ML20067B694

Application for Amend to License DPR-35,proposing Changes to TS by Revising Wording for Page 3 of License DPR-35,adding Clarifying Words to Aid Operators & Removing Obsolete Mechanical Snubber Acceptance Criterion

ML20067B694
Person / Time
Site:	Pilgrim
Issue date:	02/09/1994
From:	Boulette E BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20067B697	List: ML20067B694 ML20067B711
References
BECO-94-17, NUDOCS 9402250059
Download: ML20067B694 (7)

Text

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. BOSTON EDISON ,

Pilgnm Nuclear Power Station 10CFR50.90 Rocky Hill Road Plymouth. Massachusetts 02360 >

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E. T. Boulette, PhD l semor Vice President-Nuclear i February 9, 1994 BECo 94- 17  ;

U. S. Nuclear Regulatory Commission Document Control Desk  ;

Washington, DC 20555 License DPR-35 Docket 50-293 i

Proposed Chanaes to the Pilarim Nuclear Power Station i Technical Snecifications and License DPR-35 Boston Edison Company proposes three changes to the Pilgrim Nuclear Power Station Technical Specifications and License DPR-35. These changes consist -

of, (1) revised wording for page 3 of License DPR-35, (2) addition of ,

clarifying words to aid operators in selecting the correct pressure / temperature curve during startup and shutdown operations, and (3) ,

removal of an obsolete mechanical snubber acceptance criterion. j The proposed changes are described in Attachment A, the revised pages are in Attachment 8, and the marked-up existing technical specification pages are in Attachment C. .

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E. T. Boulette, PhD ETB/GGW/dmc/3 ADMIN ,

Commonwealth of Massachusetts)  !

County of Plymouth )

I Then personally appeared before me, E. T. Boulette, who being duly sworn, did state that he is Senior Vice President - Nuclear of Boston Edison Company and .

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that he is duly authorized.to execute and file the submittal contained herein l in the name and on behalf of Boston Edison Company and that the statements in i said submittal are true to the best of his knowledge and belief. ,

My commission expires: d@kt.f /995" c (C ? ~ gI/c u DATE ' ~~ NOT RY PUBLIC 9402250059 940209 PDR ADOCK 05000293 p PDR 4 j

. a BOSTON EDISON COMPANY U. S. Nuclear' Regulatory Commission -l

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J Attachments: (A) Description of Changes  :

(B) Revised Technical Specification Pages 't (C) Marked-up Existing Technical Specification Pages l

cc: Mr. R. Eaton, Project Manager  !

Division of Reactor Projects - I/II l Office of Nuclear Reactor Regulation i Mail Stop: 1401  :

U. S. Nuclear Regulatory Commission '

1 White Flint North 11555 Rockville Pike Rockville, MD 20852  ;

V. S. Nuclear Regulatory Commission Region I 47S Allendale Road King of Prussia, PA 19406 j Senior NRC Resident inspector Pilgrim Nuclear Power Station Mr. Robert M. Hallisey, Director  !

Radiation Control Program Mass. Dept. of Public Health j 305 South Street  !

Jamaica Plain, MA 02130 i i

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4 4 9 ATTACHMENT A Proposed Administrative Changes to Pilgrim Nuclear Power Station '

Technical Specifications Description of Changes:

Boston Edison Company (BECo) proposes the following three administrative changes to the Pilgrim Nuclear Power Station (PNPS) Technical Specifications  !

(TS):

1) Revise wording on page three to License DPR-35 to eliminate the need for the license to reference the latest approved amendment.

2) Add a clarifying phrase to TS section 3.6. A.2 as an aid to '

operators when selecting the correct pressure / temperature curve  :

during startup and shutdown operations.

3) Remove an obsolete surveillance requirement for increased mechanical snubber functional testing after a 50% increase in drag force since the last functional test.  ;

The first change will reword Section 3.B of License DPR-35 to eliminate .

reference to the last amendment number issued for the license.

For every TS Amendment issued, page 3 of the Operating License must be changed to signify the latest amendment. This practice is unnecessary because the

• License does not identify the latest TS page in effect. We are proposing to modify the paragraph to eliminate the need to issue this page with every amendment. This will reduce an administrative burden and preclude a possible +

administrative error if the correct number is somehow missed.

The second change is to T.S. Section 3.6 Primary System Boundary, paragranh 3.6.A.2. This change will correct an omission made in an earlier amendmeu by  !

adding the words "subtritical or" to the list of operational phases referenced  ;

in the paragraph. This will enhance the reference to the appropriate .j pressure / temperature curves presented in Figures 3.6.1, 3.6.2, and 3.6.3. j The first Figure, 3.6.1, provides a set of pressure / temperature curves used i during hydrostatic and/or leakage testing. The second Figure, 3.6.2, is used during startup and shutdown operations with the reactor subcritical. The third Figure, 3.6.3, is used when the reactor is critical. Adding a reference l to subcritical operations within paragraph 3.6.A.2 will aid operators when I they are in startup or shutdown mode of operation.

This TS change is being proposed at the request of Operations personnel due to an event reported to the NRC in LER 93-004, " Automatic Scram Resulting from Load Rejection at 100 Percent Power", dated April 12, 1993. During this event, the reactor vessel pressure temperature limit was exceeded during shutdown. The engineering evaluation of the event concluded the reactor vessel did not exceed the allowable limits of ASME Sections III and XI. Short term corrective actions included procedure changes and training. Long. term corrective actions include the proposed clarification to paragraph 3.6. A.2.

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i In NRC' Inspection Report No. 93-15, dated October 20, 1993, the resident NRC i inspectors made the following observation

" Control room staff demonstrated strong command and control-following i the weather related reactor trip on September 10, during which the  ;

effectiveness of recent training concerning reactor vessel '

pressure / temperature monitoring during cooldown was evident."

