ML20092G932

From kanterella
Jump to navigation Jump to search
Rev 0 to Suppl Reload Licensing Rept for Pilgrim Nuclear Power Station Reload 10 Cycle 11
ML20092G932
Person / Time
Site: Pilgrim
Issue date: 02/28/1995
From: Hetzel W, Klapproth J
GENERAL ELECTRIC CO.
To:
Shared Package
ML19311B798 List:
References
24A5172, 24A5172-R, 24A5172-R00, NUDOCS 9509200151
Download: ML20092G932 (45)
v • d • e


Text

.

GE Nuclear Energy 24A5172 Revision 0 Class I February,1995 24A5172, Rev. O Supplemental Reload Licensing Report for Pilgrim Nuclear Power Station Reload 10 Cycle 11 Approved 3 *Q',

Approved J.F. Klap ger W.H. Hetze Fuel and Facility Licensing Fuel Project Manager 9509200151 950912 PDR ADOCK 05000293 i

PILGRIM 24A5172 Rev.O Reload 10 Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by General Electric Company (GE) solely for Boston Edison Company (BECo). The information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of GE respecting information in this document are contained in the contract between BECo and GE for fuel bundle fabrication and related services for j

Pilgrim Nuclear Power. Station, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; -

and with respect to any such unauthorized use, neither GE nor any of the contributors.

to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information cotdained in this document or that such use of such information may not infringe privately owned rights; nor do they j

assume any responsibility for liability or damage of any kind which may result from such i

use of such information.

j I

e i

l l

i Page 2

PIlfsRIM 24A5172 Reload 10 Rev.0 i

Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by D.C. Serell, A. Alzaben, and T.A. Terrio, with assistance from F.T. Bol-ger and G.A. Galloway. The Supplemental Reload Licensing Report was prepared by D.C. Serell. This docu-ment has been verified by C.W. Smith.

Page 3

PILGRIM 24A5172 Reload 10 Rev.O The basis for this teport is General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-10, February 1991; and the U.S. Supplement, NEDFe240ll-P-A-10-US, March 1991.

1.

Plant-unique Items Appendix A: Analysis Conditions and Bases Appendix B: Increased Core Flow Appendix C: Decrease in Core Coolant Temperature Events 2.

Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

BP8DRB300(BP8x8R) 8 136 GE8B-P8DQB323-10G2' 80M-4WR-145-T(GE8x8EB) 9 168 c

GE10-P8H XB355-11 GZr100M-145-T (GE8x8NB-3) 10 140 Hem GE11-P9 H UB 378-15G2c10(TT-141-T (GE l l) 11 136 Total 580 3.

Reference Core Loading Pattern Nominal pn:vious cycle core average exposure at end of cycle:

24245 mwd /MT

( 21994 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 23804 mwd /MT from cold shutdown considerations:

( 21594 mwd /ST)

Assumed reload cycle core average exposure at beginning of 17222 mwd /MT cycle:

( 15623 mwd /ST)

Assumed reload cycle core average exposure at end of cycle:

28520 mwd /MT

( 25873 mwd /ST)

Reference core loading pattem:

Figure 1 4,

Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C Beginning of Cycle, keg,cuv.

Uncontrolled 1.103 Fully controlled 0.960 Strongest control rod out 0.986 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.003 Page 4

PILGRIM 24A5172 Reload 10 Rev.0

5.

(ppm)

(20 C, Xenon Free) 675 0.042 6.

Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parametersl Exposure: BOC11 to EOCll-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR

( M W ()

(1000ib/hr)

Gell 1.45 1.85 1.31 1.035 6.209 91.9 1.30 Exposure: EOC11-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5% CORE FLOW Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)

(1000lb/hr)

Gell 1.45 1.75 1.18 1.035 5.859 101.2 1.37 7.

Selected Margin Improvement Options Recirculation pump trip:

No Rod withdrawallimiter:

No Thennal power monitor:

No improved scram time:

Yes (ODYN Option B)

Measured scram time:

No Exposure dependent limits:

Yes j

Exposure points analyzed:

2

1. 'Ihe delta CPR response for Gell bounds all other 8uel types in the core.

Page 5

~ _ -., -

= _.

PILGRIM 24A5172 Reload 10 Rev.0 8.

Operating Flexibility Options Single-loop operation:

Yes2 Imadline Smit:

Yes Extended load line limit:

Yes 3

Maximum extended loadlinelimit:

Yes Increased core flow thmughout cycle:

Yes4 Increased core flow at EOC:

Yes t

BOC to 5000 MWD /STU Flow point analyzed:

102 %

5000 MWD /STU to EOC Flow point analyzed:

107.5 %

Feedwater temperature reduction throughout cycle:

No Final feedwater temperatuit reduction:

No ARTS Program:

Yess Moisture separator reheater OOS:

No

'nzrbine bypass system OOS:

No Safety / relief valves OOS:

No ADS OOS:

No 6

One Main steam isolation valve OOS:

Yes

2. " Pilgrim Nuclear Power Station Single-toop Operation", NEDo-24268)une,1980
3. fl.XJIoang," Maximum Extended toad tine timit Analyses for Pilgrim Nuclear Power Station Reload 9 Cycle 10",

NEDC-32306P,MARCil 1994.

4. " Safety Review of Pilgrim Nuclear Power Station Unit No. I at Core Flow Conditions Above Rated Flow Throughout Cycle 6",

NEDO-30242, August,1983.

5. " ARTS Improvement Pmgram Analysis for Pdgrim Nuclear Power Station", NEDO-31312P, September,1987.
6. MSIV Out of Service Report, NSE-82-0982, DRF B21-00238, September 1982.

Page 6

-.. _. -~

PILGRIM 24A5172 Reload 10 Rev.0 78 9.

Core-wide AOO Analysis Results Methods used: GEMIN1; GEXL-PLUS Exposure range: BOC11 to EOC11-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102%

CORE FLOW Uncorrected ACPR Event Flux Q/A Gell Fig.

(%NBR)

(%NBR)

FW ControllerFailure 225 115 0.23 2

Load Reject w/o Bypass 230 107 0.18 3

Exposure range: EOCll-5787 mwd /MT (5250 mwd /ST) to EOCll I NALYZED AT 107.5%

CORE FLOW Uncorrected ACPR Event Flux Q/A Gell Fig.

(%NBR)

(%NBR)

FW Controller Failure 304 123 0.30 4

Load Reject w/o Bypass 332 116 0.27 5

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary' Rod withdrawal error (RWE) is analyzed in General Electric BWR Licensing Report, Average Power Range Monitor, Rod Block monitor and Technical Specification improvement (ARTS) Program, NEDC-30474-P, datedDecernber1983. Acycle-specificrodwithdrawalanalysisfoundthe AMCPRisboundedbythegeneric RWE analysis irported in the referenced report. For a setpoint of up to 116%, the rated MCPR limit is 1.35.
7. The Gell delta CPR response bounds all other fuel types in the core.
8. Analysis at 107.5% increased core flow was conservatively assumed for EOCll analysis; 102% increased core flow was assumed as the BOC to EOCll-5250 mwd /ST carly cycle analysis basis.

