ML20115F748

Addendum to Mnps,Units 1,2 & 3 Annual Rept Millstone Unit 3 Section

ML20115F748
Person / Time
Site:	Millstone
Issue date:	12/31/1995
From:	NORTHEAST UTILITIES SERVICE CO.
To:
Shared Package
ML20115F744	List: B15771, Forwards Rev 9 to Updated FSAR in Form of Replacement Pages. Rev Accurately Presents Changes Completed Since Previous Submittal,Per 10CFR50.71(e)(2)(i).Addendum to Annual Rept Dtd 960228,also Encl ML20115F748
References
NUDOCS 9607180087
Download: ML20115F748 (4)

Text

3 Docket No. 50-423 B15771 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Addendum to Millstone Nuclear Power Station, Unit Nos.1,2, and 3 Annual; Report dated February 28,1996 Millstone Unit No. 3 Section June 1996 fIk"$00!K K

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l U. S. Nucirr Regul::. tory Commission

! B15771\ Attachment 1\Page 1 Addendum to Millstone Nucbar Power Station, Unit Nos.1,2, and 3 Annual; Report dated February 28,1996 Millstone Unit No. 3 Section FSARCR Number Title 95-MP3-55 Control Building Battery Rooms No. 3 and No. 4 Minimum Temperature Change Description of Chance The change documents the reduction of the minimum ambient temperature for battsry rooms 3 and 4 in the control building.

Reason for Chance Historical temperature measurements have noted that the room temperatures have dropped below the present design condition. This design condition was revised in DCN DM3-S-1093-95 and is reflected in the battery calculations and in this FSAR change.

Safety Evaluation Summary The minimum temperature of the batteries has been evaluated in calculation NE-188E Revision 2, CCN #5 and the batteries will have remaining spare capacity if the battery electrolyte temperatures drop to the minimum recorded room temperature. Actual battery electrolyte temperatures were monitored during the room temperature excursions and they did not drop below the analyzed temperature of 60*F. This provides added assurance of margin between the room temperatures and possible reduction of battery performance.

The changes proposed by FSARCR 95-MP3-55 will not contribute to any previously analyzed accident, or its consequences, nor will it contribut e to any accident outside of those already analyzed.

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U.S. Nuclear Regulitory Commission l B15771\Page 2 Addendum to Millstone Nuclear Power Station, Unit Nos.1,2, and 3 l

Annual; Report dated February 26,1996 Millstone Unit No. 3 Section l FSARCR Number Title 95-MP3-2 Update of Local Site Flooding Analysis Description of Chance i

j This change involves updating of the Local Site Flooding Analysis and the FSAR to l

reflect updated extreme precipitation flood design considerations for new buildings

added to the site. Each building had been evaluated and engineered to site  ;

requirements to consider the impact on intense precipitation analysis.

1 Reason for Chance l

The Local Site Flooding Analysis and the FSAR description of the effect of local l

intense precipitation had been outdated by changes made to the site.

Safety Evaluation Summarv l

The Safety Evaluation concluded that the revised flooding analysis which addresses l the updated site configuration is considered safe and does not increase the probability l of a malfunction of equipment important to safety. The revised flooding analysis resulted in changes to water levels at some of the safety-related buildings. These flood l

! levels are short-term events which would result in minor additional leakage to safety-related buildings. The inleakage has been quantified and evaluated to determine there are not effects on safety-related equipment.

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U.S. NuclTr Regulatory Commission B15771\Page 3 i Addendum to Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 l

Annual; Report dated February 26,1996 l Millstone Unit No. 3 Section Procedure Number Title f i

FP-NEU-2, Rev. O Fuel Reconstitution .

I Description of Channe i

This procedure provides guidance and direction for the disassembly and inspection, repair, and reassembly of a top reconstitutable fuel assembly.

Reason for Channe l

Reactor coolant activity from the fifth operating cycle shows indication that at least one l fuel pin has a cladding defect. This procedure provides for disassembly, inspection of individual fuel rods, replacement of failed fuel rods, and reassembly for future use in  ;

the reactor.

l Safety Evaluation Summarv l Fuel assembly, disassembly, fuel rod removal, fuel rod replacement and assembly were i

performed with fuel assembly in the new fuel elevator. The physical arrangement of the elevator, coupled with tool design and administrative controls, ensured that an acceptable level of water was maintained over the fuel assembly and/or removed rod.

Radiation levels were maintained consistent with ALARA practices. Therefore, there was no adverse affect on radiation levels in and around the spent fuel pool.

The assembly was transported to and from the elevator using the normal Spent Fuel Handling tool, therefore, this evolution did not increase the probability of a design basis fuel handling accident, nor create the probability of a new type of accident. Handling l

l on an individual fuel rod does not create the possibility of a new type of accident due to its very small mass and radionuclide inventory compared to an entire fuel assembly.

Since only one fuel assembly was out of its normal storage position at any given time during this process, the consequences of a design basis fuel handling accident had it occurred would not have been affected.

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