RCS pressure.

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

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 3-5 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 TM methodology is addressed in Section 6.7.2.

AREVA Inc. ANP-10338NP Revision 0 AREA T M -ARCADIA Rod Eject i on Accident Topi c al Report Figure 3-1 Prompt Critical Power Excursion

-Prompt --H FP -~---------~------

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

I --~--------------*

-s u b-prompt --HF P Time, Sec Page 3-6 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision O Page 4-1 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 TM 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 4.2 DNBR ANP-10338NP Revision 0 Page 4-2 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 TM 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Table 4-1 PIRT Plant Transient Analysis Subcategory Phenomenon Calculation of Ejected control rod worth power history Rate of reactivity insertion during pulse Moderator feedback (includes pulse Fuel temperature feedback width) 13 Reactor trip reactivity Fuel cycle design Calculation of rod Heat resistances in high burnup fuel, gap, fuel enthalpy and clad (including oxide layer) increase during Transient clad-to-coolant heat transfer pulse (includes coefficient clad temperature)

Heat capacities of fuel and clad Fractional energy deposition in pellet Pellet radial power distribution Rod peaking factors Notes:

• Importance ratio IR>75 important

• • Knowledge ratio KR<75 not completely understood IR* 100 61 38 100 95 0 92 58 56 94 4 63 97 ANP-10338NP Revision 0 Page 4-3 KR** 100 88 93 96 96 96 100 67 64 90 93 88 100 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 Heat resistances in fuel, gap, and clad 75 77 temperature Transient clad-to-coolant heat transfer 50 58 changes coefficient (oxidized clad) 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 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 4-4 Table 4-3 Parameters Directly Addressed by AREA TM Methodology Neutronic Thermal Other (Neutro111ic and Detailed Model) Ejected rod worth Fuel conductivity Computational accuracies*

13 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 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 5.0 ANALYTICAL MODELS ANP-10338NP Revision 0 Page 5-1 The AREAŽ 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: 11 GALI LEOŽ (Reference

7) {COPERN IC (Reference

16) can also be used if the outlined validations are performed}

11 ARTEMISŽ ( References 11 and 12), a coupled 3-D kinetics solution with neutronics, fuel rod thermal model, and 3-D thermal hydraulic model

• COBRA-FLXŽ (Reference

13) as the 3-D thermal hydraulic model implemented in Reference 12 o 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 ARTEMISŽ 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. . [ l l_ AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report [ ] ANP-10338NP Revision 0 Page 5-2 _J I AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report . [ 5.1 GALILEOŽ ] ANP-10338NP Revision 0 Page 5-3 GALILEOŽ is the fuel performance code that provides the following information pertinent to the AREA TM methodology:

o Enthalpy rise criteria functionalized by clad corrosion is converted to enthalpy rise limits versus burnup o Fuel thermal properties with burnup dependencies for the time dependent solutions of temperature 11 Fuel pin internal pressure to determine fuel enthalpy limits for high clad temperature failure criteria COPERNIC can also provide this information.

For the AREA TM topical report, whenever GALILEOŽ is used, COPERNIC can also be used with differences noted.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 5.1.1 Enthalpy Rise Limits ANP-10338NP Revision 0 Page 5-4 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 GALILEOŽ 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 GALILEOŽ 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 ARTEMISŽ.

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 ARTEMISŽ 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 [ ] ]. . I I AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report 5.1.3 Fuel Pin Pressure ANP-10338NP Revision 0 Page 5-5 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 ARCADIAcode 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 TM. It is capable of calculating all neutronics and thermal I effects discussed in Section 4.0 that are needed to demonstrate complian~e 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 AREAŽ. The thermal-hydraulic model in ARCADIA is COBRA-FLXŽ (Reference

13) as described in Section 5.3.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-6 5.2.2 Verification of Gap Conductance and Thermal Conductivity Models Comparisons between GALILEOŽ and the ARTEMISŽ 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 ARTEMISŽ fuel thermal model and a more detailed treatment of the fuel rod thermal properties in GALILEOŽ.

[ ] [ ]

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-7 Summary statistics are shown in Table 5-2, Table 5-3, and Table 5-4 for U02 fuel, 2 wt% gadolinia fuel, and 8 wt% gadolinia fuel, respectively.

The average, minimum, and maximum ratios of ARTEMISŽ to GALILEOŽ 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 U02 fuel at EOL and HZP both the centerline and the surface fuel temperatures have minimum ratios of [ ] respectively.

For the HZP transfent 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 ARTEMISŽ 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.

______ _J AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-8 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.

[ ] ARTEMISŽ 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-FLXŽ The COBRA-FLXŽ core thermal-hydraulic code is AREVA's latest development for performing nuclear core thermal-hydraulic simulations.

COBRA-FLXŽ is the hydraulic code module used in the core simulator ARTEMISŽ.

COBRA-FLXŽ is incorporated into the ARTEMISŽ code in its entirety.

Within ARCADIA, COBRA-FLXŽ can be used as part of ARTEMISŽ or stand-alone.

The AREA TM methodology uses COBRA-FLXŽ through ARTEMISŽ.

COBRA-FLXŽ is used for both the nodal simulator and the detailed model. The detailed model is a [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 5.3.1 COBRA-FLXŽ Validation ANP-10338NP Revision 0 Page 5-9 Validation of COBRA-FLXŽ is provided in Reference

13. COBRA-FLXŽ is an integral part of the ARTEMISŽ moderator feedback solution in References 11 and 12. Table 5-5 contains the COBRA-FLXŽ validation matrix specific to AREA TM. 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 TM 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 CROM 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. _J AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report 5.4.1 S-RELAP5 Code and Model ANP-10338NP Revision 0 Page 5-10 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 I 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 ARTEMISŽ.

ARTEMISŽ 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 ARTEMISŽ are coupled via a Message Passing Interface (MPI).

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-11 Sectorized Reactor Vessel: The S-RELAP5 model used by the AREA TM 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 [ ]

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page 5-12 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 ARTEMISŽ (Section 5.2).

