ML20151A422

Rev 0 to Reload Info & SAR for Oyster Creek Cycle 12 Reload

ML20151A422
Person / Time
Site:	Oyster Creek
Issue date:	03/31/1988
From:	Alammar M, Dougher J, Furia R GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML19302D386	List: ML20151A408 ML20151A411 ML20151A415 ML20151A418 ML20151A422
References
TR-049, TR-049-R00, TR-49, TR-49-R, NUDOCS 8804070011
Download: ML20151A422 (48)

Text

TR 049 Rev. O RELOAD INFORMATION AND SAFETY ANALYSIS REPORT FOR OYSTER CREEK CYCLE 12 RELOAD TR-049 Res. O u

BA No. 335400 M. A. ALAMMAR J. D. DOUGHER MARCH. 1988 APPROVALS:

W 3fZ/f38 Manager,OysterCreek/FuelProjects '

6 ATE 8M AN ,P)

Nuclear Analysis & Fuels Dire'ctor

~

DATE CPU NUCLEAR 1 Upper Pond Road Parsippany. New Jersey 07054

, jar 4S88uBi88l8e g

TR No. 049 Rev. O Page 2 of 47 ABSTRACT This report presents design information and analysis results pertinent to the Oyster Creek Nuclear Generating Station Cycle 12 Reload. This includes the Cycle 12 fuel design and core loading pattern descriptions; nuclear and tharmal hydraulic characteristics of the core; shutdown capability and reactivity functions; and the results of severhl transient and accident analyses. It is concluded that the Cycle 12 core can be operated safely, but will require a change to the plant technical spR.fications.

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TR No. 049 Rev. O Page 3 of 47 TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND

SUMMARY

.......................................... 6 2.0 CYCLE 12 REFERENCE CORE DESIGN .................................... 8 2.1 Operating History (Cycle 11) ................................. 8 2.2 Coro Loading Pattern (Cycle 12) .............................. 8 2.3 Fuel Design .................................................. 9 2.3.1 Hechanical ........................................... 9 2.3.2 Operating Experience .................................. 9 3.0 NUCLEAR CHA RACT ERISTICS OF TH E CORE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.1 Core Power Distributions ..................................... 14 3.2 Core Reactivity .............................................. 15 3.3 Shutdown Margin .............................................. 15 3.4 Reactivity Coefficients .................... ................. 15 3.5 Scram Reactivity .................................... ........ 16 3.0 Liquid Poison System ......................................... 16 4.0 T H E RMA L A N D HY O RAU L I C D E S I G N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 4.1 Steady State Evaluation ...................................... 23 4.2 Hot Channel Evaluation ....................................... 24 4.3 Local Linear Heat Generation Rate ............................ 24 4.4 Fuel Cladding Integrity Safety Limit ......................... 24 5.0 ACCIDENT AND TRANSIENT ANALYSIS ................................... 27 5.1 Accident Analysis ............................................ 27 5.1.1 Loss-of-Coolant Accident .............................. 27 5.1.2 Fuel Hisloading ....................................... 27 5.1.3 Control Rod Orop Accident ............................. 28 5.2 Transient Analysis ........................................... 28 5.2.1 Turbine Trip Without Bypass ........................... 29 5.2.2 Loss of Feedwater Heating ..,.......................... 29 5.2.3 Feedwater Controller Failure (Max. Deinand) ............ 30 5.2.4 Main Steam Isolation Valve Closure with No Scram ...... 30 5.2.5 Control Rod Withdrawal Error .......................... 30 6.0 OPERATING LIMIT MCPF ...... ............................... .... . 44 7.0 STABILITY ANALYSIS ....... .. .... .. . .. . . .. ........... . . 44 8.0 TECHNICAL SPECIFICATIONS .... ... ........... ..................... 44

9.0 REFERENCES

. .. . . . . . .. .. . 46 3100C

TR No. 049 Rev. O Pags 4 of 47 LIST OF TABLES Table Title Page Number 2.1 Design Basis Cycle 11 and Cycle 12 Exposures 10 2.2 Cycle 12 Reference Cor! Description 11 2.3 Nominal Fuel Mechanical Design Parameters 12 Shutdown Margin Calculation 17 3.1 3.2 Cycle 12 Reactivity Coefficients 18 4.1 Hot Channel Transient Analysis Initial 26 Condition Parameters 5.1 CPR Analysis Summary 32 5.2 Core-Wide Transient Analysis Results 33 5.3 CPR and MLLHGR for RWE 34  !

