ML23019A253

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Draft of Revision 2 of Traveler TSTF-576, Revise Safety/Relief Valve Requirements
ML23019A253
Person / Time
Site: Technical Specifications Task Force
Issue date: 06/08/2022
From:
Technical Specifications Task Force
To:
Office of Nuclear Reactor Regulation
References
TSTF-576, Rev. 2
Download: ML23019A253 (1)
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Text

DRAFT TSTF-576, Rev. 2 Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler Revise Safety/Relief Valve Requirements NUREGs Affected: 1430 1431 1432 1433 1434 2194 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Improvement NRC Fee Status: Not Exempt Benefit: Increases Equipment Operability Changes Marked on ISTS Rev 5.0 PWROG RISD & PA (if applicable): N/A N/A See attached.

Revision History TSTF Revision 0 Revision Status: Closed Revision Proposed by: BWROG LC Revision

Description:

Original Issue On August 27, 2019, the TSTF provided for NRC comment draft traveler TSTF-576, "Revise Safety/Relief Valve Requirements." A presubmittal meeting was held on September 12, 2019 and on October 21, 2019 the TSTF provided a revised draft. An additional presubmittal meeting was held on December 2, 2019.

On December 13, 2019, the TSTF submitted for NRC review Revision 0 of TSTF-576 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19347A726).

Owners Group Review Information Date Originated by OG: 12-Jul-19 Owners Group Comments Presubmittal meeting held September 12, 2019. Revised traveler distributed to BWROG on October 7.

Owners Group Resolution: Approved Date: 02-Aug-19 TSTF Review Information TSTF Received Date: 03-Dec-19 Date Distributed for Review 03-Dec-19 TSTF Comments:

A presubmittal meeting was held with the NRC on September 12, 2019. A revised draft was developed and submitted to the NRC on October 21. A presubmittal teleconference was held on December 2. The traveler was finalized addressing the NRC comments.

TSTF Resolution: Approved Date: 12-Dec-19 NRC Review Information NRC Received Date: 13-Dec-19 08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2 TSTF Revision 0 Revision Status: Closed NRC Comments:

A presubmittal meeting was held with the NRC on September 12, 2019. A revised draft was developed and submitted to the NRC on October 21. A presubmittal teleconference was held on December 2. The traveler was finalized addressing the NRC comments and submitted December 13, 2019.

The traveler was revised to reflect responses to the NRC's May 11, 2020 Request for Additional Information.

Final Resolution: Superceded by Revision TSTF Revision 1 Revision Status: Closed Revision Proposed by: BWROG Revision

Description:

On May 11, 2020, the NRC provided a Request for Additional Information (RAI) regarding TSTF-576 (ADAMS Accession Number ML19351D783).

On August 11, 2020, the TSTF provided a draft RAI response and revised traveler for NRC comment. A teleconference to discuss the NRC's comments as held on October 13, 2020, followed by a teleconference audit on December 7, 2020.

On February 1, 2021, a revised RAI response and traveler were provided to the NRC for comment. The NRC provided comments on April 12, 2021 and a teleconference to discuss the NRC comments was held on May 13, 2021.

TSTF-576 was also revised to be based in the completed but not yet published Revision 5 of the Standard Technical Specifications. This did not result in any changes.

The significant changes in Revision 1 are:

1.Searched the ITS for "Safety/Relief" and "S/RV" and replaced statements that the S/RVs are used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits with a statement that the Overpressure Protection System is used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

2.Expanded justification of elimination of manual actuation test and added statement that licensees adopting the traveler have an adequate FME program.

3.Removed the BWR/4 NUREG addition of an SR on relief mode S/RVs that is only applicable to Dresden and Quad Cities. Added an acceptable variation for those plants to retain their existing SR and Actions.

4.Restored an LCO Bases description of the ASME Code requirements.

5.Revised the Required Action A.1 removal justification based on a review the S/RV TS for all BWRs.

6.Added a paragraph to the Bases regarding the removal of the SR 3.4.3.1 low tolerance value.

7.Revised and expanded the justification for removal of the manual actuation SR based on a review the S/RV TS for all BWRs.

8.Many editorial revisions and improvements.

08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2 TSTF Revision 1 Revision Status: Closed Owners Group Review Information Date Originated by OG: 21-May-21 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 23-Jun-21 TSTF Review Information TSTF Received Date: 07-Jun-21 Date Distributed for Review 07-Jun-21 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 23-Jun-21 NRC Review Information NRC Received Date: 23-Jun-21 NRC Comments:

First draft provided to NRC on 8/11/2020. The draft was revised based on NRC comments on the draft RAI response.

An audit was held on December 17, 2020. The NRC provided additional questions following the audit.

On February 1, 2021, a revised RAI response and traveler were provided to the NRC for comment. The NRC provided comments on April 12, 2021 and a teleconference to discuss the NRC comments was held on May 13, 2021.

A teleconference was held on November 26, 2021 in which the NRC stated they cannot accept the traveler as written.

Final Resolution: NRC Requests Changes: TSTF Considering TSTF Revision 2 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

The TSTF and NRC held a workshop on Feburary 1, 2022 and the NRC suggested an alternative approach, which is shown in this revision:

The as-left S/RV setpoints are retained in an SR.

The as-found S/RV setpoints are moved to the COLR.

Owners Group Review Information Date Originated by OG: 06-May-22 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 23-May-22 TSTF Review Information 08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2 TSTF Revision 2 Revision Status: Active TSTF Received Date: 24-May-22 Date Distributed for Review 24-May-22 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 08-Jun-22 Affected Technical Specifications S/A 2.1.2 Bases RCS Pressure SL S/A 3.1.4 Bases Control Rod Scram Times S/A 3.3.1.1 Bases RPS Instrumentation S/A 3.3.4.2 Bases ATWS-RPT Instrumentation S/A 3.4.12 Reactor Steam Dome Pressure NUREG(s)- 1433 1434 Only Bkgnd 3.6.1.6 Bases LLS Valves SR 3.3.6.3.7 Bases LLS Instrumentation NUREG(s)- 1433 Only 3.4.3 S/RVs NUREG(s)- 1433 Only Change

Description:

Specification renamed Overpressure Protection System 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Specification renamed Overpressure Protection System Bkgnd 3.4.3 Bases S/RVs NUREG(s)- 1433 Only S/A 3.4.3 Bases S/RVs NUREG(s)- 1433 Only LCO 3.4.3 S/RVs NUREG(s)- 1433 Only LCO 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Appl. 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Action 3.4.3.A S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.A Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.B S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.B Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted 08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2 Action 3.4.3.C S/RVs NUREG(s)- 1433 Only Change

Description:

Revised and renamed "A" Action 3.4.3.C Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Revised and renamed "A" SR 3.4.3.1 S/RVs NUREG(s)- 1433 Only SR 3.4.3.1 Bases S/RVs NUREG(s)- 1433 Only SR 3.4.3.2 S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted SR 3.4.3.2 S/RVs NUREG(s)- 1433 Only Change

Description:

New SR SR 3.4.3.2 Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted SR 3.4.3.2 Bases S/RVs NUREG(s)- 1433 Only Change

Description:

New SR Ref. 3.4.3 Bases S/RVs NUREG(s)- 1433 Only S/A 3.4.11 Bases Reactor Steam Dome Pressure NUREG(s)- 1433 Only Bkgnd 3.3.6.5 Bases Relief and LLS Instrumentation NUREG(s)- 1434 Only 3.4.4 S/RVs NUREG(s)- 1434 Only Change

Description:

Specification renamed Overpressure Protection System 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Specification renamed Overpressure Protection System Bkgnd 3.4.4 Bases S/RVs NUREG(s)- 1434 Only S/A 3.4.4 Bases S/RVs NUREG(s)- 1434 Only LCO 3.4.4 S/RVs NUREG(s)- 1434 Only LCO 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Appl. 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Action 3.4.4.A S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.A Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.B S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted 08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2 Action 3.4.4.B Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.C S/RVs NUREG(s)- 1434 Only Change

Description:

Renamed A Action 3.4.4.C Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Renamed A SR 3.4.4.1 S/RVs NUREG(s)- 1434 Only SR 3.4.4.1 Bases S/RVs NUREG(s)- 1434 Only SR 3.4.4.2 S/RVs NUREG(s)- 1434 Only SR 3.4.4.2 Bases S/RVs NUREG(s)- 1434 Only SR 3.4.4.3 S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted SR 3.4.4.3 S/RVs NUREG(s)- 1434 Only Change

Description:

New SR SR 3.4.4.3 Bases S/RVs NUREG(s)- 1434 Only Change

Description:

New SR SR 3.4.4.3 Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Ref. 3.4.4 Bases S/RVs NUREG(s)- 1434 Only 08-Jun-22 Copyright(C) 2022, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

DRAFT TSTF-576, Rev. 2

1.

SUMMARY

DESCRIPTION The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The proposed change modifies NUREG-1433, "Standard Technical Specifications, General Electric BWR/4 Plants," and NUREG-1434, "Standard Technical Specifications, General Electric BWR/6 Plants."1

2. DETAILED DESCRIPTION 2.1. System Design and Operation The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requires the reactor pressure vessel to be protected from overpressure during upset conditions by self-actuated safety valves. The overpressure protection requirements dictate the size and number of S/RVs that are needed such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB) under the most severe transients. Section 5.2.2, "Overpressure Protection," of NUREG-0800, "Standard Review Plan,"

describes the typical requirements for the overpressure protection system for boiling water reactor (BWR) plants.

Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, "General Design Criteria," (GDC), criterion 15 "Reactor coolant system design," states, "The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences." While many of the operating plants are not committed to the Appendix A GDC, most plants are committed to a similar design requirement as described in their Updated Final Safety Analysis Report (UFSAR).

The overpressure protection system for a BWR utilizes the S/RVs. The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded disk or pilot valve opens when steam pressure overcomes the spring force holding the valve or pilot valve closed. For S/RVs with pilot valves, opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. In the relief mode of operation, pneumatic pressure is used to open the valve, initiated by switches located in the control room or by pressure-sensing instrumentation. Some plants credit a percentage of the total installed S/RV capacity operating via the relief mode for overpressure protection, as permitted by the ASME Code.

1 NUREG 1433 is based on the BWR/4 plant design, but is also representative of the BWR/2, BWR/3, and, in this case, BWR/5 designs. NUREG 1434 is based on the BWR/6 plant design.