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-13 Table 5-1 ARCADIA Validation Test Matrix for AREA TM Parameter Benchmark Accuracy Comparison Reference and Section --... Ejected rod worth Reference 11 Section 6.2 ITC Reference 11 Section 9.3 OTC Reference 11 Section 6.3 Trip worth Reference 11 Section 9.2 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 Reference 11 Section 7.1 fast reactivity insertion

-NEACRP rod ejection Core power versus time for Reference 11 Section 7 .3 fast reactivity insertion

-SPERT comparisons Static fuel temperatures, Reference 12 Section 9.0 transient fuel temperatures, and heat fluxes Excore power versus time Reference 11 Section 7.2 for dropped rod transients

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AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-14 Table 5-2 GALILEOŽ/

ARTEMISŽ Transient Comparisons for U02 Fuel AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page5-15 Table 5-3 GALILEOŽ/

ARTEMISŽ Transient Comparisons for 2 wt% Gadolinia Fuel AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-16 Table 5-4 GALILEOŽ/

ARTEMISŽ Transient Comparisons for 8 wt% Gadolinia Fuel AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-17 Table 5-5 COBRA-FLXŽ Validation Test Matrix for AREA TM Parameter Benchmark Comparison Accuracy Reference and Section --~ Conservation of mass Reference 13 Section 5.1 and energy Fluid flow solution Reference 13 Section 5.2 Validity of steady state Reference 13 Section 5.3.2 CHF correlations for transients Supports approved CHF Reference 13 Appendix C correlations

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AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure 5-1 Coupling of the Time Dependent Models ANP-10338NP Revision 0 Page5-18 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-19 Figure 5-2 U02 HZP EOL Transient Fuel Centerline Temperature AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-20 Figure 5-3 U0 2 HZP EOL Transient Maximum Rim Temperature AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-21 Figure 5-4 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Centerline Temperature AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-22 Figure 5-5 2 wt% Gadolinia Fuel HFP MOL Transient Fuel Surface Temperature AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-23 Figure 5-6 8 wt% Gadolinia Fuel HZP EOL Transient Fuel Centerline Temperature AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 5-24 Figure 5-7 8 wt% Gadolinia Fuel HZP EOL Transient Maximum Rim Temperature AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report 6.0 AREA TM METHODOLOGY DESCRIPTION ANP-10338NP Revision 0 Page 6-1 This section provides an overview of the AREA TM methodology that is used to demonstrate compliance with the regulatory guidance addressed in Section 2.0. The AREA TM 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 TM as they define the limits of the analysis.

6.2 GALILEOŽ As shown in Figure 6-1, GALILEOŽ 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.

S.2.1 PCMI Failure Criteria for Clad GALILEOŽ 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 TM methodology, two options are available to calculate fuel clad corrosion (or internal pin pressure).

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 1. [ ] ANP-10338NP Revision 0 Page 6-2 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 fror:n GALILEOŽ in conjunction with the regulatory corrosion based failure criteria is used to determine the failure limit, typically given in terms of enthalpy rise (Lical/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:

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-3 A conservative fuel pin pressure versus burnup relationship is generated for the AREAŽ methodology.

Either option from Section 6.2.1 is used to obtain limiting pin pressure information versus burnup. To simulate overheated conditions

[ ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report 6.2.3 Fuel and Rim Melt ANP-10338NP Revision 0 Page 6-4 Fuel melt temperature for U02 and gadolinia fuel versus burnup, rim melt temperature (same as fuel melt at a higher localized burn up), uncertainty of the melt temperature, and the predicted fuel temperature uncertainty are obtained from GALILEOŽ.

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 GALILEOŽ.

If COPERNIC is used, there are two alternatives.

[ 6.2.5 Fuel Pellet to Clad Gap Conductance

[ ] ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-5 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. [ 1 The verification of the gap conductance model is shown in Section 5.2.2. 6.3 ARTEMISŽ Models for REA Event Analysis Once the GALILEOŽ data have been generated, the next phase includes several ARTEMISŽ 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 [ ] consiste11t with application of ARTEMISŽ presented in References 11 and 12 . ., Static ARTEMISŽ calculations are used to establish the initial conditions for the REA event analysis.

.. An ARTEMISŽ 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report

• An ARTEMISŽ [ ANP-10338NP Revision 0 Page 6-6 ] Fuel temperatures are calculated using the ARTEMISŽ fuel rod thermal model while DNBR calculations are performed using the thermal-hydraulic model COBRA-FLXŽ.

6.4 ARTEMISŽ (Steady State Nodal Solution)

The ARTEMISŽ steady state analysis defines a matrix of cases at various power levels and core burn ups 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: . [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-7 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. [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.5 ARTEMISŽ (Transient Nodal Solution)

ANP-10338NP Revision 0 Page 6-8 ARTEMISŽ performs 3-D neutronics kinetics simulations with a time dependent, fuel thermal model and a nodal based COBRA-FLXŽ 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 ARTEMISŽ 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. ARTEMISŽ has the following models to implement a trip function:

• An excore core detector model

• Signal processing

• Control rod drop (SCRAM) function *-__J AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-9 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 ARTEMISŽ 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 in core assembly powers are multiplied by excore weighting factors to translate the incore conditions to ex~ore signals. [ ] 6.5.1.2 Signal Processing ARTEMISŽ 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.5.1.3 Control Rod Drop (SCRAM) ANP-10338NP Revision 0 Page 6-10 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 CROM 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. [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ] 6.5.2 Enthalpy Rise ANP-10338NP Revjsion 0 Page 6-11 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 ).

AREVA Inc. ANP-10338NP Revision O AREA TM -ARCADIA Rod Ejection Accident Topical Report Page 6-12 For power excursions where the ejected rod worth (p) is less than 13, 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 TM methods [ ] 6.5.3 Adjustment Factors ARTEMISŽ adjustment factors are used to account for uncertainty and conservative allowances.

These adjustment factors are used in the AREA TM methodology on the following parameters: . [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ANP-10338NP Revision 0 Page 6-13 ] 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-FLXŽ Calculations In addition to being the nodal simulator, ARTEMISŽ is also the driver for the COBRA-FLXŽ solution.

The COBRA-FLXŽ solution is directly coupled with the time-dependent fuel thermal model in ARTEMISŽ.

[ ] 6.6.1 Adjustment Factors In ARTEMISŽ, 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-FLXŽ.

