ML18100A883

Application for Amend to License DPR-75,modifying TS 2.2, Changing Table 2.2-1,reactor Trip Sys Instrumentation Trip Setpoints & Table 3.2-1,DNB Parameters

ML18100A883
Person / Time
Site:	Salem
Issue date:	02/03/1994
From:	Hagan J Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18100A884	List: ML18100A883 ML18100A885
References
LCR-94-06, LCR-94-6, NLR-N94016, NUDOCS 9402220069
Download: ML18100A883 (11)

Text

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Public Service Electric and Gas Company Joseph J. Hagan Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-1200 Vice President - Nuclear Operations fEB 03 1994 NLR-N94016 LCR 94-06 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REQUEST FOR AMENDMENT REACTOR COOLANT SYSTEM FLOW RATE SALEM GENERATING STATION UNIT NO. 2 DOCKET NO. 50-311 In accordance with the requirements of 10CFR50.90, Public Service Electric & Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating License DPR-75 for Salem Generating Station Unit No. 2. In accordance with 10CFR50.91(b) (1) requirements, a copy of this request has been sent to the State of New Jersey.

The proposed amendment modifies Technical Specification 2.2, Limiting Safety System Settings, Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints, Functional Unit 12 Loss of Flow.

The Reactor Coolant System (RCS) Loop design flow is reduced 1%

to 86,430 gpm per loop. The Note (*) associated with the Trip Setpoint and the Allowable Value for Loss of Flow is changed to reflect the new value.

The proposed amendment also modifies the Technical Specification 3.2.5, Power Distribution Limits, Table 3.2 DNB Parameters.

The RCS minimum required total flow rate is reduced 1% to 353,700 gpm. Table 3.2-1 is modified for Reactor Coolant System flow to provide a limit of greater than or equal to 353,700 gpm.

Attachment 1 includes a description, justification, and significant hazards analysis for the proposed change.

Attachment 2 contains the Technical Specification pages revised with pen and ink changes.

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Document Control Desk* 2 FEB 0 3 1994 NLR-N94016 LCR 94-06 PSE&G is requesting a 60 day implementation period after amendment approval.

Should there be any questions with regard to this submittal, please do not hesitate to contact us.

gan sident -

Operations c Mr. J. c. Stone Licensing Project Manager Mr. c. Marschall Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Manager IV New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625

Ref: NLR-N94016 STATE OF NEW JERSEY SS.

COUNTY OF SALEM J. J. Hagan, being duly sworn according to law deposes and says:

I am Vice President - Nuclear Operations of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Salem Generating Station, Unit No. 2, are true to the best of my knowledge, information and belief.

me 1994 KIMBERLY JO BROWN My commission expires on ~~~-~~~T~c~~~~!i~:.u~i!~~~~~~~~-~~~~-i'~;;~.R~~i~~.._~~~~~

ATTACHMENT 1 NLR-N94016 LCR 94-06 REACTOR COOLANT SYSTEM FLOW RATE I. DESCRIPTION OF THE PROPOSED CHANGE A. Change Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints as follows:

1. For Functional Unit 12, Loss of Flow, change the Note to read:

"*Design flow is 86,430 gpm per loop."

B. Change Table 3.2-1 DNB Parameters as follows:

1. Change the third parameter from "Reactor Coolant System" to read, "Reactor Coolant System Total Flow Rate".

2. Change the limit for Reactor Coolant System Total Flow Rate to read: ~353,700 gpm.

II. REASON FOR THE PROPOSED CHANGES Salem Unit No.2 has experienced a decrease in calculated RCS total flow over the past several refueling cycles. This has not been confirmed by RCS elbow tap data, which tends to indicate flow has remained basically constant. PSE&G is investigating the differences that have occurred between the calculated RCS flow and the elbow tap indications.

Following the Unit 2 eighth refueling outage, RCS total flow was calculated to be slightly above the minimum required by Technical Specification 3.2.5. Recently, a review of the flow calculation procedure identified a non-conservatism that may reduce the calculated flow by approximately 1000 gpm. PSE&G believes the decrease in calculated RCS total flow is based on changes in the indications and inputs to the calculation, not actual RCS flow.

PSE&G is investigating the low RCS flow calculation to resolve the discrepancy between calculated flow and elbow tap indications. The impetus for the proposed revision is the small margin between the calculated RCS total flow and the Technical Specification limit.

