ML20100N115
| ML20100N115 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities, 05000000 |
| Issue date: | 03/31/1984 |
| From: | Lindholm U, Norris E SOUTHWEST RESEARCH INSTITUTE |
| To: | |
| Shared Package | |
| ML17195A753 | List: |
| References | |
| SWRI-7484-002-1, SWRI-7484-2-1, NUDOCS 8504180324 | |
| Download: ML20100N115 (60) | |
Text
__ _ __
l SwRI-7484-002/1 i
l QUAD CITIES NUCLEAR POWER STATION UNIT 2 REACTOR VESSEL IRRADIATION
! SURVEILLANCE PROGRAM ANALYSIS OF CAPSULE NO.18 by E. B. Norris FINAL REPORT '
SwRI Project No. 06-7484-002 for Commonwealth Edison Company Nuclear Stations Division P. O. Box 767, Room 1230 Edison Chicago, Illinois 60690
- l March 1984 i
FC SOUTHWEST RESE ARCH INSTITUTE I
l }
-}I
, SAN ANTONIO HOUSTON l$0"$00$k$!!o$$37 p PDR
SOUTHWEST RESEARCH INSTITUTE Post Office Drawer 28510, 6220 Culebra Road San Antonio, Texas 78284 0.UAD CITIES NUCLEAR POWER STATION UNIT 2 REACTOR VESSEL IRRADIATION SURVEILLANCE PROGRAM ANALYSIS OF CAPSULE NO.18 by ;
E. B. Norris FINAL REPORT SwRI Project No. 06-7484-002 for '
Commonwealth Edison Company Nuclear Stations Division P. O. Box 767, Room 1230 Edison Chicago, Illinois 60690 March 1984 Approved:
l U. S. Lindholm, Director
! Department of Materials Sciences
l
' TABLE OF CONTENTS Page
! LIST OF TABLES- iii LIST OF FIGURES .iv I.
SUMMARY
-0F RESULTS 1 II. BACKGROUND 2 III. DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM 4
- IV. : TESTING 0F IRRADIATED SPECIPENS 5 A. Shipment,- Opening, and Inspection of Capsule 5 B. Neutron Dosimetry 5 4 C. Mechanical Property Tests 15 D. Check Chemical Analyses 18 V. HEATUP AND C00LDOWN LIMIT CURVES FOR OPERATION OF
, QUAD CITIES UNIT-NO. 2 -27 VI. REFERENCES 29 APPENDIX A - SURVEILLANCE ~ SPECIMEN DRAWINGS AND CAPSULE -
PHOTOGRAPHS' 31 APPENDIX B - PHOTOGRAPHS OF. TESTED SPECIMENS AND TENSILE TEST RECORDS 34
4
?
. x" -
LIST OF TABLES
. Table Page I:
-I' Operations. Summary - Quad Cities 2 7 II Results of-Discrete. Ordinates Sn Transport JAnalysis,-Q0ad Cities 2 12 i -III Summary of Neutron Dosinietry Results, -Quad
. Cities: 2, . Vessel Wall Surveillance Basket
- No.18 (215 ) 14-L
- IV ' Hardness Properties = of Surveillance Materials, Quad Cities 2- 17
'; V- Charpy V-Notch Impact Data on Surveillance
- Specimens ~ Removed from Quad Cities Unit 2 20
~
VI: Effect of Irradiation on the Charpy V-Notch Upper Shelf Energies of the Quad Cities Unit 2
' ~
Vessel. Surveillance flaterials-Basket No.18 (215 ) 24 VII, Tensile Properties of Surveillance Materials Basket No.~ 18 (215 ), Quad Cities 2' 25
- VIII Check Chemical Analysis Results 26 I
,a e
.l t
~
~
, iii l u
~. . .. . - - . .. ._~ . - _ - - - -
+
a
. LIST OF FIGURES
- Figure Page 1: Discrete Ordin'ates Calculational Models, Quad Cities Unit 2 11
-2 Calculated Flast Flux:(E > 1 MeV) Distribution,
. Quad Cities. Unit 2 - 13 3 Vessel Wall- Fluence as a Function of Operation of Quad' Cities 2 16 4' Charpy V-Notch Impact Properties of Quad Cities Unit 2' Surveillance Base Plate 21
=53 - Charpy V-Notch -Impact Properties of Quad Cities Unit.2. Surveillance. Electroslag: Weld Metal': 22
-6 -Charpy V-Notch Impact Properties of Quad Cities..
--Surveillance Electroslag Weld Heat'-Affected Zone Material 23 7
'L-for Pressure-Temperature Up'.to-10 EFPY Limits 'for Levels A & _ B . Service . 28 i
m A
T
+
4 1 . jy .
4 h
5
.l 6
- ). ',
I.
SUMMARY
OF RESULTS
'The analysis of the vessel material surveillance capsule basket No.-18_ removed from the 215 position in the Guad Cities 2 pressure vessel during the 1981 refueling outage led to the following results.
(1) Based on a calculated neutron spectral distribution, the capsule received a fast fluence of 6.6 x 1016 n/cm2 , E '> 1 MeV, in 5.63 Effective Full Power Years
- (EFPY) of operation.
(2) The shifts in-RT NDT were small, from -2*F for the base plate to +43 F for the weld metal, as a result of the above exposure. These data are consistent with the level of the exposure and may reflect data scatter rather than-real changes in RTNDT*
'(3 ) The' Charpy upper shelf energy of the base plate and the electroslag heat-affected zone -(HAZ) surveillance materials increased, but that of the electroslag weld
-metal ~ decreased after exposure. This may also reflect data scatter..