These adjustment factors are multipliers or adders to the following parameters:

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident T apical Report . [ ] [ ANP-10338NP Revision 0 Page 6-14 ] 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.6.2 DNBR Critical Heat Flux Correlations ANP-10338NP Revision 0 Page 6-15 The AREA TM method for DNBR calculations uses approved CHF correlations which are used in COBRA-FLXŽ 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-FLXŽ 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.7.1 RCS Pressure Evaluations ANP-10338NP Revision 0 Page 6-16 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 CROM 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 TM 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 TM methodology is [ 1 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 1. [ ] ANP-10338NP Revision 0 Page 6-17 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 TM 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.

[ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.8 Data Processing ANP-10338NP Revision O Page 6-18 Results from the transient nodal and detailed model solutions are processed to provide tables and figures for the AREA TM 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 TM application.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.8.2 Total Enthalpy for High Clad Temperature Failure Criteria ANP-10338NP Revision 0 Page 6-19 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 TM 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 TM methodology.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 6.8.3 Fuel Melt Failure Criteria ANP-10338NP Revision 0 Page 6-20 The fuel melt temperature is a function of burnup and fuel pellet material (i.e., in GALILEOŽ, burnup and gadolinia content reduces the melt temperature).

For [ ] the melt temperature for the pellet and rim is calculated.

The melt temperature function from GALILEOŽ 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 TM 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.

AREVA Inc. ANP-10338NP Revision 0 AREAŽ -ARCADIA Rod Ejection Accident Topical Report Page 6-21 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 TM 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.

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report 4. Exceeding CHF causes the fuel to overheat.

[ 6.9 Fuel Failures The AREA TM methodology uses [ ] ANP-10338NP Revision 0 Page 6-22 ] the failure criteria defined. The number of failures is used in determining dose consequences.

6.10 Radiological Consequences The AREA TM 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 TM methodology.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ] 6. 11 Update Process ANP-10338NP Revision 0 Page 6-23 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 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 6-24 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 TM methodology For all codes supporting AREA TM, test cases are included in the code test suite qualification with respect to its application to AREA TM. For those codes that have an NRC approved update process, those processes are followed to support the AREAŽ methodology.

Codes supporting AREA TM 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 TM

• Generation of a summary report documenting the code updates impacting AREA TM along with the validation testing results to be provided to the NRC. Code updates are allowed for any code supporting AREA TM_ AREVA maintains a quality program (including software quality) that is compliant with 1 OCFR50 Appendix B requirements.

This quality program assures updates are made within the bounds of NRC licensing requirements for safety evaluations.

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-25 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 AREATM_

• 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, a*nd 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.

11 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 6-26 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 TM 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 TM. If future regulatory commitments are made relative to the approved codes supporting AREA TM, the changes affecting AREA TM 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: . [ ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 6-27 These conditions are consistent with the biasing method described in Section 7.0. 6.13 Method Summary The AREA TM 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 TM uses the 3-D ARTEMISŽ nodal transient code with [ ] Capability of analyzing fuel pin internal pressure has been incorporated to evaluate coolability issues when fuel failures occur. The AREAŽ methodology also provides the [ ] for dose evaluations.

The methodology also evaluates the RCS pressure criterion for an REA.

AREVA Inc. ANP-10338NP AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure 6-1 REA Analysis Code and Data Links Revision 0 Page 6-28 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure 6-2 SCRAM Position versus Drop Time ANP-10338NP Revision 0 Page 6-29 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report

• ANP-10338NP Revision 0 Page 6-30 Figure 6-3 Pulse Width Definition for Prompt versus Non-Prompt AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure 6-4 DNBR for a Prompt Pulse at 20% Power ANP-10338NP Revision 0 Page 6-31 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7.0 UNCERTAINTY AND BIASING METHODOLOGY ANP-10338NP Revision 0 Page 7-1 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 TM 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 OTC (less negative)

• Decrease in

• Increase in MTC (more positive)

• Increase in fuel pin power peaking (detailed model calculation only) [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-2 This initial biasing is necessary to obtain sensitivities that are representative of limiting I 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

]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-3 Each of these defined dependent variables is used to meet a regulatory requirement.

For each sensitivity case, [ ] [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7 .1.1 Sensitivity Evaluation Method ANP-10338NP Revision 0 Page 7-4 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. [ ]

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report [ ] ANP-10338NP Revision 0 Page 7-5 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7.1.2 Onset of Trip ANP-10338NP Revision 0 Page 7-6 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

[ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7.1.4 Core Biasing Values ANP-10338NP Revision 0 Page 7-7 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.POil specified in the COLR. Basis: [ l 7.1.4.2 Effective Delayed Neutron Fraction (f3) Definition:

B 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.

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page 7-8 Impact: REAs with reactivity insertions less than 13 rely on delayed neutron production to maintain power escalation and have doubling times of seconds or longer. For ERWs greater than ~. the doubling time can be 1000 times smaller. Hence, a lower~ produces a higher power pulse as described in Section 3.1.1. Value: The AREA TM methodology uses a [ ] Basis: [ ] 7.1.4.3 Doppler Temperature Coefficient (DTC) Definition:

The OTC is the reactivity change per fuel temperature change with all other conditions held constant.

impact: The OTC is the major feedback mechanism to mitigate prompt critical transients.

Value: The value used is [ since it is a negative quantity.

] which is a reduction in the magnitude of the OTC AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Basis: [ l 7 .1.4.4 Moderator Temperature Coefficient (MTC) ANP-10338NP Revision 0 Page 7-9 Definition:

The MTC is the reactivity change per unit change in moderator temperature with all other conditions held constant.

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page 7-10 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 t.H) and 3-D (Fa) 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.

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Value: [ ANP-10338NP Revision 0 Page 7-11 ] 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Basis: [ . 7 .1.4. 7 Core Power ] ANP-10338NP Revision 0

• Page 7-12 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.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-13 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 GALILEOŽ (Reference 7). A range of values are used based on sensitivity calculations.

Basis: [ l AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7.1.4.9 Fuel Conductivity ANP-10338NP Revision 0 Page 7-14 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 GALILEOŽ 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Value: The uncertainty on heat capacity from GALILEOŽ [ Basis: [ 7.1.4.11 Burnup Definition:

Burn up is a measure of the depletion of the fuel. ] ANP-10338NP Revision 0 Page 7-15 ] 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Value: [ Basis: [ 7 .1.4.12 Critical Heat Flux (CHF) Correlation

] ANP-10338NP Revision 0 Page 7-16 ] 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-17 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: [ l 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Basis: [ 7.1.4.15 Core Pressure ANP-10338NP Revision O Page 7-18 ] 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.