III. JUSTIFICATION FOR THE PROPOSED CHANGES Technical Specification 2.2 Limiting Safety System Settings -

Reactor Trip System Instrumentation Setpoints requires the Trip

NLR-N94016 Attachment 1 Setpoint and Allowable Value for Functional Unit 12 Loss of Flow to be based on the design RCS Flow per loop. If the setpoint is less conservative than the required value from Table 2.2-1, the channel must be declared inoperable and the appropriate Action Statement from Chapter 3/4.3 Instrumentation must be entered.

Design RCS flow identified for calculation of this setpoint is 87,300 gpm per loop. This flow is being reduced 1% to 86,430 gpm per loop. Since the loss of flow setpoint must be greater than or equal to 90% of the specified the design flow, reducing the design flow value would not require a physical change to the setpoint.

Technical Specification 3.2.5 Power Distribution Limits - DNB Limits requires Reactor Coolant System (RCS) Total Flow Rate to be greater than or equal to 357,200 gpm. If RCS flow is less than the required limit, flow must be restored to within limits within two hours or Thermal Power must be reduced to less than 5%

of Rated Thermal Power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The flow is determined by precision heat balance measurements at least once per 18 months and the elbow tap flow indications are correlated to the calculated value. PSE&G does not re-calibrate flow tap indications, but provides administrative limits that correlate to the calculated RCS flow values. The elbow tap meters provide continuous flow indication to ensure total RCS flow is greater than that required by Technical Specification 4.2.5.1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The Loss of Flow Reactor Trip is set at ~90% of design RCS flow per loop (87,300 gpm) and thus prevents operation with significant flow reductions. The RCS total flow rate is limited to greater than or equal to 357,200 gpm to ensure DNB limits are not exceeded. The proposed Technical Specification revision reduces the RCS loop flow rate to 86,430 gpm and RCS total flow rate minimum value to 353,700 gpm. This flow rate is a 1%

reduction in the minimum required RCS loop and total flow rates.

EFFECTS OF REDUCED RCS FLOW ON FSAR ANALYSIS PSE&G has confirmed that adequate margin exists in the LOCA, Non-LOCA and Containment analyses to justify the proposed Technical Specification changes. The accidents which rely on adequate RCS flow have been evaluated for the 1% reduction.

LOCA Analyses The following UFSAR LOCA analyses were evaluated for effects of a 1% reduction in RCS flow:

Large Break LOCA (UFSAR Section 15.4.1)

Small Break LOCA (UFSAR Section 15.3.1)

Steam Generator Tube Rupture (UFSAR Section 15.4.4)

NLR-N94016 Attachment 1 The Large Break LOCA analyses assume an RCS flow of 345,600 gpm.

The Small Break LOCA and Steam Generator Tube Rupture analyses assume a flow of 330,000 gpm. These values conservatively bound the proposed reduction in RCS flow. Therefore, the proposed change has no adverse impact on these analyses. The peak cladding temperature and hydrogen generation criteria of 10CFR50.46, and the offsite dose criteria of 10CFRlOO would continue to be met.

Long Tenn LOCA (Containment Integrity) Analyses (UFSAR Section

15. 4. 8 .1)

Because the proposed reduction in RCS flow does not affect the Tavg limits, the overall energy available in the RCS would not increase, and the limiting LOCA cases for mass and energy releases would not be affected. However, because there is a slight increase in reactor vessel delta-T, the hot leg break cases were evaluated to assess the effect of the redistribution of available energy towards the hot leg side. This evaluation concludes that the containment pressure resulting from a double ended hot leg break would increase by an upper bounding limit of 0.15 psi. The hot leg breaks would remain bounded by the limiting double-ended pump suction break cases, which are not adversely affected by the proposed reduction in RCS flow.

Subcompartment Analyses (UFSAR Section 15.4.8.3)

The proposed reduction in RCS flow is estimated to result in a 0.4 degree F reduction in vessel inlet temperature. Reduced temperature results in increased fluid density. The penalty associated with the change in temperature would result in approximately a 0.1% increase in critical flow. The licensing basis analysis model is not readily available to allow performance of a sensitivity evaluation for the 0.1% increase in critical flow. However, review of the current subcompartment analysis results show that evaluations were performed for breaks as large as 100 square inches at the reactor pressure vessel inlet and outlet piping. Based on piping displacements resulting from LOCA, and gap sizes for pipe whip restraints, it has been determined that the largest break consistent with the RCS piping configuration, is 75 square inches at the vessel inlet and outlet locations, and a single-ended break at all other RCS locations.

This reduction in break size offsets any penalty associated with the reduced RCS flow.

