Text

Reload Transition Report, Attachment 4
	ML093350100
Person / Time
Site:	Calvert Cliffs
Issue date:	11/23/2009
From:	; Calvert Cliffs
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML093350100 (92)
	v • d • e

ATTACHMENT ATTACHMENT (4) (4)

RELOAD TRANSITION TRANSITION REPORTREPORT Calvert Calvert Cliffs Nuclear Power Plant, LLC November 23, 2009 November

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION REPORT TABLE OF CONTENTS CONTENTS Page LIST OF TABLES TABLES ......................................................................................................................................

...................................................................................................................................... III LIST OF FIGURES .....................................................................................................................................

..................................................................................................................................... IV 1V

1.0 INTRODUCTION

IN TRODUCTION AND

SUMMARY

SUMM A RY ...............................................................................................

............................................................................................... 11 1.1 Introduction ... :................................................................ ;......................................................... 1 Introduction...............................................................................................................................

1.2 Fuel Features Features ............................................................................................................................

............................................................................................................................ 2 2.0 MECHANICAL M ECHAN ICAL DESIGN FEATURES ................................................................... ~ ......................... 7 2.1 Introduction and Summary Sum mary .......................................................................................................

................................................................................................... 7 2.2 M echanical Com Mechanical patibility ........................................................................................................

Compatibility ................................................................................................... 8 2.3 M echanical Performance Mechanical Perform ance ........................................................................................................

................................................................................................... 12 2.4 Operational Experience .....................................................................................................

Operational .......................................................................................................... 15 Operational Experience with HTP Fuel Assemblies 2.4.1 Operational A ssem blies .................................................

............................................ 16 2.4.2 Operational Operational Experience Experience w withith M5 Cladding .............................................................

M 50 Cladding ........................................................ 19 2.4.3 Operational Operational ExperienceExperience with FUELGUARD FUELGUARD Lower Tie Plate ............................

................................. 21 2.4.4 Operational Operational Experience Experience with MONOBLOCTM MONOBLOCTM Comer Guide Comer Guide Tubes ....................

........................ 22 2.4.5 HTP Fuel Assembly A ssem bly Designs Designs in CE 14x14 Plants ...............................................

..................................................... 22 3.0 3.0 N EUTRON ICS .................................................................................................................................

NEUTRONICS ................................................................................................................................. 24 3.1 Introduction and Sum Introduction Summary m ary .....................................................................................................

................................................................................................ 24 3.2 N eutronics Acceptance Neutronics Acceptance Criteria ........................................................................................

Criteria .............................................................................................. 24 24 3.3 M ethodology ..........................................................................................................................

3.3 Methodology .......................................................................................................................... 25 3.4 N uclear Design Evaluation Nuclear .................................................................................................

Evaluation ..................................................................................................... 26 26 3.5 Conclusions ............................................................................................................................

3.5 Conclusions ............................................................................................................................ 27 27 4.0 THERMAL-HYDRAULICS THERM AL-HY DRAULICS ............................................................................................................

....................................................................................................... 37 37 4.1 Introduction and Sum Introduction Summary mary .....................................................................................................

................................................................................................ 37 37 4.2 M ethodology ..........................................................................................................................

Methodology .......................................................................................................................... 37 37 4.3 Com patibility .........................................................................................................

Hydraulic Compatibility ................................................................................................... 38 38 4.4 Transition Core Perform Performance ance .................................................................................................

............................................................................................ 42 4.4.1 Transition Core DNB Performance Performance ..................................

.......................................................................... *...42

.42 4.4.2 Fuel Rod Bow ............................................................................................................

........................................................................................................ 42 4.4.3 DNB Propagation ...................................................................................................... 42 4.4.4 Im pact Impact of Crud on DNB Perform ance .................................................................

Performance ...................................................................... .43 43 4.4.5 Verification ofTMLL Verification of TMLL ............

.................................................................................. 43 4.5 Fuel Rod Therm Thermal al Perform Performance ance .......................................................................................

............................................................................................. 44 4.5.1 Fuel Centerline Centerline Melt M elt ............................................................................................

.................................................................................................. 44 4.5.2 Fuel Rod Bow ................................................................................

'............................ 44 4.6 Conclusion ............................................................................................................................

.............................................................................................................................. 44 i

ATTACHMENT (4)

ATTACHMENT RELOAD RELOAD TRANSITION TRANSITION REPORT 5.0 PL A N T SY PLANT ST EM S ...........................................................................................................................

SySTEMS ........................................................................................................................... 44 6.0 ACCIDENT A C C ID EN T ANALYSES ..................................................................................................... :.......... 46 A N A LY SE S ................................................................................................................ 46 6.1 6 .1 Introduction ............................................................................................................................

............................................................................................................................ 4466 6.2 Computer Codes C om puter C odes .....................................................................................................................

..................................................................................................................... 47 47 6.3 Transient ANALYSIs A N A LY SIs .............................................................................................................

....................................................................................................... 48 48 6.3.1 A nalysis Methodology Analysis M ethodology ..............................................................................................

......................................................................................... 51 6.3.2 Control Control Element Assembly Withdrawal Withdrawal Event (UFSAR Section 14.2) 14.2) ....................

............... 52 52 6.3.3 Boron Dilution Event (UFSAR Section 14.3) 14.3) ...........................................................

...................................................... 53 53 6.3.4 Excess Load Event Event (UFSAR Section 14.4) 14.4) ...............................................................

.......................................................... 54 54 6.3.5 Loss of Load Event (UFSAR Section 14.5) 14.5) ..............................................................

.......................................................... 55 55 6.3.6 Loss of Feedwater Feedwater Flow Event (UFSAR (UFSAR Section Section 14.6) 14.6) .............................................

........................................ 56 56 6.3.7 Excess Feedwater Heat Removal Event (UFSAR Section 14.7)

Excess Feedwater 14.7) ...............................

........................... 57 57 6.3.8 Reactor Reactor Coolant System Depressurization Depressurization (UFSAR (UFSAR Section 14.8) 14.8) ...........................

...................... 58 58 6.3.9 Loss-of-Coolant Loss-of-Coolant Flow Event (UFSAR Section 14.9) ................................................

.......................................... 59 59 6.3.10 Loss-of-Non-Emergency Loss-of-Non-Emergency AC Power (UFSAR (UFSAR Section 14.10) .............................

Section 14.10) .................................. 60 60 6.3.11 Control ElementElement Assembly Drop Event (UFSAR (UFSAR Section 14.11) 14.11) ........................

............................. 61 6.3.12 Asymmetric Steam Generator Event (UFSAR (UFSAR Section 14.12) 14.12) .............................

.................................. 62 6.3.13 Control ElementElement Assembly Ejection (UFSAR (UFSAR Section 14.13) 14.13) .............................

.................................. 65 6.3.14 6.3 .14 Steam Line Break Event (UFSAR Section Section 14.14) 14.14) ....................................................

.............................................. 66 66 Generator Tube Rupture 6.3.15 Steam Generator Rupture Event (UFSAR (UFSAR Section 14.15) 14.15) ...............................

.......................... 68 6.3.16 Seized Rotor Event (UFSAR (UFSAR Section Section 14.16) 14.16) ............................................................

....................................................... 70 70 Loss-of-Coolant Accident (UFSAR Section 14.17) ..................................................

6.3.17 Loss-of-Coolant ............................................ 71 6.3.18 Fuel Handling Handling Incident (UFSAR Section 14.18) 14.18) ......................................................

................................................ 72 72 6.3.19 Turbine-Generator Turbine-Generator Overspeed Overspeed Incident (UFSAR (UFSAR 14.19) 14.19) .....................................

.......................................... 73 6.3.20 Containment Response Response (UFSAR Section 14.20) .................................................

Section 14.20) ....................................................... 73 6.3.21 Waste Gas Incident (UFSAR Section 14.22) 14.22) ............................................................

....................................................... 74 74 6.3.22 Waste Processing System Incident (UFSAR (UFSAR Section Section 14.23) 14.23) ....................................

.............................. 75 6.3.23 Maximum HypotheticalHypothetical Accident (UFSAR (UFSAR Section 14.24) 14.24) ................................. 76

...................................... 76 6.3.24 Excessive Excessive ChargingCharging Event (UFSAR (UFSAR Section Section 14.25) 14.25) .................................................

............................................ 76 76 6.3.25 Feedline Break Break Event (UFSAR Section 14.26) 14.26) .........................................................

................................................... 77 77 7.0 RE FE REN C E S .................................................................................................................................

REFERENCES ................................................................................................................................. 78 78 ii

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT LIST OF TABLES TABLES Table Table Page 2-1 COMPARISON OF MECHANICAL MECHANICAL DESIGN FEATURES FEATURES .........................................................

................................................... 9 2-2 GENERIC MECHANICAL MECHANICAL DESIGN CRITERIA ..................................................................

CRITERIA ............................................... ,........................ 13 13 2-3 OPERATIONAL OPERATIONAL EXPERIENCE EXPERIENCE ...................................................................................................

............................................................................................... 16 2-4 OPERATIONAL OPERATIONAL EXPERIENCE EXPERIENCE WITH M5 M5 CLADDING MATERIAL ................................... ............................... 20 20 2-5 OPERATIONAL EXPERIENCE WITH FUELGUARD OPERATIONAL EXPERIENCE FUELGUARD LOWER TIE PLATE .......................... ...................... 22 2-6 OPERATIONAL EXPERIENCE OPERATIONAL EXPERIENCE AND DESIGNS DESIGNS OF 14X14 HTP FUEL ASSEMBLIES IN CE PLANTS CE P L A N T S .............................................................................................................................

............................................................................................................................. 23 3-1 KEY KE Y PARAMETERS PA RA M E T ER S ....................................................................................................................

.................................................................................................................... 26 26 4-1 THERMAL-HYDRAULIC THERMAL-HYDRAULIC DESIGN PARAMETERS ................................................................. 39 PARAMETERS ............................................................. 39 6-1

SUMMARY

OF EVENT DISPOSITION .................................................................................... ~ ........ 50 DISPOSITION ............................................................................ 50 iii

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT LIST OF FIGURES FIGURES F~n Figure h~

Page 1-1 1-1 CE 14X14 14X14 FUEL ASSEMBLY ASSEMBLY FOR CALVERT CALVERT CLIFFS ............................................................

...................................................... .44 1-2 1-2 CE 14X14 14X 14 FUELGUARD FUELGUARD LOWER LOWER TIE PLATE ........................................................................

............................................................................ 5 1-3 1-3 CE 14X14 14X 14 RECONSTITUTABLE RECONSTITUTABLE UPPER UPPER TIE PLATE ..........................................................

................................................................ 5 1-4 1-4 CE 14X14 14X 14 CAGE A ASSEMBLy SSEM BLY .......................................................................................................

................................................................................................. 6 1-5 CE 14X14 14X 14 HTP H TP SPACER SPA CER ................................................................................................................

......................................................................................................... 6 1-6 1-6 MONOBLOC CORNER MONOBLOC CORNER GUIDE TUBE DESIGN .........................................................................

................................................................... 7 2-1 BURNUP DISTRIBUTION BURNUP DISTRIBUTION OF THE HTP FUEL ASSEMBLIES ASSEMBLIES ...............................................

.......................................... 17 2-2 BURNUP DISTRIBUTION OF FUEL ASSEMBLIES BURNUP ASSEMBLIES FEATURING FEATURING AN HMP 1IMP AT LOWERMOST LO WERM O ST POSITIONPO SITION ...........................................................................................................

....................................................................................................... 18 2-3 BURNUP DISTRIBUTION BURNUP DISTRIBUTION OF HTP FUEL ASSEMBLIES ASSEMBLIES HAVING HAVING FUEL FUEL RODS W ITH M5 WITH M 5 CLADDING CLADDING M MATERIAL ......................................................................................... 19 ATERIAL .....................................................................................

2-4 BURNUP DISTRIBUTION BURNUP DISTRIBUTION OF AREVA AREVA FUEL ASSEMBLIES FEATURING FEATURING M5 FUEL ROD CLADDING MATERIAL M ATERIAL ...................................................................................

......................................................................................... 21 3-1 FIRST TRANSITION FIRST TRANSITION CYCLE LOADING PATTERN PATTERN WITH BOC AND EOC ASSEMBLY B U RN U PS .....................................................................................................................................

BURNUPS ..................................................................................................................................... 28 28 3-2 SECOND TRANSITION TRANSITION I

CYCLE LOADING PATTERN PATTERN WITH BOC AND EOC A SSEM B LY BURNUPS ASSEMBLY B U RN U PS ...............................................................................................................

............................................................................................................... 29 3-3 TRANSITION CYCLE THIRD TRANSITION CYCLE LOADING LOADING PATTERN PATTERN WITH BOC AND EOC ASSEMBLY ASSEMBLY B U RNU P S .....................................................................................................................................

BURNUPS ..................................................................................................................................... 30 30 3-4 FIRST TRANSITION TRANSITION CYCLE ASSEMBLY POWERS POWERS AT BOC, MOC, AND EOC ................ ........... 31 31 3-5 TRANSITION CYCLE ASSEMBLY POWERS SECOND TRANSITION POWERS AT BOC, MOC, AND EOC ........... ..... 32 32 3-6 THIRD TRANSITION TRANSITION CYCLE CYCLE ASSEMBLY ASSEMBLY POWERS POWERS AT BOC, MOC, MOC, AND EOC ........... ............... 33 33 3-7 Fl FT COMPARISON COMPARISON VERSUS VERSUS CYCLE CYCLE EXPOSURE EXPOSURE FOR THE TRANSITION TRANSITION CYCLES CYCLES ......... ......... 34 34 3-8 LHR COMPARISON COMPARISON VERSUS CYCLE EXPOSURE EXPOSURE FOR THE TRANSITION TRANSITION CYCLES ..... ..... 35 35 3-9 CRITICAL CRITICAL BORON CONCENTRATION CONCENTRATION COMPARISON COMPARISON VERSUS CYCLE EXPOSURE EXPOSURE FOR THE TRANSITION TRAN SITION CYCLES ..........................................................................................

CYCLES .............................................................................................. 36 36 3-10 AXIAL OFFSET COMPARISON VERSUS OFFSET COMPARISON VERSUS CYCLE EXPOSURE EXPOSURE FOR THE TRANSITION TRANSITION C Y CL E S ........................................................................................................................................

CYCLES ........................................................................................................................................ 36 36 iv IV

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION REPORT RELOAD TRANSITION

1.0 INTRODUCTION

INTRODUCTION AND

SUMMARY

AND

SUMMARY

1.1 INTRODUCTION

INTRODUCTION The purpose of this report is to facilitate the transition of the Calvert Cliffs Nuclear Power Plant Units 1 and 2 from the use of Westinghouse Westinghouse Turbo fuel to AREVA AREVA Advanced Performance Advanced CE-14 High Thermal Performance (HTP) fuel. Calvert Cliffs plans to refuel AREVA refuel and operate with AREV A Advanced Advanced CE-14 HTP fuel in Units 1 and 2 starting with Unit 2 in the spring of 2011. 2011. The AREVA fuel design will be the Advanced CE-14 Advanced CE-14 HTP fuel consisting of a 14x14x14 assembly configuration with M5 14 assembly MS fuel rods, Zircaloy-4 fuel rods, MONOBLOC TM Zircaloy-4 MONOBLOCTM comer guide tubes, Alloy 718 High Mechanical cOll\er Mechanical Performance Performance (HMP) spacer at the lowermost axial elevation, Zircaloy-4 Zircaloy-4 HTP spacers in all other axial elevations, elevations, a FUELGUARD lower tie plate, and the AREVA reconstitutable AREV A reconstitutable upper tie plate.

The AREVA AREVA Advanced Advanced CE-14 HTP fuel design is similar to the lead fuel assemblies introduced assemblies that were introduced at Calvert Cliffs Unit 2 in Cycle IS 15 (Reference 1). They operated operated in both Units 1 and 2 and are currently operating in their third cycle. Their Their expected expected discharge discharge pin burnups are greater than the AREV AREVA A fuel rod average burnup licensing limit of 62 MWd/kgU.

MWd/kgU.

AREVA The AREV A Advanced Advanced CE-14 HTP fuel assembly design offers two improvements improvements relative to the lead fuel assembly design -

- Alloy 718 HMP lower end spacer spacer grid

- MONOBLOCTM corner MONOBLOCTM corner guide tubes The Alloy 718 HMP spacer offers improved protection protection against against fuel rod fretting damage. The MONOBLOCTM corner guide tube design has increased lateral MONOBLOCTM lateral fuel assembly stiffness. Both these features have already been exposed to considerable operating experience considerable operating experience at other nuclear facilities in the United States and world-wide.

world-wide .

. The fuel rod design consists of a 0.440 inch outer outer diameter diameter M5 MS clad rod containing an Alloy-X750 Alloy-X7S0 plenum spring and enriched U0 U0 22 or Gd2 00 3 fuel pellets. The fuel rod end caps are made of MS M5 material welded to the fuel rod cladding cladding using the Upset Shape Welding process. The main differences differences between between the fuel rod design used in the lead fuel assemblies and the AREV AREVA A Advanced CE-14 HTP fuel rod design are the use of Gd 0 3 fuel pellets Gd 2 0 pellets in some fuel assemblies, and the use of axial blankets in the AREVA AREV A Advanced CE-14 HTP design.

Section Section 1.2 of this report provides provides a more detailed discussion discussion of the design design features of the AREV AREVA A Advanced CE-14 HTP fuel assembly. Section 2 of the report outlines ARENAs ARE)VAs mechanical mechanical and structural evaluation methodology methodology for the fuel design. Section Section 3 discusses the nuclear design bases and the methodologies for transitioning from Westinghouse Turbo fuel design to the AREV methodologies AREVA A Advanced CE-14 Advanced CE-14 HTP fuel for Calvert Calvert Cliffs. Section 4 provides the thermal and hydraulic design of the reactor that ensures the core can meet steady state and transient performance performance requirements requirements without violating the acceptance criteria. Section S5 discusses the impact acceptance impact of changing changing the fuel design on plant systems.

Section 6 provides information information related to assessing the Calvert Cliffs transient and accident analyses for the proposed proposed transition. Also, summary summary reports of sample analyses analyses for the non-loss-of-coolant non-loss-of-coolant accident (LOCA) and realistic large large break LOCA (RLBLOCA) methodologies are enclosed.

(RLBLOCA) analyses methodologies Note that demonstration demonstration of the evaluation methodologies has been performed performed with a submittal submittal core design. The submittal core design was developed to provide provide key safety parameters parameters to support the transition from Westinghouse Turbo fuel to AREVA Advanced CE-14 HTP fuel prior to the development AREVA Advanced development 1

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION cycle-specific designs.

of cycle-specific designs. This provides assurance assurance that the plant licensing bases are are met for the anticipated operation of of the AREV AREVA A Advanced CE-14CE- 14 HTP HTP fuel during the transition and full core core cycles.

1.2 1.2 FUEL FEATURES The AREV The AREVA A Advanced CE-14 HTP fuel assembly for Calvert Cliffs is of a Combustion Engineering, Inc.

(CE) 14x14 lattice design. Combustion Engineering 14x14 lattice fuel designs contains 176 176 fuel rods, 4 comer guide tubes, and 1 center guide tube. The comer and center guide tubes each occupy four fuel rod positions. The fuel rods are positioned within the fuel assembly by nine spacer grids that are attached to the guide tubes.

The fuel The fuel assembly assembly design incorporates incorporates several several proven design features to enhance performance.

Figure 1-1 Figure 1-I is a drawing of the AREV AREVA A Advanced CE-14 HTP fuel assembly. The fuel rod design in this assembly uses M5 M5 cladding and end caps. The M5 M5 material has very low corrosion and hydrogen hydrogen pickup rates; providing substantial margin for end of life corrosion and hydrogen content. This material was developed in Europe and has been used extensively both in Europe and the United States for fuel rod was cladding. The material has been generically reviewed and accepted accepted by the Nuclear Regulatory Commission (NRC) for use in CE fuel assembly designs (Reference (Reference 2). Reloads with M5 M5 cladding have been provided in the United States since 2000 and in CE 14x14 designs since 2006. Performance has been demonstrated for fuel rod exposures in excess of 80 MWd/kgU. The fuel rod design includes includes uranium dioxide fuel rods and Gadolinia bearing uranium dioxide fuel rods, both with axial blankets of of lower enriched lower enriched uranium dioxide. Also, multiple uranium-235 (U-235) enrichments are used within an assembly.

The lower tie plate design is a FUELGUARD FUELGUARD structure. This structure uses curved curved vanes to provide non-line-of-sight line-of-sight flow paths for the incoming coolant to protect the fuel assembly flow paths assembly from debris that may be be present.

present. ThisThis design design is veryvery efficient at preventing preventing debris, including including small pieces of wire, from reaching reaching the fuel. The design design uses the same vane configuration configuration and spacing that has been used on CE 14x14, 14x14, CE 15x15, 15x15, Westinghouse Westinghouse 14x14, Westinghouse 15x15, Westinghouse Westinghouse 15x15, Westinghouse 17x17, 17x17, and Babcock Babcock & Wilcox Wilcox (B&W) 15x15 (B&W) 15x15 designs designs in in the United States.

the United States. This FUELGUARD FUELGUARD design has been used in reloads in the United States since 1991 and on CE 14x14 designs since 2001. 2001. A drawing of the CE 14x14 14x14 FUELGUARD lower tie plate is provided in Figure 1-2.

FUELGUARD The upper The upper tie tie plate plate design design is is the standard AREVA AREV A Advanced CE-14 HTP fuel reconstitutable reconstitutable design. The basic configuration configuration is the same as that used for other CE 14x14 plants supplied by AREVA, but the height height of the corner comer and centercenter posts, and the thickness thickness of the reaction reaction plate plate are adjusted to be compatible compatible with the core the core plate plate separation separation at Calvert Calvert Cliffs. Figure 1-3 shows the upper tie plate configuration.configuration. This reconstitutable reconstitutable designdesign uses uses the comer comer locking nuts to engage with the upper sleeves on the guide guide tube.

The design The allows the reaction plate to be depressed to a setting well beyond design allows beyond the end of life deflections, and the corner comer nuts rotated rotated to disengage the upper tie plate from the locking nuts. The upper tie plate can then be removed. This then be removed. This design design does not create create any loose or disposable parts during the reconstitution. The design has been used design has been used for AREV Afor AREVA CE 14x 14 reloads 14x14 reloads in the United States since 1982. 1982. The The reconstitution reconstitution capabilities capabilities of the AREVA of the A REV A fuel assemblies assemblies have already successfully demonstrated already been successfully demonstrated in thethe Calvert Calvert Cliffs spent fuel pool pool (SFP)

(SFP) during during the lead lead fuel assembly program.

program.

The cage The cage oror skeleton skeleton uses four Zircaloy-4 Zircaloy-4 corner comer guide tubes, one one Zircaloy-4 Zircaloy-4 center guide tube, seven Zircaloy-4 Zircaloy-4 HTP spacers, spacers, and one Alloy one Alloy 718 HMP spacer HMPspacer at the lowest lowest spacer spacer position. Figure Figure 1-4 1-4 shows the cage configuration. The HTP spacers the cage configuration. The HTP spacers are welded are welded directly to the five guide tubes.

guide tubes. The HMP spacer spacer is attached to attached to the the guide tubes by mechanically mechanically capturing capturing the HMP spacer spacer between between rings that are welded welded to the guide tubes. BecaUseBecause the guide tubes are of a zirconiumzirconium alloy, they cannot cannot be directly directly welded to the 2

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT Alloy 718 material used in the HMP spacer. The HTP spacer design was developed in the late 1980s and has been been used on CE 14x14,14x14, CE 15x15, Westinghouse Westinghouse 14x14, Westinghouse 15x15, Westinghouse Westinghouse 15x15, Westinghouse 17x17, and B&W 15x15 reloads in the United 17x17, United States. The initial use was in 1991, 1991, and the initial CE 14x14 use was in 2001.2001. The design provides eight line contacts contacts as the interface between between the fuel rod and the spacer grid, and is therefore very resistant to fuel rod failures from flow-induced vibration fretting.

The HTP spacer designdesign provides the line contact for the rods, but also is configured configured to improve heat transfer. As seen in Figure 1-5,1-5, the spring structure structure provides a flow path. This flow path is at an angle relative to the fuel rod longitudinal direction, causing causing the water to swirl around the fuel rod without creating creating a large pressure pressure drop across the spacer. The HMP spacer has the same line contact configuration configuration but the channel channel is not angled. Because Because this spacer spacer is at the lowermost position, the improved improved heat transfer transfer is not necessary. As stated previously, previously, the HMP spacer material is Alloy 718. This material is very stable stable in irradiation irradiation environments, environments, and provides additional additional assurance assurance that the fuel rod/spacer rod/spacer contact contact will be be maintained maintained throughout the design lifetime.

lifetime. As of mid-2009, ten reloads reloads of the HTP/HMP fuel assembly design have operated operated in CE 14x14 Pressurized Water Reactors (PWRs) without fuel failures.

The assembly uses a MONOBLOCTM MONOBLOCTM comer corner guide tube design for the comer corner guide tubes (Figure 1-6) 1-6) and a constant outer diameter diameter and wall thickness design for the center guide tube. The batch implementation implementation at Calvert Calvert Cliffs will be the first application application of the MONOBLOCTM MONOBLOCTM corner comer guide tube '

design in the AREVA Advanced Advanced CE-14 HTP fuel. The MONOBLOCTM MONOBLOCTM design design maintains the same inner diameters in the dash dashpot pot and non-dashpot non-dashpot regions as the Westinghouse Turbo fuel, but has a constant constant

, outer diameter the full length of the tube. Therefore, Therefore, the wall thickness in the dash dashpot pot region (about the bottom 12 inches of the guide tube) is increased.

increased. The Westinghouse Westinghouse Turbo fuel maintains the same wall thickness instead of maintaining the same outer diameter as the MONOBLOCTM MONOBLOCTM design. Therefore, Therefore, the Westinghouse Westinghouse Turbo fuel has the same inner diameters, the same outer diameter in the non-dashpot non-dashpot region, and a smaller smaller outer diameter in the dashpot region. The MONOBLOCTM MONOBLOCTM corner guide tube design has been used for fuel reload batches in Europe Europe and in lead fuel assemblies in the United States.

3

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT Figure 1-1, CE 14x14 Fuel Assembly for Calvert Cliffs Figure 1-1,

\

4

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION REPORT Figure 1-2, 1-2, CE 14x14 FUELGUARD FUEL GUARD Lower Tie Plate Figure 1-3, CE 14x14 Reconstitutable Reconstitutable Upper Upper Tie Plate 5

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT Figure 1-4, 1-4, CE 14x14 Cage Cage Assembly 1-5, CE 14x14 Figure 1-5, 14x14 HTP Spacer Spacer 6

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION REPORT 1-6, MONOBLOC Corner Figure 1-6, Corner Guide Tube Design MONOBLOCTM MONOBLOCTM Original Configuration Original Configuration

-- - -- - -- --- - -- - -- ---- -- - -- - -- ---------- -- - - ~

2.0 MECHANICAL DESIGN DESIGN FEATURES FEATURES

2.1 INTRODUCTION

AND INTRODUCTION AND

SUMMARY

This section evaluates evaluates the mechanical mechanical design of the AREVA AREV A Advanced Advanced CE-14 HTP fuel design intended for batch implementation implementation at Calvert Calvert Cliffs and its compatibility compatibility with the Westinghouse Westinghouse Turbo fuel during the transition from mixed-fuel type core populations populations to cores with only AREVA AdvancedAdvanced CE-14 HTP fuel. AREVAs AREV As ongoing lead fuel assembly assembly program with Calvert Calvert Cliffs has demonstrated, demonstrated, through operating operating experience, compatibility of the lead fuel assembly design with Calvert Cliffs reactor core experience, compatibility internals, fuel handling equipment, equipment, fuel storage storage racks, and Westinghouse Westinghouse Turbo fuel. The batch fuel intended for Calvert Calvert Cliffs is mechanically mechanically similar to the lead fuel assemblies and will continue to be mechanically compatible mechanically compatible with the Westinghouse Westinghouse Turbo fuel fuel,, and the plant equipment equipment and reactor reactor core internals.

internals. A summary summary of the mechanical compatibility evaluations mechanical compatibility evaluations performed by AREVA AREV A for the lead fuel assembly program is provided in Section 2.2.

The lead fuel assemblies were analyzed in accordance accordance with the NRC-approved NRC-approved generic mechanical design criteria contained in EMF-92-116 (Reference 3) in conjunction EMF-92-116 (Reference conjunction with NRC-approved NRC-approved topical report BAW-10240 (Reference BA W-10240 (Reference 2). Reference 2 incorporates Reference incorporates the MS cladding material properties that were M5 cladding previously approved by the NRC in BAW- BAW-1022710227 (Reference 4) into the Reference (Reference Reference 3 methodology. All the mechanical mechanical design criteria were shown to be met up to the licensed fuel rod burnup limit of of MWd/kgU in EMF-2807 62 MWd/kgU (Reference 1).

EMF-2807 (Reference I). The design improvements improvements that are mentioned in Section 1.1 Section 1.1 relative to the lead fuel assembly design do not significantly influence the fuel assembly structural significantly influence characteristics that were determined characteristics determined by prior mechanical mechanical testing of the lead fuel assemblies. Therefore, Therefore, the AREV AREVA A fuel design, design, with expected structural behavior and projected performance, projected performance, is designed to meet the applicable applicable design requirements throughout the life of the fuel. The generic mechanical design criteria that were used for the lead fuel assemblies are detailed in Section 2.3. These These criteria will also be be 7

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION applied to the licensing of the AREV AREVA A Advanced CE-14 HTP fuel design. The Reference Reference 3 methodology methodology is used to evaluate the design improvements for the AREVA AREV A Advanced CE-14 HTP fuel as approved by by the NRC.

Section 2.4 provides provides an overview overview of both the overall operating experience gained by AREVA operating experience AREVA with the various components of the AREVA AREV A Advanced CE-14 HTP fuel design as well as the specific operatingoperating experience experience in CE 14x14 plants.

2.2 MECHANICAL COMPATIBILITY MECHANICAL COMPATIBILITY AREVA AREV A and Calvert Cliffs have an on-going on-going lead fuel assembly program using AREV AREVA A fuel. Prior to insertion, the lead fuel assemblies were shown to be compatible compatible with Calvert Cliffs reactor core internals, fuel handling equipment, and fuel storage racks as well as the Westinghouse Westinghouse Turbo fuel in Reference 1. 1.

The lead fuel assembly operating experience experience has confirmed confirmed the results of the AREVA AREV A compatibility compatibility evaluations. The batch AREVA Advanced Advanced CE-14 HTP fuel to be used at Calvert Cliffs is mechanically mechanically equivalent equivalent to the lead fuel assemblies and will continue continue to be mechanically mechanically compatible with Calvert Cliffs reactor core internals, fuel handling handling equipment, fuel storage racks, and Westinghouse Westinghouse Turbo fuel. A comparison comparison of the mechanical parameters of the AREVA mechanical design parameters AREV A Advanced Advanced CE-14 HTP fuel assembly assembly to the lead fuel assembly and to the Westinghouse Westinghouse Turbo fuel is presented in Table 2-1.

2-1. A summary summary of the lead fuel assembly program program mechanical mechanical compatibility compatibility evaluations evaluations is provided provided below.

The hydraulic compatibility compatibility is discussed in detail within Section 4 of this report. Hydraulic compatibility analyses for the AREVA Advanced CE-14 HTP fuel in a transition core with the WestinghouseWestinghouse Turbo Turbo fuel have calculated bounding crossflow velocity profiles by assuming a mixed-core configuration configuration that results in more severe crossflow velocities than in a realistic mixed-core mixed-core configuration.

configuration. These crossflow velocity velocity magnitudes magnitudes are within the AREV AREVA A experience experience base of transition cores with fuel designs having having HTP spacer spacer grids. The AREVAREVA A Advanced CE-14 HTP fuel assembly design maintains very similar hydraulic characteristics characteristics as the lead fuel assemblies by having the same axial grid elevations, elevations, grid strip heights, and fuel assembly pitch and envelope. The lead fuel assemblies are currently currently operating in their their third cycle cycle and have, to date, demonstrated over over two cycles of failure-free operation.

