ML20206J212

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Proposed Tech Spec Changes to Sections 2.3.A.1,2.3.B & 3.10.C & Figures 3.10-4 & 3.10-5 to Support Cycle 11 Reload
ML20206J212
Person / Time
Site: Oyster Creek
Issue date: 06/17/1986
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20206J177 List:
References
NUDOCS 8606270036
Download: ML20206J212 (22)
v • d • e


Text

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APPENDIX 0 CYCLE 11 RELOAD LICENSE SUBMITTAL D-1/D-11 8606270036 e60617 ADOCK0500g9 PDR P

U l

D.1 REFERENCE CORE LOADING PATTERN (1.0, 2.7. 3.3.1 AND 4.0)*

Fuel Type Group Number Irradiated Exxon Type VB A

196 Exxon Type VB B

28 P80RB239 C

112 P80RB265H D

60 New P80RB265H E

4 P80RB299ZA F

48 P80RB299Z G

112 Total 560 Nominal previous cycle core average exposure at end of cycle:

14,856 MHd/t Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:

14,856 mwd /t Assumed reload cycle core average exposure at end of cycle:

15,899 mwd /t Core loading pattern:

Figure 0-1 Sources of non-GE bundle characteristics:

Local Thermal-Hydraulic Nuclear Peaking Model (Fuel and Bundle Type _

Libraries Factor R-Factor Channels)

E4xon Type VB Developed via GE See Apo. 8 See App. B lattice methods NE00-24195 NED0-24195

0-1 I

D e

0.2 CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH -

NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

BOC k.ae Uncontrolled 1.104 Fully Controlled 0.946 Strongest Control Rod Out 0.975 R, Maximum Increased in Cold Core Reactivity with-0.003 Exposure Into Cycle, ak D.3 STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak) ggm (20*C, Xenon Free).

600 0.044 0.4 TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)

Vold Coefficient N/A* ( /% Rgo)

-6.37/-7.96 Void Fraction (%)

36.2 Doppler Coefficient N/A ( /*F)

-0.221/-0.210 Average Fuel Temperature (*F) 1138 Scram Worth N/A ($)

-37.64/-30.11 Scram Reactivity vs Time Figura 0-2 0.5 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

Peaking Factors (local, 1.20 1.28 radial and axial 1.683 1.609 1.40 1.40 R-Factor 1.051 1.098 Bundle Power (MWt) 5.653 5.416 Bundle Flow (10' lb/hr) 89.34 91.02 Initial MCPR I.35 1.32

  • N = Nuclear Input Data; A - Used in Transient Analysis 0-2

0.6 CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Exposure 4

Q/A ACPR Transient (mwd /T)

(% NBR)

(%)

P8x8R Ex8 Figure Turbine Trip E0C 520 122 0.28 0.25 Figure 0-3 without Bypass Loss of 100*F BOC to EOC 115 114 0.13 0.12 Figure 0 4 Feedwater Heating Feedwater Con-EOC 334 121 0.23 0.20 Figure 0-5 troller Failure 0.7 LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(5.2.1)

Limiting Rod Pattern: Figure 0-6 Reactor Rod Position ACPR MLHGR (kW/ft)

Power (%)

(Feet Withdrawn)

P8x8R Ex8 P8x8R Ex8 104 5.0 0.19 0.25 17.3 14.6 105 6.5 0.26 0.31 17.8 15.1 106 8.5 0.31 0.35 17.8 15.1 107 8.5 0.31 0.35 17.8 15.1 108**

9.0 0.32 0.36 17.8 15.1 109 9.5 0.34 0.36 17.8 15.1 110 12.0 0.40 0.37 17.8 15.1 0.8 CYCLE MCPR VALUES *** (5.2)

Non-pressurization Events Exposure Range:

BOC to EOC P8x8R Ex8 Loss of Feedwater Heater 1.20 1.19 Fuel Loading Error 1.21 N/A Rod Withdrawal Error 1.39 1.43

~

    • Indicates APRM rod block setpoint selected.

0-3

TJ Pressurization Events Exposure Range:

SOC to EOC Option A Option B P8x8R Ex8 P8x8R Ex8 Turbine Trip w/o Bypass 1.41 1.38 1.36 1.33 Feedwater Controller Failure 1.35 1.35 1.27 1.25 0.9 OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

's1

'v Transient (psig)

(psig)

Plant Response MSIV Closure 1261 1299 Figure 0-7 D.10 STABILITY ANALYSIS RESULTS (5.4)

Not required by reference.