Page 1

DRAFT TSTF-576, Rev. 2 2.1.1. S/RV Inservice Testing The S/RVs are tested in accordance with the Inservice Testing (IST) Program, as required by 10 CFR 50.55a(f). Periodic testing is described in Appendix I of the ASME Operations and Maintenance (OM) Code, "Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices," Section I-3300, "Periodic Testing." This testing is performed during a plant shutdown and aspects are performed as a bench test at nominal operating temperatures and pressures. The inservice test verifies each S/RV opens within the required "as-found" tolerance around the setpoint.

Safety/Relief Valve nominal setpoints, as-left tolerance limits, and as-found tolerance limits that appear in the current TS are also established and controlled by the ASME OM Code. ASME OM Code Appendix I Section I-1310(e), states, "The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this Mandatory Appendix." The 2015 edition of the OM Code, section I-1320, "Test Frequencies, Class 1 pressure Relief Valves," paragraph (c), "Requirements for testing additional valves," states, "Additional valves shall be tested in accordance with the following requirements:

(1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of the plus/minus tolerance limit of the Owner-established set-pressure acceptance criteria of sub-para. I-1310(e) or +/- 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group." Other editions of the OM Code have similar requirements. Therefore, additional testing is required by the OM Code if an S/RV fails to open within established acceptance criteria (the owner specified limits or +/-3%) or the as-found tolerance established in the TS.

The ASME Code permits testing 20% of the S/RVs each cycle prior to startup, with the test population expanded if failures are found. Alternatively, all of the S/RVs or pilot valves may be removed and replaced, and the as-found testing is performed within one year after removal.

Following testing, the S/RVs or pilot valves are refurbished, tested, and certified for use. The valves are set to the "as-left" tolerance, which is typically narrower than the as-found criteria to allow for drift during the period of operation.

If an S/RV fails to open within the IST tolerance during as-found testing, the failure is entered into the Corrective Action Program and, according to licensee procedures, evaluated, corrected, and tracked. The extent of condition is also evaluated. Depending on the nature of the failure, the extent of condition could include an evaluation of the ability of the S/RVs to perform their function in the current cycle.

As an example of evaluation of S/RV performance under the Corrective Action Program, in 2016 Southern Company discovered unexpected damage during testing of the S/RVs for Plant Hatch Unit 1. After examination, it was determined that the damage was similar to damage reported in a previous 10 CFR Part 21 report. Extensive extent of condition evaluations were performed on Unit 1 and Unit 2 (see NRC Reactive Inspection Report 05000321/2016009 dated June 10, 2016), which determined the Hatch S/RVs were susceptible to fretting as described in the 10 CFR Part 21 report. As a result, in May of 2016 Southern Company performed a mid-cycle outage on Plant Hatch Unit 2 to replace all eleven S/RVs and to inspect the main valve internals.

Page 2

DRAFT TSTF-576, Rev. 2 2.2. Current Technical Specifications Requirements In addition to the ASME Code requirements, the current TS contain multiple specifications that govern the S/RVs depending on the function they are fulfilling.

  • Safety Limit 2.1.2, "Reactor Coolant System Pressure SL," states, "Reactor steam dome pressure shall be 1325 psig." The pressure limit is plant specific. The S/RVs are credited for meeting this safety limit. Safety Limit 2.1.2 limits the reactor steam dome pressure to the lowest transient overpressure allowed in order to ensure the maximum transient pressure allowable in the RCS pressure vessel is less than the ASME Code,Section III, limit of 110%

of design pressure.

  • BWR/4 and BWR/6 TS 3.6.1.6, "Low-Low Set (LLS) Valves," requires the S/RVs operating in relief mode to be operable. In the LLS mode, a subset of the S/RVs are signaled to open at a lower pressure than the relief or safety mode pressure setpoints and to stay open longer, so that reopening more than one S/RV is prevented on subsequent actuations. The LLS function prevents excessive short duration S/RV cycles.
  • BWR/4 TS 3.3.6.3, "Low-Low Set (LLS) Instrumentation," and BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation," provide instrumentation requirements that support the S/RVs in the LLS mode of operation. For plants that credit S/RVs in relief mode to prevent overpressurization, the LLS Instrumentation TS also provide the instrumentation requirements to support that function.
  • BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4, both titled, "Safety/Relief Valves," require the S/RVs to prevent RCPB overpressurization. For most plants, the most severe pressurization transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). For most BWR/2, BWR/3, BWR/4, and BWR/5 plants, the S/RVs in the safety mode ensure the Safety Limit is not exceeded during normal operation and Anticipated Operational Occurrences (AOOs). For BWR/6 plants and two non-BWR/6 plants (Dresden 2 and 3 and Quad Cities 1 and 2), some S/RVs in relief mode in addition to the S/RVs in safety mode are credited to ensure the ASME Code overprotection limit is protected.

The BWR/4 TS 3.4.3 S/RV LCO typically states, "The safety function of XX S/RVs shall be operable," with the required number of S/RVs (XX) corresponding to the minimum number needed to accommodate the limiting pressure transient without exceeding the Safety Limit Page 3

DRAFT TSTF-576, Rev. 2 using only the safety mode of operation. The LCO may specify fewer S/RVs than are installed in the plant.

The BWR/6 TS 3.4.4 LCO (which is also applicable to two non-BWR/6 plants) requires the safety function of XX S/RVs and the relief function of YY additional S/RVs to be operable.

The required number of S/RVs in the safety mode (XX) and relief mode (YY) varies by plant and may specify fewer S/RVs than are installed.

BWR/4 Surveillance Requirement (SR) 3.4.3.1 and BWR/6 SR 3.4.4.1 require verification of the safety function lift setpoints of the required S/RVs. These SRs reflect the performance of the ASME Code inservice testing and state the number of valves required to open within a specified tolerance (typically 3%) of the given setpoint. The SRs also specify the as-left tolerance (typically 1%) after testing.

BWR/6 SR 3.4.4.2 requires verification that each relief function S/RV actuates on an actual or simulated automatic initiation signal. The two Non-BWR/6 plants that credit the S/RV relief mode have a similar SR.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 verify that each S/RV opens when manually actuated.

2.3. Reason for the Proposed Change The S/RV LCO is written in terms of individual valves, but the specified safety function is based on the combined pressure relieving capacity of a group of the S/RVs. The failure of some valves to open within the SR tolerance typically would not result in the inability of the S/RVs as a group to perform the specified safety function. Therefore, the LCO should be revised to align with the specified safety function.

Testing of the safety mode of each S/RV to ensure actuation within the Code or licensee-established tolerance is required by the IST Program, which is required to be followed by 10 CFR 50.55a(f). It is unnecessary to duplicate this regulatory requirement for each valve in the TS when the result of any individual valve test is not required to meet the specified safety function of the system.

A review of Licensee Event Reports over the last ten years found over forty events in which S/RVs failed to lift within the SR lift pressure tolerance when bench tested. In all cases in which the SR was not met due to setpoint drift, the Licensee Event Reports concluded that the S/RVs as a group would have retained the capability to protect Safety Limit 2.1.2. This represents an unnecessary reporting burden on the licensees for failures that did not affect the ability to perform the specified safety function.

2.4. Description of the Proposed Change The proposed change renames BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 from "Safety/Relief Valves (S/RVs)" to "Overpressure Protection System (OPS)." This title change requires revision to the Table of Contents and a reference in the Bases of BWR/4 TS 3.3.6.3, "Low-Low Set (LLS)

Instrumentation," and BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation."

Page 4

DRAFT TSTF-576, Rev. 2 The proposed change revises the S/RV LCO to require the Overpressure Protection System (OPS) to be operable. The LCO Bases describes an operable OPS as being capable of preventing reactor steam dome pressure from exceeding Safety Limit 2.1.2.

BWR/4 LCO 3.4.3 is revised to state (deletions are struck through; insertions are in italics):

The OPS safety function of the [11] S/RVs shall be OPERABLE.

BWR/6 LCO 3.4.4 is revised to state:

The OPS safety function of the [seven] S/RVs shall be OPERABLE, AND The relief function of [seven] additional S/RVs shall be OPERABLE.

BWR/4 SR 3.4.3.1 is revised to state:

NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Verify the as-left safety function OPS lift pressures setpoints of the [required]

safety/relief valves (S/RVs) are within +/- 1% of the nominal setpoint as follows:

Nominal Number of Setpoint OPS S/RVs (psig)

[4] [1090 +/- 32.7]

[4] [1100 +/- 33.0]

[3] [1110 +/- 33.3]

Following testing, lift settings shall be within +/- 1%.

Page 5

DRAFT TSTF-576, Rev. 2 BWR/6 SR 3.4.4.1 is revised to state:

NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Verify the as-left OPS lift pressures setpoints of the [required] safety/relief valves (S/RVs) are within +/- 1% of the nominal setpoint as follows:

Nominal Number of Setpoint OPS S/RVs (psig)

[8] [1165 +/- 34.9]

[6] [1180 +/- 35.4]

[6] [1190 +/- 35.7]

Following testing, lift settings shall be within +/- 1%.

The current frequency has three options: In accordance with the Inservice Testing Program, [18]

months, or in accordance with the Surveillance Frequency Control Program. The [18] month and Surveillance Frequency Control Program options are deleted.

BWR/6 SR 3.4.4.2 is revised to state:

Verify each [required] relief function S/RV acting in the relief mode S/RV actuates on an actual or simulated automatic initiation signal.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3, which state, "Verify each [required] S/RV opens when manually actuated," are deleted.

A new BWR SR 3.4.3.2 and a new BWR/6 SR 3.4.4.3 is added. It states:

Verify the as-found OPS lift pressures of the [required] S/RVs are within the limits specified in the COLR.

The Frequency is "In accordance with the INSERVICE TESTING PROGRAM."

The changes to the LCO and SRs result in changes to the TS Actions.

BWR/4 and BWR/6 Condition A, "One [or two] [required] S/RV[s] inoperable," and "One

[required] S/RV inoperable," respectively, are deleted.

Condition B, the default action when the Condition A Required Action and associated Completion Time is not met, is no longer required after deletion of Condition A and is removed.