(4) Based on a calculated neutron spectral and spatial dis-tribution, the maximum fluence rate (flux) incident on-the pressure vessel wall is 4.89 x 108 n/cm2 /sec, E > 1 MeV. Therefore, the projected maximum vessel wall. neutron fluence after 32 EFPY of operation is 4.9 x 1017,-E > 1 MeV.
i l
l I
2 II. BACKGROUND The allowable loadings on nuclear pressure vessels are determined i by-applying the rules in Appendix G, " Fracture Toughness Requirements,"
of 10CFR50 [1]. In the case of pressure-retaining components made of ferritic materials, the allowable loadings depend on the reference stress intensity factor (KIR) curve indexed to the reference nil duc-tility temperature (RTNDT) presented in Appendix G, " Protection Against i Non-ductile Failure," of Section III of the ASME Code [2]. Further, the materials in the beltline region of the reactor vessel must be moni-tored for radiation-induced changes in RTNDT per the requirements of Appendix H, " Reactor Vessel Material Surveillance Program Requirements,"
of 10CFR50.
The RTNDT is defined in Paragraph NB-2331 of Section III of the ASME Code as the highest of the following temperatures:
(1) Drop-weight nil ductility temperature (Dti-NDT) per ASTM E 208 [3]
(2) 60 deg F below the 50 ft-lb Charpy V-notch (Cy) temperature
.(3) 60 deg F below the 35 mil Cy temperature.
The RTNDT must be established for all materials, including weld metal and heat-affected zone '(HAZ) material as well as base plates and forgings, which comprise the reactor coolant pressure boundary.
It is well established that ferritic materials undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron fluences in excess of 1017 neutrons per cm2 (E > 1 MeV) [4]. Also, it has been established that tramp elements, particular-ly copper, and nickel can affect the radiation embrittlement response of l
.ferritic materials [5-7]. The relationship between increase in RTNDT and copper content is defined in Regulatory Guide 1.99. Although this docu-ment is being revised by. the NRC to reflect a'more recent evaluation of l neutron embrittlement data from reactor surveillance programs , estimates of shifts in RTNDT in this report are based on the current Revision 1 of Regulatory Guide 1.99 [8].
--g
1 3
In general, the only ferritic pressure boundary materials in a nu-clear plant which are expected to receive a fluence sufficient to affect RTilDT are those which are located in the core beltline region of the re-actor pressure vessel. Therefore, material surveillance programs include specimens machined from or representative of the plate or forging material and weldments which are located in the core beltline region of high neu-tron flux density. ASTM E 185 [9] describes the recommended practice for monitoring and evaluating the radiation-induced changes occurring in the mechanical properties of pressure vessel beltline material.
General Electric has provided such a surveillance program for the Quad Cities 2 Nuclear Power Station. Six baskets of test specimens were located near the I.D. surface of the pressure vessel where the fast neu-tron flux density is slightly higher than that at the adjacent vessel wall surface. However, because of azimuthal variations in neutron flux density, these vessel wall basket fluences may lead or lag the maximum vessel fluence in a corresponding exposure period. Three sets of accel- !
erated-exposure baskets were included to augment the surveillance pro-gram. These baskets, located near the core, would have much larger lead factors to provide early information on long-term behavior of the sur-veillance materials. The baskets also contain several dosimeter materials for.. experimentally determining the average neutron flux density at each capsule location during the exposure period.
The Quad Cities 2 mechanical property surveillance capsules also in-clude tensile specimens as recommended by ASTM E 185. At the present time, irradiated tensile properties are used to indicate that the materials tested continue to meet the requirements of the appropriate material speci-fication and to judge credibility of the surveillance capsule Charpy data.
This report describes the results obtained from testing the contents of Basket No.18 from the 215* position in Quad Cities Unit 2 (the second vessel wall surveillance capsule tested). The results were analyzed to de-termine the radiation-induced changes in the mechanical properties' of the surveillance materials at the time of the refueling outage and the fluence rate at the vessel wall.
i 4
l
\
l III. DESCRIPTION OF MATERIAL SURVEILLANCE PROGRAM The G.E. vessel material surveillance program is described in de-tail in NED0-10115 [10]. Six vessel wall baskets, each located opposite i
the. vertical center of the core and containing encapsulated Charpy V-notch
! and tensile: specimens, were placed in the Quad Cities 2 vessel at the pres-sure vessel wall prior to startup. In addition, three baskets were placed near.the core. One vessel wall basket (No.13) and two near-core baskets (Nos.12 and 14) had been previously removed, tested, and the results re-o ported [11,12].
The vessel wall Basket No.18, removed during the 1981 refueling 4 outage from the 215 position, contained two impact capsules (each hold-ing 12 Charpy V-notch specimens plus an iron, a nickel, and a copper flux wire) and three tensile capsules (each holding two tensile specimens).
The capsules did not contain thermal monitors. Drawings of the impact and tensile specimens, and photographs of the impact and tensile capsules and the surveillance basket are shown in Appendix A.
According to NEDO-10ll5 [10], the Quad Cities 2 base metal specimens were made from flat slabs cut parallel to and one-quarter plate thickness from both of the plate surfaces, and were machined with their longitudinal axes parallel to the plate rolling direction. The electroslag weld (ESW) metal and ESW heat-affected zone (HAZ) specimens were cut from a test weld representing a vessel welded joint which had been fabricated from vessel base material, the weld and HAZ Charpy V-notch specimens being oriented with the long axis transverse to the weld direction and the weld HAZ tensile specimens with the long axis taken parallel to the weld direction.
The notches of all Charpy V-notch specimens were perpendicular to the orig-inal plate surfaces.
The mechanical properties of unirradiated (baseline) surveillance .
specimens for the Quad Cities 2 vessel were determined and reported at the time of the testing and analysis of the specimens from the first baskets [13]. These data were used to determine the shifts in RTHDT experienced by the materials in the 215 basket No.18.
l
? j 5
l
'IV. TESTING OF IRRADIATED SPECIMENS The capsule shipment, capsule opening, specimen testing, and report-ing of results were carried out in accordance with the Project Plan for the Quad Cities 2 Nuclear Power Plant Reactor Vessel Irradiation Surveil-lance Program. Each of these activities is discussed below.
