ML20149K103
| ML20149K103 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/23/1997 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20149K099 | List: |
| References | |
| NUDOCS 9707290251 | |
| Download: ML20149K103 (27) | |
Text
(.-.--.-~...-....
4-I' i
I i
4 i
- i i
)
. ENCLOSURE 6 4
1 5-i!
Joseph M. Farley Nuclear Plant i
~
j
' Unit I
.t 3'
Pressure Temperature Limits Report I
i 1-9707290251 970723 PDR ADOCK 05000348 P
Joseph M. Farley Nuclear Plant Unit 1 i
l Pressure Temperature Limits Report l
4 j
. ~.
PRESSURE TEMPERATURE LIMITS REPORT Tabic Of Contents 1
l ListofTables...............................................................................................~............iii l
Li s t of Fi gu re s...........
............................................iv
+
1 1.0 RCS Pressure Temperature Limits Report (PTLR)...............
...................................1 2
4, 2.0 O p e ra t i n g Li m i t s............................................................................................................. 1
- 2. I RCS Pressurefremperature (P.T) Limits (LCO - 3.4.10.1)..
.1 2.2 RCP Operation Limits.
.1 3.0 Reactor Vessel Material Su rveillance Program................................................................ 6 4.0 Reacto r Vessel Su neillan:e Dat a Credibility.......................................................'........................... 7 5.0 S u p p le me n t a l D a t a Ta h les.................................................................................................. 14 5
6.0 References.....................................................................................................................21 i
n4 4
1 1
4 1
1 FARLEY UNIT 1 ii AEVISION 0
I 1
PRESSURE TEMPERATURE LIMITS REPORT List of Tables 2-1 Fariey Unit 1 36 E FP Y Hcatup Cu rve Data Points............................................................ 4 2-2 Fa riey Unit 1 36 E FP Y Cooldow n Cu n e D ata Poin1s................................................. -.......... 5 i
31 Surveillance Capsule Withd rawal Schedule................................................................6 i
4-1 Surveillance Capsule Data Calculation of Best-Fit Line as Ducribed in Position 2.1 cf Regula t o ry G u i d e 1.99, Revi sio n 2.................................................................................... 10 4-2 Scatter of ART 33r Values About a Best-Fit Line for Surveillance Plate Material...............12 i
i 4-3 Scatter of ART mr Values About a Best-Fit Line for Surveillance Weld Material............13
)
8 5-1 Comparison of Surveillance Material 30 Ft-L.h Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions.........15 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data........
..................16 5-3 Reacto r Vessel Tou gh nt.ts Table (Unirradiated)................................................................... I 7 5-4 Reactor Vessel Fluence Projections for 36 EFPY...............................................................17 5-5 Summary of Adjusted Reference Temperatures (ARTS) for Reactor Vessel Beltline Materials at the 1/4-T and 3/4-T Locations for 36 EFPY............................
.... 18 5-6 Calculation of Adjusted Reference Temperature at 36 EFPY for the Limiting Reactor Vessel Material - Lower Shell Plate B6919-2.............................................................19 5-7 Pressurized Thermal Shock (RTrrs) Values for 36 EFPY................................................ 20 FARLEY UNIT 1 iii kEVISION 0
PRESSURE TEMPERATURE LIMITS REPORT 1
1ist of Figures
^
2-1 Farley Unit i Reactor Coolan t System lleatup Limitation s.................................................. 2 2-2 Farley Unit i Reactor Coolant System Cooldown Limitations................................
.3 FARLEY UNIT 1 iv AEVISION 0
PRESSURE TEMPERATURE LIMITS REPORT 1
1.0 RCS Pressure Temperature Limits Report (PTLR)
This PTLR fcr Farley Nuclear Plant - Unit I has been prepared in accordance with the requirement of Technical Specification (TS) 6.9.1.15. Revisions to the PTLR shall be provided to the NRC after issuance.
This report affects TS 3.4.10.1, RCS PressureRemperature (P-T) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are cor.tained in TS 3.4.10.3, RCS Overpressure Protection Systems.
