ML20082N592

Cycle 18 Core Performance Analysis Rept

ML20082N592
Person / Time
Site:	Vermont Yankee
Issue date:	01/20/1995
From:	Sironen M YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20082N570	List: ML20082N567 ML20082N580 ML20082N592
References
YAEC-1908, NUDOCS 9504250302
Download: ML20082N592 (100)

Text

wJ wi 45'.,.

1 +,

m'

.e. m., ,,

m'--

~ r

~'M...

G:' r- 2 ' ,

m .,

em ' '"

< ,., M b.9: -

,- ' .j k s #

4' 4 , , 7

' s' ' '

,  : ,A . _ ,  ;,~  ? s

,w >

f y

r. r .3 +h rg  ;,.r

. ' w j C, O fR ..g jf' * , ' s M -> >

f,r l.)

i

• , , > y y:q

}- 7 h ,, ;eg , , .g ., ,

5

-... z y ..,r i e s 4 c a-C+E"; MAN. KEEin, TOMICVELECTRICiCOMPANY w

>4

's :: -

r 1.: , ,

4 v- .

~

pt in 9. 9

')

' ,j.; '. .y}. ' f ' 7, c . ..

39 . .y

.A " -

w' , 3 .,

, c:q fr .,  ?

~~f

,I *v s r'. s , L -- j r'. > *

. .,i, ;

-u s g. 3 - ' >

s' n,'

1 sy> g

,; - w ' -

, ,31 rmm.

1 -

m. . ..

y,,g _

y

. ~ ~.

.-..-n ,:: ,u. n- ;s,

6

^ * + ' ij 1 4 &" ,

G '.- , . g-Q

. , , . ,, ai , c ,,, i 4

e.#a'";m ' ' -, -,, .; -

L.k?, , .

.9 1

.%? r + -

i y

• t

+ 40 -

g{l

[i ;).. ',( i 4

, .. < m ,u ,, ~.

.- '{ ; ' e ,, j $ , '_J_, jg  :/

e ,.A .s n r >;

7 . /v a

<3.

g'g' 3 ,-sp '_ \ '4. '

q 4

}

n,  ;

^hl f.k

. g. g.,.,

s ,u s ' u , 4 .4 , ..}

W,',[ '

y;. .

.W* .d

,a ^ + , ,

... ,i :

., . , '..., M' :3

[ )f } #

.( 1 j , i.

S t

y'! L.

g \

,t  ?,.I'?'

n s Q;( ,

N:

+q,..-

. ., e { w' r-z, ,

._ -]

3 ,. ,

. : 4 ,~

Q~p., '

'l ' h' - >

g ly

~

m.

-. 4

, .;>)

,i;-Q < 2 t' ,

Q, . . - ;4_-  : V '.

<y

_ iyy l,-

I .-...

i ^; 'f'E_'f .;. ;

• -; J',V-i ' - .s ,g 4.,.f'..,,t, ,

5,' 1 g . 'Y,

( i,,'* ,,

t ,

i 4'(> 3. , < i.,

n/ r

, L.

s s

xQ  :

f l, , . a i .t

~. ^Mp& :

+ O -- .

e

' . . . H) mya c

,J e

. u. .s, ,. -

• 4

' ' T

-4 <-

1 n '_, :[t s 1

..)

/m W

> s s '.

?' c .z + .g; , 4 4.j ; x3 av', ' 3 . ,~

g" 4

-. . , ' YA %IK. ,EE /;'

4

;c 1

x , . , ,

n f' ,.

a .

n.py.

.\ , v.

< , L .:

z , n.

g 4

. a t' 4 p- w, -: g} -'-

)l' * .{ .

t i p

. i -

M r, .,  :. 8

}l c

l-

r ,

4 1

+  : e- , , <

t;

.v 1

i l i 7

'b'9fb y h-

.H_- , j;Q a:

'g.

. . . t s ep

",I'.

.' I

.g i , , .

[- 4 r d'

J t m e u x- 9  ?

~ l'r-

?

.- em.

i . .

< 4 ' [' 'l

.. i I

. r

" ^ !f

/

3

- d--

d I

?

'f_'

'A' .f L 9504250302 950419 1

?

PDR ADOCK 0500 P

..1-

I Ia-O I%> t I

Vermont Yankee Cycle 18 Core Performance Analysis Report i

I  :

I I

January 1995 Major Contributors: C. Otiu N. Fujita I B. Hubbard D. Kapitz M. LeFrancois i

D. Morin i K. Morrissey l J. Neyman L. Schor I F. Seifaec R. Smith l K. St. John l R. Wochlke I

I R4N3 I

I '

1 OI oj Prepared by: J d ((

M. A. SIronen (Date)

VY Nuclear Engineering Coordinator Appmved by: 4 d 7 / b /'7 C R.

R6[ actor Physics GroupCacciapouti[p(snager # (Date) ll E

Appmved by: - IID~# ,

P. A. Bergeron, ager (Date)

Transient Analysi Group 1 Approved by:

M .~ qu/es g R. K. Sundaram, M. nager (Date) g LOCA Analysis Gmup Appmved by: . "N# # ff J. R. Chapman irector '(Date)'

Nuclear Engin ring Depanment I

I I

I, II Yankee Atomic Electric Company I

Nuclear Services Division 580 Main Street 3

E Bolton, Massachusetts 01740 R4h53 I

i

I, DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company (" Yankee"). The use of infonnation contained in this document by anyone other than Yankee, or the Organization for which this document was prepared under contract, is not authorized and, with n spect to any unauthorized I ujg, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any wananty or representation as to the accuracy or completeness of the material contained in this document i

I I

ll i i

I E

i i

I I i RAM 3

,jjg, I

I

O O

ABSTRACT This report presents design information, calculational results, and operating limits pertinent to the operation of Cycle 18 of the Vennont Yankee Nuclear Power Station. These include the fuel I design and core loading pattem descriptions; calculated reactor power distributions, exposure distributions, shutdown capability, and reactivity data; and the results of safety analyses performed to justify plant operation throughout the cycle.

I, I

I l

1 I

I i

i I

l I

I a=m -iv-I E

=

1 L_____________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________________ - _

E ACKNOWLEDGEMENTS I De author and major contributors would like to acknowledge the contributions to this work by the Vermont Yankee Reactor & Computer Engineering Depanment for their review of input data and I guidance.

lt I

E 1

I lI 1

I I

l I .

'I I

l

n a

O, 1 1

TABLE OF CONTENTS Pace l DISCLAIMER OF RESPONSIBILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111 i

A B STRA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i v ACKNOWLEDGEMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v TABLE OF CONTENTS ..............................................vi LIST OF TAB LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii LIST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x l 1.0 I NTRO D UCTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 2.0 RECENT REACTOR OPERATING HISTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 g 2.1 Operating History of the Cununt Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 {1 2.2 Operating History of Past Applicable Cycle . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.0 RELOAD CORE DESIGN DESCRII* TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1 Core Fuel Loading ............................................. 5 3.2 Design Reference Cort Loading Pattem . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.3 Assembly Exposure Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4.0 FUEL MECHANICAL AND THERMAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1 Mechanical Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.2 'Ihennal Design ............................................... 9 4.3 Ope rating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.0 N UCLEAR DES IG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.1 Core Power Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.1.1 Haling Power Distribution . . . . . . . . . . . . . . . .................. 15 5.1.2 Rodded Depletion Power Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . 15 5.2 Core Exposure Distributions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.3 Cold Shutdown Margin . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 5.4 Maximurn K,,, for the Spent Fuel Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Re33 -V5-i

=

I ,

I TABLE OF CONTENTS (Continued)

I .P, age i

6.0 TIIERMAL-IIYDRAULIC DESIGN .... ................................ 26 6.1 Steady-State Thermal Ilydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.2 Reactor Limits Determination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 7.0 ABNORMAL OPERATIONAL TRANSIENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . 28 7.1 Transients Analyzed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 7.2 Pressurization Transients Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.2.1 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.2.2 Initial Conditions and Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 7.2.3 One. Dimensional Cmss Sections and Kinetics Parameters 31 I

7.2.4 Turbine Trip Without Bypass Transient (TTWOBP) . . . . . . . . . . . . . . . . 33 7.2.5 Generator lead Rejection Without Bypass Transient (GLRWOBP) . . . . . 33 7.2.6 Pressurization Tmnsient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . 33 7.3 less of Feedwater IIeating Transient (LOFWII) Results . . . . . . . . . . . . . . . . . . 34 7.4 Overpressurization Analysis Results ................................ 35

< 7.5 Local Rod Withdrawal Error Transient Results . . . . . . . . . . . . . . . . . . . . . . . . . 35 7.6 Misloaded Bundle Ermr Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 j 7.6.1 Rotated B undle Ermr . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 7.6.2 Mislocated Bundle Ermr . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 7.7 Transient Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 8.0 DESIGN B ASIS ACCIDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 8.1 Control Rod Drop Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 8.2 Loss-of-Coolant Accident Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 8.3 Refueling Accident Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 I 9.0 STARTUP PROG R A M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 10.0 CO NCLUSIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 r APPENDIX A . . . . . ..........................................,,,,, g7 I i wn -vii.

I

_ s

rj O!

LIST OF TABLFJ Numtel Title g I 2.1.1 VY CYCLE 17 OPERATING IIIGHLIGHTS .......................... 3 Il 2.2.1 VY CYCLE 16 OPERATING HIGHLIGHTS .......................... 4 3.1.1 ASSUMED VY CYCLE 18 FUEL BUNDLE TYPES AND NUMBERS . . . . . . . . 7 3.3.1 DESIGN BASIS VY CYCLE 17 AND CYCLE 18 EXPOSURES . . . . . . . . . . . . 7 4.1.1 NOMINAL FUEL MECHANICAL DESIGN PARAMETERS . . . . . . . . . . . . . . I1 4.2.1 VY CYCLE 18 CORE AVERAGE GAP CONDUCTANCE VALUES . . . . . . . . 12 4.2.2 VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES FOR HALING AXI AL POWER DISTRIB UTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 13 4.2.3 VY CYCLE 18 HOT CHANNEL GAP CONDUCTANCE VALUES FOR 1.4 CHOPPED COSINE AXIAL POWER DISTRIBUTION . . . . . . . . . . . . . . . . . . 14 5.3.1 VY CYCLE 18 Kpfp VALUES AND SHUTDOWN MARGIN CALCULATION 18 5.4.1 VY CYCLE 18 MAXIMUM COLD K., OF ANY ENRICHED SEGMENT . . . . I8 7.2.1 VY CYCLE 18

SUMMARY

OF SYSTEM TRANSIENT MODEL INITIAL P CONDITIONS FOR TRANSIENT ANALYSES . . . . . . . . . . . . . . . . . . . . . . . . 41 7.2.2 VY CYCLE 18 PRESSURIZATION TRANSIENT ANALYSIS RESULTS . . . . . 42 7.3.1 VY CYCLE 18 LOSS OF FEEDWATER HEATING TRANSIENT ANALYSIS l

RES ULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 5 7.4.1 VY CYCLE 18 OVERPRESSURIZATION ANALYSIS RESULTS . . . . . . . . . . 44 7.5.1 VY CYCLE 18 ROD WITHDRAWAL ERROR ANALYSIS RESULTS . . . . . . . 44 7.6.1 VY CYCLE 18 ROTATED BUNDLE ANALYSIS RESULTS . . . . . . . . . . . . . . 45 7.6.2 VY CYCLE 18 MISLOCATED BUNDLE ANALYSIS RESULTS . . . . . . .... 45 7.7.1 VY CYCLE 18 LIMITING TRANSIENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 8.1.1 CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER . . . . . . . 77 8.1.2 VY CYCLE 18 CONTROL ROD DROP ANALYSIS RESULTS . . . . . . . . . . . . 77 um -viii-I a

a ;

s

LIST OF TABLES r (Continued)

L f Number Title Paze L

8.2.1 LOCA ANALYSIS ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 I

A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 18 MCPR L OPERATING LIMITS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . 88 A.2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR L B P8 DWB 31 1 - 10GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 A.3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR B P8 DWB 31 1 - I l GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 A.4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR

- B P8 DWB 3 35- 10GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91

- A.5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR L B P8 DWB 3 3 5- i l GZ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 r

U n

l ,

L w

r b

[

[ ,_, .ix.

E L --- -

p b

d LIST OF FIGURES Number Title Page 3.2.1 VY CYCLE 18 DESIGN REFERENCE LOADINO PA'ITERN, LOWER RIGHT Q UA D RANT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5.1.1 VY CYCLE 18 HALING DEPLETION, EOFPL BUNDLE AVERAGE RELATIVE POWERS................................................... 19 5.1.2 VY CYCLE 18 HALING DEPLETION, EOFPL CORE AVERAGE AXIAL POWER DI STR IB UTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.1.3 VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL, BUNDLE AVERAGE RELATI VE PO WERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1.4 VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL, CORE AVERAGE AXIAL POWER DISTRIB UTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5.2.1 VY CYCLE 18 IIALING DEPLETION, EOFPL BUNDLE AVERAGE E XPO S U RES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.2.2 VY CYCLE 18 RODDED DEPLETION, EOFPL BUNDI.E AVERAGE I EXPO S URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 5.3.1 VY CYCLE 18 COLD SIIUTDOWN MARGIN, IN %AK, VERSUS CYCLE EXPO S URE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 7.2.1 FLOW CHART FOR THE CALCULATION OF ACPR USING THE RETRAN/TCPYA01 CO DES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 7.2.2 TURBINE TRIP WITHOUT BYPASS, EOFPL18 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 7.2.3 TURBINE TRIP WITIIOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 51 7.2.4 TURBINE TRIP WITIIOUT BYPASS, EOFPL18 2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . . . . . . . . . . . 54 '

7.2.5 GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18 TRANSIENT RESPONSE VERSUS TIME, MEASURED" SCRAM TIME . . . . . . . . . . . . . . 57 1

7.2.6 GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . 60 f

)

! aum -x-I ,

l

I .

LIST OF FIGURES (Continued)

I Number Title Page 7.2.7 GENERATOR LOAD REIECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST 1 TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME . . . . 63 I 7.3.1 LOSS OF 100*F FEEDWATER HEATING (LIMITING CASE) TRANSENT RESPONSE VERS US TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 7.4.1 MSIV CLOSURE, FLUX SCRAM, EOFPL18 TRANSIENT RESPONSE VERSUS -

1 TIME, "hEASURED" SCRAM TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 7.5.1 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEPTS DETERMINED BY TIE lI A AND C CH ANNELS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.5.2 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEI'rS DETERMINED BY TIE B AND D CHANNELS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 8.1.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEQUENCES . . . . . . . . . . 79 8.1.2 FIRST FOUR ROD ARRAYS PULLED IN TiiE B SEQUENCES . . . . . . . . . . . 80 8.2.1 LOCA ANALYSIS RESULTS, PEAK CLADDING TEMPERATURE VERSUS 8 B REA K S I ZE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 I

,1 I

!I I .

I nau -xi-f Illu

O O

1.0 INTRODUCTION

%w This repon pmvides infonnation to suppon the operation of the Vermont Yankee Nuclear Power Station through the fonhcoming Cycle 18. In this repon, Cycle 18 will be referred to as the Reload Cycle. The preceding Cycle 17 will be referred to as the Current Cycle. The Cycle 17/18 refueling will involve the discharge of 120 irradiated fuel bundles and the insenion of 120 new fuel bundles. The resultant core will consist of 120 new fuel bundles and 248 irradiated fuel bundles. The General Electric Company (GE) manufactured all the bundles. Some of the irradiated fuel was also present in the reactnr in Cycle 16. This cycle will be referTed to as the Past Cycle.

This repon contains descriptions and analyses results pertaining to the mechanical, thermal-hydraulic, physics, and safety aspects of the Reload Cycle. The MAPLHGR and MCPR operating hmits calculated for the Reload Cycle are given in Appendix A. These limits will be included in the Core Operating Limits Repon.

