Text

ANP-2834(NP), Revision 0, Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report, Enclosure 4
	ML093350101
Person / Time
Site:	Calvert Cliffs
Issue date:	09/30/2009
From:	; AREVA NP
To:	; Office of Nuclear Reactor Regulation
References
	ANP-2834(NP), Rev 0
	Download: ML093350101 (126)
	v • d • e

ENCLOSURE (4)

RLB LOCA Evaluation Evaluation (Non-Proprietary Version)

Calvert Cliffs Nuclear Calvert Nuclear Power Plant, LLC November November 23, 2009 2009

/

ANP-2834(NP)

Revision Revision 000 000 Calvert Cliffs Nuclear Plant Calvert Cliffs Nuclear Plant Unit Unit 11 Cycle Cycle 21 21 &

& Unit Unit 22 Cycle Cycle 19 Realistic Realistic Large Large Break Break LOCA LOCA Summary Summary Report Report September September 2009 2009

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Nuclear Plant .

Revision 000 Unit 1 Cycle 21 & Unit 22 Cycle 19

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Plant Revision Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 1919 Eoiei.ILarge Realistic Break ILOCA

-rm.lraol, flflA (Z.-nrs Summary Dnrt Report

.. . "Page

... .. . . . . Page ii ii Changes Nature of Changes Nature Item Page Description Description and Justification Justification

1. All This is a new document.

AREVA AREVA NP Inc.

Plant ANP-2834(NP)

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & Unit 2 Cycle Cycle 19 19 Revision 000 Revision

& Unit 2 Realistic Large Break Realistic Break LOCA Summary Report Report Page iii Page iii Contents Contents 1.0 Introduction .................................................................................................................

Introduction ................................................................................................................. 1-1 1-1 2.0 Sum m ary ........................................................................................

Summary .....................................................................................................................

, ............................. 2-1 3.0 Analysis .......................................................................................................................

Analysis ....................................................................................................................... 3-1 3.1 Description of the LBLOCA Event Description ....................................................................

Event .................................................................... 3-2 3-2 3.2 Description of Analytical Models ......................................................................

Description ...................................................................... 3-3 3-3 3.3 Plant Description and Summary Sum m ary of Analysis Parameters Param eters ..................................

.................................. 3-6 3-6 3.4 SER Com pliance ..............................................................................................

Compliance .............................................................................................. 3-7 3-7 3.5 Realistic LargeLarge Break LOCA Results ................................................................

............................................................... 3-8 3-8 4.0 Generic Support for Transition Package ......................................................................

Package ...................................................................... 4-1 4.1 Reactor Power .................................................................................................

Power ................................................................................................. 4-1 4.2 Q uench .....................................................................................................

Rod Quench ..................................................................................................... 4-1 4.3 Rod-to-Rod Therm Rod-to-Rod Thermal al Radiation ........................................................................

........................................................................ 4-2 4.4 Film Boiling Boiling Heat Transfer Transfer Limit Lim it .......................................................................

....................................................................... 4-8 4.5 Downcom er Boiling ..........................................................................................

Downcomer .......................................................................................... 4-8 4-8 4.6 Break ......................................................................................................

Break Size ...................................................................................................... 4-24 4.7 inform ation for Containment Detailed information Containm ent Model Model ..................................................

................................................... .4-354-35 4.8 Cross-References Cross-References to North Anna ....................................................................

....................................................................4-39 4.9 G DC 35 - LOOP GDC LOO P and No-LOOP No-LOO P Case Sets ...................................................

.................................................... .4-404-40 4.10 Input Variables Statem ent ...............................................................................

Variables Statement ................................................................................4-41 5.0 Conclusions .................................................................................................................

Conclusions ................................................................................................................. 5-1 6.0 ..................................................................................................................

References .................................................................................................................. 6-1 AREVA NP Inc.

Inc.

Plant Calvert Cliffs Nuclear Plant ANP-2834(NP)

ANP-2834(NP)

Cycle 19 Unit 22 Cycle 19 Revision 000000 Unit 11 Cycle 21 &

& Unit Realistic Large Realistic Large Break Break LOCA LOCA Summary Summary Report Report Page iv Pageiv Tables Tables Table 2-1 Summary of Major Parameters Parameters for Limiting Transient...: Transient............................................

......................................... 2-1 Table 3-1 Sampled LBLOCA Parameters ..............................................................................

Parameters .............................................................................. 3-10 3-10 Table 3-2 Plant Operating Operating Range Supported Supported by the LOCA Analysis .....................................

..................................... 3-11 3-11 Table 3-3 Statistical Distributions Distributions Used for Process Parameters Parameters ...........................................

........................................... 3...:13 3-13 Table 3-4 SER Conditions Conditions and LimitationsLimitations ............................................................................

............................................................................ 3-14 3-14 Table 3-5 Summary of Results for the Limiting PCT Case ....................................................

.................................................... 3-163-16 Table 3-6 Calculated Calculated Event Times for the Limiting PCT Case ...............................................

.......................... 3-16 3-16 Table 3-7 Parameters for the Limiting Case ...................................................

Heat Transfer Parameters ................................................... 3-173-17 Table 3-8 Containment Initial and Boundary Conditions ........................................................

Conditions ........................................................ 3-18 3-18 Table 3-9 Passive Heat Sinks in Containment .......................................................................

in Containment... .................................................................... 3-19 3-19 Table 4-1 Typical Measurement Measurement UncertaintiesUncertainties and Local Peaking Peaking Factors .............................

............................. 4-4 Table 4-2 FLECHT-SEASET FLECHT-SEASET & & 17x17 17x17 FA Geometry Geometry ParametersParameters ...........................................

........................................... 4-54-5 Table 4-3 FLECHT-SEASET FLECHT-SEASET Test Parameters Parameters .......................................................................

....................................................................... 4-6 4-6 Table 4-4 Minimum Break Area for Large Break LOCA Spectrum Spectrum ......................................... 4-26

........................................ .4-26 Table 4-5 Minimum PCT Temperature Temperature Difference Difference - True Large and Intermediate Intermediate Bre a ks ..................................................................................................................

Breaks " ................................................................. '" ............................................4-28 4 -2 8 AREVA NP Inc.

Calvert Cliffs Nuclear Nuclear PlantPlant ANP-2834(NP)

ANP-2834(NP)

Unit 1 Cycle 21 & & Unit Unit 2 Cycle 19 19 Revision 000 Revision 000 Realistic Large Break LOCA Summary Summary Report Report Page v Page v Figures Figures Figure 3-1 Primary Prim ary System Noding Noding .......................................................................................

....................................................................................... 3-20 3-20 Figure 3-2 Secondary System Noding Noding ...................................................................................

...................................................................................3-21 Figure 3-3 Reactor R eactor V essel N Vessel oding ........................................................................................

Noding ........................................................................................3-22 3-22 Figure 3-4 Core Core Noding Noding DetailDetail ..............................................................................................

......................................... :.....................................................3-23 3-23 Figure 3-5 Upper Plenum Noding Detail Detail.................................................................................

...............................................................................3-24 3-24 Figure 3-6 Scatter Plot of Operational Operational Parameters ................................................................

................................................................ 3-25 3-25 Figure 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations Calculations .....................................

..................................... 3-273-27 Figure 3-8 PCT versus Break Break Size Scatter .................................... 3-28 Scatter Plot from 59 Calculations .................................... 3-28 Figure 3-9 Maximum Oxidation versus PCT Scatter Plot from 59 Calculations ..................... ..................... 3-29 3-29 Figure 3-10 Total Oxidation versus PCT Scatter Scatter Plot from 59 Calculations ........................... 3-30 Calculations ........................... 3-30 Figure 3-11 Peak Cladding Temperature Temperature (Independent(Independent of Elevation) Elevation) for the the Lim iting C Limiting a se .....................................................................................................

Case .....................................................................................................3-31 3-3 1 Figure 3-12 Break Flow for the Limiting Case .......................................................................

....................................................................... 3-32 3-32

............ i .......................................... 3-33 Figure 3-13 Core Inlet Mass Flux for the Limiting Case .........................................................

Figure 3-14 Core Outlet Mass Flux for the Limiting Case ......................................................

...................................................... 3-34 Figure 3-15 Void Fraction Fraction at RCS Pumps for the Limiting ............................................ 3-35 Limiting Case ............................................

Figure 3-16 ECCS Flows (Includes SIT, LPSI and HPSI) HPSI) for the Limiting Case ..................... ..................... 3-36 Figure 3-17 Upper Upper Plenum Pressure for the Limiting ....................................................

Limiting Case .................................................... 3-37 Figure 3-18 Collapsed Collapsed Liquid Level in the Downcomer Downcomer for the Limiting Case ....................... ....................... 3-38 3-38 Figure 3-19 Collapsed Collapsed Liquid Level in the Lower Plenum for the Limiting .................... 3-39 Limiting Case .................... 3-39 Figure 3-20 Collapsed Collapsed Liquid Level in the Core for the Limiting Case .................................. .................................. 3-403-40 Figure 3-21 Containment Containment and Loop Pressures Pressures for the Limiting Case .....................................

..................................... 3-41 Figure 3-22 GDC 35 LOOP versus No-LOOP Cases ............................................................

............................................................3-42 3-42 Figure 4-1 R2RRAD R2RRAD 5 x 5 Rod Segment.. Segment ................................................................................

.............................................................................. 4-5 Figure 4-2 Rod Thermal Radiation FLECHT-SEASET Bundle and in a 17x17 Radiation in FLECHT-SEASET 17x17 F A ..........................................................................................................................

FA .......................................................................................................................... 4 -7 4-7 Figure 4-3 Reactor Reactor Vessel Downcomer Downcomer Boiling Diagram ........................................................

........................................................ 4-9 4-9 Figure 4-4 S-RELAP5 versus Closed Form Solution Solution .............................................................

.............................................................4-12 4-12 Figure 4-5 Downcomer Downcomer Wall Heat Release Release - Wall Mesh Mesh Point Sensitivity ............................. 4-13

............................ .4-13 Figure 4-6 PCT independent Independent of Elevation - Wall Mesh Mesh Point Sensitivity ..............................

............................... .4-14 4-14 Figure 4-7 Downcomer Downcomer Liquid Level- Level - Wall Mesh Point Sensitivity .......................................

........................................ .4-15 4-15 Figure 4-8 Core Liquid Level - Wall Mesh Point Sensitivity ...................................................

.................................................. .4-16 4-16 Figure 4-9 A zim uthal Noding Azimuthal Noding .................................................................................................

................................................................................................. 4-18 4-18 AREVA NP Inc.

Plant Nuclear Plant Calvert Cliffs Nuclear ANP-2834(NP)

Unit 1 Cycle 21 & Unit 22 Cycle 19

& Unit 19 Revision Revision 000 000 Realistic Large Break Realistic Break LOCA Summary Report Report Page vi Page vi Figure 4-10 Lower Compartment Compartment Pressure Pressure versus Time .......................................................

........................................................ .4-19 4-19 Figure 4-11 Downcomer Wall Heat Release Release - Axial Noding Noding Sensitivity Study ......................

....................... .4-20 4-20 Figure 4-12 PCT Independent Independent of Elevation Elevation - Axial Noding Sensitivity Study ........................ ......................... .4-21 4-21 Figure 4-13 Downcomer Downcomer Liquid Level - Axial Noding Level- Noding Sensitivity ................................. .4-22 Sensitivity Study ................................ 4-22 Figure 4-14 Core Liquid Level - Axial Noding Sensitivity Sensitivity Study ...........................................

............................................ .4-234-23 Figure 4-15 Plant A - Westinghouse Westinghouse 3-Loop Design .............................................................

............................................................ .4-29 4-29 Figure 4-16 Plant B - Westinghouse 3-Loop Design ............................................................ 4-30

.4-30 Figure 4-17 Plant C - Westinghouse Westinghouse 3-Loop 3+00p Design .............................................................

Design ............................................................ 4-31

.4-31 Figure 4-18 Plant D D - Combustion Engineering 2x4 Design .................................................

.................................................. .4-32 4-32 Figure 4-19 Plant E - Combustion Engineering 2x4 Design .........................

Combustion Engineering ......................... :.......................

........................ .4-33 4-33 Figure 4-20 Plant F - Westinghouse Westinghouse 3-loop Design ..............................................................

............................................................. .4-34 4-34 Figure 4-21 PCT vs. Containment Containment Volume ............................................................................

............................................................................ 4-36

.4-36 Figure 4-22 PCT vs. Initial Containment Containment Temperature ............................................................

Temperature ............................................................ 4-37 4-37 Figure 4-23 Containment Pressure as function of time for limiting case ........

4-23 Containment ........ ,:.......................

.........................4-38 4-38 This document document contains a total of 97 pages.

AREVA AREVA NP Inc.

ANP-2834(NP)

ANP-2834{NP)

Calvert Cliffs Calvert Cliffs Nuclear Nuclear Plant Plant Revision Revision 000 000 Unit 1 Cycle Unit Cycle 2121 &

& Unit 22 Cycle 19 Realistic Realistic Large Large Break Break LOCA Summary Report LOCA Summary Paae Page vii Nomenclature Nomenclature ASI Axial Shape Index Shape Index CCTF CCTF Cylindrical Core Test Facility Cylindrical Facility CE CE Combustion Engineering Inc.

Combustion Engineering CFR CFR Federal Regulations Code of Federal Regulations CSAU Code Scaling, Scaling, Applicability, and and Uncertainty Uncertainty DC Downcomer Downcomer DEGB DEGB Double-Ended Double-Ended Guillotine Guillotine Break DNB DNB Departure Departure from Nucleate Boiling Nucleate Boiling ECCS ECCS Emergency Core Cooling Emergency Cooling System EFPH Effective Full Power Hours Effective Hours EM EM Evaluation Model Evaluation FQ FQ Total Peaking Peaking Factor Factor FAH F6H Nuclear Enthalpy Rise Factor Nuclear HPSI High Pressure Pressure Safety Injection Injection HFP HFP Power Hot Full Power LANL LANL Los Alamos National National Laboratory LHGR LHGR Generation Rate Linear Heat Generation Rate LOCA LOCA Loss of Coolant Accident Accident LOOP LOOP Loss of Offsite Power LPSI Injection Low Pressure Safety Injection MSIV MSIV Steam Isolation Valve Main $team Valve NRC U. Nuclear Regulatory Commission U. S. Nuclear Commission NSSS NSSS Nuclear Steam Supply System Nuclear PCT Temperature Peak Clad Temperature PIRT Phenomena Identification and and Ranking Ranking Table Table PLHGR PLHGR Planar Linear Planar Linear Heat Heat Generation Generation Rate Rate PWR Pressurized Water Reactor Pressurized RAS Recirculation Actuation Signal Recirculation Signal RCP RCP Reactor Coolant Pump RCS RCS Reactor Coolant System Reactor RHR RHR Residual Heat Removal Residual RLBLOCA Realistic Large Break Loss of Coolant Accident RV Reactor Vessel RWST Storage Tank Refueling Water Storage SIAS SIAS Injection Activation Signal Safety Injection SIT SIT Injection Tank Safety Injection SER Evaluation Report Safety Evaluation wlo w/o Percent Weight Percent AREVA NP NP Inc.

. ANP-2834(NP)

ANP-2834(NP)

Plant Nuclear Plant Calvert Cliffs Nuclear Unit 22 Cycle 19 Revision 000000 Unit 11 Cycle 21 && Unit Cycle 19 Realistic Large Break LOCA Summary Report Realistic Page 1-1 Page 1-1 Introduction 1.0 Introduction This report report describes describes and provides results from a RLBLOCA RLBLOCA analysis for the Calvert Cliffs Calvert Cliffs Nuclear Nuclear Plant Unit 1 Cycle 21 and Unit 2 Cycle 19. The plant is aaCE-designed CE-designed 2737 MWt plant MWt plant with aa large dry containment. AREVA NP will be the fuel supplier, starting starting with Unit 2 Cycle 19. 19.

The plant is aa 2X4 loop designdesign - two hot legs and four cold legs. The loops contain four RCPs, two U-tube generators and one pressurizer. The ECCS is provided by two independent U-tube steam generators independent safety injection injection trains and four SITs.

The analysis supports operation for Unit 1 Cycle 21 as well as Unit 2 2 Cycle 19 and beyond beyond with AREVA NP's HTP 14X14 14X14 fuel design using standard U0 U0 2 fuel with 2, 4, 6 and 8 wlo w/o Gd Gd 220 33 and M5 cladding, cladding, unless unless changes in the Technical Specifications, Specifications, Core Operating Operating Limits Report, core design, fuel design, plant hardware,hardware, or plant operation invalidate invalidate the results presented presented herein. The analysis was performed performed in compliance compliance with the NRC-approved NRC-approved RLBLOCA RLBLOCA EM (Reference 1) with exceptions (Reference exceptions noted below. Analysis results confirm the 10CFR50.46 10CFR50.46 (b) (b)

~cceptance acceptance criteria presented presented in Section 3.0 are met and serve as the basis for operation of the the Calvert Cliffs Nuclear Nuclear Plant Units Units 1 and 2 with AREVA NP fuel.

The non-parametric non-parametric statistical methods inherent inherent in the AREVA AREVA NP RLBLOCA methodology methodology consideration of a full spectrum of break sizes, break configuration (guillotine or provide for the consideration or split break), axial shapes, shapes, and plant operational operational parameters.

parameters. A conservative loss of a diesel assumption is applied in which LPSI inject into the broken broken loop and one intact loop and HPSI inject into all four loops. Regardless Regardless of the failure assumption, all containment containment pressure-reducing --systems

. pressur:e."r:educing --systems are are.... conservatively assumed,,.. Jully conservatively assumed fully,-__ functional-._

functiQI];3L_ T Th~h.e .. effe.gt~

effects 9f of Gadolinia-bearing fuel rods and peak fuel rod exposures are considered.

Gadolinia-bearing considered.

deviations from the approved The following are deviations RLBLOCA EM (Reference approved RLBLOCA (Reference 1) that were requested by the NRC.

requested NRC.

The assumed reactor core power for the Calvert Cliffs realistic large break loss-of-coolant loss-of-coolant accident is 2754 MWt. This value represents accident represents the plant rated thermal power (Le., (i.e., total reactor core heat transfer rate to the reactor coolant system) of 2737 MWt reactor coolant with aa maximum MWt \.yith maximum power measurement measurement uncertainty of 0.62% 0.62% added to the rated thermal power. The power power was not sampled in the analysis. This is not expected expected to have aa noticeable effect on the PCT results.

AREVA NP Inc.

Plant ANP-2834(NP)

ANP-2834(NP)

Nuclear Plant Cliffs Nuclear Calvert Cliffs Unit 1 Cycle 2121 & 19 Unit 22 Cycle 19

& Unit Revision Revision 000 000 Realistic Large Break Break LOCA Summary Report Page 1-2 Page 1-2 The RLBLOCA analysisanalysis was performed performed with aa version of S-RELAP5 S-RELAP5 that requires requires both the void void fraction to be less than 0.95 and the clad temperature temperature to be less than 900'F before the rod is allowed allowed to quench.

quench. This may result in aa slight increase increase in PCT results when compared compared to an analysis not subject subject to these constraints.

The RLBLOCA RLBLOCA analysis was performed performed with aa version of S-RELAP5 S-RELAP5 that limits the contribution contribution of the Forslund-Rohsenow Forslund-Rohsenow model to no more than 15 percent percent of the total heat transfer transfer at and and above a void fraction of 0.9. This may result in aa slight increase increase in PCT results when compared compared to previous analyses for similar plants.

The split versus double-ended break type is no longer related to break versus double-ended break area. InIn concurrence concurrence with with Regulatory Regulatory Guide 1.157, both the split and the double-ended Guide 1.157; double-ended break will range in area between between the minimum break area (Amln) (Amn) and an area of twice the, size of the the broken pipe. The The determination of break configuration, determination configuration, split versus double-ended, double-ended, will be made after the break break area is selected based on a uniform probability for each occurrence.occurrence. Amin was calculated calculated to bebe 28.7 percent of the DEGB area (see Section Section 4.6 for further discussion). This is not expected expected to to noticeable effect on PCT results.

have a noticeable In concurrence In interpretation of GDC 35, aa set of 59 cases was run with aa LOOP concurrence with the NRC's interpretation LOOP assumption and a second set with a No-LOOP assumption. The set of 59 cases that predicted predicted the highest PCT is reported in Section 22 and Section Section 3, herein. The results from both case sets sets are shown in Figure 3-22. The effect effect on PCT results is expected expected to be minor.

During recent recent RLBLOCA EM modeling modeling studies, itit was noted that cold leg condensation condensation efficiency may be under-predicted.

under-predicted. Water entering entering the DC post-SIT injection remained remained sufficiently subcooled to absorb absorb DC wall heat release release without significant significant boiling. However, tests (Reference 7) indicate that the steam and water entering (Reference entering the DC from the cold leg, subsequent subsequent to the end of SIT injection, injection, reach near near saturation saturation resulting from the condensation efficiency condensation efficiency ranging between 80 to 100 percent. To assure that cold leg condensation ranging condensation would not be under-under-predicted, a RLBLOCA EM update was made. Noting that saturated fluid entering entering the DC is the the most conservative modeling scheme, steam and liquid multipliers conservative modeling multipliers were developed so as to to approximately approximately saturate saturate the cold leg fluid at cold leg pressure before itit enters the DC. The The multipliers were developed developed through scoping studies using a number number of plant configurations-configurations-Westinghouse-designed Westinghouse-designed 3- and 4-loop plants, and CE-designed CE-designed plants. The results of the the AREVA NP Inc.

ANP-2834(NP)

Nuclear Plant Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & & Unit 2 19 Cycle 19 2 Cycle 000 Revision 000 Realistic Realistic Large Break LOCA Summary Report Page 1-3 Page 1-3 scoping study indicated that multipliers of 10 and 150 seoping 150 for liquid and steam, respectively, were were appropriate to produce saturated saturated fluid entering the DC. This RLBLOCA EM departure departure was was recently discussed with the NRC and the NRC agreed that the approach described immediately approach described immediately above was satisfactory satisfactory in the interim. The modification modification is implemented implemented post-SIT injection, injection, 10 seconds after the vapor void fraction in the bottom of the SIT becomes becomes greater greater than 90 percent.

Thus, the SITs have have injected injected all their water into the cold legs, and the nitrogen cover gas has has entered entered the system ,and and been mostly discharged through the break before the condensation condensation efficiency is increased increased by the factors of 10 and 150, for liquid and vapor respectively.

respectively. Providing Providing saturated saturated fluid conditions at the DC entrance conservatively conservatively reduces reduces both the DC driving head and the core flooding rate. Recall Reca" that test results indicate indicate that fluid conditions entering the DC DC range from saturated to slightly subcooled.

subcooled. Hence, Hence, itit is conservative conservative to force an approximation approximation of saturated saturated conditions for fluid entering the DC.