8

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION Comparison of Mechanical Table 2-1, Comparison Mechanical Design Features Features AREVA Advanced AREVA AREV Lead Fuel A Lead AREVA Advanced Westinghouse TurboTurbo CE-14 HTP Fuel Assembly CE-14 HTP Fuel Fuel Assembly Assembly Assembly Fuel Assembly Overall Length, 156.872 156.872 157 156.872 156.872 157 inch Fuel Rod Overall Length, inch 146.67 146.67 146.67 147.229 147.229 Nominal Assembly envelope at 8.109 8.109 8.109 8.109 8.109 8.109 Lower Tie Plate, inch Fuel Rod Pitch, inch 0.580 0.580 0.580 Number of Fuel 176 176 176 176 176 176 Rods/Assembly Rods/Assembly Corner Guide Number of Comer 4 4 4 4 4 4 Tubes/Assembly Tubes/Assembly Number of Center Guide (Instrumentation Tubes (Instrumentation 1 1 11 Tubes)/Assembly Tubes)/ Assembly Fuel Rod Cladding Material M5@

M5 M5@

M5 Zircaloy-4/ZIRLO Fuel Rod Cladding Outer 0.440 0.440 0.440 0.440 0.440 0.440 Diameter, inch Fuel Rod Cladding Thickness, 0.0265 0.0265 0.026 inch 0.0265 0.0265 0.026 Fuel Cladding Radial Gap, mil 3.25 3.25 3.5 3.5 Fuel Pellet Diameter, inch 0.3805 0.3805 0.3810 0.3810 Fuel Stack Fuel Stack Height Height (beginning (beginning 136.70 136.70 136.70 136.70 136.70 136.70 of life, cold,), inch inch Axial Blanket Length Length (top, N/A 6.00 (U0 (U022),

), 6.0,6.0 bottom), inch N/A 6.0,6.0 12.00 (Gad) 12.00 (Gad)

Corner Guide Tube Material Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Standard - Reduced Reduced MONOBLOCTM -

MONOBLOCTM Standard Standard - Reduced Reduced Outer Diameter in the Constant Outer Outer Diameter in the Comer Guide Tube Comer Guide Tube Type dashpot dashpot region, Diameter, increased dashpot dashpot region, region, constant constant wall wall thickness thickness in constant constant wall wall thickness dashpot dash pot region region thickness Corner Comer Guide Tube Outer Diameter 1.115 1.115 1.115 Diameter (upper), inch Corner Comer Guide Tube Wall 0.040 0.040 0.040 Thickness 0.040 0.040 0.040 Thickness (upper), inch Corner Comer Guide Tube Outer 1.048 1.115 1.048 1.048 1.115 1.048 Diameter Diameter (lower), inchinch Comer Guide Corner Guide Tube Wall 0.040 0.0735 0.040 Thickness (lower), 0.040 0.0735 0.040 Thickness (lower), inch 9

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION Table 2-1, 2-1, Comparison of Mechanical Mechanical Design Features (Continued)

AREA AREV CE-14AHTP Advanced AREVA AREV A Lad uel Lead Fuel AREVA Fuel Advanced Westinghouse Westinghouse Turbo CE-14 HTP Fuel Assembly Assembly Fuel Assembly Assembly Center Guide Tube Material Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Constant Outer Constant Constant Outer Outer Outer Constant Outer Center Guide Tube Type Diameter and wall Diameter and wall Diameter and wall thickness thickness thickness Center Guide Tube Outer Center Diameter Diameter (upper and lower), 1.115 1.115 1.115 1.115 inch Center Guide Tube Wall Thickness Thickness (upper and lower),

lower), 0.040 0.040 0.040 inch Number Number of Fuel Rod Spacer 8 8 8 8 8 8 Grids Grids Spacer Grid Material Fuel Rod Spacer Material . Zircaloy-4 HTP Zircaloy-4 Zircaloy-4 Zircaloy-4 HTP Zircaloy-4 and Type Zircaloy-4 HTP Alloy 718 HMP Grid Fabrication Fabrication Laser weld joining of Laser weld joining of Preformed interlocked Preformed interlocked Zircaloy-4 Zircaloy-4 Grid interlocking stamped stamped interlocking interlocking stamped egg crate crate fashion and straps straps welded together welded together Laser weld joining of of Alloy 718 Grid , N/A interlocking interlocking stamped N/A straps Grid/Guide Grid/Guide Tube Attachment Attachment Resistance-welded Resistance-welded Resistance-welded Resistance-welded Spot welded directly Zircaloy-4 Grid Zircaloy-4 Grid directly to guide tube directly to guide tube to guide tube Axially captured captured by by Zircaloy-4 rings on top top and of bottom of and bottom N/A Alloy 718 Grid N/A N/A grid; rings resistance-welded directly to guide tube Fuel Assembly Assembly The lead fuel assembly overall compatible with the dimensions of the reactor confirmed to be compatible overall length was confirmed reactor core internals (spacing between core support core plate) at beginning of life in both support plate and fuel alignment plate) cold and hot conditions. Additionally, positive engagement center/locking nuts and fuel alignment engagement of the center/locking alignment plate was demonstrated. An axial growthgrowth analysis confirmed assembly to reactor core confirmed adequate fuel assembly internals and shoulder gap margins up to the fuel rod and fuel assembly burnup limits. The lead fuel applicable for the AREVA assembly and fuel rod overall lengths remain applicable Advanced CE-14 HTP fuel AREV A Advanced design.

The array type, the number of fuel rods and guide tubes, and the fuel rod pitch dimensions dimensions are the same as for the Westinghouse Westinghouse Turbo fuel.

10 10

ATTACHMENT (4)

RELOAD TRANSITION RELOAD TRANSITION REPORT REPORT The square and diagonal widths of the lead fuel assembly assembly at the upper and lower lower tie plates and the spacer grids were confirmed to be be compatible compatible withwith the reactor core core internals, fuel storage racks, fuel elevator, and Westinghouse Turbo fuel. Further, the axial elevations of the nine spacer grids were confirmed to to have adequate overlap with the Westinghouse Turbo fuel. fuel. These dimensions and spacer elevations remain applicable for the AREV AREVA Advanced CE-14 HTP fuel design.

A Advanced design.

These evaluations These evaluations confirmed confirmed that the lead fuel assemblies were compatible with the reactor components and Westinghouse Turbo fuel in the reactor core. Additional evaluations of individual fuel assembly components were also performed performed as described below.

Upper Tie Plate The holes in the fuel alignment plate in the reactor core mate with the fuel assembly locking nuts. There are three are three basic alignment plate patterns - locations with no control rods and no instrumentation, instrumentation locations, instrumentation locations, and control rod locations. In the three cases, the holes form a 4.640" 4.640" square array, which matches the locking nut layout. The diameter of the locking nuts and the center nut (for instrumentation) are established instrumentation) established to allow sufficient clearance clearance with the fuel alignment alignment plate holes. The length of the locking nuts are also set to allow engagement engagement of the fuel assembly and the fuel alignment alignment plate at beginning of life hot conditions conditions and provide adequate clearance at end of life.

The underside The underside of the fuel alignment plate exhibits some protrusions in the form of socket head screws.

These screws These screws are lock-welded with a square stock lock-bar. The compatibility between are lock-welded between the fuel alignment plate and the upper reaction reaction plate of the fuel assembly assembly was demonstrated demonstrated by showing showing that the lock-bars and screws screws do not prevent the seating of the reaction plate against the fuel alignment plate. To ensure that the reaction plate does not interact with the screws, screws, eight notches are machined on the reaction plate to accommodate the height of the screw head.

The upper The upper tie tie plate plate was also evaluated was also evaluated withwith respect respect to compatibility compatibility with the fuel grapplesgrapples for fuel movement. Three types of grapples were evaluated: evaluated: the spent fuel grapple, refueling machine grapple, and the and the new fuel grapple.

new fuel grapple. It was shown that the grapples can fit over the center center hole in the reaction plate and that the reaction reaction plate arms fit within the grapples. The reaction reaction plate will not interfere with any part of the grapples. Review of the grapple grapple designs did not show any protrusions protrusions or unusual geometry that must be accommodated accommodated by the reaction reaction plate.

These These evaluations remain remain applicable applicable for the AREVA Advanced CE-14 HTP fuel design AREV A Advanced design since the upper upper tie plate design design is unchanged.

unchanged.

Lower Lower Tie Plate The The core plate and core plate and lower lower support support assembly assembly within the reactor reactor vessel provide four alignment alignment pins per per assembly assembly forming a 4.640" 4.640" square array. The mating holes in the fuel assembly lower tie plate are also on aa 4.640" square array. The 4.640" square array. The diameter diameter dimensions of the mating mating holes are sized to provide adequate clearance clearance with with the the alignment alignment pins.

pins. The The lower lower support support plate does not exhibit any protrusions protrusions within the confines confines of the core shroud other other than than the alignment alignment pins. The lower lower tie plate plate instrumentation instrumentation tube support was was also sized to be be smaller than than the maximum maximum dimension dimension provided provided by CE.CEo Since the the same same lower tie plate plate design is maintained maintained for the AREVA AREV A Advanced Advanced CE-14 HTP HTP fuel, the lower tie plate compatibility plate compatibility is assured.

11

ATTACHMENT (4) (4)

RELOAD TRANSITION TRANSITION REPORT REPORT Guide Tubes The radial locations of the guide tubes laterally within the lead lead fuel assembly, the inner diameters of of the guide tubes, and the weep holes diameters diameters were chosen to be the same as the Westinghouse Turbo fuel.

The axial locations of the guide tube dash pot and weep holes are also similar to the Westinghouse Turbo fuel. These critical dimensions assure that control element assembly (CEA) drop times and guide tube cooling are not affected by the introduction of the AREV AREVA A Advanced CE-14CE- 14 HTP fuel assemblies.

MONOBLOCTM corner guide tubes in the AREV The MONOBLOCTM AREVA A Advanced CE-14 HTP fuel assemblies have the same guide tube locations and dimensions as the lead fuel assemblies. Therefore, the compatibility evaluations performed for the lead fuel assemblies remain applicable.

2.3 MECHANICAL PERFORMANCE PERFORMANCE The AREVAREVA A fuel design planned for introduction at Calvert Cliffs is similar to the AREV AREVA A lead fuel assemblies that were introduced introduced at Calvert Cliffs Unit 2 in Cycle 15 (Reference 1) 1) which have since operated in both Units 1I and 2. They are currently operating operating in their third cycle with expected discharge discharge pin burnups burn ups greater than the AREVA fuel rod average burnup licensing limit of 62 MWd/kgU. The lead fuel assemblies assemblies were analyzed in accordanceaccordance with the NRC-approved generic mechanical design criteria criteria contained contained in EMF-92-116 (Reference (Reference 3) in conjunction conjunction with NRC-approved NRC-approved topical report BA BAW-10240 W-10240 (Reference (Reference 2). Reference incorporates the M5 Reference 2 incorporates M5 cladding cladding material properties that were previously previously approved approved by the NRC in BAW-I0227 BAW-10227 (Reference (Reference 4) into the Reference 3 methodology. All the mechanical mechanical design criteria were shown to be met up to the licensed fuel rod burn burnupup limit of 62 MWd/kgU MWd/kgU in EMF EMF-2807

-2807 (Reference 1).

1). The design improvements improvements that are mentioned in Section Section 1.1 relative to the lead fuel assembly design do not significantly significantly influence the fuel assembly structural characteristics characteristics that were were determined by prior mechanical testing of the lead fuel assemblies. Therefore, the AREV AREVA A Advanced Advanced CE-14 HTP fuel design, with expected expected structural behavior and projected projected performance, will meet design design requirements requirements throughout the life of the fuel.

The NRC-approved NRC-approved generic design criteria criteria used to assess the performance of the lead fuel assemblies were developed were developed to satisfy certain satisfy certain objectives (Reference 3).

(Reference These objectives objectives are used for designing fuel assemblies assemblies so as to provide the following assurances:

- The fuel assembly (system) (system) shall not fail as a result of normal operation operation and anticipated anticipated operational occurrences (AOO). The fuel assembly (system) operational occurrences (system) dimensions dimensions shall be designed designed to operational tolerances remain within operational tolerances and the functional capabilities of the fuels shall be functional capabilities be established established to either meet, or exceedexceed those assumed assumed in the safety analysis.

- Fuel Fuel assembly (system)

(system) damage damage shall never prevent control control rod insertion when it is required.

- The The number number of of fuel fuel rod rod failures shall be failures shall be conservatively conservatively estimated for postulated accidents.

accidents.

- Fuel coolability coolability shall always be maintained.

- The The mechanical mechanical design design of fuel assemblies shall be compatiblecompatible with co-resident co-resident fuel and the reactor reactor core internals.

- Fuel Fuel assemblies shallshall be designed designed to withstand withstand the loads from in-plant handling and shipping.

The The generic criteria criteria are applied applied to the fuel rod and fuel assembly assembly designs. These These criteria are listed in Table Table 2-2 2-2 along with the corresponding corresponding section section number number from Reference Reference 3. As noted noted in the specific items, some some of the criteria criteria specified below are for analyses other analyses other than the mechanical design mechanical design evaluations.

evaluations.

12

ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT REPORT Table 2-2, 2-2, Generic Mechanical Mechanical Design Criteria Reference 3 Reference Criteria Description Description Criteria Criteria Section Section 3.2 Fuel Rod Criteria Criteria Hydrogen Hydrogen content in components controlled to a minimum 3.2.1 3.2.1 Internal Internal Hydriding Hydriding level during manufacture to limit limit internal internal hydriding.

level during manufacture to hydriding.

3.2.2 Cladding Collapse Sufficient Sufficient plenum spring deflection and cold radial gap to 3.2.2 Cladding Collapse prevent axial gap formation during densification.

3.2.3 Overheating of Overheating confidence that fuel rods do not experience 95/95 confidence experience departure departure 3.2.3 Cladding from nucleate boiling (DNB) during steady state or AGOs. AOOs.

3.2.4 Overheating of Pellets Fuel ,

No centerline melting during normal operation and AOOs.

Overheating of Fuel 3.2.4 No centerline melting during normal operation and AOOs.

Pellets 3.2.5 Stress and Strain Limits Pellet / Cladding For M5 cladding, strain < 1% and no centerline melting.

For M5 cladding, strain < 1% and no centerline melting.

Interaction Interaction American Society of Mechanical Engineers (ASME)

American (ASME)

Section Section III, Appendix III Article 111-2000, III-2000, in combination combination with the specified 0.2%

0.2% offset yield strength and ultimate Cladding Stress strength of Zircaloy-4. M5 M5 stress limit based on bi-axial burst strength of cladding and buckling criteria at limiting overpressure transient at beginning overpressure beginning of life.

Not underestimated underestimated during LOCA and used in determination determination 3.2.6 Cladding Rupture 3.2.6 Cladding Rupture of 10 CFR 50.46 criteria. I 3.2.7 Fuel Rod Mechanical Mechanical ASME Section 1I1, Appendix F.

3.2.7 ASME Section III, Appendix F.

Fracturing Fracturing Models included Models in NRC-approved included in NRC-approved fuel performance performance codes Fuel Densification and 3.2.8 Swelling and taken into account account in analyses analyses contained contained in Swelling Sections Sections 3.2.2, 3.2.4, 3.2.5, 3.2.5, and 3.3.7 of this table.

3.3 Fuel System Criteria Criteria 3.3.1 Stress, strain, and loading limits on assembly components. (See (See 3.3.9 for handling and 3.3.1 3.4 for accident conditions.)

3.4 for accident conditions.)

Spacer Spacer Grid Lateral load < load limit.

Upper and Lower Tie Limiting loads occur during handling handling and postulated postulated Plates accidents.

3.3.2 Cladding Fatigue Cumulative usage factor for M5 cladding < 0.90 M5i!Y cladding 3.3.3 Fretting wear No fuel rod failures due to fretting wear.

Acceptable maximum oxide thickness. For M5 M5 cladding, 3.3.4 Oxidation, Hydriding, best estimate oxide < 100 microns. Effects of oxidation and 3.3.4 and Crud Buildup Buildup crud included in thermal and mechanical mechanical fuel rod analyses.

Stress analysis to include metal loss due to oxidation.

Lateral displacement displacement of the fuel rods shall not be of of 3.3.5 Rod Bow sufficient sufficient magnitude magnitude to impact thermal margins.

1 The swelling and rupture of the cladding is addressed in the approved LOCA models.

I The swelling and rupture of the cladding is addressed in the approved LOCA models.

13 13

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT Table Table 2-2, 2-2, Generic Mechanical Mechanical Design Criteria (Continued)

Reference 3 Reference Criteria Description Criteria Section Section 3.3.6 Axial Irradiation Growth Clearance Clearance remains between between fuel rod and upper tie Fuel Rod plate/lower plate/lower tie plate at end of life.

The fuel assembly assembly length shall not exceed exceed the minimum Fuel Assembly space space between upper and lower core plates in the cold condition condition at end of life.

Acceptable Acceptable maximum internal rod pressure. AllowableAllowable internal pressure not to exceed exceed system pressure pressure plus 3.3.7 Rod Internal Pressure 800 psia. When internal pressure exceeds exceeds system pressure, pressure, pellet-to-clad pellet-to-clad gap does not open during steady state or or increasing power.

increasing 3.3.8 Assembly Assembly Liftoff No liftoff from core lower support.

3.3.9Fuel Assembly Fuel Assembly 3.3.9 Handling Assembly withstands 2 1/2 112 times the weight as a static force.

Handling 3.4 Fuel Coolability Coolability Structural Maintain Maintain coolable geometry and ability to insert control coolable geometry control Deformations Deformations rods. ASME Section Section III, Appendix F.

Cladding 3.4.1 Cladding Include in LOCA analysis.

Embrittlement Embrittlement V iolent Expulsion of 3.4.2 Violent Expulsion of < 280 cal/gm energy energy deposition.

Fuel 3.4.3 Fuel Ballooning Ballooning Consider impact of flow blockage in LOCA analysis.

4.1 Thermal and Hydraulic Hydraulic Criteria Criteria Hydraulic 4.1.1 Hydraulic Hydraulic Compatibility Comp_atibility Hydraulic flow flow resistance similar to resident fuel assemblies.

resistance similar to resident fuel assemblies.

Thermal Margin 4.1.2 Thermal Margin 95/95 confidence confidence that fuel rods do not experience experience DNB.

DNB.

4.1.2_______ Performance 4.1.3 Centerline Fuel Centerline No centerline melting.

4.1.3 No centerline melting.

Temperature 4.1.4 Rod Bow Protect Protect thermal limits.

5.0 N eutronics Criteria Neutronics _______-accordance _withTechnicalSpecifications.

5.1 Power Distribution In accordance accordance with Technical Specifications.

Technical Specifications.

5.2 Kinetic Parameter Parameter Doppler Reactivity Negative.

Negative.

Coefficient Power Coefficient Negative Negative relative to hot zero power (HZP).

Moderator Temperature In accordance accordance with Technical Specification.

Specification.

Coefficient (MTC) 5.3 Control Rod Reactivity Reactivity Technical Specifications' margin maintained.

The fuel design objectives stated earlier include assurance assurance of fuel coolability and control rod insertability after a postulated postulated accident accident events. Reference 3 sections 3.2.7 and 3.4 pertain to these objectives.

objectives. Seismic 14 14

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT and LOCA LOCA analyses were performed for the lead fuel assemblies in order to verify that these criteria were analyses were (Reference 1) demonstrated satisfied. Analyses reported in EMF-2807 (Reference demonstrated that these criteria were met for the lead fuel assemblies. The lead fuel assembly analyses examined the various possible mixed row row configurations for the lead fuel assemblies along with the Westinghouse configurations Westinghouse Turbo fuel. The maximum grid impact forces occurring from Safe Shutdown Earthquake and LOCA events were combined using the Shutdown Earthquake square root sum of squares method and compared compared to the allowable strength for the HTP spacer grid.

allowable grid strength Results indicated that the combined maximum impact force was less than the allowable allowable grid strength.

established at a 95 percent confidence level on the true mean from the The allowable grid strength is established distribution of experimentally operating temperature.

determined grid crush data at the operating experimentally determined temperature. In addition, stresses in the fuel rods, guide tubes, and other fuel assembly components results also indicated that the stresses components seismic and LOCA-induced resulting from seismic LOCA-induced deformations are within acceptable limits. Therefore, Therefore, fragmentation of the fuel rod will not occur and the reactor can be safely shutdown under faulted condition loading. These conclusions were also shown to be valid under an Operating Basis Earthquake Earthquake event.

LOCA analyses performed for the lead fuel assemblies, accident loadings for Similar to the seismic and LOCA AREV A Advanced CE-14 HTP fuel assembly the AREVA assembly in the Calvert Cliffs core under mixed row and full row row configurations will be analyzed. These analyses serve to demonstrate fuel coolability and control configurations control rod insertability AREV A Advanced CE-14 HTP fuel assemblies in transition insertability for the AREVA transition cores with the Westinghouse Turbo fuel as well as in a full core configuration.

Westinghouse configuration.

AREVA AREV mechanical design criteria A intends to apply the generic mechanical criteria contained contained in EMF-92-116 (Reference 3)

EMF 116 (Reference to evaluate the design improvements to the lead fuel assembly design already operatingoperating at Calvert Calvert Cliffs.

AREVA In addition, AREV Gadolinia-specific fuel properties and design criteria contained in A will apply the Gadolinia-specific References 25 and 26 to evaluate References evaluate fuel rods containing containing Gd20 3 fuel pellets.

2.4 OPERATIONAL EXPERIENCE OPERATIONAL EXPERIENCE experience is an indispensable knowledge Operational experience knowledge base to demonstrate demonstrate the reliability and the relevance of such operational performance of a fuel assembly design. The relevance operational experience experience increases all the more in the case of a design with technical significantly different from all other designs.

technical features significantly Thermal Performance, or in short HTP, represents such a design. Whereas fuel assemblies equipped High IhermalE,erformance, equipped spacers employ springs and dimples to support each fuel rod in its spacer cell and have with traditional spacers mixing vanes along the top edges of the spacer strips which significantly significantly enhance thermal-hydraulic thermal-hydraulic performance, the HTP spacer represents concept in spacer represents an entirely different concept spacer design for PWR fuel. The spacer features strip doublets which are shaped HTP spacer shaped such that they not only serve as spring elements elements to produce curved firmly hold the fuel rods in radial alignment but also produce curved internal flow channels to achieve achieve the thermal-hydraulic performance.

desired thermal-hydraulic primarily the designation of a special type of spacer but is also used to performance is primarily High thermal performance denote a fuel assembly design in which this type of spacer is the major component.

component. The first insertion insertion was into a United States plant in 1988; 1988; the HTP design now possesses possesses 20 years of operational operational experience.

The AREV AREVA Advanced CE-14 HTP fuel assembly intended for implementation A Advanced implementation at Calvert Calvert Cliffs is an HTP type fuel assembly design with M5 FUELGUARD lower tie plate, and M5 fuel rod cladding, FUELGUARD MONOBLOCTM corner guide tubes. An overview MONOBLOCTM overview of both the overall operating experience experience gained with assembly design as well as the specific the various components of the fuel assembly experience in specific operating experience CE 14x14 plants is provided below.

15 15

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION 2.4.1 Operational Experience Operational Experience with HTP Fuel Assemblies As of December December 2008, the operational experience with HTP fuel assemblies comprises operational experience comprises a total of 10,502 fuel assemblies irradiated in 45 nuclear power plants. From these, 6,415 are in 27 European European plants (Belgium, France, Germany, Germany, Spain, Sweden, Switzerland, Switzerland, United Kingdom, Kingdom, and Netherlands), 4,003 assemblies in 15 United States plants, 80 assemblies in 2 Japanese assemblies Japanese plants and 4 assemblies in a Brazilian plant.

experience spans the entire range of fuel rod arrays from 14x This experience 14x14 14 to 18x18, 18x18, as well as reactors supplied supplied by various vendors, vendors, such as CE, Framatome, Framatome, Westinghouse, Westinghouse, Siemens, and B&W. The largest largest share, 4,405 fuel assemblies has been loaded into 12 ft Framatome/Westinghouse Framatome/Westinghouse plants with a 17x17 17x17 array, followed by the 16x16 16xl6 array for Siemens plants with 1,200 assemblies. The operational experience experience gained with HTP fuel covers a variety of core formations. Table 2-3 provides an overview. overview.

(

2-3, Operational Table 2-3, Operational Experience Experience Fuel of Fuel

1. of Maximum FuelFuel

of Fuel

of

of First First Assemblies Assemblies Assemblies Assemblies Plant type Assemblies Assemblies plants Insertion Insertion in operation accumulated burnup in operation accumulated in oeraion accuulaed MWd/kgU]

[MWd/kgU]

CE 14x14 5 1988 571 1,059 54 54 15xl5 CE 15x15 1 1988 204 724 53 53 CE 16x16 16x16 1 2008 8 8 <5

<5 Westinghouse 14x14 Westinghouse 3 1994 233 777 54 Westinghouse 15x15 Westinghouse 1 1991 157 702 57 Westinghouse 117x17, Westinghouse 7x 17, 6 1994 713 1,691 57 6 1994 713 1,691 57 12 ft Framatome 17x17, 12 ft 8 1993 550 2,714 67 B&W 15x15 5 2003 654 719 50 50 I

Siemens 15x15 15xl5 3 2001 324 388 70 Siemens 16x16 9 1989 889 1,200 59 59 Siemens 18x18 3 1992 407 520 61 Total 45 I 4710 10,502 70 70 performance or just HTP fuel assemblies High thermal performance assemblies have been loaded into reactors which are operated in significantly significantly different strategies strategies ranging from 6 to 24 month cycles. As of December December 2008, more than 4,500 HTP fuel assemblies equipped with Gadolinia rods have been loaded worldwide into 25 nuclear nuclear power plants. The number of Gadolinia rods within a fuel assembly varied between between 4 and 28 with Gd203 03 concentrations concentrations from 2 up to 8 wt%. 15xl5 15x15 and 17x17 HTP fuel assemblies with configurations ranging configurations ranging from 4 Gadolinia Gadolinia rods of 2 wt% to 24 Gadolinia rods of 8 wt% have been prepared prepared for Westinghouse Westinghouse type plants. A maximum maximuni fuel assembly averageaverage burnup bumup of 67 MWd/kgU MWd/kgU has been achieved achieved with HTP fuel assemblies containing Gadolinia poisoned assemblies poisoned rods.

16 16

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT The largest share of the HTP fuel assemblies up to now feature the bimetallic spacers spacers (Zircaloy-4 (Zircaloy-4 strips with Alloy 718 springs) at the outermost positions, all Zircaloy-4 HTP spacers spacers intermediate positions, at intermediate Zircaloy-4 Zircaloy-4 cladding and structural material, material, and FUELGUARD FUELGUARD debris filters. The so-called so-called bimetallic upper and lower grids are being phased out in the United States and replaced replaced with the Alloy 718 HMP grid.

With 5,783 5,783 fuel assemblies, more than half of all inserted HTP fuel assemblies have achieved a bumup burnup of of higher than 40 MWd/kgU. The maximum assembly burnup burnup is 70 MWd/kgU. The burnup distribution of of the HTP fuel assemblies as of December December 2008 is shown in Figure 2-1.

2-1.

2-1, Burnup Figure 2-1, Burnup Distribution of the HTP Fuel Assemblies Assemblies Number of Fuel Fuel Assemblies Total Number Total of Fuel Number of Fuel Assemblies:

Assemblies: 10.502 10.502 3.000-2.500-2.000-1.500.

1.000 500 '2K o

0 5 10 15 20 25 30 35 40 45 45 50 55 60 65 70 75 Bumup [MWd/kgU]

Assembly Burnup [MWdlkgUj 17 17

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION REPORT first insertion of the HTP fuel design with HMP Alloy The first Alloy 718 grids (straight flow flow channels) at the lower grid position was in 1998.1998. Today, significant operational experience with the HTP fuel fuel assembly featuring an HMP spacerspacer is available. Altogether, 4,463 4,463 such HTP fuel assemblies assemblies have have been loaded worldwide into 31 plants. Figure 2-2 shows the burnup distribution of HTP fuel assemblies featuring an HMP at the lowermost position as of December 2008. A maximum assembly burnup of 70 MWd/kgU has been achieved.

Figure 2-2, Burnup Distribution of Fuel Assemblies Featuring an HMP at Lowermost Position Number of Fuel Number Fuel Assemblies Total Number of Fuel Total Fuel Assemblies:

Assemblies: 4.463 700 600 500 400 300 300 200 200 100 100 0 5 10 15 20 25 30 35 40 45 45 50 55 60 65 70 75 Assembly Bumup [MWd/kgU]

Assembly 18

ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT The first HTP fuel assemblies equipped M5 fuel rod cladding were inserted into four plants in 2003 equipped with M5 assemblies into a South American

- four lead test assemblies American plant, four lead test assemblies into a United States consisting of 36 assemblies into a German plant with a 16x16 array, and one reload with plant, a reload consisting 85 assemblies into a United States plant of a 15x1515xl 5 B&W B& W design. As of December December 2008, 2,726 HTP fuel assemblies with M5 M5 cladding have been irradiated in 25 plants in Brazil, Germany, the Netherlands, Netherlands, Sweden, Switzerland, Sweden, South-America, and in the United States. The operational Switzerland, South-America, operational experience experience of the M5 cladding covers all arrays from 14x14 assembly and M5 combination HTP fuel assembly 14x14 up to 18x18. At this maximum assembly average burnup of 61 MWd/kgU has been achieved. Figure 2-3 shows the point, a maximum burnup distribution of assemblies equipped with M5 HTP fuel assemblies ofHTP cladding material as of December M5 cladding December 2008.

Figure 2-3, Burnup Distribution of Figure 2-3, HTP Fuel Assemblies having Fuel Rods with M5 ofHTP M5 Cladding Cladding Material Material Number of Fuel Assemblies Number Total Number Total Number of of Fuel Fuel Assemblies:

Assemblies: 2.726 400.

400 350 300-300 250-250 a

200-200 150-150 100-100 50-50 0 7 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 Assembly Assembly Bumup [MWd/kgU]

2.4.2 Operational Experience with M5 Operational M5 Cladding M5 alloy is the reference The M5 AREVA reference alloy of AREV A for fuel rod cladding cladding material. M5M5 is the result of a vast development which started at the end of the 1980's program of optimization and industrial development 1980's and reached reached completion at the beginning beginning of this millennium.

millennium.

nearly two and a half million fuel rods having 1993, nearly Since 1993, having M5 M5 cladding have completed operation completed their operation 10,141 fuel assemblies in 64 commercial reactors worldwide. These include 42 operating in 10,141 or are operating reactors in Europe France, Germany, Netherlands, Europe (Belgium, France, Switzerland, and United Netherlands, Spain, Sweden, Switzerland, Kingdom), 15 in the United States, 4 in China, 2 in South-Africa, and 1 in Brazil (Table 2-4).

The irradiation experience covers all fuel assembly irradiation experience assembly arrays ranging from 14x14 to 18x18, and different fuel assembly designs as AFA3G, HTP, Mark-B and Mark-BW. enriched natural uranium and Mark-BW. It includes enriched uranium fuel, both with and without Gadolinia. The range of enrichment extends reprocessed uranium enriched reprocessed particularly in Germany and in France.

from 3.2 to 4.95 wt% U-235. Mixed oxide fuels are also included, particularly 19

ATTACHMENT (4)

RELOAD TRANSITION REPORT RELOAD TRANSITION REPORT Table 2-4, Operational Experience Experience with M5 M5 Cladding Cladding Material Material Maximum Maximum Number of Maximum Fuel Number of First Status 12/2008 Fuel Array Reactors Irradiation Fuel Assemblies FIR Burnup F/R Burnup (MWd/kgU)

Assemblies Burnup Reactors Irradiation Assemblies (MWd/kgU) Burnup (MWd/kgU)

(MWd/k2u) 14x14 14x14 1 1993 2 54 49 49 Belgium 15x15 15xl5 1 1998 476 55 ,

50 50 17x17 17x17 1 2000 336 59 53 53 Brazil 16x16 16x16 1 2003 2003 4 49 44 44 China 17x17 17x17 44 1999 1999 1164 1164 53 48 France 900 MWe France 17x17 17x17 99 ' 1993 1993 158 80 58 58 France 1300 MWe MWe 17x17 8 1997 841 64 58 58 France N4 17x17 17xl7 44 2005 2005 692 42 39 39 15x15 1 2004 200 64 58 58 Germany 16x16 7 1993 1285 64 58 58 18xl8 18x18 33 1993 483 68 62 Netherlands Netherlands 15x15 15xl5 1 2004 120 56 50 50 South Africa 17x17 17x17 22 2002 312 62 56 56 Spain Spain 17x17 17x17 1 1999 4 51 46 46 15x15 15xl5 1 2000 232 65 61 Sweden 17x17 2 1998 378 64 58 58 Switzerland Switzerland 15x15 15xl5 1 2005 55 58 52 52 Kingdom United Kingdom 17x17 1 2008 84 14 13 13 14x14 2 2003 92 50 50 45 15x15 15xl5 7 1995 2033 2033 68 56 United States 16x16 1 2008 8 17x17 55 1997 1232 72 68 TOTAL 1 64 1 10141 1 1 20 20

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION shows the fuel Figure 2-4 shows fuel assembly assembly burnup distribution distribution with status as of December 2008. More than half half of the assemblies have achieved achieved burnups in excess of of 30 MWd/kgU, while 40 percent have achieved burnups in excess of of 40 MWd/kgU. Thus far, far, the maximum fuel assembly average average burnup achieved is is 68 MWd/kgU while the maximum fuel rod burnup achieved is 80 MWd/kgU.

Figure 2-4, Burnup Distribution of AREV AREVA A Fuel Assemblies Featuring M5 M5Fuel Rod Cladding Material Number of Fuel Number Fuel Assemblies 2000 Total number Total number of FuelFuel Assemblies: 10141 10141

~

1800 1600

~

1400

--=

1200 ~

1000 r~

800 hiiiiiiii'.i

~"...= ~

600 ~

400 200 0 ,-~

0-10 10-15 10-15 15-20 20-25 20*25 25-30 30-35 30-35 35-40 40-45 40-45 45-50 50-55 50*55 55-60 60-70 60-70 Assembly Burnup [MWd/kgU]

Assembly Bumup 2.4.3 Operational Experience with FUELGUARD Operational Experience FUELGUARD Lower Tie Plate The AREVA AREVA Advanced Advanced CE-14 HTP fuel assembly assembly design features the robust FUELGUARD FUELGUARD lower tie plate as an effective effective anti-debris filter to capture significant significant debris, thereby reducing thereby reducing the potential for fretting failures.

fretting failures. First introduced introduced in 1993 in the United United States at Robinson Robinson Unit 2, the FUELGUARD FUELGUARD lower lower tie plate design has now been been used at 14 United United States plants in batch quantities, and at another 5 United States plants as lead fuel assemblies. Over four-thousand four-thousand FUELGUARD FUELGUARD lower lower tie plates have been delivered delivered to date date in the United States States as shown in Table 2-5 below. Worldwide, Worldwide, 10,412 fuel assemblies have have been shipped with the FUELGUARD FUELGUARD lower lower tie plate.

21 21

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION To date, no known debris-related debris-related fuel rod failures have been attributed to debris passing through through the FUELGUARD lower tie plate.

FUELGUARD Table 2-5, Operational Experience with FUELGUARD Operatio':1al Experience FUELGUARD Lower Lower Tie Plate

Fuel Assemblies with Power Plant FUELGUARD FUELGUARD Comanche Comanche Peak 1 177 177 Comanche Comanche Peak 2 267 267 Kewaunee Kewaunee 172 172 Palisades 475 Millstone 2 352 352 St. Lucie 1 360 360 Ft. Calhoun 269 269 Robinson Robinson 2 606 Shearon Harris 1 687 ANO1 ANO 1 177 177 Crystal River 3 242 242 Davis Besse Besse 76 76 Oconee 2 68 68 Oconee Oconee 3 68 Palo Verde 1 8 Braidwood Braidwood 1 8 Calvert Calvert Cliffs 1 2 Calvert Calvert Cliffs 2 2 Sequoyah 1 4 Total 4020 2.4.4 Operational Experience Experience with MONOBLOCTM MONOBLOCTM Corner Corner Guide Guide Tubes The MONOBLOCTM MONOBLOCTM corner corner guide tube represents a new design feature for Calvert Cliffs, incorporating incorporating a solid tube design that features a constant outer diameter for the full length of the guide tube, and two inner diameters. The larger inner diameter at the top spans most of the tube length, transitioning to the smaller smaller inner diameter in the lower lower region of the guide tube where the smaller inner diameter diameter serves as the dashpot mechanism for control rod insertion. The thicker wall in the lower region increases bundle stiffness.

Worldwide, 20,818 fuel assemblies have been shipped with MONOBLOCTMMONOBLOCTM corner guide tubes made Zircaloy-4 material, and an additional 2,951 fuel assemblies have been shipped with the from Zircaloy-4 MONOBLOCTM MONOBLOCTM corner guide tube made from M5 M5 material. The MONOBLOCTM MONOBLOCTM corner guide tube design has also been utilized for guide tubes in multiple multiple lead fuel assembly programs programs in the United States States and is used for instrument tubes in all seven B&W plants in the United States. Calvert Cliffs will be the first application of the MONOBLOCTM MONOBLOCTM comer corner guide tube design in CE 14x14 fuel.