Thus, adding the clarifying words to the paragraph will enhance the already >

effective training and revised procedures.

The third change will modify TS Section 4.6.1.2 paragraph C.1. " Mechanical I Snubbers Functional Test Acceptance Criteria", by removing the following i sentence; " Drag Force shall not have increased more than 50% since the last functional test." Removal of this sentence will also address an NRC unresolved item (UNR 93-09-01) regarding the removal or revision of the 50% ,

drag force change requirement.  ;

i Correspondence with the vendor for the mechanical snubbers in use at PNPS, i Pacific Scientific, indicates they do not require this acceptance criteria for-  :

their snubbers and they recommend removal of this technical specification f requirement.

The requirement to monitor the increase in drag force from the last functional i test has no meaningful engineering basis. Vendor information indicates i snubbers of the same capacity can have initial drag force rates that vary '

considerably. For example, a unit whose rated load is 15000 lbs 'can be tested '

at 30 lbs of drag force while another 15000 lb. unit can be tested at 100 lbs  ;

of drag force, with both snubbers being acceptable for shipment by the vendor.  !

Therefore, the first unit would fail the less than 50% drag force test criteria if the subsequent test resulted in any drag force above 45 lbs.

Whereas, the second unit would be allowed to remain in service if it tested at ,

149 lbs. In reality, the second unit, with a drag force of 149 lbs could be i used to replace the snubber that " failed" the acceptance criteria with a drag force of 46 lbs.

Functional testing of mechanical snubbers at' PNPS will' continue to include i drag force testing in accordance with the ASME Boiler and Pressure Vessel  :

Code, Section.XI, Subsection IWF-5000. '

The improved Standard Technical Specifications, (ISTS), when referencing the Inservice Testing Program, requires adherence to the ASME Boiler and Pressure  :

Vessel Code,Section XI, for testing pumps, valves and snubbers. There is no i requirement in the ISTS or in the ASME Code'for the greater than 50% drag force criterion. Also, a review of peer plant T.S. (Vermont Yankee,  ;

Millstone, Hatch, Duane Arnold, Cooper, Oyster Creek, Peach Bottom and .

Monticello) revealed that none of these plants require demonstration of less '

than 50% drag force since the last functional test.

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r Complete conformance to the ISTS regarding snubber testing requirements cannot yet be requested due to the future NRC rulemaking regarding changes to

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10CFR50.55(a) and the incorporation by reference of the 0&M Code into Inservice Testing requirements. (Ref: 57FR 34666, dated August 6, 1992) ,

Determination of No Sionificant Hazards Considerations  !

The Code of Federal Regulations,10CFR50.91 requires that at the time a l licensee requests an amendment, it must provide to the Commission its analysis, using the standards in 10CFR50.92, about the issue of no significant j hazards consideration. Therefore, in accordance with 10CFR50.91 and r 10CFR50.92 the following analysis has been performed.

1. The operation of Pilgrim Station in accordance with the proposed r amendment will not involve a significant increase in the probability or  ;

consequences of an accident previously evaluated.

The first proposed change will modify License DPR-35 to eliminate the i need to issue a new page 3 to identify the latest amendment number. The second change will provide the correction of an error of omitting the a reference to the subcritical mode of operation, in relation to the pressure / temperature curves. The third change will remove the unnecessary mechanical snubber functional test acceptance criterion to '

determine if drag force has increased more than 50% since the last functional test.

Modification of License DPR-35 for Pilgrim Nuclear Power Station to l remove the need to update page 3 whenever a new amendment is approved ,

will reduce an administrative burden. This license change also  !

precludes a possible administrative error if the correct reference is somehow missed. This change does not affect plant operation or design and is considered an administrative change and as such does not involve l a significant increase in the probability or consequences of an accident l previously evaluated. 1 The second change corrects an error of omission made in an earlier  !

amendment by inserting a reference to the subcritical reactor operation l phase. This proposal will enhance the procedure changes and training already accomplished as short term corrective actions. This change does i not affect plant operation or design and is considered an administrative '

change and therefore does not involve a significant increase in the i probability or consequences of an accident previously evaluated. t i

The third change removes an acceptance criterion for mechanical ' snubber i testing not required by the ASME Boiler and Pressure Vessel Code, ,

Section XI, Subsection IWF nor recommended by the vendor for mechanical- =;

snubbers in use at Pilgrim. j This _ change will not result in any physical modification to Pilgrim.

The mechanical snubbers will continue to be tested in accordance with existing plant procedures which reference the ASME Code Section XI, i

Subsection IWF. Therefore, this is considered an administrative change ,

and as such, operation of Pilgrim will not involve a significant '

increase in the probability or consequences of an accident previously evaluated.  ;

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2. The operation of Pilgrim Station in accordance with the proposed l Amendment will not create the possibility of a new or different kind of .!

accident from any accident previously evaluated. l l

The proposed changas do not create the possibility of a new or different kind of accident than previously evaluated because they are '

administrative in nature and require no physical alterations of plant l

, configuration or changes to setpoints or operating parameters. '

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3. The operation of Pilgrim in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. j Because these changes do not alter plant operation or design and are i considered administrative in nature, they do not involve a significant  !

reduction in a margin of safety.

Summary 1

The proposed changes do not pose any significant hazards considerations as  !

discussed above. The changes were reviewed and recommended to the Station Director for approval by the Operations Review Committee and reviewed by the  :

Nuclear Safety Review and Audit Committee. '

i Schedule of Chance ,

i This change will be effective within 30 days of receipt of approval from the  :

NRC. l I

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ATTACHMENT B Bevised Technical Specification Pages 2

License Page 3 Appendix A Page 123 Appendix A Page 137c k

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