ML18354A431

Submittal of Relief Requests Associated with the Fifth Lnservice Inspection (ISI) Interval

ML18354A431
Person / Time
Site:	Calvert Cliffs
Issue date:	12/19/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18354A431 (30)

Text

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www.exeloncorp com 10 CFR 50.55a December 19, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

Subject:

Submittal of Relief Requests Associated with the Fifth lnservice Inspection (ISi)

Interval Attached for your review are relief requests associated with the fifth lnservice Inspection (ISi) interval for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The fifth interval complies with the 2013 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. The fifth ISi interval begins on July 1, 2019 and is currently scheduled to end June 30, 2029. We request your approval of this package by July 1, 2019.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Relief Requests Associated with the Fifth Ten-Year Interval for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, CCNPP Project Manager [CCNPP] USNRC D. A. Tancabel, State of Maryland

Attachment Relief Requests Associated with the Fifth Ten-Vear Interval for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 15R-03 15R-04 15R-05

10 CFR 50.55a RELIEF REQUEST: I5R-03 Revision 0 (Page 1 of 5)

Request for Relief for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10 CFR 50.55a(z)(2)

--Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety--

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV),

Section XI, Class 2 and 3 components that meet the operational and configuration limitations of ASME Code Case N-513-4 (N-513-4), "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l," paragraphs l(a), l(b),

l(c), and l(d).

2. Applicable Code Edition and Addenda

The fifth 10-year interval of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Inservice Inspection (ISI) Program is based on the ASME BPV Code,Section XI, 2013 Edition.

3. Applicable Code Requirement

IWC-3120 and IWD-3120 of ASME Section XI, require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. IWC-3130 and IWD-3130 of ASME Section XI, require that relevant conditions be subject to supplemental examination, corrective measures or repair/replacement activities, or evaluated and accepted by analytical evaluation.

4. Reason for Request

In accordance with 10 CFR 50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded Class 2 and 3 piping whose maximum operating temperature does not exceed 200°F and whose maximum operating pressure does not exceed 275 psig for Calvert Cliffs Nuclear Power Plant, Unit 1 and 2.

Moderately degraded Class 2 and 3 piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow Exelon to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary.

10 CFR 50.55a RELIEF REQUEST: 15R-03 Revision 0 (Page 2 of 5)

Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current Code requirements results in a hardship without a compensating increase in the level of quality and safety.

ASME Code Case N-513-3 (N-513-3) does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. N-513-4 provides guidance for evaluation of flaws in these locations.

5. Proposed Alternative and Basis for Use Exelon is requesting approval to apply the evaluation methods of N-513-4 to Class 2 and 3 components that meet the operational and configuration limitations of N-513-4, paragraphs l(a), l(b), l(c), and l(d) for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, in order to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements.

The Nuclear Regulatory Commission (NRC) issued Generic Letter 90-05 (Reference 1),

"Guidance for Performing Temporary Non-Code Repair of Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping.

The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed ASME Code Case N-513 (N-513). NRC approval ofN-513 versions in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l," Revision 18 (Reference 4 ), allows acceptance of partial through-wall or through-wall leaks for an operating cycle provided all conditions of the code case and NRC conditions are met. The code case also requires the Owner to demonstrate system operability due to leakage.

The ASME recognized that the limitations in N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the code case. Attachment 2 of the Reference 2 letter provides a marked-up N-513-3 version of the code case to highlight the changes compared to the NRC approved N-513-3 version. Attachment 3 of the Reference 2 letter provides the ASME approved N-513-4. The following provides a high level overview of the N-513-4 changes:

1) Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.

10 CFR 50.SSa RELIEF REQUEST: ISR-03 Revision 0 (Page 3 of 5)

2) Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot) 112 from the centerline of the attaching circumferential piping weld.

3) Expanded use to external tubing or piping attached to heat exchangers.

4) Revised to limit the use to liquid systems.

5) Revised to clarify treatment of Service Level load combinations.

6) Revised to address treatment of flaws in austenitic pipe flux welds.

7) Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.

8) Other minor editorial changes to improve the clarity of the code case.

Detailed discussion of significant changes in N-513-4 when compared to NRC approved N-513-3 is provided in Attachment 4 of the Reference 2 letter.

The design basis is considered for each leak and evaluated using the Exelon Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgement. As required by the code case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding.

Leakage rate is not typically a good indicator of overall structural stability in moderate energy systems, where the allowable through-wall flaw sizes are often on the order of inches.

The periodic inspection interval defined using paragraph 2(e) of N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size.

The effects of leakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4 ). The critical leakage rate is determined as the lowest leakage rate that can be tolerated and may be based on the allowable loss of inventory or the maximum leakage that can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based on ASME Code Case N-705 (N-705) (Reference 3), which is accepted without condition in Regulatory Guide 1.147, Revision 18. Paragraph 2.2(e) of N-705 requires a safety factor of two (2) on flaw size when estimating the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for nonplanar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable based upon the precedent of N-705. Note that

10 CFR 50.SSa RELIEF REQUEST: ISR-03 Revision 0 (Page 4 of 5) the alternative herein does not propose to use any portion of N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage.

During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of N-513-4 to confirm that analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the code case. Any re-inspection must be performed in accordance with paragraph 2(a) of the code case.

The leakage limit provides quantitative measurable limits which ensure the operability of the system and early identification of issues that could erode defense-in-depth and lead to adverse consequences.

