ML20077E780

TS Change Request NPF-38-162 to License NPF-38,revising Plant Protection Sys Trip Setpoint & Several Allowable Values for Consistency W/Current Setpoint/Uncertainty Methodology Being Implemented at Facility

ML20077E780
Person / Time
Site:	Waterford
Issue date:	12/09/1994
From:	Barkhurst R ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20077E783	List: ML20077E780 ML20077E795
References
W3F1-94-0170, W3F1-94-170, NUDOCS 9412130126
Download: ML20077E780 (10)

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. K iuna L A M M 0731 Tet 504 n9 b601 Ross P. Barkhurst w nm a.,

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W3F1-94-0170 A4.05 PR December 9,1994 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-38-162 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 revising a Plant Protection System (PPS) Trip Setpoint and several Allowable Values such that'they 'will be consistent with the current setpoint/ uncertainty methodology being implemented at Waterford 3.

The proposed change has been evaluated in accordance with 10CFR50.91(a)(1)'

using criteria in 10CFR50.92(c) and it has been deternined that the proposed change involves no significant hazards considerations. The Plant Operations Review and Safety Review Committees have reviewed and' accepted-the proposed change based on the evaluat' ion mentioned above.

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Technical Specification Change Request NPF-38-162 W3F1-94-0170 Page 2  ;

December 9,1994 Should you have any questions or comments concerning this request, please contact Paul Caropino at (504)739-6692.

1 Very truly yours, l i

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3L h' R.P. Barkhurst Vice President, Operations l Waterford 3 j RPB/PLC/ssf

Attachment:

Affidavit NPF-38-162 cc: L.J. Callan, NRC Region IV l C.P. Patel, NRC-NRR R.B. McGehee N.S. Reynolds NRC Resident Inspectors Office Administrator Radiation Protection Division (State of Louisiana)

American Nuclear Insurers 1

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l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1

In the matter of )

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

AFFIDAVIT l

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-38-162; 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.

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• R.P. Barkhurst Vice President, Operations - Waterford 3 STATE OF LOUISIANA )

) ss PARISH OF ST, CHARLES )

Subscribed and sworn to before me, a Notary Public in and for the Parish and State above nar:d this 9 T" day of 'D W '" 3 E R. , 1994.

Mem.b (c %

Notary Public My Commission expires v .r u tirc ,

DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-162 The proposed change affects Technical Specification (TS) Table 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS and Table 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES as follows:

1. Table 2.2-1 Item 3) Logarithmic Power Level - High: the ALLOWABLE VALUE of s 0.275% of Rated Thermal Power is changed to s 0.280% of Rated Thermal Power.

2. Table 2.2-1 Item 5) Pressurizer Pressure - Low: the ALLOWABLE VALUE of 2 1644 psia is changed to 2 1649.7 psia.

3. Table 2.2-1 Item 6) Containment Pressure - High: the ALLOWABLE VALUE of s 17.3 psia is changed to s 17.4 psia.

4. Table 2.2-1 Item 7) Steam Generator Pressure - Low: the ALLOWABLE VALUE of 2 748 psia is changed to 2 749.9 psia.

5. Table 2.2-1 Item 8) Steam Generator level - Low: the ALLOWABLE VALUE of 2 26.7% is changed to 2 26.48%.

6. Table 2.2-1 Item 11) Steam Generator Level - High: the ALLOWABLE VALVE of s 88.4% is changed to s 88.62%.

7. Table 2.2-1 Item 16) Reactor Coolant Flow - Low: the TRIP SETPOINT of 2 23.8 psid and the ALLOWABLE VALUE of 2 23.6 psid are changed to 2 19.00 psid and 2 18.47 psid respectively.

8. Table 3.3-4 Item 5.b) Refueling Water Storage Pool - Low: the ALLOWABLE VALUE of 2 9.3% is changed to 2 9.08%. l 1

The proposed changes are necessary to accommodate recalculated uncertainties and will have no effect on the original safety analysis values, i.e.,

Analytical Limits.

In order to clarify the relationship between PPS Trip Setpoints and Allowable Values, the TS Bases are revised to include additional information. l I

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'Existina Soecification See Attachment A.

