ML20214N887
| ML20214N887 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/08/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214N880 | List: |
| References | |
| NUDOCS 8609170148 | |
| Download: ML20214N887 (4) | |
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UNITED STATES 1
NUCLEAR REGULATORY COMMISSION h
WASHINGTON. D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.100 TO FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUB'LIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285
1.0 INTRODUCTION
By letter dated April 25, 1986, the Omaha Public Power District (the licensee) submitted a proposed amendment to Section 2 of the Fort Calhoun Station (FCS) Technical Specifications to update the reactor coolant system pressure-temperature limits for heatup and cooldown.
During its review, the staff identified technical concerns with the submittal and related these concerns to the licensee during telephone conversations held on June 16 and 19, 1986.
The licensee responded to the' issues raised by the staff by a revision to their submittal which was transmitted by letter dated July 10, 1986.
The revised submittal provides new heatup and cool-down limit curves for operation through 15 effective full power years (EFPY). The last surveillance capsule report submitted to the staff by
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the licensee was Omaha Public Power District Report, " Fort Calhoun Unit No. 1, Evaluation of Irradiated Capsule W-265," dated March 1984.
- 2. 0 DISCUSSION Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G, 10 CFR Part 50, which became effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirerrents of Appendix G are dependent upon the initial RT for the limiting materials in the beltline and closure flangereg1S[softhereactorvesselandtheincreaseofRT resulting g
NDT from neutron irradiation damage to the beltline materials The FCS reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness testing to determine the RT f r each of the materials needed to make the reactor vessel. Hence,tNSTinitial RT for l
materialsintheclosureflangeandbeltlineregionoftheFCSreac50Ivessel could not be determined in accordance with the test requirements M the ASME Code. Therefore, the initial RT for these materials is estimated from test data from other similar mathIals used for fabrication of reactor vessels in the nuclear industry. The licensee has estimated the RT I"
thesematerialsinaccordancewithBranchTechnicalPositionMTEBS-kT" Fracture 8609170148 860908 PDR ADOCK 05000285 P
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. Toughness Requirements," which are contained in NUREG-0800, "USNRC Standard Review Plan 5.3.2, Pressure-Temperature Limits." This branch technical position provides conservative estimates of RT f r reactor i
ND vessel materials and uses a RT f60*Ffortheclosureflangeforgings.
NDT The limiting materials in the FCS reactor vessel beltline are weld metals.
The initial RT for these weld materials was estimated by the licensee as-56*Fwith$0[tandarddeviationof17*F. These initial RT and un standard deviation values were recommended by the staff in Coinm ssion Report SECY 82-465, " Pressurized Thermal Shock" for welds fabricated by Combustion Engineering using Linde 1092 flux. This recommendation applies to the FCS reactor vessel welds.
The increase in~RT resulting from neutron irradiation damage was estimatedbytheIbnseeusingthemethoddocumentedinDraftRegulatory Guide 1.99, Revision 2, " Radiation Damage to Reactor Vessel Materials."
Although this Regulatory Guide is only a draft, its methodology is con-l sidered by the staff to be the most up-to-date method for predicting i
neutron irradiation damage. This method of predicting neutron irradiation I
damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and nickel in the beltline material.
When determining the Limiting Conditions for Operation in the FCS Technical Specifications, the impact of the initial RT the fluence induced RT shiftofreactorvesselbeltlineweldsandaNDIp,propriatemarginmustbUDT 4
considered. The fluence induced temperature shift is a function of fluence I
and the chemical composition of the limiting reactor vessel beltline material.
In the past, the absence if specific weld chemical composition data required assumption of upper bound values for beltline weld copper and nickel content.
The chemical composition of all FCS reactor vessel beltline welds was recently documented (the licensee's submittal on pressurized thermal shock l
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dated January 23, 1986) through searches of Combustion Engineering welding records and through analysis of physical weld samples removed from identical welds traced to the reactor vessel head.
