NOC-AE-14003143, Supplement to License Amendment Request Proposed Revision to Technical Specification 3.3.1, Functional Unit 20, Reactor Trip Breakers

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Supplement to License Amendment Request Proposed Revision to Technical Specification 3.3.1, Functional Unit 20, Reactor Trip Breakers
ML14184B363
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 06/09/2014
From: Gerry Powell
South Texas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-14003143, STI: 33884848, TAC MF3319, TAC MF3320
Download: ML14184B363 (39)
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Text

Nuclear Operating Company South Te.as Pro/ect ElectricGencrating Station PO. Box 289 Wadsworth. Texas 77483 June 9, 2014 NOC-AE-14003143 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Supplement to License Amendment Request Proposed Revision to Technical Specification 3.3.1, Functional Unit 20, "Reactor Trip Breakers"(TAC Nos. MF 3319 and MF 3320)

Reference:

Letter from G.T. Powell, STP Nuclear Operating Company, to NRC Document Control Desk, "License Amendment Request Proposed Revision to Technical Specification 3.3.1, Functional Unit 20, 'Reactor Trip Breakers'," dated January 6, 2014. (NOC-AE-13003031)(ML14035A075)

This supplement supersedes the above referenced letter in its entirety. The information contained in this submittal is no longer considered to be proprietary by Westinghouse, the owner of the information. All proprietary markings and requested restrictions have been removed in this supplement. STP Nuclear Operating Company (STPNOC) requests that the preceding letter (ML14035A075) be removed from the Agencywide Documents Access and Management System (ADAMS).

In accordance with the provisions of 10 CFR 50.90, STPNOC hereby requests a license amendment to South Texas Project Operating Licenses NPF-76 and NPF-80. This proposed license amendment revises Technical Specification (TS) 3.3.1, "Reactor Trip System Instrumentation," with respect to the required actions and allowed outage times for inoperable reactor trip breakers, Functional Unit 20.

The proposed changes would revise the required actions and allowed outage times for inoperable reactor trip breakers to be consistent with those generically approved in NUREG- 1431, Standard Technical Specifications, Westinghouse Plants, Revision 4. Justification for the proposed changes is based on Westinghouse Topical Report, WCAP-15376-P-A, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times".

STI: 33884848

NOC-AE-14003143 Page 2 of 3 provides a technical and regulatory evaluation of the changes. An Applicability Determination of WCAP-15376-P-A to the South Texas Project Units, ST-WN-NOC-13-46, "STP Nuclear Operating Company South Texas Project Electric Generating Station Units I and 2 Implementation of Technical Specification Changes Justified in WCAP-15376-P-A, Rev.1", is provided as an Attachment to Enclosure 1.

Proposed TS page markups are included as Enclosure 2 to this letter. The associated TS Bases change is included for information only as Enclosure 3 to this letter.

The STPNOC Plant Operations Review Committee has reviewed and concurred with the proposed change to the Technical Specifications.

STPNOC requests approval of this license amendment application by December 30, 2014 and requests 90 days for implementation of the amendment.

In accordance with 10 CFR 50.91(b), STPNOC is notifying the State of Texas of this request for license amendment by providing a copy of this letter and enclosures.

Licensing commitments for the proposed change are described in Enclosure 4 to this letter.

If there are any questions regarding the proposed amendment, please contact Wendy Brost at (361) 972-8516 or me at (361) 972-7566.

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

Executed on TAI-. I , zi4l Date G. T. Powell Site Vice President web

Enclosures:

1. Evaluation of the Proposed Change Attachment to Enclosure 1:

A Topical Report Applicability Determination, ST-WN-NOC-13-46

2. Technical Specification Page Markups
3. Technical Specification Bases Inserts (for information only)
4. List of Commitments

NOC-AE- 14003143 Page 3 of 3 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 1600 East Lamar Boulevard Arlington, TX 76011-4511 Balwant K. Singal U.S. Nuclear Regulatory Commission Balwant K. Singal John Ragan Senior Project Manager Chris O'Hara U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8 B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector Kevin Pollo U. S. Nuclear Regulatory Commission Cris Eugster P. 0. Box 289, Mail Code: MN1 16 L.D. Blaylock Wadsworth, TX 77483 CPS Energy Jim Collins Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 City of Austin Richard A. Ratliff Texas Department of State Health Services Robert Free Texas Department of State Health Services

NOC-AE-14003143 Enclosure 1 ENCLOSURE Evaluation of the Proposed Change

Subject:

Proposed Revision to Technical Specification 3.3.1, Functional Unit 20, "Reactor Trip Breakers" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENT:

A Topical Report Applicability Determination, ST-WN-NOC- 13-46

NOC-AE-14003143 Enclosure 1 Page 1 of 17 1.0

SUMMARY

DESCRIPTION The proposed license amendment revises the Technical Specification (TS) 3.3.1, "Reactor Trip System Instrumentation," with respect to the required actions and allowed outage times for inoperable reactor trip breakers.

The proposed changes will revise the required actions to enhance plant reliability by reducing exposure to unnecessary shutdowns and increase operational flexibility by allowing more time to make required repairs for inoperable reactor trip breakers consistent with allowed outage times for associated logic trains. No modifications to setpoint actuations, trip setpoint, surveillance requirements or channel response that would affect the safety analyses are associated with the proposed changes.

The proposed changes are consistent with requirements generically approved as part of NUREG-143 1, Standard Technical Specifications, Westinghouse Plants, Revision 4 (TS 3.3.1, "Reactor Trip System Instrumentation"). Justification for the proposed changes is based on Westinghouse Topical Report, WCAP- 15376-P-A,. Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times" (Reference 6.1). to this letter provides the markups for the TS pages. Enclosure 3 to this letter provides the associated TS Bases inserts for information. Attachment A to this enclosure provides the Topical Report Applicability Determination.