I I

I I

I I

I I

num I I  :

2.0 RECENT REACTOT., j.PER ATING HISTORY 2.1 Operating History of the Current Cycle The current operating cycle is Cycle 17. To date, the Current Cycle has been operating at, or near, full power with the exception of sequence exchanges, several power reductions, and four short repair outages. 'Ihe operating history highlights and control rod sequence exchange schedule of the Current Cycle are found in Table 2.1.1.

2.2 ,0,,,g,:pfne History of Past Applicable Cycle The irradiated fuel in the Reload Cycle includes some fuel bundles initially inserted in Cycle

16. This Past Cycle operated at, or near, full power with the exception of sequence exchanges, several short power reductions, one short repair outage and a coastdown to the end of cycle. The operating history highlights of the Past Cycle are found in Table 2.2.1. The Past Cycle is described in detail in the Cycle 16 Summary Report [1].

num .. .

O O

TABLE 2.1.1 VY CYCLE 17 OPERATING HIGHLIGHTS Beginning of Cycle Date October 24,1993 End of Cycle Date March 18,1995*

Weight of Uranium As-Loaded (Short Tons) 72.02 Beginning of Cycle Core Average Exposure ** (mwd /St) 11547 End of Full Power Core Average Exposure ** (mwd /St) 21997*

End of Cycle Core Average Exposure ** (mwd /St) 21997*

Number of Fresh Assembl.*es 128 Number of Irradiated Assemblies 240 Control Rod Sequence Exchange Schedule:

t Sequence Date From To January 9.1994 A2-1 B2-1 March 15,19M B2-1 Al-1 May 17.1994 Al-1 B1-1 July 19,1994 B1-1 A2-2 October 6,1994 A2-2 B2-2 December 2,1994 B2-2 Al-2  :

January 24,1995* Al-2 BI-2 I,

I e,oiec,ed eates d ex,es.,es.

I ew-s basme ein emss Cape -eg. l mm3 I a'

TABLE 2.2.1 VY CYCLE 16 OPERATING HIGHLIGHTS

. . Beginning of Cycle Date April 19,1992 End of Cycle Date August 28,1993 Weight of Uranium As-Loaded (Short Tons) 72.06

(; Beginning of Cycle Core Average Exposure * (mwd /St) 11417 End of Full Power Core Average Exposure * (mwd /St) 21103 End of Cycle Core Average Exposure * (mwd /St) 21878 Number of Fresh Assemblics 128 Number ofIrmdiated Assemblies 240 l

. Contml Rod Sequence Exchange Schedule:

Sequence Date - From To  !

June 14,1992 A2-1 B2-1 August 9,1992 B2-1 Al-1 October 15,1992 Al-1 B1-1 December 7,1992 B1-1 A2-2 February 9,1993 A2-2 B2-2 April 6,1993 B2-2 Al-2

( June 6,1993 Al-2 B1-2 L,

Exposures based on the Plant Process Computer accounting.

R*h53 -4*

I i-

O o

3.0 REl.OAD CORE DESIGN DESCRIPTION l

3.1 Core Fuel Loading i

The Reload Cycle core will consist of both new and irradiated assemblies. All tle assemblies have bypass flow holes drilled in the lower tie plate. Table 3.1.1 characterizes the core by fuel type, batch size, and first cycle loaded. A description of the fuel is found in tar GE Standard Application for Reactor Fuel [2] and the GE Fuel Bundle Design Reports [3][4].

3.2 Design Reference Core leading Pattern I:

The Reload Cycle assembly locations are indicated on the map in Figure 3.2.1. For the sake of legibility only the lower right quadrant is shown. The other quadrants are mirror images with I

W bundles of the same type having nearly identical exposures. The bundles are identifled by the reload number in which they were first introduced into the core. Table 3.1.1 p Uvides the key, called bundle ID, which identifies what explicit fuel type is found in each bundle location.

If any changes are made to the loading pattem at the time of refueling, they will be evaluated under 10CFR50.59. The final loading pattem with specific fuel bundle serial numbers will be supplied in the Startup Test Report. '

3.3 Assembly Exposure Distribution The assumed nominal exposure on the fuel bundles in the Reload Cycle design reference loading pattem is given in Figure 3.2.1. To obtain this exposure distribution, the Past Cycle was depleted with the SIMULATE-3 model[5],[6] using actual plant operating history. For the CurTent Cycle, plant operating history was used through April 22,1994. Beyond this date, the exposure was accumulated using a best-estimate rodded depletion analysis to End of Cycle (EOC).

I Il num Il !

Il

=

b' Table 3.3.1 gives the assumed nominal exposure on the Current Cycle and the Begenmg of Cycle (BOC) core average exposure that results from the shuffle into the Reload Cycle loading pattem.

1he Reload Cycle End of Full Power Life (EOFPL) core average exposure and cycle capability are provided. ,

i 1

g n

y c

c L

b E

ne.n -

O O

TABLE 3.1.1 ASSUMED VY CYCLE 18 FUEL BUNDLE TYPES AND NUMBERS Fuel Type Reload Cycle Number Designation Bundle ID Imaded of Bundles Irndiated BP8DWB311-10GZ RISA 16 40 BP8DWB311-IlGZ RISB 16 80 BP8DWB335-10GZ R16A 17 96 BP8DWB335-1IGZ R16B 17 32 I

New BP8DWB335-10GZ R17A 18 88 BP8DWB335-IIGZ R17B 18 32 i

TABLE 3.3.1 DESIGN BASIS VY CYCLE 17 AND CYCLE 18 EXPOSURES

• Assumed End of Current Cycle Core Average Exposure with an 21.921.6 GWd/St Exposure Window ofi 600 mwd /St[7]

Assumed Beginning of Reload Cycle Core Average Exposure 12.13 GWd/St Italing Calculated End of Full Power Life Reload Cycle Core Average 22.16 GWd/St Exposure Reload Cycle Full Power Exposure Capability Olaling) 10.035 GWd/St I

expes.,es sa,ed en se SimUtArE.3 acceuntin,. I. 1 aum l I

E-

[

R168 R17A R198 R14A R16A R17A R16A R178 R16A RinA R188 30 866 0 000 21.400 11381 21.794 0.000 21423 0 000 23.110 13 432 24 238 k R17A RieA R17A R14A R17A RieA R17A R178 RieA R188 RIGA 30 0 000 11248 OA00 11.738 0.000 12.475 0 000 0 000 12.730 13002 24.000 R168 R17A R198 R17A R188 R17A R188 R178 RieA R198 R168 21286 0 000 21.441 0.000 21.001 0.000 22.367 0.000 13061 13 383 26 048 RisA R14A R17A R14A R17A RieA R17A R17A R14A R188 11 4s7 11.720 0 000 11280 0 000 12.701 0 000 0.000 13 198 24104 R168 R17A R158 R17A R158 R17A R168 RIM R198 14 21.001 0 000 21402 0 000 21.o63 CD00 23D31 13170 13042 R17A R14A R17A R16A R17A R1eA R178 Rie8 R15A 0(c0 12410 0 000 12.010 0.000 13 263 0 000 13 038 24 124 Rt&A R17A R158 R17A R16A R178 R14A R188 R158 0

21.230 0 000 22 443 0 000 22.038 0.000 13 424 13 67e 26 130 R178 R178 R178 R17A R14A R188 RIGA R168 na 0 000 0 000 0.000 0 000 13.167 13.573 13.737 26 000 R16A R16A R1eA R14A R14A R16A R198 na 23 304 12000 13366 13213 13 set 23.387 26.30s L R16A R188 Rie8 R188 13 446 13 077 13292 24.136 H

R1u R168 R168 _ m=E o n.

24 442 24 Geo 26478 . .. SOC EXPCGURE (GWDST) 23 26 27 20 31 33 36 37 30 41 43 1

)

FIGURE 3.2.1 VY CYCLE 18 DESIGN REFERENCE LOADING PATTERN. LOWER RIGHT OUADRANT i'

mass i J

___ ___ ____ _o

in l

[] !

Di I

4.0 FUEL MECHANICAL AND THERMAL DESIGN 4.1 Mechanical Design All of the fuel to be inserted into the Reload Cycle was fabricated by GE. De major mechanical design parameters am given in Table 4.1.1 and Reference 2. Detailed descriptions of the fuel md mechanical design and mechanical design analyses are provided in Reference 2. Dese design analyses remain valid with respect to the Reload Cycle operation. Mechanical and chemical compatibility of the fuel bundles with the in-service reactor environment is also addressed in Reference 2.

4.2 Thermal Design The fuel thermal effects calculations were performed using the FROSSTEY-2 computer code [8],[9],[10]. The FROSSTEY-2 code calculates pellet-to-cladding gap conductance and g

fuel temperatures fmm a combination of theoretical and empirical models including but not limited to e fuel and cladding thermal expansion, fission gas release, pellet swelling, pellet densification, pellet cracking, and fuel and cladding thermal conductivity, ne thermal effects analysis included the calculation of fuel temperatures and pellet-to-cladding gap conductance under core average and hot channel conditions. De core average calculations integrate the responses of individual fuel batch average operating histories over the core average exposure range of the Reload Cycle. Dese gap conductance values are weighted axially into 12 axial nodes by power d*stributions and radially by volume. The core-wide gap conductance values for the RETRAN system simulations, described in Sections 7.1 and 7.2, are from this data set at the corresponding exposure statepoints. Table 4.2.1 provides the core average response of gap conductance.

ne hot channel gap conductance values, which are input to the hot channel transient I  ;

calculations (Section 7.1), were evaluated for the limiting fuel bundle type as a function of the assembly exposure for two axial power shapes, a 1.4 chopped cosine and the Reload Cycle's Haling.

The hot channel calculations assumed the following as required by the NRC Safety Evaluation for nam  ;

l i

l FROSSTEY-2[ll): 1) appropriate allowances to account for manufacturing uncertainties and 2) the worst axial power shape prior to the transient. The peak power node was placed at the maximum average planar linear heat generation rate (MAPLIIGR) limits. Gap conductance values for the hot channel analysis were determined using the limiting bundle exposure. The limiting bundle is defined as the bundle with the lowest MCPR or the highest power, if different, within the exposure range of

[ interest. The limiting exposure for the bundle is defined by the exposure which produces the highest bundle average gap conductance within the interval of interest. The SIMULATE-3 rodded depletion (Section 5.1.2) provided predictions of the limiting bundle exposure for each exposure interval. Table 4.2.2 provides the hot channel gap conductance values for the two axial power shapes. Results are presented for the bounding exposure for the chopped cosine shape and at the four exposure statepoints for the IIaling shape.

4.3 Operating Experience All irradiated fuel bundles scheduled to be reinserted in the Reload Cycle have operated as expected in past cycles of Vermont Yankee. Off-gas measurements in the Current Cycle indicate no .

fuel rod failure. .

l 4

l 1

i L

Ruh53 -

O O,

TABLE 4.1.1 NOMINAL FUEL MECITANICAL DESIGN PARAMETERS I

Fuel Bundle

• 1rradiated Fuel Type New & Irradiated Fuel Tvoes Bundle Types GE8X8NB GE8X8NB I

Vendor Designation BP8DWB311-10GZ & BP8DWB335-10GZ &

BP8DWB311-1IGZ BP8DWB335-1IGZ E

W Initial Enrichment.w/o U 235 3.11 3.35 Rod Array 8X8 8X8 Fuel Rods per Bundle 60 60 I

Outer Fuel Channel Material Zr-2 Zr-2 Wall Thickness, inches 0.080 0.080 I

I cemplete sendie. md, and peliet descriptions are round in aererences 2 through 4.

l aum I a.

TABLE 4.2.1 VY CYCLE 18 CORE AVERAGE GAP CONDUCTANCE VALUES I Gap Conductance (BTU /hr-ft2 , p)

Axial BOC EOFPL-2000 EOFPL-1000 EOFPL Node mwd /St mwd /St 1 1190 1830 1960 2115 2 m5 - m5 m0 g

3 2445 3810 3875 4140 4 2455 3820 3895 4185 5 2495 3860 3955 4325 6 2610 3945 4150 4560 7 2600 3940 4140 4555 8 2625 3960 4180 4565 9 2525 3880 4005 4455 10 2420 3760 3850 4080 I "

12 675 1015 1120 1225 I

I I

I I

I

_ 12 I

[]

O' TABLE 4.2.2 VY CYCLE 18 IIOT CIIANNEL GAP CONDUCTANCE VALUES

• FOR HALING AXIAL POWER DISTRIBUTION Gap Conductance (BTU /hr-ft2 op)

Axial BOC" EOFPL-2000 EOFPL-1000 EOFPL" Node mwd /St" mwd /St" 10.95 GWd/St " 10.016 GWd/St"* 10.95 GWd/St"* 11.472 GWd/St"*

I 3360 3040 3360 3590 2 7850 7520 7850 8330 3 9630 9530 9630 9500 4 9630 9780 9630 9500 5 9630 9880 9630 9500 6 9650 9890 9650 9500 7 9650 9890 9650 9500 8 9650 9890 9650 9500 9 9650 9890 9650 9500 ,

10 9450 8150 9450 9500 11 6580 6130 6580 6930 12 1530 1460 1530 1570 I

I I.

The hot channel gap conductance values am derived for tim BP8DWB335 fuel type because it is conservadve compared to the other fuel types.

" Core Average Exposure.

"* Peak Bundle Exposure.

men I

{ .

l l

TABLE 4.2.3

^

' VY CYCLE 18 HOT CHANNEL GAP CONDUCTANG VALUES

• FOR 1.4 CHOPPED COSINE AXIAL POWER DISTRIBUTION )

{ '.

Gap Conductance (BTU /hr-ft2 ,.p)

Axial EQC" EOFPL-2000 EOFPL-1000 EDEEL" ligdg mwd /St

• mwd /St 12.15 GWd/St 10.008 GWd/St 11.472 GWd/St*** 12.15 GWd/St 1 750 790 760 750 2 1510 1410 1480 1510 3 5040 3690 4680 5040 4 7920 8000 7500 7920 f' 5 9730 10450 9970 7930 6 9780 10450 10020 9780

)

7 9810 10450 10020 9810 8 9810 10450 10020 9810 j 9 9810 8570 -9150 9810 7230 f 10 6210 7070 7230 11 2640 2260 2520 2640 12 950 960 950 950 De hot channel gap conductance values are derived for the BP8DWB335 fuel type because it is consesvative compared to the other fuel types.

Core Average Exposure.

• • • Peak Bundle Exposure, a=m - - - - _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ .

i O

Ol l

5.0 NUCLEAR DESIGN 5.1 Core Power Distributions ne Reload Cycle was depleted using SIMULATE-3 to give both a mdded depletion and an All Rods Out (ARO) IIaling depletion.

5.1.1 Ifaling Power Distribution De IIaling depletion serves as the basis for defining core reactivity characteristics for most transient evaluations. His is primarily because its flat power shape has conservatively weak scram g

characteristics. Sensitivity studies have shown that the limiting pressurization transient results are 5 more conservative when calculated using the IIaling power distribution as the initial power shape.

The IIaling power distribution is calculated in the ARO condition. He llating iteration I

converges on a self-consistent power and exposure distribution for the bumup step to EOFPL in principle, this should provide the overall minimum peaking power shape for the cycle. During the actual cycle.11atter power distributions might occasionally be achieved by shaping with control rods.

Ilowever, such shaping would leave underbumed regions in the core which would peak at another point in time. Figures 5.1.1 and 5.1.2 give the IIaling radial and axial average power distributions for the Reload Cycle.

5.1.2 Rodded Depletion Power Distribution I The rodded depletion was used to evaluate the misloaded bundle error and the rod withdrawal error because it provides the initial rod pattems and more accurately defines the local characteristics prior to the transient evaluations. It was also used in the rod drop worth and shutdown margin calculations because it depletes the top of the core more realistically than the llaling depletion. He rodded depletion also pmvides the hot channel bundle exposures for the gap conductance calculation.

To generate the rodded depletion, control rod pattems were developed which give critical eigenvalues at several points in the cycle and peaking similar to the IIaling calculation. The resulting ,

I aos Il El

(

patterns were frequendy more peaked than the Haling, but were below expected operating limits.

