ML20106G487

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Rev 11C to Pilgrim Nuclear Power Station PNS Colr
ML20106G487
Person / Time
Site: Pilgrim
Issue date: 02/16/1996
From:
BOSTON EDISON CO.
To:
Shared Package
ML20106G485 List:
References
NUDOCS 9603050202
Download: ML20106G487 (25)
v • d • e


Text

_. _ _

L. -

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.- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 l '

(CYCLE 11) l APPROVED: - 'O'perations support Division Manager M /f/fd

' Date  !

APPROVED: dA ./ I suclear Engineering Sedices Dnanment Manager 'Date .

1 I

l APPROVED: AA'b.y Q OA eeNG 94-ot a/sr/9g Operat(pfs'Reviewtommittee Date

)

APPROVED: -

19 # b A // 94 D'p' 'ent Manager

[D/te '

APPROVED:

StAion eW ~ ~

eA 2//6['T /r-

' 'Dite j 4

s COLR REVISION llc PAGE I OF 25 9603050202 960227 PDR ADOCK 05000293 p PDN

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. O PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 TABLE OF CONTENTS EASC TITLE / SIGNATURE PAGE.. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 TAB LE OF C ONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . ......2  !

RECORD OF REVISIONS... . .. . .. .. . . . . . .. . . .. .3 .

1 LIST OF TABLES . . .... .... ... .. .. .. .. . . . ..... . . . . . .4 LIST OF FIGURES . . . . . . . . . . . . . . .. . . . . .. .. .5

1.0 INTRODUCTION

... .. ... .. . . ... . .. .. 6 i

2.0 INSTRUMENTATION TRIP SETTINGS . . . . .. .......7 2.1 APRM Flux Scram Trip Setting (Run Mode) .... . . .. . ... .7 2.2 APRM Rod Block Trip Setting (Run Mode).. .. ... .. . ......7 1

I 2.3 Rod Block Monitor Trip Setting., . . . . . . . . . . . . . . . . . . . . . . . . . . .8 3.0 CORE OPERATING LIMITS . . . .. .. .. .. . ... .9 3.1 Average Planar Linear Heat Generation Rate (APLHGR) .. . .. . . ... . . . .9 3.2 Linear Heat Generation Rate (LHGR) .. .. . .. . .. .9 3.3 Minimum Critical Power Ratio (MCPR) . . . . . . . .17 3.4 Power / Flow Relationship.. . . . . . 23 4.0 REACTOR VESSEL CORE DESIGN.. . . . . . .. . 23

5.0 REFERENCES

. . ... . . .. . . . . . . . 23 1

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l COLR REVISION 11C PAGE 2 OF 25 i

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a -,r ------ , , - -- y

J PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 l

RECORD OF REVISIONS Revision Effective Date Descriotion 8A Effective date based on Applicable for use during issuance officense Cycle 8 Operation amendment by NRC -

9A Effective date based on Applicable for use during issuance of ficense Cycle 9 operation amendment by NRC for ARTS and SAFER /GESTR 10A Effective date based on Applicable for use during initial startup of Cycle 10 Cycle 10 Operation 11 A Effective date based on Applicable for use during initial startup of Cycle 11 Cycle 11 Operation 1IB Effective upon final Applicable for use during approval Cycle 11 Operation 1IC Effectivt upon final Applicable for use during approval Cycle 11 Operation a

COLR REVISION 11C PAGE 3 OF 25

. 1

.- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 l LIST OF TABLES 1 l

Number Tjills Egg i

3.2-1 LHGR Operating Limits 10 3.3-1 MOC MCPR Operating Limits 19 l 1

3.3-2 EOC MCPR Operating Limits 20  :

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i i COLR REVISION 11C PAGE 4 OF 25 l

.- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 ,

i LIST OF FIGURES Number Iltli Ems 3.1-1 Maximum Average Planar Linear Heat Generation 11 Rate (MAPLHGR) for Fuel Type BP8DRB300 3.1-2 Maximum Average Planar Linear Heat Generation 12  ;

Rate (MAPLHGR) for Fuel Type BP8DQB323 3.1-3 Maximum Average Planar Linear Heat Generation 13 .

Rate (MAPLHGR) for Fuel Type BP8HXB355

  • 3.1-4 Maximum Average Planar Linear Heat Generation 14 ,

Rate (MAPLHGR) for Fuel Type BP9 HUB 378 .

