TSTF-17-16, Technical Specifications Task Force - Transmittal of TSTF-568, Clarify Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2

From kanterella
Jump to navigation Jump to search

Technical Specifications Task Force - Transmittal of TSTF-568, Clarify Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2
ML17353A437
Person / Time
Site: Technical Specifications Task Force
Issue date: 12/19/2017
From: Gullott D, Miksa J, Sparkman W, Vaughan J, Linda Williams
Technical Specifications Task Force
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TSTF-17-16
Download: ML17353A437 (30)
v • d • e


Text

TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY December 19, 2017 TSTF-17-16 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Transmittal of TSTF-568, "Clarify Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2" Enclosed for NRC review is TSTF-568, "Clarify Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2."

The following information is provided to assist the NRC staff in prioritizing their review of TSTF-568:

  • Applicability: TSTF-568 is applicable to NUREG-1433, which is used for Boiling Water Reactors of the BWR/2, BWR/3, and BWR/4 design.
  • Specialized Resource Availability: The TSTF requests approval of the traveler within one year.

The Technical Specifications Task Force should be billed for the review of the traveler.

11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation

TSTF 17-16 December 19, 2017 Page 2 Should you have any questions, please do not hesitate to contact us.

James P. Miksa (PWROG/CE) Lisa L. Williams (BWROG)

David M. Gullott (PWROG/W) Jordan L. Vaughan (PWROG/B&W)

Wesley Sparkman (APOG)

Attachment cc: Michelle Honcharik, Technical Specifications Branch Robert Tjader, Technical Specifications Branch Victor Cusumano, Technical Specifications Branch

BWROG-135, Rev. 1 TSTF-568, Rev. 0 Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler Clarify Applicability of BWR/4 TS 3.6.2.5 and TS 3.6.3.2 NUREGs Affected: 1430 1431 1432 1433 1434 2194 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Correction NRC Fee Status: Not Exempt Changes Marked on ISTS Rev 4.0 See attached.

Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: NRC Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 18-May-17 Owners Group Comments (No Comments)

Owners Group Resolution: Superceeded Date: 29-Sep-17 OG Revision 1 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

Replaced with Revision 1 based on BWROG comments.

Owners Group Review Information Date Originated by OG: 11-Sep-17 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 02-Oct-17 TSTF Review Information TSTF Received Date: 02-Oct-17 Date Distributed for Review 02-Oct-17 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 19-Dec-17 19-Dec-17 Copyright(C) 2017, 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.

BWROG-135, Rev. 1 TSTF-568, Rev. 0 OG Revision 1 Revision Status: Active NRC Review Information NRC Received Date: 19-Dec-17 NRC Comments:

Presubmittal discussion held with NRC on November 9, 2017. No changes identified.

Affected Technical Specifications Appl. 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure Appl. 3.6.2.5 Bases Drywell-to-Suppression Chamber Differential Pressure SR 3.6.2.5.1 Drywell-to-Suppression Chamber Differential Pressure SR 3.6.2.5.1 Bases Drywell-to-Suppression Chamber Differential Pressure Appl. 3.6.3.2 Primary Containment Oxygen Concentration Appl. 3.6.3.2 Bases Primary Containment Oxygen Concentration SR 3.6.3.2.1 Primary Containment Oxygen Concentration SR 3.6.3.2.1 Bases Primary Containment Oxygen Concentration 19-Dec-17 Copyright(C) 2017, 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.

TSTF-568, Rev. 0

1.

SUMMARY

DESCRIPTION The Applicability of Technical Specification (TS) 3.6.2.5, "Drywell-to-Suppression Chamber Differential Pressure," and TS 3.6.3.2, "Primary Containment Oxygen Concentration," requires the associated limiting conditions for operation (LCO) to be met when the unit is in Mode 1 during the time period: a. from [24] hours after Thermal Power is > [15]% Rated Thermal Power (RTP) following startup, to [24] hours prior to reducing Thermal Power to < [15]% RTP prior to the next scheduled reactor shutdown. This change revises the Applicability presentation of these Specifications in NUREG-1433, "Standard Technical Specifications General Electric BWR/4 Plants,1" to clearly state that these Specifications are applicable in MODE 1 with Thermal Power

> [15]% RTP while maintaining the [24]-hour allowance prior to power exceeding [15]% RTP, and prior to reducing power to below [15]% RTP as Notes in the respective Surveillance Requirements.