Vallie: 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 measuren,ent uncertainty on pressure.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Basis: [ ] 7.2 RELAP5 Biasing for Pressure Calculations AN P-10338N P Revision 0 Page 7-19 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7.2.1 RELAPS Peak RCS Pressure Calculations ANP-10338NP Revision 0 Page 7-20 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: [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 7 .2.1.2 Pressurizer Safety Valve Capacity/Setpoints/Tolerance ANP-10338NP Revision O Page 7-21 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: [ l 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: [ l AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report 7.2.1.4 Non-Safety Systems ANP-10338NP Revision 0 Page 7-22 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 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.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident ANP-10338NP Revision 0 Topical Report Page 7-23 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 ONBRs. 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 CROM that could remain open after a CROM is removed. impact: [ ] Value: The area for the breach is available in plant licensing documentation and drawings.

Basis: [ ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-24 Table 7-1 Criteria Applicability to Initial Conditions for Sensitivity Calculations AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Table 7-2 Core Biasing Strategies for the Key Parameters ANP-10338NP Revision 0 Page 7-25 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 7-26 Table 7-3 Parameters Considered For Biasing for RCS Pressure Scenarios AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure 7-1 Doppler Test Result Comparisons ANP-10338NP Revision 0 Page 7-27 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report 8.0 AREA TM PLANT SPECFIC APPLICATION ANP-10338NP Revision 0 Page 8-1 This section describes the intended use of the AREA TM methodology for plant specific applications.

As defined in this section, there is the initial use of the AREA TM 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 TM Methodology The initial application of the AREA TM methodology consists of the following:

1. Applicable regulatory requirements establish appropriate fuel limits. These fuel limits are used as the bases for the AREA TM analyses performed.

2. Define the fuel performance code (GALILEOŽ or COPERNIC) to be used for the analyses.

This topical report defines two fuel performance codes that could be used; COPERNIC or GALILEOŽ.

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 GALILEOŽ are defined and provided in this topical report. 3. The AREA TM 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 GALILEOŽ.

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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ] ANP-10338NP Revision 0 Page 8-2 If the peak values with minimal biasing are already higher than the high burn up 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 TM 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: II ERW

• AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report

• OTC

• MTC

• Initial Fa (at power cases only)

• Initial F t.H (at power cases only)

• Static post ejection Fa

• Static post ejection F t.H

• Fuel failures (see Section A.3.8) ANP-10338NP Revision 0 Page 8-3 If the key parameters established in the initial application of AREA TM are not exceeded in future cycles, then no additional analysis is requ.ired.

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.

AREVA Inc. AREA T M -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 8-4 Figure 8-1 Increased Biasing for Cycle Verification

---

, \ \ \ \ 0 20 \ \ \ \ \ \ \ 40 Burnup ----NRC Criteria ---Imposed Limit

• ERW increased

• Max rod .!\cal/g 60 80 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report 9.0 SAMPLE PROBLEMS ANP-10338NP Revision 0 Page 9-1 The AREA TM 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. [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 9-2 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Table 9-1 Core Biasing Parameters and Values ANP-10338NP Revision 0 Page 9-3 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 9-4 Table 9-2 Biasing Parameters and Values for Overpressure AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report

10.0 CONCLUSION

S ANP-10338NP Revision 0 Page 10-1 The AREA TM 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 TM methodology is applied to three different PWR plant types that result in very similar conclusions.

The AREA TM 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 TM methodology as it applies to specific applications.

AREVA Inc. AN P-10338N P Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page11-1

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," ML 14188C423, 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," ML 111890397, 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-231 OPA Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," May 2004.

AREVA Inc. ANP-10338NP Revision 0 AREAŽ -ARCADIA Rod Ejection Accident Topical Report Page 11-2 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-FLXŽ:

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&VV Designed Pressurized Water Reactors," January 2000. 16. BAW-10231 P-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 197 4. 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. AREA TM -ARCADIA Rod Ejection Accident Topical Report APPENDIX A W 4-LOOP 193 FA PLANT WITH 17X17 FUEL LATTICE ANP-10338NP Revision 0 Page A-1 This sample problem is for a Westinghouse 4-Loop plant. GALILEOŽ (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 ARTEMISŽ 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 GALILEOŽ 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 U02 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 U0 2 and gadolinia fuel, respectively.

Figure A-7 and Figure A-8 contain the fission gas release for M5 and Zr4 fuel, respectively.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report A.2 Boundary Conditions For this sample problem, [ ANP-10338NP Revision 0 PageA-2 trip at [ ] The high flux trip of 118% of rated power and the high pressurizer pressure ] 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

[ l 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-3 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, Ffl.H, and Fa are shown in Figure A-10. The maximum cal/gin 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, Ffl.H, and Fa are shown in Figure A-14. The maximum ~cal/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 ~cal/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.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report [ A.3.3 Minimum Margin to Fuel Melt Limits ] ANP-10338NP Revision 0 PaqeA-4 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, Ft.Hand Fa are shown in Figure A-18. The fuel, fuel rim, and clad temperatures with time are shown in Figure A-19. The difference for [ the [ ] between its fuel temperature and melt limit is shown in ] 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, Ft.Hand Fa 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-5 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, Ft.H and Fa 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report A.3.7 Sensitivity to High Flux Trip ANP-10338NP Revision 0 PageA-6 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 [ for the [ ] While the fuel temperature

[ ] ] 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 [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report A.4 Peak RCS Pressure Assessment ANP-10338NP Revision 0 PaqeA-7 The peak RCS pressure analysis is performed with S-RELAP5 automatically coupled with ARTEMISŽ.

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.

