ML20202F442
| ML20202F442 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/10/1986 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20202F421 | List: |
| References | |
| NUDOCS 8607150166 | |
| Download: ML20202F442 (23) | |
Text
'
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).
For this plant, based upon surveillance materials tests, weld chemical composition data, and the ef-fect of a reduced vessel fluence rate provided by core load designs beginning with fuel Cycle 8, the predicted surface fluence at the critical reactor vessel beltline weld material for 40 years at 1500 MWt and an 80% load factor is 2.9x10 9 n/cm2, 1
The flux reduction applied to the fluence calcula-tions was based on a Cycle 1-9 average azimuthal flux distribution plot generate ~d using 00T 4.3.
The predicted transition 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 movsd parallel to the temperature axis (horizontal) in the direction of increasing temperature a dis-tance equivalent to the transition temperature shift during the period since the curves were last con-structed. The boltup temperature limit line shall remain at 82*F as it is set by the NDTT of the reac-tor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 182*F because components related to this temper-ature are also not subject to fast neutron flux.
(d)
The Technical Specification 2.3(3) shall be revised each time the curves of Figure 2-1A and 2-1B are revised.
Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads du to reactor coolant system temperature and pressure changes. 1) These cyclic loads are introduced by normal unit load transients, reactor trips and startup 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.
8607150166 860710 2-4 PDR ADOCK 05000285 P
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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 III(2) limitations are derived by using the rules contained erations. Thege 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 temp-erature for transverse direction (RTNDT) of -12*F.
The mean initial RTNDT value for the Fort Calhoun submerged arc ves-sel weldments was determined to be -56*F with a standard deviation of 17'F.
By applying the shift prediction methodology of the proposed Regulatory Guide 1.99, Revision 2, a weld material adjusted reference temperature (RTNDT) was established at 10*F based on the mean value plus two standard deviations.
The standard deviation was determined by using the root-mean-squares method to combine the margin of 28'F for uncertainty in the shift equation with the margin of 17'F, for uncertainty in the initial RTNDT value.
Similar testing was not performed on all remaining material in the reactor coolant system. However, sufficient impact testing was per-formed to meet appropriate design code requirements (3) and a con-servative 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-niques used to predict the integrated fast neutron (E > 1 Mev) fluxes of the reactor vessel are described in Section 3.4.6 of the USAR ex-cept that the integrated fast neutron flux (E > 1 MeV) is 2.9x1019 4
n/cm2, including tolerance at the inside surface of the critical 4
reactor vessg] beltline weld material, over the 40 year design life of the vessel. W Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in 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 2-5
s 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatuo and Cooldown Rate (Continued) maximum integrated fast neutron (E 2 1 MeV) exposure of the reactor vessel at the critical r.eactor vessg belt}ine location, including tolerance, is computed to be 2.9x10 n/cm at the vessel inside surface for 40 years operation at 1500 MWt and 80% load factor. The predicted shift at this location at the 1/4t depth from the inner sur-face is 332*F, including margin and was calculated using the shift pre-diction equation of the proposed Regulatory Guide 1.99, Revision 2.
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 T nT N
caused by irradiation, limits on the pressure-temperature reTationship are periodically changed to stay within the stress limits during heat-up and cooldown. Analysis o vessel surveillance specimen {8jhe second removed irradiated reactor
, combined with weld chemical composition data and reduced fluence core loading designs initiated in Cycle 8, indicates that the fluence at the end of 15 Efgetive Full 2
Power Years (EFPY) at 1500 MWt will be less than 1.4x10 n/cm on the inside surface of the reactor vessel.
This results in a total shift to the initial RT UDT f 285'F shift, including margin, for the area of greatest sensitivity (weld metal) at the 1/4t location as determined from Figure 2-3.
Operation through fuel Cycle 16 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, conserva-tive to consider an isothermal heatup or KIT = 0.
2-6
(
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatue and Cooldown Rate (Continued)
For plant cooldown thermal and pressure stress are additive.
MM EE KIM t
ASME III, Figure G-2214-1 Mg Pressure, psia P
=
Vessel Radius - in.
R
=
Vessel Wall Thickness - in.
t
=
KIT " MT aTy T
ASME III, Figure G-2214-2 M
=
ATg= Highest Radial Temperature Gradient Through Wall at End of Cooldown KIT is, therefore, calculated at a maximum gradient and is considered a constant - A for cooldown and zero for heatup.