1 l

9.

References:

" ARTS Improvement Program Analysis for Pilgrim Nuclear Power Station", NEDC-31312-P. Septernher.1987 and H.X.

Hoang," ARTS Verification for Pdgrim Nuclear Power Station Reload 8 Cycle 9 GE-NE-187-11-0591, DRP A00-03980, June 1991. These documents were verified applicable to cyde 11.

Page 7 i

~ - -

PILGRIM 24A5172 Reload 10 Rev.0 to 1112

11. Cycle MCPR Values Safety limit:

1.07 Single loop operation safety limit:

LO8 Non-oressurization events:

Exposure Range: BOC11 to EOCll Rod Withdrawal Error (Setpoint can be selected up to 116%), All Fuels 1.35 Fuel Loading Error, Gell Reload 10 Fuel 1.12 Fuel Loading Error, GE10 Reload 9 Fuel 1.27 Pressurization events:

Exposure range: BOCll to EOCll-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102%

CORE FLOW Exposure point: EOC11-5787 mwd /MT (5250 mwd /ST)

Option A Option B Gell Gell FW Controller Failure 1.40 1.32 Load Reject w/o Bypass 1.35 1.27 Exposure range: EOCll-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5%

CORE FLOW Exposure point: EOCll Option A Option B Gell Gell FW Controller Failure 1.45 1.39 Load Reject w/o Bypass 1.43 1.37

12. Overpressurization Analysis Summary Psi Py Plant Event (psig)

(psig)

Response

MSIV Closure (Flux Scram) 1286 1302 Figure 8

10. The minimum MCPR operating limit required by the SAFER /GESTR analysis is 1.20.

I1. See Appendix C for discussion of decnase in core coolant temperature e.ents.

12. For single-loop operation. the MCPR operating limit is not greater than the two-loop value.

Page 8

.PilliRIM 24A5172 Reload 10 Rev.0

13. Loading Error Results From a misoriented bundle analysis with variable water gap, including a 0.02 penalty due to variable water gap R-factor uncertainty, the AMCPR for the fresh reload 10 Gell fuel bundle is 0.05. The AMCPR for the reload 9 GE10 fuel bundle is 0.20.

1 1

14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-240ll-P-A-US.

4 1

15. Stability Analysis Results Pilgrim Nuclear Power Station is exempt fmm the cunent requirement to submit a cycle-specific stability analysis as documented in the letter, C. O. Thomas (NRC) to H. C. Pfefferien (GE), AcceptanceforReferenc-ing ofLicensing Topical Report NEDE-240ll Rev. 6, Amendment 8, " Thermal Hydraulic Stability Amend-ment to GESTAR 11," April 24,1985.

Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscillations in Bolling Water Reactors (BWRs), and will comply with the recommendations contained there-in. Pilgrim Station is also complying with the NRC Bulletin No. 94-02, Long-Term Solutions and Upgrade 1

ofInterim Operating Recommendationsfor Thermal-Hydraulic Instabilities in Bolling Water Reactors.

Gell fuel has been demonstrated to have equivalent or better stability characteristics than BP8x8R fuel by the GESTAR Amendment 22 licensing analysis (

Reference:

NEDE 31917P, Gell Compliance with Amendment 22 ofNEDE-240ll-P-A (GESTRAR il), April 1991), and no unique or special actions are nceded

- to comply with the above NRC Bulletins.

16. Loss-of-Coolant Accident Results LOCA method used: SAFER /GESTR-LOCA The LOCA analysis results are presented in Sections 5 and 6 of Pilgrim Nuclear Power Station SAFER /

i GESTR-LOCA Loss-of-Coolant Accident Analysis, NEDC-31852P, April,1992 (Revision 1) as amended.

The Gell LOCA analysis for Pilgrim was performed using the same SAFER /GESTR analysis basis used for the previously analyzed BP/P8x8R and GE8x8EB/NB fuel types. Addition of the Gell fuel will not signifi-cantly affect the overall system response of the plant for the various operating modes, and the Gell analysis confirmed that the limiting break type and size and limiting ECCS failure (DBA recirculation suction line break with LPCIIV failure) do rot change. The gel l fuel analysis yielded a licensing basis peak PCT of 1815 F and a peak local oxidation fraction of <0.3%, and all licensing basis criteria are met. The Gell results are bounded by the 1825 'F licensing basis PCT for BP/P8x8R fuel and the overalllicensing basis results reported in Table 6-1 of the Reference analysis.

The Gell SAFER /GESTR results are applicable for a peak enriched lattice MAPLHGR of 12.16 kw/

ft..which bounds the MAPLHORs for the reload 10 fuel. Therefore, the MAPLHGR limits reflect the ther-mal-mechanicallimits for the reload fuel rather than LOCA/ECCS considerations. The most limiting and the least limiting MAPLHGRs for the new fuel are as follows:

1 Page 9

i PILGRIM 24A5172 Reload 10 Rev.0

16. Loss-of-Coolant Accident Results (cont)

Bundle Type: Gell-P9 HUB 378-15GZ-10(TT-141-T Average Planar Exposure MAPLHGR(kW/ft)

(GWd/ST)

(GWd/MT)

Most Limiting Least Limiting 0.00 0.00 9.95 10.46 0.20 0.22 10.04 10.53 1.00 1.10 10.19 10.61 2.00 2.20 10.41 10.78 3.00 3.31 10.64 10.98 4.00 4.41 10.88 11.21 5.00 5.51 11.09 11.35 6.00 6.61 11.19 11.55 7.00 7.72 11.30 11.63 8.00 8.82 11.40 11.70 9.00 9.92 11.52 11.79 10.00 11.02 11.65 11.91 12.50 13.78 11.64 11.92 15.00 16.53 11.48 11.71 17.50 19.29 11.25 11.47 20.00 22.05 11.02 11.22 25.00 27.56 10.55 10.75 30.00 33.07 10.07 10.22 35.00 38.58 9.39 9.53 40.00 44.09 8.72 8.87 45.00 49.60 8.06 8.23 50.00 55.12 7.40 7.60 55.00 60.63 6.72 6.96 57.02 62.86 6.44 6.69 57.10 62.94 6.68 57.92 63.84 6.56 58.02 63.96 6.55 Page 10