6.1 .ycle Operating Limit MCPRS 45 i

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I TR No. 049 Rev. O Page 5 of 47 l

LIST OF FIGURES Figure Title Page 2.1 Cycle 12 Reference Loading Pattern 13 3.1 EOC 12 Average Axial Exposure 19 I

3.2 Cycle 12 Hot Excess Reactivity 20 3.3 Cycle 12 Minimum Shutdown Margin 21 3.4 Cycle 12 Scram Reactivity 22 5.1 Turbine Trip Hithout Bypass, Power, Flow 35 Heat Flux 5.2 Turbine Trip Without Bypass, Dome Pressure 36 Rise, Relief Valve Flow 5.3 Feedwater Controller Failure (Max. Demand), 37 Dome Press. Rise, Relief Valve Flow, Bypass ,

Flow l 5.4 Feedwater Controller Failure (Max. Demand), 38 Power, Heat Flux, Core Inlet Subcooling 5.5 Feedwater Controller Failure (Max. Demand), 39 Dome Press. Rise, Relief Valve Flow, Bypass Flow i 5.6 Main Steam Isolation Valve Closure No 40 l Scram, Dome Pressure Rise, Safety Valve ,

Flow l 1

5.7 Main Steam Isolation Valve Closure No 41 i Scram, Heat Flux, Power, Core Flow 5.8 Control Rod Pattern for RHE Analysis 42 5.9 APRM Response to Control Rod Withdrawal 43 3100C i

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TR No. 049  !

Rev. O l Page 6 of 47

1.0 INTRODUCTION

AND

SUMMARY

This report justifies the operation of the Oyster Creek Nuclear Generating Station through the upcoming fuel reload Cycle 12. It is planned to operate the Oyster Creek reactor in Cycle 12 beginning in December 1988 with a partial core loading of fresh P8X8R and GE8X8EB fuel bundles supplied by General Electric Company (GE).

The Cycle 12 reference loading pattern is designed to ensure compliance with Technical Specification limits and safety analysis criteria. The reload dependent analyses are perfor e d with methodology developed by GPUN. These methods were previously submitted to the NRC for approval (References I through 7). The loss-of-coolant accident, rod drop j accident and stability analyses were performed by the fuel vendor using previously approved methods.

I The GE8X8EB fuel design will be introduced for the flist time into the j Oyster Creek core for Cycle 12. This fuel has been designed to accommodate higher enrichments, longer fuel cycles and will reduce or eliminate PC! related fuel failures. The Cycle 12 fuel design has an average enrichment of 3.217 and is described in Appendix B of Reference 9.

1 The transient and accident analyses presented in this report demonstrate I that based on the Turbine Trip Without Bypass transient, the Technical Specification CPR operating limit will have to be raised from 1.45 to 1.50 for Cycle 12. The safety limit MCPR is determined using the General Electric Company Thermal Analysis Basis, GETAB' 'vith the GE critical quality (X) boiling length (L) GEXL correlation.

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TR No. 049 l

'Rev. O i Page 7 of 47 I The proposed Cycle 12 core loading pattern and safety analyses presentec in this report are based on projected EOC 11 conditions and will be reevaluated based upon actual conditions if necessary.

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TR No. 049 Rev. O Page 8 of 47 2.0 CYCLE 12 REFERENCE CORE DESIGN 2.1 Operating History (Cycle 11)

This section discusses operating experience in the current cycle (Cycle 11) which may affect the fuel / core characteristics in the reload cycle (Cycle 12).

A total of 188 fresh GE P8X8R reload fuel assemblies were loaded into the reactor core during the 1986 outage. The residual fuel assemblies were relocated in the core to obtain adequate shutdown margin and acceptable cycle power peaking. Cycle 11 was designed for a full power energy production of 874 GWD.

The Cycle 11 power generation began on December 21, 1986. To date, March 1, 1988, no anomalies in core performance characteristics such ,

as core K.,,, control rod density or power distribution have been observed. Cycle 11 shutdown is scheduled for October 1, 1988. The l

current projected end-of-cycle energy generation is 860 GWD assuming an 857. capacity factor for the remainder of the cycle.

2.2 Core Loading Pattern (Cycle 12) figure 2.1 provides the Cycle 12 reference core loading pattern which was used for all analyses reported in this document. The figure includes the assemblies locations by fuel type and the projected beginning-of-cycle (BOC) 12 radial exposure distribution.

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TR No. 049 Rev. O Page 9 of 47 The design energy production for Cycle 12 is 1050 GWD (10.716 GWO/MT) which corresponds to a cycle length of about 21 months of reactor operation assuming 857. capacity factor (Table 2.1). Table 2.2 provides a summary of the Cycle 12 core.

2.3 Fuel Design 2.3.1 Mechanical The fresh fuel assemblies to be loaded in Cycl? 12 are of the GE P8X8R and GE8X8E8 designs. NRC approval of the GE8X8E8 fuel design is provided in reference 8. All exposed GE fuel assemblies are of the P8X8R design except for twenty-eight exposed ENC VB fuel which will be loaded on the core periphery in Cycle 12. The detailed design features of each of these fuel type are provided in References 9,10, and 11.

Table 2.3 provides a summary of key design parameters for each fuel type.

2.3.2 Operating Experience Both the ENC VB and the GE P8X8R had been loaded in previous Oyster Creek cycles. The adequacy of their designs have been demonstrated through their operating performance.

Fourteen GE plants are scheduled to receive and load the GE8X8EB fuel design prior to its application at Oyster Creek. By the time Oyster Creek's fuel reaches its expected discharge exposure (=33,600 MW0/MT), GE will have a great I deal of experience with these bundles in the extended exposure regime.

