ML20099J203
| ML20099J203 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/16/1992 |
| From: | Biemiller E, Cacciapouti R, White R Yankee Atomic Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20099J207 | List: |
| References | |
| YAEC-1853, NUDOCS 9208200159 | |
| Download: ML20099J203 (63) | |
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4 i 3 7 l:? n ANALYSIS OF SEABROOK STATION UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE U p, " c > >
. 'Ya..,.mumm Prepared By E. C. Biemiller R. J. Cacciapouti YANKEE ATOMIC ELECTRIC COMPANY June 26,1992
4 Prepared by: '
~J r ~ 7 // 2 !
E. C. Biemiller (Date) Principal Enginetr Prepared by: ddld /N% R//0. CacciapouM' [!'6 92-(date)
.Rfeactor Physic / Group Manager Reviewed by: d d Y2-R .' G (bate)
- SenYo/ ta'rter r Engineer Approved by: /. O 7 [0 b
'R . (, W hl# ~ ~~' ~ ' / (D4te)
Seabbook Project Engineering Manager Yankee Atomic Electcic Company - Nuclear Services vivision-580 Main Street Bolton, Massachusetts 01740 l R77\253 -ii-
-9 DISCLATHfR OF RESPONSIBillTY This document was prepared by Yankee Atomic Electric Company (" Yankee'). The use of information contained in this document by anyone other than Yankee. or the Organization for which this document was prepared under contract, is not authorized, and with resDect to any unauthorized use, neither Yankee nor its officers directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document. R77\253 -iii-
1 TABLE OF CONTENTS Page DISCLAIMER OF RESPONSIBILITY ,.................. iii . TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . iv LIST OF TABLES .......................... v LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . vi 1.0
SUMMARY
OF RESULTS . . ...................... 1
2.0 INTRODUCTION
........................... 3 3.0 SURVE!LLANCE MATERIALS AND CHEMISTRY ,.............. 4 4.0 PRE AND POST-IRRADIATION TEST RESULTS .............. 5 4.1 Baseline Data. WCAP 10110 . . . . . . . . . . . . . . . . . . 5 4.2 .!rradiated Data. B&W Report BAW 2157 ....., ..... 5 4.3 Tahh Curve Fits . . . . . . . . . . . . . . . . . . . . . . . 5 4.4 .Results . . . . .-. . . . . . . . . . . . . . . . . . . . . . 6 5.0 RT NDT CALCULATION ............, ........... 17 5.1 Calculated RT NDT Values . . . . . . . . . . . . . . . . . . . 17 5.2 Comparison-of Surveillance Capsule Results to Regulatory Guide 1.99. Revision 2 Predictions ....-,........ 18 5.3 Effect of Surveillance Results on Plant Current Technical Specifications . ........... ,......... 1B 6.0 RADIATION ANALYSIS AND NEUTRON 00SIMETRY ............. 21
.6 .1. Introduction . . ...................... 21 6.2 Neutron Calculations .................... 21 6.3 Neutron 00simetry Analysis ................. 22 6.4 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
7.0 REFERENCES
............................ 31 i
p i R77\253 -iv- _ - _ _ _ ;.~. . _ .,_ . - - . _ _ _ - _ . - _ . , . . _ . _ _ . . ,-- . ,_
p . O (IST OF TABLES l
)
i Number Tjtle Page i 41 Seabrook Surveillance Capsule Data Per ASTM E 185 and Regulatory Guide 1.99. Initial Values (Fluence - 0) Per WCAP-10110. 7 42 Seabrook Station Unit 1 Surveillance Program Tensile ; Properties Base Line Properties Data Per WCAP 10110 Irradiated Data Per BAW 2157 8 43 A Comparison of Tanh Ccmputed Charpy Values to Westinghouse and B&W Reported Values 9 61 Nuclear Parameters for Neutron Flux Monitors 25 , 1 62 Irradiation History of Neutron Sensors Contained in Capsule U 26 63 Measured Monitor Activities and Reaction Rates for Capsule U 27 6-4 Spectrum Averaged Reaction Cross Sections For Use In ; Fast Neutron Dosimetry Evaluations 28 65 Results of Fast Neutron Dosimetry for Capsule U 29 66 Comparison of Average Fluence for Capsule U to Other Westinghouse Reactors 30 R77\253 -v-
LIST OF FIGURES _ Number Title Page 4-1 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel Plate R1608 3 Longitudinal Prop. Unirradiated and Surveillance Capsule U Data 10 42 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel Plate R1808 3 Transverse Prop. Unirradiated and Surveillance Capsule U Data 11 i 43 Tensile Properties for Seabrook Station Unit 1 Reactor Vessel Weld Metal Unirradiated and Surveillance Capsule U Data 12 44 Seabrook Station Unit No. 1 Capsule U Surveillance Results Longitudinal Charpy impact Data 13 45 Seabrook Station Unit ha. 1 Capsule U Surveillance Results Transverse Charp; Impact Data 14 46 Seabrook Station Unit No. 1 Capsule U Surveillance Results Weld Metal Charpy impact Data 15 4 4+7 Seabrook Station Unit No. 1 Capsule U Surveillance Results Heat-Affected-Zone Charpy impact Data 16 51 Seabrook Station Unit No. 1 Capsule U Surveillance Results for Piate Versus Regulatory Guide 1.99 Predictions 20 k R77\253 -vi-
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'4 ANALYSIS OF SEABROOK STATION y, NIT 1 REACTOR VESSE1 SURVilLLANCE CAPSUtE U 1.0
SUMMARY
OF RESULTS Capsule U, the first Seabrook Station, Unit 1. Reactor vessel l Surveillance Capsule was removed from the Seabrook vessel in August 1991, after 333.37 effective full power days (EfPD) of operation. The analyses of the capsule test specimens and neutron dosimetry led to the following conclusions:
. The capsule received an average neutron fluence (E>1Mev) of 3.11 X 10 18 n/cm2 . This is equivalent to the fluence which will be received at the reactor vessel inner diameter after approximately four (4.0) effective full power years (EFPY) of operation.
. The reactor vessel _ lower shell plate material, R1808 3, was included in the surveillance capsule as the limiting plate material. -For its Charpy specimens oriented in the longitudinal direction (LT), the 30 and 50 ft lb transition temperatures increased by 36'F and 34'f, respectively. The plate's transversely (TL) oriented Charpy specimens experienced increases in the 30 and 50 ft lb transition temperatures of 28'F and 20'f, respectively. The shift in the 35 mils-lateral expansion (MLE) index. temperature was 24.5'F for LT specimens and 15'F for the TL specimens.
! .- The weld metal irradiated to 3.11 X 1018 n/cm2 ' experienced 30 ft-lb and 50 ft lb transition temperature increases of 10'F and 15'F, respectively.
. The' average upper shelf energy for transversely oriented soecimens from lower shell plate R1808 3 decreased from 79 ft-lbs to 72 ft-lbs after irradiation _to the fluence of 3.11 X 1018 n/cm2 . The weld metal, exposed to the same fluence as. the plate material, experienced a decrease in upper shelf energy from 160 ft lbs to
~
129_ft lbs. The plate and weld materials exhibit upper shelf energies for continued safe plant operation. The upper shelf-energy for these materials is expected'to be maintained above 50 lb throughout vessel life as required by 10CFR50 Appendix G. R77\253 1-
. The adjusted RTNDT values for the plate and weld material, based on the surveillance capsule data, are within the two standerd deviations of Regulatory Guide 1.99 Revision 2, predictions.
a k R77\253 l
_ . _ _ _ . _ . . _ _ . _ __ __.__-___.______m t
2.0 INTRODUCTION
i The Seabrook Station Unit No.1 Reactor Vessel Radiation Surveillance Program is described in Westinghouse Report WCAP-10110, dated March 1983.1
-The program utilizes six surveillance capsules. Each capsule contains 60 Charpy V-notch specimens, 9 tensile specimens and 12 1/2T compact test specimcas. The capsules contain vessel plate material R1808 3 with Charpy specimens oriented in the longitudinal ar.d transverse airection, weld metal and HAZ material. Each capsule provides accelerated data relative to concurrent reactor vessel inner wall material condition, since the capsules are located in the reactor on the neutron shield pad between the core barrel and the reactor vessel wall, opposite the center of the core. The surveillance program meets the requirements of ASTM E 185 79, *Standare Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels."
l The first surveillance capsule in this program, designated Capsule U. was removed during the plant's first refueling outage in August of 1991. The capsule was irradiated for 333.37 effective full
- power-days (EFPD) of operation. The capsule specimens, with the exception of the compact specimens, were tested by B & W Nuclear Services Co.2 The 1/2T compact test specimens are being saved for future test needs. The capsule data is attached as Appendix B to this report. The analysis of the specimen data and dosimetry was performed by the Yankee Atomic Electric Co. This analysis is the subject of this report. l
[ i l L L l !~ l
- 11. Superscripts denote references loacted in Section 7.0 of this report. '
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l l l 3.0 SURVElllANCE MATERIALS AND CH[H13TRY B6 sed on an evaluation of the vessel plate materials, considering initial RTNDT values _ chemistry and the irradiation prediction methods of Regulatory Guide 1.99,-Revision 1, vessel lower shell plate R1808 3 was expected to have the highest end*of life RTND7. This reactor vessel surveillance material was supplied by the vessel f abricator Combustion Engineering, Inc. Additionally, Combustion Engineering, Inc. supplied a ' weldment made up of sections of lower Shell Plate R1808-3 and the adjacent Lower Shell Plate R18081. This weldment was made using Weld Wire Heat No. ' 4P6052 and Linde flux 0091, Lot No. 0145. The recctor vessel beltline weld, intermediate and lower shell longitudinal weld seems, and the intermediate to i lower shell girth welds, were all fabricated using the above weld wire / flux combination. Therefore .the weld supplied for the surveillance program is the limiting weldment. The chemical analyses heat treatment history, drop weight and RTNDT values for the materials used in the beltline region of the Seabrook Station Unit No. I are provided in Appendix A to this report. The tables are reproduced from the Westinghouse description of the Surveillance Program. WCAP 101101 . I I ( l 1
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4.0 PRE AND POST !RRADI ATION TEST RESULTS 4.1 Baseline Data. WCAP 10110 The surveillance prrgram materials were tested in the unirradiated or baseline condition by Westinghouse and these results are reported in Westinghouse report WCAP 10110.I The Charpy impact test results of the base line data reported in WCAP 10110 are shown in Table 4 1 as the zero fluence values. The tensile data is shown in Table 4 2 as the zero fluence values. 4.2 Irradiated Data. B&W Report BAW 2157 The Capsule U test specimens and dosimetry were tested by B & W. as mentioned previously. The test data is provided as Appendix B to this report. The Charpy impact results from the irradiated capsule is listed in Table 4 1 eith the unirradiated data. The irradiated Charpy data were analyzed using the EPRI Tanh Curve fitting Routine, Version 1.8 (see Section 4.3). The upper shelf energy. values were calculated using the averaging technique described in ASTM E 185 82 3 , and its current revision due for issuance in 1992. The tensile data from the surveillance capsule is listed in Table 4 2. The effects of irradiation on the tensile properties are shown graphically in Figures 4 1 through 4 3. All data were analyzed in accordance with the 1982 revision to ASTM E 1853 as specified by 10CFR50. Appendix H. " Reactor Vessel Material Surveillance Program Requirements." 4.3 Tanh Curve Fits 3 ASTM E 185-82 defines the Charpy V notch impact test transition temperature as "the difference in the 30 ft lbf (41J) index temperatures for
.the best fit (average) Charpy-curve measured before and after irradiation."