2.0 DETAILED DESCRIPTION TS 3.3.1, Table 3.3-1, "Reactor Trip System Instrumentation," directs entry into Action 9 in the event that the Minimum Channels Operable column requirements are not met for Functional Unit 20, "Reactor Trip Breakers", when in Modes 1 and 2. Action 9 states:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

The proposed amendment revises Action 9 to increase the allowed outage time in the event that the Minimum Channels Operable column requirement is not met and increases the amount of time that one channel may be bypassed for surveillance testing provided the other channel is OPERABLE. The proposed revised Action 9 states:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, h u s V o a

,-- --

  • s- be in at least HOT STANDBY within.I,¶) hours;

NOC-AE-14003143 Enclosure 1 Page 2 of 17 however, one channel may be bypassed for up to .;A hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

TS 3.3.1, Table 3.3-1, "Reactor Trip System Instrumentation," directs entry into Action 12 in the event one of the diverse trip features of a reactor trip breaker is inoperable for Functional Unit 20, "Reactor Trip Breakers", when in Modes 1 and 2. Action 12 states:

With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

The proposed amendment revises Action 12 rather than refer to Action 9 because the proposed change to Action 9 would no longer be consistent with the requirement for an inoperable diverse trip feature. The proposed revised Action 12 states:

With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERALE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> M 42 The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

TS 3.3.1, Table 3.3-1, "Reactor Trip System Instrumentation," directs entry into Action 10 in the event that the Minimum Channels Operable column requirements are not met for Functional Unit 20, "Reactor Trip Breakers", when in Modes 3, 4 and 5 and the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal. Entry into Action 10 is also required for Functional Unit 1, "Manual Reactor Trip", and Functional Unit 21, "Automatic Trip and Interlock Logic", in the event that the Minimum Channels Operable column requirements are not met for these Functional Units when in Modes 3, 4 and 5 and the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

Action 10 states:

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor Trip System breakers within the next hour.

The proposed amendment revises Action 10 to be consistent with Standard TS in the event that the Minimum Channels Operable column requirements are not met. The proposed revised Action 10 states:

NOC-AE- 14003143 Enclosure 1 Page 3 of 17 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE atus fa New Action 12A is added to be consistent with NUREG-1431 to address the condition where one diverse trip feature for a reactor trip breaker is inoperable when in Modes 3, 4 and 5 and the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal.

New Action 12A states:

The proposed changes to Technical Specification 3.3.1 are consistent with generically approved requirements provided in NUREG-143 1, "Standard Technical Specifications, Westinghouse Plants," Revision 4 (TS 3.3.1, "Reactor Trip System Instrumentation").

3.0 TECHNICAL EVALUATION

3.1 Description of Reactor Trip Breakers Two reactor trip breakers (RTB) arranged in series connect three-phase ac power from the control rod drive motor generator sets to the rod drive power cabinets supplying power to the control rod drive mechanisms (CRDM). Opening either of the RTBs interrupts power to the CRDMs and allows the shutdown rods and control rods to fall into the core by gravity. Each RTB is equipped with a bypass breaker to allow testing of the RTB while the unit is at power.

During normal operation the output from the solid state protection system (SSPS) provides a direct voltage signal to the undervoltage coil on each reactor trip breaker and bypass breakers, if in use. Direct current holds a trip plunger out against its spring, allowing ac power to be available at the rod drive power cabinets. SSPS consists of two logic trains, each capable of opening a separate and independent reactor trip breaker.

SSPS takes binary inputs (i.e. voltage or no-voltage) from the process and nuclear instrumentation channels corresponding to the conditions of plant parameters. When a required logic combination is completed, a reactor trip signal (i.e. no voltage) is generated to the undervoltage trip coil. In addition, the reactor trip signal energizes the shunt trip auxiliary relay coils of the RTBs to trip the breakers open. The shunt trip auxiliary relay coils provide a diverse means to trip the RTBs.

NOC-AE- 14003143 Enclosure 1 Page 4 of 17 A list of reactor trips, respective coincidence logics, and interlocks are provided in Table 7.2-1 of the South Texas Project (STP), Units 1 and 2 Updated Final Safety Analysis Report (UFSAR). A functional diagram of reactor trip signals including the arrangement of the reactor trip switchgear is provided by Figure 7.2-2 of the UFSAR.

The reactor trip system is designed to permit periodic testing during power operation without initiating a protective action unless a trip condition actually exists. Where only parts of the system are tested at any one time, the testing sequence provides the necessary overlap between the parts to assure complete system operation. The testing device is semiautomatic to minimize testing time.

3.2 Background Amendment No. 136 to Facility Operating License No. NPF-76 and Amendment No. 125 to Facility Operating License No. NPF-80 for South Texas Project, Units 1 and 2, respectively were approved on March 19, 2002 (Reference 6.2). These amendments permitted relaxation of the allowed outage times (AOT) and bypass test times (BTT) for limiting conditions of operation specified in TS 3.3.1, "Reactor Trip System Instrumentation," and TS 3.3.2, "Engineered Safety Features Actuation System Instrumentation." The AOT and BTT relaxations were in accordance with the requirements of Westinghouse Owner's Group (WOG) Topical Report WCAP-14333-P-A, Revision 1 (Reference 6.3). The license amendment request that resulted in Amendments 136 and 125 did not propose a relaxation of the AOT and BTT for Functional Unit 20, "Reactor Trip Breakers".

Amendment No. 188 to Facility Operating License No. NPF-76 and Amendment No. 175 to Facility Operating License No. NPF-80 for South Texas Project, Units 1 and 2, respectively were approved on October 31, 2008 (Reference 6.4). These amendments relocated specified TS Surveillance Test Intervals (STI) to a licensee-controlled program.

These amendments included Reactor Trip System Instrumentation Surveillance Requirements.

On December 20, 2002 (Reference 6.5), the NRC accepted referencing Westinghouse Topical Report WCAP-15736-P, Revision 0 for performing risk-informed assessments to support proposed changes to plant Technical Specification reactor trip system and engineered safety features actuation system surveillance test intervals and to reactor trip breaker test and completion times. Westinghouse WCAP-15376-P-A, Revision 1 was issued in March 2003 to reflect the NRC approval of the Topical Report and incorporate information requested by the NRC in the December 20, 2002 letter including the NRC Safety Evaluation. The proposed changes in this License Amendment Request only impact reactor trip breaker test and completion times.

NOC-AE- 14003143 Enclosure 1 Page 5 of 17 3.3 Analysis The proposed change for RTB AOT and BTT relaxation is based on WCAP-15376-P-A.

The term "AOT" used in this license amendment request is synonymous with the term "completion time (CT)" used in Standard Technical Specifications.