However, as stated above, the underbumed regions of the core can exhibit pealdng in excess of the

~

Haling peaking when pulling ARO at EOFPL. Figures 5.1.3 and 5.1.4 give the ARO radial and axial average power distributions for the Reload Cycle rodded depletion at EOFPL.

5.2 Core Exoosure Distributions The Reload Cycle exposures are summarized in Table 3.3.1. The pmjected BOC radid

- exposure distribution for the Reload Cycle is given in Figure 3.2.1. 'Ihe Haling calculation pmduced the EOFPL radial exposure distribution given in Figure 5.2.1. Since the Haling power shape is constant, it can be held fixed by SIMULATE-3 to give the exposure distributions at various mid-cycle points. BOC, EOFPL-2000 mwd /St, EOFPL-1000 mwd /St and EOFPL exposure distributions were used to develop reactivity input for the core wide transient analyses.

'Ihe rodded depletion differs from the Haling during the cycle because the rods shape the power differently. However, rod sequences are swapped frequently and the overall exposure distribution at end of cycle is similar to the Haling. Figure 5.2.2 gives the EOFPL radial exposure ,

distribution for the Reload Cycle rodded depletion.

j 5.3 Cold Shutdown Marrin Technical Specifications [12] state that, for sufficient shutdown margin (SDM), the core must be subcritical by at least 0.25% AK + R (defined below) with the strongest worth control rod withdrawn. Using SIMULATE-3, a search was made for the strongest worth control rod at various exposures in the cycle. This is necessary because rod worths change with exposure on adjacent assemblies. Then the cold K,g with the strongest rod out was calculated at BOC and at the end of each contml md sequence. Subtracting each cold K,g with the strongest rod out from the cold critical K,g defines the SDM as a function of exposure. Figure 5.3.1 shows the results.

The cold critical K,g was defined as the average calculated critical K,g minus a 95%

confidence level uncertainty. Then all cold results were normalized to make the critical K,g equal to 1.000.

nm #

1

E us Because the local reactivity may increase with exposure, the SDM may decrease. To account for this and other uacertainties, the value R is calculated. R is defined as3R plus R 2 . R3 is the difference between the cold K,g with the strongest rod out at BOC and the maximum cold K,g with the strongest rod out in the cycle. R 2is a measurement uncertainty in the demonstration of SDM associated with the manufacture of past control blades. It is presently set at 0.07%

AK[13],[14]. The shutdown margin results, summarized in Table 5.3.1, show that the shutdown margin for the Reload Cycle is greater than the Technical Specification limit of 0.32% AK.

5.4 Maximum K_ for the Spent Fuel Pool Section 5.5E of the Technical Specifications requires that the K,, for any bundle stored in ,

either the new fuel vault or the spent fuel pool not exceed 1.31 to ensure compliance with the K,g safety limit of 0.95. The bundles used in the Reload Cycle do not exceed the specifications in Section 5.5E, as shown in Table 5.4.1. These values are obtained from CASMO-3G[15].

I I

I I

I I'

Il I!

RKh33 -17 I

!l

I L

g TABLE 5.3.1 L

VV CYCLE 18 K g VALUES AND SHUTDOWN MARGIN CALCULATION

[

Cold Critical K,g 1.0000 L

BOC K,g - Contmiled With Strongest Wonh Rod Withdrawn 0.9872 f

L Cycle Minimum Shutdown Margin Occurs at BOC With I Stmngest Wonh Rod Withdrawn 1.28% AK R t. Maximum Increasc in Cold K,g With Exposure 0.00% AK N

L TABLE 5.4.1 VY CYCLE 18 MAXIMUM COLD K_ OF ANY ENRICHED SEGMENT Bundle Type Maximum K_

BP8DWB311-10GZ 1.20 BP8DWB311-ilGZ l.20

( BP8DWB335-10GZ l.22 BP8DWB335-11GZ l.22 e

f L

mes3 <

O O

I I

R168 R17A R168 R16A R16A R17A R16A R178 R16A R16A R168 1 076 1.393 1 A91 1 218 1Dl6 1364 1 042 1243 0 638 0.720 0J06 -

R17A RISA R17A R18A R17A R16A R17A R178 RIGA R168 R16A 1J93 1266 1410 1261 1.392 1220 1.340 1248 0364 0.722 0.378 R168 R17A R168 R17A R168 R17A R168 R178 RISA R168 R168 1893 1.410 1.121 1.420 1.100 1.376 1 030 1229 0 913 0 083 0.342 R16A R16A R17A RIGA R17A R16A R17A R17A R16A R168 1216 1261 1 420 1271 1.394 1.194 1200 1.162 0827 0 483 R168 R17A R168 R17A R168 R17A R168 R16A R188 lbs4 1.391 1.107 1.393 1.070 1.304 0 936 0 918 0 848 R17A RISA R17A RIGA R17A RIGA R178 R188 R16A 1363 1219 1 374 1.192 1.304 1De6 ***6 0 791 0 470 R16A R17A R168 R17A R16A R17B R16A R168 R168 1 044 1.330 1 037 1 279 0 938 1.115 0 663 0 468 0.366 R178 R178 R178 R17A R16A R188 916A R168 g na 1242 12de 1 228 1.161 0 918 0792 0657 0.364 R16A R16A R16A R16A R16A R16A R168 nn 0 $30 0 963 0 011 0 827 0 See 0 471 0.364 R16A R168 R188 R168 0.726 0 720 0 862 0 467 R16A R168 R168 BUNDLE D ns 0 300 0 376 0.342 .. EOFPL RELATIVE POWER D M 27 3 31 33 2 37 2 41 4 I

I FIGURE 5.1.1 VY CYCLE 18 HALING DEPLETION.

EOFPL BUNDLE AVERAGE RELATIVE POWERS Rm3 I a

= ,

F M '

N  %

24

N L as  %

b

,0 (I 1e 10 17 l 18 l g,ie g f I

his 1 l

R1 l

[ 3,,

10 <

y q ,

e o 5 >

4 3

2 1 W 0

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.8 RELATIVE POWER 1

l i

FIGURE 5.1.2 VY CYCLE 18 HALING DEPLETION. EOFPL CORE AVERAGE AXIAL POWER DISTRIBUTION l

1 mes3 <

O I

I R158 R17A R158 R14A R16A R17A R1M R178 RtSA RisA R158 22 1.1M 1 500 1.142 1200 1.007 1.387 1.048 1221 0 813 0 001 0.370 R17A RIAA R17A RisA R17A R1M R17A R178 R16A R188 R15A 20 1.000 1348 1 496 1294 1,420 1228 1.334 1223 0.327 0 0e4 0.366 R168 R17A R168 R17A RtSB R17A R168 Rt7B R14A R148 R168 18 1.143 1 400 1.173 1.470 1.127 1200 1D30 1.190 0.M0 0 828 0310 RisA RisA R17A RiaA R17A Ri4A R17A R17A RISA R158 g 18 1 274 1.204 1.479 1298 f413 1.186 1250 1.114 0 792 0 438 R158 Rt7A R168 R17A R168 R17A R158 RisA R188 14 1.101 1 441 1.137 1.417 1 078 1290 0 917 0 M6 0860 R17A R14A R17A RisA R17A RieA R178 R188 R16A 1200 1232 1.3h6 1.148 1.200 1.064 1.0M 0 750 0 444 RISA R17A R158 R17A R16A R178 R16A RisO R168 10 1 040 1.330 1232 1201 0 019 1D88 0 434 0827 0 343 R178 R17B Rt?B R17A R16A R168 R16A R168 g na 1 225 1228 1204 1.120 08# 0 781 0 826 0 384 RISA R16A R16A R16A R18A RISA R168 ,

nn 0 800 0 931 0 sa2 0.794 0 861 0 448 0.342 R16A R188 R188 RtS8 n4 0 ear 0.e87 0 830 0 442 R16A R168 R168 .. . 8UNDLE D ns 0374 0253 0 321 ...... EOFPL REl.ATNE POWER D N 27 3 31 M M 37 2 41 4 I

I FIGURE 5.1.3 VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL.

BUNDLE AVERAGE RELATIVE POWERS wa I s

=

E 25

{ 24 g

\

I 21

- \

l

I ..

17 o

h

{ss v,,

r 44 j

)is 4

$in 11 F fso

=

L b 7 <,

I 6

i 4

3 3 ,

l o.o o.2 o4 0s e 50 3 ~4 1.s

{

\

FIGURE 5.1.4 I

I VY CYCLE 18 RODDED DEPLETION - ARO AT EOFPL, l

)

CORE AVERAGE AXIAL POWER DISTRIBUTION m*w -22 l

I l_______-___________-____---__----_ _ _ -

)

].j Oi Ill l E. I R158 R17A R158 RISA R15A Mt?A R16A R178 R15A R16A R158 '

22 l 31.704 13 910 32297 23.640 32 414 11.823 31.913 12 410 31477 20.700 20.170 1 1

R17A R16A Rt7A R14A R17A RteA R17A R178 RteA R168 R16A 13.922 23487 14 Dec 24227 13A07 24 000 13.379 12 479 22.366 2023 28.300 R158 R17A RISS R17A R158 R17A RtSB R178 R14A R198 R168 18 32 191 14.001 32 643 14 168 32 006 13.733 32.726 12278 22.071 19 966 29 000 R14A R16A R17A RitA R17A R10A R17A R17A R16A R168 18 23 622 24228 14.186 23 963 13 922 24830 12.741 11.601 21.482 28514 R158 R17A R158 R17A R168 R17A R168 R14A R188 14 32 403 13 901 22533 13 917 32.425 13 024 32 364 22.344 20896 R17A R16A R17A R14A R17A R14A R178 R188 R16A 13818 24.702 13727 24.723 13 024 24 096 11.13e 21.840 20814 RISA R17A R158 R17A R16A R178 R16A R188 R158 10 31.744 13 373 32 787 12.779 32288 11.130 22.040 20153 20 757 R178 R178 R178 R17A R1aA Ries R16A R188 I na 12 404 12 471 12271 11 498 22.330 21 787 20.302 20 926 R15A R16A RisA R18A R18A Rt5A R158 rm 31.684 22 228 23.057 21.472 20 848 20 087 20 03B Ri4A Rie8 Rie8 R158 20 701 20278 10909 28 790 R15A R158 RtS8 .. BUNDLE D ns 28.3e7 as s2s se ses , EOFPL EXPOSURE $3WD/ST) 23 25 27 29 31 33 36 37 39 41 43 I

I FIGURE 5.2.1 VY CYCLE 18 HALING DEPLETION. EOFPL BUNDLE AVERAGE EXPOSURES I, num I I.

=

I  ! I i

)

l E l l

I R168 33785 R17A 11.900 R158 31.713 R10A 23.419 R16A 32339 Rt7A 12.843 R16A 31D52 R17B 12.073 R15A 32.103 R14A 21.618 rte 8 28.801 R17A R16A R17A R18A R17A RieA R17A R178 R10A R198 R15A 12.124 23270 12829 24226 13237 24 R21 12.906 12 223 23.144 21.100 26D88 R158 R17A R158 R17A R158 RITA RitB R17B RieA R188 R158 31487 12.796 32.325 13378 22.708 13.318 22.983 12 078 23 710 20.785 20419

'I R14A RiaA R17A RieA R17A R16A R17A R17A R14A R158 23 421 23.060 13000 23362 13279 26 006 12.575 11.400 22.164 20304 R1&S R17A R188 R17A R188 R17A R188 RieA Ries 32333 12806 32371 13 214 32.793 12 000 32.824 23.070 21.464 R17A R16A R17A R14A R17A R1M R178 RIOS R16A 12.866 24600 13228 25 163 12 A24 24 540 10 S38 22 544 29 379 R16A R17A R158 R17A Rt5A Ri?B R14A R188 RtS8 10 I

31A14 12 857 33.017 12.580 32.742 10 937 22 804 20 013 29283 R178 R175 R175 R17A R14A R188 RiaA R158 na

• 12D15 12.141 12.003 11 379 23.083 22.496 20.983 20 301 I RISA 32.146 R14A 22 982 R14A 23.740 R14A 22.178 R14A 21.002 R16A 28 823 R158 20 434 na 7

R16A R168 R198 R158 21.577 21.118 20043 29302 RISA R168 R188 - SUNOLE D ns to 00s 2e ter 29 627 ..-.. EOFPt EXPOSURE {GWD/ST) 23 25 27 29 31 33 35 37 30 41 43 I

I FIGURE 5.2.2 I VY CYCLE 18 RODDED DEPLETION, EOFPL BUNDLE AVERAGE EXPOSURES I l I

1 l

f] I Oi E'

2.6 2.4 2'

2.0 ^

< I N

~

1.8

/~ g

\

,.. >  % I z

l 1.2 o Minimum seudown uarsin

-- Tm ewion uma 1.0 0.8 I

0.4

......... . . . . . . , ............... ......................, .......,........ . . . . . . , ....... g 0.2 0,0 I

0 1000 2000 3000 4000 $000 6000 7000 8000 0000 10000 11000 CYCLE EXPOSURE (MWD /ST)

FIGURE 5.3.1 VY CYCLE 18 COLD SHUTDOWN MARGIN. IN %AK, VERSUS CYCLE EXPOSURE num -25 I

I E

I 6.0 TIIERMAL-IIYDRAULIC DESIGN 6.1 Steady-State Thermal livdraulics Core steady-state thennal-hydraulic analyses for the Reload Cycle were performed using the FIBWR[16],[17],[18] computer code. The FIBWR code incorporates a detailed geometrical representation of the complex flow paths in a BWR core, and explicitly models the leakage flow to the bypass region and water rod flow. The FIBWR geometric models for each GE bundle type were benchmarked against vendor-supplied and plant thermal-hydraulic information.

j Using the fuel bundle geometric models, a power distribution calculated by SIMULATE-3 and core inlet enthalpy, the FIBWR code calculates the core pressure drop and total bypass flow for several power and flow combinations. The core pressure drop and total bypass flow predicted by the FIBWR code were then used in setting the initial conditions for the system transient analysis model.

6.2 Reactor Limits Determination Section 3.11 of the Technical Specifications requires that the plant assure the perfonnance of the fuel rods by not exceeding the Minimum Critical Power Ratio (MCPR), the Maximum Linear 11 eat Generation Rate (MLliGR), and the Maximum Average Planar Linear lleat Generation Rate (MAPLIIGR).

I The Reload Cycle fuel has MCPR operating limits, shown in Appendix A. The MCPR is a combination of the Fuel Cladding Integrity Safety Limit (FCISL) and the change in a Critical Power Ratio (ACPR) which occurs during an anticipated operational transient. For Vermont Yankee FCISL is 1.07 [2]. CPR is defined as the ratio of the critical power (bundle power at which some point within the assembly experiences onset of boiling transition) to the operating bundle power. The objective for normal operation and anticipated transient events is to maintain nucleate boiling.

Avoiding a transition to film boiling protects the fuel cladding integrity. Both the transient and normal MCPR operating limits are derived with the GEXL-Plus conclation[19), with appropriate coefficients reptrsentative of the Reload Cycle's fuel types. For core flows other than rated, the I

a.w I

O O

MCPR limits must be adjusted by a generic factor, ly19]. The analysis, described in the Section 7.0, determines the Reload Cycle MCPR operating limits.

The Reload Cycle fuel has a Linear IIcat Generation Rate (LilGR) limit of 14.4 kW/ft for all I

bundle types. The basis for this limit can be found in Reference 2.

The Reload Cycle fuel has Average Planar Linear IIcat Generation Rate (APLIIGR) limits, shown in Appendix A. The Maximum APLIIGR (MAPLilGR) values are the most limiting of the fuel rod thermal-mechanical MAPLiiGRs[20] and the LOCA analysis MAPLilGRs (Section 8.2).