A. Shipment, Opening, and Inspection of Capsule l After visually inspecting and photographing the surveillance cap-
! sule basket, SwRI personnel ~ opened the basket and photographed the con-tents. There was no evidence of damage to the capsules, see Appendix A.
SwRI loaded the impact and tensile capsules into a radioactive material shipping cask and' transported them to San Antonio, Texas.
The capsules were opened and the contents identified and stored in accordance with SwRI Procedure XIII-MS-103-1. The end plugs were cut from each capsule with'a band saw set up in the hot cell, then the test specimens and dosimeter wires were removed from the shell and placed in indexed receptacles.
Each inechanical test specimen was inspected for identification num-ber, which was checked against the master list . NED0-10ll5 [10], and no discrepancies were.found. The neutron dosimeter wires were identified and placed in tagged containers.
.B. Neutron Dosimetry The gamma activities of the dosimeters were determined in accordance with SwRI Procedure XI-MS-101-1 using an IT-5400 multichannel analyzer and a Ge(Li)_ coaxial detector _ system. The calibration of the equipment was ac-complished with 54M n,.60Co, and 137C s radioactivity standards obtained from the U.S. Department of Commerce National Bureau of Standards. All activ-
- ities were corrected to the time-of-removal (TOR) at reactor shutdown.
The dosimeter wires were weighed on a balance having a sensitivity l of 1 microgram. Infinitely dilute saturated activities (ASAT)were I
p_. rc 6
. calculated for each of the ' dosimeters- because ASAT is directly related to Lthe product of the energy-dependent' microscopic activation cross section L and the neutron flux density. ' The relationship between ATOR and ASAT is Tgiven by:
A "
TOR ; **7 p[j_g-ATm e -Atm\
A V' SAT m=1'
- V / \
where: 'Pm ':= fraction of-full power (2511 nth) for the
. operating period m; A, /= decay' constant for the activation product, day-l;.
.Tm = ' number of operating days for the period m; tm . = ; decay: time afterJt he operating period m,
' days.
Tbv Quad Cities 2 operating h'istory up to the 1981 refueling outage, which was used in the calculation of ATOR, is presented in Table I. -
.The primary result desired from the dosimeter analysis 'is the ' total fast neutron. fluence (> 1 MeV) which the surveillance specimens received.
The average flux density at full power.is given by:
^ SAT 4=Na n where: .4 -= energy-dependent neutron flux density, n/cm4/sec;
-ASAT ' = saturated activity, dps/mg target element; 5 = spectrum-averaged-activation cross section,
'cm2; and N.O
= number of target atoms per. mg.
- The total neutron fluence .is then equal to the product' of the average-neu-tron fluxfdensity and the equivalent reactor-operating time at full power.
. A discreteLordinates Sn transport analysis.for the Quad Cities 2-reactor vessel was performed using the DOT-IV code' to determine the radial,.
-si- i.
3
7
. 7 TABLE I OPERATIONS
SUMMARY
- QUAD CITIES 2 i
No. of Effective Fraction Decay '
-Operating Days - Full Power Full Power Days Period (Tm) Days (Pm) (tm) 6/72- 30 4.08. .136 3355
- 7/72 31 9.76 .315 3324 8/72 31 12.34 .398 3293 9/72 30 -
.000 3263 10/7.2 31 9.89 .319 3232
- 11/72 30 22.23 .741 3202 12/72 31 21.98 .709 3171 1/73 31 21.61 .657 3140 l 2/73 28- 22.43 .801 3112 3/73 - 31 29.36 .947 3081 4/73 30 12.57 .419 3051 l 5/73 ' 31 20.12 .649 3020 l 6/73 '30 24.39 .813 2990 l 7/73 31 24.64 .795 2959 8/73- 31 23.84 .769 2928 .
9/73 30 21.30 .710 2898 10/73 -31 21.76 .702 2867 11/73 30 24.84 .828 2837 12/73 31 27.40 .884 2806 1/74 31 23.59 .761 2775
'2/74 28 24.86- .888 2747 3/74 31 20.27 .654 2716-4/74 30 22.86 .762 2686 e 5/74 31 25.14 .811 2655 6/74 30 11.13 .371 2625 7/74 31 22.66 .731 2594 L
8/74 31 24.27 .783 2563 9/74 30 5.31 .177 2533 10/74- 31 21.76 .702 2502 11/74 30 20.58 .686 2472-
.12/74 31- 13.55 .437 2441 1/75- 31 -
.000 2410 2/75 -28 -
.000 2382 3/75 31 -
.000 2351 4/75 30 0.03 .010 2321
.5/75 31 20.96 .676 2290 6/75 30- 20.34 .678 2260 7/75 31 :
18.66 .602 2229
?;. >
- ~
8 TABLE.I CONT.-
OPERATIONS
SUMMARY
- QUAD CITIES 2
'No. of- Effective Fraction Decay
- . Operating Days Full Power Full Power Days i Period '(Tm) Days (Pm) _{.tgl !
8/75' 31 .10.70' .345 2198 l 9/75 50- 17.85 .595 2168 1 10/75' :31 2.17 .070- 2137 111/75 30 17.55- .585 2107 ,
-12/75- 31- 27.68 .893 2076
~1/76 31 29.70 .958 2045-2/76 :29 19.37 .668 2016 3/76 31 23.31 .752 l1985 4/76 '30 26.76- .892 1955 5/76 31 23.56 .760 - 1924 6/76 .30 20.25 .675 - 1894
.7/76 31 21.39 .690 1863 8/76 31 20.37 .657 1832 9/76 30- 6.42 . 21 4 1802 10/76- -
.000 1771 11/76 30' 20.13 .671 1741 12/76 31- 22.66 .731 . 1710 1/77 31 29.02 .936 1679 2/77 28 17.81 .636 1651
!3/77 31 8.52 .275 1620 4/77 30 '27.78 .926- 1590 5/77 31- 23.56 .760 1559 6/77 30 22.02 .734 1529 7/77 31' 18.44 .595 .1498 8/77 31 12.74 411 1467 9/77f 30 13.65 .455- 1437
- 10/77 31 20. 58 .664 1406 11/77 30 17.85 .595 - 1376 12/77 -31 2:2 48 .725 1345 1/78- 31 8.99 .290, 1314 2/78- 28- -
.000. 1286 3/78 31 14.10 . 455 1255.