PILGRIM 24A5172 Reload 10 Rev.0 M M Es+sM M M MMMMMMMMM HMMMMMMMMMM

MMMMMMMMMMMMM
M M M M M M M M M M M M M
H M M M M M M H H H H H H ll: H M M M M M M M M M M M M l::MMMMMMMMMMMMM
H M M M M M M M M M H H H
H M M M M M M M M M H H H MMMMMMMMMHH MMMMMMMMM HMMMMMM IIIIIIIII IIIII 1 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 Type A=GE8B-P8DQB323-10GZ-80M4WR-145-T (Cycle 9)

C=GE11-P9 HUB 378-15GZ-100T-141-T (Cycle 11) l B =G E 10'-P8HXB355-I I GZ-100M-145-T (Cycle 10)

D=BP8DRB300 (BP8x8R)

(Cycle 8)

Figure 1 Reference Core Loading Pattern Page 11 i

PILGRIM 24A5172 Reload 10 Rev.O Neutron Flux Vessel Press Rise (psi)

- - - - Safety Valve Flow

- - - - Ave Surface Heat Flux 150.0 - --- Core Inlet Flow 125.0 - --- Relet Valve Flow

- -- core inlet Subcooling

--- Bypass Valve Flow A

~

p.~~

g 100.0 _ -_ k : : ~. C

- - -l ' * ' ' ' '

p 75.0 E

W C

E Y

e~

l i

50.0 25.0

-l l

t...L I

0.0

-25.0 O.0 20.0 40.0 0.0 20.0 40.0 Time (sec)

Time (sec)

Level (incNREF -SEP-SKRT)

Void Reactivity f

- - - - - Vessel Steam Flow

- - - - - Doppler Reactivity 150.0 - --- Turbine Steam Flow 1.0 - --- Scram Reactivity

--- Feedwater Flow

--- Total Reactivity

~

'/

y s

1 0.0

- - ~ w -.=: -, g,-,,,,,

g 100.0

- ~ ~ ~

E j'

O I

~

( P',,

,g.

I h>',

.2 50.0 l

N-1.0 c

e e

C

,e r

N

(,'

{

I' I

0.0

-2.0 O.0 20.0 40.0 0.0 20.0 40.0 mme (sec)

Vme (sec)

Figure 2 Plant Response to FW Controller Failure (BOC11 to EOC11-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW) l Page 12

PILGRIM 24A5172 1

Reload 10 Rev.0 Nsutron Flux Vessel Press Fhse (psi)

Are Surface Heat Flux

- - - - - Safety Valve Flow 150.0

---.Cxe Inlet Flow 300.0 - - - Relief Valve Flow

--- Bypass Valve Flow

, ~

.-~.----

100.0 200.0 m

~

m m

g g

50.0 100.0

/

I 0.0 0.0 O.0 2.0 4.0 0.0 2.0 4.0 mme (sec) mme (see)

Level (inch-REF-SEP-SKRT)

VoMactivity Doppler Reactivity

- - - - - Vessel Steam Flow 200.0

-- Turbine Steam Flow 1.0 Scram Reactivity

-- Feedwater Flow Total Reactivity G

ug N -- ;=: '. /. -

t g/

. \\.V__,__..-

o 00 g tw.0,-, \\.

\\.

E

\\ ',

3

\\

\\.

\\

.f

] -1.0 g

0.0

-p,'-'-----

x

\\

k

~

\\

\\

\\

II

- 100.0

-2.0 O.0 2.0 4.0 0.0 20 4.0 Time (sec)

Time (sec)

Figure 3 Plant Response to Load Reject w/o Bypass (BOC11 to EOC11-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW)

Page 13

l PILGRIM 24A5172 R' load 10 Rev.0

'\\

Vessel Press Rise (psi)

Neutron Flux

- - - - Ave Surface Heat Flux f

- - - - Safety Valve Flow 150.0 - --- Core Inlet Flow 125.0 - --- Relief Vahe Flow

- -- Core inlet Subcooling

--- Bypass Valve Flow

\\

q, --,

"~

100.0 30 75 3

~

m me a

Y Y

~

I 25.0

~N 50.0 t

l 1

l

> [...t I

I 0.0

~ 25.0 O.0 20.0 40.0 0.0 20.0 40.0 mme (sec) mme (sec)

Level (ir&REF-SEP-SKRT)

Void Reactivity

- - - - - Vessel Steam Flow

- - - - - Dopper Reacevity 150.0 - --- Turbine Steam Flow 1.0 - --- Scram Reactivity I

--- Feedwater Flow

--- Total Reacevity l

V' y

/

~-)

0.0

- - w.+, -,g, ;,,7; ;,,,,e g 100.0 lii l', '

C O

l.e..

I O

so g

l o'

k *["' '.

g

.2 N -1.0 I

50.0 a) 1 g

c.

i.

I k'

d 0.0

-2.0 0.0 20.0 40.0 0.0 20.0 40.0 Time (sec)

Time (sec)

Figure 4 Plant Response to FW Controller Failure (EOC11-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5% CORE FLOW) i Page 14 i

~

PILGRIM 24A5172 Reload 10 Rev.0

>eutron Flux Vessel Press Rise (psi)

-Am Surface Heat Flux

- - - - - Safety Valve Flow

-Core inlet Flow 300.0 - --- Relief Valve Flow 150.0 - --

--- Bypass Valve Flow

/ 5;, C u ~~ ~ K s

N g200.0 g 100.0 m

s w

\\

C C

p 50.0 100.0

/

  1. I I

0.0 0.0 O.0 2.0 4.0 0.0 2.0 4.0 mme (sec) mme (sec)

Level (inch-REF-SEP--SKRT) oid

'vity

- - - - - Vessel Steam Flow pp Reactivity 200.0 - --- Turbine Steam Flow 1.0 Reactivity

--- Feedwater Flow Total Reacevity

?

{\\

E

  • '~

0 E.0

\\.

e 100.0 m.

s 1.x.

\\ .

