RS-13-240, License Amendment Request to Administratively Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program

License Amendment Request to Administratively Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program

ML13284A106
Person / Time
Site:	Braidwood
Issue date:	10/10/2013
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-13-240
Download: ML13284A106 (12)

Text

RS-1 3-240 October 10, 2013 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN-50-456 and STN-50-457

Subject:

License Amendment Request to Administratively Revise Technical Specification 5.5.16, "Containment Leakage Rate Testing Program" In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2. The proposed change would revise the date for the performance of the Braidwood Station, Unit 2, Type A or integrated containment leakage rate test (ILRT) described in Technical Specification 5.5.16, "Containment Leakage Rate Testing Program." The proposed change revises the date for the performance of the Type A test from "no later than May 4, 2014," to "prior to entering MODE 4 at the start of Cycle 18." Additionally, EGC is proposing to establish a requirement for Braidwood Station, Unit 2, to exit the MODEs of applicability for Containment as described in Technical Specification 3.6.1, "Containment" (i.e., MODEs 1 - 4), no later than May 4, 2014.

The attached request is subdivided as follows:

• provides an evaluation of the proposed change.

• provides the current TS page with the proposed change indicated.

The proposed amendment has been reviewed by the Braidwood Station Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed license amendment by April 18, 2014. Once approved, the amendment will be implemented within 14 days.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

October 10, 2013 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mitchel Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10th day of October 2013.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments:

1)

Evaluation of Proposed Change 2)

Markup of Technical Specifications Page cc:

Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT I Evaluation of Proposed Change

Subject:

License Amendment Request to Administratively Revise Technical Specification 5.5.16, "Containment Leakage Rate Testing Program" 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

5.0 REGULATORY EVALUATION

5.1 Applicable Regulatory Requirements/Criteria 5.2 No Significant Hazards Consideration 5.3 Conclusions

7.0 REFERENCES

Page 1 of 8

ATTACHMENT I Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2.

Exelon Generation Company, LLC (EGC) proposes to revise the date for the performance of the Braidwood Station, Unit 2, Type A, or integrated containment leakage rate test (ILRT) described in Administrative Controls Technical Specification 5.5.16, "Containment Leakage Rate Testing Program." The proposed change revises the date for the performance of the Type A test in Technical Specification 5.5.16, Exception 2, from "no later than May 4, 2014," to "prior to entering MODE 4 at the start of Cycle 18." Additionally, EGC is proposing to establish an additional requirement in Technical Specification 5.5.16, Exception 2, for Braidwood Station, Unit 2, to exit the MODEs of applicability for Containment as described in Technical Specification 3.6.1, "Containment" (i.e., MODEs 1 - 4), no later than May 4, 2014.

Approval of this amendment application is requested by April 18, 2014, in order to support the completion of the Braidwood Station, Unit 2 ILRT during the next upcoming refueling outage currently scheduled to commence on May 3, 2014. Once approved, the amendment will be implemented within 14 days.

2.0 DETAILED DESCRIPTION Technical Specification 5.5.16, Exception 2, currently states:

"NEI 94 1995, Section 9.2.3: The first Unit 2 Type A test performed after the May 4, 1999 Type A test shall be performed no later than May 4, 2014."

The revised Technical Specification 5.5.16, Exception 2, would state:

"NEI 94-01 -1995, Section 9.2.3: In support of the Spring 2014 refueling outage, Unit 2 shall be placed in a MODE of operation where Containment is not required to be OPERABLE as described in Technical Specification 3.6.1, "Containment," no later than May 4, 2014. The first Unit 2 Type A test performed after the May 4, 1999, Type A test shall be performed prior to entering MODE 4 at the start of Unit 2, Cycle 18." provides a mark-up of the affected Technical Specifications (TS) page for the proposed change.

3.0 BACKGROUND

On April 2, 2008, the NRC issued Operating License Amendment No. 149 for Braidwood Station, Units 1 and 2 (i.e., NRC Accession No. ML080640290). Under Amendment No. 149, the NRC approved a one-time ILRT interval extension of 5 years (i.e., from 10 to 15 years). The revised Technical Specification 5.5.16, "Containment Leak Rate Testing Program," notes a specific date by which each unit's next Type A test must be performed. The due date for the Page 2of8

ATTACHMENT 1 Evaluation of Proposed Change Unit 2 Type A test described in Technical Specification 5.5.16, Exception 2, May 4, 2014, is based on the last Braidwood Unit 2 ILRT, which was performed on May 4, 1999.

The current Braidwood Station, Unit 2 refueling outage schedule does not support completion of the Unit 2 Type A test by May 4, 2014; however, Unit 2 is currently scheduled to exit the MODEs of applicability for Containment as described in Technical Specification 3.6.1, "Containment," no later than May 4, 2014.

