ML20141J347
| ML20141J347 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/17/1986 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20141J328 | List: |
| References | |
| NUDOCS 8604280105 | |
| Download: ML20141J347 (18) | |
Text
,
a 7
t ATTACHMENT A 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)
(a)
The curve in Figure 2-3 shall be used to predict the increase in transition temperature based on integrat-ed fast neutron flux.
If measurements on the irradi-ation specimens indicate a deviation from this curve, a new curve shall be constructed.
(b)
The limit line on the figures shall be updated for a new integrated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equiv-alent integrated fast neutron exposure (E > 1 MeV).
I For this plant, based upon surveillance materials tests, weld chemical composition data, and the re-duced vessel fluence rate provided by core load de-signs beginning with fuel Cycle 8, the predicted sur-face fluence at the reactor vessel belt-line weld material for 40 years at 1500 MWt and an 80% load 2
factor is 2.9x1019 n/cm. The predicted trans-ition temperature shift to the end of the new period shall then be obtained from Figure 2-3.
(c)
The limit lines in Figures 2-1A and 2-1B shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curver were last constructed.
The boltup temperature limit line shall remain at 82*F as it is set by the NDTT of the reactor vessel flange and not subject to fast neutron flux. The low-est service temperature shall remain at 162*F because components related to this temperature are also not subject to fast neutron flux.
(d)
The Technical Specification 2.3(3) thall be revised each time the curves of Figure 2-1A and 2-1B are re-vised.
8: sis All components in the reactor coolant system are designed to with-stand the effects of cyclic loads (dye to reactor coolant system temperature and pressure changes. 1 1 These cyclic loads are in-troduced by normal unit load transients, reactor trips and start-up and shutdown operation.
During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon a rate of 100*F in any one hour period and for cyclic operation.
8604280105 860425 PDR ADOCK 05000285 P
PDR 2-4 AmendmentNo.//,ff,pf,79,77
t 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)
The maximum allowable reactor coolant system pressure at any tempera-ture is based upon the stress limitations for brittle fracture consid-in-Section IIItc) limitations are derived by using the rules contained erations. Thesg of the ASME Code including Appendix G, Protection Against Nonductile Failure, and the rules contained in 10 CFR 50, Appendix G, Fracture Toughness Requirements. This ASME Code assumes that a crack 10-11/16 inches long and 1-25/32 inches deep. exists on the inner surface of the vessel. Furthermore, operating limits on pressure and temperature assure that the crack does not grow during heatups and cooldowns.
The reactor vessel belt-line material consists of six plates. The nilductility transition temperature (TNDT) of each plate was estab-lished by drop weight tests. Charpy tests were then performed to de-termine at what temperature the plates exhibited 50 ft-lbs. absorbed energy and 35 mils lateral expansion for the longitudinal direction.
NRC technical position MTEB 5-2 was used to establish a reference temperature for transverse direction (RTNDT) of -12*F.
The mean RTNDT value for the Fort Calhoun submerged arc vessel weld-ments was determined to be -56*F with a standard deviation of 17*F.
In accordance with the methods identified in "NRC Staff Evaluation of Pressurized Thermal Shock," SECY 82-465, Appendix E, a weld material reference temperature (RTNDT) was established at -22*F based on a mean value plus two standard deviations.
Similar testing was not performed on all remaining material in the reactor coolant system. However, sufficient impact formedtomeetappropriatedesigncoderequirements(gpstingwasper-1 and a conser-vative RTNDT of 50*F has been established.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT with operation. The tech-i niques used to predict the integrated fast neutron (E > 1 MeV) fluxes l
of the reactor vessel are described in Section 3.4.6 oT the USAR ex-cept that the integrated fast neutron flux (E > 1 MeV) is 2.9x10I9 n/cm2, including tolerance at the year design life of the vessel.(5) vessel inside surface, over the 40 Since the neutron spectra and the flux measured at the samples and re-I actor vessel inside radius should be nearly identical, the measured transittcq shift for a sample can be applied to the adjacent section of the reaitor vessel for later stages in plant life equivalent to the difference te calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calibrated azimuthal neutron flux variation. The Amendment No. 22,47,64,7#,77 2-5 l
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t 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) maximum integrated fast neutron (E 2 1 MeV) exposurp of the reactor vessel including tolerance is computed to be 2.9x1019 n/cmz at the vessel inside surface for 40 years operation at 1500 MWt and 80% load The exposure at ghe ;/4t depth from the inner surface is com-factor.