In summary, Exelon will apply N-513-4 to evaluation of Class 2 and 3 components that are within the scope of the code case. N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. The application of this code case, in concert with safety factors on leakage limits, will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only.

6. Duration of Proposed Alternative The proposed alternative is for used of N-513-4 for Class 2 and Class 3 components within the scope of the code case. An ASME Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first.

Relief is requested for the fifth ISI interval for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, or until the NRC approves N-513-4, or a later revision, in Regulatory Guide 1.147 or other document during the interval. If a flaw is evaluated near the end of the interval for Calve11 Cliffs Nuclear Power Plant, Units 1 and 2, and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage.

7. Precedents Calvert Cliffs Nuclear Power Plant, Units 1 and 2, fourth ISI interval Relief Request was authorized by NRC Safety Evaluation (SE) dated September 6, 2016 (Reference 5). This Calvert Cliffs Nuclear Power Plant, Units 1 and 2, relief request was part of an Exelon fleet-wide submittal, and the alternative for the use of N-513-4 was authorized for various stations.

This relief request for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, fifth ISI interval, utilizes a similar approach to the previously approved relief request.

10 CFR 50.SSa RELIEF REQUEST: ISR-03 Revision 0 (Page 5 of 5)

8. References

l. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Cade Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.

2. Letter from D. T. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Proposed Alternative to Utilize Code Case N-513-4,

'Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l,' "dated January 28, 2016.

3. ASME Section XI Code Case N-705, "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and TanksSection XI, Division l,"

dated October 12, 2006.

4. NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l," Revision 18, dated March 2017.

5. Letter from G. E. Miller (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

Proposed Alternative to Use ASME Code Case N-513-4, dated September 6, 2016 (CAC NOS. MF7301-MF7322) (ADAMS Accession No. ML16230A237).

10 CFR 50.55a RELIEF REQUEST: ISR-04 Revision 0 (Page 1 of 10)

Request for Relief for Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality or Safety--

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel (BPV) Code,Section XI, ISI ferritic piping butt welds requiring radiography during repair/replacement activities.

2. Applicable Code Edition and Addenda The fifth 10-year interval of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Inservice Inspection (ISi) Program is based on the ASME BPV Code,Section XI, 2013 Edition.

3. Applicable Code Requirement 10 CFR 50.55a(b)(2)(xx)(B) requires that "The NDE provision in IW A-4540(a)(2) of the 2002 Addenda of Section XI must be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(l)(ii) of this section." IW A-4540(a)(2) of the 2002 Addenda of ASME Section XI requires that the nondestructive examination method and acceptance criteria of the 1992 Edition or later of ASME Section III be met prior to return to service in order to perform a system leakage test in lieu of a system hydrostatic test.

The examination requirements for ASME Section III, circumferential butt welds are contained in the ASME Code,Section III, Subarticles NB-5200, NC-5200 and ND-5200.

The acceptance standards for radiographic examination are specified in ASME Section III, Subarticles NB-5300, NC-5300 and ND-5300.

IW A-4221 requires that items used for repair/replacement activities meet the applicable Owner's Requirements and Construction Code requirements when performing repair/replacement activities. IW A-4520 requires that welded joints made for installation of items be examined in accordance with the Construction Code identified in the Repair/Replacement Plan.

4. Reason for Request In accordance with 10 CFR 50.55a(z)(l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Replacement of piping is periodically performed in support of the Flow Accelerated Corrosion (FAC) program as well as other repair and replacement activities. The use of encoded Phased Array Ultrasonic Examination Techniques (PAUT) in lieu of

10 CFR 50.55a RELIEF REQUEST: ISR-04 Revision 0 (Page 2 of 10) radiography (RT) to perform the required examinations of the replaced welds would eliminate the safety risk associated with performing RT, which includes the planned exposure and the potential for accidental personnel exposure. PAUT minimizes the impact on other outage activities normally involved with performing RT such as limited access to work locations and the need to control system fill status because RT would require a line to remain fluid empty in order to obtain adequate examination sensitivity and resolution. In addition, encoded PAUT has been demonstrated to be adequate for detecting and sizing critical flaws.

Exelon Generation Company, LLC (Exelon) requests approval of this proposed alternative to support anticipated piping repair and replacement activities for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, during the fifth ISi interval.

5. Proposed Alternative and Basis for Use Calvert Cliffs Nuclear Power Plant, Units 1 and 2, is proposing the use of encoded PAUT in lieu of the Code-required RT examinations for ISi Class 1 and 2 ferritic piping repair/replacement welds. Similar techniques are being used throughout the nuclear industry for examination of dissimilar metal welds and overlaid welds, as well as other applications including ASME B31.1 piping replacements. This proposed alternative request includes requirements that provide an acceptable level of quality and safety that satisfy the requirements of 10 CFR 50.55a(z)(l). The examinations will be performed using personnel and procedures qualified with the requirements of Section 5.1 below.

The electronic data files for the PAUT examinations will be stored as part of the archival-quality records. In addition, hard copy prints of the data will also be included as part of the PAUT examination records to allow viewing without the use of hardware or software.

5.1 Proposed Alternative Calvert Cliffs Nuclear Power Plant, Units 1 and 2, is proposing to perform encoded PAUT examination techniques using demonstrated procedures, equipment and personnel in accordance with the process documented below:

( 1) The welds to be examined shall meet the surface conditioning requirements of the demonstrated ultrasonic procedure.