1There are two methods employed in the industry for determining the best fit Charpy curve. .The first-is to ' eye" the data and draw a best fit curve through it; the second is to use a hyperbolic tangent function (Tanh) and coF34ter fit the data to determine the best curve shape. EPRI, in conjunction i t' ndust,y experts, developed a-computer routine for fitting Charpy test s/ ith the Tanh function. The advantage to using this computer routine is
,p , .!.he curve fits are performed in a consistent manner. This reduces scatter
.itesne reported shift data. To generate the irradiated Charpy results reported in Table 4 1, the EPRI Tanh Curve Fitting Routine Version 1.8 was used on the-data reported in BAW 2157, i
j R77\261 5-l
For the $eabrook Station Unit 1 Reactor Vessel Surveillance Program, the initial (unirraciated) Charpy values for the 30 ft lb and 50 ft lb fixes (T3o an.r T$o) and for the 35 mils-lateral expansion (MLE) are documented in Westinghouse Report WCAP 10110. The irradiated values will be those generated by usint ;5e Tanh computer fit. Table 4 3 provides a comparisaq of the Westinghouse (baseline) reported values, B & W (irradiated) reported values, and the Tanh fit valuet. 4.4 Results The tensile results are presented in Table 4 2 and graphically in Figures 4 1, 4 2, and 4 3. The irradiated yield and tensile strengths increased slightly over the unirradiated values, The properties which measure ductility, reduction in area and elongation, decreased with irradiation These changes are a result of irradiation induced microstructural changes and were expected. The tensile property changes will not effect reactor vessel operation. The Charpy data, presented in Table 4 1. showed increases in the Charpy transition temperatures (T3o, TSO, MLE) and slight decreases in the Charpy _ upper shelf energies of the various materials. These results are shown graphically for the surveillance plate R1808 3 specimens in Figure 4 4 (longitudinal orientation) and figure 4-5 (transverse orientation). The weld metal results are shown' graphically in Figure 4-6 and the heat affected zone (HAZ) material is shown graphically in Figure 4 7 . All graphs of Charpy data show the curve fit using the Tanh function for both the unirradiated and irradiated data. This was done in order to produce the graphs with computer software, R77\253' -- _.- _ . - - - - --
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l TABLE 4 ?. Seabrook Station Unit 1 Surveillance Program Tensile Pr0Derties - Base line Properties Data Per WCAP 10110 Irradiated Data Per BAW-?!57 FLUENCE TEST RED. TOTAL UNIFORM E+1g TEMP. UTS .21 Y5 IN ELONG. ELONG. MATERIAL n/cm 'F K51 K51 AREA % % % PLATE 0.00 75.00 92.00 74.00 65.00 24.00 14.50 {f8083 0.00 75.00 91.00 70.00 68.00 27.00 14.50 0.00 300.00 85.00 64.00 68.00 24.00 13.00
'0.00 300.00 .85.00 65.00 66.00 24.00 13.00 0.00 550.00 89.00 63.00 61.00 24.00 14.00 O.00 550.00 89.00 63.00 63.00 24.00 13.00 0.30 70.00 94.20 73.60 63.50 23.70 9.80 0.30 300.00 86.60 67.60 65.40 20.50 8.30 0.30 550.00 91.10 67.30 62.10 19.40 7.80 PLATE 0.00 75.00 91.00 71.00 55.00 26.50 15.50 h[ O.00 75.00 91.00 71.00 55.00 26.50 15.50 0.00 300.00 86.00 66.00 53.50 21.00 12.00 0.00 300.00 86.00 66.00 f5.00 22.00 12.00 0.00- 550.00 88.00 63.00 51.50 21.00 13.00 0.00 550.00 88.00 64.00 47.50 24.00 15.50 0.30 70.00 94.10 73.30 54.30 21.40- 9.50 0.30- 300.00 85.70 67.20 55.90 17.80 7.80 0.30 550.00 91.30 66.40 48.30 15.90 7.90 WELO- 0.00 75.00 87.00 75.00 71.00 27.00 15.00
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l TABtf 4 3 A Comparison of Tanh Computed Charpy Values to Westinchouse and B&W Reported Values MATERIAL PARAMETER Tanh WESTINGHOUSE B&W Unirradiated Data Plate T 30 28'F 25'F -- R1808 3 T 4'F 0*F - LT 3bkLE -6*F O'F -- Plate T 30 9'F 10'F -- R1808 3 T50 58'F 60*F -
.TL- 35MLE 4 6'F 50'F -
HAZ- T 30 199'F 160'F - (a) T -139'F -120*F - 3bMLE 126'F 105'F - NELD- T 30 7 5'F 60'F - T 50*F -4 5'F- - 3bbLE -47'F -50*F - Irradiated Datjt Plate. T 30 II*f ** 3*I R1808 3 To $ 34'F -- 40* F LT 35MLE 24*F - 26'F Plate- T 3o- 38'F -- 39'F R1808 3 To 60'F -- 78'F TL 3bMLE 65'F - 68'F
- HAZ T 30- -104'F -
-85'F L (a)- T -60'F -
41'F 3bMLE - 57'F - 47'F
- SELD 'T 30 50*F -
- 56*F T 50 -30'F --
38'F 35MLE -32'F -- 40'F
- Note (a) HAZ data exhibit significant scatter which is typical of HAZ material,
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1 I d TENSILE PROPERTIES FOR SEABROOK STATION UNIT 1 , Figure 41. REACTOR VESSEL PLATE R1808 3 LOllGITUDINAL PROP. l UNIRRADIATED AND SURVEILLANCE CAPSULE U DATA 100 95 q {85 W_ NN/S 3 A h80
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s v) 75 - - - - y 70 65 < - e 60 0 100 200 300 400 500 600 TEMPERATURE Degrees F "O YS UNIRRADIATED UTS UNIRRADIATEDYS IRPA]ED UTS IR" TENSILE PROPERTIES FOR SEABROOK STATION ' NIT 1 REACTOR VESSEL PLATE R1808 3 LONGITUDINAL PROP. UNIRRADIATED AND SURVEILLANCE CAPSULE U DATA 80 70 , . f F- __ __ . i e 60 -
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g.30 g S 20 ek.- .,. o C -: G l 10 : - 4 , j i 0 l y 0 100 200 300 400 _ 500 600 ; L TEMPERATURE Degrees F i L UN!RR RED,IN AREA UNIRR. TOTAL ELONG. UNIRR UNIFORM ELONG. SBTLTC.DRW_ IRR T ELONG. IRR UN! FORM ELONG. IRR R$1N AMEA l l . . -- . ., u.._.- ....-._.__-.-_,;..-_.~._.-.__.__.____..._.__._.__.._....-_-
TENSILE PROPERTIES FOR SEABROOL ?TATION UNIT 1 Figur; 4 2 REACTOR VESSEL PLATE R1800 3 TRArv5 VERSE PROP. UNIRRADIATED AND SURVEILLANCE CAPSULE U DATA 100 95 e q}
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W 85 - - - - 6 0 60 -- - ~ e li> 75 - -
.s 70 H% - - -
65 M- t %_ 60 0 100 200 300 400 500 600 TEMPERATURE Dogrees F A UNiRRAt% TED UTS UNiRRADtATED YS 1RRADIAJED LTTS IRRAD TED YS TENSILE PROPERTIES FOR SEABROOK STATION UNIT 1 REACTOR VESSEL PLATE R1808 3 TRANSVERSE PROP. UNIRRADIATED AND SURVElLLANCE CAPSULE U DATA 6 I g l 1 l w~- _ 50 f M' c-h 40 0 G:, h* 2
- l y'
5 D 20 *= "- N + -
~
O
-+ !
G- _- J. i f 10 s=- j ! , 0 - 0 100 200 300 400 500 600 TEMPERATURE Degrees F UNI 9R. rep. IN AREA UNiRR. TOTAL ELONG. UNiRR. UNIFORM ELON3 IRR RED lN AREA SBTTLC DRW _g IRR. TOTgl ELONG. IRR. UN{0RM ELONG.
F TENSILE PROPERTIES FOR SEABROOK STATION UNIT 1 Figure 4 3 REACTOR VESSEL WELD METAL UNIRRADIATED AND SURVEILLANCE CAPSULE U DATA 95 - 90 As
- W -s A 85 -x %
- -/ -
T 4.
. 6 80 - -+
-q a
= f16 v>
=
70 -- x , p 65 1 45 1 60 O 100 200 300 400 500 000 TEMPERATURE Degrees F UN!RRADIATED UTS UNiRRADIATED YS IRRADIAJED UTS IRRAgTED YS TENSILE PROPERTIES FOR SEABROOK STATION UNIT 1 REACTOR VESSEL WELD METAL UNIRRADIATED AND SURVEILLANCE CAPSULE U DATA 80 .
- g. I
[ 7 60 p 8 k
, g40 -- -- 3 l F
O. O 20 f_P'-- 'w-j,
.+ I
_j [ - - - - - e% l h a ' '- --- l e i 3 0 -- 0 ieJ 200 300 400 500 600 TEMPERATURE Degrees F UN!RR rep. IN AREA UNIRR. TOTAL ELONG. UNIRR. UN1 FORM ELONG. SSTWC DRW IRR R] g AREA IRR. TO ELONG. IRR. UN! FORM ELONG.
SEABROOK STATION UNIT NO.1 Figure 4-4 CAPSULE U SURVEILLANCE RESULTS LONGITUDINAL CHARPY IMPACT DATA MILS LAT. EXP. 200 < l 180 b t 4
# 16C 2
7 140 - - -+ q ' h120 100 m -t ; 1 ., i l h 80 I - n a-# - A+~-T o-~~~ s " ~-- 7------' 60 4-] i - -pr7
}
40 1 l
-- t
--{4/NO <--
i t I l j'414 c l L } 20 1-+%V -+- t
-**. M P 0
'l=I =1. ! l (250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F UNIRRADIATED DATA CAPSULE U DATA Tanh Fit Tanh Fit o a SEABROOK STATION UNIT NO.1 CAPSULE U SURVEILLANCE RESULTS LONGITUDINAL CHARPY IMPACT DATA
'400 = ,
l ! l 180 j
-q j- m- _
160 -+ 1 l I I 140 ---; -+ t . p
-+
g 120 t i o a e 100 - 80 g 'Nwl i vi , 60 . 40 ki ci ; g , , l 20
/ 7 I I g:,W '
i 0 (250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F UNIRRADIATED CATA CAPSULE U DATA SBULDRW Tang Fit Tang Fit
t> s l SEABROOK STATION UNIT NO.1 Figure 4-5 CAPSULE U SURVEILLANCE RESULTS TRANSVERSE CHARPY IMPACT DATA MILS LAT. EXP. 200 j. 180 r 4 160 2 _,140 o g 120 .-- 100 80 60 t #$ ~ ~ "[ ~ ~~~d~ T ~~~ - 40 c^ 20 ,2 T T~ 0
' - * - " ~ ' '! '- ' ' '
(250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F 5 UN!RRADIATED DATA CAPSULE U DATA e Tanh Fit Tanh Fit O O SEABROOK STATION UNIT NO.1 CAPSitLE U SURVEILLANCE RESULTS TRANSVERSE CHARPY IMPACT DATA 200 1 l -I l { ! 180 , ;
\ \ -
160 . 140 - - L 120
$100 .
