Text

Pressure & Temp Limits Rept
	ML20141K018
Person / Time
Site:	Byron
Issue date:	05/21/1997
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20141J982	List: ML20141J978; ML20141K001; ML20141K018; ML20141K024; ML20141K037; ML20141K043;
References
	NUDOCS 9705280286
	Download: ML20141K018 (23)
	v • d • e

. . _ . - .

-)

, J l

r-I ATTACHMENT F l l

t i

BYRON UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT O

l f

L 4

i 9705290286 970521 PDR ADOCK 05000454 P

PDR ..

l

. _ . _ _ . _ . - . . _ .. _ _ m . . - _ . . . m BYRON - UNIT 1 l

PRESSURE AND TEMPERATURE LIMITS REPORT I 4

Table of Contents Section Page 1.0 Introduction 1 2.0 Operating Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits 1 3 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints 2 2.3 LTOP Enable Temperature 2 2.4 Reactor Vessel Boltup Temperature 3 2.5 Reactor Vessel Minimum Pressurization Temperature 3 3.0 Reactor Vessel Material Surveillance Program 9 4.0 Supplemental Data Tables 11 ;

t 5.0 References 18 Attachment WCAP-14824, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwodd," Revision 1, April 1997.

4 i

i l

i 4

BYRON - UNIT 1 o

PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4' 100 F/hr) Applicable for the First 12 EFPY (Without Margins for Instmmentation Errors) 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 5 100 F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) 2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low 7 Temperature Overpressure Protection (LTOP) System Applicable for the First 12 EFPY

)

I ji .

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT

List of Tables Table Page 2.1 Byron Unit 1 Heatup and Cooldown Data Points at 12 EFPY 6 (Without Margins for Instrumentation Errors) 2.2 Data Points for Byron Unit 1 Maximum Allowable PORV Setpoints for the 8 LTOP System Applicable for the First 12 EFPY 3.1 Byron Unit 1 Capsule Withdrawal Schedule 10 4.1 Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data 12 4.2 Byron Unit 1 Reactor Vessel Material Properties 13 4.3 Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTS) at the 1/4T 14 and 3/4T Locations for 12 EFPY 4.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures (ARTS) at 15 12 EFPY at the Limiting Reactor Vessel Material Intermediate Shell Forging SP-5933 (Based on Surveillance Capsule Data) 4.5 RTns Values for Byron Unit I for 32 EFPY 16 4.6 RTns Values for Byron Unit I for 48 EFPY 17 iii

.-~. - . - - . . - - - _ - . . _ - __ - .- - - - . . - - .

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction Reference to Technical Specifications (TS) numbers are given in both the Byron Station current Technical Specifications (CTS). and Improved Technical Specifications (ITS). The CTS number is presented first, followed by the ITS number in brackets [ ]. i l

i This PTLR for Unit I has been prepared in accordance with ti.: requirements of !

TS 6.9.1.11/[ITS-5.6.6]. Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this repon are listed below:

LCO 3.4.9.1 Pressure / Temperature Limits; and LCO 3.4.9.3 Overpressure Protection Systems.

l

[ITS-LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System]. !

2.0 Operating Limits The PTLR limits were developed using a methodology specified in the Technical Specifications.

The methodology listed in WCAP-14040-NP-A (Reference 1) was used with four exceptions:

a. Use of ENDF/B-IV neutron transport cross-section library and ENDF/B-V dosimeter I reaction cross-sections,

b. Optional use of ASME Code Section XI, Appendix G, Article G-2000,1996 Addenda,

c. Use of ASME Code Case N-514, and

d. Use of RELAP computer code for calculation of LTOP setpoints for Unit I replacement steam generators.

