NOC-AE-05001874, Transmittal of the 10CFR50.59 Summary Report

Transmittal of the 10CFR50.59 Summary Report

ML051260215
Person / Time
Site:	South Texas
Issue date:	04/28/2005
From:	Head S South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G25, NOC-AE-05001874, STI: 31876002
Download: ML051260215 (9)

Text

Nuclear Operating Company South Trs ProjdElcrk GCSahSt3on R.O Bh289 WedM, Tre= 77483 April 28, 2005 NOC-AE-05001874 STI: 31876002 10CFR50.59 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 10CFR50.59 Summary Report Pursuant to the requirements of 10CFR50.59, the attached report contains a brief description and summary of the 10CFR50.59 evaluations of changes, tests and experiments conducted at the South Texas Project.

There are no commitments in this letter.

If there are any questions regarding this summary report, please contact J. A. Loya at (361) 972-7922 or me at (361) 972-7136.

Scott M. Head Manager, Licensing jal

Attachment:

2005 10CFR50.59 Evaluation Summaries catA/7

NOC-AE-05001874 Page 2 of 2 cc:

(paper copy) (electronic copy)

Bruce S. Mallett A. H. Gutterman, Esquire Regional Administrator, Region IV Morgan, Lewis & Bockius LLP U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 David H. Jaffe Arlington, Texas 76011-8064 U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Jack A. Fusco Attention: Document Control Desk Michael A. Reed One White Flint North Texas Genco, LP 11555 Rockville Pike Rockville, MD 20852 C. A. Johnson AEP Texas Central Company Richard A. Ratliff C. Kirksey Bureau of Radiation Control City of Austin Texas Department of State Health Services 1100 West 49th Street Jon C. Wood Austin, TX 78756-3189 Cox Smith Matthews Jeffrey Cruz J. J. Nesrsta U. S. Nuclear Regulatory Commission R. K. Temple P. O. Box 289, Mail Code: MN116 E.Alarcon Wadsworth, TX 77483 City Public Service C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704

NOC-AE-05001874 Attachment Page 1 of 7 Summaries of the following 10CFR50.59 Evaluations are provided in this attachment:

1. 00-10937-6 Main Feedwater Control Valves - Energize-to-Actuate.

2. 00-10937-96 Main Steam Isolation Valve - Energize-to-Actuate

3. 00-10937-88 Main Feedwater Isolation Valve (MFIV) - Energize to Actuate.

4. 01-14284-6 Change of Updated Final Safety Analysis Report (UFSAR) Chapter 15 for OPAT/ OTAT two-second time constant modification.

5. 02-1134-5 Replace the calculation basis document (WCAP-1 1273) for RPS &

ESFAS Setpoints with STP calculations.

6. 02-10367-7 Change UFSAR 15.2.3 to reflect revised Accident Analyses with Pressurizer Safety Valve Setpoint Tolerance of +2/-3 %.

7. 02-13601-2 Revision of the UFSAR Chapter 6, 7 & 15 LOCA Radiological Analysis.

8;k'. 02-13601-4 Revision of the UFSAR Chapter 6, 7 & 15 LOCA Radiological Analysis.

9.: 03-3600-03 Revise UFSAR Section 3.5.1.4 to delete the missile barrier function of the Essential Cooling Water (ECW) door and add new 3.7.4 Technical Requirements Manual (TRM) section for ECW system intake structure

10. 03-4113-2 Control Rod Insertion for Hot Leg Recirculation Criticality Analysis.

11. 03-6248-21 Corrosion Analysis for the Reactor Bottom Mounted Instrument (BMI) Nozzle Leak Repairs.

12. 03-6248-41 ASME Code analysis for the Reactor BMI Nozzle Leak Repair.

13. 03-17841-14 Revise Long Term Cooling Analysis in supporting Auxiliary Feedwater Storage Tank (AFST) requirements.

14. 04-1238-88 Class 1 NSSS Inverter Modification

15. 04-3921-1 Revise TRM 3.7.7 for Defueled condition to allow Control Room Makeup and Cleanup Filtration system to be functional instead of operable.

NOC-AE 05001874 Attachment Page 2 of 7 10CFR5O.59 Evaluation Summaries

1. 00-10937-6

==

Description:==

This change revises the Feedwater Isolation circuit design for the Main Feedwater Control Valves from a "Deenergize-to-Actuate" Scheme to an "Energize-to-Actuate" Scheme.

