ML20212L348

Forwards Document Re Replacement SG Water Level Trip Setpoint Differences to Be Placed in PDR

ML20212L348
Person / Time
Site:	South Texas
Issue date:	10/01/1999
From:	Alexion T NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
TAC-MA2500, TAC-MA2501, NUDOCS 9910080022
Download: ML20212L348 (6)

Text

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I October 1, 1999 MEMORANDUM TO: File FROM: Thomas W. Alexion, Project Manager, Section 1 ORIGINAL SIGNED BY Project Directorate IV & Decommissioning Division of Licensing Project Management

SUBJECT:

DISCUSSIONS WITH LICENSEE ON ITS FUTURE SUPPLEMENT TO APPLICATION FOR TECHNICAL SPECIFICATION CHANGES TO REFLECT REPLACEMENT STEAM GENERATOR WATER LEVEL TRIP SETPOINT DIFFERENCES (TAC NOS MA2500 AND MA2501)

The U. S. Nuclear Regulatory Commission (NRC) staff has had discussions with the licensee on its July 22,1998 (as supplemented by letter dated June 16,1999), application for a license amendment for technical specification changes to reflect replacement steam generator water level trip setpoint differences. The licensee is planning a supplement to the application and has requested further discussions with the NRC staff.

In order to facilitate future discussions between the licensee and the staff, the licensee provided the information in the attachment. The licensee has informed me that this information is preliminary and it plans to follow-up this information with a formal submittal. The purpose of this memorandum is to place the attachment in the Public Document Room.

Docket Nos. 50-498 and 50-499

Attachment:

As stated DISTRIBUTION: h kly . h hg EDocket Files (50-498 and 50-499)

PUBLIC PDIV-1 RF 10 receive a copy of Inis doc 0 hex I

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. . . . . ,# October 1, 1999 MEMORANDUM TOj File FROM: [ Thomas W. Alexion, Project Manager, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management

SUBJECT:

DISCUSSIONS WITH LICENSEE ON ITS FUTURE SUPPLEMENT TO APPLICATION FOR TECHNICAL SPECIFICATION CHANGES TO REFLECT REPLACEMENT STEAM GENERATOR WATER LEVEL TRIP SETPOINT DIFFERENCES (TAC NOS. MA2500 AND MA2501)

The U. S. Nuclear Regulatory Commission (NRC) staff has had discussions with the licensee on its July 22,1998 (as supplemented by letter dated June 16,1999), application for a license amandment for technical specification changes to reflect replacement steam generator water level trip setpoint differences. The licensee is planning a supplement to the application and has requested further discussions with the NRC staff.

In order to facilitate future discussions between the licensee and the staff, the licensee provided the information in the attachment. The licensee has informed me that this information is preliminary and it plans to follow-up this information with a formal submittal. The purpose of this memorandum is to place the attachment in the Public Document Room.

Docket Nos. 50-498 and 50-499

Attachment:

As stated

P ,

SEP-26-1939 13:16 NUCLEf-fi LICENSING NOC-AE-000657 P. 01/or'

( Attachment 1 Page 2 cf 5 Verbal Request for Additional Information Replacement Steam Generator Water Level Trip Setpoint Differences South Texas Project. Units 1 and 2

1. Attachment 1, Page 10 of13, 2" par, l' sentence says, "STPNOC does not believe the requirements of10CFR50.36(c)(2)(ii)(B) apply to STP because STP is excluded by the provisions of10CFR50.36(c)(2)(iii).

Mr. Weiss disagrees.

Response to VRAI 1.

This issue was resolved during a July 22,1999, meeting between STPNOC and NRC at NRC White Flint offices. The NRC's decision is expressed in an NRC letter (Reference 1), thus will not be repeated here.

2. Specipcation 3.4.4.a (RCS Relief Valves), says, "With one or both POR V(s) inoperable, because ofexcessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s);

otherwise, be in at least HOTSTANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andin HOT SHUTDOWN within thefollowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."

This allows the block valve upstream ofthepressurizer PORVs to remain closedfor an unlimited time. How can STP take creditfor thepressurizerPORVs in mitigating accidents?

Response to VRAI 2.