WCAP-14824, Revision 1 is included as an attachment for reference. WCAP-14824 contains the P/T curves for Byron Unit 1, along with the weld metal data integration for Byron and Braidwood Units 1 and 2 and the Byron /Braidwood fluence methodologyjustification for ,

ENDF/B-VI cross sections. '

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.9.1/ [ITS-LCO 3.4.3]) ,

i 2.1.1 The RCS temperature rate-of-change limits defined in Reference 2 are:

a. A maximum heatup of100 F in any 1-hour period, !

b. A maximum cooldown of 100 F in any 1-hour period, and

c. A maximum temperature change ofless than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and i i

cooldown limit curves.

f.

l l

BYRON - UNIT 1 i PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (Conticued) 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are {

specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in i Figure 2.2 and Table 2.1. These limits are defined in Reference 2. Consistent with the l

methodology described in Reference 1, the RCS P/T limits for heatup and cooldown !

shown in Figures 2.1 and 2.2 are provided without margins for instrument error. In l

determining compliance with Figures 2.1 and 2.2 and Table 2.1, instrument uncertainties !

need not be considered since appropriate station operating procedures ensure that the limits contained in the figures and table are not exceeded. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G. l l

The P/T limits for core operation (except for low power physics testing) are that the ;

reactor vessel must be at a temperature equal to or higher than the minimum temperature !

required for the inservice hydrostatic test, and at least 40 F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.9.3/

[ITS-LCO 3.4.12]).

The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 2.3 and Table 2.2. These limits are based on References 5,13, I and 14.

The LTOP setpoints are based on the P/T limits established by 10 CFR 50, Appendix G l without allowance for instrumentation error in accordance with the methodology l described in Reference 1. The LTOP PORV maximum lift settings shown in Figure 2.3 and Table 2.3 account for appropriate instmment error.

2.3 LTOP Enable Temperature The as-analyzed LTOP enable temperature is 210 F (Reference 5).

The TS required enable temperature for the PORVs shall be ;t 350 F RCS temperature (Byron Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350 F and below and disarming of LTOP for RCS temperature above 350 F).

Note that the last LTOP PORV segment in Table 2.2 extends to 450 F where the pressure setpoint is 2350 psig This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

2

.. . ~ -- -- .. - _ - - -

BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT -

l 2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification) !

t I

The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60 F. Boltup l is a condition in which the Reactor Vessel head is installed with tension applied to any l stud, and with the RCS vented to atmosphere (Reference 2).

l l

2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification) l The minimum temperature at which the Reactor Vessel may be pressurized (i.e., in an unvented condition) shall be 2 60 F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994. 3 4

4 a

l d

f f

r i

3 1

BYRON - UNIT 1 h L5SURE AND TEMPERATURE LIMITS REPORT RATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORG't4G SP-5933 (using surv. capsuie cLata)

LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60T 2500 , 6 i ,

~ :Tessessor:*t! 'l 'l I. ) !I --

!I - '

. / i' f_

. il I- I. .

2250 m .

.i ,. i, . , i l' '

m i LEAK TEST LIMIT CI 'li li' i ' ' ' ' ' ' '

ii it . In , e i ,

tno . i i Ili I i'! i i if . ,f .j . .

2000 , i .i . .

cn ,

e

, i . 'fi i i ! . . i e '

i i i '

. . .i ,i i

/i ii / i i i . .i i i, ii ! ie i i i , i ! i , ! ii J i i i e i i i! i i ! i ! i y , , ,, i i :/ .

i fi . i i i . i 6 i , e .

1750 ! UNACCEPTABLE t !/ i ! !/4 ' i i i ' ' i ' ! i ! '

OPERATION ' ' / i ' Vi s ' ' ' i i

I ' '

c , . I i ! Ii i . i i ! I i .

i i i i i , . . i i /. i i/ ! ! ' s ! i ! I i i '

b i i +i i z 1500' e ii i i i e e i i i/ i i If I i i i i i i i i i ' i e i e ' I

, ,i , , , ;

.f. y , i , i . , , i . i ,

, , , , i , . i ,

EO ie '

i

~

i ,i i i , i ! i i , iR! /> i i i 6 i i ! I i i i e i i i i 6 i i , ei i Vl/ l i

/4 I I ie i i e i ie i i cc i i ii /t 1! . !I i i i i t 6 i i i i l i !