AREVA Inc. has acknowledged conductivity degradation acknowledged an issue concerning fuel thermal conductivity degradation as a function of burnup burnup as raised by the, the NRC. In order to manage manage this issue, AREVA Inc. is modifying modifying the way RODEX3A temperatures are compensated in in the RLBLOCA Revision Revision 0/Transition package OlTransition package methodology. In the current process, the RLBLOCA computes PCTs at many different different times during an operating cycle. For each specific time in cycle, the fuel conditions conditions are computed using RODEX3A RODEX3A prior to starting the S-RELAP5 S-RELAP5 portion of the the analysis. A steady steady state condition for the given time in cycle using S-RELAP5 is established. A using S-RELAP5 A base fuel centerline centerline temperature two-transformation temperature is established in this process. Then two-transformation adjustment to the base fuel centerline temperature adjustment temperature is computed.

computed. The first transformation transformation is a linear adjustment-for an-exposure linear adjustment-for an-exposure of 10 MWd/M:rU MWd/MTU or-higher. In In the.

the. new process, aa polynomial new.process, transformation is used in the first transformation transformation transformation instead of a linear transformation.

transformation. The rest of S-RELAP5 fuel rod temperature the RLBLOCA process for initializing the S-RELAP5 temperature should not be altered and the rest of LOCA LOCA transient transient should also continue continue in the original fashion. This approach approach hashas been requested requested by the NRC. NRC.

AREVA NP Inc.

Plant Nuclear Plant ANP-2834(NP)

ANP-2834(NP)

Calvert Cliffs Nuclear Calvert

& Unit 22 Cycle 19 Revision 000 Revision Unit 1 Cycle 21 & 19 Realistic Large Break LOCA SummarySummary Report Page 2-1 Page 2-1 2.0 Summary The limiting PCT analysis is based based on the parameter specification specification given in Table 2-1. The 2-1. The limiting PCT is 1670°F1670°F for a 4 w/o Gd Gd 200 3 Rod in a case LOOP conditions. U0 case with LOOP and U0 2 rods and Gadolinia Gadolinia bearing rods of 2, 6 and 8 w/o were were also analyzed, but, were found to be bounded.bounded.

This RLBLOCA result is based on aa case set of 59 individual transient transient cases for LOOP and 59 59 individual transient casescases for No-LOOP conditions. The core is composed composed of AREVA AREVA NP HTP HTP 14x14 thermal 14x14 thermal hydraulically hydraulically compatible fuel designs designs with existing existing Westinghouse Westinghouse fuel designs.

The analysis analysis assumed assumed full core power operation at 2754 MWt. The value represents the nominal core power including measurement uncertainty including measurement uncertainty of 0.0062. The analysis assumedassumed a steam generator generator tube plugging plugging level of 10 10 percent percent in all steam generators, 15.0 kW/ft generators, aa total LHGR of 15.0

/

(no axial dependency), a total peaking peaking factor factor (FQ)

(Fa) up to aa value of 2.384, and a nuclear enthalpy enthalpy rise factor rise factor (F(FAH) aH) up to aa value vallje of 1.81 (including 6% uncertainty).

1.81 (including uncertainty). This analysis addresses addresses typical operational ranges or technical operational technical specification specification limits (whichever is applicable) applicable) with regard regard to Pressurizer Pressurizer pressure pressure and level; level; SIT pressure, pressure, temperature, and level; core average average temperature; core flow; containment containment pressure pressure and temperature; temperature; and RWST.

The AREVA AREVA RLBLOCARLBLOCA methodology methodology explicitly analyzes only fresh fuel assemblies (see explicitly analyzes (see Reference Reference 1, Appendix B). Previous analyses analyses have shown shown that once- twice-burnt fuel will once- and twice-burnt not be limiting up to peak peak rod average exposuresexposures of 62,000 MWd/MTU.

MWd/MTU. The analysis analysis demonstrates that the 10 CFR 50.46(b) criteria listed in Section demonstrates Section 3.0 are satisfied.

satisfied.

" Table 2 "Summary-of Table-2;.1- Summary-of Major Major Parameters Parameters for-Limiting Transient Core Average Burnup (EFPH)

Average Burnup (EFPH) 15085.66 15085.66 Core Power (MWt)(MWt) 2754 Hot Rod LHGR, kW kW/ft 1ft 14.5450 14.5450 Total Hot Rod Radial Peak (F (FrT) r T) 1.810 1.810 ASI (Axial Sha~e Shape Index) -0.0932

-0.0932 Break Type Guillotine Guillotine Break Size (ft22/side) 3.6978 Offsite Power Power Availability Availability available Not available Decay Heat Decay_ Heat Model ANS 1979 1979 Nominal Decay Heat Multiplier Heat Multiplier 0.99364 0.99364 AREVA NP Inc.Inc_

Plant Nuclear Plant Calvert Cliffs Nuclear Calvert ANP-2834(NP)

ANP-2834(NP) 21 & 19 Unit 22 Cycle 19 Revision 000 000 Unit 1 Cycle 21 & Unit Realistic Large Break LOCA Summary Summary Report Report Page 3-1 Page 3-1 3.0 Analysis Analysis The purpose of the analysis is to verify typical technical specification peaking factor limits and technical specification the adequacy adequacy of the ECCS by demonstrating demonstrating that the following 10CFR 10CFR 50.46(b) criteria are met:

(1)

(1) The calculated maximum fuel element element cladding temperature temperature shall not exceed 2200'F.

(2)

(2) The calculated calculated total oxidation of the cladding nowhere exceed cladding shall nowhere exceed 0.17 times the total

. cladding cladding thickness before oxidation.

(3)

(3) The calculated calculated total amount amount of hydrogen generated generated from the chemical reaction.

reaction of the the cladding with water or steam shall not exceed cladding exceed 0.01 times the hypothetical amount that would be generated would* generated if if all of the metal in the cladding cladding cylinders surrounding the fuel excluding the cladding surrounding surrounding the plenum volume were to react.

(4) The calculated changes in core geometry shall be such that the core remains amenable (4) The calculated changes in core geometry shall be such that the core remains amenable cooling.

to cooling.

(5)

(5) Long-term cooling is not addressed Long-term addressed in this calculation.

The analysis analysis did not evaluate evaluate core coolability coolability due to seismic events, nor did itit consider the consider the 10CFR 50.46(b) long-term cooling criterion.

10CFR RLBLOCA analysis conservatively The RLBLOCA conservatively considers considers blockage effects due to clad swelling and rupture rupture in the prediction of the hot fuel rod PCT. AREVA NP has previously performed performed an an analysis analysis which demonstrates demonstrates that for all cases of horizontal horizontal seismic and LOCA LOCA loads, the the resulting loads are below the spacer spacer grid elastic load limit and thus the grids sustain no no permanent deformation.

permanent deformation.

The ECCS performance performance analysis analysis for Calvert Cliffs Units 1 and 2 assures the core remains remains amenable to cooling amenable cooling from the effects effects of fuel cladding rupture and swelling, and the effects of LOCA and seismic loads. The RLBLOCA RLBLOCA analysis conservatively conservatively considers considers blockage effects blockage effects due to clad swelling and rupture in the prediction prediction of the hot fuel rod PCT. The effects of combined loads (seismic and LOCA) on the fuel assembly components components have have been evaluated by been evaluated by demonstrating that the resulting loads are below the allowable AREVA demonstrating allowable stress limit for all the the components, thus preventing components, preventing permanent permanent deformation.

deformation. Therefore, the analysis demonstrates demonstrates compliance compliance with Criterion (4).

Section Section 3.1 of this report describes the postulated LBLOCA event. Section Section 3.2 describes thethe models used in in the analysis. Section 3.3 describes the 2X4-loop PWR plant and summarizes summarizes AREVA NP Inc.

Plant ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant 2 Cycle Cycle 1919 000 Revision 000 Cycle 21 &

Unit 1 Cycle & Unit Unit 2 Realistic Large Break LOCA Summary Summary Report Page 3-2 Page 3-2 the system parameters parameters used in the analysis. Compliance to the SER is addressed addressed in in Section 3.4. Section 3.5 summarizes the results of the RLBLOCA analysis.

3.1 Descriptionof the LBLOCA Event Description A LBLOCA postulated large rupture of the RCS primary piping.

LBLOCA is initiated by a p'ostulated piping. Based on on deterministic studies, the worst break location is in the cold leg piping between deterministic between the reactor reactor coolant pump and the reactor vessel for the RCS loop containing the pressurizer. The break initiates a rapid depressurization depressurization of the RCS. A reactor trip signal is initiated when the low low pressurizer pressure trip setpoint pressurizer setpoint is reached; reached; however, reactor trip is conservatively neglected neglected in in the analysis. The reactor is shut down by coolant voiding in the core.

The plant is assumed assumed to be operating normally at full power power prior to the accident. The cold leg leg break is assumed assumed to open instantaneously. For this break, a rapid depressurizationdepressurization occurs, along with aa core flow stagnation and reversal. This causes causes the fuel rods to experience experience DNB.

Subsequently, the limiting fuel rods are cooled by film convectionconvection to steam. The coolant voiding voiding creates negative reactivity effect and core criticality ends. As heat transfer from the creates a strong negative the.

fuel rods rods is reduced, reduced, the cladding cladding temperature temperature increases.

Coolant in all regions of the RCS begins to flash. At the break break plane, the loss of subcooling in in the coolant results in substantially reduced break flow. flow.* This reduces the depressurization depressurization rate, and leads to a period period of positive core flow or reduced downflow as the RCPs in the intact loops loops continue to supply water to the RV (in (in No-LOOP conditions). Cladding temperatures temperatures may be be reduced and some portions of the core may rewet during during this period. The positive core flow or reduced downflow downflow period ends as two-phase two-phase conditions occur in the RCPs, reducing reducing their their effectiveness.

effectiveness. Once again, the core flow reverses reverses as most of the vessel vessel coolant inventory flows flows out through through the broken cold leg.

Mitigation of the LBLOCA begins when the SIAS is issued. This signal is initiated initiated by either high high containment pressure or low Pressurizer containment Pressurizer pressure.

pressure.* Regulations Regulations .require that a worst worst single-failure be considered. This single-failure single-failure single-failure has been determined to be the loss of one one ECCS pumped pumped injection train. The AREVA RLBLOCA RLBLOCA methodology methodology conservatively conservatively assumes an an on-time start on-time start and and normal normal lineups lineups of of the containment spray to conservatively the containment conservatively reduce containment containment pressure pressure and increase.

increase break break flow. Hence, Hence, the analysis assumes assumes loss of a diesel generator in in AREVA NP Inc.

ANP-2834(NP)

Nuclear Plant Calvert Cliffs Nuclear Plant 21 &

Unit 11 Cycle 21 Unit 22 Cycle

& Unit 19 Cycle 19 000 Revision 000 Realistic Large Break LOCA Summary Summary Report Page 3-3 Page 3-3 which LPSI inject inject into the broken loop and one intact loop, HPSI inject into all four loops, and all containment spray pumps are operating.

containment operating.

When the RCS pressure falls below the SIT pressure, fluid from the SITs is injected into the cold cold In the early delivery of SIT water, high pressure and high break legs. In break flow will drive some of this this fluid to bypass the core. During this bypass period, period, core heat transfer transfer remains poor and fuel rod rod temperatures increase. As RCS and containment cladding temperatures containment pressures equilibrate, ECCS water begins to fill the lower plenum and eventually the lower lower portions of the core; thus, core heat transfer transfer improves and cladding cladding temperatures decrease.

Eventually, the relatively large volume of SIT water is exhausted exhausted and core recovery continues continues relying solely on pumped ECCS injection. injection. As the SITs empty, the nitrogen gas used to to pressurize pressurize the SITs exits through the break. This gas release may result in a short period period ofof improved core heat transfer as the nitrogen gas displaces water in the downcomer. After the the nitrogen nitrogen gas has been expelled, the ECCS temporarily may may not be able to sustain full core core cooling because of the core decay decay heat and the higher temperatures created by higher steam temperatures quenching quenching in the lower portions portions of the core. Peak fuel rod cladding temperatures increase temperatures may increase for a short period until more energy removed from the core by the HPSI and LPSI while the energy is removed the decay decay heat continues continues to fall. Steam generated generated from fuel rod rewet will entrain entrain liquid and pass pass through the core, vessel vessel upper upper plenum, the hot legs, the steam generators, generators, and the reactor reactor coolant coolant pumps before itit is vented out the break. Some steam flow to the upper upper head head and pass pass pressurizer spray nozzles, which provide a vent path to the break. The resistance through the pressurizer resistance of this flow path to the steam flow is balanced balanced by the the* driving force of water filling the the downcomer. This resistance resistance may act to retard the progression progression of the core reflood postpone reflood and postpone core-wide cooling.

cooling. Eventually Eventually (within a few minutes of the accident), the core reflood will sufficiently to ensure core-wide progress sufficiently quench occurs within a few minutes core-wide cooling. Full core quench minutes after after core-wide cooling. Long-term cooling is then sustained sustained with coolant coolant provided provided by LPSI.

Descriptionof Analytical Models 3.2 Description The RLBLOCA methodology is documented documented in EMF-2103 EMF-2103 Realistic Realistic Large Large Break Break LOCA LOGA Methodology (Reference (Reference 1). 1). The methodology methodology follows the Code Scaling, Scaling, Applicability, Applicability, and Uncertainty (CSAU) evaluation approach (Reference Uncertainty (Reference 2). This method outlines an approach for defining and qualifying best-estimate thermal-hydraulic qualifying a best-estimate thermal-hydraulic code and quantifies uncertainties quantifies the uncertainties in a in a LOCA analysis.

AREVA AREVA NP Inc.

Plant Nuclear Plant Calvert Cliffs Nuclear ANP-2834(NP)

ANP-2834(NP) 19 Cycle 19 000 Revision 000 Unit 1 Cycle 21 && Unit 22 Cycle Realistic Large Break Break LOCA Summary Summary Report Page 3-4 Page 3-4 The RLBLOCA RLBLOCA methodology methodology consists of the following computer codes:

RODEX3A RODEX3A for computation of the initial initial fuel stored energy, fission gas release, release, and fuel-cladding gap conductance.

- S-RELAP5 S-RELAP5 for the system calculation (includes (includes ICECON for containment response).

- generation of ranged parameter AUTORLBLOCA for generation parameter values, transient input, transient transient runs, and general general output documentation.

documentation.

The governing two-fluid two-fluid (plus non-condensibles) non-condensibles) model with conservation conservation equations for mass, energy, and momentum transfer transfer is used. The reactor core is modeledmodeled in in S-RELAP5 S-RELAP5 with heat generation generation rates determined determined from reactor kinetics equations (point kinetics) with reactivity feedback, and with actinide actinide and decay heating.

heating.

The two-fluid formulation uses a separate separate set of conservation conservation equations constitutive equations and constitutive relations for each phase. The effects effects of one phase on the other are accounted accounted for by interfacial friction, and heatheat and mass transfer interaction interaction terms in the equations. The conservation conservation equations equations have have the same form for each phase; only the constitutive relations and physical properties properties differ.

modeling of plant components is performed by following guidelines developed to ensure The modeling accurate accurate accounting accounting for physical dimensions dimensions and that the dominant phenomena phenomena expected during expected during the LBLOCA event are captured. The basic building blocks for modeling modeling are hydraulic volumes volumes for fluid paths and heat structures structures for heat transfer. In addition, special purpose components components exist to represent represent specific components components such as the Reps RCPs or the steam steam generator generator separators.

All geometries are modeled modeled at the resolution necess~ry necessary to best resolve resolve the flow field and thethe phenomena being modeled within practical practical computational computational limitations.

System nodalization nodalization details are shown in Figures 3-1 through 3-5. A point of clarification:

clarification: in Figure 3-1,3-1, break modeling uses two junctions regardless regardless of break type-split or guillotine; guillotine; for for guillotine breaks, Junction 151 is deleted, itit is retained fully open for split breaks. Hence, total break area is the sum of the areas areas of both break junctions.

A typical calculation using S-RELAP5 begins with the establishment establishment of a steady-state steady-state initial condition with all loops intact. The input parameters parameters and initial conditions conditions for this steady-state steady-state calculation are chosen to reflect plant technical specifications measured data.

specifications or to match measured AREVA AREVA NP Inc.

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Unit 22 Cycle

&Unit 19 Cycle 19 Revision 000 Revision 000 Unit 1 Cycle 21 &

Realistic Large Break LOCA Summary Report Report Page 3-5 Page 3-5 Additionally, the RODEX3A code provides initial conditions for the S-RELAP5 fuel models.

Specific Specific parameters are discussed discussed in Section 3.3.

Following the establishment establishment of an acceptable acceptable steady-state steady-state condition, the transient transient calculation is initiated initiated by introducing a a break into one of the loops (specifically, the loop with the pressurizer).

pressurizer).

The evolution of the transient through blowdown, blowdown, refill and reflood is computed continuously continuously S-RELAP5. Containment using S~RELAP5. pressure is also calculated by S-RELAP5 Containment pressure S-RELAP5 using containment containment models models derived derived from ICECON (Reference 4), which is based on the CONTEMPT-LT ICECON (Reference CONTEMPT-LT code code (Reference (Reference 3) and has been updated updated for modeling ice condenser containments.

condenser containments.

The methods used in the application of S-RELAP5 S-RELAP5 to the LBLOCA are described described in in

. Reference Reference 1. 1. A detailed assessment assessment of this computer code was made through comparisons to experimental data, many benchmarks with cladding temperatures experimental temperatures ranging from 1 1,700'F

,700'F (or less) to above 2,200'F.

2,200$. These assessments assessments were used to develop d evelop quantitative quantitative estimates of the the ability of the code to predict predict key physical phenomena phenomena in a PWR LBLOCA. Various models-for models-for example, the core heat heat transfer, the decay heat model and the fuel cladding oxidation correlation-are correlation-are defined based on code-to-data comparisons code-to-data comparisons and are, hence, plant independent.

independent.

The RV internals internals are modeled in in detail (Figures 3-3 through 3-5) based on specific inputs inputs supplied by Constellation Energy. Nodes Nodes and connectivity, flow areas, resistancesresistances and heat heat structures are all accurately accurately modeled. The location location of the hot assembly/hot assembly/hot pin(s) is unrestricted; however, the channel is always modeled to restrict appreciable unrestricted; appreciable upper plenum plenum c liCfuid-falioack:

licq iiid-fall ba.c.k*.

k ... ".

The final step of the best-estimate best-estimate methodology methodology is to combine all the uncertainties uncertainties related to the the code and plant parameters, parameters, and estimateestimate the PCT at aa high probability probability level. The steps taken to derive the PCT POT uncertainty uncertainty estimate estimate are summarized summarized below:

1. Base Plant Input File Development Development First, base RODEX3A and S-RELAP5 input files for the plant (including (including the containment containment input file) are developed.

developed. Code input development guidelines development guidelines are applied to ensure ensure that model nodalization nodalization is consistent with the model nodalization nodalization used in the code validation.

validation.

2. Sampled Case DevelopmentDevelopment AREVA NP Inc.

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Calvert Cliffs Nuclear Plant Calvert 2 Cycle

& Unit 2 19 Cycle 19 Revision 000 Revision 000 Unit 1 Cycle 21 &

Realistic Large Break Realistic Large Break LOCA Summary Report LOCA Summary Page 3-6 Page 3-6 The non-parametric non-parametric statistical approach approach requires that many "sampled" cases be created created and processed.

processed. For every set of input created, each each "key LOCA LOCA parameter" is randomlyrandomly sampled over a range established established through code uncertainty assessment or expected (provided by plant technical operating limits (provided technical specifications specifications or data). Those Those parameters parameters considered "key LOCA parameters" are listed in Table 3-1. 3-1. This list includes includes both both parameters related to LOCA phenomena parameters phenomena (based on the PIRT provided in Reference Reference 1) and to plant operating parameters.

3. Determination of Adequacy Determination Adequacy of ECCS The RLBLOCA methodology methodology uses a non-parametric non-parametric statistical statistical approach approach to determine determine values of PCT at the 95 percent probability level. Total oxidation and total hydrogen hydrogen are based on the limiting PCT case. The adequacy adequacy of the ECCS is demonstrated demonstrated when when these results satisfy the criteria set forth in in Section 3.0. '

PlantDescription 3.3 Plant Descriptionand and Summary of Analysis Parameters Parameters The plant analysis analysis presented in this report is for aCE-designed a CE-designed PWR, which has 2X4-loop 2X4-loop arrangement. There are two hot legs each with a U-tube U-tube steam generator and four cold legs legs RCP 1. The RCS includes one Pressurizer connected each with a RCp1. connected to a hot leg. The core containscontains 217 thermal-hydraulic thermal-hydraulic compatible compatible AREVA HTP 14X14 14X14 fuel assemblies with 2, 4, 6 and 8 wlo w/o gadolinia pins. Both units of Calvert Cliffs' core contain co-resident co-resident Westinghouse Westinghouse and AREVA Advanced Advanced CE14 HTP fuel. The two assembly assembly types have have different different form loss coefficients the coefficients for the grid spacers and the upper upper and lower tie plates. Adjustments were made to these losses in the the basedeck to model basedeck model the mixed mixed core configuration.

configuration. The ECCS includes one HPSI, HPSI, one LPSI and one SIT injection path per RCS loop. The break is modeled in the same loop as the pressurizer, pressurizer, as directed by the RLBLOCA methodology. The RLBLOCA transients transients are of sufficiently short duration that the switchover switchover to sump cooling water (i.e.,(Le., RAS) for ECCS pumped injectioninjection, need not be considered The S-RELAP5 S-RELAP5 model explicitly describes describes the RCS, RV, Pressurizer, Pressurizer, and ECCS. The ECCS ECCS includes a SIT path and a LPSI/HPSI includes LPSI/HPSI path per RCS loop. loop. The HPSI and LPSI feed into a common header header that connects to each cold leg pipe downstream downstream of the RCP discharge.

discharge. The The pumped injection is modeled as a ECCS pumped a table of flow versus versus backpressure. This model also describes the secondary-side secondary-side steam generator that is instantaneously instantaneously isolated (closed MSIV MSIV 1 The RCPs are Byron-Jackson Type DFSS pumps are specified by Constellation Energy. The 1 The RCPs are Byron-Jackson Type DFSS pumps are specified by Constellation Energy. The homologous pump homologous pump performance performance curves curves were input to the the S-RELAP5 S-RELAP5 plant model; model; the the built-in built-in S-RELAP5 S-RELAP5 curves were not used.

curves AREVA NP Inc.Inc,

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Plant Nuclear Plant Calvert Cliffs Nuclear 19 Revision 000 000 Unit 1 Cycle 21 & &Unit 22 Cycle 19 Realistic Large Realistic Large Break LOCA Summary Summary Report Page 3-7 Page 3-7 and feedwater trip) at the time of the break. A symmetric symmetric steam generator generator tube plugging level of generator was assumed.