2.4.5 HTP Fuel Assembly Designs in CE 14x14 Plants As described in Section Section 1.2, the AREVA AREV A Advanced Advanced CE-14 HTP fuel assembly assembly design intended for application at Calvert Cliffs will incorporate incorporate Zircaloy-4 Zircaloy-4 HTP spacers, Alloy 718 HMP bottom spacer, M5 fuel rod cladding with Zircaloy-4 guide tubes, and the FUELGUARD M5 FUELGUARD lower tie plate design. As of of December 2008, 1,059 HTP fuel assemblies have been irradiated irradiated in CE 14x14 plants. Table 2-;6 2.6 provides provides 22 22

ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION details of the various configurations that were used in these HTP assemblies along with the proposed fuel details of the various configurations that were used in these HTP assemblies along with the proposed fuel configuration for Calvert Calvert Cliffs.

Table 2-6, Operational Experience and Designs of 14x14 Operational Experience 14x14 HTP Fuel Assemblies in CE Plants Characteristics Characteristics Zr-4 HTP Advanced Advanced Zr-4 HTP Status 12/2008 12/2008 spceHI Zr-4HTP Zr-4 HTP M5 cladding CE-14 CE-14 spacer at M5 cladding spacer at all all material All All M5 M5 Fuel HTP Fuel uppermost uppermost spacer at material positions positions (Calvert (Calvert position; HMP Cliffs)

Cliffs)

Zr-4/Inconel Zr-4/Inconel Uppermost Uppermost 718 Spring Zr-4 Zr-4 Zr-4 M5 M5 Zr-4

_ _ (Bimetallic)

(Bimetallic)

<a Q)

Inter-mediate Inter-mediate

~

S Flow MixerMixer

"'0 In-between

.~

.... In-between Zr-4 Zr-4 Zr-4 Zr-4 M5 M5 Zr-4 Q) Zr-4 Zr-4 Zr-4 Zr-4 Zr-4

~ (HTP) 0..

r:/) ~Zr-4/Inconel Zr-4/Inconel Inconel 718 Inconel 718 Inconel Loemst 78Sr-ingoe Inconel718 Zr-4 Zy-4 Inconel718 Inconel Lowermost (Bimetallic) 718 Spring (HMP) Zr-4 Zy-4 (HMP) 718 (HMP)

(HMP) (HMP) 718 (HMP)

(Bimetallic)

Cladding material Zr-4 Zr-4 Zr-4 M5 M5 M5 M5 M5 M5 Guide tube Zr-4 Zr-4 Zr-4 Zr-4 M5 M5 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 material FUEL-FUEL- FUELGUARD FUELGUARD FUEL-FUEL-Debris filter ---

--- FUELGUARDI1 FUELGUARD GUARD GUARD FUELGUARD FUELGUARD GUARD First insertion 1988 2001 2002 2003 2006 2011 2011 Maximum Assembly Burnup 46 54 54 54 54 45 45 21 21 5822 58

[MWd/kgU]

Number Number of Plants 1 2 I1 1 1 1 Total number of 2 613 352 4 88 962 2 613 352 4 88 96 2 assemblies Includes 72 assemblies without FUELGUARD.

FUELGUARD.

2 Subject to change based on actual operation and final core loading plan.

Of all the CE 14x14 HTP fuel reloads provided thus far, Millstone Millstone is the only plant to feature an all-Zircaloy spacer configuration with Zircaloy-4 spacer configuration Zircaloy-4 HTP spacers at the top and bottom locations along with all all intermediate locations. In the intermediate In addition, the lead fuel assemblies operating at Calvert Cliffs feature an all Zircaloy spacer configuration as well. Fuel failures due to grid-to-rod spacer configuration grid-to-rod fretting were observed at Millstone Millstone 23

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT Unit 2 Cycle 17 in two different different assemblies operating operating in their third cycle at locations adjacent adjacent to the baffle baffle wall. The cause of the failures was determined to be rod spinning spinning due to a loss of contact of the fuel rods with the grid structure at every primarily due to the fast relaxation rate of the Zircaloy-4 every grid elevation primarily Zircaloy-4 creepdown of the fuel rod cladding thereby leading to the formation of gaps between grids along with the creepdown implemented the Alloy 718 corrective action, AREVA subsequently implemented the fuel rods and the grids. As a corrective 718 HMP spacer at the lowermost location for this design at Millstone Millstone Unit 2 in order to prevent the recurrence of this failure mode. Of the 701 CE-14 HTP fuel assemblies recurrence assemblies supplied to date with the Alloy 718 HMP lower spacer spacer design, none have experienced grid-to-rod grid-to-rod fretting failures. In fact, the failures experienced at Millstone Unit 2 are the only instance of grid-to-rod experienced observed with the grid-to-rod fretting failure observed CE-14 HTP fuel assemblies to date. The AREV AREVA A Advanced Advanced CE-14 HTP fuel design for Calvert Cliffs Cliffs spacer at the lowermost include the Alloy 718 HMP spacer will include lowermost location.

AREVA has acquired AREVA examination data on all four of the lead fuel assemblies post-irradiation examination acquired post-irradiation assemblies that have have includes two cycles of irradiation for each of these assemblies.

Calvert Cliffs. The data includes been irradiated at Calvert Two of the lead fuel assemblies are currently undergoing a third cycle of irradiation and are expected expected to discharged with peak rod burnups of around 70 MWd/kgU.

be discharged MWd/kgU. These high bumup burnup lead fuel assemblies will be discharged in 2010 and inspected prior to loading a full batch of the AREVA Advanced CE-14 AREVA Advanced CE-14 HTP fuel. demonstrated failure-free fueL The four lead fuel assemblies have demonstrated operation up to fuel assembly failure-free operation assembly MWd/kgU. Results from the latest post-irradiation average burnups of around 45 MWd/kgU.

average examination show that post-irradiation examination M5 fuel rods as well as the Zircaloy-4 the M5 exhibited growth behavior that is consistent with Zircaloy-4 cage have exhibited AREV As models and predictions. This result was expected AREVAs expected since no unusual fuel rod or cage growth behavior has been noted to date on any CE 14x 14x14 supplied by AREV 14 fuel supplied AREVA. A. In addition, M5 fuel addition, the M5 excellent cladding corrosion rods indicate excellent performance consistent with the significantly corrosion performance significantly superior historical performance of the M5 corrosion and hydrogen uptake performance M5 alloy relative to Zircaloy-4.

Zircaloy-4.

3.0 3.0 NEITIIRONICS NEIITRONICS

3.1 INTRODUCTION

AND INTRODUCTION

SUMMARY

AND

SUMMARY

transitioning from Westinghouse Turbo fuel to AREV The effects of transitioning AREVA Advanced CE-14 HTP fuel on the A Advanced nuclear design bases and the methodologies for Calvert Cliffs are evaluated in this section.

safety parameters, such as power distributions, peaking factors, reactivity The specific values of core safety loading-pattern dependent. The variations in concentrations are primarily loading-pattern coefficients, and critical boron concentrations the loading-pattern parameters are expected to be typical of normal loading-pattern dependent safety parameters cycle-to-cycle normal cycle-to-cycle standard core variations in a standard variations parameters are also to be expected core reload. Slight variations in parameters expected due to the change in the methodology and codes used. The standard AREV change AREVA A codes and methodologies methodologies (References 5, 6, and 7), accurately predict (References Westinghouse Turbo fuel and predict the neutronics behavior of the Westinghouse AREVA AREV A Advanced CE-: CE- 14 HTP fuel during the transition period.

AREV A Advanced CE-14 HTP fuel design has significant nuclear design and operating experience The AREVA experience in the CE 14x14 fleet, including Millstone Unit 2, St. Lucie Unit 1, including Millstone I, and Ft. Calhoun. Further discussion discussion provided in Section 2.4.

experience is provided of operating experience 3.2 3.2 NEUTRONICS ACCEPTANCE NEUTRONICS ACCEPTANCE CRITERIA reactor is to ensure that fuel design limits will not be exceeded The purpose of the nuclear design of the reactor exceeded during normal operation anticipated operational operation or anticipated reactivity accidents will operational transients and the effects of reactivity damage to the reactor coolant pressure boundary or impair the capability to cool the significant damage not cause significant conformance with the requirements core and to assure conformance General Design Criteria (GDC).

requirements of General 24 24

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT The following GDCs address the transition to AREV AREVA A Advanced CE-14 HTP fuel describeddescribed in this section:

" acceptable fuel design limits be specified that are not to be exceeded GDC 10 requires that acceptable exceeded during normal operation, including the effects effects of AOOs.

GDC 11 requires, requires* that, in the power operating range, the prompt inherent nuclear feedback feedback characteristics tend to compensate characteristics compensate for a rapid increase in reactivity.

- GDC 12 requires that power oscillations that could result in conditions exceeding exceeding specified specified acceptable fuel design limits (SAFDLs) acceptable (SAFDLs) are not possible or can be reliably and readily detected detected and suppressed.

" GDC 28 requires requires that the effects effects of postulated reactivity reactivity accidents accidents neither neither result in damage damage to the reactor coolant coolant pressure boundary boundary greater than limited local yielding, nor cause sufficient sufficient damage damage to impair significantly the capability to cool the core.

To meet the GDC requirements requirements the following acceptance acceptance criteria are established as reflected reflected in References 3, 8, and 9:

References

1. Power distributions shall be in accordance

1. accordance with the plant Technical Specifications/Core Specifications/Core Operating Limits Report (COLR) (GDC 10). 10).

2. Linear heat rate (LHR) shall be in accordance accordance with the plant Technical Specifications Specifications (GDC 10).

3. Doppler coefficient shall be negative negative at all operating conditions (GDC 11). 11).

4. Power coefficient coefficient shall be negative negative at all operating operating power levels relative to HZP (GDC 11). 11).

5. Moderator temperature Moderator coefficient shall be in accordance temperature coefficient accordance with the plant specific Technical Specifications Specifications (GDC 11).

11).

6. The fuel design design and loading loading shall be such that uncharacteristic uncharacteristic power oscillations due to fuel design and loading loading do not occur occur (GDC 12). 12).

7. Margin to the Technical Specification Specification value value for minimum shutdown margin, with an allowanceallowance for a stuck most reactive rod, shall be maintained throughout the cycle (GDC 28).

3.3 METHODOLOGY METHODOLOGY The submittal core design was developed developed to provide, priorprior to the development development of cycle-specific cycle-specific designs, key safety safety parameters parameters to support the transition from Westinghouse Westinghouse Turbo fuel to AREVA Advanced AREVA Advanced CE-14 HTP fuel (see Table 3-1). These safety safety parameters will be used to provide provide an analysis-of-record analysis-of-record

, (AOR) for future reload specific analyses in order to assure assure that extensive extensive re-analysis will not be required required on a cycle-to-cycle cycle-to-cycle basis or require require an alternate alternate loading pattern just prior to installing the cycle-specific cycle-specific core design. It also provides assurance assurance that the plant licensing basis in the Technical Specifications, Technical Specifications, COLR, and Updated Final Safety Safety Analysis Analysis Report anticipated operation of the Report (UFSAR) are met for the anticipated AREVA AREV A Advanced Advanced CE-14 HTP fuel during transition and future cycles.

The nuclear nuclear design methodology and codes codes are changed to the standard AREVA methodology and code AREV A methodology package for the transition cycles and future operation of AREVA AREV A Advanced Advanced CE-14 HTP fuel.

References 5, 5, 6, and 7 are the NRC-approved NRC-approved topical reports outlining the approved AREVA AREVA neutronics methodology and codes. The following Safety Evaluation Evaluation Report (SER) constraints apply to the AREV AREVA A neutronics code code and methodology:

- The SAV95 SAV95 application will be supported by additional additional code validation validation to ensure that the methodology and uncertainties are applicable applicable for plant designs and incore monitoring systems differing from those listed below:

25 25

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION o Westinghouse Westinghouse reactors with 157 fuel assemblies with either 15x15 15x15 or 17x17 fuel rod arrays, and CE reactors with 217 fuel assemblies with a 14x14 fuel rod array.

o Incore monitoring with INPAX-2.

INPAX-2.

- Modifications Modifications to the code and methodology will be validated validated using the criteria approved approved in Reference Reference 5.

- The validation validation will be maintained maintained by AREVA AREV A and be available for NRC audit.

The above SER constraints have been met for the Calvert Cliffs transition to AREVA Advanced CE-14 AREV A Advanced CE-14 HTP fuel.

RTP Benchmarking Benchmarking of the AREVA methodology and codes was performed AREV A neutronics methodology demonstrated performed and demonstrated acceptable acceptable modeling modeling of previous and current Calvert Cliffs cores. Additional Additional benchmarking benchmarking of the 2009 2009 refueling refueling outage startup test procedure data confirmed confirmed accurate predictions by the AREVA AREV A code package.

AREV A predicts critical AREVA critical boron concentrations concentrations based on raw code predictions predictions with an additional boron bias based on the difference between between raw code predictions predictions and core follow data from previous cycles.

Key parameters parameters are calculated calculated as part of the submittal submittal core design neutronics analysis. These parameters are then biased in the safety analysis in order to create an analysis for recordrecord for the reload cycles. Key neutronics parameters calculated for the cycle-specific parameters are then calculated cycle-specific reload and compared with the values used in the AOR. If the key parameters parameters are not within the AOR, then the transient will be re-analyzed re-analyzed or re-evaluated evaluated on a cycle-to-cycle cycle-to-cycle basis using the stated methods. The results are reported in the UFSAR for that cycle.

Table Table 3-1, 3-1, Key Parameters Parameter Parameter Value Expected Limit Expected Limit F~ (without uncertainties)

Peak FrT 1.559 1.65 Peak LHR LHR (kW/ft) (with uncertainties) uncertainties) 12.81 14.3 14.3 Coefficient, (pcm/°F)

Doppler Coefficient, :::;-1.22

<- 1.22 <0.0

<0.0 Power Coefficient, (pcm/% Rated Thermal Thermal <0.0 ~ <0.0 Power)

Shutdown Shutdown Margin, Margin, (pcm)

(pcm) >_3500 2:3500 >3400 Moderator Temperature Moderator Temperature Coefficient Coefficient HFP, (pcm/oF)

Most Positive RFP, (pcm/°F) -3.71 <+1.5

<+1.5 Most Positive, <70% Rated Thermal Thermal +4.72 <+7.0 Power, (pcm/IF)

(pcm/°F)

Most Negative, Negative, (pcm/°F)

(pcm/oF) >-28.0 2:-28.0 >-30.0

>-30.0 3.4 NUCLEAR DESIGN NUCLEAR DESIGN EVALUATION EVALUATION A transition or submittal core design and two additional follow-on core designs have been developed for Calvert Cliffs Unit 2 to model the transition to AREVA AREVA Advanced CE-14 HTP RTP fuel.

The loading patterns patterns were were developed based on projected projected cycle energy requirements for Calvert Cliffs.

The loading patterns patterns have incorporated incorporated the approved Appendix Appendix K power up uprate rate and have been depleted at 2737 MWt. These These cycles were developed to be representative of future cycle designs to demonstrate 26 26

ATTACHMENT (4)

RELOAD TRANSITION RELOAD TRANSITION REPORT acceptable margins. The first transition cycle contains fresh AREV AREVA A Advanced Advanced CE-14 HTP fuel with once-burnt and twice-burnt Westinghouse Turbo fuel. The second transition cycle cycle contains fresh and AREVA once-burnt AREV once-burnt A Advanced CE-14 HTP fuel with twice-burnt Westinghouse Turbo fuel. The third twice-burnt Westinghouse AREVA transition cycle contains only AREV Advanced CE-14 HTP fuel. These cycles A Advanced developed to cycles were not developed be bounding of future cycle designs.

parameters were verified for the submittal core design in Table 3-1.

Key parameters 3-1. Figures 3-1 through 3-3 provide beginning of cycle (BOC) and end of cycle (EOC) fuel assembly the fuel loading pattern map, as well as beginning transition cycles. Figures 3-4 through 3-6 provide BOC, middle of cycle average burnup for the three transition cycle (MOC), and EOC assembly power maps for the transition cycles. The radial peaking (F!) and LHR for peaking (FJT) expected COLR limits in Figures 3-7 and 3-8, respectively.

transition cycles are plotted, along with the expected respectively.

compare the critical Figures 3-9 and 3-10 compare concentration and axial offset versus time in cycle for the critical boron concentration transition cycles.

The changes cycle-to-cycle variations changes in peaking factors are due to the normal cycle-to-cycle variations in these parameters. In Figures 3-7 and 3-8, F!FrT and LHR LHR (with uncertainties) expected COLR limits of 1.65 uncertainties) remain below their expected 14.3 kW/ft. The standard methods of fresh fuel enrichment and 14.3 integrated burnable poisons enrichment loading and integrated applied to control the peaking factors and maintain compliance will be applied Specifications compliance with the Technical Specifications and COLR.

andCOLR.

Changes Changes in boron concentration cycle-to-cycle variations in the core concentration and axial offset are typical of normal cycle-to-cycle design.

Incore monitoring is performed POWERTRAX system ,(Reference 5). Operation performed by the POWERTRAX Operation with POWERTRAX requires monitoring of the DNB overpower POWERTRAX overpower margin with the excoreexcore DNB axial shape index (ASI), thermal power, and CEA insertion limits. This is the same method currently allowed by Technical Technical Specifications. POWERTRAX system will operate in parallel with the existing core Specifications. The POWERTRAX approximately one year prior to loading monitoring system for approximately AREV A Advanced CE-14 HTP fuel into loading AREVA Calvert Cliffs reactors.

3.5 CONCLUSION

S CONCLUSIONS nuclear core design analysis of the submittal core design The nuclear transition from Westinghouse Turbo design for the transition fuel to AREV AREVA A Advanced CE-14 HTP fuel has confirmed peaking peaking factor and key safety parameters can safety parameters maintained within their specified limits using AREVA be maintained AREV A methodologies and codes. The key safety parameters generated with the submittal core design are used in the applicable parameters generated applicable analyses and evaluated to meet the acceptance acceptance criteria.

27

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT Figure 3-1, 3-1, First Transition Cycle Loading Pattern Pattern with BOC and EOC Assembly Burnups L N R sS T V w W x x z7 I1 Burnt Burnt I1 Burnt Burnt Feed Feed 11Burnt Feed Feed 1I Burnt Burnt I1 Burnt Burnt Feed Feed Reinsert (270i1 R15 (270") II T17 (90*I)

IT17 (90") X11I (0r xi (0") V18 (180")

(1308) 11 00 29520 28020 28020 00 28410 0 23440 20010 20010 1I Burnt Burnt 48330 48330 47290 47290 28660 2860 52020J 52020 28750 28750 _'47290

47290 43400 43400 23760 23760 (90,,)

R17 (90* 12 1 Burnt 1 Burnt 12 1 Burnt Feed Feed 11Feed Burnt Feed 1 Burnt Feed Feed 1 Burnt Feed Feed 30040 30040 1 R15 (180-) 39430 R15 (180") N18 (180")

N1w(130c) S16 (0")

S16 (0j) IH20 (180")

N20 (18o0l 39430 13 1 28020 28020 o0 29370 100 29660 0 21870 00 28urnt 2Burnt 147290 47290 27580 27580 52580 52580 280 28830 52960 52960 28170 28170 44990 44990 22670 22670 Vll Vll (90*)

(90") 14 11Burnt 48350 14 Feed Feed 1 Burnt Feed 1IBurnt Burnt Feed Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt 48550 V13 (180")

(180.) 54140 V13 W15 (90") W17(O")

W17 (0P Rll1 (0")

RI (0-) 54140 1510 15 29030 0 29030 o0 27130 27130 10 0 20420 204201 00 30790 30790 28660 52330 52330 28810 28810 51020 51020 250 28500 45720 27230 27230 44770 45720J 44770 11 Burnt Burnt Feed Feed 1 Burnt Feed 1I Burnt Feed Feed Feed 11 Burnt Burnt T17 R19 (270,,) S13 (90*)

16 16 T17 (0")(0*) R19(2Q70t1 S18 (270,,)

S18 (270c3 S13 (90")

28410 28410 52020 I28890o0 27300 51160 0

28810 29450 52770 00 Feed 0 . *30420 30420 52020 28890 51160 28810 52770 28930 28930 25900 41300 41300 Feed 0 Feed Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed 2Burnt 2Burnt N13 N13 (0")(0*) V16Q(90I)

V16 (90,,) Sig (0*)

S19 (0") R13 (0*)

R13 (0")

17 1710 29810 26340 48840 0 29810 0o 29400 0 26340 o 48840 28750 28750 53120 53120 28490 28490 52d740 .

52740 28490 48110 21160 54890 28490 48110 21160 54890

{1 Burnt Feed0 1 Burnt Feed 1 Burnt 1 Feed 1I Burnt Feed Feed 2Burnt 2Burnt XII (270") T19(0*)

T19 (0") (0 W16 (0") R20 (270")

R20 (270*)

18 XI1 (270*d 18 40790 1 [472901 23440 0 20460 0 26230 26230 00 40790 47290 28240 28240 457 45780 80J 28960 28960 48040 48040 21570 21570 48890 48890 I2Burnt 11 Burnt Burnt 11 Burnt Burnt Feed Feed Feed 2Burnt 19 V18 (90*)

V18 (90,,) X1 X133 (180,,)

(180.) X15 (90*)

X15 (90,,)

19 00 40660 20010 20010 43400 21070 21070 45030 27350 :296'0]

o00 21 0 40660 43400 45030 27350 25960 21180 48780 48780 Feed Feed Feed Feed 1 Burnt 1I BTunt Burnt 2Burnt 20 T11 T1OI (90,,)

(90) N16 (270")

N16 (270) N15 (0")

(0) 20 0 00 30220 30440 48870 48870 23750 23750 22880 22880 44400 41370 54940 1I Burn-t Burnt 2Bumt 2Burnt Residency Residency 21 T15 (270,,)

T1(2701 Wll (0I Wiit (0") Prev. Loc. (Count-Clock Rot)Rot.)

21 129950 29950 44840 44840 BOCBurnup BOC Burnup 39430 39430 50720 50720 EOCBurnup EOC Burnup M P 28 28

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT 3-2, Second Figure 3-2, Second Transition Transition Cycle Loading Pattern Pattern with BOC and EOC Assembly Burnups L N R S T V IIBurnt W xX zz 11 Burnt Burnt 11 Burnt Burnt Feed Feed 1 Burnt 1 Burnt Feed Feed 1 Burnt 1I Burnt Burn Feed Feed 11 11 Reinsert Reinsert 29520 Ttl TIT (0")

(W) 23750 N13 (0")

NI13 (0) XII (180")

X11 (180l V18 (130"I)I IV13 (180")

29520 28750 0 27530 27580 00 23750 1215501101 21550 0 1237501 I2BurntI 2Burnt 49130 49130 50430 50430 28600 51380J 51380 28990 28990 47670I 47670 4383043330 23480 23480 R20 (270") 12 IR209(701 Int 12 1 Burnt Feed Feed 1I Burnt jBurnt Feed Feed 1 Burnt Feed Feed 1 Burnt Feed Feed 444001 44400 51870I 13 13T Ttl (270*)

11 ~

(270,,) I R19 (270,,)

(9702) S13 (0`)um S13 (0") Tt9 (270")

T19 (2701) 51870 13 28750 0 27350 27350 00 23330 28830 00 21170 0 1I Burnt Burnt 50430 50430 27740 27740 50930 50930 28870 28870 52260 52260 28390 28390 43780 43780 22640 22640 (S3r 14 R15 (0")

f1 Burnt 14 Feed 1I Burnt Burnt Feed 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 1I Burnt Burnt 28810 23310 W5(0 W15 (90") Rll (180,,)

RI I (1301 N18 (270")

N18 (270") R17 (90")

R17Y(90°) 35530 15 15 o

0. 27230 0 23650 28650 00 282]40 28240 00 28490 23490 28600 28600 50870 50870 28330 52170]

52170 29130 29130 51530 51530 26160 26160 42480 42480 f1 Burnt Feed 1I Burnt Feed Feed 1 Burnt Feed Feed 1 Burnt Feed Feed 1 Burnt Burnt N13 (2701)

(270") T17 (130"j Tt7 (180") N20 (13o0 (180") s 18 (9o*)

S18 (90")

16 N13 27580 27530 0 28490 0 223301 22880 00 00 28960 23960 51380 51380 28930 28930 52080 28840 47410 47410 28730 25900 25900 39800 39800 1 Burntl Feed Feed 1I Burnt Burnt Feed Feed 11 Burnt Burnt Feed Feed 1 Burnt Feed 2Burnt 17 N16Q<o3l N16 (0") X13 (180,,)

(130D1 S19 (180") R13 (270")

R18 (270*)

17 0 28890 23390 00 22670 22670 00 25950 25950 I 00 45780 45730 28990 28990 52340 52340 29160 29160 47270 47270 27620 27620 47480I 47480 21370 21370 51710 51710 1 Burnt Feed Feed 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 2Burnt 2Burnt 18 XII (90")

Xl (9oo (90t V13 (90,,) (13o0 W16 (180") W13 (0")

(0Q) 23750 0 23170 28170 00 259001 25900 00 44990 47670 47670 28480 28480 51520 51520 28770 28770 47450 47450] 21100 21100 52440 11 Burnt Burnt I1 Burnt Burnt Feed Feed Feed Feed Feed Feed 2Burnt 2Burnt V18 (90") W17(90") N19 (]0' N19 (0")

19 V13(902) W17 (90I 21570 21160 21160 00 0 00 45030 45030 43840 43910 43910 26220 26220 25930 21390 21390 52490 52490 Feed Feed Feed 1 Burnt 11 Burnt Burnt 2Burnt 2011 T15 (270)

Tt5 (270,,) V16 (270")

(2701 V15 (90O)

V15 (90")

20 240 45730 0 0 235001 28500 28930 23930 45730 23480 23480 22700 22700 42520mt 42520 39790 39790 51670 51670 2Bumt 2Burnt 1 Burnt Residency Residency 21 X15 (90")

X15(90) (Q)

S16 (0") Prev. Loc.

Prevo Loc. (Count-Clock Rot)Rot.)

44760 23310 28810 BOC Burnup (MWdMTU)

(MWclMTU) 52220 35520 EOC Burnup Burnu (MWd/MTUL WdJMT M

M P 29

ATTACHMENT ATTACHMENT (4) (4)

RELOAD TRANSITION RELOAD TRANSITION REPORT Fignre Figure 3-3,3-3, Third Transition Transition Cycle Cycle Loading Loading Pattern with BOC and EOC Assembly Assembly BurnupsBnrnnps L N R S T vV w W xx zz 1I Burnt Burnt 1I Burnt Burnt Feed Feed I1 Burnt Burnt Feed Feed 11[Burt Burnt 1 Burnt 1 Burnt Feed Feed 11 11 (0' Tt5 (0`

T15 R15 (0' R15 (0)I Rll (0' RI I (0°)

00 xiI X11 (180)

(180, V 18 (270, V18 (2701 0o 22 Burnt 29130 28330 0 28600 28600 234801 23480 21100 21100 473o00 Burnt 49190 49190 50100 50100 28320 51740 51740 28550 28550 47300 44430 44430 23480 V11 (0' Vl47670 1(0l 12 12 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 11 Burnt S18 Burnt (1801 Feed Feed 1 Burnt]

Burnt 1 Feed[

Feed 47670 R15 (270`i (270, R19 (270°I (270, S18 (180, W17(O, 54730 54730 13 R15 2 Burnt 28330 28330 00 26220 0 28770 28770 0o 21370 21370 0o 28150 498 10Brt 51850 27740 44830 22650 2 Burnt 50090 50090 28150 49810 B urnt f1 28520 51850 27740 44830 22650 N15 (90`)

N15 (90' 14 I30I IBurnt Feed 14 Feed Feed 11 Burnt Burnt Feed 1 Burnt Feed Feed 1 Burnt Feed 11 Burnt Burnt 50870 151 15 1 0

w15 W15(90, 26160

o0) 0 NIS (180, N18 (180ý 284J80 28480 00 N13 (0)I N13 27740 27740 (0' o N'T16 (90, N16 (90ý 28930 55740 55740 28320 28320 49760 49760 28120 51780 51780 28890 28890 50990 50990 26490 42660 42660 1Burnt 11 Burnt Burnt Feed Feed 1 Burnt Feed

! 1 Burnt Feed Feed Feed 11Burnt Burnt 16 Rll (270, RIR1Q7oI 28600 28600 00 V13 (0' V13 (0")

28390 28390 0 N20 Burt 1N09701 (270, 122700 22700 0 o S16 S16 M) 28840 28840 (0'

51740 51740 28510 28510 51700 51700 28960 48080]

48080 28520 25240 39450 39450 Feed Feed Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed 22 Burnt Burnt 171 17 1 0

V16 (180")

V16 (180, 28730 28730 00 X13 (90'(90) 22640 22640 0 S19 S19 (0ý 25930 (0' o V17 V17 (90'(90o 47480 28550 28550 51800 51800 28850 28850 48030 48030 27940 47500J 47500 21400 53370 Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed Feed 2 Burnt 18 X11 (90' X1 IMD0I 23480 23480 00 Tt7 (0' T17 (0")

27620 27620 0 W16 (0I W16(O, 25900 25900 lol o R20 (180, 42510 47300 27780 50810 4480 21690 47300 27780 50810 28520 47480 21690 50440 50440 11 Burnt Burnt 1I Burnt Burnt Feed Feed Feed Feed Feed 22 Burnt Burnt 19 V18 (180")

V18 (180, T19 (180")

Tt9 (180, X15 X15 (180°)

(180, 19 21100 21390 21100 21390 0 0 3900 00 42480 44430 44430 44760 44760 26340 25220 21400 21400 50410 Feed Feed Feed Feed 11 Burnt Burnt 1 Burnt 22 Burnt Burnt 20 S$13 (2]70ý S13 (270, Ttl2T1 (180, T Tt818 (270°)

(270, 20 (180I 0 00 28870 28870 28990 23990 47450 47450 23480 23480 22750 22750 42580 42580 L29600 39600 53350 53350 1I Burnt Burnt 22 Burnt Burnt Residency Residency R17 (270, R17 Q2701 R13 (270')

(270, Prey. Loc. (Count-Clock Prevo (Count-Clock Rot.)