3.1-5 Flow-Dependent MAPLHGR Factor (MAPFACF) 15 3.1-6 I Power-Dependent MAPHLGR Factor (MAPFACp) 16 3.3-1 Flow-Dependent MCPR I.imits (MCPRr) 21 3.3-2 Pcwer-Dependent MCPR Limits (MCPRe) 22 3.4-1 Power /Flo.v Operating Map 24 4.0-1 Reactor Vessel Core Loading Pattern 25 1

I COLR REVISION 1IC PAGE S OF 25 l

4

. .- PILGRIM NUCLEAR POWER STATION  !

PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02

1.0 INTRODUCTION

This report provides the cycle-specific limits for operation of the Pilgrim Nuclear  !

Power Station (PNPS) during Cycle 11. In this repon, Cycle 11 will frequently be i referred to as the present cycle.

Although this repon is not pan of the PNPS Technical Specifications, the Technical Specifications refer to this repon for the applicable values of the following fuel-related parameters:

4 Reference Technical Specification ,

APRM Flux Scram Trip Setting (Run Mode) Table 3.1.1 APRM Rod Block Trip Setting (Run Mode) Table 3.2.C-2 Rod Block Monitor Trip Setting Table 3.2.C-2 Average Planar Linear Heat Generation Rate 3.11. A Linear Heat Generation Rate (LHGR) 3.11.B l Minimum Critical Power Ratio (MCPR) 3.11.C l Power / Flow Relationship 3.11.D Reactor Vessel Core Design 5.2 If any of the core operating limits in this repon are exceeded, actions will be taken as defined in the referenced Technical Specification.

The core operating limits in this repon have been established for the present cycle using the NRC-approved methodology provided in the documents listed both in Section 5.0, References, and in Technical Specification 6.9.A.4. These limits are established such that the applicable limits of the plant safety analysis are met.

COLR REVISION 11C PAGE 6 OF 25 l

)

i

. .- PILGRIM NUCLEAR POWER STATION ,

PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02

, 2.0 INSTRUMENTATION TRIP SETTINGS:

1 2.1 APRM Flux Scram Trio Settina (Run Model Reference Technical Specifica. lions: Table 3.1.1,3.1.B.1 When the mode switch is in the run position, the average power range monitor (APRM) flux scram trip setting (S.) shall be:

Ss 0.66 W + 69%

with a clamp at 120% of rated core thermal power and S= APRM flux scram trip setting in percent of rated thermal I power (1998 MW). i W= Percent of drive flow required to produce a rated core flow of 69 Mlb/hr.

The APRM flux scram trip setting is valid only for operation using two recirculation loops. Operation with one recirculation loop out of service is restricted by License Condition 3.E.

In accordance with Technical Specification Table 3.1.1, Note 15, for no combination ofloop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

2.2 APRM Rod Block Trio Settine (Run Mode)

Reference Technical SpeciScations: Table 3.2.C-2,3.1.B.1 -

When the mode switch is in the mn position, the average power range l monitor (APRM) rod block trip setting (Sas) shall be:

Saa s 0.66 W + 62%

with a clamp at 115% of rated core thermal power and Sam = APRM rod block trip setting in percent of rated thermal power (1998 MW). i W= Percent of drive flow required to produce a rated l core flow of 69 Mlb/hr.

COLR REVISION 1IC PAGE 7 OF 25 i I

l PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 1

2.2 APRM Rod Block Trio Setting mun Mode) (Continued) l The APRM rod block trip setting is valid only for operation using two recirculation loops. Operation with one recirculation loop out of service is restricted by License Condition 3.E.

2.3 Rod Block Monitor Trio Setting l

R_eference Technical Specification: Table 3.2.C-2 j l

Allowable values for the power-dependent Rod Block Monitor trip '

)

setpoints shall be:  ;

1 Reactor Power, P Trip Setpoint j

(% of Rated) (% ofReference Level)

P s 25.9 Not applicable (All RBM Trips Bypassed) )

25.9 < P s 62.0 120 ]

62.0 < P s 82.0 115 l 82.0 < P 110 i i

The allowable value for the RBM downscale trip setpoint shall be 2 94.0%

of the reference level. The RBM downscale trip is bypassed for reactor power s 25.9% of rated.