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TR No. 049 Rev. O Page 10 of 47 TABLE 2.1 Design Basis Cycle 11 and Cycle 12 Exposures Projected E0C 11 Core 17.436 GWD/MT Average Exposure Projected BOC 12 Core 9.716 GWD/HT Average Exposure Haling Calculated E0C 12 20.432 GWO/MT Core Average Exposure Projected Cycle 12 Full 10.716 GWD/HT Power Energy Capability i 3100C

i TR No. 049 Rev. O I Page 11 of 47 TABLE 2.2 Cycle 12 Reference Core Description Number of Control Rods 137 Number of Fuel Bundles 560 Total Weight of U in Core (MT) 98.11 Bundle Batch Avg Number Type Bundle Description Exposure (GWD/MT) In Core A EXXON VB 17.96 28 8 P80RB239-5G2.0-80M-145 18.33 112 C P80RB265-6G3.0-80M-145 16.30 64 0 GE7-P80RB299-7GZ2-80M-145 10.45 136 E GE7-P80RB299-7GZl-80M-145 8.76 48 F GE8-P8DQB321-8GZ-80M-145 0.0 152 G GE7-P80RB299-7GZ2-80M-145 0.0 20 9.72 560 l

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TR No. 049 Rev. O Page 12 of 47 TABLE 2.3 Nominal Fuel Machanical Design Parameters FUEL TYPE GE8X8EB P8X8R ENC VB Fuel Pellets fuel Material (sintered) UO: UO U0:

Pellets)

Average Enrichment, 3.21 2.99/2.65/2.39 2.50 w/o U-235 Pellet Density, 96.5 95.0 93.5

7. theoretical Pellet Diameter, inches 0.411 0.410 0.4195 Fuel Rod Active Length, inches 145.24 145.24 144.0 Plenum length, inches 9.23 9.23 10.62 Fuel Rod Pitch, inches 0.640 0.640 0.642 Diametral Gap (cold), 0.008 0.009 0.010 inches Fill Gas Helium Hellum Helium Cladding Material Zr-2 Zr-2 Zr-2 Outside Olameter, inches 0.483 0.483 0.5015 Thickness, inches 0.032 0.032 0.036 Inside Olameter, inches 0.419 0.419 0.4295 Fuel Channel Material Zr-4 Zr-4 Zr-4 Inside Olmension, Inches 5.278 5.278 5.278 Wall Thickness, inches 0.080 0.080 0.080 Fuel Assembly Fuel Rod Array 8X8 8X8 8X8 Fuel Rods per Assembly 60 62 60 Spacer Grid Haterial Zr-4 Zr-4 Zr-4 Flow Area, FT' O.1099 0.1099 0.1047 Heat Transfer Area. FT 2 91.83 94.90 94.53 3100C

TR No. 049 '

Rev. O FIGURE 2,1 Page 13 of 47 CYCLE 12 REFERENCE LO ADING P ATTERN l

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4 0 0 P G P G 0 P O 22.3 1a7 98 00 00 00 00 1a7 l 00 134 1 0 0 C C P E 8 C P O 8 117 13 7 118 igg 00 87 14.2 1&E 00 1a6 1&7 mmma C 0 C 8 0 P 8 Y E P T 1&S 10 1E8 118 11.0 00 1&7 1&G 18 Q0 11 0 C P T T G 0 T T P T P j 00 00 00 00 140 Q0 00 1&S 11.0 11.0 17.5 amme 8 E O E P O C 0 P E P O 115 80 00 as QO 11.0 1a8 147 00 14 00 tai l

B C E P 0 0 P E D 0 8 P l 0 8 112 118 3 G0 1&2 1&E 00 g 1&1 115 00 y 1&S A E D G C B D P S S 0 P 8 1a4 as as no its 1s7 1a1 g a0 1a3 tid 1a4 5 c0 its A 0 P e O P ATT TT T T T P 00 ist 13 0 00 167 Q0 S8 QQ to 00 1ce C0 1&2 eurm 4 F 0 F 0 F C F 0 F C P T 1&2 40 11.0 g 40 Ee 3 C0 17.2 00 19 QQ ita 00 M7 8 9 0 8 O P O 8 0 P O 8 40 Me 110 14 1 111 11 8 GO 147 its _14 5 _00 _ _118 FUEL TYPES i

A - EXXON VB E - P8X8R (GE199)

B - P8X8A (E 139) F - 28X8EB (GE 3.21)

C - P8X8A (E 165) G - P8X8R (GE 199) o - P8X8A (GE 199)

TR No. 049 Rev. O Page 14 of 47 3.0 NUCLEAR CHARACTERISTICS OF THE CORE This section presents the results of core calculations on power distribution, shutdown margin and core average reactivity coefficients.

The nuclear evaluation methods used in this section include fuel lattice ( and 3-dimensional core steady-state analyses (. The fuel lattice methods are used to calculate fuel bundle nuclear parameters such as reactivities, relative rod powers and 2 or 4 group cross sections.

The 3-dimensional reactor simulator code, N00E-B/ THERM-8, calculates power and exposure distributions and core thermal-hydraulic characteristics. The reactor simulator code also calculates cold shutdown margin and hot excess reactivity. The reactor simulator code and other codes are used to calculate the core reactivity parameters (.

