ML053190242

0000-0029-7705-MCAR, Rev. 0, Mixed Core Analysis Report for Hope Creek Reload 12 Cycle 13.

ML053190242
Person / Time
Site:	Hope Creek
Issue date:	04/30/2005
From:	Harding M, Kingston R Global Nuclear Fuel
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H05-01, LR-N05-0330 0000-0029-7705-MCAR, Rev 0
Download: ML053190242 (190)

Text

Attachment 3 LR-N05-0330 LCR H05-01 0000-0029-7705-MCAR, Rev. 0 Mixed Core Analysis Report (MCAR) for Hope Creek Reload 12 Cycle 13

Global Nuclear Fuel A Joist Vcmorm of GE, Toshibd, & Hitachi 0000-0029-7705-MCAR Revision 0 April 2005 0000-0029-7705-MCAR, Rev. 0 Mixed Core Analysis Report (MCAR) for Hope Creek Reload 12 Cycle 13 Approved: Approved: __ _ _

M. E. Hr R. E.Ki(gston Fuel Engineerin rvrs Customer Account Leader

PSEG Hope Creek Mixed Core Analysis Report Proprietary Information Notice This document is the GNF non-proprietary version of the GNF proprietary report. From the GNF proprietary version, the information denoted as GNF proprietary (enclosed in double brackets) was deleted to generate this version.

ii

PSEG Hope Creek Mixed Core Analysis Report Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by Global Nuclear Fuel - Americas, LLC (GNF-A) solely for PSEG Nuclear, LLC. The information contained in this report is believed by GNF-A to be an accurate and true representation of the facts known, obtained or provided to GNF-A at the time this report was prepared.

The only undertakings of GNF-A respecting information in this document are contained in the Contract between PSEG Nuclear, LLC, Global Nuclear Fuel - Americas, LLC and General Electric Company for Fuel Fabrication and Related Components and Services for Hope Creek Generating Station, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GNF-A nor any of the contributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

iii

PSEG Hope Creek Mixed Core Analysis Report Table of Contents 1.0 Introduction and Summary .1

1.1 REFERENCES

.2 2.0 Lattice Physics Comparison. 3 2.1 TGBLA LATTICE PHYSICS TO MONTE CARLO COMPARISON . . . 3 2.1.1 Water-CrossModels. 3 2.1.2 Q0icahfcalionMethod.3 2.1.3 TGBLA 061MCNP ComparisonResults. 4 2.1.4 Boron Worth Accuracyfor Standby Liquid ControlAnalysis .20 2.1.5 3D SimulatorAccuracy........................................................................................ 21 2.1.6 References.22 2.2 LATTICE PiYSICS RESULTS . . .22 2.2.1 Description .22 3.0 Cycle 13 Base Point Determination - Cycle 9-12 Simulation .29 3.1 CYCLE 9-12 PERFORMANCE TRACKING .29 3.2 3D PROCESS COMPUTER TRANSVERSE IN-CORE PROBE (TIP) COMPARISONS. 38 3.2.1 Statistics Summary .. 38 3.2.2 Plots of Bundle andNodal TIP Conparisons . .39 3.2.3 ComparisonofAxial TIPplotsfor Cycle 11 . .40 4.0 Fuel Rod Thermal-Mechanical Report Description .47 4.1 FUEL ROD THERMAL-MECHANICAL PERFORMANCE LIMITS FOR STEADY-STATE OPERATION.................................................................................................................... 47 4.1.1 U02 FuelRod Steady-State Limit versus Exposure . .47 4.1.2 Gd.0. Fuel Rod Steady-State Limit versus Exposure.......................................... 47 4.2 FUEL ROD THERMAL-MECHANICAL LIMITS FOR ANTICIPATED OPERATIONAL OCCURRENCES............................................................................................................... 47 4.2.1 Design and LicensingLimits.48 4.2.2 Thermal andMechanicalOverpowers. .48 4.2.3 Limiting ThermalandMechanicalOverpowrersforAQOs at Rated Powver and Flouw....................................................................................................................... 55 5.0 GE14 / SVEA 96+ Demonstration Cycle Analysis Description - Cycle 13 .61 5.1 RELOAD BUNDLE DESIGN DESCRIPTION .61 5.2 CYCLE 13 CORE DESIGN DESCRIPTION .62 5.2.1 Core Configuration Description .62 5.2.2 Design Limits and Targets.................................................................................... 62 5.3 CYCLE 13 PERFORMANCE

SUMMARY

.63 6.0 Safety Limit Minimum Critical Power Ratio (SLMCPR) .82 6.1 DISCuSSION .82 iv

PSEG Hope Creek Mixed Core Analysis Report Table of Contents 6.2

SUMMARY

...................................................................................................................... 84

6.3 REFERENCES

...................................................... 84 7.0 Cycle 13 Supplemental Reload Licensing Report (SRLR) ......................................... 90 v

PSEG Hope Creek Mixed Core Analysis Report List Of Tables TABLE 2.1 -

SUMMARY

OF MCNP SIMULATION OF TRX AND B&W CRITICAL EXPERIMENTS ...... 4 TABLE 2.2 - HOT UNCONTROLLED BEGINNING OF LIFE K.. AND DYNAMIC VOID COEFFICIENT COMPARISONS SVEA 96+ AND GE14 LATTICES .................................................... 6 TABLE 2.3 - HOT UNCONTROLLED K. AND DYNAMIC VOID COEFFICIENT COMPARISONS FOR EXPOSED CONDITIONS SVEA 96+ LATTICE 6026 .................................................. 8 TABLE 2.4 - HOT UNCONTROLLED BEGINNING OF LIFE PIN POWER COMPARISON SVEA 96+ AND GE14 LATTICES .................................................................. 15 TABLE 2.5 - HOT UNCONTROLLED PIN POWER COMPARISON VERSUS EXPOSURE FOR SVEA 96+

LATTICE 6026 .................................................................. 20 0

TABLE 2.6 - COLD (1 60 C) UNCONTROLLED BORATED BEGINNING OF LIFE K. COMPARISON SVEA 96+ AND GE 14 LATnCES .................................................................. 20 TABLE 2.7 - REPRESENTATIVE LATTICE EVALUATION .................................................................. 22 TABLE 2.8 - SVEA 96+ BUNDLE 2657 Km ..................................................................................... 24 TABLE 2.9 - GEI4 BUNDLE 2695 K. ......................... .. . . . . . . . 27 TABLE 3.1 - RM VALUES S ......................... 38 TABLE 4.1 - UO 2 ROD BWREDBFUEL LIMITS .................................................................. 56 TABLE 4.2 - U0 2 AND (U,GD)0 2 FUEL ROD MAXIMUM POWER, KW/Ff ....................................... 57 TABLE 4.3 - U0 2 AND (U, GD)O 2 ROD/SECTION BWREDBFUEL LIMITS ................................. 58 TABLE 4.4 - LFWH, INADVERTENT HPCS, HPCI, RCIC INJECTION, RWE-OUTSIDE ERROR CELL OVERPOWER LIMITS FOR U0 2 RODS ....................................................... 58 TABLE 4.5 - LFWH, INADVERTENT HPCS, HPCI, RCIC INJECTION, RWE-OUTSIDE ERROR CELL OVERPOWER LIMITS FOR GD RODS .................................................................. 59 TABLE 4.6 - OVERPOWER TRANSIENT MAGNITUDE GUIDELINE LIMITS FOR SHORT DURATION PRESSURIZATION TRANSIENTS USING GEMINI METHOD FOR U0 2 RODS ................ 59 TABLE 4.7 - OVERPOWER TRANSIENT MAGNITUDE GUIDELINE LIMITS FOR SHORT DURATION PRESSURIZATION TRANSIENTS USING GEMINI METHOD FOR GD RODS .................. 60 TABLE 5.1 - CORE DESIGN LIMITS .................................................................. 64 TABLE 5.2 - CORE DESIGN MARGIN TARGETS .................................................................. 64 TABLE 5.3 - CYCLE 13 RLP

SUMMARY

OF ROD PATTERN RESULTS ............................................. 65 TABLE 5.4 - CYCLE 13 RLP HOT EXCESS REACTIVITY ................................................................. 66 TABLE 5.5 - CYCLE 13 RLP COLD SHUTDOWN REACTIVITY MARGIN .......................................... 67 TABLE 5.6 - CYCLE 13 RLP STANDBY LIQUID CONTROL SHUTDOWN MARGIN ............................ 68 TABLE 6.1 - COMPARISON OFTITEHOPE CREEK GENERATING STATION CYCLE 13 AND CYCLE 12 SLMCPR .................................................................. 86 TABLE 6.2 - STANDARD UNCERTAINTIES .................................................................. 87 TABLE 6.3 - EXCEPTIONS TO TIIE STANDARD UNCERTAINTIES USED IN HOPE CREEK CYCLE 13 AND CYCLE 12 .................................................................. 87 vi

PSEG Hope Creek Mixed Core Analysis Report List Of Figures FIGURE 2.1 - SVEA 96+ LATTICE 6019 TGBLAO6V/MCNP K.OCOMPARISON ........................... 10 FIGURE 2.2 - SVEA 96+ LATTICE 6020 TGBLAO6V/MCNP K. COMPARISON ........................... 10 FIGURE 2.3 - SVEA 96+ LATTICE 6022 TGBLAO6V/MCNP KROCOMPARISON ........................... 11 FIGURE 2.4 - SVEA 96+ LATTICE 6023 TGBLAO6V/MCNP K.DCOMPARISON ........................... 11 FIGURE 2.5 - SVEA 96+ LATTICE 6024 TGBLAO6V/MCNP K,, COMPARISON ........................... 12 FIGURE 2.6 - SVEA 96+ LATTICE 6026 TGBLAO6V/MCNP K,, COMPARISON ........................... 12 FIGURE 2.7 - GE14 LATTICE 4963 REGULAR ZONE TGBLA06V/MCNP K. COMPARISON .......... 13 FIGURE 2.8 - GEI4 LATTICE 4966 VANISHEDZONETGBLAO6V/MCNP K. COMPARISON ........ 13 FIGURE 2.9 - SVEA 96+ LATTICE 6019 TGBLAO6V/MCNP FISSION DENSITY COMPARISON .... 16 FIGURE 2.10 - SVEA 96+ LATTICE 6020 TGBLAO6V/MCNP FISSION DENSITY COMPARISON.. 16 FIGURE 2.11 - SVEA 96+ LATTICE 6022 TGBLAO6V/MCNP FISSION DENSITY COMPARISON.. 17 FIGURE 2.12 - SVEA 96+ LATTICE 6023 TGBLAO6V/MCNP FISSION DENSITY COMPARISON.. 17 FIGURE 2.13 - SVEA 96+ LATTICE 6024 TGBLAO6VINCNP FISSION DENSITY COMPARISON.. .18 FIGURE 2.14 - SVEA 96+ LATTICE 6026 TGBLAO6V/MCNP FISSION DENSITY COMPARISON.. 18 FIGURE 2.15 - GEI4 LATTICE 4963 TGBLA06V/MCNP FISSION DENSITY COMPARISON ........... 19 FIGURE 2.16 - GE] 4 LATTICE 4966 TGBLAO6V/MCNP FISSION DENSITY COMPARISON ........... 19 FIGURE 2.17 - SVEA 96+ BUNDLE 2657 CONFIGURATION ........................................................... 23 FIGURE 2.1 8 - SVEA 96+ BUNDLE 2657 2D DOMINANT LATTICE POWER PEAKING .......... .......... 25 FIGURE 2.19 - GEI4 BUNDLE 2695 CONFIGURATION .............................................................. 26 FIGURE 2.20 - GE14 BUNDLE 2695 2D DOMINANT LATTICE POWER PEAKING ............................ 28 FIGURE 3.1 -CYCLE 9 - CYCLE 12 HOT CRITICAL EIGENVALUE TRACKING ................................. 30 FIGURE 3.2 - CYCLE 9 - CYCLE 12 COLD CRITICAL EIGENVALUE TRACKING ............................... 31 FIGURE 3.3 - CYCLE 10 HOT CRITICAL EIGENVALUE .............................................................. 32 FIGURE 3.4 - CYCLE 10 MFLCPR .............................................................. 32 FIGURE 3.5 - CYCLE 10 MFLPD .............................................................. 33 FIGURE 3.6 - CYCLE IO MAPRAT .............................................................. 33 FIGURE 3.7 - CYCLE 11 HOT CRITICAL EIGENVALUE .............................................................. 34 FIGURE 3.8 - CYCLE 11 MFLCPR .............................................................. 34 FIGURE 3.9 - CYCLE 1 MFLPD.............................................................. 35 FIGURE 3.10 - CYCLE 11 MAPRAT .............................................................. 35 FIGURE 3.11 - CYCLE 12 HOT CRITICAL EIGENVALUE .............................................................. 36 FIGURE 3.12 - CYCLE 12 MFLCPR .............................................................. 36 FIGURE 3.13 - CYCLE 12 MFLPD .............................................................. 37 FIGURE 3.14 - CYCLE 12 MAPRAT .............................................................. 37 FIGURE 3.15 - PLOT OF TIP NODAL RMS % VERSUS CYCLE EXPOSURE GWD/ST FOR CYCLES 9-12 ................................................... 39 FIGURE 3.16 - PLOT OF TIP BUNDLE RMS % VERSUS CYCLE EXPOSURE GWD/ST FOR CYCLES 9-12 ................................................... 40 FIGURE 3.17 - PLOT OF AXIAL TIP COMPARISON FOR CYCLE 11 AT A SELECTED EXPOSURE POINT NEARBOC ................................................... 41 FIGURE 3.18 - PLOT OF CORE AVERAGE AXIAL TIP COMPARISON FOR CYCLE 11 AT A SELECTED EXPOSURE POINT NEAR BOC ................................................... 42 Vii

PSEG Hope Creek Mixed Core Analysis Report FIGURE 3.19 - PLOT OF AXIAL TIP COMPARISON FOR CYCLE I I AT A SELECTED EXPOSURE POINT NEAR MOC ............................................................... 43 FIGURE 3.20 - PLOT OF CORE AVERAGE ANAL TIP COMPARISON FOR CYCLE 1 1 AT A SELECTED ExPosuRE POINT NEAR MOC ........................... .................................... 44 FIGURE 3.21 - PLOT OF AXIAL TIP COMPARISON FOR CYCLE I I AT A SELECTED EXPOSURE POINTNEAREOC ............................................................... 45 FIGURE 3.22- PLOT OF CORE AVERAGE AXIAL TIP COMPARISON FOR CYCLE 11 AT A SELECTED EXPOSURE POINT NEAR EOC ........................... .................................... 46 FIGURE 4.1 - GRAPH OF THERMAL AND MECHANICAL OVERPOWERS ............................................ 50 FIGURE 5.1 - FRESH GE14 RELOAD BUNDLE 2757 CONFIGURATION ............................................ 69 FIGURE 5.2 - FRESH GEI4 RELOAD BUNDLE 2758 CONFIGURATION ............................................ 70 FIGURE 5.3 - CYCLE 13 (QUARTER CORE) .................... ........................................... 71 FIGURE 5.4 - CYCLE 13 REFERENCE LOADING PATTERN CONTROL ROD OPERATING SEQUENCE . 72 FIGURE 5.5 - CYCLE 13 RLP ROD PATTERN THERMAL DESIGN RATIO RESULTS ........... ............... 80 FIGURE 5.6 - CYCLE 13 RLP HOT EXCESS REACTIVITY ................................................................ 80 FIGURE 5.7 - CYCLE 13 RLP COLD SHUTDOWN MARGIN .............................................................. 8 1 FIGURE 5.8 - CYCLE 13 RLP STANDBY LIQUID CONTROL SYSTEM SHUTDOWN MARGIN ............. 81 FIGURE 6.1 - REFERENCE CORE LOADING PATTERN- CYCLE 12 .................................................. 88 FIGURE 6.2 - REFERENCE CORE LOADING PATTERN- CYCLE 13 .................................................. 89 Viii

PSEG Hope Creek Mixed Core Analysis Report 1.0 Introduction and Summary The implementation of a new fuel design for a General Electric (GE) Boiling Water Reactor (BWR) follows a two-step process. First, the new fuel design is submitted to and approved by the Nuclear Regulatory Commission (NRC) (( {3))) via the GESTAR II Amendment 22 process. Then, plant-specific analyses are performed to justify use of the new fuel design in an upcoming plant reload. The (( {f3)I analyses consist of one-time (( (3))) analyses and (( (3))) analyses. The ((

(3))) analyses have been performed to support introduction of the GE14 fuel design at Hope Creek Generating Station (HCGS) for the Current Licensed Thermal Power of 3339 MWt. The (( (3))) analyses are performed for each reload regardless of fuel design.

HCGS will be loading GE14 fuel for Cycle 13 operation. Currently, the plant is operating with non-GE14 fuel assemblies (SVEA 96+) in the core. (( (3)) analyses have been performed and documented in the Fuel Transition Report for Hope Creek Generating Station.[11 This report summarizes the results of the (( 3J] analyses and evaluations for the HCGS Cycle 13 mixed core of GE14 and SVEA 96+ fuel. The Cycle 13 mixed core will consist of approximately 20% GE14 and 80% SVEA 96+ fuel. The cycle dependent analyses are documented in the plant and cycle unique Supplemental Reload Licensing Report (SRLR),

which is included in this report as Section 7.0. The following information is provided in the SRLR:

Plant-unique Items

• Reload Fuel Bundles Reference Core Loading Pattern

• Calculated Core Effective Multiplication and Control System Worth Standby Liquid Control System Shutdown Capability

• Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Selected Margin Improvement Options

• Operating Flexibility Options

• Core-wide AOO Analysis Results

• Local Rod Withdrawal Error AOO Summary Cycle MCPR Values

• Overpressurization Analysis Summary

• Loading Error Results

• Control Rod Drop Analysis

• Stability Analysis Results

• Loss-of-Coolant Accident Results I

PSEG Hope Creek Mixed Core Analysis Report In addition to the SRLR, this report also presents the following information that supports the analyses:

• Lattice Physics Benchmark and Results

• Cycle 13 Base Point Determination and Previous Operating Cycle Benchmark Results

• Fuel Rod Thermal-Mechanical Performance Limits for SVEA 96+ Fuel

• Cycle 13 Mixed Core Reload Bundle Design, Core Design and Performance Summary

• Cycle 13 Safety Limit Minimum Critical Power Ratio Summary The conclusion of the lattice physics benchmark and results evaluation is that the TGBLA06/PANCI I models are acceptable to establish the design and licensing parameters for SVEA 96+.

The benchmark of previous operating cycles with GNF methods has been utilized to determine appropriate hot and cold eigenvalues and thermal limit design margins for the Cycle 13 core design work as well as resulting in the establishment of the Cycle 13 base point.

The fuel rod thermal-mechanical performance limits for SVEA 96+ have been established.

Demonstration that individual bundle and core designs meet these performance limits ensures compliance with the fuel rod thermal-mechanical design and licensing limits.

Cycle 13 mixed core reload bundle and core design has been completed. As indicated by the performance summary, all core operating and design margins have been dispositioned to be acceptable based on the Cycle 13 reload bundle and core design.

The Cycle 13 SLMCPR calculations, including a comparison to the SLMCPR calculated for Cycle 12 using GNF methods, have been completed. The calculated Cycle 13 SLMCPR values of 1.06 for dual loop operation and 1.08 for single loop operation are appropriate for the Hope Creek Cycle 13 mixed core.

The results presented in the SRLR have been determined using NRC approved methods in accordance with the basis provided in GeneralElectric StanidardApplication forReactorFuel, NEDE-2401 l-P-A-14, June 2000 and the U. S. Supplement, NEDE-2401 1-P-A-14-US, June 2000. The results of the analyses and evaluations contained in the SRLR support the conclusion that HCGS can safely load and operate using GE14 fuel with SVEA 96+ fuel in HCGS Cycle 13.

1.1 References

1. Fuel TransitionReport ForHope Creek GeneratingStatic,:,

NEDC-33158P, Revision 4, March 2005.