The fuel rod thennal-mechanical MAPLilGRs are the result of the GE fuel md thermal-mechanical design analyses, described in Reference 2. These results assume that during steady-state: 1) the maximum LilGR is 14.4 kw/ft,2) the maximum peak pellet exposure is 60.0 GWd/Mt. and 3) maximum operating time is 7.0 years. These results also assume that, during an anticipated operational transient, the thennal and mechr.nical overpower limits [21] are not exceeded. The transient analysis, described in Section 7.0, assures that the thennal and mechanical overpower limits are not exceeded. The LOCA analysis, described in Section' 8.0, determines the LOCA analysis MAPLIIGRs.

I I

I I

I I

R4M3 i I

E

7.0 ABNORMAL OPERATIONAL TRANSIENT ANALYSIS 7.1 Transients Analyzed Transient simulations are perfonned to assess the impact of certain transients on the heat transfer characteristics of the fuel. The purpose of this analysis is: 1) to detennine the MCPR operating limit so that the FCISL is not violated for the transients considered, 2) to assure that the thermal and mechanical overpower limits are not exceeded during the transient, and 3) to demonstrate I compliance with the ASME vessel code limits.

Past licensing analyses have shown that these transients result in the maximum MCPR:

I

1. Pressurization transients, including the generator load rejection with complete failure of the turbine bypass system and the turbine trip with complete failure of the turbine bypass system;

2. Loss of feedwater heating;

3. Local md withdrawal erTor; and I 4. Misloaded bundle ermr, including the rotated bundle ermr and the mislocated bundle ermr.

l To demonstrate that the fuel rod thermal and mechanical overpowers are not exceeded, the

~

maximum powers resulting fmm the pressurization, loss of feedwater heating and rod withdrawal error l transients were compared to the criteria. To demonstrate compliance with ASME vessel code limits, I the main steam isolation valves (MSIV) closing with failure of the MSIV position switch is also analyzed. Brief descriptions and the results of the transients analyzed are pmvided in the following i sections.

'I I mum  !

I

I

i O

O 7.2 Pressurization Traients Analysis 7.2.1 Methodology I

The analysis involves two types of sanalations. A system level simulation is perfonned to detemnine the overall plant response. Transient core inlet and exit conditions and normalized power I from the system level calculation are then used to perform detailed thermal-hydraulic simulations of the fuel, referred to as " hot channel calculations." he hot channel simulations provide the bundle transient ACPR (the initial bundle CPR minus the MCPR experienced during the transient).

The system level simulations are perfonud with the one dimensional (1-D) kinetics RETRAN mode![22],[23],[24]. The hot channel calculations are performed with the RETRAN[25),[26] and TCDYA0llT/],[18],[23] computer codes. The GEXL-Plus correlation [19], contained in TCPYA01, evaluates tin transient critical power ratio.

De hot channel transient ACPR calculations employ a two-pan process, as illustrated by the flow chart in Figure 7.2.1. The first part involves a series of steady-state analyses perfonned with the FIBWR, RETRAN, and TCPYA01 computer codes. The FIBWR analyses utilize a one-channel model for each fuel type being analyzed, with bypass and water rod flow also modeled. The steady-state FIBWR analyses were performed at several power levels with other conditions (i.e., core pressure drop, system pressure, and core inlet enthalpy) held constant. The FIBWR code results provide a steady-state CPR, active channel flow (AF) and bypass flow (BPF) for each active channel power (AP).

The FIBWR conditions for channel power, channel flow, and bypass flow were then used as I

input to steady-state RETRAN/TCPYA01 hot channel calculations. Other assumptions are consistent with those in the FIBWR analysis. De Initial Critical Power Ratio (ICPR) is the result of the steady-state RETRAN/TCPYA01 analysis. Dese results allow for the development of functional relationships, describing AP as a function of ICPR, and AF and BPF as functions of AP for each fuel type. These relationships are used in the iterative process for determining the transient CPR, as shown in Figure 7.2.1.

am I I

= ,

The second part of the hot channel calculations determines the transient CPR perfonnance.

Because the ACPR for a given transient varies with Initial Critical Power Ratio (ICPR), the hot channel analysis is an iterative process. The objective of the hot channel iteration for each transient is to determine the hot channel initial conditions which result in reaching the FCISL. Each iteration requires a RETRAN hot channel mn to calculate the transient enthalpies, flows, pressure and saturation properties at each time step. Rese are required for input to the TCPYA01 code. TCPYA01 is then used to calculate a CPR at each time step during the transient, from which a transient ACPR is derived.

I In response to Reference 11. NRC Safety Evaluation for FROSSTEY-2, the hot channel methodology has considered the assumption of both fixed and time-varying power shapes. The fixed power shape assumes a 1.4 chopped cosine axial distribution which remains constant throughout the transient. The initial power shape for the time-varying power shape methodology is the Haling axial distribution used in the core wide analysis. He time-varying hot channel power distribution is assumed to be the same as that in the core wide analysis to account for the effects of transient power feedbacks and the scram. The transient MCPR limits are defined as the more conservative results from the fixed and varying shape analyses.

7.2.2 Initial Conditions and Assumptions -

The initial conditions for the Reload Cycle are based on a reactor power level of 1664 MWg which includes a 4.5% margin on the current licensed reactor power level of 1593 MW m . This margin conservatively bounds the expected 2% calorimetric uncertainty. The reactor core flow is assumed to be 100% of rated. The core axial power distribution for each of the exposure points is based on the 3<iimensional SIMULATE-3 predictions associated with the generation of the reactivity data (Section 7.2.3). The core inlet enthalpy is set so that the amount of carTyunder from the steam separators and the quality in the liquid region outside the separators is as close to zero as possible. For fast pressurization transients, this maximizes the initial pressurization rate and results in a more severe neutron power spike. A summary of the initial operating state used for the system simulations is provided in Table 7.2.1.

I

, _ 30 I

g 0

During the cycle, Vermont Yankee can adjust the core flow to account for reactivity changes rather than using the contml rods. During this type of operation, core flow may be as low as 87%

while at 100% power. To ensure the safety analysis bounds these conditions, transients are also analyzed at the limiting exposure statepoint at 1654 MW power and 87% flow. Limiting exposure is defined as the exposure which had the highest ACPR.

Assumptions specific to a particular transient are discussed in the section describing the transient. In general, the following assumptions are made for all transients:

1. Scram se@ints we at Technical Specification [12] limits.

2. Protective system logic delays are at equipment specification limits.

3. Safetyhelief valve and safety valve capacities are based on Technical Specification rated values.

I

4. Safetyhelief valve and safety valve setpoints are modeled as being at the Technical Specification upper limit. Valve responses are based on slowest specified response values.

5. Control md drive scram speed is based on the Technical Specification limits. The I

analysis addresses a dual set of scram speeds, referred to as the " Measured" and the "67B" scram times. " Measured'* refers to the faster scram times given in Section 3.3.C.I.1 of the Technical Specifications. "67B" refers to the slower scram times given in Section 3.3.C.I.2 of the Technical Specifications.

7.2.3 One-Dimensional Cmss Sections and Kinetics Parameters I

The one-dimensional (1-D) cross sections and kinetics parameters are generated as functions of fuel temperature, moderator density, and scram. The method [28) is outlired below.

mm E l

_ - __ _ -____-___ -__. l

l I

A complete set of 1-D cross sections, kinetics parameters, and axial power distributions are generated from base states using the IIaling depletion established for EOFPL, EOFPL-1000 mwd /St.

EOFPL-2000 mwd /St, and BOC exposure statepoints. These statepoints are characterized by exposure and void history di.e.ributions, control rod patterns, and core thermal-hydraulic conditions. The latter  !

are consistent with the assumed system transient conditions pmvided in Table 7.2.1.

The BOC base state is established by shuffling from the previously defined Current Cycle endpoint into the Reload Cycle loading pattem. A criticality search pmvides ai estimate of the BOC critical rod pattem. The EOFPL and intermediate core exposure and void hira>ry distributions are calculated with a Haling depletion as described in Section 5.2. The tOFPL state is unrodded. 'Ihe EOFPL-1000 mwd /St and EOFPL-2000 mwd /St exposure statepoints require base control rod pattems. These are developed to be as " black and white" as possible to minimize the scram reactivity, maximize the core average moderator density reactivity coefficient and, therefore, maximize the transient power response. Beginnmg with the rodded depletion configuration, all contml rods which are more than half inserted are fully insened, and all control rods which are less than half insened are fully withdrawn. If the SIMULATE-3 calculated parameters are within operating limits, then this configuration becomes the base case. If the limits are exceeded, a minimum number of contml rods are Ldjusted a minimum number of notches until the parameters fall within limits. -

At each exposure statepoint, a SIMULA*IE-3 initial contml state reference case is run. A series of perturbation cases are run with SIMULATE-3 to independently vary the fuel temperature, moderator temperature, and core pressure. All other variables normally associated with the SIMULATE-3 cross sections are held constant at the reference state. To obtain the effect of the contml md scram, another SIMULATE-3 reference case is run with all-rods-in. The perturbation cases described above are run again fmm this reference case. For each control state, a data set of kinetics parameters and cross sections is generated as a function of the perturbed variable. There is a table set for each of the 27 neutronic regions,25 regions to represent the active core and one region each for the bottom and top reflectors.

l I 1 men I ,

____________.._-.---__l

O O

7.2.4 Turbine Trl9 Without Bynass Transient (TrWOBP)

De transient is initiated by a rapid closure (0.1 second closing time) of the turbine stop i valves. It is assumed that the steam bypass valves, which normally open to relieve pressure, remain closed. A reactor pmtection system signal is generated by the turbine stop valve closure switches.

Contml rod drive motion is conservatively assumed to occur 0.27 seconds after the start of turbine stop valve motion. The ATWS recirculation pump trip is assumed to occur at a setpoint of 1150 psig dome pressure. A pump trip time delay of 1.0 second is assumed to account for logic delay and M-G set generator field collapse. In simulating the transient, the bypass piping volume up to the valve chest is lumped into the control volume upstream of the turbine stop valves. Predictions of the salient system parameters at the three exposure points are shown in Figures 7.2.2 thmugh 7.2.4 for the

" Measured" scram time analysis.  !

7.2.5 Generator Load Rejection Without Bypass Transient (GLRWOBP) j The transient is initiated by a rapid closure (0.3 seconds closing time) of the turbine control valves. As in the case of the turbine trip transient, the bypass valves are assumed to fail. A reactor protection system signal is generated by the hydraulic fluid pressure switches in the acceleration relay EI of the turbine contml system. Control md drive motion is conservatively assumed to occur 0.28 seconds after the start of turbine control valve motion. ne same modeling regarding the ATWS pump trip and bypass piping is used as in the turbine trip simulation. The influence of the accelerating main turbine generator on the recirculation system is simulated by specifying the main turbine generator electrical frequency as a function of time for the M-G set drive motors. The main turbine generator frequency curve is based on a 100% power plant startup test and is considered representative for the simulation. The system model predictions for the three exposure points are shown in Figures 7.2.5 through 7.2.7 for the " Measured" scram time analysis.

7.2.6 Pressurization Transient Analysis Results The transients selected for consideration were analyzed at exposure points of EOFPL, EOFPL-1000 MW4/St, and EOFPL-2000 mwd /St. The transient results, reported in Table 7.2.2, correspond to the limiting bundle type in the core. The MCPR limits, in Table 7.2.2, are calculated by num a

=

I adding the calculated ACPR to the FCISL. The worst ACPR for the pressurization transients intlude an adjustment to allow for the exposure window ofi600 mwd /St on Current Cycle and the exposure uncertainty on the Reload Cycle [7).

7.3 IAss of Feedwater Heating Transient (LOFWH) Results A feedwater heater can be lost in such a way that the steam extraction line to the heater is shut off or the feedwater flow bypasses one of the heaters. In either case, the reactor will receive cooler feedwater, which will produce an increase in the core inlet subcooling, resulting in a reactor power increase.

I De response of the system due to the loss of 100*F of the feedwater heating capability was analyzed. This represents the maximum expected feedwater temperature reduction for a single heater or group of heaters that can be tripped or bypassed by a single event. The system model used is the same as that used for the pressurization transient analysis (Section 7.2.1). The initial conditions and modeling assumptions discussed in Section 7.2.2 are applicable to this simulation.

1 Vermont Yankee has a scram setpoint of 120% of rated power as part of the Reactor Protection System (RPS) on high neutron flux. In this analysis, no credit was taken for scram on high I neutron flux, thereby allowing the reactor power to reach its peak without scram. This approach was selected to provide a bounding and conservative analysis for events initiated from any power level.

The transient response of the system was evaluated at several exposures during the cycle.

EOFPL-1000 mwd /St, EOFPL-2000 mwd /St, and BOC. The transient results, corresponding to the limiting bundle type in the core, are listed in Table 7.3.1. The MCPR limits in Table 7.3.1 are calculated by adding the calculated ACPR to the FCISL. The transient evaluation at EOFPL-1000 mwd /st was found to be the limiting case between BOC to EOFPL-1000 mwd /St. The results of the system response to a loss of 100 F feedwater heating capability evaluated at EOFPL-1000 mwd /St as predicted by the RETRAN code are presented in Figure 7.3.1.

I

{I l R WJ3 -N I

O 7.4 Overpressurization Analysis Results Compliance with ASME vessel code limits is demonstrated by an analysis of the Main Steam Isolation Valves (MSIV) closing with failure of the MSIV position switch scram. EOFPL conditions  !

were analyzed. The system model used is the same as 01at used for the transient analysis (Section 7.2.1). The initial conditions and modeling assumptions discussed in Section 7.2.2 are applicable to this simulation.

The transient is initiated by a simultaneous closure of all MSIVs. A 3.0 second closing time, which is the minimum time in Technical Specification Table 4.7.2 is assumed. A reactor scram signal is generated on APRM high flux. Control rod drive motion is conservatively assumed to initiate 0.28 seconds after reaching the high flux setpoint. The system response is shown in Figure 7.4.1 for the

" Measured" scram time analysis.

The maximum pressures at the bottom of the reactor vessel calculated for the " Measured" scram time analysis and for the *67B" scram time analysis are given in Table 7.4.1. These results are a within the ASME code overpressure design limit which is 110% of the vessel design pressure.

Vennont Yankee's design pressure is 1250 psig so the maximum pressure limit is 1375 psig.

7.5 local Rod Withdrawal Error Transient Results ,

The rod withdrawal error (RWE) is a local core transient caused by an operator erroneously  ;

withdrawing a control rod in the continuous withdrawal mode. If the core is operating at its operating limits for MCPR at the time of the enor, then withdrawal of a control rod could increase both local and core power levels with the potential for overheating the fuel.

There is a broad spectrum of core conditions and control rod patterns which could be present I

at the time of such an error. For most normal situations it would be possible to fully withdraw a control rod without violating the FCISL.

The MCPR operating limit for the RWE is defined at each Rod Block Monitor (RBM) System setpoint so that the FCISL is not violated. The consequences of the error depend on the local power num Ii

i I

increase, the initial MCPR of the neighboring locations and the ability of the RBM to stop the withdrawing md before MCPR reaches the FCISL.

I The most severe transient postulated begins with the core operating according to normal pmcedures and within normal operating limits. The operator makes a pmcedural error and attempts to fully withdraw the maximum worth contml rod at maximum withdrawal speed. The core limiting locations are close to the error rod. They experience the spatial power shape transient as well as the overall core power increase.

The core conditions and control rod pattem are conservatively modeled for the licensing bounding case by specifying the following set of concurrent worst case assumptions:

I

1. The rod should have high reactivity worth. The worst rod is identified by running the full RWE analysis for the control rods as found in the normal contml rod pattems of the rodded depletion. Every control rod sequence is checked. From this examination, the control rods that result in the highest worth and highest ACPR are identified.

Licensing test case rod pattems are then developed to further exaggerate the worth and ACPR impact of the rod to be withdrawn.

The test pattems are developed with xenon-free conditions. The xenon-free condition and the additional control rod inventory needed to maintain criticality exaggerates the worth of the withdrawn control rod when compared to normal operation with normal xenon levels.

