ML20080A723

Application to Amend Licenses DPR-53 & DPR-69,revising Tech Specs to Allow Operation of Hydrogen Sample & Hydrogen Sample Return Containment Isolation Valves During Testing Under Administrative Control

ML20080A723
Person / Time
Site:	Calvert Cliffs
Issue date:	01/27/1984
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	John Miller Office of Nuclear Reactor Regulation
Shared Package
ML20080A727	List: ML20080A723 ML20080A737
References
GL-83-43, NUDOCS 8402060234
Download: ML20080A723 (4)

Text

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r BALTIMORE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALTIMORE, MARYLAND 21203 ARTHUR E. LUNDVALL,JR.

vice e c. os."

January 27,1984 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 ATTENTION: Mr. James R. Miller, Chief Operating Reactors Branch #3 SUB3ECT: Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Amendment

REFERENCE:

(a) Letter from D. G. Eisenhut to All Licensees dated 12/19/83, Generic Letter 83-43 Gentlemen:

The Baltimore Gas & Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos. I and 2, respectively, with the submittal of the enclosed proposed change to the Technical Specifications.

CHANGE DESCRIPTION (BG&E FCR 84-015)

Remove the existing pages and replace with the attached marked up pages.

UNIT 1 UNIT 2 Page 1-2 Page 1-2 Page 3/4 3-43 Page 3/4 3-43 Page 3/4 4-10 Pcge 3/4 4-10 Page 3/4 4-12 Page 3/4 4-12 Page 3/4 4-19 Page 3/4 4-19 Page 3/4 6-9 Page 3/4 6-9 Page 3/4 7-66 Page 3/4 7-58 Page 3/4 7-69 Page 3/4 7-61 Page 3/4 7-72 Page 3/4 7-64 Page 3/4 7-73 Page 3/4 7-65 Page 3/4 7-75 Page 3/4 7-67 Page 3/4 7-77 Page 3/4 7-69 ol I tlO 8402060234 840127 PDR ADOCK 05000317 P PDR w .

s Mr. James R. Miller January 27,1984 Page 2 UNIT 1 UNIT 2 Page 6-7 Page 6-7 Page 6-10 Page 6-10 Page 6-12 Page 6-12 Page 6-14 Page 6-14 Page 6-16 Page 6-16 Page 6-17 Page 6-17 Page 6-18 Page 6-18 Page 6-19 Page 6-19 DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS These proposed changes are being submitted as requested by Reference (1). Because they are administrative changes made to conform to changes in 10 CFR 50, Paragraph 50.73, they do not result in any change in plant operating philosophy, design, or operating restrictions.

The Baltimore Gas & Electric Company will submit a Special Report gtime Reactor Coolant System Specific Activity exceeds 1.0 ACi/gm Dose Equivalent I for in excess of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> for one continuous interval or five (5) percent of the Unit's total yearly operating time pursuant to ACTION 3.4.8.a.

We believe our interpretation of this is consistent with the intent of the revision to 10 CFR 50.73, thereby eliminating unnecessary reports of Reactor Coolant System Iodine Activity " spikes" which are currently within the acceptable limits of the Technical Specification Figure 3.4-1. The proposed threshold requires a Special Report at 50% of the operating time which would lead to a Licensee Event Report under the new requirements. Since this proposed change is being made to conform to changes in the regulations, it has been determined to involve no significant hazards considerations, in that it will not:

l (1) Involve any increase in the probability or consequences of an accident l previously evaluated, or i

I (2) Create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) Involve any reduction in the margin of safety.

This determination is consistent with an Example of an Amendment Considered Not Likely to Involve Significant Hazards Considerations under the guidance provided by item (vil)of Federal Register Notice dated April 6,1983, page 14870.

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• r Mr. James R. Miller January 27,1984 Page 3 CHANGE NO. 2 (BG&E FCR 84-016)

Remove existing pages 3/4 6-23 and 6-24 of the Unit I and 2 Technical Specifications and replace with attached marked-up pages, (Attachment 2) to this transmittal.

DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change allows operation of the Hydrogen Sample and Hydrogen Sample Return containment isolation valves for brief periods during testing under administrative control. These penetrations are described in Section 5.2.2 and Table 5-3 of the Updated Final Safety Analysis Report as Type 11 isolation valve configuration which correlates to General Design Criterion 56 of 10 CFR 50.

Each penetration line is 1/4" in diameter as described in Table 5-3 of the Updated Final Safety Analysis Report. These valves must be operated to obtain or return a gaseous sample for the Post Accident Sampling System (PASS) or the containment hydrogen analyzers.

The proposed change will not increase the probability of occurrence or consequences of an accident previously evaluated in the Updated Final Safety Analysis Report. This is l because neither the operation of the system as proposed, nor any credible single failure  ;

in the system would initiate any accident previously evaluated in the Updated Final Safety Analysis Report. Similarly, the consequences of an event previously evaluated in the Updated Final Safety Analysis Report would not be more severe because operation of the system as proposed would not result in loss of containment isolation or intolerable leakage.

The proposed change does not create the possibility of a new or different kinci of l accident from those previously evaluated. It simply providas a temporary pathway under administrative control from the containment atmosphere to a closed sampling system and back to the containment atmosphere.

l The proposed change does not result in any reduction in the margin of safety, as described in Technical Specification Bases 3/4 6-4. This is because the valves will remain closed with the exception of brief periods when they must be open to obtain or return the gaseous sample to the containment atmosphere.

Based upon the above considerations, the proposed change has been determined to involve no significant hazards considerations.

SAFETY REVIEW COMMITTEE This proposed change to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.

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Mr. James R. Miller January 27,1984 Page 4 PEE DETERMINATION For fee determination purposes, we request you consider this change to be a supplement

• to our request dated December 22,1983.

Very truly ours,

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Attachments AEL/LES/gla cc: 3. A. Biddison, Esquire G. F. Trowbridg'.., Esquire D. H. Jaff e, NRC R. E. Architzel, NRC T. Magette, DNR

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