2

PSEG Hope Creek Mixed Core Analysis Report 2.0 Lattice Physics Comparison GNF methods, namely TGBLA04/PANAClO and TGBLA06/PANACI 1, have been used to design and license GNF/GE bundle designs for 8x8, 9x9 and lOxlo lattices with and without water rods. Currently they are being used to license the GE12 and GE]4 designs. No changes to the GNF design system have been made to adapt to the SVEA 96+ geometry. The purpose of this analysis is to document the accuracy of TGBLA06/PANACI I for the SVEA 96+ application in Hope Creek Generating Station Cycle 13. Extensive comparisons between TGBLA06 and the more accurate benchmark Monte Carlo code MCNP show that the accuracy of TGBLA06/PANAC1 1 for the SVEA 96+ designs is equivalent to the accuracy for GE12/GE14 designs. Therefore, the TGBLA06/PANACI I models are acceptable to establish the design and licensing parameters for the SVEA 96+ fuel designs in Hope Creek Generating Station Cycle 13 and for all future Hope Creek Generating Station cycles that exhibit design characteristics consistent with the benchmark bases in which these SVEA 96+ fuel designs are utilized.

2.1 TGBLA Lattice Physics to Monte Carlo Comparison 2.1.1 Water-Cross Models

((I

{3)}]

2.1.2 Qualification Method In this study, the benchmark model is the MCNP Monte Carlo neutron transport program. The MCNP program is a Monte Carlo neutron transport code developed at Los Alamos National Laboratory. The cross sections used in MCNP are derived from the ENDF/B-V data and are represented on a continuous energy mesh. A full scattering model developed for water and other scattering material is employed in the thermal energy region. The MCNP program is widely used as a nuclear benchmark tool throughout the world. GNF/GE was instrumental in formulating the original qualification results forBWR applications. GNF/GE has qualified MCNP against critical data. For BWR applications, the most important critical data are the TRX and B&W critical experiments Ill because they consist of U0 2 fuel in water moderated fuel pins.

Table 2.1 contains a summary of the critical eigenvalues obtained by MCNP for these experiments. Note that the calculated criticality for these experiments is quite uniform regardless of the fuel type, uranium metal or uranium oxide. Hence, the MCNP program can be used to determine the accuracy of the design tool TGBLA for BWR fuel applications.

3

-- PSEG Hope Creek Mixed Core Analysis Report Table 2.1 - Summary of MCNP Simulation of TRX and B&W Critical Experiments Experiment Description Eig envaluie Uranium metal in Al clad 1.3% enriched Triangular pitch lattice 1.0008 +/- .0013 Pitch = 1.086 cm Uranium metal in Al clad TRX-2 1.3% enriched 0.9997 +/-.0013 TRX-2 ~Triangular pitch lattice 099 01 Pitch = 2.174 cm U0 2 in Al clad B&W ~~Square pitch lattice 099+/-01 B&W 2.5% enriched 0.9995+ .0014 Pitch =1.626 cm The MCNP results can be used to compare both criticality and pin power distributions. In the past, the Monte Carlo comparisons have been restricted to beginning of life configurations. In this study, the comparisons have been extended to lattices at various stages of burn-up. The TGBLA06 code is used to establish the isotopic inventory at a number of exposure points. This isotopic inventory includes the depleted values of uranium and gadolinium, as well as the amounts of plutonium and fission products generated during the bum-up process. The isotopic inventory is then input to the MCNP code and the eigenvalue and power distribution determined.

These TGBLA06/MCNP comparisons have been carried out for the SVEA 96+ designs for three void values, 0%, 40% and 70%. Similar comparisons have been made for a conventional U0 2 design to determine the difference in model accuracy between GE14 fuel and SVEA 96+ fuel.

2.1.3 TGBLAO06MCNP Comparison Results To perform a review of the ability of TGBLA06 to model a new fuel design, three figures of merit, the infinite lattice k,,, the pin fission density, and a lattice dynamic void coefficient, have been chosen. These figures of merit provide screening functions such that the lattice average reactivity characteristics and the individual pin power generation can be assessed, and the lattice transient response to moderator density changes (voids) can be reviewed. Through a use of these global figures of merit, comparisons of the TGBLA06 analysis and MCNP analysis can be used to gain confidence in the TGBLA06 lattice physics solution.

These figures of merit and associated criteria have been applied to all fuel designs within the application range of TGBLA06. The current application range includes several geometric configurations of 8x8, 9x9 and lOxlO fuel designs.

For screening purposes, the agreement for the infinite lattice ko, between TGBLA06 and MCNP for uncontrolled conditions is expected to be within +/- 1% Ak. For hot controlled and cold conditions, the expected agreement is to be within +/- 1.5% Ak;. The RMS (root mean square) of 4

PSEG Hope Creek Mixed Core Analysis Report the pin fission density differences of all powered pins in the hot uncontrolled condition is expected to be less than 3%. The RMS of the fission density differences for the hot controlled and cold conditions are reviewed for reasonable agreement but may exceed 3% in some lattice designs. The expected agreement of the Dynamic Void Coefficient difference between TGBLA06 and MCNP is a +2% bias and a standard deviation of 8.0%.

Comparisons between the design model TGBLA06 and the Monte Carlo model MCNP have been carried out at beginning of life configurations and at exposed configurations. A summary of the TGBLA06/MCNP koo comparisons can be found in Tables 2.2 and 2.3. In these tables, the K1o values are compared for three void points, 0%, 40% and 70% voids. The percent difference in these tables is defined as I0O*(MCNP-TGBLA)/MCNP.

Tables 2.2 and 2.3 also give comparisons for a dynamic void coefficient. Given the KI, differences, an estimate can be made of the accuracy of the lattice void coefficient generated by TGBLA06. The lattice dynamic void coefficient is defined as:

dynamic void coeff = - Psffo, av 0.4 The derivative above can be calculated by fitting k. for the three void points as a second order polynomial in v, differentiating, and evaluating at v = 0.4. The result is:

04 3k (0.0) I 4k (0.7) 1 dynamic void coeff=--- o _-+ X l Peff 2.8k (0.4) 1.2 2.1k} (0.4) 5

PSEG Hope Creek Mixed Core Analysis Report Table 2.2 - Hot Uncontrolled Beginning of Life k, and Dynamic Void Coefficient Comparisons SVEA 96+ and GE14 Lattices

((

(3)))

6

PSEG Hope Creek Mixed Core Analysis Report Table 2.2 - Hot Uncontrolled Beginning of Life k. and Dynamic Void Coefficient Comparisons SVEA 96+ and GE14 Lattices (3)))

7

PSEG Hope Creek Mixed Core Analysis Report Table 2.3 - Hot Uncontrolled k. and Dynamic Void Coefficient Comparisons for Exposed Conditions SVEA 96+ Lattice 6026

((

{3)))

8

PSEG Hope Creek Mixed Core Analysis Report Table 2.3 - Hot Uncontrolled k. and Dynamic Void Coefficient Comparisons for Exposed Conditions SVEA 96+ Lattice 6026 (3)))

The results in Tables 2.2 and 2.3 show that the agreement between the design model TGBLA06 and MCNP is consistent with this expectation. A difference of (( (3j)) at 4.0 GWd/ST is above the two-sigma level of the expected results but is similar to differences seen in previous studies for GNF fuel and for the GE14 GNF fuel presented. The maximum difference of

(( (3)) at 25.0 GWd/ST is a result of differences between small void coefficients.

The analysis for SVEA 96+ Lattice 6026 in Table 2.3 shows that the dynamic void coefficient agreement between TGBLA and MCNP improves for exposures greater than 4.0 GWd/ST and is significantly within the two-sigma level at exposures greater than 8.0 GWd/ST.

In general, the Dynamic Void Coefficient comparison shows good agreement between MCNP and TGBLA06 and is within expected bounds except as described in the previous paragraph.

From these results, it is concluded that TGBLA06 can be used to model the SVEA 96+ designs present in the Hope Creek Generating Station core.

Figures 2.1 through 2.8 show Beginning of Life (BOL) TGBLA06/MCNP k. comparisons for cold, 0% void hot, 40% void hot, 70% void hot and borated cases for SVEA 96+ and GE14 lattice configurations.

9

PSEG Hope Creek Mixed Core Analysis Report

{3)))

Figure 2.1 - SVEA 96+ Lattice 6019 TGBLAO6V/MCNP k. Comparison (3)))

Figure 2.2 - SVEA 96+ Lattice 60207TGBLA06V/MCNP k, Comparison 10

PSEG Hope Creek Mixed Core Analysis Report

[3)}))

Figure 2.3 - SVEA 96+ Lattice 6022 TGBLA06V/MCNP k,,, Comparison

[33) ]

Figure 2.4 - SVEA 96+ Lattice 6023 TGBLA06V/MCNP k. Comparison 11

PSEG Hope Creek Mixed Core Analysis Report

[1 (31))

Figure 2.5 - SVEA 96+ Lattice 6024 TGBLA06V/MCNP kX Comparison R[

Figure 2.6 - SVEA 96+ Lattice 6026 TGBLA06V/MCNP k. Comparison 12

PSEG Hope Creek Mixed Core Analysis Report 13P)]

Figure 2.7 - GE14 Lattice 4963 Regular Zone TGBLA06V/MCNP k.,Comparison

[

{3)-

Figure 2.8 - GE14 Lattice 4966 Vanished Zone TGBLAO6VIMCNP k. Comparison 13

-. PSEG Hope Creek Mixed Core Analysis Report The accuracy of pin power distributions can also be determined from Monte Carlo comparisons.

In this case the basis for comparison is the standard deviation of the difference in pin power between TGBLA06 and MCNP. The standard deviation is given by:

S=I IZ (pM -PTj)2 i-I where PM. and pT. are the MCNP and TGBLA06 pin peaking factors for fuel rod j and the J J summation is taken over all fuel rods in the lattice.

In addition to the use of these comparisons as a figure of merit for TGBLA06 range of application review, the values are also utilized as the pin power uncertainty value in the SVEA 96+ GEXL correlation development. The beginning of life standard deviations, s, are summarized in Table 2.4 for the several SVEA 96+ lattices and two GE14 lattices. The standard deviations, s, for exposed condition are summarized in Table 2.5 for SVEA 96+ lattice 6026. The standard deviation in pin power is less for the SVEA 96+ lattices compared to the GE14 lattices. Plots of the standard deviation for uncontrolled, controlled, and borated states as a function of moderator density from the cold to hot, 70% void state are shown in Figures 2.9 through 2.16. The expected results should be less than a two-sigma uncertainty of 2.88%. 121 All evaluations for the SVEA 96+ fuel meet this requirement. The weighted average of 18 beginning of life state points, shown in Table 2.4, and the 21 exposed state points, shown in Table 2.5, was found to be 1.63%. This is above the one sigma value for the fleet average of 1.44% but is consistent with GNF lOxlO products. The major contributors to the higher uncertainty are the fuel rods at location (4,5), (5,4), (6,4), (4,6), (7,5), (5,7), (7,6), and (6,7).

These rods typically show a negative bias of 4-5% (TGBLA is low) as a result of the approximation of the large diamond shaped water mass in the center of the lattice.

14

. PSEG Hope Creek Mixed Core Analysis Report Table 2.4 - Hot Uncontrolled Beginning of Life Pin Power Comparison SVEA 96+ and GE14 Lattices 15

PSEG Hope Creek Mixed Core Analysis Report Figure 2.9 - SVEA 96+ Lattice 6019 TGBLA06VWMCNP Fission Density Comparison (3)))

Figure 2.10 - SVEA 96+ Lattice 6020 TGBLA06V/MCNP Fission Density Comparison 16

PSEG Hope Creek Mixed Core Analysis Report Figure 2.11 - SVEA 96+ Lattice 6022 TGBLA06V/MCNP Fission Density Comparison

{3}))

Figure 2.12 - SVEA 96+ Lattice 6023 TGBLA06V/MCNP Fission Density Comparison 17

PSEG Hope Creek Mixed Core Analysis Report

((

(3)))

Figure 2.13 - SVEA 96+ Lattice 6024 TGBLA06V/MCNP Fission Density Comparison

[I Figure 2.14 - SVEA 96+ Lattice 6026 TGBLA06VIMCNP Fission Density Comparison 18

PSEG Hope Creek Mixed Core Analysis Report (3)

}))

Figure 2.15 - GE14 Lattice 4963 TGBLA06V/MCNP Fission Density Comparison

{3)

A))

Figure 2.16 - GE14 Lattice 4966 TGBLA06V/MCNP Fission Density Comparison 19

PSEG Hope Creek Mixed Core Analysis Report Table 2.5 - Hot Uncontrolled Pin Power Comparison versus Exposure for SVEA 96+ Lattice 6026

[I (31))

2.1.4 Boron Worth Accuracy for Standby Liquid Control Analysis The Standby Liquid Control Analysis uses a statistical uncertainty for the worth of soluble boron in the moderator coolant derived from TGBLA06/MCNP comparisons. Table 2.6 contains the results of this comparison for SVEA 96+ lattices compared in this qualification study.

While this analysis is not used in the SVEA 96+ qualification review of TGBLA06, it is used to provide support information to the statistical uncertainty for Standby Liquid Control Analysis.

Table 2.6 - Cold (160 0C) Uncontrolled Borated Beginning of Life k1 Comparison SVEA 96+ and GE14 Lattices 20

PSEG Hope Creek Mixed Core Analysis Report Table 2.6 - Cold (1600C) Uncontrolled Borated Beginning of Life k, Comparison SVEA 96+ and GE14 Lattices (3)jj 2.1.5 3D Simulator Accuracy The 3D simulator PANAC1 1 receives the cross section input form TGBLA06, which has been shown to be equally accurate for the SVEA 96+ fuel and GE12/GE14 fuel. Therefore no changes in reactivity and power distribution accuracy are expected when SVEA 96+ bundles are introduced.

21

PSEG Hope Creek Mixed Core Analysis Report 2.1.6 References

1. MCNP: Light WaterReactor CriticalBenchmarks,NEDO-32028, March 1992.

2. Methodology and UncertaintiesforSafety Linit MCPR ELvaluations, NEDC-32601P-A, August 1999.

2.2 Lattice Physics Results 2.2.1 Description This section provides lattice physics results forrepresentative SVEA 96+ and GE14 lattice designs. The GE14 design is the bottom-most enriched lattice in the bundle that was developed to support the reference core loading pattern analyses for the MCAR.

Note: Table 2.7 shows a tabulation of the data that is presented for each lattice in this Section.

Table 2.7 - Representative Lattice Evaluation Design BWREDB BWREDB Variable Condition Voids Exposures

_____ Bundle # Lattice #________(GN~d/ST)

SVEA 96+ 2657 6022 k HOT, Uncontrolled All All SVEA 96+ 2657 6022 k COLD, All All Uncontrolled SVEA 96+ 2657 6022 2-D Local Peaking HOT, Uncontrolled 40% 0, 5, 10 GE14 2695 6203 kl HOT, Uncontrolled All All GE14 2695 6203 k. COLD, All All Uncontrolled GE14 2695 6203 2-D Local Peaking HOT, Uncontrolled 40% 0, 5,10 The Bundle and Lattice descriptions for the SVEA 96+ design and the GE14 design used in this study are presented in Figures 2.17 and 2.19, respectively.

The SVEA 96+ Hot Uncontrolled and Cold Uncontrolled k. is presented in Table 2.8. The SVEA 96+ power peaking data is shown in Figure 2.18.

The GE14 Hot Uncontrolled and Cold Uncontrolled kI, is presented in Table 2.9. The GE14 power peaking data is shown in Figure 2.20.

22

PSEG Hope Creek Mixed Core Analysis Report (3) jj Figure 2.17 - SVEA 96+ Bundle 2657 Configuration 23

PSEG Hope Creek Mixed Core Analysis Report Table 2.8 - SVEA 96+ Bundle 2657 k.

(3)))

24

PSEG Hope Creek Mixed Core Analysis Report

{3)))

Figure 2.18 - SVEA 96+ Bundle 2657 2D Dominant Lattice Power Peaking 25

PSEG.Hope Creek Mixed Core Analysis Report (I

Figure 2.19 - GE14 Bundle 2695 Configuration 26

PSEG Hope Creek Mixed Core Analysis Report Table 2.9 - GE14 Bundle 2695 km

(( (3)]

27

PSEG Hope Creek Mixed Core Analysis Report

((

Figure 2.20 - GE14 Bundle 2695 2D Dominant Lattice Power Peaking 28

PSEG Hope Creek Mixed Core Analysis Report 3.0 Cycle 13 Base Point Determination - Cycle 9-12 Simulation The operating history of the Hope Creek reactor has been tracked by the 3D simulator (PANACI 1). The results of this tracking are used to determine appropriate hot and cold eigenvalues for core design work as well as to evaluate thermal margin biases, which may exist between the simulator and the process computer. The tracking simulations also provide the base point (starting point) for core design work for Cycle 13.

3.1 Cycle 9-12 Performance Tracking This section contains several figures summarizing the results of the core tracking for Cycle 9, which was the last full loading of GE fuel, and Cycles 10, 11 and 12 where SVEA 96+ fuel was loaded. Figures 3.1 and 3.2 summarize the hot and cold eigenvalues for all these cycles. Also included are eigenvalues used for the reference fuel cycle (RFC) design of Cycle 13. Figures 3.3, 3.7 and 3.11 show hot eigenvalues for the individual cycles.

The hot eigenvalue selected as Cycle 13 design basis is based on a combination of the data for previous cycles at Hope Creek as well as GNF's methods experience with similar size and power BWVRs. The eigenvalue data for Cycles 9 through 12 is well behaved and relatively tightly packed. GNF would expect the eigenvalue to behave as shown by the "GE14 Equilibrium" curve as the fraction of GE14 fuel is increased in future cycles.

The cold eigenvalue selected as the Cycle 13 design basis is again based on a combination of cold critical measurements in the previous cycles as well as GNF's method experience with its BWR fleet. Generally the cold eigenvalue basis is selected so as to conservatively bound the measured data rather than fit through the data as with the hot eigenvalue.

Figures 3.4, 3.8 and 3.12 show the evaluation of MFLCPR for each of the cycles. The simulated results are compared to measured results from the process computer. Figures 3.5, 3.9 and 3.13 show the evaluation of MFLPD for each of the cycles and comparison to the process computer.

Figures 3.6, 3.10 and 3.14 show the evaluation of MAPRAT for each of the cycles and comparison to the process computer.

The process computer data comparisons of MFLCPR, MFLPD and MAPRAT are the basis for selecting the design margins shown in Table 5.2.

29

PSEG Hope Creek Mixed Core Analysis Report

((

(3)))

Figure 3.1 - Cycle 9 - Cycle 12 Hot Critical Eigenvalue Tracking 30

I PSEG Hope Creek Mixed Core Analysis Report

{3)))

Figure 3.2 - Cycle 9 - Cycle 12 Cold Critical Eigenvalue Tracking 31

PSEG Hope Creek Mixed Core Analysis Report Figure 3.3 - Cycle 10 Hot Critical Eigenvalue (3)))

Figure 3.4 - Cycle 10 MFLCPR 32

PSEG PSEG Hope Creek Mixed Core Analysis Report

((

13) ))

Figure 3.5 - Cycle 10 MFLPD (3)))

Figure 3.6 - Cycle 10 MAPRAT 33

PSEG Hope Creek Mixed Core Analysis Report Figure 3.7 - Cycle 11 Hot Critical Eigenvalue

((

{3) ii Figure 3.8 - Cycle 11 MFLCPR 34

PSEG Hope Creek Mixed Core Analysis Report (3)))

Figure 3.9 - Cycle 11 MFLPD

{3)))

Figure 3.10 - Cycle 11 MAPRAT 35

- PSEG Hope Creek Mixed Core Analysis Report 13)))

Figure 3.11 - Cycle 12 Hot Critical Eigenvalue Figure 3.12 - Cycle 12 MFLCPR 36

..PSEG Hope Creek Mixed Core Analysis Report

{3ijj Figure 3.13 - Cycle 12 MFLPD Figure 3.14 - Cycle 12 MAPRAT 37

PSEG Hope Creek -

Mixed Core Analysis Report 3.2 3D Process Computer Transverse In-core Probe (TIP) Comparisons Comparisons between measured TIPs and predicted TIP responses provide a benchmark of bundle and nodal power distribution capability. A comparison of radial (bundle) TIPs are directly proportional to accuracy of bundle powers used for determination of core MCPR. A comparison of nodal TIPs are directly proportional to the accuracy of nodal powers used for determination of core MAPRAT and MFLPD.