4/78- 30. 125.95 .865 1225 5/78 31 17.08 .551 1194 6/78: 30 15.15 .505~ 1164 p' 7/78 -31 24.92- .804 :1133 8/78 :. 31 27.87.- .899 1102 c9/78 '30 25.17 ~ .839 1072 O
m
9 TABLE I CONT.
' OPERATIONS
SUMMARY
- QUAD CITIES 2
. No.1 o f - Effective Fraction Decay 10perating Days Full Power Full Power Days i Period (Tm) Days (Pm) (tm)
)
- 10/78 31 23.84 .769 1041 11/78; 30 23.40 .780 1011
- l. '12/78 31 26.72 .862 980 l 1/79 31 26.69 .861 949 l 2/79 28 21.39 .764 921 3/79- 31 23.44 .756 890 4/79' 30 20.49 .683 860 5/79 31 23.47 .757 829 6/79 30 21.93 .731 799 7/79 31 21.24 .685 768-
.8/79 31 19.13 .617 737 9/79 30 15.24 .508 707 ,
10/79 31 1 5.31 .494 676 I 11/79: 30 10.71 .357
.000 646 )
12/79 31 -
61 5 1/80 31- -
.000 584 2/80- 29 -
.000 555 3/80- 31 , -
.000 524 4/80 30 2.79 .093 494 5/80- 31 24.02 .775 463 6/80 30 27.42 .914 433 7/80 31' 25.36 .818 402 8/80 31 27.56- .889 371 9/80 30 28.62- .954 341 L 10/80 31 22.69 .732 310 l
11/80- 30 21.66 .722 280 12/80 31 22.10 .713 249 .
1/81 31 27.53 .888 218
, 2/81 28 24.25 .866 190 3/81 31- -29.05 .937 159 4/81 :30 25.98 .866- 129.
-5/81 31 -28.24 .911 '98 6/81 25.20 .840 68 7/81- 31 22.72 .733 37 8/81 31 21.48 .693 6-9/81 6 _3.92 .653 0 Total irradiation time = 5.63 EFPY.
m 10 vertical, and azimuthal dependence of the fast neutron (E > 1.0 MeV) flux density and energy spectrum incident on the reactor vessel and surveil-lance basket capsules. The R-0 and R-Z calculations were made using 15 of the 22 neutron group DLC-23/ CASK cross section library, a P1 expansion of the scattering matrix, and an S8 order of angular quadrature. The R-0 model utilized a one-eighth segment, and the R-Z model was represented as a one-half segment (see Figure 1) because of the symmetry involved. The various material zones (e.g., core, coolant, internals, nressure vessel, l
etc.) were described by homogenizing the major elements within each zone. {
1 I f The resulting spectra were used to calculate the spectrum-averaged j reaction cross sections for the threshold detectors and the lead factors neeied to relate the neutron exposure of the pressure vessel to that of the surveillance basket capsules. The pertinent factors obtained from this transport analysis are summarized in Table II. The capsule flux is less than the peak fast flux incident on the vessel I.D. because of azimuthal and vertical flux variations related to the core geometry and power distribution.. The calculated azimuthal and vertical flux maps are shown in Figure 2.
The calculated cross section for the 54 Fe(n,p)S4fin reaction is some-what less than the 236 millibarn (mb) value calculated for vessel wall Basket fio.13. [11]. It -is believed that the values given in Table II are reasonable because SwRI-computed reaction cross sections for Browns Ferry Unit 3, which used the same methodology, agree well with those measured by General Electric during the same core cycle [14].
The dosimetry results obtained with the calculated spectra are pre-sented in Table III. If a fission-spectrum energy distribution is assumed at the capsule location, the cross section for the 54Fe(n p)S4Mn reaction (E > 1.0 MeV) would be 98.26 mb [4], and the resulting value for fast flux at the capsule location would be 4.9 x 108 n/cm2/sec. This value is re-ported for reference only and has not been used in the analysis of results.
l The discrepancies in the peak vessel flux values determined from the
-several dosimeter materials are attributed primarily to the uncertainties
.in the calculated spectra, in the spatial distribution, and in the reaction
r
, 45 0 h
Shroud , Capsule
,A Jet ' /
ss um
. - - -- t - - . P.V.
l ; -
l
......_.._2.____.
l H0 2
- 0
.i . .
.s i
.. _ . ; . - _ . . ;_ _ . . 3- - _ . - ;
, Corej l l
.._y.... 3_____3-_.._3..___4...___
,R l l ; ; l
. . i , , i i . . . , ,
t 1 e
. . . . . ;i . . . . _ , [ _ , _ . ,, r -- -- -, i - - - -- i. - - -- -. -I -- -- J s
. . . . . . . J 00 (a) R-0 Model i
Shroud l [
9 i
N '
i i
, e I e
, g l .e i i - I i . i
, , e 6 f
, , i l
' i i ilet' P. V.
. l Core-r
! t .
71 - , ,
,, g l 8 ,
' . . t ,
i , ; ,
e . i e l -
' 1 '
i i i
i l
l l
, e i HO2 g
8 '
l e' ' . i l 4
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9 I
I l \ /
(b) R-Z Model FIGURE 1. DISCRETE ORDINATES CALCULATIONAL MODELS, QUAD CITIES UNIT'.2
I %
12 TABLE II RESULTS OF DISCRETE. 0RDINATES Sn TRANSPORT ANALYSIS QUAD CITIES 1'AND 2
~ A.- Calculated Reaction Cross Sections for Analysis of Fast Neutron Monitors (E > 1.0 fieV)
' Reaction '5 (barns) l 54Fe(n.p)S4ph 0.197
', 58Ni(n.p)58o' O.237 ;
C 63Cu(n,a)60Co 0.00330 i
! B.- Calculated Basket Lead Factors -(E > 1 MeV)
+
Location within
-Position (a) Vessel Wall- Lead Factor (b)
-215 I.D. Surface 0.755 215 1/4T 1.07.