, ;; EEATUP RATI f i f, , , 6 i i 1 ca 1250 l

, , ,, or To aao r/Rr. ;, , , , ,f,i ,,, , ,ce,,,,,t, , , , I w . ii /I t i; i iii OPERATION I et i ! i i ii ! I i i i i i e i /i ! ! !/ i ie i I !

1 , a i . . . e i i i i i e /i . i / i i e i e i i i i i i e 6 i i e e i e i i i e i iii i i i i e e i e i i i i i i i i i i i 1000 , , , , , , , , , , , , ,f i i t/s

, ,, , , , 6 , , , , , I , , , , , i , i ,

': , , , i i , i i , , ii /! , vi i !ii! i i i ! ! i i i ! ! i gj i ! ii } t ! ! i e i i i if ! I i 8 i i i i i i ! I i i I I i i ,e i i i i i i i i ! ! / ! t i i i e i t i i i ii i ! i

! i !

i i i i i i e e i i e a i i 's i . . I i i i i e6 i i i 6 i i i a e i gg i . i i e i i i < ' e i i . i i ! i ! i ! ! ' I i i ! ! 6 , e i i ! ! ! ! I i '

! I e i I ! i I i i II ! i i i I II U i a i ! i 1 i i i i i i i i i i i i ! l

. 4 i i i i i i e i i i i ie i i e

! ! i !i ei i i ! I i i i . ! I i i i i i i i i i i i i i + i e i is i i 6 4 e i i ei i i i 500 i

-e , , , , , , , , , , , , , , , , , , , , , , , , , , , , , , ,

g i e i ei i i i i e i i i tit i i i i . i l i i .i l I 6 i i ei i l ! Ni i ! l ! i ie i i i ! l 250 ,i,,

l % MINIMUM BOLTUP TEMP. AT WF I!') l' , ! !l, , !!l f l i ' '

CBITICALITT LIMIT SASED ON INBERVICS ETBROSTAPIC TEST I ei TEMPBBATDBR sRETIC5 PERI 3 SOS rp'OFOR TER i! ! i i a i i e i i i ,,,

^ ~

i ;

i i ; ~

j j j UP &S.8 BFFT i i i i i i i > T i , i i , a6 0 # e i #

0 50 100' 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F)

Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 F/hr) applicable for the First 12 EFPY (Without Margins for Instmmentation Errors) 4 ew --t

4 BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using swv. capsuse data)

LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60'F 2500 i ,

_ esssassessesel n .I, .

_, i , y . '

. 6 i ** i . I

i - -

l:

, me i / -

2000 , ,

,,, ,; , ,, , /,,, . , , . . .. , . . . ,

U2 i! i +ii6 ' l' ii i 11ii' i '

i i 'A !J ' I6 '

l, ',

~

, , . i /i. . . . . t' e i e i ,

& , , , , i, , e i, . .i , i i, e .i 6a - a i ii ! !t t i '

w , i : j , i . i . , i 4 , . eii i i .

,750 i

unaccurTAnts ' / i - - -

t - - - > -

~~

0PERATION ' / i

i ! ' i i 't i i ! '

o i j ,.

i i i i e i , . . i i ,ii 6

i , i i e /t i6 6 ! ' i i It 6 e i lii e i e t

! 8

, e 1500 ,

i, , ', . i

,' , , i . . '.

, .i ! i 66t '

i t !l

. i e i i ! i t t i e y) l t ! l4 l , t I '

i I I

e e ii i 6 i i 6 6 i i e 1250 ,, , , ,,, , , , , i acca,Tants .

6 l' ! t I '

l' !! OPERATION t

, . , tiii ,i a6. ,

e i

- iii i . ' i

, . a i i l i i !*

1000

, , i i, ' ' ' .'

i . . '. i . .

. i; . # ,

l l l l l 1 i' i ! ! ' It i i ! I t i g i 6 e !I ' i it

t i i i l C00LDOWN ,,;. , , ,

i s i . , , , , . , ; , i RATBB i ; , , , , , . , i i , ii ,iii ; i j, r/ar-

, 750 -: , , i i , , .

