ML16256A211

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Revision 309 to Final Safety Analysis Report, Chapter 4, Reactor, Appendix 4.3A - Fuel Cycle 21
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WSES-FSAR-UNIT-3 4.3A-1 Revision 309 (06/16)

APPENDIX 4.3A (DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302, EC-13881, R304; EC-30663, R307, LBDCR 1 4-008, R308, LBDCR 15-035, R309) 4.3A FUEL CYCLE 21 The following subsections discuss the fuel system des ign, nuclear design, thermal-hydraulic design and reactor protection and monitoring system changes fo r the subject fuel cycle at Waterford 3. (DRN 04-502, R13;05-508, R14; 06-1059, R15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

Operating conditions for this cycle were assumed to be consistent with those of previous cycles and are summarized as full power operat ion under base load conditions. (DRN 02-1477, R12)

Cycle 2 information was submitted to the NRC via References 1 and 2. The NRC's Safety Evaluation Report for Cycle 2 was provided in Reference 3.

4.3A.1 GENERAL DESCRIPTION

(DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

The Waterford-3 Cycle 21 core will consist entirely of assemblies of the Next Generation Fuel (NGF) design; specifically, Fresh R egion EE assemblies, once burned Region DD, and twice burned Region CC and AA assemblies. See Sections 4.

2.2.1 and 4.2.2.2 for details of t he NGF fuel assembly and fuel rod designs. (DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

Control element assembly patterns and in-core inst rument locations are shown in Figure 4.3A-4 and Figure 4.3A-5 respectively.

4.3A.2 FUEL SYSTEM DESIGN

4.3A.2.1 Mechanical Design

4.3A.2.1.1 Fuel Design

(DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

The Cycle 21 core consists of those assembly ty pes and numbers listed in Table 4.3A-1. All fuel assemblies in the Cycle 21 core are of the NGF design. (DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

(DRN 02-1477, R12;04-502, R13;05-508, R14)

(DRN 02-1477, R12;04-502, R13;05-508, R14)

(DRN 04-502, R13)

(DRN 04-502, R13)

(DRN 00-1820, R10)

(DRN 00-1820, R10) 4.3A.2.1.2 Clad Collapse

(DRN 06-1059, R15; EC-9533, R302; EC-13881, R304)

The NGF fuel (UO

2) and IFBA rods in this cycle are initially pressurized with helium to the amount determined to be sufficient to prevent any gross clad deformation under the combined effect of external pressure and long term creep. The analyses of t hese rods credit the suppor t of pellets and/or the holddown spring to prevent gross deformation (s ee also Sections 4.2.

1.2.1 and 4.2.1.2.5). (DRN 06-1059, R15; EC-9533, R302; EC-13881, R304, LBDCR 15-035, R309)

WSES-FSAR-UNIT-3 4.3A-2 Revision 309 (06/16) 4.3A.2.2 Mitigation of Guide Tube Wear All fuel assemblies have stainless steel sleeves installed in the gui de tubes to prevent guide tube wear.

4.3A.2.3 Thermal Design (DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881 , R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

The thermal performance of composite fuel rods that envelope the rods of fuel batches present in Cycle 21 have been evaluated using the NRC approved FATES3B version of the C-E fuel evaluation model (References 6, 7 and 32) and the Zirconium Diboride (ZrB

2) burnable absorber methodology described in Reference 35. The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak pin at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at t he end of the Cycle. (EC-30663, R307, LBDCR 14-008, R308)

Reference 35 describes Westinghouse's 15 year f abrication and operational experience with ZrB 2 IFBA and the implementation and effect of using the coat ing on the C-E fuel assembly design and safety analyses. The neutronics effect, the helium produc tion effect on internal gas pressure, and the mechanical and thermal effects of the coating thick ness are all taken into account in the design and safety evaluations for C-E designed PWRs as described in that Reference. (DRN 02-1477, R12;04-502, R13;05-508, R14; 06-1059, R15; EC-9533, R302)

(EC-9533, R302)

The methodology for modeling the NGF design is de scribed in the CE 16x16 Next Generation Fuel Topical Report, Reference 43. (EC-13881, R304) 4.3A.2.4 Chemical Design (EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

The metallurgical design specifications of the f uel cladding and other fuel assembly components for the NGF fuel used in Cycle 21 are essentially the same as those of the fuel regions included in Cycle 1. The NGF design of Region EE, Region DD, Region CC and Region AA include Optimized ZIRLO TM for the cladding and spacer grids (Reference 44) and ZIRLO TM for the CEA guide tubes (Reference 43). The introduction of these material changes does not impose any new water chem istry requirements relative to those employed for the standard fuel assembly. (EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308) 4.3A.2.5 Shoulder Gap Adequacy (DRN 02-1477, R12;04-502, R13;05-508, R14; 06-1059, R15; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