3.1 Core Power Distributions The cycle was depleted using N00E-8/ THERM-B( to give both a rodded depletion and an All-Rods-Out (ARO) Haling depletion. The Haling depletion serves as the basis for defining core reactivity characteristics for most transient and accident evaluations. This is due primarily to its flat power shape which has conservative scram characteristics. Because of the more realistic prediction of initial CPR values, the rodded depletions were used to evaluate the Fuel Assembly Mislocation error and Loss of Feedwater Heating").

The rod patterns were developed such that the overall exposure distribution at EOC 12 is similar to the Haling. A comparison between the EOC Hal.ing and R0dded depletion exposure distribution is shown in Figure 3.1.

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TR No. 049

.Rev. O Page 15 of 47 Thermal limits evaluations were performed at each exposure point throughout the rodded depletion analysis. The results demonstrate that Cycle 12 reference core design is operationally manageable throughout the cycle with no indications of a power derate.

i 3.2 Core Reactivity The core reactivity increases during the first part of the cycle due to gadolinium burnup resulting in a maximum uncontrolled reactivity to occur at about a cycle energy of 6.0 GHD/MT. Thereafter, the reactivity decreases to expected E'OC 12 energy of 10.716 GHD/MT.

Figure 3.2 provides a hot excess reactivity curve as a function of cycle exposure for Cycle 12.

3.3 Shutdown Margin The plant Technical specification establishes a shutdown margin (SOM) design requirement of 1.0% aK to ensure that the core could be made suberitical at any time during the operating cycle with the strongest operable control rod fully withdrawn and all other rods 1

fully inserted. The Cycle 12 reference loading pattern fully meets the SOM design requirements. The minimum SOM for Cycle 12 is 1.6%

AK and occurs at BOC. Table 3.1 tabulates data for the SDM )

j calculation and Figure 3.3 provides SDM as a function of cycle l

exposure.

3.4 Reactivity Coefficients The Cycle 12 delayed neutron fraction (/5eff), neutron lifetime (t*), fuel temperature and moderator void reactivity coefficients are presented in Table 3.2.

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TR No. 049 Rev. O Page 16 of 47 3.5 Scram Reactivity The calculated full power EOC 12 scram curve is presented in Figure 3.4. The scram curve is based on an initial all-rods-out control rod configuration.

3.6 Liquid Poison System The shutdown margin of 600 ppm boron calculated for Cycle 12 core is 3.4% SK. This value applies to the core at cold, uncontrolled, xenon free, and peak reactivity conditions. This meets the design requirement of 1.0% aK SDM.

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TR No. 049 Rev. O Page 17 of 47 TABLE 3.1 Shutdown Margin Calculation BOC K rr - Uncontrolled 1.10861 BOC K.rr - Controlled 0.94511 Cold Critical K cr (with uncertainty) 0.99633 BOC K rr - Controlled 0.98021 (Strongest Worth Rod Withdretn)

BOC Minimum SOM 1.61% AK R-Value, Maximum Increase 0.0 in Cold K.cr witi Exposure s

=

k j 3100' I

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TR No. 049 Rev. 0 ,

Page 18 of 47 .I TABLE 3.2 Cycle 12 Reactivity Coefficient Parameters Time in Cycle Value Delayed Neutron BOC 0.00612 Fraction (6,r,) EOC 0.00529 Neutron Lifetime BOC 38.0 (t* - p sec) EOC 40.7 Doppler Coefficient BOC -1.82 E-5

((AK/K)/'F) EOC -1.83 E-5 Void Coefficient BOC -12.9 E-4

((aK/K/% void) at EOC -11.1 E-4 Core Average Volds l

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TR No. 049 Rev. O Page 23 of 47 4.0 THERMAL AND HYDRAFLIC DESIGN The objective of the thermal and hydraulic design analyses is to assure that the reload fuel can meet normal steady-state and transient performance requirements without exceeding thermal and hydraulic design limits, and that the reload fuel is compatible with the existing fuel in the core.

The transient performance of the reload fuel is discussed in Section 5.0.

4.1 Steady State Evaluation Core steady state thermal-hydraulic analyses were performed using the NODE-8/ THERM-B ccmouter code . The code utilizes bundle specific thermal-hydraulic characteristics to calculate bundle power and flow distributicns. The 3 dimensional bundle power and ficw distributions were performed at different exposures throughout the cycle. As stated in Section 3.1, the core can be operated at full power within thermal limits. This demonstrates the thermal-hydraulic compatibility of the different Cycle 12 fuel types under steady-state conditions.

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TR No. 049 Rev. O Page 24 of 47 4.2 Hot Channel Evaluation for system transients analyzed with RETRAN, a hot channel model is used to determine the fuel bundle thermal hydraulic condition during the course of the transient for use in the CPR calculation. Hot channel models for P8X8R and GE8X8E8 fuel types were set up to distinguish thermal-hydraulic behavior between different fuel types, The results are given in Section 5.0. A hot channel for EXXON VB fuel was not used since these bundles will reside on the core periphery in a low power region. The initial steady-state conditions for the hot channel model are shown in Table 4.1 which are based on EOC NODE-B/ THERM-B calculations.

4.3 Local Linear Heat Generation Rate (LLHGR)

The LLHGR operating limit for the new GE8X8EB fuel design is 13.4 KH/FT, the same as previous GE fuel designs used in the Oyster Creek Core.