10 percent per steam generator assumed.

described in the AREVA RLBLOCA As described RLBLOCA methodology, many parameters associated many parameters associated with LBLOCA LBLOCA phenomenological uncertainties and plant operation phenomenological uncertainties operation ranges are sampled. A summary of those parameters is given in in Table 3-1.

3-1. The LBLOCA phenomenological phenomenological uncertainties are uncertainties are provided provided in Reference Reference 1. 1. Values for process process or operational operational parameters, including ranges of sampled parameters, and fuel design parameters process parameters, sampled process parameters used in the analysis are given in in Table 3-2. Plant data are analyzed to develop uncertainties uncertainties for the process parameters parameters sampled in the analysis.

sampled Table 3-3 presents presents a summary summary of the uncertainties uncertainties used in the the analysis. Where applicable, the sampled parameter parameter ranges are based on technical specification specification limits or supporting plant plant calculations calculations that provide more bounding values.

For the AREVA NP RLBLOCA RLBLOCA EM, EM, dominant containment parameters, as well as NSSS dominant containment NSSS parameters, parameters, were established established via aa PIRT PIRT process. Other model inputs are generally generally taken as as nominal nominal or conservatively biased. The PIRT outcome yielded yielded two important (relative to PCT)

Containment Containment parameters-containment parameters-containment pressure and temperature. In many instances, the the conservative guidance of CSB 6-2 (Reference conservative guidance 5) was used in setting the remainder (Reference 5) remainder of the the containment model input parameters.

parameters. As noted in Table 3-3, containment temperature is a sampled sampled parameter. Containment Containment pressure response is indirectly ranged by sampling the the containment volume (Table 3-3). Containment heat sink data is given in Table 3-9. In accordance accordance with Reference Reference 1, the condensing condensing heat transfer transfer coefficient coefficient is intended to be closer a *t-e~tirf-at* ihste-d of a "boLihdin~rliigh"*value.

to- a**D"est-estrrifa'te'iiisfead'of b6Jnhding high-value. AlA((- ' "]Uchida Uchida heat heat transfer coefficient coefficient specifically validated multiplier was specifically validated for-use for'use in Calvert Cliffs through application of the process process used in the RLBLOCA RLBLOCA EM (Reference 1) sample sample problems.

The containment initial conditions conditions and boundary boundary conditions are given in Table Table 3-8. The building building spray is modeled at maximum heat removal capacity. All spray flow is delivered delivered to the the containment.

3.4 SER Compliance Compliance requirements on the methodology A number of requirements methodology are stipulated in in the conclusions conclusions section of the the SER for the RLBLOCA RLBLOCA methodology methodology (Reference (Reference 1). These requirements requirements have all been fulfilled fulfilled during the application application of the methodology methodology as addressed addressed in Table 3-4.

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Unit 1 Cycle 21 & Unit Cycle 19 Unit 22 Cycle 19 Realistic Large Break LOCA Summary Realistic Large Break LOCA Summary Report Report Page 3-8 Page 3-8 3.4.1 3.4.1 Item 7:

Item Blowdown Quench 7: Blowdown Quench Three cases Three cases were candidates for potential candidates were potential blowdown quench for blowdown quench and and was inspected. For closely inspected.

was closely For this calculation, no this calculation, evidence of no evidence blowdown quench of blowdown quench was observed. Therefore, was observed. compliance to Therefore, compliance to the the SER restriction has SER restriction has been been demonstrated.

demonstrated.

3.4.2 Item 3.4.2 Item 8: Top-down Quench 8: Top-down Quench provisions have Several provisions Several have been implemented inin the been implemented S-RELAP5 model the S-RELAP5 model to prevent the to prevent the top-down top-down quench.

quench. The upper plenum The upper nodalization features plenum nodalization features include:include:

- the homogenous option the homogenous selected for option isis selected for the junction that the junction connects the that connects first axial the first axial level level node node above the above the hot channel to hot channel to the second axial the second level node axial level above the node above the hothot channel; channel; P no cross-flow is allowed between the first axial level Upper Plenum nodes above the hot

no cross-flow is allowed between the first axial level Upper Plenum nodes above the hot channel to channel the average to the average channel; channel;

- the CCFL model the CCFL model is is applied applied on on all all core exit junctions.

core exit junctions.

Fifteen cases Fifteen cases were examined for closely examined were closely top-down quench.

for top-down quench. No evidence of No evidence top-down quench of top-down quench was observed. Therefore, was observed. compliance to Therefore, compliance to the the SERSER restriction restriction has been demonstrated.

has been demonstrated.

3.5 Realistic Large 3.5 Realistic Break LOCA Large Break LOCA Results Results Two case sets Two case sets of 59 transient of 59 calculations were transient calculations performed sampling were performed sampling the parameters listed the parameters listed inin Table 3-1.

Table 3-1. For For each set, PCT was case set, each case calculated for was calculated for aa U0 rod and U022 rod and for Gadolinia-bearing rods for Gadolinia-bearing rods concentrations of with concentrations with of 2, 2, 4, and 88 w/o 4, 66 and wlo Gd Gd2200 33.* The limiting case The limiting case set, set, that contained the that contained the PCT, PCT, 0 F) occurred in Case 5 for 4 was the set was set with with no offsite available. The power available.

offsite power limiting PCT The limiting peT (1670 (1670°F) occurred in Case 5 for 4 w/o--Gd 2203-rod..The-

-w/o--Gd 0 3 -rod;-The- major -parameters for--the major"parameters limiting -transient for-the -limiting -transient are are presented presented in in Table Table 2-1.2-1.

Table 3-5 lists Table lists the results of the limiting case. The the limiting fraction of The fraction of total total hydrogen generated was hydrogen generated was notnot directly however, it calculated; however, directly calculated; it is conservatively bounded is conservatively bounded by by the calculated total the calculated total percent percent oxidation, which is oxidation, is well well below below the percent limit.

the 11 percent limit. The best-estimate PCT The best-estimate PCT casecase is Case 27, is Case 27, which which corresponded to the median case corresponded case out out of 59-case set the 59-case of the set with with no rio offsite offsite power available. The power available. The nominal PCT PCT was 1435°F for was 14351F for an an 88 w/o wlo Gd Gd 2200 33 rod.

rod. This result can This result can be used to be used quantify the to quantify the conservatism in relative conservatism in the limiting case the limiting result. In case result. analysis, it this analysis, In this it was was 235°F.

235°F.

case results, The case event times results, event times and analysis plots and analysis plots for for the limiting PCT the limiting PCT case case are shown in are shown in Table Table 3-5, Table 3-6, 3-5, Table 3-6, andand in Figure 3-11 in Figure 3-11 through through Figure Figure 3-21. Figure 3-6 3-21. Figure 3-6 shows linear scatter shows linear scatter plots of the plots of the key parameters sampled key parameters sampled for for the the 59 calculations. Parameter 59 calculations. Parameter labels appear to labels appear to the left the left of each of individual plot.

each individual These figures plot. These show the figures show the parameter parameter ranges ranges used used in in the the analysis. Figure analysis. Figure AREVA NP AREVA Inc.

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Unit 1 Cycle 21 & Unit Cycle 19 Unit 22 Cycle 19 000 Revision 000 Realistic Large Realistic Large Break Break LOCA Summary Report Page 3-9 Page 3-9 3-7 and Figure 3-8 show the time of peT PCT and break break size versus PCT peT scatter plots for the 59 59 calculations, respectively. Figure 3-9 and Figure 3-10 show the maximum oxidation and total oxidation versus peTPCT scatter scatter plots for the 59 calculations, calculations, respectively. Key parameters parameters for the the PCT case are shown in Figure 3-11 through Figure 3-21.

limiting peT 3-21. Figure Figure 3-11 is the plot of peT PCT independent of elevation; this figure clearly indicates independent indicates that the transient exhibits a sustained and comparison of peT stable quench. A comparison PCT results from both case sets is shown in Figure Figure 3-22. As As seen in Figure 3-22, the peak PCT peT is from LOOP case.

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& Unit 22 Cycle 19 Revision 000 Revision 000 Unit 1 Cycle 21 &

Unit Cycle 19 Realistic Large Break LOCA Summary Report Realistic Page 3-10 Page 3-10 Table 3-1 Sampled Sampled LBLOCA Parameters LBLOCA Parameters Phenomenological Phenomenological Time in cycle (peaking factors, axial shape, rod properties, burnup) burnup)

Break Break type (guillotine versus split)

Critical flow discharge discharge coefficients coefficients (break)

Decay Decay heatheat discharge coefficients (surgeline)

Critical flow discharge Initial upper head temperature temperature Film boiling heat transfer Dispersed film boiling heat transfer Dispersed transfer Critical heat flux flux T min (intersection Tmin (intersection of film and transition transition boiling)

Initial stored energy energy Downcomer hot wall effects Downcomer Steam generator generator interfacial dragdrag interphase heat transfer Condensation interphase Metal-water reaction Metal-water reaction Plant'1 Plant 2

availability Offsite power availabilitl Break size Pressurizer pressure Pressurizer Pressurizer liquid level Pressurizer SIT pressure SIT liquid level temperature (based on containment SIT temperature containment temperature) temperature)

Containment temperature Containment volumevolume Initial RCS flow rate Initial operating operating RCS temperature Diesel start (for loss of offsite power only)

Uncertainties for plant parameters are based on typical plant-specific data with the exception of Uncertainties for plant parameters are based on typical plant-specific data with the exception of "Offsite power "Offsite power availability," which is a binary result that is specified by the analysis methodology.

availability," which 2

2 Not sampled, see Section 4.9.

Not sampled, see Section 4.9.

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Calvert Cliffs Nuclear Nuclear Plant Revision Revision 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report Realistic 3-11 Page 3-11 Paae Table 3-2 Plant Operating Table Operating Range Supported Supported by the LOCA Analysis Analysis Event Operating Operating Range Range 1.0 1.0 Description Plant Physical Description 1.1 1.1 Fuel a) Cladding outside diameter a) 0.440 in.

b) Cladding inside diameter b) diameter 0.387 in.

c) Cladding thickness 0.0265 in.

d) Pellet outside diameter 0.3805 in.

e) Pellet density density 96 percent of theoretical f) Active fuel length 136.7 in.

Resinter densification g) Resinter densification [I ]

h) Gd concentrations Gd 2 0 33 concentrations 2, 4, 6, 8 wlo 2,4,6,8 w/o 1.2 RCSRCS a) Flow resistance resistance Analysis Analysis b) Pressurizer Pressurizer location Analysis assumes location giving giving most limiting limiting_ PCT (broken loop) loop) c) Hot assembly cl assembly location Anywhere in core An}"!Jhere d) d) Hot assembly type 14X14 AREVA NP HTP fuel e) SG tube plugging <10 percent percent 2.0 Conditions Plant Initial Operating Conditions 2.1 Reactor Reactor Power a) Nominal Nominal reactor power MWt1 2754 MWt b) LHR 15.0 kW/ft 15.0kW/ft c) Fa FQ 1.6385 1.6385 2.2 Fluid Conditions

.2.2 Conditions a) Loop flow 139.5 Mlbm/hr Mlbm/hr <~ M M ~< 159.2 Mlbm/hr Mlbm/hr b) RCS Cold Leg temperature temperature 535.0°F 535.0°F ~< T ~ < 550.0°F 550.0°F Pressurizer pressure c) Pressurizer pressure 2164 psia::;;

psia < P::;;

P < 2336 psia psia d) Pressurizer Pressurizer level 32.2 percent percent ::;;< L::;;

L < 67.2 percent

~ercent e) SIT pressure 194.7 psia ~< P::;;

194.7 P < 264.7 psia psia ft3 < V

1179 ft 3 f) SIT liquid volume 1080 1080 ft3 ::;; V ::;; 1179 ft3 g) g) SIT temperature temperature 125 0 F 60°F ~< T 5~ 125°F containment (It's coupled with containment temperature) temperatu re >-

h) SIT hl SIT resistance resistance fL/D fUD As-built piping configuration configuration i)i) Minimum ECCS boron >2300

~2300 ppm 1 Includes 0.62%

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& Unit 000 Revision 000 19 Realistic Large Break LOCA Summary Report Paae 3-12 Page 3-12 Table 3-2 Plant Operating Operating RangeRange Supported Supported by the LOCA Analysis (Continued) (Continued)

Event Event Operating Range Operating Range 3.0 Accident Boundary Boundary Conditions _

a) Break Break location Cold leg pump discharge piping piping b) Break b) Break type Double-ended guillotine or split Double-endedguillotine split c) Break Break size (each side, relative to cold 0.2869 S;_ A S;_ 1.0 full pipe area (split) leg pipe area of 4.91 ft 2) ft2) 0.2869 S;< A 5S; 1.0 full pipe area (guillotine) d) Worst single-failure' d) single-failure' Loss of aa diesel generator generator e) Offsite power On or Off f) LPSI flow Minimum flow Minimum flow g) HPSI Q) HPSI flow Minimum flow flow h) ECCS pumpedpumped injection injection temperature temperature 100'F 100'F i) i) HPSI pump delay (w/ offsite power) 30.0 (wI 30.0 (w/o offsite power) j) LPSI pump delay 45.0 (wI (w/ offsite power) 45.0 (w/o offsite power)

Containment pressure k) Containment 14.7 psia, nominal value 14.7 value I)I) Containment temperature Containment temperature 125 0 F 60°F <S; T <:::; 125°F Containment spra~s m) Containment sprays del~

delay 20 s AREVA NP Inc.

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Nuclear Plant Calvert Cliffs Nuclear Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Realistic Large Large Break Break LOCA Summary Report Page 3-13 Page 3-13 Statistical Distributions Table 3-3 Statistical Distributions Used Used for Process Parameters11 Process Parameters Operational Parameter Parameter Uncertainty Uncertainty Parameter Range Parameter Range Distribution Distribution Pressurizer Pressurizer Pressure (psia) Uniform 2164 -2336

- 2336 Pressurizer Pressurizer Liquid Level (percent) Uniform Uniform 32.2 -- 67.2 67.2 SIT Liquid Volume (ft3)

Volume (fe) Uniform Uniform 1080.0 - 1179.0 1080.0 1179.0 SIT Pressure (psia)

(psia) Uniform 194.7 - 264.7 194.7 Containment Temperature ('F) Uniform Uniform 60--125 60 125 3

Containment Volume (ft Containment (fe)) Uniform Uniform 1.989E+6 - 2.148E+6 1.989E+6 Initial RCS Flow Rate (Mlbm/hr)

(Mlbm/hr) Uniform 139.5 - 159.2 159.2 Initial RCS Operating Operating Temperature Temperature Uniform 535.0 - 550.0 Uniform 535.0 - 550.0 (TCold) (17)

(Tcold) ('F)

RWST Temperature for ECCS (7F) ('F) Point 100 100 Offsite Availabilitl2 Offsite Power Availability Binary Binarv o1 0,1 Delay for Containment Spray (5) (s) Point 20 20 (s) Point 30 (wI offsite power)

(w/ offaite power) .

LPSI Pump Delay (s) Point offsite power) 30 (w/o offsite Point 45 (wI offsite power)

(w/ offsite power)

HPSI Pump Delay (s) (s) Point 45 (w/o w/o offsite offsite power) 1 Note that core power power is not sampled, see Section 1.0. 1.0.

22 This is no longer a sampled sampled parameter. One set of 59 cases is run with LOOP and one set of 59 59 cases is run with No-LOOP.

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.Page 3-14 Table 3-4 SER SER Conditions Conditions and Limitations Limitations SER Conditions and Limitations Limitations Res~onse

Response

1. A CCFL violation violation warning will be added to alert the analystanalyst There was no significant occurrence occurrence of CCFL violation in the the to CCFL violation in the downcomer downcomer should such occur. downcomer for this analysis. Violations of CCFL were noted downcomer in aa statistically insignificant number of time steps.

statistically insignificant

2. AREVA AREVA NP has agreed that itit is not to use nodalization Hot leg nozzle gaps were not modeled.

m6deled.

with hot leg to down downcomer comer nozzle gaps.

nozzle gg~s.

3. If If AREVA NP applies the RLBLOCA methodology to plants The PLHGR for Calvert Calvert Cliffs Units 1/2 is lower than that using aa higher planar linear linear heat generation rate (PLHGR) heat generation used in the development development of the RLBLOCA (Reference RLBLOCA EM (Reference than used in the current analysis, or ifif the methodology is 1). An end-of-life calculation was not performed; the performed; thus, the to be applied applied to an end-of-life end-of-life analysis analYSis for which the pin need for a a blowdown blowdown cladding rupture model was not not pressure is significantly significantly higher, then the need for a reevaluated.

reevaluated.

blowdown clad rupture rupture model model will be reevaluated.

reevaluated. The The evaluation may be based based on relevant engineering engineering experience experience and should be documented documented in either the the RLBLOCA guideline or plant specific calculation calculation file.

4. Slot breaks on the top of the pipe have have not been evaluated.

evaluated. For the Calvert Cliffs Units, the elevation of the cross-over cross-over These These breaks could cause the loop seals to refill during late piping top (I(ID) piping* D) relative to the cold leg center line is 55 reflood and the core to uncover uncover again. These break inches, and the elevation of the top of the active core core locations locations are an oxidation concern as opposed to aa PCT peT relative to the cold leg center line is -66.925 inches.

concern concern since the top of the core core can remain uncovered uncovered for Therefore, Therefore, no evaluation is required.

required.

extended extended periods periods of time. Should Should an analysis be be performed performed for a plant with loop seals with bottombottom elevations elevations that are below the top elevation elevation of the core, AREVA AREVA NP will evaluate evaluate the effect of the deep loop seal on the slot breaks.

The The evaluation evaluation may may be based on relevant engineering engineering experience experience and should be documented in either the the RLBLOCA guideline or plant-specific plant-specific calculation file.

5. The model applies to 33 and 4 loop Westinghouse- and The plant is aaCE-designed CE-designed 2X4 loop plant.

CE-designed nuclear steam systems.

CE-deslgned

6. The model model applies to bottom reflood plants only (cold side The plant is a bottom reflood plant.

injection into the cold legs at the reactor coolant discharge discharge piping).

7. The model model is valid as long as blowdown quench quench does not The limiting limiting case did not show show any evidence of aa blowdown blowdown occur. IfIf blowdown blowdown quench quench occurs, additional additional justification justification quench.

for the blowdown blowdown heat heat transfer transfer model and uncertainty uncertainty are needed or the calculation calculation is corrected.

corrected. blowdown A blowdown quench is characterized characterized by aa temperature temperature reduction reduction of thethe peak cladding temperature temperature (PCT) node to saturation temperature temllerature during the blowdown blowdown period.

8. The reflood model applies to bottom-up quench behavior. quench initiated at the bottom of the core and Core quench and If a If a top-down top-down quench occurs, the quench occurs, the model model isis to to be be justified or proceeded upward.

proceeded corrected corrected to remove top quench. A top-down quench is characterized characterized by the quench quench front moving from the top to the bottom of the hot assembly.

AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Cliffs Nuclear Nuclear Plant Plant Revision Revision 000 000 Unit 1 Cycle 21 21 && Unit Unit 22 Cycle Cycle 19 Realistic Realistic Large Large Break Break LOCA LOCA Summary Summary Report Page Page 3-15 3-15 Table 3-4 SER Conditions Conditions and Limitations Limitations (Continued)

(Continued)

SER Conditions Conditions and Limitations and limitations Response

Response

9. The The model model does does not determine determine whether whether Long-term cooling Long-term cooling was notnot evaluated evaluated in in this analysis.

analysis.

Criterion 55 of 10 10 CFR 50.46, long term term cooling, cooling, has has been satisfied.

satisfied. This will be will be determined determined by by each each applicant applicant or licensee licensee as as part of its application application of this methodoloQY.

methodology.

guidelines must be

10. Specific guidelines be used to develop develop The nodalization nodalization inin the plant plant model is is consistent consistent with the the CE-designed CE-designed plant-specific nodalization.

the plant-specific Deviations nodalization. Deviations 2X4 loop sample calculation calculation that was submitted submitted to the NRC NRC for review.

reference plant must be addressed.

from the reference from Figure 3-1 Figure 3-1 shows the loop noding used in this analysis. (Note (Note only Loop Loop 1 is shown shown in the figure; figure; Loops 22 and 33 are identical to loop 1, except except that that only Loop 1 contains contains thethe pressurizer pressurizer and and the break.)

break.) Figure Figure 3-2 shows the steam steam generator model.

model. Figures 3-3, 3-3, 3-4, and 3-5 show the reactor reactor vessel vessel noding noding diagrams.

11. A table that contains

11. contains the plant-specific plant-specific Simulation of clad temperature response clad temperature response is aa function of parameters and and the the range range of the values phenomenological phenomenological correlations correlations that that have been derived either analytically analytically considered considered for the selected selected parameter during during or experimentally.

experimentally. The important correlations have have been validated for the the the topical report report approval process process must be RLBLOCA methodology RLBLOCA methodology and a statement statement of the range of applicability applicability has provided. plant-specific parameters provided. When plant-specific parameters been documented.

documented. The correlations of interest are the set of heat transfer transfer are outside the range used in demonstrating demonstrating described in Reference correlations as described Reference 1. Table 3-7 presents presents the the acceptable code acceptable code performance, the licensee licensee or summary of the full range of applicability applicability for the important heat transfer transfer applicant applicant will submit sensitivity sensitivity studies to correlations, as well as the ranges ranges calculated calculated in the limiting case of this this show show the effects effects of that deviation.

deviation, analysis. Calculated values for other parametersparameters of interest are also provided. As is evident, the plant-specific plant-specific parameters parameters fall within the the methodology's range of applicability.

methodology's

12. The licensee licensee or applicant using the approved approved Analysis results are are discussed discussed in Section 3.5.

methodology must submit the results of the methodology the plant-specific analyses, including the the calculated worst break size, PCT, and local calculated and total oxidation. .-.....

13. The licensee or applicant wishing to apply The plant will request an exemption exemption for its operating with M5 clad fuel.

AREVA NP realistic large break loss-of-coolant accident (RLBLOCA) methodology methodology to MS clad fuel must request an exemption M5 exemption for its use until the planned rulemaking to modify 10 CFR 50.46(a)(i) to include M5 cladding cladding

.material has been completed.

