Text

Analysis of Capsule Y from AL Power Co Farley Unit 1 Reactor Vessel Radiation Surveillance Program
	ML20002E350
Person / Time
Site:	Farley
Issue date:	06/30/1980
From:	Shaun Anderson, Kaiser W, Yanichko S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20002E348	List: ML20002E347; ML20002E349; ML20002E350;
References
	WCAP-9717, NUDOCS 8101270526
	Download: ML20002E350 (78)
	v • d • e

{{#Wiki_filter:. "? f WESTINGHOUSE CLASS 3 i 7 CUSTOMER-DESIGNATED OISTRIBUTION a m ' b-w ANALYSIS OF CAPSULE Y FROM %, w-THE ALABAMA POWER COMPANY y FARLEY UNIT NO. 1 REACTOR VESSEL RADIATION T.' a SURVEILLANCE PROGRAM y R s[& \\ l f/2 ~ # 73 ~ 3# M S. E. Yanichko S. L. Anderson . h.' 9/2-273 T/f'3 , k. W. T. Kaiser y ,4 S c June 1980 ~ l- [ 7 k /b u AA2.,h APPROVED: 4 J)J N.'Chirigos, Manager i! 4 l Structural Materials Engineering o i t Work performed under Shop Order No. WLA 6320 F I. Prepared by Westinghouse for the Alabama Power Company f i t Althoug'i information contained in this report is nonproprietary, no distribution li shall bt made outside Westinghouse or its j licensees without the customers's approval. f i i) il WESTINGHOUSE ELECTRIC CORPORATION ll Nuclear Energy Systems P. O. Box 355 I Pittsburgh, Pennsylvania 15230 8101270 g

,. P i ' q'. s e TABLE OF CONTENTS s' 1 ,T i,t,l e, P,aje t Section l-1 1

SUMMARY

2-1 ~ 2 INTRODUCTION 3-1 3 BACKGROUND 4-1 4 OESCRIPTION OF PROGRAM 5-1 ? 5 TESTING SPECIMENS FROM CAPSULE Y 5-1 ( 5.1. Test Procedure 5.2. Charpy V-Notch Impact Test Results 5-2 5.3. Tensile Test Results 5-2 5-3 5.4. Compact Tension Test Results E. 6-1 f 6 NEUTRON 00SIMETRY ANALYSIS L 6.1. Radiation Analysis and Neutron r. 6-1 Dosimetry 6.2. Discrete Ordinates Analysis 6-1 I 6-4 6.3. Neutron Dosimetry ~ 6.4. Transport Analysis Results 6-8 (: 6-9 6.5. Dosimetry Results I t Appencix A HEATUP AND COOLDOWN LIMIT CURVES r $-1 kl FOR NORMAL OPERATION 5 f t 'i E.. t i f i; 5 iii L 8

J LIST OF ILLUSTRATIONS Figure Ti t'le Page 4-1 Arrangement of Surveillance Capsules in the Farley Unit No.1 Reactor Vessel (Updated l.ead Factors for the Capsules Shown in Parentheses) 4-4 I (L 4-2 Capsule Y Diagram Showing Location of Specimens, Thermal Monitors, and Dosimeters 4-5 5-1 Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower Shell PlateB6919-1(TransverseOrientation) 5-8 h 5-2 Irradiated Charpy Y-Notch Impact Properties for ~ l Farley Unit 1 Reactor vessel Lower Shell Plata BC919-1 (Longituoinal Orientation) 5-9 ~~ 5-3 Irradiated Charpy V-Notch Impact Properties for Farley Unic 1 Reactor Pressure Vessel Weld p Metal 5-10 Il e 5-4 Irradiated Charpy V-Notch impact Properties for ~{ Farley Unit 1 Reactor Pressure Vessel Weld-( Heat-Affected-Zone Metal 5-11 i f; 5-5 Charpy impact Specimen Fracture Surfaces for Farley Unit 1 Pressure Vessel Lower Shell Plate B6919-1 (Transverse Orientation) 5-12 h 5-6 Charpy Impact Specimen Fracture Surf aces for Farley Unit 1 Pressure Vessel Lower Shell Plate B6919-1 (Longitudinal Orientation) 5-13 f s v l l 1 i

LIST Of ILLUSTRATIONS (Cont) Figure Title Page 5-7 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Weld Metal 5-14 5-8 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Weld Heat-Affected-Zone Metal 5-22 5-9 Irradiated Tensile Properties for Farley Unit 1 ( Reactor Pressure Vessel Lower Shell Plate 86919-1 (Transverse Orientation) 5-16 5-10 Irradiated Tensile Properties for' Farley Unit 1 ~ y Reactor Pressure Vessel Lower Shell Plate B6919-1 (Longitudinal Orientation) 5-17 .bh 5-11 Irradiated Tensile Properties for Farley Unit 1 Reactor Pressure Yessel Weld Metal 5-18 g w,. (M 5-12 Typical Stress-Strain Curve for Tension Specimens 5-19 Tg t 5-13 Fractured Tensile Specimens From Farley Unit 1 Pressure Vessel Lower Shell Plate 36919-1 [ (Transverse Orientation) 5-20 .c If3 ) j, 5-14 Fractured Tensile Specimens From Farley Unit 1 "Ai Pressure '~issel Lower Shell Plate B6919-1 (Longitudinal Orientation) 5-21 44 em 1.D 5-15 Fractured Tensile Specimens From Farley Unit 1 4Q, 'l

q Prt_ssure Vessel Weld Metal 5-22 57)

?! 6-1 Farley Unit 1 Reactor Geometry L1 md -w vi

'[ 1 1 LIST-0F ILLUSTRATIONS (Cont)

1 J

Fic_ure Title Page 4 6-2 Plan View of a Reactor Vessel Survelilance 6-22 Capsule 6-3 Calculated Azimuthal Distribution of Maximum I Fast-Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel-Surveillance Carsule Geometry 6-23 l 6-4 Calculated Radial Distribution of Maximum Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessel 6-24, I 6-5 Relative Axial' Variation of Fast Neutron Flux l (E > 1.0 Mev) Within the Pressure Vessel 6-25 i 6-6 Calculated Radial Distribution of Fast Neutron r Flux (E > 1.0 mev) Witisin the Reactor Yessel Surveillance Capsules 6-26 ~ j f; 6-7 Calculated Variation of Fast Neutron Flux Monitor I Saturated Activity Within Capsules U, X, and Y 6-27 't 6-8 Calculated Variation of Fast Neutron Flux Monitor. I Saturated Activity Within Capsules W, V and I 6-28 i l I i I I i t I ( i i vii j 4 e v-

~ I .y ' LIST OF TABLES ,q Title _

Page, j

Table _ Pu n Chemistry and Heat Treatment of Material 4-1 ,] l Representing the Core Region Lower Shell 3 Plate and Weld Metal from the Farley Unit { a 4-3 No. 1 Reactor Vessel {f, Charpy V-Notch Impact Data for Farley Unit 5-1 a' g No. 1 Pressure Vessel Lower Shell Plate I f t B6919-1 Irradiatad at 550'F, Fluence 5.83 c 18 5-4 x 10 n/cm2 (E > 1 Mev) Charpy V-Notch Imp'act Data for Farley Unit E 5-2 l No.1 Pressure Vessel Weld and Heat Affected i Zone Metal Irradiated at 55G'F, Fluence 5.83 18 5-5 f x 10 n/cm2 (E > 1 Mev) L 18 Effect of 550*F 1rradiation at 5.83 x 10 E 5-3 n/cm2 (E > 1 Mev) on Notch Toughness Proper-f 5-6 ties of Farley Unit No. 1 Reactor Vessel 2 I L Irradiated Tensile Properties for Farley Unit 5-4 [ No.1 Pressure Vessel Lower Shell Plate and 18 2 . r Weld Metal, Fluence 5.83 x 10 n/cm 5-7 (E > 1 Mev) [ 6-11 6-1 21-Group Energy Structure s I 6-12 Nuclear Parameters For Neutron Flux Monitors I 6-2 E y' Calculated Fast Neutron Flux (E > 1.0 Mev) and i 6-3 Lead Factors for Farley Unit No.1 Surveillance 6-13 g Capsules ix l i

LIST OF TABLES (Cont) ) Table Title Page 6-4 Calculated Neutron Energy Spectra at the Center of Farley Unit 1 Surveillance Capsules 6-14 .i 6-5 ' Spectrum Averaged Reaction Cross Sections at the Center of Farley Unit 1 Surveillance Capsules 6-15 6-6 Irradiation History of Farley Unit 1 Reactor Vessel Surveillance Capsule Y 6-16 l-6-7 Comparison of Measured and Calculated Fast Neutron Flux Nonitor Saturated Activities for Capsule Y 6 17' 6-8 Results of Fast-Neutron Oosimetry for Capsule Y '6-18 1 6-9 Results of Thermal Neutron Dosimetry for Capsule Y 6-19 t2 [3 6-10 Summary of Neutron Oosimetry Results for Capsule Y 6-20 ' h? Ia- _n n nab. .w i.c i .e 6 i.Ws sa .,T I i1*l e ~J4 \\ :.*::; .9 l x l I' ( 1 (

i

ii-T { = SECTION 1 i' t SLNMARY 9 I The analysis of the reactor vessel material contained in Capsule Y, a s','rveillance capsule from the Farley Unit I reactor pressure vessel, led t to the following conclusions: i ~ s The capsule received an average fast-neutron fluence of 5.83 x 18 10 n/cm2 (E > 1.0 Mev) compared to a calculated value of 18 6.47 x 10 n/cm. Based on the fluence measurements for Capsule Y, the vessel

1/4-thickness fluence af ter 1.13 ef fective-full-power years of 17 2

operation is 9.85 x 10 n/cm compared to a calculated 18 2 fluence of 1.09 x 10 n/cm, 18 ofe,2 resulted in The fast-neutron fluence of 5.83 x 10 -f the following increases in transition temperature and decreases in upper shelf energy: 50 Ft Lb 30 Ft Lb t. Temperature Temperature Shelf Energy Increase Increase Decrease -. f Material (*F) (*F) (ft lb) f 5 l Plate B6919 60 55 0 (transverse g i orientation) I j Plate B6919-1 85 85 12 a (longitudinal I orientation). I ? 1-1 l

50 ft Lb 30 Ft Lb Temperature Temperature Shelf Energy Increase Increase Decrease Material (*F) (*F) (ft Ib) e1d Metal 80 80 19 Heat-Affected-65 60 17 Zone Metal The results of surveillance tests indicate that the reacter pressure vessel is tough enough to continue to operate safely. 'e tr The probable end-of-life fluences at various locations through pg;4 the vessel wall are"as follows: Er$3 g3 Fluence t'-j Vessel Wall Location Measured N/CM Calculated N/CM2 m$ Inner Surface 4.72 x 10 5.22 x 10 19 19 ^h(; N 19 19 1/4 Thickness 2.78 x 10 3.08 x 10 18 19 3/4 Thickness 5.65 x 10 6.27 x 10 =, v Mk N'i

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r :3 IM w h.ll. 7, . ~.: .7 3e 1-2

= i I y SECTION 2 INTRODUCTION I This repcrt presents th3 results of the examination of Capsule Y, the f' first capsule of the continuing surveillance program for monitoring the ~ effects of neutron irradiation on the Alabama Power Company Farley Unit 1 1 reactor pressure vessel materials under actual operating conditions. I The surveillance program for the Farley Unit I reactor pressure vessel I materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preir-radiation mechanical properties of the reactor vessel materials are presented by DavidsonEI3. Tne surveillance program was planned to cover the 40-year life of the reactor pressure vessel and is based on g l ASTM E-185-73, " Recommended Practice for Surveillance Tests for Nuclear ReactorVessels..[2] Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the first W capsule from the reactor and its shipment to the Westinghouse Research 5 and Development Center, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed. s ? This report summarizes testing and the postirradiation data obtained from the first material surveillance capsule (Capsule Y) removed from the Farley Unit i reactor vessel and discusses the analysis of these Using current methods,[3] heatup and cooldown pressure-tempera-data. ture operating limits were established for the nuclear power plant. The heatup ano cooldown pressure-temoerature operating limits are presented I in appendix