Proposed Soecification See Attachment B Backaround The proposed change modifies a setpoint and several allowable values as part of implementing the Waterford 3 revised PPS Setpoint Uncertainty Calculation (Reference 1). The revised calculation was generated using a new Waterford 3 Setpoint_and Uncertainty Determination 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 use of improved guidelines is intended to enhance plant safety and in no way invalidates the previous analysis which was also performed using approved methodology.

The Setpoint and Uncertainty Determination guideline was generated to identify )

and define additional uncertainty factors that have evolved in the industry l since the original Waterford 3 PPS Setpoint Analysis-(Reference 3) was developed. EC-192-019 Rev.A (Reference 1) included the maximum or minimum i Trip Setpoint values for those parameters affected by this proposed change.  !'

In all cases with the exception of Reactor Coolant Flow - Low, the existing TS Trip Setpoint is retained. Subjecting these Trip Setpoint values to the new method of determining Periodic Test Error (PTE) resulted in the proposed changes to the Allowable Values.

The Trip Setpoint for Reactor Coolant Flow - Low was reevaluated because the current Trip Setpoint is too close to the normal operating band. The new Trip Setpoint and Allowable Value for Reactor Coolant flow - Low'was established as follows in accordance with Reference 1. The Total Loop Uncertainty (TLU) accounting for instrumentation uncertainties, instrument drift, loop I calibration tolerances, and instrument errors due to accident harsh environments, is calculated for each.PPS channel loop. A nominal Trip Setpoint value of 19.00 psid was established by moving away from the analytical limit in the conservative direction by the amount of the (TLU).

The Allowable Value of 18.47 psid was established by adding a tolerance to the nominal Trip Setpoint to account for the random measurement errors encountered while performing the CHANNEL FUNCTIONAL TEST. This is the Periodic Test Error that includes RPS cabinet accuracy, measuring and test equipment (M&TE) accuracy, and RPS cabinet bistable drift. The Reactor Coolant Flow - Low change represents an improvement over the existing TS by allowing for an

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increased operating band with sufficient allowance between the Analytical Limit and the Trip Setpoint.

This proposed change also accommodates a change in plant components. Several transmitters that provide input to the PPS have been upgraded. This upgrade was completed during Refuel 6 and the revised 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. Therefore, current TS values (including those affected by this proposed change) are providing adequate protection.

Description The proposed change modifies the identified parameter TRIP SETPOINT and ALLOWABLE VALUES to reflect the following: (N/C - no change)

PARAMETER TRIP SETPOINT ALLOWABLE VALUES Logarithmic Power Level - High N/C s 0.280 % PWR Pressurizer Pressure Low N/C 2 1649.7 psia Containment Pressure - High N/C s 17.4 psia Steam Generator Pressure - Low N/C 2 749.9 psia Steam Generator Level - Low N/C 2 26.48 %

Steam Generator Level - High N/C s 83.62%

Reactor Coolant Flow - Low 2 19.00 psid 2 18.47 psid Refueling Water Storage Pool - Low N/C 2 9.08 %

The Reactor Protection System (RPS) Instrumentation and Engineered Safety Feature Actuation System (ESFAS) Instrumentation Trip Setpoints listed in the Technical Specification Tables 2.2-1 and 3.3-4 prescribe those settings for critical parameters that will avoid exceeding any Analytical Limit stated in the Final Safety Analysis Report (FSAR) for postulated Design Bases Accidents (DBAs) or Anticipated Operational Occurrences (A00s).

As stated previously, the selection of these Trip Setpoints is such that adequate protection is provided when all sensor and signal processing l component uncertainties are taken into account. To allow for calibration  !

tolerances, instrumentation uncertainties, instrumentation drift, and severe I environment effects for those RPS and ESFAS channels that must function in j harsh environments as defined by 10 CFR 50.49, Trip Setpoints 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 Trip Setpoints, including their explicit uncertainties, is documented in Reference 1 and 2.

