BSEP 09-0077, Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program

Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program

ML092160405
Person / Time
Site:	Brunswick
Issue date:	07/23/2009
From:	Waldrep B Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 09-0077, TSC-2009-07
Download: ML092160405 (24)

Text

Progress Energy Benjamin C.Waldrep Vice President Brunswick Nuclear Plant Progress Energy Carolinas, Inc.

July 23, 2009 SERIAL: BSEP 09-0077 10 CFR 50.90 TSC-2009-07 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit No. 1 Docket No. 50-325/License No. DPR-71 Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc.,

is requesting a revision to the Technical Specifications (TSs) for the Brunswick Steam Electric Plant (BSEP), Unit No,. 1. The proposed license amendment revises TS 5.5.12, "Primary Containment Leakage Rate Testing Program," to add an exception for leakage testing associated with a repair modification being made to the Unit 1 X-2 primary containment penetration sleeve. This is the penetration sleeve associated with the drywell personnel airlock.

The planned modification will involve installation of a new, concentric sleeve inside of the existing penetration. Following the modification, the new sleeve will become the primary containment liner for this penetration. The exception being added to TS 5.5.12 will allow performance of an as-left local leak rate test on the new containment metallic liner welds in lieu of the integrated Type A test specified by Nuclear Energy Institute (NEI) 94-01, Revision 0, Industry Guidelinefor Implementing Performance-BasedOption of 10 CFR 50 Appendix J. An evaluation of the proposed license amendment is provided in Enclosure 1.

CP&L has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1), using the criteria in 10 CFR 50.92(c), and determined that this change involves no significant hazards considerations.

In accordance with 10 CFR 50.91(b), CP&L is providing a copy of the proposed license amendment to the designated representative for the State of North Carolina.

In order to support planning activities for the upcoming Unit 1 refueling outage (i.e.,

designated by CP&L as Outage B 118R1), which is currently scheduled to begin February 20, 2010, CP&L requests approval of the proposed license amendment by January 16, 2010. Once approved, the Unit 1 license amendment will be implemented prior to start-up from the 2010 Unit 1 refueling outage.

PO. Box 10429 Southport, NC28461 T> 910.457.3698 Aiýýr7 kAZK

Document Control Desk BSEP 09-0077 / Page 2 Regulatory commitments associated with the proposed amendment are documented in the . Please refer any questions regarding this submittal to Mr. Gene Atkinson, Supervisor - Licensing/Regulatory Programs, at (910) 457-2056.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on July 23, 2009.

Sincerely,

,njamin C. Waldrep WRM/wrm

Enclosures:

1. Evaluation of License Amendment Request

2. Marked-up Technical Specification Pages - Unit 1

3. Typed Technical Specification Pages - Unit 1

4. Regulatory Commitments

Document Control Desk BSEP 09-0077 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Acting Section Chief Radiation Protection Section Division of Environmental Health North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221

BSEP 09-0077 Enclosure 1 Page 1 of 12 Evaluation of Proposed License Amendment Request

Subject:

Request for License Amendment to Revise Technical Specification 5.5.12, Primary Containment Leakage Rate Testing Program 1.0 Description This letter is a request by Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., to amend the Technical Specifications (TS) for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2. The proposed license amendment revises Technical Specification (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," to add an exception for leakage testing associated with a repair modification being made to the Unit 1 X-2 primary containment penetration sleeve. This is the penetration sleeve associated with the drywell personnel airlock.

As a result of identifying corrosion on the external surface of the Unit 1 X-2 penetration sleeve, CP&L is planning a repair that involves installation of a new, concentric sleeve inside of the existing penetration sleeve. Following the modification, the new sleeve will become the primary containment pressure boundary for this penetration and thus requires leakage rate testing to be performed.