2. DETAILED DESCRIPTION 2.1. System Description and Operation There are three containment designs used in the various boiling water reactor (BWR) plants.

BWR/2, BWR/3, and early model BWR/4 plants have the Mark I containment. Later model BWR/4 and BWR/5 plants have the Mark II containment. BWR/6 plants have the Mark III containment.

The Mark I containment consists of a drywell (in the shape of an inverted light bulb), a suppression chamber (in the shape of a toroid), and a network of vents which extend radially outward and downward from the drywell to the suppression chamber. The Mark II containment consists of a drywell (in the shape of a truncated cone), a suppression chamber directly below the drywell (in the shape of a right circular cylinder), and a network of vertical vents extending downward from the drywell to the suppression chamber. The Mark III containment is cylindrical with a domed head, and surrounds the drywell and suppression pool.

The Mark I and II containment designs are inerted with nitrogen gas during normal operation to prevent an explosive mixture of hydrogen and oxygen from forming during accident conditions.

Long term control of post LOCA hydrogen gas concentration is accomplished by adding additional nitrogen gas and then venting the primary containment through the standby gas treatment system. The Mark III containment atmosphere is not inerted with nitrogen due to its large volume, and hydrogen ignitors are typically used for long-term combustible gas control following a postulated design basis event. The proposed change in only applicable to plants with Mark I and Mark II containment designs.

The Mark I and II primary containment inerting system consists of a nitrogen (N2) purge supply and an N2 makeup supply. The N2 purge supply is used to initially inert the atmosphere in the 1

NUREG-1433 is based on the BWR/4 plant design, but is also applicable of the BWR/2, BWR/3, and, for some requirements, to the BWR/5 plant designs.

Page 1

TSTF-568, Rev. 0 primary containment and is typically provided simultaneously to the suppression chamber air space and the drywell. At times, the suppression chamber pressure may increase at a rate faster than drywell pressure resulting in a differential pressure between the two volumes that is less than the TS 3.6.2.5 limit. The inerting process continues until primary containment oxygen concentration is less than 4% (or a plant-specific limit), as required by TS 3.6.3.2. The inerting process takes approximately eight to twelve hours.

In a Mark I or II containment, the drywell is immediately pressurized when a postulated line break occurs within the primary containment. As drywell pressure increases, drywell atmosphere (primarily nitrogen gas) and steam are blown down through the vents into the suppression pool via the downcomers. The steam condenses in the suppression pool which suppresses the peak pressure in the drywell. Noncondensible gases discharged into the suppression pool collect in the free air volume of the suppression chamber, increasing the suppression chamber pressure. As steam is condensed in the suppression pool, drywell pressure decreases until the suppression chamber pressure exceeds the drywell pressure and the suppression chamber-drywell vacuum breakers open and vent noncondensible gases back into the drywell.

3. CURRENT TECHNICAL SPECIFICATIONS REQUIREMENTS The Applicability of NUREG-1433 TS 3.6.2.5 and TS 3.6.3.2 both state:

MODE 1 during the time period:

a. From [24] hours after THERMAL POWER is > [15]% RTP following startup, to
b. [24] hours prior to reducing THERMAL POWER to < [15]% RTP prior to the next scheduled reactor shutdown.

Only four BWR units: Browns Ferry, Dresden, Fitzpatrick, and Quad Cities, contain a Drywell-to-Suppression Chamber Differential Pressure Specification (ISTS 3.6.2.5).

All BWR/2, BWR/3, BWR/4, and BWR/5 units contain a Primary Containment Oxygen Concentration specification (ISTS 3.6.3.2).

The 24-hour allowance above 15% RTP is provided in the Primary Containment Oxygen Concentration specification to delay inerting the primary containment in a plant startup and to accelerate de-inerting for a plant shutdown. This is for personnel safety considerations so that plant personnel can access the primary containment without breathing apparatus. This allowance is also needed for the Drywell-to-Suppression Chamber Differential Pressure specification due to the pressure fluctuations in the drywell and suppression chamber during the primary containment inerting process that make it difficult to maintain the differential pressure within the required limit.

4. REASON FOR THE PROPOSED CHANGE The TS requires primary containment oxygen concentration to be less than 4.0 volume percent (or a plant-specific limit) and drywell pressure to be at least [1.5] psid above the suppression chamber pressure when in Mode 1 during the time period from [24] hours after Thermal Power is Page 2

TSTF-568, Rev. 0

> 15% RTP following startup to [24] hours prior to reducing Thermal Power to < 15% RTP prior to the next scheduled reactor shutdown.