, . . , i ,,

, i i.i Ii! !

i i i i i i i

> O l l

._ 7vi . ., , i . ,

e 500 - -

,g: / '

': : l l c '

t i

, i ,

i I i 6I I

~

, , w . c i > i i s i 250 i ! i N1 i ! t t t

' X N BOLTUP '

i i '

I i

i TEMP. AT wF I i

6 i 9 'W8-88 i 0

0 50 100 150 200 250 300 350 400 450 500 4

Indicated Temperature (Deg,F)

Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) 5

y 1

BYRON - UNIT 1 i .

] PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit 1 Heatup and Cooldown* Data Points at 12 EFPY" (Without Margins for Instrumentation Errors)

Cooisown Curves Hestup Curve

Steesy State 25F SOF 100F 100F Cstlicainy. Lirrut Leek Test Lirrut

T P T P T P T _P T P T P T P l 80 821 to 585 50 554 80 470 to 821 203 0 182 2000 j 86 821 86 810 65 570 85 489 85 821 203 0 203 2485 j 70 821 70 821 70 587 70 500 70 821 203 0 75 821 75 821 75 805 75 531 75 821 203 0 {

3 80 821 80 821 80 821 80 554 80 821 203 871

86 821 M 821 86 821 85 579 86 8 21 203 867

, 90 821 00 821 90 821 90 807 90 821 203 845 1 95 821 95 821 95 821 95 821 85 8L*1 203 830 100 821 100 821 100 821 100 821 100 821 203 834 I 105 821 105 821 106 821 106 821 105 821 203 832 l

) 110 821 110 821 110 821 110 821 110 821 203 833 _ i j 115 821 115 821 115 821 115 821 118 821 203 857 i

120 821 120 821 120 821 120 821 120 821 203 842

125 021 126 821 125 821 125 821 128 821 203 851

] 130 821 130 821 130 821 130 821 130 821 203 881 1 135 821 135 821 135 821 135 821 138 8 21 203 874

! 140 821 140 821 140 821 140 821 140 821 203 SW l

) 145 821 145 821 146 821 145 821 203 707 l 100 821 150 821 ~ ISO 821 203 727

} 155 821 i 185 821 203 740

. 100 821 ISO 821 203 774 185 821 185 8 21 205 801

, 170 821 170 821 210 831 1 175 821 175 821 215 854 i i

180 821 100 821 220 000 1 100 1483 180 900 225 935

] 185 1550 106 938 230 000 -

l 100 1640 190 900 235 1938 I 195 ' 1728 105 1035 240 1975 200 1821 300 1975 345 1138 205 1921 205 1128 230 1188 210 2019 210 1188 285 1847 215 2143 til 1947 200 1313 230 2300 220 1813 205 1305 225 2307 225 1306 270 1481 230 1401 275 1543 235 1543 30 1830 240 1830 205 1734 345 1734 230 1835 200 155 205 1833 255 1833 300 2048 300 2048 305 2171 265 2171 310 2302 270 2302 315 2441 .

275 2441 Heatup and cooldown data include vessel flange requirernents of 180 F and 621 psig per 10 CFR 50, Appendix G.

- For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.

6

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 650 ; , ; ;

1 l

UNKCCEPTABEE - ' '

? "" Ii l !_

flilli.

h PCX-45.6 il l

=/[

l/rf I

I* I I

[;,

acx A ss e i i /

h i. ^

[

[500 1!1 iib !j

! .1. ! . .

/

2 l

! //

k ; j7jj ACCEPTAB LE--

450

j_,,g-i ; ; ;

_ j j iii i i i l ! ! ! I i I ji j

! ! i l i i ! l l I l i l Il .l l l l ;

400 , ! ! I I' II I ! I O 50' 100 150 200 250 300 350 AUCTIONEERED LOW RCS TEMPERATURE (DEG.F)

Figure 2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 12 EFPY 7

l BYRON - UNIT 1 l

PRESS 11RE AND TEMPERATURE LIMITS REPORT l Table 2.2 1 Data Points for Byron Unit 1 Maximum Allowable PORV Setpoints i

for :he LTOP System Applicable for the First 12 EFPY j i

PCV-455A PCV-456 (1TY-0413M) (1TY-0413P)

AUCTIONEERED RCS AUCTIONEERED RCS LOW PRESSURE LOW PRESSURE RCS TEMP. (PSIG) RCS TEMP. (PSIG)

(DEG. Fl (DEG.F) l 50 497 50 514 70 497 70 514 100 497 100 514 ;

120 446 120 463 !