4.4 Fuel Cladding Integrity Safety Limit The minimum critical power ratio (MCPR) fuel cladding integrity safety limit is determined using the General Electric Company Thermal Analysis Basis, GETAB('*', which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the GE critical quality (X) boiling length (L) GEXL correlation.

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!R No. 049 Rev. O Page 25 of 47 A safety limit MCPR of 1.07 was established for both P8X8R and ENC VB fuel designs for the Cycle 10 reload (. A MCPR safety limit of 1.07 can be conservatively applied to the Cycle 12 core with the GE8X8EB fuel design. A generic 1.04 MCPR fuel cladding integrity safety limit was approved' to be applied to the second successive reload core of P8X8R, BP8X8R, GE8X8R or GE8X8EB fuel designs with an initial bundle R factor >l.04. Gyster Creek Cycle 12 reload will meet this criteria which will make the safety mit of 1.07 conservative.

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TR No. 049 Rev. O Page 26 of 47 Table 4.1 Hot Channel Transient Analysis Initial Condition Parameters Operating Conditions:

100.0% Power - 100.0% Flow Core Power, MHT 1930.0 Core Flow, MLB/HR 61.0 Reactor Mid-core Pressure, PSIA 1050.0 Inlet Enthalpy, BTU /LB 518.2 Axial Peaking Factor 1.40 APTM Exposure: EOCil P8X8R GE8X8E8 Peaking Factors: (Local) 1.20 1.28 (Radial) 1.683 1.609 R-Factor 1.051 1.10 Bundle Poucr. MHT 5.21 4.96 Bundle Flow, 1000 Lb/Hr 90.22 92.75 Initial MCPR 1.446 1.446 3100C

TR No. 049 Rev. O Page 27 of 47 5.0 ACCIDENT AND TRANSIENT ANALYSIS 5.1 Accident Analysis The incidents and accidents discussed and analyz?d in the following sections are initiated by events external to the fuel core. In each case, the ar.z. lysis addresses the response of the fuel and/or the reactor to the occurrence of the initiating event which has been previously identified as an appropriate incident for analysis for this reactor.

5.1.1 Loss-of-Coolant Accident Cycle 12 introduces new LOCA 'ethodology and limits for the GE fuel designs. Reference 9 is included as part of the reload submittal package.

5.1.2 Fuel Hisloading

a. Fuel Assembly Hisorientation The fuel assembly misorientation analysis evaluates the consequences of changes in the local pin power distribution 7.nd bundle reactivity due to a fuel assembly loaded in its correct location, but 180' from its proper orientation. Table 5.1 provides the limiting change in critical power ratto (aCPR) for the fuel types used in Cycle 12. The oCPR for this accident includes a 0.02 adder due to uncertainty in the avtal R factor.

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TR No. 049 Rev. O Page 28 of 47

b. Fuel Assembly Mislocation The fuel assembly mislocation analysis evaluates the consequences of changes in the local pin power distribution and bundle reactivity due to a fuel assembly load in an improper location of the core. Eighteen control cells were analyzed for this event as described in Reference 6. The largest change in CPR for this accident in Cycle 12 is 0.2;.

5.1.3 Control Rod Drop Accident Startup rod withdrawal sequences performed at Oyster Creek during Cycle 12 will comply with the requirements of the General Electric banked position withdrawal sequence (.

Adherence to BPHS ensures that the worth of any in-sequence control rod is limited such that, during a rod drop accident, the calculated peak fuel enthalpy is not greater than the 280 cal /gm design limit.

5.2 Transient Analysis The transient analysis establishes the limiting values of overpressure and overpower for the Oyster Creek plant during Cycle 12. The potentially limiting events to be evaluated are identified in Reference 11 for GE Reload Fuel. Reference 10 documents the specific events for the Oystar Creek plant. Analyzed events are the loss of 100*F feedwater heating (LFWH), Turbine Trip Hithout Bypass (TTWOBP), Main Steam Isolation Valve Closure with No 3100C l

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TR No. 049 Rev. O Page 29 of 47 Scram (MSIVN), Feedwater Controller Failure (increasing flow),

(FHCF), and Control Rod Withdrawal Error (RHE). The models used for the various transients are as follows: the N00E-B Static Simulator Model is used for the LFWH and RHE(, the RETRAN model with one dimensional kinetics is used for the TTHOBP, MSIVN and FHCe t')

The analyses assuna limiting initial conditions and equip?t ,

performance determined by GPUN based on Reference 14.

5.2.1 Turbine Trip Hithout Bypass The TTHOBP characterized by the sudden closure of the turbine 4 stop valve and the failure of the bypass valves to open, resulting in a void collapse and a sharp power increase. The response of significant plant parameters during the transient is shown in Figures 5.1 and 5.2 while peak power, heat flux, and pressure are summarized in Table 5.2. The limiting ACPR is shown in Table 5.1 for end of cycle conditions.