Adequate shoulder gap is predicted for all NGF Regions of fuel in Cycle 21. This conclusion is based upon the fuel rod growth models of Reference 34 for Zircaloy, Reference 45 for ZIRLO TM , and Reference 43 for Optimized ZIRLO TM. The shoulder gap evaluation for Regi ons with the NGF design demonstrates that the initial shoulder gap reduction of approximatel y 0.5 inches relative to the non-NGF design is accommodated by the improved dimensional stabilit y of the NGF cladding and CEA guide tube materials (Optimized ZIRLO TM and ZIRLO TM, respectively). (DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308) 4.3A.3 NUCLEAR DESIGN

4.3A.3.1 Physics Characteristics

4.3A.3.1.1 Fuel Management

(DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308)

The Cycle 21 core consists of those assembly types and numbers listed in 4.3A-1. Twenty-one (21)

Region BB and seventy-six (76) Region CC assemblies irradiated during Cycle 20 will be removed from the core and replaced with ninety-six (96) fres h Region EE assemblies and one (1) twice-burned Region AA assembly that was not loaded in the Cycle 20 core. One hundred (100) Region DD and twenty (20)

Region CC assemblies in the core during Cycle 20 will be retained for Cycle 21. (DRN (00-1820, R10; 02-1477, R12;04-502, R13; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

WSES-FSAR-UNIT-3 4.3A-3 Revision 309 (06/16)

(DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 1 4-008, R308, LBDCR 15-035, R309) The Cycle 21 core makes use of a low-leakage fuel management scheme in which eight (8) previously burned Sub-Region CA, CC, DA, DD, DF, and DG asse mblies are each placed on the core periphery.

One (1) previously burned Sub-Region AD assembly is placed in the center location. The ninety-six (96) fresh Region EE (Sub-Regions EA, EB, EC, ED, and EE) assemblies are located throughout the interior of the core, where they are arranged with other previously burned Region DD and CC fuel assemblies in a pattern that minimizes power peaking, and reduces both core leakage and the total neutron fluence to the reactor vessel.

The Cycle 21 center assembly is a twice-burned Regi on AA assembly that was held in the Spent Fuel Pool during the Cycle 19 and Cycle 20 operations. It had previously been loaded in the core for the Cycle 17 and Cycle 18 operation.

Fuel rod enrichment and Zirc Diboride configurat ions for the Region EE fuel are presented in Figure 4.3A-1.

The Cycle 21 reload fuel enrichment and region size w ill provide a nominal best estimate cycle length of 485 EFPD (498 EFPD with coastdown) based on operation at 3716 MWth and a Cycle 20 nominal endpoint of 507 EFPD. Depending on the actual Cycle 20 endpoint, the Cycle 21 core could deliver as much as 498 EFPD (511 EFPD with coastdown) or as little as 477 EF PD (490 EFPD with coastdown) on a best estimate basis. The Cycle 20 termination burnup has been assumed to be between 482 and 522

EFPD.

Figures 4.3A-3a and 4.3A-3b display the beginning of Cycle 21 and the end of Cycle 21 (502 EFPD) assembly average burnup distributions. These bur nup distributions are bas ed on Cycle 20 endpoints of 482 and 522 EFPD, respectively.

Table 4.3A-2 provides a comparison of characteri stic physics parameters for Cycle 21 to the same parameters for Cycle 20, the Referenc e Cycle. The values in this table are intended to represent nominal core parameters. Those values used in the sa fety analyses (see Chapter

15) contain appropriate uncertainties, or incorporate val ues to bound future operating cycles, and in all cases are conservative with respect to the values calculated for Cycle 21.

Table 4.3A-3 presents a summary of CEA reactivity worths and allowances for the end of Cycle 21 full power steam line break transient. The full pow er steam line break was chosen as a reasonable illustration of the CEA reactivity worth. (DRN (00-1820, R10; 02-1477, R12;04-502, R13; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309) (DRN 02-1477, R12;04-502, R13)

The CEA core locations and group identifications are s hown in Figure 4.3A-4. At the end of Cycle 11, the eight (8) Part-Length CEAs comprising Bank P were replaced with full-length, full strength CEAs and reassigned to Bank A. Four (4) full-length CEAs in Shutdown Bank A were reassigned to Bank P.