5 80 , _ ,m, _ __ 2 _s ___ _ _ 7_ _ _ _ 60 j pt
] ]'
40 f 20 o
-A 0
' "#"M'I y
'! ' ' 'l' (250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F UNiRRADIATED DATA CAPSULE U DATA SBUT.DRW Tanh Fit Tanh Fit O O
- _ - - _ _ _ - _ _ _ _ __ ._%. I
SEABROOK STATION UNIT NO.1 Figure 4-6 CAPSULE U SURVEILLANCE RESULTS WELD METAL CHARPY IMPACT DATA MILS LAT. EXP. 200 , l 180 4 4
- +- T
,@ 160 s
140 z , h120 - r ~ g a,100 -
! l -]r : I j ;
y yf ^l - c -- _ , 80 7 g gt7gi-p ;
@ 60 -
[ ] f 40 M'o o Met 0 l1 50 100 150 200 250 300 350 400 450 500 i (250)(200)(150)(100) (50) TEMPERATURE Degrees F UN1RRACIATED DATA CAPSULE U DATA Tanh Fit Tanh Fit O O SEABROOK STATION UNIT NO.1 CAPSULE U SURVEILLANCE RESULTS WELD METAL CHARPY IMPACT DATA 200 :
! ! ! l l l
--~--
180 - -
- +- ~
~
i l l l i l l l ! 160 + -- 140 ~ ~'~ ~~~ ~ ~ ~~~ ~~~
.8 {--- j , 46 T ' l 7 120 g
[i100 d' E ,,e z w " ouQ il W El 60 7
/ -
0 40 ! + o 20 4 /. -- l l 0 r- Y? .. , . (250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 50( TEMPERATURE Osgrees F UN RRADIATED DATA CAPSULE U DATA SBl.AVC.DRW Tenh Fit Tanh Fit O O
SEABROOK STATION UNIT NO.1 Figure 4 7 CAPSULE U SURVElLLANCE RESULTS HEAT AFFECT ZONE MAT'L Cv DATA MILS LAT. EXP. 200 180
/2 160 7
lE
+-
140 - i O !: g 120 100 h b go _. An0 4m r - - o- , 60 1
,-+rp l T ! I
} 40 lop j - y ~:.,cro 5 4,#!U -
8 l: l { 20 ---Ln j' /S - T~-N +
'1-
; 1 ! l ' '
0 (250)(200)(150)(100) (50) 0 50 100 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F UNIRRADIATED DATA CAPSULE U DAT4 Tanh Fit Tanh Fit o o SEABROOK STATION UNIT NO.1 CAPSULE U SURVEILLANCE RESULTS HEAT AFFECTED ZONE MAT'L. CHARPY IMPACT DATA 200 i I l
'! l '
180 -4 - l l C 160 - + O *
,; __I-140 t ;
.s !
i - o 1 g 120 i ],,'O. - ,l O a p' O f , 2 u' f' /
/ &'n l fa --+
@ 80 w 60 - -I '
* '1/ . ,' '
,O '
- 40 20
,'3 'o 0
'j' 'l '
(250)(200)(150)(100) (50) 0 50 1c; 150 200 250 300 350 400 450 500 TEMPERATURE Degrees F UNIRRADIATED DATA CAPSULE U DATA SBUHO.DRW Tanh Fit Tanh Fit O D I
5.0 SDT CALCUtATION 5.1 Calculated RT NDT Values Adjusted RTNDT values of 96*F for the R1808-3 surveillance plate and 40'F for the surveillance weld metal (see Table 4-1) were calculated f rom the surveillance capsule test results using the methodology of Regulatory Guide 1.99, Revision 2. Regulatory Guide 1.99, Revision 2,4 requires that two or more credible surveillance data sets be available before aC ual surveillance data can be used for licensing purposes. Since this is the analysis of the first capsule (only one data set), these calculated RTNDT values are presented for information only. In the absence of two or more credible surveillance data sets Section C.1 of Regulatory Guide 1.99, Revision 2 requires that the adjusted RT NDT be based on a delta RTNDT value calculated from fluence and derived chemistry factors for the material. To determine the adjusted RTFDT numbers in Table 4-1, the delta RTNDT value is the T3o shift value measured from Charpy specimen testing. This allows the surveillance data from this first capsule to be compared with Regulatory Guide 1.99 predictions (discussed in Paragraph 5.2). The adjusted RTNDT calculation used to develop Table 4-1 values is as follows: The adjusted RTNDTs are determined using R. G.1.99 equations, where, Adjusted RTNDT - Ini ti al RTNDT + T30 Shift + Margin where, Margin - 2*(sigma 42 + si gn'a3 2).5 For the plate (TL) and weld initial RTNDTs, the initial sigma margin (sigmaj) is set to zero because the initial RTNDTs were determined in accordance with ASME Code Section Ill, NB2300 which is a conservative (upper bound) method based on drop weight and Charpy data. Per Regulatory Guide 1.99, Revision 2. sigma 3 is 28*F for welds and 17'F for base metal except that sigma, need not exceed 0.5 times the mean value of T30 Shift. Therefore, sigmas for plate is 14*F (28'F + 2) and sigma3 for weld is 5'F (10'F + 2). Thus, the Table 4-1 values are: Plate: Adj. RTNDT - 40*F + 28'F + 2(14*F) - 96*F Weld: Adj. RTNDT - - 60'F + 10*F + 2 ( 5'F ) = - 4 0* F R77\253 5.2 Comparison of Surveillance Capsule Results to Regulatory Guide 1,99. i Revision 2 Predictions As described in Section 5.0, the adjusted RThDT values reported in Table 4-1 were based on Charpy T30 shift results rather than Regulatory Guide 1.99 delta RTNDT values calculated using material chemistry f actors and the fluence function. To compare the capsule results to Reg. Guide 1.99 predictions, chemistry f actors for the surveillance plate and weld were calculated using Tables 1 and 2 of Reg, Guide 1.99. Applying the chemistry factors to the Reg. Guide 1.99 fluence function, delta RTNDT curves were generated to graphically show the Reg. Guide 1.99 predictions against the T30 shif t data from the surveillance capsules. Figure 5-1 depicts the T30 shif ts for the plate L-T and T-L oriented Charpy specimens and weld specimens plotted against the Reg. Guide 1.99 prediction curve. The Reg. Guide curve has the form of: delta RT NOT (shi f t) - (CF)f(0.28 - 0.10109 O where, f - fluence (E+19), and CF - chemistry factor. For the plate, the CF is 37 (Cu .06wt%, Ni - 57wt%). For the weld, the CF is 23.5 (Cu .02wt%, Ni .10wtt). The surveillance capsule, transverse plate data (orientation used for RT NDT determination) shown at the top of Figure 5-1 agrees well with the Reg. Guide 1.99 prediction line. The longitudinal data point falls above the prediction line. Both points fall below the two sigma margin limit specified by Regulatory Guide 1.99 indicating good agreement with Reg. Guide 1.99 predictions. The weld metal results are shown at the bottom of Figure 5-1. Fiere the weld Charpy shif t result f alls slightly below the prediction line. Again, the test data point falls below the two sigma margin indicating good agreement with Regulatory Guide 1.99 predictions. The +2 sigma margins are shown as the dashed lines in Figure 5-1. As explained in Section 5.1, the one sigma values for the analysis are 14*F for plate and 5'F for weld metal . 5.3 Effect of Surveillance Results on Plant Current Technical SDecifications The current Seabrook Technical Specifications contain heatup, cooldown, and LTOP curves which are based on the predicted RTNDT shif t in Regulatory R77\253 . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
k$ Guide 1.99. Revision 1, at 16 EFPY, A Technical Specification amendment which bases the RTNOT prediction on Revision 2 of the Regulatory Guide will be submitted in the third quarter of 1992. The pressure temperature curves in the amendment are identical to those in the current Technical Specifications, however, the applicability of the curves has been reduced to 11 EFPY due to the conservatism in the revised Regulatory Guide. These curves will be re-evaluated upon analysis of the second surveillance capsule which is scheduled for removal af ter approximately 5 EFPY of operation. At that time, two sets of irradiated material data will be available which is the minimum number of data sets required by Regulatory Guide 1.99, Revision 2, to establish credibility of the RTNDT shif t values.
+s R77\253 _ _ __ _ _ - - __ - __________________ - __________- ___ - _ - __ _ ____________ - ________________A
SEABROOK STATION UNIT 1 Figure 5-1 . CAPSULE U SURVEILLANCE RESULTS FOR PLATE VERSUS R. G.1.99 PREDICTIONS 100 u. g 80 I + 2; sigma margin
; t
~~,,,_,,4 p l j ' p ___ )... -- - -
o : ;
\ -Y___ s ! \
c 60 d-- I
- u. ' - R. G.1.99 Prediction !
g ,- i i ! E , A,- ,-- j g 40 -+
/
tr A/ a 20 -+ + T w O . ! i
! I O
O 0.5 - 1 1.5 2 2.5 3 3.5 4 FLUENCE E+19 n/cm2, E>1Mev PLATE PLATE LT TL a A PLATE CF = 37 SEABROOK STATION UNIT 1 CAPSULE U SURVElLLANCE RESULTS FOR WELD METAL VERSUS R. G.1.99 PREDICTIONS 100 , l !' *
; , i
- u. ! I
! l
{ 80 h ! ! [ 6
- 4
- o. i 60 - 4- b
$ + 2 sig a margin R. G.1.99 Prediction 5 c .. L -- !-- ----r------'
] . 40 j -*_ i6__j_ r. l !'
g ,y . l - - -l / I s
$a
. 20 ,
,- ' ' { ___ -'F i l w '
, I i
C e l i l
\ \ l 0
O 0.5 1 1.5 2 2.5 3 3.5 4 FLUENCE E+19 n/cm2, E>1Mev WELD SH FT
- SBRGPW.DRW WELD CF = 23.5
6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6,1 introduction The characterization of the neutron environment with the reactor pressure vessel and surveillance capsule geometries is necessary to properly
>ssess the effects of neutron irradiation on these components. The neutron energy spectrum, flux and fluence to which these components have been exposed are necessary elements in the analyses of neutror damage to the steels used in these componca s. In addition, a relationship must be determined which will relate the changes observed in the test specimens to present and future conditions of the reactor. To characterize the flux in the capsule requires a combination of analysis and measurement of the neutron flux monitors ~
(dosimeters) contained in each surveillance capsule. The prediction of future conditions of the reactor is based on analysis. The conversion of the measured dosimeter activity to flux for Seabrook Capsule U will be based on the use of spectrum averaged cross-sections derived from calculations performed on plants similar to Seabrook. This approach provides a good approximation to the flux experienced by the dosimeters. ' A Seabrook-specific two dimensional discrete ordinate transport model is being developed. This Seabrook specific model will be utilized to characterize the flux seen by Capsule V, to analyze future surveillance capsules, and to predict future conditions of the ri, 'or vessel. This section describes the analysis of the dosimeters and the development of the two-timensional model.
