Text

Supplemental Reload Licensing Submittal for Fermi Plant Unit 2,Reload 1,Cycle 2
	ML20248J507
Person / Time
Site:	Fermi
Issue date:	03/31/1989
From:	Charnley J, Plotycia G, Rash J; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML19297H484	List: ML20248B419; ML20248J493; NRC-89-0052, Supplemental Reload Licensing Submittal for Fermi Plant Unit 2,Reload 1,Cycle 2;
References
	CON-NRC-89-0052, CON-NRC-89-52 23A5949, NUDOCS 8904140451
	Download: ML20248J507 (30)
	v • d • e

1 a

NRC-89-0052

]

i 23A5949 j

Revision 0 j

class I i

March 1989 l

23A5949 REV. O SUPPLEMENTAL RE1 DAD LICENSING SUBMITTAL g

i FOR FERMI POWER PIANT UNIT 2 RELOAD 1, CYCLE 2 1

t i

Prepared:

.L.' Rash

{

pFuelLicensing j

Verified:

'G.D. Plotyci[

Fuel Lice j

C' Approv d: e.

5 J

rnley ager, Fuel Licensi

(

O GENeckerEnergy c

175 Curw Anna SnJm. CA 95125 8904140451 890403 PDR ADOCK 05000341-PDR 1/2 p

23A5949 R:v.' O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by the General Electric Company (GE) solely for Detroit Edison Company (DECO) for Deco's use with the United States Nuclear l

Regulatory Commission (USNRC) for amending Deco's operating license of the Fermi Power Plant Unit 2.

The information contained in this report is I

believed by CE to be an accurate and true representation of the facts known, obtained or provided to CE at the time this report was prepared.

The only undertakings of GE respecting'information in this document are b

contained in the contract between DECO and GE for nuclear fuel and related services for the nuclear system for Fermi 2, 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 docu-ment makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this O

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 a.ny kind which may result from such use of such information.

O I

r l

I 3/4 1

\\

23A5949 Rov O ACKNOWLEDGEMENTS The engineering and reload licensing analyses, which form the technical l

basis of this Supplemental Reload Licensing Submittal, were performed in the Fuel Engineering Section by J.L. Casillas and M.W. Thompson, and H.L. Hubeny of Deco.

I l

t i

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5/6

23A5949 Rov. 0 1.

PLANT-UNIOUE ITEMS (1.0)*

d Appendix A: ADO Analysis Initial Conditions Appendix B: Analyzed Operating Domain l

Appendix C: Application of CEMINI Methods Appendix D: GEXL-PLUS Correlation Appendix E: Rod Withdrawal Error Analysis 2.

RELOAD FUEL BUNDLES (1.1 and 2.0)

Fuel Tyne Cvele Loaded Number Irradiated l

I 8CR183**

1 112 8CR233**

1 432 New 0

BC318D 2

72

~

BC318E 2

Ift).

]

Total 764

(

C)

I l

3.

REEERENCE CORE LOADING PATTERN (3.2.1)

Nominal previous cycle core average exposure 9,515 mwd /St C'

at end of cycle:

Minimum previous cycle core average exposure at 9,215 mwd /St end of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at 14,769 mwd /St

f end of cycle

Core loading pattern:

Figure 1

( ) Refers to area of discussion in General Electric Standard Application l

for Reactor Fuel, NEDE-24011-P A-9 (dated September 1988); a letter "S"

'r preceding the number refers to the United States Supplement.

- As identified in Fermi 2 Technical Specifications.

Denoted as P8CRB176 and P8CRB219, respectively, in NEDE-24011-P-A 8, Table D-3.

7

23A5949 Rsv. 0 4.

CAlfUIATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS. 20*C (3.2.4.1 and 3.2.4.2)

Beginning of cycle, k-effective Uncontrolled 1.108 Fully Controlled 0.950 Strongest Control Rod Out 0.971 R, Maximum Increase in Cold Core Reactivity with 0.019 g

Exposure into Cycle, ak 5.

STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.2.4.3)

PPM Shutdown Margin (ak) g (Natural Boron)

(20*C Xenon Fregl 660*

0.046 g'

6.

RETAAD-UNIOUE ANTICIPATED OPERATIONAL OCCURRENCE (AOO) ANALYSIS INPUT (3.2.3 and 4.3.1.2.3)

(Cold Water Injection Events only)

EQG Void Fraction (%)

43.1 Average Fuel Temperature (*F)'

1082 Void Coefficient N/A** (g/% Rg)

-7.78/-9.73 Q

Doppler Coefficient N/A** ( /'F)

-0.246/-0.233 l

Scram Worth N/A ($)

1 IC

or equivalent enriched Boron

- N - Nuclear Input Data, A - Used in Transient Analysis c

- - Ceneric exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE 24011 P-A-9, dated September 1988.

t 8

23A5949 R:v. 0 7.