Additionally, the four 4 Element CEAs in Shutdown B ank A, that span two fuel assemblies at the core periphery's major axes, were removed from the core. The Waterford 3 CEA Bank configurations are shown in Figure 4.3A-4. Commencing with Cycle 12, t he Waterford 3 core has a total of 87 CEAs, all of the standard five element design. The assumed pow er dependent insertion limits (PDIL) for regulating groups and CEA Group P are shown in Figures 4.3A-6 and 4.3A-7 respectively. Table 4.3A-4 shows the reactivity worths of various C EA groups calculated at full power conditions for this cycle and the Reference Cycle. (DRN 02-1477, R12;04-502, R13;05-508, R14; 06-1059, R15)

WSES-FSAR-UNIT-3 4.3A-4 Revision 309 (06/16) 4.3A.3.1.2 Power Distribution

Figures 4.3A-8 through 4.3A-10 illustrate the calc ulated All Rods Out (ARO) planar radial power distributions during this cycle. The one-pin pl anar radial power peaks presented in these figures represent the middle region of the core. Time poi nts at the beginning, middl e, and end of cycle were chosen to display the variation in maximum planar radial peak as a function of burnup.

The calculated radial power distributions described in this section do not include any uncertainties or allowances. The calculations performed to determi ne these radial power peaks explicitly account for augmented power peaking which is characteristic of fuel rods adjacent to the water holes. (DRN 02-1477, R12;04-502, R13)

The following endpoints apply to Figures 4.3A-8 through 4.3A-10: (DRN 05-508, R14; 06-1059, R15; EC-9533, R302; EC -13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

BOC21 values based on EOC20 = 482 EFPD, BOC21 = 0 EFPD MOC21 values based on EOC20 = 522 EFPD, MOC21 = 240 EFPD

EOC21 values based on EOC20 = 522 EFPD, EOC21 = 502 EFPD (DRN 02-1477, R12;04-502, R 13;05-508, R14; 06-1059, R 15; EC-9533, R302; EC-13881, R304, LBDCR 14-008, R308, LBDCR 15-035, R30

9) 4.3A.3.1.3 Maximum Fuel Rod Burnup (DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302; EC-13881, R304, LBDCR 14-008, R308, LBDCR 15-035, R309)

The Cycle 21 length will be limited to assure the ma ximum projected fuel rod burnup is less than the 60,000 MWD/T limit presented in Reference 34. The physics data which are input to cycle safety and fuel performance analyses are developed from explicit fine mesh calculations of fuel rod power and exposure. Burnup dependent physics data (e.g., maximum fuel rod fluence and fuel rod power histories) conservatively envelope core and fuel rod behavior at maximum burnups as well as lower burnups. (DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R 14; 06-1059, R15; EC -9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 1 4-008, R308, LBDCR 15-035, R309) 4.3A.3.2 Safety Related Data

4.3A.3.2.1 Augmentation Factors

As indicated in Reference 5, the increased power peaking associated with the small interpellet gaps found in modern fuel rods (non-densifying fuel in pre-pressurized tubes) is insignificant compared to the uncertainties in the safety analys es. The report concluded that augm entation factors can be eliminated from the reload analyses of any reactor loaded exclusivel y with this type of fuel. Therefore, augmentation factors have been eliminated for Waterford 3.

4.3A.3.3 Physics Analysis Methods

4.3A.3.3.1 Analytical Input To In-Core Measurements (EC-9533, R302)

In-core detector measurement constants to be used in evaluating the reload cy cle power distributions were calculated in accordance with Reference 42. (EC-9533, R302) 4.3A.3.3.2 Uncertainties In Measured Power Distribution (EC-9533, R302)

The planar radial power distribution measurement uncertainty based upon Reference 42 is applied to the COLSS and CPC on-line calculations which use pl anar radial power peaks. The axial and three dimensional power distribution measurement uncer tainties were determined using the values in Reference 42 in conjunction with other monitoring and protection system measurement uncertainties. (EC-9533, R302)

WSES-FSAR-UNIT-3 4.3A-5 Revision 307 (07/13) 4.3A.3.3.3 Nuclear Design Methodology (DRN 06-1059, R15)

Beginning with Cycle 15, the Advanced Nodal Code (ANC) (References 37, 38, and 39) was implemented in the reload design analysis. ANC is an advanced nodal analysis theory code capable of two- or three-dimensional calculations. Also , beginning with Cycle 15, PARAGON (Reference 40) computer code was implemented in the reload des ign analysis. PARAGON is a two-dimensional transport theory based code that calcul ates lattice physics constants.

