ML20069Q023

Application for Amend to License NPF-38,consisting of TS Change Request NPF-38-152,adjusting 3 Plant Protection Sys Trip Setpoints & Allowable Values

ML20069Q023
Person / Time
Site:	Waterford
Issue date:	06/22/1994
From:	Barkhurst R ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20069Q025	List: ML20069Q023 ML20069Q038
References
W3F1-94-0121, W3F1-94-121, NUDOCS 9406270210
Download: ML20069Q023 (11)

Text

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Ross P. Barkhurst v< e Rovym Qvut n v.ww 3 W3F1-94-0121 A4.05 PR June 22, 1994 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Subject:

Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-152 Gentlemen:

The attached description and safety analysis supports a change to the Waterford 3 Technical Specifications (TS). The proposed change will modify the TS by adjusting 3 Plant Protection System (PPS) trip setpoints and allowable values. The change adjusts the affected TS values in a more conservative direction such that they will be consistent with the current setpoint/ uncertainty methodology being implemented at Waterford 3.

The new setpoints and allowable values are based on a revised PPS Setpoint .

Analys i s . The new analysis will become effective as Waterford 3 restarts following the current refueling outage. Therefore, Waterford 3 will adhere to the revised conservative values during startup and operation following Refuel 6.

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4 I I 9406270210 940622 PDR ADOCK 05000382

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. I Technical Specification Change Request NPF-38-152  !

W3F1-94-0121 Page 2 June 22, 1994 Should you have any questions or comments concerning this request, please

> contact Paul Caropino at (504)739-6692.

Very truly yours, t

( JoJ N AJJ d)

R.P. Barkhurst Vice President, Operations Waterford 3 RPB/PLC/ssf

Attachment:

Affidavit NPF-38-152 Reference List Figure 1 cc: L.J. Callan, NRC Region IV D.L. Wigginton, NRC-NRR R.B. McGehee N.S. Reynolds NRC Resident Inspectors Office Administrator Radiation Protection Division (State of Louisiana)

American Nun. ear Insurers

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the matter of )

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Entergy Operations, Incorporated ) Docket No. 50-382 Waterford 3 Steam Electric Station )

AFFIDAVIT R.P. Barkhurst, being duly sworn, hereby deposes and says that he is Vice President Operations - Waterford 3 of Entergy Operations, Incorporated; that '

he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-152; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.

JJD N R.P. Barkhurst Vice President Operations - Waterford 3 STATE OF LOUISIANA )

) sr PARISH OF ST. CHARLES )

Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 22*" day of /oNa , 1994.

Tk C7 Notary Public k k l

My Commission expires W s rn u rc. .

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-152 The proposed change affects Technical Specification (TS) Table 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS, Table 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES and associated Bases by modifying three parameter setpoints as follows: ,

1. Table 2.2-1 Item 2) Linear Power Level-High: the TRIP SETPOINT of s 110% of RATED THERMAL POWER and the ALLOWABLE VALUE of s 110.7% RATED THERMAL POWER are changed to reflect s 108% and s 108.76% respectively.

2. Table 2.2-1 Item 4) Pressurizer Pressure-High: the TRIP SETPOINT of i s 2365 psia and ALLOWABLE VALUE of s 2372 psia are changed to reflect s 2350 psia and s 2359 psia respectively.

3. Table 3.3-4 Item 7 c.& d.) Emergency Feedwater, Steam Generator AP-High:

the TRIP VALUE of s 127.6 psid and ALLOWABLE VALUE of s 136.6 psid are changed to s 123 psid and s 134 psid respectively.

The proposed changes are necessary to accommodate additional uncertainty and will have no effect on the original safety analysis setpoints (analytical limits).

Existina Specificatiqn See Attachment A Proposed Specification See Attachment B fLiLCEEoMnd The proposed change updates the TS consistent with the Waterford 3 revised PPS Setpoint Uncertainty Calculation (Reference 1). The revised calculation was i genera'ad using a new Waterford 3 Setpoint and Uncertainty Determination I guideline (Reference 2) that provides plant personnel with a description of the governing codes and standards, plant specific criteria, and concepts involved in instrument loop uncertainty analysis and setpoint determination.

The Setpoint and Uncertainty Determination guideline was generated to identify and define additional uncertainty factors that have evolved in the industry since the original Waterford 3 PPS Setpoint Analysis (Reference 3) was developed.

The use of improved guidelines i.s in+ ended to enhance plant safety and in no way invalidates the previous analysis which was also performed using approved methodology.

This proposed change is also driven by a change in plant components. Several transmitters that provide input to the PPS have been upgraded. This upgrade was completed during our recent refueling outage and the new PPS Setpoint Calculation incorporates the new transmitter performance data. An engineering calculation (Reference 4) has demonstrated that the specifications for the original transmitters bound those for the upgraded transmitters.

l De_scriotion

! The proposed change modifies the identified parameter TRIP SETPOINTS and ALLOWABLE VALUES to reflect the following. .

1 PARAMETER TRIP SETPOINT ALLOWABLE VALUES HI Linear Power s 108 % PWR s 108.76 % PWR HI Pressurizer Pressure s 2350 PSIA s 2359 PSIA ,

HI Steam Generator AP s 123 PSID s 134 PSID l The Reactor Protection System (RPS) Instrumentation Trip Setpoint Limits and the Engineered Safety feature Actuation System (ESFAS) Instrumentation Trip Values listed in the Technical Specification Tables 2.2-1 and 3.3-4 prescribe those settings for critical parameters that will avoid exceeding any j analytical limit for postulated Design Bases Accidents (DBAs) or Anticipated Operational Occurrences (A00s).