RELOAD-UNIOUE GETAB'A00 ANALYSIS INITIAL CONDITION PARAMETFDR (S.2.3) l J

Exposure: BOC2 to EOC2-2000 mwd /St

]

i Fuel Peakina Factors Bundle Power Bundle Flow Initial

Design Igtgal Radial gigi R-Factor (MWT)

(1000 lbAri MCPR P8x8R 1.20 1.54 1.40 1.051 6.494 103.6 1.22 GE8x8EB 1.20 1.54 1.40 1.051 6.497 105.1 1.22 Exposure:

EOC2-2000 mwd /St to EOC2-1000 mwd /St Fuel Peakina Factors Bundle Power Bundle Flow. Initial

df Design Jaggal Radial gigi R-Factor (MWT)

(1000 lbAri MCPR P8x8R 1.20 1.52 1.40 1.051 6.398 104.3 1.24 GE8x8EB 1.20 1.52 1.40 1.051 6.401 105.8 1.24 Exposure:

EOC2-1000 mwd /St to EOC2 Fuel Peakina Factors Bundle Power Bundle Flow Initial

Design Jaggal Radial higi R-Factor (MWT)

(1000 lbAri MCPR I

P8x8R 1.20 1.49 1.40 1.051 6.305 105.0 1.26 GE8x8EB 1.20 1.50 1.40 1.051 6.309 106.5 1.26 Exposure BOC2 to EOC2 with Turbine Bypass System out of Service **

O Fuel Peakina Factors Bundle Power Bundle Flow Initial

Design

]gtgal Radial higl R-Factor (MWT)

(1000 lbAr)

MCPR P8x8R 1.20 1.51 1.40 1.051 6.340 104.2 1.30-CE8x8EB 1.20 1.51 1.40 1.051 6.318 105.9 1.32 Exposure:

BOC2 to EOC2 with Turbine Bypass System and Moisture Separater Reheater Out-of-Service l

Fuel Peakina Factors Bundle Power Bundle Flow Initial

Desien Lggal Radial Axial R-Factor (MWT)

(1000 lb/hr)

MCPR P8x8R 1.20 1.41 1.40 1.051 5.956 107.6 1.34 GE8x8EB 1.20 1.42 1.40 1.051 5.975 108.9 1.34

Initial MCPR may not exactly equal ACPR (Section 10) + 1.07 due to rounding and calculational convergence.

- With up to 150*F Feedwater Reduction (270'F Feedwater Temperature) 9

I 23A5949 Rov. 0 8.

ERTR M ED MARGIN IMPROVEMENT OPTIONS (S.S.1)

Recirculation Pump Trip:

No Rod Withdrawal Limiter:

No 4

Thermal Power Monitor:

Yes Improved Scram Time:

Yes (ODYN - Option B)

Exposure Dependent Limits:

Yes Exposure Points Analyzed:

3 (EOC2-2000 mwd /St.

EOC2-1000 mwd /St, EOC2) l 9.

OPERATING FLEXIBILITY OPTIONS (S.S.2)

El Single-Loop Operation:

Yes Imad Line Limit:

No Maximum Extended Load Line Limit:

Yes Increased Core Flow:

No C

Flow Point Analyzed:

N/A Feedwater Temperature Reduction:

Yes ARTS Program:

No Maximum Extended Operating Domain:

No G

10. CORE-VIDE A00 ANALYSIS RESULTS (S.2.2)*

C Methods Used: CEMINI and GEXL-PLUS**

Exposure Range: BOC2 to EOC2 Flux Q/A ACPR I

Transient

(% NBR)

(2 NBR) Z6xBR GE8x8EB Figure Inadvertent HPCI Activation 120 117 0.15 0.15 2

See Fermi-2 FSAR for Fermi A00 Descriptions

- See Appendices C and D 10

23A5949' Rov. 0 Exposure Range: BOC2 to EOC2 2000 W d/St Flux Q/A ACPR Transient (1 NBR)

(1 NBR) 18281 GE8x8EB Zigugg Load Rejection w/o Bypass 434 118 0.15 0.15 3

Exposure Range: EOC2-2000 W d/St to EOC2-1000 E d/St ACPR Flux Q/A.