The approach used in WCAP-15376-P-A is consistent with the Nuclear Regulatory Commission's (NRC) approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the current licensing basis as presented in NRC Regulatory Guides 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis,"(Reference 6.6) and 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" (Reference 6.7). The approach addresses the impact on defense'in-depth and the impact on safety margins, as well as an evaluation of the impact on risk.

The proposed change considers the three-tiered approach as presented in RG 1.177 for the extension to the RTB AOT. The first tier addresses PRA insights and includes the risk analyses and sensitivity analyses to support the allowed outage time and bypass test time changes. The second tier addresses avoidance of risk-significant plant configurations. The third tier addresses risk-informed plant configuration control and management.

3.3.1 Tier 1, PRA Capability and Insights Risk analysis results for WCAP-15376 are discussed in Section 8.4 of that topical report.

Comparisons are presented in Tables 8.29 (ACDF) and 8.32 (ALERF) to a base case which represents the changes previously approved under WCAP-14333. In response to an NRC request for additional information (RAI) letter, RAI Questions 4 and 11 in Westinghouse Owners Group (WOG) letter OG-02-002 (Reference 6.8), the WOG provided the impact of the requested Completion Time change (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reach MODE 3, for a total of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />) on ICCDP and ICLERP for a reactor trip breaker (RTB) in preventive maintenance (PM) or in corrective maintenance (CM), with the associated logic train inoperable, for the bounding 2/3 logic. Since these incremental risk metrics are met for a 30-hour maintenance time, they will also be met for a 4-hour bypass test time.

NOC-AE- 14003143 Enclosure 1 Page 6 of 17 The risk metric results are:

Risk Acceptance Change from Metric Criterion WCAP-14333 to WCAP-15376 ACDF <lE-06 2/4 logic 2/3 logic per year 10E-08 1.0E-08 ICCDP <5E-07 RTB in PM 3.20E-07 RTB in CM 3.22E-07 ALERF <lE-07 2/4 logic 2/3 logic per year 1.66E-09 6.25E-10 ICLERP <5E-08 RTB in PM 2.41E-08 RTB in CM 2.42E-08 CDF - core damage frequency ICCDP - incremental conditional core damage probability LERF - large early release frequency ICLERP - incremental conditional large early release probability The acceptance criteria for the risk metrics are satisfied. The applicability of WCAP-15376 to the proposed changes of this license amendment request is demonstrated below.

3.3.2 Tier 2, Avoidance of Risk-Significant Plant Configurations Tier 2 requires an examination of the need to impose additional restrictions when operating under the proposed Allowed Outage Times in order to avoid risk-significant equipment outage configurations.

Recommended Tier 2 restrictions for WCAP- 15376 are provided in Section 8.5 of the topical report when a RTB train is inoperable for maintenance. These restrictions apply during pre-planned evolutions during power operation. For emergent conditions, the Tier 3 Configuration Risk Management Program discussed below will assess the emergent condition and direct activities to restore the inoperable RTB train and place risk management actions in place, as appropriate.

The following are restrictions on equipment removal when a RTB is out-of-service for planned maintenance:

The probability of failing to trip the reactor on demand will increase when a RTB is removed from service. Systems designed to mitigate an Anticipated Transient Without Scram (ATWS) event should be available. Activities that degrade the availability of reactor coolant system pressure relief, auxiliary

NOC-AE- 14003143 Enclosure 1 Page 7 of 17 feedwater flow, ATWS Mitigating System Actuation Circuitry (AMSAC), and turbine trip would not be scheduled when an RTB is out-of-service.

  • Due to the increased dependence on the available reactor trip train when one logic cabinet or one RTB is removed from service, activities that could degrade other components of the reactor protection system including master relays, slave relays, and analog channels would not be scheduled concurrently with a logic cabinet out of service.
  • Activities on electrical systems (e.g. AC and DC power) that support the systems or functions listed in the first two bullets should not be scheduled when a RTB is unavailable.

The restrictions described above are Licensing Commitments provided in Enclosure 4.

3.3.3 Tier 3, Risk-Informed Configuration Risk Management Tier 3 requires a proceduralized process to assess the risk associated with both planned and unplanned work activities. The objective of the third tier is to ensure that the risk impact of out-of-service equipment is evaluated prior to performing any maintenance activity. As stated in Section 2.3 of Regulatory Guide 1.177, "...a viable program would be one that is able to uncover risk-significant plant equipment outage configurations in a timely manner during normal plant operation." The third-tier requirement is an extension of the second-tier requirement, but addresses the limitation of not being able to identify all possible risk-significant plant configurations in the second-tier evaluation. Programs and procedures are in place at STP which serves to address this objective.

The STP currently has in place a risk-informed on-line maintenance tracking and control process. The Configuration Risk Management Program (CRMP) was incorporated into the South Texas Project Technical Specifications via amendments 85 (Unit 1) and 72 (Unit 2), issued on October 31, 1996 (Reference 6.9). In the Safety Evaluation for Amendments 85 and 72, the NRC Staff concluded that STP had "provided the necessary assurances that appropriate assessments of the overall impacts on safety functions will be performed prior to any maintenance or other operational activities, including removal of equipment from service". The CRMP also ensures that risk is reassessed if an emergent condition results in a plant configuration that has not been previously assessed. When administrative limits are exceeded, increasing levels of management approval are required prior to initiating work. Additional risk management actions are initiated, as appropriate. This risk-informed on-line maintenance tracking and control process is implemented and governed by a plant procedure (Configuration Risk Management Program, OPGP03-ZA-0091).

NOC-AE-14003143 Enclosure 1 Page 8 of 17 3.3.4 NRC Safety Evaluation Conditions NRC approval of WCAP-15376 was subject to the following conditions requiring plant-specific information:

Condition 1 Confirm the applicability of WCAP-15376-P-A to the STP facility and perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposed changes.

In order to address Safety Evaluation (SE) Condition 1 of WCAP-15376-P-A, Westinghouse issued implementation guidelines for licensees to confirm the analyses are applicable to their plant. See Attachment A of this enclosure.

Condition 2 The Tier 2 and Tier 3 analyses in the NRC Safety Evaluation needs to be addressed including risk significant insights with confirmation that these insights are incorporated into the plant-specific configuration risk management program.

SE Condition 2 of WCAP-15376-P-A is addressed above under the Tier 2 (Section 3.3.2) and Tier 3 (Section 3.3.3) discussions.