~
6.2 Neutron Calculetions A Seabrook-specific two-dimensional model is being developed for future flux and fluence calculations. The calculation will be a forward transport calculation ir r,0 geometry using the DORT two-dimensional discrete ordinates codes and the SAILOR 6cross section library. The SAILOR library is a coupled, self-shielded, 47 neutron. 20 gamma-ray, P ,3 '_NDF/B-IV based cross-section library for light water reactor apolications. In the 00RT analysis, anisotropic scattering will be treated with a P3 expansion of the cross-sections and an S8 angular quadrature. The reactor geometry being developed includes a description of the radial regions internal to the primary concrete (core barrel, neutron pad. pressure vessel, and water annuli) as well as the surveillance capsule and the appropriate reactor core fuel loading pattern and power distribution. Thus, distortions in the fission spectrum due to the attenuation of the reactor internals will be accounted for in the analytical approach. R77\253 - -_ _ _ _ _ - _ _ _ _ _ _ _ __ _ _ __ - _ _ _ _ _ - - _ _ _ _ _ _ _ - - _ _ _ _ _ - _ - _ _ _ _ - _ _
6,3 Neutron Dosimetry Analysis In order to effect a correlation between fast neutron (E > 1.0 MeV) exposure and the radiation induced properties changes observed in the test specimens, a number of fast neutron flux monitors are included as an integral
- part of the reactor vessel surveillance program. In particular, the surveillance capsules contain dosimeters employing the following reactions.
Fe54 (n. P) Mn54 Ni S8 (n. P) CoS8 Cu63 (n, a) Co60 Np237 (n f) CsI37 U238 (n f) CsI37 The activity of each dosimeter is determined using established ASTM procedures. 7 20 Measurements of the dosimeters activity is reported in BAW-2157.2 Given the measured activity, the determination of the neutron flux proceeds with the calculation of the saturation activity from: A=A 3 ( )(1-e )e -td where: A - the measured activity corrected to end of power operation A s
- saturation actit Ry P
3
- core thermal power during irradiation period i P - reference core thermal power or 3411 MWt 1- - decay constant for the radioactive nuclide th t$ -
duration of the i period td - decay time following i th period Using the saturation activity, the reaction rate, R is determined from: A R- Y R77\253 _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . .. ._
where Y is the yield and N is the number density and is calculated as: N N _g. f . l l where: ) N, - Avogadro's number, M - atomic weight of parent nuclide, f - abundance of the parer t nuclide. As provided in Reference 12, the neutron fluence rate (flux) can be calculated as follows:
+-Ji0 where E is the spectrum averaged cross section. The spectrum averaged cross section for E > 1 Mev is calculated as:
fo(E)$(E)dE 30 - [o(E)dE 1 Hey 6.4' Results Using the procedures in Section 6.3, the flux for E > 1 Mev can be calculated.1The nuclear parameters- for the dosimeters are provided in Table 6-1. The irradiation period. power ratio for the irradiation period and [ decay time-are.provided in Table 6-2. The decay time reflects the correction
-of the measured activity to reactor shutdown. Table 6-3 provides the measured =
activity. saturation activity and reaction rate for -the dosimeters. Seabrook l is the same class as a number of Westinghouse reactors.21-25 In this case, class is defined by the location and dimensions of the core barrel, reactor
.R77\253 i.
4
-vessel, and surveillance capsules. Spectrum average cross sections for this class of plants has been calculated by Westinghouse 21 25 and are provided in
- Table 6 4 for the dosimeters of interest. Using the measured reaction rate
-and the spectrum-averaged cross -section, the flux is calculated and provided in Table 6-5. With the flux and the Seabrook irradiation period of 333.37 effective full-power days or 0.91 effective-full-power-years the fluence of each dosimeter was calculated and-is presented in Table 6 5. The_ average 18 value of all dosimeters is 3.11 x 10 This value compares favorably to the other Westinghouse reactors 21 25 in the Seabrook class as presented in Table 6 6.
R7 7 \2.53 -4"-
-TABLE 6-1 Nuclear Parameters for Neutron flux Monitor!
Target Fission Monitor Reaction Weight Response Product - Yield
- Material of Interest Fraction Range Half-Life (%) ,_
-Copper Cu63(n u)Co60 0.6917 E>4.7 MeV 5.272 years -Iron Fe54(n.p)Mn54 0.058 E>1.0 MeV 312.5 days Nickel NiS8(n.p'CoSB 0.6827 E>1.0 MeV 70.82 days Uranium-238* U233(n f)CsI37 1.0 E>0.4 MeV 30.17 years 6.0 ' Neptunium-237* Np237(n.f)CsI37 1.0 E>0.0R MeV 30.17 years 6.5
'cDenotes that monitor is cadmium shielded.
-R77\253
(. , i TABLE 6 2 Irradiation History of Neutron Sensors Contained in Capsule U 1rradiation . Irradiation Deca) Period ,P,,i , Pi/P Titne_ ( d a y s ) Tirne (days) 3/90- 41 0.012 11' 481 4/90 174 0.051 30 451 5/90. 48 0.014 31 420 6/90 771 0.226 30 390-17/90 2060 0.604 31 359 8/90 .2702 0.792 31 328-9/90 3217 0.943 30 298
'10/90 -2981 0.874 31 267 11/90 1569 0.460 30 237 12/90 3408 0.999 31 206 1/91 3408 0.999 31 175 2/91 2453 0.719 28 147 3/91 3138 0.920 31 116 4/91 2688 0.788 30 86 5/91 3408 0.999 31 55 6/91 2736 0.802 30 25 7/91 2159- 0.633 25 0 P - 3411'MWt R77\253 - _ ,
f 4
-TABLE 6 3
-Measured Monitor Activities and Reaction Rates for CaDsule U Moiiitor_ and ' Measured Activity Saturated Activity Reaction Rate
' Axial-location (dis /sec-om) (dis /sec-am) (reactions /see atom)
Cu63(n.a)Co68 Top 5.15 x 10 4 4.71 x 10 5 7.18 x 10'17 Middle- 4.67 x 10 4 4.27 x 10 5 6.51 x 10'17 4 Bottom 4.60 x 10 4.20 x 10 5 6.40 x 10'17 4 5 Average 4.81 x 10 4.39 x 10 6.69 x 10'17 54 Fe34(n.D)Mn _ Top 2.13 x 10 6 4.43 x 10 6 ~ 7.08 x 10-15 Middle 1.92 x 10 6 3.99 x 10 6 6.37 x 10-15 6 6 Bottom 1.87 x 10 3.88 x 10 6.20 x 1015
-Average 1.97 x 10 6 4.10 x 10 6 6.55 x 1015
-N1 58(n.D)CoS8 Top 5.39 x 10* 6.80 x 10 7 9.71 x 10-15 Middle -4.97 x 10 6 6.27 x 10 7 8.95 x 10-15 Bottom 4.88 x 10 0 6.16 x 10 7 8.79 x 1015 5.08 x 10 6 Average 6.41 x 10 7 9.15 x 10-15
'U238(n.f)Csl37 Middle. 1.25 x 10 5 . .
Corrected
- 1.07-x 10 5 5.25 x 10 6 3.46 x 10'14 N 237(n' f)Csl37 Middle - - -
ocorrected by 0.85 for U-235 fissions and Pu build in. R77\253 ,
TABLE 6 4 SDectrum Averaced Reaction Cross-Sections For Use in Fast Neutron Dosimetry Evaluations Reaction o (barns)* Cu63(n.a)-Co60 0.00070 Fe64(n',p)Mn54 0.0583 NiS8(n.p)CoS8 0.0790 U238(n.f)Csl37 0.320 Np237(n,f)Csl37 3.30
- Values from References 21-25 R77\253- .
-c -. 8
. j TABLE 6-5 Results of Fast Neutron ,
00simetry for Caosule U Measured 6 (E > 1.0 MeV) (n/cm2 -sec) o (E > 1.0 MeV) (n/cm2 ) Reaction Rate Reaction (reactions) Measured Measured Cu63(n.a:Co 60 6.69 x 10"17 0.96 x 10 11 2.75 x 10 18 Fe64(r. ,)Mn54 6.55 x 10;15 1.12 x 10 11 3.24 x 10 18 NiS8(n.p)CoS8 9.15 x 1015 1.16 x 10 11 3.34 x 10 18
.U238(n,f)Cs137 3.46 x 10'14 1.08 x 10 11 3.11 x 10 18 Np237(n f)Cs137 - - -
Average 1.23 x 10 11 3.11 x 10 18
~R77\253 .
4 TABLE 6-6 , Comparison of-Average Fluence for Capsule U to Other Westinghouse Reactors IRRADIATION TIME FLUENgE U<' (EFPY) (n/cm ) Callaway Unit 1 1.05 3.27 x 10 18 Catawba Unit 1 0.79 3,08 x 10 18 Wolf Creek 1.08 3.39 x 10 18 Byron Unit 1 1.15 3.50 x 10 18 Braidwood Urit 1 1.10 3.79 x 10 18 Seabrook 0.91 3.11 x 10 18 l l l i i I i l l l R77\253 : l' l
s. J7 . 0 REFERENCES
- 1. Singer, L. R., "Public Service Company of New Hampshire Seabrook Station Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP 10110.
Westinghouse Electric Corporation, March 1993.
- 2. Lowe, A. L., et. al., " Test Results of Capsule U Public Service Company of New Hampshire, New Hampshire Yankee Division Seabrook Station Unit No.1," BAW 2157, B&W Nuclear Service Company, May 1992.
- 3. ASTM Designation E185-82, " Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 4. U.S. Nuclear Regulatory Commission Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2. May 1988.
- 5. "An Updated Version of the 00T One- and Two-Dimensional Neutron / Photon Transport Code," ORNL-5851, July 1982.
- 6. "0RNL RSIC Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma Ray. P3, Cross Section Library for Light Water Reactors".
- 7. ASTM Designation E482-89,
- Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance," in ASTM Standards.
Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984
- 8. ASTM Designation E560-84, "Stendard Recommended Practice for Extrapolating Reactor Versel Surveillance Dosimetry Results " in ASTM Standards, Section 12, /merican Society for Testing and Materials, Philadelphia, PA, 1984, i 9. ASTM Designation E693 79, " Standard Practice of Characterizing Neutron Exposures:in Ferritic Steels in Terms of Displacements per Atom (dpa)."
in ASTM Standards Section 12 American Society of Testing and Materials.-Philadelphia, PA, 1984.
- 10. ASTM Designation E706-87, " Standard Master Matrix for Light Water
( Reactor Pressure Vessel Surveillance Standards," in ASTM Standards, ! Section 12, American Society for -Testing and Materials. Philadelphia, PA, 1984. L
- 11. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results," in ASTM Standards, Section'12 American Society for Testing and Materials, Philadelphia, PA, 1984.
R77\253 31- ! l l
12 -. ASTM Designation E261-90,
- Standard Method for Determining Neutron Flux, fluence. and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Matericis, Philadelphia, PA, 1984,-
- 13. -ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American-Society for Testing and Materials, Philadelphia, PA, 1984.
- 14. ASTM Designation E263-88, Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron," in ASTM Standards. Section 12 '
American Society for Testing and Materials, Philadelphia, PA, 1984.
- 15. A5fM Designation E264-87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 16. ASTM Designation E481-86, " Standard Method for Measuring Neutron Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materi:li, f hiladelphia, PA, 1984.
-17 . ' ASTM Designation E523 87, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivatici of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 18. ASTM Designation E704-90, " Standard Method for Meduring Reaction Rates by Radioactivation of Uranium-238,' in ASTM Standards, Section 12
-American Society for Testing and Materials, Philadelphia, PA,1984.
- 19. ASTM Designation E705-90, " Standard Method-for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American-Society for Testing and Materials, Philadelphia,
'PA, 1984.
- 20. ASTM Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12 American Society for Testing and Materials, L Philadelphia, PA. 1984.
