L-99-002, Requests NRC Appproval of Enclosed Relief Requests RR-31 Through RR-39 for Farley Nuclear Plant,Units 1 & 2 Isi. Approval Requested by 991231

Requests NRC Appproval of Enclosed Relief Requests RR-31 Through RR-39 for Farley Nuclear Plant,Units 1 & 2 Isi. Approval Requested by 991231

ML20207H661
Person / Time
Site:	Farley
Issue date:	03/03/1999
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NEL-99-0027, NEL-99-27, NUDOCS 9903160062
Download: ML20207H661 (50)

Text

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Dave Mc9 /

S uthernNuct:ar Vice Presment Operating Compa:y farley Project P.O. Box 1295 Birmingham. Alabama 35201 Tel 205.992.5131 l

SOUTHERN March 3, 1999

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Energyto ServeYour%rld" l

l Docket Nos.:

50-348 NEle99-0027 l

50-364 U. S. Nuclear Regulatory Commission XITN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Inservice Inspection Relief Reauest Nos. 31 Throunh 39 Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.55, Southern Nuclear Operating Company (SNC) is requesting NRC approval of enclosed Relief Requests RR-31 through RR-39 for Farley Nuclear Plant Units 1 and 2. Ap royal is requested by December 31,1999 to support activities to be performed during the Unit I refueling outage scheduled for the spring of 2000.

There are no new commitments contained in this lenei. If you have any questions, please ad.ise.

Rest.ectfully submitted,

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U. S. Nuclear Regulatory Commission cc:

Southern Nuclear Operatina Compant Mr. L. M. Stinson, General Manager - Farley U. S. Nuclear Regulatory Commission. Washinaton. D. C.

Mr. J. I. Zimmerman, Licensing Project Manager - Farley U. S. Nuclear Rem >1atarv Cor. mission. Region II Mr. L. A. Rtyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley l

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ENCLOSUREI Farley Nuclear Plant-Unit 1 Third Ten Year Interval Reauest for Relief No. RR-31 1

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l SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEAR INTERVAL i

l REQUEST FOR RELIEF NO. RR-31 I.

System / Components (s) for Which Reliefis Reauested: Seals (including 0-rings) and gaskets of

lass MC (Metallic Contamment) pressure retaining components, Examination Category E-D, item Numbers E5.10 and E5.20.

'Ihis request for relief applies to the following components that incorporate seals and gaskets as the containment pressure boundary.

Electrical penetrations.

3 Two personnel airlock doors with seals, including door operating mechanism penetrations that are part of the containment pressure boundary and the contamment equipment hatch.

Containment penetrations whose design incorporates resilient seals, gaskets, or sealant compounds.

II, Code Reauirement: 10 CFR 50.55a was amended in the Federal Register on August 8,1996, to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing i

contamment examinations 'Ihe 1992 Edition wah 1992 Addenda of ASME Section XI, Table IWE 2500-1, Examination Category E-D, item Numbers E5.10 and E5.20, requires seals and gaskets on airlocks, hatches, and other devices that are required to assure containment leak-tight integrity to be visually examined once cach interval Ill.

Code Requirement for Which Reliefis Requested Reliefis requested from performing the Code-required VT-3 visual exammation on the above identified containment seals and gaskets.

IV.

Basis for Relief: Practical VT-3 visual examination considerations of these seals and gaskets would require thejoints to be Aa=== bled since many of the surfaces of seals and gaskets are normally inaccessible. 'Ihe ASME Code Committee recognized that disassembly of thejoints to perform visual examinations was not warranted, and the 1998 Edition of ASME Section XI removed the examination requirement.

The proposed altemate examination (Appendix J, Option B) provides a periodic, non-intrusive test method which will ensure that the integrity of the seals and gaskets is being maintained. As noted in 10 CFR 50, Appendix J, the purpose of the testing is to ensure that leakage of-containment Fa.tions whose design incorporates resilient seals, gaskets, sealant compounds, and electrical penetrations fitted with seal assemblies remains below established limits. Damage to seals or gaskets, which could affect contamment integrity, is best detected with this type of test and will be performed as follows:

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Electrical Penetrations And Containment Penetrations Whose Design Incorporates Resilient Seals, Gaskets, Or Sealant Compounds Those penetrations that are not disassembled during the 10-year interval will receive an Appendix J, Optma B test at least once in the 10-year interval. For those penetrations that are disassembled or opened, an Appendix J test is required upon final assembly prior to start-up. Additionally, if a seal including O-rings or gasket is replaced, it will be visually inspected by maintenance personnel before re-assembly or closure. 'Ihese tests and inspections will assure the leak tightness of primary containment and provide an ec~atable level of quality and safety.

Airlocks and the Containment Equipment Hatch The personnel airlocks are opened as needed during maintenance outages and refueling outages.

Prior to final closure, the accessible portions of gaskets and the door sealing faces are inspected for damage that could affect the leak tightness of the seal. Ifgasket replacement is nectwsan/, the new gasket will be visually inspected by maintenance personnel before re-assembly or closure.

Door seals will be tested in accordance with Appendix J within seven days of opening and once every 30 days during periods of frequent opening.

The containment equipment hatch is normally removed during refueling outages. Ifgasket replacement is necessary, the new gasket will be visually inspected by maintenance personnel before re-asrmbly or closure. Prior to establishing containment integrity following the refueling outage, the watainment equipment hatch is leak rate tested in accordance with Appendix J.

'Ihese tests and inspections will assure the leak tightness of primary contamment and provide an acceptable level of quality and safety.

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Alter==*a Examiu*ian: 'Ihe leak-tightness of the seals (including O-rings) and gaskets will be confirmed in accordance with 10 CFR 50, Appendix J as described above. If a seal (including 0-rings) or ga- ~ - t is replaced, it will be visually inspected by maintenance personnel before re-assembly or closure. Also, an as-left Appendix J leakage test will be performed after installation I

to ensure leak-tightness VI.

Ju=*ine=* ion for Grantina Relief: 'Ihe functionality of the containment penetration seals and gaskets (including those of electrical penetratens) will contmue to be verified during the Type B testing as required by 10 CFR 50, Appendix J. 'Ihe alternative examinations are adequate to ensure the integrity of the Farley containment penetration seals and gaskets, and will provide an acceptable level of quality and safcty. 'Iherefore, relief should be granted per 10 CFR 50.55a(a)(3)(i).

VII.

Imnlementation Schedule: 'Ihis relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

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l ENCLOSUREII Farley Nuclear Plant-Unit 1 Third Ten Year Interval Reauest for Relief No. RR-32 l

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SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1

'nilRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-32 i

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System /Comoo.ngnt(s) for Which Reliefis Reauested: Class MC (Metallic Coatainment) pressure-retaining bolting, Examination Category E-G, Items E8.10 and E8.20 associated with bolted flanges on two electrical penetrations built by Westinghouse. Specifically these

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penetrations are identified as:

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J QIT52-B009-A and QIT52-B011-B SNC has determined that these are the only bolted connections which are under tension due to containment pressure and which are not routinely disassembled during refueling outages.

II.

Code Reauirement: 10 CFR 50.55a was amended in the Federal Register on August 8,19%, to require the use of ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. The 1992 Edition with 1992 Addenda of ASME Section XI, Table IWE 2500-1, Examination Category E-G, Item Number E8.10 requires a VT-1 examination of the bolted connections and Item Number E8.20 requires a bolt torque or tension test for bolted connections i

that have not been disassembled, inspected, and reassembled during the inspe: tion interval.

III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from performing the Code-required visual examination and the torque or tension test on the above identified pressure l

retaiaing bolting.