Note that the reduction in break size is based on a mechanistic evaluation of RCS piping, but does not rely upon leak-before break technology. A Salem specific leak-before-break submittal was made to the NRC on July 6, 1993, justifying further relaxations in primary loop pipe break postulations for the purposes of evaluating dynamic effects.

NLR-N94016 Attachment 1 Blowdown Reactor Vessel and Loop Forces (UFSAR Section 3.9.1.5}

The forces created by a postulated RCS pipe break are a function of RCS operating conditions, including reactor vessel inlet and outlet temperatures. The proposed 1% reduction in RCS flow would have an effect on RCS temperature. However, the impact on operating temperature is small enough to be considered negligible relative to the calculation of forces from a postulated RCS pipe break.

Post LOCA. Long Term Subcriticality (Related to UFSAR Section 15.4.1)

Hot Leg Switchover Time and Adequate Core Cooling Flow Verification (UFSAR Sections 6.3.2 and 15.4.1)

A review of the analyses associated with these transients indicates that the reduced RCS flow would have no adverse impact on the assumptions for the respective calculations.

Non-LOCA Analyses The following current Salem Non-LOCA analyses were evaluated for the effects of a 1% reduction in RCS flow:

Excessive Load Increase (UFSAR Section 15.2.11)

Four cases of a 10% step load increase from nominal full power conditions are analyzed, based on automatic versus manual rod control, and minimum versus maximum reactivity feedback parameters. Evaluation of this transient using the proposed reduction in RCS flow has concluded that the Departure From Nucleate Boiling Ratio (DNBR) design basis is met.

Excessive Heat Removal Due to Feedwater System Malfunctions (UFSAR Section 15.2.10)

Two cases for this transient are analyzed; a full power case is evaluated to show that the DNBR remains above the safety limit, and a zero power case is shown to be bounded by the Rod Withdrawal From Subcritical (RWFS) transient. Steam generator overfill is considered also evaluated for this transient.

Evaluation of the proposed reduction in RCS flow has determined that the DNBR for the full power case would remain above the safety limit, the zero power case remains bounded by RWFS, and steam generator overfill would not be affected.

Accidental Depressurization of the Main Steam System (UFSAR Section 15.2.13)

This event is initiated by the full opening of a single steam dump, relief or safety valve from hot zero power conditions. A reduction in RCS flow potentially decreases the minimum DNBR.

NLR-N94016 Attachment 1 Evaluation of this event shows that the DNBR analysis of record remains valid given the proposed reduction in RCS flow.

Major Secondary Side Pipe Rupture (UFSAR Section 15.4.2)

A reduction in RCS flow would potentially decrease the minimum DNBR for this event, which considers a double-ended rupture of the main steam piping at hot zero power, with and without offsite power. The DNB penalty would be partially offset because the lower flow would reduce the primary to secondary heat transfer, resulting in a reduced power increase. Evaluation of this event concludes that sufficient DNB margin exists to offset the reduced flow penalty, without taking credit for the offsetting reduction in power increase.

Steam Line Break Mass/Energy Release (UFSAR 15.4.8.2)

A decrease in RCS flow reduces primary to secondary heat transfer, thereby reducing steam pressure and temperature during normal operation. Any reduction in secondary temperature and pressure would tend to lessen the mass/energy release following a steam line break. Therefore, the proposed reduction in RCS flow would not adversely affect the steamline break mass/energy releases.

Loss of External Electrical Load and/or Turbine Trip (UFSAR Section 15.2.13)

Four cases of a total loss of steam demand at full power, without a direct reactor trip, are analyzed, based on automatic rod control versus no rod control, and minimum versus maximum reactivity feedback parameters. Evaluation of this transient using the proposed reduction in RCS flow has concluded that the Departure From Nucleate Boiling Ratio (DNBR) design basis is met.

An evaluation of the maximum primary and secondary system pressures following this transient was also performed. This evaluation concluded the pressures would not be significantly affected by the proposed flow reduction.

Loss of Offsite Power (UFSAR Section 15.2.9)

An evaluation of the proposed reductions in RCS flow for the Loss of Offsite Power event shows that natural circulation core cooling would not be significantly affected. The DNBR remains above the safety limit, the licensing basis criteria for primary and secondary system pressure continues to be met, and pressurizer filling would not occur.

Loss of Normal Feedwater (UFSAR Section 15.2.8)

Evaluation of this event shows that sufficient margin exists relative to primary and secondary peak pressures and pressurizer

NLR-N94016 Attachment 1 filling, such that the conclusions stated in the UFSAR remain valid.