2. The core is modeled at 104.5% power and 100% flow.

I 3. The core power distribution is adjusted with the available control rods to place the lot ms within the four by four array of bundles around the error rod as close to the operating limits as possible.

I 1 Of the many patterns tested. the pattem with the most limiting ACPR results is I

selected as the bounding case.

R*M) I

C 0

The RBM System's ability to terminate the bounding case is evaluated on the following bases:

1. Technical Specifications allow each of the separate RBM channels to remain operable if at least half of the Local Power Range Monitor (LPRM) inputs on each level are operable. For the interior locations tested in this analysis, there are a maximum of four LPRM inputs per level. One RBM channel averages the inputs from the A and C levels; the other channel averages the inputs from the B and D levels. Considering the inputs for a single channel, there are eleven failure combinations of none, one and two failed LPRM strings. The RBM channel responses are evaluated separately at these eleven input failure conditions. Then, for each channel taken separately, the lowest response as a function of error rod position is chosen for comparison to the RBM setpoint.

2. 'Ihe event is analyzed separately in each of the four quadrants of the core due to the I

differing LPRM string physical locations relative to the error rod.

Technical Specifications require that both RBM channels be operable during nonnal operation. ,

Thus, the first channel calculated to intercept the RBM setpoint is assumed to stop the rod. To allow for control system delay times, the rod is assumed to move two inches after the intercept and stop at the following notch. '

The analysis is performed using SIMULATE-3. The two separate cases presented here are selected from numerous explicit SIMULATE-3 analyses. Case I analyzes the bounding event with zero xenon, initiated from 104.5% power and 100% flow. This case also assumes the worst case W abnormal rod pattem configuration which results in the initial MCPR being as low as possible. Case 2 is the worst of all the 'od withdrawal transients analyzed from 100% power,100% flow, equilibrium xenon, and normal rod pattems used in the rodded depletion. The worst transient ACPR results for both cases are shown in Table 7.5.1. The ACPR values are evaluated such that the implied MCPR operating limit equals FCISL + ACPR This is done by conserving the figure of merit (ACPR/ICPR) shown by the SIMULATE-3 calculations. The transient ACPR results for Case I will be used to set the operating MCPR limits. Case 2 results are bounded by the Case I results by at least 0.02 ACPR margin to assure that the exposure uncertainties on the Current Cycle and the Reload Cycle are um '

=

accounted for. This method also provides valid operating MCPR values that bound expected operating conditions.

The Case 1 (bounding event) RBM channel responses are shown in Figusts 7.5.1 and 7.5.2.

Wey also show the control rod position at the point where the RBM channel response first intercepts

{ the RBM setpoint.

7.6 Misloaded Bundle Frror Analysis Results

{

7.6.1 Rotated Bundle Ermr The primary result of a bundle rotation is a large increase in local pin peaking and the associated R-factor as higher enrichment pins are placed adjacent to the surrounding wide water gaps.

In addition, there may be a small increase in reactivity, depending on the exposure and void fraction states. The R-factor increase results in a CPR reduction. The objective of the analysis is to ensure that, in the worst possible rotation, the FCISL is not violated with the most limiting bundles on their operating limits.

To analyze the CPR response, rotated bundle R-factors as a function of exposure are developed by adding the largest possible AR-factor resulting from a rotation to the exposure dependent F

L R-factors of the properly oriented bundles. Using these rotated bundle R-factors, the MCPR values resulting from a bundle rotation are determined using SIMULATE-3. This is done for each control

[ rod sequence throughout the cycle. The process is repeated with the K-infinity of the limiting bundle modified slightly to account for the increase in reactivity resulting from the rotation. For each

[ sequence, the MCPR for the properly oriented bundles is adjusted by a ratio necessary to place the corresponding rotated bundle's CPR on its FCISL. The adjusted MCPRs at each exposure is the

{ rotated bundle operating limit for the rotated bundle error.

Because the BP8DWB335 fuea designs exhibit a significant increase in R-factor with rotation early in exposure, the impact upon the rotated bundle ACPR is high at BOC. His effect soon drops off with exposure. Therefore, the operating MCPR limit resulting from a rotation is presented in Table 7.6.1 versus cycle exposure.

1 l

num 1 l

. _ _ _ _ _ _ _ _ ______ _ _________ ______-___. _____________________________________________________L

n 0

7.6.2 Mislocated Bundle Error Mislocating a high reactivity assembly into a region of high neutron importance results in a location of high relative assembly average power. Since the assembly is assumed to be properly oriented (not rotated), R-factors used for the mislocated bundle are the standard values for the given fuel type.

The analysis uses multiple SIMULATE-3 cases to examine the effects of explicitly mislocating I

every older interior assembly in a quarter core with a fiesh or once-bumed assembly. Because of symmetry, the results apply to the whole core. Edge bundles are not examined because they are never limiting, due to neutron leakage.

The effect of the successive mislocations is examined for every control md sequence throughout the cycle. For each sequence, the MCPR for the pmperly loaded core is compared to the MCPR of the misloaded core at the mistoaded location. The MCPR for the properly loaded core is adjusted by a ratio necessary to place the mislocated assembly on the FCISL. The maximum of these adjusted MCPRs is the mislocated bundle operating limit. 7he results of the mislocated bundle g

analysis are given in Table 7.6.2. E 7.7 Transient Analysis Results The results of this transient analysis has: 1) determined the MCPR operating limit so that the FCISL is not violated for the transients considered, 2) assured that the thermal and mechanical overpower limits are not exceeded during the transient, and 3) demonstrated compliance with the ASME vessel code limits.

The MCPR operating limits for the Reload Cycle are calculated by adding the calculated ACPR to the FCISL at each of the exposure statepoints for each transient. Table 7.7.1 lists the limiting transient for each statepoint. For an exposure interval between statepoints, the highest MCPR limit at either end is assumed to apply to the whole interval. The highest calculated MCPR limits for the Reload Cycle for each of the exposure intervals for the various scram speeds and for the various rod block lines are provided in Appendix A. These MCPR operating limits are valid for operation of me.n a

=

I the Reload Cycle at full power up to 10644 mwd /St and for operation during coastdown beyond EOFPL.

I I

I .

lI ,

i l

1 l

l

e i

e

!I

O O

TABLE 7.2.I VY CYCLE 18

SUMMARY

OF SYSTEM TRANSIENT MODEL INTTIAL CONDITIONS FOR TRANSIENT ANALYSES Core 'Ihennal Power (MWm) 1664.0 Turbine Steam Flow (10$bm/hr) 6.75 Total Core Flow (10$b,/hr) 48.0 Core Bypass Flow (10$b,/hr)* 6.28 Core Inlet Enthalpy (BTU /lb,)

I 523.2 Steam Dome Pressum (psla) 1034.7 I

Tushine Inlet Pressure (psia) 985.7 Total Recirculation Drive Flow (10$b,/hr) 23.7 I ,

Core Plate Differential Pressure (psi) d 20.4 Nanow Range Water Level (in.) 162.0 ,

Average Fuel Gap Conductance (See Section 4.2)

I I

includes water rod 110w.

g RAN3 MI-E

I  !

l I

TABLE 7.2.2 )

VY CYCLE 18 PRESSURIZATION TRANSIENT ANALYSIS RESULTS Peak Prompt Power Peak Average Heat Exposure (Fraction of Flux (Fraction of Transient Transient Stateroint Initial Value) Initial Value) ACPR* MCPR Limits ,

I Turbine Trip Without EOFPL 2.80378 1.19541 0.25 132 I Bypass, " Measured" Scram Time EOFPL 1000 EOFPL-2000 2.24381 1.25154 1.142 %

1.00000 0.20 0.04 1.27 1.11 I Turbine Trip Without EOFPL 3.12172 1.23971 0.27 134 Bypass, "67B" Scram EOFPL-1000 2.63805 1.19536 0.23 130 Time EOFPL-2000 1.64655 1.04145 0.08 1.15 Generator Load EOFPL 2.81519 1.17663 0.23 130 Rejection Without EOFPL-1000 234636 1.12670 0.19 1.26 Bypass. " Measured" Scram Time EOFPL-2000 1.13637 1.00000 0.02 1.09 Generator Load EOFPL 330268 1.23750 0.27 134 Rejection Without EOFPL-1000 2.92611 1.19323 0.23 130 Bypass. "67B" Scram Time EOIPL-2000 1.60336 1.01760 0.05 1.12 I

• The worst ACPR for'ITWOBP and GLRWOBP includes a 0.01 ACPR adjustment to allow for the exposure window ofi600 mwd /St on Current Cycle and the exposure uncertainty on the Reload Cycle. I am3 i I

O O

TABLE 7.3.1 VY CYCLE 18 LOSS OF FEEDWATER HEATING TR ANSIENT ANALYSIS RESULTS Peak Prompt Power Peak Average Heat I

Exposure (Fraction of Flux (Fraction of Transient Transient Statenoint Initial Value) Initial Value) ACPR MCPR Limits Loss of 100 F EOFPL-1000 1.24680 1.15373 0.12 1.19 Feedwater Heating EOFPL-2000 1.14354 1.14440 0.11 1.18 BOC 1.17041 1.14530 0.11 1.18 I

I I

I REh53 I E

=

I  ;

TABLE 7.4.1 I

VY CYCLE 18 OVERPRESSURIZATION ANALYSIS RESULTS I Maximum Pressure at Reactor i Conditions Vessel Bottom (psig)

" Measured" Scram Time 1251 t "67B" Scram Time 1278 TABLE 7.5.1 1 VY CYCLE 18 ROD WITHDRAWAL ERROR ANALYSIS RESULTS I Rod Block Monitor Transient MCPR Setpoint Bounding Case ACPR Worst Normal ACPR Limits 104 0.15 0.13 1.22 105 0.16 0.14 1.23 g 10e 0.ie 0.14 1.23 107 0.20 0.18 1.27 I 108 0.26 0.18 1.33 I

I I - .

a l 5'

TABLE 7.6.1 VY CYCLE 18 ROTATED BUNDLE ANALYSIS RESULTS Exrosure (GWd/St) Transient MCPR Limit 0.0 1.39 4.0 1.35 5.5 1.29 6.5 1.25 10.0 1.25 I

TABLE 7.6.2 VY CYCLE 18 MISLOCATED BUNDLE ANALYSIS RESULTS I

Transient MCPR Limit 1.15 I

I Il I'

RWJ3 E E

TABLE 7.7.1 VY CYCLE 18 LIMITING TRANSIENTS I Rod Block Monitor Setpoint Scram Time Exticeure (GWd/SO Limiting Transient Transient MCPR Limit 108 Measured 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 5.5* Rod Withdrawal Error 133 108 "67B" 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 I I

5.5 Rod Withdrawal Error 133 9.035* Turbine Trip 134 107 Measured 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 I, Rotated Bundle 1,29

)j 5.5 6.5 Rod Withdrawal Error 1.27 9.035* Turbine Trip 132 )

107 "67B" 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 I 5.5 6.5 Rotated Bundle Rod Withdrawal Error 1.29 1.27 8.035 Turbine Trip 130 8 9.035* Terbine Trip 134 106 Measured 0.0 Rotated Bundle 139 4.0 Rotated Bundle 135 l

5.5 Rotated Bundle 1.29 l 6.5 Rotated Bundle 1.25 8.035 Turbine Trip 1.27 9.035* Turbine Trip 132 l "67 B" 0.0 Rotated Bwdle 139 I I

106 4.0 Rotated Bundle 135 5.5 Rotated Bundle 1.29 6.5 Rotated Bundle 1.25 8.035 Turbine Trip 130 9.035* Turbine Trip 134 I -

'< ' " ce" ' >>-'<> > << <"> xe

< e.

<= < eo er'-

I __

O N

Part I Ii,

)

e Part 11 i non and

'

RET W /

s' TCPYA01 e' Estimat,e, Ini,tial '

' p go e

5tg, te Transient Analysis

g e

Le1YtionYhtb AP = f(ICP l*

RETRAN/TCPYA41 Transient H6%

= gA Channel Analysis Estiente New ICPRI as:

ICPRg =

CP 1= ICPaa no natent FCist?

l: Tor l t

1 1!  !

i FIGURE 7.2.1 Ii

_F_ LOW CHART FOR THE CALCULATION OF ACPR USING THE RETRANfrCPYA01 CODES num I I

E' e

i

v. -

1 I

I TTWOBP, EOFPL, MST TTWOBP, EOFPL, MST 10F 5 2OF5 I- 6 1.5 I 5-1.25 -

@4- j '.,

E $ >  ! '.

e 4 l '

O E -

Iz w Z

3-z g 1.0 0

'. . /

l @32-o P

h u.

\. .

0.75 - ',,

3 I CORE INLET FLOW l NORM POWER l ---- AVE. HEAT FLUX I 0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.5 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) TIME (SEC)

I FIGURE 7.2.2

,. TURBINE TRIP WITI-IOUT BYPASS EOFPL18 TR ANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME I

E

E E

I I

l' TTWOBP, EOFPL, MST TTWOBP, EOFPL, MST I

3 0F 5 4 0F 5 1500 2000

---

g.....,,..,

p 1500 -  ! s s 8

i kf 1400 -

i i*

iA

!r ii*1 1000 -

!ji ii i5 V

R 1 i ! 1I i r**ii**

ih i ! i 50

! ! I  :

E;. _

O l

t  !!!

i!? \!

w 1i i

@ 1300 -

$ 500 - l ji lli i V 3 j y w $ I I {l; cr M *

• i E 0

$ l 1 \.  ;

i & !I' 8 1200 - 3:

O -500 -

• !I 8 E 5

g m

W fl ,

1000 -

1100 -

' ~

S/R VALVE (NEG)

---

FEED WATER 15 l STEAM DOME PRES. l ~~**~ VESSEL STM. Ct;T 1000 , , , , , 2000 , , , , ,

0 1 2 3 4 5 6 0 2 3 1 4 5 6 TIME (SEC) TIME (SEC) l FIGURE 7.2.2 l

(CDntinued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18 TR ANSIENT RESPONSE VERSUS TIME, " MEASURED" SCR AM TIME l num l gj E

i i

i I TrWOBP, EOFPL, MST l 2.0 5OF5 t

[

I 1.8 i 1.6 i 1.4 4 /

t

[

t 1.2 i .

, l i 1.0 : / \  !

! \  !

0.8 .  :

l

. i n i

• m 0.6 _

1  :

5 0.4-i a

/

I t

Q~

0.2 4 d'

l *'

> 0.0 ,

b s ' .' --- -------~~~~. **,

2 -0.2 : 's -

I H  %.

O- '.

@ 0.4 : '*s e -0.6 .  %,

i I

-0.8 i 's, s

-1.0 i i i

1.2 : 's, I

i

'l 4 5 TOTAL -

1.6-i * * *

• DOPPLER k i

.......* WDERATOR i 1.8 i I

..... SCUM i.

2.0 , ,. , , , .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 I

TIME (SEC)

I FIGURE 7.2,2 (Continued)

TURBINE TRIP WITHOUT BYPASS. EOFPL18 g TR ANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME 1

aum I

\ ll zo g zw g Oz 2 *wc 0 1 2 3 4 5 6 n 0 0

a s

l 0 . N

. 5 . O R

M

. T T

T .

P O W R 1

. W A T .0 E R

O N T MI .

l B

S U I

E 1

,P E R

(

S 1 0 E N BI E

C 5 F O T N )

5 F

R E P L-E S T 2 ,1 P R i

I 0 O P M N S S W T E IT 2 i 5

V H E O R U F S T G I

3 .

U B

- S U 0 5

T P Y R 1

- I E M A 7 S

,E ,S 2 gQn z 0 zE4 >w$

3 0 E 0 1 1

M" O .

5 7

5 0 2

5 1

5 E F 0 A P .

S L 0 U 18 -

R 1- -

E 0 -

D" 0 i AC .',.

0 5 VO -

0 ER -

S . / /

H E T C M EN I

/l T R W 1 A L T E .