The current convoluted presentation in the Applicability can be misunderstood to mean that the requirements are applicable in all of Mode 1, including when reactor power is below 15% RTP, except when the 24-hour allowance is being utilized during a startup or a scheduled plant shutdown. Also, the term "scheduled" plant shutdown is ambiguous and could be misinterpreted to mean only a scheduled refueling outage instead of any planned outage, such as a mid-cycle shutdown for maintenance. The specification is intended to be applicable in MODE 1 with Thermal Power > [15]% RTP, except during the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding [15]% RTP following a startup and before reducing power to less than or equal to [15]% RTP prior to a shutdown).

The proposed change clarifies the intent of the Applicability.

4.1.Description of the Proposed Change The Applicability of both Drywell-to-Suppression Chamber Differential Pressure and Primary Containment Oxygen Concentration Specifications are revised as shown. Deleted text is identified with strikethrough and inserted text is identified in italics.

MODE 1 during the time period: with THERMAL POWER > [15]% RTP

a. From [24] hours after THERMAL POWER is > [15]% RTP following startup, to
b. [24] hours prior to reducing THERMAL POWER to < [15]% RTP prior to the next scheduled reactor shutdown.

Surveillance Requirement (SR) 3.6.2.5.1 and 3.6.3.2.1 are revised to incorporate the following Notes:

1. Not required to be met until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER > [15]% RTP.
2. Not required to be met 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to THERMAL POWER being reduced [15]% RTP.

The proposed change is supported by changes to the TS Bases. The Bases for these specifications are revised to clarify the Applicability is Mode 1 with Thermal Power > [15]%

RTP. The SR Notes allow deferral of drywell-to-suppression chamber differential pressure and primary containment oxygen concentration limits for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding [15]% RTP, or prior to reducing power to less than or equal to [15]% RTP. 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 2, 1993 (58 FR 39132).

Therefore, the Bases changes are provided for information and approval of the Bases is not requested.

Page 3

TSTF-568, Rev. 0 A model application is attached. The model may be used by licensees desiring to adopt the traveler following NRC approval.

5. TECHNICAL EVALUATION The drywell-to-suppression chamber differential pressure limit ensures the containment conditions assumed in the safety analyses are met by limiting the length of the suppression chamber downcomer water leg. High water level in the downcomers could result in excessive forces on the suppression chamber from the downcomer vents and higher pressure buildup in the drywell.

The primary containment oxygen concentration must be maintained below the limit to ensure that an accident that produces hydrogen does not result in a combustible mixture inside primary containment.

Revising the presentation of Applicability of the Drywell-to-Suppression Chamber Differential Pressure and Primary Containment Oxygen Concentration Specifications to state that these requirements are applicable in MODE 1 with thermal power above [15]% RTP clarifies the existing intent of the specification. In both of these Specifications, the existing Action to follow if the LCO is not met and compliance is not restored within the Completion Time is to reduce Thermal Power to less than or equal to [15]% RTP. The Bases for these Actions state, "the plant must be placed in a MODE in which the LCO does not apply. This is done by reducing power to

[15]% RTP." This clearly states the intent of the specification is to be applicable in Mode 1 with power greater than [15]% RTP.

General Electric Nuclear Energy Safety Communication 02-10 (Reference 1), discussed the ISTS LCO 3.6.2.5 Action when drywell-to-wetwell differential pressure is not within limit and the allowance for extended plant operation with reactor power below 15% RTP. The Safety Communication concluded that operation below 15% RTP with the differential pressure limit not met is acceptable for BWRs with Mark 2 containments and for BWRs with Mark 1 containments that have demonstrated acceptable loads with zero differential pressure. Only four BWR plants, all of which have Mark I containments, have a specification on drywell-to-suppression chamber differential pressure: Browns Ferry, Dresden, Fitzpatrick, and Quad Cities. Fitzpatrick, Dresden, and Quad Cities performed containment analyses, approved by the NRC in Safety Evaluations dated December 12, 1984, September 18, 1985, and February 15, 1986, respectively, that demonstrated that operation at less than 15% RTP without meeting the differential pressure limit was acceptable. A similar evaluation for Browns Ferry could not be identified and the proposed change to TS 3.6.2.5 is not applicable to that plant.