2.0 Proposed Change TS 5.5.12 requires that a Primary Containment Leakage Rate Testing Program be established and that the program be in accordance with NRC Regulatory Guide 1.163, Performance-BasedContainmentLeak-Test Program, dated September 1995. Regulatory Guide 1.163, Section C, "Regulatory Position," states that NEI 94-01, Revision 0, Industry Guidelinefor Implementing Performance-BasedOption of 10 CFR 50 Appendix J, provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50, subject to certain exceptions delineated in Section C of the Regulatory Guide. These requirements are further modified by certain exceptions that are listed in TS 5.5.12.

During the upcoming BSEP Unit 1 Refueling Outage 17 (i.e., designated by CP&L as Outage B 118R1), CP&L plans to install a new, concentric sleeve inside of the Unit 1 X-2 primary containment penetration to repair and replace the existing penetration sleeve.

Following the planned repair, the new sleeve will become the primary containment pressure boundary for this penetration and leakage rate testing of the penetration is required.

In lieu of performing a local Type A leakage rate test of the modified penetration, CP&L is proposing the addition of an exception to TS 5.5.12 that will allow performance of local leakage rate testing for the new containment penetration metallic liner welds and deferral

BSEP 09-0077 Enclosure 1 Page 2 of 12 of leakage rate testing of the entire penetration until the next regularly scheduled primary containment Type A test.

The proposed change will add an additional entry to the list of exceptions in TS 5.5.12 which modify the Primary Containment Leakage Rate Testing Program. The proposed new entry will read as follows:

Following modification of the X-2 primary containment penetration during Refueling Outage 17, as-left local leak rate testing of the new containment penetration metallic seam welds shall be performed at greater than or equal to Pa and Type A testing of the penetration may be deferred until the next scheduled primary containment Type A test.

For convenience, Enclosure 2 contains a marked-up version of the Unit 1 TSs showing the proposed change. Enclosure 3 provides a typed version of the affected Unit 1 TS pages.

These typed TS pages are to be used for issuance of the proposed amendment.

3.0 Technical Evaluation

Background

Figure 1 depicts a plan view of the drywell at elevation 25 foot and shows the locations of the large drywell penetrations - the equipment access hatch (i.e., penetration X-1) and the personnel air lock (i.e., penetration X-2).

N Figure D 1

/ _I_

BSEP 09-0077 Enclosure 1 Page 3 of 12 The primary difference between these two penetrations is that the equipment access hatch is sealed on the inside of the drywell shell, so that the penetration sleeve is not a part of the pressure boundary. The personnel air lock, however, is sealed on the outside of the drywell shell, where the airlock is attached, making the penetration sleeve part of the pressure boundary. Therefore, the inside surface of the sleeve is exposed to the drywell environment during plant operation. This is more clearly shown on Figure 2, which shows a horizontal section cut through the personnel air lock.

Figure 2

/

BSEP 09-0077 Enclosure 1 Page 4 of 12 Photograph of the X-2 penetration The penetration sleeve for the personnel air lock consists of a 10 foot, 2-5/8 inch inside diameter, 3/8-inch thick ASTM A516, Grade 70 steel sleeve welded to a 3-inch thick sleeve that bolts to the personnel air lock on the outside of the drywell.

During the Bi 17R1 outage in 2008, a VT-I visual inspection revealed two bulged areas in the Unit 1 X-2 primary containment penetration sleeve. The discovery of thinned areas on the bulges led to the decision to perform ultrasonic (i.e., UT) examinations of the entire Unit 1 X-2 penetration sleeve. These additional UT inspections identified many discrete locations, which exist over less than 3% of the surface area, that were below the minimum wall thickness established by the Containment Liner Specification (i.e., Reference 1).

During construction, the outside diameter of the sleeve was wrapped with two layers of 1/4-inch felt and the felt was covered with a layer of 60 mil ethylene propylene film. The felt was intended to permit the sleeve to expand when subjected to thermal loading.