.material AREVA NP AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Plant Revision 000 Revision 000 Unit 1 Cycle 21 &Unit

&'Unit 2 Cycle 19 Realistic Realistic Large Break LOCA Summary Summary Report Page 3-16 Paqe 3-16 Table 3-5 Summary of Results for the Limiting PCT Case Case Case # 5 Case Rank 1 PCT Temperature Temperature 1670°F 1670°F Time 35.0 s 35.0 Elevation Elevation 9.430 ft Metal-Water Reaction Metal-Water Reaction Percent Percent Oxidation Oxidation Maximum 0.907%

Percent Percent Total Oxidation 0.011%

0.011%

Table 3-6 Calculated Event Times for the Limiting PCT Case Case Event Time (s)(s)

Break Break Opened 0.0 0.0 RCP Trip N/A N/A SIAS Issued 0.6 0.6 Start of Broken Loop SIT-Injection SIT.Injection 16.4 16.4 Start of Intact Loop SIT Injection Injection 18.8, 18.8 and 18.8 18.8,18.8 and 18.8 (Loops 2, 3 and 4 respectively)

Broken Loop Delivery Bellan Loo~ HPSI Delive_ry_ Began 30.6 30.6 Intact Loop HPSI Delivery Delivery Began Began 30.6, 30.6 and 30.6 30.6, 30.6 and 30.6 (Loops 2, 3 and 4 respectively)

Beginning of Core Recovery Recovery (Beginning of Reflood)

Reflood) 31.0 31.0 PCT Occurred 35.0 35.0 Broken Loop LPSI Delivery Began 45.6 45.6 Intact Loop LPSI Delivery Began Began 45.6, N/A and N/A (Loops 2, 33 and 4 respectively) respectively)

Intact Loop SITs Emptied 70.8, 68.1 and 68.6 70.8, 68.1 and 68.6 (Loops 2, 3 and 4 respectively)

JLoo~s respectively)

Broken Loop SIT Emptied 72.1 Transient Transient Calculation Terminated Calculation Terminated 346.9 346.9

. AREVA AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Cliffs Nuclear Nuclear Plant Plant Revision Revision 000 000 Unit 11 Cycle Cycle 21 && Unit Unit 22 Cycle 19 Realistic Realistic Large Large Break Break LOCA LOCA Summary Summary Report Report Pace Page 3-17 3-17 Table Table 3-7 3-7 Heat Heat Transfer Transfer Parameters Parameters for the Limiting Limiting Case Case AREVA NP AREVA NP Inc.

Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Revision 000 Unit 1 Cycle 21 &

& Unit 2 Cycle Cycle1919 Realistic Large Break LOCA LOCA Summary Summary Report Paoe 3-18 Page 3-18 Table 3-8 Containment Containment Initial and Boundary Conditions and Boundary Conditions 3

Containment Net Free Containment Free Volume (ft )

(fe) 1,989,000 1,989,000 - 2,148,090 Conditions Initial Conditions Containment Pressure (nominal)

Containment psia 14.7 psia Containment Temperature Temperature 125 0 F 60°F - 125°F Temperature Outside Temperature 10°F 10°F Humidity Humidity_ 0.9 0.9 Containment ~pray_Spray Number of Pumps operating operating 2 Spray Flow Rate (Total, both pumps) 4,600 gpm Minimum Spray Temperature Temperature 40°F Post~LOCA initiation of spray Fastest Post-LOCA 20 s Initial Time for:

for:

(minimum) a) Spray Flow (minimum) a).

a) 20 sec Fans (minimum) b) Fans(minimum) b) 0o sec AREVA AREVA NP Inc.

ANP-2834(NP)

- Calvert Cliffs Nuclear Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Realistic Large Break Break LOCA Summary Report Page 3-193-19 Table 3-93*9 Passive Heat Heat Sinks in Containment Containment Slab Material Thick. 2 Description Slab Material Material (ft)Thick. Area (ft )

(fe)

Material (tt)

Paint 2.50E-04 Shell and Dome C Steel 2.08E-02 2.0BE-02 73230 Concrete 3.00E+00 3.00E+00 Unlined Unlined Concrete Concrete Concrete 4.OOE+00 4.00E+00 53000 Zinc 3.17E-04 Galvanized Steel Zinc 3.17E-04 100800 100BOO C. Steel C. 8.33E-03 B.33E-03 Paint 2.50E-04 Painted Thin Steel Paint 2.50E-04 70250 C. Steel C. 2.07E-02 2.07E-02 Paint 2.50E-04 Painted Steel Painted Steel C.Paint Steel 2.50E-04 5.25E-02 55000 55000 C. Steel 5.25E-02 Painted Thick Steel Paint Paint 2.50E-04 2.50E-04 2966 Painted Thick Steel 2966 C. Steel C. 2.01E-01 Paint Paint 2.50E-04 2.50E-04 Containment Penetration Area C. Steel 6.25E-02 3000 Concrete 3.75E+00 S. Steel S. Steel 11.56E-02

.56E-02 Stainless Steel Lined Concrete Cocrte Concrete 1.56E+02 4.00E+00 7925 Concrete 4.00E+00 Paint 2.50E-04 2.50E-04 Containment Liner Plate Stiffeners Stiffeners C. Steel 6.67E-01 4000 Concrete 2.OOE+00 2.00E+00 Base Slab Concrete 8.OOE+00 B.OOE+OO 13300 13300 Sump Strainer 1 S. Steel 1.31E-02 1.31 E-02 308.774 30B.774 Sump Strainer Sump Strainer 2 S. Steel 1.97E-02 1.97E-02 161.338 161.33B Sump Strainer 3 S. Steel 9.83E-03 9.B3E-03 3 Sump Strainer Sump Strainer 4 S. Steel 4.08E-03 4.0BE-03 3433.5' 3433.5*

Additional HIS Additional H/S 11 C. Steel 1.OOE-02 1.00E-02 193.05 193.05 Paint 2,50E-04 2.50E-04 Additional Additional HIS 2 C.Paint Steel 2.08E-02 42.79 42.79 C. Steel 2.0BE-02 2.50E-04 2.50E-04 Additional H/S 3 Additional HIS C.Paint Steel 4.17E-02 56.54 56.54 C. Steel 4.17E-02 Improvised Improvised HISH/S S. Steel S. 8.33E-02 B.33E-02 10000 Volumetric Heat 3 Capacity Volumetric Heat Capacity Material Properties Material Thermal Conductivity Conductivity (BTUIhr-ft-°F)

(BTU/hr.tt-OF) (BTUIft .OF)

(BTU/tt3 .oF)

Concrete 2.5 35 Carbon Steel Carbon 35 55 55 Stainless Steel Stainless Steel 10 10 62 62 Zinc 70 45 Paint 1.5 1.5 32 32 AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Plant 000 Revision 000 Unit 1 Cycle 21 & Unit 2 Cycle 19 Unit Realistic Large Break LOCA Summary Report 3-20 Paqe 3-20 Page Figure 3-1 Primary System Noding Noding AREVA NP Inc.

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Calvert Cliffs Nuclear Nuclear Plant Revision 000 Revision 000 Unit 1 Cycle 21 && Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report Page 3-21 3-2 Secondary Figure 3*2 Secondary System Noding Noding

\

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Inc.

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Calvert Calvert Cliffs Nuclear Nuclear Plant 000 Revision 000 Unit 1 Cycle 21 && Unit 2 Cycle 19 Realistic Large Break Break LOCA Summary Summary Report Report Page 3-22 3-22 Reactor Vessel Noding Figure 3-3 Reactor Noding AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Unit 1I Cycle 21 & Revision 000 000

& Unit 2 Cycle 19 Realistic Large Large Break LOCA Summary Summary Report Paqe 3-23 Page 3-23 Figure 3-4 Core Noding Detail AREVA NP Inc.

AREVA

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Nuclear Plant Calvert Cliffs Nuclear Revision 000 Unit 1 Cycle 21 && Unit 22 Cycle 19 Realistic Large Break Realistic Break LOCA Summary Report Page 3-24 3-24 Figure 3-5 Upper Plenum Plenum Noding Noding Detail AREVA AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Revision 000 000 Unit 1 Cycle 21 && Unit 2 Cycle 19 19 Large Break LOCA Summary Report Realistic Large Page 3-25 Page 3-25 One-Sided Break Area (ft 2

(ft'/side)

Area

/side) I -..~ -~-... *~-*l 10 1~

cOO..

2.0 2~

emm mmcl 3.0 3~

emo__5.0 4!0 4.0 50 5~

-~-~-:

Burn Time Time (hours) E .. :..

0.0 5000.0 10000.0 10000.0 15000.0 15000.0 1

20000.0 Core Core Power Power (MW)

(MW) t: : : : ' : ' ~ : : : 1j 2752.0 I

2752.5 2753.0 2753.5 2753.5 5

2754.0 r

2754.5 2754.5 2755.0 LHGR (KW/ft)

(KW/ft) I.. ~-.:,.. ~.~.: ]1

.i0e 0001 0009 640me 0m 13.0 1 3.0 14.0 14.0 15.0 15.0 16.0 16.0 ASI ASI

! :-.. .. :* -~ .. : 1

-(0.2

-0.2 -0.1

-0.1 0!0 0.0 0.1 0.2 Pressurizer Pressurizer Pressure Pressure (psia) ! -- ~ _. ~---~ .*. 1 2150.0 2150.0 n =W 2200.0 me mOmM 2250.0 m mMW 2300.0 emO 2350.0 Pressurizer Pressurizer Liquid Level

(%)

Level

! ._. ~-..~- .'_ .. 1 L.

30.0 30.0 em 40.0 e mne am 50.0 50.0 ml m W 60.0 O0 70.0 IRCS (Tcold)

Res Temperature Tem~o~ature

~,,'d)

( 0F) r ' : : _. ~- ---3

-w 1 Omemtl II fO 0 aemO 4

535.0 540.0 545.0 545.0 550.0 Figure 3-6 Scatter Scatter Plot of Operational Operational Parameters Parameters

\

AREVA NP Inc.

Calvert ANP-2834(NP)

ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Revision 000 Unit 1 Cycle 21 & & Unit Unit 22 Cycle 19 Realistic Large Break LOCA Summary Summary Report Page 3-26 3-26 Total Loop Flow Flow (Mlb/hr)

~ :::~-: :":**~:*~-l

_e.:_. :..... :.~..~...l 130.'0 130.0 135.0 135.0 140.0 145.0 150.0 155.0 160.0 L

~

SIT Liquid Volume Volume 4"sos SUMa

so0oese m (ft3)

(fe) 1080. .0 1080.0 1100.0 1120.0 1120.0 1140.0 1160.0 1180.0 1180.0

~ . : ~ ..'-.:-- ~-:- :1 SIT -

Pressure Pressure I I I (psia)

'-:..~.. ~...--~-.:..,

180.C0 180.0 200.0 220.0 240.0 260.0 260.0 280.0

~

Containment Containment -

Volume.

(fl3) 1.95e+ 06 1.95e+06 2.00e+06 2.00e+06 2.05e+06 2.05e+06 2.10e+06 2.1Oe+06 2.15e+06 2.15e+06 Tem~~~w" ~_.:~.- -~--~

SIT Temperature (OF) 1 60.0 80.0 100.0 120.0 140.0 Figure Figure 3-6 Scatter Plot of Operational Operational Parameters (Continued)(Continued)

AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Cliffs Nuclear Nuclear Plant Plant Unit Revision Revision 000 000 Unit 1 Cycle Cycle 21 & Unit 2 CycleCycle 19 Realistic Realistic Large Large Break Break LOCA Summary Summary Report Page Page 3-27 3-27 PCT vs Time Time of PCT 2000 1800 I E5

,EI 1600 1400 Ii oG:'

0 t='

CL I-U 1200 D_

()

13-c...

o omE 1000 -0 1000 - ° 800 800 ° U Split Break

_ Split Break I El Guillotine Break]

I o Guillotine Breakl 600 1 400 400 L -_ _~_ _L -_ _~_ _L -_ _~_ _L -_ _~_ _L -_ _~_ _

o0 100 100 200 300 400 500 (s)

Time of PCT (s)

Figure 3*7 3-7 PCT versus PCT Time Scatter Plot from 59 Calculations Calculations AREVA NP Inc.

AREVA

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Nuclear Plant Revision 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break Realistic Break LOCA Summary Report Page 3-28 3-28 peT PCT vs One-sidedOne-sided Break Area Area 2000 .---~---,----~---,----~---.----~---.

1800 1800 o 0 1600 1600

EE 0

. DO DO El0 0 0 "

-- IItJ ** *~.i1 1400 1400

O*

.ME IDI 0 Li:'

0 Fl]

0 0El *

0 % D j:' 1200 1200 0 (j

a..

0 0

0 0

.0 1000 0 800 0 600 *MSplit Break Break I I oECGuillotine Break Breakl 400

+UU ,l ~--~----~--~--~~--~--~----~--~

1.0 2.0 2.0 3.0 3.0 4.0 4.0 5.0 5.0 (ft2/side)

Break Area (fe/side)

Figure 3-8 PCT versusBreak versus "Break Size Scatter Scatter Plot from 59 Calculations Calculations AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Calvert Plant Revision Revision 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Summary Report Paqe 3-29 Page 3-29 Maximum Oxidation vs peT Maximum PCT 1.0 U* Split Break Break I o El 0.9 0.9 I ]o Guillotine

- Break Guillotine Breakl 0.8 F 0.7 k 0.6 I C

c

20ro 0.5 I oEl

'C

'xX o0

oEl 0

0.4 0.4 0.3 I ~

.~~

OJ El~

~iJ 0.2 k 0.1

EloV.***-.

0.0 0.0

~

-.--r

~

400 600 800 1000 1200 1400 1200 1400 1600 1800 1800 2000 0

PCT ((F)F)

Maximum Oxidation Figure 3-9 Maximum Oxidation versus PCT Scatter PlotPlot Calculations from 59 Calculations AREVA NP Inc.

ANP-2834(NP)

Calvert Nuclear Plant Calvert Cliffs Nuclear Plant Revision 000 Unit Unit 1 Cycle 21 & & Unit 22 Cycle 19 Realistic Large Break LOCA Summary Realistic Summary Report Pa~qe 3-30 Page 3-30 Total Oxidation Oxidation vs peT PCT 0.10

Split Break D Guillotine Break 0.08

-::,g e.....

ýO0 c::

0.06 0CU

+='

Ctl "0

x 00 0.04 0.04 D

0.02 0.02 D

o. 00 '--~-'-~-t'j---~-t:'III:t::I'::r--=i=F-~-I=l' 0.00 L 400 600 800 1000 1200 1400 1600 1800 2000 PCT (F)

('F)

Figure 3-10 Total Oxidation versus PCT Scatter Plot from 59 59 Calculations Calculations AREVA NP Inc.

ANP-2834(NP)

Cliffs Nuclear Calvert Cliffs Calvert Nuclear Plant Plant Revision 000 Revision 000 Cycle 21 Unit 1 Cycle Unit Unit 2 Cycle 21 & Unit Cycle 19 Realistic Large Realistic Large Break Summary Report LOCA Summary Break LOCA Page Page 3-31 3-31 PCT Trace for Case Case #5

1669.7 OF, PCT = 1669.7

at Time = 34.98 of, at 5, on 4% Gad 34.98 s, Gad Rod Rod 2000 2000 .---~---r--~---.---~---,---~---,

1500 1500 00 u::-

~

- ::J IL-M2 a>

CL a.

II E 1000 o a> 1000 I-c:

'0 a..Cn

.c:

(D

(/)

-c a>

~

500 k 0

0 100 200 300 300 400 (s)

Time (s)

ID:55570 25Jun2009 17:30:03 R50MX 10:55570 R5DMX Temperature (Independent of Elevation)

Figure 3-11 Peak Cladding Temperature for the Limiting Case AREVA NP Inc Inc....

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Plant Revision 000 Revision 000 I Cycle 21 &

Unit 1 & Unit Unit 2 Cycle 19 Realistic Realistic Large Break LOCA Summary Summary Report Page 3-32 Page 3-32 Break Flow Break 80

- - Vessel Side Side

- - Pump Side Side

- - - Total 60 60 0

.0 CIl E

40 40 Q) 4-L

~0 ro a:::

3:

0 u:

LL 20 20 o01 L

-200L---~---1~0-0--~----20LO----~--3~0~0--~--~400

-20 0 100 200 300 400 Time (s)

ID:55570 25Jun2009 25Jun2009 17:30:03 R5DMX R5DMX Figure 3-12 Break Flow for the Limiting Case Case AREVA NP Inc. Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Realistic Large Break LOCALOCA Summary Summary Report Page 3-33 Paae Core Inlet Mass Flux Mass Flux 1000 1000

-- Hot Assembly

Surround Assembl~

- - - Average Core

- -- Outer Core 500 500 en I

N

~

E

-c u::(n

l en en (II

~

0

,I

-500

-500 L -_ _~_ _ _ _L -_ _~_ _ _ _L -_ _~_ _ _ _L -_ _~_ _~

o0 100 200 300 300 400 (s)

Time (5)

ID:55570 25Jun2009 10:55570 25Jun2009 17:30:03 17:30:03 R5DMX R50MX Figure 3-13 Core Inlet Mass Flux for the Limiting Case Ca~e AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Nuclear Plant Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break Break LOCA Summary Report Page Pa~qe 3-34 Core Outlet Mass FluxFlux 900 900 -- Hot Assembly

Surround Assembl\

- - - Average Core

- -- Outer Core 700 500 (j)

I N

E 300 g

X

J u:::

(/)

cc, 100 C\l 100

~

-100

-300

-500

-500 L -_ _~_ _ _ _L -_ _~_ _ _ _~_ _~_ _ _ _~_ _ _ _~_ _~

o0 100 100 200 300 300 400 400 Time (s)

ID:55570 10:55570 25Jun2009 17:30:03 R50MX 25Jun2009 17:30:03 R5DMX Figure 3-14 Core Outlet Mass Flux for the Limiting Case AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Summary Report Paae 3-35 Page 3-35 Pump Void Fraction 1.0 IIV 0.8 0.6 0.6 cC o0C,

~

L1.

U-

-c 0

0 0.4 0.4

-___ - Broken Loop 1 Broken

- - Intact Intact Loop 2

- -- - Intact Intact Loop 3 0.2 0.2 - -- Intact Loop 4 Intact 0.0 0.0 "----~--'---~--'---~--'---~--

o0 100 100 200 200 300 300 400 400 Time (s)

ID:55570 25Jun2009 10:55570 25Jun2009 17:30:03 17:30:03 R50MX R5DMX Figure 3-15 3*15 Void Fraction Fraction at RCS Pumps for the Limiting Case AREVA AREVA NP Inc.

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Calvert Cliffs Nuclear Nuclear Plant Revision 000 000 Unit 1 Cycle Cycle 21 & & Unit 2 Cycle 19 Realistic Realistic Large Large Break LOCA Summary Report Page 3-36 Pacie 3-36 ECCS Flows Flows 2000 ~--~----~--~----~--~--~----~--~

- - Loop 1 (broken)

- - - - Loop 2

~ ~ - Loop 3 Loop 4 1500 1500 U"

'E

.$ 1000 1000 CIl c:::

=o0 u: 50 500 500 0

0 100 200 300 400 400 Time (s)

ID:55570 25Jun2009 17:30:03 10:55570 R5DMX 17:30:03 R50MX Figure Figure 3-16 ECCS Flows (Includes SIT, LPSI and HPSI) for the the Limiting Case Case AREVA AREVA NP Inc.

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Calvert Nuclear Plant Calvert Cliffs Nuclear Plant Revision 000 Revision 000 Cycle 21 &

Unit 1 Cycle Unit Cycle 19

& Unit 22 Cycle Realistic Large Realistic Summary Report Break LOCA Summary Large Break Paae Page 3-37 3-37 Upper Plenum Pressure Upper Plenum Pressure 3000 ,---~----r---~----r---~----r---~---,

2000 2000 ro

'w S.R,

~

J (J)

(J)

~

CL Cl..

1000 1000

\

0O~==~~==~===C======~==~~

o0 100 100 200 300 400 400 (s)

Time (s)

ID:55570 25Jun2009 17:30:03 17:30:03 R5DMX Case Limiting Case Figure 3-17 Upper Plenum Pressure for the Limiting AREVA NP NP Inc, Inc.

ANP-2834(NP)

Nuclear Plant Calvert Cliffs Nuclear Revision 000 000 Unit 1 Cycle 21 & & Unit Unit 2 Cycle 19 Realistic Realistic Large Large Break LOCA Summary Report Page 3-38 Downcomer Liquid Level Downcomer 30 30 r---~----'---~----,---~-----,----~--~

- - Sector 1 (broken)

Sector 2

, - - - - Sector 3 I - - - Sector 4

I

- - Average f':

, I 20 20 g

Q)

-J Q)

-.J "0

':5 0-

.J 10 o0 ~~~ ____ L -_ _~_ _ _ _~_ _~_ _ _ _~_ _ _ _~_ _~

o0 100 200 300 400 Time (s)

ID:55570 25Jun2009 17:30:03 10:55570 25Jun2009 17:30:03 R5DMX R50MX Figure 3-18 Collapsed Collapsed Liquid Level in the Downcomer Downcomer for the Limiting Case Limiting Case AREVA NP Inc, Inc.

ANP-2834(NP)

Calvert Nuclear Plant Calvert Cliffs Nuclear Revision Revision 000 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Large Break LOCA Summary Realistic Large Summary Report Paae Page 3-39 3-39 Lower Vessel Liquid Level 14 12 10 g *=88

~ CD

.-J

.....J 6

44 22 0 100 200 300 400 Time (s)

ID:55570 ID:55570 25Jun2009 17:30:0317:30:03 R5DMX R5DMX Figure 3-19 Collapsed Collapsed Liquid Level Level in the Lower Plenum for the Limiting Limiting Case Case AREVA NP Inc.

ANP-2834(NP)

Calvert Calvert Cliffs Cliffs Nuclear Nuclear PlantPlant Revision Revision 000 000 Unit 11 Cycle 21 Unit 21 & Unit Unit 22 Cycle Cycle 19 19 Realistic Large Realistic Large Break Break LOCA LOCA Summary Summary Report Report Paqe Page 3-40 3-40 Core Liquid Liquid Level 15 ~--~--~----~---.----~---.--------,

-- Hot Assembly

Center Core

- - - Average Core

- -- Outer Core 10 g

Q) 2>

Q)

.-J

-I "0 I

'5

.-o 0- (I

J 55 I-

, 'I I,

" I I '

, AI, I

1 o0 \\ 1.1 o0 100 200 300 400 Time (s) 10:55570 ID:55570 25Jun2009 17:30:03 17:30:03 R50MX R5DMX Figure 3-20 Collapsed Liquid Level in the Core Figure for the Limiting Case Case AREVA NP Inc,Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 22 Cycle 19 Realistic Large Break LOCA Summary Summary Report Report Paae 3-41 Page Containment Containment and Loop Pressures Pressures 100

-- Containment 90 90 ---- SG Outlet (primary side)

- - - Upper Plenum

- -- Oowncomer Inlet 80 80 70 70 60 60 ro CI,

'iii 0.