(2) The welds to be examined shall be conditioned such that transducers properly couple with the scanning surface with no more than a 1/32 in. (0.8 mm) gap between the search unit and the scanning surface.

(3) The ultrasonic examination shall be performed with equipment, procedures, and personnel qualified by performance demonstration.

(4) The examination volume shall include 100% of the weld volume and the weld-to-base-metal interface.

10 CFR 50.55a RELIEF REQUEST: ISR-04 Revision 0 (Page 3 of 10)

(a) Angle beam examination of the complete examination volume for fabrication flaws oriented parallel to the weld joint shall be performed.

(b) Angle beam examination for fabrication flaws oriented transverse to the weld joint shall be performed to the extent practical. Scan restrictions that limit complete coverage shall be documented.

(c) A supplemental straight beam examination shall be performed on the volume of base metal through which the angle beams will travel to locate any reflectors that can limit the ability of the angle beam to examine the weld.

Detected reflectors that may limit the angle beam examination shall be recorded and evaluated for impact on examination coverage. The straight beam examination procedure, or portion of the procedure, is required to be qualified in accordance with ASME Section V, Article 4 and may be performed using non-encoded techniques.

(5) All detected flaw indications from (4)(a) and (4)(b) above shall be considered planar flaws and compared to the preservice acceptance standards for volumetric examination in accordance with IWB-3000, IWC-3000 or IWD-3000. Preservice acceptance standards shall be applied. Analytical evaluation for acceptance of flaws in accordance with IWB-3600, IWC-3600 or IWD-3600 is permitted for flaws that exceed the applicable acceptance standards and are confirmed by surface or volumetric examination to be non-surface connected.

(6) Flaws exceeding the applicable acceptance standards and when analytical evaluation has not been performed for acceptance, shall be reduced to an acceptable size or removed and repaired, and the location of the repair shall be reexamined using the same ultrasonic examination procedure that detected the flaw.

(7) The ultrasonic examination shall be performed using encoded UT technology that produces an electronic record of the ultrasonic responses indexed to the probe position, permitting off-line analysis of images built from the combined data.

(a) Where component configuration does not allow for effective examination for transverse flaws, (e.g., pipe-to-valve, tapered weld transition, weld shrinkage, etc.) the use of non-encoded UT technology may be used for transverse flaws.

The basis for the non-encoded examination shall be documented.

(8) A written ultrasonic examination procedure qualified by performance demonstration shall be used. The qualification shall be applicable to the scope of the procedure, e.g. ,

flaw detection and/or sizing (length or through-wall height), encoded or non-encoded, single and/or dual side access, etc. The procedure shall:

(a) contain a statement of scope that specifically defines the limits of procedure applicability (e.g., minimum and maximum thickness, minimum and maximum diameter, scanning access);

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(b) specify which parameters are considered essential variables, and a single value, a range of values or criteria for selecting each of the essential variables; (c) list the examination equipment, including manufacturer and model or series; (d) define the scanning requirements; such as beam angles, scan patterns, beam direction, maximum scan speed, extent of scanning, and access; (e) contain a description of the calibration method (i.e., actions required to ensure that the sensitivity and accuracy of the signal amplitude and time outputs of the examination system, whether displayed, recorded, or automatically processed, are repeated from examination to examination);

(f) describe the method and criteria for discrimination of indications (e.g.,

geometric indications versus indications of flaws and surface versus subsurface indications); and (g) describe the surface preparation requirements.

(9) Performance demonstration specimens shall conform to the following requirements:

(a) The specimens shall be fabricated from ferritic material with the same inside surface cladding process, if applicable, with the following exceptions:

(i) Demonstration with shielded metal arc weld (SMAW) single-wire cladding is transferable to multiple-wire or strip-clad processes; (ii) Demonstration with multiple-wire or strip-clad process is considered equivalent but is not transferable to SMAW type cladding processes.

(b) The demonstration specimens shall contain a weld representative of the joint to be ultrasonically examined, including the same welding processes.

(c) The demonstration set shall include specimens not thicker than 0.1 in. (2.5 mm) more than the minimum thickness, nor thinner than 0.5 in. (13 mm) less than the maximum thickness for which the examination procedure is applicable. The demonstration set shall include the minimum, within V2 inch of the nominal pipe size (NPS), and maximum pipe diameters for which the examination procedure is applicable. If the procedure is applicable to outside diameter (O.D.) piping of 24 in. (600 mm) or larger, the specimen set must include at least one specimen 24 in. 0.0. (600 mm) or larger but need not include the maximum diameter.

(d) The demonstration specimen scanning and weld surfaces shall be representative of the surfaces to be examined.

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(e) The demonstration specimen set shall include geometric conditions that require discrimination from flaws (e.g., counterbore, weld root conditions, or weld crowns) and limited scanning surface conditions for single-side access, when applicable.

(f) The demonstration specimens shall include both planar and volumetric fabrication flaws (e.g., lack of fusion, crack, incomplete penetration, slag inclusions) representative of the welding process or processes of the welds to be examined. The flaws shall be distributed throughout the examination volume.

(g) Specimens shall be divided into flawed and unflawed grading units.

(i) Flawed grading units shall be the actual flaw length, plus a minimum of 0.25 in. (6 mm) on each end of the flaw. Unflawed grading units shall be at least 1 in. (25 mm).

