ML20237E672
| ML20237E672 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/21/1987 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 LIC-87-692, NUDOCS 8712290120 | |
| Download: ML20237E672 (6) | |
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Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 2247 402/536 4000 December 21, 1987 LIC-87-692 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
References:
1.
Docket No. 50-285 2.
Federal Register, Vol. 50, No. 141, Tuesday, July 23, 1985 3.
Letter from OPPD (R. L. Andrews) to NRC (A. C. Thadani) dated January 23, 1986 (LIC-86-024) 4.
Letter from OPPD (R. L. Andrews) to NRC (A. C. Thadani) dated May 7, 1986 (LIC-86-286) 5.
Letter from NRC (Walter A. Paulson) to OPPD (R. L. Andrews) dated March 5, 1987 6.
Letter from OPPD (W. C. Jones) to NRC (D. G. Eisenhut) dated April 25, 1984 (LIC-84-124) 7.
Letter from LeBoeef, Lamb, Lefby & MacRae (Attorneys for OPPD) to NRC (H. R. Denton) dated July 17, 1986 8.
Letter from OPPD (W. C. Jones) to NRC (James R. Miller) dated May 30, 1984 (LIC-84-150) 9.
Letter from NRC (J. A. Calvo) to OPPD (R. L. Andrews) dated November 17, 1987 Gentlemen:
l SUBJ ECT :
Revised Pressurized Thermal Shock Submittal In accordance with Reference 2, Omaha Public Power District (OPPD) submitted to the Nuclear Regulatory Commission the projected values of RTPTS (at the inner vessel surface) for the Fort Calhoun Station Unit No. 1 reactor vessel beltline materials (References 3 and 4).
NRC approvals for these submittals were ob-tained in Reference 5.
This revised pressurized thermal shock analysis is being submitted since it has been proposed that Reg. Cuide 1.99, Draft Rev. 2 be imple-mented for use with the current 10 CFR 50.61 PTS screening criteria.
The Refer-ence 4 fluence prediction equations have been reevaluated to remove excess con-8712290120 871221 DR ADOCK 050 5
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.s Docuent Control Desk December 16, 1987 Page 2 servatism and provide a more accurate representation of Fort Calhoun's PTS sta-tus.
Projected values of RT f r the reactor vessel beltline materials were FTS determined using the current methodology of 10 CFR 50.61 as well as the proposed revision using Reg. Guide 1.99, Draft Rev. 2.and are provided in the attached tables. The revised fluence prediction equations were used to examine the im-pact of a license extension from 2008 to 2013 considering both the current 10 CFR 50.61 requirements and Reg. Guide 1.99, Draft Rev. 2.
Other information and data' presented in the original submittal remain unchanged.
RTPTS calculations were performed in accordance with 10 CFR 50.61, to deter-mine the RT for each weld and plate in the reactor vessel beltline at 32 PTS EFPY and at the projected end of license life.
The end of license life was cal-culated assuming a capacity factor of 77% beyond Cycle 9 through the year 2008 and is projected to be 26 EFPY. With the proposed five year license extension (Reference 7) to 2013 the end of license life is projected to be 30 EFPY.
The fluence values used in the RT calculations are based largely on values reportedintheanalysisof'survebancecapsuleW-265(Reference 6),whichwas removed after Cycle 7 (5.92 EFPY). The Combustion Engineering analysis of thi$8 survgillancecapsulereportedapeakvessol/cladintergcef1penceof8.8x10 n/cm and projected an end-of-life fluence of 4.8 x 10 n/cm after 32 EFPY.
Cycles 1-7 used standard symmetric (i.e., out-in-in) fuel management, whereas low radial leakage fuel management initiated in Cycle 8 significantly reduces the pro-jected peak end-of-life fluence.
No credit was assumed for axial flux distribu-tion or for an asymmetric core loading pattern, with even greater flux reduction that was used in Cycle 10.
Azimuthal flux distributions for the reactor vessel werepreviouslycalculatedusingDOT4.3assummarizep0 plotted on the attached figure relative to 4.73 x 10 n/cm -sec.
The Cycle 1-7 curve was used to determine the peak fluence to reactor vessel beltlf.ne welds through 5.92 EFPY (Cycles 1-7).
The Cycle 8 curve was conservatively applied to all successive cycles. The limiting longitudinal welds (i.e., 3-410) are located at 60*, 180*, and 300* which correspond to the octant folded locations of 30*,
0*, and 30*, respectively.
The 2 410 welds are located at 0*,
120*, and 240' which correspond to the octant folded locations of 0*,
30*, and 30*, respect-ively.
For these longitudinal welds, a peak flux multiplier of 0.68 (based on the 0* location).was assumed through Cycle 7 and a factor of 0.50 (based on the 30* location) was used beyond Cycle 7 in accordance with the attached figure. A factor of 0.68 (based on the 45* location) was also assumed for circumferential welds beginning in Cycle 8.
Should the Reg. Guide 1.99, Draft Rev. 02, equation be implemented for use as the PTS screening criteria in 10 CFR 50.61, considera-tion will be given to taking further credit for the axial flux distribution and other flux reductions observed after Cycle 8.
hsed on these assumptions, the following revised EFPY dependent fluence equations were developed:
Longitudinal Veld Seams if - [ 8. 8 x 1018](0.68) + [(EFPY-5.92)(4.8 x 1019)] pg 2
g 32 Circumferential Veld Seams and Plate J - [8.8 x 1018) # [(EFPY-5.92)(4.8 x 1019)] g g g I
2 32
Document Control Desk December 16, 1987 Page 3 The results of the revised RTPTS calculations using the current 10 CFR 50.61 rule and the above fluence equations are presented in Table 1.