Summary: This change will improve plant reliability by eliminating an unwanted Feedwater Isolation due to single point solenoid valve coil, wiring or relay coil failures during normal plant operations without compromising safety functions.

2. 00-10937-96

==

Description:==

This modification changes the Main Steam Isolation Valve's (MSIV)

Solenoid Control Circuit operation and their testing circuitry (previously de-energized) to energize-to-actuate to close the MSIV.

Summary: This change eliminates the reliability problem of having the solenoid valve energized during normal operation, that results in inadvertent coil failures, hardening of the hydraulic fluid inside the solenoid valve causing malfunction and reduced life of the components due to high temperature.

Increasing the reliability of the solenoid valves is obtained by changing from normally open to normally closed. This change is considered an improvement and enhancement to the system and also a better method of accomplishing the originally intended design function of the MSIV to close upon receipt of the actuation signal.

3. 00-10937-88

==

Description:==

This modification changes the MFIV Solenoid Control Circuit operation and their testing circuitry (previously de-energized) to energize-to-actuate to close the MFIV.

Summary: This change eliminates the reliability problem of having the solenoid valve energized during normal operation, that results in inadvertent coil failures, hardening of the hydraulic fluid inside the solenoid valve causing malfunction and reduced life of the components due to high temperature.

Increasing the reliability of the solenoid valves is obtained by changing from normally open to normally closed. This change is considered an improvement and enhancement to the system and also a better method of accomplishing the originally intended design function of the MFIV to close upon receipt of the actuation signal.

NOC-AE-05001 874 Attachment Page 3 of 7 4.01-14284-6

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Description:==

This change revises the UFSAR to reflect the analyses for an increase in the lag time constant for the measured Tavg and measured AT (from 0 to 2 seconds) in the Over Temperature Delta Temperature (OTAT) and Over Pressure Delta Temperature (OPAT) reactor trip circuit.

Summary: The change will provide a filter to screen out spurious spikes in hot leg temperature and eliminate spurious OTAT and OPAT alarms but still provide a steady signal for use in the reactor trip circuit. The results of the analyses with the change described above show that all acceptance limits continue to be met and that the design basis limit for the fission product barriers is not exceeded.

5.02-1134-5

==

Description:==

This change replaces WCAP-11273 as the basis for the RPS and ESFAS setpoints/allowable values with internal STP RPS and ESFAS setpoint basis calculation documents in the UFSAR.

Summary: This change places the basis calculations under direct STP control and responsibility. There are no changes in actual setpoints, allowable values, total allowances or safety analysis limits. Additionally, there are no changes in calculation methodology that are outside the constraints and limitations associated with the NRC approved method.

6.02-10367-7

==

Description:==

This change revises the Turbine Trip/Loss of Load transient analyses described in UFSAR Section 15.2.3 to reflect a reduction in pressurizer safety valve uncertainty used in the analysis from +/-3% to +2/-3%.

Summary: This change revises the Turbine Trip/Loss of Load transient analyses, which is used to establish the pressurizer safety relief valve setpoint tolerances in Technical Specification 3.4.2.2. The proposed change meets all acceptance limits for the Turbine Trip event.

NOC-AE-05001874 Attachment Page 4 of 7 7.02-13601-2

==

Description:==

This evaluation addresses the proposed changes to Chapter 6, 7 and 15 of the UFSAR resulting from the revision of the LOCA dose analysis from the Control Room, Technical Support Center (TSC) and Offsite locations.

Summary: The revision of calculation NC-6013 replaced ICRP-2 Dose Conversion Factors with ICRP-30 Dose Conversion factors. The use of ICRP-30 constitutes the use of an approved NRC methodology. All doses remain below 10CFR100/GDC 19 Limits.

8.02-13601-4

==

Description:==

This evaluation addresses the proposed changes to Chapter 6, 7 and 15 of the UFSAR resulting from the revision of the LOCA dose analysis from the Control Room, TSC and Offsite locations.

Summary: The change updates the design inputs for the supplementary purge purge mass release, containment building sump volume and various changes to minor dose contributors. All increases in dose consequence are less than ten percent of the differential between the current UFSAR value and 10CFR100 regulatory limit and remain below the applicable Standard Review Plan guideline.