IV.

Basis for Relief: Each of these electrical penetrations receive a periodic 10 CFR 50, Appendix J, Option B test at least once per interval. 'Ihe performance of the Type B test proves that the bolt torque or tension remains adequate under simulated accident pressure conditions to restrict leakage to acceptable limits.

Once a bolt in a containment penetration is torqued or tensioned, it should not be subject to l

dynamic loading that could cause it to experience significant change. 'Ihe Appendix J testing is j

l adequate to demonstrate that the design function is met.

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Torque or tension testing is not required for ASME Section XI, Class 1,2, or 3 bolted connections or their supports as part of the inservice inspection program. The ASME Code Committee recognized that these tests were not warranted, and the 1998 Edition of the ASME Section XI Code has removed the examination requirement.

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'Ihe alternate examination will ensure the bolt torque or tension remains adequate. 'Ihis will l

ensure the structural integrity and leak-tightness of this pressure retaining bolting.

V.

Alternate Examination' 'Ibc electrical penetrations shall receive an Appendix J test at least once l

every interval to ensure the bolt torque or tension remains adequate. This will ensure the stmetural integrity and leak-tightness of Class MC (Metallic Containment) pressure retaining bolting.

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VI.

Justification for Grantina Relief: Leak testing per 10 CFR 50, Appendix J will provide assurance of the integrity of pressure-retaining bolting and is an acceptable alternative to the 1992 Code required visual and bolt torque or tension test. Public health and safety will not be endangered; therefore, this relief request should be granted pursuant to the requirements of 10 CFR l

50.55a(aX3Xi).

VII.

Imnlementation Schedule: This relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI duri.'s the current inspection interval.

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VIII.

Eglief Reauest Status: Awaiting NRC approval.

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~s ENCLOSUREIII 1

l Farley Nuclear Plant - Unit 1 Hird Ten Year Interval Reauest For Relief No. RR-33 1

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l SCUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-33 I.

System /Comoonentds) for Which Reliefis Reauested: All Class MC (Metallic Containment),

Paragraphs IWE-2420(b) and IWE-2420(c) successive examination requirements for components found acceptable for continued service.

11.

Code Reauirement: 10 CFR Part 50.55a was amended in the Federal Register on August 8,1996, to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. 'Ihe 1992 Edition with 1992 Addenda of ASME Section XI, requires i

that when component examination results require evaluation of flaws, evaluation of areas of degradation, or repairs in accordance with Article IWE-3000, and the component is found to be acceptable for continued service, the areas containing such flaws, degradation, or repairs shall be i

scexamined dering the next inspection period.

III.

Code Reauirement for Which Reliefis Raaua=taA: Reliefis requested from the requirement of Paragraphs IWE-2420(b) and IWE-2420(c) to perform successive examination of components that have been repaired.

IV.

Basis for Relief: 'Ibe purpose of a repair is to restore the component to an acceptable condition for continued service in accordance with the acceptance standards of Article IWE-3000. When making repairs, paragraph IWA-4150 requires the owner to conduct an evaluation of the suitability of the repair including consideration of the cause of failure. Successive examinations after repair do not provide an additional safety benefit.

Repairs are performed in accordance with IWA-4000, the intent of which is to use the construction code to restore the component to its original condition where practical. If a repair has restored the component to an acceptable condition, soccessive examinations are not.

warranted. If the repair was not suitable, then the repair does not meet Code requirements and the component is not acceptable fcr continued service; further repair work would be necessary. No similar requirement is found for ASME Class 1,2, or 3 Serion XI repairs. Conducting successive examinations on components that have been repaired would result in hardship without a compensating increase in the level of quality and safety. Additionally, if the repair area is subject to accelerated degradation, the repair would require augmented examination in j

accordance with Table IWE-2500-1, Examination Category E-C.

V.

Alternate Examination Repairs will be performed in accordance with IWA-4000 to restore the c=p=aa' to its original condition and successive examinations as required by IWE-2410(b) and

- (c) will not be perfur...cd. Successive examinations will continue to be done on those flaws or assas of degradation which have been accepted for continued service by evaluation.

VI.

Justification for Grantina Relief: Repairing components to restore the component to its original condition provides adequate assurance of the integrity of the repair. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increate in the level of quality and safety; therefore, relief should be granted under 10 CFR 50.55a(a)(3)(ii).

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VII.

Imolementation Schedule: This relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI daring the current inspection interval.

VIII.

Relief Reauest Status: Awaiting FiRC approval.

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ENCLOSURE IV Farley Nuclear Plant-Unit 1 Third Ten Year Interval Reauest For Relief No. RR-34

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FNP-1-M-096 I

I SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT I

- THIRD 10-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-36 I.

System /Comoonent(s) for Which Reliefis Reauested: All ASME Class 1,2,3, and contamment pressure boundary piping and components (Categories B-P, C-H, D-A, D-B, D-C, and E-P).

l II.

Code Reauirement: ASME Code,Section XI, IWA-2313 requires personnel performing l

examinations to be qualified by examination and certified in accordance with SNT-TC-1A. Level I and II personnel shall be recentified by qualification examinations every 3 years. Ievel III personnel shall be recertified by qualification examinations every 5 years.

III.

Code Recuirement for Which Reliefis Requested: Reliefis requested from qualification by examination and certification by examination in accordance with SNT-TC-1A for personnel performing leakage examinations (VT-2) of piping and components.

i IV.

Basis for Relief: The ASME Section XI Code Committee recognized that personnel that are performing examinations for evidence ofleakage (VT-2) should not be required to satisfy the same stringent requirements for qualification and certification as personnel performing other

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types of visual examinations. Personnel performing leakage examinations should be familiar

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with the plant's specific configuration, systems, and procedures for VT-2 visual examination, and j

the Owner should be able to develop an acceptable program for training personnel to perform i

VT-2 leakage examinations.

V.

Alternate Examination: Plant Farley will implement a training program that satisfies the l

requirements of ASME XI Code Case N-546 for personnel to perform VT-2 leakage examinations. Personnel that are qualified and certified in accordance with ASME XI IWA-2300 requirements may also be utilized to perform VT-2 leakage examinations; however, personnel that meet the requirements of the Owner's training requirements in accordance with Code Case N-546 will also be considered qualified to perform VT-2 examinations.

I VI.

Lustification for Grantina Relief: Code Case N-546 was published in Supplement 2 of the ASME XI Code,1995 Edition. This Code Case provides alternative requirements to those ofIWA-2300 for the qualification of VT-2 examination personnel. 'Ihe ASME XI Code Committee determined that such training in accordance with this Code Case would ensure that an adequate level of quainy and safety was being maintained. Therefore, the proposed alternative is justified per 10 CFR 50.55a(aX3)(i). Code Case N-546 has not yet been endorsed by the NRC in Regulatory Guide 1.147; therefore, SNC is requesting to apply the Code Case via this relief request.

VII.

Imal-wion Schedule: This request for reliefis applicable to exanunations per ormed using the 1989 r

Edition (1992 Edition with 1992 Addenda for Containment) of Section XI durin',the current inspection interval.

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Relief Reauest Status: Awaiting NRC approval.

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SNC has concluded that, similar to the consideration used for the IWE examinations, the use of the VT-3 requirements found in IWA-2210 and Table IWA-2210-1 when performing VT-3C examinations of the concrete surfaces is also excessively stringent and should not be applied.