. W A D FT i

0 L ULF O

M /

B T T I

S T IM E

XO W ,.'.' 2

,P E

M (

5 1 i ,.' 0 O E S -

- F F E

)

C .

,- 5 P L-2 - ,1 0 ,-

\ M S

.- T 2 . -

5 . .

3 -

0 El I l I g 3g I I I E l lil i1l

I I

I I

I TTWOBP, EOFPL-1, MST TTWOBP, EOFPL-1, MST l

3 OF 5 4 0F 5 1500 2000 I

g 7. ....,,,,

i i

L 1500 - i: { j:

ii I

!  ! i !i 14 " -

1000 -

i ilh  ! I ii li i jfi ! i ! P; !i

% i i* i i l 1 r*

E  : 8

j 1:Y i! i Im% 6 y 500 -

l: l! ) il{i

• Il il il tl 1

$13#- s i i  !! y m i:  ::

d i!  !!

Ic@mn.

m M

W 0

! '.i 2  :

I *.

I 8 1200 - 3:

0 -500 - i- l' 1000 - l 1100 - 1 l

l 1500 -

S/R VALVE (NEG)

---

FEED WATER l

l l STEAM DOME PRES.l *a-

- VESSEL STM. OUT 1000 .

, .. -2000 .

0 1 2 3 4 5 6 0 2 3 4 1 5 6 I TIME (SEC)

I TIME (SEC) l l

l t FIGURE 7.2.3

'I (Continued) l i

TURBINE TRIP WITHOUT BYPASS EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCR AM TIME i

I am O

5 I

I I

TTWOBP, EOFPL-1, MST 5 OF S 2.0 ,

1.8 i l' 1.6 i  !

1.4 i p

[

1.2 i l \ /

1.0 -i

[ ..' =

0.8 i  !

[0.6" a

0.4 i  !

80.2; l **

0.0 g t.0.,s  !. .. ... ....................

\

-0.4 i i g ~.

g -0.6 -i 'N

-0.8 4 '

1.0 -i '\,

1.2 : 1,

'l*4 TOTAL 's 1.6 : -*** DOPPLER i g MODERATOR \'

1.8 : ....- SCRAM i'

2.0 . ,,, .,

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC)

FIGURE 7.2.3 I

(Continued)

TURBINE TRIP WITHOUT BYPASS, EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCR AM TIME am g

1

-l m I l

b g 1 E ,

i 1

k.-

l TTWOBP, EOFPL-2, MST TTWOBP, EOFPL-2, MST 10F 5 2 0F 5 6 1.5 5- i 1.25 - i

{:.

$ 4- g l

> =

k E

3- E 1 .0 .

. B - .,

z g '.. , ,, ' '.,

\

?c g ,

@2- g u.

0.75 - ',

1

\ .,'~

CORE INLET FLOW l NORM. POWER l = == = AVE. HEAT FLUX 0 .

... 0.5 ...., ..

i. .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) TIME (SEC)

FIGURE 7.2.4 TURBINE TRIP WITHOUT BYPASS. EOFPL18-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME i

aan l F _ -

C,.,

LI I

I TTWOBP, EOFPL-2, MST TTWOBP, EOFPL-2, MST 3OF5 4 OF 5 1500 1 2000 J -,---...........,"

f. ..,,

1500 - 4 i:

r.

l

"*~

1  !  ! \ l\

j.;I): n}

_ 1000 -  : j i lg e

i -

.: :: :  : : : i; l i.s et -

.!  ! iF  : :

w g  : . :J  ::  : I i.

• 500 - !I :Y 55 $I ~

@ 1300 - E i j S i ii y m 2  :  :  : .

m Q i i :i i w d  ;  :  : +

E A

m 0

I W  : i '!:

m i: ji 2  : i 8 1200 - 3 lji 2 0 -500 -

s s' iji ::j:

b h <

-1000 - l 1100 - I S/R VALVE (NEG)

I,

---

FEED WATER l STEAM DOME PRES.l ... ... VESSEL STM. OUT 1000 ,,, , ,, , ,

, 2000 ,,, ,,,, , , ,

0 1 2 3 4 5 6 0

, 1 2 3 4 5 6 '

TIME (SEC) TIME (SEC)

I FIGURE 7.2.4 I

(Continued) l TURBINE TRIP WITHOUT BYPASS EOFPL18-2000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCR AM TIME

.- .ss.

g

7 l

I t

l I

i l

TTWOBP, EOFPL-2, MST S OF 5

/

1.8 : '..'

1.6 :

/

l 1.4 i l I

l 1.2 '

1.0 :

1 f

f t

0.8 i  !

{0.6i 50.4:  !

1 J g 0.2 -i j **.*...

i b

0 .0 a'.. ... .........- *** .....

l 2 -0.2 :

I

\

O -0.4 :

b a -0.6 i 5

\

\,

-0.8 4 i i 1.0 : i s

! -1.2 -i '5 s 1

• I'4 ~ TOTAL 1.6 : ---- DOPPLER I MODERATOR 1.8 -i ... SCRAM

-2.0 . .

0.0 0.5 1.0 1.5 2.0 2.5 3.0 u {

TIME (SEC) l 4

F I FIGURE 7.2.4 (Continued) I TURBINE TRIP WITHOUT BYPASS. EOFPL18-2000 MWD /ST l TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME '

ami l

-- ______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ---a

I z83, zugoz 2a@ s 0 ' 2 3 4 5 6 a 0 - -

c 0

u l

s 0

5

, N O

R M

T P G O L R 1 W R AG T .0 ,

N E I

M E

R W S N E l

O I B E E

( 1 S 1 ,P N R E 5 0

T A )

C F E R T 5 O E O F

S R 2 P P L 0

,L O O M N A S S D T E 2 ,

V E R 5 E J R E F S C G I

3 U T

- S O

I U 0 5

7 T N R

- I E M W 7

.E I 2 gai z: OEyy4 s T .

5 H 0 0 1 M" O 1 .

7 . 2 1 E U 5 5 0 5 5

A 0 -

S T 0

U B =

R Y -

E P - -

D A S 0,

. AC ,.

" 5 VO '

S .S E. ER .-

H C E E I

N

/,. G R O A L L A F 1

0 T E FT  :: / -

R M P L F UL

'.'.- W T

L 1

TI XO O I 8 M W .'. 2 B M E 1 ,.' ,P E

(

S 5

' 0 E E F C ' 5 O

) F 2 , ,.- P 0 , ,L

_ ' M S

T 2,

5 '

_ 3 0

EE I I t *E I I I I Ea i  ;

,. i

-]

[

[

E

[

GLRWOBP, EOFPL, MST GLRWOBP, EOFPL, MST 3 OF 5 4 0F 5 1500 I 2000

-g...............,,,,......

j .<

1500 - i  !!

,e !o:S I:

1400 -  !  !:  ! I  ?

f {

IMO-l l '.': i i  : i:t 8

I

!i l 1 !T  !:

7

! i .!

i 5

it; i i

@ 1 i si

g. 1300 - @ !i !:: :~~ if1: i!

• i l 500 - g l ! l!.

3  :

gn  : ! I:

d I i li,i f 8 0 i j W

ii 1200 - ) i!

o -500 -  :; I

$ ['

b i 1 1000 - .

1100 -

1500 -

S/R VALVE (NEG)

.... FEED WATER l STEAM DOME PRES.l -- VESSEL STM. OUT 1000 ., , , ,. , .. -2000 , , ,

1 ., .

p 0 1 2 3 4 5 6 0 1 2 3 4 5 6 L TIME (SEC) TIME (SEC) j h

r FIGURE 7.2.5 (Continued)

GENERATOR LOAD REJECTION WITIIOUT BYPASS. EOFPL18 TR ANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME anus F _ . -

E E

I' I

I GLRWOBP, EOFPL, MST 1.8 -i I '

j 1.6 :

[

1.4 i  !

1.2 : j

. i. [

/  !

1.0 ; i 0.8 : [ \. ,,,/

[ 0.6 ; [

a 0.4 -i l

Q 0.2 ;

~

j ,,.'

0.0 ' d. -

{a-0.2 4 0 -0.4 i

<C s

E-0.6-i .s.'s', .

-0.8 d '.'

1.0 :

-1.2 i i 4

-1.4 -i TOTAL g 1.6 : ---- DOPPLER i, m

........ wogywn

-1.8 t

....- scuu  !

-2.0 , ,.

^

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC)

FIGURE 7.2.5 I

(Continued)

GENERATOR LOAD REJECTION WITilOUT BYPASS, EOFPL18 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME a.m B E

n fj

[:

0 GLRWOBP, EOFPL 1, MST GLRWOBP, EOFPL-1, MST

{

10F 5 2 0F 5 6 1.5 5-1.25 -

w E 4- y t > l ~,

[ W  !

E  : .

3- M 1 0- , , \.,

w z

8 g

\. /

g, s .,

y 9 -

k '..,

0.75 - \. .

1- '

CORE INLET FLOW l NORM. POWER l ---- AVE. HEAT FLUX 0 ...., , ,.. , .

, . 0.5 . . ... ,, , ..... ,,

0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 - 2.0 2.5 3.0 TIME (SEC) TIME (SEC)

FIGURE 7.2.6 GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME

. num .. .. .

l l

f~a i c_s i UI  :

l i

I' I

GLRWOBP, EOFPL-1, MST GLRWOBP, EOFPL-1. MST 3 0F 5 4 0F 5 1500 2000 p-----------.....,,,,_ 4 i  !!

1500 - i :i i 2 !i t !i 1400 - i !i Ii !! r.

! ik i : !i  ! $

1000 - i !i ! ! !% !i R i ii : i ! *

. /.5 5 i !i i i/ il 5 E o i  : i r! !! i!

m

• C 500 - i ! !!  !! 'i 3 1300 - @

s  : ! :: #

!3 m 9

- i  !!

e m 0 8

CL W .' . 1 W

2 km  !*

8 1200 - 3: i :

O -500 - i .!

$ i a t 1000 -

1100 -

~ ~

S/R VALVE (NEG)

---

FEED WATER l STEAM DOME PRES.l * - VESSEL STM. OUT 1000 ,

, , ,,. , , -2000 .., .

0 1 2 3 4 5 6 0 2 3 4 1 5 6 TIME (SEC) TIME (SEC)

FIGURE 7.2.6 I

(Continued)

GENERATOR LOAD REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME nam l

=

I I

I I

, GLRWOBP, EOFPL-1, MST 5 OF 5 2.0 ,

1.8 i f 1.6 i f

1.4 i ,

j E

1.2 i 1.0 i f

/ '.,/

0.8 i I

[0.62

$a 0.4 i  !

f i

80.2i l I 0.0 a -0.2 i

~l-

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

l 0 -0.4 i i 6.

a -0.6 . '..,~.

-0.8 i 's 1.0 i ',i

-1.2 i I.

I -1.4 i

-a TOTAL

-1.6 i ---- DOPPLER MODERATOR 1.8 . ..... SCRAM

't g

i

-2.0 i

, , , l, ,

0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC)

I FIGURE 7.2.6 l

(Continued) lI GENER ATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST l

.am TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME l

Rum l

O 0

E1 Il GLRWOBP, EOFPL-2, MST GLRWOBP, EOFPL-2, MST 10F 5 20F5 6 1.5 5-I 1.25 -

w 4~

s >

2 d z F 0

2 g 3- - 1.0 ., ,

m o ',-

z ,...

s. 8 , .- -

cr *,

)' .

z2- g \.,,,

0.75 - N.,

CORE INLET FLOW l NORM. POWER l -==- AVE. HEAT FLUX 0 , , , , , 0.5 0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.0 0.5 1.0 1.5 2.0 2.5 3.0 TIME (SEC) , TIME (SEC) g g,

r FIGURE 7.2.7 I

GENERATOR LOAD REJECTION WITHOUT BYPASS, EOFPL18-1000 MWD /ST i

TRANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME I

_ .e.

g

I k

I I

I' I

GLRWOBP, EOFPL-2, MST GLRWOBP, EOFPL-2, MST I

3 OF 5 4 0F 5 l

t 1500 2000

-g-----...........,,,_

I l

1500 -

i i

i i

T.

$\

I:

I

a'  ::  ::

'40-  ! ;!

1000 -

i

l%  !! li

!  !! !\ !i l

E i i  ! ! ! \ '! \.'

i i ! ij it Im5E'300 ~

o

• .i!  !!  !!

500 -  :  ! i: i!

l m $

m i !!i

'l

~

\

$ d  ! *!

I e m o

n. W i i j!;

! w i :

2  ? !

@ 1800 - 3:  : [ 1 2 O -500 - i :

/

l <

w d y b

1000 -

l 1100 -

l I 1500 -

W VALVE (NEG)

---

FEED WATER

)

l STEAM DOME PRES.l '

• VESSEL STM.OUT 5000 , -2000 l

j 0 1 2 3 4 5 6 0 2

3 1 4 5 6 TIME (SEC) TIME (SEC)

R

( FIGURE 7.2.7 (Continued)

GENERATOR LOA _D,_, REJECTION WITHOUT BYPASS. EOFPL18-1000 MWD /ST TR ANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME am3 L .. -__ _ __ _ ______ _ __ __ .- - - - J

l Ol U

I I

I GLRWOBP, EOFPL-2, MST 5 OF 5 2.0 .e*

1.8 i /

g 1.6 1.4 -:  !

f 3 1.2 i 1.0 i  !

l 0.8 :  ;

g

[0.6- / E 5

a 0.4: i 0.2 i ,! ~,,,...* E b o.o

.c... .. ..........

E 2 -0.2 -i b-0.4i b "

a -0.6 i .,

-0.8 !

i.

1.0 : 5 E 3

1.2 -! i

-1.4 -i TOT L 1.6 :

• * *

• DOPPLER i

... .. MODERATOR  !

1.8 i ..... SCRAM i

2.0 , ) ,

l

, , I 0.0 0.5 1.0 1.5 2.0 2.5 3.0 i

TIME (SEC)

FIGURE 7.2.7 I

(Continued) I GENERATOR LOAD REJECTION WITIIOUT BYPASS. EOFPL18-1000 MWD /ST TR ANSIENT RESPONSE VERSUS TIME " MEASURED" SCRAM TIME

.., 3i

=1

1 I

l-i I

i I

i f

I LOFWH, EOFPL-1 LOFWH, EOFPL-1 10F 4 20F4 1.5 1.0 1 0.8 2 l

0.64 0.4 2 1.25 -

0.2 : "- ="" -

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

U.!

f ' ,,

$J -0.2 -' -------- -------*-

d '

IE s8 ' ** * " " ~ ~ ~ ~ ~

8 .0.4 .:

1.0 -

o -0.6 J 68-C' -1.0 -

-0 0.75 -

• 1.4 -

I ---- CORE INLET FLOW NORM. POWER 1.6 -

1,8 .: "

TOTAL

---

DOPPLER MODERATOR

- ~~ AVE. HEAT FLUX - - - SCRAM 0.5 ., ,, , , ,. -2.0 , .

l 0 25 50 75 100 125 150 0 25 50 75 100 125 150 TIME (SEC) TIME (SEC) l l

f I FIGURE 7.3.1 LOSS OF 100 F FEEDWATER HEATING 0.lMITING CASE)

TR ANSIENT RESPONSE VERSUS TIME I

I

(

men

C D

E I

LOFWH, EOFPL-1 LOFWH, EOFPL-1 3OF4 4 0F 4 100 2500 90 -

80 -

F 2250 - "

70 -

O g 60 - 3 3 S 50 - 2000 -

8 w w -

tc g 40 - g ...,,.. ------e % - ......, ,,,,,,,,,,

y 30 -

8 1750 -

20 -

10 -

FEED WATER FLOW l CORE IN SUBCOOUNG l ---- VESSEL STM. OUTLET 0 , , , , , 1500 , ,

0 25 50 75 100 125 150 0 25 50 75 100 125 15 TIME (SEC) TIME (SEC)

FIGURE 7.3.1 I

(Continued)

LOSS OF 100 F FEEDWATER FIEATING RIMITING CASE)

TRANSIENT RESPONSE VERSUS TIME non l

?

