ML20044G863

Rev 10A to Pilgrim Nuclear Power Station Colr,Cycle 10

ML20044G863
Person / Time
Site:	Pilgrim
Issue date:	05/13/1993
From:	BOSTON EDISON CO.
To:
Shared Package
ML20044G862	List: BECO-93-070, Forwards Rev 10A to Pilgrim Nuclear Power Station COLR, Cycle 10 ML20044G863
References
NUDOCS 9306040366
Download: ML20044G863 (23)

Text

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PILGRIM NUCLEAR POWER STATION CORE OPERATING LIMITS REPORT (CYCLE 10)

Y[,2 7 /73 Approved n

Nuclear / Analysis Division Manager Date Approved

/Ndhtc4 M/2-7/f.3

,r Nuclear Engineering Department Manager Date O li.-C.

Reviewed d [' b M 93** 5//tl93 T

Opera ns Review / Committee Date Approved f\\

fxwbJac. b 65K 5[b[93

[

5tation Director i Date j

)

[

Revision 10A Page.1 of 23 i

9306040366 930526 PDR ADOCK 05000293 PDR P

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., g PNPS CORE OPERATING LIMITS REPORT l

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TABLE OF CONTENTS o

Pace' i

TITLE / SIGNATURE PAGE....................................... 1 i

' TABLE OF CONTENTS......................................... 2 RECORD OF REVISIONS..........................

................ 3 g

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LIST OF TABLES....................

..................... 4 1

LIST OF-FIGURES............................................. 5 a

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1.0 INTRODUCTION

2.0 INSTRUMENTATION TRIP SETTINGS......................... 6 2.1 APRM Flux Scram Trip Setting (Run Mode)......... 6 -

2.2

-APRM Rod Block Trip Setting (Run Mode).......

... 7 1

2.3 Rod Block Monitor Trip Setting.................... 7 j

3.0 CORE OPERATING LIMITS.............................. 8 3

3.1 Average Planar Linear Heat Generation

l Rate (APLHGR)................................. 8 3.2 Linear Heat Generation Rate (LHGR)....

......... 8

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3.3 Minimum Critical Power Ratio (MCPR)............ 16 ;

3.4 Powe r/Fl ow Rel ati ons h i p........................ 21 4.0 REACTOR VESSEL CORE DESIGN............................ 21

5.0 REFERENCES

21

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Revision 10A Page 2 of 23 i

t PNPS CORE OPERATING LIMITS REPORT.

RECORD OF REVISIONS Revision Effective Date Descriotion 8A Effective date based ~

Applicable for'use during on issuance of license Cycle'8 operation.

amendment by NRC.

9A Effective date based Applicable'for use on issuance of license during Cycle 9

-amendment by NRC operation.

l for ARTS and SAFER /GESTR.

i 10A Effective date Applicable for based on initial startup use during of Cycle 10.

Cycle 10 operation.

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PNPS CORE OPERATING LIMITS REPORT i

LIST OF TABLES Number Title Pace 3.2-1 LHGR Operating Limits 15 3.3-1 MCPR Operating Limits 18 t

t t

t t

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Revision 10A i

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PNPS CORE OPERATING LIMITS REPORT-

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LIST OF FIGURES-j!

Number Title Pace 3.1 1 Maximum Average Planar Linenr Heat Generation 9

j Rate (MAPLHGR) for Fuel Type PSDRB282 5

3.1-2 Maximum Average Planar Linear Heat Generation 10 Rate (MAPLHCR) for: Fuel Type BP80RB300 j

3.1-3 Maximum Average Planar Linear Heat Generation ~

11.

Rate (MAPLHGR) for Fuel Type BP80QB323 3.1-4 Maximum Average Planar Linear Heat Generation.

12 Rate-(MAPLHGR) for Fuel Type BP8HXB3c:

t

-5 3.1-5 Flow-Dependent MAPLHGR Factor-(MAPFAC )

13

'l F

3.1-6 Power-Dependent MAPLHGR Factor (MAPFAC )

14 p

3.3-1 Flow-Dependent MCPR Limits (MCPR )

19 ~

F 3.3-2 Power-Dependent MCPR Limits.(MCPRp).

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3.4-1 Power / Flow Operating Map 22 4.0-1 Reactor Vessel. Core loading' Pattern 23

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1.0 INTRODUCTION

This report provides the cycle-specific limits for operation of the Pilgrim Nuclear Power-Station (PNPS) during Cycle 10.

In this report, Cycle 10 will frequently be referred to as the present cycle.