MM R is also a constant = B.
t Therefore:
KIR = AP + B P=KIR - B A
K p is then varied as a function of temperature from Figure Gi2110-1 of ASME III and the allowable pressure calculated.
Hydrostatic head (48 psi) and instrumentation errors (12*F and 32 psi) are considered when plotting the curves.
B.
System Hydrostatic Test - The system hydrostatic test curve is developed in the same manner as in A above with the exception that a safety factor of 1.5 is allowed by ASME III in lieu of 2.
C.
Lowest Service Temperature - 50*F + 120*F + 12*F = 182*F.
As indicated previously, an RTNDT for all material with the exception of the reactor vessel belt-line was established at 50*F.
10 CFR 50, Appendix G, IV.a.2, requires a lowest service temperature of RTNDT + 120*F for piping, pumps and valves.
Below this temperature a pressure of 20 percent of the system hydrostatic test pressure (.20)(3125) 32 psi = 545 psia cannot be exceeded.
2-7
l 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatuo and Cooldown Rate (Continued)
D.
Boltup Temperature = 10*F + 60*F +'12*F = 82*F.
At pressure below 545 psia, a minimum vessel temperature must be maintained to comply with the manufacturer's specifications for tensioning the vessel head. This temperature is based on previous NDTT methods.
This temperature corresponds to the measured 10*F NDTT of the reactor vessel flange, which is not subject to radiation damage, plus 60*F data scatter in NDTT measurements, plus 12*F instrument error.
References (1)
USAR, Section 4.2.2 (2)
ASME Boiler and Pressure Vessel Code,Section III (3)
USAR, Section 4.2.4 (4)
USAR, Section 3.4.6 (5)
Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Irradiated Capsule W-225, Revision 1, August,1980 (6)
Technical Specification 2.3(3)
(7)
Article IWB-5000, ASME Boiler and Pressure Vessel Code,Section XI (8)
Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Irradiated Capsule W-265, March 1984 2-7a
RCS PRESS-TEMP LIMITS HEATUP 15 EFPY f
REACTOR NOT CRITICAL 1500 MWt PRESSURIZER PRESS (PSIA) 3200 l
3000 E
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2800 E
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LEAK CHECK S 1600 HYDil0 TEST N
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TiHPERATURE 545
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50 100 150 200 250 300 350 400 450 500 RC INLET TEMP (DEG F) It FORT CALHOUN FIGURE TECHNICAL SPECIFICATIONS 2-1A
RCS PRESS-TEMP LIMITS C00LOOWN 15 EFPY REACTOR NOT CRITICAL 1500 MWt PRESSURIZER PRESS (PSIA) 3200 i
3000
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0 50 100 150 200 250 300 350 400 450 500 RC INLET TEMP (DEG F) Ic FORT CALHOUN FIGURE TECHNICAL SPECIFICATIONS 2-1B
PREDICTED RADIATION INDUCED NDTT SHIFT FORT CALHOUN REACTOR VESSEL BELTLINE SRhd 500 400
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NeutronFluence,n/c:1 FORT CAIlIOUN FIGURE i
TECHNICAL 2-3 SPECIFICATIONS
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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 i
valve, or a PORV is removed.
I Whenever the reactor coolant system cold leg temperature is be-low 320*F, at least one (1) HPSI pump shall be disabled.
Whenever the reactor coolant system cold leg temperature is be-low 312*F, at least two (2) HPSI pumps shall be disabled.
Whenever the reactor coolant system cold leg temperature is be-low 271*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-l up, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and 4
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 minimum of(2S3,000 gallons of usable water containing at least 1700 ppm boron. 1) This is sufficient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn 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 analyses. 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 i
appropriate valves locked.
Since the system is used for shut down l
cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.
2-22
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued) be available for emergency core cooling, but the contents of one of i
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 one low pressure operate while only one of each type is assumed to operate in the small break anal-ysis(5); and also that 25% of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot i
spot fuel clad temperatures for the break sizes considered are shown on FSAR, Appendix K, Figures 1-19 (Amendment No. 34).
j 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 320*F on Figure i
2-18.
Thus, at least one HPSI pump is disabled at 320*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 312*F on Figure 2-18.