2.0 Operating Limits The limits for TS 3.4.10.1 are presented in the subsection which follows. These limits have been developed using the NRC-approved methodologies specified in TS 6.9.1.15 with the exception that low temperature overpressure protection (LTOP) is provided by the RHR relief valves (RHRRVs) in lieu of the PORVs. Therefore, the operability requirements associated with LTOP will be retained in TS LCO 3.4.10.3. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent witn the pressure corrections incorporated in the PH limits for flow losses associated with the RCPs.
2.1 RCS PressurcHemperature (P-T) Limits (LCO - 3.4.10.1) 2.1.1 The minimum boltup temperature is 60 F.
2.1.2 The RCS temperature rate-of-change limits are:
a.
A maximum heatup of 100 F in any one hour period.
b.
A maximum cooldown of 100 F in any one hour period.
c.
A maximum temperature change ofless than or equal to 10 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.3 The RCS P-T limits for heatup and cooldown are specified by Figures 2-1 and 2-2, respectively.
2.2 RCP Operation Limits 2.2.1 The number of operating RCPs is limited to one at RCS temperatures less than 110 F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.
FARLEY UNIT I 1
hVISION 0
PRESSURE TEMPERATURE LIMITS REPORT l
t 2,500
^
~
Leak Test Limit 7
q.
- 'i
~
~
' ~
Criticahty Unit for
[
60 F/hr Heatup 2.250 Limitng Matenal.
Cnticahty umit for Lower Shell Plate B6919-2 Umstng ART Values at 36 EFPY:
100 F/hr Heatup 1/4T,161 F l
2.000 3/4T.136 F
..s...
.. a t
1.750
'Bi
.g,550 1
a.
4 u-g g
Unacceptable Operation Acceptable Operation 4
2 1,250 Q,
2 V
s e
3 j
q
.st v
c 1,000 g
e-4
-4
.6 A, e.
4 1
.2 750 l-
,..,?). ',
60
. r,
.a.
.1 J.
Masti 7 mmu (degree F/hr)
-,, a y, a._..
3 6d
/I 500
~
~
~
~~
..100 100h I
250 Critu:ahty Limit Based on inservice I
Hydrostatx: Test Temperature (289 F) !.. -....
for the Service Penod up to 36 EFPY q.
0 50 100 150 200 250 300 350 400 450 500 indicated Temperature (Degree F)
Figure 2-1 Farley Unit 1 Reactor Co0lant System Heatup Limitations (Heatup Rates up to 100 F/hr)
Applicable to 36 EFPY (Without Margins for Instrument Errors). Includes vessel flange requirements of 180*F and 561 psig per 10 CFR 50, Appendix G. I'l FARLEY UNIT 1 2
AEVISION 0
PRESSURE TEMPERATURE LIMITS REPORT 2,500
.7 Urrhting Material:
2.250 Lower Shell Plate B6919-2 Umsting ART Values at 36 EFPY:
1/4T,161 F
...L 3/4T,136 F 2,000 L
- 4... _
. g...
.._m.
j a.
.4 4
k
-4
.a e
.y...-
k 1,750 1
e h
i
'?
6 4
.9 1,500 g
a.
, ~._.
. i..
g m
g 1,250 n.
Unacceptable Operation
~
Acx:eptable Operation
" ~ ~ ~
3 i
i.
g j 1,000 l
750 r
4.
. cm nat. A
.-.-m
_.e.. ;,
2 (degree F/hr) 7
-.7 500 3-7 i
f.
yy 4
100 e0 i
250 i
' 100
,. =
.... h. -,
.4...
.w..
4
-l.
7-
.. ~...
0 1
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Degree F) l Figure 2-2 Farley Unit i Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable to 36 EFPY (Without Margins for Instrument Errors) Includes vessel flange requirements of 180 F and 561 psig per 10 CFR 50 Appendix G.VI j
FARLEY UNIT 1 3
AEVISION 0
- - - ~ _.. ~ -... -. - -. - -
4 l
PRESSURE TEMPERATURE LIMITS REPORT s
)
60 'F 60 'F Cnticahty 100 'F 100 'F Criticality Limit Leak Test o
Limit T
l P
T l
P T
l P
T l
P T
l P
60 490 189 0
60 457 289 0
267 2000 65 490 289 480 65 457 289 481 298,
2485 70 490 289 469 70 457 289 466 75 490 289 462 75 457' 289 454 80 490 289 45i 80 457 289 444 85 490 2o9 455 85 457 289 436 90 490 289 456 90 457 289 430
?.