COLR REVISION 1IC PAGE 8 OF 25  ;

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. .- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATLNG LIMITS REPORT RTYPE: G4.02 3.0 CORE OPERATING LIMITS 3.1 Average Planar Linear Heat Generation Rate (APLHGR)

Reference Technical Specification: 3.11. A During power operation, APLHGR for each fuel type as a function of axial location and average planar exposure shall not exceed the applicable limiting value. The applicable limiting value for each fuel type is the smaller of the flow- and power-dependent APLHGR limits, MAPLHGR, j and MAPLHGR,. The flow-dependent APLHGR limit, MAPLHGRr, is the product of the MAPLHGR flow factor, MAPFACF, shown in Figure 3.1-5 and the MAPLHGR for rated power and flow conditions. The power-dependent APLHGR limit, MAPLHGRe, is the product of the MAPLHGR power factor, MAPFACp, shown in Figure 3.1-6 and the MAPLHGR for rated power and flow conditions. The MAPLHGR for rated power and flow conditions for each fuel type as a function of axial location and average planar exposure are based on the approved methodology referenced in Section 5.0 and programmed in the plant ,

process computer. The MAPLHGR for rated power and flow conditions l for the limiting lattice in each fuel type (excludir,3 natural uranium) are presented in Figures 3.1-1 through 3.1-4.

1 The core loading pattern for each type of fuel in the reactor vessel is shown for the present cycle in Figure 4.0-1.

3.2 Linear Heat Generation Rage (LHGR)

Reference Technical Specification: 3.11.B During reactor power operation, the LHGR of any rod in any fuel assembly at any axial location shall not exceed the limits presented in Table 3.2-1.

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! COLR l REVISION 11C PAGE 9 OF 25 l

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. .- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 4

Table 3.2-1 LHGR Operatinn Limits LHGR Operating Fuel Tvoe Limit (KW/fD BP8DRB300 13.4 BP8DQB323 14.4 l BP8HXB355 14.4 BP9 HUB 378 14.4 1

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l COLR REVISION 11C PAGE 10 OF 25

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123 Q2.1

'12.0 l

11.5 A

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$ 10.9 13 0 '

$ 10.5 o g 103 *t!

O 10 $

s! N :1 h 95 ' b y 1000 ,

9 g

200 g 8.5 , 7.

B 8 5 0 5,000 10,000 15,000 20,000 25,000 3(.000 35,000 40,000 45,000 PLANAR AVERAGE EXPOSURE (MWD /ST) 2o a

_g

{g FIGURE 3.1-1 Maximum Average Planar Linear Heat Generation Rate

{

"I (MAPLHGR) for Fuel Type BP8DRB300 9 I3

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/'" N,m i 11.6 7 2 11.5 T u.4 5

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4 Q.2 c 10 K 1000 9.5 if 9

8 8.5 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Planar Average Exposure (MWD /ST) do

  • 19 . FIGURE 3.1-2 h k Maximum Average Planar Linear Heat Generation Rate a M (MAPLHGR) for Fuel Type BP8DQB323 g

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,- 12.5 .

123j

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"hia -fi t.4 11.4 U 5 god 10.5 N N l

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5.5 yn 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 50,000 gO Planar Average Exposure (MWD /ST) m

- m k;e FIGURE 3.1-3: Maximum Average Planar Linear Heat Generation @

g Rate (MAPLHGR) for Fuel Type BP8HXB355 S 1

p 13 4 '.

E'. 12.5 8

11.6j 11 4 11.5 g3 11 8

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11.0 10.5 m

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5.5 m 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 50,000 55,000 60,000 W

$n@n Planar Average Exposure (MWD /ST) m k FIGURE 3.1-4: Maximum Average Planar Linear IIcat Generation Q g Rate (MAPLHGR) for Fuel Type BP9 HUB 378 8

I x 1.1 ii- -

e-a C

?" /

/

/ ,

y /// / MAPLHGRp = MAPFACp

  • MAPLHGRSTD i

O MAPLIIGRSTD = Standant MAPLHGR Limits from Figures 3.1-1 through 3.1-4 O b 0.8 r ,

u. MAPFACp = MINIMUM (1.0, Ap + Bp F) p Q [ [ scoop tube setpoint calibrauon postiM F = Fraction of Rated Core Flow d 0.7 f > such that Flowmax=

A and p Bp are fuel-type dependent O g f u gi n W l'*

Flowmax All FuelTypes

_ N 11 0 % (% Rated) Ap Bp 0.6 g 102.5 0.4861 0.6784 g 107.0 0.4574 0.6758 g 112.0 0.4214 0.6807 H I17.0 0.3828 0.6886 I I 0.5 30 50 70 90 110 %

%y @n CORE FLOW (% RATED) y

W rn R. FIGURE 3.1-5 h a

w Flow I)cpendent M API.IIGR Factor (MAPFACp)

1.0  ?