21 29160 50940 50940 BOC Burnup (MWdMTU)

Burnup (fulWdMTU) 37960 56030 56030 EOC Bumup (MWdWMTU)

EOCBurnu WdlMT M P 30

.ATTACHMENT ATTACHMENT (4)

RELOAD RELOAD TRANSITION TRANSITION REPORT Figure 3-4, First Transition Cycle Assembly powers at BOC, MOC, MOC, and EOC L N R sS T v V w W xx z 1Burntt IBBurn~t Feed 1 Burnt 1 Burnt Feed Feed 1 Burnt Feed 11FBurnt Burnt 1 Burnt Feed 0.841 0.862 08621 1.244 1.117 1.117 11291 1.291 111931 1.193 1.236 1236 1.134 1.134 11 0.841 1 10981 0.885 0.8851 0.915 09151 1.407 1.098 11339 1.339 1.072 1.072 1035 1.035 1075 1.075 1I Burnt Burnt 0.876 088 1.290 1.055 1.339 1.077 1.143_

0.876 0.888 1.290 1.055 1.339 1.077 1.046 1[046 1.143 0437 1 0.437 12 1 Burnt Burnt Feed 1I Burnt Feed Feed 1 Burnt Feed Feed 1 Burnt Feed 0.409 0 .409 Feed 0.496 0.862 1.184 1.184 1.078 1.078 1.275 1.275 1094 1.094 1.255 1.255 11771 1.177 11661 1.166 0.496 13 13 0862 10341 10915 0.915 1.351 1.088 1.088 1.362 1.362 1.063 11063 1.298 1.298 1.034 1.010 1.010 2Burnt 2Burnt 1.028 1.333 1BM3n 0.888 0.88 1.5 1.256 1.028 1.331 1.331 1.0611.061 1.333 1.040 1.055 1.05Ll 0.253 1 14 Feed Feed Feed 1I Burnt Burnt Feed 1 Burnt Feed r 1 Burnt Feed Feed 1 Burnt 0.244 0.306 15 1.244 1.244 1.083 1.273 11251 1.125 1.226 1.210 1.210 1.203 0.644 0.306 15 0.644 1.407 1.407 1.092 1.367 11041 1.104 1.331 1 1.170 170 1,298 1.298 10.6291 10.629 Burnt 1.290 101 1.324 1.349 1.170 1.264 1.290 1.031 1.324 1.078 1.349 1.136 1.264 0.685 1 Burnt Feed 1 Burnt Feed Feed I1 Burnt Feed Feed Feed Feed 1 Burnt 1.117 11.117 1.279 1.124 1.124 1.276 1.276 1.104 1.1041 1285 1.285 1.112 1.112 0.464 16 113551 1.098 16098 1.365 1.103 1.355 1.077 1.077 13661.366 1236 1.236 0.496 1.334 1.049 132 1,216 0.J55 1.055 1 1.331 31f 1.076 1.334 1.049 1.328 1.216 0.555 Feed 2Burnt Feed Feed 1I Burnt Burnt Feed 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 2Burnt 17 1.291 1.291 1.096 1.227 1.106 1.106 1.293 1 1.077 077 1024 1.024 0.260 17 1.339 1.339 1.064 1.331 1.347 1.078 1.078 1.3611 0988 1.361 0.988 0,968 0.968 0.274 1.339 100 1.049 1.283 0. 1 0,982 0.324 W&J 1.339 1.060 1.060 [JiJ1.347 1.049 1.283 0.971 0.982 2Burnt 0.324 1 Burnt Feed 1 Burnt Feed Feed 1 Burnt Feed 2Burnt Feed 18 1.192 1.192 1.262 1.215 1.215 1289 1.289 1.080 1080 1.108 1.108 0.381 0381 18 1 072 1.072 1.302 1.171 1.369 1.369 0.990 10990 0.969 10.969 0.358 0,358 1.077 1.333 1.134' 1.327 0 972 0.980 0.4088 1.077 1.333 1.134 1.327 0.972 0.980 0.408 1I Burnt 1I Burnt Feed Feed eed FFeed Feed 2Burnt 2Burnt 19 1.236 1.236 1.195 1.195 1.215 1.215 1.20 1.120 1.028 10382 0.382 19 1:2399 1035 1.035 1.043 1303 1.303 1.239 0.969 0.359 0.359 0.8 1.046 1.044 1.26 1.215 0.981 1.046 1.044 1.264 1.215 0.981 0.408 Feed Feed Feed 1 Burnt Burnt 11 Burnt 2Burnt 2Burnt 1.134 1.134 0.658 0.469 0.469 10.2611 1.075 1.184 1.019 0.638 0.261 20 0.499 10.2751 1.075 1.019 0.638 0.499 0.275 1.143 0 1.061 0.691 0.555 1.143 1.060 0.691 0.555 0.324 Urn 1 Burnt 2Burnt Residency Residency 0.443 0.443 0270 0.270 BOC Power Power 21 21 0.413 0.258 0.413

.500 0.258 0.321 MOCPower MOC Power 0.500 0.321 FE)C Power EOC Power M

M p P

31 31

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION Figure Figure 3-5, Transition Cycle Assembly powers 3-5, Second Transition powers at BOC, MOC, MOC, and EOC lL N R s S T vV w W x x zz I1 Burnt Burnt I1 Burnt Burnt Feed Feed I1 Burnt Burnt Feed Feed I1 Burnt Burnt I1 Brn Burnt Feed Feed 1.180 0.870 1110.870 0.935 0.985 1265 1.265 1.095 1.095 1.254 1.254 1.180 1.112 1.178 11731 1.217 1.217 11 0.909 1.017 1.017 1363 1.368 1.142 1.436 1.112 0.985 1.018 1.013 22 Burnt Burnt 0.919 0.919 0.993 0.993 1.298 1.298 1.05-5 1.055 1.298 1.298 1.060 1.060 1.017 1.113 1.113 0.351 0.3511 12 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed 11.017 Burnt Feed Feed 0.312 0.312 109351 12691 1.184 0.410 0.410 13 0.8 0.985 1.171 1.077 1.077 1.221 1.0311 1.081 1.269 1.157 1.184 13 113731 1.017 1.017 1351 1.351 1.120 1.120 1.430 1.430 1.112 1.112 1.373 1.017 0.979 11 Burnt Burnt 0.993 0.993 1.282 1.282 1.061 1.061 1.304 1.304 1.040 1.040 1.292 1.292 1.041 1.041 1.080 1.080 0.311 0.3111 14 Feed Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt 0.283 0.233 1.265 10311 1.081 1.192 1.070 1070 111.252 2521 1.094 1.094 1.165 1.165 0.660 0.378 15 15 1.265 1.368 1.368 1.122 1.359 1.119 1.119 14341 1.434 1.092 1.092 1.232 1.232 0.614 0.614 1.062

.1359 1.2053 1.298 1.29[ 1.062 1.316 1.060 1.060 1.318 1.050 1.0L50 1.253 0.712 0.712 1Burnt 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 1 Burnt Feed Feed Feed Feed 11 Burnt Burnt 1.095 1.225 10751 1.075 1.241 1.241 1.165 1.285 1.285 1.244 1.244 0.497 16 1.095 1.198 1.143 1.434 1.434 11221 1.122 1.375 1.1-52 1.152 1.377 1.377 1.198 0.480 04301 1.055 1.305 1.062 1.062 1.335 1.112 1.112 1.323 1.323 1.213 1.213 0.571 0.571 Feed Feed I1 Burnt Burnt Feed Feed 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed Feed 22 Brn Burnt 1.254 1.033 1.083 1.257 1.257 1.171 1.171 1.241 1.241 1.085 1.0385 1.117 0.274 0.274 17 1.254 12901 1.436 1.122 1.122 1.43-5 1.435 1.15-5 1.155 1.2751 1.275 0.962 0.944 0.257 0.257 1.298 1.298 1.040 1.040 1.317 1.317 1.113 1.310 1.310 1.002 0.328 1.113 1.002 1.018 0.328 11 Burnt Burnt Feed Feed 11 Burnt Burnt Feed 11Burnt Burnt Feed Feed 22 Burnt Burnt 1.101 1.0331 0.359 1.180 11301 1.277 1.277 1.101 1.290 1.088 1.088 1.033 0.359 18 1.112 1.112 1.377 1.094 1.094 1.379 0.963 0.921 0.318 0.313 1.060 1.060 1.292 1.292 1.0L50 1.050 1.322 1.002 1.002 1.015 1.015 0.397 0.397 Feed 11 Burnt Burnt I1 Burnt Burnt Feed Feed Feed 22 Burnt Burnt .

1.178 1.171 1.170 1.172 12491 1.249 11201 1.120 0.360 19 0.985 0.935 1.016 1.016 1.022 1.043 1.043 1.235 1.252 IBurnt L.199 1.199 1.212 10.9441 0.944 1.017 0.318 0.313 0.397 0.397 Feed Feed Feed Feed 1 Burnt 1I Burnt Burnt 22 Burnt Burnt 20 1.216 1.216 1.191 0.664 0.500 0.500 0.275 20 1.018 0.981 0.480 1.018 0.981 0.615 0.615 0.480 0.2531 0.258 1.113 1.113 1.080 1.080 0.711 0.570 0.570 0.328 0.328 22 Burnt Burnt 11 Burnt Burnt Residency Residency 21 0.350 0.350 0.312 0.312 BOC Power BOCPower 0.311 0.233 0.283 MOC Power MOCPower 0.409 0.377 0.377 EOC Power EOCPower M P 32

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION Figure Figure 3-6, Transition Cycle Assembly powers at BOC, MOC, 3-6, Third Transition MOC, and EOC lL N R S T V V W xX z Z

I1 Burnt Burnt 1I Burnt Burnt Feed 11lBurnt Burnt Feed I1 Burnt Burnt 1[ Burnt Burnt Feed Feed 0.964 1.029 1.029 1.203 1.045 12311 1.231 1.194 1.194 1.254 1254 1.172 1.172 11 0.964 .1203 1.888 0.949 1.043 1.428 1.422 1.131 1.409 1.088 1.017 1.017 1.028 22 Burnt Burnt 0.906 0.906 0.968 0.968 1.277 1.036 1.036 1.306 1.083 1.083 1.064 1.064 1.122 0.331 1 Feed 12 11 Burnt Burnt Feed 1 Burnt Feed 11 Burnt Burnt Feed 11 Burnt Burnt Feed Feed 0.299 11971 112601 11.237 1.196 0.390 0.390 1.028 1.028 1267 1.267 1.085 1.197 1.067 1.260 1.237 1.196 13 13 14331 112911 1.043 1389 1.389 1.145 1.433 1.102 1.291 1.035 1.35 0.975 22 Burnt B urnt 0.968 1.253 1.253 1.050 1.297 1.046 1.305 1.077 1.077 1.072 1.072 0.224 1 14 Feed Feed 11 Burnt Burnt 1Feed Burnt 11 Burnt Burnt Feed 11 Burnt Burnt Feed 11 Burnt Burnt 0.206 15 Ljj 1.085 1.102 1.061 0.277 0.277 1.203 1.203 1.085 1.189 1.061 12451 1.245 .11111 1.111 1.241 0.662 15 1.428 1.120 14241 1.083 1.428 1.145 1.145 1.372 1.372 1.119 1.119 1.424 1.083 Fee 1.227 0.601 1.277 1.277 1.050 1.050 1.303 1.053 1.053 1.318 1.062 1.062 1.252 0.697 0.697 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed 11 Burnt Burnt Feed Feed Feed 11 Burnt Burnt 16 1.045 1.045 1.196 1.196 1.061 1.254 1.254 1.217 1.217 1.266 1.266 1.159 0.479 16 1.131 1*1201 1.131 1.433 1.120 1.396 1.396 1.200 1.374 1.374 1.177 0.471 1.036 1.297 1.297 1.053 1.334 1.334 1.141 1.141 1.321 1.321 1.213 0.564 0.564 Feed 1Burnt Feed 11 Burnt Burnt Feed 1 Burnt Feed Feed 11 Burnt Burnt Feed 22 Burnt Burnt 12411 1.0885 1.230 1.230 1065 1.065 1.241 1.217 1.217 1.260 1.260 1.085 1.099 0.267 0.267 17 1.409 1*4221 1.305 0.974 1.024 1.409 1.101 1.422 1.200 1.200 1.305 0.974 0.954 10.954 00.258 258 1.306 1.306 1.046 1.046 1.318 1.142 1.313 1.313 1.005 1.005 1.024 0.328 0.328 11318 11 Burnt Burnt Feed Feed 1 Burnt Feed Feed *1*1Burnt Burnt Feed Feed 22 Burnt Burnt 1.194 1.261 1.103 1.264 1.264 1.085 1.085 1.124 1.124 0.383 0.383 18 1.194 1.088 1.088 1.293 1.293 1.079 1.375 1.375 0.975 0.955]

0.955 0.341 0.341 1.083_ 1.*307 10791 1.323 1.005 1.030 1.083 1.307 1.060 1.323 1.005 1.030 0.416 11 Burnt Burnt 11 Burnt Burnt Feed Feed Feed Feed 22 Bturnt Burnt 19 1.254 1.254 1226 1.226 1.226 1.155 1.155 1098 1.098 0.383 0.383 19 1.017 1.177 :Feed 0.342 1.017 1032 1.032 1.220 1.177 0.954 0954 0.342 1.064 1.064 1.076 1.076 1.251 1.215 1.215 1.026 1.026 0.417 0.417 Feed 1Burnt Feed Feed Feed 1 Burnt 1 Burnt 228urnt Burnt 20 1.172 1.172 1.196 1.196 0.656 0.478 0.266 0.266 20 1.028 0.258 1.028 1.122 0.979 106001 0.600 0.472 0.258 1.122 1.81 1.081 0.700 0.567 0.567 0.328 0.328 110.417 Burnt Burnt 22 Burnt Burnt Residency Residency 21 0.234 0.417 0.234 BOCPower BOC Power 21 0.372 0.215 0.372 0.215 MOCPower MOC Power 0.478 0.290 EOC Power EOCPower M

M P P

33 33

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION REPORT Figure 3-7, F~T Comparison 3-7, F Comparison versus Cycle Exposure for the Transition Transition Cycles 1.7 1.65 1.6 1.55 I-

~

LL LL 1.5 1.45 -

1.45 Cce1

--+-- Cycle 19

--- Cycle 20 1.4 - Cycle 1.4 ---..- Cycle 21 21

- Proposed COLR Limit Limit 1.35 1.35 1 2 3 4 5 6 00 100 100 200 200 300 300 400 400 500 500 600 600 700 700 EFPD 34

ATTACHMENT ATTACHMENT (4) (4)

RELOAD TRANSITION TRANSITION REPORT REPORT Figure Figure 3-8, LHRLHR Comparison versus Cycle Exposure for the Transition Transition Cycles 15 ~------------------------------------------------~

15 ..

14

-~ 13

-.::ac:

0:::

~.1212 a..

-+- -- 19 Cycle 19 11 - ----Cycle 11 20 Cycle 20 Cycle 21

.......- Cycle

- Proposed

-Proposed COLR Limit Limit 10 +=====~====~==~~----~----~----~----~

1 1 1 1 1 1 0o 100 200 300 400 500 600 700 700 EFPD 35 35

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION REPORT Figure 3-9, 3-9, Critical Boron Concentration Concentration Comparison Comparison versus Cycle Exposure Exposure for the Transition Cycles 2000 ~--------------------------------------------,

-8:

E

-0 c

o CL 1500 1500 -+- Cycle 19

--- Cycle 20

-.- Cycle 21

~

-C1000

1000

(.)

0co u

0c M 2 500 o

m 0o +-------r------.-------r------.-------,------.----~~

0o 100 200 300 400 500 600 700 EFPD Figure 3-10, Axial Offset Comparison Figure Comparison versus Cycle Exposure Exposure for the Transition Cycles 10 8

-+- Cycle 19 6

--- Cycle Cycle 20 20 00~

4

-.- Cycle 21 21

<<0 2 0

-2

-4 0 100 200 300 400 500 600 700 EFPD 36

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION 4.0 THERMAL-HYDRAUI.JCS THERMAL-HYDRAULICS

4.1 INTRODUCTION

INTRODUCTION AND AND

SUMMARY

The purpose of the thermal and hydraulic hydraulic design of the reactor is to ensure that the core can meet steady steady state and transient performance state requirements without violating performance requirements violating the acceptance acceptance criteria.

addressed in this section:

The following GDC are addressed

- GDC 10 requires requires that the reactor core and associated associated coolant, control, and protection systems be designed with appropriate designed appropriate margin to assure that SAFDLs are not exceeded exceeded during any condition condition operation, including the effects of AOOs.

of normal operation,

"* GDC 12 requires requires that the reactor core and associated associated coolant, control, and protection systems be designed to assure that power oscillations designed oscillations that result in conditions exceeding exceeding SAFDLs are not possible or can be reliably possible reliably and readily detected detected and suppressed.

To meet the GDC requirements the following acceptance acceptance criteria are established:

1. There There should a 95%95% probability probability at the 95% confidence level that the hot rod in the core 95% confidence core does not experience a DNB or boiling experience boiling transition conditions during normalnormal operation operation or AOOs.

Problems affecting departure from nucleate boiling ratio (DNBR) limits, such as densification

2. Problems densification and rod bowing, are accounted accounted for by an appropriate appropriate design penalty.

melting point will not be reached during steady state operation and AOOs.

3. The fuel melting design should address core

4. The design core oscillations oscillations and thermal-hydraulic thermal-hydraulic instabilities.

Technical Specifications

5. The Technical Specifications should ensure that the plant can be safely operated at steady steady state conditions under all expected combinations conditions combinations of system system parameters.

parameters.

6. The thermal-hydraulic thermal-hydraulic design should account for the effects effects of crud in the critical heat flux (CHF) calculations in the core calculations core or in the pressure pressure drop throughout throughout the Reactor Coolant System (RCS). (RCS).

4.2 METHODOLOGY METHODOLOGY The XCOBRA-IIIC XCOBRA-IIIC computer code (Reference(Reference 11)11) and mixed-core mixed-core methodology methodology (Reference 12) are used (Reference 12) used to calculate DNB performance, crossflow velocity, core pressure drop, and guide tube flow parameters. parameters.

A brief summary of the XCOBRA-IIIC code inputs is described described below:

The XCOBRA-IIIC model represents represents the full core with each fuel assembly modeled modeled as a hydraulic hydraulic channel.

channel.

The loss coefficients coefficients for the AREVA AREVA Advanced CE-14 CE-I4 HTP fuel and Westinghouse Westinghouse Turbo fuels are derived derived from pressure pressure drop tests performed performed in the AREVA AREV A Portable Hydraulic Hydraulic Test Facility. The pressure drop testing characterized characterized the bare bare rod friction factor and the component coefficients of the inlet component flow loss coefficients region (including (including the lower lower core plate, lower end fitting, and first s'pacer spacer grid), the spacer spacer grids, and the exit region (including (including last spacer grid, upper tie plate, and upperupper core plate).

Other inputs inputs for XCOBRA-IIIC XCOBRA-IIIC models include the axial and radial power power distributions and plantplant operating operating conditions conditions similar to those in Table 4-1. 4-1. The thermal-hydraulic compatibility analysis is thermal-hydraulic compatibility performed performed assuming either either severe severe operating operating conditions conditions or nominal operating operating conditions. The DNB performance performance and guide tube heating XCOBRA-IIIC XCOBRA-IIIC analyses are performed assuming performed assuming severe severe operating operating parameters, parameters, while the crossflow velocity, core pressure drop, and bypass flow XCOBRA-IIIC XCOBRA-IIIC analyses are performed performed assuming nominal nominal operating operating conditions. A top-skewed top-skewed axial power profile that bounds the 37 37

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT Calvert Cliffs DNB Limiting current Calvert Limiting Condition for Operation at 100% 100% power is assumed as the severe condition, while a centrally peaked shape is assumed as the nominal condition. condition. The fuel assembly and representative of the anticipated fuel pin radial power distributions used in the analyses are representative anticipated transition cycles. For analyses analyses performed under severe severe conditions, conditions, the hot assembly radial power factor is set such such Specification maximum Fl that the hot rod operates at the Technical Specification F,' plus uncertainties. The inlet mass assembly and the four face adjacent assemblies are penalized fluxes for the limiting assembly penalized 5% for analyses performed operating conditions.

performed under severe operating conditions .

. Several Several different reactor configurations are examined to support the various thermal-hydraulic reactor core configurations thermal-hydraulic analysis evaluations. XCOBRA-IIIC XCOBRA-IIIC models are created to reflect several several different reactor core configurations, including a reactor configurations, reactor core composed of all Westinghouse Turbo fuel, all AREVA Advanced AREVA Advanced CE-14 HTP fuel, 96 AREV AREVA A Advanced CE-14 HTP fuel assemblies, and 121 Westinghouse Westinghouse Turbo fuel mixed core), and a hypothetical assemblies (a first transition mixed scenario with one AREVA hypothetical scenario AREVA Advanced CE-14 CE-14 HTP fuel assembly and 216 Westinghouse Turbo fuel assemblies. The appropriate appropriate core configurations configurations are used to support the individual evaluations.

NRC-approved HTP CHF correlation The DNB analyses utilize the NRC-approved correlation with a 95/95 limit of 1.141 l.141 (Reference 14). The HTP Cl-F (Reference CHF correlation correlation was generically approved for use on CE-14 HTP fuel generically approved conditions fall outside of the approved range of applicability assemblies. If local conditions applicability for the HTP CHF correlation (MSLB) analysis], the Biasi or Modified

[e.g., main steam line break (MSLB) correlation [e.g., Modified Bamet Barnet C1IF correlations CHF correlations (Reference 13). A mixed core penalty are used (Reference Reference 12.

penalty is applied as required by Reference In DNB analyses, uncertainties on plant operating parameters (power, flow, pressure, temperature, In DNB analyses, uncertainties on plant operating parameters (power, flow, pressure, temperature, peaking) are handled in one of two ways:

- Deterministically applied in the most adverse condition, Deterministically condition, or

- Statistically Statistically combined using the methodology from Reference Reference 16.

The method used is dependent on the amount of DNB margin available for a particular transient transient event.

deterministic analysis is performed on each event. If the results demonstrate Typically a deterministic demonstrate unacceptable unacceptable re-analyzed using the statistical margins, then the event is re-analyzed statistical approach.

approach.

Uncertainties on fuel assembly/fuel Uncertainties parameters (cladding dimensions, pellet dimensions, rod pitch, assembly/fuel rod parameters etc.) are included in an engineering channel factor. This factor also includes engineering hot channel includes the effects of fuel densification.

The submittal core design was developed development of cycle-specific developed to provide, prior to the development cycle-specific designs, key safety parameters safety parameters to support the transition from Westinghouse Turbo fuel to AREVA Advanced AREVA Advanced Thermal-hydraulic analyses are performed CE-14 HTP fuel. Thermal-hydraulic provide an AOR performed for this design to provide AOR for future specific analyses. It also provides assurance that the applicable acceptance reload specific acceptance criteria criteria and the plant licensing licensing basis in the Technical Specifications, COLR, and UFSAR are met for the anticipated operation Technical Specifications, operation of the AREVA Advanced CE-14 HTP fuel during transition and future cycles.

4.3 HYDRAULIC COMPATIBILITY HYDRAULIC COMPATIBILITY thermal-hydraulic compatibility analysis of AREV This section documents the results of the thermal-hydraulic AREVA Advanced A Advanced CE-14 HTP fuel assemblies with Westinghouse Turbo fuel assemblies in the Calvert Calvert Cliffs reactor cores.

The thermal-hydraulic thermal-hydraulic compatibility compatibility analysis for Calvert Cliffs includes evaluations of: of:

Core pressure pressure drop

- Impact of crud on core core pressure drop 38 38

ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION REPORT

- Total bypass Total bypass flow

" Crossflow velocity Crossflow

- RCS flow RCS flow rate rate

- Guide tube heating Guide

" Control rod Control rod drop drop times times

" Thermo-hydrodynamic instability Thermo-hydrodynamic Table 4-1 listslists the thermal-hydraulic thermal-hydraulic design parameters for Calvert Cliffs. These parameters have been thermal-hydraulic analyses described in the following sections.

used in the thermal-hydraulic 4-1, Thermal-Hydraulic Table 4-1, Thermal-Hydraulic Design Parameters General Characteristics Characteristics Units Value - Unit 1 Value - Unit 2 Total Heat Output MWt 2737 2737 2737 (core only) MBTU/hr 9341 9341 Fraction of heat generated generated in fuel 0.975 0.975 Primary pressure, nominal psia 2250 2250 Inlet temperature Inlet temperature (HFP) OF 548 548 Total Total Reactor Coolant flow gpm 383550 382480 382480 (steady statel (steady state) Mlbm/hr 144.64 144.64 144.24 144.24 Core Coolant Flow Mlbm/hr Mlbmlhr 139.0 138.6 138.6 Hydraulic Diameter Diameter (nominal channel)

(nominal channel) ft 0.044 0.044 Average Mass Velocity Mlbm/hr/ft 2 Mlbm/hr/ft2 2.62 2.61 Core pressure Core pressure drop psid 28.89 28.89 Core average heat flux BTU/hr/ft22 181,728 181,728 181,728 181,728 Heat transfer Heat transfer area ft2 ft2 50116 50116 50116 Film coefficient Film coefficient at average average conditions BTU/hr/ft22/°F BTU/hr/ft /OF 6123 6109 Average film temperature Average film temperature drop OF 30 30 Average LHR of undensified undensified fuel rod kW/ft 6.29 6.29 6.29 Average core enthalpy rise BTU/Ibm 67 67 67 Maximum clad Maximum surface temperature clad surface temperature OF 648 648 648 Calculation Calculation Factors Factors Value Value Engineering heat flux factor (Fen,) (F eng) 1.03 Engineering factor on hot channel channel heat heat input 1.06 1.06 Rod Rod bowbow Cycle-specific Cycle-specific Fuel densification, Fuel densification, rod pitch, clad clad diameter diameter Included in Feng Feng Peak linear heat generation rate (kW/ft)

Peak 15.0 15.0 1I Azimuthal Azimuthal power tilt (Tq) (To) 1.03 DNBR Limit DNBRLimit 1.141 I Greater Greater than than the anticipated anticipated COLR COLR limit.

Core Pressure Pressure Drop Drop The Westinghouse The Westinghouse Turbo Turbo fuel assemblies assemblies have a higher overall overall resistance resistance to flow than the AREVA AREVA Advanced Advanced CE-14CE-14 HTP HTP fuel fuel assemblies; assemblies; therefore, therefore, as the plant plant transitions from a full core of Westinghouse Westinghouse 39

ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION Turbo fuel to a fullfull core core of AREV AREVA A Advanced CE-14 HTP HTP fuel, the core pressure drop decreases. An analysis was performed to assess the change change in core pressure drop associated with with the fuel transition.

The core pressure drop for a full reactor core of AREVA Advanced CE-14 HTP fuel assemblies is is presented in Table 4-1.4-1.

The total pressure drop associated with the full reactor core of AREV AREVA A Advanced CE-14 HTP fuel is is 1.0 psi lower than the total pressure drop of the Westinghouse Turbo fuel core.

Impact of Crud on Core Pressure Drop The impact of M5 M5 cladding oxide and crud deposition on AREV AREVA A Advanced CE-14 HTP HTP fuel assemblies are captured within the use of the as-designed inputs for pressure drop calculations.

Therefore, no adjustments to the pressure loss coefficients coefficients and surface surface roughness input are needed to cover the impact of the expected oxide buildup and crud deposition when performing pressure drop calculations based on as-designed inputs.

Total Bypass Flow The change in total bypass flow was examinedexamined to determine if the active heat transfer coolant flow will be adversely impacted by the fuel transition. The bypass flow includes the following flow paths: guide tubes, vessel upper head, inlet-to-exit nozzle, and core barrel/baffle.

barrel/baffle. The change in total bypass flow was determined by examining the change due to non-guide tube paths and guide tube paths. Bypass flow for determined the non-guide non-guide tube paths is affected by changes changes in reactor reactor core pressure drop, while guide tube bypass flow is dependent on both reactor core pressure drop, and assembly geometry. The reactor core pressure drop for a full core of AREV AREVA A Advanced CE-14 HTP fuel is lower than the reactor core pressure pressure drop for a Westinghouse Westinghouse Turbo fuel reactor core. As a result, the driving force for bypass flow decreases and the total bypass flow decreases.

The analysis analysis indicates indicates that bypass flow will decrease for both non-guide non-guide tube and guide tube paths. The total reactor core reactor core bypass flow will decrease by an insignificant amount as a result of the thermal-hydraulic thermal-hydraulic changes associated associated with the fuel transition. The active heat transfer transfer coolant flow will not be adversely adversely impacted.

Crossflow The crossflow The crossflow velocities velocities affecting the AREVAREVA A Advanced Advanced CE-14 CE-14 HTP fuel assemblies were analyzed analyzed to satisfactory hydraulic assure satisfactory hydraulic and mechanical mechanical performance performance during the transition. Different Different core configurations were considered configurations considered in the analysis; the configuration configuration consisting consisting of only one AREVA one AREV A Advanced CE-14 Advanced CE-14 HTP fuel assembly and 216 WestinghouseWestinghouse Turbo Turbo fuel assemblies results in more severe crossflow velocities velocities than a realistic mixed-core mixed-core configuration.

Although Although other geometries and operating other geometries operating conditions conditions may result in different different crossflow crossflow velocity profiles, this scenario scenario provides representative representative crossflow crossflow velocities velocities to cover cover core configurations configurations associated associated with the fuel transition. The results are representative representative of anticipated anticipated operating operating conditions conditions and are used used to develop bounding inputs inputs for mechanical mechanical analyses.

RCS RCS Flow Flow Rate An An analysis analysis was performed performed to assess assess the change change in primary system loop flow attributed attributed to the fuel transition.

transition. The The analysis analysis indicates indicates that the the transition transition from a full core of Westinghouse Westinghouse Turbo Turbo fuel to a full core of AREVA Advanced AREVA Advanced CE-14 HTP fuel results results in a 0.6% increase 0.6% increase in the RCS loop flow due to the 40 40

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION REPORT lower pressure drop drop of the AREV AREVA A Advanced CE-14 HTP HTP fuel. Thus, the thermal-hydraulic thermal-hydraulic changes resulting from the fuel transition will not impact the Technical Specification minimum loop flow rate resulting requirement requirement.

Guide Tube Heating Boiling of coolant within the guide tubes has the potential to increase corrosion rates and be detrimental for neutron moderation. An analysis was performed performed to demonstrate that boiling will not occur within the guide tubes of the AREVAREVA A Advanced CE-14 HTP fuel assemblies.

The guide tube boiling analysis is typically concerned with long-term, steady state conditions that result in long-term corrosion; for conservatism, severe operating conditions were used in the analysis.

Guide tube heating is most severe when a neutron absorbing material is inserted inserted into the guide tube. The analysis considered a high poweredpowered assembly with the control rods at the 100%100% power insertion limit.

The analysis demonstrates that control rod linear heat generation rates less than or equal to 4.69 kW/ft will preclude boiling within the guide tubes of the AREVA Advanced CE-14 HTP fuel assemblies. This linear heat generation cycle-specific basis.

generation rate is verified on a cycle-specific Control Rod Drop Time An analysis was performed to validate that the Technical Specification requirement for the control rod Technical Specification drop time is not challenged as a result of the fuel transition. The control rod drop time is primarily primarily dependent on the number, size, and location of the guide tube weep holes, as well as the inner diameter diameter and height of the guide tube dashpot region.

In the Calvert Cliffs fuel assembly design, the corner comer guide tubes account account for the majority majority of the frictional ofthe losses responsible responsible for the slowing down of the control rods. The configuration Westinghouse configuration of the Westinghouse Turbo fuel and AREVA AREV A Advanced Advanced CE-14 HTP fuel comer corner guide tubes is very similar. The most notable difference difference is the presence presence of a drain hole in the AREVA AREVA design and a fifth weep hole in the Westinghouse Westinghouse Turbo fuel design. The size of these holes is identical and the holes are located near the bottom dashpot region in each design. The difference in the placement placement of the holes will not significantly significantly impact the control control rod drop times.

Due to the similarities between between the Westinghouse Westinghouse and AREVAAREVA guide tube designs, designs, the control rod drop times are not significantly significantly impacted by the fuel transition.

Thermo-Hydrodynamic Thermo-Hydrodynamic Instability Instability Flow in heated boiling channels susceptible to several channels is susceptible several forms of thermo-hydrodynamic thermo-hydrodynamic instability.

These These instabilities are undesirable because because they may cause thermal-hydraulic thermal-hydraulic conditions that reduce reduce the margin margin to CHF during steady state flow conditions conditions or induce the vibration vibration of core components.

components.

Calvert Cliffs was evaluated for its susceptibility to a wide range of potential thermo-hydrodynamic Calvert Cliffs was evaluated for its susceptibility to a wide range of potential thermo-hydrodynamic instabilities. The features thatthat enhance enhance stable fluid flow conditions conditions include:

- Rod Rod bundle bundle core core configuration configuration -- resists parallel parallel channel channel instability.

"* Highly Highly subcooled subcooled operation operation -- aa power/flow powerlflow margin margin to to saturation saturation avoids avoids bulk boiling, thus preventing preventing two-phase two-phase driven driven dynamic dynamic instabilities.

instabilities.

- High High pressure pressure operation operation -- reduces density-driven effects associated reduces density-driven associated with localized localized steam formation.

41

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT

- Core channel pressure pressure drop-flow curve has a positive slope while the RCS pump head-flow head-flow curve negative,- prevents Ledinegg flow excursion instability.

is negative-

" Margin to CHF - avoids avoids boiling crisis and film-boiling induced induced instabilities.

- Low boiling boiling number number - provides margin to the inception inception threshold for acoustic or density waves.

- U0 U02 fuel with a long time constant - resists void-reactivity void-reactivity feedback coupling coupling with thermo-hydrodynamic oscillations.

The transition to AREVA AREV A Advanced Advanced CE-14 HTP fuel will not adverselyadversely impact any of thesethese features.

Consequently, the thermo-hydrodynamic thermo-hydrodynamic stability stability of the core will not be affected by the transition transition to AREV A Advanced CE-14 HTP fuel assemblies.

AREVA 4.4 TRANSITION TRANSITION CORE PERFORMANCE PERFORMANCE 4.4.1 Transition Transition Core DNBDNB Performance XCOBRA-IIIC XCOBRA-IIIC was used to analyze the effect effect of the fuel transition transition on the DNB performance performance of the AREVA AREV A Advanced Advanced CE-14 HTP fuel assemblies. The power level was selected selected to achieve a minimum DNBR close to the HTP CHF correlation correlation limit (Reference (Reference 14). A mixed core penalty penalty (Reference 12) was (Reference 12) applied to all core configurations, configurations, including the full core of AREVA AREV A Advanced CE-14 HTP fuel.

AREVA The AREV A Advanced CE-14 associated with less overall flow resistance CE-14 HTP fuel assembly is associated resistance than the Westinghouse Turbo fuel. This results in flow transferring from the Westinghouse Turbo fuel into the Westinghouse AREVA AREV A Advanced Advanced CE-14 HTP fuel, which is beneficial beneficial for the DNB performance performance of the AREVAAREV A Advanced Advanced CE-14 CE-14 HTP fuel. As a result of a mixed mixed core configuration, configuration, the DNB performance performance of the AREVA AREV A Advanced Advanced CE-14 HTP fuel will improveimprove by as much as 6.7% 6.7% relative to the CHF correlation correlation limit. This improvement improvement in DNB performance performance does not apply to core configurations configurations consisting of all AREVA AREV A Advanced CE-14 HTP fuel.

A supplemental supplemental analysis examines the DNB performance performance at various elevations.

elevations. The AREVA Advanced AREV A Advanced CE-14 HTP fuel is characterized characterized by higher pressure pressure loss coefficients than the Westinghouse Westinghouse Turbo Turbo fuel at the inlet region and first several several spacer grids, and lower loss coefficients coefficients for the remainder remainder of the assembly. Although Although the overall overall flow resistance resistance of the AREVA AREV A Advanced Advanced CE-14 CE..; 14 HTP fuel is lower lower than the Westinghouse Turbo fuel, severely bottom-skewed bottom-skewed axial shapes can force minimumminimum DNBRDNBR to occur occur at elevations where difference in loss coefficients where the difference coefficients will force flow from the AREVAREVA A Advanced CE-14CE-14 HTP fuel into the Westinghouse Westinghouse Turbo fuel. Under these conditions, conditions, the DNB performance of a transition core is more limiting than a full core AREVA core of AREV A Advanced CE-14 HTP fuel. However, these conditions conditions are non-limiting with respect to DNB performance due to the small enthalpy rise at the location of of maximum heat flux. Exit peaked shapes of similar magnitude magnitude are much more limitingli~iting from a DNB perspective.

4.4.2 Fuel Rod Bow Bow The impact of fuel rod bowing on DNB performance is addressed addressed in the NRC-approved NRC-approved methodology methodology in (Reference (Reference 15). The effect of fuel rod bow is manifest as a burnup-dependent bumup-dependent penalty on minimum DNBR. Fuel rod bow penalties are analyzed and applied on a cycle-specific cycle-specific basis.

4.4.3 DNB DNB Propagation Propagation Propagation Propagation of DNB failures needs to be considered considered for PWRs when two conditions exist simultaneously:

- When the DNB limiting rod of a bundle is calculated calculated to have a minimum DNBR DNBR below the 95/95 limit value of the CHF correlation correlation being used; and, 42

ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION REPORT

- When the the internal pressure of of the DNB limiting rod exceeds exceeds core pressure at the time of of minimum DNBR.

Departure from nucleate boiling propagation is addressed by the NRC-approved NRC-approved methodology in (Reference 8). For Calvert Cliffs, DNB propagation is not a concern due to the low power of fuel rods rod internal pressure greater than the RCS pressure. This is verified on a cycle-specific basis.

with a rodintemal 4.4.4 Impact of Crud on DNB DNB Performance The HTP CHF correlation has been developed from CHF testing of electrically electrically heated rods with no simulation of crud deposition. This has been the standard procedure for CHF testing. The HTP CHF correlation is applied in ONB DNB analyses with no adjustment for the possible presence presence of crud. The presence of crud on the fuel rod surface adds a small, additional re~istance resistance to the heat transfer from the fuel rod surface surface to the coolant. However, the heat transfer mechanisms from the surface of the crud to the coolant are the same as those for a fuel rod surface surface without crud. Since the heated perimeter is not changing significantly with typical crud deposition, the temperature drop from the crud outer surface to the coolant is essentially essentially the same as that for a fuel rod surface without crud. Consequently, Consequently, the probability of the formation of a vapor film does not change significantly. The accumulation accumulation of crud is is a

generally small for PWR fuel as a result of the chemistry controls and core design constraints constraints currently in use for managing the risk for crud accumulation and its secondary consequences consequences (crud-induced (crud-induced power shifts and crud-induced localized corrosion). However, if during the course of plant operation operation a severe crud deposition event was detected or projected that would lead to a significant RCS flow reduction, then this flow rate 'reduction reduction could reduce the thermal margin predicted by the CHF correlation. Appropriate action would then be taken to restore or ensure adequate adequate thermal margin. Such an event is unlikely in light of the chemistry controls currently in use for managing the risk for crud related issues.