~

a)

Q) 50 L-

J 50 U,

CJ)

U)

CJ) a)

Q)

L-a..

0~

40 40 30 20 10 0

0 100 100 200 300 400 Time (s)

ID:55570 25Jun2009 10:55570 25Jun2009 17:30:03 R5DMX 17:30:03 R50MX Figure 3-21 3-21 Containment and Loop Pressures for the Limiting Case Case AREVA AREVA NP Inc.Inc,

Nuclear Plant ANP-2834(NP)

Calvert Cliffs Nuclear Unit 1 Cycle 21 & 19 Cycle 19

& Unit 2 Cycle Revision 000 000 Realistic Large Break LOCA Summary Report Report Paqe 3-42 Page 3-42 1800 1800 1800 1800

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Calvert Cliffs Nuclear Plant Calvert Unit 1 Cycle 21 & 2 Cycle 19

& Unit 2 19 Revision Revision 000 Realistic Large Break LOCA Summary Summary Report Page 4-1 Page 4-1 4.0 Generic Generic Support for Transition Package Package The following sections sections are responses to typical typical RAI questions posed by the NRC on EMF-2103 EMF-2103 Revision Revision 00 plant applications. In some In instances, these requests cross-referenced cross-referenced documentation provided documentation provided on dockets other than those for which the request is made. AREVA discussed these and similar questions discussed questions from the NRC draft SER for RevisionRevision 1 of EMF-2103 EMF-21 03 in in a meeting with the NRC on December 12, meeting 12, 2007. AREVA agreed to provide the following following additional information within new submittals of a Realistic Realistic Large Large Break LOCALOCA report.

4.1 Reactor Reactor Power Question:

Question: It is indicated indicated in the RLBLOCA analyses analyses that that the assumed assumed reactor reactor core power "includes uncertainties."

"includes uncertainties." The use of a reactor power assumption reactor power assumption other other than than 102 102 percent, percent, regardless of BE or Appendix K methodology, regardless methodology, is permitted permitted by Title Title 10 of the Code of Federal Federal Regulations Regulations (10 CFR) CFR),, Part 50, Appendix K.IA Part 50, K.L.A, "Required "Required and Acceptable Features Features of TheThe Evaluation Models, Evaluation Models, 'Sources

'Sources of Heat Heat During During a LOCA." However, Appendix K.l.A LOCA." However, also states:

K.I.A also states: ""...

An assumed assumed power level lower lower than than the level specified specified in this this paragraph paragraph[1.02

[1.02 times the licensed power level], (but not less than levelj, (but than the licensed level) may be used provided licensed power level) provided.. ... ."" Please Please explain.

explain.

Response: As indicated in Item 2.1 of Table 3-2 herein, the assumed reactor core power for

Response

the Calvert Calvert Cliffs Units 1 and 2 Realistic Large Break Loss-of-coolant Loss-of-coolant Accident is 2754 MWt.

This value represents represents the plant rated thermal power power of 2737 MWt with a maximum power measurement uncertainty of 0.62% added to the rated thermal power measurement uncertainty power....

4.2 Rod Quench Quench Question:

Question: Does the version version of S-RELAP5 used to perform perform the computer computer runs runs assure that the assure that void fraction fraction is less than than 95 percent percent and the fuel cladding cladding temperature temperature is less less than 9007F before than 900F it allows allows rod rod quench?

Response

Response: Yes, the version of S-RELAP5 employedemployed for the Calvert Cliffs Units 1 and 2 requires requires that both the void fraction is less than 0.95 and the clad temperature temperature is less than the the minimum minimum temperature temperature for film boiling heat transfer (T (Tmin) min) before before the rod is allowed to quench.

quench.

min is a sampled Tmin T parameter in the RLBLOCA sampled parameter RLBLOCA methodology that typically does does not exceed 755 0 K exceed 755°K AREVA NP Inc.

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Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 && Unit 2 Cycle 19 19 Revision 000 Revision 000 Realistic Large Realistic Large Break LOCA Summary Report Break LOCA Summary Page 4-2 Page 4-2 (900°F). This*

(900'F). This is aa change to the approved RLBLOCA EM (Reference 1). This feature feature is carried carried UAPR09 version of S-RELAP5.

forward into the UAPR09 4.3 Rod-to-Rod Thermal Thermal Radiation Radiation Question: Provide Question: justification that Providejustification that the S-RELAP5 rod-to-rod rod-to-rod thermal thermal radiation applies to radiationmodel applies the Calvert Units 1 Calvert Cliffs Units I and 2 core.

core.

Response: The Realistic LBLOCA methodology, (Reference Response: (Reference 1), does not provide modelingmodeling of of rod-to-rod rod-to-rod radiation. The fuel rod surface heat transfer processes included in the solution at high processes included high temperatures are: film boiling, convection convection to steam, rod to liquid radiation and rod to vapor radiation.

radiation. This heat transfer transfer package package was assessed against various experimental experimental data sets involving involving both moderate (1600'F - 2000'F) and high (2000'F to over 2200'F) 2200'F) peak peak cladding temperatures and shown to be conservative conservative when applied nominally.

nominally. The normal distribution of experimental data was then determined.

the experimental determined. During the execution execution of an RLBLOCA RLBLOCA evaluation, evaluation, the heat transferred transferred from a fuel rod is determined determined by the application application of aa multiplier to the the nominal heat transfer transfer model.

model. This multiplier multiplier is determined determined by a random sampling of the normal distribution of the experimental experimental data benchmarked.

benchmarked. Because the data include the effects of rod- rod-to-rod radiation, it is reasonable reasonable to conclude conclude that the modeling implicitly implicitly includes an allocation allocation for rod-to-rod radiation radiation effects. As will be demonstrated, demonstrated, the approach is reasonable because reasonable because the conditions within actual limiting fuel assemblies assure that the actual rod-to-rod rod-to-rod radiation is allocation provided through normalization larger than the allocation experiments.

normalization to the experiments.

FLECHT-SEASET tests evaluated covered aa range of PCTs from 1,651 The FLECHT-SEASET 1,651 to 2,239cF 2,239'F and thethe THTF tests covered a range of PCTs from 1,000 1,000 to 2,200'F.

2,200'F. Since the test bundle in either either FLECHT-SEASET or THTF is surrounded FLECHT-SEASET surrounded by a test vessel, which is relatively relatively cool compared compared to the heater rods, substantial radiation radiation from the periphery periphery rods to the vessel wall can occur. The The rods selected selected for assessing the RLBLOCARLBLOCA reflood heat transfer package were chosen transfer package chosen from the the interior of the test assemblies assemblies to minimize the impact of radiation heat heat transfer transfer to the test vessel.

The result was that the assessment rods comprise comprise aa set which which is primarily isolated from cold wall effects by being surrounded surrounded by powered powered rods at reasonably reasonably high temperatures.

temperatures.

As a final assessment, three benchmarks independent three benchmarks independent of THTF and FLECHT FLECHT-SEASET

-SEASET were benchmarks were selected performed. These benchmarks selected from the Cylindrical Cylindrical Core Test Facility (CCTF),

LOFT, and the Semiscale Semiscale facilities. Because these facilities are more integral Because integral tests and AREVA NP Inc.

Calvert Cliffs Nuclear Plant Calvert ANP-2834(NP)

Unit 1 Cycle 21 & & Unit 22 Cycle 19 Revision 000 Realistic Realistic Large Large Break LOCA Summary Report Report Page 4-3 Page 4-3 together cover aa wide range of scale, they also serve to show that scale effects together effects are accommodated within the code calculations.

accommodated The results of these calculations calculations are provided provided in Section 4.3.4, Evaluation of Code Biases, page page Reference 1. The CCTF results are shown in Figures 4.180 through 4.192, the LOFT 4-100, of Reference results in Figures 4.193 through 4.201, 4.201, and the Semiscale' Semiscale results in Figures 4.202 through through 4.207. As expected, these figures demonstrate demonstrate that the comparison between between the code code calculations and data is improved calculations improved with the application application of the derived biases. The CCTF, LOFT, and Semiscale Semiscale benchmarks further indicateindicate that, whatever whatever consideration consideration of rod-to-rodradiation rod-to-rod radiation is implicit implicit in the S-RELAP5 S-RELAP5 reflood heat transfer transfer modeling, modeling, it does not significantly effect codecode predictions under conditions where radiation is minimized. minimized. The measured measured PCTs in these these assessments ranged from approximately assessments approximately 1,000 to 1,540'F.

1,540cF. At these temperatures, temperatures, there is little little rod-to-rod radiation. Given the good agreement agreement between between the biased code calculations and the the CCTF, LOFT, and Semiscale Semiscale data, it can be concluded concluded that there is no significant significant over prediction of the total heat transfer coefficient.

coefficient.

Notwithstanding any conservatism Notwithstanding conservatism evidenced evidenced by experimental benchmarks, the application experimental benchmarks, application of the model to commercial commercial nuclear power plants provides some additional additional margins margins due to to limitations within the experiments.

limitations The benchmarked benchmarked experiments, FLECHET SEASET and ORNL ORNL Thermal Hydraulic Hydraulic Test Facility (THTF), used to assess assess the S-RELAP5 heat transfertransfer incorporated constant rod powers across the experimental model, Reference 1, incorporated experimental assembly.

Temperature differences that occurred were the result of guide Temperature differences guide tube, shroud shroud or local heatheat transfer transfer effects. In the operation pressurized water operation of a pressurized water reactor (PWR) and in the RLBLOCA peaking factor is present, creating power differences evaluation, a radial local peaking evaluation, differences that tend to enhance the temperature enhance temperature differences differences between between rods. In turn, these temperature temperature differences differences lead to increases in net radiation heat transfer from the hotter rods. The expected expected rod-to-rod rod-to-rod radiation will likely exceed exceed that embodied embodied within the experimental experimental results.

4.3.1 Rod-to-Rod Radiation Implicit in the RLBLOCA Assessment of Rod-to-Rod RLBLOCA Methodology Methodology As discussed above, the FLECHT-SEASET FLECHT-SEASET and THTF tests were selected selected to assess assess and determine the S-RELAP5 determine S-RELAP5 code code heat transfer transfer bias and uncertainty.

uncertainty. Uniform radial power power distribution was used in these test bundles. Therefore, distribution Therefore, the rod-to-rod temperature variation in temperature variation the rods away away from the vessel wall is caused primarily by the variation sub-channel fluid variation in the sub-channel fluid conditions. In the real operating conditions. operating fuel bundle, on the other hand, therethere can be 5 to 10 percent AREVA AREVA NP Inc.

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Calvert Cliffs Nuclear Nuclear Plant Unit 1 Cycle 21 & & Unit 22 Cycle 19 Revision 000 Realistic Realistic Large Break LOCA Summary Large Report Summary Report Page 4-4 Page 4-4 rod-to-rod power variation. In addition, addition, the methodology methodology includes includes aa provision to apply the the uncertainty measurement uncertainty measurement to the hot pin. Table 4-1 provides the hot pin measurement Table measurement uncertainty uncertainty and aa representative representative local pin peaking factor for severalseveral plants. These factors, however, relate the pin to the assembly average. To more properly assess the conditions under under which rod-to-rod rod-to-rod radiation heat heat transfer occurs, a more local peaking assessment is required.required.

Therefore, the plant rod-to-rod Therefore, assessments herein set the average rod-to-rod radiation assessments average pin power for those pins surrounding the hot pin at 96 percent of that of the peak pin. For pins further removed the average average power is set to 94 percent.

Table 4-1 Typical Typical Measurement Uncertainties and Measurement Uncertainties and Local Peaking Peaking Factors Factors F

FAH Measurement

~H Measurement Local Pin Peaking Local Pin Peaking Plant Uncertainty Uncertainty Factor Factor (-)

(percent) 1 4.0 1.068 1.068 2 4.0 1.050 1.050 3 6.0 1.149 1.149 4 4.0 1.113 1.113 5 4.25 1.135 1.135 6 4.0 1.058 1.058 4.3.2 Quantification of the Impact Quantification Impact of Thermal Radiation R2RRAD Code Radiation using R2RRAD Code The R2RRAD radiative heat transfer model model was developed developed by Los Alamos National Laboratory (LANL)

(LANL) to be incorporated incorporated in the BWR version of the TRAC code. The theoretical theoretical basis for this this code is given in References References 8 and 11 and is is similar to that developed developed in the HUXY rod heatup code (Reference (Reference 10, Section Section 2.1.2) used by AREVA for BWR LOCA applications. The version of R2RRAD R2RRAD used herein was obtained from the NRC NRC to examine examine the rod-to-rod radiation radiation characteristics characteristics of aa 5x5 rod segment of the 161 rod FLECHT-SEASET FLECHT-SEASET bundle. The output provided provided by the R2RRAD R2RRAD code includes an estimateestimate of the net radiation radiation heat transfer transfer from each rod in in the defined array. The code allows the input of different temperatures temperatures for each each rod as well as for a boundary surrounding surrounding the pin array. No geometry geometry differences differences between between pin locations are allowed. Even though this limitation affects the view factor calculations calculations for guide tubes, R2RRAD R2RRAD is aa reasonable reasonable tool to,to estimate rod-to-rod rod-to-rod radiation heat transfer.

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Calvert Cliffs Nuclear Plant Plant Unit 1 Cycle 21 & & Unit 2 Cycle 19 000 Revision 000 Realistic Large Break LOCA Summary Summary Report Page 4-5 PaQe 4-5 The FLECHT -SEASET test series was intended FLECHT-SEASET assembly and there is 17x17 fuel assembly intended to simulate a 17x17 a close similarity, Table 4-2, between the test bundle and a modern modern 17x17 assembly.

FLECHT-SEASET & 17x17 Table 4-2 FLECHT-SEASET 17x17 FA Geometry Geometry Parameters Parameters Design Parameter FLECHT-SEASET FLECHT -SEASET 17x17 17x17 Fuel Assembly Assembly Rod Pitch (in)

(in) 0.496 0.496 0.496 Fuel Rod Diameter (in) (in) 0.374 0.374 0.374 Guide Tube Diameter Diameter (in)

(in) 0.474 0.482 Five FLECHT-SEASET FLECHT*SEASET tests (Reference (Reference 6) were selected for evaluation and comparison comparison with expected plant behavior. Table expected Table 4-3 characterizes selected rod characterizes the results of each test. The 5x5 selected comprises the hot rod, 4 guide tubes and 20 near adjacent rods. The simulated array comprises simulated hot rod is rod 7J 7J in the tests.

00000 0 0 000 00 Guide Tube

---.-0000 0 0 Hot Rod Rod 8

08 0 0 8~

Adjacent Rods Adjacent Rods 0 0 000 00000 0 Figure 4-1 R2RRAD Segment R2RRAD 5 x 5 Rod Segment Two sets of runs were made Simulating experiments and one set of cases was simulating each of the five experiments was run to simulate the RLBLOCA evaluation of aa limiting fuel assembly assembly in an operating plant. For the simulation of Tests 31805, 31504, 31021, 31021, and 30817, the thimble tube (guide tube) temperatures were set to the measured 34420, the thimble tube temperature measured values. For Test 34420, temperature temperature. For the first experimental measured vapor temperature.

was set equal to the measured simulation set, the experimental simulation the temperature of all 21 rods and the exterior temperature measured PCT of the exterior boundary was set to the measured the temperature was set experimental set, the hot rod temperature simulated test. For the second experimental set to the PCT AREVA NP Inc.

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Unit 1 Cycle 21 & & Unit 2 Cycle 19 19 Revision 000 000 Realistic Large Break LOCA SummarySummary Report Page 4-6 Page 4-6 value and the remaining remaining 20 rods and the boundary boundary were set to a temperature temperature 25F25'F cooler providing a reasonable reasonable measure of the variation in in surrounding temperatures.

temperatures. To estimate estimate thethe rod-to-rod rod-to-rod radiation in a real fuel assembly at LOCA conditions and compare it the it to the experimental experimental results, each of the above above cases was rerun with the hot rod PCT set to the the experimental result and the remaining experimental conservatively set to temperatures remaining rods conservatively temperatures expected withinwithin the bundle. The guide tubes (thimble tubes) were removed for conservatism conservatism and because peak rod powers powers frequently occur at fuel assembly corners away from either either guide tubes or instrument instrument tubes. In In line with the discussion discussion in in Section 4.3.1, 4.3.1, the surrounding 24 rods were set to aa temperature temperature estimated for rods of 4 percent percent lower power. The boundary temperature was was estimated based an average estimated average power 6 percent below the hot rod power. For both of these, the the temperature estimates estimates were achieved using a ratio of pin power difference in temperature power to the difference temperature between the saturation between.the temperature and the PCT.

saturation temperature T24rods T24 =0.96*

rods = 0.96 (PCT - Tsar)

T Tsat sat) + T sat and Tsurrounding region Tsurrounding region =

= 0.94 *° (PCT - T sat) ++ T Tsat) sat.

Tsat.

sat was taken as 270 F.

Tsat T

Figure 4-2 shows the hot rod thermal radiation radiation heat heat transfer for the two FLECHT-SEASET FLECHT -SEASET sets and for the plant set. The figure shows that for PCTs greater greater than about 1700'F, 1700cF, the hot rod thermal radiation radiation in the plant cases exceeds exceeds that of the same component within the the experiments.

experi ments.

Table 4-3 FLECHT-SEASET Test Parameters Parameters htc at Steam Thimble Thimble Rod 7J PCT PCT Test t at (FGT Tie ( PCTtime Temperature -at Temperature Temperature Temperature at 6-ft ('F) Time (s) (Btu/hr-ft 22 .-F) 71 (6-ft) ('F) at 6-ft (1F)

(Btu/hr-ft -'F) 71 (6-ft) ('F) at 6-ft ('F) 34420 34420 2205 34 10 1850 1850*

31805 31805 2150 110 10 1800 1800 1800 31504 31504 2033 100 10 1750 1750 1750 31021 1684 1684 29 9 1400 1350 1350 30817 1440 1440 70 . 13 900 750

set to steam temp I

se.t to steam temp AREVA NP Inc.

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Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & & Unit 2 Cycle 19 Revision 000 Realistic Large Break Break LOCA SummarySummary Report Page 4-7 Page 4-7 4.5 4.5 44 Cl"'" ... ~

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Figure 4-2 Rod Thermal Radiation Radiation in FLECHT-SEASET FLECHT -SEASET Bundle and in a 17x17 17x17 FA FA 4.3.3 Rod-to-Rod Radiation Rod-to-Rod Radiation Summary In In summary, the conservatism of the heat transfer modeling modeling established by benchmark benchmark can be be reasonably reasonably extended extended to plant applications, applications, and the plant plant local peaking peaking provides a physical reason why rod-to-rod rod-to-rod radiation radiation should be more substantial substantial within a plant environment environment than in in the test environment. Therefore, the lack of an explicit rod-to-rod radiation model, in the version of S-RELAP5 S-RELAP5 applied applied for realistic LOCA calculations, does not invalidate invalidate the conclusion that the the temperature and local cladding temperature local cladding cladding oxidation oxidation have demonstrated to meet the criteria have been demonstrated criteria of 10 CFR 50.46 with a high level of probability.

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Calvert Cliffs Nuclear Plant 19 Revision 000 Revision 000 Unit 1 Cycle 21 &

Unit & Unit 2 Cycle 19 Realistic Realistic Large Break Break LOCA Summary Summary Report Page 4-8 Page 4-8 Boiling Heat Film Boiling 4.4 Film Heat Transfer TransferLimit Question:

Question: Unit 1 Calvert Cliffs Unit In the Calvert I Cycle 21 and Unit 2 Cycle 19 calculations, calculations, is the Forslund-Rohsenow model contribution Forslund-Rohsenow contribution to the heat heat transfer limited to less than transfer coefficient limited than or equal to 15 percent equal percent when the void fraction fraction is greater greaterthan than or equal equal to 0.9?

Response

Response: Yes, the version of S-RELAP5 employed for the Calvert Cliffs Units 1 and 22 S-RELAP5 employed RLBLOCA analysis limits the contribution RLBLOCA contribution of the Forslund-Rohsenow Forslund-Rohsenow model to no more more than-15 than-15 percent of the total heat percent heat transfer at and above a void fraction of 0.9. Because the limit is applied at a void fraction Forslund-Rohsenow within the 0.7 to 0.9 interpolation fraction of 0.9, the contribution of Forslund-Rohsenow interpolation range is limited to 15 15 percent percent or less. This is aa change to the approved RLBLOCA RLBLOCA EM EM (Reference 1). This feature is carried forward (Reference forward into the UAPR09 UAPR09 version version of S-RELAP5.

S-RELAP5.

Downcomer Boiling 4.5 Downcomer Boiling Question: If the PCT PCT is greater greater than 1800'F 1800'F or the contain containmen men t pressure than 30 psia, pressure is less than psia, has the Calvert has Calvert Cliffs Units Units 1I and and 2 downcomer model been rebenchmarked rebenchmarked by performing performing sensitivity studies, sensitivity studies, assuming assuming adequate adequate downcomer noding noding in the water water volume, vessel wall and other heat other heat structures?

structures?

Response: The downcomer model for Calvert Calvert Cliffs Units Units 1 and 2 has has been established generically as adequate adequate for the computation of downcomer downcomer phenomena phenomena including the prediction of potential local boiling effects.

effects. The model was benchmarked benchmarked against against the UPTF tests and the the LOFT facility in the RLBLOCA methodology, Revision 00 (Reference 1).

methodology, Revision 1). Further, AREVA addressed the effects of boiling in the downcomer downcomer in a letter, from James James Malay to U.S. NRC, NRC, April 4, 2003. The letter cites the lack of direct experimental evidence direct experimental evidence but contains sensitivity studies on high and low pressure pressure containments, containments, the impact of additional azimuthal azimuthal noding within noding within the downcomer, and the influence of flow loss coefficients. Of these, the study on azimuthal noding is most germane to this question; indicating that additional azimuthal nodalization question; indicating nodalization allows higher liquid buildup downcomer away from the broken cold leg and increases buildup in portions of the downcomer increases the liquid liquid driving head. Additionally, Additionally, AREVA AREVA has conducted downcomer has conducted downcomer axial noding and wall heat release studies. Each of these these studies supports the Revision 0 methodology methodology and is documented documented later in this section.

This question is primarily concerned with the phenomena primarily concerned phenomena of downcomer downcomer boiling and the the AREVA NP Inc.