8 5- 0.9 '

/ -

~

^ "

g l P > 45%

U 0.8 '

I

% ,e 0.7 PBypass< P s 45% , e i MAPLIIGRp = MAPFACp

  • MAPLHGRSTD m

~

h #

l MAPLilGRSTD = Standrad MAPLllGR Limits from Figwes 3.1-1 i) i erugh 3.14 -

g

( 0.6  :

P = % Rated Core Hermal Power g ,,,

o W

g yp= % Rated Core hermal Power that Cei,,~4 to the Setpoint o ' pI I j for Bypass of Scram Signals Generated by Closwe of Turbine

]4 0.5 Core Flow s 50% Rated [ 8 l Stop Valves or Fast Closure of Turbine Control Valves -

h I l . I y (Maximum ihypass = 45%) _ g

$ l Core Flow >50% Rated 8 y

.4 y

i For P < 25%: No Hermal Monitoring is Required (No Umits Specified)_ g g 0.4 For 25% s P s 45% : P <Synast and Core Flow s 50% Rated:

2 y . .

_ O g i M APFACp = 0.55 + 0.005 (P - 45%)

[

" 0'3 -

k N l 8

For 25% s P s 45% : P < PBynamn and Core Flow > 50% Rated:

h" l M APFACp = 0.50 + 0.006 (P - 45%) -

to

  • 0 .2  ! For 25% s P s 45% and P >PBYDass

~

h l I MAPFACp = 1.0 + 0.005224 (P - 100%) O x

g

! For P > 45%: M APFACp = 1.0 + 0.005224 (P - 100%)

0.1 i ,

I i i i l 1 o 0.0 - =

5b n r 0 20 40 Max PBypass 6() 80 100 W gW ,  % RATED CORE TIIERMAL POWER (P) Q y

O N O

v. FIGURE 3.1-6 9 Power Dependent MAPLHGR Factor (MAPFACp ) IS

i .

PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 3.3 Minimqm. Critical Power Ratio (MCPR)

Reference Technical Specification: 3. l l.C 4  !

During power operation, the MCPR shall be greater than or equal to the operating limit MCPR. The operating limit MCPR is the greater of the flow- and power-dependent MCPR operating limits, MCPR, and MCPRe.

The flow-dependent MCPR operating limit, MCPRr, is provided in Figure 3.3-1. For core thermal powers less than or equal to Psy, , the power-dependent MCPR operating limit, MCPRp, is provided in Figure 3.3-2.

  • Above Psy, , MCPRe is the product of the rated power and flow MCPR operating limit presented in Tables 3.3-1 and 3.3-2 and the Kp factor presented in Figure 3.3-2. Figure 3.3 2 also specifies the maximum value i for Psyp . The rated power and flow MCPR operating limits presented in Tables 3.3-1 and 3.3-2 are functions of t for the indicated MOC and EOC cycle exposures.

The value of x in Tables 3.3-1 and 3.3 2 shall be equal to 1.0, unless it is calculated from the results of the surveillauce testing of Technical Specification 4.3.C, as follows:

rm - r, 7

1.2 5 2 - r, Where: ,

tm = Average scram time to drop out of Notch 34

[ N,v

,i

=

~

[ N,

,.i ta = Adjusted analysis mean scram time NI

= p + 1.65cr ,,

1[..#'

COLR REVISION 11C PAGE 17 OF 25

. .- PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 3.3 Minimum Critical Power Ratio (MCPR) (Continued) n =