(ii) The number of unflawed grading units shall be at least 1-1/2 times the number of flawed grading units.

(h) Demonstration specimen set flaw distribution shall be as follows:

(i) For thickness greater than 0.50 in. (13 mm); at least 20% of the flaws shall be distributed in the outer third of the specimen wall thickness, at least 20% of the flaws shall be distributed in the middle third of the specimen wall thickness and at least 40% of the flaws shall be distributed in the inner third of the specimen wall thickness. For thickness 0.50 in. ( 13 mm) and less, at least 20% of the flaws shall be distributed in the outer half of the specimen wall thickness and at least 40% of the flaws shall be distributed in the inner half of the specimen wall thickness.

(ii) At least 30% of the flaws shall be classified as surface planar flaws in accordance with IW A-3310. At least 40% of the flaws shall be classified as subsurface planar flaws in accordance with IW A-3320.

(iii) At least 50% of the flaws shall be planar flaws, such as lack of fusion, incomplete penetration, or cracks. At least 20% of the flaws shall be volumetric flaws, such as slag inclusions.

(iv) The flaw through-wall heights shall be based on the applicable acceptance standards for volumetric examination in accordance with IWB-3400, IWC-3400 or IWD-3400. At least 30% of the flaws shall be classified as acceptable planar flaws, with the smallest flaws being at least 50% of the maximum allowable size based on the applicable all aspect ratio for the flaw. Additional smaller flaws may be included in the specimens to assist in establishing a detection threshold, but shall not be counted as a missed

10 CFR 50.SSa RELIEF REQUEST: ISR-04 Revision 0 (Page 6 of 10) detection if not detected. At least 30% of the flaws shall be classified as unacceptable in accordance with the applicable acceptance standards.

Welding fabrication flaws are typically confined to a height of a single weld pass. Flaw through-wall height distribution shall range from approximately one to four weld pass thicknesses, based on the welding process used.

(v) If applicable, at least two flaws, but no more than 30% of the flaws, shall be oriented perpendicular to the weld fusion line and the remaining flaws shall be circumferentially oriented.

(vi) For demonstration of single-side-access capabilities, at least 30% of the flaws shall be located on the far side of the weld centerline and at least 30% of the planar flaws shall be located on the near side of the weld centerline. The remaining flaws shall be distributed on either side of the weld.

(10) Ultrasonic examination procedures shall be qualified by performance demonstration in accordance with the following requirements.

(a) The procedure shall be demonstrated using either a blind or a non-blind demonstration.

(b) The non-blind performance demonstration is used to assist in optimizing the examination procedure. When applying the non-blind performance demonstration process, personnel have access to limited knowledge of specimen flaw information during the demonstration process. The non-blind performance demonstration process consists of an initial demonstration without any flaw information, an assessment of the results and feedback on the performance provided to the qualifying candidate. After an assessment of the initial demonstration results, limited flaw information may be shared with the candidate as part of the feedback process to assist in enhancing the examination procedure to improve the procedure performance. In order to maintain the integrity of the specimens for blind personnel demonstrations, only generalities of the flaw information may be provided to the candidate.

Procedure modifications or enhancements made to the procedure, based on the feedback process, shall be applied to all applicable specimens based on the scope of the changes.

(c) Objective evidence of a flaw's detection, length and through-wall height sizing, in accordance with the procedure requirements, shall be provided to the organization administering the performance demonstration.

(d) The procedure demonstration specimen set shall be representative of the procedure scope and limitations (e.g., thickness range, diameter range, material, access, surface condition).

10 CFR 50.55a RELIEF REQUEST: ISR-04 Revision 0 (Page 7 of 10)

(e) The demonstration set shall include specimens to represent the minimum and maximum diameter and thickness covered by the procedure. If the procedure spans a range of diameters and thicknesses, additional specimens shall be included in the set to demonstrate the effectiveness of the procedure throughout the entire range.

(f) The procedure demonstration specimen set shall include at least 30 flaws and shall meet the requirements of (9) above.

(g) Procedure performance demonstration acceptance criteria (i) To be qualified for flaw detection, all flaws in the demonstration set that are not less than 50% of the maximum allowable size, based on the applicable all aspect ratio for the flaw, shall be detected. In addition, when performing blind procedure demonstrations, no more than 20% of the non-flawed grading units may contain a false call. Any non-flaw condition (e.g., geometry) reported as a flaw shall be considered a false call.

(ii) To be qualified for flaw length sizing, the root mean square (RMS) error of the flaw lengths estimated by ultrasonics, as compared with the true lengths, shall not exceed 0.25 in. (6 mm) for diameters of NPS 6.0 in.

(DN150) and smaller, and 0.75 in. (18 mm) for diameters greater than NPS 6.0 in. (DN 150).

(iii) To be qualified for flaw through-wall height sizing, the RMS error of the flaw through-wall heights estimated by ultrasonics, as compared with the true through-wall heights, shall not exceed 0.125 in. (3 mm).

(iv) RMS error shall be calculated as follows:

RMS=

n where:

mi= measured flaw size n = number of flaws measured ti = true flaw size (h) Essential variables may be changed during successive personnel performance demonstrations. Each examiner need not demonstrate qualification over the entire range of every essential variable.

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(11) Ultrasonic examination personnel shall be qualified in accordance with IW A-2300.

