ML20107B685

Forwards Response to Request for Addl Info Re Proposed Change to TS on Setpoint Tolerances & TS Table 3.7-1

ML20107B685
Person / Time
Site:	Catawba
Issue date:	04/10/1996
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20107B688	List: ML20107B685
References
NUDOCS 9604170032
Download: ML20107B685 (6)

Text

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DukePbwerCornpany M.S ncowv P.O. Box 1006 Senior Vice hesident Owlotte,NC282011006 NuclearGeneration (704)382-2200 OWice (704)3824360 Fat DUKEPOWER April 10, 1996 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

Catawba Nuclear Station Docket Numbers 50-413 and -414 Proposed Change to MSSV Technical Specification Setpoint Tolerances; Supplemental Information j

1 By letter dated November 15, 1995, Duke Power Company j

submitted proposed changes to main steam safety valve setpoint tolerance values.

The NRC staff responded with a request for additional information (RAI) dated March 19, 1996.

Attached are responses to the RAI.

If there are any questions, or more information is needed, please call Scott Gewehr at (704) 382-7581.

1 Very truly yours, o Ec M. S. Tuckman l

xc:

Mr. R. E. Martin, Project Manager Office of Nuclear Reactor Regulation 3

U. S. Nuclear Regulatory Commission j

Mail Stop 14H25, OWFN Washington, D. C.

20555 Mr. S. D. Ebneter, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 R.

J.

Freudenberger Senior Resident Inspector Catawba Nuclear Station 1110039 1

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Question 1 Provide the details of your reanalyses for the transients and accidents that are affected by the proposed change of MSSV setpoint tolerance including the following:

a) Major transient curves for each of the events 3

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reanalyzed including primary and secondary pressure, DNBR, steam generator water level, subcooling margin, i

etc.

l b) Major assumptions used in these analyses (especially the ones different from the analyses in record).

i

Response

With the exception of the steam generator tube rupture event, all of the transients which could potentially I

challenge the main steam safety valves were previously i

analyzed with the increased setpoint tolerance in anticipation of a planned Technical Specification amendment.

Therefore, the details of these analyses, including the major assumptions and transient curves, are already in place in Chapter 15 of the Catawba FSAR.

The affected transients are listed below, along with the j

applicable acceptance criteria.

The tube rupture transient is discussed separately in the response to Question 4.

t 15.2.3 -

Turbine trip peak primary and peak

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secondary pressure i

15.2.6 -

Loss of non-emergency AC power - long-term core cooling capability j

15.2.8 -

Feedwater system pipe break - short and long-term core cooling capability 15.4.2 -

Uncontrolled bank withdrawal at power -

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short-term core cooling capability (DNBR) 15.6.5 -

Small-break LOCA - peak clad temperature s

Question 2 Provide the peak clad temperature following a SBLOCA from your reanalysis and compare it with your analysis of record.

Response

As discussed in the response to Question 1, the SBLOCA analysis of record already incorporates the increased main steam safety valve setpoint tolerance of i3%.

The peak clad temperature results are summarized in FSAR Table 15-39.

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1 Question 3 Provide the amount of dose release following a SGTR from your reanalysis and compare them with your analysis of record.

Respong 2

The NRC-approved transient analysis methodology for the steam generator tube rupture event (Section 7.2.2 of DPC-1 NE-3002) specifies that the main steam code safety valves are modeled with lift, accumulation, and blowdown assumptions which maximize secondary pressure.

These i

assumptions conservatively delay the operator identification and closure of the failed-open steam line PORV.

As can be seen from Figure 3-1 at the end of this attachment, a large fraction (~90%) of the total steam j

release from the ruptured generator is through this failed PORV.

This phenomenon dominates the small l

decrease in primary-to-secondary leakage resulting from the higher secondary system pressure.

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An evaluation of the offsite dose release was performed I

to determine the impact of the higher MSSV setpoint i

tolerance.

This evaluation also utilized a more explicit j

modeling of the tube bundle uncovery in the ruptured steam generator.

The revised model predicts that the tube bundle will remain covered for the duration of the j

transient.

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The net effect of these two modifications is a increase in the whole body dose and a significant decrease in the I

thyroid dose.

This result illustrates the impact of tube f

bundle uncovery, which affects only the partitioning of the iodine release and not the noble gasses.

i FSAR 15.6.3 Evaluation 10CFR100 Results Results Limit Whole body dose:

0.198 0.273 2.5 Thyroid dose:

?

l (pre-existing iodine spike) 82.6 61.1 300 Question 4 j

Confirm that the methodology used for reanalyses of each event are consistent with that used in the original licensing analyses.

Identify any differences in methodology and provide justification.

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Response

j The methodology used in the reanalysis of the above non-i LOCA events is consistent with the NRC approved methodology described in topical reports DPC-NE-3000 and DPC-NE-3002.

Any identified differences from the original approved methodology have been submitted to and approved by the NRC by SERs dated December 27th and 28th of 1995, respectively.

Question 5 Provide clarification for the changes made to TS Table 3.7-1.

Response

The change to Table 3.7-1 as submitted included a typographical error.

The intent of the change was to-remove both the footnote and the final line of the table which allowed operation in Mode 3 with 4 or 5 inoperable MSSVs.

A replacement page for that Technical Specification change is included as Attachment 2.

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1 3% drift,3% accumulation, 3% blowdown a

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-3% drift,1% accumulation,10% blowdown 3

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1000 1500 2000 2500 3000 3500 4000 4500 5000 Time (Seconds)

Revised Technical Specification Table 3.7-1 l

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