0.11 LOADING ERROR RESULTS (5.5.4)

Rotated Bundle:

P8x8R ICPR:

1.21 (before rotating)

MCPR:

1.09 (after rotating) aCPR:

0.14*

  • Includes 0.02 ACPR penalty.

D-4

lag 52 523 j

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.S O = c00cE';c a =:G:.E;d-E = P80AS:65H l 8 = ECON-VS e = G*C.cE2Wa l

l C = :?O430%

Figure D-1.

Reference Core Loading Pattern for Cycle 11 f

0-5

4g j

00 CCNToot eCO CRIVE VS 7tVE SCRAM AEACTIVITV VS TIME 40 40 I

SC 35 l

57 A CRO IN PERCENT 70 30 60

  • S 3

y 3

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SCR AM CURVE TO USE3 :N ANALY$t$

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1 2

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TtVE inct Figure 0-2.

EOC Scram Reactivity 0-6 I

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. -. -. - - -, -.,. - - - -,. ~,.. -,. -

...,,.n..-..-

t VESSEL PRE E RISEtPSI)

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TIME ISEC08CSI ILEVELCINCH. PEF.SEP.SMRT) 1 V010 REACT;9 TV 2 VESSEL STEA FLOW 2 DCPPLER RE AE VITY 3 TURa!NE STE MFLOW 3 SCR apt REACTilv! Y I

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4. 8 8.(

f!PE (SECC@$1 TIME (SECO*CS)

Figure 0-3.

Plant Response to Turbine Trip Without Bypass D-7

u I WESiEL PRESS RISE (PSI)

I NEUIRON FLUX 2 AVE SURFACE HEAT FLUX 2 RELIEF VALVE Flow 3 BYPLSS VALVE FLOW 3 CORE [NLET FLOW

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0. 0 100.0 200.0 TIME (SECO2SI TIME (SECCN051 I vol ) REACTivlTY 1 lev &LC INCH. AEF-SEP.SnRT) 2 00P'LER REACTiv!TY 2 YE55EL STEAMFL0w 3 SCR LM RE ACT!v t TY 3 Tumb!NE STE AMFLOW is s. s

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e. e i e.e 2.e time (SECCND$1 TIME (SECCN05) l Figure D 4.

Plant Response to Loss of 100*F Feedwater Heating

)

i D-8 l

150.0 1NEurRON FLUX tVESEELPRES$ RISE (PSI) 2 AVE $LHFACE HEAT FLL) 2 $AFETY VALVE F(QW 3 RELIEF VALVE FLOW 3 CCR E INLET FLCW 4 87P 453 VALVE FLOW

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0. 0 20.0 40.0
0. 0 20.0 4 0. (-

TIME (SECCNOS)

TIME (SECONOS) 1 LEwEL(INCW GEF.SEP-SnRT) t v0!b REACT!v!TV 2 VES$EL STEeFLCW 2 COPPLER RE ACT!v!TY 3 TLEBINE STE AWLov 3 SCAPM QEacTgyg Ty

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0. 0 20.0 40.0
0. 8 20.0 40 IIME (SECON011 ftME (SECOND$1 Figure 0-5.

Plant Response to Feedwater Controller Failure D-9

2 6

10 14 18 22 26 30 34 38 42 46 50 51 10 10 10 47 10 8

8 10 43 10 20 8

20 10 39 10 8

24 24 8

10 35 10 20 20 31 8

24 0

0 24 8

27 10 8

8 10 23 8

24 0

0 24 8

19 10 20 20 15 10 8

24 24 8

10 11 10 20 8

20 10 7

10 8

8 10 3

10 10 10 NOTES:

1.

Rod pattern is full core.

2.

No. Indicates number of notches withdrawn out of 48.

Blank is a withdrawn rod.

3.

Error rod is (22,31).

Figure D-6 Limiting RWE Rod Pattern i

D-10

d

.t

' UTRCN duX 1 VE55EL PRESS RISE (PSI) 2A E SURFACE M AT FLUX 2 SAFETY valve FLOW 3 C RE INLET FLOW 3 RELIEF VALVE FLOW 308.8

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TIME (SECO W S)

ILEVELCINChDEF-SEP-SnRT) 1 VO!D REACN!vtTV 2 VESSEL STEAMFL0s 2 DCPPLER SE ACTIvtTY 3 fur 8 tNE SPEAMFLCW 3 SCR AM R[ AtitvlTY crrr-i'Em et:-

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Figure D-7.