These are the same methods and models that have been used in other Westinghous e reload cycle designs. These codes are replacements for the ROCS/DIT computer codes.

The primary purpose of PARAGON is to provide input data for use in three dimensional core simulator codes. This includes macroscopic cross sections, mi croscopic cross sections for feedback adjustments to the macroscopic cross sections, pin factors for pin power reconstruction calcul ations, and discontinuity factors for a nodal method solution. PARAGON can be used as a standalone or as a direct replacement for all the previously licensed Westinghouse Pressuri zed Water Reactor ("PWR") lattice codes, such as PHOENIX-P, as approved by the NRC in Reference 40. (EC-9533, R302)

PARAGON is a two-dimensional multi-group neutron (and gamma) transport code. The PARAGON flux

solution calculation uses Collision Probability theory wi thin the interface current method to solve the integral transport equation. Throughout the whole ca lculation, PARAGON us es the exact heterogeneous geometry of the assembly and the sa me energy groups as in the cross-section library to compute the multi-group fluxes for each micro-region location of the assembly.

In order to generate the multi-group data that will be used by a core simulator code, PARAGON goes through four steps of calculations: resonance self-shielding, flux solution, homogenization, and burnup calculation.

ANC (for Advanced Nodal Code) is the three-dim ensional core simulator code in the Westinghouse nuclear design code system. The ANC nodal flux so lution is based on a set of two-group diffusion theory nodal balance equations that are solved using a solution method based on the nodal expansion method (NEM). This method and the specific approximat ions made in the ANC implementation provide an accurate representation of the core nodal neutronics. A NC is used to calculate core reactivity, reactivity coefficients, critical boron, rod worths, and core, a ssembly, and rod power distributions for normal and off-normal conditions for use in design and safety analyses. The ANC computer code is also used in the

COLSS/CPC uncertainty analysis, as a replacement for the ROCS code, which in turn was a replacement for the FLAIR computer code. (DRN 06-1059, R15; EC-9533, R302)

4.3A.4 THERMAL-HYDRAULIC DESIGN

4.3A.4.1 DNBR Analysis (DRN 02-523, R12; 03-2058, R 14; EC-9533, R302; EC-13881, R304; EC-30663, R307)

Steady state DNBR analyses at the rated power level of 3716 MWT have been performed using the TORC computer code described in Reference 11, the WSSV-T and ABB-NV critical heat flux correlations applicable to NGF assemblies described in Reference 41 and 46, respectively, the TORC modeling methods described in References 11 and 13, and the CETOP code described in Reference 14. (DRN 02-523, R12; 03-2058, R14; EC-9533, R3 02; EC-30663, R307)

Table 4.3A-5 contains a list of pertinent thermal-hy draulic design parameters. The Modified Statistical Combination of Uncertainties (MSCU) methodol ogy presented in Reference 15 was applied with Waterford 3 specific data using the calculational fa ctors listed in Table 4.3A-5 and other uncertainty factors at the 95/95 confidence/probability level to def ine a design limit of 1.24 over a DNBR range of 1.0 to 1.24, applicable to both the ABB-NV and WSSV-T correlations. (EC-13881, R304)

WSES-FSAR-UNIT-3 4.3A-6 Revision 304 (06/10)

The DNBR limit includes the following allowances: (EC-13881, R304)

1. NRC specified allowances for TORC code uncertainty.
2. Rod bow penalty as discussed in Section 4.3A.4.2 below. (EC-9533, R302)

(EC-9533, R302; EC-13881, R304) 4.3A.4.2 Effects Of Fuel Rod Bowing on DNBR Margin (DRN 03-2058, R14; EC-9533, R302; EC -13881, R304)

Effects of fuel rod bowing on DNBR margin have been in corporated in the safety and setpoint analyses in the manner discussed in Reference 19. The penalty used for this analysis is valid for bundle burnups up to 33,000 MWD/T. This penalty is included in t he 1.24 DNBR limit, applicable to both the ABB-NV and WSSV-T correlations. (EC-9533, R302; EC-13881, R304)

For assemblies with burnup greater than 33,000 MWD/T suffi cient available margin exists to offset rod bow penalties due to the lower radial power peaks in these higher burnup batches. Hence the rod bow penalty based upon Reference 19 for 33,000 MWD/T is applicable for all assembly burnups expected. DRN 03-2058, R14) 4.3A.5 REACTOR PROTECTION AND MONITORING

4.3A.5.1 Introduction

The Core Protection Calculator (CPC) System is designed to provide the low DNBR and high Local Power Density (LPD) trips to (1) ensure that the s pecified acceptable fuel design limits on departure from nucleate boiling and centerline fuel melting are not exceeded during Anticipated Operational Occurrences (AOOs) and (2) assist the Engineered Safety Features System in limiting the consequences of certain postulated accidents. The CPCS is furt her described in subsection 7.2.1.1.2.5.