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Calvert Nuclear Plant Unit 1 Cycle 21 & Unit 22 Cycle Cycle 19 Revision 000 Revision Realistic Large Realistic Large Break LOCA Summary Report Page 4-9 Paae extension of the Revision 0 methodology methodology and sensitivity studies to plants plants with low containment containment pressures and high cladding cladding temperatures. Boiling, wherever wherever it occurs, is a phenomenon that codes like S-RELAP5 developed to predict. Downcomer S-RELAPS have been developed Downcomer boiling is the result of the the release release of energy stored in vessel metal mass. Within S-RELAPS, S-RELAP5, downcomer downcomer boiling is is simulated simulated in the nucleate nucleate boiling regime with the Chen correlation.

correlation. This modeling modeling has been validated through the prediction validated prediction of several phenomenon provided in the several assessments on boiling phenomenon the S-RELAP5 S-RELAP5 Code Verification and Validation document document (Reference (Reference 12).

12).

ECC Figure 4-3 Reactor Reactor Vessel Downcomer Downcomer Boiling Diagram Hot downcomer walls penalize penalize PCT by two mechanisms:

mechanisms: by reducing subcooling subcooling of coolant coolant entering the core and through throughthe the reduction reduction in downcomer downcomer hydraulic head which is the drivingdriving force for core reflood. Although Although boiling in the downcomer downcomer occurs during blowdown, blowdown, the biggest AREVA NP Inc.

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Unit 1 Cycle 21 & & Unit 2 Cycle 19 Unit 2 19 Revision 000000 Realistic Realistic Large Break LOCA Summary Summary Report Page 4-10 Page 4-10 potential for impact potential impact on clad temperatures temperatures is during during late reflood following the end of SIT injection.

injection.

At this time, there is aa large step reduction in coolant flow from the ECC systems. As a result, coolant entering the downcomer downcomer may may be less subcooled. When the downcomer downcomer coolant approaches saturation, boiling on the walls initiates, reducing the downcomer approaches downcomer hydraulic hydraulic static static level.

level.

With the reduction of the downcomer downcomer level, level, the core inlet flow rate is reduced reduced which, which, depending depending on the existing core inventory, may result in a cladding temperature temperature excursion excursion or a slowing of the core cooldown cooldown rate.

While downcomer downcomer boiling may impact impact clad temperatures, temperatures, it is somewhat somewhat of aa self-limiting self-limiting process. If cladding temperatures If temperatures increase, less energy is transferred in the core boiling boiling process and the loop steam flows are reduced.

process reduced. This reduces the required driving head to support continued core reflood and reducesreduces the steam available available to heat the ECCS water within within the cold legs resulting resulting in greater greater subcooling of the water water entering entering the downcomer.

The impact downcomer boiling is primarily impact of downcomer dependent on the wall heat release rate and on primarily dependent the ability to slip steam steam up the downcomer downcomer and out of the break. The higher higher the downcomer downcomer wall heat release, the more steam is generated generated within the downcomer downcomer and the larger the impact on core reflooding.

reflooding. Similarly, the quicker quicker the passage of steam steam up the downcomer, the less less resident volume volume within the downcomer downcomer is occupied occupied by steam and the lower the impact impact on the the downcomer average density.

downcomer Therefore, the ability to properly simulate downcomer Therefore, downcomer boiling boiling depends on both the heat release (boiling) depends model and on the ability to track steam rising through (boiling) model through the downcomer.

downcomer. Consideration of both of these is provided in the following text. The heat release modeling in S-RELAP5 is validated by aa sensitivity study study on wall mesh point spacing and through benchmarking benchmarking against a closed form solution. Steam tracking is validated validated through both an axial and an azimuthal azimuthal fluid control control volume sensitivity sensitivity study done at low pressures. The The results indicate that the modeling accuracy accuracy within the RLBLOCA RLBLOCA methodology is sufficient to resolve the effects of downcomer downcomer boiling and that, to the extent that boiling occurs; the the methodology properly resolves the impact methodology impact on the cladding temperature temperature and cladding cladding oxidation rates.

4.5.1 Release Rate Wall Heat Release Rate AREVA NP Inc.

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Calvert Cliffs Nuclear Calvert Revision 000 Revision Unit 1 Cycle 21 && Unit 22 Cycle 19 Realistic Large Break LOCA Summary Report Realistic Large Report Page 4-11 Page 4-11 downcomer wall heat The downcomer heat release rate during during reflood is conduction limited and depends depends on the the vessel wall mesh spacing used used in the S-RELAP5 S-RELAP5 model.

model. The following two approaches approaches are used evaluate the adequacy of the downcomer to evaluate downcomer vessel wall mesh spacing spacing used S-RELAP5 used in the S-RELAP5 model.

4.5.1.1 Exact Solution In this benchmark, the downcomer downcomer wall is considered considered as a semi-infinite semi-infinite plate. Because the the benchmark benchmark uses a closedclosed form solution to verify the wall mesh spacing used in S-RELAP5, spacing used S-RELAP5, it is assumed that the material assumed material has constant constant thermal temperature Ti, thermal properties, is initially at temperature T j , and, at time zero, has one surface, the surface simulating contact with the downcomer dowricomer fluid, set to aa constant temperature, To, representing representing the fluid temperature. Section Section 4.3 of Reference Reference 9 gives gives exact solution for the temperature the exact profile as a function of time as temperature profile as (T(x,t) - To) / (T =

(Tij -- To) = erf {x / (20(0 5 )},

t)o.s)},

(2.(a t)° (1)

. where, a0 is the thermal thermal diffusivity diffusivity of the material material given by a0== k/(p Cp),

k = thermal conductivity,

=

p = density, Cp ==specific specific heat, and and erf{} is the Gauss error function (given in Table A-1 of Reference Reference 9).

The conditions conditions of the benchmark are T Tij == 500°F 500'F and To == 300°F. 3000 F. The mesh spacing in S-mesh spacing S-RELAP5 is the same as that used RELAP5 used for the downcomer vessel wall in the RLBLOCA model.

downcomer vessel temperature distributions in the metal at 0.0, 100 and 300 seconds shows the temperature Figure 4-4 shows seconds as as calculated calculated by using using Equation 1 and S-RELAP5, S-RELAP5, respectively. The solutions solutions are identical confirming confirming the adequacy of the mesh spacing used used in the downcomer downcomer wall.

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Calvert Cliffs Nuclear Plant Unit 11 Cycle Cycle 21 & & Unit 2 Cycle 19 Revision 000 Realistic Large Break LOCA Summary Report Page Paae 4-12 4-12 550 550 500 1-

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-()-O - S-RELAP5, S-RELAP5, 300 5 s 250 I 1 0.0 0.0 0.1 0.2 0.2 0.3 0.4 0.4 0.5 0.5 0.6 0.6 0.7 0.7 0.8 0.8 Distance Distance from Inner Inner Wall, Wall, feet Figure Figure 4-4 S-RELAP5 S-RELAP5 versus Closed Form Solution Solution AREVA AREVA NP NP Inc.

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Calvert Cliffs Nuclear Nuclear Plant Revision 000 Unit 1 Cycle 21 & & Unit 22 Cycle 19 Realistic Large Break LOCA Summary Summary Report Page 4-13 Paqe 4-13 4.5.1.2 Plant Model Sensitivity Sensitivity Study As additional verification, verification, a typical 4-loop plant case was used to evaluate the adequacy adequacy of the the mesh spacing within the downcomer downcomer wall heat structure. Each mesh interval interval in in the base case case downcomer vessel wall was divided into two equal intervals. Thus, a new input model downcomer model was was created by increasing the number of mesh intervals intervals from 9 to 18. The following four figures figures show the total downcomer downcomer metal heat release rate, peT PCT independent independent of elevation, downcomer downcomer liquid level, and the core liquid level, level, respectively, for the base case and the modified modified case.

These results confirm the conclusion from the exact solution study that the mesh spacing used in in the plant model for the downcomer downcomer vessel wall is adequate.

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Calvert Cliffs Calvert Nuclear Plant Cliffs Nuclear Plant Revision000 Revision 000 Cycle 21 Unit 11Cycle Unit 21 && Unit Cycle19 Unit22Cycle 19 RealisticLarge Realistic Large Break Summary Report LOCASummary BreakLOCA Report Page4-14 Paqe 4-14 240000 2400.00 .. -~- ..--...- ............,.. _._......... __

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Calvert Cliffs Nuclear Nuclear Plant Revision 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Summary Report Page 4-15 Page4-15 3000 30.00 ci, Q) ci, 20,00 2000

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Plant ANP-2834(NP)

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Calvert Cliffs Nuclear Nuclear Plant Cycle 19 19 Revision 000 Unit 1 Cycle 21 & & Unit 22 Cycle Realistic Large Break LOCA Summary Report Page 4-16 Page 4-16 12.00 , - - _ . ,....._ _ _ _ ...., . - _........ _ _- .....- , - -.................. ,.................. .

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Nuclear Plant Calvert Cliffs Nuclear Plant ANP-2834(NP)

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Calvert Cycle 21 &

Unit 1 Cycle Cycle 19 Unit 22 Cycle

& Unit 19 Revision 000 Revision 000 Unit Realistic Large Realistic Large Break Break LOCA LOCA Summary Summary Report Report Page 4-17 Page 4-17 4.5.2 4.5.2 Downcomer Fluid Downcomer Fluid Distribution Distribution To justify the adequacy To adequacy of the downcomer nodalization in calculating downcomer nodalization the fluid distribution in the calculating the the downcomer, two studies varying downcomer, axial and the azimuthal separately the axial varying separately resolution with azimuthal resolution which the with which the downcomer is downcomer is modeled modeled have been been conducted.

conducted.

4.5.2.1 Azimuthal 4.5.2.1 Azimuthal Nodalization Nodalization 1), AREVA (Reference 1),

In aa letter to the NRC dated April, 2003 (Reference documented several studies AREVA documented studies on on downcomer boiling.

downcomer Of significance Of here is the study on significance here on further azimuthal break break up of the the downcomer noding.

downcomer noding. The study, based on a 3-loop plant plant with a containment containment pressure of approximately 30 psia during reflood, consisted approximately consisted of several examining the affects several calculations examining affects onon clad temperature temperature and and other parameters.

parameters.

The base model, with 6 axial by 3 azimuthal base model, expanded to 66 axial by 9 azimuthal azimuthal regions, was expanded regions (Figure 4-9). The base base calculation simulated the limiting PCT calculation given in the calculation simulated the EMF-2103 three-loop problem. This case was then repeated with the revised 6 x 9 three-loop sample problem.

downcomer noding.

downcomer The change resulted in an alteration alteration of the blowdown blowdown evolution of the transient with little little evidence of any affect evidence affect during reflood. To isolate any possible reflood impact impact that might have an an downcomer boiling, the case was repeated influence on downcomer adjusted vessel-side repeated with a slightly adjusted vessel-side break flow. Again, little evidence of impact on the reflood reflood portion of the transient was observed.

The study concluded that blowdown or near blowdown impacted by refining the blowdown events could be impacted azimuthal resolution in the downcomerdowncomer but that reflood would not be impacted.

impacted. Although the the study was performed for a somewhat somewhat elevated system system pressure, the flow regimes within the downcomer downcomer will not differ for pressures as low as atmospheric. Thus, the azimuthal azimuthal downcomer downcomer modeling employed for the RLBLOCA methodology modeling employed reasonably converged methodology is reasonably converged in its ability to represent downcomer boiling phenomena. phenomena.

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Calvert Cliffs Nuclear Plant Revision 000 Unit 1 Cycle 21 && Unit 2 Cycle 19 Realistic Large Break Summary Report Break LOCA Summary Paae 4-18 Page 4-18 Base model (HCY 4ýC7D H Revised 9 Region Model Figure 4-9 Azimuthal Noding Noding 4.5.2.2 4.5.2.2 Axial Nodalization Nodalization The RLBLOCA methodology divides the downcomer RLBLOCA methodology downcomer into six nodes axially. In In both 3-loop 3-loop and 4-loop 4-loop models, the downcomer downcomer segment at the active core elevation elevation is represented represented by two equal length nodes. For most operating plants, the active core length is 12 feet and the downcomer downcomer segments at the active core elevation elevation are each 6-feet high. (For a 14 foot core, these nodes nodes would would be 7-feet high.)

high.) The model for the sensitivity study presented here comprises a 3-loop 3-loop plant with an ice condenser condenser containment and aa 12 12 foot core. For the study, the two nodes nodes spanning spanning the active core height are divided in half, revising the model to include eight axial nodes. Further, the refined noding noding is located within the potential boiling region of the the AREVA NP Inc.

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Calvert Unit Unit 11 Cycle Cycle 21 21 & & Unit Unit 22 Cycle Cycle 19 Revision 000 Revision 000 Realistic Large Realistic Large Break Break LOCA LOCA Summary Summary Report Page 4-19 PacQe 4-19 downcomer downcomer where, where, if if there is is an an axial axial resolution resolution influence, influence, the sensitivity sensitivity to to that that impactimpact wouldwould be be greatest.

The results show show that that the axial noding noding used in in the base methodology methodology is sufficient sufficient for plants plants experiencing exp~riencing the very low low system pressures pressures characteristiccharacteristic of ice condenser containments.

ice condenser containments.

Figure Figure 4-10 4-10 provides provides the the containment containment back pressure pressure for the the base modeling. modeling. Figure Figure 4-11 4-11 through through Figure Figure 4-14 show the total downcomer downcomer metal heat heat release rate, PCT peT independent independent of elevation, elevation, downcomer downcomer liquid level, and the core liquid liquid level, respectively, for the base base case and the the modified case.

The results demonstrate demonstrate that the axial resolution provided in the base case, 6 axial downcomer downcomer node node divisions divisions with 2 divisions divisions spanning spanning the core active active region, region, are sufficient to accurately accurately resolve void distributions distributions within within the downcomer. Thus, this modeling modeling is sufficient sufficient for the the prediction of downcomer downcomer driving head and the resolution resolution of downcomer downcomer boiling effects.

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Calvert Cliffs Calvert Nuclear Plant Cliffs Nuclear Plant Revision 000 Revision 000 Unit Cycle 21 Unit 11 Cycle 21 &

& Unit Cycle 19 Unit 22 Cycle 19 Realistic Break LOCA Large Break Realistic Large Summary Report LOCA Summary Report Page 4-20 Paae 4-20 300o0,,

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Calvert Cliffs Nuclear Plant Calvert Cliffs Plant Revision 000 000 Revision Unit 11 Cycle Unit Cycle 21 21 & Unit 22 Cycle

&Unit 19 Cycle 19 Realistic Large Realistic Large Break Break LOCA Summary Report LOCA Summary Report Paqe 4-21 Page 4-21 2400.CO 2400.CQ I--"~"-'--"--'-"'''''''------ ,

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Calvert Cliffs Nuclear Nuclear Plant Revision Revision 000 Unit 1 Cycle 21 & & Unit 2 Cycle 19 Realistic Large Break LOCA Summary Report Report Page 4-22 Paqe 4-22 2 2000 V

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ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 22 Cycle 19 Realistic Large Break LOCA Summary Report Report Page Paae 4-23 4-23 12.00 r-'---~---"---r-""''''''''''''''''''''''''''''''''''''''''''''''''''''''-',........................*.......... **--r---~*---*--,--------......--**- . -..*. -r..*-.. - ..*---..- ..*,

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Figure Figure 4-14 Core Liquid Level- Level - Axial Noding Noding Sensitivity Study 4.5.3 Downcomer Boilinq Downcomer Boiling Conclusions Conclusions To further justify justify the ability of the RLBLOCA methodology methodology to predict the potential potential for and impact impact of downcomer downcomer boiling, boiling, studies were performed downcomer wall heat release modeling performed on the downcomer modeling methodology and on the ability of S-RELAP5 within the methodology predict the migration of steam through S-RELAP5 to predict through the downcomer. Both azimuthal and axial noding sensitivity studies were performed. performed. The axial noding study was based based on, on* an ice condenser plant that is near atmospheric atmospheric pressure during during reflood. These studies demonstrate demonstrate that S-RELAP5 delivers energy to the downcomer liquid liquid volumes at an appropriate appropriate rate and that the downcomer downcomer noding detail is sufficient to track the the distribution of any steam formed. required methodology for the prediction of Thus, the required downcomer boiling at system pressures approximating downcomer approximating those achieved achieved in in plants with pressures pressures as low as ice condenser containments containments has been demonstrated. demonstrated.

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Unit 1 Cycle 21 & & Unit 2 Cycle 1919 Revision 000 Revision Realistic Large Break Realistic Large Break LOCA Summary Summary Report Page 4-24 Page 4-24 4.6 Break Size 2

Question: Were all break break sizes assumed greater than or equal greater than equal to 1.0 fttf? ?

Response: Yes.

Response

The NRC has requested that the break break spectrum for the realistic LOCA evaluations be limited limited to accidents that evolve through a range of phenomena accidents phenomena similar to those encountered encountered for the larger break area accidents. This is a change to the approved approved RLBLOCA EM (Reference (Reference 1). The The larger break area LOCAs are typically larger characterized by the occurrence typically characterized occurrence of dispersed dispersed flow filmfilm boiling at the hot spot, which sets them apart from smaller break LOCAs. LOCAs. This occurs generally generally in the vicinity vicinity of 0.2 DEGB (double-ended guillotine break) size (i.e., (Le., 0.2 times the total flow area area of the pipe on both sides of the break). However, this transitional break size varies from plant to plant and is verified verified only after the break break spectrum spectrum has been executed.

executed. AREVA NP has sought to develop develop sufficient sufficient criteria for defining defining the minimum large break flow area prior to performing performing the break spectrum. The purpose purpose for doing so is to assure a valid break spectrumspectrum is performed.

4.6.1 4.6.1 Break /I Transient Break Phenomena Transient Phenomena In determining determining the AREVAAREVA NP NP criteria, the characteristics characteristics of larger break area area LOCAs are examined. These LOCA characteristics involve a rapid and chaotic LOCA characteristics depressurization of the chaotic depressurization the reactor coolant system (RCS) during which the three three historical historical approximate approximate states of the system can be identified.

Blowdown The blowdown blowdown phase is defined defined as the time period from initiation of the break until flow from the SITs begins. This definition is somewhat different from the traditional definition definition of blowdown which extendsextends the blowdown blowdown until the RCS pressure approachesapproaches containment pressure.' The blowdown containment pressure: blowdown phase typically lasts about 12 12 to 25 seconds, depending on the break size.

depending Refill is that period that starts starts with the end of blowdown, blowdown, whichever whichever definition definition is used, and ends when when water is first forced upward into the core. During this phase the core core experiences experiences a near adiabatic heatup.

Reflood Reflood is that portion of the transient transient that starts with the end of refill, follows through through the the filling of of the core with water and ends with the achievement achievement of complete complete core quench.

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Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 &&Unit 22 Cycle 19 Revision 000 Realistic Large Break LOCA Summary Report Page 4-25 Page 4-25 Implicit in this break-down Implicit inventory has been completely, or nearly so, break-down is that the core liquid inventory primary system leaving the core in a state of near core-wide dispersed flow expelled from the primary floW subsequent adiabatic film boiling and subsequent heatup prior to the reflood phase. Although this break adiabatic heatup approaches and is valid deterministic LOCA evaluation approaches down served as the basis for the original deterministic for most LOCAs that would classically be termed termed large breaks, as the break area decreasesdecreases thethe depressurization rate decreases such that these three phases overlap depressurization overlap substantially. During During break events, the core liquid inventory is not reduced as much as that found in these smaller break in larger breaks. Also, the adiabatic core heatup is not as extensiveextensive as in in the larger larger breaks which which cladding temperature results in much lower cladding temperature excursions.

excursions.

4.6.2 New Minimum Break Size Determination Determination determination of the lower limit can be exact. The values of critical phenomena No determination phenomena that control transient will overlap and interplay.

the evolution of a LOCA transient This is especially true in aa statistical evaluation parameter values are varied randomly evaluation where parameter randomly with a strong expectation that variations will affect results. In the variations In selecting the lower area of the RLBLOCA break spectrum, spectrum, AREVA sought to preserve the generality. complete or nearly complete core dry out generality, of a complete out substantially reduced lower accompanied by a substantially accompanied lower plenum liquid inventory. It was reasoned that conditions would be unlikely ifif the break flow rate was reduced to less than the reactor such conditions coolant pump flow. That is, if coolant pumps are capable of forcing more coolant if the reactor coolant reactor vessel than the break can extract toward the reactor extract from the reactor vessel, the downcomer downcomer and core must maintain some degree degree of positive flow (positive in the normal operations sense).

circumstance is, of course, transitory. Break flow is altered as the RCS blows down and The circumstance decrease as the rotor and flywheel slow down ifif power is lost. However, the RC pump flow may decrease if the core flow was reduced to zero or became if became negative immediately after the break initiation, negative immediately then the event was quite likelylikely to proceed with sufficient inertia to expel most of the reactor reactor vessel liquid to the break. The criteria base, thus established, consists of comparing comparing the break flow to the initial flow through all reactor coolant pumps and setting the minimum break area area flows' match. This is done as follows:

such that these flows'match.

Wbreak =

Wbreak = Abreak =

Gbreak = N Abreak ** Gbreak pump

Npump

W WRCP.

RCP .

gives This gives AREVA NP Inc.

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Nuclear Plant Calvert Cliffs Nuclear Revision 000 Revision 000 Unit 1 Cycle 21 & Unit 2 Cycle 19 19 Realistic Large Large Break LOCASummary LOCA Summary Report Page 4-26 4-26 Abreak = (Npump

WRcP)/Gbreak.

The break mass flux is determined determined from critical flow. Because Because the RCS pressure in the broken broken cold leg will decrease decrease rapidly during the first few seconds of the transient, the critical mass flux flux is averaged between between that appropriate appropriate for the initial operating operating conditions conditions and that appropriate appropriate for the initial initial cold leg enthalpy enthalpy and the saturation pressure of coolant at that enthalpy.

Gbreak = (Gbreak(PO, HCLO) + Gbreak(PCLsat, HCLO))/ 2 .

The estimated estimated minimum LBLOCA LBLOCA break area, A min , is 2.817 ft Amin, ft22 and the break areaarea percentage, percentage, based on the full double-ended double-ended guillotine break break total area, is 28.69 percent.

Table 4-4 provides a listing of the plant type, initial condition, condition, and the fractional minimum minimum RLBLOCA break area, for all the plant types presented as generic representations in the next section.

section.