I l

tI 1. Davidson, J. A., et al., " Southern Alabama Power Company Joseph M. f Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveil- [ lance Program," WCAP-8810, December 1976. F 2. ASTM Ouignation E185-73, " Surveillance Tests for Nuclear Reactor [ Vessels" in ASTM Standards (1974), Part 10, pp. 286-292, American 'I i Societry for Testing and Materials, Philadelphia, PA,1974. k

3..U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor t

Regulation, Standard Review Plan, NUREG-75/087, Sect. 5.3.2, j [ " Pressure-Temperature Limits," November 1975. 2-1 f f. t

3 2. a 1 SECTION 3 44 BACKGROUND i r i n f .The ability of the large steel pressure vessel containing the reactor No ,l core and its primary coolant to resist fracture constitutes an important y:. ] factor in insuring safety in the nuclear industry. The beltline region Q of the reactor pressure vessel is the most critical region of the vessel I{l i because it is subjected to significant fast-neutron bombardment. The- 'g overall effects of fast-neutron irradiation on the mechanical properties y; of low-alloy ferritic pressure vessel steels such as A533 Grade B Class E ~ I 1 (base material of the Farley Unit I reactor pressure vessel beltline) k} are well documented in the literature. Generally, low-alloy ferritie ~g f materials show an increase in hardness and tensile properties and a, e y decrease in ductility and toughness under certain conditions of f 3 E irradiation. p g 9 p A method of performing analyses to guard against fast fracture in { reactor pressure vessels has been presented in " Protection Against {. NonDuctile Failure," appendix G to section III of the ASME Boiler and Pressure Vessel Code. The method utilizes fracture mechanics concepts } and is based on the reference nil-dictility temperature, RTNDT* l RT is the greater of either the drop weight nil-ductility transi-l 1 NDT 1__ $ tion temperature (NOTT per ASTM E-208) or the temperature 60*F less than I the 60 ft (and 35-mil lateral expansion) temperature as determined from f Charpy specimens oriented normal to the major working direction of the material. The RT of a given material is used to index that NDT curve) { l material to a reference stress intensity factor curve (Xyg which appears in appendix G of the ASME Code. The X;g curve is a lower bound of dynamic, crack arrest, and static fracture toughness t - t results obtained from several heats of pressure vessel steel. When a j, 4 s cune, aHowaMe stress intenshy f-given material is indexed to the Xyg I factors can be obtained for this material as a function of tempe.-ature. -i t Allowable operating limits can then be determined utilizing these allow-f able stress intensity factors. t-ij 3-1 t

I 41P RTNDT.nd, in turn, me operaung limits of nuc, lear power plants, can y 5 be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement or changes in 4 mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program such as the om developed for the Farle; Unit I reactor vessel. In this program a strveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens are tested. The increase in the average Charpy V-notch 50 ft Ib temperature (aRTNOT) due to irradiation is be ( ' added to the original RT to adjust the RT f r radiation NOT NOT initial + ARTNOT) IS embrittlement. This adjusted RTNOT (RTNOT used to index the material to the X;g curve and, in turn, to set y> operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials. N rt I'd. M

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) SECTION 4 DESCRIPTION OF PROGRAM I. i Six surveillance capsules for monitoring the effects of neutron exposure j on the Farley Unit I reactor pressure vessel core region material were [ inserted in the reactor vessel prior to initial plant startup. The six capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall at locations shown in figure 4-1. The lead factors shown in figure 4-1 are higher than reported in the original surveillance program report. These lead factors represent updated values resulting frem improved analytical techniques developed since the original report. These techniques are discussed in paragraph 6-4. The vertical center of the capsules is opposite the vertical t center of the, core. Capsule Y was removed after 1.13 effective full power years of' plant [ operation. This capsule contained Charpy V-notch impact specimens from I the limiting core region plate (lower shell plate 86919-1), core region weldment, and weld-heat-affected-zone material. Also contained in the capsule were l'/2 T-CT and tensile specimens from lower shell plate 86919-1 and we'idment, and a bend bar specimen from lower shell plate f I 86919-1. The chemistry and heat treatment of the surveillance material 7 are presented in table 4-1. ~ All test specimens were machined from the 1/4-thickness location of the l plate. Test specimens represent material taken at least one plate thickness frord the quenched eno of the plate. Some base metal Charpy V-n5tch and tensile specimens were oriented with the longitudinal axis of the specimens normal to and some parallel to the major working direc-l tion of the plate. The CT test specimens were machined such that the + f crack of the specimen would propagate normal to (longitudinal specimens) f and parallel to (transverse specimens) the major working direction of the plate. All specimens were fatigue precracked per ASTM E399-70T. The precracked bend bar was machined in the transverse crientation. l g 4-1 I l l i

3 5 E J9 l l Charp. 7-notch specimens fecm the Weld metal were oriented with the i 51 ,g longitudinal axis of the specimens transverse to the weld direction. g Tensile specimens were oriented with the longitudinal axis of the speci-ll men normal to the weld direction. Capsule Y contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15-w/o-cobalt (cadmium-shielded and unshielded). In 238 addition, cadmium-shielded dosimeters of Np and U were wg contained in the capsule along with the specimens and located as shown y in figure 4-2. The austenitic stainless steel capsule containing the hu specimens and dosimeters also prevents corrosion from contact with th'e M pressurized water environment during irradiation. 4;.4 ve M s Thermal monitors made fron; two low-melting eutectic alloys and sealed in st. Pyrex tubes were incidded in the capsule and were located as shown in .s> g - figure 4-2. The two eutectic alloys and their melting points are: ,jj 2.5 Ag, 9.75 Pb melting point, 579'F .y 1.75 Ag, 0.75 Sn, 97.5 Pb _ melting point, 590*F E3! >Tp W mei 4M .;r M g- .W N1 G1 T@ ~

21 Au

"*J .0 s O a (I =:4 ~ ~. ~ S 4-2

,l, TABLE 4-1 g CHEMISTRY AND HEAT TREATMENT OF MATERIAL REPRESENTING THE k CORE REGION LOWER SHELL PLATE AND WELD METAL FROM THt! I FARLEY UNIT 1 REACTOR VESSEL Chemica'l Analyses (Percent) . E Plate 86919-1 Element Westinghouse E Weld Metal ~ C 0.20 0.13 Mn 1.40 1.39 1.06 P 0.015 0.015 0.016 5 0.013 0.015 0.009 Si 0.28 0.18 0.27 I Ni 0.56 0.55 0.19 Cr 0.13 0.063 V <0.001 0.003 I Mo 0.51 0.56 0.50 Co 0.016 0.008 0.018 Cu 0.10 0.14 0.14 5 Sn 0.008 0.005 h Al 0.025 0.009 [ N 0.003 0.005 2 i i Heat Treatment g Lower shell plate 86919-1: 1550*/1650*F -- 4 hr, Water quenched l, 1225'F + 25'F - 4 hr, Air cooled 1150*F + 25*F' - 40 hr, Furnace cooled to 600*F Weldment: 1150*F + 25'F - 16 he, Furnace cooled i r I I 4-3 i I i. m.

l i 16646-1 i Weor::ve,,0 go s s .5) Y REACTOR VESSEL f (2.9) Z CORE BARREL M F-NEUTRON PAD i } \\- I 7 (2.9) V ~ W, k (3.5)X 3 y f~

!W 2700 e

E I 900 g, ) '/ / r. U(3.5)fr h W (2.9) 4 cp l 1 l .g 6 .55 ~ 9-j , :.ya 180* $4)3 NOTE:

3~Q CAPSULE IDENTIFICATIONS

.4 HAVE CHANGED FROM THOSE

W IDENTIFIED IN WCAP-8810

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i M

l Figure 4-1. Arrangernent of Surveillance Capsules in Farley Unit 1 l J Reactor Vessel (Updated Lead Factors for the Capsules Shown in Parentheses.) ,.;.n ~ 4-4 i i

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v a SECTION 5 .g Y TESTING OF SPECIMENS FROM CAPSULE Y 4 xm l g 5.1. TEST PROCEDURE -+ a 6 1 s The postirradiation mechanical testing of the Charpy V-notch and tensile 9.f specimens was performed at the Westinghoue Research and Development } l, Center Hot Cell Laboratory with consultation by Westinghouse Nuclear f g Energy Systems personnel. Testing of the CT test specimens was delayed .{ upon request by Alabama Power Company pending clarification of testing y procedures from the Nuclear Regulatory Comission. , 61' -7

L Upon receipt of the capsule at the Laboratory, the specimens and spacer b'

n i blocks were carefully removed, inspected for identification number, and M g checked against the master list in the report by Davidson aj No h ~ discrepancies were found. Examination of the two low-melting (579'F and 590*F) eutectic alloys Based on this exami-p showed that melting did not occur in either alloy. nation it was concluded that the maximum temperature to which the test {k __J U specimens were exposed during irradiation was less than 579'F. M e l f. f A TM1 Model TM 52004H impact test machine was used to perform tests on th e irradiated Charpy V-notch specimens. Before. initiating tests on the j, } irrad:ated Charpy V specimens, the accuracy of the impact machine was checked with a set of standard specimens obtained from the Army Material I The results and Nchanics Research Center in Watertown, Massachusetts. of the calibration testing showed that the machine was certified for Charpy V-notch impact testing. I e Davidson, J. A., et al., " Southern Alabama Power Company Joseph M. i j 1. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveil-- lance Program," WCAP-8810, December 1976. a k 5-1 I l " _ sum 2[.'ee

w" b

g4 ~~_ '*~ ~ '

..i ! P -1 h The tr,ile tests were conducted on a screw-driven Instron testing machine having a 20,000-pound capacity. A crosshead speed of 0.05 ~ in./ min was used. The deformation of the specimen was measured using a q strain gage extensometer. The extensometer was calibrated before

y testing with a Sheffield high-magnification drum-type extensometer calibrator.

Elevated temperature tensile tests were conducted using a split-tube 1 @4 furnace. The specimens were held at temperature a minimum of 30 minutes h to stabilize their temperature prior to testing. Temperature was moni-y' tored using a chromel-alumel thermocouple in contact with the upper and g lower clevis-pin specimen grips. Temperature was controlled within plus 6Ei or minus 5'F. T;g u -; V The load-extension data were recceded on the testing machine strip .w g chart. The yield strength, ultimate tensile strength, and uniform FM elor.gation were determined from these charts. The reduction in area and ((3{ total elongation were determined from specimen measurements. d.el -y* 5.2. CHARPY V-NOTCH IMPACT TEST RESULTS h .A -yy The irradiated:Charpy V specimens represented reactor pressure vessel %.,w beltline plate material, weld material, and heat-affected-zone (HAZ) 9 material. The results are presented in tables 5-1 and 5-2 and figures 5-1 through 5-4. A summary of the increases in transition temperature y$ and the decrease in the upper shelf energy is presented in table 5-3. l The fractured specimen appearance of all the Charpy specimens tested .4 from Capsule Y are shown in figures 5-5 through 5-8. G +8 5.3. TENSILE TEST RESULTS ?il fd i Q The results of the tensile tests are presented in table 5-4 and figures -si 5-9 through 5-11. Tests were performed on specimens from plate 95919-1 Vi and the weld metal at room temperature, at 250*F, and at 550*F. A .,.q typical load-displacement curve obtained for the tensile tests is shown 3 in figure 5-12. The broken tensile specimens from the surveillance ~ capsule are shown in figures 5-13 through 5-15. 5-2 1

