ML20206F105
| ML20206F105 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/10/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206F103 | List: |
| References | |
| NUDOCS 8606240225 | |
| Download: ML20206F105 (6) | |
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UNITED STATES E
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REG RELATED TO AMENDMENT NO. 75 TO FACILITY OPERATING LICEN PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY, AND ATLANTIC CITY ELECTRIC COMPANY SALEM NUCLEAR GENERATION STATION, UNIT NO. 1 DOCKET NO. 50-272 INTRODUCTION In a letter from E. A. Liden to S. A. Varga dated October 15, 1984, Public J
Service Electric & Gas Company (the licensee) proposed an amendment to their Facility Operating License DPR-70 for Salem Generating Station, Unit I (Salem-1). The amendment proposed to revise the reactor coolant system pressure / temperature limits, which are contained in Section 3.4.9.1 of the Technical Specifications and was based on the analysis of Capsule T data.
During the staff review of this information, the results of the analysis of Capsule Y data were made available for use as input for the staff evaluation. These data were provided in letters dated October 16, 1985 and January 30, 1986 from C. A. McNeill to S. A. Varga. The revised curves are The bases for to be applicable for 10 effective full power years (EFPY).
these changes were the test results from the Salem-1 surveillance program, which are contained in Report WCAP-10694, " Analysis of Capsule Y From The Public Service Electric and Gas Company Salem Unit 1 Reactor Vessel 8606240225 860610 hDR ADOCK 05000272 PDR
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Radiation Surveillance Program" and Report WCAP-9678, " Analysis of Capsule T t
From The Public Service Electric and Gas Company Salem Unit 1 Reactor Vessel Radiation Surveillance Program". The information provided in the October 16, 1985 and the January 30, 1986 cubmittals served to enhance the i
accuracy of the revised pressure / temperature limits.
EVALUATION AND
SUMMARY
Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G,10 CFR 50, which became effective on July 16, 1983. Pressure-temperature limits that are calculated in accordance with the
'requirements of Appendix G,10 CFR 50 are dependent upon the initial RTNDT I#
the limiting materials in the beltline, and closure flange regions of the reactor vessel and the increase in RT resulting from neutron irradiation NDT damage to the limiting beltline material. The Salem-1 reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness f r each vessel material.
The licensee testing to determine the initial RTNDT indicates that the initial RT f r the limiting materials in the closure NDT flange and beltline regions of the Salem vessel were estimated using the method recommended by the staff in Branch Technical Position MTEB 5-2,
" Fracture Toughness Requirements." This method result in an initial RTNDT for the limiting beltline base metal and weld metal of 45'F and O'F, for the limiting closure flange material of respectively and an initial RTNDT 50*F.
The increase in RT resulting fr m neutron irradiation damage was estimated NDT by the licensee using the empirical relationship documented in Regulatory Guide 1.99, Rev.1, April 1977, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of residual elements (copper and phosphorus) in the beltline material. The neutron fluence predictions were verified by e
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j measurements from passive neutron flux monitors and by analysis, which was Inputs into the made with the DDT two-dimensional discrete ordinates code.
analysis included 47 neutron energy groups, P3 expansion of the scattering cross section, and power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of The cross sections used in the analysis were Westinghouse 4-loop plants.
Using this method of obtained from the SAILOR cross section library.
analysis, the measured saturated activity and neutron fluences (E>1MeV) fo five foil reactions, which were calculated from neutron dosimetry in Capsules T and Y, were less than that predicted from the design basis calculated The authors of WCAP 10694 recommended that projections of neutron fluences.
vessel toughness into the future be based on the design calculated fluence levels, since the calculated fluence levels were based on conservative representations of core power distributions derived for long-term operation The staff while the Capsule data are representative only of past operation.
agrees.with this recommendation.
' The predicted amounds of neutron irradiation damage are based on d calculated neutron fluences and the increase in reference temperatur The prediction curves in using the curves in Regulatory Guide 1.99, Rev 1.
Regulatory Guide 1.99, Rev. 1 are dependent upon the amounts of re elements in the beltline material. The licensee in a Report entitled,
" Fracture Toughness Analysis For Salem Unit 1 and 2 Reactor Pressure to Protect Against Pressurized Thermal Shock Events 10 CFR 50.61" has identified the residual elements in each weld and plate in the Salem-1 This report was contained in a letter from C. A. McNeill, Jr., to beltline.
Based on the chemical composition of the 20, 1986.
S. A. Varga, dated January beltline materials that were reported in this report, the limiting beltline In Table 1 we have compared the increase material would be Plate No. B2402-1.
predicted fr m the surveillance material to the increase in ART NDT in ARTNDT This comparison indicates that the by Regulatory Guide 1.99, Rev.1.
f r the plate material is greater than the predicted increase in ARTNDT Thus, the measured fro'm the surveillance material.
increase in ARTNDT prediction method in Regulatory Guide 1.99, Rev.1, should conser f r the Salem-1 beltline plate material.
predict the increase in ARTHDT e
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The staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev. 1. July 1981 to evaluate the proposed pressure-temperature limits. The amount of neutron irradiation damage was calculated using design basis calculated neutron fluences and the Regulatory Guide 1.99, Rev.1, prediction curves. Our conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G,10 CFR 50 for 10 EFPY and may be incorporated into the plant's technical specifications.
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 Comii;sion has previously issued a proposed finding that' ihis amendment involves no
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significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sec 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLU ION We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner,
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5-and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Dated: June 10, 1986 PRINCIPAL CONTRIBtfTOR:
B. Elliot e
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e Table 1:
Increase in Reference Temperature for Capsules T and Y Surveillance Material Increase in Reference Temperature (*F)
Surveillance Neutron Fluence From Surveillance Predicted by Reg.
2 Capsule Guide 1.99, Rev. 1 i
Material (E>1MeV)x10880/cm Base Metal
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Plate B2402-1 2.40 100 100 Plate B2402-2 2.40 100 113 Plate B2402-3 2.40 75 98 l
Plate B2402-3 8.91 110 189 Weld Metal Weld 9-042 8.91 165 156 Correlation Monitor HSST Plate 02 2.40 60 59 HSST Plate 02 8.91 125 113 1
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