'Ihis is based on the recognition that due to the, nature of concrete, a concrete containment will have numerous, small " shrinkage-type" surface cracks or other imperfections that are not detrimental to the structural integrity of the containment. The application ofIWA-2210 and Table IWA-2210-1 " minimum illumination requirements,"" maximum direct visual examination distance requirements," and "maxunum procedure demonstration lower ca:e character height requirements" to attempt to identify these small "shrmkage-type cracks" or other imperfections is considered to be ua-y and could result in a large number of man-hours erecting scaffolding, using lifts, evaluating insignificant indications, etc.

Per the requirements ofIWL-2320, the Registered Professional Engineer (RPE) is experienced in evaluating the inservice condition of structural concrete and is knowledgeable of the design and Construction Codes and other criteria used in design and construction of concrete containments.

The RPE will use experience and training to determine the necessary requirements to detect indications that are detrimental to the containment integrity. Using knowledge of the degradation processes that could Wentially be occurring and knowledge of high stress and critical areas of I

the contamment structure, the RPE performed a detailed inspection / assessment of essentially all areas of the Farley Unit I contamment surface, to determine the need for auxiliary lighting, scaffolding, binoculars, etc. This inspection / assessment has been documented and forms the bases of the demonstration that the Farley Nuclear Plant VT-3C examinations will meet the intent of the required IWL examinations. The findings of the inspection / assessment for separate portions of he containment surface (e.g., individual auxiliary building iooms that adjoin containment and outside " daylight" surfaces) will establish the requirements for additional lighting, scaffolding, and any necessary viewing aids for those areas.

V.

Alternate Examination Vr-3C examinations will be performed as r: quired by IWL-2310 except i

that instead of using the minimum illumim. tion, maximum direct examination di :ance, and maximum procedure demonstration lower case character 'neight requirements specified in IWA-2210 and Table IWA-2210-1 for VT-3 examinations, the recommendations of the RPE for illumination and distance will be implemented.

VI.

Justification for Grantina Relief: Section XI relies on the knowledge and experience of the RPE as a key element for an IWL visual inspection program. Examining the concrete surfaces using distances and illumination requirements, established by a knowledgeable RPE, would provide for j

detection of flaws of sufficient size to assure that the structural integrity of the concrete.

containment is being maintained. 'Iherefore, an acceptable level of quality and safety will be maintained and relief should be granted per 10 CFR 50.55a(a)(3)(i).

i VII.

Imnlementation Schedule: 'Ihis relief reques is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

VIII.

Relief Request Status: Awaiting NRC approval.

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i ENCLOSURE V Farley Nuclear Plant-Unit 1 Third Ten Yer.r Interval Reauest for Relief No. RR-35 1

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FNP-1-M 096 SOUDIERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-35 1.

System / Components (s) for Which Reliefis Reauested: 'Ihis relief request applies to IWE and IWL examinations performed by the Registered Professional Engineer (RPE).

II.

Code Reauirement: 10 CFR 50.55a was amended to require the use of the ASME Section XI, 1992 Edition,1992 Addenda, when performing containment examinations. Per the 1992 Edition of ASME Section XI with the 1992 Addenda, the visual examinations (VT-1, VT-1C, VT-3, and VT-3C) am performed using certified personnel per IWA-2310.

III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from the IWA-2310 requirement to certify the RPE performing VT-1, VT-lC, VT-3, ard VT-3C examinations related to IWE and IWL.

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Basis for Relief: SNC evalus'ed how best to perform the VT-1, VT-1C, VT-3, and VT-3C examinations at FNP and detamined that the most efficient method is, as a general rule, to have the RPE perform the actual examinations. Therefore, for many cases, the RPE would perform the following:

1.

VT-3 examination of Category E-A.

2.

VT-1 examination of Category E-C (if required).

3.

VT-3C examination of Category L-A.

4.

VT-1 and VT-IC of Category L-B.

Per the requirements ofIWL-2320, the RPE shall be experienced in evaluating the inservice condition of structural concrete and is knowledgeable of the design and Construction Codes and other criteria used in design and construction of concrete contamments. 'Ihe RPE shall also be responsible for development of plans and procedures for examination of concrete surfaces; approval, instruction, and training of concrete examination personnel; evaluation of examination results for concrete; preparation of repair procedures; and submittal of a report documenting examination and repairs. Additionally, per the requirements ofIWE-3510, the RPE is requited to be knowledgeable in the requirements for design, inservice inspection, and testing of metallic liners.

The purpose of a non-destructive examination (NDE) certification program is to assure that NDE personnel are qualified to perform specific examinations. 'Ibe certification process involves traming, testing, and written testunony = wing to the qualifications of the personnel. 'Ihe use of such a pogram is an appropriate approech for NDE examiners that are performing containment examinations under the direction of the RPE.

However, when the RPE is performing the visual examinations, the RPE has already been deemed qualified by designation as the Registered Professional Engineer, therefore, the application of the certification process to the RPE is redundant and provides little, if any, additional benefit.

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V.

Alternate Exammation When VT-1, VT-lC, VT-3, or the VT-3C examinations are performed by the RPE, the "non-certified" RPE shall perform the examinations based on knowledge and experience of contamment design and degradation mechanisms. Additionally, the "non-certified" RPE will be responsible for determining the qualifications (testing requirements, test questions, experience, etc.) of NDE personnel who may also perform containment examinations VI.

Judineation for Grantina Relief: Section XI relies on the knowledge and experience of the RPE as a key element for the IWE and IWL visual inspection programs. A knowledgeable RPE can perform the requhed exammations more effectively. 'Iberefore, an acceptable level of quality and safety will be maintained and relief should be granted per 10 CFR 50.55a(aX3Xi).

VII.

Imolementation Schedule: 'Ihis relief request is applicable to examinations performed using the i

1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

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Vill.

Relief Reauest Status: Awaiting NRC approval.

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1 ENCLOSURE VI Farley Nuclear Plant-Unit i Third Ten Year Interval Reauest for Relief No. RR-36

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SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-36 L

System / Component (s) for Which Reliefis RaquataA: All ASME Class 1,2,3, and containment pressure boundary piping and components (Categories B-P, C-H, D-A, D-B, D-C, and E-P).

II.

Code Requirement: ASME Code,Section XI, IWA-2313 requires personnel performing examinations to be qualified by examination and certified in accordance with SNT-TC-1 A. Level I and II perecunel shall be recertified by qualification examinations every 3 years. Level III personnel sha.1 be recertified by qualification examinations every 5 years.

III.

Code Requirement for Which Reliefis Reauested: Reliefis requested from qualification by examination and certification by examination in accordance with SNT-TC-1 A for personnel performing leakage examinations (VT-2) of piping and components.

IV.

Basis for Rel!:f: "Ihe ASME Section XI Code Committee recognized that personnel that are performing examinations for evidence ofleakage (VT-2) should not be required to satisfy the same stringent requirements for qualification and certification as persanel performing other types of visual examinations. Personnel performing leakage examinations should be familiar with the plant's specific configuration, systems, and procedures for VT-2 visual examination, and the Owner should be able to deve!op an acceptable program for training personnel to perform VT-2 leakage examinations, j

V.

Alternate Examination: Plant Farley will implement a training program that satisf:es the requirements of ASME XI Code Case N-546 for personnel to perform Vf-2 leakage examinations. Personnel that are qualified and certified in accordance with ASME XI IWA-2300 requirements may also be utilized to perform VT-2 leakage examinations; however, personnel that meet the requirements of the Owner's training requirements in accordance with Code Case N-546 will also be considered qualified to perform VT-2 avaminations.

VI.