Number of surveillance tests performed to date in the present cycle

= Total number of active control rods Ni i

= Number of active control rods measured in the i*

Ni i surveillance test )

i ti =

Average scram time to drop out of Notch 34 position of all rods measured in the i* surveillance test I

p = Mean of the distribution for average scram )

insertion time to drop out ofNotch 34 J

0.937 sec l c

Standard deviation of the distribution for average scram insertion time to dropout of Notch 34 )

i

= 0.021 sec I l

l l

l l

l l

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COLR REVISION 1IC PAGE 18 OF 25

. .- - - _ - = - - - . - . _. _

PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 Table 3.3-1 MOC MCPR Operating Limits The MCPR operating limits (OLMCPR) for operation from the Beginning of Cycle (BOC) to the End of Cycle (EOC) - 5250 MWD /ST as a function ofI and core flow are:

OLMCPR(t) 102% Rated < Core Flow I Core Flow s102% Rated s107.5% Rated t s 1.0 1.40 1.45 COLR REVISION 11C PAGE 19 OF 25 l

PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 i

Table 3.3-2 EOC MCPR Oper.gting Limits The MCPR operating limits (OLMCPR) as a function of T for operation from the End of Cycle (EOC) - 5250 MWD /ST to the EOC with core flow s107.5% of rated are:

I OLMCPR (t) l 1.39  ;

I s 0.0 0.0 < t s 0.1 1.40 0.1 < t s 0.2 1.40 l 0.2 < t s 0.3 1.41 l 0.3 < t s 0.4 1.41 0.4 < t s 0.5 .1.42 0.5 < t s 0.6 1.43 0.6 < t s 0 7 1.43 0.7 < t s 0.8 1.44 i 0.8 < t s 0.9 1.44 0.9 < t s 1.0 1.45  ;

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COLR REVISION 11C PAGE 20 OF 25 I . ~ . - .

y l l l l l l l  :

For F2 0.40:

h..

( ._

D MCPRy= Maximum (1.20 A p F + Bp) b 1.6

\ \ For F < 0.40-MCPRp = ( ApF + g B )(1 + 032 (0.40- F))

(

Where: F = Fraction of Rated Core Flow g2 I

\ < \ '

Ap and By are Flow Dependent Constants Given Below:

\ m Z \ \ N Flowmax Ay By yg N N N 1023 -0.571 1.655 8;c D1 g I^ '

if 107.0 -0.586 1.697 O

F 112.0 -0.602 1.747 Scoop tube setpoint 117.0 -0.632 1.809 k M

calibration postioned / , \

// / \

33 3 5s 107.0 %

[

\

A \ O llI/ \Nhh x x s ' b 1.2 8

4 1.1 20 30 40 50 66 70 80 90 100 110 y a

Q g '

CORE FLOW (% RATED) Q

M
3 FIGURE 3.3-1 9 Flow Dependent MCPR Limits (MCPRp ) y

3.6 g g y

[ Operating Lirr.it MCPHp -

$. gi, 3.4 -- for Cwe Flow > 50% Rated Operating Limit MCPRp=Kp

U- g 3.2 i / OLMCPR(100) = Operating Limit MCPR Values for 100% Rated Core Dennal Pbwer

@ gi, i /

U i g fmm Table 33-1 and Table 33-2 8 2 3.0 P = % Rated Core Hermal Power h

h 2.8 l ,

i PBypass = % Rated Core Hermal Power that Corresponds to the Setpoint for Bypass of Scram Signals Generated by Closure of hrbine Stop Valves or Fast Closure of i

~, 3 I TurbineControlValves (Maximum g O 2.6 3

K p = Rated MCPR Multiplier as Defined Below. yp,3, = 45% )

Z hating Limit MCPRp l T 2.4 -- for Cwe Flow s 50% Rated 'q For P < 25%: No hermal Monitoring is Required (No Limits are Specified) g

' ' M M [

g 2.2 'w '

For 25% s P s 45% and P < pnvnm :

O gp-_ Kayo Ksio (45% - P) 2.0 i

i N_w i OLMCPR(100)

W fU i

i Where K Byp= 1.95 and K3;o= 0.0125 for Core Flow s 50% Rated or -

o m

1.8 K Byp = 235 and K ;o= 0.043 for Core Flow > 50% Rated M 3

Fw 25% s P s 45% and P >Ihynsss : Kp = 1.28 + 0.0134 (45% - P) d l.6 U 2 I I For 45% < P < 60%: Kp = 1.15 + 0.00867 (60% - P) C N

8 5 F' For P 2 60%: Kp = 1.0 + 0.00375 (100% - P)

I4 i

[ h i g M

\i 13 C' W m

KpfmPBypass< P s 45% l K > for P > 60% l

, . W / '.2 m b

i y N L- g g

, lK forp 45% < P< 60% l _

3,3 y I i l  % ^

t t I  % % M

' E T g 20 30 40 Max $YPass 50 60 70 80 90 100 y

%h  % RATED CORE THERMAL POWER (P) M U* .

