ULNRC-05287, CFR 50.59 Summary Report

CFR 50.59 Summary Report

ML061450122
Person / Time
Site:	Callaway
Issue date:	05/18/2006
From:	Keith Young AmerenUE, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05287
Download: ML061450122 (9)

Text

AmerenUE PO Box 620 Cal/away Plant Fulton, MO 65251 May 18, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 ULNRC-05287 WAmerenI Ladies and Gentlemen:

UF DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT I UNION ELECTRIC CO.

FACILITY OPERATING LICENSE NPF-30 10 CFR 50.59

SUMMARY

REPORT In accordance with 10 CFR 50.59( dX2), this letter transmits a report which summarizes the evaluations performed pursuant to 10 CFR 50.59(c)(1) for changes, tests, and experiments approved and implemented for activities at Callaway Plant.

This report covers all 10 CFR 50.59 evaluations that were implemented from July 1, 2004 through December 30, 2006.

This letter does not contain new commitments.

Sincerely, Keith D. Young Manager - Regulatory Affairs Enclosure 1~4 a subsidiary of Ameren Corporation

ULNRC-05287 May 18,2006 Page 2 Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-7D1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street PO Box 360 Jefferson City, MO 65102-0360 Mr. Ron Reynolds Director Missouri State Emergency Management Agency P.O. Box 116 Jefferson City, MO 65102

DOCKET NO. 50-483 w"AmereflUE UNION ELECTRIC COMPANY CALLAWAY PLANT 10CFR 50.59

SUMMARY

REPORT July 2004 - December 2005

CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 EXECUTIVE

SUMMARY

In accordance with 10 CFR50.59 (d)(2), a summary report has been prepared, which provides summaries of the 10 CFR 50.59 evaluations of changes, tests, and experiments approved and implemented for activities at Callaway Plant.

This report covers all 10 CFR 50.59 evaluations that were implemented from July 1, 2004 through December 31, 2005. During this period there were 3 changes implemented that required a 10 CFR 50.59 evaluation.

1

CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 REFERENCE/ABBREVIATION KEY CN - FSAR Change Notice.

MODIFICATION PACKAGES (Design Changes)

• CMP - Callaway Modification Package

• MP - Modification Package RFR - Request for Resolution CARS - Callaway Action Request System TM - Temporary Modification 2

CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 CN 04-056 Identification of a failure mode not previously described in the FSAR.

Activity

Description:

A change to the FSAR was made to revise the described effect of either of two particular breakers failing closed in the plant electrical power supply system, as addressed in the failure modes and effects analysis (FMEA) summarized in FSAR Table 8.3-4. Although no physical changes to the plant were made, the change was to an FSAR described design function, i.e., the onsite electrical power system's designed capability to tolerate electrical component failures per the FMEA presented in Table 8.3-4, which was done in accordance with IEEE 308-1974. In particular, the total system effect of either of the subject breakers failing closed was not adequately described in the FSAR. The FSAR change provided a more complete description of the effect which, when more accurately described, had a greater (though still acceptable) impact than what was described, since the subject breaker failure would impact both 4.16-kV safety buses (but does not result in a loss of power to both buses). The proposed change thus incorporated a failure mode not previously evaluated and described in the FSAR. The change was determined to be adverse and thus required evaluation pursuant to 10 CFR 50.59.

Summary of Evaluation:

The particular breaker failure that was identified is more severe than other component failures described in FSAR Table 8.3-4. The FMEA design objective of ensuring that auxiliary power system component failures do not prevent satisfactory performance of plant Class IE loads required for safe shutdown of the facility is still met. This is based on the fact that the subject breaker failure would not cause a loss of power to both 4.16-kV safety buses. Likewise, if the failure is assumed to occur with a design-basis accident, one train of required safety equipment would still be available to perform required functions, as supported with electric power from the associated diesel generator. The subject breaker failure may be compared to or associated with a loss of offsite power (LOOP), but analysis using a conservative frequency of occurrence for the subject breaker failure (which is a highly unlikely failure mode and is assumed to occur without cause) demonstrates that this failure is an insignificant contributor to LOOP frequency. The subject breaker failure would not physically or directly affect other plant equipment except with regard to electric power support which has been satisfactorily evaluated. No new or additional methodology was required to identify or evaluate the expanded breaker failure effect since the effect is based on the existing FMEA.