Justification for Grantina Relief: Code Case N-546 was published in Supplement 2 of the ASME XI Code,1995 Edition. 'Ihis Code Case provides alternative requirements to those ofIWA-2300 for the qualification of VT-2 examination personnel. 'Ihe ASME XI Code Coramittee determined that such training in accordance with this Code Case would ensure that an adequate level of quality and safety was being maintained 'Ihereforr., the proposed alternative is justified per 10 CFR 50.55a(a)(3)(i). Code Case N-546 has not yet been endorsed by the NRC in Regulatory Guide 1.147; therefore, SNC is requesting to apply the Code Case via this relief request.

VII.

Implementation Schedule This request for reliefis applicable to examinations performed using the 1989 Edition (1992 Edition with 1992 Addenda for Containnent) of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

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3 ENCLOSURE VII Farley Nuclear Plant-Unit 1 11 ird Ten Year Interval Request For Relief No. RR-37 i

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SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-37 l

I.

Systein/Comnonent(s) for Which Reliefis Reauested: De pressurizer-to-skirt weld.

II.

Code Reauirem-nt: Item No. B8.20, Category B-H, Table IWB-2500-1 of ASME Section XI, requires a surface examination of the pressurizer-to-skirt weld from the inside and outside diameter of the skirt. De required examination area for the Farley Unit-1 pressurizer skirt weld is shown in ASME Section XI Figure IWB-2500-13. His category does not require examination during the third interval per the 1989 ASME Code. However, since SNC is using Code Case N-509, this item is included as past of the examination scope.

III.

Code Reauirement for Which Relielis Reauested: Reliefis requested from performing the surface examination from the inside A=e of the pressurizer skirt.

IV.

Basis for Relief: he pressurizer heater penetrations restrict personnel access to " Area C-D" of l

the pressurizer support skirt shown in ASME Section XI Figure IWB-2500-13.

j V.

Alternate Exammation: In addition to the surface examination of the outside diameter, a "best l

effort" ultrasonic examination will be performed from the outside diameter for a limited portion of the " Area C-D."

VI.

hetifie=*iaa for Gr=atia= Relief: He heater penetrations of the bottom head restrict personnel access to the inside of the pressurizer support skirt. To obtain access to the bottom of the pressurizer would require a modified design and would be very expensive. He alternate exammation proposed in Section V will provide reasonable assurance of the continued structural integrity of this weld. Denial of this relief request would cause an excessive burden upon SNC, as modification of the pressurizer to perform tids Code required examination is impractical; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).

VII.

Imolementation Schedule: Dis relief request is applicable to examinations performed using the 1989 Edition of ASME Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

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ENCLOSURE VIII Farley Nuclear Plant-Unit 1 Third Ten Year Interval Reauest for Relief No. RR-38 l

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SOUIRERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-38 l

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System /Comnonent(s) for Which Reliefis Reauested: Regenerative heat exchanger welds and l

component supports.

II.

Code Reanirement: W 1989 Edition of the Section XI Code, Table IWC-2500-1, Examination Category C-A, Item No. C1.20 requires a volumetric examination of head-to-shell welds and Item No. C1.30 requires a volumetric exammation of tubesheet-to-shell welds. Table IWF-2500-1, Category F-A, requires visual, VT-3 examination of component supports.

III.

Cgdg.Reauirement for Which Reliefis Reauested: Reliefis requested from performing the examination of the regenerative heat exchanger welds and component supports.

IV.

Basis for Relief: h regenerative heat exchanger is a Class 2 heat exchanger that is designed to reduce unnecessary heat losses by heating the reactor coolant system (RCS) charging flow with the letdown flow. The 3" charging inlet / outlet lines are connected to the heat exchanger on the l

tube side, and the 3" letdown inlet / outlet lines are enanaded on the shell side. All of the 3" lines are exempt from non-destructive examinations per IWC-1220(c); however, the heat exchanger requires examination. b examination of the regenerative heat exchanger is considered to constitute an naa-a - y hardship without an associated increase in the level of quality and safety 'Ihis conclusion is based on the following:

1. Previous dose rate surveys and data for the Unit I regenerative heat exchanger examinations indicate a contact dose rate of approximately 2800 mrem /hr with a cumulative whole body dose of approximately 2500 mrem associated with the examination of one weld. The whole body cumulative dose to accomplish the required Code examinations for this heat exchanger will be in excess of 8 Rem. SNC considers this cumulative dose to constitute a hardship with no increase in the level of quality and safety for this system.

2. 'Ihe regenerative heat exchanger shell is fabricated from material which restricts ultrasonic examination to a half-node technique. Using a half-node technique, the geometric configuration of the weld surface limits volumetric examinations to approximately half of the required examination volume. SNC considers this a minimal examination for the amount of corresponding dose.

3. The subject welds and piping supports are located on a component where all of the numerous welds and supports on the connecting lines are exempt from non-destructive examination.

Not performing the examination of these heat exchanger welds and suppoits in a system where almost all of the welds and supports do not require examination should have no effect on the level of quality and safety for this system.

V.

Alternate Evamination: No alternative examinations will be performed.

EVill-1 u

FNP-1-M-096 s

VI.

Justification for Grantina Relief: A cumulative radiation dose in excess of 8 rem for the required Code examinations, where the ultrasonic examination of the welds is limited to approximately one-half of the required volume, is considered a hardship by SNC. The level of quality and safety should not be decreased by deletion of the subject examinations, since it is located in piping exempt from nondestructive examinations. The pressure tests which are performed on this section of piping will provide adequate assurance of the integrity of the component and piping in the flow path; therefore, approval is requested per the requirements of 10 CFR 50.55a(aX3Xii).

VII.

Imolamantation Schedule: This relief request is applicable to the ISI examinations performed using the 1989 Edition of ASME Section XI during the current inspection interval.

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VIII.

Relief Reauest Status: Relief request RR-56, which climinated the volumetric examination of l

one weld and visual examination of two supports for the Second Interval, for Unit I was approved by NRC SER dated November 16,1998. RR-38 is awaiting NRC approval.

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EVIII-2

1 ENCLOSURE IX I

l Farley Nuclear Plant-Unit 1 hird Ten Year Interval for Relief No. RR-39 l

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FNP-1-M 0%

SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 1 THIRD 10-YEARINTERVAL REQUEST FOR RELIEF NO. RR-39 I.

System /Comoonent(s) for Which Reliefis Requested

• ne nozzle inside radius section of the pressurizer nozzles.

II.

Code Reauiremqp_t: Item No. B3.120, Category B-D, Table IWB-2500-1 of ASME Section XI, 1989 Edition, no addenda, requires a volumetric examination of the nozzle inside radius section of all pressurizer nozzles.

III.

Code Requirement for Which Reliefis Reauested Reliefis requested from performing the volumetric examination of the nozzle inside radius section of the pressurizer nozzles.

l IV.

Basis for Relief: He ASME Section XI Code Committee recognized that, based upon inspection data and fracture mechanics evaluations, the pressurizer nozzles are unlikely to crack under any anticipated service conditions. Extremely small probabilities of failure cause the benefit ofin-service inspection to be negligible. Further, these examinations are difUcult to perform because oflimited access, rough metal surfaces, and long metal paths. In most cases radiation dose levels ace high.

V.

Alternate Exammation: No examinations will be performed on these nozzle inside radius sections.

VI.

Justification for Grantina Relief: Code Case N-619 has been approved and published by the ASME, but has not yet been endorsed by the NRC in Regulatory Guide 1.147; therefore, SNC is requesting to apply Code Case N-619 via this relief request. As part of the preparation of the Code Case, it was determined that aAer more than 25 years of plant operation and inspections, the industry has found no cracking incidents or service-induced flaws of any kind in these nozzle inner radius sections. Fracture mechanics evaluations, based on conservative assumptions, demonstrated that these nozzles have a large tolerance for flaws. Probabilistic risk assessment calculations, with and without in-service inspection, gave such small probabilities of failure that i

any gain from inspection is meaningless.