5.4. COMPACT TENSION TEST RESULTS The 1/2 T compact tension fracture mechanics specimens that.were con-tained in Capsule Y have been stored at the Westinghouse Research Loboratory on the recommendation of the United State Nuclear Regulatory Commission and will be tested at a later date. The results of these tests will be reported upon their completion. ~ I I l I F i L 5 P f 5-3 m

k M }{ 'ABLE 5-1 + N _} CHARPY V-NOTCH IMPACT DATA FOR FARLEY UNIT 1 { g PRESSURE VESSEL LOWER SHELL PLATE B6919-1 { $2 IRRADIATED AT 550*F, FLUENCE I 5.83 x c. 'I' n/cm2 (E > 1 Mev) uy Transverse Orientation i ^2 i 3-Specimen Test Energy Lateral Shear Number Temo. (*F) (ft 1b) Expansion (mils) (%) EC N + AT74 25 10 19.0 5 I AT64 50 15 13.0 15 is AT75 50 23 23.0 15 i j AT66 75 41 29.5 35 su AT73 79 34 30.5 35 ] AT63 100 ' 46 36.0 40 AT65 125 49 40.5 50 . <l. 7 AT71 126 40 46.0 40 ?4 AT62 149 56 43.5 40 T] AT67 174 58 48.5 60 [jf AT69 175 70 58.0 60 gj AT70 225 77 68.5 75 y AT61 249 88 64.0 100 hj AT72 301 87 69.0 100 lg AT68 350 95 54.5 100 NdM Longitudinal Orientation +3 ij AL68 0 22 21.5 5 2f AL66 25 13 10.5' 5 '. 3' AL64 50 25 21.0 15 AL63 50 27 21.0 20 AL62 75 64 40.5 40 G AL71 79 50 44.5 30 'l AL72 99 59 42.5 40 AL67 100 65 46.5 45

i AL70 125 53 40.0 35 2

AL75 125 83 61.0 55 lj AL69 174 99 69.5 70 l ;# AL73 175 98 74.0 90 l 1 AL74 224 110 79.0 90 i AL65 300 127 79.0 100 AL61 350 129 79.0 100 5-4

s TABLE 5-2 CHARPY V-NOTCH IMPACT DATA FOR FARLEY UNIT 1 g PRESSURE VESSEL llELO AND HEAT-AFFECTED ZONE METAL IRRADIATED AT 550*F, FLUENCE 5.83 x 10 n/em2 (g,y g,y) 18 Wela Metal p Specimen Test Energy lateral Shear Number Te.o. (*F) (ft-lb) Expansion (mils) (%) AW73 -26 9 5.5 10 AW62 0 26 25.0 35 AW64 0 32 24.0 40 AW75 25 47 35.5 35 AW67 25 55 41.0 40 AW66 50 61 47.0 60 AW69 79 85 68.5 65 AW61 124 117 81.0 85 AW71 125 85 78.0 70 AW72 174 1( 77.0 80 AW68 209 11, 84.5 85 AW74 226 113 79.0 85 { AW65 251 133 87.0 100 l AW70 299 124 82.0 100 AW63 350 137 84.5 100 Weld HAZ Metal AH66 -110 20 13.0 10 AH68 -108 27 16.5 20 l AH62 -75 35 23.5 2a $q j' AM72 -50 83 54.0 AH65 -25 59 36.0 AH70 -25 74 42.5 33 AH69 0 86 68.5 33 AH64 0 124 67.0 43 AH74 25 95 57.0 53 AH71 79 50 69.5 60 Ail61 80 127 c.10. 0 73 AH67 100 123 82.0 80 AH73 151 112 73.0 7s AH63 199 141 83.5 100 l AH75 250 137 85.0 100 t l 5-5

IAntt s.J !!1E.C.8..'.8. 9ff..!"?.HI A' 8.9. 48..S.8.!.a.8?'" al=#.l t..t.l..t.'.! ?.3!'S!.33E?."I.D. [?.I.#.I.E.I,II.S. 88..I A88 I I..bk,1,1, 8 N( At timt Vt SSit mal (RI AL Average Average J5-=41 Average Average Baergy 50 ft It h $ t ater al t ar,tas,6,ua leisp,M {tt,f t,1,m ),ey Q gserbt, lea f ell, Sne,ar,Jf11g Material Un teredQ Led Irradiated g Unirradi,a t,ed Irradle,ted Q Un f rradij Qr adjat ed,' g un6cradlated irradiated a It 14 86119 1 (Irans.) 10 130 60 45 800 SS 15 70 55 90 90 0 se119 1 (Leag.) 0 85 85 50 10 80 30 ti SS 140 123 Ig W id Metal 5% 7% 80 50 ?O /H no o 80 149 1 10 19 <n 8m hAl Metal -125 6u 6% 505 50 SS 150 90 4.0 ISt l}s 37 I I i t' s** * **** # # ,,,..,....=w**'"***"'*,,.wme4 ,. g .m.-*-e*~***

-_.as w. T Atti.t 8.,-4 IRRADI ATED TENSILE PROPERTI_ES. FOR FARLEY UNIT 1 PRESSURE VESSEt. = y_/cm, (E > 1 H_ed LOWER SilELL PLATE AND WELD HETAL, FLilENCE 5.83 x 10 n Ultimate 0.2 Yield Tensile Fracture Fracture Total Uniform Reduction Temp. Strength Strength Load Strength Elong. Elong. in' Area (ksi) (ksi) _ (IbL_ (ksi) (%)__ (%) (%) Spec. No. (*F) 28.9 *I I AT-13 78 74.35 92.89 3640 91.67 10.30* 9.83 AT-14 250 70.18' 06.58 3070 62.74 19.35 9.90 55.1 ~ AT-15 550 65.19 86.58 3340 68.24 16.35 8.55 46.4 AL-13 78 76.90 96.80 2990 61.11 23.10 10.35 72.3 AL-14 250 71.30 89.02 2890 59.00 21.85 10.13 70.1 m AL-15 550 71.55 88.92 2870 58.57 22.05 10.43 68.5 AW-13 78 88.10 100.22 3080 62.84 25.00 10.58 67.9 AW-14 250 82.50 '94.93 3020 61.72 22.55 10.05 69.0 AW-15 550 80.46 96.46 3390 69.26 19.65 9.00 61.8 Specimen broke outside the gage length, a. l l \\ l ....._.3,,,

'? I. I E~ 16646-3 120 3 g i t iOO e-e-o l 80 /, l 5 m E

/e 40 3

ese e i-O 20 n / i 0 f ')- n d 80 3 A_ g .?$ 5 60 e?* ^ ill - 5 age

e-[55 3

~ 0 m -m a .94 20 g 4 g e m o M 5 o tw 100 I 'A UNIRRADIATED e e.',5 e-g 44 80 o 12 E e o a 2] / -E I gy [ 60 ge IRRADIATED (550 F) 0 j.- [ obGO 5.83 x 1018 N/CM2 e /g

c e

40 O o 20 O C s'o . a o o e 9 ld 1 I I I 0 -100 0 100 200' 300 400 4 r.j TEMPERATURE (OF) ~~ y Figure 5-1. Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B6919-1 (Transverse i Orientation) ~- 5-8 e

Ij t

F x'i 16646 4 l

120 l l 3 l 3 l 3

0"O 100 i

O 9 80 -6 e k 60 of x 99' 40 2 O'. M 20 O O'9 2 0 i 100 [2 8 6 g p .o e-o - o 9 7 e t 9 s-60 d ::' / 40 O g$ g j b d 80, i e /- w 2$e/k Q 20 2 a' l n s t UNIRRADIATED I 140 O /e-O 120 /e a 9 4 100 H u. 9 s 80 / 0 IRRADIATED (550 F) a g[S E 60 g 850 3 40 n = - 85* 9 /* 20 Te O C -100 0 .100 200 '300 400 TEMPER ATURE (OF) Figure 5-2. Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Vessel Lower Shell Plate B6919-1 (Longitudinal Orientation) \\ i I 5-9 l

{ l 16646 5 i ~20 I l 3 13 3I I j. 100 b b e-o-e 5-p 80 O 4 2 9 Q 60 O*O .s / r".

.+

G 40 O 20 [ g p 0 I d 100 h 80 O~O . T4 G O '31. 60 R^ O / a g 40 Sj-700 20 2 g / (w.3 = 3 0 C 180 160 -4 O M9 140 O e i e e "_7 l A lA ED .-Q g 120 rO 4 eO ee 5-n $ 100 e

3. fi c

80 oe .r. x c O / .; ~. - w IRRADIATED (550 F) 0 .i 60 g1 5.83 x 1018 i 9 Boo N/CM2 '.N*f 40 - l/ 4 M 9 800 gg 20 - /w : 'b

1 I

I I .-ut O 3Q: -200 -100 0 100 200 300 400 1 '.tl TEMPERATURE (OF) r' -I Ficure 5-3. T,a Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Pressure Vessel Weld Metal s. l 5-10 P D .S { e l

y bl 16646-6 l l ,l t l 120 l dl i l l 2, j 3 e 100 2 G 80 g g i g 9/ I; 60 O z o 9-I 9 gg i m 40 o 0.8, y O 20 f 2 0 100 2 z ~ h 80 z< $ ~3 O O 60 ~ mO p 40 55* of e c: O { -fp 20 0 a 0 180 160 UNIRRADI ATED 9 ro

140 0

,0 E 120 9 9 0 O 100 IR R ADIATED (55*F) c 80 5.83 x 1o18 n/cm2 e' 5 60 o J 9 65* 40 W 20 - 9 i i i l i 0 -200 -100 0 100 200 300 400 TEMPERATURE (OF) Figure 5-4. Irradiated Charpy V-Notch Impact Properties for Farley Unit 1 Reactor Pressure Vessel Weld-Heat-Affected-Zone Metal l 5-11

lg> 2 iR lb646(1 ) g- .r-F .u m. I ) v \\ n -- r - - -J __: _y n . :.__ _,_=.w... wE - - m; c4 E. _ 1 M G. 1 I .m,: AL13 8 78 F 4 -{ { P- .-., g-. ry.w y, a 1 .'3 @6

k 2

i.; I $ T a-q - i Q .. ww,,,ep g g- = g .h

p. 2. 2.- -,x:,c

~ -- x.... si [* ~~ A i

c 5 R:19 n

1 ?U[ D.,* AL14 t m$b 250 F 0 7 r ~.. c-7.m au., - +.v g g g .~ g k.- - ....., v.,.. n n. m t. .;a,9 -~~- N ~., Yk E.,.

.c

- ~. - __ WN ..g z.. ., 3 - - at.

==. [hl ___-~% k3 Mi 'M mi y-5.* g - L: - a, AL15 0 550 F ~33: Figure 5-5. .;4 Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Pressure Vessel Lower Shell Plate B6919-1 (Transverse " ~, Orientation) u. l l 5-12 4

s 16646 23 I l f,. ,.s %. < w w_ -=- ca-1 , y;;;Gy- ~ s ..n w, - - .b Q y _ _.._o - g7 e-_ _ _... _ 2,.g - + ; ar d TJ .y- . ~ _ _. o i-0 AW13 78 F r- >:-rn,; w _'*~\\ - 1'7 d +

. hQ, y

+ e ,p . ~xw>r f.:e,- -* ' Hi.: 4.:ll

m. v %,.. %.... e

.m h-A..K-} .+ \\- [ -- w --T ^T- _ s. m _3f y - -6 m3 = _- __, - 3 ;__q ^~y e '.i nP t- - 'Mm .B 0 AW14 250 F fI}nki?$Q$${,f?.Yl:r~- [~". - * -r 5: ~ ...z mss':w. - wp n;;2.rms _ =;.. 4mi:9,,g e.,, f flId{ k <Y MWg{ ^^ l 1- .L - y. e [ l 0 AW15 550 F i l Figure 5-6. Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Pressure Yessel Lower Shell Plate 86919-1 (Longitudinal Orientation) 5-13 4 m

@p

R 1L 16646-24 l

l -,-.., irtir;mmg m b,.7, - M. t. - ay= ,j 3, [ .w 3 A . t. 7 t' p. N g i 1 L. 255 l 9 g l j ',, -}, [ _ W, ~ ~~ Q [ t'"w 4 61 $w.w S Q B r -1 m 4 6 AT13 73 p 0 a h,.n& g- ~-w,{# M-- "?- 3f _..t. ~q ~- d'."?"f $.f Q f-' & f ' 'M c y L. : s. w.:- m m. ::M',,=; 'H N -1 [*+, r N .+ y W.. + f.-i,_R-- - _ A n _: -Q ~ ~ ~ t.:. f Q Db~ A w

1 O* g

?n i M 9 E. AT14 250 F 'a i. r. . 4.. -=.._.a._. ~. .w re s9. 2..$s.u_-w- ?.m.

e-m

,} t -} - ',gQ-T.' x: 5 !,w;'. I .f .'r .n:, . - s. s ++ ; n:w n 1 ~ h 6-1 i 7 J ~ ) J 6 l m _ 3 7 y .y ~; _ i .l. l AT15 550 F 0 1i Figure 5-7. Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Weld Metal I l ( 5-14 m

mm nPD.db )\\Il f 16646 29 O JP. 10N odJ c rM

===q =r v t ,p 5t4 .. im i > 1 $iM' -kg' [ j A. l - 3 g.;5is.. _._.ag: -_. '.'"5 ,. m y 1 p .. --m 3_, ..:-t a &=. - p. r s 3 7,, ,7 ~ t

c. m.