STP does not take credit for pressurizer PORVs in mitigating accidents. However, as with all control systems, STPNOC models Pressurizer PORVs in non.LOCA transient analyses if their function would tend to increase the severity of the transient. This is the reason that we model pressurizer PORVs in analyses of the loss of norinal feedwater (LONF) and feedline mpture (FLB) events.

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sTI: 30955646 Attachment i

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NOC-AE-000657 4 Attachment 1 Page 3 of 5 f

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[ 3. Attachment 1, Page 6 of13, shows Table 4, " Time Sequence ofEventsfor the FLB Event

\ with Maximum Reactivity Feedback and Offsite Power Available. " The last line ofthe j table lists, " Pressurizer Reaches Water 4elid Condition," and shows a corresponding time l of2796 seconds.

Respr,nse to VRAI 3.

2796 seconds is 46 minutes and 36 seconds. The event analysis calculates this value, and we provided it to demonstrate that it is significantly greater than 30 minutes (1800 seconds). The importance of this is that our Condition IV FLB analysis shows that the pressurizer does not go water solid during the first thirty minutes following initiation of the design basis transient.

This clearly demonstrates that STP operators have sufficient time following event initiation to prevent pressurizer overfill.

d. STPNOCsupplement to TSC-205, NOC-AE-000462, " Auto Action ofSGPORVs."

Response to VRAI 4.

(Question withdrawn by NRC.)

5. Address each ofthe three conditions listedin the conclusions section ofthe NRCstaffSER

.for WCAP-14882.

The NRCstafconcludes in the SERfor WCAP-14882 (Ref 3) thet the "use ofRETRAN as describedin WCAP-14882 is acceptableforlicensing calculations andRETRANmay be used to replace the LOFTRAN computer code in Westinghouse reload methodology provided that thefollowing conditions are met:

a. The transients and accidents that Westinghouseproposes to analyze with RETRANare listed in this SER (fable 1) and the NRC staffreview ofRETRAN usage by Westinghouse was limitei to snis set. Use ofthis codefor other analyticalpurposes will require additionaljushycation.

b. WCAP-14882 describes modeling of Westinghouse designed 4,3, and 2-loopplants of the type that are currently operating. Use ofthe code to analyze other designs, including the Wesdaghouse AN00, willrequire additionaljustsycation.

c. Conservative safety analyses using RETRANare dependent on the selection of consernadveinput. Acceptablemethodologyfordevelopingplant-specsficinputis L discussedin WCAP-14882 andin Reference 4. Licensing applications using RETRAN

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should include the source ofandjushycationfor the input data used in the analysis. "

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' l Page 4 of 5 Response to VRAI 5.

l Each of these conditions are addressed below, as they relate to the South Texas Project Model A94 Replacement Steam Generator Program.

a. The non-LOCA transients explicitly analyzed with RETRAN for this program include the following: feedwater system malfunctions, steam system piping failures, turbine trip, loss  !

of ofhite power, loss of normal feedwater flow, and feedwater system pipe break. All of I these events are listed in Table 1 of the SER; therefore, no additional justification is required.

b. The South Texas Project nuclear units are 4-loop, Westinghouse-designed, pressurized l water reactors that are currently in commercial operation. Therefore, no additional justificationis required.

c. The non-LOCA RETRAN analyses were' performed in accordance with the methodologies discussed in WCAP-14882 (Ref. 2) and WCAP-9272-P-A (Ref. 4).

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,SEP-28-1999 13:18 NUCLEAR LICENSING P.05/05 NOC-AE-000657 4 Attachment 1 Page 5 of 5 References

1. USNRC Letter " South Texas Project, Units 1 and 2 (STP)- Proposed Technical Specification (TS) Change on Replacement Steam Generator (SG) Water Level Trip Setpoint (TAC Nos.

MA2500 and MA2501)," Thomas W. Alexion (USNRC) to William T. Cottle (STPNOC),

dated August 26,1999

2. WCAP-14882 Revision 0, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," Huegel, D. S., et al., June 1997.

3. USNRC Letter, " Acceptance for Referencing of Licensing Topical Report WCAP-14882,

'RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analysis' (TAC NO. M99107)," Akstulewicz, F. (USNRC) to Sepp, H. (E, February 11,1999. ,

4. WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology," Bordelon, F. M.,

et al., Approved July 1985.

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i STI: 30958646 TOTAL P.05