The CPC/CEAC in conjunction with the balance of t he Reactor Protection System (RPS) must be capable of providing protection for certain specified design bas is events, provided that at the initiation of these occurrences the Nuclear Steam Supply System, its sub-systems, com ponents and parameters are maintained within operating limits and Limiting Conditions for Operation (LCOs).

4.3A.5.2 CPCS Software Modifications

The CPC/CEAC software for Waterford 3 was modified prior to Cycle 2. This modification implemented the CPC Improvement Program, including algorithm s and plant specific dat a base changes, changes to the list of addressable constants and implementation of the Reload Data Block (RDB).

The Waterford 3 CPC/CEAC algorithms are the same as those implemented at SONGS-2 and -3 (Cycle

3) and at ANO-2 (Cycle 6) and described in Refer ences 21 and 22. The revised list of addressable constants are defined in Reference 23. The software modifications are described in References 23, 24, 25, and 29. All changes were implemented per the established software change procedures, References 26 and 27.

WSES-FSAR-UNIT-3 4.3A-7 Revision 304 (06/10) 4.3A.5.3 Addressable Constants Certain CPC constants are address able so that they can be changed as required during operation.

Addressable constants include (1) constants that are measured during startup (e.g., shape annealing matrix, boundary point power correlation coeffici ents, and adjustments for CEA shadowing and planar radial peaking factors), (2) uncertainty factors to account for processing and measurement uncertainties in DNBR and LPD calculations (BERRO through BERR4

), (3) trip setpoints and (4) miscellaneous items (e.g., penalty factor multipliers, CEAC penalty factor time delay, pre-trip setpoints, CEAC inoperable flag, calibration constants, etc.).

Trip setpoints, uncertainty factors and other addre ssable constants have been det ermined consistent with the software and methodology established in the CPC Improvement Program (Reference 23, 24 and 25)

and the cycle design, performance, and safety analyses.

4.3A.5.4 Digital Monitoring System (COLSS)

The Core Operating Limit Supervisory System (COLSS) is a monitoring system that initiates alarms if the LCO on DNBR, peak linear heat rate, core power, axial shape index, or core azimuthal tilt are exceeded.

The COLSS is further described in subsection 7.7.

1.5. The COLSS data base and uncertainties have been updated to reflect the current core design.

4.3A.6 REFERENCES TO APPENDIX 4.3A

1. W3P86-1686 dated August 29, 1986.
2. W3P86-3328 dated October 1, 1986.
3. J.H. Wilson (NRC), to J.G. Dewease (LP&L), "Reload Analysis Report for Cycle 2 at Waterford 3," January 16, 1987.
4. C.O. Thomas (NRC), to A.E. Scherer (C-E), "Acceptance for Referencing of Licensing Special Report LD-84-043, CEA Guide Tube Wear Sleeve Modification," September 7, 1984.
5. EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding Volume 5:

Evaluation of Interpellet Gap Formation and Cl ad Collapse in Modern PWR Fuel Rods," April 1985.

6. CENPD-139-P-A, "C-E Fuel Evaluat ion Model Topical Report," July 1974.
7. CEN-161(B)-P-A, "Improvements to Fuel Evaluation Model," August 1989.
8. R.A. Clark (NRC) to A.E. Lundvall, Jr., (B G&E), "Safety Evaluation of CEN-161 (FATES3)," March 31, 1983.
9. ENEAD-02-NP, "Verification of CECOR Coeffici ent Methodology for Application to Pressurized Water Reactors of the Entergy System," September 1994.
10. CENPD-266-P-A, "The ROCS and DIT Comput er Codes for Nuclear Design," April 1983.

10B. CENPD-275-P-A, "C-E Methodology for Core Designs Containing Gadelinia-Urania Burnable Absorbers," May 1988.