4.4.5 Verification Verification of TMLL The thermal margin limit lines (TMLL) are a series of isobars in power and core inlet or average average temperature temperature that establish establish the operating operating frontiers for these parameters such that DNB ONB in the core and hot leg saturation are leg saturation are both both avoided. The lower power, or flatter region of the TMLL, is established established by the requirement requirement that reactor hot leg saturation must be prevented. The steeper sloped, higher power power region is is established established by the requirement requirement that the ONBR DNBR limit must not be exceeded.exceeded.

I Thermal Thermal margin limit lines are not necessarily necessarily verified cycle-by-cycle basis. The TMLL verified on a cycle-by-cycle TMLL are not not affected by most cycle-specific affected by most cycle-specific changes, but need to be verified if:

if:

- there there are changes which may affect the DNB performance of the fuel, ONB performance

- there there are changes in uncertainties associated with the inlet temperature, uncertainties associated temperature, core core pressure, pressure, and power,

- there there are changes changes to other plantplant variables variables or uncertainties uncertainties (other (other than inlet temperature, temperature, RCSRCS pressure, pressure, and power) that affect the DNB ONB or core exit temperature temperature (such as flows, radial peaking peaking factors, ONB DNB correlation, correlation, etc.).

etc.).

The The validation validation of the the Calvert Calvert Cliffs TMLL TMLL was performed performed using XCOBRA-IIIC XCOBRA-IIIC and and the HTP CHF correlation correlation within the NRC-approved methodology in Reference 16. The analysis NRC-approved methodology Reference 16. The analysis demonstrated that demonstrated that the current TMLL current TMLL remain valid valid for AREVA Advanced AREV A Advanced CE-14 HTP fuel assemblies in both full-core and and transition transition core configurations.

configurations.

43

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT REPORT 4.5 FUEL ROD THERMAL PERFORMANCE PERFORMANCE 4.5.1 Fuel Centerline Centerline Melt Melt centerline melt (FCM)

Fuel centerline calculated using the RODEX2 code (References (FCM) limits are calculated (References 17 and 18), and the NRC-approved NRC-approved methodology from Reference Reference 16. The purpose of the FCM calculation calculation is to generate generate a maximum allowed kW/ft LHR LHR limit such that melting in all pin types of any composition composition in the core is precluded throughout throughout the cycle. This analysis is performed each cycle cycle for all reloads utilizing Gadolinia-bearing fuel.

4.5.2 Fuel Rod BowBow The impact impact of fuel rod bowing bowing on fuel rod thermalthermal performance performance is evaluated evaluated with the NRC-approved NRC-approved methodology in Reference Reference 15.

15. The effect of fuel rod bow is manifest as a burnup-dependent burn up-dependent penalty penalty on on the local peaking factor (F (Fq).

q). Fuel rod bow penalties penalties are analyzed and applied on a cycle-specific cycle-specific basis.

4.6 CONCLUSION

CONCLUSION The thermal-hydraulic thermal-hydraulic evaluation of the fuel transition at Calvert Calvert Cliffs indicates indicates that the AREVAAREV A Advanced CE-14 HTP fuel assemblies and Westinghouse Westinghouse Turbo fuel assemblies are hydraulically hydraulically compatible.

5.0 PLANT PLANT SYSTEMS The potential effects of the AREVA fuel transition transition were evaluated for the following:

- Normal Operation Shielding Shielding and Personnel Exposure Exposure

- Qualification (EQ)

Radiological Environmental Qualification (EQ)

- Post-LOCA Access to Vital Areas

- Radioactive Waste Systems

- Fuel Storage Racks The impact of transition transition to AREVA Advanced CE-14 HTP fuel on Chapter 14 accidents or transients AREV A Advanced transients is provided in Section Section 6.

Normal Operation Operation Shielding and Personnel Exposure AREVA The transition to AREV A Advanced Advanced CE-14 HTP fuel will not increase expected radiation levels and will increase expected will not affect radiation radiation zoning or shielding requirements requirements in the various areas of the plant. It is expected expected that the reduction/elimination of grid-to-rod-fretting reduction/elimination grid-to-rod-fretting induced peripheral fuel failures as a result of the adoption of the HTP fuel grid design will result in reduced personnel exposure reduced personnel exposure during outages by reducing reducing the noble gas source term.

Radiological Environmental Environmental Oualification Oualification In accordance accordance with 10 CFR 50.49, safety-related safety-related electrical equipment must be qualified qualified to survive the environment at their specific location during normal operation and during an accident.

radiation environment The Containment Containment and Auxiliary Buildings are divided into various rooms for environmental environmental zoning purposes. The radiological environmental conditions environmental conditions noted for these rooms are the maximum conditions expected to occur. The current normal operation operation values represent 60 years years of operation, while the AOR AOR post-accident post-accident radiation exposure exposure levels are determined determined for a one-year one-year period following a LOCA using 44 44

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT Regulatory Guide 1.89 source term tenn assumptions and a core power level of 2738 MWt (bounds Technical Specification Specification limit).

For the transition to AREVA AREVA Advanced Advanced CE-14 HTP RTP fuel, the EQ accident source reanalyzed source term was reanalyzed with the same corecore power level and release assumptions as before. The AREVA transition source term AREV A transition* tenn was compared to the AOR to develop integrated integrated energy ratios from various sources (airborne, sump, iodine filters, etc.)

etc.) for each Containment Containment and Auxiliary Building rooms. The AOR was found to be bounding for the first cycle of AREVA AREV A fuel in each unit. Additional analyses analyses will be perfonned performed to ensure that equipment equipment remains within qualified doses as further reloads increase the quantity of AREV AREVA A fuel to a full core core load. As discussed discussed above, the normal operation contribution to the EQ dose is not impacted nonnal operation contribution impacted by the transition to AREV AREVA A Advanced Advanced CE-14 RTP HTP fuel.

Post-LOCA Access to Vital Areas Vital access dose considerations considerations are evaluated evaluated against GDC 19, 10 CFR Part 50, Appendix Appendix A, as amplified in NUREG-0737, NUREG-0737, Item II.B.2. Specifically, Specifically, the design design dose for personnel personnel in a vital area should not exceed 5 rem whole body, or its equivalent equivalent to any part of the body, for the duration of design basis accidents. The source source terms tenns utilized in existing time-motion time-motion study dose analyses and Chapter 11 time-dependent radiation dose rate maps are the same as those used for equipment dependent equipment EQ. Therefore, Therefore, for the same saine reasons indicated above for EQ dose, the transition transition to AREVA AREV A Advanced Advanced CE-14 HTP RTP fuel will not have an impact impact on vital access requirements.

Radioactive Waste Radioactive Waste Systems The waste processing systems systems are designed to provide provide controlled controlled handling and disposal of radioactive liquid, gaseous, gaseous, and solid wastes from both units. Design criteria were establishedestablished to maintain maintain the release release of radioactive radioactive material from the plant to the environment environment at levels which are as low as reasonably reasonably achievable.

The design of the waste processing systems was based based upon processing processing reactor coolant and miscellaneous miscellaneous waste during operation With operation with 1%1% failed fuel. The annual radioactive radioactive waste releases releases for this design were shown to meet the dose guidelines guidelines of 10 CFR Part 50, Appendix I.

All releases meet the Offsite Dose Calculation Manual (ODCM) limits. The ODCM is a program Manual (ODCM) governed by requirements governed requirements described in Technical Technical Specifications.

Specifications. It provides limits for offsite radioactive waste releases, calculational methods to detennine determine those releases, and alternative alternative methods of accounting accounting for and controlling release of radioactive materials. By meeting the ODCM limits, the guidelines guidelines ofof 10 CFR Part 50, Appendix I will be met. This is confinnedconfirmed by the effluent data and doses reported to the NRC in the Radioactive' Radioactive' Effluent Release Release Reports required by the Technical Specifications and Technical Specifications 10 CFR 50.36a. The transition to AREVA Advanced CE-14 RTP AREVA Advanced HTP fuel is not expected to adversely adversely impact the the quantities quantities of liquid, gaseous, or solid radioactive wastes, and will not alter the ODCM requirements.

Fuel Storage Racks New AREVA AREV A Advanced CE-14 HTP RTP fuel arriving at the site will be removed removed from its shipping container and transferred to the new fuel storage racks. These dry storage racks are for both units and are constructed to provide storage for two-thirds constructed two-thirds of a core (144 (144 assemblies). These racks are currently currently licensed licensed to store new CE 14x14 14xl4 fuel with a maximum maximum enrichment enrichment of 5.0 wt% U-235 while maintaining maintaining the maximum maximum effective neutron multiplication multiplication factor (kefr)

(keff) less than 0.95, including all biases and uncertainties for full flood and aqueous foam conditions, in accordance with 10 CFR 50.68(b 50.68(b)(2).

)(2).

45 45

ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION Analyses have demonstrated that the CE 14x14 14x14 fuel modeled in the current current analyses is more reactive than than enriched AREV similarly enriched AREVA A Advanced CE-14 HTP fuel in the dry new fuel storage racks, and therefore therefore the above conclusions remain valid.

AREVA Spent AREV A Advanced CE-14 HTP fuel assemblies will be stored in the SFP following discharge from the reactor. The pool, designed in two two halves, can accommodate accommodate 1830 assemblies and and one spent fuel shipping cask. The Unit 1 half of the SFP contains storage racks in six lOxlO, 10xI0, two 8xl0, 8x10, and one 7xl0 7x10 array. The Unit 2 half of the SFP contains racks in ten 10xlO 10x10 arrays. The racks are fabricated from steel and consist of vertical cells grouped in parallel rows with a center-to-center stainless steel center-to-center distance of 10-3/32" in both Units. Sandwiched between the inner and outer walls of each storage cell is a 6.5" wide sheet of B44C neutron absorber material. Unit 1 storage racks use a B44C composite material, carborundum, and Unit 2 racks use Boraflex. The Boraflex is no longer credited in criticality calculations.

The Unit 11 SFP racks are currently licensed to meet the requirements of 10 CFR 50.68(b)(4) for storage 14x14 of CE 14x 14 fuel enriched to 5.0 wt% U-235 assuming an infinite axial and radial array of storage cells of of nominal dimensions with credit for the carborundum neutron absorber neutron absorber sheets and no credit for assembly burnup. At the worst case temperature of 40°F, bumup. 40'F, the maximum unborated keff value with all biases and uncertainties is less than the 10 CFR 50.68 regulatory limit of 1.0. The maximum kerr keff at a moderator moderator concentration of350 boron concentration of 350 ppm with all biases and uncertainties uncertainties is less than the 10 CFR 50.68 regulatory limit of 0.95. Analyses have demonstrated that the CE 14x14 fuel modeled in the current analyses is is more reactive than similarly enriched AREVA enriched AREV A Advanced CE-14 HTP fuel in the Unit 1 SFP storage racks, and therefore the above conclusions remain valid.

The Unit 2 SFP racks are currently licensed to meet the above mentioned requirements of 10 CFR mentioned requirements CFR 50.68(b)(4) for storage ofCEof CE 14x14 fuel enriched enriched to 5.0 wt% U-235 assuming assuming an infinite radial array arr~y of of storage cells of nominal dimensions dimensions with credit for burnup bumup in lieu of the boraflex boraflex neutron absorber sheets and credit credit for 350 ppm soluble boron. Irradiated assemblies to be stored in the Unit 2 SFP must meet the burmup bumup requirements requirements of Technical Technical Specification 3.7.17. Fresh fuel may also be stored in the Unit 2 pool provided provided that the fuel assembly assembly is surrounded on all four adjacent faces by empty empty rack cells or other nonreactive nonreactive materials (e.g., wall, water, etc ... ).). Analyses have demonstrated that the CE 14x14 fuel materials (e.g.,

assumed assumed in the current analyses is more reactive reactive than similarly enriched AREVA AREV A Advanced Advanced CE-14 HTP fuel in the Unit 2 SFP storage racks, and therefore the above conclusions remain valid.

6.0 6.0 ACCIDENT ANALYSES ACCIDENT ANAI,YSES

6.1 INTRODUCTION

INTRODUCTION This This section section provides information information related to assessing assessing the Calvert Calvert Cliffs Cliffs Nuclear Nuclear Power Power Plant transient transient and accident analyses analyses for the transition to AREVA AREV A Advanced Advanced CE-14 CE-14 HTP fuel. It includesincludes a brief brief description of methodology description methodology used to evaluate evaluate the Calvert Calvert Cliffs UFSAR Chapter Chapter 14 14 events affected affected by the transition transition to AREVA Advanced AREV A Advanced CE-14 HTP fuel. Summary reports of sample analyses analyses for the non-LOCA LOCA and RLBLOCA RLBLOCA analysesanalyses methodologies methodologies are attached attached in accordance accordance with the NRC requirements ni'quirements for application application of the respective respective topical reports.

Recently, Recently, Calvert Cliffs received approval approval for a measurement measurement uncertainty uncertainty recapture power uprate (Reference (Reference 27), which increased increased the licensed licensed power level from 2700 MWt to 2737 MWt (1.38 (1.38 percent percent increase). The increase). The corecore thermal power measurement uncertainty measurement uncertainty is limited to 0.62 percent of actual reactor reactor thermal power thermal power whenwhen the Caldon CheckPlusTM Caldon CheckPlus' system is in operation. The operation. The same same analytical core power power is achieved achieved with the the measurement measurement uncertainty uncertainty recapture recapture power power level level of 2737 2737 MWt and and an an uncertainty uncertainty of of 0.62 0.62 percent.

percent. Therefore, Therefore, thethe existing analytical analytical power power level level of 2754 2754 MWt MWt remains remains valid valid and and is used for all analyses analyses at hot full power.

46 46

ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION 6.2 COMPUTER CODES Descriptions of the principal computer codes used in the LOCA and non-LOCA transient analyses are provided below.

S-RELAP5 The S-RELAP5 (References 13, 13, 19, and 20) code is an AREVA AREVA modification modification of the RELAPSIMOD2 RELAP5/MOD2 S-RELAP5 is used for simulation of the transient system response to LOCA as well as non-LOCA code. S-RELAPS events. Control volumes and junctions are defined which describe all major components in the primary and secondary systems that are important for the event being analyzed. The S-RELAP5 hydrodynamic hydrodynamic two-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture. S-model is a two-dimensional, RELAP5 uses a six-equation model for the hydraulic solutions. These equations include two-phase RELAPS two-phase continuity equations, two-phase momentum equations, and two-phase energy equations. The six-equation model also allows both non-homogeneous non-homogeneous and non-equilibrium situations encountered in reactor problems to be modeled.

RODEX2-2A RODEXl-2A RODEX2-2A (References RODEX2-2A (References 17 and 18) 18) was developed to perform calculations for a fuel rod under normal operating conditions. The code incorporates models to describe the thermal-hydraulic thermal-hydraulic condition of the fuel rod in a flow channel; densification and cracking in the pellet; the gap channel; the gas release, swelling, densification conductance; the radial thermal conduction; conductance; conduction; the free volume and gas pressure internal to the fuel rod; the fuel and cladding cladding deformations; and the cladding cladding corrosion. RODEX2-2A RODEX2-2A has been extensively extensively benchmarked; benchmarked; its predictive predictive capabilities were correlated correlated over a wide range of conditions conditions applicable applicable to light water reactor reactor fuel conditions. For non-LOCA non-LOCA applications, RODEX2-2A is used to validate validate the gap conductance used conductance used in the in the analyses and to establish the FCM LHR as a function of exposure.

exposure. For small break LOCA (SBLOCA)

(SBLOCA) applications, RODEX2-2A RODEX2-2A is used to establish establish the burnup-dependent burnup-dependent initial fuel (i.e., gap dimensions, gas composition, and gas inventory) for the S-RELAPS conditions (i.e., S-RELAP5 calculations.

calculations.

RODEX3A / RODEX3 The RODEX3A code (Reference (Reference 22)22) simulates the thermal and mechanicalmechanical response of a fuel rod in a coolant channel channel as a function of exposure for the normal and power ramp conditions encountered in conditions encountered pressurized pressurized (PWR) and boiling boiling water reactors. Phenomenological Phenomenological rate-dependent rate-dependent models are used to evaluate the temperature-, exposure-dependent changes in the state of the fuel and cladding temperature-, stress-, and exposure-dependent cladding materials and in the materials and in the release release of the inert gaseous fission products. A quasi-steady state computational procedure procedure is is used used to analyze the to analyze response ofa the response of a fuel fuel rod rod as as aa function function of time for ofti~e for RLBLOCA RLBLOCA applications.

applications.

The RODEX3A The RODEX3A code has been code has been benchmarked benchmarked to realistically model the mechanical mechanical response, the thermal response, and response, and the the fission gasgas release release during steady steady state state and normal power maneuvering maneuvering operations operations for light light water reactor fuel rods (Reference water reactor (Reference 21). The RODEX3A RODEX3A code uses pseudo-steady pseudo-steady state solution algorithms to solvesolve for the thermal and mechanical mechanical responses responses that neglect the effects effects of the thermal thermal heat heat capacity and capacity and the mechanical mechanical inertia. Hence, RODEX3A Hence, RODEX3A applicationsapplications are limited to analyses analyses where where the effects of the thermal and and mechanical mechanical inertia can be neglected.

neglected. For the analysis analysis of of rapid transients transients and/or and/or accidents, the thermal accidents, thermal capacitance capacitance of the fuel and cladding cladding must be modeled modeled toto correctly correctly predict predict the thermal thermal response response of of the fuel and the cladding. For these these transients, transients, the RODEX3A RODEX3A code code is used to establish initial initial fuel rod conditions conditions at the start start of the transient transient and/or accident accident and at the burnup burnup of of interest.

interest. Variables Variables defining defining the fuel state at the the start of the transient transient are transferred transferred to transient thermal-thermal-hydraulic hydraulic codes codes (such (such asas the the S-RELAP5 S-RELAPS code) code) for for the initiation of the transient/accident transient/accident analysis.

47 47

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION The RODEX3A code calculational calculational models are identical to the RODEX3 models. The RODEX3A RODEX3A codecode differs from the RODEX3 code in the following areas:

The RODEX3 RODEX3 input format was changed to reorganize reorganize the input into a more logical format and to use the S-RELAP5 input processing routines. The S-RELAP5 input processing processing routines permit permit dimensions for code variables to be expanded contracted without modifying the input description expanded or contracted description and permit permit input for more than one fuel rod.

1.

1. The RODEX3A input was changed changed to allow up to 52 fuel rods to be analyzed in a single calculation. This change was made to enable RODEX3A RODEX3A to write a binary binary data file for transferring fuel rod data at a given burnup bumup to the S-RELAP5 code.

2. The input/output routines were rewritten so the output fully describes all print variables.

3. A binary data file containing containing plot data was created.

4. The calculations calculations were expanded to permit dished annular pellets to be analyzed.

ICECON The ICECON code (Reference (Reference 20) is an AREVA modification modification of the CONTEMPT/LT-022 CONTEMPTILT-022 code to which which an ice condenser model was added. (Note (Note that Calvert Cliffs does not have an ice condenser condenser containment.) ICECON containment.) ICECON predicts the long-term behavior of PWR nuclear reactor containment containment systems systems subjected subjected to postulated LOCA conditions. It calculates the time variation of compartment compartment pressures, temperatures, temperatures, mass and energy energy inventories, inventories, heat structure temperature distributions, distributions, and energy exchange exchange with adjacent compartments. Models are provided to describe fan cooler cooler and cooling spray engineered safety systems. ICECON can be used to model from one to four compartments,compartments, and any compartment compartment except the reactor system may have both a liquid pool region and a vapor atmosphere atmosphere region above the pool. Each region is assumed to have a uniform temperature, temperature, but the temperatures temperatures of the two regions may be different. ICECON can be used to model PWR dry and ice condenser containments, sub-atmospheric sub-atmospheric containments, containments, and dual containments with an annular region. Specifically, ICECON ICECON is used to establish containment containment backpressure backpressure for RLBLOCA applications.

XCOBRA-IIIC XCOBRA-IIIC The XCOBRA-IIIC XCOBRA-IIIC code (Reference 11) 11) is a steady state thermal-hydraulics thermal-hydraulics code that calculates calculates the axial and radial flow and enthalpy enthalpy distribution within assemblies and sub-channels sub-channels for non-LOCA non-LOCA events.

When used in conjunction with core boundary conditions from the S-RELAP transient analysis and the HTP DNB correlation HTP (Reference 14), XCOBRA-IIIC correlation (Reference XCOBRA-IIIC also calculates corresponding minimum calculates the corresponding Minimum DNBR calculations DNBR. Minimum calculations are performed performed in a two-step process. Calculations Calculations are first performed core-wide basis to calculate the axially varying flow and enthalpy performed on a core-wide enthalpy distribution in the peak powered powered fuel assembly. Next, these flow and enthalpy boundary boundary conditions are applied to a sub-channel model of the peak powered assembly assembly to determine the local conditions for the calculation calculation of minimum DNBR. The XCOBRA-IIIC analyses are performed as part of the thermal-hydraulicsthermal-hydraulics portion of the AREVA Advanced CE-14 HTP fuel transition, which is discussed in Section 4.0.

6.3 TRANSIENT ANALYSIS TRANSIENT The Calvert Cliffs UFSAR Chapter 14 analyses are listed in Table 6-1.

Calvert CliffsUFSAR 6-1. Although Calvert Cliffs is a pre-Standard Standard Review Plan (SRP)(SRP) plant, a cross-reference cross-reference to the corresponding corresponding SRP section was provided provided to assist the NRC with their review. A review of each event was conducted relative to the transition to assist AREVA Advanced Advanced CE-14 HTP fuel. These events are listed below with a more detailed event-by-event event-by-event disposition of the challenge challenge to the SAFDLs included in following sections.

48

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION 1.

1. Events With System Transient Response Response - A computer code (such as S-RELAP5) S-RELAPS) is used to generate the transient response to be assessed against the acceptance acceptance criteria.

a. Numerous Numerous Calvert Cliffs UFSAR Chapter Chapter 14 events are affected by the transition to AREVAAREVA Advanced Advanced CE-14 HTP fuel, specifically specifically because of changes in thermal-hydraulic thermal-hydraulic performance and neutronics inputs to the safety analyses. The events requiring analysis using performance the AREV AREVA A non-LOCA non-LOCA safety analysis methodology (Reference 13) methodology (Reference 13) are as follows:

- Control Element Assembly Withdrawal Withdrawal Event (UFSAR Section 14.2) 14.2) o From Hot Full PowerPower o From a Subcritical Subcritical or Low Power Condition

- Excess Load Event (UFSAR (UFSAR Section 14.4)14.4)

- Excess Feedwater Feedwater Heat Removal Event (UFSAR (UFSAR Section 14.7) 14.7)

- Reactor Reactor Coolant Coolant System Depressurization (UFSAR System Depressurization (UFSAR Section 14.8)

Section 14.8)

- Loss-of-Coolant Loss-of-Coolant Flow Event (UFSAR Section 14.9)

- Control Control Element Assembly Drop Event (UFSAR Section 14.11) 14.11)

- Asymmetric Steam Generator Event (UFSAR Section 14.12) 14.12)

- Control Element Assembly Ejection (UFSAR (UFSAR Section 14.13) 14.13)

- Steam Line Break Event (UFSAR Section 14.14)

- Seized Rotor Event (UFSAR Section 14.16)

" Base Large Break Break LOCA Analysis, using Reference 20 methodologies (UFSAR 14.17.2)

Section 14.17.2)

Base Small Break Break LOCA Analysis, using Reference 19 methodologies methodologies (UFSAR (UFSAR 14.17.3)

Section 14.17.3)

b. The following UFSAR Chapter 14 events are not affected by the transition to AREV AREVA A Advanced Advanced CE-14 HTP fuel because the key parameters for these events are plant related related system responses [e.g.,

system [e.g., core power, decay heat, auxiliary auxiliary feedwater feedwater (AFW) capability, offsite power availability, safety injection and/or charging capability, etc.] rather than the fuel design parameters.

parameters. As such, these events will not be analyzed analyzed at this time.

" Loss of Load Event (UFSAR Section 14.5) 14.5)

- Loss of Feedwater Flow Event (UFSAR Section 14.6) 14.6)

- Loss-of-Non-Emergency AC Power (UFSAR Section 14.10)

Loss-of-Non-Emergency 14.10)

- Steam Generator Tube Rupture Event (UFSAR Section 14.15)

Section 14.15)

- Turbine-Generator Overspeed Turbine-Generator Overspeed Incident Incident (UFSAR Section 14.19)

- Containment Response (UFSAR Section 14.20)

Containment

- Excessive Charging Event (UFSAR Section 14.25) 14.25)

- Feedline Break Event (UFSAR Section 14.26)

Section 14.26) 49

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION

2. Events Events Without System Response - The remaining UFSAR Chapter Chapter 14 events are evaluated evaluated using approved methodologies.

methodologies.

a. The following event is analyzed analyzed in accordance accordance with AREVA AREV A safety analysis methodology methodology 13):

(Reference 13):

- Boron Dilution Event (UFSAR (UFSAR Section 14.3)

Section 14.3)

b. Events with radiological consequences consequences only are evaluated to ensure that the AOR remains remains bounding.

- Fuel Handling Incident (UFSAR Section 14.18)

Section 14.18)

- Waste Gas Incident (UFSAR Section 14.22) 14.22)

- Waste Processing System Waste System Incident (UFSAR Section 14.23) 14.23)

- Hypothetical Accident (UFSAR Section Maximum Hypothetical 14.24)

Section 14.24)

To comply with the NRC Safety Evaluation requirements requirements of AREV AREVA A safety analysis methodology methodology (Reference 13), a sample application (Reference application of the non-LOCA non-LOCA safety analysis is enclosed. The sample analysis sample analysis presented is the Loss-of-Coolant Loss-of-Coolant Flow event (UFSAR (UFSAR Section 14.9) since this event challenges challenges minimum DNBR correlation correlation limit. The analysis provides the required elements to demonstrate demonstrate applicability applicability of the method to Calvert Cliffs and addresses the SER restrictions as discussed in Section Section 6.3.1.

6.3.1. The remaining remaining analyses will be available event analyses available for NRC audit when the analyses have been completed.

6-1, Summary of Event Disposition Table 6-1, Disposition UFSAR SRP Event Description Disposition Section Section Section Event Description Disposition 14.2 Control Control Element Assembly Withdrawal Withdrawal Analysis Required Required 15.4.1 Subcritical or Low Power Startup Condition 15.4.2 At Power Power 14.3 15.4.6 Boron Dilution Analysis Required Required 14.4 15.1.3 Excess Excess Load Analysis Required Required 14.5 Loss of Load AOR Remains Bounding Bounding 15.2.1 Loss of Electric Load 15.2.2 Turbine Turbine Trip Trip 15.2.3 Loss of Condenser Condenser Vacuum 15.2.5 . Stearm Pressure Regulator Failure Steam 14.6 15.2.7 Loss of Feedwater Feedwater Flow AOR Remains Bounding 14.7 Excess Feedwater Heat Removal Excess Feedwater Analysis Required Required 15.1.1 Decrease Decrease in Feedwater Temperature Temperature 15.1.2 Increase Increase in Feedwater Feedwater Flow 14.8 15.6.1 Reactor Reactor Coolant System Depressurization Depressurization Analysis Required Required 14.9 15.3.1 Loss-of-Coolant Loss-of-Coolant Flow Analysis Required' Required 1 14.10 14.10 15.2.6 Loss-of-Non-Emergency Loss-of-Non-Emergency AC Power AOR Remains Bounding 14.11 15.4.3 Control Element Assembly Drop Analysis Required Ana!ysis R~guired 14.12 Asymmetric Asymmetric Steam Generator (SG) (SG) Events Excess Excess Feedwater to One SG Bounded by Other Events Loss of Feedwater Feedwater to One SG Bounded by Other Bounded Other Events Events 15.1.4 15.1.4 Excess Load to One SG Bounded by Other Events 15.2.4 Loss of External Load to One SG Analysis Required Required 50

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION Summary of Event Disposition (Continued) 6-1, Summary Table 6-1, UFSAR UFSAR SRP Event Description Disposition Section Section Section Event Description Disposition 14.13 15.4.8 Element Assembly Ejection Control Element Analysis Analysis Required 14.14 15.1.5 Steam Line Break Steam Analysis Required 14.15 15.6.3 Steam Generator Tube Rupture Steam Generator Rupture AOR Remains Bounding 14.16 15.3.3 Reactor Coolant Pump (RCP) Seized Rotor Analysis Required 14.17 15.6.5 Loss-of-Coolant Accident Loss-of-Coolant Accident Small Break Analysis Required 2 Large Break Analysis Required 2 Analysis Required 14.18 14.18 15.7.4 Fuel Handling Incident Incident AOR Remains Remains Bounding 14.19 ---

--- Turbine-Generator Overspeed Turbine-Generator AOR Remains AOR Remains Bounding 14.20 6.3 Containment Containment Response AOR Remains Remains Bounding 14.22 --- Waste Gas Incident Waste Incident AOR Remains AOR Remains Bounding 14.23 15.7.3 Waste Waste Process System Incident AOR Remains Bounding 14.24 15.6.5 Maximum Hypothetical AccidentAccident AOR Remains Remains Bounding 14.25 15.5.2 Excess Charging AOR Remains Bounding 14.26 15.2.8 Feedline Feedline Break AOR Remains Bounding Notes:

1 The Loss-of-Coolant Flow The Loss-of-Coolant event is the AREVA non-LOCA safety analysis methodology sample application Flow event is the AREV A non-LOCA safety analysis methodology sample application analysis (refer to Enclosures Enclosures 2 and 5).

5).

2 2

RLBLOCA event The RLBLOCA The is the event is the AREV AREVA large break A large LOCA analysis break LOCA methodology sample application (refer to analysis methodology sample application (refer to Enclosures 1 and 4).

6.3.1 6.3.1 Analysis Analysis Methodology AREV A methodology for evaluating non-LOCA transients is described The AREVA described in (Reference 13). The non-(Reference 13).

methodologies to be applied for the Calvert Cliffs fuel transition LOCA analysis methodologies transition have been previously approved by the NRC. This report includes the required elements reviewed and approved reviewed elements for approval of the application of application these these methodologies to Calvert Calvert Cliffs.

each non-LOCA transient event analysis, the nodalization, For each parameters, conservative input and nodalization, chosen parameters, sensitivity studies sensitivity studies are reviewed for applicability compliance with the SER for non-applicability to the fuel transition in compliance 13).

(Reference 13).

LOCA topical report (Reference

- The nodalization nodalization used for the calculations calculations supporting the fuel transition transition is specific to Calvert accordance to the (Reference Cliffs Units 1I & 2 and is in accordance (Reference 13)13) methodology. Nodalization diagrams used for the fuel transition analyses diagrams analyses are included with the sample non-LOCA non-LOCA application in Enclosures application Enclosures 2 and 5. 5.

- The parameters parameters and equipment states are chosen to provide provide a conservative conservative estimate of the acceptance criteria. The biasing and assumptions for key input parameters challenge to the acceptance challenge parameters are consistent with the approved Reference consistent Reference 13 methodology.

- The S-RELAP5 S-RELAP5 code assessments in Reference Reference 13 validated ability of the code to predict the validated the ability primary and secondary systems to UFSAR Chapter 14 non-LOCA response of the primary non-LOCA transient transient and accidents. No additional accidents. model sensitivity sensitivity studies are needed for this application.

application.

Reference 2 incorporates Reference incorporates M5M5 properties into the S-RELAP5S-RELAP5 based non-LOCA methodology. No restrictions or requirements were identified in the SER for the Reference restrictions Reference 2 methodology relative to its methodology relative application S-RELAP5 non-LOCA application to S-RELAP5 non-LOCA analyses.

51 51

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION The methodology methodology for performing DNB calculations XCOBRA-IIIC code is described in calculations using the XCOBRA-IIIC Reference Reference 12 topical Reference 12. The SER for the Reference topical report states that the use of XCOBRA-IIIC is limited ofXCOBRA-IIIC "snapshot" mode. Thus, minimum to the "snapshot" calculations are performed minimum DNBR calculations performed using a steady state XCOBRA-IIIC model with core boundary conditions at the time of minimum XCOBRA-IIIC minimum DNBR from the S-RELAP5 transient analyses.

Reference 16 topical report describes the method for performing statistical The Reference statistical DNB analyses. No restrictions identified in the SER for this methodology.

restrictions or requirements were identified methodology.

calculating the enthalpy deposition for a CEA ejection accident is given in approved methodology for calculating The approved in requirements were identified in the SER for this methodology.

Reference 7. No restrictions or requirements Reference 6.3.2 Control Element Assembly Withdrawal Event (UFSAR Section Control 14.2)

Section 14.2)

The uncontrolled withdrawal of a CEA bank could be caused by a malfunction in the reactor control uncontrolled withdrawal control or rod control systems or by operator error. The malfunction could lead to a large and rapid positive reactivity addition, resulting in a power reactivity power transient challenges the DNBR and FCM SAFDLs.

transient that challenges HZP CEA Withdrawal Withdrawal neutron flux which results from the bank withdrawal is countered The rapid increase of the neutron countered by the reactivity feedback effect of the negative Doppler coefficient. This inherent self-limitation of the power inherent self-limitation power excursion excursion is of primary importance, importance, because because it limits the power to a tolerable level during the delay time for protective action. Although nuclear power peaks at a very high level during the rapid Although the nuclear excursion, the duration is short enough to preclude significant energy deposition. The fuel rod surface heat flux lags behind the nuclear power level but still peaks behind peaks at a significant fraction of the rated-power value. The increase in the primary coolant temperatures, temperatures, in tum,turn, lags behind the increase in the fuel rod heat flux.

Protective System (RPS) is designed to terminate the transient The Reactor Protective transient before the DNBR limit is reached. The principal protective trip for this event is the variable reached. high power trip (VHPT).

