ML051950127

From kanterella
Jump to navigation Jump to search
WCAP-16461-NP, Rev 0, Ginna Station Extended Power Uprate Supplemental Information.
ML051950127
Person / Time
Site: Ginna Constellation icon.png
Issue date: 07/31/2005
From: Dominicis D, Dunsavage D, Sechrist J
Westinghouse
To:
Office of Nuclear Reactor Regulation
References
WCAP-16461-NP, Rev 0
Download: ML051950127 (31)
v • d • e


Text

Westinghouse Non-Proprietary Class 3 WCAP-16461-NP July 2005 Revision 0 Ginna Station Extended Power Uprate Supplemental Information S Westinghougse

Westinghouse Non-Proprietary Class 3 WCAP-1 6461 -NP Revision 0 Ginna Station Extended Power Uprate Supplemental Information July 2005 Approved: Official record electronically aDDroved In EDMS D. R.Dunsavage, Jr.

Nuclear Services Major Programs Group Approved: Official record electronically aDproved in EDMS D. R Dominicis, Project Manager Nuclear Services Major Programs Group Approved: Official record electronically aDDroved in EDMS J. P.Sechrist, Manager Nuclear Services Major Programs Group Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2005 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 Table of Contents Section Title Page

1.0 INTRODUCTION

....................................... 1 2.8.1.2.2.5 REQUESTS FOR ADDITIONAL INFORMATION (RAls) ....................................... 2 2.8.1.2.3.4 RAls ........................................ 3 2.8.3.

2.3 DESCRIPTION

OF ANALYSES AND EVALUATIONS ...................................... 6 2.8.3.2.4 RESULTS ...................................... 8 List of Tables Table Title Page 2.2.2.1-1 MAXIMUM RCL PIPING STRESS

SUMMARY

............................................................... 11 2.2.2.2-1 STRESS

SUMMARY

AT EPU CONDITIONS ............................................................... 12 2.2.2.3-3 REACTOR VESSEL SUPPORTS FAULTED LOADS COMPARISON (KIPS) ....................... 15 2.2.2.3-4 REACTOR VESSEL SUPPORTS NORMAL/OPERATING LOADS COMPARISON (KIPS).. 15 2.2.2.4-1 CUMULATIVE USAGE FACTORS FOR CRDM JOINTS, APPLICABLE FOR GINNA STATION EPU ............................................................... 16 2.2.2.5.1-1 RCL PRIMARY EQUIPMENT SUPPORT LOADS STRESS MARGIN

SUMMARY

............... 17 2.2.2.6-1 RCP PRESSURE RETAINING COMPONENT STRESSES AND USAGE FACTORS .......... 18 2.2.2.6-2 RCL PRIMARY EQUIPMENT SUPPORT LOADS STRESS MARGIN

SUMMARY

............... 19 2.2.3-1 LOWER INTERNAL CRITICAL COMPONENT STRESSES DUE TO FIV .............................. 20 2.2.3-2 UPPER INTERNAL CRITICAL COMPONENT STRAINS DUE TO FIV ................................. 20 2.2.3-3 REACTOR INTERNAL COMPONENTS STRESSES AND FATIGUE USAGE FACTORS....21 2.2.7-1 STRESS

SUMMARY

OF ANSI B31.1-1973 EQUATIONS 11 THROUGH 14 ........................ 22 2.8.3-4 RTDP UNCERTAINTIES .............................................................. 23 2.8.3-5 DNBR MARGIN

SUMMARY

.............................................................. 24 List of Figures Figure Title Page 2.8.3-1 CROSSFLOW VELOCITY-WESTINGHOUSE 14X14 422V+ 14X14 OFA FUEL ............... 25 2.8.3-2 FUEL AVERAGE TEMPERATURES ......................................................... 26 2.8.3-3 FUEL SURFACE TEMPERATURES ......................................................... 27 2.8.3-4 FUEL CENTERLINE TEMPERATURES ......................................................... 28 I

Westinghouse Non-Proprietary Class 3 1.0 Introduction The purpose of this topical Is to provide information of a proprietary nature to support the License Amendment Request (LAR) by Ginna Nuclear Plant, LLC for the Extended Power Uprate of the Ginna Station. Included in this topical are sections, tables and figures from the LAR that contain Information that is designated Westinghouse Proprietary. Bracketed [ ]a.C information designates data that is Westinghouse Proprietary.

1

Westinghouse Non-Proprietary Cass 3 2.8.1.2.2.5 Requests for Additional Information (RAls)

To facilitate the NRC review, the RAls applicable to mechanical compatibility and performance that were received in prior power uprating submittals are addressed below for the Ginna EPU specifics.

"in Section 7.1 of the Application Report, the licensee states the level of fuel rod fretting, oxidation and hydriding of thimbles and grids, fuel rod growth gap, and guide thimble wear was acceptable. Provide a reference to the document which provides the analytical results, and lists the numerical values for these parameters along with their acceptable limit for the SPU conditions. Also, explain how the analysis performed for IP2 SPU conditions met the applicable regulatory criteria and indicate whether the methodology used has been previously approved by the staff."

This RAI discusses several issues as they apply to fuel rods and to fuel assembly structures. All design criteria have been shown to be met and are documented in proprietary calculation notes and test reports that can be made available for audit.

A series of hydraulic tests and analyses were performed by Westinghouse to confirm fuel assembly vibration and fretting performance. Based on these tests and analyses, the 14x14 422V+ nine-grid design has adequate margin.