5.2.2 Loss of Feedwater Heating The loss of feedwater heating analysis determines the change in the critical power ratio (ACPR) as a result of cooler water entering the core. The analysis considers a 100'F decrease in feedwater inlet temparature and a 101 increase in feedwater flow. Reactor power is increased to 115.7 percent (APRM scram limit) and thermal margins are evaluated assuming steady state conditions. Table 5.1 provides the aCPR for this transient. The peak LHGR is 17.d KH/ft which is less than the transient limit"'

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TR No. 049 Rev. O Page 30 of 47 5.2.3 Feedwater Controller Failure (Max. Demand)

This transient is characterized by the sudden failure of the feedwater controller resulting in a maximum feedwater demand of 120% rated followed by a Turbine Trip on high water level. The response of significant plant parameters during the transient is shown in Figures 5.3, 5.4 and 5.5 while peak power, heat flux and pressure are summarized in Table 5.2.

The limiting ACPR is shown in Table 5.1 for EOC conditions.

5.2.4 Main Steam Isolation Valve Closure with No Scram In this limiting overpressurization transient the MSIV closes in 3 seconds but the reactor fails to scram. The void collapse results in a power increase followed by a pressure increase which is controlled by the safety valves. The response of significant plant parameters is shown in Figures 5.6 and 5.7 . The peak vessel pressure of 1305 psia is well below the vessel pressure safety limit of 1390 psia.

5.2.5 Control Rod Withdrawal Error The control rod withdrawal error (RHE) event is initiated by centinuously withdrawirig a control rod at its maximum withdrawal rate.

M Two withdrawal error analyses were performed to determine which control rod gave the largest change in thermal limits.

The first analysis looks at the highest worth control rod as the transient rod.

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TR No. 049 Rev. O Page 31 of 47 The second analysis uses a transient rod location that will result in a poor response of the .,vRM system and also has a high rod worth. The most limiting results for Cycle 12 were obtained in the highest worth control rod analysis. Figure 5.8 provides the-control rod pattern used in this analysis.

The APRM response, and hence the rod block effectiveness, versus transient rod position will vary based upon the number of available LPRMs feeding the APRM. Three APRM status conditions have been defined and the APRM response-to control rod withdrawal is displayed in Figure 5.9. The results of the most limiting APRd response (Status 1) are shown in Table 5.3. Since the RHE is not the limiting CPR transient for Cycle 12 all three APRM status conditions will have the same operating CPR limit (1.50). The peak MLLHGR remains well within limits.

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1 TR No. 049 Rev. O Page 32 of 47 Table 5.1 CPR Analysis Summary DELTA CPR TRANSIENT Exposure SOC to E0C Fuel Loading Error (Mislocated) 0.21 Fuel Loading Error (Rotated) 0.25*

Loss of 100*F FH Heating 0.13 Exposure: Peak Cycle Reactivity Rod Withdrawal Error (108%) 0.38 Exposure: EOC FH Controller Failure 0.27**

Turbine Trip H/0 Bypass 0.37**

Includes 0.02 adder.

• • Does not include statistical multi.' lier.

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TR No. 049 Rev. O Page 33 of 47 Table 5.2 Core-Wide Transient Analysis Results Peak Neutron Flux in % Peak Heat Flux Peak Steamline Peak Vessel Transient of Initial 1 of Initial Pressure (PSIA) Pressure (PSIA)

LFHH 115.7 113.4 1035. 1065.

TTNBP 627 136 1277. 1284.

FHCF 320 125 1148, 1175.  :

l l

l l

I

\

l l

l I

3100C l i

TR No. 049 Rev. O Page 34 of 47 TABLE 5.3 aCPR AND MLLHGR FOR CONTROL R00 WITHDRAWAL ERROR (MAXIMUM ALLOWABLE FAILURE COMBINATION)

APRM POWER HLLHGR ROD WITHDRAWAL (PERCENT) SCPR (KW/FT) (FEET) 100.0 0.0 13.4 0.0 101.0 0.09 14.0 2.5 102.0 0.13 14.3 3.0 103.0 0.17 14.5 3.5 104.0 0.21 14.8 4.0 105.0 0.24 14.8 4.5 106.0 0.27 14.7 5.0 107.0 0.31 14.5 6.0 108.0* 0.38 14.0 8.5 109.0 0.39 13.9 9.0 110.0 0.40 13.9 9.5 APRM Rod Block  !

l l

i  ;

l l

i 3100C l l

m = 627 % CYCLE 12 TURBINE TRIP

+-+

8 ' ' '

l '

i '

m l l

~

FMIID VALUES

+ POWER 1930 M. W. .

~

8

~

X CORE FLOW 61 M.LB. / HR.

S 2 A HEAT FLUX 1.218 x 10 BTil / HR -FT -

g o ,

W 88 -

U a .

E g - u I g '

i ,' l 222

% ." 2 o

1.6 3.2 4.8 6.4 8.0 o 0.0 TIME o,, $

FIGURE 5.1 - TURDINE TRIP WITHOUT BYPASS. POWER CORE FLOW. HEAT FLUX.

___ ___a _ _ _ _ . . _ _ _ . _,__u______

-_ ___m .___ _ _._.____- _- _ _ - - - - - ._ - -. ______m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

CYCLE 12 TURBINE TRIP g

g

..___.__7_____.7-.____,._.______..-_.---r-----p- -. -------g --r---

RATTD VALIN S

+ PRESS RIS E -

g 2- X RELIEF VALVE FLOW 2015.1 LBM / SEC

~

m m 1 T T E $

N N . .