ForHope Creek, summary statistics are presented in section 3.2.1. Section 3.2.2 contains graphical summaries of root mean square (RMS) differences between predicted and measured TIPs for both bundle average behavior and nodal (3D) performance. As further evidence, section 3.2.3 provides string-by-string performance for BOC, MOC, and near EOC for Cycle 11. These figures demonstrate that both core wide radial and axial behavior are captured as well as individual string behavior. These TIP comparisons for Hope Creek demonstrate that the GNF methodology is capable of predicting the mixed core environment well. Additionally, both the summary statistics for all cycles and trends within cycles do not present a departure from the GNF experience base.

3.2.1 Statistics Summary The average over the cycles 9-12 for bundle RMS is (( 3)3)) and nodal RMS

(( 3)) . The standard deviations are (( 13))) and (( (3))) , respectively.

For cycle 11 specifically, the numbers are essentially the same. Overall, these numbers are not unreasonable for a gamma TIP plant. Bundle RMSs < (( {3})) and nodal RMSs < ((

131)) are generally exceptionally good. The Bundle RMSs shown in Table 3.1 demonstrate that the use of a (( 13))) integrated effective TIP reading in the SLMCPR calculation (Table 6.2) is fully applicable.

A plot of the RMS values for the bundle data is provided in Figure 3.15.

Table 3.1 - RMS Values 38

PSEG Hope Creek Mixed Core Analysis Report 3.2.2 Plots of Bundle and Nodal TIP Comparisons

((

Figure 3.15 - Plot Of TIP Nodal RMS % Versus Cycle Exposure GWdIST For Cycles 9-12 39

PSEG Hope Creek Mixed Core Analysis Report

[31]

Figure 3.16 - Plot Of TIP Bundle RMS % Versus Cycle Exposure GWd/ST For Cycles 9-12 3.2.3 Comparison of Axial TIP plots for Cycle 11 Axial TIP plots are shown in the following figures for three selected exposure points for Cycle 11 near DOC, MOC, and EOC, respectively. In the figures below, PCTIP is the process computer TIP readings and CALTIP is the PANACI I calculated TIP readings. The exposure shown is in MWd/ST.

40

PSEG Hope Creek Mixed Core Analysis Report II (( 13)))

Figure 3.17 - Plot Of Axial Tip Comparison For Cycle 11 At A Selected Exposure Point Near BOC 41

PSEG Hope Creek Mixed Core Analysis Report 13)))

Figure 3.18 - Plot Of Core Average Axial TIP Comparison For Cycle 11 At A Selected Exposure Point Near BOC 42

PSEG Hope Creek Mixed Core Analysis Report

(( {3]3 Figure 3.19 - Plot Of Axial TIP Comparison For Cycle 11 At A Selected Exposure Point Near MOC 43

PSEG Hope Creek Mixed Core Analysis Report

((

Figure 3.20 - Plot Of Core Average Axial TIP Comparison For Cycle 11 At A Selected Exposure Point Near MOC 44

PSEG Hope Creek Mixed Core Analysis Report 11 (3)]

Figure 3.21 - Plot Of Axial TIP Comparison For Cycle 11 At A Selected Exposure Point Near EOC 45

PSEG Hope Creek Mixed Core Analysis Report Figure 3.22- Plot Of Core Average Axial TIP Comparison For Cycle 11 At A Selected Exposure Point Near EOC 46

PSEG Hope Creek Mixed Core Analysis Report 4.0 Fuel Rod Thermal-Mechanical Report Description This section of the MCAR documents the fuel rod thermal-mechanical performance limits for the SVEA 96+ fuel design for application in the Hope Creek Generating Station. The performance limits are applied in the nuclear bundle and core design process. Demonstration that individual bundle and core designs meet these performance limits ensures compliance with the fuel rod thermal-mechanical design and licensing limits. The steady-state LHGR performance limits for SVEA 96+ have been supplied by PSEG Nuclear LLC as part of the transfer of information for the SVEA 96+ bases. The steady-state limits are not being replaced with steady-state limits based on GNF thermal-mechanical methodology. The SVEA 96+ steady-state LHGR performance limits are being used to define the initial conditions for the subsequent Anticipated Operational Occurrence (AOO) evaluations. These initial conditions are described in Section 4.1. The AOO limits for the SVEA 96+ fuel design, as specified in Section 4.2, have been determined using the GNF thermal-mechanical methodology using the limits provided in Section 4.1 as starting points for the AOO limit evaluation. The GNF thermal-mechanical methodology has been applied to determine the allowable overpowers during an AOO that assure pellet centerline melting will not occur and the cladding strain will not exceed the 1% circumferential plastic strain criterion.

The fuel rod thermal-mechanical performance limits for GE14C are documented in the GNF Design Basis documents.

4.1 Fuel Rod Thermal-Mechanical Performance Limits for Steady-State Operation

• Maximum Steady-State Linear Heat Generation Rate [(3

• Maximum Peak Pellet Exposure (3))]

• Maximum Operating Time (( (31))

4.1.1 UO2 Fuel Rod Steady-State Limit versus Exposure The maximum peak pellet power allowable for a U0 2 fuel rod at a given U0 2 rod peak pellet exposure can be calculated from Table 4.1.

4.1.2 Gd20 3 Fuel Rod Steady-State Limit versus Exposure For the purposes of the AOO evaluations described in Section 4.2, gadolinia bearing fuel rods are assumed to have the same steady-state allowable limits as the U0 2 fuel rods (see Tables 4.2 and 4.3). This is a conservative assumption for the initial condition relative to AOO evaluations.

4.2 Fuel Rod Thermal-Mechanical Limits for Anticipated Operational Occurrences The purpose of this section is to present the criteria to be applied in the core design process to ensure consistency with the General Electric fuel rod thermal-mechanical design and licensing basis with respect to Anticipated Operational Occurrences (AOOs).

47

PSEG Hope Creek Mixed Core Analysis Report 4.2.1 Design and Licensing Limits 4.2.1.1 Overheating of Fuel Pellets The fuel rod is evaluated to ensure that fuel rod failure due to fuel melting will not occur.

Evaluations are performed for whole core AQOs to ensure that fuel melting does not occur. For local AOs, such as the Rod Withdrawal Error, a small amount of calculated fuel pellet centerline melting may occur, but the event is limited by the 1% cladding circumferential plastic strain criterion.

4.2.1.2 Pellet Cladding Interaction The fuel rod is evaluated to ensure that fuel rod failure due to pellet-cladding mechanical interaction will not occur. Evaluations are performed for the limiting AQOs to ensure that the circumferential cladding plastic strain during the event does not exceed 1%.

4.2.1.3 Cumulative Performance Limits Other fuel performance considerations are included in fuel rod design and licensing analyses to address the cumulative effects (e.g., fatigue) of AOOs and other reactor operational behavior.

No specific constraints on core design are required to ensure consistency with this cumulative effects design and licensing basis, because the input to these evaluations are based on periodically updated actual operational experience of all General Electric BWR/2-6 reactors.

4.2.2 Thermal and Mechanical Overpowers 4.2.2.1 Thermal Overpower The thermal overpower is used to evaluate the potential for the fuel entering the molten state at the fuel centerline. Temperature at the fuel centerline is proportional to either the fuel rod linear power or the fuel rod surface heat flux, so the magnitude of these quantities reached during the AOO are the parameters of interest. The measurement of an AOO and its approach to fuel centerline melting also depends on the nuclear methods used for evaluation of the transient.

Three methods are used. They are:

1. Point model. The core is represented by a point model and all changes in power are assumed to be the same percentage at all locations in the core. The current point model used by General Electric is the REDY model.

2. 1-D model. The core is represented through a one-dimensional model in the axial dimension. The core power distribution in the radial or r-O plane is collapsed at each axial node. All fuel bundles are therefore assumed to experience the same percentage change in power at the same axial elevation. The current l-D model used by General Electric is the ODYN model.

3. 3-D model. The core is represented by a model of each fuel bundle, which are all represented with a number of axial nodes. Changes in power are calculated for each 48

PSEG Hope Creek Mixed Core Analysis Report axial node of each individual fuel bundle throughout the core. The current 3-D model used by General Electric is the PANACEA model.

For use with the point model or the l-D model, the thermal overpower is defined as:

-XIOO p[ pTh pTh o (4.1) where:

OpTh = The thermal overpower for a particular fuel design during an AOO, %. A fuel design is the quantity loaded in the core that has the same fuel rod thermal-mechanical limits for both steady-state operation and AOs .

plyh = The maximum steady-state heat flux in the fuel bundle of a particular fuel design prior to the event.

pmTh = The maximum heat flux in a fuel bundle of the same fuel design during the event. This may occur at a different axial node than Pos when evaluated based on the 1-D model results.

49

v PSEG Hope Creek Mixed Core Analysis Report A graph of this thermal overpower is shown in Figure 4.1.

piT K- MezX I-I rd2 i.,

Axial Location Figure 4.1 - Graph of Thermal and Mechanical Overpowers 50

PSEG Hope Creek Mixed Core Analysis Report For use with the 3-D model, the thermal overpower is defined as:

[ ]MFLPD0

] (4.2) where:

OpTh = The thermal overpower for a particular fuel design during an AOO, %. A fuel design is the quantity of fuel loaded in the core which has the same fuel rod thermal-mechanical limits for both steady-state operation and AQOs and the OPTh is the maximum value for the event considering all fuel bundles of that design present in the core.

MFLPDTh = The maximum fraction of linear power density (MELPD), relative to the steady-state thermal-mechanical limit prior to the event for any fuel bundle in the core which is located no more than two (2) positions away from the fuel bundle for which the OpTh is being calculated.

MFLPDn = The maximum fraction of linear power density relative to the steady-state thermal-mechanical limit during the event for any fuel bundle in the core.

Alternately, for use with the 3-D model, the thermal overpower can also be defined as:

OpTh =[ MAP R.TTjh}° X100 (4.3) where:

opTh The thermal overpower for a particular fuel design during an AOO, %. A fuel design is the quantity of fuel loaded in the core which has the same fuel rod thermal-mechanical limits for both steady-state operation and AQOs and the OPTh is the maximum value for the event considering all fuel bundles of that design present in the core.

MAPRATon= The maximum ratio of the average planar linear heat generation rate (APLHGR) relative to the APLHGR limit (MAPLHGR) prior to the event for any fuel bundle in the core which is located no more than two (2) positions away from the fuel bundle for which the OPTh is being calculated.

MAPRA¶1 Th = The maximum ratio of the APLHGR relative to the APLHGR limit (MAPLHGR) during the event for any fuel bundle in the core.

51

PSEG Hope Creek Mixed Core Analysis Report Equations (4.2) and (4.3) differ in that equation (4.2) will include any differences and changes in the local power and exposure peaking in the nodes being evaluated as compared with that used for establishing the MAPLHGR.

4.2.2.2 Mechanical Overpower The mechanical overpower is used to evaluate the potential for overstraining of the cladding.

The incremental cladding strain during an AOO is proportional to the change in fuel volume average temperature, which is proportional to the change in either the fuel rod linear power or the fuel rod surface heat flux at a particular axial location or cross-section of the fuel rod. This overpower is therefore evaluated based on the change in heat flux at a specific fuel rod axial location. As with the thermal overpower, the fuel volume average temperature change is evaluated differently depending on the nuclear methods used to evaluate the transient.

For use with the point model or the l-D model, the mechanical overpower is therefore defined as:

PMe _PMC op m e I ] X100 (4.4) where:

oPme = The mechanical overpower for a particular fuel design during an AOO, %.

A fuel design is the quantity of fuel loaded in the core, which has the same fuel rod thermal-mechanical limits for both steady-state operation and AOOs.

Pal = The steady-state heat flux prior to the event in the fuel rod of a particular fuel design for the axial location that experiences the largest heat flux increase during the event. A fuel design is the quantity of fuel loaded in the core, which has the same fuel rod thermal-mechanical limits for both steady-state operation and AQOs . The largest surface heat flux increase at a particular node in a fuel rod is measured as the absolute value for the magnitude of the surface heat flux change at that node and not as a percentage value.

plM. = The maximum heat flux reached during the event in the same fuel rod at the same axial location.

p~ax = The maximum steady-state heat flux in the same fuel rod prior to the event. This may occur at a different axial node than pMe and P,' when evaluated based on the 1-D model results.

A graphical illustration of the mechanical overpower is shown in Figure 4.1.

52

'PSEG Hope Creek Mixed Core Analysis Report For use with the 3-D model, the mechanical overpower is defined as:

opM e [FLPDmC-FLPDN' 100 (4.5) where:

OPM = The mechanical overpower for a particular fuel design during an AOO,

%. A fuel design is the quantity of fuel loaded in the core which has the same fuel rod thermal-mechanical limits for both steady-state operation and AQOs and the Opie is the maximum value for the event considering all fuel bundles of that design present in the core.

Evaluation of the peak} power rod in a node can be assumed to bound all rods in that node if the controlled state of the node has not changed during the event.

FLPD' - The fraction of linear power density (FLPD), relative to the steady-state thermal-mechanical limit prior to the event for any fuel bundle and rod at the axial location which experiences the largest increase in fuel rod power during the AOO. The largest power increase at a particular node in a fuel rod is measured as the absolute value for the magnitude of the power change at that node and not as a percentage value.

FLPD" - The fraction of linear power density relative to the steady-state thermal-mechanical limit during the event for any fuel bundle and rod at the axial location which experiences the largest increase in fuel rod power during the AOO for a particular fuel design. This will therefore be for the same fuel bundle and rod and at the same axial node as FLPDo when evaluated based on the 3-D model results.

MFLPD, = The maximum fraction of linear power density (MFLPD), relative to the steady-state thermal-mechanical limit prior to the event for any fuel bundle in the core which is located no more than two (2) positions away from the fuel bundle for which the OPM ' is being calculated.

53

PSEG Hope Creek Mixed Core Analysis Report Alternately, for use with the 3-D model, the mechanical overpower can also be defined as:

RAPLHGRm-APLHGRM ]xl (4.6) where:

OPEN = The mechanical overpower for a particular fuel design during an AOO, %. A fuel design is the quantity of fuel loaded in the core which has the same fuel rod thermal-mechanical limits for both steady-state operation and AOOs and the OP' is the maximum value for the event considering all fuel bundles of that design present in the core.

RAPLHGR" e = The ratio of the average planar linear heat generation rate (APLHGR) relative to the APLHGR limit (MAPLHGR limit) prior to the event for any fuel bundle at the axial node which experiences the largest increase in nodal power during the AOO.

The largest power increase at a particular node is measured as the absolute value for the magnitude of the power change at that node and not as a percentage value.

RAPLHGRm 4 ' = The ratio of the APLHGR relative to the APLHGR limit (MAPLHGR limit) during the event for the same fuel bundle and at the same axial node as RAPLHGR'!C.

MAPRATJm, = The maximum ratio of the APLHGR relative to the APLHGR limit (MAPLHGR limit) at the axial node with the largest value for RAPLHGR prior to the event for any fuel bundle in the core which is located no more than two (2) positions away from the fuel bundle for which the OPM' is being calculated.

Equations (4.5) and (4.6) differ in that equation (4.5) will include any differences and changes in the local power and exposure peaking for the rod being evaluated as compared with that used for establishing the MAPLHGR.

54

I PSEG Hope Creek Mixed Core Analysis Report 4.2.3 Limiting Thermal and Mechanical Overpowers for AOOs at Rated Power and Flow 4.2.3.1 Rod Withdrawal Error(RWE)

The U0 2 and U0 2 - Gd2O3 RWE mechanical overpower (MOP, % above the steady-state envelope) is 50% above the steady-state envelope defined in Table 4.1 for U0 2 Rods and 13%

above the steady-state envelope defined in Table 4.3 for Gd Rods, for any axial node of any fuel rod located in the Error Cell. The RIVE mechanical overpower for any axial node of any fuel rod located outside of the Error Cell shall not exceed the value specified in Table 4.4 or Table 4.5. No thermal overpower limit is applied to the RWE.

4.2.3.2 Loss of Feedwater Heater/Inadvertent Actuation of Auxiliary Cold Water Supply Systems (H1CI, HPCS, RCIC)

These events are characteristically similar and are of sufficient duration that the fuel thermal response is steady-state. Therefore, the surface heat flux values from the steady-state or transient BWR Simulator models can be compared to the surface heat flux values determined acceptable by the steady-state fuel rod thernal-mechanical analysis methods. Table 4.4 and 4.5 show the fuel rod thermal and mechanical overpowers corresponding to design and licensing limits.

4.2.3.3 Short Duration Pressurization Transients The remaining AQOs (e.g., Load Rejection with Bypass Failure, Feedwater Controller Failure) occur quite rapidly relative to the fuel thermal time constant such that the calculated surface heat flux is not a valid indicator of the thermal and mechanical consequences of these events. The limits for these short duration events are therefore expressed in terms which can be related to the transient analysis method used for the evaluations of the events.

4.2.3.3.1 Short Duration Pressurization Transients with GE Methods The limits for these short duration events are expressed as an allowable thermal and mechanical overpower based on the surface heat flux values from the BWR transient model. These overpowers are calculated by the transient model based on the definitions for thermal and mechanical overpower presented in Section 4.2.2 and are compared with the limits for these overpowers presented in Tables 4.6 and 4.7.

The short duration pressurization transients are subcategorized according to the nature of the event and the specific analyses performed to evaluate those event types.

55

PSEG Hope Creek Mixed Core Analysis Report 4.2.3.3.1.1 Generator Load Rejection with Bypass Failure (LRNBP) Type This category of pressurization transient events is characterized by a rapid increase in neutron flux followed by a prompt reactor scram resulting in a total event duration of <3 seconds.

Included in this category are the following events:

• Generator Load Rejection with Bypass Failure

• Turbine Trip with Bypass Failure

• Generator Load Rejection

• Turbine Trip

• Loss of Normal Condenser Vacuum

• MSTV Closure-Position Scram

• Inadvertent Closure One MSIV The limits for thermal and mechanical overpower for this type of event are presented in Tables 4.6 and 4.7.

4.23.3.1.2 Feedwater Controller Failure (FWCF) Type This category of pressurization transient events is characterized by a slow (approximately 13-30 seconds) increase in power followed by a rapid (<1.5 seconds) increase in neutron flux and prompt reactor scram. The only event in this category is the feedwater controller failure.

The limits for thermal and mechanical overpower for this type of event are presented in Tables 4.6 and 4.7.

Table 4.1 - U02 Rod BWREDBFUEL Limits

{3)))

56

• PSEG Hope Creek Mixed Core Analysis Report Table 4.2 - U0 2 and (U,Gd)O, Fuel Rod Maximum Power, kW/ft

((

(3)))

57

PSEG Hope Creek Mixed Core Analysis Report Table 4.3 - U02 and (U, Gd)02 Rod/Section BWREDBFUEL Limits c[

To convert linear power (P. k-W/ft) to heat flux at the cladding outer surface (Q/A, Btu/hr-fte):

[lrQ[Btul i -i__ IkW [t 3412.14xl2iFII Btu ][ in II -8 A 2 LPft 7cx 0.3 787 j Lhr-kW]iL n-ft J=34416.2xP (4.7)

Table 4.4 - LFWH, Inadvertent HPCS, HPCI, RCIC Injection, RWE-Outside Error Cell Overpower Limits for U02 Rods Maximum Allowable Surface Heat Flux Increase, %

Thermal I Mechanical 11 (3' I(3)11 For a definition of Thermal and Mechanical Overpowers, please see Sections 4.2.2.1 and 4.2.2.2.