All BWR/2, BWR/3, BWR/4, and BWR/5 plants with Mark I or II containments have a specification on primary containment oxygen concentration. The Bases for the Primary Containment Oxygen Concentration states, "As long as reactor power is < 15% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert." Therefore, this change is applicable to all BWR/2, BWR/3, BWR/4, and BWR/5 plants with this specification.

Page 4

TSTF-568, Rev. 0 Therefore, revising the specification to clarify via Surveillance Requirement Notes that allowing operation at less than [15]% RTP without meeting the differential pressure or primary oxygen concentration specifications is acceptable for the applicable plants.

Moving the 24-hour exceptions from the Applicability to Notes in the SRs is functionally equivalent. Example 1.4-4 in the Use and Application section of NUREG-1433 states that such Notes constitute an "otherwise stated" exception to the Applicability of this Surveillance. A more direct example of this type of Note being used is with TS 3.4.8 where SR 3.4.8.1 allows deferral of part of the LCO requirements for a two-hour period.

Additionally, the existing Applicability statements refer to "following startup" and "prior to the next scheduled reactor shutdown." These phrases are ambiguous as the terms "startup" and "next scheduled reactor shutdown" are undefined. The use of these ambiguous terms is unnecessary.

The term "after startup" is implied and not needed in a Note that states "after THERMAL POWER > [15]% RTP," as power will only transition from less than 15% RTP to greater than 15% RTP during a startup. The need to de-inert the containment prior to shutdown to permit safe personnel access is applicable to any shutdown that permits planning such an activity. For unplanned shutdowns, such as a reactor trip or rapid power-down, Thermal Power will be below

[15]% RTP so TS 3.6.2.5 and TS 3.6.3.2, as revised, will not be applicable. Therefore, the proposed SR Notes do not include the phrases "after startup" and "prior to the next scheduled reactor shutdown."

In summary, moving the Applicability exceptions of TS 3.6.2.5 and TS 3.6.3.2 to Notes in SR 3.6.2.5.1 and SR 3.6.3.2.1 and simplifying the presentation does not alter the technical requirements. The requirements will continue to apply during the period from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entry into Mode 1 and power exceeds 15% RTP, and until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a power reduction to below 15% RTP. The basis for this conclusion is that the change is technically justified for the applicable plants, and the proposed change is consistent with the current TS Actions and associated TS Bases.

6. REGULATORY EVALUATION Section IV, "The Commission Policy," of the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 Federal Register 39132), dated July 22, 1993, states in part:

The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.

[T]he Commission will also entertain requests to adopt portions of the improved STS, even if the licensee does not adopt all STS improvements.

The Commission encourages all licensees who submit Technical Specification related submittals based on this Policy Statement to emphasize human factors principles.

Page 5

TSTF-568, Rev. 0 In accordance with this Policy Statement, improved STS have been developed and will be maintained for [BWR designs]. The Commission encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications.

[I]t is the Commission intent that the wording and Bases of the improved STS be used to the extent practicable.

As described in the Commissions "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," recommendations were made by NRC and industry task groups for new STS that include greater emphasis on human factors principles in order to add clarity and understanding to the text of the STS, and provide improvements to the Bases of STS, which provides the purpose for each requirement in the specification. Improved vendor-specific STS were developed and issued by the NRC in September 1992.

The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TS in accordance with the requirements of 10 CFR 50.36. The applicant must include in the application a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls." However, per 10 CFR 50.36(a)(1), these technical specification bases "shall not become part of the technical specifications." The Final Policy Statement provides the following description of the scope and the purpose of the Technical Specification Bases:

Appropriate Surveillance Requirements and Actions should be retained for each LCO

[limiting condition for operation] which remains or is included in the Technical Specifications. Each LCO, Action, and Surveillance Requirement should have supporting Bases. The Bases should at a minimum address the following questions and cite references to appropriate licensing documentation (e.g., FSAR, Topical Report) to support the Bases.

1. What is the justification for the Technical Specification, i.e., which Policy Statement criterion requires it to be in the Technical Specifications?
2. What are the Bases for each LCO, i.e., why was it determined to be the lowest functional capability or performance level for the system or component in question necessary for safe operation of the facility and, what are the reasons for the Applicability of the LCO?
3. What are the Bases for each Action, i.e., why should this remedial action be taken if the associated LCO cannot be met; how does this Action relate to other Actions associated with the LCO; and what justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?
4. What are the Bases for each Safety Limit?
5. What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed Page 6

TSTF-568, Rev. 0 to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?