"",.*4 j g e ?. i { g -f 5' T -I'

f
,==.,#

h ? I* .. ] h *I. - 3 '.f ,z N l ' W 7*. 4. 7 .,. v. a ~ -. . y==-g I .t - 'I ] 6', L 3 !. l! \\ U.i. AH66 AH68 AH62 AH72 AH65 ~ u ~ e,; ., wW k ' ' f.~J i,% ... r( . t.. . -g. - -,e. s. j y, ~ g { ,9 [g

4,

.\\. 9-r .t v._ a-1A u c; i.; 7. 2,.J .,g. e y \\ s ~ ~ . k~W hm I - h #\\ j h..h, 5 l { ._..z. 1 I.p g.- { . L.g. s AH70 AH69 AH64 AH74 'AH71 I i, L.... - 2., 1 t, ;, 3,, A F;y; Y (. .' A. H --). 4-l' V k~ [] M

.e,.. -

-) p

l

~ :. ; ~.-. f 'd M>: - - l41.'4N ~ P' p. t f r' *,.,.N n F-y / ~8%

'd I

9. :,b ,,. i y - " ~- gti -7.. 7,32 . r: au mm i r.. 4,- )I is t- ,A I 's ?. M.. ... u ; i. .J. : = AH61 AH67 AH73 AH63 AH75 Figure 5-8. Charpy Impact Specimen Fracture Surfaces for Farley Unit 1 Weld-Heat-Affected-Zene Metal 5-15'

i ? 16646-7 120 l l l l { l LEGEND: g. O UNIRRADIATED 9 IRRADIATED 5.83 x 1018 2 100 N/CM TENSILE STRENGTH g = 80 e l h o 0- -g

60 m

u f 40 t 3 e I 80 i ir 9-o i 60 h o REDUCTION IN AREA ~ g N. -] 40 t; t 8 i TOTAL ELONGATION D D 20 g O h. _gUNIFORM ELONGATlON 9-o I O j 1 100 200 300 400 500 600 i TEMPERATURE (OF) i F'gure 5-9. Irradiated Tensile Properties for Farley Unit 1 Reactor Pressure Vessel Lower Shell Plate B6919-1 (Transverse Orientation) l 5-16

MV i s. 16646-8 O l l l l l 120 TENS'LE STRENGTH 100 S 5 e o k _! S 5 a 80 N -o -e g f 0 -- / l m O 0.2% YlELD STRENGTH ~ e 40 80 ^ O s-e g v REDUCTION IN AREA 60 LEGEND: g O UNIRRADIATED g 18 2 b 40 9 IRRADIATED 5.83 x 10 N/CM T-TOTAL ELONGATION g g A v A .g c v g-g UNIFORM ELONGATION 20 D a ne e o 0 0 100 200 300 400 E00 600 TEMPERATURE ( F) E Figure 5-10. Irradiated Ter.sile Properties for Farley Unit 1 Reactor [ Pressure Vessel Lower Shell Plate 86919-1 (Longitudinal Orientation) 5-17

P ') ~9!l w . ?q = M b 16646 9 } .g - i e j 120 ] l l l l . l ,1 1 0.. j 100 e 2 -O Q TENSILE STRENGTH s m 80 -e ], 0.2% YlELD STRENGTH f t; S- ^ x v j 60 .? LEGEND: i ' i.-. O UNIRRADIATED i bh' 8 IRRADIATED 5.83 x 10 N/CM2 18 40 .13 [2 3 80 t t 3. 6 A 3 e-

~

~

$3 60 REDUCTION IN AREA

..l p .1 7 N b 40 '9 p d S d TOTAL ELONGATION 20 h 2 UNIFORM ELONGATION, - O 0% .g n l l l l l

j 0

0 100 200 300 400 500 600 'i

TEMPERATURE (OF)

Figure 5-11. Irradiated Tensile Properties for Farley Unit 1 Reactor Pressure Vessel Weld Metal 5-18 6 1

TF d

16646 10 w a 4 i i O ,= N C i. 3 N ,= O' E

3 8.

o e C 8 O e 2 c N ~ v-2 O -C -< E c: 3 >= U C .E l' 5M C n C U ~ C t v3 3-a> l-to C. N O im 8mS W M C. O I l l C O O O O o o O CD v N r= (is'Al ss38.l.s 5-19 \\ l

p i 1 m n q' % Q i i 6 1664628 l D g g L& ' bra% u u ay a m ?. pg wurisse I e - 1

j s.

M,,.

j 1

i lI q p N s.. 4 +4,7 ~ -' i ), ;-. ,-r r*. b -j

t. t i

{!-Q,g- '% -s',, i,q % ] - 3 4 --. e.. l l li' *- pi M, - ) i ?u' h. j i ..r -i.

tc

! $ 9; .~. E ri SS a a m=m a l ? ? '.{, l1 1 J' ta ju i I. .4 n '{ { AW73 AW62 AW64 AW75 AW67 9 I, t

h..!

- ~g G

Fi

a w{n9 R

~~! L,.. - i O .i j Zi. . : si 4 --.i ,1 i:t i bw h. 0,) L M.. m< - ik W'. 1. - <",f ^ ' i.7 p%+ l y y f,: 7 i 34 ,a 4 . 1' i

s..~.),,

'1 gr. f. f%w - ~ a l r~ ta .;,n . y~ j fLM j f UM f , ' " ~

]i

.t i Ti la j'i 1 -? .f i I db 1 AW66 AW69 AW61 AW71., AWJ2 4 1 \\ ~ k& - Ph' d b ..,s.) - ?._b [ ~~ i .3 I r. ~ y-/ 1.0 i a

'i!. <

1 l d N_ j if

C Q

.s P'% 1 al

i p -

L.:p py R,, s c. y~. :. e. u k g LL I. ,t{. ..,..,5-, ya l 3re&J Iy. e :'....e,;-.,. p[ e

i.

.,3 i ,1 ; Q ;3 f_.. i b-.~ O :f,E :-?,,.

j o

AW68 AW74 AW65 AW70 AW63 t 1 i l i Figure 5-13. Fractured Tensile Speci. mens From Farley Unit 1 Pressure Vessel Lower.Shell Plate B6919-1 (Transverse Orientation) i Ye 5-20 'l

.MY m D Sa r gst g? Eg a $~ YM g,, y, t 4 e.- 4 i T H ~ - --- -. i ,. c, : I I f' 'q f 3 t \\ '~. r, L.s t <= ~ l ~ .3 4 - - )'

c. c-J

~. < p K y" f{ } 1 l *. "s E '. .f.j b_fy k d.' .i 6-I t a. n i l AT74 AT64 AT75 AT66 AT73 ,n. m -r r n, e, r.a

e.,

^s i4 l

m:

1 [;j, ]% v .G t'."; 4 (. y i- {!j:{_ v

-i

59.
' d {

e-

8. p

_-f ..'e ,,( y -.... f. ,a ...."I r$ .] 6 r j ; ?"...*< , $:.M.: } } .ia;/m!E . -;, i cd. j ,. ]

_r_. Q 3

g r,,.... .j (; il l} P*! 5 *c. ch .*\\ di 3 iI Ci is .4 AT63 AT65 AT71 AT62. AT67 t 4 3...- 1 a 1:,;..

o v.

t --t

e.. -

r L.1 r

. y4 Q, J..

y : T;f..~0 '~ l ....d .. g-

. 4

1- '.

..c - f_., s t 1 w-e.,cca g. e.; c

w.

- l..=.~~u. t p +-. 9 , i..f.,,_.;,,,3 i w 7: .. i !.s 1 1L i 5 Oi sli AT69 AT70 AT61 AT72 AT68 Figure 5-14. Fractured Tensile Specimens From Farley Unit 1 Pressure Vessel Lower Shell Plate B6919-1 (Longitudinal Orientation) 5-21.

w9 P {' O bf0 16646-26 k o vN b Ju J -s If>f Y }. y-s <f. $

.4er

= t* h ,q; cp

- Era

. h+. 5 ? [.., (n s'

1 p'. I ed I

t i. [,.' ), . h. .,P""~~ ,? ^.,! I L i . W. c !.t L'*3.* h. ' 'q i. m.s ) ' y'. s ',( i W s c. .. Y, p ! ' L- ' ' ' N I?

e. - s

. } ~, ) { 8 .I .a .l .r s. 1i -t ss AL68 AL66 AL64 AL63 AL62 t I a-t !P i jj ,I

.d i

.w.... N $5f b. i

i y~p* J g
D{

'5

t. ~

=). %.*j 'g L. j g r a* y Vy (g '\\ ,. p k fO

- :.lL '

j Lh 0

,.Yf L..-Y 1

e -s-t .m. ij E .! '. ] 1( Fi ' ~w ,.t m- - m h

I

=. j { GF.sy. .. Y.. !

4 g

i .!i. dt g .o = v. .m i3 AL71 AL72 AL67 AL70 AL75 ,G Un S LN GO h5r.- ?a f7 Jh.'-3 Ed 3 s-A N e.6 r.}' '!.A.. N

3.,

A .t . W} vr\\ r:n h ~ c A ,G'"- .7 $. J t> df'..] - . ~...., a.,+ t- . ;) iA y /-b . f$'

j

'l-ed I i b **-"' f M9 'e s'*"' .3 .i y '.i s

j kL i

n .c .i t k 3 AL69 AL73 AL74 AL65 AL61 1 -i f f i. .k 4 Figure 5-15. Fractured Tensile Specimens From Farley Unit 1 Pressure l Vessel Weld Metal ' I ? 5-22 e i 1

W I @D T N@ 9 Q )) SECTION 6 I 4 M 1

e NEUTRON DOSIMETRY ANA1.YSIS 6.1.

RADIATION ANALYSIS AND NEUTRON 0051 METRY Knowledge of the neutron environment within the pressure vessel surveil-lance capsule geometry is required as an integral part of L'4R pressure vessel surveillance programs for two reasons. First, in the interpreta-tion of radiation-induced property changes observed in materials test specimens, the neutron environment (fluence, flux) to which the test specimens were exposed must be known. Second, in relatin; the changes observed in the test specimens to the present and future condition of the reactor pressure vessel, a relationship between the environment at e various positions within the reactor vessel and that experienced by the test specimens must be established. The former requirement is normally met by employing a ccmhination of rigorous anal ~ytical techniques and measurements obtained witt. passive neutron flux monitors contained in i each of the surveillance capsules. The latter information, on the other hand, is derived solely from analysis. This section describes a discrete ordinates Sa transport analysis per-formed for the Farley Unit I reactor to determine the fast-neutron (E > 1.0 Mev) flux and fluenca as well as the neutron energy spectra within the reactor vessel and surveillance capsules, and, in turn, to develop lead factors for'use in relating the neutron exposure of tne pressure vessel to that of the surveillance capsules. Based on spectru.n-aver, aged t i reaction cross sections derived from this calculation, tne analysis of the neutron dosimetry contained in Capsule Y is discussee and co par-isons with analytical predictions are presented. ti 6.2. DESCRETE ORDINATES ANALYSIS A plan view of the Farley Unit i reactor geometry at the core midplane fl is shown in figure 6-1. Since the reactor exhibits 1/8th core symmetry, only a 0* to 45* sector is depicted. Six irradiation capsules attached 1 i 6-1 I

e ~5 ' to the neutron pad are included in the design for use and the reactor j vessel surveillance program. Three capsules are located symmetrically i at 17* and 20' from the cardinal axes, as shown in figure 6-1. j ] A plan view of surveillance capsules attached to a neutron pad is shown { _ in figure 6-2. The stainless steel specwen container has a 1.25 inch 1 by 1.07 inch cross section and is approximately 56 inches high. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 4-2/3 feet of the ] 12-foot-high reactiar core. 1 x From a neutronic standpoint, the surveillance capsule structures are significant. In fact, as will be shown later, they have a market impact I on the distributions of neutron flux and energy spectra in the water a annulus between the neutron pad and the reactor vessel. Thus, in order l 1 to properly ascertain the neutron environment at the test specimen loca- + }, tions, the capsules themselves must be included in the analytical f i model. Use of at least a two-dimensional computation is, therefore, .a7 mandatory, s 4 I In the analysis of the neutron environment within the Farley Unit 1 -t 2 reactor geometry, predictions of neutron flux magnitude and energy spectra were made with the DOT two-dimensional discrete ordinates code. The radial and aximuthal distributions were obtained from an R,e g o p computation wherein the geometry shown in figures 6-1 and 6-2 was - j described in the analytical model. In addition to the R, o computation, I j a second calculation in R, Z geometry was also made to obtain relative. 2 i ~ axial variations of neutron flux throughout the geometry of interest. f In the R, Z analysis the reactor core was treated as an equivalent volume cylinder and, of course, the surveillance capsules were not included in the model. [; i' j 1. Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation-Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970. I l 6-2

74 I h 'D C 6 00 & D lER M N y a@ & n l e! l a: 'Y Soth the R, e and the R, 2 analyses employed 21 neutron energy groups, ..y ,,G an 5 angular quadrature, and a P cross-section expansion. The f 3 y cross sections were generated via the Westinghcuse GAMBITEI) code k system, with broad-group processing by the APROP05[2] and ANISHb .) u.:. ~* codes. The energy group structure used in the analysis is listed in 3

2. -j 1 r table 6-1.