95 490 289 458 95 457 289 426 100 490 289 461 100 457 289 423 105 491 289 466 105 457 289 422 110 493 289 473 110 457 289 423.
)
110 458 289 480 110 422 289 424 i
115 461 289 488 115 422 289 427 l
120 466 289 498 120 422 289 432 125 473 289 509 125 423 289 437
{
130 480 289 521 130 424 289 443 135 488 289 534 135 427 289 451 140 498 289 547 140 432 289 460 145 509 289 563 145 437 289 469 150 521 289 579 150 443 289 481 155 534 289 597 155 451 289 493 160 547 289 617 160 460 289 506 165 561 289 637 165 469 289 521 170 561 289 660 170 481 289 537 175 561 289 684 175 493 289 555 180 561 289 710 180 506 289 574 180 617 289 738 185 521 289 595 185 637 289 768 190 537 289 617 190 660 289 800 195 555 289 641 195 684 289 835 200 574 289 668 200 710 289 873 205 595 289 696 205 738 289 913 210 617 289 726 210 768 289 956 215 641 289 759 215 800 289 1002 220 668 289 794 220 835 289 1051 225 696 289 832 225 873 290 1105 230 726 290 872 230 913 295 1162 235 759 295 916 235 956 300 1223 240 794 300 963 240 1002 305 1288 245 832 305 1013 245 1051 310 1258 250 872 310 1068 250 1105 315 1434 255 916 315 1125 255 1162 320 1515 260 963 320 1188 l
260 1223 325 1601 265 1013 325 1254 265 1288 330 1694 270 1068 330 1326 270 1358 335 1793 275 1125 335 1402 275 1434 340 1899 280 1188 340 1484 280' 1515 345 2012 285 1254 345 1572 285 1601 350 2120 290 1326
-350 1666 290 1694 355 2233 295 1402 355 1767 295 1793 360 2354 300 1484 360 1874 300 1899 305 1572
- 365 1988 305 2012 310 1666 370 2111 310 2120 315 1767 375 2241 315 2233 320 1874 380 2380 320 2354 325 1988 330 2111 335 2241 340
.2380 Table 21 Farley Unit 136 EFPY Heatup Curve Data Points (Without Margins for Instrument Errors)l4 FARLEY UNIT 1 4
AEVISION 0
l PRESSURE TEMPERATURE LIMITS REPORT 0 'E 20 *E 40*F 60 'F 100'F T
l P
T l
P T
l P
T l
P T
l P
60 508 60 471 60 434 60 396 60 317 65 512 65 475 65 438 65 400 65 321 70 516 70 479 70 442 70 404 70 326 75 521 75 484 75 447 75 409 75 331 80 525 80 489 80 452 80 414 80 337 85 531 85 494 85 457 85 420 85 343 90 536 90 500 90 463 90 426 90 349 95 542 95 506 95 470 95 433 95 357 i
100 549 100 513 100 477 100 440 100 365 105 556 105 520 105 484 105 448 135 373 110 561 110 528 110 492 110 456 110 382 110 528 110 493 110 457 110 421 110 347 115 536 115 501 115 466 115 430 115 357 120 545 120 510 120 475 120 440 120 368 125 554 125 520 125 485 125 451 125 380 130 561 130 530 130 496 130 462 130 393 135 561 135 541 135 508 135 475 135 407 140 561 140 553 140 521 140 488 140 422 145 561 145 561 145 534 145 502 145 438 150 561 150 561 150 549 150 518 150 455 155 561 155 561 155 561 155 535 155 474 160 561 160 561 160 561 160 553 160 494 165 561 165 561 165 561 165 561 165 516 170 561 170 561 170 561 170 561 170 5 40 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 715 180 689 180 664 180 640 180 593 185 737 185 713 185 689 185 666 185 623 190 760 190 738 190 716 190 694 190 655 195 786 195 765 195 744 195 725 195 690 200 813 200 794 200 775 200 758 200 727 205 842 205 825 205 808 205 793 205 767 210 874 210 858 210 844 210 831 210 811 215 908 215 894 215 882 215 872 215 857 220 944 220 933 220 923 220 916 220 908 225 983 225 974 225 968 225 963 225 962 230 1025 230 1019 230 1015 230 1014 230 1020 235 1070 235 1067 235 1066 235 1069 240 1119 240 1118 245 1171 250 1226 255 1286 260 1351 265 1420 270 1494 275 1573 280 1658 285 1749 290 1846 295 1951 300 2062 305 2182 310 2309 Table 2-2 Farley Unit 136 EFPY Cooldown Curve Data Points (Without Margins for Instrument Errors)VI FARLEY UNIT I 5
AEVISION 0
PRESSURE TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with 10 CFR 50, Appendix if, and is described in Section 5.4.3.6 of the Farley FSAR. The removal schedule is provided in Table 3-1. The results of these examinations shall be used to update Figures 2-1 and 2-2 if the results indicate that the adjusted reference temperature (ART) for the limiting beltline material exceeds the ART used to generate the Pfr limits shown in Figures 2-1 and 2-2 for the specified fluence period.