BSEP 09-0077 Enclosure 1 Page 5 of 12 Corrosion of the Unit 1 X-2 penetration sleeve was first identified during performance of inspections in accordance with plant procedure 0PT-20.5.1 during the B 115RI outage in 2004. The corrosion was identified by bulges that were caused by the accumulation of corrosion products between the sleeve and the concrete backing. These bulges were identified by general visual examinations, not VT-I exams as were performed during the B1 17R1 outage in 2008. This corrosion is believed to have been caused by the felt wrapping on the outside of the sleeve, which became wet during original construction.

A general visual inspection of the Unit 1 X-2 penetration sleeve was performed during the B1 16R1 outage in 2006, and no problems were identified.

Boat samples of degraded areas of the sleeve, including corrosion products and felt wrapping, were taken during both the B 115RI and B 117R1 outages and sent to CP&L's Harris Environment & Energy Center for evaluation of potential ongoing corrosion mechanisms. These evaluations identified that the pitting and corrosion on the concrete side of the sleeve was caused by under-deposit corrosion, and the dry condition of the corrosion products and the felt indicated that the corrosion mechanism is no longer active.

During the B 115RI outage (i.e., calendar year 2004), all areas of the sleeve that were found to be below the minimum wall thickness defined in the Containment Liner Specification were repaired. In addition, a Type A integrated leakage rate test (ILRT),

which was already scheduled for performance, was successfully conducted after the repairs. The degradation identified during the BI 17R1 outage was evaluated with the determination that the Unit 1 X-2 penetration is operable, but degraded, with an interim use-as-is qualification until the next refueling outage which is scheduled to begin in February 2010.

Discussion The Code of Record for the fourth 10-year inservice inspection interval at the Brunswick Plant is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2001 Edition with 2003 Addenda. Paragraph IWE-5221 of the ASME Code,Section XI states:

Except as noted in IWE-5222, repair/replacement activities performed on the pressure retaining boundary of Class MC or Class CC components shall be subjected to a pneumatic leakage test in accordance with the provisions of Title 10, Part 50 of the Code of Federal Regulations, Appendix J, Paragraph IV.A.

10 CFR Part 50, Appendix J includes two options, A and B, either of which may be chosen for meeting the leakage testing requirements contained in the appendix. Option A provides prescriptive requirements, whereas Option B provides performance-based requirements.

BSEP 09-0077 Enclosure 1 Page 6 of 12 By letter dated November 20, 2008, the NRC approved Brunswick-specific 10 CFR 50.55a request CIP-01 as an alternative to the paragraph IWE-5221 of the ASME Code,Section XI. CIP-01 allows repairs, replacements, or modification activities that affect the containment leakage integrity to be subjected to the applicable leakage rate testing requirements specified in Nuclear Energy Institute (NEI) 94-01. As such, CIP-01 allows the use of the leakage testing requirements contained in Option B of Appendix J to 10 CFR Part 50 (i.e., NEI 94-0 1) for the Unit 1 X-2 penetration repair/replacement modification.

Regulatory Guide 1.163, Section C, "Regulatory Position," states that NEI 94-01, Revision 0, provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50, subject to certain exceptions delineated in Section C of the Regulatory Guide. For Brunswick Unit 1, these requirements are further modified by certain exceptions that are listed in TS 5.5.12.

NEI 94-01, Revision 0, Section 9.2.4, "Containment Repairs and Modifications," states the following:

Repairs and modifications that affect the containment leakage integrity require leakage rate testing (Type A testing or local leakage rate testing) prior to returning the containment to operation. Testing may be deferred to the next regularly scheduled Type A test for the following repairs or modifications:

• Welds of attachments to the surface of steel pressure-retaining boundary; 0 Repair cavities, the depth that does not penetrate required design steel wall by more than 10%; or 0 Welds attaching to steel pressure-retaining boundary penetrations, where the nominal diameter of the welds or penetrations does not exceed one inch.

During the BI 18R1 outage, CP&L plans to install a new, concentric sleeve inside of the Unit 1 X-2 containment penetration to repair the existing containment penetration sleeve.