Note: In answering these questions the Bases for each number (e.g., Allowable Value, Response Time, Completion Time, Surveillance Frequency), state, condition, and definition (e.g., operability) should be clearly specified. As an example, a number might be based on engineering judgment, past experience, or PSA [probabilistic safety assessment] insights; but this should be clearly stated.

Additionally, 10 CFR 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.

The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TSs will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

The regulation at 10 CFR 50.36(c)(3) requires TSs to include items in the category of SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

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.

The NRC staffs guidance for the review of TSs is in Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STS for each of the light-water reactor nuclear designs.

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 Page 7

TSTF-568, Rev. 0 approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

7. REFERENCES
1. Safety Communication SC02-10, "Drywell-to-Wetwell Differential Pressure Control Technical Specification for some Mark I Containments."

Page 8

TSTF-568, Rev. 0 Model Application

TSTF-568, Rev. 0

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

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-568, "Clarify Applicability of BWR TS 3.6.2.5 and TS 3.6.3.2" 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 568, "Clarify Applicability of BWR TS 3.6.2.5 and TS 3.6.3.2." TSTF-568 revises the Applicability presentation of Technical Specification (TS)

[applicable PLANT TS numbers and titles] to state that the[se] Specification[s are][is] applicable in MODE 1 with Thermal Power > [15]% RTP while maintaining the [24]-hour exception after power exceeds [15]% RTP and prior to reducing power to below [15]% RTP as a Surveillance Requirement Note.

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.

Approval of the proposed amendment is requested by [date]. Once approved, the amendment shall be implemented within [ ] 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

TSTF-568, Rev. 0 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

[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

TSTF-568, Rev. 0 ENCLOSURE DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

[LICENSEE] requests adoption of TSTF-568, "Clarify Applicability of BWR TS 3.6.2.5 and TS 3.6.3.2." TSTF-568 revises the Applicability presentation of Technical Specification (TS)

[applicable PLANT TS numbers and titles] to state that the[se] Specification[s are][is] applicable in MODE 1 with Thermal Power > [15]% RTP, while maintaining the [24]-hour allowance after power exceeds [15]% RTP and prior to reducing power to below [15]% RTP as Notes to SR 3.6.2.5.1 and SR 3.6.3.2.1.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation

[LICENSEE] has reviewed the safety evaluation for TSTF-568 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-568. [As described herein,]

[LICENSEE] has concluded that the justifications presented in TSTF-568 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 the TSTF-568 or the applicable parts of the NRC staffs safety evaluation dated [DATE].] [LICENSEE is proposing the following variations from the TS changes described in the TSTF-568 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-568 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-568 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-568 to the [PLANT]

TS.]

[The [PLANT] TS provide a different limit on primary containment oxygen concentration than the 4.0 volume percent limit shown in NUREG-1433. This difference does not affect the applicability of the proposed change.]

[The [PLANT] TS contain requirements that differ from the Standard Technical Specifications on which TSTF-568 was based, but are encompassed in the TSTF-568 justification. [Describe differences and why TSTF-568 is still applicable.))

Page 3

TSTF-568, Rev. 0

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

[LICENSEE] requests adoption of TSTF-568, "Clarify Applicability of BWR TS 3.6.2.5 and TS 3.6.3.2." TSTF-568 revises the Applicability presentation of Technical Specification (TS)

[3.6.2.5, "Drywell-to-Suppression Chamber Differential Pressure," and TS] 3.6.3.2, "Primary Containment Oxygen Concentration," to state that the[se] Specification[s are][is] applicable in MODE 1 with Thermal Power > [15]% Rated Thermal Power (RTP), while maintaining the [24]-

hour allowance after power exceeds [15]% RTP and prior to reducing power to below [15]%

RTP as Notes to SR 3.6.2.5.1 and SR 3.6.3.2.1.

[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 Applicability presentation of the [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration Technical Specification[s]. [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration are not initiators to any accident previously evaluated. [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration are assumptions in the mitigation of some accidents previously evaluated. [Analysis has demonstrated that the Drywell-to-Suppression Chamber differential pressure limit is not required to be met at less than 15%

power.] [When in Mode 1, but at less than 15% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert.]