Thus, at least two HPSI pumps will be disabled at 312*F.
Inadvertent actuation of one (1) HPSI and three (3) charging pumps, i
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 271*F on Figure 2-18.
Thus all three HPSI pumps will be disabled at 271*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-i tem pressure of 160 psia.
160 psia would correspond with a minimum allowable temperature that is less than the 82*F boltup temperature i
limit on Figure 2-1B.
Therefore, operation of the charging pumps need not be restricted.
j Removal of the reactor vessel head, one pressurizer safety valve, or one PORV provides sufficient expansion volume to limit any of the de-l 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
ATTACHMENT B JUSTIFICATION, DISCUSSION, AND SIGNIFICANT HAZARDS CONSIDERATIONS FOR HEATUP AND C00LD0WN 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), the fluence induced RTNDT shift of reactor vessel beltline welds and an appropriate margin must be considered. The fluence induced temperature shift is a func-tion of fluence and the chemical composition of the limiting reactor vessel beltline material.
In the past, the absence of specific weld chemical compo-sition data required assumption of upper bound values for beltline weld cop-per 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 sam-ples removed from identical welds traced to the reactor vessel head. Addi-tional information on this project is provided in the attached discussion sec-tion. With specific weld chemical composition data, it is no longer neces-sary 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 projected fluence at the
4 i
critical reactor vessel beltline weld was determined using the following equation:
pl.D. = [8.8x1018 + (EFPY - 5.92) (4.8x1019)/32] (.60) n/cm2 This equation was developed using results from the analysis of surveillance l
i capsule W-265 which was removed after 5.92 EFPY, the end of Cycle 7.
The i
predicted reactor vessel I.D. fluence was reported to be 8.8x1018 n/cm2
]
by Combustion Engineering in the W-265 analysis report and was projected to l
reach 4.8x1019 n/cm2 by the end-of-life. However, following Cycle 7, 3
reduced fluence core loading patterns were initiated. To take credit for i
i this reduction, an azimuthal flux distribution plot for Cycles 1-9 average, j
generated using DOT 4.3, was used to determine the azimuthal flux reduction i
value at the critical weld location (.60) for use in the fluence equation above.
I The shift prediction equation developed in the proposed Revision 2 of Regula-i tory Guide 1.99 was chosen for use in this submittal, however, future shift calculations will re-evaluate its applicability to the Fort Calhoun Station and consider other correlations which may better fit the available data. The corrected attenuation equation developed in the CE Owners Group Response to Proposed Revision 02, Reg. Guide 1.99 was used to predict the fluence into i
the reactor vessel wall. The root-mean-squares method was used to combine the margin associated with the initial RTNDT (17'F) and the margin associ-ated with the shift prediction equation (28'F). The predicted temperature shift plus a 2 sigma nargin were applied to -56*F initial RTNDT baseline curves to generate the proposed 15 EFPY heatup and cooldown limit curves.
Likewise, Figure 2-3 has been revised to predict the fluence induced temper-ature shift, including margin, for the limiting reactor vessel beltline ma-
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terial. These curves will ensure that adequate fracture toughness is main-tained throughout all conditions of normal operation, including anticipated operational transients and system hydrostatic tests.
The proposed 15 EFPY heatup and cooldown limit curves are required for opera-t tion beyond approximately August 25, 1986.
Commission approval of the pro-i' posed Technical Specification is, therefore, requested prior to August 10, 1986 to ensure adequate curves are available for continued operation through-out the cycle without interruption.
2
)
Significant Hazards Considerations This Technical Specification amendment will not 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 be-cause the change maintains conservative restrictions on pressure-temperature limits for the reactor vessel based on recently obtained beltline weld chem-ical composition data and the Cycle 1-9 average azimuthal flux distribution.
The probability of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be created because this application only revises the heatup and cooldown curves which are bounded by the existing Safety Analysis Report.
The margin of safety as defined in the basis for the Technical Specifications will not be reduced because the shift prediction equation of the proposed Revision 2 of Regulatory Guide 1.99 has been used to determine the value of the RTNDT shift.
I i
e I
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ATTACHMENT B l
DISCUSSION The following is a discussion of recently obtained reactor vessel beltline weld chemical composition data for the Fort Calhoun Station and its impact on i
Fort Calhoun's position with regard to the pressurized thermal shock (PTS) i l
l issue.
t In an effort to better address the PTS issue, a search of Combustion Engineer-f ing's, (CaE's) weld records was performed in 1984 to determine beltline weld f.