The fuel assembly structure formerly had a hydriding pickup limit of [ Jac. This hydride pickup limit was replaced as indicated in the review and approval by the NRC of WCAP-12488-A, Addendum 1-A, January 2002. The upper bound value is [

]a" Maximum grid strap and thimble thinning at the Ginna Station is calculated at EPU conditions to be ]8C, thus the 14x14 422V+ assembly meets this design criterion.

The Westinghouse criteria for fuel rods are [ ]l for clad hydriding, and [ a for clad oxide steady-state interface temperature. All design criteria have been shown to be met and are documented in proprietary calculation notes that are available for audit.

These criteria were approved by NRC in WCAP-12610-P-A, which is applicable to the ZIRLOQm cladding used on the 14x14 422V+ design.

The space between the fuel rod end plugs and the fuel assembly nozzles must be sufficient to prevent interference of these members. All aspects of the 14x1 4 422V+

nine-grid design that affect this requirement are similar to the 14x14 422V+ seven-grid design features currently in the Point Beach and Kewaunee cores and have been shown to be acceptable. These criteria were approved by NRC in WCAP-1261 0-P-A, which is applicable to the ZIRLOm cladding used on the 14x14 422V+ nine-grid design.

2

Westinghouse Non-Proprietary Cass 3 The Westinghouse design bases and criteria for guide thimble wear are that no localized perforation of the tube wall should occur and the integrity of the guide thimble tube should be maintained throughout the normal life of a fuel assembly. These criteria were approved by NRC in WCAP-12610-P-A, which is applicable to the ZIRLOTh guide thimble tube used on the 14x14 422V+ design. The tube wall thickness, material, initial clearances, and thimble bypass flow do not differ significantly between the 14x14 422V+

nine-grid and 14x14 OFA Ginna fuel assembly designs. Thus, no changes are expected in the guide thimble wear performance for the EPU.

"in Section 7.1 of the Application Report, the licensee states that analyses verified the fuel assembly holddown spring's capability to maintain contact between the fuel assembly and the lower core plate at normal operating conditions for the SPU. Describe the analyses performed to justify this statement.

Additionally, provide the numerical values that show the design criteria are met."

The fuel assembly holddown spring analysis was performed on the 14x1 4 422V+ nine-grid assembly using the same standard holddown spring methodology approved in WCAP-12488. The analysis that was completed evaluates the net holddown force on the fuel assembly throughout its design lifetime, taking into account fuel assembly growth and spring relaxation on a cycle-by-cycle basis. The analysis accounts for the opposing forces that act on each fuel assembly due to assembly weight, buoyancy, spring forces, and lift force. The analysis ensures that there is a positive net fuel assembly holddown force on the bottom core plate at all times except during a pump over-speed at hot conditions. During a postulated pump over-speed event, the assembly holddown force acceptance criterion allows assemblies to lift off the lower core plate but not enough to plastically deform the holddown spring during the event. This criterion is satisfied for the 14x14 422V+ nine-grid fuel assembly design under the Ginna Station EPU conditions.

The holddown spring for the 14x1 4 422V+ nine-grid design satisfies all of the standard fuel assembly holddown spring requirements and provides [

holddown during normal operation.

2.8.1.2.3.4 RAls To facilitate the NRC review, the RAls applicable to seismic/LOCA analysis that were received in prior power uprating submittals are addressed below for the Ginna EPU specifics.

"State whether the core is being treated as a mixed core during the transition cycles. Also, explain how fuel damage was analyzed in a seismic event for the mixed core as it transitions to a homogeneous 15x1 5 Upgraded fuel loading and describe the worst case scenario analyzed. In addition, provide the technical justification that shows structural integrity at the SPU condition for the mixed core 3.

Westinghouse Non-Proprietary Cass 3 is maintained In a loss-of-coolant accident (LOCA) coincident with a seismic event at IP2."

The licensing basis for fuel structural integrity requires that the loading conditions address seismic loading, LOCA loading, and the combination of LOCA and seismic loading as required by the NRC. The seismic and LOCA analysis of the reactor pressure vessel system was performed for the EPU conditions, including the generation of the core plate seismic motions that were used in the Ginna Station analysis of 14x14 422V+

nine-grid and 14x1 4 OFA 9-grid fuel assembly designs. The LOCA analysis used LOCA hydraulic forcing functions calculated using the MULTIFLEX computer code and crediting leak-before-break (LBB) for the reactor c6olant loop piping.

Detailed site-specific fuel assembly analyses for Ginna Station have been performed under EPU conditions in accordance with approved methodologies. These methodologies were approved by NRC in WCAP 9401-P-A (Reference 5), WCAP-9500-A (Reference 4), WCAP-1261 0-P-A (Reference 6), and WCAP-1 2488-A (Reference 1).

Results from these analyses demonstrate that for the limiting-loading condition (combined seismic and LOCA loading), the fuel assembly structural integrity is maintained and the grid impact loads and component stresses remain below the allowable limits. Therefore, the requirements to maintain a coolable core geometry are met. These analyses were performed for homogenous cores of 14x14 422V+ nine-grid fuel and transition cores with both 14x1 4 422V+ nine-grid fuel and 14x1 4 OFA nine-grid (current resident) fuel. The transition core analyses considered various fuel assembly loading combinations to determine the limiting conditions. The transition core-loading pattern that is limiting for the upgrade fuel occurs when the 14x14 422V+ nine-grid fuel is located at [ ]a-c and the 14x14 OFA nine-grid fuel is located at [ ]3c. The transition core loading pattern that is limiting for the 14x14 OFA nine-grid fuel occurs when the 14x14 OFA nine-grid fuel is located at [ ]aC and the 14x14 422V+

nine-grid fuel is located at [ J4C In both limiting cases, significant margins remain for both the 14x14 422V+ nine-grid and 14x14 OFA nine-grid fuel assemblies, considering combined seismic and LOCA loading.