58

PSEG Hope Creek Mixed Core Analysis Report Table 4.5 - LFWH, Inadvertent HPCS, HPCI, RCIC Injection, RWE-Outside Error Cell Overpower Limits for Gd Rods (3)))

Table 4.6 - Overpower Transient Magnitude Guideline Limits for Short Duration Pressurization Transients using Gemini Method for U02 Rods

((

(3)))

59

PSEG Hope Creek Mixed Core Analysis Report Table 4.7 - Overpower Transient Magnitude Guideline Limits for Short Duration Pressurization Transients using Gemini Method for Gd Rods

[

(3)))1 60

PSEG Hope Creek Mixed Core Analysis Report 5.0 GE14 / SV'EA 96+ Demonstration Cycle Analysis Description - Cycle 13 This section of the MCAR provides the results of the Reference Loading Pattern (RLP) core operation simulation of the first reload of GE14 into the Hope Creek Generating Station. The RLP is developed to meet all design bases set for the upcoming cycle (Cycle 13) and is used as the basis for deriving the actual loading pattern. The RLP is the basis for the licensing calculations that are documented in the Supplemental Reload Licensing Report (SRLR) that is reported in Section 7.0 of this report.

5.1 Reload Bundle Design Description The reload bundle nuclear design process is closely coupled with the core nuclear design process in demonstrating compliance with safety and performance criteria. An iterative process was used between bundle design and core design to obtain an optimal balance among performance objectives while satisfying all safety criteria.

This process resulted in a two-stream GE14 reload design strategy using GEI4 bundles with axial and radial isotopic configurations shown in Figures 5.1 and 5.2. The average content and specific distributions of gadolinium and enriched uranium used for the GE14 bundle designs were selected to accomplish the following goals:

1. Meet PSEG specified cycle energy and operating strategy for an 18-month operating cycle. The average enrichment of the fuel bundles was 4.02 wvt% U235. The gadolinium loading of 4.0 and 6.0 wt% Gd 2Q3 was chosen to compensate for the natural decrease in hot excess reactivity of the legacy fuel resulting in a relatively flat overall core hot excess reactivity throughout the majority of the operating cycle and to control radial and axial power shapes without leaving significant amounts of undepleted gadolinium at the end of the cycle.

2. Maintain adequate thermal margins. Lattice enrichment and gadolinium distributions were optimized to obtain desired relative rod-to-rod thermal performance. This included analysis of the local power peaking factors used to calculate linear heat generation rates and the bundle R-factors used to calculate critical power ratios. These parameters were minimized, consistent with other goals, throughout the bundle exposure range associated with expected high power operation for these GE14 designs. Relative powers for gadolinia rods were suppressed to provide adequate margin to meet thermal-mechanical design requirements.

3. Maintain adequate reactivity margins. To demonstrate one stuck rod sub-criticality, design margin to criticality is calculated with the 3D simulator (PANACEA) in conjunction with critical eigenvalue determinations at the reactor during plant startup.

Reactivity control of the fresh fuel is accomplished through the choice of gadolinia design. Cold shutdown margin at beginning of cycle is influenced primarily by the number of gadolinia rods used, while cold shutdown margin later in the cycle is influenced primarily by the concentration of gadolinia used.

61

PSEG Hope Creek Mixed Core Analysis Report

4. Minimize the number and complexity of unique pellet and rod types. The number of pellet and rod types are determined primarily to optimize relative rod-to-rod thermal performance. There are 51 unique (non-symmetric) lattice locations for fueled rods in the GEM4 bundle configuration. The two GEM4 reload bundles both utilize 11 unique pellet types in a total of 15 unique rod types (including two types of tie rods).

5.2 Cycle 13 Core Design Description 5.2.1 Core Configuration Description Changing the design of the fuel utilized in a nuclear power reactor requires a wide range of analyses to support acceptance relative to operational and safety requirements. The purpose of the core design analysis is to demonstrate feasibility of operation, assure compliance with safety limits and provide operating state points for further safety analyses.

Hot operating analyses with projected control rod patterns were performed at different burn-up points through Cycle 13 to demonstrate that the specified operating strategies can be supported and that all operating limits can be satisfied. These analysis conditions also provide the beginning state points for other safety analyses. Cold shutdown calculations have been performed throughout the cycle to demonstrate compliance with the stuck control rod criteria.

5.2.2 Design Limits and Targets The target core flow range is 98.0 - 103.0% rated flow. The critical ker design target for hot, rated operation is shown in Figure 3.1. The distributed critical krf design target for cold shutdown evaluations is shown in Figure 3.2. Core design limits are provided in Table 5.1 and parameters for tracking the core design limits are provided in Table 5.2.

The cold critical kff values are based on the local, cold, critical kfr predicted for Cycle 13 operation. The local cold critical kdrr= (distributed cold critical ffr) - 0.003, where the distributed cold, critical kff are based on observed plant data from in-sequence cold critical cases.

MCPR margin is tracked via the parameter MFLCPR; MLHGR (pellet power margin) is tracked via the parameter MFLPD; and, nodal power margin is tracked via the parameter MAPRAT, where:

MFLCPR = MCPR Operating Limit (5.1)

WLCPR= MCPR 51 MFLPD= Peak LHGR (5.2)

LHGR Operating Limit MAPRAT = Maximum Average Planar LHGR (5.3)

MAPLHGR Operating Limit 62

PSEG Hope Creek Mixed Core Analysis Report 5.3 Cycle 13 Performance Summary The resultant Cycle 13 RLP for the upper left quarter core loading configuration is provided in Figure 5.3.

• The table below Figure 5.3 lists all fuel types and how many of each type are included in the Cycle 13 core configuration.

Table 5.3 compares the calculated thermal limit core performance parameters to the Table 5.2 thermal limit design margin targets. Table 5.4 provides hot excess reactivity vs. cycle exposure.

Tables 5.5 and 5.6 compare cold shutdown and standby liquid control system (SLCS) reactivity performance parameters, respectively, to the Table 5.2 reactivity limit design margin targets.

Figure 5.4 provides the Cycle 13 core control blade configuration for the upper left quadrantb, calculated thermal marginsc and kfreigenvalue as a function of cycle exposure. Figure 5.5 plots the thermal limit parameters vs. cycle exposure. Figure 5.6 plots core hot excess reactivity vs.

cycle exposure. Figures 5.7 and 5.8 plot cold shutdown and SLCS reactivity margins, respectively, versus cycle exposure.

As is seen in the above referenced tables and figures, all core operating and design margins are met by the Cycle 13 RLP, except for MFLPD at the beginning of cycle (BOC). The MFLPD exception at BOC has been dispositioned to be acceptable based on the MFLPD comparisons at BOC shown in Section 3.1.

a'he RLP was evaluated on a full-core basis.

bAll control blade patterns are quartcr-core mirror symmetric.

c Mininum margin in full-corc reported.

63

PSEG Hope Creek Mixed Core Analysis Report Table 6.1 - Core Design Limits Minimum Critical Power Ratio (MCPR) - Design GE14 operating limit for RLP core design (Actual operating (( 3))) BOC to 8022 MWd/ST limits as determined by reload analyses are presented (( (311 after 8022 MWd/ST in Section 7.0) SVEA 96+

(( {] BOC to 8022 MWd/ST

[F '3]1i after 8022 MVd/ST Maximum Lincar Heat Gencration Ratc (MLHGR) Fucl Dcpendcnt Limit in kW/ft 13)]I kV/ft (GE 14)

((

13 kW/ft (SVEA 96+)

Cold Shutdown Margin - One Stuck Control Rod 1.0% Al Boron Injection Shutdown Margin 1.0% Ak Peak Pellet Exposure (( Il)) GWd/MTU (GE14)

(( ~13I] GWd/MTU (SVEA 96+)

Table 5.2 - Core Design Margin Targets MFLCPR 0.93 MFLPD 0.85 MAPRAT 0.89 Cold Shutdown Margin - One Stuck Control Rod 1.3% Ak Boron Injection Shutdown Margin 1.0% Ak Peakl Pellet Exposure (( 13]1 GWd/MTU (GE14) ffr '1)II GWd/MTU (SVEA 96+)

64

PSEG Hope Creek Mixed Core Analysis Report Table 5.3 - Cycle 13 RLP Summary of Rod Pattern Results 65

PSEG Hope Creek Mixed Core Analysis Report Table 5.4 - Cycle 13 RLP Hot Excess Reactivity

[I (3)))

66

PSEG Hope Creek Mixed Core Analysis Report Table 5.5 - Cycle 13 RLP Cold Shutdown Reactivity Margin CARI AND SDM RESULTS

• CASE CONVERGENCE: PASSED DESIGN CRITERIA: MET (3)))

67

PSEG Hope Creek Mixed Core Analysis Report Table 5.6 - Cycle 13 RLP Standby Liquid Control Shutdown Margin SLCS ANALYSIS - PANACEA SLCS RESULTS PLANT NAME  : HOPE CREEK I EIS CODE  : KTI CYCLE NUMBER :13 PANACEA VERSION: PANAC11V NITER: 15 ANALYSIS TEMP: 160. C ANALYSIS BORON: 726. PPM SLCS SDM REQUIREMENT DETERMINATION:

IAT NO. BUNDLES INTHE PRODUCT LINE SDM REQ NO. CORE 1 89 SVEA 96+ 0.010 2 38 SVEA96+ 0.010 3 166 SVEA96+ 0.010 4 69 SVEA96+ 0.010 5 164 SVEA 96+ 0.010 6 62 SVEA 96+ 0.010 7 56 GE14C 0.010 8 108 GE14C 0.010 9 2 SVEA96+ 0.010 10 2 SVEA 96+ 0.010 11 2 SVEA96+ 0.01D 12 4 SVEA 96+ 0.01D 13 2 SVEA 96+ 0.010 SDM REQUIREMENT (MOST RESTRICTIVE VALUE): 0.010

((

(3)))

68

PSEG Hope Creek Mixed Core Analysis Report (3)))

Figure 5.1 - Fresh GE14 Reload Bundle 2757 Configuration 69

-PSEG Hope Creek Mixed Core Analysis Report

((

13) 1i Figure 6.2 - Fresh GE14'Reload Bundle 2758 Configuration 70

PSEG Hope Creek Mixed Core Analysis Report (3)))

Figure 5.3 - Cycle 13 (Quarter Core) 71

PSEG Hope Creek Mixed Core Analysis Report I Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence I

{3)j3 72

PSEG Hope Creek Mixed Core Analysis Report Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence

((

73

PSEG Hope Creek Mixed Core Analysis Report I Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence

((

(3) ))

74

PSEG Hope Creek Mixed Core Analysis Report i Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence l

[I 75

PSEG Hope Creek Mixed Core Analysis Report I Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence 76

PSEG .Hope Creek Mixed Core Analysis Report Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence l 3[

(3) ))

77

I .PSEG Hope Creek Mixed Core Analysis Report I Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence

((

(3) ))

78

PSEG Hope Creek Mixed Core Analysis Report I Figure 5.4 - Cycle 13 Reference Loading Pattern Control Rod Operating Sequence

((31

{3}]

79

PSEG Hope Creek Mixed Core Analysis Report (31))

Figure 5.5 - Cycle 13 RLP Rod Pattern Thermal Design Ratio Results Figure 5.6 - Cycle 13 RLP Hot Excess Reactivity 80

PSEG Hope Creek Mixed Core Analysis Report Figure 5.7 - Cycle 13 RLP Cold Shutdown Margin

((

Figure 5.8 - Cycle 13 RLP Standby Liquid Control System Shutdown Margin 81

PSEG Hope Creek Mixed Core Analysis Report 6.0 Safety Limit Minimum Critical Power Ratio (SLMCPR)

This section of the MCAR provides the results of the SLMCPR evaluation of the Reference Loading Pattern containing legacy SVEA 96+ and the first reload of GE14 in the Hope Creek Generating Station, as reported in Section 5.0 of this report. SLMCPR information developed with GNF NRC approved methodologies and uncertainties1 ' is also included to allow for a comparison to the cycle previous to the introduction of GE14 (Cycle 12). The purpose of the evaluation is to determine the minimum allowable MCPR during the most limiting full core transients under which at least 99.9% of the rods in the core would be expected to avoid boiling transition. The minimum allowable MCPR established in this way is defined as the safety limit minimum critical power ratio (SLMCPR).

6.1 Discussion The Safety Limit Minimum Critical Power Ratio (SLMCPR) evaluations for Hope Creek Cycle 13 were performed using NRC approved methodology and uncertainties. Table 6.1 summarizes the relevant input parameters and results for Cycle 13. Additional information is provided in response to NRC questions related to similar submittals regarding changes in Technical Specification values of SLMCPR. NRC questions pertaining to how GE14 applications satisfy the conditions of the NRC SER111 have been addressed in Reference 2. Other generically applicable questions related to application of the GEXL14 correlation, and to the applicable range for the R-factor methodology, are addressed in Reference 3. Items that require a plant/cycle specific response are presented below.

Previously, the SLMCPR was calculated on the upper boundary of the power/flow operating map only at 100% flow / 100% power (rated flow/rated power) with limiting control blade patterns developed at the rated flow/rated power point. This approach had been shown in NEDC-32601P-A to result in conservative SLMCPR evaluation values. As reported in Reference 4, recent SLMCPR evaluations performed by GNF have shown that limiting control blade patterns developed for less than rated flow at the rated power condition sometimes yield more limiting bundle-by-bundle MCPR distributions and/or more limiting bundle axial power shapes than the limiting control blade patterns developed at the rated flow/rated power evaluation point. Consequently, in addition to the rated flow/rated power evaluation point, an SLMCPR calculation has been performed for Hope Creek at a lower flow/rated power evaluation point. The current Hope Creek licensing basis minimum allowable core flow at rated power is 87% rated flow. However, to account for future operation at lower flow/rated power conditions, SLMCPR evaluations were performed at a reduced core flow rate of 76.6% rated flow at the rated power condition for the same exposure points that were previously calculated for the rated flow/rated power evaluations. The SLMCPR results for Hope Creek Cycle 13 at the 76.6% rated flow condition are equivalent to or bound the SLMCPR results calculated at the rated flow condition and the 87% flow condition.

The core loading information for Hope Creek Cycle 13 is provided in Figure 6.2.

82

PSEG Hope Creek Mixed Core Analysis Report In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR distributions, and (2) flatness of the bundle pin-by-pin power/

R-factor distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. The value of these parameters for Hope Creek Cycle 13 is summarized in Table 6.1 as the MIP (MCPR Importance Parameter) and the RIP (R-factor Importance Parameter), respectively.

The impact of the fuel loading pattern differences on the calculated SLMCPR is correlated to the values of MIP and RIP. The calculated MIP value for the Hope Creek Cycle 13 core at EOR using a limiting rod pattern is (( (3)))

Pin-by-pin power distributions are characterized in terms of R-factors using the NRC approved methodology. 151 For the Hope Creek Cycle 13 limiting case analyzed at EOR, the weighted RIP value, considering the participation of the contributing bundles, was calculated to be ((

3)))

The revised power distribution methodology was used for the Hope Creek Cycle 13 analysis.

This methodology has been justified, reviewed and approved by the NRC (reference NEDC-32601P-A). When applying the revised model to calculate a lower SLMCPR, the conservatism that remains was reviewed, approved and documented by the USNRC. It was noted on page A-24 of NEDC-32601P-A (( {3}))

The SLMCPR was calculated for Hope Creek Cycle 13 using the reduced power distribution uncertainties described in Reference 1.

Table 6.1 summarizes the relevant input parameters and results of Cycle 13 evaluated at the condition of 76.6% rated flow/rated power. The SLMCPR values were calculated for Hope Creek using uncertainties that have been previously reviewed and approved by the NRC as listed in Table 6.2 and described in Reference 1 and, where warranted, higher plant-cycle-specific uncertainties as listed in Table 6.3. A (( {'})) consistent with current GNF fuel operation. For the Hope Creek Cycle 13 lower flow evaluations, the Core Flow Rate and Random effective TIP reading uncertainties were (( (31))

These calculations use the GEXL14 correlation for GE14 fuel and GEXL80 correlation for SVEA 96+ fuel (Reference 6). (( (3)))

83

PSEG Hope Creek Mixed Core Analysis Report The Two Loop and SLO SLMCPR values calculated for Hope Creek Cycle 13 are shown in Table 6.1.'

6.2 Summary The calculated 1.06 SLMCPR and 1.07 SLO SLMCPR for Hope Creek Cycle 13 are consistent with expectations given the ratios for MIP and RIP that have been calculated and the use of the reduced uncertainties described in Reference 1. Correlations of MIP and RIP directly to the calculated SLMCPR have been performed for this plant/cycle which show that these values are appropriate when the approved methodology and the reduced uncertainties given in NEDC-32601P-A and NEDC-32694P-A are used.

Based on all of the information and discussion presented above, it is concluded that a 1.06 SLMCPR and 1.08 SLO SLMCPR for the Hope Creek Cycle 13 core are appropriate.

6.3 References

1. Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601P, Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations;NEDC-32694P, Powver Distribution UncertaintiesforSafety Limit MCPR Evaluation;and Amendment 25 to NEDE-24011-P-A on Cycle Specific Safety Limit MCPR," (TAC Nos. M97490, M99069 and M97491), March 11, 1999.

2. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to R. Pulsifer (NRC), "Confirmation of 10x10 Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies",

FLN-2001-016, September 24, 2001.

3. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Confirmation of the Applicability of the GEXL14 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GE14 Fuel", FLN-2001-017, October 1, 2001.

4. Letter, Jason S. Post (GE Energy) to U.S. Nuclear Regulatory Commission Document Control Desk, "Part 21 Reportable Condition and 60-Day Interim Report Notification:

Non-conservative SLMCPR", MFN-04-081, August 24, 2004.

aThe calculated SLO SLNICPR is 1.07, however, the Hope Creek Cycle 13 NRC approved SLO value is 1.08. Hope Creek requested a SLO SLUCPR or l.OS in Iope Creek's initial license change request forthe Cycle 13 SLNICPR Due to an earlier than planned shutdown at the end of Cycle 12 and the removal oftwo fuel defects found during core offload, the Cycle 13 core was redesigned after the initial Cycle 13 SLMCPR license change request SLMICPR information based on the redesigned core was provided to the NRC as a supplement to the initial license change request. In the supplement Hope Creek chose to keep the SLSLMICPR value of l.O that was initially calculated since tie value is bounding for the redesigned core. The SLO SLNICPR value of 1.07 is presented in this report to provide an exact numerical base for comparison with Hope Creek constant pressure power uprate SLMCPR results.

84

PSEG Hope Creek Mixed Core Analysis Report

5. Letter, Thomas H. Essig (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Report NEDC-32505P, Revision 1, R-Factor CalculationMethod for GE)), GE12 and GEJ3 Fuel," (TAC Nos. M99070 and M95081), January 11, 1999.

6. GEXL80 Correlationfor SVEA 96+ Fuel, NEDC-33107P, Revision 0, Class HI, September 2003.

7. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Final Presentation Material for GEXL Presentation - February 11, 2002", FLN-2002-004, February 12, 2002.

85

PSEG Hope Creek Mixed Core Analysis Report Table 6.1 - Comparison of the Hope Creek Generating Station Cycle 13 and Cycle 12 SLMCPR DESCRIPTION Hope Creek Hope Creek Hope Creek DESCRIPTION Cycle 12 Cycle 13 Cycle 13 Number of Bundles in Core 764 764 764 Limiting Cycle Exposure Point' EOR EOR EOR 12020 10472 10472 Cycle Exposure at Limiting Point (EOR- 1102) (EOR- 1467) (EOR- 1467)

(MWdIMTU) 100.0 100.0 76.6 Core Flow, % Rated Reload Fuel Typc SVEA 96+ GE14 GE14 Latest Reload Batch Fraction, % 31.4 21.5 21.5 Latest Reload Avcrage Batch 3.61 4.02 4.02 Weight % Enrichment Core Fuel Fraction for GE14 (%o) 0.0 21.5 21.5 Core Fuel Fraction for GE9B (%o) 6.9 0.0 0.0 Core Fuel Fraction for SVEA 96+

93.1 78.5 78.5

(%/o)

Core Average Weight %

3.44 3.63 3.63 Enrichment Core MCPR (for limiting rod 1.38 1.42 1.38 pattern)

[r (3)jj Revised NEDC- Revised NED C- Revised NEDC-Power distribution methodology 32601P-A 32601P-A 32601P-A Reduced NEDC- Reduced NEDC- Reduced NEDC-Power distribution uncertainty 32694P-A 32694P-A 32694P-A Revised NEDC- Revised NEDC- Revised NEDC-Non-power distribution uncertainty 32601P-A 32601P-A 32601P-A Calculated Safety Limit MCPR 1.09 1.05 1.06 (Two Loop)

Calculated Safety Limit MCPR 1.10 1.06 1.07 (SLO) 0

'End of Rated (EOR) is defined as end-of-cycle all rods out, 100 /powcrl/ 00%' flowand normal feedwatertempcrature. Thc actual analysis is performed prior to EOR in order to have sufficient control rod denshy to force some bundles near to the OLSCPR.