Therefore, the consequences of an accident previously evaluated are not significantly increased.

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

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 Applicability presentation of the [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration Technical Specification[s]. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed).

No credible new failure mechanisms, malfunctions, or accident initiators that would have been considered a design basis accident in the UFSAR are created.

Page 4

TSTF-568, Rev. 0 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 Applicability presentation of the [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration Technical Specification[s]. No safety limits are affected. No Limiting Conditions for Operation or Surveillance limits are affected. The [Drywell-to-Suppression Chamber Differential Pressure and] Primary Containment Oxygen Concentration Technical Specification requirements assure sufficient safety margins are maintained, and that the design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants' licensing basis. The proposed change does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant 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 The proposed change does not 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 does not change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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 change.

Page 5

TSTF-568, Rev. 0 Technical Specifications and Bases Changes

TSTF-568, Rev. 0 Drywell-to-Suppression Chamber Differential Pressure 3.6.2.5 3.6 CONTAINMENT SYSTEMS 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure LCO 3.6.2.5 The drywell pressure shall be maintained [1.5] psid above the pressure of the suppression chamber.

APPLICABILITY: MODE 1 with THERMAL POWER > [15]% RTP except within following during the time periods:

a. From [24] hours after THERMAL POWER is >exceeds [15]% RTP following startup, to; and
b. [24] hours prior to reducing THERMAL POWER to

<below [15]% RTP prior to the next scheduled planned reactor shutdown.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell-to-suppression A.1 Restore differential 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> chamber differential pressure to within limit.

pressure not within limit.

B. Required Action and B.1 Reduce THERMAL 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion POWER to [15]% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.5.1 ----------------------------------- NOTES ----------------------

1. Not required to be met until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER > [15]% RTP.
2. Not required to be met 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to THERMAL POWER being reduced to [15]%

RTP.

General Electric BWR/4 STS 3.6.2.5-1 Rev. 4.0

TSTF-568, Rev. 0 Drywell-to-Suppression Chamber Differential Pressure 3.6.2.5 SURVEILLANCE FREQUENCY Verify drywell-to-suppression chamber differential [ 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure is within limit.

OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/4 STS 3.6.2.5-2 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration 3.6.3.2 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxygen Concentration LCO 3.6.3.2 The primary containment oxygen concentration shall be < 4.0 volume percent.

APPLICABILITY: MODE 1 during the time period:with THERMAL POWER > [15]% RTP except within following time periods:

a. From [24] hours after THERMAL POWER is >exceeds [15]% RTP following startup, ; toand
b. [24] hours prior to reducing THERMAL POWER to

<below [15]% RTP prior toduring planned the next scheduled reactor shutdown.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 Restore oxygen 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> oxygen concentration concentration to within limit.

not within limit.

B. Required Action and B.1 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion POWER to [15]% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 ----------------------------------- NOTES ----------------------

1. Not required to be met until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER > [15]% RTP.
2. Not required to be met 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to THERMAL POWER being reduced to [15]%

RTP.

General Electric BWR/4 STS 3.6.3.2-1 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration 3.6.3.2 SURVEILLANCE FREQUENCY Verify primary containment oxygen concentration is [ 7 days within limits.

OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/4 STS 3.6.3.2-2 Rev. 4.0

TSTF-568, Rev. 0 Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.5 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure BASES BACKGROUND The toroidal shaped suppression chamber, which contains the suppression pool, is connected to the drywell (part of the primary containment) by [eight] main vent pipes. The main vent pipes exhaust into a continuous vent header, from which [96] downcomer pipes extend into the suppression pool. The pipe exit is [4] ft below the minimum suppression pool water level required by LCO 3.6.2.2, "Suppression Pool Water Level." During a loss of coolant accident (LOCA), the increasing drywell pressure will force the waterleg in the downcomer pipes into the suppression pool at substantial velocities as the "blowdown" phase of the event begins. The length of the waterleg has a significant effect on the resultant primary containment pressures and loads.

APPLICABLE The purpose of maintaining the drywell at a slightly higher pressure with SAFETY respect to the suppression chamber is to minimize the drywell pressure ANALYSES increase necessary to clear the downcomer pipes to commence condensation of steam in the suppression pool and to minimize the mass of the accelerated water leg. This reduces the hydrodynamic loads on the torus during the LOCA blowdown. The required differential pressure results in a downcomer waterleg of [3.06 to 3.58] ft.