Table 31 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE
- Capsule Capsule Location Lead Removal Fluence *
(Degree)
Factor EFPY "
(n/cm )
2 Y*
343 3.33 1.13 5.80 x 10" U*
107 3.34 3.02 1.69 x 10" X*
287 3.38 6.12 2.95 x 10 ~
W "
110 3.13 12.43 3.82 x 10
V 290 3.11 Standby Z
340 3.11 Standby NOTES:
(a) WCAP-14689 'l I
(b) Effective Full Power Years (EFPY) from plant startup (c) Fluence recalculated using methodology contained in WCAP 14040-NP-A, Revision 2.141 (d) Plant-specific evaluation (c) Final capsule withdrawal required by ASTM E-185-82.
FARLEY UNIT 1 6
AEVISION 0
PRESSURE TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for
{
light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date, there have been four surveillance capsules removed from the Farley Unit I reactor vessel.
In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.
4 The purpose of this evaluation is to apply the credibihty requirements of Regulatory Guide 1.99, Revision 2, to the Farley Unit I reactor vessel surveillance data and determine if the Farley Unit i surveillance data is credible.
Criterion 1:
Materials in the capsules should be those judged most !ikely to be controlling with regard to radiation embrittlement.
The beltline region of the reactor vessel is defmed in Appendix G to 10 CFR 50, Fracture Toughness Requirements, December 19,1995, to be:
the reactor sessel (shell material including welds, heat afTected zones, and plates or forgings) that directly si rrounds the efTective height of the active core and adjacent regions of the reactor sessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.
The Farley Unit I reactor vessel consists of the following beltline region materials:
d Intermediate shell plates B6903-2 and B6903 3; e
Lower shell plates B6919-1 and B6919-2; e
Intermediate shell longitudinal weld seams19-894 A & B, heat number 33A277, Linde 1092 flux, flux lot 3889; Lower shell longitudinal weld seams20-894 A & B, heat number 90099, Linde e
0091 flux, flux lot 3977; and Circumferential weld I l-894, heat number 6329637, Linde 0091 flux, flux lot i
3999.
.I FARLEY UNIT I 7
REVISION 0 s
PRESSURE TEMPERATURE LIMITS REPORT P3 Per WCAP-8810, the Unit I surveillance program was based on ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. Per Section 4.1 of ASTM E185-73, the base metal and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime. The test material should be selected on the basis ofinitial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper and phosphorus) and neutron fluence.
At the time the Farley Unit I surveillance capsule program was developed, lower shell plate B6919-1 was judged to be most limiting based on the above recommendations and was, therefore, utilized in the surveillance program.
The surveillance program weld for Farley Unit I was fabricated using the same heat of weld wire used to fabricate the middle shell axial seams19-894 A & B (heat 33A277). The results of mechanical property tests performed on the surveillance weld are considered to be representative of the property changes expected in the reactor vessel beltline seams.
Therefore, the materials selected for use in the Farley Unit I surveillance program were those judged to be most likely controlling with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was deseloped The Farley Unit I surveillance program meets the requirements of Criterion 1.
Criterion 2:
Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy, unambiguously.