Transient (1 NBR)

(1 NBR) E8x8B GE8x8E3 Ziggga Load Rejection w/o Bypass 429 120 0.17 0.18 4

3 Exposure Range: EOC2-1000 Wd/St to EOC2 Flux Q/A ACPR Transient

(% NBR)

(% NBR) E8x1R GE8x8EB Iiggga Load Rejection w/o Bypass 437 122 0.19 0.20 5

Feedwater Controller Failure 245 114 0.11 0.12 6

0 Exposure Range: BOC2 to EOC2 with Turbine Bypass System Out-of-Service

Flux Q/A' ACPR Transient (1 NBR)

(1 NBR) IBx8E GE8x8EB Firure

[l Feedwater Controller Failure 449 128 0.24 0.24 7

Exposure Range: BOC2 to EOC2 with Turbine Bypass System and Moisture Separator Reheater Out-of-Service 1

Flux Q/A ACPR Transient

(% NBR)

(% NBR) Egg 8E GE8x8EB Eiggga Load Rejection w/o Bypass 637 (25 0.22 0.23 8

Feedwater Controller Failure 686 132 0.27 0.28 9

l i

k

With up to 150*F feedwater temperature reduction (270*F feedwater i

temperature).

i 11 4

I

(

23A5949 R3v. 0

11. IDCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) ADO

SUMMARY

(S.2.2.1.5) i Base (for application with only Control Cell Core control rods inserted; 1

see Appendix E).

j Limiting Rod Pattern: Figure 10 Rod Block Rod Position ACPR ACPR Readine (%)

(Feet Withdrawn) f.616E GElgBJJ l

104 4.5 0.14 0.12 105 5.5 0.15 0.16 106 6.0 0.16 0.18 107 10.0 0.19 0.22 108 11.0 0.20 0.22 109 11.5 0.20 0.22 g

110 12.0 0.20 0.22 Setpoint Selected: 106 C

Alternate (for application with c n-Control Cell Core rods inserted; see l

Appendix E).

Limiting Rod Pattern: Figure 11 l

Rod Block Rod Position ACPR ACPR Readine (%)

(Feet Withdrawn)

P8x8R GE8x8EB 104 4.5 0.14 0.20 0

105 5.0 0.15 0.21 106 5.5 0.16 0.23 107 7.5 0.18 0.26 108 9.5 0.20 0.27 j

109 10.5 0.20 0.27 110 11.0 0.20 0.27 Setpoint Selected: 106 l

l i

-(

6 12

l 23A5949 Rov. 0

,)

b L-:

12.

CYCLE MCPR VALUES (4.3.1. S.2.2)

Non-Pressurization Events Exposure Range:

BOC2 to EOC2 EB181 GE8x8EB Inadvertent HPCI Activation 1.22 1.22 Rod Withdrawal Error (Alternate)

(See Appendix E) 1.23 1.30 3

Rod Withdrawal Error 1.23 1.25 Pressurization Events p

Exposure Range: BOC2 to EOC2-2000 mwd /St Ontion A*

Ontion B*

ElsBE GE8x8EB E3XBE GE8x8EB 3

- Load Rejection w/o Bypass 1.32 1.32 1.25 1.25 I

Exposure Range:

EOC2-2000 mwd /St to EOC2 1000 mwd /St

-)

Ontion A*

Ontion B*

P8x8R GE8x8EB P8x8R GE8x8EB Load Rejection w/o Bypass 1.34 1.34 1.27 1.27 t

0ption A and B adjustment factors were transmitted to the NRC in Reference C-3, 13

23A5949 Rav. O j

Exposure Range: EOC2-1000 MWD /St to EOC2 Detion A Detion B j

Z8x8B GE8x8EB Z118R GE8x8EB j

lead Rejection w/o Bypass 1.34 1.34 1.28 1.28 r

Feedwater Controller

]

Failure 1.22 1.23 1.19 1.20 Exposure Range:

EOC2 with Turbine Bypass System (or Moisture

[

Separator Reheater) Out-of-Service Dotion A Dotion B ZBXBE GE8x8EB Z1KHE GE8x8EB Feedwater Controller 0

Failure 1.35 1.36 1.32 1.33 Exposure Range:

EOC2 with Turbine Bypass System and Moisture Separator Reheater Out-of-Service Ootion A Ontion B

(

Z1EEE GE8x8EB P8x8R GE8x8EB Load Rejection w/o.

Bypass 1.35 1.35 1.31 1.31 9

Feedwater Controller Failure 1,39 1.39 1.36 1.36

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.3) t Steam Line Vessel Pressure Pressure Plant Transient (osie)

(osie)

Resconse HSIV Closure (Flux Scram) 1252 1276 Figure 12 C

14.

IDADING ERROR RESULTS (S.2.2.3.7)

The mislocated loading error analysis is not performed for reload cores.

I The disoriented loading error analysis is not performed for C-lattice plants.