Page 6

DRAFT TSTF-576, Rev. 2 BWR/4 and BWR/6 Condition C, "[Three] or more [required] S/RVs inoperable," and "[Two] or more [required] S/RVs inoperable," respectively, are replaced with a new Condition, "OPS inoperable." The new Condition retains the existing Required Actions to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The TS Bases are revised to reflect the changes to the TS. The Bases of the following specifications state that S/RVs are used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

  • 3.3.1.1, RPS Instrumentation
  • 3.3.4.2, ATWS-RPT Instrumentation
  • BWR/4 3.3.6.3, LLS Instrumentation
  • BWR/6 3.3.6.5, Relief and LLS Instrumentation
  • BWR/4 3.4.11, Reactor Steam Dome Pressure
  • BWR/6 3.4.12, Reactor Steam Dome Pressure
  • 3.6.1.6, LLS Valves These Bases are revised to state that the Overpressure Protection System is used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Part 50.36, states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." A licensee may make changes to the TS Bases without prior NRC review and approval in accordance with the Technical Specifications Bases Control Program.

The proposed TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). Therefore, the Bases changes are provided for information and approval of the Bases is not requested.

A model application is attached. The model may be used by licensees desiring to adopt the traveler following NRC approval.

3. TECHNICAL EVALUATION Specification Name Change.

As discussed in Section 2.2, there are several specifications which provide requirements on the S/RVs. It is confusing to title the specification "Safety/Relief Valves," because that name implies it is the only specification that governs the equipment. Just as TS 3.5.1 refers to the "Automatic Depressurization System (ADS)" function of the S/RVs, it is more appropriate to title BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 "Overpressure Protection System (OPS)," to represent the functional capability required by the specification. Renaming the specifications is consistent with the STS convention that an LCO requires a system to be operable and the LCO Page 7

DRAFT TSTF-576, Rev. 2 Bases describe what is required for the system to capable of performing its specified safety function. The term "overpressure protection system," is not new. The NRC Standard Review Plan (NUREG-0800), Section 5.2.2, is titled, "Overpressure Protection," and many BWR plants have a similar Updated Final Safety Analysis Report (UFSAR) section. In addition, the existing BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 "Applicable Safety Analysis" section of the Bases begins, "The overpressure protection system must accommodate the most severe pressurization transient." As a result, referring to the S/RV overpressure protection function as the "Overpressure Protection System (OPS)," is a clearer representation of the requirement.

The traveler revises the NUREG-1433 and NUREG-1434 Table of Contents to reflect the name change. Plants adopting the traveler will also need to update the TS Table of Contents. For many plants, the Table of Contents is not part of the TS and is not required to be included in an amendment request.

LCO Changes Title 10 of the CFR, Paragraph 50.36(c)(2)(i) states, "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility." However, the existing S/RV LCO does not represent the lowest functional capability required for safe operation. The existing S/RV LCO requires each specified valve to be operable, but the specified safety function is based on the combined pressure relieving capacity of the credited S/RVs (which may be less than the installed complement of valves). The failure of a particular valve or valves to open within the SR tolerance may not (and based on historical performance, is unlikely to) result in the inability of the S/RVs as a group to perform the specified safety function. As a result, the existing LCO does not represent the lowest functional capability or performance level of equipment required for safe operation of the facility. Therefore, the LCO is revised to ensure the safety function of providing overpressure protection, which is consistent with 10 CFR 50.36(c)(2)(i).

TSTF-GG-05-02, "Writers Guide for Plant-Specific Improved Technical Specifications,"

(ADAMS Accession No. ML070660229) Section 4.1.4, "Chapter 3 LCO Content," states, "The LCO describes as simply as possible the lowest functional capability or performance levels of equipment required for safe operation of the facility. ... It is acceptable to generically refer to the system, subsystem, component or parameter which is the subject of the LCO and provide the specific scope/boundaries in the Bases." Following this guidance, the BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 LCOs are revised to require the OPS to be operable. The LCO Bases are revised to state, "The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety [and relief] mode[s] of the S/RVs. The OPERABILITY of the OPS is only dependent on the ability to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs." The phrase "and relief" is bracketed (i.e., plant-specific) in the BWR/4 TS Bases since it is applicable to only two plants. The phrase is not bracketed in the BWR/6 TS since it is applicable to all BWR/6 plants.

The terms "safety function" and "relief function" are used in the existing BWR/4 LCO 3.4.3 and BWR/6 LCO 3.4.4. However, the TS Bases, "Background" section uses the terms "safety mode" and "relief mode." For example, the Bases state, "The S/RVs can actuate by either of two modes: the safety mode or the relief mode," and "The S/RVs that provide the relief mode are the Page 8

DRAFT TSTF-576, Rev. 2 low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves." The term "safety function" could be easily confused with the term "specified safety function" used in the definition of operability. For clarity and for consistency, the TS and Bases are revised to use the terms "safety mode" and "relief mode." This is an administrative change with no change in intent.

The LCO is revised to no longer specify the number of credited operable S/RVs. As stated previously, the overpressure protection function is provided by the collective action of the credited S/RVs, not individual S/RVs. This change is consistent with the required function and the 10 CFR 50.36 requirement that the LCO represent the lowest functional capability required for safe operation of the facility. The proposed Applicable Safety Analysis Bases describe the number of S/RVs credited in the overpressure analyses.

The BWR/6 LCO requires relief mode of operation for a subset of S/RVs. These plant designs permit crediting some of the pressure relieving capability of electrically operated pressure relief valves in the overpressure analysis. Two non-BWR/6 plants, Dresden 2 and 3 and Quad Cities 1 and 2, also credit electrically operated pressure relief valves and changes are proposed to accommodate that design. The revision to the LCO to require the OPS to be operable includes a change to the Bases to describe the role of the S/RVs in relief mode. The Applicability Safety Analysis Bases are revised to state that the S/RVs credited for the relief mode are in addition to those credited in the safety mode portion, consistent with the term "additional" which appears in the current LCO. The required relief mode operation of the S/RVs is verified in an SR.

BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1 Changes The proposed BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1 are similar to the existing SR. The revised SR requires verification that the as-left S/RV lift pressure is within 1% of the nominal setpoint, which is consistent with the existing SR. However, the SR is revised to only address the as-left lift pressure setpoint. The as-found lift pressure verification is moved to the new proposed BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3.

The current SR has an optional Note which states, " [2] [required] S/RVs may be changed to a lower setpoint group." That Note only appears in the TS of one plant and, therefore, should not appear in the standard TS. It is removed.

The current SRs have three optional frequencies: In accordance with the Inservice Testing Program, [18] months, and in accordance with the Surveillance Frequency Control Program.

The as-left S/RV lift pressure testing is required by the ASME Code. Therefore, the Frequency is revised to only reference the Inservice Testing Program.

BWR/6 SR 3.4.4.2 Changes BWR/6 SR 3.4.4.2 states, "Verify each [required] relief function S/RV actuates on an actual or simulated automatic initiation signal." The SR is revised to refer to the "S/RV acting in the relief mode" instead of the "relief function" as previously discussed. The brackets around the word "required" are removed. Brackets indicate a plant-specific option. The equivalent SR in all four BWR/6 plants contains the word "required." Therefore, the brackets are unnecessary and are removed to make the STS consistent with the plant TS.

Page 9

DRAFT TSTF-576, Rev. 2 BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 Elimination The existing SRs state, "Verify each [required] S/RV opens when manually actuated." The Bases for the SRs state that a manual actuation is performed to verify that, mechanically, the valve is functioning properly, and no blockage exists in the valve discharge line. This SR is removed from the TS.

The STS SR is not representative of the plant-specific TS. Ten of the thirty BWR units have no equivalent SR. An additional thirteen of the thirty units have different requirements, such as verifying the actuator strokes (eleven units) or verification that the valve is capable of being opened (two units). Only seven of the thirty units contain an SR equivalent to the STS SR.

There is no safety analysis assumption that the safety mode S/RVs will open when manually actuated to limit overpressure. As a result, the ability to open manually is not required to demonstrate that the valve can perform its safety function. This is supported by the existing SR Bases statement, "If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered operable."

TS SR 3.0.1 states that when an SR is not met, the associated LCO is not met (i.e., the system is inoperable). Therefore, the existing TS Bases are contrary to SR 3.0.1 and state that the LCO is met even if the manual actuation SR isn't met. The manual actuation SR is removed to be consistent with the TS requirements.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 are not needed to verify the S/RVs are mechanically functioning properly. The as-left test of the S/RVs verifies each valve is functioning properly.

Following installation in the plant, post-maintenance testing is performed. For example, the actuator is typically tested in relief mode with the actuator disengaged. This test not only verifies operability of the actuator, but also verifies the pneumatic and electric connections. The actuators are then re-engaged to the valve with appropriate independent verification and sign-offs to verify the S/RV will function properly. The STS does not include post-maintenance or post-installation testing, as described in the SR 3.0.1 Bases. Therefore, an SR is not needed for that purpose.

The SR is not needed to verify the downstream piping is unobstructed. Licensees have robust Foreign Material Exclusion (FME) programs to ensure systems are not contaminated during maintenance. Those programs are routinely inspected under NRC Inspection Manual Procedure 71111.20. Further, in licensing actions the staff agreed that an SR to verify that S/RV downstream piping is not obstructed is unnecessary because of licensee FME controls. The safety evaluation for license amendment 116 for the Hope Creek Generating Station, dated February 10, 1999, stated:

Another difference between the current TS-required stroking and the licensee's proposal is that, when performing the testing in-situ as required by the current TS, the testing verifies that the SRV discharge line is not blocked. However, the licensee stated that there is a Foreign Material Exclusion Program in place at the plant which minimizes the potential of debris blocking the discharge lines such that the possibility of blockage is extremely remote. The staff agrees that there is a very small possibility of blockage of an Page 10

DRAFT TSTF-576, Rev. 2 SRV discharge line as demonstrated by operational history and finds that the licensee has acceptably addressed this concern.

As discussed above, there are nine additional BWR unit TS that do not require manual actuation of the safety mode S/RVs. There is no industry operational history of S/RV downstream piping being obstructed by foreign material and the staff's conclusion regarding the Hope Creek FME program is equally applicable to other plants adopting the proposed change.

Testing of S/RV actuation in relief mode is unaffected by the proposed change. BWR/6 SR 3.4.4.2 requires verification that each required S/RV acting in the relief mode actuates on an actual or simulated signal. The initiating instrumentation of the S/RVs in relief mode will continue to be tested by BWR/6 TS 3.3.6.5, "Relief and Low-Low Set (LLS) Instrumentation."

The two BWR/4 plants that credit S/RVs in relief mode will retain similar requirements.