The Trip Setpoints and Allowable Values are based on analytical limits stated in the Final Safety Analysis Report (FSAR). The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties, instrumentation drift, and severe environment effects for those RPS and ESFAS channels that must function in harsh environments as defined by 10 CFR 50.49, Allowable Values specified in Tables 2.2-1 and 3.3-4 are conservatively adjusted with respect to the analytical limits. A detailed description of the methodology used to calculate the proposed trip setpoints, including their explicit uncertainties, is documented in Reference 1 and 2.

The specific safety analysis applicable to each protective function affected by the proposed change are identified below:

Linear Power Level-Hiah This function provides a reactor trip on high linear neutron flux level . This trip provides protection against core damage during the following:

. Uncontrolled CEA Withdrawal from Low Power (A00);

. Uncontrolled CEA Withdrawal at Power (A00); and

. CEA Ejection (Accident).

Pressurizer Pressure-Hiah This function provides a reactor trip on high pressurizer pressure. This trip provides protection for the high Reactor ,

Coolant System pressure Safety Limit. In conjunction with the pressurizer safety valves and the main steam safety valves, it provides protection against overpressurization of the Reactor Coolant Pressure Boundary during the l following events:

. Loss of Electrical Load Without a Reactor Trip Being Generated by the l Turbine Trip (A00); l

. loss of Condenser Vacuum (A00); ,

l . CEA Withdrawal From Low Power Conditions (A00); i j . Chemical and Volume Control System Malfunction (A00); and l l . Main Feedwater System Pipe Break (Accident). I Steam Generator Differential Pressure-Hiah This function of the Plant Protection System enables emergency feedwater flow only to the intact Steam Generator (SG) when the high SG differential pressure condition exist. ESFAS logic includes SG Differential Pressure-High (SG-1 > SG-2 or SG-2 > SG-1) to determine if a rupture in either generator has occurred. Rupture is assumed if the affected generator has a low pressure condition, unless that generator is significantly higher in pressure than the ether generator. This latter feature allows feeding the intact SG, even if both are below the Main Steam l Isolation Signal (MSIS) setpoint, while preventing the ruptured generator from j being fed. Not feeding a ruptured generator prevents containment  !

nyerpressurization during analyzed events. The most limiting accident condition for the High Steam Generator Differential Pressure function is a ,

Steam Line Break event. {

The methodology used in the revised PPS Setpoint Calculation is generelly unchanged. However, several factors are now considered or clarified in the j revised analysis (e.g. cable insulation resistance degradation, steam generator elongation, containment conditions during specific accidents, and more precise definition of process measurement errors). Figure 1 is provided to illustrate the proposed setpoint change.

i SE ETY ANALYSIS l

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No Implementing the proposed change will not affect any design basis accidert. The revised trip and actuation setpoints are based upon the same analytical limits that form the basis for the current trip and actuation setpoints. The design basis for each trip and actuation setpoint was verified to be consistent with the apprcpriate accident analyses as part of the process of revising the PPS setpoint analysis.

The proposed changes in trip and actuation setpoints are all in the conservative (away from the analytical limits) direction. Therefore, the proposed change will not involve a significant increase in the probability or consequences of any previously analyzed accident.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or uifferent kind of accident from any accident previously analyzed?

Response: No Plant operation and the manner in which the plant is operated will not be altered as a result of implementing the proposed change since no new system or design change is being implemented. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in margin of safety?

Response: No The current safety margins of the affected trip setpoints and allowable values is preserved by the proposed change. This is assured by retaining the current analytical limit for the affected parameters.

Since the analytical limits are not affected and the total channel uncertainty is increased the margin of safety for the affected trip setpoints and allowable values is preserved. Therefore, the proposed change will not involve a significant reduction in margin of safety.

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Safety and Sionificant Hazards Determination Based on the above safety analysis, it is concluded that: (1) the proposed l change does not constitute a significant hazards consideration as defined by '

10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC final environmental statement.

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4 NPF-38-152 l 1

Figure 1 1 ORIGINAL CALCULATION REVISED CALCULATION Hich Pressurizer Pressure Tu &aw AN Alvi! CAL LIMIT (2t.22 Pst A ) g, Uncertainty Uncertainty 57 PS!A 0 3 PSIA setoonnt: 2365 setooint: 2350 Hich Linear Power ANALYTICAL LIMIT (115 %) g, Uncertainty uncertainty 4.9% setooint: 110.1 % 6.24%

letooint: 108 % Round of f Allowsnce Hich S/G D/P g, g, ANALYTICAL LIMlf (230 PSID) gg,g ,g Uncertainty Uncertainty 102.4 P51D setooint: 127.6 PSID 104.76 PSID setooint: 123 PSID Round off Allowance

P References

1. Plant Protection System Setpoint Uncertainty Calculation (EC-192-019 Rev.A) dated August 30,1993.

2. I&C Engineering Design Guide for Setpoint and Uncertainty Determination Rev. 1 dated February 26,1993

3. Combustion Engineering "LP&L WSES-3 PPS Setpoint Analysis" (9270-ICE-36182 Rev.2) dated August 16,1983.

4. Rosemount Transmitter Comparison Calculation (EC-192-030 Rev.1) dated 06/11/93 4

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