'./.

o 1

O l

\\

g

\\

\\-

T N N, 0.0 - - p

,L-- - -.-- - - e u

~

g-1.0

\\

cc

\\

~

\\

\\

\\

\\

I E

- 100.0

-2.0 0.0 2.0 4.0 0.0 2.0 4.0 mme (sec)

Eme (sec) i 1

1 Figure 5 Plant Response to Load Reject w/o Bypass (EOC11-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5% CORE FLOW) 1 i

Page 15

g esu

$ Cm RP eII 1

1 2

1 1

laE o

0 0

0 5

0 5

0 0

0 0

0 0

0 0

0 0

0 0

0 0

dR 0

0 I

0 0

0M 1

' A'\\. N A'

FTVL

'9

\\

t F

CAN euee

\\

s uew i

N er s v u

a el ni m

i Srt g

dibs e e

w n e (l r o nu l

n r

t Sc T

ef e

T eSt aR i

r t eh i

N Fe u t

e a -

m c

m

(

0 Ra mR e4 p

\\

6 om E

g l

x e 4 o

wH w

FF

(

0 e

P s

Fl -

S a

wS e

N ol o l

e t

a c

~

w E

c

\\

uR n

)

)

P s

t x

R S

K e

p R

N s

T

)

on se t

o 8

8 M

0 0

S x@E. k 3 gum G

$ Cm I

a 1

2 3

V 0

0 0

C 2

1 0

1 0

0 0

0 0

0 0

0 0

0 o

0 0

l

~

s 0

0 ur

(

e F

l u

.s TSDV I,

BRSV x

3 r

i o cooi aetes y eae S

t

\\s-S RmpeR a ap d p

l f s c

l s fy e r

l l

err Ti VaVP e

s V a

T a

l ar m

i

\\. 3 a

c' e R c m

l v

l e

a m

i aeti e4 eRF v

v e s

e

)

e 4 vtcav s

t i

7 0

i i ct y i vi s

ow we s

tvtiy 0

(

Fol R

(

y l

oi s

t e

y e

w c

c g

p(

P

)

)

si a

)

1 2

e 4

RA 6

e5

.v 1 7

O2 8

8 0

0

',l!

l l

Il I

I

PILGRIM 24A5172 Reload 10 Rev.0 Appendix A Analysis Conditions and Bases To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle. The use of the increased core flow for the analysis produces bounding results for the flow range down to 75% of rated core flow. Justification of operation at 100% powerdown to 75% poweris provided in Reference A-1. The cycle 11 licensing analysis has verified the applicability of the MELLL flow range.

Table A-1 Parameter 107.5% Flow Analysis Value Thermal power, MWt 1998.0 Core flow, Mlb/hr 74.2 Reactorpressure, psia 1066.5 Inlet enthalpy, BTU /lb 528.4 Non-fuel power fraction 0.038 Steam flow analysis, Mlb/hr 7.98 Dome pressure,psig 1035.8 Turbine pressure, psig 975.7 No. of Safety / Relief Valves 4

No. of Single Spring Safety Valves 2

Relief mode lowest setpoint, psig 1126.0 Safety mode lowest setpoint, psig 1253.0 For the overpressurization analysis, the MSIV closure (flux scram) case was analyzed at 102% licensed pow-er and steamflow. Also, the maximum possible initial steam dome pressure of 1085 psig was used, which corresponds to the high pmssure scram analyticallimit. The mostlimiting end of cycle core conditions were utilized at 107.5% core flow, which produces a bounding result.

For the first introduction of Gell fuel in Pilgrim, a plant specific evaluation was made of the Gell fuel

" Scram Speed Adjustment Factors"(SSAF) that adjust the option B MCPR limit to optain the Option A MCPR limit. This evaluation concluded that use of a 0.06 EOC scram speed adjustment factor is justified for the load rejection, turbine trip, and feedwater controller failure pressurization events. For Pilgrim gel 1 fuel application, this supercedes the " generic" EOC value from the letter, J.F. Klapproth to USNRC, " GEM-INI/ODYN Statistical Adders for gel 1 fuel for BWR/2 and 3", September 23,1992. The 0.08 generic mid-cycle adders are still applicable to Gell fuel in Pilgrim.

A-1. H.X. Hoang, " Maximum Extended LoadLine Limit Analysesfor Pilgrim Nuclear Power Station Re-load 9 Cycle 10", NEDC-32306P, March,1994.

Page 17

.~ -

PILGRIM 24A5172' Reload 10 Rev.O Appendix B Increased Core Flow The analyses perfonned for Cycle 11 included increased core flow throughout the cycle and after the all-mds-out condition is reached. There are no concerns regarding reactor intemals pressure drop or flow-in-duced vibration as discussed in the increased core flow analysis document for the EOC-6 (NEDO-30242)..

l Page 18

PILGRIM 24A5172 Reload 10 Rev.0 Appendix C Decrease in Core Coolant Temperature Events The loss-of-feedwater heating (LFWH) and the high pressure coolant injection (HPCI) inadvenent stan-up anticipated operational occurmnces (AOO) are the only cold water injection events checked on a cycle-by-cycle basis. For both the LFWH and HPCI events, the delta CPR is not limiting when compared to the delta CPR of the liming pressurization AOO. This is based on the results of calculations performed with consider-ation of the cycle-to-cycle differences such as ARTS. Therefore,the LFWH and HPCIinadvenent stan-up AOOs are not reponed for Cycle 11.

Page 19 I

GE Nuclear Energy 24A5172 Revision 0 Class I February,1995 i

24A5172, Rev. O Supplemental Reload Licensing Report for Pilgrim Nuclear Power Station Reload 10 Cycle 11 Approved 3 *y""*,

Approved j

J.F. Klapproth, Manager W.H. Hetze Fuel and Facility Licensing Fuel Project Manager

PILGRIM 24A5172 Reload 10 Rev.0 Important Notice Regarding Contents of This Report

{

Please Read Carefully This report was prepared by General Electric Company (GE) solely for Boston Edison Company (BECo). The information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of GE respecting information in this document are contained in the contract between BECo and GE for fuel bundle fabrication and relat d services for Pilgrim Nuclear Power Station, and nothbg contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither GE nor any of the centributors to this document makes any representation or warranty (expressed or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

1 Page 2

PILGRIM 24A5172 Reload 10 Rev.0 Acknowledgement The engineering and reload licensing analyses, which fonn the technical basis of this Supplemental Reload Licensing Report, were performed by D.C. Screll, A. Alzaben, and T.A. 'Ibrrio, with assistance from F.T. Bol-1 ger and G.A. Galloway. The Supplemental Reload Licensing Report was prepared by D.C. Screll. This docu-ment has been verified by C.W. Smith.

l

  • D Page 3

PIIERIM 24A5172 Reload 10 Rev.0 I

~

The basis for this report is General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-10, February 1991; and the U.S. Supplement, NEDF-24011-P-A-10-US, March 1991.

1.

Plant-unique Items Appendix A: Analysis Conditions and Bases Appendix B: Increased Core Flow Appendix C: Decrease in Core Coolant Temperature Events 2.

Reload Fuel Bundles Cycle Fuel Type Loaded Number Irradiated:

BP8DRB300(BP8x8R) 8 136 GE8B-P8DQB323-10GZ-80M-4WR-145-T (GE8x8EB) 9 168 GE10-P811XB355-11GZ-100M-145-T (GE8x8NB-3) 10 140 NcE l

G E 1 1 -P911 U B 378-15 GZ-10(TT-141-T (G E l l )

11 136 t

Total 580 3.