HFP CEA Withdrawal corresponding HZP CEA response, withdrawal is slower than a corresponding The transient response for the HFP bank withdrawal increase is actively coupled to CEA movement. The increase since the power increase since increase of the neutron flux flux withdrawal is following by a rise in thermal power, with the thermal power resulting from the HFP bank withdrawal reactivity insertion rate of the CEA withdrawal.*

lag determined by the reactivity withdrawal. The positive reactivity addition results in a power transient, increasing the primary temperatures and core heat flux, and primary coolant temperatures margin to the DNB and FCM SAFDLs.

decreasing the margin decreasing transient before the DNBR limit is reached. The principal designed to terminate the transient The RPS is designed VHPT and the Thermal Margin/Low Pressure (TMlLP) protective trips for this event are the VHPT protective (TM/LP) trip.

The key parameters for this event are:

- Initial operating conditions

- Maximum differential worth for CEAs moving in sequence Maximum sequence

- Maximum Maximum CEA withdrawal withdrawal rate

- Doppler Doppler reactivity reactivity feedback

- Moderator Moderator reactivity feedback (HFP only) 52 52

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION

- Trip setpoint(s),

Trip setpoint(s), uncertainty and delay time

- Number of RCPs running (sub (subcritical critical and low power cases)

- Fuel rod gap conductance (subcritical (subcritical and low power cases)

- Maximum Fq Fq predicted for the purpose of calculating the peak (hot spot) fuel centerline temperature (subcritical and low power cases)

This event is classified as an AOO which may occur during the lifetime of the plant. This event does not This provide a significant challenge to peak pressure. Therefore, the principally challenged acceptance criteria for this event are:

1. Fuel cladding integrity shall be maintained by ensuring that SAFDLs are not exceeded (i.e.,

1. (i.e., the minimum calculated DNBR shall remain above the 95/95 DNB correlation limit).

2. FCM shall not occur.

Some of the key parameters parameters listed for this event, such as maximum differential CEA worth and Doppler reactivity feedback, are potentially impacted by the transition to AREV AREVA A Advanced CE-14 CE-I4 HTP fuel. As such, the acceptance acceptance criteria specified for this event must be evaluated to support the fuel transition.

Consequently, the event will be reanalyzed in accordanceaccordance with the NRC-approved NRC-approved methodology described in Section 6.3.1, 6.3.1, using the AREV AREVA A non-LOCA methodology (Reference 13). Departure from nucleate boiling ratio analyses will employ appropriate NRC-approved NRC-approved CHF correlations in accordance with AREVA methodology.

AREV A methodology. The The results of the CEA Withdrawal Withdrawal event analysis will be available for NRC audit when the analysis has been completed.

6.3.3 Boron Dilution Event (UFSAR Section 14.3)

A Boron Dilution event is defined as any event caused by a malfunction or an inadvertent operation operation of the Chemical and Volume Control System that results in a dilution of the active portion of the RCS. The active portion active portion ofof the RCS is defined as that volume of water that circulates the RCS circulates through the core. For example, when in shutdown cooling, no credit is allowed allowed for the volume of water in the SG and other other stagnant stagnant portions of the RCS. A dilution of the RCS can be the result of adding adding borated borated water, which has a boron concentration concentration that is less than the system boron concentration, concentration, or by the removal of boron using a purification purification ion exchanger exchanger with a deborating resin.

The analysis of The analysis of the Boron Dilution event covers the Boron covers the six modes of operation listed in the UFSAR UFSAR

" Section Section 14.3 and defined defined in Technical Specifications. A Boron Dilution Technical Specifications. Dilution event can approach approach the DNBR DNBR and FCM SAFDLs and the RCS pressure pressure limit. In all cases, operator operator action is required to prevent prevent exceeding exceeding these limitslimits by securing securing the dilution dilution and borating, if necessary, necessary, to maintain the required required shutdown shutdown boron concentration.

concentration.

Under the worst conditions, the operator has 30 minutes in the refueling mode and 15 15 minutes in the other other modes modes of of operation from the operation from the time time of initiation of the event to secure of initiation secure the dilution to prevent prevent losing the minimum shutdown margin. The DNBR and FCM SAFDLs and the RCS pressure limit criteria criteria will be met if the entire shutdown margin entire shutdown margin is not lost.

The key parameters parameters for this event are:

are:

- Initial Initial operating operating conditions conditions

- Boron worth Boron worth and dilution dilution rate rate 53 53

ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION

- Mixing volume Mixing volume

- Shutdown cooling flowrate (Modes 4, 5, Shutdown 5, 6) 6)

This event This event is is classified as an AOO which may occur during the lifetime of the plant. As long as the reactor remains reactor remains sub-critical sub-critical then overpressure and event progression are not limiting. Therefore, the principally challenged acceptance criterion for this event is: is:

1. Fuel cladding integrity should be maintained by ensuring SAFDLs are not exceeded. This is 1.

demonstrated by assuring that the minimum calculated DNBR is not less than the 95/95 DNB correlation limit.

thermal-hydraulic Nuclear Steam Supply This event is not analyzed with a thermal-hydraulic Supply System (NSSS) transient code such as S-RELAP5.

S-RELAP5. This event is evaluated each cycle as part of the reload licensing process using AREVA AREVA non-LOCAnon-LOCA safety analysis methodologymethodology (Reference 13, Section 5.6). The results of the Boron Dilution event analysis are available for NRC audit.

6.3.4 Excess Load Event (UFSAR (UFSAR Section 14.4)

An Excess Load An Excess event is defined as any rapid, uncontrolled Load event uncontrolled increase in SG steam flow other than a Steam Line Break. The Line Break. The full full opening of the turbine control valves, atmospheric atmospheric dump valves, or turbine bypass valves during steady state operation would result in an Excess Load event.

The increase in steam flow creates a mismatch between the energy being generated in the reactor core and the the energy energy beingbeing removed by the secondary secondary system and results in a cooldown of the primary system. A power increase will occur if the moderator moderator temperature temperature reactivity reactivity feedback coefficient is negative. If the power increase is sufficiently sufficiently large either either the overpower overpower limit or the thermal thermal margin margin limit will be reached reached and and the event will the event will be be terminated terminated byby aa reactor reactor trip. If the power increase increase is less significant, the reactor will stabilize at an elevated power level without reaching a reactor trip.

The event is The event is primarily primarily protected protected by the upper setpoint setpoint on the VHPT, which terminates the moderator moderator feedback feedback driven power excursion. excursion. As the cold water front enters the core, over-moderation over-moderation will result in the core core power distribution distribution shifting towards the bottom of the core.

Depending Depending upon upon thethe ASI ASI at which the plant was operating operating at transient initiation, as well as this transient transient ASI ASI shift, shift, the the TM/LP TMiLP and and linear linear power power density density limiting limiting safety system system setting trips may also intercede.

intercede.

Depending Depending upon the response response of the SGsSGs and the feedwater feedwater system, the plant plant may also potentially potentially trip on on a low SG SG level or low SG pressure as well. The HFP excess load transient represents represents a trip design-basis design-basis event for the TM/LP TMiLP limiting safety system setting and the linear power power density limiting safety safety system setting setting verification verification analyses.

The key parameters parameters for this event are:

"* Initial Initial operating conditions operating conditions

- Magnitude Magnitude of of the the step step increase increase in load load (i.e.,

(i.e., the event initiator) initiator)

- Moderator Moderator reactivity reactivity feedback feedback

- Doppler reactivity feedback Doppler reactivity feedback (HZP (HZP case) case)

- Fuel rod gap conductance (HZP Fuel rod gap conductance (HZP case) case)

- Trip Trip setpoint(s),

setpoint(s), uncertainty uncertainty and delay time

- Core Core power power (NI & AT) d T) signal signal decalibration decalibration 54

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION

- Maximum Fq predicted predicted for the purpose of calculating calculating the peak (hot spot) fuel centerline temperature temperature (HZP case)

- Technical Specifications minimum shutdown Technical SpeCifications shutdown margin (post-scram return-to-power case)

(post-scram return-to-power case) classified as an AOO which may occur during the life of the plant. This event does not This event is classified provide significant challenge provide a significant Therefore, the principally challenge to peak pressure. Therefore, principally challenged acceptance criteria challenged acceptance criteria for this event are:

maintained by ensuring SAFDLs are not exceeded

1. Fuel cladding integrity shall be maintained

1. (i.e., the exceeded (i.e.,

minimum calculated DNBR shall remain above the 95/95 DNB correlation correlation limit).

2. FCM shall not occur.

parameters listed for this event, such as Doppler reactivity feedback and fuel rod gap Some of the key parameters impacted by the transition to AREV conductance, are potentially impacted conductance, AREVA A Advanced CE-14 HTP fuel. As such, acceptance* criteria the acceptance' criteria specified event must be evaluated to support the fuel transition.

specified for this event accordance with the NRC-approved Consequently, the event will be reanalyzed in accordance Consequently, NRC-approved methodology described methodology described in Section 6.3.1, using the AREVA Section 6.3.1, non-LOCA methodology AREVA non-LOCA methodology (Reference Departure from nucleate (Reference 13). Departure appropriate NRC-approved boiling ratio analyses will employ appropriate boiling NRC-approved CHF C-F correlations correlations in accordance with AREV A methodology. The Excess Load event AREVA event analysis will be available for NRC audit when the analysis has been completed.

6.3.5 Loss of Load Event (UFSAR Section Section 14.5)

The Loss of Load event decrease in heat removal by the secondary characterized by a decrease event is characterized secondary system caused by either a direct turbine trip or a fast turbine runback following loss of external electrical electrical load. A major difference between the two events difference Termination of events is the rate at which turbine steam flow is reduced. Termination of main turbine steam flow following a design basis Loss of Load event occurs due to rapid closure of the Termination of steam flow for a turbine trip event occurs due to turbine stop valve turbine throttle valves. Termination closure. The stop valves are designeddesigned for turbine overspeed protection, and they close faster than the throttle valves. Following electrical load, a runback is initiated and the turbine throttle valves Following a loss of electrical close at a moderately fast rate, but not instantaneously. In the event of a turbine trip, the turbine stop valves close close almost instantly (typically within 0.1 second). A transient instantly (typically constructed that transient scenario is constructed bounds the results of both the Loss of Load and turbine trip events. The event is typically initiated by by near-instantaneous turbine stop valve closure coincident near-instantaneous coincident with loss of non-safety-related non-safety-related steam dump capability.

A Loss of Load event can result from an electrical electrical disturbance which causes the generator to separate electrical grid. The Loss of Load scenario necessarily assumes that offsite power from the external electrical power connected to the station, thereby allowing the reactor to stay on-line until a RPS trip setpoint remains connected setpoint is challenged.

The key parameters for this event are:

- Initial core power power

- Initial operating operating conditions

- uncertainty and delay time Trip setpoint(s), uncertainty

- Primary safety safety relief valve setpoint setpoint and capacity capacity (for the RCS overpressurization overpressurization case) case)

"* Main steam safety valve (MSSV)

(MSSV) setpoints and capacities capacities

- Moderator Moderator temperature coefficient temperature coefficient 55 55

ATTACHMENT (4)

ATTACHMENT RELOAD RELOAD TRANSITION TRANSITION REPORT classified as an AOO which may occur during the life of the plant.

This event is classified The principally challenged acceptance challenged criteria for this event are:

acceptance criteria

1. The pressures in the reactor coolant and main steam systems should

1. should be less than 110%

110% of design values.

2. Fuel cladding cladding integrity shall be maintained maintained by ensuring SAFDLs are not exceeded exceeded (i.e.,

(i.e., the minimum calculated calculated DNBR shall remain above the 95/95 DNB correlation limit).

3. An incident of moderate frequency should not generate generate a more serious plant condition condition without independently.

other faults occurring independentlY.

Per UFSAR SectionSection 14.5, 14.5, the AOR for this event assumes assumes an initial core power level of 2754 MWt, and the most positive MTC allowed at HFP by Technical Specifications Specifications was used. These values are not impacted impacted by the transition to AREV AREVA A Advanced CE-14 HTP fuel and remain bounding. The event behavior is predominantly predominantly a function of the primary-to-secondary primary-to-secondary heat transfer capability. Therefore, perturbations in parameters small perturbations parameters such as the core pressure drop, core bypass flow fraction, core inlet flow distribution, and reactivity feedback do not impact the parameters parameters of interest interest in assessing the acceptance acceptance criteria. The plant system characteristics characteristics that potentially potentially impact the key parameters parameters listed for this event remain unchanged for both the transition fuel cycle, and the full core implementation of of AREVA AREV A Advanced Advanced CE-14 HTP fuel at Calvert Cliffs. The cause of the event and the parameters parameters that consequences of the event are unchanged control the consequences unchanged from or bounded by the current AOR AOR presented in in UFSAR Section 14.5. 14.5. Therefore, an analysis of the Loss of Load event is not required to support the transition to AREVA AREV A Advanced CE- 14 HTP fuel.

Advanced CE-14 6.3.6 Loss of Feedwater Feedwater Flow Event (UFSAR Section 14.6)

The Loss of Feedwater Feedwater Flow event is initiated by the termination termination of main feedwater feedwater (MFW) flow which results from failures in the MFW or condensate systems. The sudden loss of subcooled subcooled MFW flow, while the plant continues continues to operate at power, results in a reduction reduction in the SG inventory, a reduction in the primary-to-secondary primary-to-secondary heat transfer capability and an increase increase in the reactor coolant temperatures.

temperatures. The reactor coolant coolant expands, resulting in an insurge into the pressurizer. The resulting increase increase in pressure actuates the pressurizer spray system and may cause the pressurizer power-operated actuates power-operated relief valves (PORVs) to open. If the PORVs are unavailable, the PSVs will lift to mitigate the pressure transient.

(PORVs)

The key parameters parameters for this event are:

- Initial Initial core power core power

- Initial operator operator conditions conditions

- Decay heat assumptions Decay heat assumptions

- Trip setpoint(s), uncertainty and delay time

- Initial SG liquid inventory inventory (affects SG liquid inventory at the low water level reactor reactor trip setpoint) setpoint)

- Minimum AFW performance Minimum performance and actuation delay time

- RCS pump RCS pump heat heat

- MSSV setpoints and capacity

" PSV setpoint and capacity

- SG blowdown flow rate (for plants that do not have early automatic isolation of SG blowdown flow on the Engineered Safety Feature Feature Actuation Actuation System signal) signal) 56

ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION

Timing of any operator actions This event is classified as an AOO which may occur during the life life of the plant. The principally challenged acceptance criteria for this event are:

1. The pressures in the reactor coolant and main steam systems should be less than 110%

1. 110% of design values.

2. Fuel cladding integrity shall be maintained by ensuring SAFDLs are not exceeded exceeded (i.e., the calculated DNBR shall remain above the 95/95 DNB correlation limit).

minimum calculated

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

Per UFSAR Section 14.6, 14.6, the AOR for this event assumes an initial core power level of 2754 MWt, and the most positive MTC allowed at HFP by Technical Specifications Specifications was used. These values are not impacted impacted by the transition to AREVA AREV A Advanced CE-14 HTP fuel and remain bounding. The event behavior is predominantly predominantly a function of theprimary-to-secondary the primary-to-secondary heat transfer capability. Therefore, small perturbations perturbations in parameters parameters such as the core pressure drop, core bypass flow fraction, core inlet flow distribution, and reactivity feedback do not impact parameters of interest in assessing the impact the parameters acceptance criteria. The plant system characteristics acceptance characteristics that potentially potentially impact the key parameters parameters listed for this event remain unchanged unchanged for both the transition fuel cycle, and the full core implementation of of AREVA AREV A Advanced CE-14 HTP fuel at Calvert Cliffs. The cause of the event and the parameters parameters which control the consequences consequences of the event are unchanged unchanged from or bounded by the current AOR presented in in UFSAR Section Section 14.6.

14.6. Therefore, an analysis of the Loss of Feedwater Flow event is not required required to support the transition to AREV AREVA A Advanced Advanced CE-14 HTP fuel.

6.3.7 Excess Feedwater Feedwater Heat RemovalRemoval Event (UFSAR Section Section 14.7) 14.7)

The Excess Feedwater Feedwater Heat Removal event is defined as an increase increase in heat removal from the primary side to the SG secondary secondary side due to a reduction in SG feedwater feedwater temperature temperature without a corresponding corresponding reduction in steam flow from the SGs, or an increase in feedwater flow. This could be caused caused by the loss of one or more of the feedwater heaters, or due to a feedwater controller malfunction at steady state power.

The system system response response to this event event is that the RCS temperature temperature and pressure will decrease. When there is a negative negative MTC, a positive reactivity feedback positive reactivity feedback occurs in the core core in response response to the decreasing decreasing core average temperature. This increases core core power power and the core core average average heat flux. Elevated cladding cladding heat fluxes and fuel temperatures temperatures in the hot assembly may approachapproach the DNB and LHR SAFDLs.

If core protection protection is needed, needed, the event is usually terminated terminated by a TM/LP TM/LP trip or a VHPT. Otherwise Otherwise a new equilibrium equilibrium state may be reached without tripping the plant.

The key parameters parameters for this event are:

- Initial Initial operating operating conditions condi~ions

- Magnitude Magnitude of the step changechange in feedwater feedwater conditions (i.e., the event conditions (i.e., event initiator) initiator)

- Moderator Moderator reactivity reactivity feedback

- Doppler Doppler reactivity reactivity feedback feedback

- Trip Trip setpoint(s),

setpoint(s), uncertainty uncertainty and and delay delay time

- Core Core power power (NI & AT) ~ T) signal signal decalibration decalibration 57

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION REPORT RELOAD TRANSITION This event is classified classified as an AOO which may occur during the life of the plant. This event does not not provide a significant challenge significant challenge to peak pressure.

pressure. Therefore, the principally challenged acceptance criteria challenged acceptance criteria for this event are:

1. Fuel cladding integrity

1. integrity shall be maintained maintained by ensuring SAFDLs are not exceeded exceeded (i.e.,

(i.e., the calculated DNBR shall remain above the 95/95 DNB correlation minimum calculated correlation limit).

2. FCM should not occur.

The event has been reanalyzed in accordance with the NRC-approved NRC-approved methodology describeddescribed in 6.3.1, using the AREVA non-LOCA methodology Section 6.3.1, methodology (Reference (Reference 13). Departure Departure from nucleate boiling ratio analyses will employ employ appropriate NRC-approved NRC-approved CHF correlations accordance with correlations in accordance AREVA AREV A methodology. The results of the completed Excess Feedwater Feedwater Heat Removal event analysis are available for NRC audit.

6.3.8 Reactor Depressurization (UFSAR Section 14.8)

Reactor Coolant System Depressurization 14.8)

Depressurization event The RCS Depressurization characterized as a rapid, uncontrolled decrease event is characterized decrease in RCS pressure other other than a LOCA. Inadvertent opening the PSVs, PORVs, or a malfunction malfunction in the pressurizer pressurizer spray spray system during steady steady state operation operation would result in an RCS Depressurization Depressurization event. This event results in a rapid rapid RCS depressurization depressurization that could reach the hot leg saturation pressure if a reactor trip did not occur, which could results in a challenge challenge to the DNB SAFDL. The RCS depressurization is typically accompanied accompanied by by an increase in the pressurizer pressurizer level due to the fluid expansion expansion caused by the decrease in system pressure.

The pressurizer level transient is typically terminated terminated by the reactor trip and does not challenge the volume of the pressurizer. The core power increases (when the MTC is positive) in response to positive moderator density feedback caused by the depressurization.

depressurization. The RPS will automatically function to scram the reactor, terminating the challenge to the DNB SAFDL. Experience Experience has shown that this event presents a mild challenge challenge to the DNB SAFDL. ReactorReactor scram is expected expected to occur on a TMiLP TM/LP trip.

The key parameters parameters for this event are:

- Initial operating conditions

- Capacity of the stuck open relief valve or safety valve

- Trip setpoint(s),

setpoint(s), uncertainty uncertainty and delay time This event is classified classified as an AOO which may occur during the life of the plant. This event does not not provide a significant significant challenge to peak pressure pressure and the FCM SAFDL SAFDL is not challenged challenged because because there is no significant increase increase in power for this event. Therefore, the principally challenged acceptance principally challenged acceptance criterion for this event is:

1. Fuel cladding I. cladding integrity shall be maintained maintained by ensuring SAFDLs are not exceeded exceeded (i.e.,

(i.e., the calculated DNBR shall remain above the 95/95 DNB correlation minimum calculated correlation limit).

Although Although the key parameters parameters listed for this event are not impacted impacted by the transition transition to AREV AREVA A Advanced Advanced CE-14 HTP fuel, the change change in subchannel subchannel model and elevation of the CHF correlation,correlation, as well as differences differences in the fuel rod and assembly design may introduce a perturbation perturbation in the minimum DNBR DNBR associated with this event. As such, the acceptance criteriacriteria specified for this event must be evaluated evaluated to support the fuel transition. Consequently, the event will be reanalyzed reanalyzed in accordance accordance with the NRC-approved approved methodology methodology described in Section 6.3.1, using the AREVA Section 6.3.1, AREVA non-LOCA methodology non-LOCA methodology (Reference (Reference 13). Departure Departure from nucleate nucleate boiling ratio analyses will employ appropriate appropriate NRC-approved NRC-approved 58 58

ATTACHMENT (4)

ATTACHMENT RELOAD RELOAD TRANSITION TRANSITION REPORT REPORT CHF correlations correlations in accordance accordance with AREVA methodology. The results of the RCS Depressurization Depressurization event analysis will be available for NRC audit when the analysis has been completed. completed.

6.3.9 Loss-of-Coolant Flow Event (UFSAR Section 14.9)

Loss-of-Coolant 14.9)

The Loss-of-Coolant Loss-of-Coolant Flow event is characterized characterized by a decrease in forced RCS flow. There may be either a partial or a total loss of RCS flow. A partial loss-of-coolant flow may be caused by a mechanical partialloss-of-coolant mechanical or or electrical failure in a pump, pump motor, a fault in the power supply to the pump motor, or a pump motor trip caused by such anomalies over-current or phase imbalance.

anomalies as over-current imbalance. A'complete A' complete loss of forced coolant flow may result from the simultaneous simultaneous loss of electrical electrical power to all pump motors. The partial partial Loss-of-Coolant Coolant Flow event is a less severe transient than the complete complete Loss-of-Coolant Loss-of-Coolant Flow event due to the smaller flow reduction.

A decrease decrease in reactor coolant flow occurring occurring while a plant is at power results in a degradation of corecore heat transfer, reduction reduction in DNBR margin, and a challenge to the DNB SAFDL. The reduction in primary system flow and associated increase increase in core coolant temperatures temperatures result in a reduction in DNBR margin.

The increasing primary system coolant coolant temperatures also results in expansion of the primary coolant coolant

volume, volume, causing an insurge into the pressurizer pressurizer and an increase in the pressure of the primary primary system.

However, However, the overpressure overpressure transient transient is bounded by the Loss of Load/turbine Load/turbine trip event (UFSAR (UFSAR Section 14.5) 14.5) due to the more rapid loss of primary-to-secondary primary-to-secondary heat transfer. This event is analyzed analyzed to verify verifY the RPS low flow trip in combination with the DNB Limiting Condition for Operation provides protection for the reactor core during these flow decreases.

The minimum DNBR is controlled by the interaction interaction of the primary primary coolant flow decay, the trip signal, the trip signal generation generation delay time, the scram delay time, the core power power decrease following reactor trip, and the rod surface surface heat flux.

flux. The power-to-flow power-to-flow ratio initially increases, increases, peaks, peaks, and then declines declines as the challenge challenge to the DNB SAFDL is mitigated by the decline in core power power due to the reactor trip.

The key parameters parameters for this event are:

- conditions Initial operating conditions

- RCP coastdown coastdown rate (pump (pump inertia and pump frictional torque)

- Trip setpoint(s), uncertainty uncertainty and delay time

- Minimum Minimum HFP scram worth

- Fraction Fraction of scram reactivity reactivity versus fraction of control rod insertion distance at HFP and delay time

- Fuel rod gap conductance conductance This event is classified as an AOO which may occur during during the life of the plant. This event event does not provide provide a significant challenge to peak pressure significant challenge pressure and the FCM SAFDL is not challenged because because there is is no significant increase in power for this event. Therefore, the principally challenged acceptance criterion challenged acceptance for this event is:

1. Fuel cladding

1. cladding integrity shall be maintained maintained by ensuring SAFDLs are not exceeded exceeded (i.e.,

(i.e., the minimum calculated calculated DNBR shall remain above above the 95/95 DNB correlation correlation limit).

Some of the key parameters parameters listed for this event, such as minimum HFP scram worth and fuel rod gap conductance, conductance, are potentially potentially impacted by the transition to AREVAREVA A Advanced Advanced CE-14 HTP fuel. As such, the acceptance acceptance criteria specified for this event must be evaluated to support the fuel transition.

Consequently, Consequently, the event has been reanalyzed reanalyzed in accordance accordance with the NRC-approved NRC-approved methodology 59

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION RELOAD TRANSITION REPORT REPORT described in Section 6.3.1, using the AREVA non-LOCA methodology (Reference (Reference 13).

13). Departure from employed appropriate nucleate boiling ratio analyses employed appropriate NRC-approved CHF correlations correlations in in accordance with AREVA methodology.

AREVA methodology. The results of the Loss-of-Coolant the Loss-of-Coolant Flow event event analysis are enclosed (Enclosures 2 and 5). 5).

Loss-of-Non-Emergency AC Power (UFSAR Section 14.10) 6.3.10 Loss-of-Non-Emergency The primary function of the AC power on the plant's ring bus is to provide power to the NSSS and the balance of plant electrical loads. Plant AC power goes to emergency emergency and non-emergency AC power Emergency power loads are classified loads. Emergency classified as those loads that are essential to safely shutdown shutdown the plant and maintain the plant in a safe shutdown condition. The response of the Res RCS to a Loss-of-Non-Emergency AC Power event is identical to a Loss-of-Coolant Flow event during the first five seconds (see Section 6.3.9). During this time interval, the secondary system has not had enough time to affect the RCS due to the loop cycle time. Consequently, the action of the low RCS flow RPS trip ensures the fuel SAFDLs will not be exceededexceeded during an Loss-of-Non-Emergency Loss-of-Non-Emergency AC Power event. As such, the analysis Loss-of-Non-Emergency AC Power event presented herein will address the approach of Loss-of-Non-Emergency approach to the RCS pressure upset limit and the approach to the site boundary boundary dose criteria in 10 CFR Part 100 guidelines guidelines precipitated by the longer term secondary secondary system response.

In addition to the loss of RCS flow initiated by the loss of power to the RCP motors, the loss of AC power also impacts the condensate condensate system pumps, which is assumed to result in a loss of feedwater flow to the SGs. The reactor trip signal on low RCS flow generates a turbine trip signal and results in termination of of steam flow due to the closure of the turbine stop valves. With no credit given to the atmospheric atmospheric steam dump and turbine bypass systems, the SG pressure pressure will rapidly approach approach the MSSV MSSVss opening pressure.

The MSSVs will become become the pathway for decay heat removal. Prior to AFW initiation, the SG liquid inventory will inventory will slowly slowly deplete deplete due to the steam blowdown through MSSV MSSVs. s. As the SG liquid inventory decreases and temperature temperature increases, the SG heat transfer capability capability will be reduced. Due to the degraded heat transfer transfer capability of the RCS, the primary RCS temperature, temperature, and then the pressure start to increase.

The pressurizer pressure The pressurizer pressure and and level control systems, as well as the pressurizer PORVs, are not creditedcredited in the analysis. In one to two minutes, the RCPs will have completely completely coasted down and the RCS will be in natural circulation, further degrading primary-to-secondary primary-to-secondary heat transfer. The pressurizer safety safetY valves (PSVs)

(PSVs) act to limit the primary RCS pressure. However, the RCS temperature will continue to increase increase until the until the steam relief capacity capacity of the MSSVs matches the decay heat generation generation rate in the core. At 600600 seconds (10 (10 minutes),

minutes), the analysis assumes the operator operator initiates initiates AFW via remote-manual remote-manual operation operation from from the the Control Control Room. The subcooledsubcooled AFWAFW decreases decreases the SG temperature temperature and starts to cool down the RCS.

At 900 seconds seconds (15(15 minutes),

minutes), the analysis assumes assumes the operator, operator, by remote-manual remote-manual operation operation of the atmospheric atmospheric dump valves, initiates plant cooldown.

The key parameters parameters for this event are:

- Initial core power Initial core power

- Initial operating operating conditions conditions

- Decay Decay heat heat assumptions assumptions

- Trip Trip setpoint(s),

setpoint(s), uncertainty uncertainty and and delay time time

- Initial Initial SG liquid liquid inventory inventory (affects (affects SG SG liquid inventory inventory at the low water levellevel reactor trip setpoint)

- Minimum Minimum AFW AFW performance performance andand actuation actuation delay delay time time

- MSSV MSSV setpoints setpoints and capacities capacities 60 60

ATTACHMENT ATTACHMENT (4)

RELOAD RELOAD TRANSITION TRANSITION REPORT REPORT

- PSV setpoint and capacity

- SG blowdown flow rate

- Timing of any operator operator actions This event is classified classified as an AOO which may occur during the life of the plant. principally The principally challenged acceptance acceptance criteria for this event are:

1. The pressures

1. pressures in the reactor coolant and main steam systems should be less than 110% 110% of design values.

2. Fuel cladding integrity shall be maintained maintained by ensuring SAFDLs are not exceeded exceeded (i.e.,

(i.e., the minimum calculated calculated DNBR shall remain above the 95/95 DNB correlation limit).

3. FCM shall not occur.

4. An incident of moderate frequency should should not generate a more serious plant condition without other faults occurring occurring independently.

Per UFSAR Section 14.10, 14.10, the AOR for this event assumes an initial core power level of2754 of 2754 MWt, and an MTC bounding the most positive MTC allowed at HFP by Technical Specifications was used. These Technical Specifications These values are not impacted impacted by the transition to AREV AREVA A Advanced Advanced CE-14 HTP fuel and remain bounding.

The event behavior is predominantly predominantly a function of the primary-to-secondary primary-to-secondary heat transfer capability.

Therefore, small perturbations in parameters parameters such as the core pressure drop, core core bypass bypass flow fraction, core inlet flow distribution, and reactivity feedback do not impact the parameters of interest in assessingassessing the acceptance acceptance criteria. The plant system characteristics characteristics that potentially impact the key parameters parameters listed listed for this event remain unchanged for both the transition fuel cycle, cycle, and the full core implementation implementation of of AREVA Advanced AREV A Advanced CE-14 HTP fuel at Calvert Cliffs. The cause of the event and the parameters parameters which control control the consequences of the event are unchanged unchanged from or bounded by the current AOR presentedpresented in UFSAR UFSAR Section 14.10. The input assumptions for the radiological consequence consequence analysis of this event also remain unaffected by the fuel transition. Therefore, an analysis of the Loss-of-Non-Emergency Loss-of-Non-Emergency AC Power event is not required to support the transition transition to AREVA AREV A Advanced Advanced CE-14 HTP fuel.

6.3.11 6.3.11 Control Element Element Assembly Drop Event (UFSAR (UFSAR Section 14.11)14.11)

The CEA Drop event is initiated by de-energizing a control rod drive mechanism or by a mechanical mechanical fault associated associated with a CEA drive during power operation. The result is that a single control rod falls into the core.

In response to the negative reactivity insertion when the control control rod drops into the core, the core power power decreases decreases rapidly rapidly at first. A decrease decrease in the moderator temperature temperature results from the initial power power reduction. At EOC conditions, a strongly negativenegative MTC can return the reactor to the full-power condition condition with elevated elevated radial power peaking corresponding corresponding to the new radial power distribution caused by the dropped control rod. Elevated cladding Elevated cladding heat fluxes and fuel temperatures temperatures in the hot assembly may result approaches to the DNB and FCM SAFDLs.

in approaches The event is postulated to occur when the plant is operating operating in manual CEA control with no load limit control.

control. As stated in UFSAR Section Section 7.4.2.2, the Calvert Calvert Cliffs units do not operate with automatic automatic control control of CEAs.

If the plant is operating in the manual manual rod control mode without load limiting control, the core power will typically typically return to a full-power condition following a dropped rod transient. The return to power full-power condition power following a dropped dropped rod transient is limited by the excess capacity capacity of the plant's turbine control valve. In response to a decrease secondary-side steam flow resulting from the initial drop in core power, the decrease in the secondary-side 61

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION turbine valve will throttle open in an attempt to maintain a constant load demand. demand. If the reactivity worth sufficiently large, the turbine valve will not have enough excess capacity for the of the dropped rod is sufficiently reactor to return to full power. The lower power level could be offset, however, by the higher peaking reactor peaking factor associated with a high worth dropped rod.

The dropped CEA event event may be terminated by a TMiLP TMILP trip, a V1-PT potentially reach a new VHPT trip, or potentially equilibrium state without resulting in a reactor trip. A reactor equilibrium reactor trip is not expected expected for this event due to the limited worth of a single or dual (shutdown group) CEA.

The key parameters for this event are:

- Initial operating operating conditions

Moderator Moderator reactivity feedback

- Trip setpoint(s), uncertainty uncertainty and delay time

- Core power (NI & AT) ~ T) signal decalibration decalibration

- Worth of dropped rodrod

- Turbine Turbine control valve operation operation classified as an AOO which may occur during the life of the plant. This event does not This event is classified not challenge to peak pressure.

significant challenge provide a significant pressure. Therefore, the principally principally challenged acceptance criteria challenged acceptance criteria for this event are:

maintained by ensuring SAFDLs are not exceeded

1. Fuel cladding integrity shall be maintained

1. exceeded (i.e.,

(i.e., the calculated DNBR shall remain above the 95/95 DNB correlation minimum calculated correlation limit).

2. FCM should not occur.

parameters listed for this event, such as dropped rod worth are potentially impacted by Some of the key parameters the transition transition to AREVA Advanced CE-14 HTP fuel. As such, the acceptance criteria AREV A Advanced criteria specified specified for this Consequently, the event has been reanalyzed event must be evaluated to support the fuel transition. Consequently, reanalyzed in NRC-approved methodology accordance with the NRC-approved methodology described in Section 6.3.1, 6.3.1, using the AREVA non-LOCA methodology (Reference Departure from nucleate (Reference 13). Departure nucleate boiling ratio analyses analyses will employ employ appropriate NRC-approved correlations in accordance NRC-approved CHF correlations AREV A methodology. The results of accordance with AREVA of the completed CEA drop event analysis are available for NRC audit.

6.3.12 Asymmetric Generator Event (UFSAR Section 14.12)

Asymmetric Steam Generator 14.12)

An Asymmetric SG event is defined as any initiator that affects only one of the two SGs. An excess excess feedwater, a loss of feedwater flow, an excess load, or a loss of load to only one SG would would result in an Asymmetric SG event. The Asymmetric Loss of Load event is the most limiting of the postulated Asymmetric postulated Asymmetric Asymmetric SG events, events, as presented presented in the UFSAR.