Condition 3 The risk impact of concurrent testing of one logic cabinet and associated reactor trip breaker needs to be evaluated on a plant-specific basis to ensure conformance with the WCAP-15376 evaluation, and RGs 1.174 and 1.177.

The response to NRC RAI Question 4 in Reference 6.8 provided the incremental conditional core damage probability (ICCDP) for this configuration (i.e., both the logic train and associated RTB train out of service) for preventive maintenance for a total time of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, which is comprised of a Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reach Mode 3. The ICCDP for 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of unavailability for this configuration is 3.2E-07, which meets the Regulatory Guide 1.177 acceptance criteria of 5E-07. Because this ICCDP value is based on the logic train and reactor trip breaker being out of service for 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> at the same time, bypassing one logic train and associated RTB train for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for testing will also meet the Regulatory Guide 1.177 ICCDP guideline.

SE Condition 3 is addressed by demonstrating that the WCAP-15376 analysis is applicable. Demonstrating the applicability of the WCAP-15376 analysis is discussed in detail in response to SE Condition I (see Attachment A of this enclosure).

NOC-AE- 14003143 Enclosure 1 Page 9 of 17 Condition 4 The model assumptions for human reliability in WCAP-15376-P, Rev. 0 should be confirmed to be applicable to the STP facility.

In order to address Safety Evaluation (SE) Condition 4 of WCAP-15376-P-A, Westinghouse issued implementation guidelines for licensees to confirm the analyses are applicable to their plant. See Attachment A of this enclosure.

Condition 5 For future digital upgrades with increased scope, integration and architectural differences beyond that of Eagle 21, the generic applicability of WCAP-15376-P, Rev. 0 should be considered on a plant-specific basis.

The proposed change does not involve a digital upgrade. Future digital upgrades will require separate evaluation.

Additional Commitment Each plant should review their setpoint calculation methodology to determine the impact of extending the Channel Operational Test (COT) Surveillance Frequency from 92 days to 184 days.

The proposed change does not involve a change to surveillance frequencies.

Overall Conclusion The Conditions and Limitations for the NRC Safety Evaluation are met so that WCAP-1 5376-P-A, Rev. 1 is applicable to changes proposed in this license amendment request for STP Units 1 and 2.

3.3.5 STP PRA The quality of the STP Probabilistic Risk Assessment (PRA) was reviewed by the Nuclear Regulatory Commission (NRC) as part of the process for approving Risk.

Managed Technical Specifications (RMTS) at the South Texas Project (Reference 6.10).

During the approval of that application, the NRC concluded that based on the licensee's assessment and the staff reviews, the staff determined that the STP PRA internal events model satisfied the guidance of RG 1.200, Revision 1, and conformed to capability category II of the American Society of Mechanical Engineers (ASME) standard for the

NOC-AE-14003143 Enclosure 1 Page 10 of 17 supporting requirements in using probabilistic risk assessments for nuclear power applications. The STP internal events PRA was determined to be of sufficient technical adequacy to support the RMTS application. In addition, the staff concluded that based on the licensee's submittals and staff reviews, the STP PRA external events models satisfy the guidance of RG 1.200, Revision 1, (Reference 6.11) and are acceptable to support the RMTS application.

The PRA reviewed by the staff for the RMTS application was Revision 5. Revisions to the PRA model are controlled by a site quality procedure. Periodic reviews are conducted for major design changes and updates are performed, if necessary, for plant changes including performance data, procedures, and modifications. The reviews and updates are performed by qualified personnel with independent reviews and approvals.

The plant-specific information for the proposed changes of this license amendment request is based on Revision 7.2 of the PRA model. Because the version of the PRA model reviewed by the staff for the RMTS application was sufficient and the PRA model for this application has been updated via a controlled process, the current PRA quality is sufficient to support this license amendment request.

3.3.6 Evaluation of Specific TS Changes TS 3.3.1, Table 3.3-1, Action 9 WCAP-1 5376-P-A provides the technical justification for extending the RTB AOT for one RTB inoperable to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the BTT for a RTB to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The AOT and BTT are consistent with the AOT and BTT for the logic cabinets (Functional Unit 21 of TS Table 3.3-1). The conditions of WCAP-15376-P-A are confirmed to be applicable to the STP facility.

The proposed AOT and BTT for the RTBs are reasonable because these times take into account the operability status of the redundant RTB, the capability of the remaining reactor trip features to provide protection, a reasonable time for repairs or replacement, and the low probability of a design basis accident (DBA) occurring during the repair period. The proposed AOT for the RTBs provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related compliance with RTB AOTs. The proposed AOT provides consistency with the AOTs for the testing of reactor protection system logic cabinets.

The proposed increases to Reactor Trip Breaker allowed outage times and bypass test times are expected to reduce the potential for human errors by personnel performing required actions, corrective maintenance and surveillance testing.

NOC-AE-14003143 Enclosure 1 Page 11 of 17 TS 3.3.1, Table 3.3-1, Action 12 The wording of Action 12 is required to change because a new Action 9 is proposed.

The current Action 12 requirement is to apply Action 9 if one of the diverse trip features is inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. With the proposed change to Action 9, the Action 12 (without revision) would allow an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the currently allowed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before initiating the requirement to be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This is contrary to the Standard TS.

The proposed Action 12 is revised to require the plant to be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if the diverse trip feature is not restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Therefore, the proposed Action 12 remains consistent with the current Action 12.

TS 3.3.1, Table 3.3-1, Action 10 The proposed Action 10 is revised to be consistent with the language of the Standard TS. The proposed Action 10 provides clarity that action to fully insert rods should be initiated and allows flexibility in the method used to place the rod control system in a condition incapable of rod withdrawal. No change is proposed to the time requirement to render rods incapable of being withdrawn if the number of operable channels is not restored to the minimum channels requirement within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

TS 3.3.1, Table 3.3-1, New Action 12A New Action 12A is added to be consistent with NUREG-143 1 to address the condition where one diverse trip feature for a reactor trip breaker in inoperable when in Modes 3, 4 and 5 and the reactor trip system breakers are in the closed position and the control rod drive system is capable of rod withdrawal. This new action addresses a condition not previously addressed by Technical Specifications.