86

s ..

PSEG Hope Creek Mixed Core Analysis Report Table 6.2 - Standard Uncertainties Hope Creek Hope Creek Hope Creek DESCRIPTION Cycle 12 Cycle 13 Cycle 13 100% Flow 100% Flow 76.6% Flow Non-power Distribution Revised NEDC- Revised NEDC- Revised NEDC-Uncertainties 32601P-A 32601P-A 32601P-A Core flow rate (derived from 2.5 Two Loop 2.5 Two Loop 2.5 Two Loop pressure drop) 6.0 SLO 6.0 SLO 6.0 SLO Individual channel flow area (( "IJ (( I3'l] I[IT]1 Individual channel friction factor 5.0 5.0 5.0 Friction factor multiplier (( {"In (( I3'l] (( "3I))

Reactor pressure (( 131j] (( 13j)) (( 13}1]

Core inlet temperature 0.2 0.2 0.2 Fecedwatcr temperature[ III] 31))

1[ "T{3]

Feedwater flow rate (( 13)jj (( 13))) (( 13))

Power Distrbuton Uncertainties Reduced NEDC- Reduced NEDC- Reduced NEDC-32694P-A 32694P-A 32694P-A GEXL R-factor (( '3}1II 13)1 13)jj 1((

Random effective TIP reading 1.2 Two Loop 1.2 Two Loop 1.2 Two Loop 2.85 SLO 2.85 SLO 2.85 SLO Systematic cffcctivc TIP reading (( 13))) 13))) [ 13))

Integrated effective TIP reading 133)) [L 3]

3)]((

13 Bundle power (( "311jj1 l3ljj Effective total bundle power ((313))) 33)1 uncertainh' __ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _

Table 6.3 - Exceptions to the Standard Uncertainties Used in Hope Creek Cycle 13 and Cycle 12 Reactor pressure 2.04 2.04 2.04 Core Flow Rate Random Effective TIP Reading 1 -3)11

- 3I 13 13 GEXL R-factor )j 1j 3 87

PSEG Hope Creek Mixed Core Analysis Report 60 EhEIEm AE LEG+

58 50 A [AE9I9 44 36n-C LI Al h DEILPILI C [El [E El1EEElID EBD L2 BElEl19 El U1 umAl 32 CGE EBA I E ABSAA DEB A BAE A 24 44 [2lA B BhE 1 B ~ B B A 1 1111 30 ASE B1E BL A1 B1E LE B1 BAA B1 BA C B1 BD1B1 E 22 F A ECDE Elif [71gp F1191 C A CE B0 B O~

28 A Um Cm B~ A BIII mA s~ B1 A~ ~ lC l

[4Eilln ~l 11 l 3 I2BE7]~ 7-E -0 28E 320 12 1MEI EITE Er1[ ER] EID E11y EIElE El I ID El El2 El 38 E] [3

'°ElS

[E E Ue]

iSE' rEv Li~mrfm rmEri m E01 To RSU[ETEii 12 E EMEA

_El__ [ _E__ E El m AE BOB UE]BAE EF E El El 1 0 IlElio AE] Bc ffliffl Cli 9I lDE EA0 FE C E 18 ElE 0Bffl r~fF[nElFl C5E nAA P1M EC1BE 0[DE] EAAEC ElB C7D bE][DIME[NumbeO]rnICyl A A FEC ElE M ID 16 E DE 9 IU11 MI Mo 95-I E1f008 4 E2 W= 0 F[

2 AAAEAlAE DE 3 5 79 11 13 15 17 1921 23 25 2729 31 33 3537 3941 43 45 47 4951 53S5557 59

. Number Cycle Code Bundle Name I I Loaded Loaded A SVEA96-PlOCASB326-1I1GZ-568U-4WR-150-T6-2654 184 10 B SVEA96-PlOCASB326-1 lG4.5-568U-4WR-150-T6-2655 48 10 C SVEA96-PlOCASB36O-12GZ-568U-4WR-150-T6-2656 167 11 D SVEA96-PIOCASB360-12G5.0-568U-4WR-150-T6-2657 72 11 E SVEA96-PlOCASB361-14GZ-568U-4WR-150-T6-2658 176 12 F SVEA96-PIOCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 64 12 G GE9B-P8CWB280-8G4.0-80U-150-T6 53 9 Figure 6.1 - Reference Core Loading Pattern - Cycle 12 88

PSEG Hope Creek Mixed Core Analysis Report 60 AE A A 58 O I E 56 56 El im M910 ii9 M 3MMErMI M E 54 EI E1ET E _E1X E- E-r]

mt 70 f E -E-1 r E_1 EI El7E I El 7E E 52 E107119IMIE]IEh 007IMIMI El 48 1_E

_OUDE1117 E 1171171MME]

46 El E11 DIE] E-1 B FE710 ff0 r1a,- 111- El E7 El E I ff-44 AE O _E _D 407_]E 1iIEII 3[E]E ]E DE ]r] EE]f] E]3E3 L3ilD!1[lDgLE]IL CE!EE 3842EE1E OM1 tEJS~IS Mtm 1R1, t UEI 19DE1F2 E12ED17E 11EtE31EIt 36 S [El[w E 1: EIM E10l EhE 0~ DLE] MLE] M-1171EP17iE] MLE DO Eli[l 32 ISE Rl - 15 EMD O2=2 lt 30 Eiin Ei EmiEi o pi [,Ei pinEaS ffE] ffSEl 21E mE lEf 2El~f DI 181 36 8EIEE 1 E5E1 l E1E 3DI 16 E cV EliD EIDE~ CEMIE US 0eE] G E] Eq[ M! E]

12 EtE EtE3 5 7 91335339143 25 rc 93135314491 m19 35575 14E l[ m m llB E 211E] ff E 28 SVEAPl 3El B1 [3ILI El1 j~=GJ ElGD-ID ElU El4 ElWE-5 ED M38 6 E3J8 W~ 10E 268 E m]E M7E] El[

]1E] EW l L E 1 1rm E E] E 34 11EA96-Pr1 ]OCAS36O12GZ01-568-4R-10-6-5G16 11M 28 EIE- EIE E]tE-l DCAESSEBEff- 2E[E-110EIE1EY- E6G

-l A96-IEO tET- -I-6m]25057-W-6El f ElEElEl B ElEl 69 ElEl 1E E VEA96-P1OS ElB6 -llgZ MIRU E4M W15- E1E64 EI ME E 12 F2 SElAl9l6l-lPEOCASB36EOEII-1-RE1 E-T-12G55/G25-68-4Ol50t6-2659 E 62E-El- 12_71D 1 GE14-P5 S A 19 G1113517 EYE-4 2 E1079 1 33 3E537 39 414547456 557E 5 13 H2 E14-1DOCNBO2-5G6H E]4G.0-IE]OT- ]E50-T6-275E8 108EE13 SVEA96-PlOCASB326-1 IGZ-568U-4WR-150-T6-2654 2E 10 K SVEA96-PlOCASB360-12G5.0-568U4WR-150-T6-2657 2 11 A SVEA96-P1IOCASB326-11 GZ-568U-4WR-150-T6-2654 89 10 J SVEA96-PIOCASB326-11 G4.5-568U-4WR-150-T6-2655 SVEA96-PlOCASB326-1 IG4.5-568U-4WR.150-T6-2655 3828 10 10 K SVEA96-PlOCASB360-12G5.0-568U-4 WR-150-T6-2657 69 11 L SVEA96-PlOCASB361-14GZ-568U-4WR-150-T6-2658 SVEA96-P1 OCASB361 -14GZ-568U-4WR-150-T6-2658 14 12 12 M

M SVEA96-P1IOCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 SVEA96-PlOCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 62 12 12 Figure 6.2 - Reference Core Loading Pattern - Cycle 13 89

PSEG Hope Creek Mixed Core Analysis Report 7.0 Cycle 13 Supplemental Reload Licensing Report (SRLR)

A copy of the Cycle 13 SRLR follows this analysis. The SRLR sections, tables, figures, appendices and page numbering are self contained as in the original report and therefore have not been modified to be consistent with Sections 1.0 - 6.0 of the MCAR. Accordingly, individual SRLR sections, tables and figures are not contained in the MCAR Table of Contents, List of Tables or List of Figures.

90

I-

tsrw

,GIobal Nuclear Fuel AJinlVeniture o?GE,'.Tohiba. Z Wileth 0000-0031 -0596-SRLR Revision I Class I December 2004 0000-0031-0596-SRLR, Rev. 1 Supplemental Reload Licensing Report for Hope Creek Unit 1 Reload 12 Cycle 13 Approved XM .2

• Approved e-4 M. E. Harding, Manager R. E. Kingston Fuel Engineering Services Customer Account Leader

Hope Creek 1 0000-0031-0596-SRLR Reload 12 Rev. 1 Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by Global Nuclear Fuel - Americas, LLC (GNF-A) solely for PSEG Nuclear, LLC and the U.S. Nuclear Regulatory Commission (USNRC). The information contained in this report is believed by GNF-A to be an accurate and true representation of the facts known, obtained or provided to GNF-A at the time this report was prepared.

The only undertakings of GNF-A respecting information in this document are contained in the contract between PSEG Nuclear, LLC and GNF-A for nuclear fuel and related services for the nuclear system for Hope Creek Generating Station Unit I and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GNF-A nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

Page 2

Hope Creek I 0000-0031 -0596-SRLR pRenod 19

.\....v..

Rev. I Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by G. N. Marrotte, R. H. Szilard, S. C. Gupta, S. J. Peters, A.

E. Horna and W. Wong. The Supplemental Reload Licensing Report was prepared by G. N. Marrotte.

This document has been verified by R. H. Szilard.

Page 3

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 The basis for this report is General Electric StandardApplicationfor Reactor Fuel, NEDE-240 11 -P-A-14, June 2000; and the U.S. Supplement, NEDE-2401 1-P-A-14-US, June 2000.

1. Plant-unique Items Appendix A: Analysis Conditions Appendix B: List of Acronyms Appendix C: Decrease In Core Coolant Temperature Events Appendix D: Basis for Kf curve Appendix E: Option B Licensing Basis Appendix F: Reactor Recirculation Pump Seizure Event

2. Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

SVEA96-PIOCASB326-1 IGZ-568U-4WR-150-T6-2654 (SVEA-96+) 10 91 SVEA96-P I1OCASB326-1 I G4.5-568U-4WR-1 50-T6-2655 (SVEA-96+) 10 40 SVEA96-PIOCASB360-12GZ-568U-4WR-150-T6-2656 (SVEA-96+) 11 166 SVEA96-PIOCASB360-12G5.0-568U-4WR-150-T6-2657 (SVEA-96+) 11 71 SVEA96-P I OCASB361 -14GZ-568U-4WR-150-T6-2658 (SVEA-96+) 12 168 SVEA96-PI OCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 (SVEA-96+) 12 64 New:

GE14-PIOCNAB402-5G6.0/14G4.0-1OOT-150-T6-2758 (GE14C) 13 108 GE14-PIOCNAB402-4G6.0/16G4.0-1OOT-150-T6-2757 (GE14C) 13 56 Total 764 Page 4

Hope Creek I 0000-0031 -0596-SRLR Relond 12 Rev. 1

3. Reference Core Loading Pattern Nominal previous cycle core average exposure at end of cycle: 25828 MWd/MT

( 23430 MWd/ST)

Minimum previous cycle core average exposure at end of cycle 25828 MWd/MT from cold shutdown considerations': ( 23430 MWd/ST)

Assumed reload cycle core average exposure at beginning of 17839 MWd/MT cycle: ( 16183 MWd/ST)

Assumed reload cycle core average exposure at end of cycle 29777 MWd/MT (rated conditions): ( 27013 MWd/ST)

Reference core loading pattern: Figure I

4. Calculated Core Effective Multiplication and Control System Worth - No Voids, 201C

[Beginning of Cycle, keffective I 1 Uncontrolled 1.103 Fully controlled 0.947 Strongest control rod out 0.983 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.001

5. Standby Liquid Control System Shutdown Capability Boron (ppm) Shutdown Margin (Ak)

(at 200 C) (at 1601C, Xenon Free) 660 0.035

' The licensing analyses are based on the actual shutdown exposure for the previous cycle.

Page 5

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I

6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters 2 Operating domain: ICF (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.57 1.40 1.040 6.688 105.5 1.32 SVEA-96+ 1.45 1.62 1.40 1.040 6.922 101.4 1.33 Operating domain: ICF (HBB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

. (MWt) (1000 lb/hr)

GE14C 1.45 1.51 1.54 1.040 6.453 109.3 1.30 SVEA-96+ 1.45 1.56 1.54 0.990 6.661 104.8 1.33 Operating domain: MELLLA (HBB)

Exposure range  : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) l__ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.43 1.34 1.040 6.109 78.7 1.32 SVEA-96+ 1.45 1.46 1.34 0.990 6.208 74.9 1.35 2 End of Rated (EOR) is defined as end-of-cycle all rods out, 100% power/100% flow, and normal feedwater temperature.

Page 6

Hope Creek I 0000-0031 -0596-SRLR Repnad 12 Rev. 1 Operating domain: MELLLA (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.39 1.44 1.040 5.922 80.9 1.33 SVEA-96+ 1.45 1.41 1.44 0.990 6.020 76.6 1.36 Operating domain: ICF (UB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13

___ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.58 1.29 1.040 6.760 104.4 1.33 SVEA-96+ 1.45 1.64 1.29 0.990 6.977 100.7 1.34 Operating domain: MELLLA (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ (MWt) (1000 lb/hr)

GE14C 1.45 1.49 1.21 1.040 6.365 76.1 1.30 SVEA-96+ 1.45 1.52 1.21 0.990 6.487 72.8 1.31 Operating domain: ICF & MFWT 3 (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/IT (2944 MWd/ST)

Peaking Factors _____

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ l_ (MWt) (1000 lb/hr) l GE14C 1.45 1.61 1.40 1.040 6.883 104.1 1.29 SVEA-96+ 1.45 l 1.66 1.40 0.990 7.060 100.1 l 1.31 3MFWT, minimum feedwater temperature, is allowed by plant Technical Specifications as low as 400 °F at rated power.

Page 7

Hope Creek I 0000-0031 -0596-SRLR V.I,-A V)

PRv 1.

Operating domain: ICF & MFWT (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.55 1.54 1.040 6.620 108.2 1.28 SVEA-96+ 1.45 1.60 1.54 0.990 6.837 103.3 1.30 Operating domain: MELLLA & MFWT (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.49 1.34 1.040 6.323 77.3 1.28 SVEA-96+ 1.45 1.51 1.34 0.990 6.402 73.6 1.31 Operating domain: MELLLA & MFWT (HBB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13

___ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.45 1.44 1.040 6.164 79.5 1.28 SVEA-96+ 1A5 1.47 1.44 0.990 6.238 75.2 1.31 Operating domain: ICF & MFWT (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 T__Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ _ (MWt) (1000 lb/hr) _

GE14C 1.45 1.63 1.29 1.040 6.949 103.1 1.30 SVEA-96+ 1.45 1.68 1.29 0.990 7.139 99.3 1.31 Page 8

Hope Creek I 0000-0031 -0596-SRLR Relnnd 19 R ev 1 Operating domain: MELLLA & MFWT (UB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ (MWt) (1 000 lb/hr) _

GE14C 1.45 1.54 1.21 1.040 6.555 75.0 1.26 SVEA-96+ 1.45 1.57 1.21 0.990 6.686 71.5 1.27 Operating domain: ICF with RPTOOS (HBB)

Exposure range

• BOC13 to EOR13-3245 MWd/MT (2944 MWdlST)

_________l__ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.54 1.40 1.040 6.586 106.2 1.35 SVEA-96+ 1.45 1.59 1.40 0.990 6.797 102.3 1.36 Operating domain: ICF with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.49 1.54 1.040 6.342 110.0 1.33 SVEA-96+ 1.45 1.54 1.54 0.990 6.583 105.3 1.35 Operating domain: MELLLA with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Peaking Factors _ _ __

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ _ _ (MWt) (1000 lb/hr) _

GE14C 1.45 1.43 1.34 1.040 6.077 78.8 1.33 SVEA-96+ 1.45 1 1.45 1.34 0.990 6.174 75.1 136 Page 9

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 Operating domain: MELLLA with RPTOOS (HBB)

Exposure range : EOR13-3245 MWdtMT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.39 1.44 1.040 5.906 80.9 1.34 SVEA-96+ 1 .45 1 .41 1 .44 0.990 6.009 76.7 1.37 Operating domain: ICF with RPTOOS (UB)

Exposure range : EOR13-3245 MWdtMT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lbthr)

GE14C 1.45 1.55 1.29 1.040 6.613 105.4 1.37 SVEA-96+ 1.45 1.60 1.29 0.990 6.838 101.7 1.37 Operating domain: MELLLA with RPTOOS (UB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13

___ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.48 1.21 1.040 6.301 76.4 1.31 SVEA-96+ 1 .45 1.51 1.21 0.990 6.428 73.1 1.33 Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range  : BOC13 to EOR13-3245 MWdtMT (2944 MWd/ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ _ _ _ _ (MWt) (1000 lbthr)

GE14C 1.45 1.58 1.40 1.040 6.758 104.9 1.32 SVEA-96+ 1.45 l 1.64 l 1.40 0.990 6.965 100.8 1.33 Page 10

Hope Creek I 0000-0031-0596-SRLR DP1-...A V) RPV I an.- ,1 Xv. ,

Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range  : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors _____

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lbthr)

GE14C 1.45 1.53 1.54 1.040 6.524 108.8 1.30 SVEA-96+ 1.45 1.58 1.54 0.990 6.733 104.1 1.32 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range  : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lbthr)

GE14C 1.45 1.48 1.34 1.040 6.289 77.5 1.29 SVEA-96+ 1.45 1.50 1;34 0.990 6.380 73.7 1.31 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lbthr)

GE14C 1.45 1.44 1.44 1.040 6.142 79.6 1.29 SVEA-96+ 1.45 1.46 1.44 0.990 6.215 75.3 1.32 Operating domain: ICF & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWdtMT (2944 MWd/ST) to EOC13

___ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

_ l_ l(MWt) (1000 lb/hr)

GE14C 1.45 1.59 1.29 1.040 6.795 104.1 1.33 SVEA-96+ 1.45 1.65 1.29 0.990 7.017 100.2 1.34 Page 11

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Operating domain: MELLLA & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13

___ _ Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)

GE14C 1.45 1.53 1.21 1.040 6.497 75.3 1.27 SVEA-96+ 1.45 1.55 1.21 0.990 6.599 72.0 1.29

7. Selected Margin Improvement Options 4 Recirculation pump trip: Yes Rod withdrawal limiter: No Thermal power monitor: Yes Improved scram time: Yes (ODYN Option B)

Measured scram time: No Exposure dependent limits: Yes Exposure points analyzed: 2 4 Refer to GESTAR for those margin improvement options that are referenced and supported within GESTAR.

Page 12

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1

8. Operating Flexibility Options s Extended Operating Domain (EOD): Yes EOD type: Maximum Extended Load Line Limit (MELLLA) 6 Minimum core flow at rated power: 76.6 %

Increased Core Flow: Yes Flow point analyzed throughout cycle: 105.0 %

Feedwater Temperature Reduction: No ARTS Program: No Single-loop operation: Yes Equipment Out of Service:

Safety/relief valves Out of Service: Yes (credit taken for 13 of 14 valves)

RPTOOS Yes

9. Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Operating domain: ICF (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

I Uncorrected ACPR J Event Flux Q/A GEI4C SVEA-96+ Fig.