% , e

- - - A fil G 1 y a et b

U aaJ O b ~

te la E=E= - -

1 l _ -em x _ *t - x x x *:

28%

o o; . x x .x x x J _

_2 I ._ t_ I i I -' 'S ? g 0.0 1.6 3.2 4.8 6.4 8.0 M O 3*

TIME '

• t l FIGURE 5.2 - TURBINE TRIP WITHOUT BYPASS DOME PRESSURE RISE. REUEF VALVE FLOW.

CYCLE 12 FWCF g g l l l '

l '

FHHD VAI.UES _

+ LIQUID LEVEL CH ANG E [ W ) -

o o _ X FEEDWATER FLOW 20151 LBM / SEC. -

e m A VESSE L STE A M FLOW 20151 LBM / SEC.

o o  : x x x x x x x x x x x x x x x x x >

2 2

_ a  :  : 2 ., .

W N s _

u.

o 8 g 8 a

5 f

_a ~

E 3 h e a '

a g -

o I , I e I i  ! ' EEE o o l 40

%5 2 8 16 24 32 ,o o

0 ~

TIE o y FIGURE 5.3 - FEEDWATER CONTROLLER FAILURE ( MAX. DEMAND ), o FEEDWATER FLOW STEAM FLOW. LEVEL CHANGE.

m CYCLE 12 FWCF ,

e 8

cu l

I i '

I

_ FM1ID val UES -

+ POWER 1930 M. W.

S 2 o - X HE AT FLUX 1218 x10 BIU / HH -FT -

so A CORE FLO W 61 htLB. / INL

_ O Cone INLET SUSCOOLING 3 2. 8 "F _

I g -

o a c c c m m m m m g ,.- _

__ _ E_ w _ P_ - _ _. _7 _7 _7 _7 7_ 1 _

O

~

Ei @

$2 I'$ -

a W

? -

\

o i i i i . I ... I i j"2 0 8 16 24 32 40 *~5 go o

TIME o FIGURE 5.4 - FEEDWATER CONTROLLER FAILURE ( MAX. DEMAND ). POWER, 2*

HEAT FLUX. CORE FLOW. CORE INLET SUBCOOLING.

CYCLE 12 FWCF l ' '

I l

$ @ I FMTED VAttES o o _ + PRESS RIS E - _

X RELIEF VALV E FLOW 20111 L B M / SEC.

A BYP ASS VALV E FLOW 20151 L B M / SEC. _

R R

~

s =

~

f

$ 'l fX m a n! a

- l E

e un

. ~

~

5 E

• i i i i J d ,  : :  :  :  :  : _ i i i i i S S -

a s l i . I i I i ygg o o i i

o<

an no 8 16 24 32 40

• g

' 0 s.> O -

o TIME FIGURE 5.5 - FEEDWATER CONTROLLER FAILURE ( MAX. DEMAND ). "

DOME PRESS. RISE. RELIEF VALVE FLOW, BYPASS FLOW.

CYCLE 12 MSIVNSCR i i

' ' ' i I l l ImitD VALUES ,

+ PRESS RI95 -

o o _

X SAFETY VALV E FLOW 20151 L B M / SEC. .

E N o o _

Z Z s =

W E

a -

g8M8

-y 5 .

g m _

d $ x x x x x x ,

<" g g _

o o  :  :  :'  :  : - x 'x x -

1 x ' 1 ' I '

2 'E "

0.0 1.6 3.2 4.8 6.4 8.0 *;5 O

TIME o 2 FIGURE 5.6 - MAIN STEAM ISOLATION VALVE CLOSURE - NO SCRAM. 2*

~

DOME PRESS. RISE, SAFETY VALVE FLOW.

CYCLE 12 MSIVNSCH BAAX. = 329 % m m m o

g '

I l '

l '

l H4lfD VALUES -

+ HE AT FLUX 1.218 x 10 BYU / HR -FT ~

g -

Y X POWER 1930 bd. W.

~

A CO RE F L O W 6181LB. / HR.

g

g  :  ; - .

G E

Ei S

=

n_

? -

o i l i l i I . l i j"O 0.0 1.6 3.2 4.8 6.4 8.0 *'8 go-TIME o g FIGURE S.7 - MAIN STEAM ISOLATION VALVE CLOSURE - NO SCRAM, 2*

HEAT FLUX, POWER, CORE FLOW.

TR No. 049 Rev. O Page 42 of 47 Figure 5.8 CONTROL R00 PATTERN FOR RHE ANALYSIS 26 30 34 38 42 46 50 20 20 51 12 04 47 28 43 00* 16 04 39 24 20 35 12 04 12 31 04 24 16 20 27 NOTES: 1. Rod pattern is 1/4 core mirror symmetric.