The maximum calculated load for the combined seismic and LOCA loads was compared to the maximum load that can be applied before plastic deformation occurs in the subject grid (called the allowable limit in the analysis). In all cases the postulated load was well below the allowable limit. The closest ratio of combined seismic and LOCA loading to limit load is [ r]c. For thimble tubes and fuel rods, there is no case for which the strength of the thimble tubes and fuel rods is not at least [ I`C the calculated loading for the combined seismic- and LOCA-loading condition. Because none of the fuel assembly components will experience loading at or above their strength limit, the fuel assembly geometry is maintained for this limiting loading combination and the coolable geometry conclusions of the LOCA ECCS analyses are not affected. The approval of the 4

Westinghouse Non-Proprietary Cass 3 methodology is discussed Inthe following RAI. The mixed core configuration resulted in the limiting loads for all loading conditions and had significant margin.

"In the Fuel Criterion Evaluation Process (FCEP) Notification of the 15x15 Upgrade Designs submitted by Westinghouse Electric Company to the NRC on February 6, 2004, Westinghouse states that evaluations of the 15x15 Upgraded fuel assembly design for seismic and LOCA loading at IP2 have been performed in accordance with the 'Reference Core Report 17x17 Optimized Fuel Assembly" methodology. Provide the technical justification showing that the 17x17 design/method referenced is applicable to the 15x1 5fuel design."

In Section 3.0, Category B, Item e, "Fuel Assembly Structural response to Seismic/LOCA Loads" of the FCEP, notification to the NRC regarding the 14x14 422V+

nine-grid design, Westinghouse states: "Evaluations of the revised 14x14 422V+ design for seismic and LOCA loading has been performed in accordance with approved methodologies(3)."

The indicated Reference 3 cites:

Reference 3. Davidson, S. L and lorii, J. A. (Eds), et al., Reference Core Report 17x17 Optimized Fuel Assembly, WCAP-9500-A, May 1982; Beaumont, M. D.

and Skaritka, J. (Eds.), et al., Verification testing and Analysis of the 17x17 Optimized Fuel Assembly, WCAP-9401-P-A March 1979; and Davidson, S. L, and loni, J.A (Eds.), et al., Supplemental Acceptance Information for NRC Approved Version of WCAP-9401/9402 and WCAP-9500, February 1983.

The references cited were approved by the NRC for the intended application in WCAP-12488-P-A (Reference 1). On page 5.3 of the SER)TER under 5.4 "Fuel Assembly Structural Damage from External Forces Evaluation," it states: "Generic analysis methods for performing combined LOCA-seismic loading analysis have been described by W in WCAP 9401-P-A (and WCAP-9402-A). These analysis methods not only include the fuel assembly structural response, but also fuel rod cladding loads.

These methods have been approved by the NRC and therefore, PNL concludes they remain acceptable for application to W fuel design changes."

In the SER for WCAP-9500-A (Reference 4) and WCAP-9401 -P-A (Reference 5), the NRC discusses the generic analysis methodology used to evaluate the 17x17 OFA. The methodology essentially consisted of four mathematical models: a system model, a detailed core model, a lateral fuel assembly model, and an axial fuel assembly model.

Details of the methodology are described in WCAP-9401-P-A.

In the NRC's SER approval for WCAP-9500-A, the following statement was made:

5

Westinghouse Non-Proprietary Class 3 "The methodology described applies not only to three- and four-loop 17x17 plants but generically for plants having other standard arrays (e.g., 14x14, 15x15 and 16x1 6)."

This methodology was captured in Chapter 18 of WCAP-9500-A (Reference 4), and included seismic and LOCA loads. The methodology was further described in WCAP-9401-P-A (Reference 5). For each fuel transition, the "new' design was compared to the previous design. For the analysis of the combined seismic and LOCA loads, there has been no change that would invalidate the original methodology that was shown and stated to be applicable to all Westinghouse fuel arrays.

WCAP-12488-A (Reference 1), Fuel Criteria Evaluation Process, is not limited to any specific fuel design or geometry, and has been in use since March 1993 based on the NRC approval of this methodology for evaluating Westinghouse fuel changes.

Westinghouse has followed the methodology described and approved for seismic and LOCA analysis. While the methodology used is the same as that referenced in WCAP-9500IWCAP-9401, separate calculations and evaluations were conducted for Ginna Station based on EPU conditions.

2.8.3.2.3 Description of Analyses and Evaluations For the fuel transition and EPU analysis, the design limit DNBR values for the 14x14 422V+ fuel is 1.24 for typical and thimble cells. After accounting for the plant-specific margin, the SAL DNBR for the 422V+ fuel is 1.38/1.38 (typicalthimble). These SALs are employed in the DNB analyses.

