ML20212J457

Application for Amend to License NPF-36,incorporating License Change Application 6 & Revising Tech Specs to Implement Design Change to Lower Reactor Vessel Water Level Setpoint of MSIVs & Steam Line Drain Valves

ML20212J457
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	01/21/1987
From:	Leonard J LONG ISLAND LIGHTING CO.
To:
Shared Package
ML20212J427	List: ML20212J457 ML20212J475 SNRC-1308, Forwards Application for Amend to License NPF-36, Incorporating License Change Application 6 & Revising Tech Specs to Implement Design Change to Lower Reactor Vessel Water Level Setpoint.W/Certificate of Svc.Fee Paid
References
NUDOCS 8701280204
Download: ML20212J457 (4)

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LONG ISLAND LIGHT COMPANY Operating License NPF-36 Docket 50-322 License Change Application 6 This License Change Application requests modification to Operating License NPF-36 for the Shoreham Nuclear Power Station to achieve consistency with the anticipated physical condition of the plant. LILCO plans to implement a design change that will lower the reactor vessel water level setpoint (for containment isolation) of the main steam isolation valves and steam line drain valves (Groups 1 and 14) from level 2 to level 1 (see Bases Figure B 3/4 3-1 of the Shoreham Technical Specifications) . The implementation of this design change will fulfill the commitment LILCO made in SNRC-816 dated January 7, 1983, and_ acknowledged by the NRC in NUREG 0420 Supplement 4 (see pp 22-4 and 5),

inspection report 50-322/83-09 (see pp 5-6) and in the Shoreham Partial Initial Decision LBP-83-57 dated September 21, 1983 (see pp 119-124 and pp 389-415)

The request and supporting documentation are contained in the Attachment to this License Change Application.

Long Island Lighting Company f

By .M $k ont D. Leonsrd, Jr.

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Vice President - Nuclear rations Subscribed and sworn to before me this [ day of January 1987.

LINM A CRATT

?$ R hmwn tyrd dyjpgo'tary Public of New %*t71T 8701280204 870121 PDR

'3 My Commission Expires: /2arcL 7 M/W ADOCK 05000322 P PDR

Attachment To License Change Application 6 1.0 pESCRIPTION OF CHANGE Shoreham Technical Specification Tables 3.3.2-1, 3.3.2-2, 3.3.2-3 and 4.3.2-1 will require the proposed changes as identified in Exhibits A through D. These changes show that

'the main steam isolation valves and steam line drain valves will isolate primary containment on reactor vessel water level low low low level 1 instead of low low level 2.

In addition, Technical Specification Bases Figure B 3/4 3-1,

" Reactor Vessel Water Level", needs to be revised as shown on Exhibit E.

2.0 REASON FOR CHANGE This change is necessary to allow LILCO to fulfill a commit-ment that was made to resolve an outstanding issue in the Shoreham Safety Evaluation Report. Outstanding issue number 57, "TMI-2 Requirements"; item II.K.3.16, " Reduction of Challenges and Failures of Relief Valves - Feasibility Study and System Modification", was resolved, in part, by LILCO's commitment to modify Shoreham's design to lower the reactor vessel water level isolation setpoint of the main steam line isolation valves and the main steam line drain valves from level 2 (478.75 inches above vessel zero) to level 1 (384.25 inches above vessel zero). This modification will result in a reduction in the number of times the reactor is isolated from the main condenser and will reduce challenges to the safety relief valves (SRVs) by eliminating isolation cycling of the SRVs resulting from transients related to loss of feedwater flow.

In addition, LILCO completed the IDCOR Individual Plant Evaluation in May 1986. This evaluation was subsequently forwarded to the NRC by IDCOR on June 26, 1986. A significant outcome of this study was to confirm the wisdom of maintaining the main condenser heat sink.

3.0 BASIS FOR NO SIGNIFICANT HAZARDS FINDING The proposed isolation setpoint change does not involve a significant hazards consideration because operation of the Shoreham Nuclear Power Station Unit 1, in accordance with this change, would not t

(1) involve a significant increase in the probability or consequences of an accident previously evaluated because this modification reduces the probability of accidents due to challenges to SRVs and analyses demonstrate that the consequences of this modification on loss of feedwater transient, and large and small break Loss of Coolant Accidents (LOCAs) are not significant.

(2) create the possibility of a new or different kind of accident from any accident previously evaluated because the equipment utilized is equal to design, function and qualifications of existing equipment and the ef fects of the change are encompassed by existing accident analyses.

The direct impact of this proposed isolation setpcint change on an accident due to the delay in MSIV closure is an increased loss in the RPV coolant inventory. As described below, the adverse affects of this modification are most noticeable in LOCAs which have been reanalyzed to incorporate the change.

(3) involve a significant reduction in a margin of safety because the analyses demonstrates that the design parameters; namely, peak cladding temperature (PCT),

total core uncovery time, minimum critical power ratio (MCPR) and MSIV closure delay, are not significantly affected.

Analyses were performed by the Nuclear Steam Supply System vendor (GE) to support the change in setpoint for MSIV closure. New calculations were performed for a loss of feedwater flow transient using the same core analytical model (USAR 15A.1.8.3.1) and changing only the conditions for the MSIV isolation setpoint from level (L2) to level (L1). The change to level (L1) avoids the isolation pressurization which occurs at level (L2). Reactor Core Isolation Cooling (RCIC) and liigh Pressure Coolant Injection (!!PCI) systems, which are still initiated at level (L2), will keep the water in the reactor pressure vessel above level (L1). Other design parameters remain unchanged.

A new limiting break (85% Design Basis Accident dischargo) analysis was performed assuming the MSIV isolation to occur at level (L1) with overything else being the same, and it made ne difference to either the Peak Cladding Temperature (PCT) or the total core uncovered time. Whereas the MSIV would be opened three seconds longer because of the change, the additional loss of coolant inventory is negligible for the large break.

The effect of the change on a small break, based on pre-vigus plants similar to Shorehag, was only approximately 30 F for a limiting PCT of 1500 F, so the effect of this proposed change on the margin of safety was concluded to be insignificant and another analysis was not performed.

! The MSIV closure related calculations were reviewed for effect of changing the reactor level initiation signal from L2 to L1 on Pipe Break Outside Containment (PBOC) and Steamline Break Analyses. The review indicated no effect on the calculations because for pipe breaks outside the containment, the controlling isolation signal does not come from low reactor water level.

In the course of examining accidents evaluated j

previously, it has been determined that this setpoint change has no impact on radioactive releases.

The Commission has provided guidance concerning the appli-cation of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards consideration. Example (vi) relates to a change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all regulatory j acceptance criteria.

In this case, the proposed change described above is similar to Example (vi) because the analyses demonstrate that the applicable critoria of 10 CFR 50.46 is still conservatively satisfied.

Thorofore, based upon the above considerations and analyses, LILCO has determined that this proposed change does not involve a significant hazards consideration.

4.0 TIMING OF CilAMGE Since this proposed technical specification change will require extensive numbers of station procedure changes and operator training, LILCO requests that it become effective upon implementation of modification and operator training.

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