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 welds. This heat was traced to the torus longitu'dinal seam welds 4
j on the Fort Calhoun reactor vessel head and it was determined that weld chip l
samples could be removed from these welds for analysis. A conference call l
involving 0 PPD, C-E and NRC personnel was held on 7/30/85 to review plans for i
sampling this material during the 1985 refueling outage. As a result of this conversation and others between C-E and NRC personnel, several items of NRC l
concern were introduced and preliminarily resolved.
These NRC concerns are addressed in the following paragraphs.
The first concern was how the location of the weld seams could be accurately I
j distinguished from the surrounding base metal.
This was accomplished by j
polishing the areas to be sampled and then etching them with a nitric acid l
solution to reveal the weld outline. After the chip samples were removed the sample areas were blended and inspected by magnetic particle testing.
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~. -,.. - -, -
--,.-c,-,
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,.,-...__,_---------,,~.-----,v~-.
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 records.
Likewise, the results of the chemical analyses for wire heat 13253 are cansistent 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 i
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 (0.72 w/o) value, indicating that 13253 was the wire heat used, and further indicating that weld metal was sampled, since the base metal normally i
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 -
4 e
0.14 w/o). This can be attributed to the variation in copper coating on the l
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-i 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 j
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 i
determined for each weld deposit in the Fort Calhoun reactor vessel belt-i 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 l
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 seam i
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 4
wire with the highest chemistry factor was used to weld the ID of the seam.
1 i
0.14 w/o). 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 seam 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 evaluation of these weld seams.
f 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) f 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 i
P 0.011 0.013 0.010 0.013 S
0.009 0.011 0.008 0.011 4
Si 0.30 0.36 0.17 0.18
!!i 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) 1 i
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O M B USTION ENGINEERING
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CL.".CICA1. & MATERI ALS 1.ABORATORY DATE:
11-13-85 P.O. NO.
C-E J03 NO.
99759617 PROJECT NO.
960001 CHEMICA1. ANAI.YSIS REPORT C-E Leb No.
D41588 D41589 D41590 D41591 Customer No.
Area 1 Arca 1 Arca 2 Arca 2 Description 2-415-A 1-415-C 2-415-A 1-415-E C
.11
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.096 th 1.10 1,39 1.14 1.50
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.010
.011
.013
,og3 5
.008
.009
.011
.011 51
.17
.30
.18
.36 Ni
.72
.20
.74
,13 Cr
.04
.08
.04
.08 PM 43
.47
.44
.52 V
003
.004
.003
.004 Cb e. 01 e.01
<.01
<.0L Ti
<.01
<. 01
<.01
.:. 01 Co
.016
.012
.016
.011 Cu
.14
.16
.14
.18 j.1
.002
.006
.002
.007 L
<.001
<.001
<.001
<.001 W
<. 01
<.01
<. 01
<. 01 As
.010
.010
.013
.011 Sn
.005
.006
.006
.006
't
<.001
<.001
<,001
<.001
.009
.015 0
.009
.009 re
.m i
Rea::or Vessel Closurc !! cad Cr.cha Public Po *cr Dist.
For: Calhoun Station I
Comeust4on Engineenng. loc.
gli West Main Street (615) 265 4631 Cna. aacoca. Tennessee 37402 Tcie, 5se:29
.... c-
,-,v,
^4 4
'i a
- t lQ k
g Table 3
.k Chemical Content of Fort Calhoun Beltline Welds 2
i Material Chemical Content Weld Seam (Wire Heat / Flux Lot)
_C_u Hj.
Comment / Source
'l3 2-410 A/C 51989/3687 0.17 0.17 Fort Calhoun closure
- 5 head longitudinal weld
.t sample.
f 3-410 A/C 27204/3774 0.22 1.02 Average of multiple weld 3
deposit records includ-I ing PG&E Diablo Canyon
{
surveillance weld.
i 13253/3774 0.21 0.73 Average of Salem #2 and Cook #1 surveillance welds, and Fort Calhoun 5
closure head torus-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)
4 ATTACHMENT C i
o i
l a
Question 1:
1 Weld metal margin: Why was no margin applied to ARTNDT?
Response: The base curves had some margin included in them.