With the SAL DNBR set, the core limit lines, axial offset limit lines, and dropped rod limit lines are generated. Based on these limit lines, the maximum FNAH limit that can be supported is 1.72 for the 422V+ fuel. This limit incorporates all applicable uncertainties, including a measurement uncertainty of 4 percent (Reference 6), and is adjusted for power level using the equation:

FNAH = 1.72 x [1 + 0.3(1-P)J where P is the fraction of full power.

Rod bow can occur between mid-grids, reducing the spacing between adjacent fuel rods and reducing the margin to DNB. Rod bow must be accounted for in the DNBR safety analysis of Condition I and Condition II events.. Westinghouse has conducted tests to determine the impact of rod bow on DNB performance; the testing and subsequent analyses were documented in Reference 7.

Currently, the maximum rod bow penalty for the OFA fuel assembly is [

]ax at an assembly average burnup of 24,000 MWD/MTU (References 7 and 8).

No additional rod bow penalty is required for burnups greater than 24,000 MWD/MTU 6

Westinghouse Non-Proprietary Cass 3 since credit is taken for the effect of FNAH burndown due to the decrease In fissionable isotopes and the buildup of fission products (Reference 9). Based on the testing and analyses of various fuel array designs documented in Reference 7, including the 14x14 STANDARD assembly, the 14x14 OFA and the 14x14 422V+ fuel assemblies should have the same rod bow penalty applied to the analysis basis as that used for 14x14 STANDARD fuel assemblies.

Two different bypass flow rates are used in the thermal-hydraulic design analysis. The thermal design bypass flow (TDBF) is the conservatively high core bypass flow used with the thermal design flow (TDF) in power capability analyses that use standard (non-statistical) methods, and is also used to calculate fuel assembly pressure drops. The best estimate bypass flow (BEBF) is the core bypass flow that would be expected using nominal values for dimensions and operating parameters that affect bypass flow without applying uncertainty factors. The BEBF is used in conjunction with the vessel minimum measured flow (MMF) for power capability analyses using the ITDP or RTDP (statistical) design procedures. The BEBF is also used to calculate fuel assembly lift forces.

Flow redistribution occurs between adjacent fuel assemblies with different hydraulic resistances, resulting in a net reduction in flow in the higher-resistance assemblies.

Crossf low can also be induced by local hydraulic resistance differences, such as differences in grid elevations and resistances. The flow redistribution affects both mass velocity and enthalpy distribution, which, in turn, affect DNB. The design procedure establishes a transition core DNBR penalty due to flow redistribution, and all further plant-specific analysis proceeds as if it were a full core analysis.

Excessive crossflow is prevented in the EPU transition cores by maintaining grid-to-grid overlap between the 14x14 OFA design and the 14x14 422V+ design, and by ensuring that the difference in grid loss coefficients between the two designs is within previous Westinghouse design experience. The only exception is the variation in the top-grid centerline elevation, which is approximately 3 inches higher in the 422V+ fuel assemblies (see Section 2.8.1 Fuel System Design), and has been explicitly considered in the axial flow distributions assumed in the thermal-hydraulic analysis. A comparison of fuel assembly crossflow velocities due to grid pressure drop mismatch between the 14x14 Westinghouse 422V+ fuel assembly design and the 14x14 OFA design was performed.

Full-scale hydraulic tests were performed on the 14x14 Westinghouse 422V+ fuel assembly design to confirm the pressure loss compatibility with the 14x14 OFA fuel design. The 14x14 Westinghouse 422V+ fuel assembly design pressure drop is approximately [ JSC lower than the 14x14 OFA design due principally to differences in the grid designs (see Section 2.8.1 Fuel System Design).

7

Westinghouse Non-Proprietary aass 3 Transition cores were analyzed as if they were full cores 6f one assembly type (full 14x1 4 422V+ and full 14x1 4 OFA), applying the applicable transition core penalty as a function of the number of each fuel assembly type in the core using the NRC-approved methodology detailed In Reference 11 and approved in Reference 10.

The thermal-hydraulic analysis was performed using VIPRE-01 with the RTDP methods described above, including fuel rod bow, bypass flow and flow redistribution effects on a full core of 14x14 422V+ fuel. The results of this analysis are presented in Section 2.8.3.2.4.

The introduction of lower resistance assemblies will influence the Ift forces on the remaining assemblies. While the flow redistribution tends to reduce the flow in the higher resistance assemblies, the lower resistance 422V+ fuel assemblies will have a higher average flow than they would in a full core situation. Thus, the lift force on these 422V+ assemblies will be higher during the transition cores, and will be greatest during the first transition cycle since this will have the highest number of co-resident 14x14 OFA fuel assemblies.

Fuel temperatures and associated rod internal pressures have been generated using the NRC-approved PAD code (Reference 12) for the 422V+ fuel. The characteristics of the 14x14 OFA and 14x14 422V+ designs are very similar except for the rod diameter. The 14x14 422V+ design also includes a larger rod plenum for gas accommodation. The performance criteria employed by Westinghouse for the 14x14 OFA and 14x14 422V+

designs are the same throughout life, in terms of fuel temperatures, rod internal pressures, and core stored energy.

The fuel rod average and surface temperatures are needed for the accident analyses. In addition, minimum fuel average and fuel surface temperatures are required by Non-LOCA Analysis. Fuel centerline temperatures were also generated for the 422V+ fuel.

These will be used for future verification, during reload design validation, that fuel melt will not occur.

In addition to the fuel temperatures and pressures, the revised core stored energy for the 422V+ fuel has been determined for use in containment analysis (refer to Section 2.6).