215* 3/4T 4.75
-(a) Azimuthal position of surveillance Basket No.18.
Capsule neutron flux density, E > 1.0 MeV.
(b), Maximum neutron flux density at vessel I.D., E > 1.0 MeV i
L N
i
.t
, s ,, ,
+L' .,. - .-. ,c , , .- - , .
13 1C9 8 -
E
. . . - ..._-..=.=_,=w. = .=~-
~
6 - 22,: ~._1~T._._C'i~C,_E f5: ~~li .1_1__~. f_ _ _ . - _ :-_ .
._ ._ _. _ _. l,
- _ . ._ a _.__ -.. . - _. .---. . s. \
3
.o .. o o _c ,
w 4 - [_~j i _ . . . _ - - - ?. ^._. _ -..~- - . _ - -. . _- .l - _ c G oCo go l
.i _ o .o . O C'
- - L+) l = = = ~~~;= _- 3 5 g O L X=;Q;.C2+d== 2.F
- =~:,5=:: -E ~
.= 0= _ .
w . - - . _ _
-.--.__.7 --._;._-
.._.-._z__
.-_-~_._.7
- u. ._. -+_------_-
c ~
n + - + - - - -
o 2 n ,
u ; _ 1 ___
m <> 1 0 I 0-'~O_L. Lo 3 ___._+_ __- _
2:
-m. -.-__
t------ --
r t- r-, t r----,----
103 0 5 10 15 20 25 30 35 40 45 l
l t
Azimuthal Position, degrees (a) R-e Calculation 250 . ,
_. o ,
.I
. : j .
- [. .. _. _i. . .
m .. o. p m.
e 200 =' ._.,...
u > -> .
2 _ . .
a L_p,.. .-. ..y .. g}...
e
, _ p_ .. .
p-.n. ... .
..g.
9 .. .. o _.
e
_p'_L h.
_. 2 m ;
150 =.}_.._t _4 7 . _ _ _ ' _ .. o j 4p . _p. _.
.q. .
p ._..L j._ ___ .. ., ..
,, . ~ .._. .i..
c _ ~_. ,.. .
.o u_ ._ 2p.
- {
u 100 .. t__p p. . . _%_ . _ -
_ m.__ .
- ,_.__. ._.t..
i
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._ ._g_o.-,: ,., a__ _4. ,._L_
g .+._ . #__ . _ .. . ._ .__ ._ _
. m,- . . _ . . _ _. _
o . . L_. __L _- '
'u m _ _.o
", 50 = _ . ' . _ ._._ _ ..
o_. _ . .
, . . . , . . ...._ # L.,_ _._ . .
. . - ___ _ a .. . r-- ..
o
.q_ . . _._ *%_ _._ ._.
_o.
0 y - q_. ++ .
._. .;4=_t._
L.
._y
._. ._.y . .&
m .._ . n _ . _
o o
4 _. .4...
_ L_ c. J .. __._
w
-. . __ ..n.
_. _..p ,o
" -50 =, -- ;
o e
.a
. , . . ..m *-.L 4
-; _- { _._ - . + . :o u
. i. . .4..
. . - ..n _.t a_,
... .._ _.s
._a.. ._ o
- 4. a . '
.0-
-100 a
g . g .q' ' ..
. .% , _ .. ; . . g, .
. TJ 1 ~:3.1 - ~Tr [.O '
v -150 = . . _ :
i 1.La. .
a'_I.__. ._ ; ; op_a _ _ ,_
Lu . ._H._4. Lg. . - ._ -
__.q ._ .t 7 ._ +. g
. $ . } .L- - . . . - . - . .. ..
-200 =. L_
. 4 .
i-.
..q
.,_2
.L.
. .. .q. . .. . .a.y
-250 ' i '
i '^ N I 2x107 3 4 56 8 108 2 3 4 5x108 Neutron Flux, E > 1 MeV (b) R-Z Calculation FIGURE 2. CALCULATED FAST FLUX (E > 1 MeV) DISTRIBUTI0tl, QUAD CITIES UtlIT 2
TABLE III~
StM4ARY OF NEUTRON 00SIMETRY RESULTS
QUAD CITIES 2, VESSEL WALL SURVEILLANCE BASKET N0.18 (215 )
Dosimeter I.D. Activation Weight ATOR ASAT Flux, E >2 1 MeV(a)
Material Capsule Reaction (mg) (dps/mg}_ (dps/mg) (n/cm /sec).
Fe' G10 54Fe(n,p)S4Mn ~ 155.2 ' 2.093 x 101 2.994 x 10I 2.47'x 108 Gil .138.3 2.136 x 101 '3.057'x 101 2.52 x 108 Average = 2.49 x 108
'Cu . G10 63C u(n,a)60Co 360.2 5.015 x.100 1.144 x 101 5.29 x 108 Gil 331.8 4.258 x 100 9.717 x 100 4.49 x 108 Average = 4.89 x 108 1
Ni G10 58tli(n p)58Co 290.2 (b) - -
Gli 247.2 (b) - -
(a) Calculated flux values subject to a 16.5% uncertainty (10).
(b) 58 Co had decayed away'at time of measurement.
E
15 cross sections. Other neutronic factors contributing to the estimated 16.5% uncertainty (la) in the. calculated fluxes are the determination of disintegration rates and the calculation of reaction rates (ASAT/NO).
For example, the iron monitors effectively measured the capsule flux for the last three years of operation, and the copper monitors effectively measured the capsule flux for the entire exposure period.