Based on industry inspection results, fracture mechanics evaluations, and probabilistic risk assessment calculations, the ASME Section XI Code Committee determined that structural integrity of these nozzles will not be reduced. SNC has determined that an adequate level of quality and safety can be maintained without inspecting these nozzle inside radius sections.

Therefore, the elimination of the inspections of these nozzle inside radius sections isjustified per 10 CFR 50.55a(aX3Xi).

VII. Imnlamantatian Sch41a: His request for reliefis applicable to examinations performed using the 1989 Edition of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

l ElX - 1 I

L.

ENCLOSURE X Farley Nuclear Plant-Unit 2 Updated Program Request for ReliefNo. RR-31 l

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FNP-2-M-0%

i SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-31 I.

System /Comnonents(s) for Which Reliefis Reauested: Seals (including 0-rings) and gaskets of Class MC (Metallic Containment) pressure retaining components, Examination Category E-D, item Numbcas E5.10 and E5.20.

This request for relief applies to the following components that incorporate seals and gaskets as the containment pressure boundary.

1 Electrical penetrations.

Two personnel eirlock doors with seals, including door operating mechanism penetrations that are p ut of the containment pressure boundary and the contamment equipment hatch.

Containment penetrations whose design incorporates resilient scais, gaskets, or sealant compounds.

II.

Code Rmuirement: 10 CFR 50.55a was amended in the Federal Register on August 8,1996, to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment exvninations. 'Ihc 1992 Edition with 1992 Addenda of ASME Section XI, Table IWE 2500-1, Fy=enia=* ion Category E-D, Item Numbers E5.10 and E5.20, requires seals and gaskets on airlocks, hatches, and other devices that are required to assure containment leak-tight integrity to be visually examined once cach interval.

III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from performing the Code-required VT-3 visual examination on the sbove identified containment seals and gaskets.

IV.

Basi lor Egligt Practical VT-3 visual examination consideratiora of:hese seals ar.d gaskets would require the joints to be disassembled since many of the surfaces of seals and gaskets are normally innere==ihle. M ASME Code Committee recognized that disassembly of thejoints to perform visual examinations was not warranted, and the 1998 Edition of ASME Section XI removed the examination requirement.

j h proposed alternate examination (Appendix J, Option B) provides a periodic, non-intrusive test meth ad which will ensure that the integnty of the seals and gaskets is being maintained. As noted in 10 CFR 50, Appendix J, the purpose of the testing is to ensure that leakage of j

containment penetrations whose design incorporates resilient seals, gaskets, scalant compounds, and electncal penet a' ions Stted with seal assemblics remains below established limits. Damage j

to seals or gaskets, which could r.ffect contamment integrity, is best detected w;th this type of test and will be performed as follows:

EX-1 i

C.-

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i, FNP-2-M-0%

Electrical Penetrations And Containment Penetrations Whose Design Incorporates Resilient Seals, Gaskets, Or Sealant Compounds Those g emnons that are not disassembled during the 10-year interval will receive an Appendix J, Option B test at least once in the 10-year interval. For those penetrations that are disassembled or opened, an Appa-Av J test is required upon final assembly prior to start-up. Additionally, if a seal including O-rings or gasket is replaced, it will be visually inspected by maintenance persocnel before re-assembly or closure. These tests and inspec as will assure the leak tightres of primary contamment and provide an =W=ble level of quality and safety.

Airlocks and the Containment Equipment Hatch

'Ihc personnel airlocks are opened as needed during maintenance outages and refueling ottages.

Prior to fmal closure, the accessible portions of gaskets and the door scaling faces are inspected for damage that could affect the leak tightness of the seal. If gasket replacement is necessary, the new gasket will be visually inspected by maintenance personnel before re-assembly or closure.

Door seals will be tested in accordance with Appendix J within seven days of opening and once every 30 days during periods of frequent opening.

t The containment equipment hatch is normally removed during refueling outages. If gasket replacement is m== y, the new gasket will be visually inspected by maintenance personnel before re-assembly or closure. Prior to establishing containment integrity following :he refueling outage, the containment equipment hatch is leak rate tested in accordance with Appendix J.

l These tests and inspections will assure the leak tightness of primary containment and provide an j

acceptable level of quality and safety.

V.

Alternate Exammation: 1hc ter.k-tightness of the seals (including O-rings) end gaskets will be confirmed in accordance with 10 CFR 50, Appendix J ar. described above. If a seal (including 0-rings) or gaskd is replaced, it will be visually inspected by maintenance personnel beforv re-assembly or closure. Also, an as-left Appendix J leakage test will be performed after installation to ensure leak-tightness VI.

Justification for Grantina Relief: 'ac feianality of the containment penetration seals and gaskets (including those c f electrical penetratices) will continue to be verified di.ritag the Type B testing as required by 10 CFR 50, Appendix J. The alternative examinations are adequate to ensure the integrity of the Farley containment penetration seals and gaskets, and will provide an acceptable level of quality and safety. 'Iherefore, relief should be granted per 10 CFR 50.55a(a)(3)(i).

VII.

Imn!-aatadan Schedule: This relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

Vill.

Rehef Request Status: Awaiting NRC mroval.

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EX-2

ENCLOSURE XI Farley Nuclear Plant ~ Unit 2 Updated Procram Recuest for Relief RR-32 l

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SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-32 I.

System /Comoonent(s) for Which Reliefis Requested: Class MC (Metallic Containment) pressure-retammg bolting, Examination Category E-G, items E8.10 and E8.20 associated with bolted flanges on two electncal penetrations built by Westinghouse. Specifically these gamions are identified as:

QIT52-B009-A and QIT52-B011-B SNC has determined that these are the only bolted connections which are under tension due to containment pressure and which are not routinely disassembled during refueling outages.

II.

Code Reauirtagnt: 10 CFR 50 35a was amended in the Federal Register on August 8,1996, to n

require the use of ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. The 1992 Edition with 1992 Adderi.ia of ASME Section XI, Table IWE 2500-1, Examination Category E-G, Item Number E8.10 requires a VT-1 examination of the bolted connections and Item Number E8.20 requires a bolt torque or tension test for bolted connections that have not been disassembled. inspected, and reassembled during the inspection interval.

III.

Code Requirement for Which Reliefis Reauested: Reliefis requested from performing the Code-required visual examination and the torque or tension test on the above identified pressure retaining bolting.

IV.

Basis for Relief: Each of these electrical penetrations receive a periodic 10 CFR 50, Appendix J, l

Option B test at least once per interval. The performance of the Type B test proves that the bolt l

torque or tension remains adequate under simulated accident pressure conditions to restrict leakage to accept.2ble limits.

Once a bolt in a contamment penetration is torqued or tensioned, it should not be subject to dynamic loading that could cause it to experience significant change. He Appendix J testing is i

adequate to demonstrate that the design function is met.

Torque or tension testing is not required for ASME Section XI, Class 1,2, or 3 bolted connections or their supports as part of the inservice inspection program The ASME Code Committee recognized that these tests were not warranted, and the 1998 Edition of the ASME j

Section XI Code has removed the examinatim requirement.

The alternate examination will ensure the bolt torque or tension remains adequate. This will ensure the structural integrity and leak-tightness of this pressure retaining bolting.

V.

Altemate Examination %c electrical penetrations shall receive an Appendix J test at least once every interval to ensure the bolt torque or tension remams adequate. This will ensure the structural integrity and leak-tightness of Class MC (Metallic Containment) pressure retaining bolting.

