ML20247F350
| ML20247F350 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/07/1998 |
| From: | Martin L HOUSTON LIGHTING & POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9805190294 | |
| Download: ML20247F350 (6) | |
Text
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Nuclear Operating Company 5
South TensIhyettE&ctreGeneratingStatkm PO Bar189 Mkdwtwth, Tem 77483 w.
May 7,1998 NOC-AE-000159 File No.: G20.02.01 1
G21.02.01 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498 and STN 50-499 Licensing Methodology for the Replacement Steam Generator Project STP Nuclear Operating Company plans on replacing the Unit 1 Steam Generators in March of the year 2000. Replacement of the Unit 2 Steam Generators is expected to occur in the year 2002. The Licensing effon associated with the Replacement Steam Generator (RSG) Project will include a number of Technical Specifications changes as well as amendments associated with Unreviewed Safety Questions.
The attachment entitled,"L.eensing Methodology for the Replacement Steam Generator Project,"
is an overview of the Licensing effort associated with the RSG project and is provided for Nuclear Regulatory Commission (NRC) review. The first Technical Specification change associated with the RSG Project will be submitted in the near future.
The Attachment constitutes the overall plan with respect to Licensing the Replacement Steam Generators. The NRC will be kept informed if any changes occur. There are no commitments to the NRC in this submittal.
If you have any questions regarding the attached plan, please contact either Scott Head at (512) 972-7136 or me at (512) 972-8686.
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Attachment:
Licensing Methodology for the Replacement Steam Generator Project i
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File No.: G20.02.01 G21.02.01 Page 2 Ellis W. Merschoff Jon C. Wood l
Regional Administrator, Region IV Matthews & Branscomb L
U. S. Nuclear Regulatory Commission One Alamo Center 611 Ryan Plaza Drive, Suite 400 106 S. St. Mary's Street, Suite 700 Arlmgton, TX 76011-8064 San Antonio,TX 78205-3692 Thomas W. Alexion Institute of Nuclear Power Project Manager, Mail Code 13H3 Operations - Records Center U. S. Nuclear Regulatory Commission 700 Galleria Parkway Washington, DC 20555-0001 Atlanta, GA 30339-5957 David P. Loveless Richard A.Ratliff Sr. Resident Inspector Bureau of Radiation Control c/o U. S. Nuclear Regulatory Commission Texas Department of Health P. O. Box 910 1100 West 49th Street Bay City, TX 77404-0910 Austin, TX 78756-3189 J. R. Newman, Esquire D. G. Tees /R. L. Balcom Morgan, Lewis & Bockius Houston Lighting & Power Co.
1800 M. Street, N.W.
P. O. Box 1700 Washmgton, DC 20036-5869 Houston,TX 77251 M. T. Hardt/W. C. Gunst Central Power and Light Company City Public Service ATTN: G. E. Vaughn/C. A. Johnson P. O. Box 1771 P. O. Box 289, Mail Code: N5012 San Antonio,TX 78296 Wadsworth,TX 77483 A. Ramirez/C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin,TX 78704 i
Attachment i
NOC-AE-000159 Page1 LICENSING METHODOLOGY FOR THE REPLACEMENT STEAM GENERATOR PROJECT BACKGROUND The STP Nuclear Operating Company (STPNOC) plans to replace the South Texas Project (STP) 1 Unit 1 Model E Steam Generators with Westinghouse A94 Steam Generators in March of the year 2000. Current plans are to replace the Unit 2 steam generators in the year 2002. *he majority of the work and analysis required to support this effort will be done in accordance with 10CFR50.59.
However, a number of issues, including a limited number of Technical Specification Changes and Unreviewed Safety Questions will require NRC approval. A general description and listing of critical aspects of these upcoming submittals are provided below. Also descriSc<1 is a number of the more imponant 10CFR50.59 evaluations that are currently undu.my or nearing completion.
It is important to note, that as currently planned, all amendment requests will be for BOTH Units.