Following the planned modification, the new sleeve will become the primary containment liner for this penetration. Based on the guidance of NEI 94-01, the Unit 1 X-2 penetration modification requires leakage rate testing following the modification, prior to returning the containment to operation. Local leakage rate testing of the new penetration sleeve metallic seam welds will be performed by welding test channels over the seam welds of the penetration sleeve and the circumferential tie-in welds to the existing primary containment liner. However, performance of a local Type A leakage rate test of the penetration is not practical because the interior side of the penetration is open to the drywell and cannot be sealed to support pressurizing the penetration to the required test pressure of Pa, 49 psig.

BSEP 09-0077 Enclosure 1 Page 7 of 12 Therefore, in lieu of a Type A leakage rate test for the penetration, CP&L proposes to only perform local leak rate testing of the new metallic pressure-boundary seam welds of the Unit 1 X-2 penetration, and to defer pressure testing of the entire Unit 1 X-2 penetration until the next regularly scheduled primary containment Type A test, which is currently scheduled for calendar 2014.

Following installation of the replacement penetration sleeve, an "as-left" local leak rate test of the welds will be performed by welding test channels over the seam welds of the new penetration sleeve and the circumferential tie-in welds to the existing primary containment liner. A local leakage rate test provides the most accurate and direct method of assuring the leak integrity of the repair welds. The local leakage rate test is a superior test for determining leakage at the installation weld areas compared to the integrated Type A test.

The local leakage rate test will directly measure leakage at the installation welds, while a Type A test will only measure total primary containment leakage and thus is a less sensitive test than a local leakage rate test. This leakage test is being performed to re-establish the leak-tight integrity of the replacement penetration sleeve and the primary containment liner tie-in welds due to installation of the replacement penetration sleeve. No other work is being performed during the 2010 refueling outage which warrants a Type A leakage rate test of the entire primary containment.

The local leak rate tests which will be performed will meet the requirements of ANSI/ANS 56.8-1994. CP&L's acceptance criterion for leakage of the new penetration sleeve welds and tie-in welds will be zero leakage. This acceptance criterion is more stringent than that of a Type A test. Thus, if there is any leakage of the replacement penetration sleeve welds and containment tie-in welds, it will be identified by the local leakage rate test and corrected. After the local leak rate tests are completed, the test channels will be left in place, unless required to be removed to permit recirculation motor removal. However, the test ports will be left unplugged in order to ensure the replacement penetration sleeve seam welds and primary containment liner tie-end welds are exposed to the containment atmosphere during the next scheduled Type A ILRT.

In order to ensure the overall integrity of the pressure-retaining plates to be used for the replacement penetration sleeve, the plates will be volumetrically tested (i.e., UT) in accordance with ASTM A 577, StandardSpecificationfor UltrasonicAngle-Beam Examinations of Steel Plates,and A 578, StandardSpecificationfor Straight Beam UltrasonicExaminationof Plain and Clad Steel Platesfor SpecialApplications. Plates with identified through-wall defects will be rejected.

The new penetration sleeve welding will be performed by qualified personnel in accordance with original design requirements. Consistent with the original design requirements, examinations will be performed on the metallic liner repair welds. As a minimum 100% surface examination (i.e., liquid penetrant or magnetic particle examination) and spot volumetric (i.e., UT or radiography) will be performed on the

BSEP 09-0077 Enclosure 1 Page 8 of 12 metallic repair welds. A 100% VT-I examination of the accessible surfaces of the new penetration sleeve and containment liner affected by the repair will be conducted after completion of the repair. Qualified individuals will perform these VT-I examinations.

