ML20077E749

TS Change Request NPF-38-160 to License NPF-38,revising Allowable Opening Tolerances on Pressurizer Code Safety Valves & Main Steam Line Code Safety Valves from Plus or Minus 1% to Plus or Minus 3%

ML20077E749
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
ML20077E752	List: ML20077E749 ML20077E760
References
W3F1-94-0171, W3F1-94-171, NUDOCS 9412130114
Download: ML20077E749 (8)

Text

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W3F1-94-0171.

A4.05 PR December 9,1994-U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 .i

Subject:

Waterford 3 SES  !

Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-160 Gentlemen:

The attached description and safety analysis supports a change to the Waterford 3 Technical Specifications (TS).

The proposed change modifies the Waterford 3 TS by revising the allowable opening tolerances on the Pressurizer Code Safety Valves and the Main Steam Line Code Safety Valves from'il% to i3%. The FSAR analyses affected by this change have been evaluated and results of the impacted events have been found to be within acceptable limits. This request is submitted as a  ;

result of an effort to improve valve performance and to ensure that the'TS '!

limits are consistent with-expected valve performance capabilities.

The proposed change has been evaluated _in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined 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 evaluation mentioned above.

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Technical Specification Change Request NPF-38-160 W3F1-94-0171 Page 2 December 9,1994 l

Should you have any questions or comments concerning this request, please l contact Paul Caropino at (504)739-6692. I Very truly yours,

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

Attachment:

Affidavit NPF-38-160 cc: L.J. Callan, NRC Region IV 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

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-38-160; 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.

1A.

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 named this CfT" day of D E C. E_m 6 ER , 1994.

% I!L - I~C (O Notary Public My Commission expires la i r o L i N .

DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-160 This proposal requests a change to Waterford 3 Technical Specification 3.4.2.1, " Reactor Coolant System: Safety Valves - Shutdown", 3.4.2.2, " Reactor Coolant System: Safety Valves - Operating" and Table 3.7-1, " Steam Line Safety Valves Per Loop."

Existina Specification See Attachment A Proposed SDecification See Attachment B Description The proposed amendment would: 1) change the allowable opening tolerances on the Pressurizer Code Safety Valves (Technical Specification 3.4.2 and 3.4.3) and the Main Steam Line Code Safety Valves (Technical Specification Table 3.7-

1) from il% to i3%, and 2) remove the 4 safety valve inoperable option of TS Table 3.7-2. If, however, the setpoint is found outside of a il% tolerance band, the setting would be adjusted to within i1% of the specified lift setting.

The impact of increasing the Pressurizer Safety Valve (PSV) and Main Steam Safety Valve (MSSV) opening setpoint tolerance from il% to i3% on Waterford 3 Final Safety Analysis Report (FSAR) analyses was evaluated.

To offset some of the impact of the increase in setpoint tolerances on the analysis results (e.g., RCS and secondary peak pressure), improved (more realistic) PSV and MSSV opening characteristics were used in the analyses.

Technical Specifications 3.4.2 and 3.4.3 contain requirements for PSV operability with lift setting of 2500 psia 1%; Technical Specification 3.7.1.1 contains the MSSVs operability requirements with reference to the lift settings specified in Table 3.7-1, which allows a il% tolerance. The i1%

allowed tolerance on the PSV and MSSV has been occasionally exceeded during

.past surveillance testing. To establish a safe and reliable setpoint tolerance during plant operation and reduce the number of LER's that may result when valve setpoints are outside of required values, Waterford 3 pursued a change to increase the setpoint tolerance from il% to'i3%.

The events impacted by the proposed change have been evaluated. The evaluation results demonstrate safe plant operation with an increased safety valve tolerance.

The setpoint tolerance change impacts FSAR analyses with respect to RCS overpressurization, steam generator overpressurization, required overpower margin and peak clad temperature criteria. The following describes the events that have been analyzed due to the change in PSV and MSSV setpoint tolerance-from fl% to i3%.

. Increase in PSV and MSSV setpoint from +1% to +3%

This change, which increases the allowable PSV and MSSV opening setpoint by 2%, will adversely impact the peak RCS and peak secondary pressure for some analyzed events. The following events were evaluated with respect to peak RCS l

and peak secondary pressure with a simultaneous increase in PSV and MSSV tolerance from +1% to +3%:

. Loss of Condenser Vacuum (LOCV) with a single failure (SF)

The analysis of this event with a +3% PSV and MSSV tolerance resulted in peak RCS and peak secondary pressures that are within the acceptable limits (i.e., <110% of the RCS and design pressure and

<110% of the steam generator design pressure). However, the results are slightly higher than the analysis of record currently documented in the FSAR. The FSAR will be updated to document the new analysis as analysis of record.

. Feedwater Line Break (FWLB) events (large and small)

The large FWLB event resulted in peak RCS and peak secondary pressures that are bounded by the current analysis of record. The decrease in the primary and secondary peak pressure, despite an increase in the PSV and MSSV opening tolerances, are due to the use of more realistic PSV and MSSV characteristics. For the small FWLB event, the peak RCS pressure exceeded the analysis of record, but was within the acceptable limit. The peak secondary pressure was below the analysis of record.