Although thi:; report is not part of the PNPS Technical Specifications, the Technical Specifications refer to this report for the applicable values of the following fuel-rated pararneters:

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-2J Rod Block Monitor Trip Setting Taolo 3.2.C-2 Average Planar Linear Heat Generation Rate 3.11'.A Linear Heat Generation Rate (LHGR) 3.11.B i

Minimum Critical Power Ratio (MCPR) 3.11.C 3.11.0 Power / Flow Relationship Reactor Vessel Core Design 5.2 If any of the core operating. limits in this report are exceeded, actions will be taken as defined in the referenced Technical Specification.

The core operating limits in this report 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.

q 2.0 INSTRUMENTATION TRIP SETTINGS 2.1 APRM Flux Scram Trio Settino (Run Mode)

Reference Technical Specifications:

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 (Ss) shall. be:

I Ss 5. 0. 58 W + 62.

Where Ss - APRM flux scram trip setting in percent of.

rated thermal power (1998 MW )-

t W = Percent of drive flow required to produce a rated core flow of 69 Mlb/hr.

Page 6.of 23 Revision 10A

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PNPS CORE OPERATING LIMITS REPORT-2.1 ApRM Flux Scram Trio Settina (Run Mode). Continued' r

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.

i In accordance with Technical Specification Table 3.1.1, Note 15, for no combination of loop' 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 Red Block Trio Settina (Run Mode)

Reference Technical Soecifications: Table'3.2.C-2, 3.1.'B.1 When the mode switch is in the run position, the average power range monitor (APRM) rod block trip setting (SRB) shall be-i SRB s 0.58 W + 50%

Where, SRB = APRM rod block trip setting in percent of rated thermal power (1998 MW )-

't t

i W = Percent of drive flow required to produce a rated

]

core flow of 69 Mlb/hr.

j 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.

't 2.3 Rod Block Monitor Trio Settina Reference Technical Soecification: Table 3.2.C l Allowable values for the power-dependent Rod Block Monitor trip-setpoints shall be:

Reactor Power, P Trip Setpoint l

(% of Rated)

(% of Reference levell' P 5 25.9 Not Applicable ~ (All RBM Trips Bypassed).

25.9 < P 1 62.0 120 62.0 < P i 82.0 115

.]

I 82.0 < P 110

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PNPS CORE OPERATING LIMITS REPORT 2.3 Rod Block Monitor Trio Settino. Continued The allowable value for the RBM downscale trip setpoint shall be >.

t 94.0% of the reference level. The RBM downscale trip is bypassed.

t for reactor power < 25.9% of rated.

3.0 CORE OPERATING LIMITS 3.1 Averace Planar Linear Heat Generation Rate ( APlHGR) o a

Reference Technical Specification:

3.11.A l

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, MAPLHGRp and MAPLHGRp. The flow-dependent APLHGR limit, MAPLHGR, is the product of the MAPLHGR flow factor, MAPFACF, F

shown in Figure 3.1-5 and the MAPLHGR for rated power and flow '

conditions. The power-dependent APLHGR limit, MAPLHGRp, is the l

product of the MAPLHGR power factor, MAPFACo, 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 for the' limiting lattice in each' fuel type (excluding natural uranium) are presented in Figures 3.1-1 l

through 3.1-4.

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

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3.2 linear Heat Generation Rate (LHGR1 Reference Technical Soecification:

3.11.8 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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0 5,(XX)

.10,(XX) 15,(XX)

20,(XX) 25,(XX)

- 30,(XX) 35,(XX) 40,(XXI PLANAR AVERAGE EXPOSURE (MWD /ST) i m

FIGUllE 3.1-1 7

. Maximum Average Planar Linear llent Generation llate

{

_(M APLIIGR) for Fuel Type P8Dittl282

W N

s.

5' 12.5 g

12.3 12.1 12.0 I1.5'

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0 5,(XX) 10,(XX) 15,(XX) 20,(XX) 25,(XX) 30,(XX) 35,(XX) 40,(X)0 45,(XX)

PLANAR AVERAGli EXPOSURE (MWD /ST)-

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m FIGUllE 3.1-2 a

h Maxiinuin' Average Planar Linear llcat Generation llate

}.