11. CENPD-161-P-A, "TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986.

WSES-FSAR-UNIT-3 4.3A-8 Revision 304 (06/10)

12. CENPD-162-P-A, "Critical Heat Flux Correlati on for C-E Fuel Assemblies with Standard Spacer Grids, Part 1, Uniform Axial Po wer Distribution," September 1976.
13. CENPD-206-P-A, "TORC Code, Verification and Simplified Modeling Methods," June 1981.
14. CEN-160(S)-P, Rev. 1-P, "CETOP Code St ructure and Modeling Methods for San Onofre Nuclear Generating Station Un its 2 and 3," September 1981.
15. CEN-356(V)-P-A, Revision 01-P-A, "Modified Statistical Combinati on of Uncertainties, Part 1, Combination of System Parame ter Uncertainties," May 1988.

(EC-13881, R304)

16. Deleted
17. Deleted (EC-13881, R304)
18. NUREG-0787, Supplement 1, "Saf ety Evaluation Report Related to the Operation of Waterford Steam Electric Station, Unit No. 3," Docket No. 50-382, October 1981.
19. CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983.
20. Robert A. Clark (NRC) to William Cavanaugh III, (AP&L), "Operation of ANO-2 During Cycle 2," July 21 1981 (Safety Evaluation Report and License Amendment No. 26 for ANO-2).
21. CEN-304-P, Rev. 01-P, "Functional Requirements for a Control Element Assembly Calculator," May 1986.
22. CEN-305-P, Rev. 01-P, "Functional Requirem ents for a Core Protection Calculator," May 1986.
23. CEN-308-P-A, "CPC/C EAC Software Modifications for the CPC Improvement Program," April 1986.
24. CEN-310-P-A, "CPC and Methodology Changes for the CPC Improvement Program," April 1986.
25. CEN-330-P-A, "Rev. 00-P, "CPC/CEAC Software Modificati ons for the CPC Improvement Program Reload Data Block," October 1987.
26. CEN-39(A)-P, Rev. 03, "CPC Protection Algorithm Software Change Procedure," January 1986.
27. CEN-39(A)-P, Supplement 1-P, Rev. 03-P, "CPC Protection Al gorithm Software Change Procedure Supplement 1," April 1986.
28. CEN-323-P-A, "Reload Data Block Constant Installation Guidelines," September 1986.
29. CEN-281(S)-P, "CPC/CEAC Softw are Modifications for San Onofre Nuclear Steam Generating Station Units No. 2 and 3," July 1984.
30. LP&L (KW Cook) letter to NRC, "Cycle 3 Shoulder Gap Evaluation," July 24, 1987.
31. Deleted.
32. CEN-161(B)-P-A Supplement 1-P-A, "Improvements to Fuel Evaluation Model," January 1992.

WSES-FSAR-UNIT-3 4.3A-9 Revision 308 (11/14)

33. CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990
34. CEN-386-P-A, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/Kg for Combustion Engineering 16 x 16 PWR Fuel," August 1992.

(EC-13881, R304)

35. WCAP-16072-P-A, "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs," August 2004. (EC-13881, R304)

(DRN 03-270, R12-B)

36. ENEAD-01-P, Revision 0, "Qualification of Reac tor Physics Methods for the Pressurized Water Reactors of the Entergy System," December 1993. (DRN 03-270, R12-B)

(EC-9533, R302)

37. WCAP-11596-P-A, "Qualification of the PH OENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988
38. WCAP-10965-P-A, "ANC: A Westinghouse Adv anced Nodal Computer Code," September 1986
39. WCAP-10965-P-A Addendum 1, "ANC: A Westi nghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery," April 1989
40. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport C ode PARAGON," August 2004
41. WCAP-16523-P-A, Rev. 0. "Westinghouse Co rrelations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles wi th Side-Supported Mixing Vanes", August 2007.
42. CENPD-153-P, Revision 1-P-A, "Evaluation of Uncertainty in the Nuclear Power Peaking Measured by the Self-Powered, Fixed In-Core Detector System", May 1980. (EC-9533, R302)

(EC-13881, R304, LBDCR 14-008, R038)

43. WCAP-16500-P-A, "CE 16x16 Next Generation Fuel Core Reference Report", August 2007.
44. WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A, "Optimized ZIRLOŽ", July 2006.
45. CENPD-404-P-A, "Implementat ion of ZIRLOŽ Material Cladding in CE Nuclear Power Fuel Assembly Designs," November 2001.
46. CENPD-387-P-A, "ABB Critical Heat Fl ux Correlations for PWR Fuel," May 2000. (EC-13881, R304, LBDCR 14-008, R308)

WSES-FSAR-UNIT-3 TABLE 4.3A-1 Revision 309 (06/16)

(DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R14; 06-1059, R15; EC-9533, R 302; EC-13881, R304; EC-30663, R307, LBDCR 1 4-008, R308, LBDCR 15-035, R309)

Waterford - 3 Cycle 21 Core Loading Description Sub-Batch ID Number of Assemblies Pattern ID UO 2 Rods per Assembly Nominal Enrichment (wt. %) ZrB 2 Rods per Assembly Shim Loading (ZrB 2) Number of Fuel Rods (Including ZrB 2 Rods) Number of ZrB 2 Rods EA 20 PAT1633IFB 164 4.38 20 2.0 X 3680 400 (60 IFBA) 12 3.98 40 2.0 X 1040 800 EB 8 PAT1649IFB 116 4.38 68 2.0 X 1472 544 (112 IFBA) 8 3.98 44 2.0 X 416 352 EC 4 PAT1632IFB 176 3.98 8 2.0 X 736 32 (48 IFBA) 12 3.58 40 2.0 X 208 160 ED 48 PAT1649IFB 116 3.98 68 2.0 X 8832 3264 (112 IFBA) 8 3.58 44 2.0 X 2496 2112 EE 16 PAT1636IFB 112 3.98 72 2.0 X 2944 1152 (124 IFBA) 0 3.58 52 2.0 X 832 832 Total 96 22656 9648 DA 16 PAT1632IFB 176 4.53 8 2.0 X 2944 128 (48 IFBA) 12 4.23 40 2.0 X 832 640 DB 4 PAT1648IFB 124 4.53 60 2.0 X 736 240 (88 IFBA) 24 4.23 28 2.0 X 208 112 DC 12 PAT1649IFB 116 4.53 68 2.0 X 2208 816 (112 IFBA) 8 4.23 44 2.0 X 624 528 DD 8 PAT1650IFB 92 4.53 98 2.0 X 1472 736 (136 IFBA) 8 4.23 44 2.0 X 416 352 DE 8 PAT1649IFB 116 3.83 68 2.0 X 1472 544 (112 IFBA) 8 3.53 44 2.0 X 416 352 DF 20 PAT1636IFB 112 3.83 72 2.0 X 3680 1440 (124 IFBA) 0 3.53 52 2.0 X 1040 1040 DG 32 PAT1650IFB 92 3.83 92 2.0 X 5888 2944 (136 IFBA) 8 3.53 44 2.0 X 1664 1408 Total 100 23600 11280 CA 8 PAT1643IFB 160 4.23 24 2.0 X 1472 192 (68 IFBA) 8 3.92 44 2.0 X 416 352 CC 8 PAT1648IFB 124 3.81 60 2.0 X 1472 480 (88 IFBA) 24 3.51 28 2.0 X 416 224 CE 4 PAT1649IFB 116 3.81 68 2.0 X 736 272 (112 IFBA) 8 3.51 44 2.0 X 208 176 Total 20 4720 1696 AD 1 PAT1635IFB 136 3.90 48 2.0 X 184 48 (100 IFBA) 0 3.50 52 2.0 X 52 52 Total 1 236 100 Grand Total 217 51212 22724 (DRN 00-1820, R10; 02-1477, R 12;04-502, R13;05-508, R14; 06-1059, R15; EC-9533, R 302; EC-13881, R304; EC-30663, R307, LBDCR 1 4-008, R308, LB DCR 15-035, R309)

WSES-FSAR-UNIT-3 TABLE 4.3A-2 Revision 309 (06/16)

(DRN 05-508, R14; 06-1059, R15; EC-9533, R302; EC -13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

NOMINAL PHYSICS CHARACTERISTICS Units Reference Cycle** Cycle 21* Dissolved Boron Dissolved Boron Concentration for Criticality, CEAs Withdrawn, Hot Full Power, Equilibrium Xenon PPM 527 583 Inverse Boron Worth Hot Full Power, Equilibrium Xenon BOC PPM/% 130 130 EOC PPM/% 103 104 Moderator Temperature Coefficients Hot Full Power, Equilibrium Xenon BOC 10-4/°F -1.5 -1.4 EOC 10-4/°F -2.8 -2.9 Doppler Coefficient Hot Zero Power, BOC 10

-5/°F -1.7 -1.7 Hot Full Power, Equilibrium Xenon BOC 10-5/°F -1.6 -1.6 EOC 10-5/°F -1.8 -1.8 Total Delayed Neutron Fraction eff BOC ----------- 0.0061 0.0061 EOC ----------- 0.0050 0.0050 Neutron Generation Time,