19 n/cm.(5s The predicted TNDT shift for an puted to be 2.0x10 integrated fast neutron (E 21 MeV) exposure of 2.0x1019 n/cm2 is 209'F, the value obtained from the curve shown in Figure 2-3.
The actual shift in TNDT will be reestablished periodically during the plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the USAR. To compensate for any increase in the TNDT caused by irra-diation, limits on the pressure-temperature relationship are periodic-ally changed to stay within the stress limits during heatup and. cool-down. Analysis of the second removed irradiated reactor vessel sur-veillance specimen, combined with a weld chemical composition data and reduced fluence core loading designs initiated in Cycle 8, indicates that the fluence at the end of 15 Effective Full Power Years (EFPY) at 19n/qm2ontheinsidesurp1ce of the reac-1500 MWt will be 1.4xL0 at the 1/4t depth.(
This re-tor vessel and 8.1x10 8 n/cmz sults in a total shift of the RTNDT of 165*F for the area of great-est sensitivity (weld metal) at the 1/4t location as determined from Figure 2-3.
Operation through fuel Cycle 13 will result in less than 15 EFPY.
The limit lines in Figures 2-1A and 2-1B are based on the following:
A.
Heatup and Cooldown Curves - From Section III of the ASME Code, Appendix G-2215.
KIR - 2 KIM + KIT KIR Allowance stress intensity factor at temperatures
=
related to RTNDT (ASME III Figure G-2100.1).
KIM Stress intensity factor for membrane stress (pres-sure). The 2 represents a safety factor of 2 on pressure.
KIT Stress intensity factor radial thermal gradient.
=
The above equation is applied to the reactor vessel belt-line.
For. plant heatup the thermal stress is opposite in sign from the pressure stress and consideration of a heatup rate would allow for a higher pressure.
For heatup it is therefore conservative to consider an isothermal heatup or KIT = 0.
Amendment No. 27, #7, $#, 7#, 77 2-6
f-RCS PRESS-TEMP LIMITS HEATUP-15 EFPY REACIOR NOT CRITICAL 1500 MWt PRESSURIZER PRESS (PSIAl 3200 I
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3000 2800
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1800
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1600 HORO TESl N
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1200 NON-CRITICAL
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LOWEST SERVEE
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TFyPERATUR E 545 400
-BOLTUP -~
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50 100 150 200 250 300 350 400 450 500 RC INLET TERP (DEG F) Tc FORT CALHOUN FIGURE TECHNICAL SPECIFICATIONS 2-1A AmendmentNo./f,77
r RCS PRESS-TEMP LIMITS C00LDOWN 15 EFPY REACTOR NOT CRITICAL 1500 MWt PRESSURIZER PRESS (PSIA) 3200 t
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RC INLET TEMP (DEG Fl Tc FORT CALHOUN FIGURE l l
TECHNICAL SPECIFICATIONS 2-1B Amendment No. 77, 77
PREDICTED RADIATION INDUCED NDTT SHIFT FORT CALHOUN REACTOR VESSEL BELTLINE A Tadt 500 400 300
/
/
200
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,/
/
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100 2
3 4 56789 2
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- iE18 iE19 SEi9 2
NeutronFluence,n/ca FORT CALHOUN FIGURE TECHNICAL 2-3 SPECIFICATIONS Amendment No. 74, 77
~:. :-
2.0 LIMITING CONDITIONS FOR OPERATION 2.3' Emergency Core Cooling System (Continued)
(3)
Protection Against low Temperature Overpressurization The following limiting conditions ~ shall be applied during sched-uled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor vessel head, a pressurizer safety valve, or a PORV is removed.