(%NBR) (%NBR) _

FW Controller Failure 272 118 0.23 0.25 2 Turbine Trip w/o Bypass 340 117 0.26 0.27 3 Load Reject w/o Bypass 327 116 0.26 0.27 4 5 Refer to GESTAR for those operating flexibility options that are referenced and supported within GESTAR.

6 MELLLA is not a licensed operating domain at Hope Creek, however these analyses results bound their current licensed operating domain, ELLLA.

Page 13

Hope Creek I 0000-0031 -0596-SRLR Relonai 12 Rev. 1 Operating domain: 1CF (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 324 122 0.21 0.24 5 Turbine Trip w/o Bypass 396 121 0.24 0.26 6 Load Reject w/o Bypass 384 121 0.24 0.26 7 Operating domain: MELLLA (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Uncorrected ACPR Event Flux QIA GE14C SVEA-96+ Fig.

(%NBR) (%NBR) _

FW Controller Failure 191 111 0.22 0.24 8 Load Reject w/o Bypass 238 112 0.26 0.29 9 Turbine Trip w/o Bypass 242 112 0.26 0.28 10 Operating domain: MELLLA (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 233 115 0.22 025 11 Load Reject wlo Bypass 286 116 0.27 0.30 12 Turbine Trip w/o Bypass 285 116 027 0.30 13 Operating domain: ICF (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux QIA GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 231 114 0.23 0.24 14 Turbine Trip w/o Bypass 306 115 027 028 15 Load Reject w/o Bypass 301 114 0.27 0.28 16 Page 14

Hope Creek 1 0000-0031 -0596-SRLR Reload 12 Rev. I Operating domain: MELLLA (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWdlST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 147 106 0.18 0.19 17 Turbine Trip w/o Bypass 200 108 024 0.25 18 Load Reject w/o Bypass 198 107 0.23 0.25 19 Operating domain: ICF & MFWT (HBB)

Exposure range  : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

_ _Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 283 119 0.23 0.25 20 Operating domain: ICF & MFWT (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux QIA GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 331 123 0.21 024 21 Operating domain: MELLLA & MFWT (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 199 112 0.22 025 22 Operating domain: MELLLA & MFWT (HBB)

Exposure range : EOR13-3245 MWdlMT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux l Q/A GE14C SVEA-96+ Fig.

(%NBR) l (%NBR)

FW Controller Failure 243 116 0.22 025 23 Page 15

Hope Creek I 0000-003 1-0596-SRLR Reload 12 Rev. I Operating domain: ICF & MFMT (UB)

Exposure rangc : EOR13-3245 MWd/MT (2944 MWd/ST) to EOCI3 Uncorrected ACPR Event Flux QA GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 236 116 0.24 0.25 24 Operating domain: MELLLA & MFWT (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 151 107 0.19 0.21 25 Operating domain: ICF with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWdlMT (2944 MWd/ST)

Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 301 121 0.25 0.27 26 Turbine Trip w/o Bypass 385 121 0.28 0.30 27 Load Reject wlo Bypass 378 120 0.28 0.29 28 Operating domain: ICF with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GEI4C SVEA-96+ Fig.

(%NBR) (%NBR) _

FW Controller Failure 355 125 0.24 0.26 29 Turbine Trip w/o Bypass 442 125 0.27 0.29 30 Load Reject w/o Bypass 436 125 0.27 0.28 31 Page 16

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Operating domain: MELLLA vith RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 200 113 0.23 0.25 32 Load Reject w/o Bypass 254 114 0.27 0.29 33 Turbine Trip w/o Bypass 256 114 0.27 0.29 34 Operating domain: MELLLA with RPTOOS (HBB)

Exposure range : EOR13-3245 MWdlMT (2944 MWdlST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 240 117 0.23 0.26 35 LoadReject w/o Bypass 305 118 0.27 1 0.30 36 Turbine Trip w/o Bypass 293 118 0.27 0.30 37 Operating domain: ICF with RPTOOS (UIB)

Exposure range : EOR13-3245 MWdlMT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 257 118 0.27 0.27 38 Load Reject w/o Bypass 355 118 0.31 0.31 39 Turbine Trip w/o Bypass 342 118 0.30 0.31 40 Operating domain: MELLLA with RPTOOS (UB)

Exposure range : EOR13-3245 MWdlMT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 167 108 0.20 0.21 41 Load Reject w/o Bypass 223 110 0.25 0.27 42 Turbine Trip w/o Bypass 221 III 0.25 0.26 43 Page 17

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 313 122 0.26 0.27 44 Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 364 126 0.24 0.26 45 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 205 114 0.23 0.25 46 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range : EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 253 118 0.23 0.26 47 Operating domain: ICF & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWdtMT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVE;A-96+ Fig.

l (%NBR) (%NBR) _ _

FW Controller Failure 263 119 1 0.27 0.28 48 Page 18

Hope Creek I 0000-0031 -0596-SRLR PRlnora 12 Rev. I Operating domain: MELLLA & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Uncorrected ACPR Event Flux Q/A GE14C SVEA-96+ Fig.

(%NBR) (%NBR)

FW Controller Failure 169 110 0.21 0.23 49

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary Assuming the worst channel response and 50% availability of the LPRM's yields a ACPR of 0.21 for all RBM setpoints including the unblocked response.

11. Cycle MCPR Values 7 Two loop operation safety limit: 1.06 Single loop operation safety limit: 1.08 ECCS OLMCPR Design Basis: See Section 16 (Initial MCPR)

Non-pressurization events:

Exposure range: BOC13 to EOC13 GE14C SVEA-96+

Loss of Feedwater Heating (1 10F) 1.20 1.20 Control Rod Withdrawal Error (unblocked) 1.27 1.27 Fuel Loading Error (mislocated) 1.21 1.21 Fuel Loading Error (misoriented) 1.18 1.28 7 For single-loop operation, the MCPR operating limit is 0.02 greater than the two-loop value.

Page 19

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Limitintg Pressurization Events OLMCPR Summarv Table: s Appl. Option A Option B Cond.9 Exposure Range GE14C SVEA-96+ GE14C SVEA-96+

I EQUIPMENT IN SERVICE BOC13 to EOR13-3245 1.45 1.47 1.34 1.36 MWd/MT (2944 MWdIST)1.5.4134.6 EOR13-3245 MWd/MT1561013914 (2944 MWd/ST) to EOC13 1.56 1.60 1.39 1.43 2 RPTOOS BOC13 to EOR13-3245 M1d4MT (2944 MWd/ST) 1.7 1.48 1.36 1.37 EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 1.59 1.60 1.42 1.43 Pressurization events: 0 Operating domain: ICF (HBB)

Exposure range : BOC13 to EOR13-3245 M`Wd/MT (2944 MWd/ST)

Application condition: 1, 2 Option A Option B GE14C SVE A-96+ GE14C SVEA-96+

FW Controller Failure 1.42 1.43 1.31 1.32 Turbine Trip w/o Bypass 1.44 1.46 1.33 1.35 Load Reject w/o Bypass 1.44 1.46 1.33 1.35 s Each application condition (Appl. Cond.) covers the entire range of licensed flow and feedwater temperature unless specified otherwise. The OLMCPR values presented apply to rated power operation.

9 One SRV out-of-service allowed.

'°The application condition number(s) shown for each of the following pressurization events represents the application condition(s) for which this event contributed in the determination of the limiting OLMCPR value.

Page 20

Hope Creek I 0000-0031 -0596-SRLR Reloadl I? Rev 1 Operating domain: JCF (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1,2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.49 1.53 1.32 1.36 Turbine Trip w/o Bypass 1.52 1.55 1.35 1.38 Load Reject wlo Bypass 1.52 1.55 1.35 1.38 Operating domain: MELLLA (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 1, 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.40 1.42 1.29 1.31 Load Reject w/o Bypass 1A5 1.47 1.34 1.36 Turbine Trip w/o Bypass 1.45 1.47 1.34 1.36 Operating domain: MELLLA (HBB)

Exposure range : EOR13-3245 MWdlMT (2944 MWd/ST) to EOC13 Application condition: 1,2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.50 1.54 1.33 1.37 Load Reject w/o Bypass 1.55 1.60 1.38 1.43 Turbine Trip w/o Bypass 1.55 1.59 1.38 1.42 Operating domain: ICF (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE14C SV'EA-96+ GE14C SVEA-96+

FW Controller Failure 1;51 1.53 1.34 1.36 Turbine Trip w/o Bypass 1.56 1.57 1.39 1.40 Load Reject w/o Bypass 1.56 1.57 1.39 1.40 Page 21

Hope Creek 1 0000-0031 -0596-SRLR V.l-4 1V Rev. 1 Operating domain: MELLLA (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE14C SVE;A-96+ GE14C SVEA-96+

FW Controller Failure 1.46 1.48 1.29 1.31 Turbine Trip w/o Bypass 1.52 1.54 1.35 1.37 Load Reject w/o Bypass 1.51 1.54 1.34 1.37 Operating domain: ICF & MlUT (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 1, 2 Option A Option B GE14C SVTEA-96+ GE14C SVEA-96+

FW Controller Failure 1.42 1.44 1.31 1.33 Operating domain: ICF & MFWT (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE_4C SVEA-96+ __GE14C SVEA-96+

FW Controller Failure 1.49 1.53 1.32 1.36 Operating domain: MELLLA & MFWT (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 1, 2 Option A J Option B GE 14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.41 1.43 1.30 1.32 Operating domain: MELLLA & MFWT (HBB)

Exposure range : EOR13-3245 MWd/MT.(2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE14C l SVEA-96+ l GE14C SVEA-96+

FW Controller Failure 1.51 1.54 1.34 1.37 Page 22 i . .

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Operating domain: ICF & MFU'T (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE14C lSVEA-96+ GE14C lSVEA-96+

FW Controller Failure 1.52 1.54 1.35 1.37 Operating domain: MELLLA & MFWT (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 1, 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.47 1.50 1.30 1.33 Operating domain: ICF with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.44 1.45 1.33 1.34 Turbine Trip w/o Bypass 1.47 1.48 1.36 1.37 Load Reject w/o Bypass 1.47 1.48 1.36 1.37 Operating domain: ICF with RPTOOS (HBB)

Exposure range : EOR13-3245 MWdIMT (2944 MWd/ST) to EOCI3 Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.52 1.55 1.35 1.38 Turbine Trip w/o Bypass 1.55 1.58 1.38 1.41 Load Reject wlo Bypass 1.55 1.57 1.38 1.40 Page 23

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 Operating domain: MELLLA with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 2 Option A Option B GE14C SVE:A-96+ GE14C SVEA-96+

FW Controller Failure 1.41 1.43 1.30 1.32 Load Reject w/o Bypass 1.46 1.48 1.35 1.37 Turbine Trip w/o Bypass 1.46 1.48 1.35 1.37 Operating domain: MELLLA with RPTOOS (HBB)

Exposure range : EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.51 1.55 1.34 .1.38 Load Reject w/o Bypass 1.56 1.60 1.39 1.43 Turbine Trip w/o Bypass 1.55 1.59 1.38 1.42 Operating domain: ICF with RPTOOS (UB)

Exposure range : EOR13-3245 MWdlMT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SN'EA-96+ GE14C SVEA-96+

FW Controller Failure 1.55 1.56 1.38 1.39 Load Reject w/o Bypass 1.59 1.60 1.42 1.43 Turbine Trip w/o Bypass 1.59 1.60 1.42 1.43 Operating domain: MELLLA with RPTOOS (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.48 1.49 1.31 1.32 Load Reject w/o Bypass 1.54 1.56 1.37 1.39 Turbine Trip w/o Bypass 1.54 1.55 1.37 1.38 Page 24

Hope Creek I

• 0000-0031-0596-SRLR Reload 12 Rev. 1 Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range: BOC13 to EOR13-3245 MWdtMT (2944 MWd/ST)

Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.44 1.46 1.33 1.35 Operating domain: ICF & MFWT with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SIEA-96+

FW Controller Failure 1.52 1.55 1.35 l 1.38 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range : BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST)

Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.41 1.44 1.30 1.33 Operating domain: MELLLA & MFWT with RPTOOS (HBB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SVEA-96+ GE;14C SVEA-96+

FW Controller Failure 1.51 1.55 1.34 1.38 Operating domain: ICF & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B fGE14C SVEA-96+ GE14C l EA-96+

FW Controller Failure 1.55 1.57 1.38 1.40 Page 25

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Operating domain: MELLLA & MFWT with RPTOOS (UB)

Exposure range : EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 Application condition: 2 Option A Option B GE14C SVEA-96+ GE14C SVEA-96+

FW Controller Failure 1.49 l 1.51 1.32 1.34

12. Overpressurization Analysis Summary Psi l Pdome Pv Plant Event (psig) l (psig) (psig) Response MSIV Closure (Flux Scram) 1244 l 1246 1265 Figure 50

13. Loading ErrorResults Variable water gap misoriented bundle analysis: Yes 11 Misoriented Fuel Bundle ACPR GE14-PI OCNAB402-4G6.0/16G4.0-l OOT- 150-T6-2757 (GE14C) 0.08 GE14-PIOCNAB402-5G6.0/14G4.0-lOOT-150-T6-2758 (GE14C) 0.12 SVEA96-P I OCASB326-1 IGZ-568U-4WR-150-T6-2654 0.17 SVEA96-P I OCASB326-1 IG4.5-568U4WR-150-T6-2655 0.17 SVEA96-PI OCASB360-12GZ-568U-4WR-150-T6-2656 0.20 SVEA96-P I OCASB360-12G5.0-568U-4WR-150-T6-2657 0.20 SVEA96-PIOCASB361-14GZ-568U-4WR-150-T6-2658 0.22 SVEA96-P I OCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 0.21

14. Control Rod Drop Analysis Results Banked Position Withdrawal Sequence is utilized at Hope Creek Generating Station Unit 1; therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-2401 1-P-A-US.

" Includes a 0.02 penalty due to variable water gap R-factor uncertainty.

Page 26

Hope Creek 1 0000-0031-0596-SRLR Reload 12 Rev. 1

15. Stability Analysis Results GE SIL-380 recommendations, BWROG Interim Corrective Actions (Reference 15.1) and Backup Stability Protection for Inoperable Option III Solution (Reference 15.2) have been included in the Hope Creek Cycle 13 operating procedures. Regions of restricted operation defined in Attachment I to NRC Bulletin No 88-07, Supplement I, Power Oscillations in Boiling Water Reactors (BWRs) and expanded in BWROG-94079 (Reference 15.1) and Backup Stability Protection for Inoperable Option III Solution (Reference 15.2) are used for Hope Creek Cycle 13 backup stability protection evaluation (Reference 15.3). The standard ICA stability regions are expanded as appropriate to offer stability protection per BWROG-02072 (Reference 15.4) and OG 02-0119-260 (Reference 15.2) for Hope Creek Cycle 13 ELLLA operation.

The Hope Creek Cycle 13 stability analyses discussed above are applicable to the ELLLA operation domain as specified in Reference 15.3 Additional analysis would be required to support operation in the expanded MELLLA operating domain.

References:

15.1. BWROG-94079, BWTR Owner's Group Guidelinesfor Stability Interim Corrective Action, June 1994.

15.2. OG 02-0119-260, GE to BWR Owners' Group Detect and Suppress II Committee, Backup Stability Protection(BSP) for InoperableOption III Solution, July 17, 2002.

15.3. GENE-0000-0029-6193-01-RI, Backup Stability Protection Evaluationfor Hope Creek Cycle 13, November 2004.

15.4. BWROG-02072, Reviewv of BFVR Owners' Group Guidelines for Stability Interim Corrective Action, November 20,2002.

16. Loss-of-Coolant Accident Results 16.1 IOCFR50.46 Licensing Results The ECCS-LOCA analysis is based on the SAFER/GESTR-LOCA methodology. The licensing results applicable to each fuel type in the new cycle are summarized in the following table:

Page 27

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Table 16.1-1 Licensing Results Licensing Local Core-Wide Fuel Type Basis PCT Oxidation Metal-Water (OF) (%) Reaction

__(%)

SVEA-96+ 1540 < 1.00 < 0.10 GE14C 1370 < 1.00 <0.10 The SAFER/GESTR-LOCA analysis results for GE14C fuel and SVEA-96+ fuel are documented in Section 5 of Reference 16.4.1.

16.2 10CFR50.46 Error Evaluation The IOCFR50.46 errors applicable to the Licensing Basis PCT are shown in the table below.

Table 16.2 Impact on Licensing Basis Peak Cladding Temperature for GE]4C and SVEA-96+

10 CFR50.46 Error Notifications Number Subject PCT Impact (OF) 2003-05 Impact of Postulated Hydrogen-Oxygen 0 Recombination on PCT Total PCT Adder (OF)I 0 The GE14C and SVEA-96+ Licensing Basis PCT remain below the IOCFR50.46 limit of 22000 F.

16.3 ECCS-LOCA Operating Limits The ECCS MAPLHGR operating limits for all fuel bundles in this cycle are shown in the tables below.

Page 28

Hope Creek I 0000-0031 -0596-SRLR Reloadt 12 Rev. 1 Table 16.3-I MAPLHGR Limits for GE]14C Bundle Types: GE14-P I OCNAB402-4G6.0/16G4.0-l OOT-150-T6-2757 (GE14C)

GE14-P I OCNAB402-5G6.0/14G4.0-100T-150-T6-2758 (GE14C)

Average Planar Exposure MAPLHGR Limit GWd/IMT GWd/ST kW/ft 0.00 0.00 12.82 21.09 19.13 12.82 63.50 57.61 8.00 70.00 63.50 5.00 Table 16.3-2 MAPLHGR Limits for SVEA-96+

Bundle Types: SVEA96-P I OCASB326-1 IGZ-568U-4WR-150-T6-2654 (SVEA-96+)

SVEA96-P I OCASB326-11 G4.5-568U-4WR-150-T6-2655 (SVEA-96+)

SVEA96-P I OCASB360-12GZ-568U-4WR-150-T6-2656 (SVEA-96+)

SVEA96-P I OCASB360-12G5.0-568U-4WR-150-T6-2657 (SVEA-96+)

SVEA96-P I OCASB361 -14GZ-568U-4WR-150-T6-2658 (SVEA-96+)

SVEA96-PI OCASB360-12G5.5/2G2.5-568U-4WR-150-T6-2659 (SVEA-96+)

Average Planar Exposure MAPLHGR Limit GWdlMT GWd/ST kW/ft 0.00 0.00 12.85 3.68 3.34 12.85 16.00 14.51 10.97 65.00 58.97 7.24 The single loop operation multiplier on LHGR and MAPLHGR, and the ECCS Initial MCPR values applicable to each fuel type in the new cycle core are shown in the table below.

Table 16.3-3 Initial MCPR and Single Loop Operation PLHGR and MAPLHGR Multiplier Single Loop Operation Fuel Type Initial MCPR PLHGR and MAPLHGR Multiplier SVEA-96+ 1.250 0.80 GE14C 1.250 0.80 Page 29

Hope Creek I 0000-0031 -0596-SRLR nplPtlv, V1 Rev. 1 16.4 References The SAFER/GESTR-LOCA analysis base report applicable to the new cycle core is listed below.

16.4.1. NEDC-33 153P, SAFER/GESTR-LOCA ECCS-LOCA Loss of CoolantAccident Analysis for Hope Creek GenerationStation, Revision 1, September 2004.