2. No. Indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.

3. Asterisk (*) denotes the transient control rod.

l Core Exposure: 6.0 GHD/MT (peak cycle reactivity)

Reactor Power: 1930 MH(th)

Recirc Flow: 61.000 MLB/HR System Pressure: 1050 PSIA Inlet Subcooling: 34.5 BTU /lb 1

3100C l

l

I FIGUFE 5.9 APFN RESPONSE TO CONTRJL ROD WITHOR4WN_

APFN FEADING 1.12 1.10 -

1.08 -

mo a rrx , [

j 7 _ _

1.06 -

1.04 -

1.02 -

1.OG x ---

2 3 4 5 6 7 8 9 0 1 FEET WITHORAWN 3eFO

=

STATUS i STATUS 2 -*- STATUS 3 ;e

TR No. 049 Rev. O Page 44 of 47 G.0 OPERATING LIMIT HCPR The operating Ilmit MCPR for each transient is presented in Table 6.1 and that the limiting event is the Turbine Trip Hithout Bypass. The required MCPR operating limit for Cycle 12 is 1.50 based on a safety limit of 1.07, a ACPR of 0.37, and a statistical multiplier of 1.042.

7.0 STABILITY ANALYSIS According to Reference 17, Oyster Creek (as a low power density BWR/2) is exempt from the current requirement to submit a cycle specific stability analysis for its reload fuel. Ample stability margins to the 1.0 decay ratto criteria, as shown in the stability analysts for Cycle 10',

are typical for the Oyster Creek Plant.

8.0 TECHNICAL SPECIFICATIONS Based on the Cycle 12 reference core des!gn and safety analysis provided in this report, the following sections in the Technical Specifications will require modification.

Section 3.10.A (Average Planar LHGR): Add new limits for GE8X8E3 fuel l

and revise limits for P8X8R fuel designs. Four and five loop operation will use same MAPLHGR figures.

l Section 3.10.8 (Local LHGR): Add reference for new fuel design (GE8X8EB) to include LHGR limit of 1 13.4 KW/ft.

Section 3.10.C (Minimum Critical Power Ratto): Change MCPR Limit from 1.45 to 1.50 for each of the three APRM status levels.

The appropriate bases sections will also require modification.

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TR No. 049 Rev. O Page 45 of 47 TABLE 6.1 Cycle Operating Limit MCPRs Transient MCPR Fuel loading Error (Mislocated) 1.28 fuel Lodding Errcr (Mi$ orientated) 1.32 Loss of Feedwater Heating 1.20 Rod Withdrawal Error 1.45 FH Controller Failure 1.39 Turbine Trip w/o Bypass 1.50 1

3100C

TR No. 049 Rev. O Page 46 of 47

9.0 REFERENCES

1.0 H. Fu, R. V. Furia, "Methods for the Analysis of Boiling Water Ractors lattice Physics," TR 020-A, Rev. O, January 1988. <

2.0 Letter from J. N. Donohew, Jr. (NRC) to P. B. Fiedler (GPUN) dated November 14, 1986, "Reload Topical Report TR 020 (TAC 60339)."

3.0 R. V. Furia, "Methods for the Analysis of Boiling Water Reactors Steady State Physics," TR 021-A, Rev. O, January 1988.

4.0 Letter from A. W. Oromerick (NRC) to P. B. Fledler (GPUN) dated

- September 27, 1987, GPU Nuclear Corporation (GPUN) Topical Report TR 021, Revision 0, "Methods for the Analysts of Bolling Water Reactors Steady State Physics."

5.0 0. E. Cabrilla, et al., "Methods for the Generation of Core Kinetics Data for RETRAN-02," TR 033 Rev. O, February 1987.

6.0 E. R. Bujtas, et al., "Steady-State and Quast-Steady-State Methods Used in the Analysts of Accidents and Transients," TR 040, Rev. O, February 1987.

7.0 H. A. Alammar, et al., "BWR-2 Transient Analysis Model Using the RETRAN Code," TR 045, Rev. O, September 3, 1987.

8.0 Letter from H. Berkow (NRC) to J. S. Charnley (GE) dated December 3, 1985, "Acceptance for Approval of Fuel Designs Described in Licensing Topical Report NEDE-240ll-P-A-6, Amendment

10 for Extended Burnup Operation."

l l

9.0 "0yster Creek Nuclear Generating Station SAFER /CORECOOL/GESTR-LOCA l Loss-of-Coolant Accident Analysis," NEDC-31462P, August 1987.

10.0 "General Electric Reload Fuel Application for Oyster Creek," l NE00-24195, (As Amended). I 11.0 "General Electric Standard Application for Reactor Fuel,"

NEDE-240ll-P-A-8, May 1986.

12.0 "General Electric BWR Thermal Analysis Basis (GETAB): Data, i Correlation and Design Application," NE00-10958-P-A, January 1977.  !

l 13.0 "Banked Position Withdrawal Sequence," NE00-21231, January 1977.

14.0 "Guidelines for Generating OPL-3 Inputs," NEDE-22061, Feb. 1982, 15.0 Letter from W. A. Paulson (NRC) to P. B. Fiedler (GPUN), dated l August 27. 1984. "Core 10 Refueling."

I 3100C j

TR No. 049 Rev. O Page 47 of 47 16.0 Letter from A. C. 1hadant (NRC) to J. S. Charnley (GE) dated December 27, 1987, "Acceptance for Referencing of Amendment 14 to General Electric Licensing Topical Report NEDE-240ll-P-A, General Electric Standard Application for Reactor Fuel (TAC No. 60113)."

17.0 Letter from C. O. Thoms (NRC) to H. C. Pfefferlen (GE) dated April 24,1985, "Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6. Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II."

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- _ _ _ _ .