4.0 TECHNICAL EVALUATION

The Unit 2 refueling outage is scheduled to begin on May 3, 2014. The Unit 2 ILRT, assuming target dates are met, will be performed on or about May 21, 2014. This would result in a brief extension from the current TS due date. Additionally, as described in Section 3.0 above, Unit 2 will not be in an operational MODE that will require containment to be operable as described in Technical Specification 3.6.1 during the entire period of the proposed extension. In operating MODES 5 and 6, the probability and consequences of design basis accidents that could cause a release of radioactive material into containment are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 to prevent leakage of radioactive material from containment. The requirements for containment during MODE 6 are addressed in Technical Specification 3.9.4, "Containment Penetrations," and are only applicable during the movement of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />).

Braidwood Station has an administrative restriction in Technical Requirements Manual Section 3.9.a, "Decay Time," that prevents the movement of recently irradiated fuel; therefore, EGC will not be moving any recently irradiated fuel in the Braidwood Station, Unit 2 reactor core during the period of the proposed extension.

Based on the short timeframe of the proposed extension, and the fact that the unit will not be in an operating MODE or condition requiring containment to be OPERABLE, administratively extending the Unit 2, Type A test interval would not result in any significant changes to the analyses described in the Referenced license amendment request dated April 4, 2007.

Furthermore, Technical Specification 5.5.16 states that the Containment Leakage Rate Testing Program is conducted in accordance with NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0. NEI 94-01, Revision 0, Section 9.2.2, "Initial Test Intervals," states that, "If the test interval ends while primary containment integrity is either not required or it is required solely for shutdown activities, the test interval may be extended indefinitely. However, a successful Type A test shall be completed prior to entering the operating mode requiring primary containment integrity." Administratively extending the Unit 2 Type A test interval past May 4, 2014, while Unit is in an outage and not in an operating MODE or condition where containment is required to be operable, is consistent with this guidance.

No performance issues exist that would preclude the NRC from allowing the administrative extension of the performance of the Braidwood Station, Unit 2 Type A containment leakage test.

Specifically, the results of the previous two Unit 2 ILRTs are shown in Table 1 below.

Page 3of8

ATTACHMENT I Evaluation of Proposed Change Table 1: Braidwood Station, Unit 2 ILRT Test Results ILRT Test Date Total Leakage

(% weight per day)

Acceptance Limit

(% weight per day) 11/94 0.053

< 0.075 5/99 0.063

< 0.075 Additionally, all surveillances related to Braidwood Station, Unit 2 containment performed since April 2007 have been completed satisfactorily:

Local leak rate tests The maximum leakage rate attributed to a penetration leakage path (MXPLR) is the larger leak rate for a containment penetration. The sum of the MXPLRs for all local leak rate tests performed for Unit 2 since April 2007 have been acceptable.

Inspections of the materiel condition of the containment structure Following the issuance of Braidwood Station, Unit License Amendment 149, a visual exam of the Unit 2 containment structure was performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Subsection IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants," (ASME IWE) and Subsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Plants,"

(ASME IWL). The results of the examination revealed no degradation that adversely affects the structural integrity of the containment. The conditions and indications identified were primarily cosmetic in nature.

Therefore, since Braidwood Station, Unit 2 will not be in a MODE where containment is required beyond May 4, 2014, and since all examination and testing results related to the operation and materiel condition of the Braidwood Station, Unit 2 containment since April 2007 have been satisfactory, the technical justification used as the basis for the approval of Braidwood Station, Units 1 and 2 Operating License Amendment No. 149 remains valid and would not be altered by operating in the proposed manner.

5.0 REGULATORY EVALUATION

5.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

General Design Criterion 16, "Containment Design," states that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

Page 4of8

ATTACHMENT 1 Evaluation of Proposed Change General Design Criterion 19, "Control Room," states that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including LOCAs, and that adequate radiation protection shall be provided. Adequate shielding is provided to maintain tolerable radiation levels in the control room under accident conditions for the duration of the accident.

General Design Criterion 38, "Containment Heat Removal," states that a system to remove heat from the reactor containment shall be provided that rapidly reduces, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintain them at acceptable low levels.

Additionally, 10 CFR 50, Appendix J, Section V, "Inspection and Reporting of Tests,"

Option B specifies that the regulatory guide (i.e., RG 1.163) or other implementing documents used to develop a performance-based leakage testing program must be included, by general reference, in the plant's TS. Deviations from guidelines endorsed in a regulatory guide are to be submitted as a revision to the plant's TS.