Plots of Charpy energy versus temperature for the unirradiated condition are presented in the Unit I reactor vessel surveillance program description contained in WCAP-8810BI Plots of Charpy energy versus temperature for the irradiated conditions are presented in the reactor vessel surveillance capsule reports for capsules Yl63, U PI, Xi'l, and Wl4 Based on engineering judgment, the scatter in the data presented in these plots is small enough to determine the 30 ft lb temperature and upper shelf energy of the Farley Unit i surveillance materials unambiguously. Therefore, the Farley Unit I surveillance program meets the requirements of Criterion 2.
FARLEY UNIT 1 8
REVISION 0 3
i
PRESSURE TEMPERATURE LIMITS REPORT Criterion 3:
When there are two or more sets of surveillance data from one reactor, the scatter of ARTm values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28 F for welds and 17 F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.
The least squares method, as described in Regulatory Position 2.1 will be utilized in determining a best-fit line for this data to determine if this criterion is met.
[ Continued on the following page]
l FARLEY UNIT I 9
REVISION 0 s
PRESSURE TEMPERATURE LIMITS REPORT TABLE 4-1 SURVEILLANCE CAPSULE DATA CALCULATION OF BEST-FIT LINE AS DESCRIBED IN POSITION 2.1 OF REGULATORY GUIDE I.99, REVISION 2 '
Material Capsule F *)
ART T
)
Ocv)
Y 0.580 0.848 85 72.0 0.718 Lower She;;
Plate B69191 U
l.69 1.14 105 120.2 1.31
" E " ""'
X 2.95 1.29 135 173.7 1.66 W
3.82 1.35 155 208.7 1.81 Y
0.580 0.848 55 46.6 0.718 Lower Shell Plate B6919-1 U
l.69 1.14 90 103.0 1.31
- "* "* I X
2.95 1.29 105 135.1 1.66 W
3.82 1.35 145 195.2 1.81 9.25 875 1054.5 10.99 Weld Metal Y
0.580 0.848 80 67.8 0.718 U
l.69 1.14 80 91.6 1.31 X
2.95 1.29 100 129.0 1.66 W
3.82 l_35 95 127.9 1.81 4.62 355 415.9 5.50 NLQTES:
3 (a) WCAP-14689 'l (b) F = Finence (10 n/cm', E > 1.0 MeV) 0 (c) FF = Fiuence Factor = F'* ' *' 5 i
i FARLEY UNIT I 10 REVISION 0 4
PRESSURE TEMPERATURE LIMITS REPORT the method ofleast squares, the values bo and b, are ons.
J nb+b,[x,=[y, o
b [ x, + b, [ x,' = [ x,y, o
These equations can be re-written as follows:
N
[y, = an + b[ x, and
'='
~
[ x,y, = aT x, + b T x' '
M g
,i Lower Shell Plate B69190 Based on the data provided in Table 4-1, these equations become:
875 = 8a + 9.25b I 054.5 = 9.25a + 10.99b Thus, b=145.2 and a=-58.5, and the equation of the straight line which provides the b sense ofleast squares is:
Y' = 145.2 (X) - 58.5 The error in predicting a value of Y corresponding to a given X value is e = Y - Y'.
FARLEY UNIT I II REVISION O 4
PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 2 SCATTER OF ARTm1 VALUES ABOUT A BEST FIT LINE FOR SURVEILLANCE PLATE MATERIAL (
Lower Shell Plate ARTer Best Fit ARTer Scatter of ARTer B6919 1 FF (30 ft-lb)
( F)
('F)
Orientation
(*F) 0.848 85 64.6 20.4 Longitudinal 1.14 105 107.0
-2.0 1.29 135 128.8 6.2 1.35 155 137.5 17.5 0.848 55 64.6
-9.6 Transverse 1.14 90 107.0
-17.0 1.29 105 128.8
-23.8 1.35 145 137.5 7.5 NOTES-IU (a) WCAP 14689 The scatter of ARTer values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17 F for base metal. However, even if the fluence range is large, the scatter should not exceed twice this value (i.e.,34 F). As shown above, the error is within 34 F of the best-fit line. Therefore, Criterion 3 is met for the Farley Unit I surveillance plate material.