In addition, examination personnel shall demonstrate their capability to detect and size flaws by performance demonstration using the qualified procedure in accordance with the following requirements:

(a) The personnel performance demonstration shall be conducted in a blind fashion (flaw information is not provided).

(i) The demonstration specimen set shall contain at least 10 flaws and shall meet the flaw distribution requirements of (9)(h) above, with the exception of (9)(h)(v). When applicable, at least one flaw, but no more than 20% of the flaws, shall be oriented perpendicular to the weld fusion line and the remaining flaws shall be circumferentially oriented.

(b) Personnel performance demonstration acceptance criteria:

(i) To be qualified for flaw detection, personnel performance demonstration shall meet the requirements of the following table for both detection and false calls. Any non-flaw condition (e.g., geometry) reported as a flaw shall be considered a false call.

Performance Demonstration Detection Test Acceptance Criteria Detection Test Acceptance False Call Test Acceptance Criteria Criteria Minimum Maximum No. of Flawed No. of Unflawed Detection Number of False Grading Units Grading Units Criteria Calls 10 8 15 2 11 9 17 3 12 9 18 3 13 10 20 3 14 10 21 3 15 11 23 3 16 12 24 4 17 12 26 4 18 13 27 4 19 13 29 4 10 14 30 5 Note 1: Flaws 2:: 50% of the maximum allowable size, based on the applicable all aspect ratio for the flaw.

(ii) To be qualified for flaw length sizing, the RMS error of the flaw lengths estimated by ultrasonics, as compared with the true lengths, shall not

10 CFR 50.SSa RELIEF REQUEST: ISR-04 Revision 0 (Page 9 of 10) exceed 0.25 in. (6 mm) for NPS 6.0 in. (DN150) and smaller, and 0.75 in.

(18 mm) for diameters larger than NPS 6.0 in. (DN150).

(iii) To be qualified for flaw through-wall height sizing, the RMS error of the flaw through-wall heights estimated by ultrasonics, as compared with the true through-wall heights, shall not exceed 0.125 in. (3 mm).

( 12) Documentation of the qualifications of procedures and personnel shall be maintained. Documentation shall include identification of personnel, NDE procedures, equipment and specimens used during qualification, and results of the performance demonstration.

(13) The pre-service examinations will be performed per ASME Section XI (Reference 1).

5.2 Basis for use The overall basis for this proposed alternative is that encoded PA UT is equivalent or superior to RT for detecting and sizing critical (planar) flaws. In this regard, the basis for the proposed alternative was developed from numerous codes, code cases, associated industry experience, articles, and the results of RT and encoded PAUT examinations. It has been shown that PAUT provides an equally effective examination for identifying the presence of fabrication flaws in carbon steel welds compared to RT (Reference 5). The examination procedure and personnel performing examinations are qualified using representative piping conditions and flaws that demonstrate the ability to detect and size flaws that are both acceptable and unacceptable to the defined acceptance standards. The demonstrated ability of the examination procedure and personnel to appropriately detect and size flaws provides an acceptable level of quality and safety alternative as allowed by 10 CFR 50.55a(z)(l).

The requirements in this relief request are based upon ASME Section XI Code Case N-831 (N-831) (Reference 4) and will apply to ISI ferritic piping butt welds requiring radiography during repair/replacement activities. N-831 was approved by ASME Board on Nuclear Codes and Standards on October 20, 2016; however, it has not been incorporated into NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Applicability, ASME Section XI, Division 1," and thus, is not available for application at nuclear power plants without specific NRC approval.

6. Duration of Proposed Alternative Relief is requested for the fifth ISI interval and the remainder of the plant life for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, or until the NRC approves N-831, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

10 CFR 50.55a RELIEF REQUEST: ISR-04 Revision 0 (Page 10 of 10)

7. Precedents

• Calvert Cliffs Nuclear Power Plant, Units 1 and 2, fourth ISI interval relief request was authorized by NRC Safety Evaluation (SE) dated June 5, 2017 (Reference 2).

This Calver Cliffs Nuclear Power Plan, Units 1 and 2, relief request was part of an Exelon fleet-wide submittal, and the use of encoded phased array ultrasonic examination techniques in lieu of radiography was authorized for various stations.

This relief request for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, fifth ISI interval, utilizes a similar approach to the previously approved relief request.

o Relief request was authorized for Millstone Power Station, Units 1 and 2, and Surry Power Station, Units I and 2 by NRC SE dated January 24, 2018 (Reference 3).

8. References

1. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," 2013 Edition.

2. Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques (CAC Nos. MF8763-MF8782 and MF9395), dated June 5, 2017 (ADAMS Accession No. ML17150A091).

3. Letter from Michael T. Markley (NRC) to Daniel G. Stoddard (Dominion Energy),

"Millstone Power Station, Units 1 and 2; North Anna Power Station, Units I and 2; and Surry Power Station, Units 1 and 2; Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination (CAC Nos. MF9923, MF9924, MF9925, MF9926, MF9927, and MF9928; EPID L-2017-LLR-0060)," dated January 24, 2018 (ADAMS Accession No. ML18019Al95).

4. ASME Section XI Code Case N-831, "Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic Pipe Section XI, Division l," ASME Approval Date: October 20, 2016.