3.4 Technical Evaluation Conclusion The proposed change for RTB AOT and BTT relaxation is based on WCAP-15376-P-A.

The approach used in WCAP-15376-P-A is consistent with the NRC approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the current licensing basis as presented in NRC Regulatory Guides 1.174 and 1.177. The approach addresses the impact on defense-in-depth and the impact on safety margins, as well as an evaluation of the impact on risk.

The proposed change considers the three-tiered approach as presented in RG 1.177 for the extension to the RTB AOT.

NOC-AE-14003143 Enclosure 1 Page 12 of 17 The Conditions and Limitations of the NRC Safety Evaluation for WCAP-15376-P-A are met so that the WCAP is applicable to changes proposed in this license amendment request for STP Units 1 and 2.

The STP PRA satisfies the guidance of NRC RG 1.200, Revision 1 for internal and external events. Revisions to the STP PRA model are controlled by a site quality procedure.

The proposed AOT and BTT for the RTBs are reasonable because these times take into account the operability status of the redundant RTB, the capability of the remaining reactor trip features to provide protection, a reasonable time for repairs or replacement, and the low probability of a design basis accident (DBA) occurring during the repair period. The proposed AOT for the RTBs provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related compliance with RTB AOTs. The proposed AOT provides consistency with the AOTs for the testing of reactor protection system logic cabinets.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants Criterion 20 - Protection system functions.

The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Criterion 21 - Protection system reliability and testability.

The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.

Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

NOC-AE-14003143 Enclosure 1 Page 13 of 17 Criterion 22 - Protectionsystem independence.

The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.

Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

Criterion 23 - Protectionsystem failure modes.

The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air),

or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

Criterion 29 - Protection againstanticipatedoperationaloccurrences.

The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

4.2 Precedent The Nuclear Regulatory Commission (NRC) approved similar changes to the allowed outage times and bypass test times for limiting conditions for operation of TS 3.3.1, "Reactor Trip System Instrumentation" and TS 3.3.2, "Engineered Safety Features Actuation System Instrumentation" for South Texas Project, Units 1 and 2 with Amendments No. 136 and 125, respectively.

NRC Standard Technical Specifications, Westinghouse Plants, Revision 4 specifies completion times and bypass test times for reactor trip breakers based on WCAP-l15376 similar to the changes proposed by this License Amendment Request. (Reference 6.12)

Extension of completion times and bypass test times for reactor trip breakers based on WCAP-15376 was approved for the Comanche Peak Steam Electric Station as Amendment 114 to the Unit 1 and 2 Operating Licenses. (Reference 6.13)

NOC-AE- 14003143 Enclosure 1 Page 14 of 17 4.3 Significant Hazards Consideration STP has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10CFR50.92, "Issuance of amendment," as discussed below.

1) Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response

No.

The overall reactor trip breaker performance will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed.

The reactor trip breakers will continue to function in a manner consistent with the plant design basis.

The proposed changes do not introduce any new accident initiators, and therefore do not increase the probability of any accident previously evaluated. There will be no degradation in the performance of or an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report.

The determination that the results of the proposed changes are acceptable was established in the NRC Safety Evaluation (issued by letter dated December 20, 2002) prepared for WCAP-15376-P-A, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times". Implementation of the proposed changes will result in an insignificant risk impact. Applicability of these conclusions has been verified through plant-specific reviews and implementation of the generic analysis results in accordance with the respective NRC Safety Evaluation conditions.

Therefore, the proposed changes do not increase the probability or consequences of an accident previously evaluated.

2) Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No.

NOC-AE-14003143 Enclosure 1 Page 15 of 17 The proposed changes do not result in a change in the manner in which the Reactor Trip Breakers provide plant protection. The proposed changes do not change the response of the plant to any accidents. No design changes are associated with the proposed changes.

The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously analyzed.

3) Does the proposed change involve a significant reduction in a margin of safety?

Response

No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria as stated in the Updated Final Safety Analysis Report are not impacted by these changes. Redundant Reactor Trip Breaker features and diverse trip features for each Reactor Trip Breaker are maintained.

All signals credited as primary or secondary, and all operator actions credited in the accident analyses are unaffected by the proposed change. The proposed changes will not result in plant operation in a configuration outside the design basis. The proposed changes should enhance plant reliability by reducing exposure to unnecessary shutdowns and increase operational flexibility by allowing more time to make required repairs for inoperable reactor trip breakers.

The calculated impact on risk is insignificant and meets the acceptance criteria contained in NRC Regulatory Guides 1.174 and 1.177.

Therefore, the proposed changes do not result in a significant reduction in a margin of safety.

Based on the above, STP concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

NOC-AE-14003143 Enclosure 1 Page 16 of 17 manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

S A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement:

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement, or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1. Westinghouse WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and.

Completion Times", March 2003.

6.2 South Texas Project, Units 1 and 2 - Issuance of Amendments Revising Allowed Outage Times and Bypass Test Times for Instrumentation, dated March 19, 2002.

(ST-AE-NOC-02000932) (ML020220399) 6.3 Westinghouse WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times", October 1998.

6.4 South Texas Project, Units 1 and 2 - Issuance of Amendments to Relocate Surveillance Test Intervals to License-Controlled Surveillance Frequency Control Program (Risk-Informed Initiative 5-b), dated October 31, 2008.

(ST-AE-NOC-08001824) (ML082830172) 6.5 Letter from William H. Ruland, NRC to Robert H. Bryan, Chairman, Westinghouse Owners Group, "Acceptance for Referencing of Topical Report WCAP-15376-P, Rev. 0, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times", dated December 20, 2002 [letter included in Reference 6.1].

6.6 NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current

NOC-AE- 14003143 Enclosure 1 Page 17 of 17 Licensing Basis".

6.7 NRC Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications'.

6.8 Westinghouse Owners Group letter OG-02-002 dated January 8, 2002 (copy included in Appendix D of the approved version of Reference 6.1).

6.9 South Texas Project, Units I and 2 - Amendment Nos. 85 and 72 to Facility Operating License Nos. NPF-76 and NPF-80, dated October 31, 1996.

(ST-AE-HL-94678) (ML021300535) 6.10 South Texas Project, Units 1 and 2 - Issuance of Amendments RE:

Broad-Scope Risk-Informed Technical Specifications Amendments, dated July 13, 2007. (ST-AE-NOC-07001652) (ML071780186) 6.11 NRC Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,.

