TR 049 Rev. O

TR 049Rev. O

RELOAD INFORMATION ANDSAFETY ANALYSIS REPORT

FOR

OYSTER CREEK CYCLE 12 RELOAD

TR-049Res. O

u

BA No. 335400

M. A. ALAMMARJ. D. DOUGHER

MARCH. 1988

APPROVALS:

W 3fZ/f38Manager,OysterCreek/FuelProjects ' 6 ATE

8M ANuclear Analysis & Fuels Dire'ctor ~

N ,P)DATE

CPU NUCLEAR1 Upper Pond Road

Parsippany. New Jersey 07054.

jar 4S88uBi88|8eg,

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TR No. 049Rev. O 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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TABLE OF CONTENTS

Section Page

1.0 INTRODUCTION AND SUMMARY .......................................... 6

2.0 CYCLE 12 REFERENCE CORE DESIGN .................................... 8

82.1 Operating History (Cycle 11) .................................82.2 Coro Loading Pattern (Cycle 12) ..............................

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 ............................ 244.4 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 ....................................... 275.1.3 Control Rod Orop Accident ............................. 28

5.2 Transient Analysis ........................................... 28

5.2.1 Turbine Trip Without Bypass ........................... 295.2.2 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. .. . . . . . .. .. .

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TR No. 049Rev. OPags 4 of 47

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LIST OF TABLES

TableNumber Title Page

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

3.1 Shutdown Margin Calculation 17

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 45i

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LIST OF FIGURES

Figure Title Page

2.1 Cycle 12 Reference Loading Pattern 13

3.1 EOC 12 Average Axial Exposure 19I

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 35Heat Flux

5.2 Turbine Trip Without Bypass, Dome Pressure 36Rise, Relief Valve Flow

5.3 Feedwater Controller Failure (Max. Demand), 37Dome Press. Rise, Relief Valve Flow, Bypass ,

Flow l

5.4 Feedwater Controller Failure (Max. Demand), 38Power, Heat Flux, Core Inlet Subcooling

5.5 Feedwater Controller Failure (Max. Demand), 39Dome Press. Rise, Relief Valve Flow,Bypass Flow i

5.6 Main Steam Isolation Valve Closure No 40 |Scram, Dome Pressure Rise, Safety Valve

,

Flow |1

5.7 Main Steam Isolation Valve Closure No 41 iScram, Heat Flux, Power, Core Flow

5.8 Control Rod Pattern for RHE Analysis 42

5.9 APRM Response to Control Rod Withdrawal 43

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TR No. 049 !

Rev. O l

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.

IThe 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.3100C

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TR No. 049 |

'Rev. O i 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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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

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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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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TABLE 2.1

Design Basis Cycle 11 and Cycle 12 Exposures

Projected E0C 11 Core 17.436 GWD/MTAverage Exposure

Projected BOC 12 Core 9.716 GWD/HTAverage Exposure

Haling Calculated E0C 12 20.432 GWO/MTCore Average Exposure

Projected Cycle 12 Full 10.716 GWD/HTPower Energy Capability

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TR No. 049Rev. O I

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TABLE 2.2

Cycle 12 Reference Core Description

Number of Control Rods 137

Number of Fuel Bundles 560Total Weight of U in Core (MT) 98.11

Bundle Batch Avg NumberType 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

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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 2.50w/o U-235

Pellet Density, 96.5 95.0 93.57. theoretical

Pellet Diameter, inches 0.411 0.410 0.4195

Fuel Rod

Active Length, inches 145.24 145.24 144.0Plenum length, inches 9.23 9.23 10.62Fuel Rod Pitch, inches 0.640 0.640 0.642Diametral Gap (cold), 0.008 0.009 0.010

inchesFill Gas Helium Hellum Helium

Cladding

Material Zr-2 Zr-2 Zr-2Outside Olameter, inches 0.483 0.483 0.5015Thickness, inches 0.032 0.032 0.036Inside Olameter, inches 0.419 0.419 0.4295

Fuel Channel

Material Zr-4 Zr-4 Zr-4Inside Olmension, Inches 5.278 5.278 5.278Wall Thickness, inches 0.080 0.080 0.080

Fuel Assembly

Fuel Rod Array 8X8 8X8 8X8Fuel Rods per Assembly 60 62 60Spacer Grid Haterial Zr-4 Zr-4 Zr-4Flow Area, FT' O.1099 0.1099 0.1047

2Heat Transfer Area. FT 91.83 94.90 94.53

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TR No. 049 '

