Text

Relief Requests Associated with the Fourth Inservice Inspection Interval
	ML19350C642
Person / Time
Site:	Clinton
Issue date:	12/16/2019
From:	Gullott D; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-19-116
	Download: ML19350C642 (103)
	v • d • e

4300 Winfield Road Warrenville, IL 60555 630 65 7 2000 Office RS-19-116 10 CFR 50.55a December 16, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

Relief Requests Associated with the Fourth lnservice Inspection Interval In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (EGC) requests NRC approval of the attached relief requests associated with the fourth lnservice Inspection (ISi) interval for Clinton Power Station (CPS),

Unit 1. The fourth interval of the CPS, Unit 1, ISi program complies with the 2013 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The fourth ISi interval at CPS will begin on July 1, 2020, and is currently scheduled to end June 30, 2030.

EGC requests approval of these requests by December 16, 2020, to support implementation of the fourth 10-year ISi interval.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

Respectfully, David M. Gullatt Director Licensing

Attachment:

Relief Requests Associated with the Fourth lnservice Inspection Interval cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector - Clinton Power Station

ATTACHMENT Relief Requests Associated with the Fourth Inservice Inspection Interval A listing of the relief requests associated with the fourth Inservice Inspection interval for Clinton Power Station is provided below.

1. I4R-01: Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds

2. I4R-02: Alternative Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section

3. I4R-03: Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves

4. I4R-04: Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs)

5. I4R-05: Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) Piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB

6. I4R-06: Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection

7. I4R-07: Alternative to Use ASME Code Case N-864, Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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1. ASME Code Component(s) Affected Code Class: 1 and 2

Reference:

Table IWB-2500-1, Table IWC-2500-1 Examination Category: B-F, B-J, C-F-1, and C-F-2 Item Number: B5.10, B5.20, B9.11, B9.21, B9.31, B9.32, B9.40, C5.11, C5.51, and C5.81

Description:

Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Component Number: Pressure Retaining Piping

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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

Table IWB-2500-1, Examination Category B-F, requires volumetric and surface examinations on all welds for Item Number B5.10 and surface examinations for all welds for Item Number B5.20.

Table IWB-2500-1, Examination Category B-J, requires volumetric and surface examinations on a sample of welds for Item Numbers B9.11 and B9.31, and surface examinations on a sample of welds for Item Numbers B9.21, B9.32, and B9.40. The weld population selected for inspection is specified in Note (2).

Note (2) Examinations shall include the following:

(a) All terminal ends in each pipe or branch run connected to vessels.

(b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the following limits under loads associated with specific seismic events and operational conditions:

(1) primary plus secondary stress intensity range of 2.4Sm for ferritic steel and austenitic steel.

(2) cumulative usage factor U of 0.4.

(c) All dissimilar metal welds not covered under Examination Category B-F.

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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(d) Additional piping welds so that the total number of circumferential butt welds (or branch connection or socket welds) selected for examination equals 25% of the circumferential butt welds (or branch connection or socket welds) in the reactor coolant piping system. This total does not include welds exempted by IWB-1220. These additional welds may be located as follows Note (2) For BWR plants (a) One reactor coolant recirculation loop (where a loop or run branches, only one branch)

(b) One branch run representative of an essentially symmetric piping configuration among each group of branch runs that are connected to a loop and that perform similar system functions (c) One steam line run representative of an essentially symmetric piping configuration among the runs (d) One feedwater line run representative of an essentially symmetric piping configuration among the runs (where a loop or run branches, only one branch)

(e) Each piping and branch exclusive of the categories of loops and runs that are part of the system piping of (a) through (d) above Table IWC-2500-1, Examination Categories C-F-1 and C-F-2 require volumetric and surface examinations on a sample of welds for Item Numbers C5.11 and C5.51 and surface examinations on a sample of welds for Item Number C5.81. The weld population selected for inspection is specified in Note (2) for both Examination Categories.