Weld Metal:
Based on the data provided in Table 4-1, the equation becomes:
355 = 4a + 4.62b and 415.9 = 4.62a + 5.5b Thus, b=35.6 and a=47.625, and the equation of the straight line which provides the best fit in the 1
sense ofleast squares is:
Y' = 35.6 (X) + 47.625 The error in predicting a value of Y corresponding to a given X value is e = Y - Y'.
FARLEY UNIT I 12 REVISION O a
PRESSURE TEMPER. PRE LIMITS REPORT i
i 8
TABLE 4-3 SCATTER OF ARTwor VALUES ABOUT A BEST-FIT LINE FOR SIJRVEILLANCE WELD MATERIAL
(*F)
(*F)
(*F)
O.848 80 77.8 2.2 Weld Metal 1.14 80 88.2
-8.2 1.29 100 93.5 6.5 1.35 95 95.7
-0.7 J
NOTES:
(a) WCAP-1468914 The scatter of ARTwm values about a best fit line drawn, as described in Regulatpry Position 2.1, is less than 28 F as shown above. Therefore, Criterien 3 is met for the Farley Unit I surveillance weld material.
1 Criterion 4:
The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding / base metal interface within 25 F.
1 The Farley Unit I capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the neutron shielding pads. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience j
equivalent operating conditions and will not differ by more than 25 F. Therefore, the Farley reactor vessel surveillance program meets the requirements of Criterion 4.
i Criterion 5:
The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.
T The Farley Unit I surveillance program does not include correlation monitor material. Therefore, Criterion 5 is not applicable to Farley Unit 1.
J i
FARLEY UNIT I 13 REVISION 0 e
i
PRESSURE TEMPERATURE LIMITS REPORT CONCLUSION:
Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Farley Unit I surveillance data is credible.
5.0 Supplemental Data Tables Table 5-1 contains a comparise f measured surveillance material 30 0-lb transition temperature shins and upper shelf energy c.
.:s with Regulatory Guide 1.99, Revision 2, predictions.
i Table 5-2 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
l Table 5-3 provides the unirradiated Farley Unit I reactor vessel toughness data.
l Table 5-4 provides a summary of the fluences used in the generation of the heatup and cooldown curves and the PTS evaluation.
{
Table 5-5 provides a summary of the adjusted reference temperatures (ARTS) of the Farley Unit 1 reactor vessel beltline materials at the 1/4-T and 3/4-T locations for 36 EFPY, a
Table 5-6 shows the calculation of the ART at 36 EFPY for the limiting Farley Unit I reactor vessel material (lower shell plate B6919-2).
Table 5-7 provides RTrrs values for Farley Unit i for 36 EFPY.
J 4
h 4
FARLEY UNIT 1 14 REVISION 0 a
PRESSURE TEMPERATURE LIMITS REPORT Table 5-1 COMPARISON OF SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFT AND UPPER SHELF ENERGY DECREASE WITH REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS"'
30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease Material Capsule (x 10n/cm,
Predicted Measured Predicted Measured 2
E > 1.0 MeV)
(*F)
( F)
(%)
(%)
Y 0.580 83 85 20 9
U l.69 112 105 26.5 21 Plate B6919-1 (Longitudinal)
X 2.95 126 135 30 19 W
3.82 132 155 32 22 Y
0 580 83 55 20 0
U l.69 112 90 26.5 9
Plate B69191 X
2.95 126 105 30 11 W
3.82 132 145 32 16 Y
0.580 101 80 34 13 Weld Metal U
l.69 136 80 44 28 X
2.95 153 100 49 23 W
3.82 160 95 52 26 Y
0.580 60 11 HAZ Metal U
l.69 120 26 X
2.95 125 19 W
3.82 110 14 NOTES:
(a) WCAP-14689 u t
FARLEY UNIT I 15 REVISION O 4
~.-...-.
.., -. - -. ~.
- _. ~. - - -.
. - - ~.