Plant Response to MSIV Closure, No Scram 0-11

T Table 5-14a MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Oyster Creek Fuel Type:

P80R8239 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)

(kW/ft)

(*F)

Fraction 200 9.5 2198 0.095 1000 9.5 2198 0.095 5000 9.5 2194 0.085 10,000 9.5 2193 0.085 15,000 9.5 2194 0.085 20,000 9.0 2092 0.169 25,000 8.9 2048 0.169 30,000 8.9 2049 0.170 35,000 8.7 2050 0.170 40,000 8.3 2049 0.170

Table 5-14b MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE 4

Plant: Oyster Creek Fuel Type:

P80R8265H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)

(kW/ft)

(*F)

Fraction 200 9.5 2198 0.095 1000 9.5 2198 0.095 5000 9.5 2198 0.086 10,000 9.5 2198 0.086 15,000 9.5 2198 0.086 20,000 8.9 2078 0.170 25,000 8.8 2050 0.170 30.000 8.8 2044 0.166 35,000 8.6 2049 0.170 40,000 8.3 2050 0.170 i

i a

m-.-.

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Table 5-14c MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Oyster Creek Fuel Type:

P80R8299 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)

(kW/ft)

(*F)

Fraction 200 9.3 2198 0.092 1000 9.4 2198 0.091 5000 9.5 2199 0.085 10,000 9.5 2197 0.085 15,000 9.5 2198 0.086 20,000 8.8 2047 0.169 25,000 8.7 2046 0.168 30,000 8.7 2042 0.166 35,000 8.5 2047 0.169 40,000 8.2 2048 0.170 45,000 7.7 2046 0.167

2.3-1 2.3 LIMITING SAFETY SYSTEM SETTINGS Applicability:

Applies to trip settings on automatic protective devices related to variables on which safety limits have been placed.

Objective:

To provide automatic corrective action to prevent the safety limits from being exceeded.

Specification:

Limiting safety system settings shall be as follows:

FUNCTION LIMITING SAFETY SYSTEM SETTINGS A.

Neutron Flux, Scram A.1 APRM When the reactor mode switch is in the Run position, the APRM flux scram setting shall be S S[(0.90 x 10-6) W + 60.8] [ FRP ]

WEPU with a maximum setpoint of 115.7% for core flow equal to 61 x 106 lb/hr and greater, where:

1 S = setting in percent of rated power W = recirculation flow (1b/hr)

FRP = fraction of rated thermal power is the ratio of core i

thermal power to rated thermal power MFLPD = maximum fraction of limiting power density where the limiting power density for each bundle is the i

design linear heat generation rate for that bundle.

3454f

2.3-2 FUNCTION LIMITING SAFETY SYSTEM SETTINGS The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0 in which case the actual operating value will be used.

This adjustment may be accomplished by increasing the APRM gain and thus reducing the flow reference APRM High Flux Scram Curve by the reciprocal of the APRM gain change.

A.2 IRM 25 38.4 percent of rated neutron flux B.

Neutron Flux, Control Rod Block j

The Rod Block setting shall be I

S 1 [(0.90 x 10-6) W + 53.1 ] [ FRP ]

3 WEPU i

i with a maximum setpoint of 108% for core flow equal to

~i l

61 x106 lb/hr and greater.

The definitions of S, W, FRP and MFLPD used above for the APRM scram trip apply.

l The ratio of FRP to MFLPD shall be set equal to 1.0 unless j

the actual operating value is less than 1.0, in which case the actual operating value will be used.

4

}

This adjustment may be accomplished by< increasing the APRM l

gain and thus reducing the flow referenced APRM rod block curve by the reciprocal of the APRM gain change.

I 4

I I

j h

3454f

. t

2.3-6 4

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, i

prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range.

The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship.

i Therefore, the worst-case MCPR, which could occur during steady-state operation, is at 108% of the rated thermal l

l power because of the APRM rod block trip setting.

The l

actual power distribution in the core is established by j

specified control rod sequences and is monitored J

continuously by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting i

power density exceeds the fraction of the rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gains.

l The settings on the reactor high pressure scram, anticipatory scrams, reactor coolant system relief valves t

and isolation condenser have been established to assure never reaching the reactor coolant system pressure safety

' limit as well as assuring the system pressure does not l

exceed the range of the fuel cladding integrity safety limit.