Excess Feedwater Feedwater An asymmetric Excess Feedwater event is initiated at HFP by a malfunction Excess Feedwater malfunction in one of the feedwater feedwater instantaneously fully opens the feedwater controllers, which instantaneously feedwater regulator valve to one SG. The full openingopening subcooled feedwater to enter the SG which lowers of the feedwater regulator valve causes additional subcooled lowers the temperature and pressure. The result is a reduction temperature reduction in the steam flow from the affected SG. The excess feedwater also causes the affected SG cold leg temperature temperature to decrease decrease because additional heat is being extracted.

extracted.

62

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION The analysis assumes the turbine demand demand remains constant, which causes the unaffected unaffected SG to pick up up part of the load by further opening the turbine control valve. The increased increased steaming rate results in lowering the temperature temperature of the SG and therefore therefore the cold leg temperatures.

temperatures.

decrease in the core inlet temperature is a temperature The result of the asymmetric decrease temperature and power tilt across the core. Since the increased increased feedwater flow rate only decreases decreases the temperature temperature slightly, there will be a small increase in radial peaks and core power. The event will be terminated by the Asymmetric Asymmetric Steam Generator Protection Protection Trip (ASGPT).

The resulting asymmetry asymmetry in the RCS cold leg temperatures temperatures between the affected affected and unaffected unaffected loops is very very small. There There would would be no significant augmented peaking peaking due to this slight asymmetry asymmetry in RCS cold leg temperatures.

temperatures. Since there is no significant significant asymmetry asymmetry in RCS cold leg coolant temperatures temperatures for this event and the decrease decrease in RCS coolant temperatures temperatures due to increased increased MFW to one SG is much smaller for this event than for the excess excess load caused by the Excess Load event (UFSAR Section 14.4), 14.4), this event event results in a much smaller smaller positive positive moderator reactivity feedback. Therefore, Therefore, this Asymmetric Asymmetric SG event is event is bounded by the Excess Load event.

Loss of Feedwater Feedwater An asymmetric Loss of Feedwater event is initiated at HFP by a malfunction in one of the feedwater feedwater controllers which instantaneously instantaneously shuts the feedwater feedwater regulator regulator valve to one SG. The closure of the feedwater regulator valve causes causes a loss of feedwater to the SG. The loss of feedwater flow will cause cause the temperature temperature and pressure pressure to increase in responseresponse to the decreasing decreasing SG level. The temperature and pressure in the unaffected (i.e., with feedwater unaffected SG (i.e., feedwater flow available) available) also increases increases in response to the increased increased turbine header header pressure. The core inlet temperature temperature from both SGs will increase with the decr6ased secondary decreased secondary heat transfer. A slight core inlet temperature asymmetry occurs with the higher inlet temperature asymmetry temperature resulting from the affected SG.

The small core inlet temperature temperature tilt will not causecause a significant radial power tilt. The slight increase in in core temperatures temperatures in conjunction with a negative MTC will result in a decrease decrease in core average average power.

The event will be terminated terminated by the ASGPT or a Low SG Level Trip.

The increase in core inlet temperature temperature for the asymmetric asymmetric Loss of Feedwater event will be less than that for the Loss of Feedwater Feedwater event (UFSAR Section Section 14.6), which considers considers the instantaneous failure of both feedwater controllers. Hence the associated associated power increase will also be less.

As noted, the presence presence of a negative MTC in conjunction with the increase in core inlet temperature temperature will cause the reactor power to decrease. Hence fuel integrity is not in question and the relevant relevant acceptance acceptance criterion is associated associated with preserving preserving the secondary secondary and primary primary pressure boundary. Loss of Feedwater Feedwater to both SGs will result in higher pressure increase increase in the secondary secondary and primary systems than this event.

Consequently, whether this event causes overcooling or overpressurization, overpressurization, this event is bounded bounded by the Loss of Feedwater Feedwater Flow event (to both SGs), describeddescribed in UFSAR Section 14.6. 14.6.

Excess Load asymmetric Excess Load event is initiated at HFP by the inadvertent An asymmetric inadvertent opening opening of a single secondary secondary safety valve on one SG. The excess load on a single SG causes its pressure and temperature temperature to decrease which results in a decrease in the core inlet temperature. temperature from only one SG temperature. Since the temperature SG decreases, a core inlet temperature temperature distribution tilt occurs occurs across the core. In the presence presence of a negative MTC, positive positive moderator reactivity reactivity feedback feedback occurs that increases the core power. A new steady state 63

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT condition is obtained obtained once the core power increases to match the excess excess load demand. The event will be terminated by the ASGPTASGPT or Low SG Level Trip.

particularly large asymmetry in steam flow from each SG. In addition, the The event does not result in a particularly increase is small in comparison to Excess Load event (UFSAR Section 14.4),

load increase 14.4), which is initiated by by opening of the two steam dump valves and the four turbine bypass valves. Hence the opening Hence this event will most most likely reach steady state condition without a reactor trip. In any case, it is bounded by the Excess reach a new steady Load event (UFSAR Section 14.4) which results in a greater challenge to the SAFDLs.

greater challenge Loss of Load An asymmetric Loss of Load event is initiated at HFP by an inadvertent inadvertent closure of a single main steam (MS1V) on one SG. The loss of load to a single SG causes the pressure and temperature isolation valve (MSrv) on the SG to increase. With the decrease in SG heat transfer, the core inlet temperature from the isolated SG will increase. The isolated SG water level drops rapidly as the increasing pressure pressure collapses the steam bubble in the liquid inventory. The pressure will continue to increase until the MSSVs open.

The analysis assumes the turbine load demand remains constant, which causes the turbine control valves valves increased load demand to open further. The increased demand will decrease the other (i.e.,(i.e., unaffected) SG pressure pressure and temperature from the SG will also decreased temperature, the core inlet temperature temperature. In response to the decreased decrease. Present operating practice maintains the turbine control control valve flow area constant, which will lessen the severity event.

severity of the .event.

The result of the outlet temperature increase and decrease from their respective temperature increase respective SGs is a severe severe core inlet temperature maldistribution. In the presence of negativenegative MTC and fuel temperature coefficient coefficient (normally negative at power), the coolant temperature tilt will cause a radial power shift toward the cold side of the coolant temperature where there is almost no mixing of the inlet flow, will outermost fuel bundles, where core. The power in the outermost increase. The power on the hot side of the core will decrease greatest local power increase.

experience the greatest experience decrease due to the negative moderator reactivity negative moderator reactor trip to terminate the event.

reactivity feedback. The ASGPT will initiate a reactor The key parameters parameters for this event are:

- Initial operating operating conditions conditions

- Trip setpoint(s), uncertainty and delay time

- PSV and/or PORV setpoints and capacities capacities

- MSSV setpoints and capacities capacities This event is classified as an AOO which may occur occur during the life of the plant. This event does not Therefore, the principally provide a significant challenge to peak pressure. Therefore, challenged acceptance principally challenged acceptance criteria for these events are:

cladding integrity shall be maintained

1. Fuel cladding

1. ensuring SAFDLs are not exceeded maintained by ensuring (i.e., the exceeded (Le.,

minimum calculated DNBR shall remain above the 95/95 DNB correlation limit).

2. FCM should not occur.

asymmetric loss of load, is not a particularly Asymmetric SG event, asymmetric The limiting Asymmetric challenging event.

particularly challenging However, this event will be analy?ed analyzed as the limiting asymmetric completeness under the asymmetric event for completeness AREVA methodology. As such, the acceptance specified for this event must be evaluated to acceptance criteria specified support support the fuel transition. Consequently, the Asymmetric Loss of Load event will be reanalyzed using the AREVA non-LOCA (Reference 13). Departure from nucleate boiling ratio analyses non-LOCA methodology (Reference 64 64.

ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION REPORT NRC-approved CHF will employ appropriate NRC-approved CHF correlations in accordance with AREVAREVA A methodology. The The results of the Asymmetric SG event analysis will be available for NRC audit when the analysis has been completed.

6.3.13 Control Element Assembly Ejection Eoection (UFSAR Section 14.13) 14.13)

The CEA Ejection event is initiated by a postulated rupture of a control rod drive mechanism housing.

Such a rupture allows the full full system pressure to act on the drive drive shaft, which ejects its control rod from consequences of the mechanical failure are a rapid positive reactivity insertion and an the core. The consequences increase in radial power peaking, which could possibly lead to localized fuel rod damage.

Doppler reactivity feedback mitigates the power excursion excursion as the fuel begins to heatup. Although the initial increase in power occurs too rapidly for the scram rods to have any effect on the power during that portion of the transient, the negative reactivity reactivity insertion from the reactor scram does affect the total energy input to the fuel and subsequently the peak fuel centerline temperature and fuel rod cladding surface heat flux.

The key parameters parameters for this event are:

- Initial core Initial core power

- Initial operating Initial operating conditions conditions

" Ejected rod worth Ejected

- Doppler reactivity Doppler reactivity feedback feedback

- Trip setpoint(s), uncertainty and delay time

- Number Number of RCPs running (HZP cases) cases)

- Fuel Fuel rod rod gap conductance conductance

- Post ejection Fq predicted predicted for the purpose calculating the peak purpose of calculating peak (hot spot) fuel centerline centerline temperature temperature

- Percentage of fuel experiencing centerline Percentage centerline melt and cladding damage damage

- Meteorology Meteorology

- Technical Specification Technical Specification RCS iodine limit

- Radiological source terms Radiological terms

- Core power level Core power level This This event event is classified classified as a Postulated Postulated Accident, which which is not expected expected to occur during the lifetime lifetime of the plant, but plant, but must must be evaluated to demonstrate be evaluated demonstrate the adequacy adequacy of the plant design. The principally principally challenged challenged acceptance acceptance criteria for this event are:

1. The

1. radial-average fuel pellet The radial-average enthalpy at the hot spot must be :s pellet enthalpy < 280 cal/g.

2.

2. The The maximum maximum RCS pressure pressure during any portion portion of the transient transient must remain below the emergency emergency condition condition stress limit as defined defined in Section Section III of the ASME ASME Boiler and Pressure Vessel Vessel Code Code (120%

(120% of the design pressure). (This(This pressure pressure limit is higher higher than the 110%

110% design pressure pressure criterion criterion used used for most other design-basis design-basis events.)

events.)

3. IfIf fuel failure is predicted, predicted, the radiological radiological consequences consequences must not exceed exceed the limits limits defined defined in Regulatory Regulatory Guide Guide 1.183, 1.183, Table Table 6.

65 65

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION The Alternate Source Term (AST) radiological radiological consequence consequence analysis for CEA ejection assumed that 8% 8%

of the fuel will reach incipient centerline centerline melt and 2% will experience experience clad damage. The 8% of the fuel that is assumed to melt is also assumed to experience clad failure. Fuel rods experiencing experiencing cladding cladding failure release release the gap inventory of noble gasses and iodines. This is conservative, but consistent with earlier earlier assumptions of 10%10% fuel failure for previous design basis TID-14484 radiological calculations. Analyses have demonstrated demonstrated that the current current AST core source terms and Calvert Cliffs specific specific non-LOCA non-LOCA gas gap fractions used in the AST Seized Seized Rotor analysis remain remain bounding for AREVAAREV A Advanced Advanced CE-14 HTP fuel burnable poison. Other inputs to this radiological analysis (i.e.,

using Gd 220 3 burnable (i.e., core power, meteorology, Technical Technical Specification RCS iodine limit) remain unchanged unchanged by the use of AREVA AREVA Advanced Advanced CE-14 CE-14 HTP fuel. Therefore, further radiological analyses will not be performed performed provided that the thermal hydraulic hydraulic analysis demonstrates that the combined percentage of fuel experiencing combined percentage centerline melt and experiencing centerline fuel experiencing experiencing cladding damage damage remains below 10% 10% for the CEA Ejection event.

parameters listed for this event, such as ejected rod worth and Doppler Some of the key parameters Doppler reactivity feedback, are potentially impacted impacted by the transition to AREVA Advanced CE-14 HTP fuel. As such, the AREV A Advanced acceptance criteria specified specified for this event must be evaluated to support the fuel transition. Consequently, Consequently, the event has been reanalyzed accordance with the NRC-approved reanalyzed in accordance NRC-approved methodology described in 6.3.1, using the AREVA Section 6.3.1, ARE VA non-LOCA methodology (Reference (Reference 13,13, Section 5.8). Departure Departure from nucleate nucleate boiling ratio analyses will employ appropriate NRC-approvedNRC-approved CHF correlations accordance correlations in accordance with AREVA methodology.

methodology. The results of the completed CEA Ejection event analysis are available available for NRC audit.

6.3.14 Steam Line Break Event (UFSAR (UFSAR Section 14.14) 14.14)

The MSLB event is analyzed analyzed for post-scram post-scram return-to-power return-to-power behavior and pre-scram behavior.

Post-Scram MSLB The post-scram MSLB event is initiated by a break break in a main steam line upstream of the MSIV. The maximum breakbreak size (i.e., double-ended guillotine break)

(i.e., a double-ended guillotine break) is limiting limiting for the post-scram return-to-power return-to-power consequences of an MSLB event because it maximizes the rate .ofcooldown consequences of cooldown and positive reactivity feedback.

The rupture of a main steam line will cause the affected SG pressure and temperature temperature to rapidly decrease.

decrease.

This in turn tum will cause a rapid cooldown in the RCS loop containing the affected affected SG and in the core sector sector cooled primarily by water from the cold legs of the affected affected loop. Other loops and related core sectors will cool at a lesser lesser rate, depending on the various mixing and/or crossflow phenomena present within the crossflow phenomena reactor vessel. The drop in SG pressure will initiate a SG isolation isolation signal. Following appropriate appropriate delays, the MSIVs on both the affected affected and unaffected unaffected SGs will close and terminate terminate the blowdown blowdown from the unaffected SG.

Due to cooldown of the RCS, the RCS coolant will contract. This may cause the pressurizer to empty and the RCS pressure to decrease decrease rapidly. Water in the reactorreactor vessel upper head may flash if this region is fairly stagnant. Upper head flashing will act to delay the RCS pressure decay once the saturation saturation pressure of the upper head is reached. This in tum turn will delay the injection injection of borated water by the safety injection actuation signal (SIAS). Higher primary system back-pressure back-pressure will also result in lower flow from the SIAS, lengthening lengthening the time it takes for boron to reach the core.

The cooldown cooldown of the RCS will insert positive positive reactivity reactivity from both moderator and fuel temperature (particularly at EOC conditions with a most-negative reactivity feedback (particularly most-negative MTC). This positive reactivity addition will erode the core shutdown margin, especiallyespecially when considering considering the most reactive CEA stuck stuck 66 66

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION out of the core. The magnitude of the reactivity transient depends depends on the minimum mlllimum Technical Technical Specification shutdown margin, Specification margin, the worth of the stuck CEA, and positive reactivity insertion from the moderator moderator and fuel temperature temperature reactivity reactivity feedback. The core will remain subcritical subcritical shortly after reactor trip - due to either a scram at power, a scram at critical HZP, or sub subcritical critical initial conditions.

Reactor trip would be expected Reactor expected to occur on one of the following reactor trips: containment containment high pressure (for breaks breaks inside Containment),

Containment), low SG pressure, or the ASGPT. No credit is taken in the post-scram analysis for reactor trip or MSIV closure on a predicted high containmentcontainment pressure. The post-scram MSLB event is analyzed analyzed with and without a loss of offsite power and considers considers the effect of a single-failure.

Pre-Scram Pre-Scram MSLB pre-scram phase of an MSLB event The pre-scram event can also challenge challenge the SAFDLs because because of the rate of the primary system coolant coolant temperature temperature decrease combined combined with power power decalibration decalibration (both NI-power ~T-power NI-power and AT-power signals) and the potential potential effect of harsh Containment conditions conditions on reactor trips for cases with the break inside Containment.

Break Break sizes ranging up to a double-ended double-ended guillotine break in a main steam line are evaluated evaluated in the pre-scram MSLB analysis. The system response for the pre-scram phase of the MSLB event is similar to that of the increase in steam flow event. If the break break is large enough, the reactor reactor will trip on a low SG SG pressure pressure signal, a low pressurizer pressurizer pressure signal, a containment containment high pressure pressure signal if the break break is inside Containment, or an ASGPT. Smaller breaks will prolong the cooldown until the reactor trips on an VHPT signal or a containment containment high pressure signal (for breaks inside the reactor containment) or until the reactor containment) reactor reaches a new steady state condition condition at an elevated elevated power level. The pre-scram pre-scram MSLB event is analyzed with a loss of offsite power occurring coincident coincident with reactor trip to further challenge challenge the DNB SAFDL.

The key parameters parameters for this event are:

Initial core power power

- Initial operating conditions

- Initial SG inventory

- Break Break size and location location

- Moderator reactivity reactivity feedback

- Doppler reactivity reactivity feedback

- Trip setpoint(s), uncertainty uncertainty and delay time

- Core power (Nl (NI & AT)

~ T) signal decalibration decalibration

Minimum Minimum HFP scram worth for HFP cases cases and Technical Specifications Specifications minimum shutdown shutdown margin for HZP cases cases

- AFW flow rate and delay time

- Safety injection flow rate and delay time

- MSIV closure time

- MFW isolation time

- Post-scram Post-scram radial power peaking factors

- Radiological Radiological source terms 67 67

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT

- Specification primary and secondary Technical Specification secondary iodine activity limits

Primary-to-secondary leak rate Primary-to-secondary

- Meteorology This event is classified as a Postulated Accident, which is not expected to occur during the lifetime of the evaluated to demonstrate the adequacy plant, but must be evaluated adequacy of the plant design. The principally challenged principally challenged acceptance criterion for this event is:

acceptance

1. If fuel failure is predicted, the radiological I. radiological consequences consequences must not exceed exceed the Regulatory 1.183, Table 6 limits.

Guide 1.183, The AST radiological consequence consequence analysis for MSLB event event currently assumes that 0.8% 0.8% fuel failure occurs releasing the gap inventory inventory of noble gasses and iodines. Analyses have demonstrated that the current AST core sourcesource terms and Calvert Cliffs specific non-LOCA gas gap fractions used in the AST AST Seized Rotor analysis remain bounding for AREVA AREV A Advanced Advanced CE-14 HTP fuel using Gd220 3 burnable poison. Other inputs to this radiological radiological analysis remain unchanged unchanged by the use of AREVA AREV A Advanced Advanced CE-14 HTP fuel. Therefore,Therefore, further radiological radiological analyses will not be performed performed provided provided that the thermal-hydraulic analysis demonstrates that the MSLB failed fuel fraction remains below 0.8%.

thermal-hydraulic 0.8%.

Some of the key parameters parameters listed for this event, such as Moderator and Doppler reactivity feedback, are potentially impacted by the transition to AREV AREVA A Advanced Advanced CE-14 HTP fuel. As such, the acceptance acceptance criteria specified for this event must be evaluated to support the fuel transition. Consequently, Consequently, the event reanalyzed in accordance has been reanalyzed accordance with the NRC-approved NRC-approved methodology described in Section 6.3.1, methodology described 6.3.1, using the AREVA non-LOCA methodology methodology (Reference (Reference 13,13, Section Section 5.4). Departure Departure from nucleate nucleate boiling ratio analyses will employ NRC-approved CHF correlations in accordance employ appropriate NRC-approved accordance with AREV AREVA A methodology. The results of the completed MSLB event analysis analysis are available for NRC audit.

6.3.15 Steam Generator Generator Tube Rupture Event (UFSAR (UFSAR Section 14.15)

The SG Tube Rupture event is initiated by a double-ended double-ended break of a single SG tube. Coolant from the RCS begins to escape escape through the break, driven by the pressure differential between the RCS and the SG secondary side, increasing the inventory and pressure in the affected secondary affected SG.

As the break flow begins to depressurize depressurize the RCS, the charging charging pumps pumps activate activate in order to make-up make-up the pressurizer heaters energize on decreasing lost inventory and pressurizer decreasing pressure.

pressure. If the RCS inventory inventory and pressure are stabilized via the charging charging pumps, no reactor trip will occur. However, if the break flow exceeds exceeds the capacity of the charging pumps, the RCS pressure pressure and inventory will continue to decreasedecrease resulting in a reactor trip on a low RCS pressure signal (TM/LP or low pressurizer pressurizer pressure).

pressure). Following Following the reactor reactor trip, the turbine will trip and, in the case where offsite power power is lost, the RCPs will coast down and make-up flow will terminate terminate until emergency emergency diesel generator power is available. If offsite power is available, a fast transfer transfer to the offsite power will keep the RCPs running and the make-up make-up flow available.

available.

The loss of offsite power results in the loss of condenser condenser vacuum and the steam dump to condenser valves valves are closed to protect the condenser. The continued mass and energy energy transfer between between the RCS and secondary side results in an increase in the affected secondary affected SG pressure pressure and discharge discharge to the atmosphere via the MSSVs and atmospheric dump valves.

As the RCS pressure continues continues to decrease, decrease, a low pressurizer pressurizer pressure signal activates the SIAS. SIAS. The emergency emergency diesels start and high pressure safety injection (HPSI)

(HPSI) flow begins begins once the shutoff head of the HIPSI HPSI pumps has been reached. For some plants, the HPSI pumps have a very high delivery head which 68

ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT re-pressurization of the RCS. In this case, a high break flow may result in an earlier and more significant re-pressurization rate is maintained maintained leading to to" a more rapid filling of the affected SG. This may lead to liquid in the undesirable since it may cause the MSSVs to fail open and potentially steamlines and MSSVs, which is undesirable steamlines potentially damage the steam piping due to the weight of the water on the pipe supports.

The operators will take a series of actions to regain control control of the plant systems and to bring the RCS to a condition allowing for initiation of the residual residual heat removal system.

The key parameters parameters for this event event are:

- Initial core power power

- Decay heat assumptions

" Initial conditions conditions

" Initial SG liquid inventory

- Trip setpoint(s), uncertainty and delay time

" SG tube break break area

" Primary-to-secondary pressure Primary-to-secondary pressure difference

" performance and actuation delay AFW performance

" Safety injection performance p~rformance

" PORV capacity

- Atmospheric dump valve capacity

- Operator actions Postulated Accident, which is not expected to occur during the lifetime of the This event is classified as a Postulated demonstrate the adequacy evaluated to demonstrate plant, but must be evaluated adequacy of the plant design. The principally challenged principally challenged acceptance criterion for this event is:

acceptance

1. The radiological consequences

1. consequences must not exceedexceed the Regulatory Regulatory Guide 1.183,1.183, Table 6 limits.

Per UFSAR Section 14.15, the AOR for this event assumes an initial core power level of 2754 MWt.

Thi§ value is not impacted This impacted by the transition AREV A Advanced CE-14 HTP transition to AREVA RTP fuel and remains bounding.

The event event behavior predominantly a function of the primary-to-secondary behavior is predominantly pressure differential, break primary-to-secondary pressure size, and atmospheric atmospheric dump valve capacity. Therefore, small perturbations parameters such as the core perturbations in parameters pressure drop, core bypass flow fraction, core inlet flow distribution, and reactivity feedback do not parameters of interest impact the parameters interest in assessing the acceptance criteria. The plant system characteristics system characteristics parameters listed for this event remain unchanged for both the transition that potentially impact the key parameters implementation of AREVA fuel cycle, and the full core implementation AREV A Advanced CE-14 RTP HTP fuel at Calvert Cliffs. The cause of the event event and the parameters parameters which control control the consequences consequences of the event are unchanged from or bounded by the current AOR presented AOR presented in UFSAR Section 14.15.

Section 14.15. The input assumptions for the radiological consequence consequence analysis of the SG Tube Rupture event listed in UFSAR Section 14.15, 14.15, also remain unaffected introduction of AREVA unaffected by the introduction AREV A Advanced Advanced CE-14 HTP RTP fuel. Therefore, an analysis of of the SG Tube Rupture Rupture event is not required to supportsupport the transition AREV A Advanced transition to AREVA Advanced CE-14 HTP RTP fuel.

69

ATTACHMENT ATTACHMENT (4) (4)

TRANSITION REPORT RELOAD TRANSITION 6.3.16 Seized Rotor Event (UFSAR (UFSAR Section 14.16)

The Seized Rotor event is postulated to be caused by the instantaneous instantaneous seizure of a RCP rotor. Flow through the faulted RCS loop rapidly decreases, causing a reactor trip on a Low RCS Loop Flow signal within 1 to 2 seconds and a turbine trip on the reactor trip.

continues to be transferred to the reactor coolant.

Following the reactor trip, heat stored in the fuel rods continues combination of the relatively high fuel rod surface heat fluxes, decreasing The combination decreasing core flow, and increasing increasing core coolant coolant temperatures challenges the DNBR safety limit.

temperatures challenges At the same time, the SG primary-to-secondary primary-to-secondary heat transfer rate decreases, because (1) the decreasing primary coolant coolant flow degrades the SG tube primary-side coefficients and (2) primary-side heat transfer coefficients (2) the turbine trip causes the secondary-side decreasing rate of heat removal in the SGs and temperature to increase. The decreasing secondary-side temperature and the decreasing flow of coolant removing heat from the reactor core cause the reactor coolant removing reactor coolant to heatup.

overpressurization of The resultant reactor coolant expansion causes fluid to surge into the pressurizer and overpressurization of pressurizer spray system and may even open the pressurizer the RCS. This may actuate the automatic pressurizer pressurizer PORVs.

The most limiting Seized Rotor event is defined as a complete (i.e. binding) of a instantaneous seizure (i.e.

complete instantaneous single RCP shaft at HFP (UFSAR Section 14.16). The reactor coolant flow through the core would be be asymmetrically reduced asymmetrically reduced to three-pump three-pump flow as the result of a shaft seizure on one pump.

The key parameters parameters for this event are:

- operating conditions Initial operating conditions

- RCS coolant inertia

- resistance RCS loop resistance

" Seized rotor impeller impeller resistance

- Trip setpoint(s), uncertainty uncertainty and delay time

- Minimum HFP scram worth worth

- Fraction of scram scram reactivity reactivity versus fraction of control rod insertion distance distance at HFP

- Scram delay time

- Fuel rod gap conductance conductance This event is classified as a Postulated Accident, which is not expected to occur during the lifetime of the adequacy of the plant design. The FCM SAFDL evaluated to demonstrate the adequacy plant, but must be evaluated SAFDL is not challenged because there is no significant increase significant increase in power for this event. The overpressure overpressure transient transient bounded by the Loss of Load (UFSAR Section response for this event is bounded 14.5) due to the rapid loss of Section 14.5) of challenged acceptance primary-to-secondary heat transfer. The principally challenged primary-to-secondary acceptance criteria for this event are:

integrity shall be maintained by ensuring SAFDLs are not exceeded

1. Fuel cladding integrity

1. exceeded (i.e.,

(i.e., the calculated DNBR shall remain above the 95/95 DNB correlation minimum calculated correlation limit.

2. If fuel failure is predicted, predicted, the radiological consequences radiological consequences must not exceed the Regulatory 1.183, Table 6 limits.

Guide 1.183, The AST radiological consequence consequence analysis currently assumes that 5%

analysis for Seized Rotor event currently 5% fuel failure occurs releasing the gap inventory inventory of noble gasses and iodines. Analyses have demonstrated that the current AST core source terms and Calvert Cliffs specific non-LOCA gas gap fractions used in the AST AST Seized Rotor analysis remain bounding for AREVA Advanced CE-14 HTP AREV A Advanced RTP fuel using Gd 20 3 burnable burnable 70 70

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION poison. Other inputs to this radiological analysis remain unchanged by the use of AREVA AREV A Advanced Advanced CE-14 HTP fuel. Therefore, Therefore, further radiological radiological analysis will not be performed provided provided that the thermal hydraulic hydraulic analyses demonstrate demonstrate that the failed fuel fraction remains below 5% for a Seized Rotor event.

parameters listed for this event, such as minimum HFP scram worth and fuel rod gap Some of the key parameters conductance, conductance, are potentially impacted by the transition to AREV AREVA A Advanced CE-14 HTP fuel. As such, acceptance criteria the acceptance criteria specified specified for this event must be evaluated evaluated to support the fuel transition.

Consequently, Consequently, the event has been reanalyzedreanalyzed in accordance accordance with the NRC-approved methodology NRC-approved methodology described described in Section 6.3.1, 6.3.1, using the AREVA non-LOCA methodology (Reference 13). Departure (Reference 13).. Departure from nucleate nucleate boiling ratio analyses analyses will employ appropriate appropriate NRC-approved NRC-approved CHF correlations correlations in accordance AREVA with AREV A methodology. The results of the completed Seized Rotor event analysis analysis are available for NRC audit.

6.3.17 Loss-of-Coolant Accident (UFSAR Loss-of-Coolant Accident (UFSAR Section 14.17)14.17)

The loss-of-coolant accident is analyzed to assure that the design bases for the ECCS satisfy the requirements requirements of 10 CFR 50.46 regarding regarding ECCS acceptance acceptance criteria.

Small Small Break Loss-of.,Coolant Loss-of-Coolant Accident Accident A SBLOCA is defined defined as a break in the RCS pressure boundary boundary which has an area of up to approximately approximately 10%

10% of a cold leg pipe area. The most limiting limiting break location is in the cold leg pipe on the discharge discharge side side of the RCP, which results in the largest amountamount of inventory loss and the largest fraction of ECCS fluid being ejected out through the break. This behaviorbehavior produces the greatest degree of core uncovery, the longest fuel rod heatup time, and consequently, consequently, the greatest greatest challenge to the 10 CFR 50.46(b)(1-4) 50.46(b)(l-4) criteria.

The SBLOCA event event is characterized characterized by a slow depressurization depressurization of the RCS with a reactorreactor trip occurring occurring on a low pressurizer pressurizer pressure signal. The SIAS occurs when the system system has further depressurized. The capacity and shutoff head of the HPSI pumps are important important parameters in the SBLOCA analysis. For the limiting break size, the rate of inventory loss from the primary primary system is such that the HPSI pumps cannot preclude significant core uncovery. The primary system depressurization depressurization rate is slow, extending the time required required to reach the safety injection tank pressure pressure or to recover recover core liquid level on HPSI flow. This tends to maximize maximize the heatup time of the hot rod which produces produces the maximum maximum peak clad temperature temperature .

(PCT) and local cladding oxidation. Core recovery for the limiting break begins when the HPSI flow that is retained in the RCS exceeds exceeds the mass flow rate out the break, followed by injection injection of safety injection safety injection tank flow. For very small break break sizes, the RCS pressure does not reach the safety safety injection tank pressure.

The AREVA S-RELAP5 SBLOCA evaluation model for event response AREV A S-RELAP5 response of the primary primary and secondary systems systems and hot fuel rod used in this analysis (Reference (Reference 19) consists of two computer computer codes, S-RELAP5 S-RELAP5 and RODEX2/2A, RODEX2/2A, which are described described in Section 6.2. The appropriate conservatisms, conservatisms, as prescribed prescribed byby Appendix K K of 10 CFR 50, are incorporated. This ReferenceReference 19 methodology methodology has been reviewed and approved approved by the NRC to perform perform SBLOCA analyses. The results of the completed completed SBLOCA SBLOCA analysis are available available for NRC audit.

Large Break Loss-of-Coolant Accident Accident The large break LOCA event characterized by a postulated event is characterized postulated large rupture in the RCS cold leg. The RLBLOCA analysis considers a break range from 2.817 ft RLBLOCA ft22 to 9.819 ft W2 (See Section 4.6 in Enclosures 1I and 4). Two scenarios scenarios are run, both with loss of offsite powerpower and no loss of offsite power. The non-parametric parametric statistical approach approach of the RLBLOCA analysis samples key plant parameters parameters such as break 71

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT REPORT size and pressurizer operational range. A mixed pressurizer pressure through an operational mixed core of AREV AREVA A Advanced CE-14 Advanced CE-14 HTP fuel and existing Westinghouse Westinghouse Turbo fuel is modeled for the analysis. The full list of sampled sampled parameters and their range of values as well as more detailed large break LOCA event description parameters description may be found in Enclosures 1 and 4.

The large break LOCA analysis is performed for Calvert Cliffs by applying the S-RELAP5, S-RELAP5, RODEX3A, and ICECON computer codes outlined in Section Section 6.2 above. The large break LOCA LOCA approach approach applied for Calvert Cliffs is based on the methodology methodology documented documented in Reference 20 with specific deviations deviations outlined in Section 1 of the RLBLOCA Summary Summary Report (Enclosures 1 and 4) These deviations are a response response to NRC inquiries related to the methodology update to Reference 20. This altered altered methodology methodology is referred to as the "transition program program or transition package."

package." This methodology methodology follows the Code Scaling, Applicability, and Uncertainty evaluation approach (Reference evaluation approach (Reference 24), which outlines outlines an approach for defining and qualifying qualifying a best-estimate best-estimate thermal-hydraulic thermal-hydraulic code code and quantifies the uncertainties for the large break break LOCA analysis. The RLBLOCA RLBLOCA methodology methodology conforms to the SRP Section 6.3 acceptanceacceptance criteria for realistic realistic evaluation models as described in Regulatory Guide 1.157. 1.157.

Results from the analysis show that the 10 CFR 50.46(b) acceptanceacceptance criteria for PCT, maximum oxide thickness, and hydrogen hydrogen generation are met with significant significant margin. Realistic large break LOCA analysis analysis limitingPCT results show that the limiting ,PCT occurred occurred for a 4 wt% Gd Gd 200 3 rod in a case with loss of offsite offsite power power conditions. This case yielded 'aa limiting PCT of 1670°F.

1670'F. See Enclosures Enclosures 11 and 4 for detailed PCT results.

6.3.18 Fuel Handling Handling Incident (UFSAR Section Incident (UFSAR Section 14.18) 14.18)

The Fuel Handling Handling incident assumes that a fuel assembly assembly is dropped during fuel handling in the containment containment refueling pool or the SFP. The analyses analyses for a Fuel Handling incident in the refueling pool and the SFP both assume that gas gap activity from all 176 fuel pins of the highest power assembly assembly is released. In the SFP the fuel assemblies are stored within the racks at the bottom of the SFP. The top of of the rack extends above the tops of the stored stored fuel assemblies. A dropped fuel assembly could not strike more than one fuel assembly assembly in the storage storage rack. Due to the design of the rack system, impact could only only occur between between the ends of the involved fuel assemblies, i.e., the bottom-end bottom-end fitting of the dropped fuel assembly impacting against against the top-end fitting of the stored fuel assembly. The results of an analysis of of the end on energy absorption capability of a fuel assembly indicate that a fuel assembly is capable of of absorbing the kinetic energy of the drop with no fuel rod failures. The worst Fuel Handling Handling incident that could occur in the SFP is the dropping of a fuel assembly to the fuel pool floor. Because of the high energy absorption required to rupture a single fuel rod, assuming the rupture of all 176 fuel pins in the highest power 14x1414x14 fuel assembly assembly represents a bounding maximum maximum number of pins that could be be damaged in any credible credible fuel handling incident scenario.