- 21. " Analysis of Capsule U from the Union Electric Company Callaway Unit 1
; Reactor Vessel Radiation Surveillance Program," WCAP-11374. November i I
1986.
- 22. " Analysis of Capsule Z from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11527, June 1987.
R77\253 l 1 L j
v
- 23. " Analysis of Capsule U from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program,' WCAP 1153, August 1987.
- 24. " Analysis of Capsule U from the Commonwealth Edison Co. Byron Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11651, Ncvember 1987.
- 25. " Analysis of Capsule U from the Commanwealth Edison Company Braidwood IJnit 1 Reactor Vessel Radiation Surveillance Program." WCAP-12685.
August 1990. l-l- 1 I I l l. l l R77\253 l l
- . . . .. , . _ . - - ~ = . . ..- - . - . . . -.
=
+ i .
--4 ., 3 k
. APPENDIX A-l Unirradiated Vessel Plate and Weld Data 3
E
. R71\253 I
Dr P ( f' T T*A'1
1 4 TABLE A-1 Chemical Analysis of the lower Shell Plates Used in the Core Region of the Seabrook Station Unit No. 1 Reactor Pressure Vessel Chemical Composition (a) (weight %) Plate R1808-1 Plate R1808-2 Plate [b] R1808-3 C .22 .22 .20 Mn 1.39 1.36 1.45 P .005 .007 .007 S .010 .012 .010 Si .22 .21 .24 Ni .58 .57 .57 Mo .58 .56 .55 Cr .04 .03 .06-Cu .05 .05 .06 Al .017 .021 .028 Co .009 .009 .010 Pb (c) (c) (c) W <.01 <.01 <.01 Ti <.01 <.01 <.01 Zr <.001 <.001 <.001 V .004 .004 .003 Sn .001 .002 .011 As .002 .001 .006 Cb <.01 <.01 <.01 N.2 .007 .008 .008 8 -<.001 <.001 < 001 (a) Chemical Analysis by Combustion Engineering. Inc. (b) Surveillance Program test plate. IC3 Not detected. R77\253 A-1
i, l TABLE A 2
' Chemical Analysis of the Intermediate Shell Plates Used in the Core Region of the Seabrook-Station Unit No. 1 Reactor Pressure Vessel Chemical Composition (a)
.(weight %)
Element Plate R1806-1 Plate R1806-2 Plate R1806-3 C .25 .22 .21 Mn 1.47 1.33 1.33 P .012 .007 .007 S .012 .009 .012 L Si .22 .21 .23 Ni .64 .65 .65 l Mo .59 .59 .58 Cr .08 .04 .03 Cu .04 .05 .07 Al .024 .018 .027 Co' .014 .012 .012 Pb (b) (b) (b) W: .02 .01 .02 Ti <.01 <.01- <.01
-Zr- .001 .001 .001 i= V .006 .004 .004 Sn .003 .004 .004 As .006 .007 .07 Cb. <.01 -< 01 <.01 L N 2
.010 .008 .010 B -<.001 <.001 <.001 l
[a] l Chemical Analysis'by Combustion-Engineering, Inc. (b). Not detected. R77\253- A-2
TABLE A-3
-Chemical Analysis of the Weld Metal hed in the Core Region of the Seabrook Station Unit No. 1 Reactor Vessel Intermediate Intermediate $Mell Lower Shell To Longitudinal Seam -Longitudinal Sear Lower Shell (101-124A.S.&C))'I (101-142A.8.1C)y'I Girth Weld 'eam (101-171)I*I Chemical Corposition (weight 1)
Element C Mn P $ St N1 Me tr ! Cu Al Co Pb W T1 Ir V Sn A* ' CD N 7 8 dire / Flux .13 1.24 .008 1009 .12 .02 . 50 .01 .07 -- -- -- -- -- --
.004 -- -- --
.009 --
id TestOpDI Sample CE Wald .15 1.35 .007 .008 .14 .10 - .53 .03 .02 <.001 .008 < 001 <.01 <.01 <.001 .004 .003 .001 <.01 .01 <.001 Test
.C*IDplect D tbid Test .14 1421 .010 .005 .14 .05 .50 .025 .02 .004 .005 <.001 <.002 <.007 <.002 .002 .004 .006 <.002 .000 <.001 k
Ilofc1
.g.
[a] Weld Wire Heat No. 4P6052. Linde 0091 flux. Lot No. 0145. (b) Combustion Engineering. Inc. Certification Repo-ts. ICI Westinghouse Analysis. R77\253 A-3 l
TABLE A-4 Heat Treatment History of the Seabrook Station Unit No. 1 Reactor Pressure Vessel Core Recion Shell Plates and Weld Seems Material Temperature ('f) Time (hr) Cooling Lower Shell Plates Austenitizing: 4 Water quenched R1808 1-2-3 1600 1 25 Tempered: 4 Air cooled 1225 i 25 Stress Relief: 16 Furnace-cooled 1150 1 50 Intermediate Shell Austenitizing: 4 Water-quenched Plates R1806 1-2-3 1600 i 25 Tempered: 4 Air-cooled 1225 1 25 Stress Relief: 16.5 Furnace cooled 1150 1 50 Lower Shell Plate Stress Relief: 16 Furnace cooled Longitudinal Seam Welds 1150 1 50 Inter. Shell Plate Stress Relief: 16.5 Furnace-cooled Longitudinal Seam Welds 1150 1 50
-Intermediate to Lower Stress Relief: 12.50 Furnace-cooled Shell Girth Seam Weld 1150 i 50 Surveillance Program Test Material Surveillance Program Austenitizing: 4 Water-quenched Test Plate R1808-3 1600 i 25 Tempered: 4 Ai r-cool ed 1225 i 25 Stress Relief:[a] 16.25 Furnace-cooled 1150 i 50
~
Weldment Stress Relief:[a] 17 Furnace-cooled 1150 1 50 I"3 The stress relief heat treatment received by the surveillance test plate and weldment have been simulated. R77\253 A-4
TABLE A 5
^
T*DTA 31NOT. and Upper Shelf Energy for the Seabrook Station Unit No. 1 Reactor Pressure Vessel Core Region Shell Plates and We'id Metal Upper Shelf [a] T hDT II RT NDT I* Energy Material (*C) ('F) ('C ) (*f) (J) (ft lb) Lower Shell Plates: R1808-1 -34 30 4 40 106 78
-29 -20 -12 10 104 77 R1808 R1808-3 2[b] 29 -20 4 40 106 78 l Intermediate
! Shell Plates: R1806-1 -34 -30 4 40 111 82
.R1806-2 -34 -30 -18 0 138 102 R1806-3 -40 -40 -12 10 156 115 Intermediate Shell and Lower Shell Long t dinal -51 -60 -51 -60 212 156 Seams-WeldMetallCyId3 and the Intermediate to
' Lower Shell Girth Seam-Weld Metal [c][d]
[a] Data by Combustion Engineering, Inc. (b) Surveillance Program test plate. ICI _W eld Metal Heat No. 4P6062, Flux Type 0091, and Lot No. 0145. L [d] Combustion Engineering Surveillance Weld Test Plate "C 'ertification Report. I R77\253 A-5
. l
-APPENDIX B BAW 2157 Test Results of Capsule U C
R77\253
9 BAW-2157 May 1992 TEST RESULTS OF CAPSULE U PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE NEW HAMPSHIRE YANKEE DIVISION SEABROOK STATION UNIT NO. 1 Reactor Vessel Material Surveillance Program -- by A. L. Lowe, Jr., PE R. E. Napolitano W. R. Stagg B&W Document No. 77-2157-00 (See Section 6 for document signatures) B&W NUCLEAR SERVICE COMPANY Engineering and Plant Services Division P. O. Box 19935 Lynchburg, Virginia 24506-0935 13W!!senihr
SUMMARY
This report describes the results of the testing of the specimens from the first capsule (Capsule U) of the Public Service Company of New Hampshire Seabrook Station Unit No. I reactor vessel surveillance program. The objective of the ! prograr H to monitor the effects of nectron irradiation on the tensile and fractur < tghness properties of the reactor vessel materials by the testing and evaluation of tension and Charpy impact specimens. The program was designed in i accordance with the requirements of 10CFR50, Appendix H, and ASTM Specification E185-79. The results of the tension tests and the Charpy impact test results indicated that the materials exhioited normal behavior relative to the estimated neutron fluence exposure.
- ii -
13W!!sWahr
~ ---- . - - - - . -- -- --. . - - - - -.--
CONTENTS
?vue
- 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1
- 3. POST-IRRADIATION TESTING . . . . . . . . . . . . . . . . . . . . . 3-1 3.1. Visual Examination and Inventory . . . . . . . . . . . . . . 3-1 3.2. Thermal Monitors ......................3-1 3.3. Tension Test Resul ts . . . . . . . . . . . . . . . . . . . . . 3-1 :
3.4. Charpy V Notch Impact Test Results . . . . . . . . . . . . . 3-2
- 4. DOSIMETER MEASUREMENTS . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Introduction ........................4-1 4.2. Dosimeter Preparation . . . . . . . . . . . . . . . . . . . 4-1 4.3.' Quantitative Gamma Spectrometry . ._. .. . . . . . . . . . . . . 4-2 5._ REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1
- 6. -CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 List of Tables i . Table -
L . l 3 1. Tensile Properties Irradiated Base Meta ~i and Weld Metal from Capsule U . . . ....., . . . . . . ... , , . ._. . . . . . . . . . 3-3
- 2. Charpy Impact Results-for Irradiated Base Metal, Longitudinal (LT) ,
! Orientation,-from Capsule U . . . . . . . . . ... . . . . . . . . 3-4 " 3-3. -Charpy Impact Results for Irradiated Base Metal, Transverse (TL) Orientation, from Capsule.U . . . . . . . . . . . . . . . . . . 3-4
.3-4. Charpy Impact Results for Irradiated Heat-Affected Zone Metal, from Capsule U .......................... 35 L 3-5. :Charpy Impact Test Results for Irradiated Weld Metal, Transverse ,
l -(TL) Orientation, from Capsule U . . . . . . . . . . . . . . . 3-5 L 41 ' Copper Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1- . .:.-. . . . . . . .- , . . . . . . . . . . . . . . . 4-4 4-2. Iron Dosiretry, Measurements from Capsule U Seabrook Station Unit Mo. 1 . . . . . . . . . . . . .. ... . . . . . . . . . . 4-5
- iii -
BWMMf%r
Tables (Cont'd) l Table Page 4-3. Cobalt Dosimetr.v Heasurements from Capsule U Seabrook Station Unit No. 1 ........................... 46 4 4. Cadmium Shielded Cobalt Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 . . . . . . . . . . . . . . . . . . . 4-7 . 4-5. Nickel Dosimetry Measurements from Capsule U Seabrook Station l Unit No. 1 ........................... 48 4 6. Uranium-238 Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 .......................4-9 4-7. Neptunium-237 Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1 ....................... 49 i List of Fiaures Figure 3-1. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal Longitudinal Orientation ... 36 3-2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal, Transverse Orientation . . . . 3-7 3-3. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Weld Metal, Transverse Orientation . . . . 3-8 3-4. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KL2, Tested at 70F . . . . . . . . . . . . . . . . 3-9 3-5. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specinen No. KL3, Tested at 300F . . . . . . . . . . . . . . . . 3-9 3-6. Tension Test Stress-Strain Curve for Base Metal Plate R1808 3, Specimen No. KL1, Tested at 550F . . . . . . . . . . . . . . . . 3-10 3-7. Tension Test Stress Strain Curve for Base Metal Plate R1808-3, Specimen No. KT2, Tested at 70F , . . . . . . . . . . . . . , . 3-10 3-8.- Tension Test Stress-Strain Curve for Base Metal Plate R1808-3, Specimen No. KT1, Tested at 300F . . . . . . . . . . . . . . . . 3-11 l 3-9. Tension Test Stress Strain Curve for Bate Metal Plate R1808-3, l Specimen No. KT3. Tested at 550F . . . . . . . . . . . . . . . . 3-11 3-10. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KW3, l Tested at 70F ,..... . . . . . . . . . . . . . . . . . . 3-12 l 3-11. Tension Test Stress-Strain t.urve for Weld Metal, Specimen No. KWl, Tested at 300F . . . . . . . . . . . . . . . . . . . . . . . . . 3-12 3-12. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. KW2, Tested at 550F . . . . . . . . . . . . . . . . . . . . . . . . . 3-13 3-13. Charpy Impact Data for Irradiated Plate Material, R1808 3, Longitudinal Orientation . . . . . . . . . . . . . . . . . . . . 3-14 3-14. Charpy Impact Data for Irradiated Plate Material, R1808-3, Transverse Orientation . . . . . . . . . . . . . . . . . . . . . 3-15 3-15. Charpy Impact Data for Irradiated Plate Material, R1808-3, Heat-Affected Zone . . . . . . . . . . . . . . . . . . . . . . . 3-16 iv - BW!!nEY5%
. 1 i
Fioures (Cont'd) Figure Page 3-16. Charpy Impact hta for Irradiated Weld Metal . . . . . . . . . . 3-17 3-17. Photographs of- ; harpy Impact Specimen Fracture Surfaces - I Plate Material'l.angitudinal Orientation . . . . . . .. . . . . . 3-18 318. Photographs of Charpy Impact Specimen Fracture Surfaces - Plate Material Transverse Orientation . . . . . . . . . . . . . 3-19 ' 3-19. Photographs of Charpy Impact Specimen Fracture Surfaces - - Plate Material, Heat-Affected Zone . . . . . . . . . . . . . . . 3-20 3 20. Photographs of Charpy Im Weld Metal . . . . . . . pact. Specimen Fracture Surfaces -
. . . . . . . . . . . . . . . . . . . 3 21 I
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- 1. INTRODUCTION This report describes the results of the testing of the specimens from the first capsule (Capsule U) of the Public Service Company of New Hampshire Seabrook Station Unit No. I reactor vessel material surveillance program (RVSP). The capsule was removed and evaluated after being irradiated in the reactor as part of the Reactor Vessel Materials Surveillance Program. The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pressure vessel materials under actual operating conditions. The surveillance program for Seabrook Station Unit I was designed and furnished by Westinghouse Electric Corporation (H). The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel. The surveillance program for Seabrook Unit I was designed in accordance with E185-79' and thus is in compliance with 10CFR50, Appendixes G' and H'.