In addition, although the Unit 1 X-2 penetration sleeve is considered to be a degraded condition, based on the guidance contained in NRC Regulatory Issue Summary 2005, Revision 1, Revision to NRC Inspection Manual Part9900 Technical Guidance, "OperabilityDeterminationsand FunctionalityAssessments for Resolution of Degradedor Nonconforming ConditionsAdverse to Quality or Safety, "the Unit 1 X-2 penetration is operable and the integrity of this portion of the primary containment pressure boundary is not being compromised by installation of the replacement penetration sleeve. As previously discussed, a Type A (i.e., ILRT) test was successfully conducted during the B 115R1 outage (i.e., calendar year 2004) following repairs of areas of the sleeve that were found to be below the minimum wall thickness defined in the Containment Liner Specification. Additional repairs of the penetration sleeve were performed in 2008 and local leakage rate testing of the affected areas was satisfactorily performed.

Conclusion The exception being added to TS 5.5.12 will allow performance of an as-left local leak rate test on the new metallic containment liner seam welds in conjunction with deferral of leakage rate testing of the entire penetration until the next scheduled Type A. Performance of local leak rate testing of the new penetration sleeve pressure-boundary welds will provide the most accurate and direct method of assuring the leak integrity of the repair welds by directly measuring the leakage at the new welds. Such testing is superior to a local Type A test, which would only measure total leakage of the entire penetration, or an integrated Type A test, which would only measure total primary containment leakage.

Installation of the new penetration sleeve will meet all design requirements and will not adversely affect the structural integrity of the existing penetration sleeve. Compliance with the original Containment Liner Specification and the associated inspections and tests of the replacement penetration sleeve are adequate to ensure the structural integrity of the primary containment will be maintained.

4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50, Appendix A, General Design Criterion (GDC) 50, Containment design basis, states that the reactor containment structure, including access openings, penetrations, and

BSEP 09-0077 Enclosure 1 Page 9 of 12 the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. The planned replacement activity for the Unit 1 X-2 penetration sleeve is being performed to restore the design margins for this containment structure penetration, and the planned local leakage rate testing will ensure that the containment will not exceed its design leakage rates. Therefore, the primary containment will continue to meet the requirements of GDC 50.

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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

CP&L 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 GDC differently than described in the Updated Final Safety Analysis Report (UFSAR).

4.2 No Significant Hazards Consideration The proposed change will revise Technical Specification (TS) 5.5.12, "Primary Containment Leakage Rate Testing Program," to add an exception associated with leakage testing following a one-time repair being made to the Unit 1 X-2 primary containment penetration sleeve. This is the penetration sleeve associated with the drywell personnel airlock. As a result of identifying corrosion on the external surface of the Unit 1 X-2 penetration sleeve, CP&L is planning a repair that involves installation of a new, concentric sleeve inside of the existing penetration sleeve. Following the repair, the new sleeve will become the primary containment liner for this penetration. The exception will allow performance of an as-left local leak rate test on the new containment metallic liner welds in lieu of the Type A test of the entire penetration as specified by Nuclear Energy Institute (NEI) 94-0 1, Revision 0, Industry Guidelinefor Implementing Performance-Based Option of 10 CFR 50 Appendix J.

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

BSEP 09-0077 Enclosure 1 Page 10 of 12 (3) Involve a significant reduction in a margin of safety.

CP&L has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, as discussed below.

I1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The proposed amendment revises Technical Specification 5.5.12, "Primary Containment Leakage Rate Testing Program," to provide a one-time exception allowing the use of certain local leakage rate testing in lieu of overall primary containment Type A leakage rate test for verifying the leak-tightness of the Unit 1 X-2 containment penetration following installation of a new replacement penetration sleeve. The proposed change does not impact any accident initiators and does not involve the addition or removal of any plant equipment. As such, no individual precursors of an accident are affected and the proposed amendment does not increase the probability of a previously analyzed event.

The consequences of a previously analyzed accident are dependent on the initial conditions assumed for the analysis, the behavior of the fuel during the analyzed accident, the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The proposed change will allow for the use of a local leakage rate test, rather than an overall primary containment Type A leakage rate test, following installation of the replacement containment penetration sleeve. A local leakage rate test provides the more accurate and direct method of assuring the leak integrity of the new welds, versus a Type A leakage test, and therefore is a superior test for determining the leak-tight integrity of the installation weld areas compared to an

• overall Type A test. As such, the function of the primary containment for mitigating the consequences of an accident is not being adversely impacted and the proposed amendment does not involve a significant increase in the consequences of an accident previously evaluated.