Manual actuation of other S/RV operating modes will continue to be tested by BWR/4 SR 3.5.1.12 and BWR/6 SR 3.5.1.7, "Verify each ADS valve opens when manually actuated,"

and BWR/4 and BWR/6 SR 3.6.1.6.1, "Verify each LLS valve opens when manually actuated."

In conclusion, the SR to perform a manual actuation of the safety mode S/RVs is not necessary to ensure the specified safety function can be performed.

Proposed BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 New surveillances are proposed to verify that the as-found S/RV lift pressures are consistent with the assumptions in the overpressure analysis. This is similar to the existing requirement in BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1; however, the as-found S/RV lift pressure limits are specified in the Core Operating Limits Report (COLR), as discussed below. The S/RV lift pressure limits specified in the COLR protect the assumptions in the overpressure accident analysis while providing the licensee with the flexibility to revise the limits to reflect actual S/RV testing performance.

The new SRs are applicable to the required S/RVs assumed to actuate as part of the OPS. The as-found testing is performed in accordance with the ASME Code, as also required by 10 CFR 50.55a(f)). The Frequency is in accordance with the Inservice Testing Program.

Under the proposed change, the licensee may justify under 10 CFR 50.59 a representative set of S/RV as-found lift pressure limits based on historical plant operation to be used as inputs to the overpressure analysis. The as-found lift pressure limits are not valve-specific. That is, the limits may be expressed as one or more S/RVs opening within a prescribed limit, and one or more other groups of valves opening at different limits. These groups and associated limits must be consistent with, or more conservative than, the assumptions of the overpressure analysis.

The COLR as-found S/RV lift pressure limits may be greater than the ASME OM Code testing acceptance criteria. However, each S/RV must continue to satisfy the ASME Code requirements, or the S/RV condition must be entered into the Corrective Action Program. The 2015 edition of the OM Code, section I-1320, "Test Frequencies, Class 1 pressure Relief Valves," paragraph (c), "Requirements for testing additional valves," states, "Additional valves shall be tested in accordance with the following requirements: (1) For each valve tested for Page 11

DRAFT TSTF-576, Rev. 2 which the as-found set-pressure (first test actuation) exceeds the greater of the plus/minus tolerance limit of the Owners-established set-pressure acceptance criteria of sub-para. I-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group." Section I-1310(e), "Acceptance Criteria," states, "The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this mandatory Appendix." Other editions of the OM Code have similar requirements.

Periodic testing of S/RVs will still be performed as required by Appendix I of the ASME OM Code, "Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices," Section I-3300, "Periodic Testing." Title 10 of the CFR, Part 50, paragraph 55a, "Codes and standards,"

requires licensees to follow the ASME OM Code. The ASME OM Code testing acceptance criteria and the required testing and associated maintenance will continue to be performed for each S/RV.

The proposed change removes the low tolerance about the as-found lift pressure limit from the TS. The low tolerance is not an assumption in the overpressure analysis and not needed to protect Safety Limit 2.1.2. However, the Inservice Testing Program will continue to confirm that the S/RVs open between the lower and upper tolerances about the setpoint.

The S/RV relief mode setpoints will continue to be specified in BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation," and in some plant-specific TS not based on the STS.

Safety/Relief Valves fall under 10 CFR 50.65 (the Maintenance Rule). Licensee Maintenance Rule programs require establishing performance criteria, monitoring and trending performance, determining the cause of failures, and taking corrective action. Those activities are available for NRC inspection.

The Boiling Water Reactor Owners' Group (BWROG) has been working to improve S/RV performance for many years and has trended the performance of problematic two-stage S/RVs.

The BWROG plans to continue to monitor and improve S/RV performance across the BWR fleet.

Safety/Relief Valves that are removed from the plant for testing are refurbished, certified, and reset to within the as-left tolerance prior to reinstallation in accordance with the ASME Code.

In summary, the proposed BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 S/RV as-found lift pressure limits specified in the COLR will verify that the collective S/RV performance is consistent with the overpressure protection analysis assumptions and will protect the safety limit.

Addition of the S/RV Lift Pressure Limits to the Core Operating Limits Report The proposed change adds the as-found S/RV lift pressure limits to the Core Operating Limits Report (COLR), described in BWR/4 and BWR/6 STS 5.5.3.

The COLR will contain the as-found S/RV lift pressure limits assumed in the overpressure protection analysis. The report provides the limits in terms of the number of S/RVs at each lift pressure limit, but a particular limit is not associated with a particular S/RV. The limits may be Page 12

DRAFT TSTF-576, Rev. 2 greater than the ASME OM Code testing acceptance criteria. This approach permits adjustment of the assumptions used in the overpressure protection analysis to reflect the actual test results while ensuring the overall S/RV performance is consistent with the analysis assumptions.

The COLR may include multiple combinations of as-found S/RV lift pressure limits as long as each combination satisfies the overpressure protection analysis assumptions. For example, one combination may assume that all S/RVs open within 3% of the lift setpoint, while another combination may assume that a one S/RV opens within 5% of the lift setpoint and the remainder open at 2.5% of the lift setpoint. The licensee may include as many analyzed combinations as needed to reflect the actual or analyzed test results.

The COLR Specification, paragraph a, contains a Reviewer's Note that directs listing the individual specifications that reference the COLR. The STS directions are not affected by the proposed change. However, plant-specific amendments to adopt the proposed change will add the Overpressure Protection System specification to paragraph a.

The COLR Specification, paragraph b, states that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. A Reviewer's Note states that the NRC-approved methods must be listed. The STS directions are not affected by the proposed change. For most licensees, the methodology for performing overpressure protection analyses is already listed in the COLR. However, if it is not the plant-specific amendment to adopt the proposed change must add the methodology reference to paragraph b.

The COLR Specification, paragraph c, states, "The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met." This paragraph will also apply to the S/RV as-found pressure limits. If a licensee should specify S/RV as-found lift pressure limits that will result in an increase in the peak reactor pressure in an overpressure event, the licensee must also consider other effects resulting from the new limits as required by paragraph c. Note that paragraph c states that "all applicable limitsof the safety analysis are met." In addition to the parenthetical examples in paragraph c, an increase in the peak reactor pressure in an overpressure event could also affect the containment, the S/RV piping, the Reactor Core Isolation Cooling System, and the Standby Liquid Control System. These effects fall within the requirement of existing paragraph c, and it is unnecessary to list every possible effect in the examples.

The COLR will not specify a particular NRC-approved methodology to perform the evaluations described in paragraph c because a licensee may not have such an NRC-approved methodology in their licensing and design basis for a particular evaluation. In addition, a licensee may never need to choose as-found S/RV lift pressure limits that result in an increase in the peak reactor pressure in an overpressure event. Therefore, establishing such a methodology is unnecessary.

The COLR, including any midcycle revisions or supplements, must be provided on issuance for each reload cycle to the NRC. This informs the NRC as to the S/RV as-found lift pressure limits.

A plant-specific amendment to adopt the proposed change will include a markup of the revised Page 13

DRAFT TSTF-576, Rev. 2 COLR, for information, illustrating the presentation of the as-found pressure limits. A revised COLR will be issued and provided to the NRC prior to implementation of the license amendment.

Action Changes Existing BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4, Condition A, apply when one or two [required]

S/RVs are inoperable. The Action requires the inoperable S/RV(s) to be restored within 14 days, followed by a plant shutdown.

NUREG-1433 and NUREG-1434 contain an Action for one or two inoperable required safety mode S/RVs, but no BWR plant TS contain equivalent requirements. Three plants (Hatch, Hope Creek, and Monticello) contain an Action that permits one or two required S/RVs to be inoperable, but the corresponding LCOs require more S/RVs to be operable than are credited in the overpressure analysis. As a result, for those plants the assumptions of the overpressure analysis can still be met when one or two safety mode S/RVs are inoperable. This is consistent with the STS Required Action A.1 Bases which states, "With the safety function of one [or two]

[required] S/RV[s] inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only." (Note that Hatch was the lead plant for the development of NUREG-1433 and this why the Required Action Bases reflect the Hatch requirements.)

The remaining BWR plants' current LCOs only require operability of the number of S/RVs assumed in the overpressure analysis. There are no additional S/RVs required to be operable.

Therefore, if one of the required S/RVs is inoperable, the assumptions of the overpressure analysis cannot be met, and a plant shutdown is required. Similar to those plants' TS, the proposed Overpressure Protection System LCO and SRs do not require additional S/RVs beyond what is required to perform the overpressure protection function. Therefore, Required Action A.1 is not applicable, and it is removed.

The existing BWR/6 TS 3.4.4 Actions are not consistent with the plant TS of the four BWR/6 plants. TS 3.4.4 for the BWR/6 plants (River Bend, Grand Gulf, Perry, and Clinton) contains a single action for one or more required safety mode or relief mode S/RVs inoperable, and it requires being in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, which is consistent with the proposed change. Therefore, the BWR/6 STS is revised to be consistent with the plant-specific BWR/6 TS.

Should it be determined that an S/RV is degraded, an evaluation under the Corrective Action Program would be required to determine whether 1) the associated SRs are met, and, if so, 2) whether the Overpressure Protection System is operable. If an SR is not met or if the OPS is otherwise inoperable, the associated Action must be taken.

Page 14

DRAFT TSTF-576, Rev. 2 Existing BWR/4 TS 3.4.3, Condition B, applies when the Required Action and associated Completion Time of Condition A are not met. As Condition A is deleted, Condition B is no longer needed and is also deleted.

Existing BWR/4 TS 3.4.3, Condition C, applies when [three] or more [required] S/RVs are inoperable. The Condition is renumbered Condition A and revised to state, "OPS inoperable."

The existing Required Actions to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> are retained.

Testing to verify that the OPS is operable is typically performed during a shutdown when the LCO is not applicable. Therefore, it is worthwhile to consider how the Actions would be applied at power. As an example, consider a unit operating at 100% power. A vendor bulletin is received that identifies several installed S/RVs have faulty parts that could change the S/RV lift pressure. Under the Corrective Action Program, the licensee must evaluate whether SRs are still met and whether the Overpressure Protection System is operable. If the SRs are not met or if the OPS is not operable, the Actions must be entered, and a plant shutdown is required.

As an additional example, consider a unit operating at 100% power. During the previous outage, all of the S/RVs were removed for testing and replaced with refurbished S/RVs. The ASME Code testing of the removed S/RVs is completed and the as-found lift pressures for some of the S/RVs was not within the SR limit specified in the COLR. If an evaluation of the deficient condition performed under the Corrective Action Program determines that the affected S/RVs did not satisfy the SR while the plant was operating within the Applicability, a Licensee Event Report documenting that the TS were not followed is required.