1-1 i IBW!!nnnifaar
- 2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled reactors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of such low alloy ferritic steels as SA533, Grade B, used in the fabrication of the Seabrook Station Unit I reactor vessel, are well characterized and documented in the literature. The low alloy ferritic steels used in the beltline region of reactor l
vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the Charpy upper shelf energy value. Appendix G to 10CFR50, " Fracture Toughness Requirements, specifies minimum fracture toughness requirements for the ferritic materials of the pressure- , retaining components of the reactor coolant pressure boundary (RCPB) of t]ater-cooled power reactors, and provides specific guidelines for determining the pressure temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix G to 10CFR50 became effective on August 13, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date. I l L 2-1 BW!!nEVn% 1
Appendix H to 10CFR50, " Reactor Vessel Materials Surveillance Program Requirements,"* defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor , vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life. A method for guarding against brittle fracture in reactor pressure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section Ill,
" Nuclear Power Plant Components.** This method utilizes fracture mechanics concepts and the reference nil-ductility temperature, RTuor, which is defined as the greater of the drop weight nil-ductility transition temperature or the temperature that is 60F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RTuo, of a given material is used to index that material to a reference stress intensity factor curve (Km curve), which appears in Appendix G of ASME Section 111. The Km curve is a lower cound of dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Km curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
The RTuor and, in turn, the operating limits of a nuclear power plant, can be adjusted to account for the effects of radiation on the properties of the reactor vessel materials. The radiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which a surveillance capsule containing prepared specimens of the reactor ves;e1 materials is periodically removed from the oper-" ng nuclear reactor and the specimens are tested. The increase in the l Charpy V-notch 30 ft-lb temperature is added to the original RTuor to adjust it l for radiation embrittlement. This adjusted RTuoi is used to index the material l l 2-2 13W!!nEV5 % t 1
to the K,n curve which, in turn, is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials. 1 2-3 S W !!s M as h v
- 3. POST-IRRADIATION TESTING l
3.1. Visual Examination and Inventory All specimens were visually examined and no signs of abnormalities were found.
-The contents of the capsule were inventoried and found to be consistent with the surveillance program report inventory. There was no evidence of rust or of the penetration of reactor coolant into the capsule. The compact fracture toughness specimens and three point bend bar were stored for future disposition.
- 3.2. Thermal Monitors Surveilla ce Capsule U contained temperature monitor sets in each of three holder blocks. Tti holder blocks each contained one thermal monitor. The monitors located at tf e top and bottom of the capsule are designed to melt at 579F and the l monitor located at the midpoint of the capsule is designed to melt at 590F, The holder blocks were radiographed for evaluation. None of the three sets of ,
thermal monitors exhibited any. signs of melting. From these data, it was concluded that the irradiated specimens had been exposed to a maximum temperature ;
- of less. than 579F during the reactor vessel operating period. This is not '
'significantly _ greater 1 than the nominal inlet temperature of 558F, and is considered acceptable for inclusion of _the data in the general pool of irradiated L . surveillance data. There appeared to be no significant signs of a temperature gradient along the capsule length. ;
. L3. Tension Test Results '
- The results of the- postirradiation tension tests are presented in Table 3-1.
Tests were performed on specimens at room temperature, 300, and 550F, They were L tested on a computer controlled 55,000-lb load capacity HTS servohydraulic test l . nachine at1a crosshead speed of 0.005 inch per minute to yield point and lthereafter 0.040 inch per minute. A 4-pole extension device with a strain gaged i extensometer was used to determine the 0.2% yield point'. Test conditions were 3-1 BWlinM%
.. = . . - . - _-._ . - :
1 in accordance with the applicable requirements of ASTM A370-77.5 for each material type and/or condition, specimens were tested at room temperature, 300, and 550F to correspond to the unirradiated material test temperatures. The tension-compression load cell used had a certified accuracy of better than +0.5% of full scale (25,000 lb). Photographs of the tension test specimen fractured surfaces are presented in Figures 3-1 through 3-3. In general, the ultimate strength and yield strength of the material increased with a corresponding slight decrease in ductility as compared to the unitradiated values; both effects were the result of neutron radiation. The type of behavior observed and the degree to which the material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were exposed. The stress-strain curves from the irradiated tension tests are presented in Figures 3-4 through 3-12. 3.4. Charov V Notch Imoact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material are presented in Tables 3-2 through 3-5 and Figures 313 through 3-16. Photographs of the Charpy specimen fracture surfaces are presented in Figures 3-17 through 3-20. The Cherpy V-notch impact tests were conducted in accordance with the requirements of ASTM E23-88' on an Satec SI-lK impact tester certified to meet Watertown standards. The data show that the materials exhibited a sensitivity to irradiation within the values to be expected from their chemical composition and the fluence to which they were exposed. 3-2 IBW!!nEY5%
Table 3-1. Tensile Properties Irradiated Base Metal and Weld Metal from Capsule U Strenath, est Fracture Properties Elongation. % Reduction Specimen Test Temp, load, Stress, Strength, in Area, ifo. F Yield Ultimate 1bs osi osi Uniform Total % Base Metal. R1808-3. Loncitudinal KL2 70 73,600 94,200 2901 162,000 59,100 9.8 23.7 63.5 KL3 300 67,600 86,600 2877 169,500 58,600 8.3 20.5 65.4 KL1 550 67,300 91,100 3009 161,600 61,300 7.8 19.4 62.1 Base Metal. R1808-3. Transverse KT2 70 73,300 94,100 3374 150,400 68,700 9.5 21.4 54.3 KTl 300 67,200 85,700 2968 137,100 60,500 7.8 17.8 55.9 KT3 550 66,400 91,300 3643 144,800 74,800 7.9 15.9 48.3 Weld Metal. Transverse KW3 70 76,000 90,000 2556 189,600 52,100 8.9 23.7 72.5 KW1 300 70,500 83,700 2565 184,600 52,300 7.5 21.2 71.7 KW2 SrC 69,400 87,700 2645 175,600 53,900 4.9 18.0 69.3
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Table 3-2. Charpy Impact Results for Irradiated Base Metal, . Lonoitudinal (LT) QI.igntation from Caosule U Test Impact lateral shear l Specimen Temperature Energy Expansion Fracture : ID F ft-lbs Inch % KL13 -40 11.0 0.009 0 KL2 0 28.0 0.025 25 + KL15 20 34.0 0.031 25 KL6 40 62.0 0.047 50 KL14 40 52.5 0.042 30 KL1 70 102.5 0.071 85 , KL4 70 60.0 0.049 50 KL7 90 88.0 0.068 80
-KL12 100 117.5* 0.077 100 '!
KL5 125 122.0* 0.085 100 KL9 125 120.0* 0.083 100 KL3 150 117.0* 0.084 100 i KLll 225 112.5* 0.076 100-KLIO 325 112.0 0.085 100 KL8 550 121.5 0.078 100
- Values used to determine upper-shelf energy value per ASTM E185.7 Table 3-3. Charpy impact Results for Irradiated Base Metal, Transverse (TL) Orientation. from Caosule U ,
Test Impact -Lateral Shear 4 Specimen Temperature Energy Expansion fracture ID F fi.lbi Inch % KTl -40 12.0 0.009 0 z KT5 -0 15.0 0.014 20- , KT9 2 0 -- 19.0 0.016 20 KT4 40 32.5 0.028 30
-KT7 '70 45.0 0.036 45 KT11 100 59.5 0.045 65 ;
-KT14 120 58.0 0.051 60 KT8- -125 71.5* 0.059 -100 .