(M APLilGlt) for Fuel Type llP8Ditll300

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_m____-____m_m___.._-____.__m__a.,

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.5,000 10,000 15,(XX) 20,(XX) 25,(XX) 30,(XX) 35,(X)0 40,(XX) 45,(XX)

Planar Average Exposure (MWD /ST) 1 FIGUltE 3.1-3 k-Maximum Average Planar Linear IIcat Generation llate bl-(MAPLIIGit) for Fuel Type HP8DQH323

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45,000 50,(XX) w%

Planar Average Exposure (MWD /ST)

U Maximum Average Planar Linear IIcat Generation p,,

FIGURE 3.1-4:

M llate (M APLIIGlt) for Fuel Type llP811X11355 m

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h1APLllGRp = h1APFACp

• h1APLilGRSTD f

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g MAPLilGRSTD = Standard MAPLIIGR Limits m

o from Figures 3.1-1 through 3.1 O b 0.8 m

R W

MAPFAC, = MINIMUM (1.0, A : + B. F) y f

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MAPFACp or g

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scoop tube scipoint F = Fraction of Rated Core Flow.

O C$

d.

calibration postioned A andUp as fuel-type dependent ucli ti at Flowmax=

p

'F constants given below:

x Fio max - Fucirvocce8x8en.neee8x8a

-s 5

io7.0 %-

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- N tn N

(% Rated)

Ap Up A

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g

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'O.4861 0.6784 m

0.6 f.

107.0 0.4574.0.6758 h-1 I2.0 0.4214 0.6807 H

I I 7.0 03828 0.6886

- 0.5 -

30 50 70 90 110 m

CORE FLOW (% RATliD).

a W

Rc FIGURE 3.1-5 u

. Flow Dependent MAPLIIGR Factor.(MAPFACp)

...=. ---...: =.-....

. -..-. ~ -.-..

-.-.a.-

=e 1.0 9-

-/

0.9

~

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l P > 45%

U 0.8 -

0.7 -

Papass< P '45% h a

7 b1 Al'l.llGRp = h1APFACp

• hlAPIIIGR m 3

s T

h1 API.llGR m = Standard hl API.llGR 1.imits from Figures 3.1 l l

3 thr""sh 31 -4 n

< 0.6 P = % Rated Cure'lhermal Powcr O

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4g thyp3, = % Raica Core' thermal Power that Corresimnds to the Sctroint for Hypass of Suam Signals Generated by Closure of Tuihine I

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Stop Valves or Fast Closure of Turbine Contrul Valves k

I Core Flow s 50% Rated

] 0.5 -

7 111 I

l (Masimum lhypass = 45%)

[

, t l Core Flow >50% Rated 4

i For P< 25% No'1hermal Alonitoring is Required (NoI.imits Specified) g g'4 Z

For 25% $ P s 45% : P <I3llypass and Core Flow s 50% Rated; Q

i gi h1 APFACp = 0.55 + 0 005 (P - 45%)

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1 g

h1 APFACp = 0.50 + 0 006 (P - 45%)

(n g

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h

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I h1 APFACp = 1.0 + 0.005224 (P - 100%)

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FBIP > 45%; M APFACp = 10 4 0 005224 (P - 100%)

A 8- -

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% IR ATi!D COltIi TilliitM AL POWillt (P)

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PNPS CORE OPERATING LIMITS REPORT I

Table 3,2-1 f

LHGR Oceratino limits t

i LHGR'0perating.

Fuel Tvoe limit (KW/ft) l P8DRB282 13.4

?

l BP80RB300 13.4 BP800B323 14.4' i

i

.BPSHXS355 14.4 t

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PNPS. CORE OPERATING LIMITS REPORT 3.3 Minimum Critical Power Ratio (MCpR)

Reference Technical Specification:

3.ll.C During power operation, MCPR shall be greater than or equal to the e

operating limit MCPR. The operating limit MCPR is the greater of.

the flow-and power-dependent MCPR operating limits, MCPRF and i

MCPRp.

The flow-dependent MCPR operating limit, MCPR, is F.

provided in Figure 3.3-1.

For core thermal powers less than or equal to PBypass, the power-dependent MCPR. operating limit, MCPRp, Above P vpass, MCPRp is.the product is provided in Figure 3.3-2.

B of the rated power and flow MCPR operating limit presented in Table 3-.3-1 and the Kp factor presented in Figure 3.3-2.

Figure The rated 3.3-2 also specifies the maximum value for PBy power and flow MCPR operating limits presente~ pass.

d in Table 3.3-1 are functions of cycle-average. exposure ano the average scram i

insertion time, T.

The value of the average scram insertion-time (r) in Table 3.3-1 shall be equal to 1.0, unless it is calculated from the 'results of the surveillance testing of Technical Specification 4.3.C, as follows:

l T ave - 73 r=

1.275 - TB Where:

n{N ri i

I"l ave-Average scram time to the =

r 30% insertion position n{N i i=1 4

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141 1/2 rs - Adjusted analysis mean - g + 1.65 c'

scram time n.