Reference Core Loading Pattern Nominal pn:vious cycle core average exposure at end of cycle:

24245 mwd /MT

( 21994 mwd /ST)

Minimum previous cycle core average exposure at end of cycle 23804 mwd /MT fmm cold shutdown considerations:

( 21594 mwd /ST)

Assumed reload cycle core average exposure at beginning of 17222 mwd /MT cycle:

( 15623 mwd /ST)

Assumed reload cycle com average exposure at end of cycle:

28520 mwd /MT

( 25873 mwd /ST)

Reference core loading pattem:

Figure 1 4.

Calculated Core Effective Multiplication and Control System Worth -No Voids,20 C Beginning of Cycle, keffecuve Uncontrolled 1.103 Fully controlled 0.960 Strongest control md out 0.986 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.003 Page 4

PILGRIM 24A5172 Reload 10 Rev.0

5. ' Standby Liquid Control System Shutdown Capability Boron Shutdown Margin (Ak)

(ppm)

(20 C, Xenon Free) 675 0.042 6.

Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parametersl Exposure: BOCll to EOCll-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)

(1000lb/hr)

Gell 1.45 1.85 1.31 1.035 6.209 91.9 1.30 Exposure: EOC11-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5% CORE FLOW Peaking Factors l

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt)

(1000lb/hr)

Gell 1.45 1.75 1.18 1.035 5.859 101.2 1.37 7.

Selected Margin Improvement Options Recirculation pump trip:

No Rod withdrawallimiter-No Thermal power monitor:

No Impmved scram time:

Yes (ODYN Option B)

Measured scram time:

No Exposure dependent limits:

Yes Exposure points analyzed:

2

1. The delta CPR response for Gell bounds an other fuel types in the core.

Page 5

PILGRIM 24A5172 Reload 10 Rev.0 8.

Operating Flexibility Options Single-loop operation:

Yes2 Load line limit:

Yes l

Extended loadline limit:

Yes Maximum extended load line limit:

Yes3 4

Increased core flow throughout cycle:

Yes Increased core flow at EOC:

Yes BOC to 5000 MWD /STU Flow point analyzed:

102 %

5000 MWD /STU to EOC Flow point analyzed:

107.5 %

Feedwater temperature reduction throughout cycle:

No Final feedwater temperature reduction:

No s

ARTS Program:

Yes Moisture separator reheater OOS:

No Turbine bypass system OOS:

No Safety / relief valves OOS:

No ADS OOS:

No 6

One Main steam isolation valve OOS:

Yes

2. "Pilgnm Nuclear Power Station Single-loop Operation", NEDO-24268)une,1980
3. II.X.lloang," Maximum Extended load line limit Analyses for Pilgrim Nuclear Power Station Reload 9 Cycle 10",

NEDC-32306P,MARCil 1994.

4. " Safety Review of Pdgrim Nuclear Power Station Unit No. I at Core Row Conditions Above Rated Flow Throughout Cycle 6",

NEDO-30242, August,1983.

$. " ARTS Improvement Program Analysis for Pdgrim Nuclear Power Station", NEDO-31312P, September,1987.

6. MSIV Out of Service Report, NSE-82-0982, DRF B21-00238, September 1982.

Page 6

PILGRIM 24A5172 Reload 10 Rev.0 78 9.

Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Exposure range: BOC11 to EOC11-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102%

CORE FLOW Uncorrected ACPR Event Flux Q/A Gell Fig.

(%NBR)

(%NBR)

FW Contmiler Failure 225 115 0.23 2

Load Reject w/o Bypass 230 107 0.18 3

Exposure range: EOC11-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5%

CORE FLOW Uncorrected ACPR Event Flux Q/A Gell Fig.

(%NBR)

(%NBR)

FW Controller Failure 304 123 0.30 4

Load Reject w/o Bypass 332 116 0.27 5

10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary' Rod withdrawal error (RWE) is analyzed in General Electric BWR Licensing Repon, Average PowerRange Monitor, Rod Block monitor and Technical Specification improvement (ARTS) Program, NEDC-30474-P, dated Decernber 1983. A cycle-specific rod withdrawal analysis found the AMCPR is bounded by the generic RWE analysis reponed in the referenced repon. For a setpoint of up to 116%, the rated MCPR limit is 1.35.
7. lhe Gell delta CPR response tuunds all other fuel types in the core.
8. Analysis at 107.5% increased core flow was conservatively assumed for EOCll analysis; 102% increased core flow was assumed as the BOC to EOC11-5250 Mwd /sT carly cycle analysis basis.
9.

References:

" ARTS improvement Program Analysis for Pilgnm Nuclear Power Station", NEDC-31312-P, septemter,1987 and II.X.

Iloang," ARTS Verification. for Pdgrim Nuclear Power Station Reload 8 Cyde 9, GE-NE-187-11-0691, DRF A0N)3980, June 1991. These documents were verified applicable to cycle 11.

Page 7

PILGRIM 24A5172 Reload 10 Rev.0 to 1112

11. Cycle MCPR Values Safety limit:

1.07 Singleloop operation safety limit:

1.08 lj Non-oressurization events:

Exposure Range: BOC11 to EOCll Rod Withdrawal Error (Setpoint can be selected up to 116%), All Fuels _

1.35 Fuel Loading Error, GEli Reload 10 Fuel 1.12 Fuel Loading Error, GE10 Reload 9 Fuel 1.27 Pressurization events:

Exposure range: BOCll to EOCll-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102%

CORE FLOW l

l Exposure point: EOC11-5787 mwd /MT (5250 mwd /ST)

Option A Option B Gell Gell FW Controll::r Failure 1.40 1.32 Load Reject w/o Bypass 1.35 L27 l

Exposure range: EOCll-5787 mwd /MT (5250 mwd /ST) to EOCll ANALYZED AT 107.5%

CORE FLOW Exposure point: EOC11 Option A Option B Gell GEli FW Controller Failure 1.45 1.39 Load Reject w/o Bypass 1.43 1.37

12. Overpressurization Analysis Summary Psi Pv Plant Event (psig)

(psig)

Response

MSIV Closun:(Flux Scram) 1286 1302 Figure 8 i

10.1he minimum MCPR operating limit required by the SAFER)GESTR analysis a 1.20.

I1. See Appendix C for discussion of decrease in core coolant temperature events.

12. For single-loop operation. the MCPR operating hmit is not greater than the two-loop value.

Page 8

PILGRIM 24A5172 Reload 10 Rev.0 l

l

13. Loading Error Results From a misodented bundle analysis with variable water gap, including a 0.02 penalty due to variable water j

gap R-factor uncertainty, the AMCPR for the fresh reload 10 GElI fuel bundle is 0.05. The AMCPR for I

the reload 9 GE10 fuel bundle is 0.20.