150 446 150 463 200 446 200 463 250 587 250 604 300 587 300 604 350 587 350 604 450 2350 450 2350 i

Note: To determine maximum allowable lift setpoints for RCS Pressure and RCS Temperatures greater than 350 F, linearly interpolate between the 350 F and 450 F data points shown above.

8

BYRON - UNIT 1 a

PRESSURE AND TEMPERATURE LIMITS REPORT -

3.0 Reactor Vessel Material Surveillance Prograrn The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The removal schedule is provided in Table 3.1. Also, the results of these analyses shall be used to update Figures 2.1 and 2.2 and Table 2.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor 4

shutdowns most closely approaching the withdrawal schedule.

The pressure vessel material surveillance program (Reference 6) is in compliance with l

Appendix H to 10 CFR 50, " Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTer, !

which is determined in accordance with ASTM E208. The empirical relationship between RTm3 and the fracture toughness of the reactor vessel steel is developed in accordance with ;

Appendix G," Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and '

Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

I I

1 I

t 1

9

BYRON - UNIT 1

. PRESSURE AND TEMPERATURE LIMITS REPORT l

l Table 3.1 l

1 i

Byron Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time

Estimated Capsule (Degrees) Factor (EFPY) 2 !

Fluence (n/cm )

U 58.5 3.85 1.15 (Removed) 3.72 x 10 8

i I

X 238.5 3.79 5.64 (Removed) 1.39 x 10

W 121.5 3.79 8.44 (EOL Wall) 2.159 x 10*) ,

12.66(1.5 EOL l Z 301.5 3.79 Wall")) 3.238 x 10

J V 61.0 3.59 Standby ---

~

4 Y 241.0 3.59 Standby ---

1 j

(a) EfTective Full Power Years (EFPY) from plant startup.

I (b) Maximum end oflicense (32 EFPY) inner vessel wall fluence.

(c) Derived from Table C-1 of WCAP-14824, (Reference 2, which is the Attachment to this report).

i, 2

i l .

l, I

i I

,. I J

4 l

10 4

( ~

r -

~ . - . - - - . . - . - - . - . - . _- -- ..

. l l

BYRON - UNIT 1 l

PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Supplemental Data Tables t The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values i i shown were used as inputs to the Pfr limits.

Table 4.1 shows the calculation of the suiveillance material chemistry factors using surveillance i capsule data. '

Table 4.2 provides the reactor vessel material properties table.

Table 4.3 provides a summary of the Byron Unit I adjusted reference temperature (ARTS) at the 1/4T and 3/4T locations for 12 EFPY. j i

Table 4.4 shows the calculation of ARTS at 12 EFPY for the limiting Byron Unit I reactor vessel material (Intermediate Shell Forging 5P-5933).

l Table 4.5 provides RTns values for Byron Unit I for 32 EFPY.

l Table 4.6 provides RTns values for Byron Unit 1 for 48 EFPY. ;

)

4 11

BYRON - UNIT 1 ,

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 l

Calculation of Chemistry Factors Using Surveillance Capsule Data l Fluence Material Capsule (n/cm2 , pp(.) Measured FF*ARTer (FF)2 E> l .0 ARTer ;

Mev), f l Inter. Shell 8

Forging 5P-5933 U 3.72x10 0.727 0 0 0.529 (Tangential)

X 1.39x10 l.091 30 32.73 1.190 Inter. Shell Forging SP-5933 U 3.72x10'8 0.727 0 0 0.529 (Axial)