.I !.c., [ g h. A key input partmeter in the analysis of the integrated fast-neutron ,[

I exposure of the reactor vessel is the core power distribution. For this IN b'

analysis, power distributions representative of time-averaged cond W ons Ib [ derived from statistical studies of long-term operation of Westinghouse ,{ four-loop plants were employed. These input distributions include { L rodby-rod spatial variations for all peripheral fuel assemolies. .T - {Si .t 1.h i It should be noted that this particular power distribution is intended g to produce accurate end-of-life neutron exposure levels for the pressure [ vessel. As such, the calculation is indeed representative of an average [ neutron flux and small (+ 15-20 percent) deviations from cycle to cycle' [l are to be expected. TG; Having the results of the R, e and R, Z calculations, three-dimensional 9; Mi variations of neutron flux may be approximated by assuming that the j g; following relation holds for the applicable regions of the reactor: 1:R 5. ~ L 19 -y. 1. Collier, G., et al, "Second Version of the GAMBIT Code," >w WANL-TME1969, November 1969.

d}

2. Soltesz, R. G., et al, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation - Volume 3, i2j Cross-Section Generation and Data Processing Techniques," , :ii WANL-PR-(LL)-034, August 1970.

jj 3.

Soltesz, R. G. et al, " Nuclear Rocket Shielding Methods, N .tdification, Updating and Input Data Preparation - Volume 4 - i.: One-Dimensional Discrete Ordinates Transport Technique," i: i WANL-PR-(LL)-034, August 1970.

l 1

9 I I : 6-3

w d 6(R,Z 0,E ) - d(R,e,E ) F(Z,E ) g g g where d(R,Z,e,E ) - neutron flux at point R,Z,e within energy group g g d(R,e,E ) - neutron flux at point R, o within energy group g g obtained from the R, e calculation 1 F(Z,E ) - relative axial distribution of neutron flux within 9 energy group g obtained from the R, Z calculation 6.3. NEUTRCN 00SIMETRY s, F2 The passive neutron flux monitors included in Capsule Y of Farley Unit 1 m. i h.t t are listed in table 6-2. The first five reactions in table 6-2 are used s as fast-neutron monitors to relate neutron fluence (E > 1.0 Mev) to $d measured material property changes. Bare and cadmium-covered cobalt-Mk aluminum monitors were also included in the table to determine the magnitude of the thermal neutron flux at the monitor location, which is y e Q necessary to account for burnout of the product isotope'~ generated by [y fast-neutron reactions,

Pg Tne relative locations of the various monitors within the surveillance 3

capsules are shown in figure 4-2. The iron, nickel, copper, and cobalt-f h aluminum monitors, in wire form, are placed in holes drilled in spacers at several axial levels within the capsules. The cadmium-shielded neptunium and uranium fission monitors are accommodated within~the dosi-8 meter block located near the center of the capsule.

'd

.~].3 The use of passive monitors such as those listed in table 6-2 does not } i yield a direct measure of the energy-dependent flux level at the point 9 of interest. G Rather, the activation or fission process is a measure of L the integrated effect that the time and energy-dependent neutron flux 5 has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the 6-4

%. f I44 ?'. .( D**D .[ 2l I b; oo o o j various monitors may be derived from the activation measurements only if k . the' irradiation parameters are well known. In particular, the follo.ving ! '-i$ I variables are of interest: y< }t' The operating history o' the reactor .1 The energy response of the monitor s The neutron energy spectrum at the monitor location . (( The physical characteristics of the monitor j -s c. - Tne analysis of the passive monitors and subse:Iuent derivation of the -i^G s average neutron flux requires completion of two procedures. First, the g desintegration rate of product isotope per unit mass of monitor must be lh~ l-determined. Second, in order to define a suitable spectrum averaged g reaction cross section, the neutron energy spectrum at the monitor location must be calculated. I

d.

1 The specific activity of each of the monitors is determined using estab-l lished ASTM proceduresD,2,3,4,0. Following sample preparation, .tQ ..y [ e.. ,i .ii? 1. A5TM Designation E261-70, " Standard Method for Measuring Neutron iy Flux by Radioactivation Techniques," in ASTM Standards (1975), Part W ' ;[ 45, Nuclear Standards, pp. 745-755, Am. Society for Testing and l.. } Materials, Philadelphia, PA,1975. I 2. ASTM Designation E262-70, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards in (1975), Part 45, Nuclear Standards, pp. 756-763, Am. Society for

4 Testing and Materials, Philadelphia, PA,1975.

3. ASTM Designation E263-70, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Iron," in ASTM Standards (1975), Part 45, Nuclear Standards, pp. 764-769, Am. Society for Testing and Materials, Philadelphia, PA, 1975. M 4 ASTM Des'ignation E481-73T, " Tentative Method of Measuring Neutron - Flux Density by Radioactivation of Cobalt and Silver," in ASTM F Standards (1975), Part~45, Nuclear Standards, pp. 887-894, Am. Society for Testing a Materials, Philadelphia, PA,1975. l i 5. ASTM Designation E264-70, " Standard Method for Measuring Fast-Neutron Flux by Radioactivation of Nickel," in ASTM Standards (1975), Part 45, Nuclear Stendards, pp. 770-774, Am. Soc. for l Testing and Materials, Philadelphia, PA,1975. l 6-5 t

e. 7 l1 l

~ i$ the ac '/ity of each monitor is determined by means of a lithium def ftec 4 g germanium (Ge(Li)) gamma spectrometer. l3 The overall standard deviation !b of the measured data is a function of the precision of sample weighing, { ,j

he uncertainty in counting, and the acceptable error in detector cali-ss bration.

For the samples removed from Farley Unit 1, the overall 2 a l$ deviation in the acasured data is + 10 percent. k The neutron energy E spectra are determined analytically using the method described in sub-j section 6-1. } is 2 Having the menured activity of the monitors and thc neutron energy 1 spectra at the locations of interest, the calculation of the neutron a flux proceeds as follows: .0 ? j The reaction product activity in the monitor is expressed as A 1, r J N 9 R=4f Yfe a(E')d(E) I n P t d ^ g (1-e-i t].), -itd (5-1) y~ p j=1 max A J where: k.. T R I = Induced product activity N = Avagadro's number g =) A - Atomic weight of the target isotope d f 3 = Weight fraction of the target isotope in the target g material l Y L = Number of product atoms produced per reaction a e(E) = Energy-dependent reaction cross-section i d(E) = Energy-depenoent neutron flux at the monitor location with I the reactor at full power Pj = Average core power level during irradiation period j E = Maximum . max e reference tore power level j ~ i - Decay constant of the product isotope t j = Length of irradiation period j } td = Decay time following irradia' ion period j { l e. i 6-6 e

l hh d( .-{ Since neutron flux distributions are calculated using multigroup l ' transport methods and, further, since the prime interest is in the [ l f ast-neutron flux above 1.0 Mev, spectrum-averaged reaction cross sections are defined such that the integral term in equation (6-1) is replaced by the following relation: l / E a(E)d(E)dE = i 6 (E > 1.0 Mev) where n

f e(E)d(E)dE l,*g#g a

n i f.d(E)dt ,. 4 1 0 Mev "gg, 1.0 Mev Thus, equation (6-1) is rewritten l t P R= f Yi 6 (E > 1.0 Mev) I 3 (3,,-It)),-Itd (6-2) j L-j=1 P,,, or, solving for the neutron flux 1 N d(E > 1.0 Mev) - ( } i gf y; 7 (g_,-It)) g-Atd N d*I max The total fluence above 1.0 Mey is then given by p d(E > 1.0 Mev) = d(E > 1.0 Mev) I J tj (6-3) j=1 P max i ~ j where f f n ~1 d "[1 t) = total effective-full-power seconds of reactor p max operation up to the time of capsule removal. 6-7

,a gfP/P)D #h'9" M' _ h ju aj\\\\ 1 1}L ~ l An ass

ment o." the thermal neutron flux levels within the surveillance capsules is obtained from the bare and cadmium-covered Co (n,y)

~ 60 Co data by means of cadmium ratios and the use of a 37-barn, 2200 m/sec cross section. Thus, l 0-1 N I T} Bare

Th

N n

'j (1-e' j) e **d (6-4)

k 7

max R M where D is defined as 88"' g Cd covered t::E:4 6.4. w _ TRANSPORT ANALYSIS RESULTS k Resuhs of the S transport calculations for the Farley Unit I reactor n th are summarized in figures 6-3 through 6-8 and in tables 6-3 through IQ 6-5. In figure 6-3, the calculated maximum neutron flux levels at the v.a g@ surveillance capsule t; ente 11ne, pressure vessel inner radius, 1/4-thick-J ness location, and 3/4-thickness location are presented as a function of a.m azimathal angle. The influence of the surveillance capsules on the na ER fast-neutron flux distribution is evident. In figure 6-4, the radial distribution of maximum fast-neutron flux (E > 1.0 Mev) through the h thickness of the reactor pressure vessel is shown. The relative axial mgg variation of neutron flux within the vessel is given in figure 6-5. qy Absolute axial variations of fast-neutron flux may be obtained by {g multiplying the levels given in figures 6-3 or 6-4 by the appropriate 20 values from figure 6-5. .ci' f.! j' ),[ In figure 6-6, the radial variations of fast-neutron flux within surveillance capsules at 17* and 20* are presented. This data, in con- ,N junction with the maximum vessel flux, is used to develop lead factors ( for each of the capsules. Here, the lead factor is defined as the ratio [ w 1 -1 2 ~._ ! 6-8 09 0 0 4

n' D**D

D b.