I I

I t

l l

l l

I MSIVC, EOFPL, MST MSIVC, EOFPL, MST l 10F 5 2 OF 5

! 6 1.5 I

5-  !

1.25 -

W

~* . .

1 e a 3- g 1.0 - ~~' '.,

w O e, z

• Izz2- 8 g

i

\.

w .

0.75 - '.,

1

• rI l

CORE INLET FLOW '.

l l NORh4 POWER l ---- AVE. HEAT FLUX

• 0 .. . . . 0.5 0 1 2 3 4 5 6 0 1 2 3 4 5 6 TIME (SEC) TIME (SEC) l I FIGURE 7.4.1 MSIV CLOSURE, FLUX SCRAM. EOFPL18 I TR ANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME I

nam

[]

U Il I E

I MSIVC, EOFPL, MST MSIVC, EOFPL, MST 3 OF 5 4 0F 5 1500 2000

---

.g..................,,,,...,'

[

, 1500 - i ~..N.

s 1400 -

).

1000 - k fij[y#i/ A " .. ..- .. ....

w G

• 500 - i  !

@ 1300 - h j ,i 8" 3 i: E.

E M

( A W 0  ?

w 2

@ 1200 - 3:

O -500 -

-1000 -

1100 -

00 -

SR VALVE (NEG)

---

FEED WATER j l STEAM DOME PRES.l l 1000 , , , , , -2000

• "**" VESSEL STM. OUT g

0 1 2 3 4 5 6 0

, , 5 1 2 3 4 5 6 TIME (SEC) TIME (SEC) l FIGURE 7.4.1 I

(Continued)

MSIV CLOSURE, FLUX SCRAM, EOFPL18 i TRANSIENT RESPONSE VERSUS TIME, " MEASURED" SCRAM TIME men l w

%i i

I I

L.

E l r-. . MSIVC, EOFPL, MST SOF5 .

p 2.0 ,

1.8 i f 1.6 i /

t

1.4-!

% .f 1.2-!  !

g 1.0-i ,/ ./

0.8 -! ./

{0.64 ,[

0.4 -i ,***

0.2 i , * .*,

g 0.0 -=v:. ,/

L 3 -0.2 i .,..f......-

O -0 .4 i g 's a -0.6 -! .,

-0.8 i I.,

1.0 i k 1.2 i i "I'4 TOTAL g 1.6 i --= = DOPPLER i

...... MODERATOR I j 1.8 i ..... SCRAM k

-2.0 , ,

0 1 2 3 4 5 6 TIME (SEC) l FIGURE 7.4.1 (Continued)

MSIV CLOSURE. FLUX SCRAM. EOFPL18 TRANSIENT RESPONSE VERSUS TIME. " MEASURED" SCRAM TIME mes , . .

l C

0l I

I I

I 1.35 ' ' ' ' '

1.3 - l 0 No instrument Failurse l I

,. . e I -

g ... . ~

I E , ,. . xx -

1.1 -

NOTE:

I

1. M Intercepts am determned ty tw BSD Channed.

,o O 8 16 24 32 40 48 I

ERROR ROD POSITION I

I I

FIGURE 7.5.1 I

VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEPTS DETERMINED BY THE A AND C CHANNELS I

num ll E l l

I

,I 1

'I l

I il

, .5 . . . . .

, .... i _ _ _ 1 _ ,.. i

^

1.25 -

g g,.2- ,

I- 1.15 -

1 I 1.1 -

NOTES:

l I 1.05 -

[ '

3 107 3 1. ABM Setpoent inoroept 6e merhed 2.

• 8

Rod es empped at noem gotowing two inches of Free Rod Monan.

I to 0

8 16 24 32 40 48 ERROR ROD POSITION I

I I FIGURE 7.5.2 VY CYCLE 18 RWE CASE 1 - SETPOINT INTERCEFTS DETERMINED BY THE B AND D CHANNELS I

.- .n.

y

O E

8.0 DESIGN BASIS ACCIDENT ANALYSIS 8.1 Control Rod Dron Accident Results The control rod sequences are a series of rod withdrawal and banked withdrawal instructions specifically designed to minimize the worths of individual contml rods. The sequences are examined so that, in the event of the uncoupling and subsequent free fall of the rod, the incremental rod worth is acceptable, incremental rod worth refers to the fact that rods beyond Gmup 2 are banked out of the core and can only fall the increment from full-in to the rod drive withdrawal position. Acceptable worth is one which produces a maximum fuel enthalpy less than 280 calories per gram.

Some out-of-sequence control rods could accrue potentially high wonhs. However, the Rod Worth Minimizer (RWM) will prevent withdrawing an out-of-sequence md, if accidentally selected.

The RWM is functionally tested before each startup.

The sequence in the RWM will take the plant from All Rods in (ARI) to well above 20% core thermal power. Above 20% power even multiple operator ermrs will not create a potential md dmp situation above 280 calories per gram [29],[30].[31]. Below 20% power, however, the sequences must be examined for incremental rod worth. This is done thmughout the cycle using the full core, xenon-free SIMULATE-3 model.

Both the A and B sequences were examined at various exposures throughout the cycle. For startup, the rods are grouped, as shown in Figures 8.1.1 and 8.1.2, and are pulled in numerical order.

All the rods in one group are pulled out before the pulling of the next group begins. The rods in the first two groups are individually pulled from full-in to full-out. Beyond Group 2, the rods are banked out using pmcedures[32],[33] which reduce the rod incremental worths.

The potentially high wonhs that occur in pulling the mds in Group 1 are ignored because the reactor is suberitical in Group 1. Therefore, if a rod drops from any configuration in the first group, its excess reactivity contribution to the Rod Drop Accident (RDA) is zero. Successive reloads of axially zoned fuel have extended this subcriticality situation to the second group as well.

ams I I

E

=

t

E

'lhe second gmup of rods was examined using the following aralysis method [34]. Both

( the A and B sequences were examined. It was found that the highest worth rod was the first rod in the second group. Any of the first four rod arrays, shown in Figures 8.1.1 and 8.1.2, may be designated as the first group pulled. However, a specific second group must follow as Table 8.1.1

[

illustrates. For added conservatism, each of the high worth rods in the second group were checked;

{ l.c., one at a time, they were assigned to be the first rod pulled. 'lhis assures that in any sequence the actual worths will always be less than those calculated here.

Only that portion of the control rod worth above the SIMULATE-3 cold critical eigenvalue contributes to the rod drop accident. For conservatism " critical" was defined as the SIMULATE-3 average cold critical Kdt minus 1% AK (reactivity anomaly criteria). The results of the Group 2 calculations, as presented in Table 8.1.2, fit under the bounding analysis of References 29 through 31.

c Beyond Group 2, the rods are banked out of the core. This generally limits the incremental worth of a single rod drop; however, virtually all of the pre-drop cases in Group 3 are critical.

Therefore, the entire dropped sud worth contributes toward the RDA excess reactivity insertion. The P

method used to evaluate Group 3 involved pulling Groups 1 and 2 out and banking Group 3 to  ?

I varying positions. The types of cases examined included:

1. Banked positions 04,08,12, and 48 (full-out).

l H

2. Group 3 rods pulled out of sequence, creating high flux regions.

E

3. Xenon-free conditions, both cold moderator and " standby" (i.e.,1020 psia).

4. Group 3 rods dropping fmm 00 (full-in) to the appropriate banked position.

5. Stuck rods from previously pulled Group 1 or 2 dmpping from 00 to 48.

The highest worth results fmm the Group 3 analysis fit under the Group 2 results, presented in Table 8.1.2.

um n.

t O

8.2 Loss-of-Coolant Accident Analysis ne LOCA analysis, performed in accordance with 10CFR50 Appendix K and the Safety Evaluation Reports [35][36), demonstrates that Vermont Yankee, operating within the assumed conditions, complies with the LOCA limits specified in 10CFR50. 46.

The LOCA analysis for the Reload Cycle is a combination of cycle specific analysis and a base analysis for Cycle 17[37]. Both analyses use the NRC-approved codes, FROSSTEY-2[11]

and RELAP5YA[38]. The base analysis provided the break spectrum and the single failure conditions. De Reload Cycle analysis provided the verification that the base analysis was valid for the Reload Cycle given changes in the reactivity and the UNIX system configuration. All other assumed initial conditions and assumptions are the same for both analyses. Table 8.2.1 lists some of the key input assumptions but Reference 37 provides a more detailed listing of the input assumptions.

De base analysis was perfonned for a combination of break size, break location, and single failure conditions. The break sizes range from 0.05 ft2 to 7.28 ft2 Five break locations were E

5 analyzed: main steam line, core spray line, feedwater line, recirculation loop suction and recirculation loop discharge. Five possible single failures were evaluated: low pressure coolant injection valve, high pressure coolant injection, DC power supply, diesel generator and one automatic depressurization system valve. De impact of the Gd 20 on 3 initial volume average temperature and material properties was included. De PCT results for the limiting break 0.6 ft2 with loss of DC power was 1778.1'F.

The Reload Cycle analysis was performed for the limiting break size and two single failure I

conditions. De PCT results for the limiting break 0.6 ft2w th loss of DC power was 1788.9 F which is a 10.8 F increase in PCT compared to the base analysis results. For the same size break with LPCI Injection valve failure, the PCT for the Reload Cycle was 1770.4*F, an increase of 26.2'F compared to the base analysis. The Reload Cycle analysis also showed that the bn:sak spectrum performed for the base analysis remains valid for the Reload Cycle.

The combined analysis results, in terms of peak cladding temperature (PCF), are shown in Figure 8.2.1. De break spectrum PCT results for Reload Cycle were obtained by increasing the base analysis results by the maximum change in PCT from the Reload Cycle analysis,26.2'F. These a*m E!

=,

I-L L

results show that the limiting break is 0.6 ft2 in the recirculation loop at the pump discharge with one r

L DC power supply as the single failure and loss of effsite power coincident with the break opening.

Overall, the calculated peak clad temperatures are wall below the 2200*F limit of 10CFR 50.46. De

( analysis also shows compliance with the other 10CFR 50.46 limits: total cladding oxidation at the peak location is less than 17%: hydrogen generated in the core is less than 1%; and the core retains a

[ coolable geometry with no clad rupture.

L r During the cycle. Vermont Yankee can adjust the core flow to account for reactivity changes L

rather than using the control rods. During this type of operation, core flow may be as low as 87%

while at 100% power. To ensure the safety analysis bounds these conditions, the LOCA analysis was -

{ ,

analyzed at 1698 MWm power and 87% flow. The results showed that the 100% flow case bounded the low flow case. I De analysis showed that the MAPLIIGR limits are not limited by a LOCA. Therefore, the L MAPLIIGR limits are set based on the thermal-mechanical analysis of the bundle from Reference 18.

They are provided in Appendix A for all the fuel types in the Reload Cycle, as a function of average ~

planar exposure. De analysis also verified that the single loop MAPLliGR multiplier,0.83,is valid for the Reload Cycle.

L 8.3 Refueline Accident Results L

If any assembly is damaged during refueling, then a fraction of the fission product inventory

( could be released to the environment. The source term for the refueling accident is the maximum gap activity within any bundle. He source term includes contributions from both noble gases and iodines.

Re calculation of maximum gap activity is based on the MAPLilGRs, the maximum operating fuel

{

centerline temperatures, and maximum bundle bumup.

The fuel rod gap activity,intemal pressure and centerline temperature for the Reload Cycle are bounded by the values used in Section 14.9 of the FSAR[39).

auw - -

Y

1 O)

O

. TABLE 8.1.1

{

l CONTROL ROD DROP ANALYSIS - ROD ARRAY PULL ORDER l The order in which rod arrays are pulled is specific once the choice of the first group is made. 1 First Group Second Group Successive Group Pulled Is: Pulled Must Be: Is Banked Out Array 1 Array 2 Arrays 3 or 4 Army 2 Array I Arrays 3 or 4 Array 3 Array 4 Arrays I or 2 Array 4 Array 3 Arrays I or 2 I

TABLE 8.1.2 I

VY CYCLE 18 CONTROL ROD DROP ANALYSIS RESULTS I

I Maximum Incmmental Rod Worth Calculated 0.80% AK Cold, Xenon-Fee l

Bounding Analysis Worth for Enthalpy Less than 1.30% AK  !

280 Calories per Gram [29),[30),[31]

I I

aum I I

m

I TABLE 8.2.1 l

LOCA ANALYSIS ASSUMirTIONS  ;

Core lhermal Power (MWm ) 1698.3 I

Total Core Flow (10hbA) 48.0 Reactor Vessel Pressure (psia) 1067.0 Recirculation loop Flow (10hbA) - Each Loop 12.3 )

Feedwater Flow (10hbA) 6.93 1

Feedwater Temperature ( F) 377.0 Water Level Above Top of Enriched Fuel (in.) 130.0 Containment Drywell Pressure (psia) 16.5 Containment Wetwell Pressure (psia) 14.7 l

Containment Wetwell Liquid Temperature (*F) 165.0 Maximum Bundle Power (MWm ) 7.3 Maximum Average Planar Linear Heat Generation Rate (kW/ft)

I Maximum Linear IIcat Generation Rate (kW/ft) 13.6*

14.4*

• l

,I I

• Plus Calorimetric and TIP Reading Uncertainties (8.9%)

Plus Calorimetric and TIP Reading Uncertainties (9.2%)

, _ 78 I

I O>

0 Ii g

l 43 3 39 2 2 3R 4 1 1 g

3 4 3 4 W 31 1 2 2 1 27 3 4 3 4 3 23--- 1 2 1 1 2 1 19 3 4 3 4 3 15 1 2 2 1 11 4 3 4 3 4 07 2 1 1 2 0'

3 e2 ee 1. ,4 ,e m 2e se 34 3. .2 3

I I

I FIGURE 8.1.1 FIRST FOUR ROD ARRAYS PULLED IN THE A SEOUENCES am3 I I

a.

l 1

, I

,g lI 49 3 3 39 2 1 2 l

l 3R 3 4 4 3 31 2 1 2 1 2 27 3 4 3 3 4 3 23-- 1 2 1 2 1 -

19 3 4 3 3 4 3 in 2 1 2 1 2 11 3 4 4 3 07 2 1 2

@ 3 3 02 06 10 14 18 22 26 30 34 38 42 I

I

!I FIGURE 8.1.2 l l'

g FIRST FOUR ROD ARRAYS PULLED IN THE B SEOUENCF .

II I

. .j

I I

l l

2400 . .

Recire. Pioe Low Pressure : Discharoe l Break Location CSCS Credited : Coefficient Appendix K Limit 2200- . .

E . .

Dischwge 2 LPC3 l 10 l

2000 -_u- g,c,g-- upu----j- og----------------

• u_- -

X Dischwge 2 LPCS 0,8 .

@ 1800 - -------------------4--------------- '

o v

A Dischwge 10 1LPCS+1LPCI l 0.) .

3a 1600 ----------------------j---------

m

---

l- .

a.1400 E -------------------- :  :

o . .

H  :  :

1200 ---------------------..-------------------.--------------------

1000 -

---

1.-------------------1.--------------------

i 800 . . . . ....; . . .....; . . . . . . . .

0.01 0.1 1 10 Break Size (ft2)

LOCA ANALYSIS RESULTS. PEAK CLADDING TEMPERATURE VERSUS BREAK SIZE m s3 M M M M M M M M M M M E E E E E E

I 9.0 STARTUP PROGRAM Following refueling and prior to vessel reassembly, fuel assembly position and orientation will be verified and videotaped by undenvater television.

I The Vennont Yankee Startup Program will include process computer data checks, shutdown margin demonstration, in-sequence critical measurement, rod scram tests, power distribution a

comparisons, TIP reproducibility, and TIP symmetry checks. The content of the Startup Test Report will be similar to that sent to the Office of Inspection and Enforcement in the past[40].

I I

I I 7 I

I I

I I  ;

I aess l I i l

I

O a

10.0 CONCLUSION

I!