Core stored energy Is defined as the amount of energy In the fuel rods in the core above the local coolant temperature. The local core stored energy is normalized to the local linear power level.

2.8.3.2.4 Results Table 2.8.3-5 summarizes the available DNBR margin for Ginna power uprate. It should be noted that the DNBR margin summaries are cycle-dependent and may vary from 8

Westinghouse Non-Proprietary aass 3 cycle-to-cycle in future reload designs. The continued satisfaction of the DNBR criterion for reload cycles is confirmed via the approved reload methodology of WCAP-9273-NP-A (Reference 14).

For the Ginna analyses, the maximum permissible TDBF is [ Ia.c percent and the maximum permissible BEBF is [ ` percent.

I The 14x14 422V+ design allows power uprating at an FNAH limit of 1.72. All the thermal-hydraulic design criteria are satisfied for the Ginna EPU fuel transition. The anticipated reduction in margin that would result from the increased power level has been offset by the following margin contributors:

  • Larger fuel rod diameter (i.e. lower heat flux) for 422V+ fuel design,
  • The use of the advanced setpoint methodology (Reference 2) for 422V+

and OFA fuel designs, and

  • The use of a lower FaH for the OFA fuel.

The uprate analysis demonstrates that the combined DNBR margin gain is enough to accommodate the extended power uprate to 1775 MWt.

The hydraulic compatibility of the 14x14 422V+ and 14x14 OFA fuel assemblies has been addressed and found to be acceptable. The difference in loss coefficients between the two designs and the respective grid locations of the two designs has been analyzed to demonstrate that crossflow-induced vibration will not result in fretting. The expected 9

Westinghouse Non-Proprietary Class 3 fuel assembly crossf low is well within Westinghouse experience with transition cores with intermediate flow mixing (IFM) vane grids. The crossflow velocity profile versus height for the Ginna fuel transition is shown in Figure 2.8.3-1.

The maximum kW/ft limit for fuel melt is [ ]C for the 14x14 422V+ fuel.

Fuel temperatures were generated for the 14x1 4 422V+ fuel for use in the safety analyses. Figure 2.8.3-2 provides representative data (based on non-IFBA fuel) for the maximum and minimum fuel rod average temperatures. Figure 2.8.3-3 provides the fuel surface temperatures corresponding to the maximum and minimum fuel average temperatures in Figure 2.8.3-2. Figure 2.8.3-4 provides the maximum and minimum fuel centerline temperatures.

Fuel rod internal pressure is important in assessing the degree of burst and blockage which may occur after a loss-of-coolant accident. Pressures are computed with the PAD codes (Reference 12). Fuel parameters for reload fuel are evaluated to confirm that the pressures used in the reference analysis remain applicable to the reload.

The core stored energy for the 14x14 422V+ fuel is 5.25 Full Power Seconds (FPS).

The evaluations demonstrate that the minimum DNBR for the static misaligned rod event is above the SAL DNBR. In addition, the maximum calculated linear heat rate for the static misaligned rod event was less than the fuel centerline melt limit at uprate conditions. Therefore, the peak fuel centerline melt temperature criterion is confirmed to be met.

The SAL DNBR is met for the HZP Large Steam Line Break, including the effects of the worst stuck rod, based on the power distributions from the reference loading plan. A confirmational DNBR calculation will be performed as part of each reload design in accordance with the WCAP-9273-NP-A reload methodology (Reference 14).

The thermal-hydraulic evaluation of the fuel upgrade for Ginna has shown that 14x14 OFA and 14x14 422V+ fuel assemblies are hydraulically compatible and that the DNB margin gained through use of the upgraded fuel is sufficient to allow an increase in the power rating to 1775 MWt. Sufficient DNBR margin in the SAL DNBR exists to cover any rod bow and transition core effects. All current thermal-hydraulic design criteria are satisfied.

10

Westinghouse Non-Proprietary Class 3 Table 2.2.2.1-1 Maximum RCL Piping Stress Summary (from Reference 4) (Based on KAverge )

ANSI B31.1 Actual Piping Code Percentage Stress For Allowable of ANSI B31.1 Code Equation RCL Piping EPU (ksi) Stress (ksi) Allowable Normal - Equation 11 Hot Leg a.c 16.8 la [ c Design Pressure + Deadweight Crossover Leg [ jac16.8 a.c Cold Leg [ lac 16.8 [ ]a.c Upset - Equation 12 Hot Leg [ ]2.C 20.1 [ ]8C Design Pressure + Deadweight + Crossover Leg [ ac 20.1 8 j8.C OBE Cold Leg [ la 20.1 [ a 1 Emergency - Equation 12 Hot Leg [ ac 30.2 [ c Design Pressure + Deadweight + Crossover Leg ] 9c 30.2 [ la c SSE _

Cold Leg I ] aC30.2 l Faulted - Equation 12 Hot Leg [ ac - 40.3 [ ]ac Design Pressure + Deadweight + Crossover Leg [ a 40.3 E 1 ac

[(SSE)2 + (DBPB*)2] 1/2 Cold Leg ( ]a. 40.3 [ ja~c Maximum Thermal - Equation 13 Hot Leg [ ]3.C 27.5 [

Maximum Thermal Stress Crossover Leg [ ]a*c 27.5 [ ]8.c Range* + Seismic Anchor Motion OBE Displacements Cold Leg [ l 27.5 [ 1 a.c Normal + Maximum Thermal - Hot Leg [ ].C 44.4 [ ]8.C Equation 14 Design Pressure + Deadweight + Crossover Leg [ ]8.c 44.4 [ pac Maximum Thermal Stress Cold Leg ]c 44.4 ac Range* + Seismic Anchor C Motion OBE Displacements .