Table 4-4 Minimum Break Break Area for Large Large Break LOCA Spectrum Spectrum Spectrum Spectrum Spectrum System Subcooled Saturated RCP flow Minimum Minimum Plant System Cold Leg Subcooled Saturated Decit Plant Pressure Enthalpy Gbreak Gbreak (HEM) (Rbm/s)

RCP flow Break Area Minimum Break Area Minimum Description Pressure (psia) Enthalpy (Btu/Ibm) (Ibm/ft2-s)

Gbreak Gbreak (HEM)

(Ibm/ft2-s) Ims)ftak2) AreaGBrekAe Description (Ibm/s) Break Area Break Area (psia) (Btullbm) (Ibm/te-s) (Ibm/fe-s)

(te) (DEGB) 3-Loop 3-Loop W 31417 A

Design W 2250 555.0 23190 5700 2.18 0.26 0.26 3-Loop 3-Loop W 28124 B

Design Design W 2250 544.5 23880 5450 28124 1.92 0.23 0.23 3-Loop 3-Loop W C W 2250 550.0 23540 5580 29743 29743 2.04 0.25 0.25 DesiQn Design 2x4 2x4 CE D CE Design DesiQn 2100 538.8 538.8 22860 5310 21522 1.53 0.24 0.24 2x4 2x4 CE CE 535.8 5230 0.27 E

Design DesiQn 2055 535.8 22630 37049 2.66 0.27 4-loopW 4-1oopW 540.9 5370 39500 0.33 F DesignIII DesiQn 2160 540.9 23290 39500 2.76 0.33 The split versus The versus double-ended double-ended break type type isis no no longer longer related related to to break break area. In In concurrence concurrence with Regulatory Regulatory Guide 1.157, 1.157, both the split and the double-ended double-ended break will range in area area between between the minimum minimum break area (Amin) (Amin) and an area of twice the size of the broken pipe. The The determination of break configuration, determination configuration, split versus double-ended, double-ended, is made after the break area is selected based on a uniform probability selected probability for each occurrence.

occurrence.

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Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & & Unit 2 Cycle 19 Revision 000 Realistic Large Break Break LOCA Summary Summary Report Page 4-27 Page 4-27 4.6.3 Intermediate Intermediate Break Size Disposition Disposition With the revision of the smaller break area for the RLBLOCARLBLOCA analysis, analysis, the break range for small breaks and large breaks are no longer contiguous. Typically the lower end of the large break spectrum occurs spectrum occurs at between 0.2 to 0.3 times the total area of aa 100 percent double-ended percent double-ended guillotine break (DEGB) and the upper end of the small break spectrum occurs occurs at approximately approximately 0.05 times the area of aa 100 100 percent percent DEGB. This leaves a range of breaks that are not specifically analyzed specifically analyzed during aa LOCA LOCA licensing analysis. The premise for allowingallowing this gap is that these breaks these breaks do not comprise comprise accidents that develop develop high high cladding temperature temperature and and thus do not comprise accidents accidents that critically critically challenge challenge the emergency core cooling systems systems (ECCS).

Breaks Breaks within this range remain large enough to blowdown to low pressures. Resolution is provided by the large provided large break break ECC systems and the pressure-dependent injection limitations that pressure-dependent injection determine determine critical small break performance performance are are avoided.

avoided. Further, these accidents accidents develop develop relatively slowly, assuring maximum effectiveness of those ECC systems.

relatively A variety of plant types for which analysis analysis within the intermediate intermediate range have been completed were surveyed.

surveyed. determinations are extracted Although statistical determinations extracted from the consideration consideration of breaks with areas above above the intermediate intermediate range, the AREVA AREVA best-estimate best-estimate methodology methodology characterize the ECCS performance remains suitable to characterize performance of breaks within the intermediate intermediate range.

Table 4-4 provides Table provides a listing of the plant type, initial condition, condition, and the fractional minimum minimum RLBLOCA break area. Figure 4-15 through Figure 4-20 provide RLBLOCA provide the enlarged enlarged break spectrum results with the upper upper end of the small break spectrum and the lower end of the large break indicated by bars. Table 4-5 provides differences spectrum indicated differences between the true large break break region l

and the intermediate intermediate break region (break areas areas between between that of the largest SBLOCA and the the smallest RLBLOCA).

RLBLOCA). difference is 141'F; The minimum difference 141'F; however, this case is not representative representative of the general general trend shown shown by the other other comparisons.

comparisons. The next minimum minimum difference is 704'F (see Figure 4-15). Considering difference Considering this point as an outlier, the table shows the the minimum difference minimum between the highest intermediate difference between intermediate break spectrum PCT and large large break spectrum PCT, for the six plants, as at least 463'F 463'F, , and including including this point would provide an average difference of 427'F average difference differenc ee of 840'F.

427cF and a maximum differenc 840cF.

Thus, by both measures, the peak cladding temperatures temperatures within the intermediate intermediate break range range will be several hundred degrees below those in the true large large break range. Therefore, these these AREVA NP Inc.

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Calvert Cliffs Nuclear 000 Revision 000 Unit 1 Cycle 21 && Unit 22 Cycle 19 Realistic Large Break LOCA Summary Report Page 4-28 Paae 4-28 breaks will not provide a limit or a critical critical measure of the EGGS ECCS performance.

performance. Given that thethe large break break spectrum bounds the intermediate spectrum, spectrum, the use of only the large break spectrum spectrum meets the requirements of 10GFR50.46 10CFR50.46 for breaks within the intermediate intermediate break LOCA spectrum, and the method LOGA demonstrates that the EGGS method demonstrates ECCS for a plant meets the criteria of 10CFR50.46 1OGFR50.46 with high probability.

Table 4-5 Minimum Minimum PCT Temperature Temperature Difference Difference - True Large and and Intermediate Breaks Intermediate Breaks Generic Generic Maximum Maximum Maximum Maximum Plant Plant (I7)

PCT ('F) PCT (7)

('F) Delta PCT Average Delta Average Delta Description Description Label Label Intermediate Intermediate Large Large Size (7)

('F) PCT ('F)

(Table 4-4) Size Break Break Break A

A 17461 17461 1887 1887 1411 3-Loop 3-Loop W B 1273 1951 678 427' 4271 B 1273 1951 678 Design C

G 1326 1326 1789 1789 463 D 984 1751 767 2x4CE 2x4 GE D 2x4 984 CE767 1751 767 767 Design Design E 869 1636 767 E 869 1636 767 3-IoopW 3-loop W F 1127 1967 840 840 F 1127 1967 840 840 Design nd Note: 1. The 2 2 nd highest PCT was 1183'F.

highest PGT 11837F. This changes the Delta Delta PC PG T to 704'F and the the increases to 615'F.

average delta increases AREVA AREVA NP Inc.

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Calvert Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 & Unit 2 Cycle 19 Revision 000 Realistic Large Break LOCA Realistic LOCA Summary Summary Report Page 4-29 4-29 2000 ---- ---r-------7----

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Calvert Cliffs Nuclear Nuclear Plant Revision 000 Unit 1I Cycle 21 & & Unit 2 CycleCycle 19 Large Break Realistic Large Break LOCA Summary Summary Report Report Page Paqe 4-30 4-30 2000 2000 - - - T -

Upper End SBLOCA -

SBLOCA Break Size End of of Large Break Spectrum Break

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Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Calvert Nuclear Plant Revision Revision 000 Unit 1 Cycle 21 &

Unit & Unit 2 Cycle 19 Realistic Large Break LOCA SummarySummary Report Page 4-31 2000 - - - - - - - - - - - I - - - - - - - I - - - - - - - - - - - - - - , - - - - - - - - r - - - - - - - - - - - - -~~~1~

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ANP-2834(NP)

Nuclear Plant Calvert Cliffs Nuclear Unit 1 Cycle 21 & Unit 22 Cycle 1919 Revision 000 Revision 000 Realistic Large Summary Report Large Break LOCA Summary Report Page 4-32 2000 2000 - - - - - - - - - - ., - - - - - - - r - - - - - - - - - - - - - - ., - - - - - - - - r - - - - - - - - - - - - - - ., - - - - - - - - r - - - - - - - -I Upper End of of Large Break Break SBLOCA Spectrum Break Size Minimum Minimum 1800 1800 Spectrum

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ANP-2834(NP)

Nuclear Plant Calvert Cliffs Nuclear Plant Revision 000 Revision 000 Unit 1 Cycle Cycle2121 && Unit 2 Cycle 19 Summary Report Realistic Large Break LOCA Summary Page 4-33 2000 2000 Upper EndEnd of of Large Break Break SBLOCAA SBLOC* Spectrum Break Break Si ze Size Minimum Minimum 1800 1800 Spectrumnr

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Calvert Calvert Cliffs Nuclear Nuclear Plant ANP-2834(NP)

AN P-2834(N P)

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ANP-2834(NP)

Calvert Cliffs Nuclear Plant Nuclear Plant Unit 1 Cycle 21 & 19

& Unit 2 Cycle 19 Revision 000 000 Realistic Large Break LOCA Summary Realistic Large Summary Report Page 4-35 Page 4-35 4.7 Detailed Detailedinformation ContainmentModel information for Containment Containment initial conditions Containment conditions and cooling information are provided in Table 3-8 and cooling system information Heat Sinks are provided in Table 3-9. For Calvert Cliffs Units 1 and 2, the scatter plots of PCT Heat containment volumes and initial atmospheric versus the sampled containment temperature are shown in atmospheric temperature in Containment pressure as a function of time for limiting case is Figure 4-21 and Figure 4-22. Containment shown in shown in Figure 4-23.

AREVA NP Inc.

AREVA

ANP-2834(NP)

Calvert Cliffs Nuclear Nuclear Plant Revision 000 000 Unit 1 Cycle 21 & & Unit 22 Cycle 19 Realistic Large Large Break LOCA Summary Summary Report Page Page 4-36 PCT vs Containment Containment Volume Volume 2000 1800 1800 0 0 F-1a 1600 1600 0

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'AREVA

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Plant Revision 000 000 Unit 1 Cycle 21 && Unit 22 Cycle 19 Realistic Large Break LOCA Summary Report Page 4-37 Paae 4-37 Containment Temperature PCT vs Containment Temperature 2000 1800 F 0 0 1600 F

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AREVA NP Inc.

ANP-2834(NP)

Calvert Nuclear Plant Calvert Cliffs Nuclear Revision 000 000 Unit 1 Cycle 21 && Unit 2 Cycle 19 Realistic Realistic Large Break LOCA Summary Report Paqe 4-38 Page Containment Containment Pressure 60 60 ~--~----~--~----~--~----,---~--~

1-- Containment 1 50 50 40 30 20 10 10 L -_ _~_ _ _ _L -_ _~_ _ _ _~_ _~_ _ _ _~_ _~_ _~

o0 100 200 300 400 Time (s)

ID:55570 25Jun2009 17:30:03 R5DMX 25Jun2009 17:30:03 Figure 4-23 Containment Containment Pressure as function of time for limiting limiting case case AREVA AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant 19 Cycle 19

&Unit 22 Cycle Revision 000 000 I Cycle 21 &

Unit 1 Realistic Large Break LOCA Summary Report Page 4-39 Page 4-39 Cross-References to North Anna 4.8 Cross-References Question: In orderorder to conduct its its review of the Calvert Units 112 Calvert Cliffs Units application of AREVA's 1/2 application AREVA's manner, the NRC staff would like to make reference methods in an efficient manner, realisticLBLOCA methods realistic reference to responses to NRC staff requests the responses requests for additional additional information that were developed for the information that the AREVA methods to the North Anna Power application of the AREVA application Power Station, Units 1I and Station, Units and found and 2, and acceptable during acceptable that review.

during that review. The NRC Staff safety evaluation issued on April 1, evaluation was issued 1, 2004 (Agency-wide Documentation Documentation and and Management Management System (ADAMS) accession number accession ML040960040). The staff would like to make use of the information ML040960040). that was provided by the information that North Anna licensee that is North applicable only to North Anna or is not applicable or only to subatmospheric subatmospheric containments. This information containments. contained in letters to the NRC from the North Anna licensee information is contained dated dated September September 26,26,2003 accession number ML032790396) 2003 (ADAMS accession ML032790396) and and November 10,2003 10, 2003 number ML033240451).

accession number (ADAMS accession ML033240451). The specific responses that the staff would like to responses that are:

reference are:

reference September 26, September Question 1I 26, 2003 letter: NRC Question NRC Question Question 2 NRC Question Question 4 NRC Question Question 6 November 10, 2003 letter:letter: NRC Question Question 1 information in these that the information Please verify that Please these letters applicable to the AREVA letters is applicable applied to AREVA model applied Units 112 Calvert Cliffs Units Calvert 1/2 except forfor that that information related specifically information related North Anna and to sub-specifically to North sub-atmospheric containments.

atmospheric containments.

Response: The responses provided Response: provided to questions questions 1, 2, 4. generic and related to the 4, and 6 are generic the ability of ICECON calculate containment ICECON to calculate containment pressures. They are applicable to the Calvert Cliffs Cliffs Units 1/2 RLBLOCA submittal.

Question Question 1 - Completely Applicable Applicable Question 2 - Completely Applicable Applicable AREVA NP Inc.

Calvert Cliffs Nuclear Plant ANP-2834(NP)

ANP-2834(NP)

Calvert Cliffs Nuclear Plant

& Unit 2 Cycle 1919 Revision 000 Revision 000 Unit 1 Cycle Cycle 21 &

Realistic Large Large Break LOCA Summary Summary Report Page 4-40 Page 4-40 Question 4 -

Question Completely Applicable reference to CSB Applicable (the reference CSB 6-1 should now be to CSB Technical Position 6-2). The NRC altered identification of this branch technical position in altered the identification Revision 3 of NUREG-0800.

NUREG-0800.

Question 6 -

Question Completely applicable.

applicable.

The supplemental supplemental request and response are applicable applicable to Calvert Cliffs Units 112. 1/2.

GDC 35 - LOOP 4.9 GDC LOOP and No-LOOPNo-LOOP CaseCase Sets Question: 10CFR50, Question: IOCFR50,Appendix A, GDC GDC [General

[GeneralDesignDesign Criterion] [Emergency core Criterion]35 [Emergency core cooling]

cooling] states states that, that, "Suitable redundancy in components "Suitable redundancy components and and features features and suitable suitable interconnections, detection, isolation, interconnections, leak detection, and containment isolation, and containment capabilities shall be provided capabilities shall provided to I

assure that assure that for onsite electric power system operation onsite electric operation (assuming offsite electric (assumingoffsite electric power is not available) and for offsite electric available) electricpower operation operation (assuming onsite power is not available)

(assuming onsite the available) the system function function cancan be accomplished, assuming a single accomplished, assuming failure. "

single failure."

The Staff interpretation interpretationis that that two cases cases (/oss (loss of offsite power with onsite power available, available,and onsite power with offsite power available) loss of onsite available) must be run independently to satisfy run independently satisfy GDC GDC 35.

Each Each of these casescases isis separate other in that separate from the other each case that each representedby a different case is represented statistical spectrum. To accomplish response spectrum.

statistical response accomplish the task task of identifying case would worst case identifying the worst require require more runs. However, for LBLOCA analyses runs. However, analyses (only), the high high likelihood likelihood of loss of onsite onsite power being the most limitinglimiting is so small that that only loss of offsite power cases cases need be run. run. (This (This is unless a particular particularplant design, e.g., CE [Combustion plant design, Engineering]plant

[Combustion Engineering] plant design, also design, is also vulnerable to a loss vulnerable power, in which situation loss of onsite power, situation the NRC may require that both cases require that cases be be separately.This would require analyzed separately.

analyzed require more case case runs runs to satisfy the statistical statisticalrequirement requirement than forjust than loss of offsite power.)

for just loss power.)

What is your basisbasis for assuming probabilityof loss of offsite power?

assuming a 50% probability power? Your statistical statistical runs need to assume assume that offsite power is lost that offsite lost (in (in an independent independent set of runs).runs). If, stated above, If, as stated above, it has been determined determined that Palisades,being of CE design, that Palisades, design, is also also vulnerable vulnerable to a loss of onsite onsite power, this also power, also should should be addressed addressed(with an independent independent set of runs).runs).

Response: In concurrence with the NRC's interpretation interpretation of GDC GOC 35, a set of 59 cases each was run with a LOOP and No-LOOP assumption. The set of 59 cases that predicted No-LOOP assumption. predicted the highest AREVA NP Inc.Inc.

Plant Nuclear Plant Calvert Cliffs Nuclear ANP-2834(NP)

ANP-2834(NP) 19 Cycle 19

& Unit 2 Cycle Revision 000 Unit 1 Cycle 21 &

Realistic Large Realistic Large Break LOCA LOCA Summary Summary Report Page 4-41 Page 4-41 figure of merit, PCT, is reported in in Section Section 2 and Section herein. The results from both case Section 3, herein. case sets are shown in Figure Figure 3-22. This is a change to the approved RLBLOCA EM (Reference approved RLBLOCA (Reference 1).1).

4.10 Variables Statement Input Variables Question: Provide a statement Question: Provide confirming that statement confirming that Constellation Constellation Energy and and its LaLOCA LBLOCA analyses analyses vendor have ongoing vendor ongoingprocesses processes that assure assure that the input variables variables and ranges ranges of parameters parameters for the LaLOCA LBLOCA analyses conservatively bound analyses conservatively bound the values and ranges ranges of those parameters parameters for operated Calvert the operated Calvert Cliffs Nuclear Nuclear Plant Unit 1I (CCA) and Plant Unit and Unit Unit 2 (CCa).

(CCB). This statement statement addressescertain addresses certainprogrammatic programmaticrequirements requirements of 10 CFR 50.46, 50.46, Section Section (c).

(c).

Constellation Energy and the LBLOCA Response: Constellation LBLOCA Analysis Vendor have have an ongoing ongoing process process to ensure that all input variables and parameter parameter ranges for the CCA'and CCA' and CCB realistic large large break loss-of-coolant loss-of-coolant accident are verified as conservative conservative with respect respect to plant operating operating and design design conditions. In accordance accordance with Constellation Constellation Energy Quality Assurance Assurance, program process involves requirements, this process involves

1) Definition Definition of the required input variables and parameter parameter ranges by the Analysis Vendor.

2) Compilation of the specific specific values from existing existing plant design input and output documents documents by by Constellation Constellation Energy and Vendor personnelpersonnel in in aa formal analysis input summary summary document issued by the Analysis Vendor and 3)

3) Formal review and approval of the input summary document document by Constellation Constellation Energy. Formal Constellation Constellation Energy Energy approval of the input document document serves as the release for the Vendor to perform the analysis.

Continuing Continuing review of the input summary summary document is performed by Constellation Constellation Energy as part of the plant design change process and cycle-specific change process cycle-specific core design process. Changes Changes to thethe input summary summary required to support modifications or cycle-specific support plant modifications cycle-specific core alternations alternations are are formally communicated to the Analysis Vendor by Constellation Constellation Energy. Revisions updates Revisions and updates to the analysis parameters are documented documented and approved in accordance accordance with the process process described described above for the initial analysis.

AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant

& Unit 2 Cycle 19 19 Revision 000 Unit 1 Cycle 21 &

Realistic Large Realistic Large Break Break LOCA LOCA Summary Summary Report Report Page 5-1 Page 5-1 5.0 Conclusions Conclusions A RLBLOCA RLBLOCA analysis was performed for the Calvert Cliffs Nuclear Plant Units using Units 1 and 2 using NRC - approved approved AREVA NP RLBLOCA methods (Reference RLBLOCA methods (Reference 1).

1). Analysis results show that the the o

limiting LOOP case has a PCT of 1670 1670'F, F, and a maximum maximum oxidation thickness and hydrogen hydrogen generation generation that fall well within regulatory requirements.

The analysis supports operation operation at a nominal nominal power level of 2754 MWt (including 0.62%

uncertainty), a steam generator generator tube plugging level of up to 10 percent percent in all steam generators, aa total LHGR of 1S.0 15.0 kWkW/ft, 1ft, a total peaking factor (Fa)(FQ) up to a value of 2.384, and aa nuclear nuclear enthalpy enthalpy rise factor (F t.H) up to a value of 1.81 (FAH) 1.81 including the 6%6% uncertainty uncertainty with no axial oror burnup burnup dependent dependent power peaking.limit peaking . limit and peak rod average average exposures of up to 62,00062,000 MWd/MTU.

MWd/MTU. For large break LOCA, the three 10CFRS0.46 10CFR50.46 (b) criteria presented in Section 3.0 criteria presented 3.0 are met and operation operation of Calvert Cliffs Units 1 and 22 with AREVA NP-suppliedNP-supplied 14x14 M5 clad fuel is justified.

AREVA NP Inc.

ANP-2834(NP)

Calvert Cliffs Nuclear Plant Unit 1 Cycle 21 && Unit 22 Cycle 19 Revision 000 Realistic Large Break LOCA Summary Report Report Page 6-1 Page 6-1 References 6.0 References

1. EMF-2103(P)(A) Revision Revision 0, Realistic Realistic Large Large Break LOCA Methodology, Methodology, Framatome Framatome ANP, Inc., April 2003.

2. Program Group, Quantifying Technical Program Quantifying Reactor Reactor Safety Margins, Margins, NUREG/CR-5249, NUREG/CR-5249, EGG-2552, October EGG-2552, October 1989.

1989.

3. Wheat, Larry L.,L., "CONTEMPT-LT A Computer Program for Predicting Computer Program Predicting Containment Containment Pressure-Temperature Response Pressure-Temperature Response to a Loss-Of-Coolant-Accident," Nuclear Loss-Of-Coolant-Accident," Aerojet Nuclear Company, TID-4500, TID-4500, ANCR-1219, June 1975.

4. XN-CC-39 (A) Revision 1, 1, "ICECON: A Computer Computer Program to Calculate Containment Calculate Containment Back Pressure Pressure for LOCA Analysis (Including (Including Ice Condenser Plants)," Exxon Nuclear Exxon Nuclear Company, October 1978. 1978.

5. U. S. Nuclear U. Regulatory Commission, NUREG-0800, Nuclear Regulatory NUREG-0800, Revision 3, Standard Standard Review March 2007.

Plan, March

6. NUREG/CR-1532, EPRI NP-1459, WCAP-9699, "PWR FLECHT SEASET NUREG/CR-1532, SEASET Unblocked Unblocked Bundle, Forced and Gravity Gravity Reflood Task Data Report," June 1980.

7. G.P. Liley and L.E. Hochreiter, "Mixing of Emergency Core Cooling Water Core Coofing Water with Steam:

1/3 - Scale Test andand Summary,"

Summary," EPRI Report EPRI-2, June June 1975.

8. NUREG/CR-0994, "A Radiative Heat Transfer Model for the TRAC Code" November

8. NUREG/CR-0994, "A Radiative Heat Transfer Model for the TRAC Code" November 1979.

9. J.P. Holman, Heat Transfer, 4th McGraw-Hill Book Company, 1976.

4 th Edition, McGraw-Hili 1976.