(

14

23A5949 R:v. 0 3

15. CONTROL ROD DROP ANALYSIS RESULTS (S.2.2.3.1) l Banked Position Withdrawal Sequence (BPWS) has been implemented at Fermi 2; therefore, the Control Rod Drop Accident Analysis is a generic l

analysis applied to Cycle 2. NRC approval is documented in NEDE-24011-P-A-9-US, September 1988.

16.

STABILITY ANALYSIS RESULTS (S.6)

GE SIL-380 recommendations and GE interim correction actions have been b

included in the Fermi 2 operating procedures. Regions of restricted operation defined in Attachment 1 to NRC Bulletin No. 88-07 Supplement 1, are applicable to Fermi 2 Reload 1.

b

17. IDSS-OF-COO 1 ANT ACCIDENT RESULTS (S.2.2.3.2)

LOCA Method Used:- SAFE /REFLOOD/ CHASTE (See Feral 2 FSAR, LOCA Analysis)

E MAPLHCR (kW/ft)

Average Planar BC318D BC318E Peak Clad Exposure Most Least Most Least Temperature Oxidation (CWd/St)

Limitinz Limiting Limiting Limiting

(*F)

Fraction 0.0 12.02 12.07 11.99 12.02 1943 0.03 1.0 12.14

'12.20 12.10 12.14 1954 0.03 5.0 12.93 13.00 12.79 12.93 2003 0.03 8.0 13.28 13.30 13.15 13.28 2018 0.04 10.0 13.34 13.35 13.34 13.35 2022 0.04 12.5 13.33 13.33 13.32 13.33 2023 0.04 15.0 13.02 13.02 13.02 13.02 1985 0.03 25.0 11,75 11.75 11.75 11.75 1850 0.02 45.0 9.05 9.05 9.04 9.05 1606

<0.01 50.0 6.64 6.64 6.63 6.64 1520

<0.01 15

k' 23A5949 Rev. 0

.HMMMMMM.

EHHHHHHHHH

..HHHHHHHHHHH.

.HHHHHHHHHHHHH.

k

HHHHHHHHHHHHHHH l

CHHHHHHHHHHHHHHH

':: H H H H H H H H H H H H H H H l' H H H H H H H H H H H H H H H o

o

':: H H H H H H H H H H H H H H H

.i;: H H H H H H H H H H H H H H H CHHHHHHHHHHHHHHH

---"HHHHHHHHHHHHH" 1:

"'H H H H H H H H H H H""

EHHHHHHHHH

'-" H H H H H H H "

IIIIIIIIIIIIII 1 3 5 7 911131517192123252729313335373941434547495153555759 p.

l l

FUEL TYPE A = 8CR183 C = BC318D B = 8CR233 D = BC318E Figure 1.

Reference Core Loading Pattern 16 l

23A5949 R v. O.

l I

1VES5EL PRESS RISE (PSI) 1 NEurRON FLUX l.

2 AVE SURFACE HEAT FLUX 2 REL IEF VALVE FLOW 3 COR : INLET FLOW 3 BYPLSS VALVE FLOW 130.0

' C "" "'"E'""

150.0

' "*E ' 'LO" " S' '"'

+-

U 12 100.0 0

200.0 E

l 8

s.

U W

i 2

.)

S 0. 0

50. e

~

0-

0. 0 e;;

S.

O0 50.0 100.0

0. 0 50.0 100.'

TIME (SECONOS)

TIME (SECONOS) 1 5

1 LEV EL(INCH REF-SEP-SKRT) 1 VO! ) REACT!v!TY 2 DOP'LER REACT!v!TY 3 SCRAM REACTIVITY I

2 VESSEL STEANFLOW 3 TUR BINE STEAMFLOW at nr *retunv a w 150.0 e err'uivre eteu t.0 3

3 g

100.0 f "' ~ ~ ~

~

""~

8. 0 A

^^

U s.

g A

M.n.0 50.0 W

m l

1

8. 0

-2.0

0. 0 50,8 100.0 0.0 50.0 100.0 TIME (SECONOS)

TIME (SEco@ s)

Figure 2.

Plant Response to Inadvertent HPCI Activation 17

t.

23A5949 Rev. 0 1 NEUTRON FLU <

1 VESSEL PRESS RISE (PSI) 2 AVE SURF ACE E AT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 15 8.0 300.0

'. ?<=*S? untur etow E

(

200.0 W 100.0 5

b i

N 300.0 50.0 g

0 i

(!