Whenever the reactor coolant system cold leg temperature is be-
~
low 231*F, at least one (1) HPSI pump shall be disabled.
Whenever the reactor coolant system cold leg temperature is be-low 220*F, at least two (2) HPSI pumps shall be disabled.
Whenever the reactor coolant system cold leg temperature is be-low 182*F, all three (3) HPSI pumps shall be disabled.
In the event that no charging pumps are operable, a single HPSI pump may be made operable and utilized for boric acid injection to the core.
Basis The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The ~ reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.
The SIRW tank contains a ulnimur. of(M3,000 gallons of usable water containing at least 1700 ppm boron.11 This is sufficient boron concentration to provide a shutdown margin of 5~,, including allowances for uncertainties, wjth all control rods withe awn and a new core at a temperature of 60*F.1 )
2 The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analysgs.
The minimum 116.2 inch level corresponds to a volume of 825 ft3 and the maximum 128.1 3
inch level corresponds to a volume of 895.5 ft,
Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked.
Since the system is used for shut down cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.
2-22 Amendment No. 17,39,43,#7 l
pf, 76, 77
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 ~ Emergency Core Cooling System (Coatinued) be available for emergercy core cooling, but the contents of one of the tanks is assumed to be lost through the reactor. coolant system.
In addition, of the three high-pressure safety injection pumps and the-two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure:and cne low pressure operate while ysis(gne of each type is assumed to operate in the small-break anal-only 3); and also that 25% of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sizes considered are shown on FSAR, Appendix ~X, Figures 1-19 (Amendment No. 34).
Inadvertent actuation of three (3) HPSI pumps and three (3) charging pumps, coincident with the opening of one of the two PORV's, would result in a peak primary system pressure of 1190 psia.
1190 psia corresponds with a minimum permissible temperature of 231*F on Figure 2-1B.
Thus, at least one HPSI pump is disabled at 231*F.
Inadvertent actuation of two (2) HPSI pumps and three (3) charging
- pumps, coincident with the' opening of one of the two PORV's, would result in a peak primary system pressure of 1040 psia.
1040 psia corresponds with a minimum permissible temperature of 220*F on Figure 1B.
Thus, at least two HPSI pumps will be disabled at 220*F.
. Inadvertent actuation of one (1) HPSI and three'(3) charging pumps, coincident with opening of one of the two PORV's, would result in a peak primary system pressure of 685 psia.
685 psia corresponds with a minimum allowable temperature of-182*F._on Figure 2-18.
Thus all three HPSI pumps will be disabled at 182*F.
Inadvertent actuation of three (3) charging pumps, coincident with the opening of one of the two PORV's, would result in a peak primary sys-
. tem pressure of 160 psia.
160 psia would correspond with a minimum allowable temperature that is less than the 82*F boltup temperature limit on Figure 2-18.- Therefore, operation of the charging pumps need not be restricted.
Removal of the reactor vessel head, one pressurizer safety valve, or one PORV provides sufficient expansion volume to limit any of the de -
sign basis pressure transients. Thus, no additional relief capacity is required.
Technical Specification 2.2(1) specifies that, when fuel is in the reactor, at least one flow path shall be provided for boric acid injection.to the core.
Should boric acid injection become necessary, and no charging pumps are operable, operation of a single HPSI pump would provide the required flow path.
2-23a Amendment No. 39, 47, 64 74,77
JUSTIFICATION, DISCUSSION, AND SIGNIFICANT HAZARDS CONSIDERATIONS FOR HEATUP AND C00LDOWN CURVES The Fort Calhoun Technical Specifications are being amended to update the cur-rent heatup and cooldown limit curves for continued safe operation of the re-actor vessel and associated primary coolant system beyond 8.5 Equivalent Full Power Years (EFPY). This application requests continued operation through 15 EFPY.