6.12 NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 4, April 2012.

6.13 Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 - Issuance of Amendment RE: Plant Protection Test Times, Completion Times, and Surveillance Test Intervals. (ML050460331)

NOC-AE-14003143 Attachment A to Enclosure 1 Attachment A Topical Report Applicability Determination, ST-WN-NOC-13-46 Attachment A contains the applicability determination of WCAP-15376-P-A Topical Report to the STPNOC License Amendment Request. The applicability determination was performed by the South Texas Project.

NOC-AE- 14003143 Attachment A to Enclosure 1 Page 1 of 6 Safety Evaluation Condition 1 for WCAP-15376-P-A In order to address Safety Evaluation (SE) Condition 1 for WCAP-15365-P-A, Westinghouse issued implementation guidelines for licensees to confirm the analyses are applicable to their plant.

1. Confirm applicability of the WCAP-15376 Analysis To demonstrate the applicability of the WCAP-15376 analysis on a plant-specific basis, a comparison between the key generic analysis parameters and assumptions, and plant-specific parameters and design is necessary. Table 1 provides a list of the key analysis parameters and assumptions along with the input used in the generic analysis. Table 2 and Table 3 of the.

Westinghouse implementation guidelines are not provided because these tables address the applicability of analysis reactor trip actuation signals and applicability of analysis engineered safety features actuation signals. Reactor trip actuation and engineered safety features actuation signal channels are not the subject of the proposed changes to the South Texas Project Technical Specifications.

Information is provided for the proposed changes on the plant's calculated core damage frequency (CDF), large early release frequency (LERF), and the contribution to CDF from ATWS events. The plant CDF and LERF values are used to show that these values meet the NRC Regulatory Guide 1.174 criteria for determining that small increases in CDF and LERF are acceptable. The ATWS contribution to CDF is necessary to understand the importance of the ATWS event to the plant's risk, since the proposed changes can impact reactor trip signal availability.

2. Confirm applicability of the Component Failure Probabilities Component failure probabilities developed as part of WCAP-15376 need not be confirmed because the proposed changes are not requesting extensions to surveillance test intervals.
3. Containment failure assessment In addition to the applicability confirmation above, a discussion of containment failure modes is requested for WCAP-15376.

The containment failure modes considered in the STP PRA include containment isolation failure; containment bypasses from Interfacing System Loss of Coolant Accidents (ISLOCA) and Steam Generator Tube Rupture (SGTR); and containment failure. The significant contributors to LERF for large dry containment designs are typically containment isolation failure and containment bypasses.

NOC-AE-14003143 Attachment A to Enclosure 1 Page 2 of 6 The LERF analysis completed to support WCAP-15376 was based on a large dry containment with LERF contributions from containment isolation failure and containment bypass. STP large dry containment is similar to that of the reference plant, Vogtle. The STP LERF is 3.72E-07 per year for the average annual model (including external events). The largest initiating event contributor to LERF are Losses of Offsite Power (LOOP) caused by external events (i.e. tornado and breach of the main cooling reservoir). The dominant contributor to LERF is containment bypass caused by induced Steam Generator Tube Rupture. Containment bypass due to ISLOCAs are not as important at STP as typical Westinghouse plants because the Residual Heat Removal system is located inside containment.

The WCAP-15376 analysis and determination of LERF is based on a large dry containment.

South Texas Project Units 1 and 2 are both large dry containments; therefore, the results are applicable to both units.

Therefore, WCAP-15376 results should be considerable applicable to the STP facility.

NOC-AE-14003143 Attachment A to Enclosure 1 Page 3 of 6 Table 1 WCAP-15376 Implementation Guidelines:

Applicability of the Analysis General Parameters WCAP-15376 Analysis Parameter Assumption (Plant) Specific Parameter 1

Logic Cabinet Type (SSPS or Relay) SSPS 2

Component Bypass Test Time

" Analog channels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

" Logic cabinets (SSPS or Relay Protection System) (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for SSPS or 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Relay Protection System)

" Master Relay (SSPS or Relay Protection System) (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for SSPS or 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Relay Protection System)

  • Reactor trip breakers 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3

Component Test Interval

  • Reactor trip breakers 2 months 18 months (Note A) 4 Typical At-Power Maintenance Intervals
  • Reactor trip breakers 12 months 18 months (refueling outage activity)

Plant procedures are in place for the following operator actions5

" Reactor trip from the main control board switches Credited Credited

" Reactor trip by interrupting power to the motor-generator sets Credited Credited

" Insertion of the control rods via the rod control system Credited Credited

" Safety injection actuation from the main control board switches Credited (Note B)

  • Safety injection by actuation of individual components Credited (Note B)

" Auxiliary feedwater pump start Credited (Note B)

NOC-AE-14003143 Attachmuent A to Enclosure 1 Page 4 of 6 WCAP-15376 Analysis Parameter Assumption (Plant) Specific Parameter 6 Credited for AFW pump start Credited for AFW pump start AMSAC Total Transient Event Frequency 7 3.6 1.24 ATWS Contribution to CDF (current PRA model) 8 1.OE-06/yr 4.30E-08/yr 9 3.60E-06/yr Total CDF from Internal Events (current PRA model) --

Total LERF from Internal Events (current PRA model) 9 -- 1.80E-07yr Notes for Table 1:

1. Both types of logic cabinet, SSPS and Relay are included in WCAP-15376, therefore the analysis is applicable to STP.
2. Since STP current Tech Spec bypass test times are equal to or less than those used in WCAP-15376, the analysis is applicable to STP.
3. Since STP current Tech Spec test interval is equal to or greater than that used in WCAP-15376, the analysis is applicable to STP.
4. Since the typical maintenance interval at STP greater than that used in WCAP-153 76, the analysis is applicable to STP.
5. Since plant procedures are in place to perform these actions, the WCAP-15376 analysis is applicable to STP. [see Note B below]
6. Since AMSAC will initiate AFW pump start, the WCAP-15376 analysis is applicable to STP.
7. This entry includes the total frequency for initiators requiring a reactor trip signal to be generated for event mitigation to assess the importance of ATWS events to CDF. Events initiated by a reactor trip are not included. Since the STP value is less than the WCAP-15376 value, this analysis is applicable to STP.
8. This entry indicates the ATWS contribution to core damage frequency (from at-power, internal events) to determine if the ATWS event is a large contributor to CDF.
9. This entry indicates the total CDF and LERF from internal events (including internal flooding) for the most recent PRA model update. This is required for comparison to the NRC's risk-informed CDF and LERF acceptance guidelines in Regulatory Guide 1.174. Sections 8.4.3.3 and 8.4.3.4 of WCAP-15376 address the ALERF and ICLERP risk metrics generically.