The Westinghouse A94 Steam Generators are similar to the Westinghouse A75 Steam Generators installed in the V. C. Summer plant. Design aspects of the A94s, including the fact that a preheater is not used, result in a larger heat transfer area and a lower resistance to Dow than the Model E Steam Generators. A number of the Technical Specification changes noted below, including the change in Reactor Coolant System Flow, are a direct result of the differences between the two models of generators. The Unit 2 replacement Steam Generators will be identical to those on Unit 1.
Technical Specification Chances 1
Significant aspects of the proposed Technical Specification changes are:
The changes are made " Steam Generator" specific instead of Unit specific. This will allow the Technical Specification changes to be applicable to either unit and will obviate the need for Technical Specification amendment requests for the Unit 2 replacement effort.
The analysis performed to support the changes associated with Reactor Coolant System flow, as well as other analysis performed to support the replacement effort, rely on the use of the Westinghouse RETRAN model. A Topical Report (WCAP 14882) was submitted to the NRC on June 6,1997 (NSD-NRC-97-5144) which details the model development. and qualification work performed by Westinghouse tojustify the RETRAN-02 code as the system transient code for licensing-basis non-LOCA safety analyses using NRC previously approved Westinghouse methodology. STPNOC has submitted an application for NRC review of this l
I topical report in a letter from T.H. Cloninger to Document Control Desk dated May 7,1998 (ST-NOC-AE-0156).
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NOC-AE-000159 Page 2 i
As currently dnvisioned, the proposed changes consist of:
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'e JReactor Coolant System Flow - The A94 steam generators have a lower resistance to flow.
As a result, total RCS flow rate will increase by a small amount. This increase is sufficient, if left uncompensated, to raise fuel lift off forces beyond design limits. Two changes are being made to accommodate this increased flow. First, the upper head flow nozzle plugs are being removed which will increase the flow of cold-leg water into the upper head region.
l-This has the added benefit of lowering the upper head temperature with a corresponding reduction in susceptibility for cracking of the Alloy 600 upper head penetrations. Second, thimble plugs will be removed from the fuel elements which result in an increase in core bypass flow and therefore decreases the differential pressure across the fuel assemblies.
. Safety Limit Figure - The Safety Limit curves are a function of core temperature, power, pressure and flow. Since the flow inemases with the new steam generators, a Safety Limit Figure has been added that is specific to the A94 steam generators. This change will be
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included as part of the RCS flow change noted above.
Steam Generator level Setooints - Current low-low and high-high steam generator water level trip setpoints are 33% and 87.5% of narrow range span, respectively.
Due to operational, as well as design considerations, these Trip Setpoints have been changed to 20%
and 83.5%, respectively. Auxiliary Feedwater Actuation will also occur at the new 20%
water level. Appropriate allowable values are also included.
Steam Generator Surveillance - A number of previously approved changes to the steam i
generator surveillance program, such as F*, voltage based repair criteria, and the use of laser welded sleeves, are not applicable to the A94 steam generators.
The surveillance requirements are revised to show non-Applicability to the A94 generators.
' Reactor Coolant System Volume - The Reactor Coolant System (RCS) volume specified in e
Section 5 of the Technical Specification is not applicable' to the A94 steam generators. A change will be proposed to move this descriptive information to the' UFS AR. This change is consistent with the NUREG 1431, " Standard-Technical Specification for Westinghouse Plants," Revision 1.
It should be noted that STPNOC has previously submitted a proposed Technical Specification change related to the automatic operation of the Steam Generator Power Operated Relief Valves (letter from T.H.
Cloninger to Document Control Desk, dated August 18,1997, L ST-HL-AE-5689). This more restrictive change is applicable to both the Model E and A94 steam generators. It is'our understanding that this proposed change is currently on schedule to be approved this summer.
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Attachment NOC-AE-000159 y
Page 3 UNREVIEWED SAFETY OUESTIONS The analysis performed to support the replacement steam generator project has identified three Unreviewed Safety Questions that will need NRC approval prior to implementation.