Note:

  • DBPB = Design-Basis Pipe Break.

Loss-of-load overtemperature transient effects are included (Reference 4).

Kaverage is the average of the support stiffnesses.

r.

11

Westinghouse Non-Proprietary Cass 3 Table 2.2.2.2-1 Stress Summary at EPU Conditions Piping Analysis Loading Existing EPU Allowable Design Description Condition Stress Stress Stress Margin (psi) (Psi) (psi)

(Note 1)

Main Steam Inside Equation 12U [ ]a* j ]8.C 16,444 [ ].c Containment Loop A (Occasional)

Equation 12F [ ]a c [ ]a c 24,665 [ la c (Occasional)

Main Steam Inside Equation 12U [ 1a Jac 16,444 [ a8c Containment Loop B (Occasional)

Equation 12F [ ]a.c [ ]aC 24,665 [ 14 (Occasional)

Main Steam Outside Equation 12U la [ JaC 16,444 [ ]ac Containment (Occasional)

(Containment Equation 12U [ ]a.c [ ]a~c 18,000 [ 8a.C Penetrations 401 & 402 (Occasional) to Anchor MSU-35)

(3" Bypass)

Equation 12F [ ]ac [ ]a 24,665 [ ]a.c (Occasional)

Equation 12F [ ]axc [ jac 27,000 a l.c (Occasional)

(3" Bypass)

Main Steam Outside Equation 12U ( ]a c [ ]a C 16,440 [ ax Containment (Occasional)

(Anchor MSU-35 to Turbine Stop Valve Nozzle Connections) 12

Westinghouse Non-Proprietary Class 3 Table 2.2.2.2-1 Stress Summary at EPU Conditions Piping Analysis Loading Existing EPU Allowable Design Description Condition Stress Stress Stress Margin (psi) (psi) (psi)

(Note 1)

Feedwater Inside Equation 12U [ ]ac [ ]a~c 21,000 [ ]a Containment Loop A (Occasional)

Equation 12F I Ja.c [ ]ac 31,500 [ ]ac (Occasional)

Feedwater Inside Equation 12U [ ] [ ]ac 21,000 [ ]c Containment Loop B (Occasional)

Equation 12F [ ]ax [ ]* 31,500 [ ]ac (Occasional)

Feedwater Outside Equation 12U [ .c [t ac 21,000 ( ].C Containment (Occasional)

(Anchor FWU-28 to Equation 12F [ JC [ ac 31,500 [ a~c Containment Penetration (Occasional) 404)

Feedwater Outside Equation 12U ]cax ]ax* 21,000 [ ]a.c Containment (Occasional)

(Anchor FWU-28 to Equation 12F [ ]ac [ Ja.C 31,500 I Containment Penetration (Occasional) 403)

Feedwater Outside Equation 12U [ ]a [ ]a 21,000 [ ]a8C Containment (Occasional)

(Feedwater Pumps Discharge to Anchor FWU-28)

Condensate (Heater 2A Equation 13 ( ]a.c ]a.c 22,500 [ ]ac to Heater 3A) (Thermal) 13

Westinghouse Non-Proprietary aass 3 Table 2.2.2.2-1 Stress Summary at EPU Conditions Piping Analysis . Loading Existing EPU Allowable Design Description Condition Stress Stress Stress Margin (psi) (psi) (psi)

(Note 1)

Condensate (Heaters Equation 13 [ ]a~c[ ] 22,500 [ ac 3A/B to Heaters 4AIB) (Thermal)

Extraction Steam to Equation 13 [ a.C ac 22,500 ]a.c Feedwater Heaters 4AB (Thermal)

Feedwater Heater Drain Equation 14 [ [ l ,*

aac 37,500 l ac Piping From Heaters 5A (Thermal +

to 4A Sustained)

Moisture Pre-separator Equation 13 [ a.c Jac 22,500 [ ]a.c Drains (Thermal)

Moisture Separator Equation 13 [ Jaxc ]a.c 22,500 [ JaC Drains (normal drains) (Thermal)

NOTES:

(1) Design Margin reported is based on the ratio of EPU stress divided by the Allowable stress.

14

Westinghouse Non-Proprietary Class 3 Table 2.2.2.3-3 Reactor Vessel Supports Faulted Loads Comparison (kips)

Load Gilbert Associates Support Load Cases Snubber EPU Direction Case 1 Case 2 Case 3 Reduction Support Program Loads Support Loads (1)

_ _ _ _ _ _ _ __ _ _ _ _ _ _ _ ___ _ __ __ _ _ __ _ _ _ _( 1)

Horizontal [ ]ac ]a.c ].c a [ ]a.c [ 8,C Vertical l[ a.c [ ]a c .c a.c 8 [ ]a c Notes:

1. The support loads from the Snubber Reduction and EPU Programs are considered acceptable because they are less than at least one of the Load Cases 1, 2, or 3.

Table 2.2.2.3-4 Reactor Vessel Supports Normal/Operating Loads Comparison (kips)

Load Gilbert Associates Support Load Case Snubber EPU Direction Reduction Support Program Loads Support Loads Horizontal [ ] [ J.C f ].C Vertical [ TIC [ ]axc [ ]ac Notes:

[1 .