Based on the above, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

BSEP 09-0077 Enclosure 1 Page 11 of 12 Response: No Creation of the possibility of a new or different kind of accident requires creating one or more new accident precursors. New accident precursors may be created by modifications of plant configuration, including changes in allowable modes of operation. The proposed amendment permits the use of local leakage rate testing in lieu of an overall Type A leakage rate test following installation of a replacement containment penetration sleeve. The proposed change does not change or introduce any new equipment, modes of system operation, or failure mechanisms; therefore, no new accident precursors are created. Based on the above information, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No The proposed change permits the use, on a one-time basis, of local leakage rate testing in lieu of an overall primary containment Type A leakage rate test following installation of a replacement containment penetration sleeve. The proposed change has no impact on equipment design or fundamental operation, and there are no changes being made to safety limits or safety system allowable values that would adversely affect plant safety as a result of the proposed change. The use of local leakage rate testing in lieu of a Type A leakage rate test following installation of a replacement containment penetration sleeve will directly measure the leakage at the new welds, while a Type A test will only measure total primary containment leakage and thus is a less sensitive test than the local leakage rate test.

The local leakage rate test will ensure that the leak-tight integrity of the penetration sleeve and primary containment liner has been re-established, thus ensuring the margin of safety in the original safety analyses is maintained. Based on the above information, the proposed amendment does not result in a significant reduction in the margin of safety.

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

6.0 Environmental Considerations 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, "Standards for Protection Against Radiation," or would change an

BSEP 09-0077 Enclosure 1 Page 12 of 12 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), "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review." Therefore, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 References

1. Progress Energy Nuclear Generation group Specification 015-001, Containment StructuralSteel Liners.

2. NRC Regulatory Guide 1.163, Performance-BasedContainment Leak-Test Program, dated September 1995

3. ASTM A 577, StandardSpecificationfor UltrasonicAngle-Beam Examinationsof Steel Plates

4. ASTM A 578, StandardSpecificationfor Straight Beam UltrasonicExamination of Plain and Clad Steel Platesfor SpecialApplications

5. Nuclear Energy Institute (NEI) 94-01, Revision 0, Industry Guidelinefor Implementing Performance-BasedOption of 10 CFR 50 Appendix J

(

BSEP 09-0077 Enclosure 2 Marked-up Technical Specification Pages - Unit 1

I I FR INFORMATION ONLY NOT BEING REVISED Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

2. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4. Other appropriate limitations and remedial or compensatory actions.

b. A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

1. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or

2. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or

3. A required system redundant to support system(s) for the supported systems described in b.1 and b.2 above is also inoperable.

c. 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.

5.5.12 Primary Containment Leakage Rate Tostinq Program A primary containment leakage rate testing program-shall establish requirements to implement the leakage rate testing of the primary 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, as modified by the following exceptions:

(continued)

Brunswick Unit 1 5.0-15 Amendment No. 209 I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Pro-gram (continued)

a. The visual examination of concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

b. The visual examination of the metallic shell, penetrations, and appurtenances intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

c. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at Pa as specified in Nuclear Energy Institute Guideline 94-01, Revision 0;

d. Reduced duration Type A tests may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

e. Performance of Type C leak rate testing of the hydrogen and oxygen monitor isolation valves is not required; and

f. Performance of Type C leak rate testing of the main steam isolation valves at a pressure less than Pa instead of leak rate testing at Pa as specified in ANSI/ANS 56.8-199-4.

Th k calculated primary containme t internal pressure for the design basis loss o coolant accident, Pa, is 49 psig.

The ma 'mum allowable primary containment leakage rate, La, shall be 0.5% of primary ntainrent air weight per day at Pa.