0.0 O.0 0.0 2.0 4.0 6.0 0.0

2. 0 4.0 6.0 TIME (SECONDS)

TIME (SECONDS)

E 1 LEVEL (!NCH-REF-SEP-SKRT) 1V REACTI VITY 2 VESSEL STEAiFLOW 2 DOP REA:TIVITY 3 TURBINE STELMFLOW 3 SCRAM R IVITY I

' 'O'*L "T

200.0

' 'T ED" T"

' D' 3

0.0 100.0 ;

p

~M. --- -e -

B t

^ ^ ^

h-1.0

^

0.0 s.

y g

l 2.0

-100.0

0. 0 2.0 4.e s.0 0.0

2. 0 4.0 s0 1

TIME (SECOND$)

TIME (SECONDS)

Figure 3.

Plant Response to Generator Load Rejection Without Bypass (EOC2 2000 MVd/St) 18

1 23A5949 Rev. 0 1

1 1 NEUTRON FLU <

1 VESSEL PRES 5 RISE (PSI) f 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW i

3 CORE IPLET TLOW 3 RELIEF VALVE FLOW l

199. 0 300.0

' ava

SS u = LuE etow I

200.0 100.0 I

e 50.0 300.0 0

r g

0.0

0. 0 I

0. 0 2.0 4.0 6.0 0.0 2.0 4.0 6.0 TIME (SECONOS)

TIME (SECONDS)

)

1 LEVEL (INCH-REF-SEP.SKRT) 1 010 REACTIVITY 2 VESSEL STEA1 FLOW 2 PPLER REACT!v!TY 3 TURBINE STEAMFLOW 3S REACTIv!TY I'0

^ " ' " " ' '

o60.0 M

0

^

100 0 y

y

)w x

4 5

-I W

.l.0

0. 0 1 -

r_

g

\\

100.0 2.0

0. 0 2.0 4.0 6.0 0.0 2.C

4. 0 6.0 TIME (SECON0s)

TIME (SECOND$)

Figure 4.

Plant Response to Generator Load Rejection Without Bypass (EOC2 1000 !Wd/St) 19 i

f

\\

L I

23A5949 Rov. 0

.I8IEUTRON FLUK 1 VESSEL PRES 5 RISE (PSI) 2 AVE ~ SURFACE TEAT FLUX 2 SAFETY VALVE FLOW 3

I OW a,,,

} gy[ { 3:[ y 333,3 E....

)

2....

g v

~

l 1

E 3...

l....

M I'

88 88 2.0 4.0 6.0

.0

2. 0 4.0 6.

TIME (SECONDS]

TIME (SECONDS)

I 1 LEVEL (INCH-REF-SEP-SKRT) 1 r0!D REACTI VITY I

!. !.'an.,.5?d.O%

!p1EEii??"

.u

~

i..

ss~

V k__

f' a*

.....L

/

' " ' ' q.h y

/

% v..

r m

_=

1 e

lb:.

y E...

L

=

g V

I

-i..

-2..

4..

1 TIME (K CON 0s)

TIME ( K C02 $1 Figure 5.

Plant Response to Generator Lead Rejection Without Bypass (EOC2) 20

23A5949 Rsv. 0 150.0 1 NEurRON FL'UX

\\

l VES 5EL PRESS RISE (PSI) 2 AVE SURFACI:

AT X

2 SAF ETY VALVE F OW 3 CORE INLE1 LOW 3 REL lEF VALVE F W 150.0

' CO*

'"?'

U?

4 SYPNS$ VALVE L 100.0 8

.Y

\\

W 100.0

" "g m.- -

b 50.0 U

M lt' k'

.h

.) '

30.0 V

3 ll 0.0 N

]

/_

0. 0 g

0. 0 10.0 20.0

0. 0 10.0 20.

TIME (SECONOS)

)1 1 LEVEL (!NCW-REF-SEP-SKRT) i VO!b REACT!d!T 2 VESSEL STEANFLOW 2 00PGLDR RE ( T! IT 3 D._ N_ N 150.0 e_

__u 1.0 0

3) i 3

1 100.0 2

T 0.0

_ 'f a

a a g

=

.C l\\

S 0. 0 J

M -1.0 t

V g

W 1

t

0. 0

-2.0

0. 0 10.0 20.0

0. 0 10.0 20.0 TIME (SECON05)

TIME (SECONOS)

Figure 6.

Plant Response to Feedwater Controller Failure (EOC2) i 21

23A5949 Rsv. 0 350.0 1 NEurRON FLUX 1 WES l P $$ RISECPSI) 2 Avl: SURFACE E AT FLUX 2 SAF :T VA VE FLOW 3 Col : INLET FLOV 3 REL llF VAL E FLOW m_!' S'?

4 BYPU $ VAL Flow 130.0

'=

100.0 l

, 100.0 <.i S

50.0 t

U k

g

=

g 50.0 V

N 0.0 h

1

,,0,

{

b' O. 0

0. 0 10.0 20.0

0. 0 10.0

20. f TIME (SECONDS)

TIME (SECONDS)

(

1 LEV *L(INCR-REF-SEP-SKRT) 1 V01 ) REA IVITY 2 VES sEL STEAMFLOW 200F 'lER R CTIV Y

.0 5^cLvi'.