_ . ,. _' ,, . ]a,c 15

WesUnghouse Non-Proprietary Class 3 2.2.2.4-1 Cumulative Fatigue Usage Factors for CRDM Joints, Applicable for Ginna Station EPU CRDM Joint and Component Cumulative Usage Factor Values Applicable for Allowable Value the EPU Program

  • Upper- Joint Canopy 3.C 1.00 Upper Joint Canopy Weld a 1.00 Upper Joint Threaded Area [ 1.00 Middle Joint Canopy Weld [ ]C 1.00 Lower Joint Canopy Weld [ ]a 1.00
  • - The values from the generic analysis performed for model L-1 06 CRDM remain applicable and bounding for the Ginna EPU program.

16

Westinghouse Non-Proprietary Cass 3 Table 2.2.2.5.1-1 RCL Primary Equipment Support Loads Stress Margin Summary (Stress Margin = Allowable/Actual) (Based on Kaw.ge) (See Note 3)

Service Normal Upset Emergency SSE Faulted Level (See Note 1)

Deadweight +

Deadweight + Deadweight Deadweight Thermal Load Deadweight Thermal (Normal + + Thermal + Thermal Normal +

Combinati + Thermal Overtemperature) Normal + Normal + [(SSE) 2 on Normal + OBE DBPB SSE +(PIBK 2)]" 2 Steam Generator Upper (See Note 2) [ ]aC [ jaCx Iac [ rac Supports Bumpers Steam Generator Upper (See Note 2) [ ]a ca[ ] axc j ]ac Supports Snubbers Steam Generator Lower (See Note 2) l ]ac[ ]ac ]8.C [ J8C Supports Lateral Steam Generator Lower [ ]axc [ ]axc ja.c a [ ja c [ ]axc Supports Columns Notes

1. PIBK (pipe break) includes DBPB (design basis pipe break) and main steam/feedwater breaks.
2. Under normal conditions no significant loads are imposed on theses lateral support elements.
3. Kaverage is the average of the steam generator upper support stiffness.

17

Westinghouse Non-Proprietary Class 3 Table 2.2.2.6-1 RCP Pressure Retaining Component Stresses and Usage Factors Component EPU Stresses and Usage Allowable Comments Factors Casing Primary [ 16,700 psi The original calculations had a Membrane Stress ]ac (Sm) zero usage factor since the Intensity calculated stresses were below Primay + 5,00 psithe lowest Sa value on the Primary + [ 25,000 psi fatigue curve. This isstill true for Secondary a'c (1.5 Sm) the Ginna Station EPU Pressure and -h. .

Mechanical Loads conditions.

Maximum Steady [ 50,100 psi Original stress intensities were State Thermal + a'c (3 Sm) 16,359 psi, 22,924 psi, and Pressure and 41,898 psi, respectively.

Mechanical Stresses Main Flange General Primary [ 20,000 psi No changes from the original Membrane Stress (S,) calculation.

Intensity Local Primary 30,000 psi Membrane Stress Iac (1.5 Sm)

Intensity Primary + [ 60,000 psi Secondary Stress Ia (3 S.)

Intensity Usage Factor ]aC 1.0 Main Flange Maximum Service [ 55,800 psi No changes from the original Studs Stress, Averaged ac (2 Sm) calculation. Allowable stresses Across Section are based on SA-193, Grade B7 material. Current drawings show Maximum Service [ 83,700 psi the stud material is SA-540, Stress at (3 Sm) Grade B24, Class 4, or Periphery of Cross Grade B23, Class 4. Both of Section from these have higher allowable Tension + Bending stresses than the SA-1 93, Grade Usage Factor ]a c 1.0 B7 material originally considered.

18

Westinghouse Non-Proprietary aass 3 Table 2.2.2.6-2 RCL Primary Equipment Support Loads Stress Margin Summary (Stress Margin = Allowable/Actual) (Based on Kverage)

Service Level Normal Upset Emergency SSE Faulted

._ (See Note 1)

Deadweight +

Deadweight + Deadweight + Thermal Deadweight Thermal (Normal Deadweight + Thermal Normal +

Load + Thermal +Overtemperature) Thermal N ormal + [(SSE) 2 Combination Normal + OBE Normal + DBPB 5SE +(PIBK)]"2 Reactor Coolant See Note 2 [ ].C [ Ja.C [ ]a.c [ ]awc Pump Lateral Tierods Reactor Coolant [ ]ac [ a.c ac[ ]ac[ ]ac a,[

Pump Columns .

Note:

1. PIBK (pipe break) includes DBPB (design basis pipe break).
2. Under normal conditions no significant loads are imposed on these lateral support elements.

19

Westinghouse Non-Proprietary Cass 3 Table 2.2.3-1 Lower Internal Critical Component Stresses Due to FIV ASME Code Endurance Limit")

Maximum Alternating Stress (high-cycle fatigue)

Component psi psi Top Support Bolts [ ]aC 23,700 Flexures [ ]8.C 23,700

1. Basis is ASME Code section NB-3222 and Figure 1-9.2.2, Curve A and Table 1-9.2.2.

Table 2.2.3-2 Upper Internal Critical Component Strains Due to FIV

1. Basis of acceptance is from

-N 20

Westinghouse Non-Proprietary aass 3 Table 2.2.3-3 Reactor Internal Components Stresses and Fatigue Usage Factors Allowable Stress Intensity (ksl) S.,.

