ML20216D433
| ML20216D433 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/11/1998 |
| From: | Alexion T NRC (Affiliation Not Assigned) |
| To: | Cottle W HOUSTON LIGHTING & POWER CO. |
| References | |
| TAC-MA0614, TAC-MA0615, TAC-MA614, TAC-MA615, NUDOCS 9803170044 | |
| Download: ML20216D433 (5) | |
Text
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NUCLEAR REGULATORY COMMISSION
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QM wasumorow, o.c. sonas.eem March 11,1998 -
Mr. William T. Cottle President and Chief Executive Officer STP Nuclear Operating Company South Texas Project Electric Generating Station P. O. Box 289 Wadsworth,TX 77483
SUBJECT:
UPDATE TO TECHNICAL SPECIFICATION BASES SECTION 3/4.4.9, SOUTH TEXAS PROJECT, UNITS 1 AND 2 (STP) (TAC NOS. MA0614 AND MA0615)
Dear Mr. Cottle:
The purpose of this letter is to respond to Houston Lighting & Power Company's (HL&P, the former licensed operator) November 5,1997, letter which requested an update to Technical Specification (TS) Bases Section 3/4.4.9, " Pressure / Temperature Limits," provided a proposed mark-up of the current Bases page, and provided background information. In addition, this letter also responds to STP Nuclear Operating Company's (STPNOC, the current licensed operator)
November 20,1997, follow-up letter which informs the Nuclear Regulatory Commission (NRC) th
.ases Section 3/4.4.9 has been changed pursuant to 10 CFR 50.59 and provides the NRC we. a copy of the revised page. The latter letter did not request any approval or response from NRC However, this letier is provided to confirm that NRC agrees with the proposed changes and to clarify the background presented in the November 5,1997, letter.
Amendments 31 and 22 were approved on November 8,1991, in response to an application dated December 21,1990, as supplemented by letter dated October 15,1991. Thess amendments approved a portion of HL&P's proposed TS and Bases changes (the unapproved portion of the application was revised by HL&P and approved later in Amendments 55 and 44).
Amendments 31 and 22 revised TS 3.4.9.3, _* Reactor Coolant System, Overpressure Protection Systems," to permit full operability testing of inoperable pressurizer power-operated relief valves (PORVs) following maintenance on the PORVs. The revised TS says, in part, that during this test period, cold overpressure protection will be provided by the residual heat removal (RHR) discharge relief valves. The related Bases pages were not revised with these amendments.
This was done later in Amendments 55 and 44, as discussed below.
Amendments 55 and 44 were approved on October 7,1993, in response to an application dated
' August 10,1992, as supplemented by letter dated September 14,1993. These amendments revised TS 3/4.4.4, " Reactor Coolant System, Relief Valves," TS 3/4.4.9, " Reactor Coolant System, Pressure / Temperature Limits," and Bases Section 3/4.4.4, " Reactor Coolant System.
Relief Valves." There were no revisions to Bases Section 3/4.4.9," Reactor Coolant System
)
(Pressure / Temperature Limits Subsection)," Amendments 55 and 44 also revised Bases Section i
3/4.4.9, " Reactor Coolant System (Low Temperature Overpressure Protection Subsection)," to add details concoming the use of the RHR relief valves to provide cold overpressure protection g{i 9903170044 990311 J
NRC IR.E CENTER COPY
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William T. Cottle during the stroke testing of inoperable PORVs consistent with TS 3.4.9.3, as approved in Amendments 31 and 22. However, this Bases revision inadvertently contained incorrect information as discussed below.
HL&P's application, dated December 21,1990, proposW, in part, revisions to TS 3/4.9.3 and revisions to the related Bases. The proposed related Bases to TS 3/4.9.3 said, in part, that the RHR relief valves will actuate at a reactor coolant system (RCS) pressure of 485 psig. However, in response to NRC questions, HL&P's October 15,1991, supplement indicated that the RHR relief valve will open when the RCS pressure is about 589 psig. This discrepancy is believed to be the reason why Amendments 31 and 22 were issued without the related Bases revision. It is also noted that the NRC's Safety Evaluation, related to Amendments 31 and 22, confirmed the
' 589 psig value.