The key parameters parameters for this event are:

- Core power level

- Cycle length Cycle length

" Maximum assembly burnupbumup

- Decay time prior to removal from reactor vessel

" assembly Water level above assembly

- Percentage of fuel rods failed 72 72

ATTACHMENT (4)

RELOAD TRANSITION RELOAD TRANSITION REPORT This event is evaluated evaluated to demonstrate adequacy of the plant design. The principally demonstrate the adequacy principally challenged acceptance criterion for this event is:

1.

I. If fuel failure is predicted, predicted, the radiological consequences consequences must not exceed exceed the Regulatory Regulatory Guide 1.183, 1.183, Table 6 limits.

This event is not analyzed with a thermal-hydraulic thermal-hydraulic NSSS transient code such as S-RELAPS. S-RELAP5. For the AST analysis, analysis, the Fuel Handling incident uses a worst case core source term (based (based on a 4.0 wt%

enriched fuel with a 73/72/7273/72172 fuel loading, 24-month 24-month cycle, and a maximum assembly burnup of of 62 GWd/MTU)

GWdIMTU) divided by 217 assemblies per core core and multiplied by a 1.02 1.02 power power measurement measurement uncertainty factor and a 1.70 power peaking factor. Additional uncertainty Additional analysis was performed to demonstrate that the AST core source term and non-LOCA gas gap fractions used in the Fuel Handling incident incident remain bounding bounding for AREVA Advanced AREV A Advanced CE-14 HTP fuel using Gd 220 33 0 burnable poison. The meteorology, water water level, minimum cooling cooling time prior prior to core offload offload are not impacted impacted by the fuel design.

Structural analyses performed performed for the AREV AREVA demonstrate that no mote A lead fuel assemblies demonstrate more than 176176 fuel pins (i.e., all fuel pins in a 14xl4 14x14 fuel assembly) would fail during a Fuel Handling incident, therefore therefore the percentage percentage of failed fuel is unchanged unchanged from the current analysis. Since all of the above inputs are bounded bounded or unchanged, unchanged, the current bounding bounding AST Fuel Handling Handling incident radiological radiological consequence consequence analysis supports the use of AREVA AREV A Advanced CE- CE-14 14 HTP fuel.

6.3.19 6.3.19 Turbine-Generator Turbiue-Generator Overspeed Incident Incident (UFSAR (UFSAR 14.19)

The turbine-generator turbine-generator oversoverspeed peed incident is postulated postulated to be caused by a failure of components that control control admission of steam to the turbine resulting in destructive shaft rotational rotational speed, which may yield turbine-generator turbine-generator produced produced missiles.

Unit I1 Turbine Missile Missile Analysis The Unit 1I turbine turbine (General (General Electric)

Electric) has a missile generation generation probability (P1)(PI) of less than 10- 1055 per year.

As long as the missile generation 5 generation probability is maintained maintained less than 10- 10-5 per year, the Unit I turbine presents an acceptably acceptably low risk and no further analysis analysis of missile risk from the Unit 1I turbine is necessary.

necessary. This also applies applies to the turbine missile risk for all equipment.

Unit 2 Turbine Missile Analysis The Unit 2 turbine (Westinghouse)

(Westinghouse) has a missile generation probability probability (PI)

(P1) of less than 10.'10-5 per year. As 5

long as missile generation generation probability is maintained maintained less than 10- per year, the Unit 2 turbine turbine presents an acceptably acceptably low risk and no further detailed analysis of missile missile risk from the Unit 2 turbine necessary.

is necessary.

Based on the above discussion, both Units' Units' turbine-generators turbine-generators have an acceptably low probability probability of of generating a missile, and Units 1I and 2 are adequately protected against against turbine missiles.

6.3.20 Containment Response (UFSAR (UFSAR Section 14.20)

The containment containment structure encloses RCS (vessel, hot and cold leg piping, RCPs, pressurizer) pressurizer) and the SGs.

Containment Containment is the final barrier barrier against the release release of radioactivity radioactivity to the environment in the event of of accidents accidents with the potential of releasing significant amounts amounts of fission products. Specifically, Specifically, the containment containment structure structure must withstand the pressure pressure and temperature temperature conditions conditions resulting from such postulated design basis accidents as LOCA or MSLB. While other events, such as a feedwater line break break also discharge mass and energy energy into the Containment, the LOCA and MSLB have been confirmed confirmed to be limiting events limiting events with respect to maximizing containment pressure and temperature. Containment maximizing the peak containment Containment heat sink structures and mass as well as active safety systems, such as the containment spray spray and fan cooler cooler trains, reduce the severity severity of the design basis accident and help maintainmaintain the structural integrity.

73

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION The key parameters for this event are:

" Initial NSSS power and core decay heat heat

- (LOCA)

RCS flow rate (LOCA)

- RCS pressure pressure (LOCA)

(LOCA)

- SG pressure pressure (MSLB)

Limiting single-failure single-failure

- Trip setpoint(s),

setpoint(s), uncertainty uncertainty and delay time

MFW flow and temperature temperature (MSLB)

- AFW flow and temperature temperature (MSLB)

- Safety injection flowrate and delay times

- containment cooling (fan coolers and related delay time)

Active containment

- Active containment cooling (sprays and related delay times)

Active containment

- Containment heat conductors conductors (passive heat sinks including shell and internals) internals)

- containment atmosphere Initial containment atmosphere pressure, temperature temperature and relative humidity

- Containment spray This event is classified as a Postulated Occurrence Occurrence which could involve a release of radioactivity.

radioactivity. The acceptance criteria for this event are:

acceptance

1. Pressure and temperature

1. temperature remain below design limits throughout the design basis accidents.

AREVA The transition to AREV A Advanced CE-14 HTP fuel does not affect the containmentcontainment response response to the limiting design basis accidents.

accidents. This is because the mass and energy transfer from both the MSLB and LOCA are primarily primarily a function of core power, the stored energy energy in the primary and secondary RCS and the stored energy in primary primary and secondary secondary RCS intemals.

internals. The initial power level for the limiting mass and energy released was based on a core power of 2754 MWt, which addresses uncertainty uncertainty for the Appendix K K measurement measurement uncertainty uncertainty recapture recapture and is applicable applicable to both the transition and full core.

core implementation cores of AREVA implementation Advanced CE-14 HTP fuel.

AREV A Advanced The cause of the event and the parameters parameters which control the consequences consequences of the event are unchanged unchanged from or bounded by the current AOR. As a result, the mass and energy energy release data for the AOR AOR continue to be applicable to the transition and full core implementation implementation of AREV AREVA A Advanced CE-14 HTP fuel.

The AOR AOR remains bounding and a new containment containment response response analysis to design basis MSLB event and LOCA are not required required to support support the transition to AREV AREVA Advanced CE-14 HTP fuel.

A Advanced 6.3.21 Waste Gas Incident (UFSAR Section 14.22) 14.22)

The most limiting waste gas incident is defined as an unexpected unexpected and uncontrolled uncontrolled release to the atmosphere of the radioactive radioactive xenon and krypton fission gases that are stored in one waste gas decay tank.

As the components of the waste gas system are subjected subjected to pressures pressures no greater than 150 psig, a failure is is not likely. However, a rupture of a waste gas decay tank is analyzed to define define the limit of the hazard that could result from any malfunction in the radioactive radioactive waste gas system.

74 74

ATTACHMENT ATTACHMENT (4)

TRANSITION REPORT RELOAD TRANSITION The key parameters parameters for this event are:

"* Core power level

- Cycle Cycle length

- Maximum assembly burnup bumup

- Percentage Percentage of failed fuel

- Meteorology Meteorology This event event is classified as a Postulated Occurrence Occurrence which could involve a release of radioactivity.

radioactivity. It is evaluated evaluated to demonstrate demonstrate the adequacy of the plant design. The principally challenged acceptance criterion for this event is:

1.

1. The radiological consequences must not exceed 10 CFR Part 100 ("Reactor radiological consequences ("Reactor Site Criteria")

Criteria") limits.

This event is not analyzed analyzed with a thermal-hydraulic thermal-hydraulic NSSS transient transient code such as S-RELAP5.

S-RELAP5. The source term is based on the RCS specific specific activity of xenon and krypton isotopes for 24-month 24-month cycle steady state operation operation with 11% % failed fuel. The fuel isotopic inventories for the currentcurrent analysis were calculated calculated for the most limiting assemblies (1.02 assemblies (1.02 power measurement uncertainty measurement uncertainty factor and 1.65 power peaking factor) irradiated to a maximum bumup burnup to 62 GWd/MTU.

GWdlMTU. The meteorology meteorology and failed fuel percentage are not impacted by the fuel design. Thus, the current bounding radiological analysis analysis supports supports the use of of AREVA Advanced CE-14 HTP fuel.

AREV A Advanced 6.3.22 6.3.22 Waste Processing Processing System Incident (UFSAR Section 14.23) 14.23)

Calvert Calvert Cliffs UFSAR SectionSection 5A.2.1.2 5A.2.1.2 identifies the seismic seismic design requirements requirements for various systems in the plant, among them the radioactive liquid waste systems. Safety Guide 29 as referenced referenced in UFSAR UFSAR Section 5A.2.1.2, 5A.2.1.2, was the original licensing licensing basis for the seismic seismic classification of the radioactive liquid waste system at Calvert Calvert Cliffs. Among the systems required by Safety Guide 29 (Section (Section C.l.i) to be designed to Seismic Category Category I standards were: "radioactive "radioactive waste treatment, handling and disposal

systems, systems, except those portions of these systems whose postulatedpostulated simultaneous failure would not result in conservatively conservatively calculated potential potential offsite exposures exposures comparable comparable to the guideline exposures of 10 CFR CFR 100."

100." In the event event of a seismically-induced seismically-induced failure of the non-Seismic non-Seismic Category Category I portions of the Reactor Reactor Coolant Waste Processing System,System, it is hypothesized hypothesized that the contents of those portions portions of the system will be released.

The key parameters for this event are:

- Core power level

- Cycle Cycle length

- Maximum assembly bumup burnup

- Percentage Percentage of failed fuel

- Meteorology Meteorology

- Non-seismic category I waste system components This event event is classified as a Postulated Occurrence Occurrence which could involve a release of radioactivity. It is evaluated evaluated to demonstrate demonstrate the adequacy of the plant design. The principally challenged acceptance acceptance criterion for this event is:

1.

1. The radiological radiological consequences consequences at the site boundary boundary must be considerably considerably below below the 10 CFR Part 100 100 limit of 25 rem whole body (i.e.,

(i.e., below 0.5 rem) originally used as the acceptance criteria for acceptance 75

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT REPORT this calculation (which is consistent consistent with the 0.5 rem site boundary dose limit established established in later later revisions of Regulatory Regulatory Guide 1.29, as well as Regulatory 1.143).

Regulatory Guide 1.143).

This event is not analyzed with a thermal-hydraulic thermal-hydraulic NSSS transient code such as S-RELAP5.S-RELAP5. The source term term is based on the RCS specific activity of iodine, xenon and krypton isotopes for 24-month cycle steady state operation with 11% % failed fuel. The fuel isotopic inventories for the current analysis were calculated for the most limiting assemblies (1.02 (1.02 power measurement measurement uncertainty power uncertainty factor and 1.65 power irradiated to a maximum burnup to 62 GWd/MTU.

peaking factor) irradiated GWdIMTU. The meteorology, non-seismic non-seismic Category I waste system components, and failed fuel percentage Category percentage are not impacted by the fuel design.

Thus, the current current bounding bounding radiological analysis supports the use of AREVA AREVA Advanced CE-14 HTP fuel.

6.3.23 Maximum Hypothetical Hypothetical Accident Accident (UFSAR Section 14.24)

The maximum maximum hypothetical accident for Calvert Cliffs is evaluated evaluated to determine compliance with the siting criteria given in 10 CFR Part 100. In general, the maximum hypothetical accident accident is a non-mechanistic scenario which evaluates evaluates the containment's containment's capability capability to contain released released radioisotopes.

effectiveness is not considered; Safety system effectiveness considered; the quantity of radioisotopes released released to the containment containment atmosphere atmosphere is dependent on the power power level (MWt) of the reactor. The criterion for this release is established established so that the magnitude magnitude of the release bounds all crediblecredible accident accident releases.

The key parameters parameters for the maximum hypothetical hypothetical accident are:

- Core power level

- Fission Fission product source term

- Containment Containment free volume and leakage rate

Containment Containment cooling, spray and filtration system effectiveness effectiveness

- Meteorology Meteorology This event is classified as a Postulated Occurrence which could involve a release of radioactivity. It is Postulated Occurrence evaluated evaluated to demonstrate demonstrate the adequacy of the plant design. The principally principally challenged acceptance acceptance criterion criterion for this event is:

1. The radiological consequences must not exceed the Regulatory radiological consequences 1.183, Table 6 limits.

Regulatory Guide 1.183, This event is not analyzed with a thermal-hydraulic thermal-hydraulic NSSS transient transient code such as S-RELAP5.

S-RELAP5. Additional analysis was performed to demonstrate demonstrate that the AST core source source term used in the maximum hypothetical accident accident radiological analyses remain bounding for AREVA AREV A Advanced CE-14 HTP fuel using Gd 220033 burnable burnable poison irradiated irradiated to a maximum burnup burnup to 62 GWd/MTU.

GWdIMTU. The meteorology, radionuclide removal removal mechanisms, and other inputs to the radiological radiological analysis are not impacted impacted by the change to AREVA AREV A Advanced Advanced CE-14 HTP fuel design. Thus, the current bounding radiological analysis supports the use of AREV AREVA A Advanced CE-14 HTP fuel.

6.3.24 6.3.24 Excessive Excessive CharpinE Charging Event (UFSAR(UFSAR Section 14.25)14.25)

The Excessive Charging event is assumed to occur by inadvertent inadvertent initiation of charging charging flow. The Excessive Excessive Charging event initiated from maximum pressurizer level is performed performed to assure that the operator operator has at least 15 minutes from initiation initiation of a high pressurizer pressurizer level alarm to take corrective corrective action and terminate the event prior to filling the pressurizer pressurizer solid.

76 76

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT The key parameters parameters for this event are:

- operating conditions Initial operating conditions

"* Pressurizer Pressurizer steam volume

" Pressurizer Pressurizer heater and spray availability

- Pressurizer Pressurizer high level alarm setpoint setpoint

Charging flow rate

- Charging flow flow enthalpy

- Single failure assumption assumption (failure (failure in the letdown valve controller)

This event is classified as an AOO which may occur during the life of the plant. The principally principally challenged acceptance challenged acceptance criterion for this event is:

1. An incident of moderate

1. moderate frequency shouldshould not generate a more serious plant condition without other faults occurring independently.

The event behavior is predominantly predominantly a function of plant system specifically the charging and system capability, specifically letdown letdown flow. The plant system characteristics characteristics that would affect the key parameters parameters listed above remain remain unchanged unchanged for the transition transition fuel cycle. The cause of the event and the parameters parameters which control the consequences of the event are unchanged consequences unchanged from or bounded by previous analysis. Therefore, an analysis of the Excessive Charging event is not required for the fuel transition.

6.3.25 Feedline Break Event (UFSAR Section Section 14.26)

Feedwater Line Break The Feedwater Break event is defined as a major break in a MFW line that is sufficiently sufficiently large to prevent maintaining maintaining the SG secondary side water water inventory inventory in the affected affected SG. This event can be be considered as a heatup event, a cooldown considered cooldown event, or a combination of both. There can be an initial, short, .

heatup transient when the feedwater flow stops. This phase is terminated terminated by a reactor reactor trip. This heatup portion of the transient produces produces an RCS response which may result in a challenge to RCS pressure limits. Following the reactor trip, the RCS begins to cooldown cooldown as a result of the heat removal from the affected SG.

The RCS pressure may decrease decrease enough to cause HPSI,HPSI to activate. The cooldown portion of the transient is terminated by dryout of the affected affected SG, which dramatically reduces the heat removal from the RCS.

The lack of MFW results in a long-term long-term heatup similar to the Loss of Feedwater Flow event. Auxiliary feedwater flow is actuated on the AFW actuation signal. The expansion expansion of the reactor coolant coolant and the potential HPSI flow will re-pressurize and refill the RCS. The RCS pressure transient which results in a second peak pressure is limited by the opening of the PSV PSVs. s. This second second peak pressure may produce the maximum RCS pressure. The AFW will eventually eventually restore the inventory in the unaffected unaffected SGs and the decay heat will be removed via steam flow through the MSSVs. As the decay decay heat levels drop, the liquid liquid level in the unaffected unaffected SGs stabilizes and then begins to rise. Also, RCS temperatures temperatures stabilize and then begin to decrease. When the unaffectedunaffected SG levels begin to increase increase and the RCS temperatures begin to decrease, the feedwater line break transient is over.

The key parameters parameters for this event are:

- Break size Break size

Unaffected SG liquid inventory inventory at the time of reactor trip 77

ATTACHMENT (4)

ATTACHMENT RELOAD TRANSITION TRANSITION REPORT

- Trip setpoint(s), uncertainty uncertainty and delay time

- AFW actuation actuation setpoint, minimum minimum flow rate and actuation delay time

" SG blowdown flow rate and isolation time

- operating conditions Initial operating conditions

- Core decay heat assumptions

- RCP heat

"* MSSV setpoints and capacities capacities

- PSV setpoint and capacities capacities

- Technical Specification primary and secondary Technical secondary iodine activity limits

- Primary-to-secondary leak rate Primary-to-secondary

- Meteorology Meteorology This event is classified classified as Postulated Accident, Accident, which is not expected to occur during the lifetimelifetime of the plant, but must be evaluated evaluated to demonstrate demonstrate the adequacy adequacy of the plant design. The principally challenged principally challenged acceptance criteria for this event are:

acceptance

1. The pressures

1. pressures in the reactor coolant and main steam systems should be less than 110% 110% of design values.

2. Although not an SRP criterion, liquid flow through the PSVs or PORVs is not desirable since the PSVs and PORVs may not be qualified for liquid flow. This is demonstrated by showing that the pressurizer pressurizer level does not reach the PORV inlet piping penetrations.

penetrations.

calculated to occur must be sufficiently

3. Any fuel damage calculated sufficiently limited such that the core will remain in in place and intact with no loss of of core core cooling capability. Preclusion of fuel failure is demonstrated by delivering sufficient sufficient AFW to remove core decay decay heat such that there is no significant heatup of of the RCS following reactor reactor trip.

4. Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR Part 100 guidelines.

The AOR AOR for this event event assumes an initial core power level of 2771 MWt, and an MTC bounding the most positive positive MTC allowed at HFP RFP by Technical Specifications was used. These values are not impacted Technical Specifications impacted by the transition to AREVAAREV A Advanced Advanced CE-14 HTP fuel and remain bounding. The event behavior is CE-I4 RTP predominantly predominantly a function of the primary-to-secondary primary-to-secondary heat transfer capability. Therefore, small small perturbations perturbations in parameters parameters such as the core pressure drop, core bypass flow fraction, core inlet flow distribution, and reactivity reactivity feedback do not impact the parameters parameters of interest in assessing the acceptance acceptance criteria. The plant system characteristics characteristics that potentially impact the key parameters parameters listed for this event remain unchanged for both the transition transition fuel cycle, and the full core implementation implementation of AREV AREVA A Advanced CE-14 HTP fuel at Calvert CE-I4 RTP Calvert Cliffs. The cause of the event and the parameters parameters which control control the consequences of the event consequences event are unchanged from or bounded bounded by the current AOR. The input assumptions assumptions for the radiological consequence consequence analysis of the Feedline Break event also remain unaffected by the unaffected transition to AREVA AREV A Advanced CE-14 RTP Advanced CE-I4 HTP fuel. Therefore, an analysis of the Feedline Break event is not required to support the transition AREVA transition to AREV CE-14 RTP A Advanced CE-I4 HTP fuel.

7.0 REFERENCES

REFERENCES 1.

1. EMF-2807(P), Volume Volume II,II, Revision 0, "Calvert "Calvert Cliffs Lead Fuel Assemblies Fuel Design Criteria Criteria Review," September Review," September 2002 78 78

ATTACHMENT (4)

ATTACHMENT TRANSITION REPORT RELOAD TRANSITION

2. BAW-10240(P)(A),

BA "Incorporation of M5 Properties W-1 0240(P)(A), Revision 0, "Incorporation Properties in Framatome ANP Approved Methods, May 2004

3. EMF-92-116(P)(A),

EMF "Generic Mechanical Design Criteria 116(P)( A), Revision 0, "Generic Criteria for PWR Fuel Designs,"

1999 February 1999

4. BAW-10227(P)(A),

BA W-I0227(P)(A), Revision 1, 1, "Evaluation "Evaluation of Advanced Cladding and Structural Material (M5) (M5) in PWR Reactor Fuel," June 2003 5.

5. EMF-96-029(P)(A) Volumes 1 and EMF-96-029(P)(A) 2, "Reactor Analysis System System for PWRs Volume 1 -

Methodology Description, Volume 2 - Benchmarking Results," Siemens Power Corporation, 1997 January 1997

6. XN-75-27(A) and Supplements 11 through 5, 5, "Exxon "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors,"

Reactors," Exxon Nuclear Company, Report and Supplement 1 dated April 1977, Supplement Supplement 2 dated December 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December 1986 (P), and Supplement 5 dated February 1987 (P)

7. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for XN-NF-78-44(NP)(A),

Pressurized Water Reactors, Exxon Nuclear Company,"Company," October 1983 1983

8. XN-NF-82-06(P)(A) Revision 1 and Supplements 2, 4 and 5, XN-NF-82-06(P)(A) 5, "Qualification "Qualification of Exxon Nuclear Fuel for Extended Burnup,"

Burnup," Exxon Nuclear Company, October 1986 1986

9. ANF-88-133(P)(A) and Supplement ANF-88-133(P)(A) Supplement 1, 1, "Qualification "Qualification of Advanced Nuclear Fuels' PWR Design Methodology Methodology for Rod Burnups of 62 MWd/kgU," Advanced Nuclear Fuels Corporation, MWdlkgU,"

December December 1991

10. Not used 11.

11. XN-NF-75-21(P)(A), Revision XN-NF-75-21(P)(A), Revision 2, "XCOBRA-IIIC:

"XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient Core Operation," Operation," Exxon Nuclear Nuclear Company Inc.,

Inc., January 1986 January 1986 12.

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

Configurations," Exxon Exxon Nuclear Company Inc.,

September September 1983 13.

13. EMF-23 I 0(P)(A), Revision EMF-2310(P)(A), Revision 1, "SRP "SRP Chapter Chapter 15 Non-LOCA Non-LOCA Methodology Methodology for Pressurized Pressurized Water Water Reactors," Framatome Reactors," Framatome ANP, May 2004 14.

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

"HTP: Departure Departure from Nucleate Nucleate Boiling Boiling Correlation Correlation for High Thermal Performance Performance Fuel,"

Fuel," Siemens Corporation, January Siemens Power Corporation, January 2005 15.

15. XN-75-32 (P)(A)

(P)(A) Supplements 1, 2, 3, and 4, "Computational Supplements 1,2,3, "Computational Procedure Procedure for Evaluating Evaluating Fuel Fuel Rod Rod Bowing,"

Bowing," Exxon Nuclear Company Nuclear Company Inc., October 1983 1983 16.

16. EMF-1961 (P)(A)

EMF-1961 Revision 0, (P)(A) Revision 0, "Statistical "Statistical Setpoint/Transient Combustion SetpointiTransient Methodology for Combustion Engineering Type Engineering Reactors," Siemens Type Reactors," Siemens Power Corporation, Corporation, July 2000 17.

17. XN-NF-81-58(P)(A),

XN-NF-81-58(P)(A), Revision Revision 2 and and Supplements Supplements I1 and 2, 2, "RODEX2 "RODEX2 FuelFuel Rod Rod Thermal-Thermal-Mechanical Mechanical Response Response Evaluation Evaluation Model,"

Model," Exxon Exxon Nuclear Nuclear Company Company Inc, March 1984 March 1984 18.

18. ANF-81-58(P)(A), Revision 22 and ANF-81-58(P)(A), Revision and Supplements Supplements 3 and 4, "RODEX2 "RODEX2 Fuel Rod Thermal-Thermal-Mechanical Mechanical Response Response Evaluation Evaluation Model,"

Model," Siemens Power Power Corporation, 1990 Corporation, April 1990 19.

19. EMF-2328(P)(A),

EMF-2328(P)(A), Revision Revision 0, 0, "PWR "PWR Small Small Break LOCA LOCA Evaluation Evaluation Model, S-RELAP5 S-RELAP5 Based" Based" March 2001 79 79

ATTACHMENT ATTACHMENT (4)

RELOAD TRANSITION TRANSITION REPORT

20. EMF-2103(P)(A), Revision 0, "Realistic EMF-2103(P)(A), "Realistic Large Methodology for Pressurized Large Break LOCA Methodology Pressurized Water Water Reactors," April 2003 Reactors,"

21.

21. ANF-90-145(P)(A),

ANF-90-145(P)(A), Supplement 1, 1, "RODEX3 Thermal-Mechanical Response "RODEX3 Fuel Thermal-Mechanical Evaluation Response Evaluation Model," Advanced Model," Advanced Nuclear Nuclear Fuels, April 1996 1996

22. ANF-90-145(P)(A),

ANF-90-145(P)(A), "RODEX3 - Fuel Rod Thermal -Mechanical-Mechanical Response Evaluation Model,"

Vol. 1,2, 1, 2, and Supplement Supplement 1, 1, April 1996 1996 23.

23. Not used used

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

25. XN-NF-85-92(P)(A),

XN-NF-85-92(P)(A), Revision 0, "Exxon "Exxon Nuclear Dioxide/Gadolinia Irradiation Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"

Examination Results," September September 1986 1986

26. XN-NF-79-56(P)(A),

XN-NF-79-56(P)(A), Revision I1 and Supplement 1, 1, "Gadolinia "Gadolinia Fuel Properties for LWR Fuel Evaluation," November 1981 Safety Evaluation,"

27. Letter from Mr. D. V. Pickett (NRC) to Mr. J. A. Spina (CCNPP), dated July 22, 2009, "Calvert 22,2009, Cliffs Nuclear Power Plant, Units 1 & 2, Amendment 2, Amendment Re: Measurement Uncertainty Measurement Uncertainty Recapture Power Uprate" Uprate" (TAC NOS. MD9554 AND MD9555)" MD9555)" (Accession ML091820366)

(Accession Number ML091820366) 80 80

ENCLOSURE (3)

AREV AREVA A Proprietary Affidavit Proprietary Affidavit Nuclear Power Plant, LLC Calvert Cliffs Nuclear November 23, 2009 November 23,2009

AFFIDAVIT AFFIDAVIT COMMONWEALTH OF VIRGINIA COMMONWEALTH VIRGINIA )

) 55.

ss.

CITY OF LYNCHBURG LYNCHBURG )

1.

1. My name is Gayle F. Elliott. I am Manager, Product licensing, Licensing, for AREVA NP Inc. and as such I am authorized authorized to execute execute this Affidavit.

2. II am familiar familiar with the criteria criteria applied by AREVA NP to determine whether whether certain certain AREVA NP information is proprietary. II am familiar with the policies policies established by

.-.- AREVA-N P-to-ensure-the- proper-application -of-these-criteria.----

AREVA-NP-to-enstlre-the-proper-application-oHhese-criteria-;---------------, ------------._--

3. I am familiar with the AREVA NP information contained contained in the report ANP-2857(P),

ANP-2857(P), Revision 0, entitled "Loss of Forced ReactorReactor Coolant Coolant Flow Analysis Analysis for Calvert Calvert Cliffs Nuclear Plant, Unit 2," dated September 2009 and referred to herein as "Document."

"Document."

Information contained Information contained in this Document has been been classified classified by AREVA AREVA NP as proprietary proprietary in accordance with the policies accordance policies established established by AREVA AREVA NP for the control and protection protection of proprietary and confidential confidential information.

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

confidential.

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

accordance with 10 CFR 2.390. The information accordance information for which withholding from disclosure disclosure is requested qualifies qualifies under 10 CFR 2.390(a)(4) 2.390(a)(4) "Trade secrets secrets and commercial commercial or financial information."

6. The following criteria are customarily applied applied by AREVA NP to determine determine whether information information should be classified as proprietary:

(a) The information information reveals details of AREVA AREVA NP's research research and development development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly significantly reduce reduce its expenditures, expenditures, in time or resources, to design, produce, or market a similar product or service.

_.__________ ____ .. ___ . ____ ~~)______ . ~_~e infor~ation inc~~~s_~_est~ata or a_na_lyti~al techniques co~:er~~ng~--------.----J (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application application of which results in a advantage for AREVA NP.

competitive advantage (d) The information information reveals certain distinguishing distinguishing aspects aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in productproduct optimization optimization or marketability.

(e)

(e) The information is vital to a competitive advantage advantage held by AREVA NP, would be helpful to competitors to AREVA AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information The information in the Document is considered considered proprietary proprietary for the reasons set forth in in paragraphs 6(b) and 6(c) above.

paragraphs

7. In accordance accordance with AREVA AREVA NP's policies governing protection and control governing the protection of information, proprietary information of information, information contained contained in this Document Document have have been made available, on a limited basis, to others outside AREVA AREVA NP only as required and under suitable agreement nondisclosure and limited use of the information.

providing for nondisclosure

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

9. The foregoing statements are true and correct to the best of my knowledge, foregoing statements information, information, and belief.

'/II't1,_

SUBSCRIBED SUBSCRIBED before me this

-\-

__"___\t' day of September September 2009.

Sherry L.

Sherry L. McFaden McFaden NOTARY COMMONWEALTH OF VIRGINIA NOTARY PUBLIC, COMMONWEALTH VIRGINIA COMMISSION EXPIRES:

MY COMMISSION EXPIRES: 10/3111010/31/10 Reg. # 7079129

_... .-.- ........... -~

SHERRV S-HERRY l.

L.MCFADIN MC-FA0EN t

- Notary PubliC N0101Y Public C:ommonweolth ot Commonwealth VlrQlnla at VRfgOINI

,7079129 7079129 ~

MV Expires oct Commission Expires My Commisson

~

Oct al. 2010

31. 20'0

AFFIDAVIT AFFIDAVIT COMMONWEALTH OF COMMONWEALTH OF VIRGINIA VIRGINIA )

) ss.

5S.

CITY OF LYNCHBURG LYNCHBURG )

1.

1. My name is Gayle F. Elliott. I am Manager, Manager, Product Product Licensing, for AREVA NP NP Inc. and as such I am am authorized authorized to execute this Affidavit.

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

proprietary. I am familiar familiar with the pOlicies policies established established by

.......... AREVANP AREVA-NPtcnmsare-the-proper-application-ofthese--criteria-:--------------

to-ensure-the-prop-er-appiication--of these criteria_ -. . . ------.-----.-

3. I am familiar with the AREVA NP information contained in the report information contained ANP-2834(P), Revision 000, entitled "Calvert Cliffs Nuclear ANP-2834(P), Nuclear Plant Unit 1 Cycle 21 & & Unit 22 Cycle Cycle 19 Realistic Realistic Large Break LOCA SummarySummary Report," dated September 2009 and referred referred to Information contained in this Document herein as "Document." Information Document has been classified by AREVA AREVA proprietary in accordance NP as proprietary accordance with the policies established by AREVA NP for the control and protection of proprietary protection proprietary and confidential confidential information.

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

kind contained

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

accordance accordance with 10 CFR 2.390. The information information for which withholding from disclosure disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determinedetermine information should be classified whether information classified as proprietary:

(a)

(a) information reveals details of AREVA NP's research and development The information development plans and programs or their results.

(b) Use of the information information by a competitor competitor would permit the competitor competitor to significantly significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information information includes test data data or analytical analytical techniques concerning a techniques concerning process, methodology, or component, the application application of which results in a competitive advantage for AREVA NP.

(d) The information information reveals certain distinguishing distinguishing aspects aspects of a process, methodology, or component, the exclusive use of which provides a advantage for AREVA NP in product competitive advantage product optimization optimization or marketability.

(e) information is vital to a competitive The information advantage held by AREVA NP, would competitive advantage be helpful helpful to competitors competitors to AREVA NP, and would would likely cause substantial harm to the competitive position of AREVA AREVA NP.

The information information in the Document is considered proprietary proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

paragraphs

7. In accordance accordance with AREVA governing the protection AREVA NP's policies governing protection and control of information, proprietary proprietary information information contained in this Document have been been made available, on aa limited basis, to others outside AREVA AREVA NP only as required and under suitable agreementagreement nondisclosure and limited use of the information.

providing for nondisclosure

8.

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

on aa need-to-know basis.

9.9. The foregoing The statements are foregoing statements true and are true correct to and correct the best to the of my best of my knowledge, knowledge, information, and information, and belief.

belief.

SUBSCRIBED before SUBSCRIBED before me me this 14 t!1 this ---A.--'-_ _

day of day September 2009.

of September 2009.

Sherry L.

Sherry L. McFaden McFaden

--- t NOTARY PUBLIC, NOTARY COMMONWEALTH OF PUBLIC, COMMONWEALTH OF VIRGINIA VIRGINIA MY COMMISSION EXPIRES:

MY COMMISSION EXPIRES: 10/31/10 10/31/10 Reg.

Reg. ## 7079129 7079129 IMIRRY L.

8"HERRY L. MCSADUt MCfADIN W~ar# PubtI@

NOtary PubltO Commonweath commonwealth

~7079129 of W of VIfQInIa IO 7079'19 COMml..lon Expires My CommIs-Ion My Expire. Oct *'.

Oct 31. 2010 2010

ML093350100

Contents

Text

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

1.0 INTRODUCTION

SUMMARY

SUMMARY

1.1 INTRODUCTION

2.1 INTRODUCTION

SUMMARY

SUMMARY

3.1 INTRODUCTION

SUMMARY

SUMMARY

3.5 CONCLUSION

4.1 INTRODUCTION

SUMMARY

4.6 CONCLUSION

6.1 INTRODUCTION

7.0 REFERENCES

Navigation menu

ML093350100

Text

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

1.0 INTRODUCTION

SUMMARY

SUMMARY

1.1 INTRODUCTION

2.1 INTRODUCTION

SUMMARY

SUMMARY

3.1 INTRODUCTION

SUMMARY

SUMMARY

3.5 CONCLUSION

4.1 INTRODUCTION

SUMMARY

4.6 CONCLUSION

6.1 INTRODUCTION

7.0 REFERENCES

Navigation menu

Search