10.

10. EMF-CC-1 30, "HUXY: A Generalized EMF-CC-130, Generalized Multirod Multirod Heatup Heatup Code for BWR Appendix Appendix K K LOCA Analysis Theory Manual,"

Manual," Framatome ANP, May May 2001.

2001.

11.

11. D. A. Mandell, D. Mandell, "Geometric View Factors for Radiative Radiative Heat Transfer Transfer within Boiling Boiling Water Reactor Fuel Bundles," Nucl.

Reactor Nucl. Tech., Vol. 52, March 1981.

1981.

12.

12. EMF-2102(P)(A) Revision 0, S-RELAP5: Code Verification EMF-2102(P)(A) Verification and Validation, Validation, Framatome Framatome ANP, Inc., August 2001.

2001.

AREVA NP Inc.

ENCLOSURE (5)

Sample Application Application for non-LOCA Analysis Analysis (Non-Proprietary Version)

(Non-Proprietary Version)

Calvert Cliffs Nuclear Calvert Nuclear Power Plant, LLC November 23, 2009 23,2009

ANP-2857(NP)

Revision 0 Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Power Plant, Unit 2 September 2009 A EVA

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ANP-2857(NP)

Revision 0 Loss of Forced Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Power Plant, Unit 2

AREVA NP Inc.

ANP-2857(NP)

dar

A AREVA AREVVA ANP-2857(NP)

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Loss of Forced Reactor Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Nuclear Plant, Revision Revision 0 Unit 2 Paae 4 Page Nature of Changes Nature Changes Item Page Description Description and Justification Justification 1.

1. All Initial release.

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Loss of Forced Reactor Coolant Flow Analysis Analysis for CalvertCalvert Cliffs Nuclear Plant, Revision 0 Revision Unit 2 Unit Page 5 Table of Contents Contents Nature of Changes ........................................................................................................................

Changes ........................................................................................................................ 4 Table Table of Contents ..........................................................................................................................

.......................................................................................................................... 5 List ooff Tables List Ta ble s .................................................................................................................................

................................................................................................................................. 5 5 List ooff Figures List F igure s ...............................................................................................................................

............................................................................................................................... 6 No me nc la tu re ................................................................................................

Nomenclature ................................................................................................................................

............................... 7 11.0

.0 Intro ductio n .......................................................................................................................

Introduction ....................................................................................................................... 88 22.0

.0 C o nc lu s ion ........................................................................................................................

Conclusion ........................................................................................................................ 99 3.0 Analytical Methodology Methodology ...................................................................................................

.............................................................................................. 10 10 3.1 Nodalization ......................................................................................................

Nodalization ......................................................................................................... 10 10 3.2 Param eters .....................

Chosen Parameters .......................................................................................

....................................................................... 11 11 3.3 Sensitivity Studies ...............................................................................................

............................................................................................ 11 11 3.4 Definition of Event Analyzed and Bounding Bounding Input ...............................................

........................................... 11 11 4.0 Complete Loss of Forced Reactor Coolant Flow (UFSAR Event 14.9) ....................... 17

........................... 17 4.1 Identification Identification of Causes and Event Description .............................................

........... ........................................ 17 17 4.2 Acceptance Criteria .............................................................................................

Acceptance ......................................................................................... 17 17 4.3 .......................................................

Analysis Results ...................................... ~ ........................................................... 18 18 5.0 5.0 Re fe re nce s ......................................................................................................................

References ...................................................................................................................... 2277 List of TablesTables Table 3.1 Key Assumptions Assum ptions .................................................................................................

............................................................................................ 12 12 Table 3.2 Key Input Parameters Param eters Biases ..............................................................................

.......................................................................... 13 13 Table 4.1 Sequence of Events ............................................................................................

Sequence ....................................................................................... 19 19 AREVA NP INC. INC,

A AREVA ANP-2857(NP)

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Loss of Forced Forced Reactor Reactor Coolant Flow Analysis for Calvert Calvert Cliffs NuclearNuclear Plant, Revision Revision 0 Un~2 Unit 2 . Page 6 List of FiguresFigures Figure 3.1 S-RELAP5 S-RELAP5 Reactor Reactor Vessel Nodalization ..............................................................

Nodalization .................................................................. 14 S-RELAP5 Reactor Figure 3.2 S-RELAP5 Reactor Coolant System Nodalization Nodalization (Loop 1) .....................................

................................ 15 Figure 3.3 S-RELAP5 S-RELAP5 Steam Generator Generator SecondarySecondary System and Steam Line Line Nodalization Nodalization (Loop (Loop 1) ......................................................................................................

............................................................................................... . . 16 Figure 4.1 RCS Flow for Loss of Forced Reactor Reactor Coolant Flow ...........................................

................................................ 20 20 Figure 4.2 Core Neutronic Neutronic and Surface Surface Power for Loss of Forced Reactor Reactor C o o la n t Flow Coolant F lo w ...................................................................................................................

................................................................................................................... 2211 Figure 4.3 Core Reactivity Reactivity for Loss of Forced Reactor Coolant Flow ........................................

Forced Reactor .................................... 22 22 Figure 4.4 Primary and Secondary Secondary Pressure for Loss of Forced Reactor Coolant Coolant F low .................................................................................................................................

Flow ................................................................................................................................. 223 3 Figure 4.5 Reactor Reactor Coolant System Fluid Temperatures Temperatures for Loss of Forced Forced R eactor Coolant Reactor C oolant Flow ......................................................................................................

............................................................................................... . . 24 24 Pressurizer Spray Flow for Loss of Forced Reactor Figure 4.6 Pressurizer ...................... 25 Reactor Coolant Flow ........................... 25 Figure 4.7 Pressurizer Pressurizer PORV Flow for Loss of Forced Reactor Coolant .....................

Coolant Flow .......................... 26 26 AREVA NP INC.

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Reactor Coolant Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Revision Unit 22 Page 7

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Nomenclature Nomenclature AOO AOO Anticipated Operational Occurrence Anticipated Occurrence BOC Beginning-of-Cycle Beginning-of-Cycle CE Engineering Combustion Engineering CHF Critical Heat Flux DTC Doppler Temperature Temperature Cofficient Cofficient DNB(R) Departure Departure from Nucleate Boiling (Ratio)

(Ratio)

HTP High Thermal Performance Performance HFP Hot Full Power HZP Hot Zero Power LOCF Loss of Forced Reactor Reactor Coolant Flow LOOP Loss Of Offsite Power MDNBR Minimum Departure from Nucleate Minimum Departure Nucleate Boiling Ratio Ratio

, MTC Moderator Temperature Coefficient Moderator Temperature Coefficient PORV PORV Operated Relief Valve Power Operated Valve RCP Reactor Reactor Coolant Pump RCS Reactor Reactor Coolant System RPS Reactor Reactor Protection System RTP Rated Thermal Thermal Power SAFDL SAFDL Specified Specified Acceptable Acceptable Fuel Design LimitLimit TS Specification Technical Specification AREVA AREVA NP INC.

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Loss of Forced Forced Reactor Coolant Flow Analysis Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Unit 2 Page 8 1.0 1.0 Introduction Introduction The analysis documented describes a LOCF analysis for Calvert Cliffs Nuclear documented herein describes Nuclear Power Power Plant. This analysis demonstrates demonstrates the application application of the Reference Reference 1 methodology to thethe Calvert Cliffs Nuclear Nuclear Power Power Plant. The methodology methodology utilizes the S-RELAP5 computer computer code for the thermal-hydraulic thermal-hydraulic analysis analysis to determine determine the plant transient response. The transient transient core boundary conditions determined boundary determined from the thermal-hydraulic thermal-hydraulic analysis are used by the XCOBRA-XCOBRA-IIIC code (Reference (Reference 2) along with the HTP correlation (Reference (Reference 3) to determine determine the MDNBR.

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Loss of Forced Reactor Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Unit 2 Page 9 2.0 Conclusion Based on the results of this analysis, margin margin exists to the DNB SAFDL.SAFDl. Because the core core power does not increase increase appreciably appreciably during this event, the challenge challenge to the fuel centerline melt centerline melt SAFDL is not limiting. The pressurization pressurization transient does not present a severe challenge to the the maximum pressure criterion since system temperatures temperatures and pressure, pressure. increase less significantly significantly for a loss of flow event compared compared to complete loss of load type events. Therefore, the event event acceptance criteria are met.

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Loss of Forced Reactor Coolant Flow Analysis Forced Reactor Analysis for Calvert Calvert Cliffs Nuclear Plant, Revision 0 Unit 2 Page 1010 3.0 Analytical Methodology Analytical Methodology The analysis performed using the approved analysis is performed approved Reference Reference 1 methodology. The S-RELAP5 S-RELAP5 codecode is used to model the primary and secondary secondary side systems of the Calvert Calvert Cliffs Nuclear Nuclear Power Plant and to calculate calculate reactor power, total reactivity and RCS fluid conditions conditions (such as coolant coolant flow rates, core inlet temperatures, pressurizer pressure and level). The MDNBR MDNBR for the event event is calculated calculated using the thermal-hydraulic thermal-hydraulic conditions from the S-RELAP5 S-RELAP5 calculation calculation as input toto the XCOBRA-IIIC XCOBRA-IIIC code (Reference (Reference 2) along with the HTP CHF correlation (Reference 3).

correlation (Reference 3.1 Nodalization Nodalization The plant configuration configuration is represented represented by an S-RELAP5 S-RELAP5 model. The S-RELAP5S-RELAP5 model nodalizes the primary and secondary sides into, control volumes nodalizes representing reasonable volumes representing reasonable homogenous interconnected by flow paths, or "junctions". The reactor vessel, RCS homogenous regions, interconnected RCS piping and steam generator generator nodalization nodalization diagrams are shown in Figures 3.1 to 3.3. The current current analysis is based on a Calvert Cliffs Nuclear Nuclear Power Power Plant specific model.

In general, the plant nodalization In nodalization is defined to be consistent wherever possible for different different plant types. Calvert Cliffs is a CE 2x4 plant. The S-RELAP5 S-RELAP5 model used for the current Calvert Cliffs Cliffs Nuclear Plant analysis is based on the sample problem Nuclear problem in in Reference Reference 1 which is for a CE 2x4 plant, with some modifications modifications to account for plant-specific plant-specific geometry.

The steam generator generator secondary and steam line models models are nodalized nodalized slightly different between the current current model for Calvert Calvert Cliffs Nuclear Nuclear Power Plant and the Reference Reference 1 sample problem model, namely, the steam generator generator downcomer downcomer and boiler regions in the current model each contain one fewer node. Although the number of nodes decreased decreased by one in each of these characteristics of the steam generator, specifically regions, the characteristics specifically the volume distribution in thethe downcomer downcomer and the heat transfer to the boiler region, are more accuratelyaccurately captured. The The overall effect of these changes on the analysis is negligible negligible for this event. Also, the MFW and AFW connections connections to the SG downcomer downcomer are one node lower than the sample problem, to match the Calvert Calvert Cliffs plant geometry.

Other plant specific differences differences include the number and location of the main steam safety valves, the geometry of the pressurizer surgeline and the pressurizer pressurizer PORV design.

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Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Plant, Revision Revision 0 Unit 2 Unit 2 Page Page 1111 3.2 Chosen Parameters Chosen Parameters The input parameters parameters and equipment equipment states are chosen to provide conservative conservative initial and boundary conditions for estimating estimating the challenge challenge to DNB. The biasing biasing and assumptions for key input parameters parameters are consistent with the approved Reference Reference 1 methodology. The key assumptions are given in Table 3.1 and the biasing parameters is provided biasing of key parameters provided in Table 3.2.

The process process of defining the biasing and assumptions for key input parameters parameters is consistent consistent with the Reference Reference 1 sample problem.

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3.3 Sensitivity Studies Sensitivity This event is controlled controlled primarily primarily by the primary system flow coast down. The S-RELAP5 S-RELAP5 codecode Reference 1 validate the model relative to this controlling assessments in Reference controlling parameter. Thus, no no additional model model sensitivity studies are needed for this application.

The biasing of input parameters parameters is chosen to produce aa conservative conservative estimate of the challenge challenge to DNB for this application.

application. Thus, no additional parameter sensitivity studies are needed.

additional input parameter 3.4 Definition of Event Analyzed and Definition and Bounding Bounding Input The event is analyzed from full power initial conditions since the margin to the DNB limit is minimized at the beginning beginning of the event. The input parameter biasing and assumptions for this this event, shown in Tables 3.1 and 3.2, are consistent with the approved methodology.

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Reactor Coolant Loss of Forced Reactor Coolant Flow Flow Analysis for Calvert Cliffs Nuclear Plant, Revision 0 Unit 22 Paqe 12 Page Table 3.1 Key Assumptions Assumptions Parameter Parameter Assumption Time of.of loss-of-offsite loss-of-offsite power Offsite Offsite power is available available Isolated at event initiation, to allow Feedwater Main Feedwater MDNBR MDNBR from this event to bound MDNBR MDNBR from LOOP event.

Mitigating systems Mitigating systems

- Primary Flow Trip Low Primary TriJ) Available Available

- Pressurizer Spray Available Available

- Pressurizer PORVs Available Available Operator Actions OQerator No operator operator actions actions credited No single failure will adversely adversely affect the the Single Failures Single Failures .consequences consequences of this event Loops Operating Loops Number of Operating Number All loops All are in loops are in operation operation consistent with NumberofOperatingLoopsI operation HFP operation AREVA AREVA NP INC.

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Loss Loss of Forced Reactor Coolant Forced Reactor Coolant Flow Analysis Analysis for Calvert Cliffs Nuclear Plant, Revision Revision 0 Unit 2 Page 13 Table 3.2 Key Input Table 3.2 Parameters Biases Input Parameters Biases Parameter Parameter Bias Bias Rated thermal power plus calorimetric calorimetric Initial reactor core power (MWt)

Initial reactor core power (MWt) uncertainty uncertainty Maximum TS value (( ] plus plus Initial RCS vessel vessel average average temperature temperature measurement and control deadband measurement and control dead band

(°F)

CF) uncertainties [ I uncertainties [ ]

Nominal value [ )) minus minus Initial RCS pressure (psi (psia) a) measurement measurement and control deadband dead band uncertainties [ I]

TS minimum accounting for measurement measurement Initial RCS flow rate (Mlbm/hr)

Initial RCS flow rate (Mlbmlhr) uncertainty uncertainty Minimum HFP worth assuming assuming the most Scram reactivity (pcm)

Scram reactivity (pcm) reactive reactive rod is stuck out of the core core Moderator temperature temperature coefficient Most positive TS value Most positive TS value (pcm/°F)

(pcmrF)

Doppler reactivity coefficient (pcm/°F) coefficient (pcmrF) Nominal BOC [ ]

Pellet-to-clad gap conductance Pellet-to-clad conductance 2and fuel (Btu/hr-ft -OF) properties (Btu/hr-ft2-0F) BOC 130C rod thermal properties RCS Low Flow RPS trip setpoint Nominal minus uncertainty AREVA NP INC.

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Loss Loss ofof Forced Forced Reactor Reactor Coolant Coolant Flow Flow Analysis Analysis for for Calvert Calvert Cliffs Cliffs Nuclear Nuclear Plant, Revision Revision 0 Unit Unit 22 Page Page 1414

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Figure Figure 3.1 3.1 S-RELAP5 S*RELAP5 Reactor Reactor Vessel Vessel Nodalization Nodalization AREVA AREVA NP NP INC.

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Loss of Forced Reactor Reactor Coolant Coolant Flow Flow Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Unit 22 Page 15

[I 1I S-RELAP5 Reactor Coolant System Nodalization Figure 3.2 S-RELAP5 Nodalization (Loop 1)

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Loss Loss of Forced Forced Reactor Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Nuclear Plant, Revision Revision 0 Unit 2 Page 16

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Figure 3.3 S-RELAP5 Figure S-RELAP5 Steam Generator Generator Secondary Secondary System and and Steam Line Nodalization Nodalization (Loop 1)

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Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Forced Reactor Nuclear Plant, Revision 0 Unit 2 Page 17 4.0 Complete Loss of Forced Reactor Reactor Coolant Flow (UFSAR(UFSAR Event 14.9) 4.1 Identification of Causes Identification Causes and Event Description Description The LOCF event is defined as a complete loss of forced reactor reactor coolant coolant through the core with offsite power power available, but without a seized Reactor Coolant Coolant Pump (RCP) rotor. The seized RCP rotor event is discussed discussed in the UFSAR (Section 14.16). A loss-of-coolant flow without without offsite power available is the same as the Loss of Offsite Power (LOOP) power available (LOOP) event discussed in the in the UFSAR (Section 14.10).

UFSAR A LOCF event may result from a simultaneous loss of electricalelectrical power to all four reactor coolant pumps. If If the reactor is at power at the time of the event, the immediate immediate effect of loss of forced reactor coolant flow is a rapid increase in the reactor coolant temperature. This increaseincrease could could result in DNB with subsequent fuel damagedamage if ifthe reactor is not tripped promptly.

A reactor trip on low primary flow is provided provided to trip the reactor reactor for the loss of flow event. The The event initiated from Mode 1 conditions with the reactor at full power, bounds other modes modes ofof operation because operation because this provides margin.

provides the least DNBR margin.

Mitigation of this event is accomplished Mitigation accomplished by timely reactor scram. Since the RPS has sufficient redundancy, no single active active failure will adversely affect the consequences consequences of the event.

The key assumption in the analysis is to use aa conservative conservative RCP coastdown rate. The fastest RCP coastdown coastdown rate is used based on benchmarking benchmarking against plant data.

4.2 Acceptance Criteria Criteria For Calvert Cliffs Nuclear Nuclear Power Plant, the LOCF event is classified as an Anticipated Anticipated Operational Occurrence Operational Occurrence (AOO). For this event, the principally challenged acceptance acceptance criterion criterion is:

The fuel cladding integrity should be maintained by ensuring that fuel design design limits are not exceeded.

exceeded. This is demonstrated demonstrated by assuring that the minimum minimum calculated departure departure from nucleate nucleate boiling ratio (DNBR) is not less than the the applicable limits of the DNBR correlation correlation being used and that fuel centerline melt does not occur.

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Loss of Forced Reactor Loss Reactor Coolant Flow Analysis Analysis for Calvert Cliffs Nuclear Plant, Revision 0 Unit 2 Page 18 The analysis documented demonstrates that the DNB SAFDL is met for this event. Fuel documented herein demonstrates challenged since there is no appreciable centerline melt is not challenged appreciable increase in in core power. System overpressure is bounded by more challenging events.

overpressure 4.3 Analysis Results The results of the analysis indicate indicate that the predicted predicted MDNBR is greater greater than the safety limit. The The critic~1 critical heat flux correlation limit ensures that, with 95% probability probability and 95%95% confidence, confidence, DNB is not expected expected to occur; therefore, no fuel is expected expected to fail. The fuel centerline centerline melt threshold is penetrated during this event. Thus, AOO acceptance not penetrated acceptance criteria criteria are met for this event.

The sequence of events is shown in in Table 4.1.

4.1. The transient transient history of key system variables are given inin Figure 4.1 to Figure 4.7.

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Loss of Forced Reactor Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Plant, Revision Revision 0 Unit 2 Page Page 19 Sequence of Events Table 4.1 Sequence Events Event Time (sec)

Initiate transient (all four pumps begin coastdown) 0.00 RCS low flow RPS Trip signal 0.91 Reactor Reactor scram (begin rod insertion) 1.40 1.40 Core power peaks 1.90 Pressurizer Pressurizer spray flow begins 2.95 MDNBR MDNBR occurs 3.15 3.15 Pressurizer PORV Pressurizer PORV opens 4.55 Pressurizer pressure peaks Pressurizer 4.60 AREVA AREVA NP INC.

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. Loss of Forced Reactor Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Unit 2 Unit Page 20 20 CCNPP LOCF Analysis Analysis 120 100 100 Q) iii 0::

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Loss of Forced Reactor Coolant Flow Analysis for Calvert Calvert Cliffs Nuclear Plant, Revision Revision 0 Unit 2 Unit Paae Page 21 CCNPP LOCF Analysis Analysis 3000 2500 2000 2000

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Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Loss of Forced Reactor Nuclear Plant, Revision 0 Unit 2 22 Page 22 CCNPP LOCF Analysis LOUF Analysis o

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Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Reactor Coolant Nuclear Plant, Revision 0 Unit 2 Paqe 23 Page 23 CCNPP LOCF Analysis Analysis 2500 2000 2000 ro Cal

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Loss of Forced Forced Reactor Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Nuclear Plant, Revision Revision 0 Unit Unit 2 Paqe 24 Page 24 CCNPP LOCF Analysis Analysis 620 600 600~ _ _ ___

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560 560 540 1 2 3 4 5 6 7 8 9 10 Time (sec)

ID:15299 5Aug2009 07:37:58 10:15299 CCNPPLOCF_02.dmx 07:37:58 CCNPP _LOCF _02.dmx Figure 4.5 Reactor Coolant Coolant System Fluid TemperaturesTemperatures for Loss of Forced Reactor Reactor Coolant Coolant Flow AREVA AREVA NP INC.

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Loss Loss of Forced Reactor of Forced Coolant Flow Reactor Coolant Analysis for Calvert Flow Analysis Nuclear Plant, Cliffs Nuclear Calvert Cliffs Revision 00 Revision Unit 2 Unit Paae Page 25 25 CCNPP CCNPP LOCF LOUF Analysis Analysis 60 I -----e mflowj-157000000 I E

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Loss of Forced Reactor Coolant Flow Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision Revision 0 Unit Unit 2 Page 26 26 CCNPP LOCU Analysis LOCF Analysis 100 I - - mflowj-185000000 I 80 80

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Loss of Loss of Forced Forced Reactor Reactor Coolant Coolant Flow Analysis for Calvert Cliffs Nuclear Nuclear Plant, Revision 0 Unit 2 Page 27 5.0 References References

1. EMF-231 EM 0(P)(A) Revision 1, SRP Chapter F-231 O(P}(A) Chapter 15 Non-LOCA Methodology for for Pressurized Pressurized Water Reactors, Water Framatome ANP, May 2004.

Reactors, Framatome

2. XN-NF-82-21 (P)(A) Revision 1, 1, Application Application of Exxon Nuclear Nuclear Company PWR Thermal Thermal Margin Methodology Margin Methodology to Mixed Core Configurations, Core Configurations, Exxon Nuclear Company, September September 1983.

1983.

3. EMF-92-153(P)(A)

EMF-92-153(P)(A) Revision Revision 1, HTP:

HTP: Departure Oeparlure from Nucleate Boiling Correlation Correlationfor High Thermal Performance Fuel, High Thermal Performance Fuel, Siemens Power Corporation, January January 2005.

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