4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(b) requires:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"]. The Commission may include such additional technical specifications as the Commission finds appropriate.

Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for an amendment must be filed with the Commission, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.

Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate.

Section IV, "The Commission Policy," of the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58FR39132), dated July 22, 1993, Page 15

DRAFT TSTF-576, Rev. 2 states in part that improved STS have been developed and will be maintained for each NSSS owners group. The Commission Policy encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications." The industry's proposal of travelers and the NRC's approval of travelers is the method used to maintain the improved STS as described in the Commission's Policy. Following NRC approval, licensees adopt travelers into their plant-specific technical specifications following the requirements of 10 CFR 50.90. Therefore, the traveler process facilitates the Commission's policy while satisfying the requirements of the applicable regulations.

The regulation at 10 CFR 50.36(a)(1) also requires the application to include a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls.

4.2. Conclusions In conclusion, based on the considerations discussed above, the proposed revision does not alter the current manner of operation and (1) there is reasonable assurance that the health and safety of the public will not be endangered by continued operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

5. REFERENCES None.

Page 16

DRAFT TSTF-576, Rev. 2 Model Application

DRAFT TSTF-576, Rev. 2

[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 PLANT NAME DOCKET NO. 50-[xxx]

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-576, "Revise Safety/Relief Valve Requirements" Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) TS to align the overpressure protection requirements with the safety limits and the regulations.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. Attachment 3 provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only. provides an example revised Core Operating Limits Report, for information, that illustrates the addition of the S/RV limits.

[LICENSEE] requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within six months of acceptance. Once approved, the amendment shall be implemented within [30] days.

There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.

[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.

Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]

Page 1

DRAFT TSTF-576, Rev. 2 If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].

Sincerely,

[Name, Title]

Enclosure:

Description and Assessment Attachments: 1. Proposed Technical Specification Changes (Mark-Up)

2. Revised Technical Specification Pages
3. Proposed Technical Specification Bases Changes (Mark-Up) - For Information Only
4. Example Updated Core Operating Limits Report - For Information Only

[The attachments are to be provided by the licensee and are not included in the model application.]

cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2

DRAFT TSTF-576, Rev. 2 ENCLOSURE DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation

[LICENSEE] has reviewed the safety evaluation for TSTF-576 provided to the Technical Specifications Task Force in a letter dated [DATE]. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-576. As described herein,

[LICENSEE] has concluded that the justifications presented in TSTF-576 and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.

2.2 Optional Changes and Variations

[LICENSEE is not proposing any variations from the TS changes described in TSTF-576 or the applicable parts of the NRC staffs safety evaluation.] [LICENSEE is proposing the following variations from the TS changes described in TSTF-576 or the applicable parts of the NRC staffs safety evaluation: describe the variations]

[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-576 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-576 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-576 to the [PLANT]

TS.]

[{The following is only applicable to Dresden 2 and 3 and Quad Cities 1 and 2} The [PLANT]

overpressure analysis credits some S/RVs in relief mode in addition to the S/RVs in safety mode to ensure the ASME Code overpressure limit is protected. Existing SR [3.4.4.2] that verifies each required relief function S/RV actuates on an actual or simulated automatic initiation signal is retained. The reference to "relief function" is replaced with "relief mode" to be consistent with the terminology in the Bases.]

[The [PLANT] TS contain requirements that differ from the Standard Technical Specifications on which TSTF-576 was based but are encompassed in the TSTF-576 justification. [Describe differences and why TSTF-576 is still applicable.))

[The Table of Contents of the [PLANT] TS is licensee-controlled and is not included in Attachments 1 and 2.]

Page 3

DRAFT TSTF-576, Rev. 2

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The Limiting Condition for Operation (LCO) is revised to replace requirement on each credited S/RV with a requirement that the Overpressure Protection System (OPS) be operable. The Surveillance Requirements are revised to move the as-found S/RV lift pressure limits to the licensee-controlled Core Operating Limits Report. An SR that tests the ability of the S/RVs to be capable of manual operation is removed as that capability is not credited in any safety analysis. [An SR that verifies the ability of credited S/RVs acting in the relief mode is revised to be consistent with the revised LCO.] The TS Actions are revised to be consistent with the changes to the LCO and SRs. Administrative changes are made to the TS for clarity and consistency. The Core Operating Limits Report specification is revised to reference the Overpressure Protection System specification.

[LICENSEE] has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position. Therefore, the probability of the event is unaffected.

The consequences of the accident are based on the peak reactor pressure vessel pressure.

Both the current and proposed TS ensure the overpressure Safety Limit is not exceeded.

The accident analyses consider the aggregate operation of the credited S/RVs, not the performance of individual valves. The proposed change moves the S/RV as-found lift pressure limits to the licensee controlled Core Operating Limits Report which uses NRC-approved methodologies. Altering the control process for these values has no effect on the accident evaluations. As a result, the consequences of the accident are not changed.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 4

DRAFT TSTF-576, Rev. 2

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

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change does not alter the design function or operation of the S/RVs. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing basis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change ensures that the S/RVs can protect Safety Limit 2.1.2. Although the as-found S/RV lift pressure limits are moved to the licensee-controlled Core Operating Limits Report, the safety margin provided by the S/RVs, which ensures the Safety Limit is protected, is not changed. The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change. The proposed change does not alter a design basis limit or a safety limit, and, therefore, does not reduce the margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, [LICENSEE] concludes that the proposed change 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.

3.2 Conclusion In conclusion, 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 manner, (2) such activities will be conducted in compliance with the Commissions 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.

4.0 ENVIRONMENTAL EVALUATION 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 Page 5

DRAFT TSTF-576, Rev. 2 amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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.

Page 6

DRAFT TSTF-576, Rev. 2 Technical Specifications and Bases Changes

TSTF-576, Rev. 2 DRAFT TABLE OF CONTENTS Page 3.3 INSTRUMENTATION (continued) 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ......................... 3.3.5.1-1 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ....................................................................................... 3.3.5.2-1 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation ................. 3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation ........................................... 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation....................................... 3.3.6.2-1 3.3.6.3 Low-Low Set (LLS) Instrumentation ............................................................. 3.3.6.3-1 3.3.7.1 [ Main Control Room Environmental Control (MCREC) ] System Instrumentation ....................................................................................... 3.3.7.1-1 3.3.8.1 Loss of Power (LOP) Instrumentation .......................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ...................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ........................................................................ 3.4.1-1 3.4.2 Jet Pumps ....................................................................................................... 3.4.2-1 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ........ 3.4.3-1 3.4.4 RCS Operational LEAKAGE ........................................................................... 3.4.4-1 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage ................................................. 3.4.5-1 3.4.6 RCS Leakage Detection Instrumentation ........................................................ 3.4.6-1 3.4.7 RCS Specific Activity ....................................................................................... 3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .................................................................................................. 3.4.8-1 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown .................................................................................................. 3.4.9-1 3.4.10 RCS Pressure and Temperature (P/T) Limits ............................................... 3.4.10-1 3.4.11 Reactor Steam Dome Pressure .................................................................... 3.4.11-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating ........................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory Control .......................................................................... 3.5.2-1 3.5.3 RCIC System................................................................................................... 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment.................................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Lock...................................................................... 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ............................................ 3.6.1.3-1 3.6.1.4 Drywell Pressure .......................................................................................... 3.6.1.4-1 3.6.1.5 Drywell Air Temperature............................................................................... 3.6.1.5-1 3.6.1.6 Low-Low Set (LLS) Valves ........................................................................... 3.6.1.6-1 General Electric BWR/4 STS vi Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS) Safety/Relief Valves (S/RVs)

LCO 3.4.3 The OPS safety function of [11] S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [ One [or two] [required] A.1 Restore the [required] 14 days S/RV[s] inoperable. S/RV[s] to OPERABLE status. [OR In accordance with the Risk Informed Completion Time Program] ]

B. [ Required Action and B.1 ---------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A not applicable when entering met. ] MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AC. OPS AC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable[Three] or more [required] S/RVs AND inoperable.

AC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> General Electric BWR/4 STS 3.4.3-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 -------------------------------NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Verify the as-left safety function OPS lift setpoints pressures of the [required] safety/relief valves [ In accordance (S/RVs) are within +/- 1% of the nominal with the setpointas follows: INSERVICE TESTING Nominal PROGRAM Number of Setpoint OPS S/RVs (psig) OR

[4] [1090 +/- 32.7] [ [18] months]

[4] [1100 +/- 33.0]

[3] [1110 +/- 33.3] OR Following testing, lift settings shall be within +/- 1%. In accordance with the Surveillance Frequency Control Program ]

SR 3.4.3.2 Verify the as-found OPS lift pressures of the In accordance

[required] S/RVs are within the limits specified in with the the COLR. INSERVICE SR 3.4.3.2 -------------------------------NOTE------------------------------ TESTING Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after PROGRAM reactor steam pressure and flow are adequate to perform the test.

Verify each [required] S/RV opens when manually actuated. [ [18] months [on a STAGGERED TEST BASIS for each valve solenoid OR In accordance with the General Electric BWR/4 STS 3.4.3-2 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.3 SURVEILLANCE FREQUENCY Surveillance Frequency Control Program ] ]

General Electric BWR/4 STS 3.4.3-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT TABLE OF CONTENTS Page B 3.3 INSTRUMENTATION (continued)

B 3.3.6.1 Primary Containment Isolation Instrumentation ......................................... B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation..................................... B 3.3.6.2-1 B 3.3.6.3 Low-Low Set (LLS) Instrumentation ........................................................... B 3.3.6.3-1 B 3.3.7.1 [ Main Control Room Environmental Control (MCREC) ]

System Instrumentation.............................................................................. B 3.3.7.1-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation ........................................................ B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................... B 3.3.8.2-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating ................................................................... B 3.4.1-1 B 3.4.2 Jet Pumps .................................................................................................. B 3.4.2-1 B 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ... B 3.4.3-1 B 3.4.4 RCS Operational LEAKAGE ...................................................................... B 3.4.4-1 B 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage ............................................ B 3.4.5-1 B 3.4.6 RCS Leakage Detection Instrumentation ................................................... B 3.4.6-1 B 3.4.7 RCS Specific Activity .................................................................................. B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System

- Hot Shutdown .......................................................................................... B 3.4.8-1 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System

- Cold Shutdown ........................................................................................ B 3.4.9-1 B 3.4.10 RCS Pressure and Temperature (P/T) Limits ............................................ B 3.4.10-1 B 3.4.11 Reactor Steam Dome Pressure ................................................................. B 3.4.11-1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating ...................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory Control ..................................................................... B 3.5.2-1 B 3.5.3 RCIC System.............................................................................................. B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment.................................................................................. B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Lock.................................................................... B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................................... B 3.6.1.3-1 B 3.6.1.4 Drywell Pressure ........................................................................................ B 3.6.1.4-1 B 3.6.1.5 Drywell Air Temperature............................................................................. B 3.6.1.5-1 B 3.6.1.6 Low-Low Set (LLS) Valves ......................................................................... B 3.6.1.6-1 General Electric BWR/4 STS vi Rev. 5.0

TSTF-576, Rev. 2 DRAFT RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, [1971 Edition], including Addenda through the [winter of 1972] (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.

