ML20210C195
| ML20210C195 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/09/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210C191 | List: |
| References | |
| NUDOCS 8609180269 | |
| Download: ML20210C195 (3) | |
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SAFETY EVALUATI N BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-395 Introduction In a letter from D. A. Nauman to H.R. Denton dated February 7,1986, the South Carolina Electric & Gas Company (the licensee) proposed changes to Surger Technical Specification Section 3/4.4.9, " Pressure / Temperature Limits-Reactor Coolant System" ar.d its bases.
The licensee requested changes to the pressure temperature limits described in Figures 3.4-2 and 3.4-3 and surveillance capsule withdrawal schedule described in Table 4.4-5.
The bases for these changes are the test results from the Summer surveillance program, which are contained in Report WCAP-10814, " Analysis of Capsule U from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program." WCAP-10814 was sucaitted for staff review in a letter from D.A. Nauman to H.R. Denton dated November 8, 1985.
Evaluation Pressure-temperature limits must be calculated in accordance with the requirerrents of Appendix G,10 CFR 50, which became effective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G, 10 CFR 50 are dependent upon the initial RT f r the limiting materials in the beltline and closure flange regions of Ne reactor vessel and the increase in RT resulting from neutron N
irradiation damage to the limiting Leltline b erial.
The Sumtrer reactor l-vessel was procured to ASME Code requirements, which specified fracture tought.ess testing to determine the initial RT f r each vessel material.
NDT I
Tne test results indicate that the initial RT for the limiting beltline and closure flange region materials are 30*F b 10*F, respectively.
A increase in RT resulting from neutron irradiation damage was estiroledbythelibnseeusingtheempiricalrelationshipdocumentedin Reguletory 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 predicated amourit of neutron fluence and the amounts of residual elements (copper and i
phosphorus) in the beltline material.
The neutron fluence used to predict neutron. irradiation damage is based on the calculated neutron flux at the
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vessel location with peak flux.
These neutron fluence predictions were l
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verified by rneasureunts f rom passive neutron flux monitors and by analysis, which was made with the DOT two-dimensional discrete ordinates code.
Inputs into the analysis included 47 neutron energy groups, P3 expansion of the scattering cross section, and power distributions represertative of time-averaged conditions derived from statistical studies of. long-term operation of Westinghouse 4-loop plants.
The cross sections used in the analysis were obtained from the SAILOR cross section library.
Using this method of analysis, the measured average fast (E > 1.0 MeV) neutronfluxderivedfromthefivetgesholdreactiondosimeters,whichwere contained in Capsule U, is 1.80 x 10 n/cm'-secwithagtandagddeviation of 27.5 percent.
The calculated flux value of 2.09 x 10 n/cm -sec exceeds all of the measured values, with calculation to experimental ratios ranging from 1.06 to 1.25.
Since the calculated flux is greater than the measured flux from the capsule dosemetry, neutron fluence calculations using the calculated flux should conservatively predict neutron fluence.
The predicted amcunts of neutron irradiation damage are based on design basis calculated neutron fluences and the increase in reference ThepredictionchT)usingthecurvesinRegulatoryGuide1.99,Rev.1.
temperature (ART ves in Regulatory Guide 1.99, Rev. I are dependent upon the amounts of residual elements in the belt'line material.
In Table B 3/4.4-1 of the Technical Specification, the licensee identified tne residual elements in the core region welds and plates.
Based on the chemical composition of the beltlir,e materials that were reported in this Table, the limiting beltline material would be Plate No. A9154-1.
Specimens from this plate were irradiated and tested as part of the Summer Surveillance program.
The increase in ART predicted for Plate-No. A9154-1 by Regulatory Guide 1.99, Rev.1 is 52 F.
gM incteases in SRT measured from longitudinal and transversely oriented specimens were 4 N and 30 F, respectively.
$1rce the predicted incresse in ART for the plate naterial is greater than the increase in ART measured frokDIhesurveillarcematerial,thepredictionmethodinRegulatchTGuice 1.99, Rev. 1, should conservatively predict the increase in ART f r the Summer NDT beltline plate material.
The NRC 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.
The NRC staff concludes that the proposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR 50 for 8 EFPY and therefore may be incorporated into the p.lant's Technical Specificatiens.
The reactor vessel material surveillance program must meet the requirements of A;4endix H,10 CFR 50, whir w ef fective on ?aly 26, 1983.
Appendix H requires that the surveillance program meet the requirements of ASTM E 185-82 to the extent practical.
ASTM E 185-82 requires that the Sur.mer surveillance program have a minimum of four capsules.
The time for capsule withdrawal recommended in the ASTM specification is dependent upon the effective full power years of operation, the capsule and vessel neutron fluences and the predicted increase in transition temperature of the encapsulated materials.
~
3 Tne Summer surveillance program contains six capsules; five are scheduled for removal and one is standby.
The Summer surveillance capsules are scheduled for withorawal at refueling outages that are either immediately before or after the ASTM recommended targets.
As capsules can cnly be scheduled for withdrawal during refueling outages, the capsule withdrawal schedule documented in Table 4.4-5 of the Summer Technical Specification meets, to the extent practical, the withdrawal schedule tabulated in ASTM E 185-82.
The proposed changes to the reactor vessel capsule withdrawal schedule meet the requirements of Apper: dix. H,10 CFR 50, are acceptable to the NRC staff, and therefore may be incorporated into the plant's technical specifications.
The deletion of the reference to Figure 3.4-4 is acceptable to the NRC staff, because there is no Figure 3.4-4 in Technical Specifications.
Environmental Consideration This amendment involves a change in the use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes in an inspection or surveillance requirement.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant 1
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 issued a proposed ~ finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sectior.
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.
Conclusion 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, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public.
Dated: September 9, 1986 Principal Contributors:
Jon B. HoM 4 nt, Project Ofre:terate #2, DPLA Barry J. Elliot, Engineering Branch, OPLA