When determining the Limiting Conditions for Operation, the impact of the ini-tial nil-ductility transition reference temperature (RTNDT) and the fluence induced RTNDT shift of reactor vessel beltline welds must be considered.
The fluence induced temperature shift is a function of fluence and the chem'i-cal composition of the limiting reactor vessel beltline material.
In the past, the absence of specific weld chemical composition data required assump-tion of upper bound values for beltline weld copper and nickel content. The chemical composition of all Fort Calhoun reactor vessel beltline welds was recently documented through searches of Combustion Engineering (CE) welding records and through analysis of physical weld samples removed from identical welds traced to the reactor vessel head. Additional information on this pro-ject is provided in the attached discussion section. With specific weld chem-ical composition data, it is no longer necessary to assume the upper bound copper and nickel values for these welds. The effect of this change has been evaluated and the lower shell longitudinal weld seam, 3-410 was determined to be the most limiting beltline material with 0.23 w/o copper and 0.95 w/o nickel.
The Guthrie Mean Curve Equation, as referenced in 10 CFR 50.61, was
applied using the weld chemical composition data and the azimuthal flux dis-tribution to yield the predicted temperature shift as reflected in the pro-posed 15 EFPY heatup and cooldown limit curves. Likewise, Figure 2-3 has
~
been revised to predict the fluence induced temperature shift for the limit-ing reactor vessel beltline material. These curves will ensure that adequate fracture toughness is maintained throughout all conditions of normal opera-tion, including anticipated operational transients and system hydrostatic tests. Thus, the changes are made using the same methodology as the previous request, but make use of the specific weld chemistry data recently cbtained.
The proposed 15 EFPY heatup and cooldown limit curves are required for opera-tion beyond approximately August 25, 1986.
Commission approval of the pro-posed Technical Specification is, therefore, requested prior to August,1986 4
to ensure adequate curves are available for continued operation throughout the cycle without interruption.
Significant Hazards Considerations 10 CFR 50.91 requires that a licensee's request for amendment provide an anal-ysis addressing the three factors of 10 CFR 50.92 relative to significant haz-ards considerations. The proposed changes to the Technical Specifications do not involve a significant hazards consideration because operation of Fort Cal-houn Station in accordance with this change would not:
(1)
Involve a significant increase the probability of occurrences or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report because the change maintains conservative restrictions on pressure-temperature
limits for the reactor vessel based on recently obtained beltline weld chemical composition data and the azimuthal flux distribution.
(2)
Create the possibility of a new or different kind of accident than any previously evaluated because this application only revises the heatup and cooldown curves which are bounded by the existing Safety Analysis Report.
(3)
Involve a significant reduction in a margin of safet'.
The method-ology of 10 CFR 50.61 has been used to determine the value of the RTNOT shift. The use of the 10 CFR 50.61 methodology, and plant specific weld chemical composition data, enhances the accuracy of the RTNDT shift calculation. This methodology provides the necessary margin of safety to assure that the limit will not be exceeded.
s
DISCUSSION m
The following is a discussion of recently obtained reactor vessel beltline weld chemical composition data for the Fort Calhoun Station and its impact on Fort Calhoun's position with regard to the pressurized thermal shock (PTS) issue.
In an effort to better address the PTS issue, a searen of Combustion Engineer-l ing's, (C-E's) weld records was performed in 1984 to determine beltline weld chemistries for the Fort Calhoun Station reactor vessel. Weld chemical compo-sition analyses were obtained for all weld wire heats used in the beltline re-gion except weld wire heat 51989, which was used in the middle shell longitud-inal seam wclds. This heat was traced to the torus longitudinal seam welds on the Fort Calhoun reactor vessel head and it was determined that weld chip samples could be removed from these welds for analysis. A conference call involving OPPD, C-E and NRC personnel was held on 7/30/85 to review plans for As a result of this sampling this material during the 1985 refueling outage.
conversation and others between C-E and NRC personnel, several items of NRC concern were introduced and preliminarily resolved. These NRC concerns are l
addressed in the following paragraphs.
l The first concern was how the location of the weld seams could be accurately distinguished from the surrounding base metal. This was accomplished by polishing the areas to be sampled and then etching them with a nitric acid solution to reveal the weld outline. After the chip samples were removed the l
l sample areas were blended and inspected'by magnetic particle testing.