Feedwater System Pipe Break (UFSAR Section 15.4.3)

Evaluation of this event shows that the proposed reduction in RCS flow would result in a small decrease in steam generator mass, and no significant impact on peak hot leg temperatures. The current licensing basis analyses contain sufficient margin to accommodate the decay heat removal penalty associated with the proposed reduction in RCS flow.

Partial Loss of Forced RCS Flow (UFSAR Section 15.2.5)

Complete Loss of Forced RCS Flow (UFSAR Section 15.3.4)

Evaluation of these transients shows that the effects of the proposed reduction in RCS flow on DNB and RCS pressure can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid.

Reactor Coolant Pump Shaft Seizure (Locked Rotor)

Evaluation of this transient shows that the effects of the proposed reduction in RCS flow on peak fuel clad temperature and RCS pressure can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid.

Rod Withdrawal from Subcritical Condition (UFSAR Section 15.2.1)

Rod Withdrawal at Power (UFSAR Section 15.2.2)

Rod Cluster Control Assembly Misalignment (UFSAR Section 15.2.3 and 15.3.5)

Startup of an Inactive Loop (UFSAR Section 15.2.6)

Evaluation of these transients shows that the effects of the proposed reduction in RCS flow on DNB can be accommodated by existing margins, such that the conclusions in the UFSAR remain valid.

Uncontrolled Boron Dilution (UFSAR Section 15.2.4)

Thermal design flow is not an input to the boron dilution analysis. Therefore, the conclusions presented in the UFSAR are not affected by the proposed reduction in RCS flow.

Rupture of a Control Rod Drive Mechanism Housing (UFSAR Section 15 .4. 7)

In order to demonstrate that gross fuel damage would not occur, the core would remain in a coolable geometry, and the RCS would remain intact, the following more restrictive criteria are applied to this event:

NLR-N94016 Attachment 1

1) The average fuel pellet enthalpy at the hot spot is less than 200 cal/gm (360 Btu/lbm) .

2) Fuel melt at the hot spot is limited to less than the innermost 10% of the fuel pellet.

3) Peak RCS pressure is less than that which would cause stresses to exceed the Faulted Condition Limits.

Evaluation of this transient shows that the effects of the proposed reduction in RCS flow on peak fuel clad temperature and RCS pressure can be accommodated by existing margins, such that the above criteria would continue to be met, and the conclusions in the UFSAR remain valid.

Spurious Operation of Safety Injection at Power (UFSAR Section 15.2.14)

Since the transient conditions of this event are not significantly altered by the proposed reduction in RCS flow, the conclusions in the UFSAR remain valid.

Accidental Depressurization of the RCS (UFSAR Section 15.2.12)

Since the transient conditions of this event are not significantly altered by the proposed reduction in RCS flow, and the OT-delta-T setpoint is not changed, the conclusions in the UFSAR remain valid.

IV. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION In accordance with 10CFRS0.92, PSE&G has reviewed the proposed changes and concluded the proposed changes do not involve a significant hazards consideration because the changes would not:

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

No component modification, system realignment, or change in operations will occur which could affect the probability of any accident or transient. The proposed reduction in RCS loop and total flow rates will not change the probability of a challenge to any Engineered Safeguard Feature or other device. The consequences of previously analyzed accidents have been found to remain within acceptable licensing basis limits when the reduced flow rates are assumed. The system transient response is not affected by the initial RCS flow assumption, unless the initial assumption is so low as to impair the steady-state core cooling capability or steam generator heat transfer capability. This is clearly not the case with a 1% reduction in RCS flow. The proposed change to the wording of the parameter title on Table 3.2-1 is editorial for clarity. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously analyzed.

NLR-N94016 Attachment 1

2. Create the possibility of a new or different kind of accident.

No component modification, system realignment, or change in operating procedure will occur which could create the possibility of a new event not previously considered. The proposed reduction in RCS loop and total flow rates will not initiate any new events. Therefore, the proposed changes would not create the possibility of a different or new kind of accident.

3. Involve a significant reduction in a margin of safety.

The proposed decrease in RCS loop and total flow rates has been analyzed and found to have an insignificant effect on the applicable transient analyses found in the FSAR. The proposed change to the wording of the parameter title on Table 3.2-1 is editorial for clarity. Therefore, the proposed changes would not involve a significant reduction in any margin of safety.

Therefore, based on the information presented above, PSE&G has concluded there is no significant hazards consideration.