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-8 The nominal and biased cases reach similar peak RCS pressures with the difference in peak pressures of less than [ ] [ ]

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report These results demonstrate that even with a prompt critical excursion of [ ANP-10338NP Revision 0 Page A-9 ] 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

[ ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision O Page A-10 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 [ the RCS. ] are needed to challenge the pressure integrity of AREVA Inc. ANP-10338NP Revision 0 AREAŽ -ARCADIA Rod Ejection Accident Topical Report Page A-11 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 [ ]

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-12 Table A-2 W 4-Loop Limiting Results Summary for Burnup 1 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-13 Table A-3 W 4-Loop Limiting Results Summary for Burnup 2 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-14 Table A-4 W 4-Loop Limiting Results Summary for Burnup 3 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-15 Table A-5 W 4-Loop Limiting Results Summary for Burnup 4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-16 Table A-6 W 4-Loop Limiting Results Summary for Burnup 5 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-17 Table A-7 W 4-Loop, Measure of Conservatism for Limiting Result Cases AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-18 Table A-8 Transient and Static Difference in Limiting Conditions AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report Table A-9 W 4-Loop Plant Overpressure Input Summary ANP-10338NP Revision 0 PageA-19 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-20 Table A-1 O W 4-Loop Plant Overpressure Results Summary (High Pressurizer Pressure Trip Modeled)

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-21 Table A-11 W 4-Loop Plant Overpressure Results Summary AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-22 Table A-12 W 4-Loop Plant Core Pressure for MDNBR Input Summary AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-23 Figure A-1 W 4-Loop Enthalpy Rise Limits for M5 Fuel Based on Relative Oxide Thickness AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-24 Figure A-2 W 4-Loop Enthalpy Rise Limits for Zr4 Fuel Based on Relative Oxide Thickness AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-25 Figure A-3 W 4-Loop Limiting Pressure Parameters for U0 2 Fuel with M5 Clad AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-26 Figure A-4 W 4-Loop Limiting Pressure Parameters for Gadolinia . Fuel with MS Clad **

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-27 Figure A-5 W 4-Loop Limiting Pressure Parameters for U02 Fuel with Zr4 Clad AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-28 Figure A-6 W 4-Loop Limiting Pressure Parameters for Gadolinia Fuel with Zr4 Clad AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-29 Figure A-7 W 4-Loop Limiting FGR for U0 2 and Gadolinia Fuel with MS Clad AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-30 Figure A-8 W 4-Loop Limiting FGR for U0 2 and Gadolinia Fuel with Zr4 Clad AREVA Inc.* AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure A-9 W 4-Loop General Depressurization Curve ANP-10338NP Revision 0 Page A-31 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-32 Figure A-10 Transient F 0 , F t.H, and Core Power for Max Enthalpy Condition AREVA Inc. AREATM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-33 . Figure A-11 Transient Maximum Enthalpy for Max Enthalpy Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-34 Figure A-12 Total Enthalpy Limit with Burnup for Max Enthalpy Condition AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision O PageA-35 Figure A-13 Enthalpy Margin to Limit Scatter Plot for Max Enthalpy Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-36 Figure A-14 Transient Fa, FAH, and Core Power for Max Enthalpy Rise. Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-37 Figure A-15 Transient Maximum Enthalpy Rise for Max Enthalpy Rise Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-38 Figure A-16 Transient Maximum Enthalpy for Max Enthalpy Rise Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-39 Figure A-17 Maximum Enthalpy Rise and Limits by Clad Type for Max Enthalpy Rise Condition AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PaqeA-40 Figure A-18 Transient FQ, FAH, and Core Power for Max Fuel Temperature Condition AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-41 Figure A-19 Transient Fuel, Fuel Rim and Clad Temperature for Max Fuel Temperature Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-42 Figure A-20 Maximum Fuel Temperature by Fuel Type -Margin to Limits for Max Fuel Temperature Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PaqeA-43 Figure A-21 Maximum Fuel Rim Temperature by Fuel Type -Margin to Limits for Max Fuel Rim Temperature Condition AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-44 Figure A-22 Transient FQ, FAH, and Core Power for MDNBR Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure A-23 Transient MDNBR for MDNBR Condition ANP-10338NP Revision 0 Page A-45 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-46 Figure A-24 SAFDL to MDNBR Ratio by Fuel Type as a Function of Burnup for MDNBR Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-47 Figure A-25 SAFDL to MDNBR Ratio by Fuel Type as a Function of Fuel Pin to Core Pressure Difference for MDNBR Condition AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-48 Figure A-26 Case 2 Power Response for High Pressurizer Pressure Trip AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision O PaqeA-49 Figure A-27 Case 2 Pressure Response for High Pressurizer Pressure Trip AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-50 Figure A-28 Case 4 Power Respo1!1se for High Pressurizer Pressure Trip AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-51 Figure A-29 Case 4 Pressure Response for High Pressurizer Pressure Trip AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 PageA-52 Figure A-30 Core Pressure for MDNBR Response Comparison AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report APPENDIX B B&W 177 FA PLANT WITH 15X15 FUEL LATTICE ANP-10338NP Revision 0 Page B-1 This sample problem is for a B&W 177 fuel assembly plant. GALILEOŽ (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 ARTEMISŽ (Reference 11 and Reference

12) is manually coupled for the maximum RCS pressure calculation.

For the manual coupling ARTEMISŽ 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 GALILEOŽ This plant is assumed to have only MS clad fuel. The PCMI limit for excess hydrogen is used for this sample problem and is calculated using GALILEOŽ.

[ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-2 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 U02 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. 0 [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-3 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 8-6 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 8.6. 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 8.4 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 8-8. Power and peak RCS pressure plots for Case 2 are provided in Figure 8-6 and Figure B-7, respectively.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ] These two cases reach similar peak RCS pressures.

[ ] [ ] ANP-10338NP Revision 0 Page B-4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report [ ] Even with the conservatisms used, the peak RCS pressure [ ANP-10338NP Revision 0 Page B-5 ] remains well below 120% overpressure limit value of 3014.7 psia for this plant. j \ ] 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.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-6 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

[ ]

AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report The MDNBR case in Section 8.3 occurs at HFP. [ B. 7 Sample Summary 1 ANP-10338NP Revision 0 Page B-7 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 exists 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 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.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-8 Table 8-1 General Timing of the Event Event Timing Time to eject rod 0.1 second

• fraction of insertion 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 Rods begin to drop Total delay time (1 second after trip actuation)