Component S.I. = (Pm + Pb+ 0) (3 Sm) ksi Fatigue Usage Upper Support Plate [ 49.2 [ ]ac Deep Beam Structure Jac 49.2 Upper Core Plate [ ]. 48.6 [

Upper Core Plate Alignment [ c 3 4 .4 4 (b) [ a~c Pins Upper Support Columns [ ]a.c 49.2 1 ]c Lower Support Plate ac 49.2 [ ja.c Lower Core Plate [ ]ac 49.2 [ ]ac Lower Support Columns [ '49.2 ( ]atc Core Barrel Assembly:

Upper Girth Weld [ 49.2 [ ]C Lower Girth Weld [ 49.2 ]a.c Outlet Nozzle [ a 34.44 ) . ]ac Thermal Shield & Flexures Thermal Shield Flexures [ 49.2 a[

]8C Thermal Shield Flexure Bolts [ ,C 57.0 [ ]axc Radial Keys and Clevis Insert Assembly

  • Lower Radial Inserts [ 69.0 ]a~c Notes:

[a.

b. Allowable based on weld quality factor 21

Westinghouse Non-Proprietary Class 3 Table 2.2.7-1 Stress Summary of ANSI B31.1-1973 Equations 11 through 14 Stress Allowable Stress Ratio Equation No. (psi) (psi) (Actual/Allowable) 11 (Design) [ 3 15,900 [ a~c 12 (Upset) [ ]C 19,080 [ ]S.c 12 (Emergency) [ ]ac 28,620 [ ]a~c 13 (Normal) [ ]ac 27,350 [ a]c 13 (Upset) [ ]S. 27,350 [.

13 (Emergency) f ]* 27,350 [ ]a.

13 (Faulted) [ ]ac 27,350 [ ]ac 14 (Normal) ]ac 43,250 [ a.c 14 (Upset) - 43,250 [ Ja~c 14 (Emergency) [ ]ac 43,250 [ ]a.C 14 (Faulted) [ 43,250 ( ]a.c 22

Westinghouse Non-Proprietary Class 3 Table 2.8.3-4 RTDP Uncertainties Parameter Uncertainty Used in EPU Safety Analysis Power I

]ax.

Reactor Coolant System Flow I Ia,c Pressure

]a.c Inlet Temperature I

]ac 23

Westnghouse Non-Proprietary Cass 3 Table 2.8.3-5 DNBR Margin Summaryvi) 14x14 OFA 14x14 422V+

DNB Correlation [ ]ac [ a DNBR Correlation Limit l ]a ja.c DNBR Design Limit (TYP) 2 ja~c (THM)(3) [ ]ac[ Ja~c DNBR SAL(6 ) (TYP) [ ]ac [ 8a'c (THM) [ jac [ ]axc DNBR Retained Margins4 ) (TYP) [ c a'c (THM) [ ]a.c [ ]'c Rod Bow DNBR Penalty ]a.c [ ]c Transition Core DNBR Penalty [ ja.c l a.c Available DNBR Margin 51) (TYP) [ c[ ]a'c

. (THM) [ ]_[c ]aC Notes:

1. Steam line break is analyzed using the W-3 correlation with STDP. The correlation limit DNBR is 1.45 in the range of 500 to 1000 psia. Rod withdrawal from subcritical is also analyzed using the W-3 correlation (w/o spacer factor) with STDP below the bottom non-mixing vane grid. The correlation limit DNBR is 1.30 above 1000 psia and the SAL DNBR is 1.447 (422V+) which provides [ Ja4c to cover the rod bow penalty and retain generic margin for operational issues. WRB-1 with RTDP is used for rod withdrawal from subcritical above the bottom non-mixing vane grid.
2. TYP = Typical Cell
3. THM = Thimble Cell
4. DNBR margin is the margin that exists between the SAL and the design limit DNBRs.
5. The margin summary for OFA corresponds to the first transition cycle. For the second transition cycle, the OFA DNBR transition penalty will increase; however, this will be offset by the FAH reduction.
6. The SAL DNBR was changed from the current value of [

]a~c in order to support the proposed Over Temperature AT (OTAT) trip setpoint revisions. Sufficient DNBR margin has been retained to offset rod bow and transition core effects.

24

Westinghouse Non-Proprietary Class 3 a, c Crossflow velocity sign convention Is as follows:

Plus (+) sign Indicates flow Into the 14x14 OFA fuel assembly and minus (-) sign Indicates flow Into the 14x14 422V+ fuel assembly 2 Figure 2.8.3-1. Crossflow Velodty - Westinghouse 14x14 422V+ 14x14 OFA Fuel 2 Scc Section 2.8.1 Fuel System Design for discussion of testing and basis for crossbow determination.

25

Westinghouse Non-Proprietary Cass 3

-1 a,c Figure 2.8.3-2. Fuel Average Temperatures 26

Westinghouse Non-Proprietary aass 3 K a, c Figure 2.8.3-3. Fuel Surface Temperatures 3 3 The labels for Minimum Fuel Surface Temperature and Maximum Fuel Surface Temperature correspond to the Maximum and Minimum Fuel Average Temperatures in Figure 2.8.3-2.

27

Westinghouse Non-Proprietary Ciass 3 a, c Figure 2.8.3-4. Fuel Centerline Temperatures 28

Retrieved from "https://kanterella.com/w/index.php?title=ML051950127&oldid=2592034"