EXI-l

FNP-2-M-0%

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VI.

Justification for Grantina Relief: Leak testing per 10 CFR 50, Appendix J will provide assurance of the integrity of prssure-retaining bolting and is an acceptable alternative to the 1992 Code required visual and bolt torque or tension test. Public health and safety will not be endangered; therefore, this relief request should be granted pursuant to the requirements of 10 CFR 50.55a(a)(3)(i).

Vll.

Imolementation Schedule: This relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

Vill.

Relief Reauest Status: Awaiting NRC approval.

1 EXI - 2

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I I

l WCLOSURE XII Farley N.a..e Plant - Unit 2 j

Uodated Program Reauest for Relief RR-33 1

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FNP-2-M-096 SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-33 I.

System / Components (s) for Which Reliefis Reauested: All Class MC (Metallic Containment),

Paragraphs IWE-2420(b) and IWE-2420(c) successive examination requirements for components found acceprable for continued service.

II.

Code Reauirement: 10 CFR Part 50.55a was arnended in the Federal Register on August 8,1996 to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations, & 1992 Edition with 1992 Addenda of ASME Section XI, requires that when component examination results require evaluation of flaws, evaluation of areas of degradation, or repairs in accordance with Article IWE-3000, and the component is found to be acceptable for continued service, the areas containing such flaws, degradation, or repairs shall be reexamined during the next inspection period.

III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from the requirement of Paragraphs IWE-2420(b) and IWE-2420(c) to perform successive examination of components that have been repaired.

IV, Basis for Relief M purpose of a repair is to restore the component to an acceptable condition for continued service in accordance with the acceptance standards of Article IWh.3000. When making repairs, paragraph IWA-4150 requires the owner to conduct an evaluatios of the suitability of the repair including consideration of the cause of failure. Successive examinations after repair do not provide an additional safety benefit.

Repatrs are performed in accordance with IWA-4000, the intent of which is to ure se construction code to restore the co.nponent to its original condition where practical. If a repair has restored the component to an mapable condition, successive examinations are not

{

warranted. If the repair was not suitable, then the repair does not meet Code requirements and the component is not acceptable for continued service; further repair work would be necessary. No similar requirement is found for ASME Class 1,2, or 3 Section XI repairs. Conducting successive examinations on components that have been repaired would result in hardship without a cow.g.seting increase in the level of quality and safety. Additionally, if the repair area is j

subject to accelerated degradation, the repair would require augmented examination m accordance with Table IWE-2500-1, Examination Category E-C.

j V.

Alternate Exammation Repairs will be performed in accordance with IWA-4000 to restore the component to its original condition and successive exandnations as required by IWE-2420(b) and (c) will not be pufurad Successive examinations will continue to be done on those flaws or areas of degradation which have been repH for continued service by evaluation.

VL Justification for Grantina Relief: Repairing components to restore the component to its original condition provides adequate assurance of the integrity of the repair. Con.pliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; therefore, relief should be granted under 10 CFR 50.55a(a)(3)(ii).

i EXII-1

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l FNP-2-M-096 o

VII.

Imolementation Schedule: This relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

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EXII-2 1

ENCLOSURE XIII Farley Nuclear Plant-Unit 2 Uodated Procram Reauest for Re'ief RR-34 1

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SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-34 I.

System /Comnonents(s) for Which Reliefis Reauested: This relief request applies to the exterior surface of the concree Containment Building.

II.

Code Reauirement: 10 CFR 50.55a was amended in the Federal Register on August 8,1996, to require the use of the ASME Section XI,1992 Edition,1992 Addenda, when performing containment examinations. Per the 1992 Edition of ASME Section XI with the 1992 Addenda, the visual examination (VT-3C) of the concrete portion of the containment buildings is subject to the rules and requirements ofIWL-2310, " Visual Examination and Pc:wnnel Qualification."

IWL-2310 subsequently requires that the minimum illumination, maximum direct examination distance, and maximum procedu~ demonstration lower case character height shall be as specified in IWA-2210 and Table IWA-2210-1 for VT-3 examinationis.

111.

Code Reauisement for Which Reliefis Reauested: Reliefis requested from the IWL-2310 requirement to use the minimum illumination, maximum direct examination distance, and j

maximum procedure demonstration lower case character height specified in IWA-2210 and Table IWA-2210-1 for VT-3 examinations when performing visual examinations (VT-3C) of the concrete containment.

IV.

Basis for Relief The VT-3 requirements specified in IWA-2210 and Table IWA-2210-1 were developed for the examination of components such as Class I pump and valve bodies, the Class I reactor pressure vessel interior, Class 3 welded a*=chments, and Class 1,2, and 3 supports. VT-3 examinations are conducted to determine the general mech==ical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements. Additionally, VT-3 examinations are conducted to detect discontinuitics and i

imperfections, such as loss ofintegrity at bolted or welded conr.ections, loose or missing parts, debris, corrosion, wear, or erosion. For these Class 1,2 and 3 components, small amounts of corrosion /crosion or small crack-like s.nface flaws may be detrimental to the structural integrity of the component; therefore, the stringent requirements ofIWA-2210 and Table IWA-2210-1 are generally appropriate However, it was recognized by the industry and the NRC during the development of the implementing 10 CFR 50.55a rules that IWA-2210 and Table IWA-2210-1 requirements were i

excessively stringent for the IWE required examination of the metal portion of the containment.

Therefore, the NRC changed the requirements to allow the following: "When performing i

remotely the visual examinations required by Subsection IWE, the maximum direct distance 4

specified in Table IWA-2210-1 may be extended and the minimuir illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indicat:ons for which the visual examination is performed can be detected at the chosen distance and

.tlumination."

EXIII-1

r FNF 2 M-0%

SNC has concluded that, similar to the consideration used for the IWE examinations, the use of the VT-3 requirements found in IWA-2210 and Table IWA-2210-1 when performing VT-3C examinations of the concrete surfaces is also excessively stringent and should not be applied.

This is based on the recognition that due to the nature of concrete, a concrete containment will have numerous, small " shrinkage-type" surface cracks or other imperfections that are not detrimental to the structuralintegrity of the containment. The application ofIWA-2210 and Table IWA-2210-1 " minimum illumination requirements,"" maximum direct visual examination distance requirements," and " maximum procedure demonstration lower case character height requirements" to attempt to identify these small " shrinkage-type cracks" or other imperfections is considered to be unnocessary and could result in a large number of man hours crecting scaffolding, using liAs, evaluating insignificant indications, etc.

Per the requirements ofIWL-2320, the Registered Professional Engineer (RPE) is experienced in evaluating the inservice condition c (structural concrete and is knowledgeable of the design and Construction Codes and other crite ia used in design and construction of concrete containments.

'Ihe RPE will use experience and i aining to determine the necessary requirements to detect indications that are detrimental to the containment integrity. Usmg knowledge of the degradation processes that could potentially be occurring and knowledge of high stress and critical areas of the containment structure, the RPE performed a detailed inspection / assessment of essentially all areas of the Farley Unit I contamment surface, to determine the need for auxiliary lighting, scaffolding, binoculars, etc. This inspection / assessment has been documented and forms the bases of the demonstration that the Farley Nuclear Plant VT-3C examinations will meet the intent of the required IWL examinations. 'Ihe findings of the inspection / assessment for separate portions of the contamment surface (e.g., individual auxiliary building rooms that adjoin contrinment and outside " daylight" surfaces) will establish the requirements for additional lighting, scaffolding, and any necessary viewing aids for those areas.

V.