HL&P's application, dated August 10,1992, proposed, in part, changes to the related Bases to TS 3/4.9.3 to be consistent with the TSs in Amendments 31 and 22. However, the Bases proposed by HL&P inadvertently contained the incorrect 485 value, and the NRC inadvertently approved the incorrect 485 value as being consistent with operation as described in the Safety
- Evaluation related to Amendments 31 and 22. Therefore, the 485 value needs to be corrected.
In the HL&P and STPNOC letters dated November 5 and 20,1997, respectively, the revised Bases states that the analysis conservatively demonstrates that the RHR relief valves limit RCS pressure to approximately 590 psig, and detail regarding the RHR pump's head, flowrate and pressure has been deleted. The NRC staff asked for the supporting calculation and the licen,ee provided Calculation No MC-6305, "RHR Relief Valves as COMS/LTOP [ Cold Overpressure Mitigation System / Low-Temperature Overpressure Protection] " Revision 1, dated November 20, 1991. To support the Bases chan9e the licensee used a general energy equation and considered the RHR relief valve setpoint of 600 psig with 10% accumulation (660 psig), RHR pump head, friction losses, flow velocity heads, and elevation heads. From these considerations, the derived RCS pressure upstrean1 of the RHR isolation valves was calculated to be 588.8 psig. The NRC staff has reviewed the calculation and agrees that h supports this change to the Bases. The staff also notes that this Bases change is consis'.ent with the discussion in the Safety Evaluation related to Amendments 31 and 22, and agrees that the other details regarding the RHR pump are unnecessary and can be deleted.
The November 5 and 20,1997, letters also replaces the phrase in the Bases which states, in part, that use of the RHR relief valves will maintain the RCS pressure below the low temperature
" endpoint of the Technical Specification limit curve (550 psig, Ref. Technical Specification Fig.
3.4-2)," and replaces the quoted portion above with " limits of ASME Section ill, Appendix G."
The NRC staff agrees with this change as well since TS Figure 3.4-2 is still applicable at all times and since the ASME Boiler and Pressure Vessel Code, Section ill, Appendix G limits are defined in Bases Section 3/4.4.9.
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Wsiliam T. Cottle Enclosed, for STPNOC's information, is revised TS Bases page B 3/4 4-15 that the NRC staff will use to update NRC's copy of the Bases. The revised page contains marginallines indicating the areas of change. The corresponding overieaf page is also provided to maintain document completeness. In addition, please notify the NRC if STPNOC does not agree with NRC's version of the background and discussion of how this Bases change came about, as set out above.
l Sincerely, l
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Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects ill/IV Office of Nuclear Reactor Regulation Docket Nos. 50-4g8 and 50-4g9
Enclosure:
TS Bases page B 3/4 4-15 cc w/ encl: See next page l
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i William T. Cottle Enclosed, for STPNOC's information, is revised TS Bases page B 3/4 4-15 that the NRC staff will use to update NRC's copy of the Bases. The revised page contains marginallines indicating the arear, of change. The corresponding overleaf page is also provided to maintain document completeness. In addition, please notify the NRC if STPNOC does not agree with NRC's version of the background and discussion of how this Bases change came about, as set out above.
Sincerely, l
ORIGINAL SIGNED BY:
Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Docket Nos. 50-498 and 50-499
Enclosure:
TS Bases page B 3/4 4-15 I
cc w/ encl: See next page DISTRIBlJilON:
Docket File PUBLIC PD4-1 r/f WBeckner EAdensam (EGA1)
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Mr. William T. Cottle STP Nuclear Operating Company South Texas, Units 1 & 2 oc:
Mr. David P. Loveless Jack R. Newman, Esq.
Senior Resident inspector MorDan, Lewis & Bockius U.S. Nuclear Regulatory Commission 1800 M Street, N.W.