In addition, the APRM neutron flux scram and the turbine bypass system also provide protection for these safety limits, e.g., turbine trip and loss of electrical load transients (5).

In addition to preventing power operation above 1060 psig, the pressure scram backs up the other scrams for these transients and other steam line isolation type transients. Actuation of the isolation condenser during these transients removes the reactor decay heat without further loss of reactor coolant thus protecting the reactor water level safety limit.

The reactor coolant system safety valves offer yet another protective feature for the reactor coolant system pressure safety limit since these valves are sized assuming no i

I i

3454f

3.10-3 C.

Minimum Critical Power Ratio (MCPR)

During steady state power operation, MCPR shall be greater than or equel to the following:

APRM Status MCPR Limit 1.

If any two (2).LPRM assemblies which 1.45 are input to the APRM system and are separated in distance by less than three (3) times the control rod pitch contain a combination of (3) out of four (4) detectors located in eithe.r the A and B or C and D levels which are failed or bypassed f.e., APRM channel or LPRM input bypassed or inoperable.

2.

If any LPRM input to the APRM system 1.45 at the B, C, or D level is f ailed or bypassed or any APRM channel is in-operable (or bypassed).

3.

All B, C, and D LPRM inputs to the 1.45 APRM system are fperating and no APRM channels are inoperable or bypassed.

When APRM status changes due to instrument failure (APRM or LPRM input failure), the MCPR requirement for the degraded condition shall be met within a time interval of eight (8) hours, provided that the control rod block is placed in operation during this interval.

For core flows other than rated, the nominal value for MCPR shall be increased by a factor of k, where kf is as shown in Figure 3.10-6.

f If at any time during power operation it is determined by nonnal surveillance that the limiting value for MCPR is being exceeded for reasons other than instrument failure, action shall be initiated to restore operation to within the prescribed limits.

If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue 3454f l

3.10-7 The APRM response is used to predict when the rod block occurs in the analysis of the rod withdrawal error transient. The transient rod position at the rod block and corresponding MCPR can be detemined.

The MCPR has been evaluated for different APRM responses which would result from changes in the APRM status as a consequence of bypassed APRM channel and/or failed bypassed LPRM inputs. The results for the reference cycle (3) indicate that the steady state MCPR required to protect the minimum transient MCPR of 1.07 is 1.23 or higher for the worst case APRM status condition ( APRM STATUS 1). This steady state limit conservatively applies to APRM status 2 and 3.

The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle.

In order to provide for a' limit which is considered to be bounding to future operating cycles, the limits for each APRM status condition have been conservatively adjusted upward to 1.45.

This is l

also the assumed value for LOCA analysis.

The time interval of eight (8) hours to adjust the steady state MCPR to account for a degradation in the APRM status is justified on the basis of instituting a control rod block which precludes the possibility of experiencing a rod withdrawal error transient since rod withdrawal is physically prevented. This time interval is adequate to allow the operator to either increase the MCPR to the appropriate value or to upgrade the status of the APRM system while in a condition which prevents the possibility of this transient occurring.

The steady-state MCPR limit was selected to provide margin to accommodate transients and uncertainties in monitoring the core operatl31ng state, manufacturing, and in the critical power correlation itself This limit was derived by addition of the CPR for the most limiting abnomal operational transient caused by a single operator error of equipment malfunction to the fuel cladding integrity MCPR limit designated in Specification 2.1.

Transients analyzed each fuel cycle will be evaluated with respect to the steady-state MCPR limit specified in this specification.

The purpose of the kf factor is to define operating limits at other than rated flow conditions. At itss than 100% flow the required MCPR is the product of the operating limit MCPR and the kf factor.

Specifically, the kf factor provides the required themal margin to protect against a flow increase transient.

l 3454f l

t FIGURE 3.10-4 MAXIMUM ALLOWABLE AVERAGE PLANAR LINEAR HEAT GENERATION RATE

(

FIVE LOOP OPERATION )

11.0 MP AL i

XA a-

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Amendment No. 75 AVERAG.E PLANAR EXPOSURE ( GWD / MT )

1 FIGURE 3.10-5 MAXIMllM ALLOWABLE AVERAGE PLANAR LINEAR HEAT GENERATION RATE

(

FOUR LOOP OPERATION )

11.0 MP i

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.m 0.0 10.0 20.0 30.0 40.0 50.0 :e Amendment No. 75 AVERAGE PLANAR EXPOSURE ( GWD / MT )

-

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