This report presented the design infonnation, calculational results, and operating limits pertinent to the operation of the Reload Cycle. The core is designed to consist of 120 new GE-9B fuel bundles and 248 irradiated GE-9B fuel bundles. The shutdown margin for the Reload Cycle is greater than the Technical Specification limit of 0.32% AK., The bundles used in the Reload Cycle do not exceed the Technical Specification limit of 1.31 K, for storage in the spent fuel pool or the new fuel storage facility. The transient analysis has: 1) determined the MCPR operating limits so that the FCISL is not violated for the transients considered, 2) assured that the thermal and mechanical overpower limits are not exceeded during the transient, and 3) demonstrated compliance with the ASME vessel code limits. The control rod drop worth is less than the bounding analysis which demonstrates a maximum fuel enthalpy less than the Technical Specification limit of 280 calories per gram. 'Ihe LOCA analysis demonstrates compliance with the acceptance criteria specified in ,

10CFR50.46. The fuel rod gap activity, internal pressure and centerline temperature are bounded by the values used in Section 14.9 of the FSAR which demonstrates the limits of 10CFR100 are not exceeded for a refueling accident.

I I

I I

I I

I am B.

lI l

REFERENCES

1. K. J. Morrissey, Vermont Yankee Cycle 16 Summary Report, YAEC-1878 (April 1994).

2. General Electric Company, General Electric Standard Anolication for Reactor Fuel (GESTARil). NEDE-24011.P-A-10, GE Company Proprietary, Febmary 1991, as amended.

3. letter, D. T. Weiss to J. M. Buchheit, "GE9B Bundle Nuclear Design Reports for Reload 14,"

DTW89168, October 6,1989.

l. 4. Letter, D. T. Weiss to R. T. Yee, " Fuel Bundle Nuclear Design Reports for Vermont Yankee Reload 16," I7TW92194, September 3,1992.

5. A. S. DiGiovine, J. P. Gorski, and M. A. Tremblay; SIMULATE-3 Validation and Verification: YAEC-1659-A (September 1988). l

6. R. A. Wochlke, et al.; MICBURN 3/CASMO-3/ FABLES-3/ SIMULATE-3 Benchmarking of  ;

l Vermont Yankee Cycles 9 through 13: YAEC-1683-A (March 1989).

7. B. Y. Ilubbard, et al.; End-of-Full. Power-Life Sensitivity Study for the Revised BWR_

Licensing Methodology: YAEC-1822 (October 1991).

l

8. VYNPC Letter to USNRC, "Vennont Yankee LOCA Analysis Method: FROSSTEY Fuel Perfonnance Code (FROSSTEY-2)," FVY 87-116 (December 16, 1987). {

J lg 9. VYNPC Letter to USNRC, " Responses to Request for Additional Infonnation - FROSSTEY-2 j Fuel Performance Code," BVY 91024 (March 6.1991).

lg 10. VYNPC Ixtter to USNRC, "FROSSTEY-2 Fuel Performance Code - Vermont Yankee j Response to Remaining Concems," BVY 92-54 (May 15,1992).

i

11. USNRC Letter to L. A. Tremblay, SER, " Vermont Yankee Nuclear Power Station, Safety Evaluation of FROSSTEY-2 Computer Code (TAC No. M68216)" NVY 92-178 (September 24, 1992).

i 12. Appendix A to Operating License DPR-28 Technical Specifications and Bases for Vermont

) Yankee Nuclear Power Station, Docket No. 50-271.

j 13. VYNPC Letter to USNRC, " Inverted Control Rod Poison Tubes at Vennont Yankee," WVY 3 75-51 (May 16,1975).

14. USNRC Letter to G. C. Andognini, " Change to Bases," (June 6,1975).

! 15. A. S. DiGiovine, et al.; CASMO-3 Validation: YAEC-1363-A (April 1988).

16. A. A. F. Ansari, Methods for the Analysis of Boiline Water Reactors: Steady-State Core Flow Distribution Code (FIBWR), YAEC-1234 (December 1980).

i I

!I

_ .s4 lI I

r C

E REFERENCES

17. A. A. F. Ansari, R. R. Gay, and B. J. Gitnick; FIBWR: A Steady-State Core Flow Distribution g Code for Boiling Water Reactors - Code Verification and Oualification Report: EPRI NP-1923; Project 1754-1 Final Report, July 1981.

g

18. USNRC Letter to J. B. Sinclair, SER, " Acceptance for Referencing in Licensing Actions for the Vermont Yankee Plant of Reports: YAEC-1232. YAEC-1238, YAEC-1299P, and YAEC-1234," NVY 82-157 (September 15, 1982).

19. General Electric Company, GEXL-Plus Correlation Application to BWR 2-6 Reactors GE6 through GE9 Fuel. NEDE-31598P, GE Company Proprietary, July 1989.

20. Letter, D. T. Weiss to R. T. Yee, " Mechanical MAPLHGRs for Vermont Yankee Reload 16,"

DTW93136, June 3,1993.

21. Letter, D. T. Weiss to R. T. Yec, " Fuel Rod Dennal-Mechanical Performance Limits,"

DTW92260, November 19,1992.

22. A. A. F. Ansari and J. T. Cronin, Methods for the Analysis of Boiling Water Reactors: A Systems Transient Analysis Model (RETRAN). YAEC-1233,(April .1981).

23. USNRC Leuer to R. L. Smith, SER, " Amendment No. 70 to Facility License No. DPR-28,"

(November 27,1981).

24. V. Chandola, M. P. LeFrarcois, and J. D. Robichaud; Application of One-Dimensional Kinetics to Boiling Water Reactor Transient Analysis Methods: YAEC-1693-A, Revision 1 l

5 (November 1989).

25. Electric Power Research Institute, RETRAN - A Program for One-Dimensional Transient Dennal ilydraulic Analysir of Complex Fluid Flow Systems. CCM-5, December 1978.

26. USNRC Letter to T. t'!. Schnatz, SER, " Acceptance for Referencing of Licensing Topical Reports: EPRI CCM-5 and EPRI NP-1850-CCM," (September 4,1984).

27. A. A. F. Ansari, K. J. Burns, ar.d D. K. Beller; Methods for the Analysis of Boiling Water Reactors: Transient Critical Power Ratio Analysis (RETR AN-TCPYA01): YAEC-1299P (Maren 1982).

28. J. T. Cronin, Method for Generation of One-Dimensional Kinetics Data for RETRAN-02.

YAEC-1694-A (June 1989).

29. General Electric Company; C. J. Paone, et al.; Rod Dmp Accident Analysis for Large Boiling Water Reactors NEDO-10527; March 1972.

30. General Electric Company; R. C Stim. et al.; Rod Dron Accident Analysis for Large Boiling Water Reactors Addendum No.1. Multiple Enrichment Core? ;h Axial Gadolinium: NEDO-10527, Supplement 1; July 1972.

m.,a Ii 1

s .

s REFEREN S 31.' General Electric Company; R. C. Stim, et al.; Rod Dmo Accident Analysis for Larne Boiline :

Water Reactre Addendum No. 2 Exoosed Cores: NEDO-10527, Supplement 2; January 1973.

32. General Electric Company, C. J. Paine, Banked Position Withdrawal h=w, NEDO-21231 January 1977, h 33. General Electric Company, D. Radcliffe and R. E. Bates, Reduced Notch Worth Pn mdure.

SIL-316, November 1979.

34. M. A. Simnen, Vermont Yankee Cycle 14 Core Performance Analysis Resort, YAEC-1706 -

[~

(October 1988).

=

35. USNRC Letter to L. A. Tremblay, SER, " Safety Evaluation for Vennont Yankee Nuclear Power Station RELAPSYA LOCA Analysis Methodology (TAC No. M74595)," NVY 92-192 (October 21,1992).

36. USNRC Letter to R. W. Capstick, SER, "Appmval of Use of 'Ihermal-Hydraulic Code RELAPSYA (TAC No. 60193)," NVY 87-136 (August 25,1987).

37. L. Schor, et al.; Vermont Yankee Loss-of-Coolant Accident Analysis; YAEC-1772 (June 1993). _

38. Report, RELAP5YA. A Computer Program for Light-Water Reactor System Thermal-Hydraulic Analysis. YAEC-1300P-A, Revision 0, October 1982; Revision 1, July '

1993. -<

b 39. Vennont Yankee Nuclear Power Station Final Safety Analysis Report, December 1991.

p 40. Letter from L. A. Tremblay to USNRC, " Cycle 17 Startup Test Report," BVY 94-03 (January -

L- 13, 1994).

r L

aam [

r 0

0 APPENDIX A CALCULATED OPERATING LIMITS Vm MCPR operating limits for the Reload Cycle are calculated by adding the calculated I

ACPR to the FCISL. This is done for each of the analyses in Section 7.0 at each of the exposure statepoints. For an exposure interval between statepoints, the highest MCPil limit at either end is assumed to apply to the whole interval.

Table A.1 provides the highest calculated MCPR limits for the Reload Cycle for each of the exposure intervals for the various scram speeds and for the various rod block lines. Timse MCPR operating limits are valid for operation of the Reload Cycle at full power up to 10644 mwd /St and for operatic during coastdown beyond EOFPL.

Tables A.2 through A.5 provide the maximum calculated MAPLIIGR limits for all the fuel types in the Reload Cycle. These values are bounding for all the lattice zones in each fuc! type.

I I

I I'

I Ii nam 87-al

=

1 TABLE A.1 VERMONT YANKEE NUCLEAR POWER STATION CYCLE 18 MCPR OPERATING LIMITS Value of "N" in Averare Control Rod Cycle Exposure Range MCPR Operating I RBM Ecuation i Scram Time Limit 2J Equal to or 0.0 to 4000 MWNSt 139 I 42% better than L.C.O. 3.3.C.I.1 4000 to 5500 MWNSt 5500 to 10644 MWNSt 135 133 Equal to or u.u to 4uuu Mwwst av hW better than L.C.O. 3.3.C.I.2 4000 to 5500 MWNSt 135 5500 to 9035 MWNSt 133 9035 to 10644 MWNSt 134 l W 41%

Equal to or better than u.u to 4uuu Mwwst IJy 4000 to 5500 MWNSt 135 i

L.C.O. 3.3.C.I.1 5500 to 6500 MWNSt 1.29 6500 to 9035 MWNSt 1.27 m 9035 to 10644 MWNSt 132 Equal to or u.u to 4uuU MWwst IJy g better than 4000 to 5500 MWNSt 135 Im L.C.O. 33.C.I.2 5500 to 6500 MWNSt 1.29 6500 to 8035 MWNSt 1.27 g 8035 to 9035 MWN3t 130 3 9035 to 10644 MWNSt 134 l Equal to or u.o to 4uuu Mwwst lav 540% better than 4000 to 5500 MWNSt 135

'I L.C.O. 33.C.l.1 5500 to 6500 MWNSt 1.29 6500 to 8035 MWNSt 1.25 8035 to 9035 MWNSt 1.27 I 9035 to 10644 MWNSt 132 l Equal to or u.u to 4uuu Mwwst lay better than 4000 to 5500 MWNSt 135

L.C.O. 3.3.C.I.2 5500 to 6500 MWNSt 1.29 6500 to 8035 MWNSt 1.25 8035 to 9035 MWNSt 130 9035 to 10635 MWNSt 134 i

(1) The Rod Block Monitor (RBM) trip setpoints are determined by the equation showTi in Table 3.2.5 of the Technical Specifications.  !

I 1

(2) The current analysis for the MCPR operating limits does not include the 7X7,8X8,8X8R or P8X8R fuel types. On this basis, if any of these fuel types are to be reinserted, they will be evaluated in accordance with 10CFR50.59 to ensure that the above limits are bounding for these fuel types.

(3) MCPR operating limits should be increased by 0.01 for the single loop operation.

a,m l l

1

r C

0 TABLE A.2 MAPLIIGR VERSUS AVERAGE PLAhAR EXPOSURE FOR BP8DWB311-10GZ Plant: Vermont Yankee Fuel Type: BP8DWB311-10GZ I

Averare Planar Exposure MAPLIIGR Limits (kW/ft)

(mwd /St) Two-laxin Operation SinRie-Imn Operation

• 0.00 10.93 9.07 200.00 11.00 9.13 1,000.00 11.13 9.35 2,000.00 11.32 9.40 3,000.00 11.52 9.56 4,000.00 11.64 9.66 5,000.00 11.77 9.77 6,000.00 11.92 9.89 7,000.00 12.I1 10.05 8,000.00 12.34 10.24 9,000.00 12.59 10.45 10,000.00 12.83 10.65 12,500.00 13.00 10.79 15,000.00 12.81 10.63 20,000.00 12.24 10.16 25,000.00 11.55 9.59 35,000.00 10.24 8.50 45,000.00 8.76 7.27 51,299.00 5.87 4.87 I

MAPLilGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLIIGR limits by 0.83. '

am3  !

a .

3 I

TABLE A.3 MAPLIIGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB311-ilGZ f

Plant: Vermont Yankee Fuel Type: BP8DWB311-11GZ I

Average Planar Expowre MAPLHGR Limits (kW/ft)

(mwd /St) Two-Loon Oncration Single-Imo Operation

• 0.00 10.93 9.07 200.00 11.00 9.13 ,

1,000.00 11.13 9.24 2,000.00 11.32 9.40 3,000.00 11.52 9.56  ;

4,000.00 11.64 9.66 I

5,000.00 11.77 9.77 6,000.00 11.92 9.89 7,000.00 I 8,000.00 12.I1 12.34 10.05 10.24 9,000.00 12.59 10.45 10,000.00 12.83 10.65 12,500.00 13.00 10.79 15,000.00 12.81 10.63 20,000.00 12.24 10.16 25,000.00 11.55 9.59 35,000.00 10.24 8.50 45,000.00 8.76 7.27 51,466.00 5.83 4.84 I

• MAPLHGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLIIGR limits by 0.83.

l I

(- T E

o; l

TABLE A.4  ;

l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-10GZ  !

Plant: Vermont Yankee Fuel Type: BP8DWB335-10GZ Avenne Planar Exposure MAPLHGR Limits (kW/ft)

(mwd /St) Two-Loon Operation Single-Imo Operation

• 0.00 11.29 9.37 200.00 11.34 9.41 1,000.00 11.48 9.53 2,000.00 11.69 9.70 3,000.00 11.92 9.89 4,000.00 12.17 10.10 5,000.00 12.43 10.32 g

6,000.00 12.68 10.52 7,000.00 12.87 10.68 8,000.00 13.05 10.84 9,000.00 13.24 10.99 '

10,000.00 13.35 11.08 12,500.00 13.20 10.% l 15,000.00 13.01 10.80 20,000.00 12.27 10.I8 25,000.00 11.43 9.49 35,000.00 9.88 8.20 45,000.00 8.38 6.96 50.593.00 5.65 4.69 MAPLliGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLilGR limits by 0.83.

ams I

3 I

TABLE A.5 MAPLliGR VERSUS AVERAGE PLANAR EXPOSURE FOR BP8DWB335-11GZ Plant: Vermont Yankee Fuel Type: BP8DWB335-1IGZ I Average Planar Exposure MAPLHGR Limits (kW/ft)

(mwd /St) Two-Loon Operation Single-Loco Operation

• 0.00 11.28 9.36 200.00 11.33 9.40 1,000.00 11.43 9.49 I 2,000.00 3,000.00 11.60 11.80 9.63 9.79 I

4,000.00 12.04 9.99 5,000.00 12.30 10.21 6,000.00 12.53 10.40 7,000.00 12.73 10.57 8,000.00 12.94 10.74 9,000.00 13.13 10.90 10,000.00 13.29 11.03 12,500.00 13.20 10.%

15,000.00 12.99 10.78 20,000.00 12.27 10.18 25,000.00 11.43 9.49 35,000.00 9.88 8.20 45,000.00 8.38 6.96 50,593.00 5.65 4.69 I

• MAPLIIGR limits for single-loop operation are obtained by multiplying the two-loop operation MAPLliGR limits by 0.83.

num I