14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence piant, therefore, the control md drop accident analysis is not required. NRC approvalis documented in NEDE-24011-P-A-US.
15. Stability Analysis Results Pilgrim Nuclear Power Station is exempt from the current requirement to submit a cycle-specific stability analysis as documented in the letter, C. O. Thomas (NRC) to 11. C. Pfcfferlen (GE), Acceptancefor Referene-ing ofLicensing Topical Report NEDE-24011 Rev. 6, Amendment 8, " Thermal Hydraulic Stability Amend-ment to GESTAR H," April 24,1985.

Pilgrim Nuclear Power Station recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscillations in Boiling Water Reactors (BWRs), and will com ply with the recommendations contained there-in. Pilgrim Station is also complying with the NRC Bulletin No. 94-02, Long-Term Solutions and Upgrade ofinterim Operating Recommendationsfor Thermal-Hydraulic Instabilities in Boiling Water Reactors.

gel 1 fuel has been demonstrated to have equivalent or better stability characteristics than BP8x8R fuel by the GESTAR Amendment 22 licensing analysis (

Reference:

NEDE 31917P, Gell Compliance with

}

Amendment 220fNEDE-240ll-P-A (GESTRAR il), April 1991), and no unique or special actions are nceded to comply with the above NRC Bulletins.

16. Loss-of-Coolant Accident Results LOCA rnethod used: SAFER /GESTR-LOCA The LOCA analysis results are presented in Sections 5 and 6 of Pilgrim Nuclear Power Station SAFER /

GESTR-LOCA Loss-of-Coolant Accident Analysis, NEDC-31852P, April,1992 (Revision 1) as amended.

The gel 1 LOCA analysis for Pilgrim was performed using the same SAFER /GESTR analysis basis used for

  • he previously analyzed BP/P8x8R and GE8x8EB/NB fueltypes. Addition of the gel 1 fuel will not signifi-cantly affect the overall system response of the plant for the various operating modes, and the Gell analysis confirmed that the limiting break type and size and limiting ECCS failure (DBA recirculation suction line break with LPCIIV failure) do not change. The gel 1 fuel analysis yielded alicensing basis peak PCT of 1815 F and a peak local oxidetion fraction of <0.3%, and all licensing basis criteria are met. The Gell results are bounded by the 1825 F licensing basis PCT for BP/P8x8R fuel and the overall licensing basis results reported in Table 6-1 of the Reference analysis.

'Ihe Gell SAFER /GESTR results are applicable for a peak enriched lattice MAPLIIGR of 12.16 kw/

ft.,which bounds the MAPLilGRs for the reload 10 fuel. Therefore, the MAPLHGR limits reflect the ther-mal-mechanicallimits for the reload fuel rather than LOCA/ECCS considerations. The most limiting and the least limiting MAPLHGRs for the new fuel are as follows:

Page 9

t-

~ PILGRIM 24A5172 Reload 10 Rev.0

16. Loss-of-Coolant Accident Results (cont)

Bundle Type: GE11-P9 HUB 378-15GZ-10(TT-141-T

- Average Planar Exposure MAPLIIGR(kW/ft)

(GWd/ST)

(GWd/MT)

Most Limiting Least Limiting 0.00 0.00 9.95 10.46 0.20 0.22 10.04 10.53 1.00 1.10 10.19 10.61 2.00 2.20 10.41 10.78 3.00 3.31 10.64 10.98 4.00 4.41 10.88 11.21 5.00 5.51 11.09 11.35 6.00 6.61 11.19 11.55 l

7.00 7.72 11.30 11.63 8.00 8.82 11.40 11.70 9.00 9.92 11.52 11,79 10.00 11.02 11.65 11.91 l

12.50 13.78 11.64 11.92 l

15.00 16.53 11.48 11.71 17.50 19.29 11.25 11.47 20.00 22.05 11.02 11.22 25.00 27.56 10.55 10.75 30.00 33.07 10.07 10.22 35.00 38.58 9.39 9.53 40.00 44.09 8.72 8.87 45.00 49.60 8.06 8.23 50.00 55.12 7.40 7.60 55.00 60.63 6.72 6.96 57.02 62.86 6.44 6.69 57.10 62.94 6.68 57.92 63.84 6.56 58.02 63.96 6.55 Page 10

- PILGRIM 24A5172 Reload 10 Rev.0 f

MMMMMMM ll MMMMMMMMM MMMMMMMMMMM

MMMMMMMMMMMMM
M M M M M M M M M M M M M
H M M M M M M M M M M M M ll: H M M M M M M M M M M M M ll: M M M M M M M M M M M M M

':: M M M M M M M M M M M M M

HMMMMMMMMMMMM MMMMMMMMMMM l

MMMMMMMMM MMMMMMM ilIIililiIIIII I 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 s5 43 45 47 49 51 l

Fuel Type A=GE8B-P8DQB323-10G7 80M-4WR-145-T (Cycle 9)

C=GE11-P9 HUB 378-15GL10(TT-141-T (Cycle 11)

B.GE10-P8HXB355-11GL100M-145-T (Cycle 10)

D=BP8DRB300 (BP8x8R)

(Cycle 8)

Figure 1 Reference Core Loading Pattern l

Page 11

Reload 10 Rev.0 l

'\\

Vessel Press Rise (psi)

Neutron Flux 7*

- - - - Safety Valve Flow

- - - Ave Surface Heat Flux 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow

- -- Core Inlet Subcooling

--- Bypass Valve Flow A

~

,. 4'^ -......-__

.l,v - s EL. ;

g 75.0 g 100.0 a

a C

E Y

~

Y

~

l 50.0 25.0

~i I

+1 1

8...L I

I 0.0

~25.0 0,0 20.0 40.0 0.0 20.0 40.0 mme (sec)

Time (sec)

Level (inch-REF-SEP-SKRT)

Void Reactivity f

- - - - Vessel Steam Flow

- - - - - Doppler R4, activity 150.0 - --- Turbine Stearn Flow 1.0 - --- Scram Reactivity

--- Feedwater Flow

- - - - Total Reactivity E

~

y I,..