Page 30

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. I 60 An A [DA EA E] + E A 58 El ElE l E ME3[3u((0 fl[o ffloem 56 AnMI S MBIC [E]D [CD 113 MJ[Dl E~[DFBI ID1[DIE)l MC1ID E]

54 0 10FE] I-OID [E ] M QM O lEN [- F19 131[C FtD M 52 56I EJ EJE D [ED [D3 [ED [ EEC][][HEtg 0 [D (G [E[GP(D EC]A El 52W 50l 36 _]1M MFl EET1ET1 EET1 EEEET1 _LEBEEEEETl 34 ENsEDm els 1T mm MFEIDID TEl etm 1 ; IT IT ETE 21 n 10TS El FD ID 44 E]3 D[C D[DFJI 2] C EmDE nH[B[]SH TmB] [Em MlHW 01[GD [H C G [CD E]PA 28K+ mE r!5Elmr El E~r Mim1E HI E 1E fC E A1 424 En3 E]E E1 ETEE] E+]; ;l Bl0l El 0l E1E~FDI 0 32 40 Enm [Dm E]IM [E] [D[H E [ED] [m 8M [E][E] M1 JCD[E [TEEE[HE]GJ[ M[F2 MTT3 [BEDE 20U 1 _ -Fl 1 "im l 1ET TEEl El T 18 EDE [9 L mImEIB R WlM EllE DE EID WIE 21E [D B T SEl BI I TBRE 22 8 El BLEElEC3 T~l ElE T31ES6TS E El OEl13rEl HIE r+E E3 28 [D 3 UE]

5 7 [D [Dff 9 11 1 l 15 17 19S 21 TE][FE 23 ffl 25 [E]M 27 29 S nC 35 En 33 0 37MHs [m41 43 45 47 39 51 (D 53 55057 5 34 0 F1 E 0TD [D E 10 M'[%g i[gfCA ET M ,

Fuel Type A=SVEA96-PIOCASB326-llIGZ-568U-4WR-1 50-T6-2654 (C10) H=GEI4-PI OCNAB4O2-5G6.0114G4.0-IOOT-150-T6-2758 (C13)

B=SVEA96-PI OCASB326-ll1G4.5-568U-4WR-150-T6-2655 (C10) 1=SVEA96-PIOCASB326-ll1GZ-568U-4WR-150-T6-2654 (C10)

C=SVEA96-PIOCASB360-12GZ-568U-4WR-150-T6-2656 (CII) J=SVEA96-P IOCASB326-ll1G4.5-568U-4WR-150-T6-2655 (C10)

D=SVEA96-P IOCASB360-12G5.0-568U-4WR-150-T6-2657(CI) K=SVEA96-P IOCASB36-12G5.0-568U-4WR-150-T6-2657(CII)

E=SVEA96-PIOCASB361-14GZ-568U-4WR-150-T6-2658 (C12) L=SVEA96-PI OCASB361-14GZ-568U-4WR-150-T6-2658 (C12)

F=SVEA96-PI OCASB36]-12G5.5[2G2.S-D68U-4WR-150-T6-2659 (C12) M=SVEA96-PIOCASB36[-12G5.51[G2.5-[68DJ-4WR-150-T6-2659 G=GE14-PIIOCNAB402-4G6.0/16G4.0-OOT-150-T6-2757(C13) (C12)

Figure 1 Reference Core Loadiug Pattern Page 31

Hope Creek I 0000-0031 -0596-SRLR P.l-r V) R ev. I

- Vessel Press Rise (psi)

Safety Valve RoN 1250 & Relief Valve Flow

-e- Bypass Valve Row X 750 25.0

-250 4 I 0.0 50 iQo 15.0 0o 5.0 IOa 150 Tine (sac) Tine (sac) osQo itco 0

, 0 V

Zu QO -

0.0 5.0 10.0 15.0 0.0 50 10.0 150 Tine (sec) Tine (sec)

Figure 2 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) ICF (HBB))

Page 32

Hope Creek I 0000-0031 -0596-SRLR PpRlenr 1F Rev. 1

.1-1.1. .-

150.0

-0 2Mo C) 50.0 mo no ao 6.0 nono ao ao 60 Tine (sec) Tine (sec) x, Imo E0 a) 0.

va Ix E C

Itz Z,

cc:U no ao 6.0 no ao no Tinm (see) line (sec)

Figure 3 Plant Response to Turbine Trip w/o Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) ICF (HBB))

Page 33

Hope Creek I - 0000-0031-0596-SRLR Reload 12 Rev. 1 15Q

,, 2MD a) irn mo MO 0.0 0t0 30 ao Tirl (sec) Tine (sec) aC IU Q0 ao 60 Q00 30 6.0 Time (sec) Time (sec)

Figure 4 Plant Response to Load Reject w/o Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/SI) JCF (HBB))

Page 34

Hope Creek I 0000-0031 -0596-SPRLR Reload 12 Rev. 1 15&01 IS co SQO 0.0

.0 5.0 100 150 5.0 100 15.0 Time (sec) Tire (sec)

OSLO IS=O 4.

0 C)

C:

E 0

0 (U

50.0 C, 0.0 50 10.0 150 00 50 1On 150 Tirre (sec) Tire (sec)

Figure 5 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF (HBB))

Page 35

Hope Creek I 0000-003 1-0596-SRLR Reloadl 12 Rev 1 m

0) co m

.1

-0 0o 30 30 6O Time (see) ime (see) m0 1mo C}

IC l0 ao0 60 00 30 TimD (sec) Tine (sec)

Figure 6 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF (HBB))

Page 36

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I 3M.0

.0 2Oao w

c .o 0-100.0 QO a3 50 8.0 Tine (sec) Time(sec) a Leve(inchREF-SEP-SKM

-*- \ssel Stean RXow

-A Turbine Steam Flow

_Feedwater Flow z;

• s 1.0Q cl 0

E 0

oo QO C.)

a:I A6 V A A nO iO 0.0 10 Tine (sec) Tirme (sec)

Figure 7 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOCI3 ICF (HBB))

Page 37

Hope Creek I 0000-0031 -0596-SRLR Reloand 19 PRv. 1 Ave Surface Elat Rwc 15Q0 6 Core Irlet FlAr

-4Core Irlet Sttcoofiri D10OD-a)

C to '

It w

-- 6 6 5Qo4 5.0 10.0 150 5.0 100 15.0 Tirre (sec) Tirre (sec) en z

4)0 SL E

0 Ma a)

Zx zs 5.0 100 150 0 100 15.0 Trine (sec) Tirre (sec)

Figure 8 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA (HBB))

Page 38

Hope Creek I 0000-0031 -0596-SRLR Oelnnd 1) Rev. 1 303.0

• B 2MO aS 0o ao 60 oo ao Tine (sec) Time (sec)

B Level(inch-REF-SEP-SKRT

-- Vessel Steam Raw

- Turbine Stear Row

- Feedwater Flow 0)

I P A-I ,

-luou oo ao 6.0 oo ao Tine (sec) Tine (sec)

Figure 9 Plant Response to Load Reject w/o Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA (HBB))

Page 39

Hope Creek 1 0000-0031 -0596-SRLR Relonadr 12 Rev. I 0 2mo 0.

Q0 3.0 6.0 no0 30 Timr (see) Tine (sec)

Z.

8

-0W~aa M i)

W E 0

r-c:

CD' QO 30 W0 Q0 ao So Tior (see) Time (sec)

Figure 10 Plant Response to Turbine Trip wlo Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/S¶J) MELLLA (HBB))

Page 40

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I 150.0 a,

co o0 SO 10.0 150 0.0 50 10.O 150 Tirr (sec) Tine (sec)

Z, a

0S 0.

E 0

U W.

a,'

0.0 so 10.0 15.0 0.0 50 10.0 150 Tirre (sec) Tire (sec)

Figure 11 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA (HBB))

Page 41

Hope Creek I 0000-0031 -0596-SRLR Reloadr 12 Rev. 1 GO ao ED ao ao Tine (sec) Tine (sec) zz 0

U 0

(U E,

o0 ao ao Q0

.0 Tinm (sec) Tine (sec)

Figure 12 Plant Response to Load Reject l/o Bypass (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 MELLLA (HBB))

Page 42

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I 150.01 10 Lu 50.0 1 0.0 3.0 60 00 ao 60 Tine (sec) Tine (sec)

S V;

a0 iv 0

0.

EL c

0 0'

0 60 0.0 30 6O Time (sec) Tine (sec)

Figure 13 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA (HBB))

Page 43

Hope Creek I 0000-0031 -0596-SRLR n-1-Alq ixvioau 11)

Ih Rev. I C)

'a ioao

-. 0T ot 0.0 50 10.o 150 MO 50 1OO 150 Tinrm (sec) TimiL (sec) 10

• C)1mO C, 0.

LU E 0

U ZU S.

0.0 SO IQ0 ISO 0o so 10.0 150 Tine (sec) Tine (sec)

Figure 14 Plant Response to FW' Controller Failure (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 ICF (UB))

Page 44

. . Hope Creek I -0000-0031 -0596-SRLR Reload 12 Rev. 1

- Vessel Press Rise (psi)

)f SafetyValve Row 3mo i ReiefVaieFlow

-Bypass Valve Flow a, 2m0 a:

1mo 1 010 0.0 A 60 as t I 10 Tihm (sec) Tine (sec) zz 0.

0 E0

.D) a)

W 0.0 30 60 00 3T 0

Tine (sec) Tirme (sec)

Figure 15 rlant Response to Turbine Trip w/o Bypass (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 ICF (UB))

Page 45

Hope Creek I 0000-0031 -0596-SRLR Rlb1ad1\.~

K 12)

Iev 1

.,Wv.-

1MQO C):

.ii 0.0 3.0 630 00 3o Tire (sec) Tirm (sec) 03 C

m 03 0.

D I=

E 0

U z.

0tr 0 3060 ao o10 ao Trne (sec) Tirm (sac)

Figure 16 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF (UB))

Page 46

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 15QO 125.0

• n 75.0

0) 0) aU IT

,A

-25.0 n0 5.0 10.0 15.0 Q0 50 Iao ISO Tunm(sec) Tern (see) r-XI, e) r-0 E

0 lu

-b a:

so 10O ISO 50 1i0 1S0 Time (sec) Tirn (sec)

Figure 17 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA (UB))

Page 47

Hope Creek 1 0000-0031 -0596-SRLR V.1n-,A 1V Rev. 1 15MO 10 Ixrm a)

't is:

0o.0 -

a0 3a0 6 00 0 ao £0 Tirm (sec) Tine (sec)

CD

., 1M.0. 0-a) 0.

E I'U 0 C.

I.5 W

Q0 ao 0. ac 60 Tim (sec) Tine (sec)

Figure 18 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA (UB))

Page 48

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I or:

0.0 1ao.

00 3o 6.0 0.0 3.0 6.0 Tine (see) Tine (sec)

Z, a) 0)

-C E C) 0 CL

-Z a

ao 30o &0 0. 3.0 60 Tirre (sec) Tifne (sec)

Figure 19 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA (UB))

Page 49

Hope Creek I 0000-0031 -0596-SRLR Reload 12 I.. _ .. _ .

Rev. I C,

C-O 10.0 15.0 0.0 10.0 15.0 Time (sec) Tirm (sec) z; r-0 a.

E 0

U C,.

0. 50 10.0 15.0 0.0 so 10.0 15.0 Tirm (sec) Time (sec)

Figure 20 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) ICF & MFWT (HBB))

Page 50

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I nO 50 100 15.0 0.0 5.0 10.0 15.0 Tine (sec) lrre (sec) 1510]

X!

a, 0

.a iao.s 0.

EL 0

0

-. 0

'3 Z,

50.0 0.0 10.0 15.0 0O 5.0 100 15.0 Tirme (sec) Tihe (sec)

Figure 21 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF & MFWT (HBB))

Page 51

Hope Creek I 0000-0031 -0596-SRLR

'PRenad 12 R ev 1 xAxeSurface tFRLa I 150.0 i Co1dRe ir.

C, C, WU

': 6 Is \

0.0 UVo 0.0 5.0 1a0 ISO 5.0 100 15.0 Time (sec) Tine (sec)

IsQo

'9 0,

a)

-0 100. 0 W~i~ a-C, a) E

'U1 0 U

EU C,

0.0 0.0 Io0 15.0 0.0 50 10O. 15.0 Tine (sec) Tine (sec)

Figure 22 Plant Response to FW Controller Failure (BOC13 to l:OR13-3245 MWd/MT (2944 MWd/ST) MELLLA & MFWT (HBB))

Page 52

Hope Creek l 0000-0031 -0596-SRLR Reload 12 Rev. 1 cy QO 1O 15.0 5.0 100 15.0 Tine (sec) Tirm (sec) 150 I s, Imo c) 0.

E 0

50.0 I.D 5.0 10.0 15.0 0.0 5.0 10.0 15.c I Tirm (sec) Tirm (sec)

Figure 23 Plant Response to FW Controller Failure (EOR13-3245 MWdAIMT (2944 MWd/ST) to E0C13 MELLLA & MFWT (HBB))

Page 53

Hope Creek I 0000-0031 -0596-SRLR ReIoncl 1), Rev. 1 150.01

. 1~0O

- T0 0.01M 00 0 5.0 100 15.0 00 50 10o 15.0 mnne (sec) Tine(sec) to 1500.0

-1sa 0.

Is ~. E-0 C, C.,

E

'U m,

0.o 5.0 10.0 15.0 0.0 50 . 1.0O 150 Tire (sec) Tinm (sec)

Figure 24 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF & MFWT (UB))

Page 54

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 a Vessel Press Rise (p) i~Safety~ veFlo\

1510 IMSO Relief Vale Rov I

_ Bypass Valve Flow 0 IMOo *0 75.0

- . a, 5 50.0 25.0 I

so 10.0 15.0 0o 5.0 10.0 15.0 Tine (sec) Tinm (sec) a Level(irch-REF-SEP-SKR1)

-- Vessel Steamn now

.5 no - Tubire Steam Flov

, EedwMierRaw, c

-,ICno a) no i)

Z-E a,

C 0

5Uo aI, -1.0 a 0r- ao

-20 ao 1T(a 150 0.0 so 10.0 15.0 Time (sec) Tine (sec)

Figure 25 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA & MFWT (UB))

Page 55

Hope Creek I - 0000-0031-0596-SRLR Reload 12 Rev. I 15so -

-0 100.o la a, a) cm ix:

50.0

.So LO 5.0 Qoo 150 to 50 100 150 Tine (sec) Tinm (sec)

MQO 0

Z; 150.0 0.

no E

0 a,

a) ob so 1o 150 n.o o50 100 15o0 Tine (sec) Time (sec)

Figure 26 Plant Response to FW Controller Failure (BOCI3 to EOR13-3245 MWd/MT (2944 MWd/ST) ICF with RPTOOS (HBB))

Page 56

Hope Creek I 0000-0031 -0596-SRLR RpInod 19 Rev. 1 co 03

-Q0 MO IQO SO 50G O 3.0 ao Tine (sec) Time (see) ci 0

, icmo 0.

o:

e E

0 0

103 0o ao 60 00 ao 60 Tine (sec) Tire (sec)

Figure 27 Plant Response to Turbine Trip w/o Bypass (BOC13 to EOR13-3245 MWdIMT (2944 MWd/ST) ICF with RPTOOS (HBB))

Page 57

Hope Creek I 0000-0031 -0596-SRLR Ro ) Rev. I 150.0 I 1C0O nC, Cow 1r:

.0-ict~

50.0 _ 10DO 0.0944 Io ED0 0.0 ao rinr (sec) Tire (sec)

C C,

0 2 0.

Ea 0: 0 eI

)'U C,

ao 60 0.0 ao 6.0

• rnm (sec) - T (sec)

Figure 28 Plant Response to Load Reject w/o Bypass (BOCI3 to EOR13-3245 MWd/MT (2944 MWd/ST) ICF with RPTOOS (HBB))

Page 58

Hope Creek I 0000-0031 -0596-SRLR ReloaId 12 Rev. 1 G)

(U 0o so 10.0 1&O 0.0 5.0 10.0 10.

Tine (sec) Tine (sec) 150.0 V;

0 c

100.0 0 v- E 0

U It MO 0),

0.0 so 10.0 15.0 - 0.0 0 10.0 150 Tire (sec) Tine (sec)

Figure 29 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF with RPTOOS (HBB))

Page 59

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 E- Vessel Press Rise (psi)

- Safetys ve FIowv 1sao & Relief Valve Flow

-Bypass Valve Flow 21MO0 - -0 2mo a) sao ao sas tmo tE o ao ELI ao 6.0 Tine (see) Tine (see) 1-1.0 +

~0.1 e Leel(inchREF-SEP-SKRT)

Vessel

- Steam FoN ao

- Turbne Steam Flo

-Feedwater FRov

_ QC a) to E

-0 q ----

• .i I. _ -1.0 30 60 0.0 3.0 Time (sec) Tirm (sec)

Figure 30 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 JCF with RPTOOS (HBB))

Page 60

Hope Creek I 0000-0031 -0596-SRLR Rel.InnH 11 Rev. I 150.0

-ic1iO *0 P.U.U W a) w 1-0 6- 0.0n ao a0 ac 6.0 Thm (sec) Timn (sec)

C Cl C

Cl to 0

id E C.,

i) 0

.?:.

0.0 ao 6o0 cao ao 6.0 The (sec) TiEe (sec)

Figure 31 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWdJMT (2944 MWdIST) to EOC13 ICF with RPTOOS (HBB))

Page 61

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 e Au n ILD

)(Ave Surfaoe Fhat FILux ISQO 6 Coe e Fo 15.0.lO

~ CoeWtScdn rmo ixr we EU1 i-t 50Lo 6~~6 UU .

QO 50 1QO ISO 0O SO Iao 15.0 Tine (sec) Tine (sec) zz

-a, 0

m1 r-0

'U 0.

i: E 0

C-ZU z,

50 10O 150 0.0 5.0 10.0 15.0 Tine (sec) Tirre (sec)

Figure 32 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA with RPTOOS (HBB))

Page 62

Hope Creek I 0000-0031 -0596-SRLR RPloadr 12 Pev I 150.0 0S 0)

IU 50.0 L0 60 0.0 ao 6.0 Tirn (see) Tine (see)

- Levd(irch-REF-SEP-SKM

- Vesse Steam Flo 00 uaa e- Turbie Steam RoN

- FeedwMter low X0 1010

'U 0-Q00 , Dp a _=

-1mn ao 60 0.0 30 6.0 Time (sec) Tife (sec)

Figure 33 Plant Response to Load Reject w/o Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA with RPTOOS (HBB))

Page 63

Hope Creek I 0000-003 1-0596-SRLR D.1-4 1)

-v~va Rev. I 150.0

.01M0 C,

'U Iet We o.0 ao Tine (sec) Tine (sec)

E- Lsel(inch-REF-SEPt-SK)

\ kessel SteamRXo 2taO 4 - Tutire Steam Flow A Feedwater FloN Z,

10O D QO (a 0 a, E E

C)

U CO

-I=.- v- ID-1.0 zo_

2.0 I6 . - 00 30 6.0 Tine (sec) Trme (sec)

Figure 34 Plant Response to Turbine Trip w/o Bypass (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA with RPTOOS (HBB))

Page 64

Hope Creek I .0000-0031 -0596-SRLR Deln-d rs.-a

1) Rev.

1.~P

- ub t FkL'c e Messel Press Rise (psi)

-Ae Suface l-lat R - Safetyk IveRON 150.0 Cae ItAd Rw Crie 1250 - Relief ae Flow

--- Core I2de Sboing I - Bypass Vahe Row 6~I

-0 750 mo 2

> ,, Dit Ct QO 4.

0.0 50 10.0 150

-250 I 0.0 5.0 10.0 15.0

]

Tuim (sec) Tine (sec)

V; 0

0.

E 0

U C,

ZU C,.

0.0 5.0 1TO 15.0 0.0 50 10.0 15.0 Tirm (sec) . Tirn (sec)

Figure 35 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (HBB))

Page 65

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I 150O

-DM a:

SQO 00 Iao ao 60 0.0 ao 60 Tine (sec) Tirne (sec)

-6 Level(irrh-REF-SEP-SKR) - *d eat

)- Vessel Steam Flow nh- r e ty o^ Tubre Steam Fkw 1.0 -t Reactivit Feedwater Flow 0.0(o

• 0 1m 0

to E 0

C -1.0 ~ no no 3.0 - o.o 3.0 6.C Time (sec) Tiffe (sec)

Figure 36 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (HBB))

Page 66

Hope Creek I 0000-003 1-0596-SRLR Reload 12 Rev. 1 150.0

.co~0 Ca)

It

'Ut o0 a so Q00 3o 60 Tine (sec) Tine (sea)

= 1100 CM C,

(U is S,

Z.

E 0

I .