P. O. Box 910 Washington, DC 20036-5869 Bay City, TX 77414 Mr. Lawrer.co E. Madin A. Ramirez/C. M. Canady Vice President, Nuc. Assurance & Licensing City of Austin STP Nuclear Operating Company Electric Utility Department P. O. Box 289 721 Barton Springs Road Wadsworth,TX 77483 Austin,TX 76704 Office of the Govemor Mr. M. T. Hardt ATTN: John Howard, Director Mr. W. C. Gunst Environmental and Natural City Public Service Board Resources Policy P. O. Box 1771 P. O. Box 12428 San Antonio, TX 78293 Austin,TX 78711 Mr. G. E. Vaughn/C. A. Johnson Jon C. Wood Central Power and Light Company Matthews & Branscomb P. O. Box 289 One Alamo Center Mail Code: N5012 106 S. St. Mary's Street, Suite 700 Wadsworth, TX 74463 San Antonio, TX 78205-3692 INPO Arthur C. Tate, Director Records Center Division of Compliance & Inspection 700 Galleria Parkway Bureau of Radiation Control Atlanta, GA 30339-3064 Texas Department of Health 1100 West 49th Street Regional Administrator, Region IV Austin, TX 78756 I
8? S. Nuclear Regulatory Commission 613 Ryan P aza Drive, Suite 400 Jim Calloway M ngton, TX 76011 Public Utility Commissica of Texas Electric Industry Analysis D. G. Tees /R. L Balcom P. O. Box 13326 Houston Lighting & Power Co.
Austin, TX 78711-3326 P. O. Box 1700 Houston,TX 77251 Judge, Matagorda County Mstagorda County Courthouse 1700 Seventh Street Bay City, TX 77414
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BASES i
LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) overshoot beyond the PORV Setpoint which can occur as a result of time delays in sigrkal processing and valve opening, instrument uncertainties, and single failure..To ensure that mass and head input transients more severe than those assumed cannot occur. Technical Specifications require lockout of all high head sdr.y - Mon pumps while in MODE 5 and MODE G with the reactor vessel head on. All bat om ogp <.Md safety injection pump are required to be locked out in MODE 4. Technics asimiu3 also require lockout of the positive displacement pump and all but one chargir.g pump while in MODES 4,5, and 6 with the reactor vesse' nead installed and disallow start of an RCP if secondary temperature is more l
than 50*F above primary temperature.
Administrative controls and two RHR relief valves wi4 be used to provide cold overpressure protection (COMS) during the ASME stroke tasting of two administratively declared inoperable PORVs. During the performance of the PORV function test, two RHR trains will be OFERABLE and in operation with the auto closure interlock bypassed (or deleted) to provide COMS.
~ With one PORV inoperable, COMS will be provided during the ASME test by the OPERABLE PORV and one RHR relief valve associated with an OPERABLE and operating RHR train which has the auto closure interlock bypassed (or deleted). Each RHR relief valve provides sufficient capacity to relieve the flow resulting from the maximum charging flow with concurrent loss of letdown. Analysis conservatively demonstrates that the RHR relief valves limit RCS pressure to approximately 59t nig.
Therefore two OPERABLE and operating RHR trains at one OPERABLE PORV and one OPERABLE and operating RHR train will provide adequate and redundant overpressure protection. Use of the RHR relief valves will maintain the RCS pressure below the low temperature limits of ASME Section lil, Appendix G.
l With regard to the MODE 6 applicability of this Technical Specification, the statement
.th the head on the reactor vessel" means any time the head is installed witn or without
>.nsioning the RPV studs.
The Maximum Allowed PORV Setpoint for the COMS will be updated based on the results of examinations of reactor vessel material irradiation survel' lance specimens performed
- as required by 10 CFR Part 50, Appendix H.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1,2, and 3 components ensure that the structuralintegrity and operational readiness of these components will be maintained at an acceptable level
[ continued on next page) p SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-15 Unit 1 - Amendment No. 4;0666 Unit 2 - Amendment No. 97;44 Revised by letter dated: March 11, 1993