5. US NRC, NUREG/CR-7204, "Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping" (ADAMS Accession No. ML15253A674).

10 CFR 50.55a RELIEF REQUEST: ISR-05 Revision 0 (Page 1 of 13)

Request for Relief for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange in Accordance with 10 CFR 50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality or Safety--

1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-G-1 Item Number: B6.40

Description:

Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange Component Number: 54 RPV threads in flange for Unit 1 and 54 RPV threads in flange for Unit 2

2. Applicable Code Edition and Addenda

The fifth 10-year interval of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2013 Edition.

3. Applicable Code Requirement

The Reactor Pressure Vessel (RPV) threads in flange, Examination Category B-G-1, Item Number B6.40, are examined using a volumetric examination technique with 100% of the flange threaded stud holes examined every ISI interval. The examination area is the one-inch area around each RPV stud hole, as shown on Figure IWB-2500-12.

4. Reason for Request

In accordance with 10 CFR 50.55a(z){l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative from the requirement to perform inservice ultrasonic examinations of Examination Category B-G-1, Item Number B6.40, Threads in Flange for Calvert Cliffs Nuclear Power Plant, Units 1 and 2. Exelon has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in the U.S. and internationally have worked with the Electric Power Research Institute (EPRI) to produce Technical Report No.

3002007626, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements" (Reference 1), which provides the basis for elimination of the requirement. The report includes a survey of inspection results from over 168 units, a review of operating experience related to RPV flange/bolting, and a flaw tolerance

10 CFR 50.55a RELIEF REQUEST: ISR-05 Revision 0 (Page 2 of 13) evaluation. The conclusion from this evaluation is that the current requirements are not commensurate with the associated burden (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) of the examination.

The technical basis for this alternative is discussed in more detail below.

Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability was performed as part of Reference 1. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenomena, dealloying corrosion, general corrosion, stress relaxation, creep, mechanical wear and mechanical/thermal fatigue. Other than the potential for mechanical/thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component.

The EPRI report notes a general conclusion from Reference 2, (which includes work supported by the U.S. Nuclear Regulatory Commission (NRC)) that when a component item has no active degradation mechanism present, and a preservice inspection has confirmed that the inspection volume is in good condition (i.e., no flaws I indications),

then subsequent in-service inspections do not provide additional value going forward. As discussed in the Operating Experience review summary below, the RPV flange ligaments have received the required pre-service examinations and over 10,000 in-service inspections, with no relevant findings.

To address the potential for mechanical/thermal fatigue, Reference 1 documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical/thermal fatigue potential. The evaluation consists of two parts. In the first part, a stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in the ASME Code,Section XI, IWB-3500. The Pressurized Water Reactor (PWR) design was selected because of its higher design pressure and temperature. A representative geometry for the finite element model used the largest PWR RPV diameter along with the largest bolts and the highest number of bolts. The larger and more numerous bolt configuration results in less flange material between bolt holes, whereas the larger RPV diameter results in higher pressure and thermal stresses.

Stress Analysis A stress analysis was performed in Reference 1 to determine the stresses at critical regions of the thread in flange component as input to a flaw tolerance evaluation. Sixteen nuclear plant units (ten PWRs and six Boiling Water Reactors (BWRs)) were considered in the analysis. The evaluation was performed using a geometric configuration that bounds the sixteen units considered in this effort. The details of the RPV parameters for

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 3 of 13)

Calvert Cliffs Nuclear Power Plant, Units 1 and 2, as compared to the values used in the evaluation of the bounding preload stress are shown in Table 1. The preload stresses for both units are bounded by the Reference 1 report. Specifically, the Reference 1 preload stress is 42,338 psi, whereas the preload stresses are 30,747 psi for Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The Calvert Cliffs Nuclear Power Plant, Units l and 2, stresses are bounded by the Reference l report which demonstrates that the report remains applicable to this relief request. Dimensions of the analyzed geometry are shown in Figure ISI-05-05-1.

For comparison purposes, the global force per flange stud can be estimated by the pressure force on the flange (p*n*r 2 , where pis the design pressure and r is the vessel inside radius at the stud hole elevation) divided by the number of stud holes. From the parameters in Table 1, this results in a value of 1088 kips per stud for the configuration used in the analysis and 1076 kips per stud for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 configurations, indicating that the configuration used in the analysis bounds that at Calvert Cliffs Nuclear Power Plant, Units, Units 1 and 2. As shown in Table 1, the preload stress used in the analysis is also bounding compared to that at Calvert Cliffs Nuclear Power Plant, Units 1 and 2.

Table 1: Comparison of Parameters to Values Used in Bounding Analysis No. of Minimum Stud RPV Inside Flange Design Preload Studs No. of Nominal Diameter at Thickness at Plant Pressure Stress Currently Studs Diameter Stud Hole Stud Hole (psig) (psi)

Installed Evaluated (inches) (inches) (inches)

Calvert Cliffs, Unit 1 54 54 7 172 16.5 2500 30.747 Calvert Cliffs, Unit 2 54 54 7 172 16.5 2500 30,747 Values Used in 54 54 6.0 173 16 2500 42,338 Bounding Analysis The analytical model is shown in Figures ISI-05-05-2 and ISI-05-05-3. The loads considered in the analysis consisted of:

• A design pressure of 2500 psia at an operating temperature of 600°F was applied to all internal surface exposed to internal pressure.

o Bolt/stud preload - Stress of 42,338 psi.

o Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100°F/hour ramp up to the operating temperature, with a corresponding pressure ramp up to the operating pressure.

The ANSYS finite element analysis program was used to determine the stresses in the thread in flange component for the three loads described above.