0.0 ao &0 QO ao 6.0 Tirn (sec) Tirre (sec)

Figure 37 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (HBB))

Page 67

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 i,

ix 00 50 10.0 150 5.0 1O 15.0 Tirn (sec) Tire (sec) is 0

0.

E 0

C.)

Or U

Eu 0) ao 50 10.0 150 0.0 5.0 ... 10.0 . . 15.0 Tirr (sec) Tirm (sec)

Figure 38 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF with RPTOOS (UB))

Page 68

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 C?

WU no 6.0 ao Time (sec) Time (sec)

L- el(irchREF-SEP-SKRT)

-- Vessel Steam Flow

- Turbine Steam Flaw iFeedwater Flow C) 0)

to 1mo 0.

G)

E 0

id: C.,Z

-0.

(in l

• 61 I! 0I*-1.0 VrX1 0.0 0.0 ao I 2.0 6.0 rim? (sec) Time (sec)

Figure 39 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF with RPTOOS (UB))

Page 69

Hope Creek I 0000-0031 -0596-SRLR RFlIAAI I2 Rev. 1

- V Messel Press Rise (psi)

-*--Safety alve lRow

- Relief Valve Flaw

-.- Bypass Vahe RFow Xz.o 0

2.0 60 co 06

-E lea Pe )( e-i S ao ao 60 Thin (sec) Tine (sec)

V; 0

0.

0)

W rr E

E 5

U Z:

'U 0.0 ao 6o0 0 ao

-Tinm (sec) Tine (sec)

Figure 40 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 ICF with RPTOOS (UB))

Page 70

Hope Creek I 0000-0031 -0596-SRLR Relonad 12

.W - . - Rv. ...

Rev 1 a,

Wr

'e 50 10 0 iso QO 50 150 150 Timr (sec) inrre (sec)

-EVod Reacivity i--Do-pper Reacinvity 1.04 ,- Scra= ReacbviIty Totl Readi*gt a

= ao , re Lk c QO E

co:.1.0 a

50 10. 150 15O 15.0 Tire (sec) Tirm (sec)

Figure 41 Plant Response to FW Controller Failure (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (UB))

Page 71

Hope Creek I 0000-0031-0596-SRLR DOelnad 12 Rev 1 15Q0 a,MO cl co W: D:

we 30 60 0.0 ao Tirm (see) Tire (sec)

• 10n0 2

Ita, no 30 00 3.0 6.0 Tiwm (sec) TirrD (sec)

Figure 42 Plant Response to Load Reject w/o Bypass (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (UB))

Page 72

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I 15QO E 10.0 o 0c 0.0 ao 60q 0.0 ao ct Tine (sec) Tine (sec) sao a Levef(inchREF-SEP-SKD xk MsseI Steam Row 6 Turbine Steam Flaw

- Feedwater Flow a)

.0 1010 0-0.

0 C.

ZU

'U UU I .^ L6 V, S.

YrE v 0.0 ao 60 0.0 ao 60

- Tine (sec) Tie (sec)

Figure 43 Plant Response to Turbine Trip w/o Bypass (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA with RPTOOS (UB))

Page 73

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 1SAO

-01 mo *0 0

5.0 10. 15.0 1- 5.0 100 15.0 Tire (sec) inre (sec) i)

0.0 S

0 0

Z-0O 5.0 10. 15.0 0o 1QO 150 Tinm (sec) Tiffe (sec)

Figure 44 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWdIMT (2944 MWd/ST) ICF & MFWT with RPTOOS (HBB))

Page 74

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 150.0 20 10110 a) a, to D WX nW

-x:

so 10O 150 0. 5.0 10.0 150 Tire (sec) Tine (sec)

Z; C) 0 0.

E a

U It aO 5.0 100 150 0.0 s0 10.0 15.0 Tirr (sec) Tinre (sec)

Figure 45 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOCI3 ICF & MFWT with RPTOOS (HBB))

Page 75

• Hope Creek I 0000-0031 -0596-SRLR ReloadI 12 ePv 1 150.0 125.0 E Vessel Press Rise (psi)

Safety Valve Flo 6 Relief Valve Fo

-N

-.- Bypass Valve FkO

• a)

W IMo sa *0 75.0 510 250 4 n~-^

.,~~ u

-W.U .

a0o 10.0 150 5.0 1a0o 15.0 Tirre (sec) Time (sec)

V;

.D 1010 CL it0

C, E 0

0 ZU

'U z,

5.0 1ao 1S0 0o0 5.0 1ao 15.0

-Tirm (sec) Tin (sec)

Figure 46 Plant Response to FW Controller Failure (BOC13 to EOR13-3245 MWd/MT (2944 MWd/ST) MELLLA & MFWT with RPTOOS (HBB))

Page 76

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 Al xAxe Surfacel-effiux 15aO 6 -d Is o I

.n 1~0.

c)

W W.

IEt c0o 0.0 so 10.0 150 0.0 5.0 10.0 15.0 Time (sec) Tirn (see) 150.

.Z 1mo Sa-Sao 0.0 50 100 150 0. 50 10.0 15.0 Tirm (see) . . . . . . . Tire (sac)

Figure 47 Plant Response to FW Controller Failure (EOR13-3245 MWdIMT (2944 MWd/ST) to EOC13 MELLLA & MFWT with RPTOOS (HBB))

Page 77

Hope Creek I 0000-0031 -0596-SRLR T?.lnnrl 12 Rev. 1 150.0 a) ca a:

0.0 so MO 160 Q0 60 100 15.0 Tine (sec) Time (sec) a)

0 E

0 U

ma a,.

5.0 l_o 15.0 o0 6o 10.0 15.0 nTe6(sEc) -noIrm(sec)

Figure 48 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOCI3 ICF & MFWT with RPTOOS (UB))

Page 78

Hope Creek I 0000-0031 -0596-SRLR Relnoad 12 Rev. 1

- Vessel Press Rise>(psi) A x SafetyValve Row 125.0 6 Relief Mahle FlAo

_-- Bypass Valve RNo

• 0 75.0 a)

I 25.0

, an,,^ ,,^

-.ZDU 0.0 50 M00 15.0 5.0 1Q0 15.0 Tine (sec) Time (sec) 0 0.

E 0

O ZU

'U 6

0.0 5.0 10.0 15.0 0.0 5.0 10.0 15.0

- Tirm (sec) - - Tiur (sec)

Figure 49 Plant Response to FW Controller Failure (EOR13-3245 MWd/MT (2944 MWd/ST) to EOC13 MELLLA & MFWT with RPTOOS (UB))

Page 79

Hope Creek I 0000-0031 -0596-SRLR Re1lnora 12 Rev. l 0.01

_ QODzBe I km of 0.0 4.0 to t0 4.0 ao Tirn (sec) Tire (sec)

S 0) 0.

E 0

CD)

U 0.0 4.0 to 00 4.0 8.0 Tine (sec) Tirre (sec)

Figure 50 Plant Response to MSIV Closure (Flux Scram)

Page 80

Hope Creek I 0000-0031 -0596-SRLR Reload 12 RIv. 1 Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-I were used this cycle.

Table A-1 Analysis Value 12 Parameter ICF ICF & MELLLA MELLLA &

MFWT MFWT Thermal power, MWt 3339.0 3339.0 3339.0 3339.0 Core flow, Mlb/hr 105.0 105.0 76.6 76.6 Reactor pressure (core mid-plane), psia 1036.0 1031.5 1030.7 1026.4 Inlet enthalpy, BTU/lb 527.2 524.0 519.5 515.3 Non-fuel power fraction 0.036 0.036 0.036 0.036 Steam flow, Mlb/hr 14.43 13.99 14.39 13.96 Dome pressure, psig 1005.0 1000.7 1005.0 1000.7 Turbine pressure, psig 961.8 960.0 962.1 960.2 No. of Safety/Relief Valves' 3 14 14 14 14 Relief mode lowest setpoint, psig 1141.2 1141.2 1141.2 1141.2 Safety mode lowest setpoint, psig - - -

12 These analysis values were also applied for RPTOOS condition for ICF and MELLLA.

13 One SRV is allowed to be out of service.

Page 81

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. I Appendix B List of Acronyms Acronym Description ACPR Delta Critical Power Ratio Ak Delta k-effective

%NBR Percent Nuclear Boiler Rated 2RPT Two Recirculation Pump Trip ADS Automatic Depressurization System ADSOOS Automatic Depressurization System Out of Service AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARTS APRM, Rod Block and Technical Specification Improvement Program BOC Beginning of Cycle BSP Backup Stability Protection BWROG Boiling Water Reactor Owners Group COLR Core Operating Limits Report CPR Critical Power Ratio DIVOM Delta CPR over Initial MCPR vs. Oscillation Magnitude DR Decay Ratio ECCS Emergency Core Cooling System EEOC Extended End of Cycle ELLLA Extended Load Line Limit Analysis EOC End of Cycle EOR End of Rated (All Rods Out I 00%Power/ 00%Flow NFWT)

ER Exclusion Region FFWTR Final Feedwater Temperature Reduction FMCPR Final MCPR FOM Figure of Merit FWCF Feedwater Controller Failure FWTR Feedwater Temperature Reduction GDC General Design Criterion GESTAR General Electric Standard Application for Reactor Fuel GETAB General Electric Thernhal An alysisBasis - -

GSF General Shape Function HAL Haling Bum HBB Hard Bottom Bum HBOM Hot Bundle Oscillation Magnitude HCOM Hot Channel Oscillation Magnitude HFCL High Flow Control Line HPCI High Pressure Coolant Injection ICA Interim Corrective Action Page 82

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 Acronym Description ICF Increased Core Flow IMCPR Initial MCPR IVM Initial Validation Matrix LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LRHBP Load Rejection with Half Bypass LRNBP Load Rejection without Bypass LTR Licensing Topical Report MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELLLA Maximum Extended Load Line Limit Analysis MELLLA+ MELLLA Plus MFWT Minimum Feedwater Temperature MOC Middle of Cycle MRB Maximal Region Boundaries MSIV Main Steam Isolation Valve MSIVOOS Main Steam Isolation Valve Out of Service MTU Metric Ton Uranium MWd Megawatt day MWd/ST Megawatt days per Standard Ton MWd/MT Megawatt days per Metric Ton MWt Megawatt Thermal NBP No Bypass NCL Natural Circulation Line NFWT Normal Feedwater Temperature NOM Nominal Bum NTR Normal Trip Reference OLMCPR Operating Limit MCPR OOS Out of Service OPRM Oscillation Power Range Monitor Pdome Peak Dome Pressure PsI Peak Steam Line Pressure Pv Peak Vessel Pressure PCT Peak Clad Temperature PHE Peak Hot Excess

-PLHGR- - Peak Linear Heat Generation Rate PLUOOS Power Load Unbalance Out of Service PRFDS Pressure Regulator Failure Downscale PROOS Pressure Regulator Out of Service Q/A Heat Flux RBM Rod Block Monitor RC Reference Cycle RFWT Reduced Feedwater Temperature RPS Reactor Protection System Page 83

Hope Creek I 0000-0031-05 96-SRLR Refnload 12 Rev. 1 Acronym Description RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out of Service RVM Reload Validation Matrix RWE Rod Withdrawal Error SC Standard Cycle SL Safety Limit SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRLR Supplemental Reload Licensing Report SRV Safety/Relief Valve SRVOOS Safety/Relief Valve(s) Out of Service SS Steady State STU Short Tons (or Standard Tons) of Uranium TBV Turbine Bypass Valve TBVOOS Turbine Bypass Valves Out of Service TCV Turbine Control Valve TCVOOS Turbine Control Valve Out of Service TCVSC Turbine Control Valve Slow Closure TLO Two Loop Operation TRF Trip Reference Function TTHBP Turbine Trip with Half Bypass TTNBP Turbine Trip without Bypass UB Under Bum Page 84

Hope Creek 1 0000-0031-0596-SRLR Reload 12 Rev. 1 Appendix C Decrease In Core Coolant Temperature Events The Loss-of-Feedwater event was analyzed at 100% rated power using the BWR Simulator Code. The use of this code is permitted in GESTAR II. The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator Code; therefore, those items are not included in this document. The OLMCPR result is shown in Section 11.

In addition, the Inadvertent HPCI start-up event without a Level 8 turbine trip was shown to be bounded by the LFWH event in accordance with Determination ofLimiting Cold Water Event, NEDC-32538P-A.

The Rev. 0 SRLR (Reference C-1) indicated the Inadvertent HPCI with a Level 8 turbine trip is non-limiting.

References:

C-I. 0000-0031-0596-SRLR, Supplemental Reload Licensing Report for Hope Creek Unit I Reload 12/Cycle 13, Rev. 0, September 2004.

Page 85

Hope Creek I 0000-0031 -0596-SRLR Reload 12 Rev. 1 APPENDIX D Basis for Kf Curve A Kr curve for two-loop operation was provided for Cycle 9 in Reference D-1. The curve was updated for Cycle 13 application regarding core flow less than 40%. The updated curve is provided in Figure D-1 based on a maximum core flow runout of 109.0% and remains valid for Cycle 13.

References:

D-1. JiI-03372SRLR, Supplemental Reload Licensing Report for Hope Creek Generating Station Reload 8 Cycle 9, Revision 0, December 1998.

Page 86

Hope Creek 1 0000-0031-0596-SRLR Reload 12 Rev. 1 1.40 -

Kf= Max [1.00, (A + B(FI100))]

. I Mode A B 1.30 \ AFC 1.4410 -0.441 109% 1.3634 -0.441 The manual flow control scoop tube setpoint calibration is positioned such that 109% is 1.20 --------- - the maximum core flow value.

Automatic Flow Control (AFC) 1.10 __ ______ --------------------

109%

1.00 I 0.90.l 0.00 20.00 40.00 60.00 80.00 100.00 120.00 Core Flow C/6)

Fi(Ture D-1 Flomv Dependent MCPR Multiplier for Hope Creek Page 87

Hope Creek 1 0000-0031-0596-SRLR Reload 12 Rev. 1 APPENDIX E Option B Licensing Basis The NRC has concluded that a statistical approach (Option B) may be used for pressurization events analyzed with ODYN (References E-1 and E-2). The GEMINI statistical scram speed is provided in Table E-1.

Table E-1 GEMINI Methods: CRD Control Fraction vs. Time in BWR/2-5 0% 5% 20% 50% 90% 100%

X (sec) 0.200 .324 .694 1.459 2.535 2.804 cr (sec) .014 .016 .031 .070 The NRC Staff requires that, "in order to take credit for conservatism in the scram speed performance, it must be demonstrated that there is insufficient reason to reject the plant-specific scram speed as being within the distribution assumed in the statistical analysis".

General Electric presents the following procedure as one that satisfies the Staff's objectives for scram conformance. It should be noted that some utilities using ODYN Option B might desire to establish their own conformance procedures.

The procedure consists of testing, at the 5 percent significance level, the scram surveillance data at the 20 percent insertion position which is generated several times each cycle as required in the Reactivity Control System Technical Specification (20 percent insertion is representative of that portion of the scram most affecting the pressurization transient). The unique rod notch position closest to 20 percent (and the appropriately adjusted time of insertion) is expected to be utilized in actual plant application of this generic concept. For most plants, the surveillance requirements are as follows:

(1) all control rods are measured at beginning of cycle (BOC), and (2) X% of control rods are measured every 120 days during cycle (X is plant-dependent and ranges from 10 to 50).

At the -completion -of each surveillance test performed -in compliance -with -the -technical specifications surveillance requirements, the average value of all surveillance data at the 20 percent insertion position generated in the cycle to date is to be tested at the 5 percent significance level against the distribution assumed in the ODYN analyses. The surveillance information which each plant, using this procedure, will have to retain throughout the fuel cycle is the number of active control rods measured for each surveillance test (the first test is at the BOC and is denoted N.; the ii test denoted Ni and the average scram time to the 20 percent insertion position for the active rods measured in test i denoted ri.

Page 88

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. 1 The equation used to calculate the overall average of all the scram data generated to date in the cycle is:

n EN,

,=,

where:

n= number of surveillance tests performed to date in the cycle; n total number of active rods measured to date in the cycle; and EN, sum of the scram time to the 20 percent insertion position of all active EN,, = rods measured to date in the cycle to comply with the Technical i-1 Specification surveillance requirements.

The average scram time, rave, is tested against the analysis mean using the following equation:

Tave~ <B (2) where:

TB -,u+1.65 N (3)

The parameters p and a are the mean and standard deviation of the distribution for average scram insertion time to the 20 percent position used in the ODYN Option B analysis.

If the cycle average scram time satisfies the Equation 2 criterion, continued plant operation under the ODYN Option B operating limit minimum critical power ratio (OLMCPR) for pressurization events is permitted. -If-not,-the OLMCPR-for-pressurization-events-must -be-re-established,-based on a -linear interpolation between the Option B and Option A OLMCPRs.

The equation to establish the new operating limit for pressurization events is given below:

OLMCPRN, = 0LMCPROPIonB + v' - r AOLMCPR (4)

TA - TB where: rv and TB are defined in Equations I and 3, respectively; Page 89

Hope Creek I 0000-0031 -0596-SRLR Reload 12._.

Rev 1._....

rA = the technical specification limit on core average scram time to the 20 percent insertion position.

AOLMCPR = the difference between the OLMCPR calculated using Option A and that using Option B for pressurization events.

The control fractions presented in Table E-l are based on a ratio of distance inserted to control rod stroke.

Alternatively, scram times are expressed as a function of notch position. Table E-2 provides notch positions that correspond to approximately 20% control fraction. These notch positions and times can be used in equations I through 4.

Table E-2 GEMINI Methods: CRD Notch Positions for Tr Determination Notchiu (pickup) p (dropout) ar (pickup) a (dropout) 39 0.655 0.672 0.016 0.016 38 0.706 0.724 0.016 0.017 37 0.759 0.777 0.017 0.018 36 0.813 0.830 0.018 0.019

References:

E-1. NEDO-24154 and NEDE-24154-P, Safety Evaluationfor the General Electric Topical Report -

Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, Volumes I, II, and III, USNRC, June 1980.

E-2. Letter, J. S. Charnley (GE) to H. N. Berkow (NRC), Revised Supplementary Information RegardingAmendment 11 to GELicensingTopicalReportNEDE-24011-P-A,January 16, 1986.

Page 90

Hope Creek I 0000-0031-0596-SRLR Reload 12 Rev. I Appendix F Reactor Recirculation Pump Seizure Event The reactor recirculation pump seizure event is analyzed for Single Loop Operation (SLO) at HCGS (Reference F-I). This analysis was performed for the HCGS Cycle 13 transition cycle with GE14 and SVEA-96+ fuel in the core and transient analysis inputs that are consistent with the Reload 12/Cycle 13 analyses.

The SLO OLMCPR of 1.51 is required so that the reference SLO SLMCPR of 1.12 is protected in the event of a seizure of the recirculation pump in the active loop. If the cycle-specific Safety Limit Minimum Critical Power Ratio (SLMCPR) changes then the SLO OLMCPR may be adjusted by the following factor:

(Cycle Specific SLMCPR/ 1.12)

Thus, for HCGS Cycle 13 with a SLO SLMCPR of 1.08 the SLO OLMCPR required is:

1.51 * (1.08/1.12)= 1.46 In order to protect the required SLO OLMCPR of 1.46 (based on a SLO SLMCPR of 1.08) the following TLO limits must be maintained:

Pre ARTS implementation:

As long as the TLO full power OLMCPR is 1.33 or greater, the current Hope Creek Kr curves bound operation in SLO. If the full power OLMCPR is lower than 1.33 and is not bounded by the cycle specific off-rated limits, then the condition specific SLO OLMCPR of 1.46 should be applied for GE14 fuel and SVEA-96+ fuel.

Post ARTS implementation:

• As long as the TLO full power OLMCPR is 1.31 or greater, the proposed Hope Creek K(p) curve bounds operation in SLO. If the full power OLMCPR is lower than 1.31 and is not bounded by the cycle specific off-rated limits, then the condition specific SLO OLMCPR of 1.46 should be applied for GE]4 fuel and SVEA-96+ fuel.

References:

F-1. NEDC-33158P, Fuel Transition Reportfor Hope Creek Generating Station, Revision 2, November 2004.

Page 91