Note A: Each RTB breaker is tested at a frequency in accordance with the Surveillance Frequency Control Program. The current frequency is every 9 months on a staggered test basis so that each RTB breaker is tested every 18 months.

Note B: These actuations are engineered safety features. The proposed changes do not impact engineered safety featuresTechnical Specifications

NOC-AE- 14003143 Attachment A to Enclosure 1 Page 5 of 6 The results in Table 1 demonstrate that the total CDF is much less than the 1E-04 limit and the total LERF is much less than the lE-05 limit that is discussed in RG 1.174 for making risk-informed changes. Therefore, the risk metrics support the change in RTB channel inoperability time to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the change in RTB bypass time from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The STP PRA, used in support of this submittal, has been extensively reviewed by both STP probabilistic safety assessment (PRA) staff and outside PRA experts, including the WOG Peer Review process. The STP PRA was also reviewed by the NRC as part of the process for approving Risk Managed Technical Specifications (RMTS) at the South Texas Project.

STP considers the issues identified as a result of these reviews, including the Peer Review process, have been adequately addressed.

Safety Evaluation Condition 4 for WCAP-15376-P-A In order to address Safety Evaluation (SE) Condition 4 for WCAP- 15365-P-A, Westinghouse issued implementation guidelines for licensees to confirm the analyses are applicable to their plant.

The model assumptions for human reliability in WCAP-15376-P, Rev. 0 should be confirmed to be applicable to the STP facility.

Tablt 4 provides the results of the applicability of the human reliability analysis.

Table 4 WCAP-15376 Implementation Guidelines:

Applicability of the Human Reliability Analysis Operator Action that results in a success path (backup to the automatic Are Plant function) prior to the Procedures in action becoming ineffective Place for the OperatorAction to mitigate the event?' Action?'

Reactor trip from the main control board switches Yes Yes Reactor trip by interrupting power to the motor- Yes Yes generator sets Insertion of the control rods via the rod control system Yes Yes Safety injection actuation from the main control board (Note A) (Note A) switches Safety injection by actuation of individual (Note A) (Note A) components Auxiliary feedwater pump start (Note A) (Note A)

NOC-AE-14003143 Attachment A to Enclosure 1 Page 6 of 6 Note for Table 4

1. Since "yes" is filled in for both questions, the analysis is applicable to STP.

See Note A.

Note A: These actuations are engineered safety features. The proposed changes do not impact engineered safety features Technical Specifications Therefore, the model assumptions for human reliability in WCAP-1 5376 is applicable to the STP facility.

NOC-AE- 14003143 Enclosure 2 ENCLOSURE 2 Technical Specification Page Markups Technical Specification pages 3-4, 3-8 and new page 3-8a reflect the proposed changes.

Part of Technical Specification page 3-6 is provided for information. No changes proposed for this page.

NOC-AE-14003143 Enclosure 2 Page 1 of 4 Table 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

18. Safety Injection Input from ESFAS 2 1 2 1,2 9A
19. Reactor Trip System Interlocks
a. Intermediate Range Neutron Flux, P-6 2 1 2 8
b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 8 or 1 P-13 Input 2 1 2 8
c. Power Range Neutron Flux, P-8 4 2 3 1
d. Power Range Neutron Flux, P-9 4 2 3 8
e. Power Range Neutron Flux, P-10 4 2 3 1,2 8
f. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8
20. Reactor Trip Breakers 2 1 2 1,2 9,12 2 1 2 3",4",5" 10302 SOUTH TEXAS - UNITS 1 &2 3/4 3-4 Unit 1 -Amendment No.9 Unit 2 - Amendment No.

NOC-AE-14003143 Enclosure 2 Page 2 of 4

[No Changes on this page]

Table 3.3-1 (Continued)

TABLE NOTATIONS

  • When the Reactor Trip Systems breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.

SOUTH TEXAS - UNITS 1 &2 3/4 3-6

NOC-AE-14003143 Enclosure 2 Page 3 of 4 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. For Functional Units with installed bypass test capability, the inoperable channel may be placed in bypass, and must be placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Note: A channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing per Specification 4.3.1.1, provided no more than one channel is in bypass at any time.

b. For Functional Units with no installed bypass test capability,
1. The inoperable channel is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and
2. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - (Not Used)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, W-i set-t! tl ffn4.era1caRan rtoA :R . a so be in at least HOT STANDBY within W 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to f hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 9A- a. With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable channel to OPERABLE status, or apply

  • the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
b. With the number of OPERABLE channels more than one less then the Minimum Channels OPERABLE requirement, within I hour restore at least one inoperable channel to OPERABLE status or apply the requirements of the CRMP, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 10- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status o nrti ec, ie Aulise f

!o.th~in 48t~ e tAaA,7iýl within the next hour.

ACTION 11 - (Not Used)

ACTION 12- With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it tl,,T** .,

to OPERABLE status within 48!hour e. '*r.ino t*

,*.*_*[

  • ,*,m*,.*#* ,

.,*'***'*~it . *.The breaker shal not be bypassdw one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

SOUTH TEXAS - UNITS 1 & 2 3/4 3-8 Unit 1 - Amendment No. 136 17AL-7R Unit 2 - Amendment No. 125 166"'!

NOC-AE-14003143 Enclosure 2 Page 4 of 4 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

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ýt ~div' Aei feMRatM Tn o~fh de Molt e hut trT tiTb5chnMMOinw Fr iao eW e'i ithrQ SOUTH TEXAS - UNITS 1 & 2 3/4 3-8a Unit 1 - Amendment No.

Unit 2 - Amendment No.

NOC-AE-14003143 Enclosure 3 ENCLOSURE 3 Technical Specification Bases Inserts (For information only)

All information except for system headers is new.

NOC-AE- 14003143 Enclosure 3 Page 1 of 3 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Reactor Trip Breakers Two reactor trip breakers (RTB) arranged in series connect three-phase ac power from the rod drive motor generator sets to the rod drive power cabinets supplying power to the control rod drive mechanisms (CRDM). Opening of the RTBs interrupts power to the CRDMs and allows the shutdown rods and control rods to fall into the core by gravity. Each RTB is equipped with a bypass breaker to allow testing of the RTB while the unit is at power.