10 CFR 50.SSa RELIEF REQUEST: ISR-05 Revision 0 (Page 4 of 13)

Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis to determine how long it would take an initial postulated flaw to reach the ASME Code,Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Code,Section XI, IWB-3600 was performed.

Stress intensity factors (K's) at four flaw depths of a 360° inside-surface-connected, partial-through-wall circumferential flaws are calculated using finite element analysis techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (alt) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole.

This creates an ellipsoidal flaw shape around the circumference of the flange, as shown in Figure ISI-05-05-4 for the flaw model with alt= 0.77 alt crack model. The crack tip mesh for the other flaw depths follows the same pattern. When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases.

The maximum K results are summarized in Table 2 for the four crack depths. From Table 2, the maximum K occurs at operating conditions (preload + heatup +pressure).

Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K vs. a profile.

Table 2: Maximum K vs. alt Kat Crack Depth (ksiv'in)

Load 0.02 alt 0.29 alt 0.55 alt 0.77 alt Preload 11.2 17.4 15.5 13.9 Preload + Heatup + Pressure 13.0 19.8 16.1 16.3 The allowable stress intensity factor was determined based on the acceptance criteria in ASME Section XI, IWB-3610/Appendix A which states that:

K1 < K1clv'l0 = 69.6 ksiv'in Where, K1 = Allowable stress intensity factor (ksiv'in)

Kie = Lower bound fracture toughness at operating temperature (220 ksiv'in)

As can be seen from Table 2, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of alt= 0.77. Hence the allowable flaw depth of the 360° circumferential flaw is at least 77% of the thickness of the flange. The

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 5 of 13) allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis.

For the crack growth evaluation, an initial postulated flaw size of 0.2 in. (5.08 mm) is chosen consistent with the ASME Section XI IWB-3500 flaw acceptance standards. The deepest flaw analyzed is alt= 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heat-up/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta Kand the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension).

An evaluation was also performed to determine the acceptability at preload condition.

Table 3 below provides the RPV flange RTNoT values and the bolt-up temperatures for Calvert Cliffs Nuclear Power Plant, Units 1 and 2. These were determined using the RTNoT value from plant records. As can be seen from this table, the minimum (T-RTNoT) is 90°F and 40°F, corresponding to Calvert Cliffs Nuclear Power Plant, Units 1 and 2, respectively. From the equations in paragraph A-4200 of ASME Section XI, Appendix A, the corresponding values of Kie are 159 and 79 ksi-Vin. Using a structural factor of

-V 10, the allowable Kie value is 50.2 and 25. l ksi-Vin. This value is more than the maximum stress intensity factor (K1) for the preload condition of 17.4 ksi-Vin shown in Table 2, thus the report evaluation is bounding for Calvert Cliffs Nuclear Power Plant, Units 1 and 2.

Table 3: RPV Flange RTNDT and Bolt-Up Temperature Flange RTNoT (°F) Minimum Preload Plant Name From Plant FromNRC T-RTNoT Temp (°F)

Records RVID2 Database (oF)

Calvert Cliffs, Unit 1 -20 -10 >70 to <130 90 Calvert Cliffs, Unit 2 30 30 >70 to <130 40 The stress analysis I flaw tolerance evaluation presented above shows that the thread in flange component at the units in the relief request is very flaw tolerant and can operate for 80 years without violating ASME Code,Section XI safety margins. This clearly demonstrates that the thread in flange examinations can be eliminated without affecting the safety of the RPV.

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 6 of 13)

Operating Experience Review Summary As discussed above, the results of the survey, which includes results from the Calvert Cliffs Nuclear Power Plant, Units l and 2, confirmed that the RPV threads in flange examination are adversely impacting outage activities (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) while not identifying any service induced degradations. Specifically, for the U.S. fleet, a total of 94 units have responded to date and none of these units have identified any type of degradation. As can be seen in Table 4 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total 3,793 examinations were conducted and for the PWR units a total of 6,869 examinations were conducted, with no service-induced degradation identified. The response data includes information from all of the plant designs in operation in the U.S. and includes BWR-2, -

3, -4, -5 and -6 designs. The PWR plants include the 2- loop, 3-loop and 4-loop designs and each of the PWR NSSS designs (i.e., Babcock & Wilcox, Combustion Engineering and Westinghouse).

Table 4: Summary of Survey Results - U.S. Fleet Number of Number of Number of Plant Type Reportable Units Examinations Indications BWR 33 3,793 0 PWR 61 6,869 0 Total 94 10,662 0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, Reference 1 discusses studies conducted in response to the issuance of the Anticipated Transient Without Scram (ATWS) Rule by the NRC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an ATWS event. In particular, the reactor coolant system (RCS) and its individual components were reviewed to determine weak links. As an example, even though significant structural margin was identified in NRC SECY-83-293 for PWRs, the ASME Service Level C pressure of 3200 psig was assumed to be an unacceptable plant condition. While a higher ASME service level might be defensible for major RCS components, other portions of the RCS could deform to the point of inoperability.

Additionally, there was the concern that steam generator tubes might fail before other RCS components, with a resultant bypass of containment. The key take-away for these studies is that the RPV flange ligament was not identified as a weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 7 of 13)

In summary, Reference 1 identifies that the RPV threads in flange are performing with very high reliability based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g., the number and magnitude of transients is small, generally not in contact with primary water at plant operating temperatures/pressures, etc.). The robust design is manifested in that plant operation has been allowed at several plants even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.