Averaging the results obtained from the iron and copper neutron dosimeters (the Co-58 in the nickel dosimeters could not be detected be-cause of the long decay period), the fast neutron flux at the surveillance
~
basket location during full power operation would be 3.69 x 108 n/cm2 /sec.
E > 1 MeV, and the peak value incident on the pressure vessel I.D. would be 4.89 x 108 . Since Quad Cities 2 operated for 5.63 effective full power years (EFPY) up to the September 1981 refueling outage, the calculated basket and vessel fluences to that time are as follows:
- Surveillance Basket -
6.56 x 1016 n/cm2
- Pressure Vessel I.D. Surface -
8.68 x 1016 n/cm2
- Pressure Vessel 1/4T -
6.13 x 10 16 n/cm2 Pressure Vessel 3/4T -
1.38 x 1016 n/cm2 The vessel wall fluence as a function of plant operation is shown in Figure 3.
C. flechanical Property Tests Hardness tests were run in accordance with ASTri Method E 18 [15] on one Charpy V-notch specimen selected from each material group. The re-suits are given in Table IV.
The irradiated Charpy V-notch specimens were tested on a calibrated
- 240-ft-lb,16-ft/sec SATEC impact machine (Model SI-lK) in accordance with SwRI Procedure XI-MS-104-1. The test temperatures, selected to develop the ductile-brittle transition and upper shelf regions, were obtained using a liquid conditioning bath monitored with a Fluke Model 2168A
- Inspected and calibrated using specimens and procedures obtained from the Army Materials and Mechanics Research Center.
16 17 6 x 10
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2 x 10 1 2 4 6 10 20 40 Plant Operation, EFPY FIGURE 3. VESSEL WALL FLUEllCE AS A FUtlCTION OF OPERATION OF QUAD CITIES 2
'17 ;
1 TABLE IV HARDNESS PROPERTIES OF SURVEILLANCE MATERIALS l- QUAD CITIES 2
~
(Fluence = 6.6 x 1016. E > l'MeV)
Charpy ,
i
. Test 'Specjmpn '
'No.(a)
-Material Hardness (HRB)
Base T4D- .92.0 92.0
' 91.0 L
I Average = 97.7 ESW TBP 88.0 f
i
! 90.0 90.0 Average = 89.3 ESHAZ TML 90.0 -
t 91.0 91.0 l '
Average = 90.7.
l f.
t e
(a) Specimen' identification code given-in Reference 10.
I I
9
_-___ _ __ __ _ _-_ = _ _
I 18 i digital thermometer. The Charpy V-notch impact data obtained on the ir-radiated materials are presented in Table V, and photographs of the frac-ture faces are included in Appendix B. The Charpy V-notch transition curves for the irradiated plate material, weld metal, and HAZ material contained in the 215 capsules are compared to the unirradiated values in Figures 4 through 6. The shifts in the 50 ft-lb, 30 ft-lb, and 35-mil lateral expansion transition temperatures were small, see Table VI. The Charpy upper shelf energy behavior, also shown in Table VI, was not en-tirely consistent with the transition temperature changes. For example, an increase in the upper shelf energy of the base metal and the HAZ mat-erial may result from data scatter rather than a real change in the prop-erty of the material. However, the small changes which were measured are consistent with the level of fluence received by capsule basket No.18.
Tensile tests were carried out on the irradiated materials in ac-cordance with SwRI Procedure XI-MS-103-1 using a 22-kip capacity MTS Model 810 tester equipped with an Instron Catalog No. G-51-13A 2-in, strain gage extensometer and Hewlett Packard Model 7004B X-Y auto-graphic recording equipment. Tensile tests were run at room tempera-ture and 550 F. The results are presented in Table VII. Each tensile load-strain record and photographs of the tested specimens are included in Appendix B.
D. Check Chemical Analyses Check chemical analyses were run on samples cut from the fracture end of selected tested Charpy specimens. All of the weld metal speci-mens were tested for copper and nickel content at SwRI using an X-ray fluorescent technique. Two each base plate and weld specimens were then sent to Westinghouse Advanced Reactors Division Analytical Labo-ratory for complete analyses using gravimetric (Si), combustion (C and S), and ICP Plasma (remainder of elenents) methods of analyses. The re-sults are summarized in Table VIII. The differences in the copper and nickel results can be attributed to at least three factors:
19
-(l ^) :The ICP ' Plasma method measures the chemical content of the full volume of sample material, while the X-ray fluorescent method-looks only at the surface
- of the sample.
"(2)l .The 1 square centimeter cross sectional area avail-
[ _
- able from the Charpy samples is smaller than desired f for detecting small amounts of residuals'with an X-ray l fluorescent technique.
- (3) The gamma activity of the sample increases the diffi-culty in accurately measuring the fluorescent peak.
4 i
i
(
4-
g; .-_ - .- . _ - . . - _ - __-
20 TABLE V CHARPY V-NOTCH IMPACT DATA ON SURVEILLANCE SPECIMENS REMOVED FROM QUAD CITIES UNIT 2
.(Fluence = 6.6 x 1016, E > 1 MeV)
Test Impact Lateral i Specjmp . Temperature Energy Expansion Shear Material No.taf ( oF) (ft-lb) (mil) ( *.' )
Base T2D -50 9.0 9 2 T27 -25 22.5 20 5 T2E O 42.0 34 10 T3P- 40 74.0 59 20 T4D 75 117.5 79 75 T3M 120 139.0 94 100 T37 160 150.5 80 100 T2B 210 145.5 95 100 ESHAZ TLS -50 16.5 17 5 TLK -25 40.0 37 10 TJ2 0 72.5 57 15 TJU 40 40.0 38 15 TML- 75 176.5 66 45 TL7 120 145.5' 88 100 TM1 160 149.5 97 100 TMA 210 153.0 78 100 ESW TAE -25 10.5 11 2 TAT - 0 31.5 27 5
- TAP 20 33.0 29 10 T72 40 74.0 60 20 TBP 75 52.0 50 25 TBM 120 86.0 65 90 TB1 160 103.0 86 100 T6K 210 78.0 63- 100
-(a) Specimen identification code given in Reference 10.