Alternate Examination: VF-3C examinations will be performed as required by IWL-2310 except that instead of using the minimum illumination, maximum direct examination distance, and maximum procedure demonstration lower case character height requirements specified in IWA-2210 and Table IWA 2210-1 for VT-3 examinations, the recommendations of the RPE for illumination and distance will be implemented.

VI.

Justification for Grantina Relief: Section XI relies on the knowledge and experience of the RPE as a key element for an IWL visual inspection program. Examining the concrete surfaces using distances and illumination requirements, established by a knowledgeable RPE, would provide for detection of flaws of sufficient size to assure that the structural integrity of the concrete containment is being maintained. 'Iherefore, an acceptable level of quality and safety will be maintained and relief should be granted per 10 CFR 50.55a(a)(3)(i).

VII.

Imolementation Schedule: 'Ihis relief request is applicable to examinations performed using the IW2 Edition,1992 Addenda, of Section XI during the current inspection interval.

VllI.

Bahtf Reauest Status: Awaiting NRC approval.

EXIII-2

9 ENCLOSURE XIV Farley Nuclear Plant-Unit 2 IJpdated Program Reauest for Relief RR-35 l

FNP-2-M-096 SOUIRERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-35 I.

System /Comnonents(s) for Which Reliefis Reauested: his relief request applies to IWE and IWL examinations puforw.cd by the Registered Professional Engineer (RPU).

II.

Code Reauirement: 10 CFR 50.55a was araended to require the use of the ASME Section XI, 1992 Edition,1992 Addenda, when performing containment examinations. Per the 1992 Edition of ASME Section XI with the 1992 Addenda, the visual examinations (VT-1, VT-lC, VT-3, and VT-3C) are performed using certified personnel per IWA-2310.

III.

Code Reauirement for Which Reliefis Ree=ted: Reliefis requested from the IWA-2310 requirement to certify the RPE performing VT-1, VT-IC, VT-3, and VT-3C examinations related to IWE and IWL.

IV.

Basis for Relief: SNC evaluated how best to perform the VT-1, VT-IC, VT-3, and VT-3C examinations at FNP and determined that the meist eflicient method is, as a general rule, to have the RPE perform the actual examinations. %crefore, for many cases, the RPE would perform the following:

1.

VT-3 examination of Category E-A.

2.

VT-1 examination of Category E-C (if required).

3.

VT-3C examinatica of Category L-A.

4.

VT-1 and VT-lC of Category L-B.

Per the requirements ofIWL-2320, the RPE shall be experienced in evaluatmg the inservice condition of structural concrete and is knowledgeable of the design and Construction Codes and other criteria used in design and construction of concrete containments. The RPE shall also be responsible for development of plans and procedures for examination of concrete surfaces; approval, mstruction, and training of concrete examination personnel; evaluation of examination results for concrete; preparation of repair procedures; and submittal of a report documenting examination and repairs. Additionally, per the requirements of'WE-3510, the RPE is required to be knowledgeable in the requirements for design, inservice inspection, and testing of metallic liners.

The purpose of a non-destructive examination (NDE) certification program is to assure that NDE personnel are qualified to perform specific examinations. The certification process involves training, testing, and written testimony attesting to the qualifications of the personnel. The use of such a program is an appropriate approach for NDE examiners that are performing containment examinations ender the direction of the RPE.

However, when the RPE is performing the visual exanunations, the RPE has already been deemed qualified by designation as the Registered Professional Engineer, therefore, the application of the certification process to the RPE is raah and provides little, if any, additional benefit.

EXIV-1

FNP-2-M-096 V.

Alternate Examination: When VT-1, VT-lC, VT-3, or the VT-3C examinations are performed by the RPE, the "non-certified" RPE shall perform the examinations based on knowledge and experience of containment design and degradation mechanisms. Additionally, the "non-certified" RPE will be responsible for determining the qualifications (testing requirements, test questions, experience, etc.) of NDE personnel who may also perfonn containment examinations.

. VI.

Justification for Grantinn Relief: Section XI relies on the knowledge and experience of the RPE as a key element for the IWE and IWL visual inspection programs. A knowledgeable RPE can perform the required examinations more effectively. 'Iherefore, an acceptable level of quality and safety will be maintained and relief should be granted per 10 CFR 50.55a(a)(3)(i).

VII.

Imolementation Schedule: ' Ibis relief request is applicable to examinations performed using the 1992 Edition,1992 Addenda, of Section XI during the current inspection interval.

VIII.

Eslief Reauest Status: Awaiting NRC approval.

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i EXIV - 2

ENCLOSURE XV Farley Nuciccr Plant-Unit 2 Uodated Program Reauest for Relief RR-36 l

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d SOUIRERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-36 I.

System / Component (s) for Which Relief!s Reg ~*eA: All ASME Class 1,2,3, and containment pressure boundary piping and components (Categories B-P, C-H, D-A, D-B, D-C, and E-P).

i II.

Code Reauir.nent: ASME Code,Section XI, IWA-2313 requires personnel performmg examinations to be qualified by examination and cestificd in accordance with SNT-TC-1 A. level I and II personnel shall be recertified by qual:fication examinations every 3 years. Level III personnel shall be recertified by qualification examinations every 5 years.

III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from qualificatica by examination and certification by exammstion in accordance with SNT-TC-1 A for personnel performing leakage examinations (VT-2) of piping and components.

IV.

Basi for Relief: h ASME Sxtion XI Code Committee recognized that personnel that are performing examinations for evidence ofleakage (VT-2) sould not be rquired to satisfy the same stringent requirements for qualification and certification as personnel performing other types of visual examinations Personnel performing leakage examinations should be familiar with the plant's specific configuration, systems, and procedures for VT-2 visual examination, and the Owner should be able to develop an acceptable program for training personnel to perform i

VT-2 leakage examinations.

V.

Alternate Examination: Plant Farley will implement a training program that satisfies the requirements of ASME XI Code Case N-545 for personnel to perform VT-2 leakage i

examinations. Personnel that are qualified and certified in accordance with ASME XI IWA-2300 i

requirements may alw be utilized to perform VT-2 leakage examinations; however, personnel that meet the requirements of the Owner's training requirerr.ents in accordance with Code Case j

N-546 will also be considered qualified to perform VF-2 examinations.

VI.

Justification for Grantina Relief: Code Case N-546 was published in Supplement 2 of the ASME XI Code,1995 Edition. 'Ihis Code Case provides alternative requirements to those ofIW A-2300 for the qualification of VT-2 examination personnel. The ASME XI Code Committee reermined that such training in accordance with this Code Case would ensure that an adequate level of quality and safety was being inaintained. hrefore, the proposed alternative isjustified per 10 j

CFR 50.55a(a)(3)(i). Code Case N-546 has not yet been endorsed by the NRC in Regulatory Guide 1.147; therefore, SNC is requesting to apply the Code Case via this relief request.

VII.

Implementation Schedule This request for reliefis apphcable to exammations performed using the 1989 Edition (1992 Edition with 1992 Addenda for Containment) of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval, i

EXV-1 e

ENCLOSURE XVI Farley Nuclear Plant-Unit 2 Updated Prozram Reauest for Relief RR-37,

FNP-2-M-096 l,

i SOUTHFRN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-37 1.

System /Como9 Dent (s) for Which Reliefis Reauested: The pressurizer-to-skirt weld.

II.

Code Reauirement: Item No. B8.20, Category B-H, Table IWB-2500-1 of ASME Section XI, requires a surface examination of the pressurizer-to-skirt weld from the inside and outside diameter of the skirt. De required examination area for the Farley Unit-2 pressurizer skirt weld is shown in ASME Section XI Figure IWB-2500-13. His category does not requir: examination per the 1989 ASME Code. However, since SNC is using Code Case N-509, this item is included as pan of the examination scope III.