5. Proposed Alternative and Basis for Use In lieu of the inservice requirements for a volumetric ultrasonic examination, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, proposes that the industry report (Reference 1) provides an acceptable technical basis for eliminating the requirement for this examination because the alternative maintains an acceptable level of quality and safety.

This report provides the basis for the elimination of the RPV threads in flange examination requirement (ASME Section XI Examination Category B-G-1, Item Number B6.40). This report was developed because evidence had suggested that there have been no occurrences of service-induced degradation and there are negative impacts on worker dose, personnel safety, radwaste, critical path time for these examinations and additional time at reduced water inventory.

Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, Calver Cliffs Nuclear Power Plant, Units 1 and 2, requests authorization to use the proposed alternative in accordance with 10 CFR 50.55a(z)(l) on the basis that use of the alternative provides an acceptable level of quality and safety.

To protect against non-service related degradation, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, uses detailed procedures for the care and visual inspection of the RPV studs and the threads in flange each time the RPV closure head is removed. Care is taken to inspect the RPV threads for damage and to protect threads from damage when the studs are removed. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then replaced and tensioned into the RPV flange. This activity is performed each time the closure head is removed, and the procedure documents each step. These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service.

Technical Procedure RV-79, "Reactor Vessel Alignment Pin Removal, Stud Hole Cleaning and Stud Installation," Step 4.6.9, requires reactor vessel stud, nut, and washer cleaning and inspection be completed prior to installation. Additionally Step 6.4 requires an inspection of the reactor vessel stud holes as part of the stud hole cleaning and lubrication.

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 8 of 13)

Technical procedure RV-62, "Reactor Vessel Stud, Nut, and Washer Cleaning and Inspection," provides guidance in Section 6.0 for the cleaning, inspection, and lubrication of the reactor vessel studs, nuts, and washers. Cleaning of the reactor vessel studs, nuts, and washers is performed to remove all loose oxidation, coating residue, and caked on lubricant prior to installation. Inspection of the reactor vessel studs, nuts, and washers is performed to identify nicks, burrs, thread deformation, or any other signs of damage.

If a nonconforming condition is identified, a corrective action issue report is initiated to document the condition in accordance with plant administrative procedures. The 10 CFR 50, Appendix B corrective action program ensures that conditions adverse to quality are promptly corrected. If the deficiency is assessed to be significantly adverse to quality, the cause of the condition is determined, and an action plan is developed to preclude recurrence.

The requirement in this relief request are based upon ASME Section XI Code Case N-864 (N-864) (Reference 5) and will apply to Examination Category B-G-1, Item Number B6.40, Reactor Vessel Threads in Flange. N-864 was approved by ASME Board on Nuclear Codes and Standards on July 28, 2017; however, it has not been incorporated into NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l ," and thus, is not available for application at nuclear power plants without specific NRC approval.

6. Duration of Proposed Alternative Relief is requested for the fifth ISI interval for Calvert Cliffs Nuclear Power Plant Units, 1 and 2, or until the NRC approves N-864, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents Calvert Cliffs Nuclear Power Plants, Units 1 and 2, fourth ISI interval relief request was authorized by NRC Safety Evaluation (SE) dated June 26, 2017 (Reference 3). This Calvert Cliffs Nuclear Power Plant, Units 1 and 2, relief request was part of an Exelon fleet-wide submittal, and the alternative for examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, threads in flange was authorized for various stations. This relief request for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, fourth IS interval, utilizes a similar approach to th\:! previously approved relief request.

Relief request was authorized for Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 by NRC SE dated January 26, 2017 (Reference 4)

[ADAMS Accession No. ML17006Al09].

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 9 of 13)

8. References

1. Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements. EPRI, Palo Alto, CA: 2016. 3002007626 (ADAMS Accession No. ML16221A068).

2. American Society of Mechanical Engineers, Risk-Based Inspection:

Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.

3. Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548), dated June 26, 2017 (ADAMS Accession No. MLl 7l70AO13).

4. Letter from M. T. Markley (NRC) to C.R. Pierce (Southern Nuclear Operating Co. Inc.), "Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 - Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads Inflange Inspection (CAC Nos. MF8061, MF8062, MF8070)," dated January 26, 2017 (ADAMS Accession No. MLl 7006A 109).

5. ASME Section XI Code Case N-864, "Reactor Vessel Threads in Flange Examinations,"Section XI, Division 1, ASME Approval Date: July 28, 2017.

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 10 of 13)

Figure ISI-05-05-1 Modeled Dimensions R86.5"  ;:

8.5" 12.0" 17.0" ~

7.0"

~L

'* 16.0" .

R83.75"  ;:

R4. 5"

- 10. 75" .

R85.69" ..

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 11 of 13)

Figure ISI-05-05-2 Finite Element Model Showing Bolt and Flange Connection 1

E:UMl*fl'S RFAL Nll-1 F1':0_Vessel_Flange

10 CFR 50.55a RELIEF REQUEST: 15R-05 Revision 0 (Page 12 of 13)

Figure ISI-05-05-3 Finite Element Model Mesh with Detail at Thread Location

10 CFR 50.55a RELIEF REQUEST: I5R-05 Revision 0 (Page 13 of 13)

Figure ISI-05-05-4 Cross Section of Circumferential Flaw with Crack Tip Elements Inserted After 10th Thread from Top of Flange