ML20012C138
| ML20012C138 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/06/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20012C126 | List: |
| References | |
| NUDOCS 9003200169 | |
| Download: ML20012C138 (4) | |
Text
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9, UNITED STATES
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NUCLEAR REGULATORY COMMISSION n
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' WASHINGTON, D. C. 20666 SAFETY EVALUATION.BY.THE OFFICE,0F NUCLEAR. REACTOR. REGULATION RELATED.TO AMENDMENT NO. 48..
TO FACILITY.0PERATING. LICENSE.NO..NpF.49 q
NORTHEAST. NUCLEAR ENERGY.COHpANY..ET.AL.
MILLSTONE. NUCLEAR p0WER STATION. UNIT NO. 3 DOCKET.NO. 50-423 1
INTRODUCTION By application for license amendment dated October 20, 1989, Northeast Nuclear Energy Company, et al. (the licensee), requested changes to Millstone Unit 3 TechnicalSpecifications(TS).
l The proposed amendment would change Millstone Unit 3 Technical Specifications (TS) Table 4.4-5, " Reactor Vessel Material Survaillance Program - Withdrawal i
Schedule" to provide a revised in-vessel material capsule withdrawal program and revised capsule lead factors.
DISCUSSION Af!D EVALUATION The Reactor Coolant System pressure / temperature limit curves for plant heatup, cooldown, and inservice leak and hydrostatic pressure testing operations are provided in the Technical Specifications. These curves-define limits to ensure the prevention of nonductile failures of materials. incorporated within the reactor coolant system (RCS). The allowable pressure / temperature for specified heatup and cooldown rates are calculated in accordance with Appendix G of Section III of the ASME Boiler and Pressure Vessel Code and 10 CFR;50, Appendix G.
The heatup and cooldown limit curves are calculated using the most limiting value of the RT (reference nil-ductility transition temperature) inherentinthereactorvesse$DItaterial. The initial value of RT is determinedfrommaterialtestsmadeatthetimeofthevessel~fabNation.
During the service life of the reactor vessel, the RT increases above the initial value because of neutron irradiation. The ambt of change (delta RT
) depends upon the neutron fluence and material chemical composition. The
-trMitiontemperatureshiftisoeterminedfromfluencemeasurements, calculations, and-trend curves based on tests of irradiated specimens that predict the effects of neutron irradiation. The irradiated specimens are actual-(or archive) reactor vessel material specimens and are positioned around the reactor vessel to provide surveillance of the irradiation levels to which the reactor vessel is subject. The specimens are maintained in an inert environment within a corrosion-resistant capsule to prevent deterioration of the surface of the specimens during radiation exposure.
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Associated with each surveillance capsule location is a lead factor, the ratio of the instantaneous neutron flux density at the location of the specimens in a surveillance capsule to the maximum calculated neutron flux density at the inside surface of the reactor vessel wall. The lead factor is thus used to extrapolate the surveillance measurement from the specimens to the reactor vessel wall, thereby the material property changes of the reactor vessel are monitored through its life. The in-vessel capsule irradiation program is described in Section 5.3.1.6 of the Millstone Unit 3 Final-Safety Analysis i
l Report. Each surveillance capsule is also subject to a withdrawal schedule, per TS 4.4.9.1.2, as specified in TS Table 4.4-5.
The specimens within the withdrawn capsule are subjected to various inspections and tests to determine the delta RTN and-any needed changes in the heat-up and cooldown limit TheNmberofcapsulestobewithdrawnoverthelifeofthereactor l
curves.
pressure vessel is required by Appendix H to 10 CFR Part 50-to meet the requirements of ASTM E185.-
The licensee has proposed a change to the number of surveillance capsules to be withdrawn (and the associated withdrawal schedule) and the lead factors as l
specified in TS Table 4.4-5.
The proposed changes result from analysis of the first capsule which was withdrawn during the first refueling outage. At the
.present time TS Table 4.4-5 describes a capsule program containing four capsules. The first capsule was withdrawn during the first refueling outage and subsequent capsules are-to be withdrawn at 5, 9 and 15 effective full power years (EFPY). The-requirements of ASTM E185 allow a program to contain only.
L threecapsulesiftheend-of-life (EOL)RT is less than 100'F. Based upon theevaluationofthefirstcapsuletobeYSkoved,thelicenseehasprojected that the E0L RT will be less than 100'F and has proposed a change to TS-Table 4.4-5.
Th0Trevised capsule program would have three capsules. The first capsule would be withdrawn during the first refueling outage (already accomplished) and subsequent capsules at 9 and 16 EFPY. Changes to the TS Bases have also been proposed.
We concur with the licensee's evaluation that the E0L RT projection of less than 100'F permits the irradiation capsule program to be Nduced from four to N
three capsules. The proposed irradiation capsule program conforms to ASTM E185 and, in this regard, meets the requirements of Appendix H to 10 CFR Part 50.
Accordingly, the proposed changes to TS Table 4.4-5 are acceptable.
ENVIRONMENTAL. CONSIDERATION This amendment changes a requirement with respect to installation or use l
of a facility component located within the restricted area as defined in i.
10 CFR Part 20. We have 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 staff has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no I
public connent 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 51.22(b), no environmental impact statement or environmental l
assessment need be prepared in connection with the issuance of the amendment.
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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 (3) the~ issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: March 6, 1990 Principal Contributor: D. Jaffe l
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. AMENDMENT.NO.48 TO FACILITY OPERATING LICENSE NO. NPF-49 DISTRIBUT10N
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