The licensee's proposed pressure-temperature limits at 15 EFPY have been calculated using a neutron fluence of 1.4 x 1018 n/cm2 (E>1MeV). The value of 2.9 x 101s n/cm2 (E>1MeV) is the end of life neutron fluence for the FCS reactor pressure vessel. The projected fluence at the critical reactor j
vessel beltline weld was determined using the following equation:
$I.D. = [8.8 x 101s + (EFPY - 5.92) (4.8 x 1018)/32] (.60) n/cm
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2 This equation was developed using results from the analysis of surveillance capsule W-265 which was removed after 5.92 EFPY, the end of Cycle 7.
The predicted reactor vessel I.D. fluence was reported to be 8.8 x 101s n/cm2 i
by Combustion Engineering in the W-265 analysis report and was projected to reach 4.8 x 1018 n/cm2 by the end-of-life. However, following Cycle 7, reduced fluence core loading patterns were initiated. To take credit for this reduction, an azimuthal flux distribution plot for Cycles 1-9 average, 1
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. generated using the transport code DOT 4.3, was used to determine the azimuthal flux reduction value at the critical weld location (.60) for use in the fluence equation above.
3.0 EVALUATION The lower shell longitudinal weld seam, 3-410, is the most limiting material. The amount of copper and nickel in the longitudinal weld is 0.23% by weight and 0.95% by weight, respectively.
The reactor vessel inside radius is 70.125 inches, and the outside radius is 77.25 inches which yields reactor vessel wall thickness of 7.125 inches.
Flaws are postulated on the inside surface and the outside surface of the vessel or weld.
Distance is measured from the inner radius of the vessel outward.
The flaws on the inside surface and outside surface are referred to by location as 1/4 thickness and 3/4 thickness, respectively.
Recently, Draf t Regulatory Guide 1.99, Revision 2, was revised to incorporate comments from the public.
No substantive changes in the Regulatory Guide occurred.
However, the formula for attenuation of neutron fluence through the vessel wall was revised. Based on Regulatory Guide 1.99, Revision 2, the shift in RT at 1/4 thickness and 3/4-thicknessforthelongitudinalweldcomhbfedbythe'staffandthelicensee differ by a nominal 3% or less.
This difference in ART is due to the methodusedtoattenuateneutronfluencethroughtheveskkiwall. The N
staff finds the adjusted RT computed by the licensee acceptable.
NDT The criteria from Section III, ASME Code, Article G-2000, Vessels, was used to determine the measured temperature during an inservice hydrostatic test (K 2 1.5K KI ).
At the flaw on the inside surface of the longitukknal welb) +the [emperature computed by the staff and the licensee are essentially the same for the hydrostatic test pressure of 2310 psig.
The curve for pressure-temoerature limits for inservice hydrostatic testing is acceptable to the staff.
The criteria from Section III, ASME Code, Article G-2000, Vessels, was also used to determine the measured temperature during various heatup and cooldown rates (K eithertheflawibRthelodhitudkba)l.weldontheinsidesurface(1/4 2 2K
+K of the reactor vessel.
For heatup, thickness) or the outside surface (3/4 thickness) is limiting. For cooldown, the flaw in the longitudinal weld on the inside surface (1/4 thickness) is limiting. The heatup and cooldown curves (20*F/hr and 100*F/hr) computed by the licensee either conservatively bound the values l
calculated by the staff or differ by 3% or less.
The differences in j
computed temperature values between the licensee and the staff are due to (1) the expressions used to attenuate the neutron fluence, (2) the methodology used to determine the temperature gradient through the vessel wall, and (3) whether K is disregarded when it would be conservative to y
do so.
Thus, the staff Iinds the pressure-temperature limits for the heatup and cooldown curves presented by the licensee to be' acceptable.
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. The staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan Section 5.3.2, NUREG-0800, Rev. 1, dated July 1981, to evaluate the proposed pressure-temperature limits. The amount of neutron irradiation damage to the beltline materials was estimated using the method documented in Draft Regulatory Guide 1.99, Revision 2.
The staff concludes that the licensee's proposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR Part 50, for 15 EFPY and may be incorporated into Section 2 of the FCS Technical Specifications.
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4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that the amendment involves no significa~nt hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR $51.22(c)(9).
Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
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5.0 CONCLUSION
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We have concluded, based on the considerations discussed above, that
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(1) there is reasonable assurance that the health and safety of the j
public will not be endangered by operation in the proposed manner, and (2) such activities will~be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: September 8, 1986 Principal Contributor:
K. Wickman
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