No changes beyond the short administrative extension of the due date to prior to entering MODE 4 at the start of Unit 2, Cycle 18 for the performance of the Braidwood Station, Unit 2 Type A test are proposed; therefore, these requirements will continue to be met following the implementation of the change as follows:

The change has no impact on the leak-tightness of Braidwood Station, Unit 2 containment or the environmental qualification of equipment within containment.

The proposed change has no affect on control room or offsite radiation levels and corresponding dose.

No changes to the design or operation of containment heat removal systems are proposed, so the proposed change has no bearing on the previously analyzed containment pressure response.

The proposed change involves a modification of a previously reported deviation from NRC guidelines, and while administrative in nature, it is required to be submitted as a revision to the Braidwood Station, Units 1 and 2 TS.

In conclusion, EGC has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements or criteria.

5.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2. The proposed change would revise Technical Specification 5.5.16, "Containment Leakage Rate Testing Program," through an extension to the due date for the performance of the Braidwood Station, Unit 2 containment integrated leak Page 5 of 8

ATTACHMENT I Evaluation of Proposed Change rate test (ILRT) (i.e., Type A test). The Braidwood Station, Unit 2 Type A test is currently required by Technical Specification, 5.5.16, Exemption 2 to be performed by May 4, 2014. EGC proposes to administratively extend this date from May 4, 2014, to prior to entering MODE 4 at the start of Unit 2, Cycle 18, an extension of approximately 17 days. Braidwood Station, Unit 2 will not be in a MODE of operation that requires Containment to be operable at any point during the extension.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change for Braidwood Station, Units 1 and 2 using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

1.

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

Response: No.

The proposed change to the Braidwood Station, Units 1 and 2 Containment Leakage Rate Testing Program does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment function is to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents.

As such, the containment itself, and the testing requirements to periodically demonstrate the integrity of the containment, exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

Implementation of the proposed change will continue to provide adequate assurance that during design basis accidents, the containment and its components would limit leakage rates to less than the values assumed in the plant safety analyses. Therefore, the consequences of an accident previously evaluated will not be increased by this proposed change.

Therefore, operation of the facility in accordance with the proposed administrative change to the date for the performance of the Unit 2, Type A containment leak rate test will not involve a significant increase in the probability Page 6of8

ATTACHMENT 1 Evaluation of Proposed Change or consequences of an accident previously evaluated.

2.

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

Response: No.

The containment, and the testing requirements to periodically demonstrate the integrity of the containment, exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is currently operated or controlled.

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

3.

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

Response: No.

This proposed change does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as proposed, will continue to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained.

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

Based on the above evaluation, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

5.3 Conclusions 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 the operation of Braidwood Station, Units 1 and 2 in the proposed manner, (2) such activities will be conducted in compliance with the NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

Page 7of8

ATTACHMENT I Evaluation of Proposed Change

6.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated the proposed amendment for environmental considerations. The review has resulted in the determination that the proposed amendment would change requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent 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, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

7.0 REFERENCES

Letter from D. M. Benyak (EGC) to NRC, "Request for Amendment to Technical Specification 5.5.16, "Containment Leakage Rate Testing Program," dated April 4, 2007 Page 8of8

ATTACHMENT 2 Markup of Technical Specifications Page Braidwood Station UNITS 1 AND 2 Docket Nos. STN-50-456 and STN-50-457 Facility Operating License Nos. NPF-72 and NPF-77 REVISED TS PAGE 5.5-21

5.5 Programs and Manuals Programs and Manuals 5.5 5.5.15 5.5.16 In support of the Spring 2014 refueling outage, Unit 2 shall be placed in a MODE of operation where containment is not required to be OPERABLE in accordance with Technical Specification 3.6.1, "Containment," no later than May 4, 2014.

Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0, as modified by the following exceptions:

1.

NEI 94 1995, Section 9.2.3: The first Unit 1 Type A test performed after the October 5, 1998 Type A test shall be performed no later than October 5, 2013.

2.

NEI 94 1995, Section 9.2.3:

he first Unit 2 Type A test performed after the May 4, 1999 Type A test shall be performed 2014.

containment internal pressure for the design t accident, Pa, is 42.8 psig for Unit 1 and containment leakage rate, La, at Pa, shall be air weight per day.

kage rate acceptance criterion is <_ 1.0 La-t unit startup following testing in accordance am, the leakage rate acceptance criteria are e Type B and C tests and < 0.75 La for Type A prior to entering MODE 4 at the start of Unit 2, Cycle 18 The peak calculated basis loss of coola 38.4 psig for Unit The maximum allowabl 0.20% of containment a.

Containment le During the fir with this prog

< 0.60 La for t tests; and BRAIDWOOD - UNITS 1 & 2 5.5 - 21 Amendment