B13357, Application for Amend to License NPF-49,changing Tech Specs Re Reactor Vessel Pressure/Temp Limits

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Application for Amend to License NPF-49,changing Tech Specs Re Reactor Vessel Pressure/Temp Limits
ML19325E087
Person / Time
Site: Millstone 
Issue date: 10/20/1989
From: Mroczka E, Romberg W
NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19325E088 List:
References
B13357, NUDOCS 8911010171
Download: ML19325E087 (5)
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E8IE P.O. Box 270 t;

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HARTFORD, CONNECTICUT 061410270 k

L J 7,((["d%C (203) 665-6000 October 20, 1989 I

Docket No. 50 421 B13357 i

i U.S. Nuclear Regulatory Commission i

Attention: Document Control Desk Washington, DC.20555 i

s Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications l

Reactor Vessel Pressure /Temocrature limits j

Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby t

proposes to amend Operating License NPF 49 by incorporating the changes identified in Attachment I into the Technical Specifications of Millstone Unit i

No. 3, i

P Table 4.4-5, Reactor Vessel Material Surveillance Program Withdrawal Schedule, j

lists lead factors and withdrawal times corresponding to reactor vessel surveillance capsule locations.

Since the removal and testing of Capsule U i

during the first refueling, the proposed changes update the information j

associated with each remaining capsule, j

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The Reactor Coolant System pressure / temperature limit curves for plant heatup, i

cooldown, and inservice leak and hydrostatic pressure testing operations are

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provided in the Technical Specifications.

These curves define limits to l

ensure the prevention of nonductile failures of material incorporated within

'l the reactor coolant system (RCS).

The allowable pressure / temperature for l

1 specified heatup and cooldown rates are calculated in accordance with Appendix l

G of Section III of the ASME Boiler and Pressure Vessel Code and 10CFR50, I

Appendix G.

The heatup and cooldown limit curves are calculated using the most limiting to the reactor Msse(l material. reference nil-ductility transition temperature) inherent i

value of the RT T

from material tests made at the time of the vessel fabricaNcIn.is determine The initial value of the RT During the service life of the reactor vessel, the RT increases above the initial i

L value because of neutron irradiation.

The a$Iant of change (ARTThhDI)ransition i

depends upon the neutron fluence and material chemical composition.

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temperature shift is determined from fluence measurements, calculations, and trend curves based on tests of irradiated specimens that predict the effects l

of neutron irradiation.

The irradiation specimetis are actual (or archive) 1 j

reactor vessel material specimens and are positioned around the reactor vessel 1

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U.S. Nuclear Regulatory Commission B13357/Page 2

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October 20,.1989 i

to provide' surveillance of the irradiation levels to which the reactor vessel

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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.

o Associated with each surveillance capsule location is a lead factor - the ratio of the instantaneous neutron flux density at the location of the speci-mens in a surveillance capsule to the maximum calculated neutron flux density t

at the inside surface of the reactor vessel wall.

The lead factor is thus t

used to extrapolate the surveillance measurements from the specimens to the reactor vessel wall, thereby the material property changes of the reactor vessel are monitored throughout its life.

Rescription of Chanae The capsule lead factors and withdrawal times are being updated based on the

.results of the capsule removed during the first refueling and are consistent with ASTM E185-82.

The requirement to remove only three capsules results from the revised shift in E0L RT which is now less than 100*F.

This is based on NDT ASTM E185-82.

The change to the bases of Technical Specification Section 3/4.4.9 reflects a revision in Regulatory Guide (RG) 1.99.

Based on issuance of RG 1.99, Revi-l sion 2, the Westinghouse Copper Trend Curves (Figure B 3/4.4-2) are no longer applicable and are therefore being removed from the Technical Specifications.

l-The proposed changes of lead factors will result in a more precise assessment of the reactor vessel materials' irradiation level s.

The calculational l

uncertainties in extrapolating the surveillance measurements are lessened.

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These changes are consistent with those(gcommended by Westinghouse Electric following analysis of capsule specimens.

Sianificant Hazards Consideration NNEC0 has reviewed the proposed changes in accordance with 10CFR50.92 and has concluded that the changes do not involve a significant hazards consideration.

The basis for this conclusion is that the three criteria of 10CFR50.92(c) are L

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(1) WCAP-11878, Analysis of Capsule U from the Northeast Utilities Service L

Company Millstone Unit 3 Reactor Vessel Radiation Surveillance Program, June 1988.

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s U.S.' Nuclear Regulatory Commission B13357/Page 3 October 20, 1983 f

not compromised.

The proposed changes do not involve a significant hazards consideration because they would not:

1.

Involve a significant. increase in the probability or consequences of an accident previously analyzed.

The changes provide. assurance that the reactor vessel integrity is maintained as assumed in the design basis analysis.

Changes to the sam)1e lead factors and withdrawal times are based on analysis of the vessel sample removed during the first refueling and are consistent with RG 1.99.

The changes do not impact plant opera-tions or the performance of any system.

The changes do not introduce any failure modes and therefore do not increase the probability or conse-quences of any event.

2.

Create the possibility of a new or different kind of accident from that i

previously analyzed.

The proposed changes do not have the potential to initiate any event nor do the changes alter plant operation. The changes do not introduce any single failures.

Thus, the changes do not create a 7

new or different kind of accident, 3.

Involve a significant reduction in the margin of safety.

The changes do not impact any of the protective boundaries.

The changes do not nega-t tively impact any of the safety systems, nor do they increase the number of challenges to the safety systems.

For these reasons the changes do not involve a reduction in the margin of safety.

Moreover, the Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6,

1986, 51FR7751) of amendments that are considered not likely to involve a signifi-cant hazards consideration.

Although the proposed changes to the reactor

- vessel lead factors and withdrawal times are not enveloped by a specific example, the changes would not involve a significant increase in the proba-bility or consequences of an accident previously analyzed.

As previously stated, the updated lead factors and withdrawal times represent a more precisa assessment of the reactor vessel materials' irradiation levels and, therefore, results in increased plant safety.

Based upon the information contained in this submittal and the environmental assessment for Millstone Unit No. 3, there are no radiological or nonradio-logical impacts associated with the proposed changes and the proposed license amendment will not have a significant effect on the quality of the human l

environment.

The Millstone Unit No. 3 Nuclear Review Board has reviewed and approved the attached proposed revisions and has concurred with the above determinations.

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r U.S. Nuclear Regulatory Commission B13357/Page 4 October 20, 1989 In accordance with 10CFR50.91(b) we are providing the State of Connecticut' with a' copy of this proposed amendment.

Very truly yours, i

NORTHEAST NUCLEAR ENERGY COMPANY

$ 5 O)(*c.t Nw s

E. J. Mroczka Senior Vice President d

By:

W. D. Romberg V Vice President cc:

W. T. Russell,: Region I Administrator D. H.'Jaffe, NRC Project Manager, Millstone Unit No. 3 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 Mr. Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, Connecticut 06116 STATE OF CONNECTICUT ss. Berlin C0VNTY'0F HARTFORD )

Then personally appeared before me, W. D. Romberg, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, a

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Licensee herein, that he is authorized to execute and file the foregoing information in' the name and on behalf of the Licensees herein, and that the statements contained in said information are true and correct to the best of his knowledge and belief.

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.J NotaryPulpfc MY COMMISSION EXPIRES MARCH 31, 1991'

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