  • BOC 10-6 sec 17.0 18.0 EOC 10-6 sec 27.9 27.9
  • values vary with cycle

WSES-FSAR-UNIT-3 TABLE 4.3A-3 Revision 309 (06/16)

LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT FULL POWER STEAM LINE BREAK, % , END-OF-CYCLE (EOC)

(DRN 05-508, R14; 06-1059, R15; EC-9533, R302; EC -13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

Reference Cycle** Cycle 21*

Net Available Scram Worth (No LOAC) 8.1 7.9

  • values vary with cycle
    • Reference cycle is Cycle 20

(DRN 05-508, R14; 06-1059, R15; EC-9533, R302; EC -13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

WSES-FSAR-UNIT-3

TABLE 4.3A-4 Revision 309 (06/16)

REACTIVITY WORTH OF CEA REGULATING GROUPS AT HOT FULL POWER, %

Beginning of Cycle End Of Cycle (DRN 05-508, R14; 06-1059, R15; EC-9533, R302; EC -13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

Reference Reference Cycle** Cycle 21* Cycle** Cycle 21* (EC-13881, R304)

Group P @ 0" 0.4 0.4 0.4 0.4 Group 6 @ 0" 0.4 0.4 0.4 0.4 Group 5 @ 0" 0.4 0.3 0.4 0.4

Note: Values shown assume sequential group insertion

WSES-FSAR-UNIT-3 TABLE 4.3A-5 Revision 309 (06/16)

(DRN 00-1820, R10;02-523, R 12; 02-1477, R12;04-502, R 13 ,03-2058, R14;05-508, R14; 06-1059, R15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

Cycle 21 Thermal-Hydraulic Parameters at Full Power General Characteristics Units Cycle 20 Cycle 21 Total Heat Output (Core Only) (MW th) 3716 3716 (10 6 Btu/hr) 12680 12680 Fraction of Heat Generated in Fuel Rod --- 0.975 0.975 Primary System Pressure (Nominal) (psia) 2250 2250 Inlet Temperature (Maximum Indicated) (°F) 543 543 Total Reactor Coolant Flow (Minimum Steady State) (gpm) 390,220 390,220 (10 6 lbm/hr) 148.0 148.0 Coolant Flow Through Core (Minimum) (10 6 lbm/hr) 144.2 144.2 Hydraulic Diameter (Nominal Channel) (ft) 0.041 0.041 Core Average Mass Velocity (10 6 lbm/hr-ft

2) 2.55 2.55 Pressure Drop Across Core (at Minimum Steady State Core Flow Rate) (psi) 20.7 20.7 Total Pressure Drop Across Vessel (Based on Nominal Dimensions and Minimum Steady State

Flow) (psi) 46.6 46.6 Core Average Heat Flux (Accounts for Fraction of Heat Generated in Fuel Rod and Axial

Densification Factor) (Btu/hr-ft

2) 198,016 (1) 198,016 (1) Total Heat Transfer Area (Accounts for Axial Densification Factor) (ft 2) 62,432 (1) 62,432 (1) Film Coefficient at Average Conditions (Btu/hr-ft 2-°F) 6092 6092 Average Film Temperature Difference (°F) 32.50 (1) 32.50 (1) Average Linear Heat Rate of Undensified Fuel Rod (Accounts for Fraction of Heat Generated In

Fuel Rod) (kw/ft) 5.67 (1) 5.67 (1) Average Core Enthalpy Rise (Btu/lbm) 88.0 88.0 Maximum Clad Surface Temperature (°F) 656.76 (1) 656.76 (1) Engineering Heat Flux Factor --- 1.03 (2),(3) 1.03 (2),(3) Engineering Factor on Hot Channel Heat Input --- 1.03 (2),(3) 1.03 (2),(3) Rod Pitch, Bowing and Clad Diameter Factor --- 1.05 (2),(3) 1.05 (2),(3) Fuel Densification Factor (Axial) --- 1.002 1.002 (1) Based on 100 shims (non fuel rods) in the core and 217 NGF assemblies. (2) These factors have been combined statistically with other uncertainty factors at 95/95 confidence/probability level and included in the design limit on ABB-NV minimum DNBR and WSSV-T minimum DNBR. (3) These values are generic based on fuel design drawing tolerances and are also applicable to NGF.

(DRN 00-1820, R10;02-523, R 12; 02-1477, R12;04-502, R 13 ,03-2058, R14;05-508, R14; 06-1059, R15; EC-9533, R302; EC-13881, R304; EC-30663, R307, LBDCR 14-008, R308, LBDCR 15-035, R309)

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