As evidenced from this table, the limiting beltline material was found to be the lower shell longitudinal weld seams, 3-410.
These seams are projected to have an RT PTS of 216*F at the end of license life (26 EFPY) and 224*F at 30 EFPY (2013).
Table 1 has been expanded to also include data for reactor vessel plate mater-ial. Using the methodology of 10 CFR 50.61, all Fort Calhoun Station reactor vessel beltline materials fall well below the PTS screening criteria throughout life.
Table 2 is provided to show the effects of using Reg. Guide 1.99, Draft Rev.
02, in place of the current rule should 10 CFR 50.61 be amended as indicated by discussions with your staff.
The limiting longitudinal weld is one of the 3-410 welds, however, the 0.22 w/o Cu, 1.02 w/o Ni weld is more limiting for Reg. Guide 1.99, Draft Rev. 02 than the 0.23 w/o Cu, 0.95 w/o Ni weld which was most limiting using 10 CFR 50.61.
Since the data in Table 2 indicates that the 270* screening criteria will be exceeded at approximately 17.6 EFPY when using the Reg. Guide 1.99, Draft Rev.
2 equntion, OPPD plans to perform additional DOT 4.3 calculations to better de-fine the accumulated fluence to the critica) 3-410 weld seams through the end of Cycle 10, considering both the r-0 and r-z components.
Once this data is available, OPPD will use it to evaluate the current Adjusted Nil Ductility keferenco Temperature for the Fort Calhoun reactor vessel and determine methods of ratisfying current and future requirements of 10 CFR 50.61, including flux reduction. Uhen these methods have been identified, OPPD will provide the sup-plemental information requested in Reference 9, regarding the five year license extension request for Fort Calhoun.
Sincetely, R. L. Andrews Division Manager Nuclear Production RLA/rh Attachments c:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, DC 20036 l
A. Bournia, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector
TABLE 1 Revised RT f r Fort Calhoun Beltline Materials FTS Using 10 CFR 50.61 Weld Cu Ni Chem.*
RT RT RT RT PTS PTS PTS PTS geam E/Q EZQ Factor 17.6 EFPY 26 EFPY 30 EFPY 32 EFPY 2-410 0.17 0.17 80.02 92*F 10l*F 104*F 106*F (longitudinal) 3-410 0.23 0.95 174.58 197'F 216*F 224'F 228'F (longitudinal) 8-410 0.21 0.73 142.36 176*F 193*F 200*F 203*F (circumferential) 9-410 0.21 0.74 143.09 177'F 194*F 201'F 204'F (circumferential)
D-4802 0.12 0.56 69.92 121*F 129*F 133*F 134'F (intermediate shell-plate)
D-4812 0.12 0.60 71.60 123'F 132*F 135'F 137'F (lower shell-plate)
- Chemistry Factor - -10 + 470(Cu) + 350(Cu)(Ni) 10 CFR 50.61 Equation:
PTS - I + M + (-10 + 470(Cu) + 350(Cu)(Ni)]f.270 0
RT Where; f-calculatedvalueofneu{gonfluenceatthereactorvessel/ clad interface divided by 10 For Weld Material:
I-generic mean value of initial reference temperature - -56*F for welds made with Linde 1092 and 124 fluxes, M-margin to cover uncertainties in initial RTNDT - 59'F since generic value of I was used.
I For Plate Material:
l I-initial reference temperature of irradiated material as defined in the ASME Code - -12*F for reactor vessel beltline plate material.
M-margin to cover uncertainties
-in initial RT M - 48' since a NDT measured value of I was used.
The FTS criteria applied to the vessel ID for longitudinal weld seams and plate material is RTPTS - 270*F and for circumferential weld seams is RTPTS - 300'F.
TABLE 2 ART for Fort Calhoun Beltline Materials hfngReg. Guide 1.99,DraftRev.02 Weld Cu Ni Chem.*
ART ART ART ART NDT NDT NDT NDT Eeam EZQ HZQ Factor 17.6 EFPY 26 EFPY 30 EFPY 32 EFPY 2-410 0.17 0.17 89.45 109"F 118'F 121*F 122*F (longitudinal) 3-410 0.22 1.02 234.50 270*F 292*F 300*F 304*F (longitudinal)-
8-410 0.21 0.73 185.45 232'F 248'F 254*F 257'F (circumferential) 9-410 0.21 0.74 187.10 234*F 250*F 256*F 259'F (circumferential)
D-4802 0.12 0.56 82.20 135'F 142*F 144'F 145'F (intermediate shell-plate)
D-4812 0.12 0.60 83.00 136*F 143*F 145*F 146*F (lower shell-plate)
Reg. Guide 1.99, Draft Rev. 02, Equation:
I + M + (CF)f(0.28 - 0.10 log f) - Adjusted Nil Ductility ARTNDT -
Reference Temperature Where;
- CF -
Chemistry Factor determined from Tables in Reg. Guide 1.99, Draft Rev. 02 f-calculatedvalueofneu{gonfluenceat the reactor vessel / clad interface divided by 10 For Weld Material:
I-generic mean value of initial reference temperature - -56*F for welds made with Linde 1092 and 124 fluxes.
M-margin to cover uncertainties in initial RTNDT - 66'F since generic value of I was used.
For Plate Material:
I-initial reference temperature of irradiated material as defined in the ASME Code - -12*F for reactor vessel beltline plate material.
M'-
margin to cover uncertainties in initial RT M - 48' since a NDT measured value of I was used.
.The proposed PTS criteria applied to the vessel ID for longitudinal weld seams and plate material is RTPTS - 270*F and for circumferential weld seams is RTPTS - 300*F.
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