8 l

g 0.0

- - -,. w.=:.cy,, q,l

~~~

g 100.0

~~

E E

l' b

i p

\\

e h.,..,

.a 50.0 ll k -1.0 i

e c

t>. ;,

U I

I' I

0.0

-2.0 O.0 20.0 40.0 0.0 20.0 40.0 mme (sec)

Time (sec)

Figure 2 Plant Response to FW Controller Failure (BOC11 to EOCil-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW)

Page 12

(_

PILGRIM 24A5172 Reload 10 Rev.0 houtron Flux Vessel Press Rise (psi)

Are Surface Heat Flux

- - - - - Safety Valve Flow

.Cwe inlet Flow 300.0 - --- Relief Valve Flow 150.0 - ---

--- Bypass Valve Flow y,

,s

~,

-~----

200.0 100.0 m

m C

C Y

~

~

50.0 100.0

/

I

'I 0.0 0.0 O.0 2.0 4.0 0.0 2.0 4.0 mme (sec)

Time (sec)

(

Level (inch-REF-SEP-SKRT)

VoMaactivity Doppler Reacevity

- - - - - Vessel Steam Flow 200.0 - --- Turbine Steam Flow 1.0 Scram Reactivity

--- Feedwater Row Total Reactivity G

.E

]" p, N,%----,',.-

g' p.h.....

8 E

\\'

o g

- \\,.

0 x

\\

0.0 - Q

,L - - -.- - -

g -1.0 g

C

\\

\\k

\\

I I

I

-100.0

-2.0 O.0 2.0 4.0 0.0 2.0 4.0 Eme (sec)

Time (sec)

Figure 3 Plant Response to Load Reject w/o Bypass (BOC11 to EOC11-5787 mwd /MT (5250 mwd /ST) ANALYZED AT 102% CORE FLOW)

Page 13

PILGRIM 24A5172 Reload 10 Rev.0 i

'\\

Vessel Press Rise (psi)

Neutron Flux

- - - - - Ave Surface Heat Flux I

- - - - - Safety Valve Flow 150.0 - --- Core inlet Flow 125.0 - --- Relief Valve Flow

- - - Core inlet Subcooling

--- Bypass Valve Flow

{

~-

p __,,

~Q

--~~-

3 3

~'

C C

Y r~

l s

~

25.0 50.0 l

I lc..L

=.-

0.0

-25.0 O.0 20.0 40.0 0.0 20.0 40.0 Time (sec)

Time (sec)

Level (inch-REF-SEP-SKRT)

Void Reactivity

- - - - Vessel Steam Flow

- - - - - Doppler Reactvity 150.0 - --- Turbine Steam Flow 1.0 - --- Scram Reactivity l

--- Feedwater Flow

--- Total Reacevity I

's' g

g 100.0

~~ ~ }

0.0

-- K ?. ;,--

,q,,,,,,,,

16 l

l', '

E

).

O F

(...

I O

o p

)

P l"'

.a

,'.l'l N -1.0 50.0 o

L x

~

~

c.

i'.

b' 0.0

-2.0 O.0 20.0 40.0 0.0 20.0 40.0 Eme (sec)

Time (sec)

Figure 4 Plant Response to FW Controller Failure (EOCll-5787 mwd /MT (5250 mwd /ST) to EOC11 ANALYZED AT 107.5% CORE FLOW) i Page 14

' PILGRIM 24A5172 Reload 10 Rev.0 L

boutron Flux Vessel Press Rise (psi)

-Ave Surface Heat Flux

- - - - Safety Valve Flow 150.0 - --

-Cwe inlet Flow W.0 - --- Relief Valve Flow

--- Bypass Valve Flow l

/ ' S;,

N-"~]

-~

N g 200.0 g 100.0 N

w s

s p

\\

C C

f i

50.0 100.0 L

7.---------

/

I

  1. I 0.0 0.0 O.0 2.0 4.0 0.0 2.0 4.0 Time (sec)

Time (sec)

Level (inch-REF-SEP-SKRT) id

'vity

- - - - - Vessel Steam Flow pp Reactivity 200.0 - --- Turbine Steam Flow 1.0 Sc Reactivity

--- Feedwater Flow Total Reactivity

?

{\\

g

, ~~~~.....-

0.0 u 100.0 v',. \\,

~. -.

~ ~ ~

\\.

S

\\

a

\\.

s O

\\

- %,i N

\\

\\,

o s..

-A

-1.0

)

0.0

~

e

\\

\\

~

\\

\\

\\

\\

I

-100.0

-2.0 O.0 2.0 4.0 0.0 2.0 4.0 Time (sec)

Time (sec)

Figure 5 Plant Response to Load Reject w/o Bypass (EOCll-5787 mwd /MT (5250 mwd /ST) to EOCil ANALYZED AT 107.5% CORE FLOW)

Page 15

PILGRIM 24A5172 Reload 10 Rev.0 I

Nwtron Flux Vessel Press Rise (psi)

- - - - - Safety Valve Flow ke Surface Heat Flux

. 300.0 - --- Relief Valve Flow 150.0 - ---

C<re inlet Flow

--- Bypass Valve Flow

'Y \\m 100.0 g200.0 E

\\

E E

g C

~

N l

N N

l s

f 50.0 100.0 I

I

'/

't..............

0.0 0.0 0.0 4.0 8.0 0.0 4.0 8.0 Time (sec)

Time (sec)

Level (inch-REF-SEP-6KRT)

Void Reactivity

- - - - - Vessel Steam Flow

- - - - - Doppler Reactivity 200.0 - --- Turbine Steam Flow 1.0 - --- Scram Reactivity

- - - Feedwater Flow

- - Total Reactivity 6

to E

e

,s'%

m 9'.

g. 0.0 t3 100.0

.s e

F

, s x...*..... _... _ _ m. 3

' T)

. \\...

. gi.

p

\\,

c

.a E -1.0 v

'r,

0.0 e

~

l

-100.0

-2.0 0.0 4.0 8.0 0.0 4.0 8.0 Time (sec)

Time (sec)

Figure 6 Plant Response to MSIV Closure (Flux Scram)

Page 16

PILGRIM 24A5172 Reload 10 Rev.0 Appendix A Analysis Conditions and Bases To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle. The use of the increased core flow for the analysis produces bounding results for the flow range down to 75% of rated core flow. Justification of operation at 100% power down to 75% power is pmvided in Reference A-1. The cycle 11 licensing analysis has verified the applicability of the MELLL flow range.

1 1

Table A-1 Parameter 107.5% Flow Analysis Value Thermal power, MWt 1998.0 Core flow, Mlb/hr 74.2 Reactorpressure, psia 1066.5 Inlet enthalpy, BTU /lb 528.4 Norbfuel power fraction 0.038 Steam flow analysis, Mlb/hr 7.98 1

Dome pressure,psig 1035.8 Turbine pressure, psig 975.7 No. of Safety / Relief Valves 4

No. of Single Spring Safety Valves 2

Relief mode lowest setpoint, psig 1126.0 Safety mode lowest setpoint, psig 1253.0 For the overpressurization analysis, the MSIV closure (flux scram) case was analyzed at 102% licensed pow-er and steamflow. Also, the maximum possible initial steam dome pressure of 1085 psig was used, which corresponds to the high pressure scram analytical limit. The most limiting end of cycle core conditions were utilized at 107.5% core flow, which producer s bounding result.

For the first intmduction of Gell fuel in Pilgrim, a plant specific evaluation was made of the Gell fuel

" Scram Speed Adjustment Factors"(SSAF) that adjust the option B MCPR limit to optain the Option A MCPR limit. This evaluation concluded that use of a 0.06 EOC scram speed adjustment factor is justified for the load rejection, turbine trip, and feedwater contmiler failure pressurization events. For Pilgrim gel 1 fuel application, this supercedes the " generic" EOC value from the letter, J.E Klappmth to USNRC, " GEM-INI/ODYN Statistical Adders for GElI fuel for BWR/2 and 3", September 23,1992. The 0.08 generic mid-cycle addem are still applicable to Gell fuel in Pilgrim.