3

I CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 CN 05-012 Reduction in the minimum Aux Feedwater flow assumed for the LONF and LOAC events Activity

Description:

This activity reduced the minimum Auxiliary Feedwater (AFW) System flow assumed for the Loss of Normal Feedwater Flow (LONF) and Loss of Non-Emergency AC Power to the Station Auxiliaries (LOAC) events described in the FSAR. Specifically, the change is a reduction in assumed AFW flow from 960 gpm to all four steam generators (see Amendment 168) to 800 gpm to all four steam generators.

Summary of Evaluation:

The 10 CFR 50.59 Evaluation that was completed found that the reduction in AFW flow from 960 gpm to all four steam generators (see Amendment 168) to 800 gpm to all four steam generators for-the LONF and LOAC events does not require prior NRC approval.

This change is a revision to an input parameter for the LONF and LOAC analyses, but is not a change to the analyses methodology itself. The methodology on which the LONF and LOAC analyses are based was described in License Amendment Requests (LARS) for Amendments 168 and 170 and was thus reviewed and approved by the NRC per those license amendments. The proposed change did not result in a departure from the methods of evaluation used in establishing the design bases or the results of the safety analyses. This change did not impact any key input assumptions that determine the radiological consequences per the analyses, nor did it impact the calculational methodology in the radiological analyses for the accident sequences. The radiological consequences calculated per the accident analyses remained unchanged. Furthermore, this analytical change involved no physical changes to the facility that could introduce a new malfunction or new accident not previously evaluated, or result in an increase in the likelihood of occurrence of a malfunction or accident previously evaluated. The evaluation concluded that all acceptance criteria for the applicable accidents continue to be met with no increase in the consequences of the accidents or malfunctions previously evaluated.

4

CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 MP 05-3077 Revised TM Average Scaling (RTAS)

Activity

Description:

RCS Hot leg temperature measurement consists of inputting three RTD signals into a summing amplifier that provides an output equivalent to the average of the three inputs.

Westinghouse developed the RTAS method for assigning different gain values to the individual RTD signals to reduce the impact on plant operation from periodic temperature fluctuations caused by the Upper Plenum Anomaly (UPA). A modification for implementing the RTAS method was thus developed. The physical change to the plant that was made for this modification was the replacement of two (of three total) input resistors on the 7300 series summing amplifier card for RCS loop 2 with different resistance values to obtain the desired gain values calculated by Westinghouse. The previous configuration assigned an equal gain to all three input signals (33.333%) by using 50K ohm resistors. The new RTD gain values used resulted in a lower indicated average temperature in RCS loop 2 relative to the other three loops and increased the loop's operating margin to the OTDT/OPDT protective functions. MP 05-3077 directed recalibration of the Delta-T, T' and T" values to loop specific values following implementation of the revised RTD gains to ensure that the margin available to the OTDT/OPDT setpoints in each loop is consistent with the supporting safety analyses.

Summary of Evaluation:

This evaluation of RTAS for RCS loop 2 concluded that the proposed modification could be performed without prior NRC approval. The RTAS modification affected the T-Hot indication for Loop 2. This provided increased margin to the OTDelta-T and OPDelta-T reactor trips during UPA events. Since the RTAS modification made no physical change to the reactor core, or to the allowable primary or secondary side contamination levels allowed by the DEI-131 Technical Specification, it was determined that the modification does not adversely impact accident consequences described in the FSAR. Implementation of the RTAS modification does not increase the likelihood of accidents evaluated in the FSAR.

The constants used in the Delta-T trip setpoint calculations for Callaway Plant are loop-specific. As a part of the implementation of the Loop 2 RTAS modification, the calculations for Callaway Plant were loop-specific and the loop-specific constants were rescaled as necessary. This ensured that the change in indicated T-Hot for loop 2 does not result in the loss of redundancy or reliability for the loop 2 Delta-T trips. Therefore, implementation of the RTAS modification will not adversely impact the likelihood or consequences of equipment failure described in the FSAR.

5

4 1,.

CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 12-May-06 The analysis performed in support of the RTAS modification did not involve a deviation from the methodologies described in the FSAR. Following implementation of the RTAS modification, the Delta-T trip functions continue to have the capability credited in Callaway's Licensing Bases safety analysis. Therefore, implementation of the RTAS modification did not adversely impact any design basis limit for fission product barriers as described in the FSAR.

6