Ei h FIGURE 3.3-2  %

Power Dependent MCPR Limits (MCPRp ) S

6 PILGRIM NUCLEAR POWER STATION PNPS CORE OPERATING LIMITS REPORT RTYPE: G4.02 3.4 Power / Flow Relationshio During Power Operation .

i Reference Technical Soccification: 3.11.D l The power / flow relationship shall not exceed the limiting values shown on L the Power / Flow Operating Map in Figure 3.4-1.

l 4.0 REACTOR VESSEL CORE DESIGN Reference Technical Soecification: 5.2 The reactor vessel core for the present cycle consists of 580 fuel assemblies of the  ;

types listed below. The core loading pattern for each type of fuel is shown for the  !

I present cycle in Figure 4.0-1.

Fuel Tvoe Cvele Loaded Number Irradiated BP8DRB300 8 136 BP8DQB323 9 168.

BP8HXB355 10 140 New BP9 HUB 378 11 136 Total 580 The reactor vessel core contains 145 cmciform-shaped control rods. The control materials u >ed are either boron carbide powder (B4C) compacted to approximately l 70% of the theoretical density or a combination of boron carbide powder and solid hafnium. -

5.0 RECERENCES 5.1 NEDE-24011-P-A-10 and NEDE-24011-P-A-10-US, " General Electric Standard Application for Reactor Fuel", February and March,1991.

l 5.2 NEDC-31852-P," Pilgrim Nuclear Power Station SAFER /GESTR-LOCA i Loss-of-Coolant Accident Analysis", September 1990.

5.3 NEDC-31312-P," ARTS Improvement Program Analysis for Pilgrim Nuclear Power Station", September 4,1987.

COLR REVISION 1IC PAGE 23 OF 25

2@  ;  ;  ;  ;  ; i 130 ,

g APRM SCRAM 0.66W + 69%

24m 120 .

Q CLAMP @ 120% 7 '

$5 -

I I I N / '

/ "

$ APRM ROD BLOCK 0.66W + 62%

8 Iws -

CLAMP @ 115%

x

/ [// r 3

1m i8= y y e y 9o c

G1600 / -

/ /

+

g 7

) D 80 g &

7 7

.1400 r

/ Operation Prohibited > 70 m

Outside this Curve g g

\

1200 26% Pump \ 60 m$ o,o lb 1000 Speed Line

'" K I 50 %E

.c:

H d

Natural "E z

g 800 40 O g Circulation Lme

, 5 u

g r

p 600 30g $

I w 400 - 20 $

/ / 3 200 . 7 / 10 y

-~

0 0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 69 75 80

c o Core Flow (MLB/IIR) y 99
$

WW '

m 2, FIGURE 3.4-1 9 O Power / Flow Operating Map I3

1 .

~

PNPS CORE OPERATING LIMITS REPORT RTYPE G4.02 l

MBiBBBBBBBiH
M H H BEM M BIBBBB i+s M H H H H H i+i H i4 M
3 BBBBBBBBBBBIE454E4BIBEM
334E+RBRBEMBBBBBBBBBBBBBE
R BIBER H BIBBBBBBBBBBBBBB
3 H B B M B B M B I N E 4 M B E M M
3 H H H B i H i4 B R M B B B i H M l3HBEMBEMBBBBBEHHHH
3 E+ M BBBEBRE+iBER BE Hif M
H H H H H BEBRM M M M
MMMMMMI4HH
MHHHHHH I 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 Fuel Types ( Cycle Loaded)

@ BP8DRB300 (Cycle 8) @ BP8HXB355 (Cycle 10)

@ BP8DQB323 (Cycle 9) E BP9 HUB 378 (Cycle 11)

FIGURE 4.01 Reactor Vessel Core Loading Pattern Revision 11C Page 25 of 25

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