X 1.39x10 l.091 30 32.73 1.190 Sum: 65.46 3.44 Chemistnr Factor (d) = 65.46 + 3.44 = 19.0*F Byron 1 Weld Metal WF336(M U 3.72x 10 O.727 0 0 0.00 0.529 X l .39x10 l.091 35 105(*) 114.56 1.190 Byron 2 Weld Metal WF447(" U 3.996x10" 0.746 0 0 0 0.557 W l.211x10 l.053 30 90) 94.77 1.110 Sum: 209.33 3.386 Chemistry Factor (d) = 209.33+3.386= 61.8'F (a) FF = Fluence Factor = fm2:4 i ios o (b) Byron Unit 1 ARTmr aluesv were obtained from the surveillance Capsule X analysis, WCAP-13880 (Reference 3). The Byron Unit I capsule fluence values were recalculated using the ENDF/B-V scattering cross sections in 1994 and are documented in WCAP-14044 (Reference 8).

(c) Byron Unit 2 capsule fluence, FF, and ARTer values were obtained from the surveillance Capsule W analysis (WCAP-14064 (Reference 10)) using the ENDF/B-V scattering cross-sections.

(d) Chemistry Factor = E (FF*ARTer) / I ((FF)2)

(e) Adjusted ARTmr per Ratio Procedure of 10 CFR 50.61 (Reference 11). Ratio = 3.0. See Table 2 of WCAP 14824, (Reference 2).

12

-~

BYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 >

Byron Unit 1 Reactor Vessel Material Properties !

Chemistry Initial J Material Cu (%)(') Ni (%)(') Factor) RT mr ( F)'M ,

Description Closure Head --

0.74 --

w Flange

]

Vessel Flange --

0.73 --

10) l Intermediate Shell 0.0364 0.747 23.8 40 l Forging SP-5933 Lower Shell 0.04 0.64 26.0 10 Forging 5P-5951 Circumferential 0.06 0.61 82.0 -30 Weld WF336 (a) Chemistry Factors are calculated from Cu and Ni values per Regulatory

Guide 1.99, Position 1.

(b) Initial RTmr alues v are measured, WCAP-14824, (Reference 2).

! (c) Closure head and vessel flange Initial RTmrvalues are used for considering ;

l flange requirements for the heatup/cooldown curves, WCAP-14824,

/

)

e t

i 13

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

i

l BYRON - UNIT 1 i

)

i PRESSURE AND TEMPERATURE LIMITS REPORT l 1

Table 4.3 I

Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTS) i j at 1/4T and 3/4T Locations for 12 EFPY j I

, 12 EFPY l

Material Description 1/4T ART ( F) 3/4T ART ( F) J Intermediate Shell 78 66 Forging SP-5933

)

l

~

(RG Position 1(*)) ,

Using credible 70*) 60*)

surveillance capsule data (RG Position 2(')) ,

J l

Lower Shell Forging 52 38 SP-5951 (RG Position 1('))

Circumferential Weld 92 58 (RG Position 1('))

Using credible 47 31 surveillance capsule data (RG Position 2()

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Positions 1 and 2 (Reference 12).

(b) These ART values were used to generate the Byron Unit I heatup and

. cooldown curves in WCAP-14824, (Reference 2).

14

- . . . . . . - . - -_ . . -. - . - . - - - - . _ - . . ~. - _ .

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT -

Table 4.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures (ARTS) at 12 EFPY at the Limiting Reactor Vessel Material Intermediate Shell Forging 5P-5933 (Based on Surveillance Capsule Data)

Parameter Values Operating Time 12 EFPY Location 1/4T ART 3/4T ART Chemistry Factor, CF ( F) 19.0 19.0 Fluence (f), n/cm2 4.86x10 ' l.75x10 8 (E>l.0 Mev))")

Fluence Factor, FF 0.799 0.538 ARTmr= CFxFF( F) 15.2 10.2 l

Initial RT mr.,1( F) 40 40 Margin. M( F) 15.2 10.2 l ART = I+(CF*FF)+M, F 70 60 per RG 1.99, Resision 2 (a) Fluenca, f, is based upon f a(E>l.0 Mev) = 8.10x10 at 12 EFPY (WCAP-14824, Reference 2).