L gr g. Following modification of the X-2 primary containment L re ac penetration during Refueling Outage 17, as-left local leak rate testing of the new containment penetration metallic

a. Primg th seam welds shall be performed at greater than or equal to Pa and Type A testing of the penetration may be deferred until program, the next scheduled primary containment Type A test.

and C te (continued)

Brunswick Unit 1 5.0-16 Amendment No. 245 I

FOR INFORMATION ONLY NOT BEING REVISED Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakagqe Rate Testing Program I (continued)

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is < 0.05 La when tested at > Pa.

2) For each air lock door, leakage rate is < 5 scfh when the gap between the door seals is pressurized to > 10 psig.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program frequencies.

5.5.13 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation (CREV) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the ,ORE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) deterroining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the CREV System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the assessment of the CRE boundary.

(continued)

Brunswick Unit 1 5.0-17 Amendment No. 248 I

BSEP 09-0077 Enclosure 3 Typed Technical Specification Pages - Unit 1

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakaqe Rate Testing Program (continued)

a. The visual examination of concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

b. The visual examination of the metallic shell, penetrations, and appurtenances intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

c. Following air lock door seal replacement, performance of door seal leakage rate testing with the gap between the door seals pressurized to 10 psig instead of air lock testing at Pa as specified in Nuclear Energy Institute Guideline 94-01, Revision 0;

d. Reduced duration Type A tests may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

e. Performance of Type C leak rate testing of the hydrogen and oxygen monitor isolation valves is not required; and

f. Performance of Type C leak rate testing of the main steam isolation valves at a pressure less than Pa instead of leak rate testing at Pa as specified in ANSI/ANS 56.8-1994.

g. Following modification of the X-2 primary containment penetration during Refueling Outage 17, as-left local ieak rate tests of the new containment penetration metallic seam welds shall be performed at greater than or equal to Pa and Type A testing of the penetration may be deferred until the next scheduled primary containment Type A test.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, Pa, is 49 psig.

The maximum allowable primary containment leakage rate, La, shall be 0.5% of primary containment air weight per day at Pa.

(continued)

Brunswick Unit 1 5.0-16 Amendment No. I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakage Rate Testing Program (continued)

Leakage rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the

.leakage rate acceptance criteria are < 0.60 La for Type B and C tests and

_ 0.75 La for Type A tests.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is < 0.05 La when tested at _>Pa.

2) For each air lock door, leakage rate is < 5 scfh when the gap between the door seals is pressurized to > 10 psig.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program frequencies.

5.5.13 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation (CREV) System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. 'The program shall ensure that adequate radiation protection is provided to permit occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

(continued)

Brunswick Unit 1 5.0-17 Amendment No. I

Programs and Manuals 5.5 5.5 Programs and Manuals Control Room Envelope Habitability Program (continued)

d. Measurement, at designated locations, of the CRE pressure relative to external areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem of the CREV System, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

Brunswick Unit 1 5.0-17a Amendment No. I

BSEP 09-0077 Enclosure 4 List of Regulatory Commitments The following table identifies those actions committed to by Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., in this document.

Any other actions discussed in the submittal represent intended or planned actions by CP&L.

They are described for the NRC's information and are not regulatory commitments. Please notify the Manager - Support Services at the Brunswick Steam Electric Plant of any questions regarding this document or any associated regulatory commitments.

Committed date or Commitment outage Local leakage rate testing of the new X-2 containment penetration metallic Prior to start-liner welds associated with installation of a replacement X-2 penetration up following sleeve will be performed. the B 118R1 outage.

As part of the X-2 containment penetration replacement, a 100% surface Prior to start-examination (i.e., liquid penetrant or magnetic particle examination) and up following spot volumetric (i.e., UT or radiography) will be performed on the metallic the BI 18R1 liner repair welds. outage.

As part of the X-2 containment penetration replacement, a 100% VT-I Prior to start-examination of the accessible surfaces of the new penetration sleeve and up following containment liner affected by the repair will be conducted. the B 118R1 outage.