3 cw o of the fast-neutron flux (E > 1.0 Mav) at the dosimeter block location (capsule center) to the maximum fast-neutron flux at the pressure vessel inner radius. Updated lead factors for all of the Farley Urif t I sur-veillance capsules are listed in table 6-3. Radial variations of analytically determined reaction rate gradients for each of the fast-neutron monitors are shown in figures 6-7 and 6-8 for capsules at 17* and 20*, respectively. In order to derive neutron flux and fluence levels from the measured disintegration rates, suitable spectrum-averaged reaction cross sections are required. Calculations of the neutron energy spectrum existing at the center of each of the Farley Unit 1 surveillance capsules a e listed in table 6-4. The associated spectrum-averaged cross sections for each ' of the fast-neutron reactions are given in table 6-5. i l 6.5. DOSIMETRY RESUt.TS 1 r jp The irradiation history of the Farley Unit I reactor up to the time of removal of Capsule Y is listed in table 6-6. Comparisons of measured and calculated saturated activity of the flux monitors contained in Capsule Y, based on the irradiation history shown in table 6-6, are 4 i given in table 6-7. Also presented are the monitor specific activities as measured on September 11, 1979. The fast-neutron (E > 1.0 Mev) flux and fluence levels derived for Capsule Y are presented in table 6-8. The theni;&l-neutron flux obtained i from the cobalt-aluminum monitors is summarized in table 6-9. Due to j the relatively low thermal-neutron flux ~ at the capsule location, no l burnup correction was made to any of the measured activities. The maximum error introduced by this assumption is estimated to be less than 1 S8 1 percent for the Ni (n,P)Co reaction and even less for all of the other fast-neutron reactions. .t Using the iron data presented in table 6-8 along with the lead factor given in table 6-3, the current and end of-life fast-neutron fluence 6-9 i

W ~ i

\\

U) [ ikk j = l (E > 1.., Mev) for Capsule Y, as wel! as for the Farley Unit I reactor j ~ 1 vessel, are summarized in table 6-10. The correspondence between cal-l 1 culation and measurement is good, with the measured fluence level of 18 2 18 I 5.83 x 10 n/cm comparing to a calculated value of 6.47 x 10 2 n/cm for Capsule Y. I' 1-1 l Since the calculated levels were based on core power distributions l derived for long-term operation while the Capsule Y data is representa- ') y tive only of cycle 1 operation, it is recomended that projections of 5 vessel toughness be based on the analytically based fluence trend curve. n ) Withdrawal of future surveillance capsules should further substantiate [ the adequacy of the analytical approach. Based on the new capsule'to vessel inner wall lead factors identified in [ table 6-3 and the new capsule withdrawal schedule identified in ASTM l'

j.

E185-79, it is,recomended that future capsules be removed from the j y reactor per the following schedule. .A i h Removal Time Estimated Fluencc N Caosule Lead Factor (EFpY) (10 n/cm2 (E>l Mev)) f 19 { j c

j M Y 3.5 24 1.13 (removed) 0.583

-{ VU 3.5 J.4 3 1.71 2 .x X 3.5 24 6 3.43 3 q wW 2.9 O c 11 5.20 .i yV 2.9 d.o 20 9.46 gZ 2.9 p.c. (standby) i j 1 5 L 4 i k i i 6-10

~NF I ~ TABLE 6-1 21-GROUP ENEP.GY STRUCTURE I I GROUP LOWER ENERGY (Mev) 1 7.79 2 6.07 3 4.72 4 3.68 5 2.87 6 2.23 7 1.74 8 1.35 I 9 1.05 10 0.821 11 0.368 ~~ ' 12 0.111 13 4.09 x 10-2 14 1.50 x 10-2 15 5.53 x 10-3 15 5.83 x 10-4 17 7.89 x 10-5 18 1.07 x 10-5 ~ 19 1.86 x 10-6 20 3.00 x 10-7 21 0.0 t 4 i 6-11

l..L-: L. 1 '.'E50.a$.bEk?!W5 555t&&hk55$55El'?bl TAllLE 6-2 NUCLEAR PARAMETERS FOR NEUTRON FLUX MONIT0_RS WT ".. of Target Product Fission syntarggt_ Monitor Material Reaction of Interest 9"' mon i t or Response Range, llalf-Life Yield (6), b 60 Copper Cu (n,=) Co 0.6917 E > 4.7 Mov 5.27 years Iron Fe g,,p) g,54 0.0585 E > 1.0 Mew 314 days 54 b0 Nickel NiS8 (n.p) Co 0.6777 E > 1.0 Hev 71.4 days OE*3 0 0 (n.f) Cs 1.0 E > 0.4 Mev 30.2 years 6.3 Uranium 237[a] 2 I Neptunium Np (n,f) Cs 1.0 E > 0.08 Hev 30.2 years 6.5 Cobal t-aluminum [a] C059(n.y) Co 0.0015 0.4.ev < < 0.015 Hev 5.27 years 60 Cobalt-aluminum C059 (n.y) Co 0.0015 E < 0.015 Hey 5.27 years 60 a. Denotes that monitor is cadmium shielded.

)I

1.,

+ r y i TABLE 6-3 CALCULATED FAST _-NEUTRON FLUX (E>1.0 Mev) AND LEAD FACTORS FOR FARLEY UNIT 1 SURVEILLANCE CAPSULES ~ Capsule Azimuthal 6 (E>1.0 Mev) Lead Identification Angle (n/cm -sec) Factor-11 U 17* 1.81 x 10 3.5 1 I1 x 17* 1.81 x 10 3.5 11 Y 17* 1.81 x 10 3.5 11 W 20* 1.51 x 10 2.9 11 V 20* 1.51 x 10 2.9 11 Z 20* 1.51 x 10 2.9 i h L 4 1 6-13

gr I Ei .h i .d i l TABLE 6-4 l .h I) i d I; l CALCULATED NEUTRON ENERGY SPECTRA AT THE CENTER OF 3 j i FART EY UNIT 1 SURVEILLANCE CAPSULES i 4 2 Neutron Flux (n/cm -sec)

f Group No.

Caosules U, X, Y Caosules W, V, Z l 1 0 8 8 1 9.13 x 10 8.48 x 10 l 9 9 2 3.03 a 10 2.82 x 10 'k 9 9 3 4.85 x 10 4.41 x 10 i 9 9 4 5.31 x 10 4.70 x 10 0 9 9 5 8.95 x 10 7.70 x 10 ,I 10 10 6 1.81 x 13 - 1.54 x 10 l 10 10 7 2.84 x 10 2.39 x 10 l k 8 4.42 x 10 3.65 x 10 10 10 i 9 6.71 x 10 5.46 x 10 10 10 10 10 10 7.53 x 10 6.09 x 10 11 I1 w 11 2.70 x 10 2.16 x 10 11 11 12 3.42 x 10 2.71 x 10 11 11 ' i' 13 1.38 x 10 1.09 x 10 I 10 10 14 8.85 x 10 7.02 x 10 0 15 6.58 x 10 5.21 x 10 10 10 3 11 11 16 1.64 x 10 1.29 x 10 i 17 1.03 x 10 8.12 x 10 11 10 11 10 18 1.08 x 10 8.36 x 10 10 10 19 7.27 x 10 5.61 x 10 10 10 20 6.15 x 10 o 4.72 x 10 10 10 21 9.52 x 10 7.00 x 10 e t i 6-14 t i i

l0 @ : ~ j h N i. C ll 42 TABLE 6-5 n .:,7 EI SPECTRUM-AVERAGED REACTION CROSS SECTIONS AT THE CENTER OF a FARLEY UNIT 1 SURVEILLANCE CAPSULES [b{\\ .j . $I i.i o (barns) j (' ' 7 I.s M! ,;5,. ~ ' Reaction Capsules U, X, Y Capsules W, V, Z t.(e p Ef, 54 Fe54 (n.p) Mn 0.0510 0.0551 f.d Lif.- O MY ~ 5 58 4, Lv i.g Ni58 (n.p) Co 0.0714 0.0753 h P. Ith 11 (i t 60 3 la Cu63 (n,a) Co 0.000359 0.000406 y; a ' _,{' +' Np237 (n,f) FP 3.21 3.17 tj m h A Di U238 (n,f) FP 0.307 0.316 f T4 l I v4 k e I 'l

h 5. f,"o(E) d(E) dE

,.? b [1Mev d(E)dE

[] {

n i- :- ? 3 IM! 9 113 % if ':li! W 9 1ED jflN 4 .Dif i !k a

j tw I' bl 1

il i . t.: 1.4

.Q
y.

ii .I 6-15 ,g i Ifi l

.1 P ) I 3 9 3 3 TABLE 6-5 J IRRADIATION HISTORY OF FARLEY UNIT 1 3 REACTOR VESSEL SURVEILLANCE CAPSULE Y h! Irradiation Decay (a) 4 D P P Time Time J Max p /pMax (days) (days) 3g Month (fel) (MW) J wa h 4/77.- 11/77 401 2766 0.145 244 657 12/77 1823 2766 0.659 31 625 h I[ 1/78 1884 2766 0.681 31 595 W 2/78 2185 2766 0.790 28 567 r,c f 3/78 2642 2766 0.955 31 535 E,, 4/78 2390 2766 0.864 30 506 5/78 '2301 2766 0.832 31 475 }3 6/78 ~2550 2766 0.922 30 445 E 7/78 2492 2766 0.901 31 414 8/78 2459 2766 0.889 31 383

E 9/78 1330 2766 0.481 30 353 w$

10/78 20f1 2766 0.756 31 322 ] 11/78 2730 2766 0.987 30 292 jfj 12/78 2478 2766 0.896 31 261 tM 3 1/79 2157 2766 0.780 31 230 2/79 .2473 2766 0.894 23 202 3/79 1814 2766 0.656 9 193 .9 -+ f Total EFPS = 3.576 x 10. a. Decay time is referenced to 9/18/79. [ l!H m -l u; e 6-16

(. p '. I i

(I TABLE 6-7 NL COMPARISON OF ME ASulED AND CALCULATED FAST-NEUTRON FLUX MONIT

{ SATURATED ACTIVITIES FOR CAPSULE Y j: ' Reaction Radial Measured Saturated Activity

t-and Axial Location Activity (dps/gm) f location (m)

(dos /gm) Capsule Y Calculated N p' 54 Fe54(n.p) Mn 6 6 Top 186.21 1.95 x 10 5.45 x 10 6 6 Middle 186.21 1.90 x 10 5.31 x 10 l 6 6 Bottom 186.21 1.99 x 10 5.56 x 10 6 6 5.44 x 10 6.00 x 10, Average l f (n p) Co' 7 Ni 7 7 Top 186.21 1.01 x 10 8.02 x 10 6 7 h Middle 186.21 9.63 x 10 7.65 x 10 7 7 Bottom 186.21 1.04 x 10 8.26 x 10 7 7 7.98 x 10 9.10 x 10 l Avera9e 60 CuS3(n.a) Co 4 5 Top 186.21 6.59 x 10 5.20 x 10 4 5 Middle 186.21 6.28 x 10 4.96 x 10 4 5 Bottom 186.21 6.81 x 10 5.37 x 10 5 5 5.18 x 10 4.30 x 10 Average 137 Np237(n,f) Cs 6 7 7 Middle 186.21 1.82 x 10 7.00 x 10 9.60 x 10 J 1 238(n,f) Cs U 6 6 6 Hiddle 186.21 2.59 x 10 9.96 x 10 8.80 x 10 O e 6-17

x..:...c4h :b.iMiniMlfidf$i4.ifiditLN5!%.Y $$I$2dI(d((iIdb NMkl TABt.E 6-8 RESULTS OF FAST-NEUTRON DOSIMETRY FOR CAPSULE Y Saturated Activity 6 (E > 1.0 Mev) d (E > 1.0 Mev) 2 2 (dps/gm) (n/cm -sec) (n/cm ) Reaction Measured Ca lcula ted Measured Calculated Measured Calculated Fe54(n.p) Mn54 6 0 II II 18 18 5.44 x 10 6.00 x 10 1.63 x 10 1.81 x 10 5.83 x 10 6.47 x 10 N158(n.p) C058 II II 18 18 7.98 x 10 9.10 x 10 1.59 x 10 1.81 x 10 5.69 x 10 6.47 x 10 Cu63(n a) Co60 5 5 II II 18 18 5.18 x 10 4.30 x 10 2.18 x 10 1.81 x 10 7.80 x 10 6.47 x 10 I Np (n f) Cs 7.00 x 10 9.60 x 10 1.32 x 10 1.81 x 10 4.72 x 10 6.47 x 10 II I 18 18 mL m 238(n,f) Cs 9.96 x 10 8.80 x 10 2.03 x 10 1.81 x 10 7.26 x 10 6.47 x 10 l37 6 6 II II 18 18 U R fi 5 1 y I I J