Initial drywell-to-suppression chamber differential pressure affects both the dynamic pool loads on the suppression chamber and the peak drywell pressure during downcomer pipe clearing during a Design Basis Accident LOCA. Drywell-to-suppression chamber differential pressure must be maintained within the specified limits so that the safety analysis remains valid.

Drywell-to-suppression chamber differential pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO A drywell-to-suppression chamber differential pressure limit of [1.5] psid is required to ensure that the containment conditions assumed in the safety analyses are met. A drywell-to-suppression chamber differential pressure of < [1.5] psid corresponds to a downcomer water leg of > [3.58] ft.

Failure to maintain the required differential pressure could result in excessive forces on the suppression chamber due to higher water clearing loads from downcomer vents and higher pressure buildup in the drywell.

General Electric BWR/4 STS B 3.6.2.5-1 Rev. 4.0

TSTF-568, Rev. 0 Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.5 BASES APPLICABILITY Drywell-to-suppression chamber differential pressure must be controlled when the primary containment is inert. The primary containment must be inert in MODE 1 with THERMAL POWER > [15]% RTP, since this is the condition with the highest probability for an event that could produce hydrogen. It is also the condition with the highest probability of an event that could impose large loads on the primary containment.

Inerting primary containment is an operational problem because it prevents primary containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in the unit startup and is de-inerted as soon as possible in the unit shutdown. As long as reactor power is < [15]% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert. Furthermore, the probability of an event that generates hydrogen or excessive loads on primary containment occurring within the first [24] hours following a startup or within the last

[24] hours prior to a shutdown is low enough that these "windows," with the primary containment not inerted, are also justified.

For the purposes of this Specification, a transition from MODE 2 to MODE 1 followed by a power ascension to above [15]% RTP is considered a startup. A reactor shutdown that results from a planned activity at least [24] hours prior to a power reduction to below [15]% RTP is considered a planned reactor shutdown for the purposes of this Specification. The [24] hour time period is a reasonable amount time to allow plant personnel to perform inerting or de-inerting.

ACTIONS A.1 If drywell-to-suppression chamber differential pressure is not within the limit, the conditions assumed in the safety analyses are not met and the differential pressure must be restored to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time provides sufficient time to restore differential pressure to within limit and takes into account the low probability of an event that would create excessive suppression chamber loads occurring during this time period.

B.1 If the differential pressure cannot be restored to within limits within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. This is done by reducing power to

[15]% RTP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to reduce reactor power from full power conditions in an orderly manner and without challenging plant systems.

General Electric BWR/4 STS B 3.6.2.5-2 Rev. 4.0

TSTF-568, Rev. 0 Drywell-to-Suppression Chamber Differential Pressure B 3.6.2.5 BASES SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS The drywell-to-suppression chamber differential pressure is regularly monitored to ensure that the required limits are satisfied. [ The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was developed based on operating experience relative to differential pressure variations and pressure instrument drift during applicable MODES and by assessing the proximity to the specified LCO differential pressure limit. Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal pressure condition.

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.

]

This Surveillance is modified by two Notes. The first Note allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to inert the primary containment after increasing THERMAL POWER above [15]% RTP. The second Note allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to de-inert the primary containment prior to decreasing THERMAL POWER to less than or equal to [15]% RTP. The Notes take exception to the requirements of the Surveillance being met (i.e., establishing the drywell-to-suppression chamber differential pressure limits for this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period). Inerting primary containment is an operational problem because it prevents primary containment access without an appropriate breathing apparatus.

Therefore, the primary containment is inerted as late as possible in a unit startup and is de-inerted as soon as possible in a unit shutdown. As long as reactor power is [15]% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert.

Furthermore, the probability of an event that generates hydrogen or excessive loads on primary containment occurring within the [24] hours following exceeding [15] RTP, or within [24] hours prior to reducing THERMAL POWER to less than or equal to [15]% RTP, is low enough that these "windows," with the primary containment not inerted, are also justified. The [24] hour time period is a reasonable amount time to allow plant personnel to perform inerting or de-inerting.

REFERENCES None.

General Electric BWR/4 STS B 3.6.2.5-3 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration B 3.6.3.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.2 Primary Containment Oxygen Concentration BASES BACKGROUND All nuclear reactors must be designed to withstand events that generate hydrogen either due to the zirconium metal water reaction in the core or due to radiolysis. The primary method to control hydrogen is to inert the primary containment. With the primary containment inert, that is, oxygen concentration < 4.0 volume percent (v/o), a combustible mixture cannot be present in the primary containment for any hydrogen concentration.