5 3 !? "

' !y"L3"5r'P So*

13 0. 0 t

a E

W 8.,

300.0 B

E' C

E 50.0

.l.0 i

c a

W. O

-2.0

0. 0 10.0 20.0

0. 0 30.0 20.0 TIME (SECOW S)

TIME (SECONOS)

(-

)

Figure 7.

Plant Response to Feedwater Controller Failure (EOC2 - Turbine Bypass System Out of-Service)

I 0

22 3

i l

23A5949 Rev. 0 I NEUTRON FLU (

1 VESSEL PRESS RISE (P$!)

2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 15 $. 0 300.0

' ev= = ?? u = Lug rLew

}l Ei n

200.0 W 10 0. 0 6

b u

b 300.0 50.0 v\\

?

O 0

2

0. 0 0.0 S

0. 0 2.0 4.0

6. 0

0. 0 2.0

4. 0 6.0 TIME (SECONOS)

TIME (SECONOS)

)

1 LEVEL (INCH-REF-SEP-SKRT) 1 10 REACT!v!TY 2 VESSEL STEAiFLOW 2 PPLER REA:TIVITY 8*0 200.0 E

/ '

t{-

l$ 0. 0 3

0.0..

Ovv w n

-e

=

H l

W

-1.0 t

0. 0 l

.i l

-100.0

-2.0

0. 0 2.0 4.0 6.0 0.0 2.0 4.0 See TIME (SECONOS)

TIME (SECONDS) l Figure 8.

Plant Response to Generator Load Rejection Without Bypass (EOC2 - Main Steam Reheater Out of-Service) l 23 l

L___._-___---_____________.

{

23A5949 Rev. 0 150.0 1NEU h0N FLd 1 VES$4L PR S RISE (PSI) 2 AVE St.RFACE AT FLUX 2 SAFE 'Y VAL E FLOW 3 COR INLET

~ OW 3 RELI EF VAL

FLOW 130.0 "O"

!NLE' c__

4 BYPA SS VALV' FLOW 100.0 g

/

\\

h td 100.0 g

7 I

50.0 g

W W

W i

S 0. 0 I

~

u V

Y

0. 0 O'C4 ^ ;

I,,

0. 0

0. 0 10.0 20.0 30.0

0. 0 10.0 20.0 30.0 TIME (SECONOS)

TIME (SECONOS)

I 1 LEVEL (INCH-REF-SEP-SKRT) 1 Voli REACTl vlTY 2 VESSEL STEA1 FLOW 2 DOPI L En REACTiv!TY STEW LOW 3 S,CR,,N It E A. CTI, V I, T,Y 3 TURBIE.,ge e 1.0 g

. - re u

v g

I'30.0 2

'egggu gu

. O ;

4 a

r, r

i 0.0 100.0 7

74 8

C

\\

n

-1.0 00.0 w

A e

I a

0. 0

. a

-2.0

0. 0 10.0 20.0 30.0

0. 0 10.0 20.0 30.0 TIME (SECONDS)

TIME (SECONOS)

Figure 9.

Plant Response to Feedwater Controller Failure (EOC2 - Turbine Bypass System and Moisture Separator Reheater Out of-Service) 24 r

L

23A5949 R:.v. 0 1

3 5

7 9

11 13 15 17 19 21 23 25 27 29 3

1 3

6 0

0 0

6 5

)

7 0

6 10 0

10 6

0 9

40 40 11 0

10 10 0

g 13 15 0

6 0

0 0

6 0

17 3

19 0

10 10 0

21 40 40 23 0

6 10 0

10 6

0 9

25 27 6

0 0

0 6

29 j

NOTES:

1.

No. indicates number of notches withdrawn out of 48.

Blank is a Withdrawn Rod.

2.

Error Rod is (15, 11).

Figure 10.

Limiting Rod Pattern (Base) 25

-- ]

i 23A5949 R:;v. 0

)

l l

l 1

3 5

7 9

11 13 15 17 19 21 23 25 27 29 l

i 1

{

3 6

0 0

6 8-5 l

7 10 6

0 0

6 10 9

0 11 6

12 0

0 12 6

l 13 15 6

6 10 0

0 10 6

6

-Q 17 19 6

12 0

0 12 6

21 0

23 10 6

0 0

6 10 25 27 6

0 0

6 0

29 c

NOTES:

1.

No. indicates number of notches withdrawn out of 48.

Blank is a Withdrawn Rod.

