TR 049 Rev. O
Transcript of 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. OPage 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. 049Rev. OPage 3 of 47
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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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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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.
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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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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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TR No. 049Rev. OPage 10 of 47
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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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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CYCLE 12 REFERENCE LO ADING P ATTERN
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mmma
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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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ist 13 0 00 167 Q0 S8 QQ to 00 1ce C0 1&2 00
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8 9 0 8 O P O 8 0 P O 8
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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. OPage 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.,
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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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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 - -
. _
i.85 - -
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1.70 --
1.55 -
1.40 -
1.25 -
1.10 -
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0.80 AIE' ' ' ' ' ' ' ' '
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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.
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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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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
. _ _ _ _ _ _ _ .
.. . . .. ._
.
TR No. 049Rev. OPage 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.3100C
r
.
.
.
.
_ _ _ _
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TR No. 049Rev. OPage 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. 049Rev. OPage 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. 049Rev. OPage 33 of 47
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\
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i
TR No. 049Rev. OPage 34 of 47
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.
TR No. 049Rev. OPage 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 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'=
-
___ _ _ _ _ _ _ . _ _ . _ _ _ ___ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
TR No. 049Rev. OPage 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|
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
_. - _ _ _ .
TR No. 049Rev. OPage 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. 049Rev. OPage 46 of 47
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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_ _ _ - - - _ _ - - - _ _ . - - - _ _ _ _ - - - - - _ - _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ - - _ __ _ ._ _ _ _ . _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..
.. ,
TR No. 049Rev. OPage 47 of 47
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
- _ _ _ _ .