An event that rapidly generates hydrogen from zirconium metal water reaction will result in excessive hydrogen in primary containment, but oxygen concentration will remain < 4.0 v/o and no combustion can occur.

This LCO ensures that oxygen concentration does not exceed 4.0 v/o during operation in the applicable conditions.

APPLICABLE The Reference 1 calculations assume that the primary containment is SAFETY inerted when a Design Basis Accident loss of coolant accident occurs.

ANALYSES Thus, the hydrogen assumed to be released to the primary containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the primary containment.

Primary containment oxygen concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The primary containment oxygen concentration is maintained < 4.0 v/o to ensure that an event that produces any amount of hydrogen does not result in a combustible mixture inside primary containment.

APPLICABILITY The primary containment oxygen concentration must be within the specified limit when primary containment is inerted, except as allowed by the relaxations during startup and shutdown addressed below. The primary containment must be inert in MODE 1 with THERMAL POWER

> [15]% RTP, since this is the condition with the highest probability of an event that could produce hydrogen.

Inerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus.

Therefore, the primary containment is inerted as late as possible in the plant startup and de-inerted as soon as possible in the plant shutdown.

As long as reactor power is < [15]% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert. Furthermore, the probability of an event that generates hydrogen occurring within the first [24] hours of a startup, or within the last [24] hours before a reactor shutdown, is low enough that these "windows," when the primary containment is not inerted, are also justified.

General Electric BWR/4 STS B 3.6.3.2-1 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration B 3.6.3.2 For the purposes of this Specification, a transition from MODE 2 to MODE 1 followed by a power ascension to above [15]% RTP is considered a startup. A reactor shutdown that results from a planned activity at least [24] hours prior to a power reduction to below [15]% RTP is considered a planned reactor shutdown for the purposes of this Specification. The [24] hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting.

General Electric BWR/4 STS B 3.6.3.2-2 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration B 3.6.3.2 BASES ACTIONS A.1 If oxygen concentration is 4.0 v/o at any time while operating in MODE 1 above [15]% RTP, with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must be restored to

< 4.0 v/o within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is allowed when oxygen concentration is 4.0 v/o because of the low probability and long duration of an event that would generate significant amounts of hydrogen occurring during this period.

B.1 If oxygen concentration cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, power must be reduced to [15]% RTP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is reasonable, based on operating experience, to reduce reactor power from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.3.2.1 REQUIREMENTS The primary containment must be determined to be inert by verifying that oxygen concentration is < 4.0 v/o. [ The 7 day Frequency is based on the slow rate at which oxygen concentration can change and on other indications of abnormal conditions (which would lead to more frequent checking by operators in accordance with plant procedures). Also, this Frequency has been shown to be acceptable through operating experience.

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.

]

This Surveillance is modified by two Notes. The first Note allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to inert the primary containment after increasing THERMAL POWER above [15]% RTP. The second Note allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to de-inert the primary containment prior to decreasing THERMAL POWER to less than or equal to [15]% RTP. The Notes take exception to the requirements of General Electric BWR/4 STS B 3.6.3.2-3 Rev. 4.0

TSTF-568, Rev. 0 Primary Containment Oxygen Concentration B 3.6.3.2 the Surveillance being met (i.e., establishing the primary containment oxygen concentration limits for this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period). Inerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in a plant startup and de-inerted as soon as possible in a plant shutdown. As long as reactor power is [15]% RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert.

Furthermore, the probability of an event that generates hydrogen occurring within [24] hours following exceeding [15]% RTP, or within

[24] hours prior to reducing THERMAL POWER to less than or equal to

[15]% RTP, is low enough that these "windows," when the primary containment is not inerted, are also justified. The [24] hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting.

REFERENCES 1. FSAR, Section [6.2.5].

General Electric BWR/4 STS B 3.6.3.2-4 Rev. 4.0

Retrieved from "https://kanterella.com/w/index.php?title=TSTF-17-16,_Technical_Specifications_Task_Force_-_Transmittal_of_TSTF-568,_Clarify_Applicability_of_BWR/4_TS_3.6.2.5_and_TS_3.6.3.2&oldid=2621562"