The SL of 1325 psig, as measured in the reactor steam dome, is General Electric BWR/4 STS B 2.1.2-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued) during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the Overpressure Protection Systemsafety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 137 x 7% = 10) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.3.

This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods.

APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the General Electric BWR/4 STS B 3.1.4-2 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) located. Each APRM channel receives two total drive flow signals representative of total core flow. The total drive flow signals are generated by four flow units, two of which supply signals to the trip system A APRMs, while the other two supply signals to the trip system B APRMs. Each flow unit signal is provided by summing up the flow signals from the two recirculation loops. To obtain the most conservative reference signals, the total flow signals from the two flow units (associated with a trip system as described above) are routed to a low auction circuit associated with each APRM. Each APRM's auction circuit selects the lower of the two flow unit signals for use as the scram trip reference for that particular APRM. Each required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channel only requires an input from one OPERABLE flow unit, since the individual APRM channel will perform the intended function with only one OPERABLE flow unit input. However, in order to maintain single failure criteria for the Function, at least one required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channel in each trip system must be capable of maintaining an OPERABLE flow unit signal in the event of a failure of an auction circuit, or a flow unit, in the associated trip system (e.g., if a flow unit is inoperable, one of the two required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channels in the associated trip system must be considered inoperable).

The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Fixed Neutron Flux - High The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The Average Power Range Monitor Fixed Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive General Electric BWR/4 STS B 3.3.1.1-10 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

RCS pressure. For the overpressurization protection analysis of Reference 5, the Average Power Range Monitor Fixed Neutron Flux -

High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 6) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip.

Four channels of Average Power Range Monitor Fixed Neutron Flux -

High with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

The Average Power Range Monitor Fixed Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Fixed Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High, Setdown Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux - High Function is not required in MODE 2.

2.d. Average Power Range Monitor - Downscale This signal ensures that there is adequate Neutron Monitoring System protection if the reactor mode switch is placed in the run position prior to the APRMs coming on scale. With the reactor mode switch in run, an APRM downscale signal coincident with an associated Intermediate Range Monitor Neutron Flux - High or Inop signal generates a trip signal.

This Function was not specifically credited in the accident analysis but it General Electric BWR/4 STS B 3.3.1.1-11 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

3. Reactor Vessel Steam Dome Pressure - High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure -

High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 5, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure

- High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 7). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

General Electric BWR/4 STS B 3.3.1.1-14 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Four channels of Reactor Vessel Water Level - Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level - Low, Level 3 Allowable Value is selected to ensure that during normal operation the separator skirts are not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water - Low Low Low, Level 1 will not be required.

The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level - Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.

5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 5, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. Additionally, MSIV closure is assumed in the transients analyzed in Reference 8 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Each inboard and outboard MSIV inputs to a main steam line channel in each trip system, and each of the two trip logics within each RPS trip system receive parallel inputs from two of the four main steam lines. Thus, each General Electric BWR/4 STS B 3.3.1.1-15 Rev. 5.0

TSTF-576, Rev. 2 DRAFT ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The specific Applicable Safety Analyses and LCO discussions are listed below on a Function by Function basis.

a. Reactor Vessel Water Level - Low Low, Level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level - Low Low, Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.

b. Reactor Steam Dome Pressure - High Excessively high RPV pressure may rupture the RCPB. An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates a RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the Overpressure Protection Systemsafety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig).

The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.

Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude General Electric BWR/4 STS B 3.3.4.2-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT LLS Instrumentation B 3.3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.3.6.3.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specified channel.

The system functional testing performed in LCO 3.4.3, "Overpressure Protection System (OPS) Safety/Relief Valves(S/RVs)" and LCO 3.6.1.8, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs)," for S/RVs overlaps this test to provide complete testing of the assumed safety function.

[ The Frequency of once every 18 months for SR 3.3.6.3.7 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES 1. FSAR, Figure [ ] .

2. FSAR, Section [5.5.17].
3. GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

General Electric BWR/4 STS B 3.3.6.3-9 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

BASES BACKGROUND The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 2) requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, tThe size and number of safety/relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. The safety mode is credited for overpressure protection. This satisfies the Code requirement.

In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures.

In the relief mode, valves may be opened manually or automatically at the selected preset pressure. [S/RVs operating in relief mode are not credited for overpressure protection.][Some S/RVs operating in the relief mode are also credited for overpressure protection.]

[Some of the S/RVs operating in relief mode also provide the low-low set relief function, specified in LCO 3.6.1.6, "Low-Low Set (LLS)

Valves," and the Automatic Depressurization System, specified in LCO 3.5.1, "ECCS - Operating." The instrumentation associated with the [relief mode and] low-low set relief function is discussed in the General Electric BWR/4 STS B 3.4.3-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 Bases for LCO 3.3.6.3, "Low-Low Set (LLS) Instrumentation," and instrumentation for the ADS function is discussed in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS) Instrumentation."]

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LCO 3.6.1.6, "Low-Low Set (LLS) Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."

APPLICABLE The OPS overpressure protection system must accommodate the most SAFETY severe pressurization transient. Evaluations have determined that the ANALYSES most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). The S/RV discharge piping is designed to accommodate forces resulting from relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity so that system integrity is maintained. For the purpose of Tthe overpressure protection analyses, (Ref. 1) assume

[eleven11] S/RVs are assumed to operate in the safety mode of operation [and an additional [seven] S/RVs operate in the relief mode]. The analysis results demonstrate that the OPS design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =

1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design design Basis basis Eventevent.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

Reference 3 2 discusses additional events that are expected to actuate the S/RVs.

The OPS satisfies S/RVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

General Electric BWR/4 STS B 3.4.3-2 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 BASES LCO The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety [and relief] mode[s] function of the

[11] S/RVs [and the relief mode of additional S/RVs]. are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2). The requirements of this LCO are applicable only toThe OPERABILITY of the OPS is only dependent on the ability capability of the S/RVs to mechanically open to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs. when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve to be set setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressureization conditions. The transient evaluations in Reference 3 the FSAR are based on these setpoints, but also include the additional uncertainties of +/- 1% of the nominal setpoint drift to provide an added degree of conservatism.

An inoperable OPS Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in Safety Limit 2.1.2 the ASME Code limit on reactor pressure being exceeded.

APPLICABILITY In MODES 1, 2, and 3, the OPS all S/RVs must be OPERABLE, since there may be considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES.

The OPS S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The OPS S/RV function is not needed during these conditions.

ACTIONS [ A.1 With the safety function of one [or two] [required] S/RV[s] inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable.

General Electric BWR/4 STS B 3.4.3-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only.

BASES ACTIONS (continued)

The 14 day Completion Time to restore the inoperable required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. [Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.] ]

B.1

REVIEWERS NOTE ----------------------------------

Adoption of a MODE 3 end state requires the licensee to make the following commitments:

1. [LICENSEE] will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision [4F].
2. [LICENSEE] will follow the guidance established in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 2, "Technical Specifications End States, NEDC-32988-A," November 2009.

If the safety function of the inoperable required S/RVs cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 3) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low risk state.

Required Action B.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is General Electric BWR/4 STS B 3.4.3-4 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 BASES ACTIONS (continued) not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The allowed Completion Time is reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

AC.1 and AC.2 If the OPS is inoperable, [three] or more [required] S/RVs are inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure. The plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance verifies that the S/RVs used by the OPS have their as-left settings within 1% of the nominal opening pressure setpoint.

The verification of the S/RV as-left settings is performed in accordance with the requires that the [required] S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safe lift settings must be performed during shutdown, since this is a bench test, [to be done in accordance with the INSERVICE TESTING PROGRAM]. The nominal setpoint lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RVs setpoint is +/- [3]% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift. [A Note is provided to allow up to [two] of the required [11] S/RVs to be physically replaced with S/RVs with lower setpoints. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.]

REVIEWERS NOTE-----------------------------------

If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.

General Electric BWR/4 STS B 3.4.3-5 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3

[ The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

BASES SURVEILLANCE REQUIREMENTS (continued)

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.4.3.2 This Surveillance verifies that the as-found lift pressures of the S/RVs used by the OPS are consistent with the assumptions of the overpressure analysis. The measurement of the S/RV lift pressures must be performed during shutdown, since this is a bench test, in accordance with the INSERVICE TESTING PROGRAM. The OPS S/RV lift pressures are specified in the COLR.

A manual actuation of each [required] S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is [920] psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to General Electric BWR/4 STS B 3.4.3-6 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

[ The [18] month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code (Ref. 4). Operating experience has shown that these components usually pass the Surveillance when General Electric BWR/4 STS B 3.4.3-7 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued) performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES 1. FSAR, Section [5.2.2.2.4].

24. ASME Code for Operation and Maintenance of Nuclear Power Plants.
32. FSAR, Section [15].
3. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.
4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

General Electric BWR/4 STS B 3.4.3-8 Rev. 5.0

TSTF-576, Rev. 2 DRAFT Reactor Steam Dome Pressure B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed initial condition of design basis accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.

APPLICABLE The reactor steam dome pressure of [1020] psig is an initial condition of SAFETY the vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the Overpressure Protection System pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified reactor steam dome pressure limit of [1020] psig ensures the plant is operated within the assumptions of the transient analyses.