KT6 150 71.0* 0.059 100-KT2 175' 73.5* 0.062 100 KT12 200 69.5* 0.061 100 KT15 200 67.0* 0.062 100 KT13 225 81.5* 0.065 100 ,
'KT3 -325 79.5 0.071 100 KT10 550 70,0 0.062 100
- Values used to determine upper-shelf energy value per ASTM ElfA.-
3-4 SWfteWihr , Il
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ = _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - . __ _- _ Table 3 4 Charpy impact Results for Irradiated Heat. Affected 7one Matal. from Caosule U Test Impact lateral Shear Specimen Temperature Energy Expansion Fracture 10 F ft lbs Inch %
-KH1 -80 32.0 0.021 45 KH4 -40 89.0 0.056 60 KHIS -40 79.0 0.052 60 KH7 -20 62.5 0.042 70 KH8 0 52.0 0.040 50 KH6 20 71.0 0.051 70 '
KH14 40 121.5* 0.074 100 KH12 70 130.0* 0.082 100 KH3 100 147.5* 0.085 100 KH9 150 142.5* 0.085 100 KHS 175- 111.0* 0.080 100 KH10 200 138.0 0.079 100 KH13 225 118.5 0.078 100 KHll- 325 161.5 0.080 100 KH2 550 132.0 0.082 100
- Values used to determine upper-shelf energy value per ASTM E185.7 Table 3-5. - Charpy impact Test Results for Irradiated Weld Metal,
-Transverse (TL) Orientation. from Caosule U Test Impact lateral Shear Specimen Temperature. . Energy Expansion Fracture ID F Lt-lbi Inch %
KW10 -80 7.0 0.004 10 KWll -50 35.0 0.026- 30 KW2 48.0 0.036- 40 KW5- -20 30.5 0.025 30 KW9- -20 69.0 0.051 55 KW1 0 101.5* 0.073 100 KW6 0 93.5 0.065 80 KW7- -40 123.0 0.083 90 KW13 70 --- 0.092 100 KW15 ~70- 133.0* 0.090 100 KW12 100 138.0* 0.091 100 KW8 150- 143.0* 0.089 100 KW14 225 142.5 0.087 100 KW3 325 157.0 0.088 100 KW4 550- 153.0 .0.085 100-
- Values used to determine upper-shelf energy value per ASTM E185.'
3-5 SWftamhv l __ -_=_-___ _: __
i ! Figure 3 1. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal. Lonaitudinal Orientation n
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i Figure 3 2. Photographs of Tested Tension Test Specimens and Corresponding Fractured Surfaces - Base Metal. Transverse Orientation
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Figure 3-3. Photographs of Tested Tension Test Specimens and Correspnnding Fractured Surfaces - Weld Metal. Transverse Orientation
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Figure 3-4 Tension Test Stress-Strain Curve for Base Metal Plate
.. R1808-3. Sgt.cj_ men No. KL2 Tested at 70f Specimen KL2 Test Temp.: 70 F( 21 C)
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Figure 3-6. Tension Test Stress-Strain Curve for Base Metal Plate R1808 3. Specimen No. KLI. Tested at 550f Specimen: KL1 Test Temp.: 550 F( 287 C) 3 3 ,, Strength , Y1 eld: 67276. . 7es.
~ UTS: 91064.
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- a. ma , se .me .aa .se .am Engine ering Strain Figure 3-7. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3. Specimen No. KT2. Tested at 70F
, 3 ,,
Speelment KT2 Test Temp.: 70 F( 21 C) Strength Yleid: 73204. . ves. UTS: 94094. ee. , ,,, ,,,, h E g . See. , es. . I b E w _ 4. , O
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. 1 es.
- p. i i a i i i i i . g,
- s. se .mo .32 .se .a4 .as Engineering Stroin 3-10 sW#fJi'ahr
Figure 3-8. Tension Test Stress Strain Curve for Base Metal Plate R1808-3. Soecimen No. KT1. Tested at 300F Specimen: KY1 Test Temp.: 300 F( 145 C) Strength Yleid: 67210.
- UTS: 65657. _ oaa.
on. . 05 . sos. E
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- a. se .e4 .se .12 .se .2a Engineerinq Stroin figure 3-9. Tension Test Stress-Strain Curve for Base Metal Plate R1808-3. SpecimRD No. KT3. Tested at 550F Specimon: KT3 Test Temp.: 550 F( 287 C)
Strength m 'd: 66442. . 7am. LsTS: 91294. se. - _ oaa. M 3 g . sea.
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Figure 3-10. Tension Test Stress Strain Curve for Weld Metal, Specimen No. KW3. Tested at 70F Specimen: KW3 Test Temp.: 70 F( 21 C) Strength Yleid: 76034. . 7am.
- tTTS: 90024.
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$ se. ,.
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. 3 mm.
- s. - I - f a ' - I -
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- a. ma . se .32 .se .24 .ma Engineering Struin Figure 3-11. Tension Test Stress-Strain Curve for Weld Metal, Soecimen No. KW1. Tested at 300F o Specimen: KW1 Test Temp.: 300 F( 148 C)
Strength Yleid: 70525.
- UTS: 63083. _ sea.
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! Figure 3 12. Tension Test Stress-Strain Curve for Weld Metal, Specimen No. Kw2. Tested at 550F Specimon: KW2 Test Temp.: 550 F( 287 C) Strength Yield: 60306. UTS: 87673- eaa. sa. . .
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=
Figure 3-13. Charpy Impact Data for Irradiated Plate Material, _ R1808-3. Lonaitudinal Orientation 100 - : ; ; : i ,
, i i 2
". n -
e* - 5 s so e e - g,, - e e - 0.10 3 i i i i i 5 . e e g0.08 , e
~
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_[0.04 E s 0.02 - 0 220 . . i i i i
- DATA SumARY -
g .T Not Determined - m Tcy (35 not) +2GF 1 80 - eT , (50 et La) +40F Tgy(M F1-u) +3F ~
, 160 c .USE (us)ll8 U ~ D
~ v RT Not Detortuned g ,
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, g100 E e 3 to -
~
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~
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;g , pg.g,,c g To be determined .
HtAT No. R1808-3 ' i i e i e i 0 100 0 100 200 3x 400 500 00 Test Temperoture, F 3-14 13W##ME'ohr bum.
Figure 3-14. Charpy impact Data for Irradiated Plate Material, R1808-3. Transverse Orientation _ 100 . , : ' ; 7 = ' ' ' 2
}75 - -
! 50 f25 -
0 . O.10 i i i i i i 0.06 -
-g0.Os - . . -
{0.04 3 5 0,02 - E . 0 I
- DATA SJTARY -
;oo -T Not Determined -
Tgy (35 mLt) +68F
*Tgy (50 FT-LB) +78F 1 80 TgyW at-ts) +3% -
g 160 Cy USE (ava) 72 It-lbs S RT,,, Not Determined
, 140 - -
?
Em - - 3 g 100 - - E [ g - * - - g . , g . 40 - Mtu pA SA333 GrB1(T) 20 - FLutact [1b dig termined - HtAt ho, _-R 1808-3 0 100 6 0 100 1
, 200 1
m
~~ ~
e6E b too Test-Tencere:ure, F 3-15 BW!!nn%%r
Figure 3 15. Charpy impact Data for Irradiated Plate Material, R1808-3. Heat Affected Zone m , :: ; : : ; e i : , i :
' " , 75 -
i
! 50 m
f 25 I t f f f I ,. 0.10 i e i i i i i Q,Qg - C , O ' kD.06 - a e e f0.04 - ** - 5 s 0.02 - 0 220 - i i i i i i
- DATA $ WARY -
200 -T Not Determined - m Tey (35 not) -47F IM -Tey (50 st-u) -41F Tey (30 et La) ~ 85F , - g IM C, ust (Ava>131 f t-Ibs y Not ,e ,
, 140 - RT,c, Determined e
& e sm S
- a e -
e Q g 100 - E o UM - e
~
1 g - e 90 y,73,g SA533 GrB1(HAZ) 20 - Fwance To be determined - put %, R1808-3 1 1 1 I 8 ' 0 SM 6M
-100 0 100 2Jo 3M 400 Test- Tercercture, F 5-16 N @ ((n#$EcN v SANY
Fiaure 3 16. Charov impact._ Data for 1rradiated Weld Metal ico :, e 2; ; ,
", 1s . .
v 3 p 50 - - be g 2, - - g i i e i i i 0.10 , , , , , i i - - 0.08 -
-E 1 0.06 a
[0.04 - - E e f_, 0.02 2 0 220 i i , i i i
- DATA SUTARY -
2g0 -T , Not Determined - icy (35 met) -40F 180
-T3 (50 st a ) -38F Tgy (30 rt u) -56F g 150 q.USE (avr,)129 f t-lbs , ,
y RT,,, Not Determined ,
. 140 - -
5
$120 g 100 -
E U 80 - - l
- 60 - -
no ,.
- PAttatat Weld Metal 20 -
FLutnet To be determined -
~~
HtAt No. t i i l i e h-100 0 100 , 2 00 300 400 500 60 Test '.empercture, F 3-17 n eswnucinn la WSERVICE COMPANY
Figure 3-17. Photographs of Charpy Impact Specimen Fracture Surfaces - Plate Material Lonaitudinal Orientation o.o ,o neo om >,..e ,am) a,c ,,,, ,u o u i as... ,a non s,u. ni otso
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ou ,..e a. utso om,sazom, ou ,.< e <o , i . 3, , o.m... . o. o m , ,,si <as uou 3-18 13W!!#H%h
Figure 3-18. Photographs of Charpy impact Specimen fracture Surfaces - Plate Material Transverse Orientation
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Figure 3-19. Photographs of Charpy impact Specimen Fracture Surfaces - Plate Material. Heat-Affected Zone
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e Figure 3-20. Photographs of Charpy Impact Specimen Fracture Surfaces - Weld Metal
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3-21 19111B&W NUCLFJ\Q 83 WSERVICE COMPANY
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- 4. DOSIMETER MEASUREMENTS 4.1. Introducti2D One set of dosimetry from the reactor vessel surveillance capsule (RVSP) denoted by the letter "U" was delivered to the Nucle . Environmental Laboratories by the Failure Analysis and Facility Operations Group of same affiliation (NES).
The set consisted of seventeen dosimeters made up of shielded and unshielded Co/Al wires, unshielded Cu, Ni, and Fe wires, and unshielded 2380 and as7Np fission powders. Cadmium sleeves were used to shield the Co/Al wires. Each dosimeter was contained in one of three stainless steel holder blocks that were l installed in various positions in the assembly. The dosimeters were-delivered in vials identified by labels consisting of the position of the holder block in the assembly, and the position of the dosimeter item in the holder (see part 4.2 for explanation). 4.2. Dosimeter Preparation Vials were prepared for the dosimeters by labeling them with identifications that indicated their positions in the holder blocks. For example, the first wire in the top block was labeled Sbl, 01-U TOP 1. When the nuclides to be analyzed were determined by gamma scanning, the identifications were appended accordingly. For example, Sbl, 01-U TOP 1 Cu. This identification code stands for Seabrook Unit 1, Cycle 1 Capsule U, Top holder block, first wire, Copper. The stainless steel fission powder capsules were clamped in a metal-working vise which was mounted by a suction cup in a hood. A flat mill-bastard file was used to file the capsules open. The cadmium-covered wires had been crimped at the ends so that the wires had to be removed by " nibbling" through the shield with diagonal cutters and removing the wires. 4-1 B WilEEVeh % v
l The dosimeter wires were cleaned by washing in reagent grade acetone, blotting dry with a laboratory towel. Each dosimeter wire diameter was measured with a certified 'nicrometer caliper, and weighed on a certified analytical balance. Each was then mounted in the center of a PetriSlide* with double-sided tape. Wires over 1/2" in length were bent in a "n" before mounting. The exact oxide compositions of the uranium and neptunium dosimeters were uncertain. It was not possible to correct for self-absorption of the powders, so it was necessary to dissolve them and put them into geometries for which our gamma spectrometer was calibrated. This was the 20 cc liquid scintillation vial geometry. The uran.am dosimeters were dissolved in 88 HNOa and diluted to ca. 20 mL with the same acid- in a pre-weighed 20 cc scintillation vial. The neptunium dosimeters were digested in 68 HC1/16H HF with addition of H,0, in increments until dissolved. These were also diluted up to ca. 20 mL in a pre-weighed 20 cc scintillation vials. 4.3. Quantitative Gamma Soectrometry Each of the dosimeters, in the PetriSlide* (point source) or 20 cc vial geometry, was given a 300 second preliminary count on the 31% PGT gamma spectrometer. This provided information with which to judge the best distance at which to count the dosimeter to get a minimum of 10,000 counts in the photopeak of interest while keeping the counter dead time below 15%. It also provided qualitative identification of the dosimeters. This identification was made from the presence of the gamma rays in the table below in the spectra, Dosimeter Analyte Cobalt "Co 01332 kev from Co, very high activity Iron "Mn 0 834 kev from Fe Nickel seCo 0 811 kev from Ni Copper "Co @ 1332 kev from Co, very low activity compared to Co wires, wire has coppery color Titanium Sc 0 889 kev from Ti 238g is7 Cs 0 662 kev, 2 'Pa 01001 kev
*7 Np
'87 Cs 0 662 kev, '"Pa 0 312 kev i
l l ! 4-2 BWunaniPuBm
c,..-; (; LThe:; spectra ~ confirmed the identifies of the dosimeters. The: spectra: were then: measured. quantitatively 'at the appropriate counting positions. and for. the appropriate count times determined from the preliminary counts.- u p. L l 4 4 i y 4 3= n S W # aff*c h v
' Table 4-1. Copper Desimetry Measurements from Capsule U Seabrook Station Unit No. I Post-Irrad.