. 1-1 j

n - Number of surveillance tests performed to date in the present cycle-I N1 - Total number of active control rods iPage 16 of 23 Revision _10A

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PNPS CORE OPERATING LIMITS REPORT i

3.3 Minimum Critical Power Ratio (MCPR), Continued

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= Number of active control rods measured in the i Ni surveillance test

= Average scram time to {ge 30% insertion position of all tj rods measured in the i surveillance test.

g - Mean of the distribution for average

= 0.945 sec scram insertion time to the 30%-

position m = Standard deviation of the distribution = 0.064 sec for average scram insertion time to the 30% position j

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PNPS CORE OPERATING LIMITS REPORT TABLE 3.3-1 MCPR OPERATING LIMITS i

Average Scram Insertion Time (r)

MCPR Ooeratino limit i

i For operation from the Beginning of Cycle (B0C) to the En'd o'f Cycle (E0C) - 2000 MWD /ST:

3 All values of r 1.35 l;

For operation from E0C - 2000 MWD /ST to EOC:

All values of r 1.38 l.'

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• OI.MCiht(100) k Operating Limit htCPRp 3A ~~

for Cine Flow > 50% Rated

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OLMCPR(I A) = Operating Lunit MCPR values for 100% Rate.1 Coic' thermal Power g

3.2 i

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2 from Tabic 3 3-l I

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4 P = % Rated Coie Thermal Powcr 30 yp,,,= % Rated Core lhcimal Power that Corresponds to the Seignint for Ilypass of h

l P

g g

Scram Signah Generaicd by Closure of Tmbine Stop Valves or Fast Closuic of -

d 18-a J

l t

Tuihine Control Valves (Maximum l(i

= 45% )

Kp = Rated MCPR Multiplier as Defined inclow:

O 24 3

,t l

Og rating Limit MCPRp

[hP < 25%: Nolhennal Mumtoring is Reatuire.l(No 1.inuts are Specified) 2.4 -

for Core Flow s 50% Rated N

un I

l fnrE% < P s 45% and P < Ellypass :

S g

2.2 -

K - UYPlblM!d3 M

P OI.MCPR(100) o 2.0 I

T Whre Knyp: 195 and K jo= 0.0125 for Core How s 50% Raica or 3

1.8 Kggyp = 2.35 and K3to= 0 083 for Cote Flow > $0% Rated f>

Fot25% s P s 45% arn! P >Divnns; Kp = 1.28 + 0 0134 (45% - P)

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For 45% < P < 60% Kp = 1.15 + 0 00867 (60% - P)

F*

8\\

l 1.5 r-)

r*

m M

i s

i liitP 2 N)%:

K p = 1.0 + 0 00375 (100% - P)

W G

s g3 L.

v)

'g n

C' g

I h

I N'

I3 L P~

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[for P > N)% f-y P or PHypass< P s 45%

K f

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for 45% < P < 60% f-h*

- Ll g

^

T b

N g,y I

t

'" IYPd55 50 60 70 80 90 100 N

20 30 40 ra 5

% It ATIID CORii TilliitM AL POWIiR (P) l IJ O

O FIGUlt E 3.3-2 m

Power Depetulent MCPit Limits (MCPit,)

tJ j

-___m

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. 1 t PNPS CORE OPERATING LIMITS REPORT t

4 3.4 Power / Flow Relationshio Durino Power Ooeration Reference Technical Soecification:

3.11.0 t

The power / flow relationship shall not exceed the limiting values j

shown on the Power / Flow Operating Map in Figure 3.41.

f 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 present cycle in Figure 4.0-1.

Fuel Tvoe Cycle loaded Number Irradiated P80RB282 7

80 BPSDRB300 8

192-BP8008323 9

168 j

New BP8HXB355 10 140-Total 580 r

The-The reactor vessel core contains 145 cruciform-shaped control rods.

control materials used are either boron carbide powder (B4C) compacted to approximately 70% of theoretical density or a combination of boron carbide powder and solid hafnium.

5.0 REFERENCES

5.1 NEDE-24011-P-A-10 and NEDE-24011-P-A-10-US, " General. Electric i

Standard Application for Reactor Fuel," February and March,.1991..

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

. l 5.3 NEDC-31312-P, " ARTS Improvement Program Analysis. for Pilgrim f

Nuclear Power Station", September 4,' 1987.

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Fuel Types ( Cycle Loaded)

P8DRB282 (Cycle 7)

$ BP8DQB323 (Cycle 9)

B BPSDRB300 (Cycle 8)

E BP8HXB355 (Cycle 10)

FIGURE 4.01 Reactor Vessel Core Loading Pattern Revision 10A Page 23 of 23

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