Code Reauirement for Which Reliefis Reauested: Reliefis requested from performing the l

surface examination from the inside diameter of the pressurizer skirt.

IV.

Basis for Relief: The pressurizer heater penetrations restrict personnel access to " Area C-D" of the pressurizer support skirt shown in ASME Section XI rigure IWB-250013.

i V.

Alternate Examination: In addition to the surface examination of the outside diameter, a "best effort" ultrasonic examination will be performed from the outside diameter for a limited portion of the " Area C-D."

VI.

Justification for Grantina Relief: De heater penetrations of the bottom head restrict personnel access to the inside of the pressurizer suppoit skirt. To obtain access to the bottom of the pressurizer would require a modified design and would be very expensive. The alternate examination proposed in Section V will provide reasonable assurance of the continued structural integrity of this weld. Denial of this relief request would cause an excessive burden upon SNC, I

as modifict. tion of the pressurizer to perform this Code required examination is impractical-l therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).

VII.

Imolementation Schedule: This relief r; quest is applicable to examinations performed using the 1989 Edition of ASME Section XI during the current inspection interval.

i VIII.

Relief Reauest Status: Awaiting NRC approval.

EXVI-1

ENCLOSURE XVII Farley Nuclear Plant-Unit 2 IJpAted Procram Reauest for Relief RR-38

FNP-2-M-0%

SOUTHERN NUCLEAR OPERATING COMPANY FC 'EY UNIT 2 UPL 1 PROGRAM REQUESTl OR ltELIEF NO. RR-38 I.

System /Comnonent(s) for Which Reliefis Requested: Regenerative heat exchanger welds and

)

component supports.

II.

Code Reauirement The 1989 Edition of the Section XI Code, Table IWC-2500-1, Examination Category C-A, Item No. C1.20 requires a volumetric examination of head-to-shell welds and Item No. C1.30 sequires a volumetric examination of tubesheet-to-shell welds. Table IWF-2500-1, Category F-A, requires visual, VT-3 examination of component supports.

III.

Code Reauirement for Which Reliefis Requested: Reliefis requested from performing the I

examination of the regenerative heat exchanger welds and component supports.

IV.

Basis for RAigf ne regenerative heat exchanger is a Class 2 heat exchanger that is designed to reduce unnecessary heat losses by heating the reactor coolant system (RCS) charging flow with the letdown flow. De 3" charging inlet / outlet lines are connected to the heat exchanger on the tube side, and the 3" letdown inlet / outlet lines are connected on the shell side. All of the 3" lines are exempt from non-destructive examinations per IWC-1220(c); however, the heat exchanger requires examination. Tne examination of the regenerative heat exchanger is considered to constitute an unnecessary Inrdship without an associated increase in the level of quality and safety. This conclusion is based on the following:

)

1. Previous dose rate surveys and data for the Unit I regenerative heat exchanger examinations indicate a contict dose rate of approximately 2800 mrem /hr with a cumulative whole body dose of approximately 2500 mrem associated with the examination of one weld. He whole body cumulative dose to accomplish the required Code examinations for this heat exchanger will be in excess of 8 Rem. SNC considers this cumulative dose to constitute a hardship with no increase in the level of quality and safety for this system.

2. He regenerative heat exchanger shell is fabricated from material which restricts ultrasonic examination to a half-node technique. Using a half-node technique, the geometric configuration of the weld surface limits volumetric examinations to approximately half of the required e.xamination volume. SNC considers this a minimal examination for the amount of corresponding dose.

3. The subject welds and piping supports are located on a component where all of the numerous welds and supports on the connecting lirees are exempt from non-destructive examination.

Not performing the examination of these heat exchanger welds and supports in a system where almost all of the welds and supports do not require examination should have no effect on the level of quality and safety for tids system.

V.

Alternate Examinaugn: No alternative examinations will be performed.

J EXVII-1

1

~

I FNP-2-M-096 VI.

JMifica' ion for Grantina Relief: A cumulative radiation dose in excess of 8 rem for the reauired Code examinations, where the ultrasonic examination of the welds is limited to approumately one-half of the required volume, is considered a hardship by SNC. He level of quality and safety should not be decreased by deletion of the subject examinations, since it is located in piping exempt from nondestructive examinations. The pressure tests which are performed on this i

section of piping will provide adequate assurance of the integrity of the component and piping in the flow path; therefore, approval is requested per the requirements of 10 CFR 50.55a(a)(3)(ii).

VII.

Imolemenigtion Schedule: This relief request is applicable to the ISI examinations performed L

using the 1989 Edition of ASME Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Relief request RR-56, which eliminated the volumetric examination of one weld and visual examination of two supports for the Second Interval, for Unit I was approved by NRC SER dated November 16,1998. RR-35 is awaiting NRC approval.

I i

l l

i EXVII-2

O ENCLOSURE XVIII Farley Nuclear Plant-Unit 2 IJpdated Program Reauest for Relief RR-39

e e.

o' FNP-2-M-0%

r SOUTHERN NUCLEAR OPERATING COMPANY FARLEY UNIT 2 UPDATED PROGRAM REQUEST FOR RELIEF NO. RR-39 I.

System / Component (s) for Which Reliefis Reauested: 'Ihe nozzle inside radius section of the pressurizer noules.

II.

Code Reauirement: Item No. B3.120, Category B-D, Table IWB-2500-1 of ASME Section XI, 1989 Edition, no addenda, requires a volumetric examination of the nozzle inside radius section of all pressurizer nozzles.

III.

Code Reauirement for Which Reliefis Requested: Reliefis requested from performing the volumetric examination of the nozzle inside radius section of the pressurizer nozzles.

IV.

Basis for Relief: 'Ihe ASME Section XI Code Committee recognized that, based upon inspection data and fracture mechanics evaluations, the pressurizer nozzles are unlikely to crack under any anticipated service conditions. Extremely small probabilities of failure cause the benefit ofin-service inspection to be negligible. Further, these examinations are difficult to perform because oflimi.cd access, rough metal surfaces, and long metal paths. In most cases radiation dose levels are high.

V.

Alternate Examination No examinations will be performed on these nozzle inside radius sections.

VI.

Juctification for Grantina Relief: Code Case N-619 has been approved and published by the ASME, but has not yet been endorsed by the NRC in Regulatory Guide 1.147; therefore, SNC is requesting to apply Code Case N-619 via this relief request. As part of the preparation of the Code Case, it was determined that after more than 25 years of plant operation and inspections, tle industry has found no cracking incidents or service-induced flaws of any kind in these nozzle inner radius sections. Fracture =~ h"cs evaluations, based on conservative assumptions, demonstrated that these nozzles have a large tolerance for flaws. Probabilistic risk assessment calculations, with and without in-service inspection, gave such small probt.bilides of failure that any gain from inspection is meaningless.

Based on industry inspection results, fracture mechanics evaluations, and probabilistic risk assessment calculations, the ASME Section XI Code Committee determined that structural integrity of these nozzles will not be reduced. SNC has determined that an adequate level of quality and safety can be maintained without inspecting these nozzle inside radius sections.

'Iherefore, the elimination of the inspections of these nozzle inside radius sections is justified per 10 CFR 50.55a(aX3Xi).-

VII. Imolemanh+ian Schedule: This request for reliefis applicable to examinations performed using the 1989 Edition of Section XI during the current inspection interval.

VIII.

Relief Reauest Status: Awaiting NRC approval.

1 EXVIII-1

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