' h-l TABLE 6-9 l l RESULTS OF THER."*L-NEUTRON 00SIMETRY FOR CAPSULE Y I Axial Satarated Activity (dps/gm) th Location Bare Cd - Covered (n/cm -sec) 8 7 10 Top 1.14 x 10 6.40 x 10 8.82 x 10 ~ 8 7 10 Middle 1.18 x 10 6.49 x 10 9.38 x 10 8 7 10 Bottom 1.15 x 10 6.42 x 10 8.96 x 10 e S S f I 1 i e s

se e

O b I 6-19 ,.m,

r. __._.__.._.__m.__ ,.. ~.,,. _,,,,,m,. F TABLE 6-10 _Ng

SUMMARY

OF NEUTRON 00SIMETRY RESULTS FOR CAPSULE Y w Current 6(E>1.0 Mev) EOL 6 (E>1.0 Mev) (n/cm ) (n/cm ) Location Measured Calculated Measured Calculated 18 18 s Capsule Y 5.83 x 10 6.47 x 10 l 4 18 18 19 19 d-. Vessel IR 1.67 x 10 1.85 x 10 4.72 x 10 5.22 x 10

4*

17 18 19 19 4 Vessel 1/4 7 9.85 x 10 1.09 x 10 2.78 x 10 3.08.t 10 tE] 3 ' 17 17 18 18 -j Vessel 3/4 7 2.00 x 10 2.22 x 10 5.65 x 10 6.27 x 10 p $N cz w e g NOTE: EOL fluences are based on operation at 2766 MWt for 32 eficctive-full-power years. e ew r;' gi 1' 4 MC ~ i IN N._, ^ .e e -%f ts a bfd 3. M, 6-20 -p - si t

~ 16646-11 0 17 CAPSULES U, X, Y 00 / / 0 20 CAPSULES W, V, Z 0 45 /// e s\\\\ / // / f REACTOR VESSEL a ' u;/ / / /, / / I,,, NEUTRON PAD ~ c / REACTOR I CORE / l / ~ / / / / Figure 6-1. Farley Unit 1 Reactor Geometry l 6-21

? m t ' 16646 12 t i I i i i / / i, //7/// p CHARPY SPECIMEN m j / / / / ///7 / g ///// ////// / x x x N w N s x x x N N N N N N N M N p +g = NEUTRON PAD W N s N N N N N N N N N N N N N N N N \\ .g w '5 5$w 8%N a 01 A.s _. y J <R <1 N... y u V \\ e q h Figure 5-2 a) Plan View Of A Reactor Vessel Surveillance Capsule _) j + 6-22 .J

~N ~. 16646 13 1012 8 6 4 2 11 10 g 6 SURVEILLANdE 8 CAPSULES 54 N 4 g M 2 X 2 z PRESSURE VESSEL IR 3 1010 z 1/4T LOCATION 8 6 4 2 3/4T LOCATION f 9 10 0 10 20 30 40 50 60 70 AZIMUTHAL ANGLE (DEGREES) Figure 6-3. Calculated Azimuthal Distribution of Maximum Fast-Neutron Flux (E>1.0 Mev) Within the Pressure Vessel Surveillance Capsule Geometry 6-23

3I 18! g 16646-14 I I I l 3 1 9 0 m .4. ?. 11 y 10 199.39 .E. 6 l G g 4 l 204.39 m i a2 l 2 1/4T "2 2 M 4 2 5-1 N 1010 214.39 3 l d 8 1 E z l g g 6 %+' + 3/4T 3 219.39 Ec 4 l c F 5 l OR 2 3

y HO PRESSURE VESSEL 2

~ 9 I I I I I I I I I I I I i 10 3 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 7 p R ADIUS (CM) s

4.

3 5 2 ts .5 3 Figure 6 4 Calculated Radial Distribution Of Maxi. mum Fast-Neutron Flux Q (E>1.0 Mev) Within The Pressure vessel M i, .2 -1 Id sy 6-24 al -h

= --=-==. ,, gg 16646-15 1 + 1 0 10 8 6 4 i 2 1 10-1 x 8 3 d 6 z 4 W i z m 2 E .a c 10-2 W 8 m 6 z .J i _ 4 _2 ~ wc 2 o u TO VESSEL

CLOSURE HEAD 10-3 I

I I I l -300 -200 -100 0 100 200 300 DISTANCE FROM CORE MIDPLANE (CM) Figure 6-5. Relative Axial Variation Of Fast-Neutron Flux (E>1.0 Mev) Within The Pressure Vessel 6-25 l

. ??M f ~ NO ~ 16646-16

A

.w Y 12 M 10 iD 9 ~ ~ ~ 8 'g 7 {Ej 6 "~ t.d ga 5 ~.4 na e +, 4 y ?~YM 3 CAPSULES 4-g E y%

u. x. Y W

S CAPSULES GQ W. v. Z l EC 2 $Yl 3 -:ww e n Nr E l -e m\\ W$ 4 E-x .i.?-% =z .c x 11 h }{' 3 10

g u.

g

a oo 2

O8 Z g O hg-- f 7 u a w 6 2 ' h9 5

.a.

- q $

4 Nk, $,.f EF ~' 44 1 3 .d p ,,sm ! ~* W5} . w. 2 ~ /. -q GUIDE TEST SPECIMENS Gul0E NEUTRON M PAD l l l l l l -1. h3 to 10 -y ..[*6 182 183 184 185 186 187 188 189 R ADIUS (cm)

-@1

~~S. Figure 6-6. Calculated Radial Distribution Of Fast-Neutron Flux (E>1.0 -9 Mev) Within The Reactor Vessel Surveillance Capsules gi{ - y ;, e: N?-[ w.. 6-26 i

it ~ M <t N. k. 16646-17 y 11 9 10 [ 10 8 8 t 6 6 11 4 Ff 4 'I It 2 l! l 2 NiS8 (n.p) CoS8 .i 10 tj, 8 l 10 { j,. 10 l 8 8

(

6 6 I i l: b 4 Oe 4 o g e !, E .4 m S: p z b $4 d g b jf,w 2 2 238 gn,g3c,137 g U + 10 5 k. o 9 FeI4 (n.f) MnS4 i 7 { 10 l 8 [i 8 6 O

.1 O

6 I}. H l F e H g:? 4 C" 3 c:: 4 3 H f-H },j a w 2 2 j Np2 I(n f) Cn' I q} h.. l 8 10 6 F; 10 g 8 g7 h. Cu63 (n, c)Co60 6 6 i e l' 4 4 g d: [D N E 2 E 2 GUIDE TEST SPECIMENS GUIDE _[ NEUTRON t PAD 7 10 5 10 182 183 184 185 186 187 188 189 RADIUS (cm) Figure 6-7. Calculated Variation Of Fast-Neutron-Flux Monitor Saturated 1 I Activity Within Capsules U, X, And Y "a l.j a 1 l 6-27 i 1 l

s:f 1 a E 16646-18

s 9

10 y 8 ->.] g. 6 i f 4 ~ l-E 2 a b NiS8(n.p)CoS8 8 I 10 8 g 3 6 l -3 9 0 4 Ez E 5 $O R E ~ S M w m m. lE < o G8 ~ 3 2 3 t 8" E m > 107 l 10g- { E U238(n,g)c31W e '- Feb,(n.p)Mny h N 8 8 d o w 6 r I 4 = .:). ~ 3$ ~I 2 ~ Ti et ro 6 Np237(n.f)CslI 8 i 10 10 I G i,d, 8 r

Cu63(n.c )Co"
f}

6 21 l 7' 4 1 3 ~ M 3 2 NEUTRON GUIDE TEST SPECIMENS GUIDE PAD 5' I I I I I d: 10 107 e' 182 183 184 185 186 187 188 189 RADIUS (CM) Ficure 6-8. Calculated variation Of Fast Neutron-Flux Monitor Saturated Activity Within Capsules W, V, And Z 6-28 l

3 APPENDIX A l HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION A-1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference all-ductility temperature). The most limiting RT of the material in the core region of the reactor NDT vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART s NDT* NDT designated as the higher of either the drop weight nil-ductility tra:fsi-tion temperature (NDTT) or the temperature at which the material exhibits at least 50 ft 1b of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F. RT increases as the material is exposed to fast-neutron radiation. NOT j_ _ Thus, to find tne most limiting RT at any time period in the NDT reactor's life, ART due to the radiation exposure associated with NDT that time period must be added to the original unfrradiated RT NDT* The extent of the shift in RT is enhanced by certain chemical NDT elements (such as copper) present in reactor vessel steels. Design curves which show the effect of fluence and copper content on ART NDT i for reactor vessel steels exposed to SE0*F are shown in figure A-1. Given 'the copper content of the most limiting material, the radiation-induced 4RT can be estimated from figure A-1. Fast-neutron fluence NDT (E > 1 Mev) at the vessel inner surface, the 1/4 T (wall thickness), and 3/4 T (wall thickness) vessel locations are given as a function of full-power service life in figure A-2. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT NDT* O A-1

4 I: Y A-2. F *.CTURE TOUGMRESS PEROPERTIES [ The preirradiation fracture-toughness properties of the Farley Unit 1 3 reactor vessel materials are presented in table A-1. The fracture-s toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory i Standard Review Plan.b13 The postirradiation fracture-toughness f-prc,perties of the reactor vessel beltline material were obtained directly from the Farley Unit i Vessel Material Surveillance Program. i A-3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS g The ASME approach for calculating the allowable limit curves for various b heatup and cooldown rates specifies that the total stress intensity i factor, K, for the combin6d thermal and pressure stresses at any time g 7 9 during heatup or cooldown cannot be greater than the reference stress A j intensity factor, Kyg, for the metal temperature at that time. Kyg is obtained from the reference fracture toughness curve, defined in f Appendix G to the ASE Code. The K curve is given by the IR y equation: 1 KIR = 26.78 + 1.223 exp (0.0145 (T-RTNDT +160] (A-1) s where K is the reference stress intensity factor as a function of gg j the metal temperature T and the metal reference nil-ductility tempera-ture RT Thus, the governing equation for the heatup-cooldown -r NDT. analysis is defined in Appendix G to the ASME Code [2] as follows: CK + kit < KIR (A-2) gg a ~ 1 j. 1. " Fracture Toughness Requirements," Branch Technical Position MTES No. 5-2, Section 5.3.2-14 in Standard Review Plan, NUREG-75/087, 1975. c. ASME Boiler and Pressure Vessel Code, Section III, Division 1 - ~ Appendices " Rules for Construction of Nuclear Vessels," Appendix G, " Protection Against No,nductile Failure," pp 461 - 469, 1977 Edition and Winter 1979 Addendum, American Society of Mechanical Engineers, New York, 1977. A-2

T w a*:. E. ,~. where p E is the stress intensity factor caused by membrane (pressure) K y. Ig =c. N stress F. is the stress intensity f actor caused by the thermal gradients f. K gg i of the material K is a function of temperature to the RTNDT gg C - 2.0 for Level A and Level B service limits ll e f k-C = 1.5, for hydrostatic and leak test conditions during which the .f p k reactor core is not critical is determined j 5 At any time during the heatup or cooldown transient, K IR [ { by the metal temperature at the tip of the postulated flaw, the appro-4 P priate value for RTNOT, and the reference fracture toughness curve.- The thermal stresses resulting from temperature gradients through the F.. h vessel wall are calculated and then tne corresponding (thermal) stress From intensity factors, Kyg, for the reference flaw are computed. a equation (A-2), the pressure stress intensity factors are obtained and, ,n from these, the allowable pressures are calculated. r I ? For the calculation of the allowable pressure-versus-coolant temperature [ during cooldown, the Code reference flaw is assumed to exist at the { inside of the vessel wall. During cooldown, the controlling location of f tne flaw is always at the inside of the wall because the thermal gradi- } ents produce tensile stresses at the inside, which increase with f Allowable pressure-temperature relations are increasing cooldown rates. l generated for both steady-state and finite cooldown rate situations. i l From these rela: ions, composite limit curves are constructed for each ( k cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary l because control of 'the cooldown procedure is based on measurement of reactor coolan. temperature, whereas the limiting pressure is actually dependent on tie material temperature at the tip of the assumed flaw. A-3