During normal operation .the output from the solid state protection system (SSPS) provides a direct voltage signal to the undervoltage coil on each reactor trip breaker and bypass breakers, if in use. Direct current holds a trip plunger out against its spring, allowing ac power to be available at the rod drive power cabinets. SSPS consists of two logic trains, each capable of opening a separate and independent reactor trip breaker. SSPS takes binary inputs (i.e. voltage or no-voltage) from the process and nuclear instrumentation channels corresponding to conditions of plant parameters. When a required logic combination is completed, a reactor trip signal (i.e. no voltage) is generated to the undervoltage trip coil. In addition, the reactor trip signal energizes the shunt trip auxiliary relay coils of the RTBs to trip the breakers open. The shunt trip auxiliary relay coils provide a diverse means to trip the RTBs. When any one train of RTBs is taken out of service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed.

The LCO for Table 3.3-1, Functional Unit 20 requires two OPERABLE channels (trains) of trip breakers. A trip breaker channel (train) consists of the normal trip breaker associated with a single RTS logic train that are racked in, closed, and capable of supplying power to the Rod Control System. Two OPERABLE trains ensure no single random failure can disable the RTS trip capability.

The LCO for Table 3.3-1, Functional Unit 20 requires both diverse trip features (undervoltage and shunt trip attachment) to be OPERABLE for each RTB that is in service. The diverse trip features are not required to be OPERABLE for trip breakers that are open, racked out, incapable of supplying power to the Control Rod Drive System, or declared inoperable. OPERABILITY of both diverse trip mechanisms on each breaker ensures that no single trip mechanism failure will prevent opening any breaker on a valid signal.

The diverse trip features for Functional Unit 20 must be OPERABLE in MODE 1 or 2. In MODE 3, 4, or 5, the diverse trip feature for Functional Unit 20 must be OPERABLE when the RTBs are closed and the Control Rod Drive System is capable of rod withdrawal.

ACTION 9 applies to the RTB trains in MODES 1 and 2. This action addresses the train orientation of the Reactor Trip System for the RTBs. With one train inoperable, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are

NOC-AE-14003143 Enclosure 3 Page 2 of 3 allowed for train corrective maintenance to restore the train to OPERABLE status or the Unit must be placed in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed to restore the inoperable RTB train to OPERABLE status is justified in WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times", March 2003. The completion time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reach MODE 3 from full power in an orderly manner and without challenging Unit systems is reasonable based on operating experience. With the Unit in MODE 3, ACTION 10 would apply to any inoperable RTB trip mechanism. ACTION 9 is modified to allow bypassing one train up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing, provided the other train is OPERABLE. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time limit for RTB testing is justified in WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times", March 2003. The Conditions and Limitations of the NRC Safety Evaluation for WCAP-15376, published as part of WCAP-15376-P-A, Revision 1, for extending the allowed outage time and bypass test time for RTBs are met.

ACTION 12 applies to the RTB diverse trip features (undervoltage and shunt trip attachment) in MODES 1 and 2. With one of the diverse trip features inoperable, it must be restored to an OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the unit must be placed in a MODE where the requirement does not apply. This is accomplished by placing the unit in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> total time). The completion time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging Unit systems. With the Unit in MODE 3, ACTION 10 would apply to any inoperable RTB trip feature. The affected RTB shall not be bypassed while one of the diverse features is inoperable except for the time required to perform maintenance to one of the diverse features.

The allowable time for performing maintenance of the diverse features is limited by the ACTION 9 allowance for one channel to be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing. The AOT of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is reasonable considering that in this Condition there is one remaining diverse feature for the affected RTB, and one OPERABLE RTB capable of performing the safety function and given the low probability of an event occurring during this interval.

ACTION 10 applies to the following reactor trip functions in MODE 3, 4, or 5 when the RTBs are in the closed position and Control Rod Drive System capable of rod withdrawal:

  • Functional Unit 20, RTBs that include diverse trip features, and
  • Functional Unit 21, Automatic Trip and Interlock Logic With one channel or train inoperable, the inoperable channel or train must be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If the affected Function(s) cannot be restored to OPERABLE status within the allowed 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> allowed outage time (AOT), the Unit must be placed in a MODE in which the requirement does not apply. To achieve this status, action must be initiated within the same 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to ensure that all rods are fully inserted, and the Rod Control System must be placed in a condition incapable of rod withdrawal within the next hour.

The additional hour provides sufficient time to accomplish the action in an orderly manner.

With rods fully inserted and the Control Rod Drive System incapable of rod withdrawal, these Functions are no longer required. The AOT is reasonable considering that in this Condition, the

NOC-AE-14003143 Enclosure 3 Page 3 of 3 remaining OPERABLE train is adequate to perform the safety function, and given the low probability of an event occurring during this interval.

ACTION 12A applies to the RTB diverse trip features (undervoltage and shunt trip attachment) in MODE 3, 4, or 5 when the RTBs are in the closed position and Control Rod Drive System capable of rod withdrawal.

NOC-AE- 14003143 Enclosure 4 Enclosure 4 List of Commitments

NOC-AE-14003143 Enclosure 4 List of Commitments The following table identifies those actions committed to by STPNOC in this document. Any statements in this document with the exception of those in the table below are provided for information purposes and are not considered commitments. Please direct questions regarding these commitments to Wendy Brost at (361) 972-8516.

Commitment Scheduled Completion Date I Procedures will be revised or created to ensure activities that Upon amendment degrade the availability of reactor coolant system pressure implementation relief, auxiliary feedwater flow, ATWS Mitigating System Actuation Circuitry (AMSAC), and turbine trip will not be CR 11-2937-4 scheduled when a reactor trip breaker is out-of-service.

2 Procedures will be revised or created to ensure activities that Upon amendment could degrade other components of the reactor protection implementation system including master relays, slave relays, and analog channels will not be scheduled concurrently with a logic cabinet CR 11-2937-5 out of service.

3 Procedures will be revised or created to ensure activities on Upon amendment electrical systems (e.g. AC and DC power) that support the implementation functionality of the following systems or components will not be scheduled when a reactor trip breaker is unavailable: CR 11-2937-6

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