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TUPE24 M itt4 FI FIGURE 4. CHARPY V-NOTCH IMPACT PROPERTIES OF QUAD CITIES UNIT 2 SURVEILLANCE PLATE o
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TUPERAfUE (N4 FI FIGURE 5. CHARPY V-NOTCH IMPACT PROPERTIES OF OUAD CITIES UNIT 2 SURVEILLANCE ELECTROSLAG WELD METAL
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TEMEMftfE (DES F1 i
FIGURE 6. CHARPY V-fl0TCH IMPACT PROPERTIES OF QUAD CITIES SURVEILLA!!CE ELECTROSLAG tlELD HEAT-AFFECTED Z0flE IMTERIAL
24 EFFECT 0F IRRADIATION ON THE CHARPY V-NOTCH PROPERTIES OF THE QUAD CITIES UNIT 2 VESSEL SURVEILLANCE MATERIALS BASKET NO. 18 (215*)
(Fluence. = .6.6 x 1016, E > 1 MeV)
Base ESW Ueld ESW HAZ Criterionb) Plate Metal Material Transition Temperature Shift (2) 0 50 ft-lb -6*F 20 F 27*F 0 30 ft-lb -2 F 43 F 9*F 0 35 mil nil -10 F 45*F ART NDT( ) (4) 43 F 9F l
Cy Upper Shelf Drop -9 ft-lb 34 ft-lb -29 ft-lb
(-7%) (27%) (-24%)
l l
(1) Refer to Figures 4-6 (2) Cyparameter.
. determined by hyperbolic tangent fit where:
Cy parameter.= A + B tanh((T - T1 )/T2 )
l (3) Transition temperature' shift 0 30 ft-lb[1]
(4) Apparent regative shift in RTNDT l
l l l
TABLE VII -
. TENSILE PROPERTIES.OF SURVEILLANCE MATERIALS-BASKET NO. 18 (215 ), QUAD CITIES 2 (Fluence .= 6.6 x' 1016. E > 1 MeV)
Fracture Fracture Uniform Total
- Test. . Temp 0.2%.YS UTS. Load ' Stress.
Elongation Elongation R.A.
Material No. Spe(c) a (ksi)- (ksi) (Ib) (ksi)
(*F) ~ (%) (%) -(%)
' Base UDA 73 71.0 92.3 2669 171 .17.5 .20.1 -68.2 00T 550 61.2- 87.3 2639 171 15.7 22.6 - 68.6 -
ESHAZ UL1 73- 63.9 84.4 2539 .185 16.1 21.6 72.1 ULK 550 56.6 78.3 L2547 1 61 13.0 19.7 67.7 ESW ~UJ1 73 '64.3 86.2 2663 .165 18.9 26.4 67.3 UJE 550 60.0 80.0 2639 143 15.0 20.5 62.5 (a) _ Specimen identification code given'in. Reference 10.
Di l
l,
~
TABLE VIII l CHECK CHEMICAL ANALYSIS RESULTS BASKET NO.18, QUAD CITIES 2 Specimen Weiaht Percent of Element l i
Identification (a) : Source (b) C S P Si Cu Ni Cr Mo- V
'T27 W .265 .019- .008 .199 .072. .494 .127 .469 .009 S .12 .60 T2D W .287 .019 .008 .233- .078 .525 .132 .488 .008 L-S .10 . .45-TAE W .209 .017 .008 .115 .129 .31 3 .079 .503 .008 S .14 .32
, TAT W .21 6 .018 .008 .126 .122 .359 .091 .522 .009 j .. S .16 .33 TBP S - - - -
.16 .34 - - --
, TB1 S - - - -
.17 .37 - - -
T6K S - - - -
.17 .39 - - -
TBM S - - - -
.17 .37 - - -
T72 S - - - -
.20 .41 - - -
TAP S - - - -
.12 .32 - - -
(a) Cut from tested Cy . specimen, see Table .V for material I.D.
(b) W = Westinghouse Atomic Power Division; S = Southwest Research Institute.
i-l L_ _
. , . =
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3 , .
2 1~
V. - HEATUP AND'C00LDOWN LIMIT CURVES FOR OPERATION OF QUAD CITIES UNIT NO. 2
. Unit No. 2 is one of four similar 251-in. I.D. boiling water reactors (Dresden Units 2 & 3 amd Quad Cities Units 1 & 2) operated i by Commonwealth Edison Company. 'Each of the simi16? units has-been f provided with a reactor -vessel material surveillance program as re-
-' quired by 10CFR50, Appendix H [1].
f .
Based on the capsule analyses, heatup and cooldown limit curves L
for LevelLA and B service and for up to 10 EFPY of operation have been. computed.s The limit curves _ developed are a generic set for all four vessels listed above. .The curves, given in Figure 7, are based on the: worst case for.the vessel fluence rate, copper / nickel content,_and initialiRT They were computed in accordance with-NDT.
. Regulato'ry Guide 1.99, Revision 1 [8] amd NUREG-75/087 [15] using the following bases:
1/4T,RTNDT = 60 F
.3/.4T RTNDT = 49 F Cooling / heating rate = 100 F/hr. .
Vessel inner radius ~= 125.5 in.
Vessel outer radius = 131.8 in.
0perating pressure. =-1030 psig Initial temperature _ = 60*F Final temperature = 550*F' Effective coolant flow rate = 1.08 x 108 lbm/hr.
Effective flow area ~ = 96.0 ft2 l Effective hydraulis
- , diameter = 24.0 in. 1 I
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'A I
29 V. REFERENCES
- 1. Title 10, Code of Federal Regulations, Part 50, " Licensing of Production and Utilization Facilities."