_. -- =g s During.ooldown, the 1/4 T vessel location is at a nigner temperature tnan the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the aT developed during cooldown results in a higher value of KIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in Kgg exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value. Tne above procedures are needed because there is no direct control on d temperature at the 1/4 T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldawn period. Three separate calculations are required to determine tne limit curves for finite heatup rates. As is done in the cooldown analysis, allowable L. pressure-temperature relationships are developed for steady-state condi-tions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the vessel wall. The thermal gradients ~ during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; there-fore, the K for the 1/4 T crack during heatup is lower than the gg K for the 1/4 T crack during steady-state conditions at tne same IR coolant temperature. During heatup, especially at the end of the tran-f sient, conditions may exist such that the effects of compressive thermal stresses and lower K do not offset each other, and the pressure-IR's temperature curve based on steady-state conditions no longer represents a lower cound of all similar curves for finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. A-4

7 b '. % , j gg t J, i The second portion of the heatup malysis concerns the calculation of i pressure-temperature limitations for the case in which a 1/4 T deep t j t.e; Unlike the situation at the vessel f outside surface flaw is assumed. inside surf ace, the thermal gradiests established at the outside surface !- I)[Il during heatup produce stresses whidi are tensile in nature and thus tend 1, l These thermal stresses are 1 Le to reinforce any pressure stresses present. 6 c dependent on both the rate of heatup and the time (or coolant tempera-j I Since the thermal stresses at the outside 'ture) along the heatup ramp. y are tensile and increase with increasing heatup rates, each neatup rate f must be analyzed on an individual basis. f 2,. .n Following the generation of pressare-temperature curves for both the i f t steady-state and finite heatup rate situations, the final limit curves j A composite curve is constructed based on a 2 are produced as follows: point-by-point comparison of the' steady-state and finite heatup rate p#C? f l At any given temperature, the allowable pressure is taken to be ~ l[ data. the lesser of the three values taken from the curves under considera-,I The use of the composite curve is necessary to set conservative ij T i tion. heatup limitations because it is possible for conditic's to exist yr r 4 3 wherein, over the course of the heatup ramp, the controlling condition w. y 4 i switches from the inside to the outside and the pressure limit must at !?.- n - 54 j

Then,

] all times be based on analysis of the most critical criterion. (,j composite curves for the heatup rate data and tne cooldown rate data are $w7 adjusted for possible errors in the pressure and temperature sensing Q ~ instruments by the values indicated on the respective curves. IM; 4.m. hi HEATUp AND C00LOOWN LIMIT CURVES '4 [ A-4. i %3 Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in l d f, l The derivation of the limit curves is presented in tne } paragraph A-3. j NRC Regulatory Standard Review Plan.EI) ' W I 3 ,0, " Pressure-Temperature Limits," Section 5.3.2 in Standard Review 31 1. O Plan, NUREG-75/087, 1975. b. c.L4b .-h i? M A-5 4

d. hi l V

r

l Transit.n temperature shifts occurring in the pressuro vess21 materials due to radiation exposure have been ootained directly from the reactor l l.1 pressure vessel survetilance program. ,h. 'f Churpy test specimens from Capsule Y indicate that the representative core region weld metal and the limiting core region plate (86919-1) 3 These shifts exhibited shifts in RThDT of 80' and 60'F, r:spectively. h _ at a fluence o' 5.83 a 10 n/cm are well within the appropriate 18 2 [ cesign curve (figure A-1) prediction. Heatup and cooldown calculations J were basec on the ARTET predicted for the core region weld material, which is the limiting vessel material. Heatup and cooldown limit curves W for normal operation of the reactor vessel are presented in figures A-3 .( and A-4 and represent an operational time period of 10 effective-full-

)

power years. $1 Allowable combinations of temperature and pressure'for specific tempera-ture change rates are below and to the right of the limit lines shown on

Cg.

The reactor must not be made critical j the heatup and cooldown curves. until pressure-temperature combinations are to the right of the criti- ~ s

?,.

This is in addition to other cality limit line, shown in figure A-3. .x ,fj 73) criteria which must be met before the reactor is made critical. s E$ M ~ The leak test limit curve shown in figure A-3 represents minimum temper-ature requirements at the leak test pressure specified by applicable @i The leak test limit curve was determined by methods of ~~ codes. Q ,j references.b1'23 v.g~ Figures A-3 and A-4 define limits for insuring prevention of nonductile ?j th failure. 4.1 YM 31 u " Pressure-Temperature Limits," Section 5.3.2 in Standard Review 1. -g Plan, NUREG-75/087, 1975. u-q ASME Boiler and Pressure Vessel Code, Section III, Division 1 - Appenoices, " Rules for Construction of Nuclear Vessels," Appendix G, M 2. 461 - 469, 1977 Edition 3' " Protection Against Nonductile Failure," pp 7 and Winter 1979 Addendum, American Society of Mechanical Engineers, ~3 New York, 1977. s.: .A A-6 -__,.,__m

r---,- w sm7wamyryrpg'g gypr y p g 0 A.sa., -0 1 Rf.A.C.i.tu. t..V.f.55f.t.10.ll.G.il.N.(.55..D.A.I.A. Material Ces l' I HWD HHWH Ri Upper Shelf (nergy gg gg NHWD C gponent_ Co;l,e No. _tyL (*. ) ('* ) j'Q Q Q (*f), ,HWD 40 *I -20 140 I Closure head dome 86901 A5338, Cl.1 0.16 0.009 -30 20 10 *I -20 138 I Closure head segment 86902-1 A5338, Cl.1 0.11 0.007 -20 -10 75 'I I O *I 60 I 60 *I -20 I Closure head flange 06915-1 A500, Cl. 2 0.10 0.012 I 106 *I I I -10 'I 60 60'! -30 Vessel flange 86913-1 8508, Cl. 2 0.17 0.0ll I 110 60 'I 45 60 0.010 Inlet norile 86917-1 A508. Cl. 2 I 80 60 'I 115 60 Inlet nosale 86917-2 A500, Cl. 2 0.008 I 98 60 *I 35 60 Inlet nozzle 86917-3 A503, Cl. 2 0.008 I 96.5 60 *I 60 60. Outlet norale 86916-1 A508. Cl. 2 0.001 I 97.5 60 *I 30 60 0.011 Outlet nogale 86916-2 A508, C1. 2 100 Outlet nozzle 86916-3 A500 Cl. 2 0.009 60(a) 50 si; 90 *I 30 148 I Nortle shell 86914-1 A500, Cl 2 0.010 30 70 Inter, shell 86903-2 A5338, Cl. 1 0.13 0.011 0 -25 40 0 151.5 97 y Inter. shell 86903-3 A5338, Cl. t 0.12 0.014 10 5 52 10 134.5 IDO N tower shell 86919-1 A5338 Cl. t 0.14 0.015 -20 -5 75 15 133 90.5 tower shell 86919-2 A5338, Cl. 1 0.14 0.015 -30 0 65 5 134 97 -5 'I 10 163.5 I 0.010 10 -25 Bottom head ring 86912-1 A508. Cl. 2 -30 'I -30 147 I Bottom head segment 86906-1 A5338, C1. 1 0.15 0.011 -30 -50 10 'I -30 143.5 I Bottom head dome 86907-1 A5338, Cl. t 0.17 0.014 -30 -10 O *I <60 0 I 0.27 0.015 Inter. shell'Iong. O 'I <60 0 I . weld seams 0.24 0.011 Inter, to lower shell O 'I 460 0 I weld seam 0.17 0.022 Lower shell long. weld seams (a) Estimated per llRC Regulatory $tandard Review rian, section 5,3,2. NWC - Major Working Direction NHVO - 140rmal to Hajor Working Direction 6 g ~ Y.* [.$

h

4, j,,,, ' ~ ~ ..... a.. ; _ ;.. ;.,, _,, f-

9.,

gg

y 1 0 16646 19 4 lil\\ j i i 400 300 f 200 e i 150 100 J o_ 80 ,s c y \\- 60 y* x 0.30% CU BASE,0.25% CU WELD e =< 5 40 0.25% CU 8ASE,0.20% CU WELD E 0.20% CU BASE,0.15% CU WELD I 0.15% CU BASE,0.10% CU WELD e. 0.10% CU BASE,0.05% CU WELD F i g E I I I lilli l I I i I lli 0 '1 19 2 4 6,8 1020 4, 1018 2 4 6 8 10 r 2 E > 1 MeV) FLUENCE (N/CM B 5 ~ .A t Ls C : ti rigure A-1. Effect of Fluence ana Copper Content on ARTY 0T for Reac ce vessel Steels Exposed :c Irradiation at 550 F 'c,' ( ha*f N A-8 1a M

> J i 16646-20 1 i i. 1 20 I 10 6 i i 8 I 6 l ~ I I i 4 s n .I 2 - t l i i. i i 10l9 - I L ~ 1/4 THICKNESS ', i i i c 8 l i p N i, j lE i Q 6 I 2 j i w o 4 2 f ua n 3 \\ s ' a \\ 3/4 THICKNESS T u. o 2 l s 2 I c: I H a 5 l w I Z l e 10 l l 18 i 8 l j t r 6 I i 2 s j ty-j j 4 e i i u i I i 'f ~ l h i i e 2 i \\ r,- g b k; 17 I 10 O 5 10 15 20 25 30 35 l SERVICE LIFE (EFFECTIVE FULL POWER YEARS) n Figure A-2. Fast-Neutron Fluence (E > 1 Mev) as a Function of Full-Power { c4$ Service Life 1&W E.C .e g A-9 UE. L@

16646-21 4000 p MAThRIAL PROPERTY BASIS fM CONTEOLLING MATERIAL: WELD METAL COPPER CONTENT: 0.24 WT% N-PHOSPHORUS CONTENT: 0.011 WT% RT 0 D NOTINITIAL: 0F RTNDTAFTER 10 EFPY: 1/4T.185 F 0 3/4T.118 F CURVE APPLICABLE FOR HEATUP RATES UP TORI F/HR'FOR THE SERVICE O PERICOUP TO 10 EFPY AND CONTAINS EAj 3000 MARGiuSOF 10 F AND 60 PSIG FOR 0 POSSIR.E INSTRUMENT ERRORS @m 3 .g-c Wh agj c 3M f u=y LEAK TEST LIMIT US W m 'S-S w m ff ?Q00 \\ ~@U. - O a. 's- ~ m O E

EM o

Ga 8 HEATUP RATES Ms 3 0 UP TO 60 F/HR g .A i f' {. .m h. 1000 f I ICAllW LIMIT .= 31 P BASED ON fd INSERVICE

3.1 HYDROSTATIC TEST 3

..1)1

3....

I TEMPE"+TURE 0 'U (325 ' ' A THE 'j'} SERVite.ERIOD UP TO 10 EFPY- ) US .i 0 .+ R 0 100 200 300 400 500

r..-,

,g INDICATED TEMPERATURE (OF) .-7 Ficure A-3. Farley Unit 1. Reactor Coolant System Heatup Limitations ~0 -I Applicable For The First 10 EFPY 1 ti "I A-10 2 Q, d. ~ ..lf ]., -i.. n..1. T L

5% ~ '& Y ' ~ -

'~

1 l: 16646-22 r.$ E l .a 3 MATERIAL PROPERTY 8 ASIS _ CONTROLLING MATERIAL: WELD METAL f. 1 e.- COPPER CONTENT: 0.24 WT%

t PHOSPHORUS CONTENT: 0.011 WT%

M 0 INITIAL:0 F

2 RTNDT 0

AFTER 10 EFPY: 1/4T,185 F ~{ RTNOT 3/4T,118 F Q 0 3 ti - CURVE APPLICABLE FOR COOLDOWN c RATES UP TO 100 F/HR FOR THE SERVICE 0 9 k PERIOD UP TO10 EFPY AND CONTAINS {p MARGINS OF 10 F AND 60 PSIG FOR N 0 [ POSSIBLE INSTRUMENT ERRORS E l; 2000 i h ,5 I s L a ?er h E [ a L M I .t z i 1000 I .r COOLDOWN RATES OF/HR ~ N 2040 - 60 100 l I i t 400 0 200 300 100 l 0 INDICATED TEMPER ATURE (O F) 1 Farley Unit 1 Reactor Coolant System Cooldown t.imitatiens Figure A 4 Applicable For The First 10 EFPY A-11 --}}

ML20002E350

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