Target Analyte Target Shielded Weight Item Dosimeter No./ Abundance (Yes/No) (Grams) No. Location Material TYDe Nuclide Nuclide Wire Cu-63 Co-60 0.692 No 0.08022 1 Sbl, 01-U Top 1 Copper Wire Cu-63 00-60 0.692 No 0.07936 2 Sbl, 01-0 Mid 1 Copper Wire Cu-63 Co-60 0.692 No 0.07954 3 Sbl, 01-0 Bot 1 Copper , l Attenuation Geometry Corrected Co-60 % Error Of fset Activity pCi/ Gram Item Coefficient Distance l fuC1) Factor _ fuci/ Gram _1 Tarcet No. u fcm.) _Co-60 { 1.01 0.9959 1.393E+00 2.012E+00 1 4.544E-01 17.536 1.097E-01 9.841E-02 0.97 0.9959 1.263E+00 1.825E+00 2 4.544E-01 17.536 9.706E-02 0.99 0.9959 1.242E+00 1.795E+00 3 4.544E-01 17.536 D EE - l @a NE 83 d a
4 _
. d:
Table 4-2. . Iron Dosimetry Measurements from Capsulo U Seabrook Station Unit No. 1 , Post-Irrad. Item ~ Dosimeter No./ ' . . Targetc. Analyte ;- Target- Shielded . Weight No. Location Material Jype Nuc)ide Nuclide . Abundance fYes/No) fGrams1 1 'Sbl,.01-U Top 2 Iron l Wire Fe-54 Mn-54' 0.058 No 'O.08056-- 2' 'Sbl,~;01-U Mid 2 Iron ~ ' Wire Fe-54 Mn-54-' 'O.058 .No 0.08135.
- 3 :.Sbl,l01-U Bot 2 . I ran . Wire Fe-54 Mn-54 0.058- No 0.07965
' Attenuation Geometry Corrected Item Coefficient Distance Co-60 .% Error' ' Offset Activity pCi/ Gram
[ No. u fcm.)' fuCi) 00 Factor fuCi/ Gram) Target 1: 5;800E-02 ~17.536 4.613E+00 0.56 0.9957 5.762E+01 9.934E402
.2 5.800E-02 17.536 4.193E+00 0.59 0.9956 5.187E+01 8.943E+02 3~ 5.800E-02 17.536 '3.993E+00' O.59 0.9956 5.045E401 8.698E402 Q.
E m a g7 nE k a i i n - , - - - _ _ - . - . - . . - _ - . _ . . . . -
'J, Table 4-3i Cobalt Dosimetry Measurements.from Capsule U
~
Seabrook Station Unit'No. l'
. Post-Irrad.
Item' Dosimeter.No./- . Target.' Analyte- Target- Shielded- .Weight No. Location- :Haterial Iypg Nuclide Nuclide Abundance: (Yes/Nol (Grams)
-1 3Sbl, 01-U. Top 3' ' Cobalt Wire. Co-59 Co-60' 0.0015 No 0.01057-All oy --
2 .Sbl, 01-0 Mid 3. Cobalt Wire Co-59 Co 0.0015- No 0.01072
; Alloy.
3- Sbl, 01-U 80t.3'- Cobalt -Wire' C0-59 Co-60 0.0015 No 0.00998-Alloy Attenuation Geometry Corrected-Item Coefficient- Distance C0-60 % Error Offset Activity -pCi/ Gram No. u (cm.) futil Co-60 Factor- (uCi/ Gram) Tarcet 1 4.532E-01 L17.536 2.995E400' O.68 0.9973 2.850E+02 'I.900E+05 2 4.532E-01 17.536~ 3 172E+00 0.68 0.9971 2.977E+02 1.984E+05 3 4.532E-01 17.536 3.173E+00 0.67 0.9971 3.198E+02 2.132E+05 D E , RR kE R2
A Table 4-4. Cadmium Shielded.Ccbalt Dosimetry Measurements from Capsule U Seabrook Station Unit No. 1
' Post-Irrad.
Item Dosimeter No./' Target Analyte Target Shielded Weight No. Location Material lyng Nuclide Nuclide Abundance (Yes/No) (Grams) 1 Sbl, 01-U Top 4 Cobalt Wire Co-59 Co-60 0.0015 Yes 0.00862 Alloy 2 Sbl 01-U Mid 5 Cobalt Wire Co-59 Co-60 0.0015 Yes 0.01110 Alloy 3 Sbl, 01-U Bot 4 Cobalt Wire Co-59 Co-60 0.0015 Yes 0.00964 Alloy P Attenuation Geometry Corrected Item Coefficient Distance Co-60 % Error Offset Activity pCi/ Gram No. u (cm.) fucil Co-60 Factor fuCi/ Gram) Taraet 1 4.532E-01 17.536 1.309E+00 0.91 0.9972 1.528E402 1.018E+05 2 4.532E-01 17.536 1.630E+00 0.61 0.9971 1.477E+02 9.849E+04 3 4.532E-01 17.536 1.537E+00 0.96 0.9971 1.604E+02 1.069E+05 C3 EE ER nn N2 8" mE
,s'
- 1. _ . _ .
Table 4-5. - Nickel Dosimetry Measurements from Capsule U Seabrook Station Unit No.'l Post-Irrad.
~ Item Dosimeter No./- .
Target Analyte- _ Target Shielded- Weight No. -Location Material- Jype Nuclide Nuclide : Abundance fYes/No) iGrams) 1- :Sbl,'01-U Top 5' Nickel. Wire Ni-58 C0-58 0.683 No 0.07773
'2 Sbl,.01-U.Mid 4 : Nickel Wire Ni-58 'Co-58 0.683 No 0.08024 3 Sbl, 01-U Bot'5- Nickel Wire' Ni-58 Co-58 0.683 No 0.08105 Attenuation- . .
Geometry Corrected-Item ' Coefficient- Distance. Co-60 % Error Offset Activity 'pCi/ Gram-No. u' fcm.) fuci) C0-60 Factor fuci/ Gram) Taraet 7 1 0.6092 28.943 1.107E+02 .0.33 0.9974 1.456E+03 2.132E+03 2 0.6092- 28.943 1.054E+02 0.34 0.9974 1.343E403 1.966E+03 3 0.6092 28.943 1.045E+02 0.33 0.9974 1.318E+03 1.930E+03 D GE MR Ei ne rd a x e' 4 _. _ . . . _ ._ _ . _ . . . s . . _ _ . - d
Table 4-6. Uranium-238 Dosimetry Measurements from Capsule U Scabrook Station Unit No. 1 Post-Irrad. Target Analyte Shielded Weight Item Dosimeter No./ Material lyRg Nuclide Euclide (Yes/No) (Grams) HL._ Location l Sbl, 01-U FIS 1 U-238 Pwder U-238 Cs-137 No 0.0105 1 l 1 Item Distance cs-137 % Error pC1/ Gram Eh_ (cm.) fact) cn-137 Taraet I 1 0.650 3.560E-02 1.10 3.387E+00 Table 4-7. Neptunium-237 Dosimetry Measurements from y Caosule U Seabrook Station Unit No.1 Post-Irrad. Item Target Analyte Shielded Weight Dosimeter No./ Location Material Iy2g Nuclide Huclide (Yes/No) (C:ams)___ h_ 1 Sbl, 01-U FIS 2 Np-237 Powder Np-237 Cs-137 No 0.0022 Item Distance ce2137 % Error pC1/ Gram fcm.) fuC1) cs-137 __Tarcet _ L 1 l Q 7.387 6.010E-02 1.30 2.689E+01 g 1 b sE FE 8 b ' a
.J
_ _ _ _ _ ~i I
v v.:
- 5. REFERENCES
= 1. ASTM- Designation- E185-79, " Standard Practice for Conducting Surveillance Tests - for- Light-Water Cooled Nuclear Power Reactor Vessels," in ' ASTM l
- Standards, American Society for Testing and Materials, Philadelphia, PA.
J2. Code: of . Federal Regulation, Title 10, Part 50, Domestic Licensing of
- Production '.and Utilization Facilittes, Appen(1x G, Fracture Toughness
; -Requirements.
I '3; Code -of Federal . Regulation, _ Title :10, Part 50, Domestic Licensing of y Production and Utilization Facilities, Appendix H, Reactor Vessel Material-
. Surveillance Program Requirements.
4.1 : American Society of _ Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section _ III,--Nuclear. Power Plant' Components, Appendix G, Protection
~Against-.Nonductile Failure (G-2000).
L, 5. ASTM Designation A370-7/, " Methods and Definitions for Mechanical Testing af Stecl- Products," in ; ASTM Standards, American_ Society _ for Testing -and
' Material s~, Philadelphia,~PA.
l t 6. ASTM Designation E23-88, " Methods for Notched Bar Impact Testing of Metallic g : Materials," in ASTM Standards, American Society for Testing and Materials,
- Philadelphia, PA. -
r l:- - 7. ; ASTM' Designation E185-XX (to be released), Standard Practice for Conducting g Surveillance Tests for Light-Water Cooled. Nuclear Power Reactor Vessels, in
' ASTM _ Standards, American Society for Testing and Materials, -Philadelphia,
' PA . --
l r 5-1 S W M "cFa h r
-+
M ,
- 6. CERTIFICATION
.The specimens-were tested,;and the data obtained from Public Service Company of
.New Hampshire Seabrook Station Unit No.1, reactor vessel surveillance Capsule l U were evaluated using-accepted techniques and established standard methods and-procedures in accordance with--the requirements of 10CFR50,-Appendixes G and H.
h
- h. L'. Ldwe','d r . , P . E :
(')[ hffff2 '
'Date Project Technical Manager This ~ report has been_ reviewed for technical content and accuracy. ,
??O bd M. J.'Devan (Material Analysis)
Gbha Date M&SA Unit-Verification of independent review. j
.Y ff Wk K. E.. Moore; Map 6ger' 2k$
Date M&SA Unit-This report:is approved for. release. Y & bSL T. L. Baldwin Date Program Manager 1 SWsTsefE h r . . - _ _ .. _.}}