The increase in MSSV lift pressure also adversely impacts the Required Overpower Margin (ROPM) for some CEA misoperation events. The increase in secondary pressure and temperature results in a lower primary to secondary heat transfer and in turn higher primary temperature. The higher primary temperature has an adverse impact on the CEA misoperation events in the presence of a positive moderator temperature coefficient (MTC). The MTC is a

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I major contributor to the severity of these events. The impact of the tolerance change on the CEA misoperation events are factored into the COLSS and CPCs setpoints. I l

The increase in MSSV lift pressure also adversely impacts the peak clad temperature during a small break LOCA (SBLOCA) event. The increase in MSSV l lift pressure results in a higher SG pressure and in turn higher RCS pressure j during the limiting SBLOCA event. The higher RCS pressure decreases the l safety injection flow and increases break flow, resulting in a higher peak l clad temperature. The limiting small break LOCA was analyzed by ABB-CE. The analysis resulted in a peak clad temperature higher than the result in the current analysis of record, but within the acceptable limit and lower than the peak clad temperature for the large break LOCA event. 1 The limiting event for the peak secondary pressure (LOCV) was analyzed with a MSSV opening setpoint tolerance of +3%. This event was analyzed with 1, 2, 3 and 4 MSSVs inoperable respectively, to confirm the validity of the Technical Specification Table 3.7-2, " Maximum Allowable Linear Power Level High Trip Setpoint With Inoperable Steam Line Safety Valves During Operation With Both Steam Generators." The analysis for the cases with 1, 2 and 3 MSSVs inoperable per operable SG, resulted in acceptable peak SG pressure, however, for the case with four inoperable MSSVs per operable SG, the secondary peak pressure slightly (1 psi) exceeded the peak SG pressure acceptance criteria (110% of the design pressure, 1210 psia). Based on the above, Technical Specification Table 3.7-2 is modified to remove the option for the four inoperable MSSVs.

. PSV setpoint tolerance change from -1% to -3%

This change does not adversely impact any of the previously analyzed events. Therefore, no event had to be reevaluated for this change.

The concern with the PSV opening at -3% of the nominal setpoint ,

(2425 psia) is that the PSV may open prior to, and interfere, with the l Pressurizer Pressure-High Reactor Trip, resulting in more severe  ;

consequences. The Pressurizer Pressure-High Reactor Trip Setpoint assumed in the analyses is the current Technical Specification limit plus a conservative instrument uncertainty based on the limiting accident conditions. The current Technical Specification limit for the Pressurizer Pressure-High Reactor Trip Setpoint is 2365 psia with an i Allowable Value of s 2372 psia. Additional separation between the PSV I opening and the Pressurizer Pressure-High Reactor Trip is provided by Technical Specification Change Request NPF-38-152, dated June 22,1994.

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NPF-38-152 proposed a Pressurizer Pressure-High Reactor Trip Setpoint of 2350 psia and an Allowable Value of s 2359 psia. Thus, sufficient separation exists between the minimum allowed PSV opening setpoint and the Pressurizer Pressure-High Reactor Trip Setpoint.

. MSSV setpoint tolerance change from -1% to -3%

This change primarily impacts the FSAR reported secondary steam release through the MSSVs due to the earlier opening of the MSSVs and the corresponding dose results. The impact of this change on all of the FSAR analyses were evaluated and found to be insignificant. The event that was impacted the most is the steam generator tube rupture (SGTR) concurrent with loss of offsite power. The total increase in offsite dose for this event is found to be about 0.22 Rem. This small increase

, in dose did not cause the results of this event to exceed the acceptance criteria of 10 CFR 100.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will the operation of the facility in accordance with these proposed changes involve a significant increase in the probability or consequence of any accident previously evaluated?

Response: No The proposed change does not involve any change to the physical characteristics of the PSVs and MSSVs and will have no impact on the PSVs and MSSVs as-left setting. This change only allows for a larger (i3% versus il%) as-found setpoint tolerance. Therefore, this change )

has no impact on the probability of occurrence of any accident previously evaluated. The impact of this change on the FSAR analyses I has been evaluated and the results of the impacted events have been l found to be within the acceptable limits.

Therefore, revising the PSV and MSSV as-found opening setpoint tolerance from 1% to i3% does not increase the probability or consequences of an accident previously evaluated.

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2. Will the operation of the facility in accordance with'these proposed changes create the possibility of a new or different kind of-accident from any accident' previously evaluated?

Response: No The proposed changes to the PSVs and MSSVs as-found opening setpoint tolerance do not modify. equipment or change the manner in which the plant will be operated. The safety valves will continue to function per their design. Since no hardware modifications.or changes in operation

, procedures will be made, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

~ 3. Will the operation of the facility in accordance with these proposed changes involve a significant reduction in the margin of' safety?

Response: No The impact of the proposed changes on the Waterford 3 FSAR analyses have been evaluated. The evaluation demonstrates that the results of the impacted events remained within the acceptable limits. The system capabilities to mitigate and/or prevent accidents will be the same as they were prior to these changes. Therefore, the proposed changes does not involve-a reduction in a margin of safety.

Safety and Sionificant Hazard 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.

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NPF-38-160 ATTACHMENT A i

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