PRESSURE TEMPERATURE LIMITS REPORT I
i r
Table 5-2 l
CALCULATION OF CHEMISTRY FACTORS USING SURVEILLANCE CAPSULE DATA (*
Material Capsule f*
Lower Shell Plate Y
0.580 0.848 85 72.0 0.718 B6919 1 U
l.69 1.14 105 120.2 1.31 (Longitudinal)
X 2.95 1.29 135 173.7 1.66 W
3.82 1.35 155 208.7 1.81 Lower Shell Plate Y
0.580 0.848 55 46.6 0.718 B6919 1 U
l.69 1.14 90 103.0 1.31 l
(Trannerse)
X 2.95 1.29 105 135.1 1.66 W
3.82 1.35 145 195.2 1.81 l-Sum:
1054.5 10.99 2
Chemistry Factor = I (FF
0.580 0.848 80 67.8 0.718 U
l.69 1.14 80 91.6 1.31 Weld Metal X
2.95 1.29 100 128.7 1.66 W
3.82 1.35 95 127.9 1.81 l'
Sum:
415.9 5.50 1
2 L
Chemistry Factor = I (FF
- ARTwor) + I (FF ) = 75.7 i
i NOTES: ~
14 t
(a) WCAP-14689 2
(b) f = 11uence (x 10" n/cm. E > 1.0 MeV) 28 ' '*D (c) FF = fluence factor = f(
5 I
i
!~-
i r
i FARLEY UNIT I 16 REVISION 0 a
i
PRESSURE TEMPERATURE LIMITS REPORT Table 5-3 REACTOR VESSEL TOUGHNESS TABLE (UNIRRADIATED)
Beltline Material Cu Weight %
Ni Weight %
IRTmn ( F) i Closure Head Flange 60 Vessel Flange 60 Intermediate Shell Plate B6903 2 0.13 0.60 0
Intermediate Shell Plate B6903-3 0.12 0.56 10 Lower Shell Plate B6919-1 0.14 0.55 15 Surveillance Plate ""
0.14 0.55 15 Lower Shell Plate B6919-2 0.14 0.56 5
4 Intermediate Shell Longitudinal Weld Scams19-894 A & B "'
O.24 0.17
-56 i
(Heat # 33 A277)
Surveillance Weld'#'
O.24 0.17
-56 Circumferential Weld Scam 11894 "'
0.21
- 0. I1
-56 (Heat # 6329637)
Lower Shell Longitudinal Weld Scams20-894 A & B
O.20 0.20
-56 (Heat # 90099)
NOTES:
14 (a) WCAP-14689 (b) The surveillance plate is representativc oflower shell plate B6919-1 (c) Generic Letter 92-01, Revision 1, Supplement I response (d) The surveillance weld is representative ofintermediate shell longitudinal welds19-894 A & B Table 5-4 REACTOR VESSEL FLUENCE PROJECTIONS FOR 36 EFPY b' EFPY O'
15*
15"
30' 30*'"
45*
36 3.97 2 45 1.96 1.83 1.77 1.23 NOTEi (a) WCAP-14689 14 2
(b) Fluence in 10 n/cm (E > 1.0 MeV)
(c) Indicates location in octants with a 26* neutron pad span.
FARLEY UNIT I 17 REVISION 0 a
PRESSURE TEMPERATURE LIMITS REPORT Table 5 5
SUMMARY
OF ADJUSTED REFERENCE TEMPERATURES (ARTS) FOR REACTOR VESSEL BELTLINE MATERIALS AT THE 1/4-T AND 3/4-T LOCATIONS FOR 36 EFPY O' Material 1/4-T 3/4-T
( F)
( F)
Intermediate Shell Plate B6903 2 147 124 Intermediate Shell Plate B6903-3 146 125 Lower Shell Plate B69191 171 146 Lower Shell Plate B6919-1 93 151 127 Using S/C Data Lower Shell Plate B6919-2 161 'd' 136 (d)
Intermediate Shell Longitudinal Weld 120 "'
89 "'
Seams19-894 A & B (Heat # 33 A277)
Intermediate Shell Longitudinal Weld Seams19-894 A & B 58 '*'
3 9 "'
(Heat # 33A277)
Using S/C Data")
Circumferential Weld Il 894 135 110 (Heat # 6329637)
Lower Shell Longitudinal Weld Seams20-106
- 80**'
894 A & B (Heat # 90099)
NOTES:
(a) WCAP-14689 'I 1
2 (b) The ARTS presented here are based on the peak reactor vessel surface fluence of 3.97 x 10* n/cm (E >
1.0 MeV) unless otherwise noted.
121 (c) Based on surveillance capsule data contained in WCAP-14196 (d) ART values used to generate heatup and cooldown curves.