Specifically:
e _ Increased dose associated with Main Steam Line Break (MSLB) - The radiological impact of the replacerrent of the Model E steam generators with the A94 Model steam generators will be submitted to the Commission. As a parallel effort, STPNOC is proposing to lower the feedwater temperature of the existing Model E steam generators to allow for long-term I
operation with one feedwater heater bypassed. Because of the shared analyses and required submittals, these proposed changes have been combined into one submittal. Effects on the I
resultant doses due to the physical changes to the plant are slight. However, this opportunity was taken to improve the analytical models used in the analyses, specifically with regard to steam generator tube uncovery during a MSLB. This added conservatism causes the bulk of the increases in offsite doses. All results remain within the acceptance criteria for the respective accidents, e.g., the Exclusion Zone thyroid dose increased from 0.963 rem to 1.37 rem for the MSLB.
STPNOC has conservatively assumed that these dose increases constitute an Unreviewed Safety Question to ensure that the current industry issues surrounding the 10CFR50.59 process do not delay the RSG schedule.
Small Break LOCA (SBLOCA) - The analysis performed on the SBLOCA indicates that i
manual operation of the steam generator PORV's will be required approximately 45 minutes into the accident.' This is consistent with similar requirements for operator actions in other accident scenarios.
Large Break LOCA Mass and Energy Releases - The methodology to calculate the mass and
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energy release for containment pressure and temperature for equipment qualification and
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containment structural integrity analysis during a LOCA is being revised. Currently, the i
mass and energy releases during the blowdown, refill and reflood phases _ up to the point L
- where suction for the safety injection system is switched from the Refueling Water Storage Tank to the containment sump is calculated using the methodology described in the WCAP 8264-P-A, Revision 1, " Westinghouse Mass and Energy Release Data for Containment j
l Design." The new analysis calculates the mass and energy releases during the blowdown, refill and reflood phases up to the point where the steam generators are cooled down and L
depressurized using the methodology described in WCAP-10325-P-A, " Westinghouse: LOCA l
Mass and Energy Release Model for Containment Design, March 1979 version." The NRC has previously reviewed and approved the methodology in WCAP-10325-P-A for this type of application. Note the change does not apply to the containment back pressure calculations used in the 10CFR50.46 analysis. The methodology to satisfy 10CFR50.46 requirements remains the same.
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Page 4 SAFETY AND LICENSING REPORT Details of the safety evaluations performed to support the replacement effort, including those noted above, will be detailed in the Safety and Licensing Report. This document is similar to the report submitted by South Carolina Electric & Gas for the V.C. Summer plant to support their replaceme.nt effort. This document is currently scheduled to be completed by the end of October, 1998, at which time it will be submitted to the NRC. This document is intended to provide details and the bases for changes made as a result of the replacement effort. No specific action is required or requested of the NRC in response.
CHANGES MADE IN ACCORDANCE WITH 10CFR50.59 Except as noted above, all other changes associated with the replacement steam generator effort will be made via analysis performed under 10CFR50.59. Examples of 10CFR50.59 evaluations being performed as part of the replacement steam generator effort are:
The Steam Generator Tube Rupture analysis has been performed using the Westinghouse RETRAN methodology described in WCAP 14882. The 10CFR50.59 analysis assumes that L
the NRC will approve the use of this WCAP.
The MSLB mass and energy releases are based on analysis performed using the e
Westinghouse RETRAN methodology described in. WCAP 14882.
The 10CFR50.59 analysis assumes that the NRC will approve the use of this WCAP.
The removal of the upper head flow plugs changes the characteristics of the plant and allows for a more rapid cooldown following a 1.oss-of-Offsite-Power event. This more rapid cooldown will be used in the analysis performed to support the water volume requirements of the Auxiliary Feedwater Storage Tank (AFWST).
As a result, the current Technical Specification requirement for AFWST volume is adequr.te.
-SCHEDULE The current schedule for planned submittals is as follows:
' RCS Flow Technical Specification Change-May 1998 e
Safety Limit Figure Technical Specification Change May 1998 e
, steam Generator Water Level Technical Specification Change June 1998 e
Reactor Coolant System Volume Technical Specification Change June 1998 e
SBLOCA Unreviewed Safety Question June 1998 e
Dose Related Unreviewed Safety Question June 1998 e
Surveillance Technical Specification Change June 1998 e
LOCA Mass and Energy Unreviewed Safety Question September 1998 To allow for timely implementation of the changes, STPNOC will request the NRC complete its review and approval of the required items by November 1999.
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