It was 2 oj -
34*F.
Since some applications of these curves also require the incorporation of c, all margin has been removed from the baseline curves. When margin a
needs to be included, all of the appropriate terms and values can be clearly included.
Question 2:
NRC Staff believes R.G.1.99, Rev. 2, is more representative than use of l
Guthrie formula.
Response: The District believes that the Guthrie formula, as defined in 10 CFR 50.61, is a representative formula not only for PTS screening criteria calculation but also for heatup/cooldown curve calculations.
As requested, the District will use proposed Reg. Guide 1.99, Rev. 2, to
]
calculate heatup/cooldown curves for this submittal.
The District still believes that the 10 CFR 50.61 equation is valid for this purpose and has not eliminated its use from future submittals.
The District further states that 10 CFR 50.61 is the only method it will use to calculate E0L fluences and fluences at screening criteria limits.
i Question 3:
j Staff calculates 20% of the pre-service system hydro test pressure as 0.2 x i
1.25 x 2500 - 625 psia vs the 545 psia in the submittal. Why the discrep-ancy?
j Response: No discrepancy exists.
The calculation from the 625 psia to 545 psia is clearly stated in the current Technical Specifications on page 2-7 in section 2.1.2.C.
l Question 4:
{
How did OPPD obtain 162*F for lowest service temperature? Staff calculates i
182*F.
Response: The District concurs that 182*F should be used and that the re-vised submittal will incorporate a lowest service temperature of 182*F.
i t
i Ouestion 5:
Submittal has omitted a page of references.
i Response: The District has included the references in the text.
i i
I i
i
d o
l 3
Ouestion 6:
i How was the fluence attenuated through the vessel wall?
Response: The attenuation was based on information contained in W-265, a i
surveillance capsule examination report prepared by Combustion Engineering.
Page 36 of the report showed ID,1/4t, and 3/4t fluences. These values were ratioed.
The ratios were used for Future attenuation calculations.
The ratios are used only when attenuation is not clearly defined.
In the current submittal using proposed R.G.1.99, Rev. 2, the District will use the following equation for attenuation:
ARTNDT/ARTNDTsurface "
- Question 7:
j The fluence in the May 7,1986 PTS submittal and the fluence in this submit-tal are not consistent. Why the discrepancy?
I Response: The fluences for each submittal were calculated with different
^
assumptions. The PTS submittal was calculated with no flux reduction credit assumed for Cycles 1 through 7.
Credit was taken for fuel loading pattern changes (low radial leakage) made for Cycle 8 and beyond. The equation stated below mathematically reflects these assumptions.
910 - 8.8x1018 + 0.7(EFPY - 5.92)(4.8x10l9)/32 n/cm2 The Technical Specification submittal utilized 00T 4.3 calculations to deter-mine a Cycle 1-9 average azimuthal flux reduction. This was incorporated into the fluence extrapolation equation as shown below.
Longitudinal Welds SID = [8.8x1018 + (EFPY - 5.92)(4.8x1019)/32] (.6) n/cm 2
Circumferential Welds 9ID = [8.8x1018 + (EfPY - 5.92)(4.8x1019)/32] (.92) n/cm 2
Question 8:
The 100*/ hour heatup curves for the 3/4t should be included.
Response: The District's submittal will include 3/4t curves when they are limiting.
Ouestion 9:
There are no pressure temperature limits for heatup or cooldown for the reac-tor critical.
Response: The District removed these curves via a previous amendment.
They l
are not restrictive since present Technical Specifications require all 4 l
1
o a
1 reactor coolant pumps to be in operation prior to criticality. Additionally, exception.is for physics testing when the power level shall be 10-1.The only the inlet temperature must be at least 515'F prior to criticality.
=
7 or less. Further, the reactor cannot be made or maintained critical unless the inlet temperature is NDTT + 120*F. These restrictions are delineated in Technical Specification sections; 2.1.1(1),1.3 Table 1.1 No. 2, 2.10.1(1),
and 2.10.1(2) which are located on pages 2-1, 1-10a and 2-48, respectively.
The District feels that incorporating additional if not less restrictive guidance could be confusing and lead to a violation of the more restrictive requirements.
96 6
e 1'
1 T
4
'.