2 (c) ARTS calculated using the peak vessel fluence of 1.23 x 10" n/cm (E > 1.0 MeV) at 45 FARLEY UNIT 1 18 REVISION 0 a
PRESSURE TEMPERATURE LIMITS REPORT Table 5-6 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE AT 36 EFPY FOR THE LIMITING REACTOR VESSEL MATERIAL - LOWER SHELL PLATE B6919-2 "
Parameter Operating Period 36 EFPY Location 1/4-T 3/4-T Chemistry Factor, CF ('F) 98.2 98.2 Fluence, f(10 n/cm ) *'
2.48 0.962 2
Fluence Factor, FF 1.24 0.989 ARTer = CF x FF ( F) 122 97 Initial RTer, I ( F) 5 5
Margin, M ("F) "'
34 34 Adjusted Reference Temperature ( ART), ( F) per 161 136 Regulatory Guide 1.99, Resision 2 NOTES:
(a) WCAP-14689 'l 1
(b) Fluence is based on f s (10 n/cm. E > 1.0 MeV) = 3.97 at 36 EFFY, The Farley Unit I reactor 2
vessel wall thickness is 7.875 inches in the beltline region.
(c) Margin is calculated as M = 2(o,2, y,2)" The standard deviation for the initial RTer margin term, o, is 0*F since the initial RTer is a measured value. The standard deviation for the ARTer term, a, is 17'F for the plate, except that a3 need not exceed 0.5 times the mean value of ARTer.
s In accordance with Regulatory Guide 1.99, Resision 2, Position 2.1, values of a may be cut in half s
when based on credible surveillance data.
4 FARLEY UNIT 1 19 REVISION 0 a
PRESSURE TEMPERATURE LIMITS REPORT Table 5-7 PRESSURIZED THERMAL SHOCK (RTers) VALUES FOR 36 EFPY
Surface Fluence ARTer 2
1 M
RTers Material CF (10 n/cm.
( 7)
E > 1.0 MeV)
(*F)
Intermediate Shell Plate B6903-2 91.0 3.97 1.35 123.3 0
34 157 Intermediate Shell Plate B6903-3 82.2 3.97 1.35 111.3 to 34 155 Lower Shell Plate B6919-1 97.8 3.97 1.35 132.5 15 17 164 Lower Shell Plate B6919-1 95.9 3.97 1.35 129.9 15 34 179 Using S/C Data Lower Shell Plate B6919-2 98.2 3.97 1.35 133.0 5
34 172 Intermediate Shell Longitniinal Welds 118.6 1.23 1.06 125.4
-56 66 135 19-894 A & B (Heat # 33 A277)
Intermediate Shell Longitudinal Welds19-894 A & B 75.7 1.23 1.06 80.1
-56 44 68 (Heat # 33A277)
Using S/C Data Circumferential Weld Il-894 100.8 3.97 1.35 136.5
-56 66 147 (Heat # 6329637)
Lower Shell Longitudinal Welds 104.0 1.23 1.06 110.0
-56 66 120 l
20-894 A & B (Heat # 90099)
NOTES:
(a) WCAP-14689 'l 1
l FARLEY UNIT 1 20 REVISION 0 l
l
PRESSURE TEMPERATURE LIMITS REPORT 6.0 References
- 1. WCAP-14689, Revision 1, Farley Units 1 and 2 Heatup and Cooldown Lindt Curves for Normal Operation and PTLR Support Documentation, E. Terek, April 1997.
2.
WCAP-14196, Analysis of Capsule W from the Alabama Power Company Farley Unit I Reactor Vessel Radiation Surveillance Program, P. A. Peters, et al., February 1995.
l
- 3. WCAP-14687, Joseph M. Farley Units I and 2 Radiation Analysis and Neutrou Dosimetry Evaluation, R. L. Bencini, June 1996.
4.
WCAP-14040-NP-A ' Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.
5.
WCAP-8810, Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et al., December 1976 j
6.
WCAP-9717, Analysis of Capsule Y from the Alabama Power Company Farley Unit No.1 I
Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et al., June 1980.
7.
WCAP-10474, Analysis of Capsule U from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, R. S. Boggs, et al., February 1984.
i FARLEY UNIT I 21 REVISION 0 4