E
Photographs were taken of the prepared surfaces after etching, after chip removal and after blending at all locations to document that only weld material was removed.
The second concern was the ability to distinguish between uncharted weld re-pairs and cosmetic welds. C-E has determined that the possibility of an un-charted weld repair was small and that cosmetic welds were not performed on the OD weld surface, rather the submerged arc weld was ground smooth to the surface. To further minimize this concern, duplicate samples were obtained for each weld from different locations.
The results of an optical emission chemical analyses performed on the chip samples, including a check analysis by x-ray fluorescence for copper and nic-kel content, are shown in Table 1.
The report for all elements determined in the optical emission analysis is attached as Table 2.
The optical emission values are an average of two analyses. The x-ray fluorescence values repre-sent a single analysis. The chemical analysis results for wire heat 51989 are consistent with the values expected for a weld made with a Mil B-4 wire and a Linde Type 124 flux, as is indicated for these welds by the weld infor-mation record..
Likewise, the results of the chemical analyses for wire heat 13253 are cons. stent with a Mil B-4 Modified wire and Linde Type 1092 flux.
The D. C. Cook and Salem 2 surveillance welds were also made with heat 13253 and a Linde Type 1092 flux. The nickel content of the Salem 2 (0.72 w/o) and D. C. Cook (0.74 w/o) surveillance welds are almost identical to the Fort Calhoun (0.73 w/o) value, indicating that 13253 was the wire heat used, and further indicating that weld metal was sampled, since the base metal normally contains less than 0.60 w/o nickel. The copper contents between these welds vary significantly (D. C. Cook - 0.27 w/o, Salem 2 - 0.23 w/o, Fort Calhoun -
0.14 w/0). This can be attributed to the variation in copper coating on the coils of the wire making a heat of weld wire. This wide variation in copper has been observed on several other heats of wire for which multiple analyses are available. A portion of the samples from each weld seam was metallograph-ically examined using a Nital solution to reveal the microstructural charac-teristics.
In all cases, the examination showed the fine-grained ferritic structure of weld metal.
The adequacy of a smaller chip sample as opposed to a full boat sample was also questioned. This has been addressed by the fact that the Fort Calhoun closure head has a relatively small allowance for the removal of a sample and anything larger than the proposed chip sample might require a UT inspection and some degree of analysis. Meaningful results have been obtained since it was possible to do metallography and chemical analysis on the same chip spec-imens.
Using the results of the closure head weld sampling for heats 51989 and 13253, and other available records, copper and nickel contents have been determined for each weld deposit in the Fort Calhoun reactor vessel belt-line. These copper and nickel contents are presented in Table 3.
The chem-istry established for wire heat 27204 resulted from a search in October 1985 of the C-E Metallurgical and Materials Laboratory chemical analysis log books for weld deposit information and a review of data for the Diablo Canyon Unit
- 1 surveillance weld made with heat 27204. The lower shell longitudinal seau welds were each made using three heats of wire (27204,12008, and 13253).
It is not known whether only one or a combination of two of the wires were used to weld the ID of the seam.
It was assumed for conservatism that the weld wire with the highest chemistry factor was used to weld the ID of the seam.
Therefore, the copper and nickel content of wire heat 12008 was used for the 4
evaluation of these weld seams.
Supporting calculations for this amendment demonstrate that the lower shell k
longitudinal weld seam, 3-410 (weld wire heat 12008) is the limiting reactor The fluence induced temperature shift at this weld vessel beltline material.
was calculated by applying the Guthrie Mean Curve Equation referenced in 10 CFR 50.61 and conservatively assuming the Cycles 1-9 average azimuthal flux The RTNOT of the limiting f
distritation throughout reactor vessel life.
reactor vessel beltline material (weld seam 3-410) after 15 EFPY w l
Figure 2-3 can also be revised to reflect a reduced fluence in-l be 165'F.
duced temperature shift as a result of lower assumed copper and nickel chem-ical content for the reactor vessel beltline weld materials.