Rods to full insertion Total drop time (2.4 seconds) Simulation ended for the event [ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-9 Table 8-2 B&W Plant Limiting Results Summary for Burnup 1 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-10 Table 8-3 B&W Plant Limiting Results Summary for Burnup 2 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-11 Table 8-4 B&W Plant Limiting Results Summary for Burnup 3 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-12 Table 8-5 B&W Plant Limiting Results Summary for Burnup 4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-13 Table 8-6 B&W Plant Limiting Results Summary for Burnup 5 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-14 Table B-7 Measure of Conservatism for Each of the Limiting Cases AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Table B-8 B&W plant Overpressure Input Summary ANP-10338NP Revision 0 Page B-15 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-16 Table B-9 B&W plant Overpressure Results Summary (no high pressure trip modeled)

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Table B-10 B&W Plant Overpressure Results Summary ANP-10338NP Revision 0 Page B-17 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-18 Table 8-11 B&W Plant Core Pressure for MDNBR Input Summary AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-19 Figure 8-1 Enthalpy Rise Limits for M5 Fuel Based on Excess Hydrogen AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-20 Figure B-2 Limiting Pressure Parameters for U02 Fuel with MS Clad AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page 8-21 Figure 8-3 Limiting Pressure Parameters for Gadolinia Fuel with MS Clad AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-22 Figure B-4 Limiting FGR for U0 2 and Gadolinia Fuel with M5 Clad AREVA Inc.* AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure 8-5 B&W General Depressurization Curve ANP-10338NP Revision 0 Page B-23 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure B-6 Reactor Power for Biased Case ANP-10338NP Revision 0 Page B-24 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure 8-7 Peak RCS Pressure for Biased Case ANP-10338NP Revision 0 Page B-25 AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report Figure 8-8 Reactor Power For Prompt Critical -No Trip ANP-10338NP Revision 0 Page B-26 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-27 Figure 8-9 Peak RCS Pressure Response for Prompt Critical Reactivity Addition -No Trip AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure 8-10 Hot Leg Pressure Comparison ANP-10338NP Revision 0 Page B-28 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-29 Figure 8-11 Verification of the General Depressurization Curve.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report APPENDIXC CE 217 FA PLANT WITH 14X14 FUEL LATTICE ANP-10338NP Revision 0 Page C-1 This sample problem is for a CE 217 fuel assembly plant. GALILEOŽ (Reference

7) is used as the fuel performance code. The previous two samples problems adequately show the AREA TM 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 GALILEOŽ 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 U02 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

[ ]

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report C.3 Fuel Integrity Sample Problem Summaries

[ ANP-10338NP Revision 0 Page C-2 ] 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 13 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 [ ]

AREVA Inc. ANP-10338NP AREA TM -ARCADIA Rod Ejection Accident Topical Report C.4 Conservatism of Biasing Method Revision 0 Page C-3 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 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 exists 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 fpr coolability and propagation.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-4 Table C-1 CE Plant General Timing of the Event Event Timing Time to eject rod 0.1 second

• fraction of insertion 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 Rods begin to drop Total delay time (1 second after trip actuation)

Rods to full insertion Total drop time (2.844 seconds) Simulation ended for the event [ 1 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-5 Table C-2 CE Plant Limiting Results Summary for Burnup 1 AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-6 Table C-3 CE Plant Limiting Results Summary for Burnup 2 .. __ J AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-7 Table C-4 CE Plant Limiting Results Summary for Burnup 3 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-8 Table C-5 CE Plant Limiting Results Summary for Burnup 4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-9 Table C-6 CE Plant Limiting Results Summary for Burnup 5 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-10 Table C-7 CE Plant Measure of Level of Conservatism for Each Limiting Parameter AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-11 Figure C-1 Enthalpy Rise Limits for MS Fuel Based on Relative Oxide Thickness AREVA Inc. AREATM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-12 Figure C-2 Limiting Pressure Parameters for U02 Fuel with MS Clad AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-13 Figure C-3 Limiting Pressure Parameters for Gadolinia Fuel with MS Clad AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page C-14 Figure C-4 Limiting FGR for U02 and Gadolinia Fuel with M5 Clad AREVA Inc. AREAŽ -ARCADIA Rod Ejection Accident Topical Report Figure C-5 CE Plant General Depressurization Curve ANP-10338NP Revision 0 Page C-15 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report APPENDIX D-ARTEMISŽ AND S-RELAP5 COUPLING DESCRIPTION ANP-10338NP Revision 0 Page D-1 ARTEMISŽ (Reference 01 and 04) and S-RELAP5 (Reference

03) can be run as a coupled system to solve time dependent multi-physics problems.

While the AREA TM methodology allows a manual coupling between the two codes and also allows ARTEMISŽ to be run standalone, the focus of this appendix is on the coupled system method of solution.

Figure D-1 provides a schematic of the major modules and the data coupling between them. The S-RELAP5 code calculates the system response.

The "ARTEMISŽ nodal calculation" calculates the neutronic response for the core. The "ARTEMISŽ Detailed Calculation" calculates the DNBR and fuel rod response [ ] COBRA-FLXŽ (Reference

02) is the name for the core thermal-hydraulics module within ARTEMISŽ that is used for both the nodal simulator and the detailed model (calculation).

The interface between S-RELAP5 and ARTEMISŽ is controlled externally using the "External Controller".

This controller provides the overall time step control for the problem and translation of data between the S-RELAP5 and ARTEMISŽ geometric models. The coupled calculation with S-RELAP5 and ARTEMISŽ includes two simultaneous ARTEMISŽ calculations; one based on a core nodal model and the other based on a detailed core model. Each model is run with its own ARTEMISŽ executable.

The nodal model couples the neutronic, thermal-hyd~aulics, fuel rod module and dehomogenization modules. The detailed model is [ ] with only the thermal-hydraulic module and the fuel rod module. The following sections describe the data sharing of the nodal solution, the detailed model solution, the time step management, parallelization, and the coupling between ARTEMISŽ and S-RELAP5.

The code GALILEOŽ (Reference

05) is not coupled with ARTEMISŽ and S-RELAP5 in the time dependent solution.

GALILEOŽ is only used to generate input data for the ARTIEMISŽ fuel rod module and detailed model. As noted in Section 5.0, COPERNIC (Reference

06) could be used in its place. System Equations The equations for the modules in ARTEMISŽ are provided in References 01 and 02. The equations for S-RELAP5 are provided in Reference

03.

AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page D-2 a. Neutronics module -See pages 3-1 to 3-11 of Reference 01 for the steady state and transient equations.

b. Thermal-hydraulics module -See pages 2-35 to 2-40 of Reference 02 for the basic equations.

c. Fuel rod module -See pages 5-1 to 5-3 of Reference 01 for the heat trarisfer equation and its formulation.

d. Oehomogenization module-See pages 3-15 to 3-19 of Reference 01 for the equations.

e. S-RELAP5 -See pages 2-1 to 2-3 of Reference 03 for the two-fluid field equations and page 2-12 for the noncondensable gas and boron concentration in the liquid field equations.

f. The algorithm for coupling between ARTEMISŽ and S-RELAP5 is described in Appendix 0.7. System State Variables The coupled multi-physics system involves many time-dependent state variables.

These variables are used both directly and to derive additional data elements.

The combined set of time-dependent state variables for S-RELAP5 model is denoted as SR(uR, tR) where tR is the time of the S-RELAP5 solution and uR is the spatial location in the S-RELAP5 model. The set SR(uR, tR) consists of the variables listed above in item e.

AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page D-3 The combined set of time-dependent state variables for the ARTEMISŽ nodal model is denoted as SA(uA-tA) where tA is the time of the ARTEMISŽ nodal solution and uA is the spatial location in the ARTEMISŽ nodal model. The set SA(uA, tA) consists of the variables listed above in items a, b, c and d. The combined set of time-dependent state variables for the ARTEMISŽ detailed model is denoted as Sv(uv, tv) where tv is the time of the ARTEMISŽ detailed solution and uv is the spatial location in the ARTEMISŽ detailed model. The set Sv (uv, tv) consists of the variables listed above in items b and c. D.1 ARTEMIS rM Nodal Transient Calculation The ARTEMISŽ nodal transient calculation begins with a steady state solution at the defined core conditions.

Then, the transient proceeds from one time step to the next as described in this section. The time steps are controlled by a time step manager as described in Appendix 0.4. For the following description, it is assumed that the current transient solution has been completed for time tA-The time step A=O refers to the initial steady state solution.

The transient determines the solution at time tA+1 using the following steps.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report The process is repeated for each time step until the end of the transient is reached. D.2 Detailed Model Transient Calculation ANP-10338NP Revision 0 Page D-4 The calculations and data flow for the ARTEMISŽ thermal-hydraulic module and the fuel rod module of the detailed model are described below. The steps are the same for the detailed model and the nodal model calculations.

Note that the time steps of the detailed model are obtained from the nodal model, thus, tv = tA AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page D-5

[ AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report D.3 S-RELAP5 Calculation Page D-6 The S-RELAP5 equations and state variables were previously provided in the sections "System Equations" and "System State Variables" earlier in this Appendix.

An overview of the transient solution process is provided in Section 1.2.4 of Reference

03. D.4 ARTEMISŽ Time Step Manager The ART EM IS TM time step control method ( see Section 3.1 O of Reference

04) is based on [ ] J AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.5 S-RELAP5 Time Step Manager ANP-10338NP Revision 0 Page D-7 The S-RELAP5 code contains an internal time step manager that provides a variety of checks on solution applicability to control the time step size. [ ] D.6 Parallel Application The solution of the ARTEMISŽ nodal model takes considerably less time than the solution of the detailed model. ARTEMISŽ has been developed for use with multiple threads (parallel calculation).

Typically, more threads are assigned to the detailed model calculation to allow it to complete over the same time frame as the nodal calculation.

This allows for faster completion of the coupled calculation.

D.7 S-RELAP5 and ARTEMISŽ Coupling The data exchange between S-RELAP5 and ARTEMISŽ is made through the External Controller.

This controller performs geometric translation of data and time step control between S-RELAP5 and ARTEMISŽ as needed. The time progression is shown in Figure D-2 and data flow is illustrated in Figure D-3. The automatic control of the calculation steps and their timing are described below.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report The process (steps 1 through 4) is repeated until the end of the transient is reached. ANP-10338NP Revision 0 Page D-8 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.8 Variable Definitions Variable C h p p Q'_coolant Q'_fuel Description precursor concentration enthalpy inlet flow pressure fuel rod power power deposited in the coolant power deposited in the fuel Q'_fuel surface heat flux transferred from the clad to the coolant SA ARTEMISŽ nodal model time-dependent state variables Sn ARTEMISŽ detailed model time-dependent state variables SR S-RELAP5 time-dependent state variables tA time in ARTEMISŽ nodal model tn time in ARTEMISŽ detailed model tR time in S-RELAP5 model Tc moderator temperature Tc1ad clad temperature T ett effective fuel temperature Tt fuel temperature Tin inlet temperature T wall wall temperature u internal energy uA spatial location in ARTEMISŽ nodal model un spatial location in ARTEMISŽ detailed model uR spatial location in S-RELAP5 model U axial velocity Ut liquid specific internal energy U 9 gas specific internal energy Vt liquid velocity v 9 gas velocity Xn noncondensable quality ANP-10338NP Revision 0 Page D-9 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Og void fraction p density Ps boron density Pm moderator density L macroscopic cross section neutron flux <f) surface heat flux ANP-10338NP Revision 0 Page D-10 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report D.9 References ANP-10338NP Revision 0 Page D-11 D1 ANP-10297P-A Revision 0, "The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," February 2013. D2 ANP-10311 P-A Revision 0, "COBRA-FLX:

A Core Thermal-Hydraulic Analysis Code Topical Report," January 2013. D3 Letter, Pedro Salas (AREVA Inc.) to Document Control Desk (NRC), "Document to Support the NRC review of EMF-2103P, Revision 3, 'Realistic Large Break LOCA Methodology for Pressurized Water Reactors'," NRC: 14:001, January 10, 2014, contains a copy of EMF-2100(P), Revision 16, "S-RELAP5 Models and Correlations Code Manual," (Accession No. ML 14016A220), December 2011. D4 ANP-10297P-A Revision 0, Supplement 1, 'The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results Topical Report," June 2015. D5 ANP-10323P Revision 0, "Fuel Rod Thermal-Mechanical Methodology for Boling Water Reactors and Pressurized Water Reactors," July 2013 D6 BAW-10231 P-A Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004.