The specific purpose applicable to each protective function affected by the proposed change is identified below:

Loaarithmic Power Level - Hiah: Its purpose is to assure the integrity of the fuel cladding and reactor coolant system boundary in the event of unplanned critically from a shutdown condition, resulting from either dilution of the soluble boron concentration or withdrawal of CEAs. This trip provides protection during the following:

. Control Element Assembly (CEA) Withdrawal Pressurizer Pressure - Low: Its purpose is to limit core damage during a postulated accident. This trip provides protection during the following:

Loss of Coolant Accident (LOCA)

. CEA Ejection Containment Pressure - Hiah: Its purpose is to protect the containment vessel integrity and minimize the radioactive release during a postulated accident.

This trip provides protection during the following:

Main Steam Line Break (MSLB)

Steam Generator Pressure - Low: Its purpose is to provide a reactor trip to assist the Engineered Safety Features (ESF) system in a Steam Line Break or a Feedwater Line Break event. This trip provides protection during the following:

. MSLB Feedwater Line Break (FWLB)

Steam Generator Level - Low: Its purpose is te assure that there is sufficient time for actuating the emergency feeddater pumps to remove decay heat from the reactor in the event of a reduction of feedwater flow. This trip provides protection during the following:

. MSLB Steam Generator Level - Hiah: Its purpose is to prevent moisture carryover from the steam generators which could result in damage to the turbines.

Reactor Coolant Flow - Low: Its purpose is to provide protection for loss of reactor coolant flow in the sheared shaft event. The Low Reactor Coolant Flow trip is credited in the Safety Analysis Report for the Sheared Shaft event and Steam Line Breaks with Loss of Offsite Power.

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~Refuelina ' Water Storaae Pool - Low: Its purpose is to allow long 'erm t cooling of the reactor core following a LOCA by recirculating water in the containment sump'using the.HPSI pumps. This trip provides protection during the 3 following: i

.- LOCA SAFETY ANALYSIS

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

.l Response: No Implementing the proposed change will not affect any design basis accident. The revised Trip Setpoint and Allowable Values are based upon the same Analytical _ Limits that form the basis for the current Trip Setpoints and Allowable Values. The design basis for each Trip Setpoint was verified to be consistent with the appropriate accident analyses.as  !

part of the process of revising the PPS. setpoint analysis. The proposed' change would implement a new Trip Setpoint for the Reactor Coolant (RC)

System Low Flow Reactor trip and new Allowable Values for RC Low Flow,  !

HI Log Power, HI Steam Generator Water Level, HI Containment Pressure, Low Pressurizer Pressure, Low Steam Generator Pressure, Low Steam Generator Water Level, and Low RWSP Level, based on the results of calculation EC-192-019. The revised Low RC Flow Trip Setpoint is based on the same analytical limit as the current setpoint. The revised ,

calculation uses the same design inputs with a similarly based methodology to calculate a smaller loop uncertainty. This results in a revised RC Low Flow Trip Setpoint that retains the original analysis limit. Therefore, the proposed change will not involve a significant . .

increase in the probability or consequences of any previously analyzed l accident.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different 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. The proposed Setpoint and l Allowable Value changes do not create any new system interactions or interfaces. All information used to calculate the new Trip Setpoint is e I

l- consistent with that of the existing accident analyses, and no new system interfaces / interactions are created. 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 proposed setpoint change revised the point at which the RCS Low Flow reactor trip initiates a reactor trip. The Trip Setpoint is based on the same Analytical Limit used to determine the current setpoint. In addition, the same basic setpoint determination methodology is employed.

That is, the Trip Setpoint is the Analytical Limit the Total Loop Uncertainty. The Allowable Valne is the Trip Setpoint the Periodic Test Error. The change in the setpoint and allowable values are due to a change in calculated TLU and PTE. The proposed Trip Setpoint and Allowable Values are based on the same Analytical Limits for the affected parameters and are determined using approved methodology.

Therefore, the proposed change will not involve a significant reduction in margin of safety.

Safety and Sianificant Hazards Determination Based on the above safety analysis, it is concluded that: (1) the proposed 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.

References

1. Plant Protection System Setpoint Uncertainty Calculation (EC-I92-019 Rev.A) dated August 30,1993.

2. Design Guide DE-IC-502, I&C Engineering Design Guide for Setpoint and Uncertainty Determination Rev. I 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.

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NPF-38-162 ATTACHMENT A i

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