2.

Error Rod is (13, 15).

Figure 11. Limiting Rod Pattern (Alternate)

C 26

4 23A5949 Rev. 0 1 NEUTRON F UX 1 VESSEL PRESS R!SECPSI) 2 AVE SURF A:E HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE IM.ET FLOW 3 RELIEF VA.LVE FLOW 130.0 300.0

_ ti-c'_cu

=v=. e e si 100.0 200.0 O

h 8

0 S 0. 0 tog,e 0

O' 0.0

0. 0 O. 0 5.0 0.0

5. 0 j

TIME (SECONOS)

TIME (SECONDS)

O 1 LEVEL (INC4REF-SEP-SMRT)

! V010 REACT!V!TY N

W E

_..V,...

___ ~.E.._.

i.0

, ~.

200.0 0

100.0 5 00 8

~~ --

~

V

' N_ _,

[* %_-

0.0 l

W

-1.0 V

W

~

-100.0

-2.0

0. 0 5.0 0.0 5.0 TIME (SECONOS)

TIME (SECONO3)

Figure 12. Plant Response to MSIV Closure (Flux Scram - EOC2) 27/28

23A5949 Rsv. O APPENDIX A A00 ANALYSIS INITIAL CONDITIONS To accurately reflect actual plant parameters, the values listed in j

Table A-1 were used instead of the values reported in NEDE-24011-P-A-9-US, l

September 1988. All of the A00s and overpressure protection analyses were I

run considering a power uncertainty of 2%.

The uncertainty in the ODYN' A00 analysis to determine the MCPR operating limit is included in the

(

statistical adjustment factors and the A00 is initiated at 100% rated power.

These initial conditions app".y to analyses performed at standard operating conditions only.

l fJ Table A-1 PLANT PARAMETER Parameter Analysis Value NEDE-24011 Value

(

Pressurization Events (ODYN)

Thermal Power, MWe 3293 3359 6

(I Rated Steamflow, 10 lb/hr 14.11 14.43 4

i Dome Pressure, psig 1006 1010 Turbine Pressure, psig 948 965 No. of S/R Valves

11 15 f

CETAd Analysis Inlet Enthalpy (Btu /lb) 526.2 526.1 l

Non Fuel Power Fraction 0.038 0.04 s

For conservatism, credit was taken for only 11 valves.

The lowest setpoint valves were assumed to be out of service.

4 A 1/2

23A5949 Rov. O APPENDIX B ANALYZED OPERATING DOMAIN The core-wide abnormal operational occurrence (A00) analysis results reported in Section 10 are the most limiting values over the entire allowable operating range. This range covers the following operating options:

i 1.

With all equipment in service, the operating domain includes the 100% power / flow map, MELLLA with 100% power and the flow range from 75% to 100% of rated, and up to 150' feedwater temperature reduction; 2.

With the turbine bypass out of service, the operating domain includes the 100% power / flow map, MELLLA with 100% power and the flow range from 75% to 100% of rated, and up to 150' feedwater temperature reduction; 3.

With the moisture separator reheaters out of service the operating 0

domain includes the 100% power / flow map, KELLLA with 100% power and the flow range from 75% to 100% of rated, and up to 150' feedwater temperature reduction; O

4.

With both the TBOOS and MSROOS the operating domain includes the 100% power / flow map and MELLLA with normal feedwater temperature heating; and G'

5.

Single Loop Operation as allowed by pending Technical Specifications.

Limiting events and conditions analyzed are based on Reference 1 and the UFSAR analytical results.

References 1.

" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-l A 9 US, September 1988.

B 1/2

23A5949 Rsv. O APPENDIX C APPLICATION OF GEMINI METHODS The GEMINI system of methods is used to perform the licensing analyses of Fermi 2 Reload 1.

The GEMINI system of methods is described in Reference 1; NRC approval of these methods is documented in Reference 2.

In Reference 3, the application of GEMINI methods in licensing analyses is described. Pressurization events that could establish the Operating Limit MCPR are analyzed at the 100% power level. Power level uncertainties speci-fied in Regulatory Guide 1.49 are accounted for by adding adjustment factors 1

to the calculated ACPR. NRC approval of this procedure is provided in Reference 4.

The GEMINI system of methods has been incorporated into the approved GESTAR-II licensing topical report, Reference 5 0

Overnressurization Analysis The MSIV Closure (Flux Scram) analysis is performed using GEMINI methods at the 102% power level to account for the power level uncertainties g

specified in Regulatory Guide 1.49.

The analysis was conservatively performed with the 4 lowest setpoint S/RVs out-of-service.