Rev. OFIGURE 2,1 of 47

CYCLE 12 REFERENCE LO ADING P ATTERN

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mmma

C 0 C 8 0 P 8 Y E P T1&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 j1&S 00 00 11.0 00 11.0 00 140 Q0 17.5 00 -

amme

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|B C E P 0 0 P E D 0 8 P | 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 81a4 as as no its 1s7 1a1 g a0 1a3 tid 1a4 5 c0 its

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8 9 0 8 O P O 8 0 P O 8

its 40 Me _14 5 110 _00 14 1 111 11 8 GO 147 _118_

FUEL TYPESi

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)

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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 3.1.3100C

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TR No. 049.Rev. O 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.,

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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 rods1

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 ),

calculation and Figure 3.3 provides SDM as a function of cyclej

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exposure.

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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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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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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 0.0in Cold K.cr witi Exposure

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TR No. 049Rev. 0 ,

of 47 .I

TABLE 3.2

Cycle 12 Reactivity Coefficient

Parameters Time in Cycle Value

Delayed Neutron BOC 0.00612Fraction (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-4Core Average Volds

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FIGURE 3.3CYCLE 12 MINIMUM SHUT [XMN MAFGIN

SHUTDCMN MAFEIN (% DELTA K)'

2.45,

2.30 -

2.15 -

2.00 - -

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i.85 - -

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1.70 --

1.55 -

1.40 -

1.25 -

1.10 -

0.95 ,-

0.80 AIE' ' ' ' ' ' ' ' '

O i 2 3 4 5 6 7 8 9 10 *;pCrCLE EXPWJRE LGWD/MTU) a3

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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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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. 049Rev. O 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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Table 4.1

Hot Channel Transient AnalysisInitial 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

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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.3100C

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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 No3100C

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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

t')dimensional kinetics is used for the TTHOBP, MSIVN and FHCe

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 4stop 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''"'3100C

. _ _ _ _ _ _ _ .

.. . . .. ._

.


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.3100C

r

.

.

.

.

_ _ _ _

,


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.

.

3100C

i.

m.m.

-- _ __- _ n

1


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.!

3100C


Table 5.2

Core-Wide Transient Analysis Results

Peak NeutronFlux 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. :

|

|

||

I\

|!

||

I

3100C |

!!

,

i


TABLE 5.3

aCPR AND MLLHGR FOR CONTROL R00WITHDRAWAL 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

|i ;

|

!

li

3100C |

:

|_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ . - - _ _ _ - - - - _ - - _ __ _ _ - _ _ _ - - _ _

_ _ _ _ _ _ _ _ _ _

m = 627 % CYCLE 12 TURBINE TRIP

+-+8 ' ' ' 'l | l i'm

~-

FMIID VALUES

+ POWER 1930 M. W. .

8 -

X CORE FLOW 61 M.LB. / HR.~

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

.-

g _-

_

o ; ;-

,

W&

-

88 -

%U .

-

aE

g - u

-

_

i ,' | 222' ' -

g ',

I ' % ." 2'o

0.0 1.6 3.2 4.8 6.4 8.0 o

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 ~

1m m-

:

- -

T T E $

. . %N N , e.

A-- -

fil G 1

y aet

bU aaJ

b ~

O

te la

E=E= - -

1

| _ -em x _ *t - x x x *:

28%I I -' 'S ? gJ I ._ t_ i

o o; . x x .x x x __2

0.0 1.6 3.2 4.8 6.4 8.0 MO 3*

TIME ' *

t

| FIGURE 5.2 - TURBINE TRIP WITHOUT BYPASS DOME PRESSURE RISE. REUEF VALVE FLOW.

_ . .

.

. _ _ _ . ._.

CYCLE 12 FWCF

| ' | l l' ' ''g g

_ FHHD VAI.UES _

+ LIQUID LEVEL CH ANG E [ W ) -

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

e mA 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. 8 g 8 -

oa

5 f_a >

E 3~

h.

'e a-

a g -

o

-

_

I ! EEEI I e i ',

o o | %5 2

0 8 16 24 32 40 o,o~

yTIE o

FIGURE 5.3 - FEEDWATER CONTROLLER FAILURE ( MAX. DEMAND ), o"

FEEDWATER FLOW STEAM FLOW. LEVEL CHANGE.