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Table 1 Wire Type:
Mil B-4 Mil B-4 Mil B-4. Mod
-. Mil B-4 Mod Heat No.:
51989 51989 13253 13253 Flux Type:
Linde 124 Linde 124 Linde 1092 Linde 1092 Flux Lot:
3687 3687 3791 3791 Weld Seam:
1-415C 1-415E 2-145/A 2-415/A (near 1-415C)
(near 1-415E)
CE Lab No.:
D-41589 D-41591 D-41588 D-41590 optical optical optical optical Emission Emission Emission Emission (X-Ray Flour.)
(X-Ray Flour.)
(X-Ray Flour.)
(X-Ray Flour.)
W/O W/O W/O W/O C
0.11 0.096 0.11 0.12 Mn 1.39 1.50 1.10 1.14 P
0.011 0.013 0.010 0.013 S
0.009 0.011 0.008 0.011 Si 0.30 0.36 0.17 0.18 Ni 0.20 (0.18) 0.13 (0.114) 0.72 (0.72) 0.74 (0.72)
Cr 0.08 0.08 0.04 0.04 Mo 0.47 0.52 0.43 0.44 Cu 0.16 (0.17) 0.18 (0.18) 0.14 (0.14) 0.14 (0.14) l l
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OMBUSTION ENGINEERING
- .L..".CICA1. & MATERIALS LABORATORY DATE:
11-13-85
. P.O. NO.
C-E JOB No.
99759617 PROJECT NO.
960001 +
CHEMICAL AllALYSIS REPORT C-E L:b No.
D41588 D41589 D41590 D41591 Cus t cime r ' So.
Area 1 Area L Area 2 Arca 2 Description 2-415-A 1-415-C 2-415-A 1-415-E C
.11
.)1
.12
.096 Mn 1.10 1.39 1.14 1.50 F
.010
.011
.013
.013 5
.008
.009
.011
.011 Si
.17
.30
.18
.36 Ni
.72
.20
.74
.13
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Cr
.04
.08
.04
.08 r.a 43
.47 44
.52 V
003
.004
.003
.004 Cb
<. 01 e. 01
<. 01
<. 01 Ti
<. 01 c. 01
<. 01 c. 01 Co
.016
.012
.016
.011 Cu
.14
.16
.14
.18 Al
.002
.006 002
.007 E
4.001
<.001
<.001 c.001 W
<.01
<. 01
<.01 e. 01 As
.010
.010
.013
.011
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Sn
.005
.006
.006
.006 Zr-e.001
<.001
<.001
<.001 N
.009
.015
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Fe-Kea::or Vessel Closure llcad s
Cr.sha Public Power Dist.
Fort Calhoun Station i
Repor:cd by:
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i Cemeussion Engineermg. Inc.
gii wes pa,n Street (615) 265 4G31 i
Cha.*anooga. Tennettee 37402 Tcie, ssetto i
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1 Table 3 Chemical Content of Fort Calhoun Beltline Welds Material Chemical Content Weld Seam (Wire Heat / Flux Lot) gg Hi gomment/ Source 2-410 A/C 51989/3687 0.17 0.17 Fort Calhoun closure 1
head longitudinal weld i
sample.
3-410 A/C 27204/3774 0.22 1.02 Average of multiple weld deposit records includ-3 ing PG&E Diablo Canyon surveillance weld.
13253/3774 0.21 0.73 Average of Salem #2 and Cook #1 surveillance welds, and Fort Calhoun closure head tcrus-to-dome girth seam weld samples.
12008/3774 0.23 0.95 Average of multiple weld deposit records of tan-dem arc welds in which second weld wire heat copper content known.
9-410 20291/3833 0.21 0.74 Cooper Station surveil-lance weld.
8-410 13253/3774 0.21 0.73 (see 3-410)
-