Control Rod Dron Accident g

The NRC approved bounding Control Rod Drop Accident analysis for Banked Position Withdrawal Sequence plants (such as Fermi 2) described in Reference 5 is applied to Fermi 2 Reload 1.

The impact of GEMINI methods on g-the results of the generic analysis is negligible.

Stability i

The NRC approved generic stability approach described in Section 16 is applied to Fermi 2 Reload 1.

The use of GEMINI methods does not impact the generic analysis.

C-1

23A5949 Rsv. 0 References 1.

Letter, J.S. Charnley (GE) to C.O. Thomas (NRC), " Amendment 11 to CE LTR NEDE 24011-P-A," February 27, 1985.

2.

Letter, G.O. Thomas (NRC).to J.S. Charnley (GE), " Acceptance for Ref-erencing of Licensing Topical Report NEDE-24011-P-A, Rev. 6, Amend-ment 11, ' General Electric Standard Application #or Reactor Fuel',"

i November 5, 1985, 3.

Letter, J.S. Charnley (GE) H.N. Berkow (NRC), " Revised Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P A," January 16, 1986.

I 4.

Letter, G.C. Lainas (NRC) to J.S. Charnley (GE), " Acceptance for Ref-

'[

erencing of Licensing Topical Report NEDE-24011-P-A, 'GE Generic Licensing Reload Report,' Supplement to Aniendment 11," March 22, 1986.

5.

"GESTAR-II, General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P A-9-US, September 1988.

)

'3 1

C-2

23A5949 R v. O APPENDIX D GEXL-PLUS CORRELATION The analyses performed for this cycle utilized the GEXL-PLUS thermal margin correlation. The Cycle 1 analyses used the CEXL correlation. The GEXL-PLUS correlation is described in Reference 1.

NRC approval of this correlation is documented in Reference 2.

Reference 3 describes the applica-tion of the GEXL-PLUS correlation which was approved by the NRC in Reference 4.

This application requires an adjustment to the MCPR values for f

bundle flows below 0.4 M1b/sq ft-hr and incorporation of a 3% adjustment

(

factor if inlet subcooling exceeds 70 Btu /lbm. The GEXL-PLLS correlation has been incorporated into Reference 5.

References g

1.

Letter, J.S. Charnley (GE) to C.O. Thomas (NRC), " Amendment 15 to General Electric Licensing Topical Report NEDE-24011-P-A," January 23, 1986.

2.

Letter, Ashok C. Thadani (NRC) to J.S. Charnley (CE), " Acceptance for g

Referencing of Amendment 15 to General Electric Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Application for Reactor Fuel'," March 14, 1988.

3.

Letter, J.S. Charnley (GE) to M.W. Hodges (NRC), " Application of GESTAR-II Amendment 15," March 22, 1988.

g 4.

Letter, Ashok C. Thadani (NRC) to J.S. Charnley (GE), " Acceptance for Referencing of Application of Amendment 15 to General Electric Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Applicatf.on for Reactor Fuel'," May 5, 1988.

t 5.

"GESTAR-II, General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A-9 and NEDE-24011 P-A-9-US, September 1988.

l t

j D-1/2

23A5949 R:v. O APPENDIX E ROD WITHDRAWAL ERROR ANALYSIS Fermi 2 will operate Cycle 2-in an A2 Control' Cell Cora (CCC) loading and operating mode.

In the CCC operating mode, only CCC control rods are utilized. The CCC rods include all A2 control rods, Al shallow rods (inserted less than or equal to notch position 36) and all peripheral rods.

If non CCC rods are used to distort the power sliape, it is possible to obtain a more severe response (ACPR at block) to a postulated RWE. Since the non CCC rods will be fully withdrawn during power operation, this unrealistic assumption results in an undue penalty on the MCPR operating limit for the reload fuel. The initial core fuel will not be limiting in Cycle 2 and the unrealistic analysis will have no practical effect.

For the reload fuel, two analyses were performed providing dual MCPR operating limits; one for use L

when non CCC rods are inserted, and the other for use when the only control rods inserted will be limited to CCC rods only.

In this way, protection against violation of the fuel cladding integrity safety limit is provided without unnecessarily limiting plant operation by applying the more i

conservative MCPR operating limit with non CCC operation mode and applying the realistic limit for the CCC operation mode.*

O t

[

Normal control rod operability checks and testing of non-CCC control rods does not require the utilization of the more restrictive non-CCC operational mode MCPR limits.

For control rod cooling, B rods may be inserted to position 46 for short periods of time without requiring use of the more t-restrictive MCPR operating limit.

f E-1/2 n

(Final)

- - _. _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _

NRC-89-0052, Supplemental Reload Licensing Submittal for Fermi Plant Unit 2,Reload 1,Cycle 2

Text

SUMMARY

SUMMARY

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