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _

m

CYCLE 12 FWCF,

e

8 | I i I' ' ' ''

cu

_ FM1ID val UES -

+ POWER 1930 M. W.S 2

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

soA CORE FLO W 61 htLB. / INL*

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

Ig -

o a c c c

m m m m m

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

_.g ,. _ ___ __ _ _

O&

~

Ei @-

$2

I'$ -

-

aW

? - \

-

_

*

j"2Ii i I . . . ii .

io

0 8 16 24 32 40 *~5go

oTIME o

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

HEAT FLUX. CORE FLOW. CORE INLET SUBCOOLING.

.

.

. . ..

.

_ _ . _ _ _ _ _ - .

_

CYCLE 12 FWCF

'$ @ l I 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 fXm a n! a - l

Ee

~.

un

5 E~

d : : : : : : i i i i i i i i i J* _

,

_

S S -

a s

.

yggi I I i

l i . i

i o<o o' ' 0 8 16 24 32 40 g*an no

s.> O -*

oTIME' *

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

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

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-

.___ _

,

CYCLE 12 MSIVNSCR

$ $ i i' | |' i' I

_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 -

-y5 .

m _g _

d $ x x x x x x ,

>--

<" g g _

_

_

"""1 1 I 2 'E "' ' '

o o : : :' : : x 'x x x--

*;50.0 1.6 3.2 4.8 6.4 8.0,O

2TIME o

FIGURE 5.6 - MAIN STEAM ISOLATION VALVE CLOSURE - NO SCRAM. 2*!~|DOME PRESS. RISE, SAFETY VALVE FLOW.|

._-___-_- - _ _ _ _ _ _ _ _ _ - _ _ _ _ ._ _ .

_ _ _ _ _ .

CYCLE 12 MSIVNSCHBAAX. = 329 % m m m

og I 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 : ; = = -

-

- .

GE

Ei S -

-

=-

n_

? -

-

_

j"OI ll l i . iiio

0.0 1.6 3.2 4.8 6.4 8.0 *'8go-

gTIME o

FIGURE S.7 - MAIN STEAM ISOLATION VALVE CLOSURE - NO SCRAM, 2*''

HEAT FLUX, POWER, CORE FLOW.


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 awithdrawn rod.

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

| 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

|

|!

__

I

FIGUFE 5.9APFN RESPONSE TO CONTRJL ROD WITHOR4WN_

,.

APFN FEADING1.12 .

1.10 -

* .

mo a rrx , [ - *,

7 j1.08 - . .

_ _

1.06 -

1.04 -

1.02 -

, , , . . . . .

1.OGx .- ---

0 1 2 3 4 5 6 7 8 9

FEET WITHORAWN 3eF*

O

STATUS i STATUS 2 -*- STATUS 3 ;e'=

-

___ _ _ _ _ _ _ . _ _ . _ _ _ ___ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _


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|

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.3100C

_. - _ _ _ .


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


9.0 REFERENCES

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

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

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

,

4.0 Letter from A. W. Oromerick (NRC) to P. B. Fledler (GPUN) datedSeptember 27, 1987, GPU Nuclear Corporation (GPUN) Topical Report-

TR 021, Revision 0, "Methods for the Analysts of Bolling WaterReactors Steady State Physics."

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

6.0 E. R. Bujtas, et al., "Steady-State and Quast-Steady-State MethodsUsed 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 theRETRAN Code," TR 045, Rev. O, September 3, 1987.

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

; 10 for Extended Burnup Operation."|

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

10.0 "General Electric Reload Fuel Application for Oyster Creek," |

INE00-24195, (As Amended).

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 |August 27. 1984. "Core 10 Refueling."

I

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:

_ _ _ - - - _ _ - - - _ _ . - - - _ _ _ _ - - - - - _ - _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ - - _ __ _ ._ _ _ _ . _ _ _ _ _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..

.. ,


16.0 Letter from A. C. 1hadant (NRC) to J. S. Charnley (GE) datedDecember 27, 1987, "Acceptance for Referencing of Amendment 14 toGeneral Electric Licensing Topical Report NEDE-240ll-P-A, GeneralElectric Standard Application for Reactor Fuel (TAC No. 60113)."

17.0 Letter from C. O. Thoms (NRC) to H. C. Pfefferlen (GE) dated April24,1985, "Acceptance for Referencing of Licensing Topical ReportNEDE-24011, Rev. 6. Amendment 8, ' Thermal Hydraulic StabilityAmendment to GESTAR II."

3100C

- _ _ _ _ .

TR 049 Rev. O

Documents

Transcript of TR 049 Rev. O