Operation above the limit may result in a transient response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam and the design basis accidents and transients are bounding.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

General Electric BWR/4 STS B 3.4.11-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT LLS Valves B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves BASES BACKGROUND The safety/relief valves (S/RVs) can actuate in either the safety mode as part of the Overpressure Protection System, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcomes the spring force and opens the pilot valve. As in the safety mode, opening the pilot valve allows a differential pressure to develop across the main valve piston and opens the main valve. The main valve can stay open with valve inlet steam pressure as low as

[50] psig. Below this pressure, steam pressure may not be sufficient to hold the main valve open against the spring force of the pilot valves. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.

[Four] of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, so that reopening more than one S/RV is prevented on subsequent actuations.

Therefore, the LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load.

APPLICABLE The LLS relief mode functions to ensure that the containment design SAFETY basis of one S/RV operating on "subsequent actuations" is met. In other ANALYSES words, multiple simultaneous openings of S/RVs (following the initial opening), and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the S/RV discharge line resulting from "subsequent actuations" of the S/RV during Design Basis Accidents (DBAs). Furthermore, the LLS function justifies the primary containment analysis assumption that simultaneous S/RV openings occur only on the initial actuation for DBAs.

Even though [four] LLS S/RVs are specified, all [four] LLS S/RVs do not operate in any DBA analysis.

LLS valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

General Electric BWR/4 STS B 3.6.1.6-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT TABLE OF CONTENTS Page 3.3 INSTRUMENTATION (continued) 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ......................... 3.3.5.1-1 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ....................................................................................... 3.3.5.2-1 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation ................. 3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation ........................................... 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation....................................... 3.3.6.2-1 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation ....................................................................................... 3.3.6.3-1 3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation ....................... 3.3.6.4-1 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation ............................................ 3.3.6.5-1 3.3.7.1 [Control Room Fresh Air (CRFA)] System Instrumentation .......................... 3.3.7.1-1 3.3.8.1 Loss of Power (LOP) Instrumentation .......................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ...................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ........................................................................ 3.4.1-1 3.4.2 Flow Control Valves (FCVs) ............................................................................ 3.4.2-1 3.4.3 Jet Pumps ....................................................................................................... 3.4.3-1 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ........ 3.4.4-1 3.4.5 RCS Operational LEAKAGE ........................................................................... 3.4.5-1 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage ................................................. 3.4.6-1 3.4.7 RCS Leakage Detection Instrumentation ........................................................ 3.4.7-1 3.4.8 RCS Specific Activity ....................................................................................... 3.4.8-1 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .................................................................................................. 3.4.9-1 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ................................................................................................ 3.4.10-1 3.4.11 RCS Pressure and Temperature (P/T) Limits ............................................... 3.4.11-1 3.4.12 Reactor Steam Dome Pressure .................................................................... 3.4.12-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating ........................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory Control ......................................................................... 3.5.2-1 3.5.3 RCIC System................................................................................................... 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment.................................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Locks .................................................................... 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ............................................ 3.6.1.3-1 General Electric BWR/6 STS vi Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 Overpressure Protection System (OPS) Safety/Relief Valves (S/RVs)

LCO 3.4.4 The OPS The safety function of [seven] S/RVs shall be OPERABLE.,

AND The relief function of [seven] additional S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [ One [required] S/RV A.1 Restore [required] S/RV to 14 days inoperable. OPERABLE status.

[OR In accordance with the Risk Informed Completion Time Program] ]

B. [ Required Action and B.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A not applicable when entering met. ] MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AC. OPS AC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable[Two] or more [required] S/RVs AND inoperable.

AC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> General Electric BWR/6 STS 3.4.4-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 -------------------------------NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Verify the as-left safety function OPS lift pressures [ In accordance setpoints of the [required] safety/relief valves with the (S/RVs) are within +/- 1% of the nominal setpoint INSERVICE are as follows: TESTING PROGRAM Nominal Number of Setpoint OR OPS S/RVs (psig)

((18] months]

[8] [1165 +/- 34.9]

[6] [1180 +/- 35.4] OR

[6] [1190 +/- 35.7]

In accordance Following testing, lift settings shall be within +/- 1%. with the Surveillance Frequency Control Program ]

SR 3.4.4.2 -------------------------------NOTE------------------------------

Valve actuation may be excluded.

Verify each [required] safety/relief valveS/RV [ [18] months acting in the relief mode relief function S/RV actuates on an actual or simulated automatic OR initiation signal.

In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.4.4-2 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs 3.4.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.4.3 Verify the as-found OPS lift pressures of the In accordance

[required] S/RVs are within the limits specified in with the the COLR. INSERVICE

NOTE------------------------------ TESTING Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after PROGRAM reactor steam pressure and flow are adequate to perform the test.

Verify each [required] S/RV opens when manually actuated. [ [18] months on a STAGGERED TEST BASIS for each valve solenoid OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.4.4-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT TABLE OF CONTENTS Page B 3.3 INSTRUMENTATION (continued)

B 3.3.6.1 Primary Containment Isolation Instrumentation ......................................... B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation..................................... B 3.3.6.2-1 B 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation........................................................................................... B 3.3.6.3-1 B 3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation ..................... B 3.3.6.4-1 B 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation .......................................... B 3.3.6.5-1 B 3.3.7.1 [Control Room Fresh Air (CRFA)] System Instrumentation ........................ B 3.3.7.1-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation ........................................................ B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................... B 3.3.8.2-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating ................................................................... B 3.4.1-1 B 3.4.2 Flow Control Valves (FCVs) ....................................................................... B 3.4.2-1 B 3.4.3 Jet Pumps .................................................................................................. B 3.4.3-1 B 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ... B 3.4.4-1 B 3.4.5 RCS Operational LEAKAGE ...................................................................... B 3.4.5-1 B 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage ............................................ B 3.4.6-1 B 3.4.7 RCS Leakage Detection Instrumentation ................................................... B 3.4.7-1 B 3.4.8 RCS Specific Activity .................................................................................. B 3.4.8-1 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System

- Hot Shutdown ......................................................................................... B 3.4.9-1 B 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System

- Cold Shutdown ....................................................................................... B 3.4.10-1 B 3.4.11 RCS Pressure and Temperature (P/T) Limits ............................................ B 3.4.11-1 B 3.4.12 Reactor Steam Dome Pressure ................................................................. B 3.4.12-1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating ...................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory Control ..................................................................... B 3.5.2-1 B 3.5.3 RCIC System.............................................................................................. B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment.................................................................................. B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Locks .................................................................. B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................................... B 3.6.1.3-1 B 3.6.1.4 Primary Containment Pressure .................................................................. B 3.6.1.4-1 B 3.6.1.5 Primary Containment Air Temperature....................................................... B 3.6.1.5-1 B 3.6.1.6 Low-Low Set (LLS) Valves ......................................................................... B 3.6.1.6-1 B 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System ...................... B 3.6.1.7-1 General Electric BWR/6 STS vi Rev. 5.0

TSTF-576, Rev. 2 DRAFT RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, the number of protective barriers designed to prevent radioactive releases from exceeding the limits would be reduced.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code,Section III, [1971 Edition], including Addenda through the [winter of 1972] (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of General Electric BWR/6 STS B 2.1.2-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) trip. Six channels of Average Power Range Monitor Flow Biased Simulated Thermal Power - High, with three channels in each trip system arranged in one-out-of-three logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located. Each APRM channel receives one total drive flow signal representative of total core flow. The recirculation loop drive flow signals are generated by eight flow units. One flow unit from each recirculation loop is provided to each APRM channel. Total drive flow is determined by each APRM by summing up the flow signals provided to the APRM from the two recirculation loops.

The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heater event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal that is proportional to the THERMAL POWER.

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Fixed Neutron Flux - High The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The Average Power Range Monitor Fixed Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 3, the Average Power Range Monitor Fixed Neutron Flux -

High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 8) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

General Electric BWR/6 STS B 3.3.1.1-10 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

3. Reactor Vessel Steam Dome Pressure - High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure -

High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 3, the reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure-High signal), along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure

- High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 4). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

General Electric BWR/6 STS B 3.3.1.1-13 Rev. 5.0

TSTF-576, Rev. 2 DRAFT RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

6. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the Nuclear Steam Supply System and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 3, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.

Additionally, MSIV closure is assumed in the transients analyzed in Reference 5 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Each inboard and outboard MSIV inputs to a main steam line channel in each trip system, and each of the two trip logics within each RPS trip system receive parallel inputs from two of the four main steam lines. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve

- Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on both of the main steam lines (MSLs) in one of the two logics in each RPS trip system must close in order for a scram to occur.

The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.

Sixteen channels of the Main Steam Isolation Valve - Closure Function with eight channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, the General Electric BWR/6 STS B 3.3.1.1-15 Rev. 5.0

TSTF-576, Rev. 2 DRAFT Relief and LLS Instrumentation B 3.3.6.5 B 3.3 INSTRUMENTATION B 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation BASES BACKGROUND The Overpressure Protection System safety/relief valves (S/RVs) prevents overpressurization of the nuclear steam system utilizing the safety/relief valves (S/RVs). Instrumentation is provided to support two modes of S/RV operation - the relief function (all valves) and the LLS function (selected valves). Refer to LCO 3.4.4, "Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)," and LCO 3.6.1.6, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs)," for Applicability Bases for additional information of these modes of S/RV operation. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the S/RV Safety/Relief valve instrumentation, as well as LCOs on other reactor system parameters, and equipment performance.

Technical Specifications are required by 10 CFR 50.36 to include LSSS for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytical Limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that an SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

REVIEWER'S NOTE-----------------------------------

The term "Limiting Trip Setpoint" [LTSP] is generic terminology for the calculated trip setting (setpoint) value calculated by means of the plant specific setpoint methodology documented in a document controlled under 10 CFR 50.59. The term [LTSP] indicates that no additional margin has been added between the Analytical Limit and the calculated trip setting.

"Nominal Trip Setpoint [NTSP]" is the suggested terminology for the actual setpoint implemented in the plant surveillance procedures where margin has been added to the calculated [LTSP]. The as-found and as-left tolerances will apply to the [NTSP] implemented in the Surveillance procedures to confirm channel performance.

