NOC-AE-03001618, Revised Response to Request for Additional Information Regarding a Proposed Amendment to Technical Specification 3.4.2.2

Revised Response to Request for Additional Information Regarding a Proposed Amendment to Technical Specification 3.4.2.2

ML032760236
Person / Time
Site:	South Texas
Issue date:	09/30/2003
From:	Parkey G South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-03001618, STI: 31660090
Download: ML032760236 (3)

Text

Nuclear Operating Company (Z777 7

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SoShh kws PrcEkdikc net Sta/on PO A828W9 Wahmflh. Texas 77483-A September 30, 2003 NOC-AE-03001618 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units I and 2 Docket Nos. STN 50-498 and STN 50-499 Revised Response to Request for Additional Information Regarding a Proposed Amendment to Technical Specification 3.4.2.2

Reference:

Letter, G. L. Parkey to NRC Document Control Desk, "Response to Request for' Additional Information Regarding a Proposed Amendment to Technical Specification 3.4.2.2," dated September 10, 2003 (NOC-AE-03001592)

This letter provides a revised response to the second question presented in the referenced letter.

We have provided additional information based on further discussion with the NRC staff. If there are any questions regarding this response, please contact Mr. Scott Head at (361) 972-7136.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on September 30.2003 /,.,

&L. Par. ey Vice President, Generation jtc

Attachment:

Revised Response to Request for Additional Information OD I STI: 31660090

NOC-AE-03001618 Page 2 of 2 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U.S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 L. D. Blaylock City Public Service U. S. Nuclear Regulatory Commission David H. Jaffe Attention: Document Control Desk U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike R. L. Balcom Rockville, MD 20852 Texas Genco, LP Richard A. Ratliff A. Ramirez Bureau of Radiation Control City of Austin Texas Department of Health 1100 West 49th Street C. A. Johnson Austin, TX 78756-3189 AEP Texas Central Company Jeffrey Cruz Jon C. Wood U. S. Nuclear Regulatory Commission Matthews & Branscomb P. 0. Box 289, Mail Code: MNI 16 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704

NOC-AE-03001618 Attachment Page I of I Revised Response to Request for Additional Information

2. Please discuss the effects of the change of PSV setting to the system downstream of the PSVs including the tail piping and pressurizer relief tank (PRT).

Response

The amendment request does not involve any physical modification of plant design, changes in plant operation, or revision of lift setting of the PSVs. There are no physical changes that would affect the design acceptability of the system downstream of the PSVs including the tail piping and PRT.

The existing definition of pressurizer safety valve (PSV) operability per Technical Specification 3.4.2.2 is +/-2% of the setpoint pressure. The proposed definition of operability is +2%, -3%. The

+2% is currently an analyzed condition in the UFSAR; the effect of-3% is discussed below.

UFSAR Section 5.4.11.1 states:

The system design is based on the requirement to absorb a discharge of steam and consequent heat input equivalent to 110 percent of the full power pressurizer steam volume. The steam volume requirement is approximately that which would be experienced if the plant were to suffer a complete loss of load accompanied by a turbine trip but without the resulting direct reactor trip. A delayed reactor trip is assumed in the design of the system for the loss of load condition.

UFSAR Section 5.4.11.3 states:

The pressurizer relief discharge system is capable of handling the design discharge of steam without exceeding the design pressure and temperature.

The analysis of a complete loss of load accompanied by a turbine trip without the resulting direct reactor trip was performed with the PSV lifting at -3% of the lift setpoint. The results of the analysis show that the reactor coolant system pressure increases at the onset of the transient, but the pressurizer does not become water solid during the transient. As a result, the full pressurizer steam volume does not discharge to the pressurizer relief tank (PRT) and the PRT design basis capacity of 110 percent of the full power pressurizer steam volume is not challenged. Therefore, the proposed Technical Specification change would not expose the PRT to an unanalyzed condition.