9.03-3600-03

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Description:==

This evaluation addresses two related changes regarding the watertight door for each compartment of the ECW Intake Structure. The changes include revising UFSAR Section 3.5.1.4 to delete the missile barrier function of the ECW door and adding a new 3.7.4 TRM section for ECW system intake structure Summary: This change takes information already in the UFSAR and consolidates it in the Technical Requirements Manual Bases. The consolidated information supports the generation of new Technical Requirement 3/4.7.4. The design requirements of the watertight door have not been changed. The new Technical Requirement only clarifies when the door must perform its support function to ensure operability of the associated ECW System.

NOC-AE-05001874 Attachment Page 5 of 7 10.03-4113-2

Description:

This change revises the method used to evaluate core criticality during hot leg recirculation for a large break LOCA.

Summary: This change revises the UFSAR to reflect the methodology used to ensure the reactor is shutdown during the hot leg recirculation phase of the large break LOCA. This change takes credit for control rod insertion in addition to borated water to ensure the reactor does not become critical during the hot leg recirculation phase of the large break LOCA. This change is consistent with an NRC approved methodology used for the D.C. Cook plant and is reflected in the NRC's SER. Additionally, there are no limitations imposed in the SER that preclude or invalidate the use of the methodology for similar PWR's, such as STP.

11.03-6248-21

Description:

Corrosion evaluation regarding the reactor vessel low alloy steel exposed to primary water due to the Unit 1 Reactor BMI Nozzle Leaks using the Half-Nozzle Repair technique.

Summary: The Half-Nozzle Repair leaves a portion of the reactor vessel low alloy steel ....  : .I exposed to primary water by way of the existing nozzle cracks and the gap between the new Alloy 690 Half-Nozzle and the existing Alloy 600 nozzle . .  ;

remnant. The evaluation concluded that minimal corrosion of the reactor vessel base material occurs and that the Half-Nozzle repair was acceptable.

12. 03-6248-41

Description:

ASME Code analysis for the Reactor BMI Nozzle Leaks using the Half-Nozzle Repair.

Summary: The BMI Half-Nozzle Repair is required to meet the ASME Code requirements, which include the effects of cyclic loading. This evaluation demonstrated that the ASME Code analysis for the Half-Nozzle Repair was acceptable.

NOC-AE-05001874 Attachment Page 6 of 7 13.03-17841-14

==

Description:==

This change revises the long term cooling analysis that forms the basis for Technical Specification 3.7.1.3.

Summary: The new long term cooling analysis incorporates changes that apply only when the RCS is being cooled down to MODE 4 under natural circulation conditions. The results of the analyses of these events show that all acceptance limits continue to be met and that the design basis limit for the fission product barriers is not exceeded.

14. 04-1238-88

==

Description:==

This modification involves changes to Channels I - IV of the vital 120V AC power distribution.

1. Channels I-IV: The 7.5kVA inverters are replaced with new 1OkVA inverters with a static transfer switch and a maintenance bypass switch for all channels. The 100A breakers in the 125V DC Switch Boards are replaced with 150A breakers. The 20A breakers are replaced with 30A breakers in the 480V Motor Control Centers (MCC).

2. Channels I & IV only: The 3OkVA voltage regulating transformers are..

replaced with 2OkVA voltage regulating transformers. New power feeds c are provided from the 480V MCC buses and new 1OkVA voltage regulating transformers are installed. Additionally, the Qualified Display Processing System (QDPS) and Steam Generator (SG) Power Operated Relief Valve (PORV) Servo loads are relocated.

3. Channels II & III only: Replace l5kVA voltage regulating transformers with 10kVA voltage regulating transformers and determinate the power feed from voltage regulating transformers to 120V AC panels including rerouting power to the 1OkVA inverter/rectifiers.

Summary: The safety related 7.5kVA inverters and bypass transformers were replaced since their increase in failure frequency and age presented a challenge to the operation of the station. The new inverters are equipped with a static transfer switch that allows auto transfer capability for certain QDPS loads, SG PORV Servos and other 120V AC panel loads including NSSS Process Cabinets that if lost due to failure of an inverter would automatically transfer to an alternate power source. This eliminated a potential plant trip and improves the equipment reliability.

NOC-AE-05001 874 Attachment Page 7 of 7 W

15.04-3921-1

Description:

Revise TRM 3.7.7 for Defueled condition to allow Control Room Makeup and Cleanup Filtration system to be functional instead of operable.

Summary: The proposed change allows an evaluation that shows that a system with a degraded boundary (i.e., points with differential pressure less than 1/8" wg with respect to adjacent areas) is still functional and capable of meeting GDC-19 design basis requirements for mitigating the dose from a design basis fuel handling accident.