A-1. II.X. Hoang, " Maximum Extended LoadLine Limit Analysesfor Pilgrim Nuclear Power Station Re-load 9 Cycle 10", NEDC-32306P, March,1994.

Page 17

PILGRIM 24A5172 -

[-

Reload 10 Rev.0 Appendix B Increased Core Flow 1

The analyses performed for Cycle 11 included increased core flow throughout the cycle and after the all-rods-out condition is reached. There are no concerns regarding reactor intemals pressure drop or flow-in-duced vibration as discussed in the increased core flow analysis document for the EOC-6 (NEDO-30242)..

l i

Page 18

PILGRIM 24A5172 Reload 10 Rev.0 Appendix C Decrease in Core Coolant Temperature Events The loss-of-feedwater heating (LFWH) and the high pressure coolant injection (HPCI) inadvenent stan-up l

anticipated operational occunences (AOO) are the only cold water injection events checked on a cycle-by-cycle basis. For both the LFWH and HPCI events, the delta CPR is not limiting when compared to the delta CPR of the liming pressurization AOO. This is based on the results of calculations perfonned with consider-ation of the cycle-to-cycle diffemnces such as ARTS. Then: fore, the LFWH and HPCIinadvenent stan-up AOOs are not reponed for Cycle 11.

f Page 19

Memo to E.L. Heinlein from G. A. Watford dated February 3,1992;

Subject:

PNPS Technical Specification Scram Time Requirements - DRF A12 - 00038 - 2

.~

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

i'

~

,o.

l February 3, 1992 cc:

J.S. Charnley i

E.G. Thacker D.C. Serell To:

E.L. Heinlein DRF A12-00038-2 From:

G.A. Watford

Subject:

PNPS Technical Specification Scram Time Requirements

Reference:

1.

Letter, R.V. Fairbank (BECo) to E.L. Heinlein (GE), same subject, 11/21/91.

2.

Letter, R.V. Fairbank (BECo) to E.L. Heinlein (GE), same subject, 12/18/91.

3.

Letter, G.A.

Watford to E.L.

Heinlein, same subject, 1/22/92.

This letter summarizes the information provided in Reference 3 and also provides additional information concerning the GEMINI scram times.

The responses are also provided in the same format as the questions of References I and 2.

1)

Average scram insertion time requirements for all operable control rods (TS 3.3.b.1) from deenergization of the scram pilot valve solenoids to dropout (DO) (reed switch opening) of Notches 04, 24, 34, and 44.

Average Notch Scram Time Position (seconds) 44 DO 0.504 34 DO 1.249

'24 00 2.013 04 00 3.575 2)

Average scram insertion time requirements for the three fastest control rods in each group of four control rods in all two-by-two arrays (TS 3.3.c.2) from deenergization of the scram pilot valve solenoids to dropout of Notches 04, 24, 34, and 44.

3 out of 4 Notch Scram Time Position (seconds) 44 DO 0.534 1

34 00 1.324 24 DO 2.134 04 DO 3.790 3)

.The y and a values based on scram insertion times from deenergization of scram pilot valve solenoids to dropout of Notch 34 which are used to calculate 78 (TS 4.11.C) consistent with GEMINI advanced physics methods.

y - 0.937 seconds a = 0.021 seconds

=_.

Page 2 E.L. Heinlein i

February 3, 1992 4)

Correction factors required to account for measurement biases and uncertainties when demonstrating compliance with the scram insertion times requested in Items 1 and 2 above.

The limits specified in the responses to Items 1, 2, and 3, explicitly account for the uncertainties in the location of the position indication probes and for the uncertainty in the control rod position when pickup or dropout of the reed switch occurs.

Any other measurement uncertainties and biases ' introduced by the BECo surveillance procedures and hardware configuration used in the measurements are specific to Pilgrim and are not included in the specified limits (e.g., determination of time zero, accuracy of measurement devices, etc.).

')

16,

i t

c verified by:

.A.

atford E.Y.

bo, LSE Systems Integration Engineering Control Rod Drive System M/C 740, Tel. 5-6136 Reactor Design Engineering M/C 771, Tel. 5-6783 l

i i

Memo to E.L. Heinlein from S. J. Peters dated September 3,1993;

Subject:

Time to Notch 34,24, and 04 Dropout for Pilgrim - RNE93-260

i Fuel Engineering General Electric Nuclear Energy

-San Jose, California 175 Curtner, San Jose, CA 95125 RNE93-260 September 3, 1993 TO:

E.L. Heinlein FROM:

S.J. Peters

SUBJECT:

Time to NOTCH 34, 24, and 04 DROPOUT for Pilgrim

REFERENCE:

Letter, E.L.

Heinlein to J.H. Paiscik, " Technical Specification Scram Time Requirements", February 5, 1993.

The referenced letter contains scram times to assure technical specification compliance for the fastest three rods in a clumped 2X2 control rod array at Pilgrim.

At BECo request, the purpose of this letter is to update the time requirement for 10%, 30%, 50% and 90%

insertion if it is determined by measuring from the NOTCH 44 DROPOUT, NOTCH 34 DROPOUT, NOTCH 24 DROPOUT and NOTCH 04 DROPOUT, respectively.

The values are shown in the table below and they supersede the values reported in the referenced letter.

1 NOTCH 44 DROPOUT 0.538 seconds NOTCH 34 DROPOUT 1.327 seconds NOTCH 24 DROPOUT 2.137 seconds NOTCH 04 DROPOUT 3.793 seconds These values are based on removing the conservative assumption that the control reed switch is at the minimum tolerance, reasonable for averaging multiple control rod drives.

All other effects discussed in the referenced letter remain conservatively included.

1 If you have any questions please call.

J V1 M

Verified by:

S. J./ Peters '

/

J.F. Casillas Reidad Nuclear Engineering 2 Reload Nuclear Engineering 1 M/C 156, Ext. 51124 M/C 171, Ext. 56910 cc.

P.J. Savoia E.G. Thacker II DRF J11-02042 4

i

S&SA Calculation 088 dated 6/28/95;

Subject:

Scram times for Tech Spec 3.3.C.1 l

l l

l l

l

Retrieved from "https://kanterella.com/w/index.php?title=ML20092G932&oldid=4332428"