(b) The Byron Unit I reactor vessel wall thickness is 8.5 inches at the beltline region.

i

)

15

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5 RTns Values for B3 Ton Unit I for 32 EFPY CF p.) pp M R T m n u) ARTns RTns Material (F) ( F) (F) ( F)

( F)

Intermediate Shell Forging 23.8 2.159 1.209 28.8 40 28.8 97.6 5P-5933 Using Surveillance 19.1 2.159 1.209 17.0 40 23.1 80.1 Capsule Data (*)

Lower Shell Forging 26.0 2.159 1.209 31.4 10 31.4 72.8 5P-5951 Weld Metal, WF336 82.0 2.159 1.209 56.0 -30 99.1 125.1 Using Sun'elliance 61.8 2.159 1.209 28.0 -30 74.7 72.7 Capsule Data (')

2 (a) 2.159 x 10 n/cm (E>l.0 Mev) for 32 EFPY from Byron 1 PTS report, WCAP-13881 (Reference 9).

(b) FF (Fluence Factor) = fa2:4 iones o (c) Calculated using a CF based on sun'eillance capsule data per RG 1.99, Position 2 (Reference 12).

16

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTns Values for Byron Unit I for 48 EFPY CF p.) . ppa) M RTsonu) ARTns RTns Material ( F) ( F) (F) (F) ( F) l~ntermediate Shell Forging 23.8 3.238 1.309 31.2 40 31.2 102.4 SP-5933 Using Surveillance 19.1 3.238 1.309 17.0 40 25.0 82.0 Capsule Data)

Lower Shell Forging 26.0 3.238 1.309 34.0 10 34.0 78.0 5P-5951 Weld Metal, WF336 82.0 3.238 1.309 56.0 -30 107.9 133.3 Using Surveillance 61.8 3.238 1.309 28.0 -30 80.9 78.9 Capsule Data (*)

2 (a) 2.159x10 n/cm (E>1.0 Mev) for 32 EFPY from Byron 1 PTS repon WCAP-13881(Reference 9).

The following calculation provides the 48 EFPY fluence value:

2 2.159x10 + ((2.159x10-3.807x10)/32-5.64 EFPY)) * (48-32EFPY) = 3.238x10 n/cm (b) FF (Fluence Factor) = fa2iaio ioso (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 1.2).

17

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References

1. Andrachek, J.D., et al., WCAP-14040-A, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996.
2. Grendys, P.A., WCAP-14824," Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"
Revision 1, April 1997.
3. Peter, P.A., et al., WCAP-13880, " Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program,"
January 1994.
4. Terek, E., et al.,WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program," August 1990.
5. Westinghouse Letter to Commonwealth Edison Company, CAE-96-106, " Byron Unit I and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits," January 17,1996.
6. Davidson, J. A., WCAP-9517, " Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program," July 1979.
7. Peter, P.A., Westinghouse Letter Report to Commonwealth Edison Company, FDRT/SPRO-009(94), " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," January 1994.
8. Lippencott, E.P., WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation," April 1994.
9. Peter, P. A., WCAP-13881, " Evaluation of Pressurized Thermal Shock for Byron Unit 1,"
January 1994.
10. Malone, M.J., et al., WCAP-14064, " Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program,"
July 1994.
11. 10 CFR 50.61," Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," January 18,1996 (PTS Rule).
12. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99," Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
18
_ ._. . _ .~ _ . _ - . . . __ _ . _ _ . . _ _ . . . _ _ . _ . . . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ .
BYRON - UNIT 1 e
PRESSURE AND TEMPERATURE LIMITS REPORT 4
References (Continued)
13. ' Comed Calculation BRW-96-906I/BYR 96-293," Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)," Revision 0.
14. Comed Nuclear Fuel Services Department, NDIT No. 960186," Maximum Allowable LTOPS.PORV Setpoints for Byron Unit I with RSGs," ReviCon 1.
)
i i
19

ML20141K018

Text

Navigation menu

Search