General Electric BWR/6 STS B 3.3.6.5-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT Relief and LLS Instrumentation B 3.3.6.5 BASES BACKGROUND (continued) has been found to be different from the [LTSP] due to some drift of the setting may still be OPERABLE because drift is to be expected. This expected drift would have been specifically accounted for in the setpoint methodology for calculating the [LTSP] and thus the automatic protective action would still have ensured that the SL would not be exceeded with the "as found" setting of the protection channel. Therefore, the channel would still be OPERABLE because it would have performed its safety function and the only corrective action required would be to reset the channel within the established as-left tolerance around the [LTSP] to account for further drift during the next surveillance interval. Note that, although the channel is OPERABLE under these circumstances, the trip setpoint must be left adjusted to a value within the as-left tolerance, in accordance with uncertainty assumptions stated in the referenced setpoint methodology (as-left criteria), and confirmed to be operating within the statistical allowances of the uncertainty terms assigned (as-found criteria).

However, there is also some point beyond which the channel may not be able to perform its function due to, for example, greater than expected drift. This value needs to be specified in the Technical Specifications in order to define OPERABILITY of the channels and is designated as the Allowable Value.

If the actual setting (as-found setpoint) of the channel is found to be conservative with respect to the Allowable Value but is beyond the as-found tolerance band, the channel is OPERABLE, but degraded. The degraded condition will be further evaluated during performance of the SR. This evaluation will consist of resetting the channel setpoint to the

[Nominal Trip Setpoint (NTSP)] (within the allowed tolerance), and evaluating the channel response. If the channel is functioning as required and is expected to pass the next surveillance, then the channel is OPERABLE and can be restored to service at the completion of the surveillance. After the surveillance is completed, the channel as-found condition will be entered into the Corrective Action Program for further evaluation.

The Overpressure Protection System, which utilizes the relief function of the S/RVs, prevents overpressurization of the nuclear steam system.

The LLS function of the S/RVs is designed to mitigate the effects of postulated thrust loads on the S/RV discharge lines by preventing subsequent actuations with an elevated water leg in the S/RV discharge line. It also mitigates the effects of postulated pressure loads on the containment by preventing multiple actuations in rapid succession of the S/RVs subsequent to their initial actuation.

General Electric BWR/6 STS B 3.3.6.5-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

BASES BACKGROUND The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 21) requires the rReactor pPressure vVessel be protected from overpressure during upset conditions by self- actuated safety valves. As part of the nuclear pressure relief system, tThe size and number of safety/relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the direct action of the steam pressure in the main steam lines will act against a spring loaded disk that will pop open when the valve inlet pressure exceeds the spring force. The safety mode is credited for overpressure protection.

In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. In the relief mode, valves may be opened manually or automatically at the selected preset pressure. Some S/RVs operating in the relief mode are also credited for overpressure protection.

Some Six of the S/RVs operating in relief mode providing the relief function also provide the low-low set relief function specified in LCO 3.6.1.6, "Low-Low Set (LLS) Valves,." and Eight of the S/RVs that provide the relief function are part of the Automatic Depressurization System specified in LCO 3.5.1, "ECCS - Operating." The instrumentation associated with the relief valve mode function and low-low set relief function is discussed in the Bases for LCO 3.3.6.5, "Relief and Low-Low General Electric BWR/6 STS B 3.4.4-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 Set (LLS) Instrumentation," and instrumentation for the ADS function is discussed in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS)

Instrumentation."

APPLICABLE The OPS overpressure protection system must accommodate the most SAFETY severe pressurizatione transient. Evaluations have determined that the most ANALYSES most severe transient is the closure of all main steam isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 12). The S/RV discharge piping is designed to accommodate forces resulting from relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity so that system integrity is maintained. For the purpose of Tthe overpressure protection analyses (Ref. 1), assume [seven]

S/RVs operate in the safety mode of operation and an additional

[seven] S/RVs operate in the relief mode. , [six] of the S/RVs are assumed to operate in the relief mode, and seven in the safety mode.

The analysis results demonstrate that the OPS design S/RV capacity is capable of maintaining reactor pressure below BASES APPLICABLE SAFETY ANALYSES (continued) the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the design basis event.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 3 discusses additional events that are expected to actuate the S/RVs.

The OPS satisfies S/RVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety and relief modes function of the seven S/RVs and the relief mode of additional S/RVs. is required to be OPERABLE in the safety mode, and an additional seven S/RVs (other than the seven S/RVs that satisfy the safety function) must be OPERABLE in the relief mode. The OPERABILITY of the OPS is only dependent on The requirements of this LCO are applicable only to the capability ability of the S/RVs to mechanically open to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs. In Reference 2, an evaluation was performed to establish the parametric relationship between the peak vessel pressure and the number of OPERABLE S/RVs. The results show that with a minimum of seven General Electric BWR/6 STS B 3.4.4-2 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 S/RVs in the safety mode and six S/RVs in the relief mode OPERABLE, the ASME Code limit of 1375 psig is not exceeded.

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve to be set at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressure conditions. The transient evaluations in Reference 3 are based on these setpoints, but also include the additional uncertainties of +/- 1% of the nominal setpoint to account for potential setpoint drift to provide an added degree of conservatism.

An inoperable OPS Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in Safety Limit 2.1.2 the ASME Code limit on reactor pressure being exceeded.

APPLICABILITY In MODES 1, 2, and 3, the OPS specified number of S/RVs must be OPERABLE since there may be considerable energy in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The OPS S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

BASES APPLICABILITY (continued)

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The OPS S/RV function is not needed during these conditions.

ACTIONS A.1 With the safety function of one [required] S/RV inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only.

The 14 day Completion Time to restore the inoperable required S/RVs to OPERABLE status is based on the relief capability of the remaining General Electric BWR/6 STS B 3.4.4-3 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. [Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.]

B.1

REVIEWERS NOTE ----------------------------------

Adoption of a MODE 3 end state requires the licensee to make the following commitments:

1. [LICENSEE] will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision [4F].
2. [LICENSEE] will follow the guidance established in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 2, "Technical Specifications End States, NEDC-32988-A," November 2009.

If the inoperable required S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a MODE in which overall plant risk is minimized. To BASES ACTIONS (continued) achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action B.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

General Electric BWR/6 STS B 3.4.4-4 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

AC.1 and AC.2 If the OPS is inoperable, [two] or more [required] S/RVs are inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure. The plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This Surveillance verifies that the S/RVs used by the OPS have their as-left settings within 1% of the nominal opening pressure setpoint.

The verification of the S/RV as-left settings is performed in accordance with the demonstrates that the [required] S/RVs will open at the pressures assumed in the safety analysis of Reference 2. The demonstration of the S/RV safety function lift settings must performed during shutdown, since this is a bench test[, and in accordance with the INSERVICE TESTING PROGRAM]. The nominal setpoint lift setting pressure shall BASES SURVEILLANCE REQUIREMENTS (continued) correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RVs setpoint is +/- [3]% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift. [A Note is provided to allow up to [two] of the required [11] S/RVs to be physically replaced with S/RVs with lower setpoints. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.]

REVIEWERS NOTE-----------------------------------

If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.

General Electric BWR/6 STS B 3.4.4-5 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4

[ The [18 month] Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.4.4.2 The OPS assumes that tThe [required] relief function mode S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify the mechanical portions of the automatic relief function mode operate as designed when initiated either by an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.5.4 overlaps this SR to provide complete testing of the relief mode safety function.

BASES SURVEILLANCE REQUIREMENTS (continued)

[ The [18 month] Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at the [18 month]

Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

General Electric BWR/6 STS B 3.4.4-6 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 This SR is modified by a Note that excludes valve actuation. The SR may be performed by removing the actuator and verifying its operation. This prevents an RPV pressure blowdown.

SR 3.4.4.3 This Surveillance verifies that the as-found lift pressures of the S/RVs used by the OPS are consistent with the assumptions of the overpressure analysis. The measurement of the S/RV lift pressures must be performed during shutdown, since this is a bench test, in accordance with the INSERVICE TESTING PROGRAM. The OPS S/RV lift pressures are specified in the COLR.

A manual actuation of each [required] S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is 950 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required General Electric BWR/6 STS B 3.4.4-7 Rev. 5.0

TSTF-576, Rev. 2 DRAFT OPSS/RVs B 3.4.4 BASES SURVEILLANCE REQUIREMENTS (continued) pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If the valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

[ The [18] month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 18 month Frequency was developed based on the S/RV tests required by the ASME (Ref. 1). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

REFERENCES 1. FSAR, Section [5.2.5.5.3].

21. ASME Code for Operation and Maintenance of Nuclear Power Plants.
2. FSAR, Section [5.2.5.5.3].
3. FSAR, Section [15].
4. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

General Electric BWR/6 STS B 3.4.4-8 Rev. 5.0

TSTF-576, Rev. 2 DRAFT Reactor Steam Dome Pressure B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents (DBAs) and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.

APPLICABLE The reactor steam dome pressure of [1045] psig is an initial condition of SAFETY the vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the Overpressure Protection Systempressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

Reference 2 also assumes an initial reactor steam dome pressure for the analysis of DBAs and transients used to determine the limits for fuel cladding integrity MCPR (see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified reactor steam dome pressure limit of [1045] psig ensures the plant is operated within the assumptions of the transient analyses.

Operation above the limit may result in a transient response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam, and the DBAs and transients are bounding.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

General Electric BWR/6 STS B 3.4.12-1 Rev. 5.0

TSTF-576, Rev. 2 DRAFT LLS Valves B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves BASES BACKGROUND The safety/relief valves (S/RVs) can actuate either in the relief mode or, the safety mode as part of the Overpressure Protection System, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcome the spring force and open the pilot valve.

As in the safety mode, opening the pilot valve allows a differential pressure to develop across the main valve piston and thus opens the main valve. The main valve can stay open with valve inlet steam pressure as low as [0] psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.

[Six] of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, such that reopening of more than one S/RV is prevented on subsequent actuations.

Therefore, the LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load.

APPLICABLE The LLS relief mode functions to ensure that the containment design SAFETY basis of one S/RV operating on "subsequent actuations" is met (Ref. 1).

ANALYSES In other words, multiple simultaneous openings of S/RVs (following the initial opening) and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the S/RV discharge line resulting from "subsequent actuations" of the S/RV during Design Basis Accidents (DBAs).

Furthermore, the LLS function justifies the primary containment analysis assumption that multiple simultaneous S/RV openings occur only on the initial actuation for DBAs. Even though [six] LLS S/RVs are specified, all

[six] LLS S/RVs do not operate in any DBA analysis.

LLS valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO [Six LLS valves are required to be OPERABLE to satisfy the assumptions of the safety analysis (Ref. 2). The requirements of this LCO are applicable to the mechanical and electrical/pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs General Electric BWR/6 STS B 3.6.1.6-1 Rev. 5.0

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