- 2. ASME Boiler and Pressure Vessel Code, Section III, " Nuclear p Power Plant Components," 1980 Edition.
- 3. ASTM E 208-69, " Standard Method for Conducting Drop-Neight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," 1975 Annual Book of ASTM Standards.
4 L. E. Steel and C. Z. Serpan, Jr. , " Analysis of Reactor Vessel Radiation Effects Surveillance Programs," ASTM STP 481, December 1970.
- 5. L. E. Steele, " Neutron Irradiation Embrittl ement of Reactor Pres-sure Vessel Steels," International Atomic Energy Agency, Techni-cal Reports Series No. 163, 1975.
- 6. ASME Boiler and Pressure Vessel Code, Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," 1980 Edition.
$ 7. J. F. Perrin, R. A. Uullaert, G. R. Odette, and P. M. Lom'brozo,
" Physically Based Re7ression Correlations of Embrittlement Data from Reactor Pressure Vessel Surveillance Procrams", EPRI MP-3319.
January 1984
- 8. Regulatory Guide 1.99, Revision 1, Office of Standards Develop-ment, U.S. Nuclea r Regulatory Commiss ion , April 1977.
- 9. ASTM L 185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," 1975 Annual Book of ASTM Standards.
- 10. J. P. Higgins and F. A. Brandt, " Mechanical Property Surveillance h
of General Electr ic BWR Vessels ," NED0 10115, July 1969.
- 11. J. S. Perrin et al, " Quad Cities Nuclear Plant Unit No. 2 Reactor x Pressure Vessel Surveillance Program: Capsule Basket No. 12 and 6 Capsule Basket No.13," Battelle Columbus Laboratories Report, Sept ember 19, 1975
; 12. T. R. Mager et al, " Analysis of the Thir d Capsule from the Common-t wealth Edison Company Ouad Cities Unit 2 Nuclear Plan! leactor Ves-i sel Radiation Surveillance Program," Westinghouse Electric Corpora-tion WCAP-10064, Electric Power Research Institute Research Project 9 1021-3 Topical Report, April 1982.
s
30
- 13. J. S. Perrin and L. M. Lowry, " Quad Cities Nuclear Plant Unit No. I and Unit No. 2 Reactor Pressure Vessel Surveillance Prograns: Unir-radiated Mechanical Properties," Battelle Columbus Laboratories Final p Report to Commonwealth Edison Comagny, February 15, 1975.
f 14. E. B. florris, " Spectral Analysis of a BWR Vessel," Proceedings of the Fourth ASTM-EURATOM Symposium on Reactor Dosimetry," NUREG/CP-0029, Vol . 2, pp.1043-1050.
- 15. US NRC Standard Review Plan, NUREG-75/087, Section 5.3.2, Pressure-Temperature Limits, November 24, 1975.
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. Offset had = f fgp/ Initial Area psi i lE M Upper Y.S. = Initial Area = b3 hfY psi - % Elongation = Final G. L. - Initial G. L. Initial G. L. * ,g/ 3 ' Initial Area - Final Area Initial Area 100 = 7 2. / - t i e / i m ^ ignature: // ' M ~~ l [ / " 4 M = h , + k N T 5 I * { g = y I 4 ,a = c w b --= , e m & 'r A k - %M
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- f Department of Materials Sciences
== TENSILE TEST DATA SHEET
- 'es t No. T- Y Est. U. T. S. psi Project No. d'8- J FM o #
. :pec. No. M.7 E Initial G. L. / O 2 _7 in. Machine No. // f/k ; ' 'emperature NO 'F Initial Dia. , J' I # in. Date / 6-- Z /- O
- train Rate Initial Thickness in. Initial Area M O NO 7M
~ Initial Width in. Top Temperature f60 'F Maxim 2m Load 3925 lb : .- 6Y7 i Bottom Temperature 'F 0. 2% Offset Load 29 WO lb f N' ?% 1 i Final Gage Length /.23 3 in. 0.02% offset Load 1706 lb O?- ; .y :. ! Final Diameter 9.II3M in. Upper Yield Point M lb k Final Area 6 O /8 Y in. 2 .a z.p s. / -
- 5. U. T. S. =
,1 = 7 9 F5' psi / 7. ?.f
- 0. 2% Y. S. =
Initial Area = h dO/ 2 psi : i -:s . . , . j#; i
- 0. 02% Y. S. -
Initial Area - 85/A3 psi f.i
- . 9.
Uccer Yield Point ', 't i Upper Y. S. = ~ = . . . . Initial Area psi ' e ~ l "?q.~A. . ' Final G. L. - Initial G. L. x 100 = 6" '.
- % Elongation .
= Initial G. L. M. h' '/% + .- [k y- .:o, yP- { Initial Area - Final Area 3 Initial Area x 100 = 07. % : ..E ;. p ? ) #:gl-A .,*' { e f ."~,. r .: h;ignature: [*
- w. , [U
.c. . , ' ' .;i; g b ;a fs P A O---$Al=4 97 hpon : 1~Si O - 8 3 .t i 3 s 1 ' ki 0 3 4 w? I g y ,. D )
- ,R , 2 to T 1
Q VI i D N R ) 1 30 5 1 7 9*
- 4) u O
- i. 3
'c k o , v w i h 4 . ( w i Q ,() 'h . 'l ~ . .] ,I I ~A i - A(' ~h, w . i- 7 . N h L g s I; $ C .y'v A ena atie.s a 9 i 4 x l 0 1 N uf ,'8- - t . \Q V _ , i = R o N n V, s ~ Y b i = ~ ; N Ii :. k ' { ,'o ~ .A g m 4 ld 'I h F h n - A J .e Q y . T f 9 ' J-d j 'N ! .? ;g i i d ,7 _. w u-t . r. M .: f s 1s O m 4 1 l. = -t l '% o ,~ . }lf f = d f9Mf f%~ r _ unumma ..}}