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure D-1 Data Flow of System and Core-Coupled Calculations ANP-10338NP Revision 0 Page D-12 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page D-13 Figure D-2 Overview of System-Core Coupled Calculation Time Steps AREVA Inc. AREA rM -ARCADIA Rod Ejection Accident Topical Report Figure D-3 Variables passed between S-RELAP5 and ARTEMISŽ ANP-10338NP Revision 0 Page D-14 Original Pages for ANP-10338NP-A, Revision 0 -Page iv -Page ix -Page 5-1 -Page 6-26 -Page A-8 -Page A-17 -Page A-20 -Page B-4 -Page B-14 -Page B-16 -Page B-28 -Page C-10 AREVA Inc. ANP-10338NP Revision 0 AREA T M -ARCADIA Rod Ejection Accident Topical Report Page iv 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 RELAP5 Biasing for Pressure Calculations

.......................................... 7-19 .2.1 RELAP5 Peak RCS Pressure Calculations

................................ 7-20 7 .. 2 RELAP5 Core Pressure for MDNBR Calculations

...................... 7-22 8.0 NT SPECFIC APPLICATION

....................................................... 8-1 8.1 Initial /S.. lication of AREA T M Methodology

....................

......................... 8-1 8.2 Cycle to cle Evaluation

....................................................................... 8-2 9.0 SAMPLE PROBLE S ...................................................................................

.... 9-1

10.0 CONCLUSION

S

............................................................................................. 10-1

11.0 REFERENCES

.................................................................

.............................. 11-1 APPENDIX A W4-LOOP 193 FA ANT WITH 17X17 FUEL LATTICE ..................... A-1 APPENDIX B B&W 177 FA PLANT TH 15X15 FUEL LATTICE .............................. B-1 APPENDIX C CE 217 FA PLANT WITH 4X14 FUEL LATTICE ................................ C-1 AREVA Inc. ANP-10338NP Revision 0 AREA TM -ARCADIA Rod Ejection Accident Topical Report Page ix B&W General Depressurization Curve ..........................

.......................... B-23 Reactor Power for Biased Case .............................................................. B-24 Peak RCS Pressure for Biased Case ......................................

................

B-25 Reactor Power For Prompt Critical -No Trip .......................................... B-26 eak RCS Pressure Response for Prompt Critical Reactivity Addition No Trip ..................................................

.............

................................ B-27 Figure B-10 Hot eg Pressure Comparison

......................

........................................ B-28 Figure B-11 Verific ion of the General Depressurization Curve ............................... B-29 Figure C-1 Enthalpy

• e Limits for M5 Fuel Based on Relative Oxide Thickness

................................................................................

............ C-11 Figure C-2 Parameters for U02 Fuel with M5 Clad ....................

C-12 Figure C-3 Limiting Pressure rameters for Gadolinia Fuel with M5 Clad ............. C-13 Figure C-4 Limiting FGR for U0 2 nd Gadolinia Fuel with M5 Clad ....................... C-14 Figure C-5 CE Plant General Depr surization Curve ............................................. C-15 AREVA Inc. ANALYTICAL MODELS ANP-10338NP Revision 0 Pa e 5-1 A TM methodology is capable of evaluating an REA to demonstrate compliance with the a eptance criteria discussed in Section 2.0. The methodology requires the

• GALILEOŽ ( eference 7) {COPERNIC (Reference

16) can also be used if the outlined validatio s are performed}

• ARTEMISŽ ( Refer ces 11 and 12), a coupled 3-0 kinetics solution with al model, and 3-0 thermal hydraulic model

• COBRA-FLXŽ (Reference

3) as the 3-0 thermal hydraulic model implemented in Reference 12

• S-RELAP5 (Reference

8) for Wes* ghouse and CE plants or RELAP5/M002-B&W (Reference

9) for B&W plants Figure 5-1 shows the coupling of the time de ndent models. The fuel performance code is the source of thermal properties of the f I , clad, and gap for the time dependent models which is why it is not shown in

• ure 5-1. The ARTEMISŽ nodal and detailed model are approved in Reference

11. Th interface with RELAP5 is introduced in this topical report. As shown in Figure 5-1 t ee distinct models can be used together with information exchange between the mode where appropriate.

A description of these models follows. . [ ]

AREVA Inc. ANP-10338NP Rev i sion 0 Pa e 6-26 F xibility to perform discretionary updates is important to maintaining modern and computer codes. For instance, making updates and improvements to physical models nd correlations (that have no more than a small impact on the results) is a necessary lement to expand the robustness of the application. This flexibility provides AREVA the a 'lity to maintain the AREAŽ methodology so that it keeps pace with subsequent upd es and improvements from new data or expanded assessments and to keep pace with p tential changes in regulatory guidance. It is foreseen that NRC proval may be granted for updates to approved codes and/or correlations that revise ore tend a code's capabilities for use with AREA T M. If future regulatory commitments are de relative to the approved codes supporting AREA T M, the changes affecting AREA TM w be incorporated without further NRC notification or request for renewal/approval.

6.12 Level of Significance The following definition is used to classify a

• nificant update as it affects the results to the dependent variables listed in Section 7.1.1, hen determining the impact of updates to computer codes, correlations or data libraries: . [ ]

AREVA Inc. AREA T M -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-8 e nominal and biased cases reach similar peak RCS pressures with the difference in ressures of less than [ [ ]

AREVA Inc. AREA T M -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-17 Table A-7 W 4-Loop, Measure of Conservatism for Limiting Result Cases AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page A-20 Table A-10 W 4-Loop Plant Overpressure Results Summary (High Pressurizer Pressure Trip Modeled)

AREVA Inc. AREA T M -ARCADIA Rod Ejection Accident Topical Report These two cases reach s1 ilar peak RCS pressures.

[ ] [ ] ANP-10338NP Revision 0 Page B-4 AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident T apical Report ANP-10338NP Revision 0 Page B-14 Table B-7 Measure of Conservatism for Each of the Limiting Cases AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report ANP-10338NP Revision 0 Page B-16 Table 8-9 B&W plant Overpressure Results Summary (no high pressure trip modeled)

AREVA Inc. AREA TM -ARCADIA Rod Ejection Accident Topical Report Figure B-10 Hot Leg Pressure Comparison ANP-10338NP Revision 0 Page B-28 AREVA Inc. ANP-10338NP Revision 0 CE Plant Measure of Level of Conservatism for Each Limiting Parameter Pa e C-10