Note (2) The welds selected for examination shall include 7.5%, but not less than 28 welds, of all dissimilar metal, austenitic stainless steel or high alloy welds (Examination Category C-F-1) or of all carbon and low alloy steel welds (Examination Category C-F-2) not exempted by IWC-1220. (Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Categories C-F-1 and C-F-2. These welds, however, shall be included in the total weld count to which the 7.5% sampling rate is applied.)

The examinations shall be distributed as follows:

(a) the examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt dissimilar metal, austenitic stainless steel and high alloy welds (Examination Category C-F-1) or carbon and low alloy welds (Examination Category C-F-2) in each system;

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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(b) within a system, the examinations shall be distributed among terminal ends, dissimilar metal welds, and structural discontinuities prorated, to the degree practicable, on the number of nonexempt terminal ends, dissimilar metal welds, and structural discontinuities in the system; and (c) within each system, examinations shall be distributed between line sizes prorated to the degree practicable.

4. Reason for Request

In accordance with 10 CFR 50.55a(z)(1), relief is requested on the basis that the proposed alternative utilizing Electric Power Research Institute (EPRI) Topical Report (TR) 112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

Revision B-A (Reference 1) along with two enhancements from ASME Code Case N-578-1 (N-578-1), "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1," (Reference 4) will provide an acceptable level of quality and safety.

As stated in "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999)"

(Reference 2):

"The staff concludes that the proposed RI-ISI Program as described in EPRI TR-112657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a for the proposed alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection."

The initial CPS Risk Informed Inservice Inspection (RI-ISI) Program was submitted during the first period of the Second ISI Interval. This initial RI-ISI Program was developed in accordance with EPRI TR-112657, Revision B-A, as supplemented by N-578-1. The initial program was approved for use by the Nuclear Regulatory Commission (NRC) via a Safety Evaluation (SE) as transmitted to Exelon Generation Company, LLC (EGC) on April 8, 2002 (Reference 5).

The CPS RI-ISI Program was resubmitted using the same approach during the Third ISI Interval. The program was approved for use by the NRC via a SE as transmitted to EGC on December 22, 2010 (Reference 6).

The transition from the 2004 Edition with No Addenda to the 2013 Edition of ASME Section XI for CPS's Fourth ISI Interval does not impact the currently approved RI-ISI evaluation methods and process used in the Third ISI Interval, and the requirements of the new Code Edition will be implemented as detailed in the CPS ISI Program Plan.

Therefore, with the exception of specific weld locations that may have changed due to maintenance or modification activities, the proposed alternative RI-ISI Program for the

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Fourth ISI Interval is the same program methodology as approved in Reference 6 for the Third ISI Interval.

The Risk Impact Assessment completed as part of the initial baseline RI-ISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI Program to the new RI-ISI methodology. For the Fourth Interval ISI update, there is no transition occurring between two different methodologies, but rather, the previously approved RI-ISI methodology and evaluation will be maintained for the new interval. The initial methodology of the evaluation has not changed, and the change in risk was simply re-assessed using the initial 1989 Edition with No Addenda ASME Section XI Program prior to RI-ISI and the new element selection for the Fourth ISI Interval RI-ISI Program. This same process has been maintained in each revision to the CPS RI-ISI assessment that has been performed to date.

Based on the Fourth ISI Interval update of this risk impact assessment, the change in risk from the pre-RI-ISI ASME Section XI Program to the Fourth Interval RI-ISI Program was below the 1.00E-06 and 1.00E-07 acceptance criteria for delta-core damage frequency (Delta-CDF) and delta-large early release frequency (Delta-LERF) as described in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." The Delta-CDF and Delta-LERF values for CPS are listed in the following table.

Change in Risk from Clinton Power Station, Unit 1 Pre-RI-ISI ASME Section XI Program to Fourth Interval RI-ISI Program Unit No. Delta-CDF Delta-LERF Unit 1 7.95E-09 1.36E-09 The following tables document the Delta-CDF and Delta-LERF for CPS over the initial ASME Section XI Program for the Fourth ISI Interval. The results for CPS are provided in the first table by system and in the second table by system for only the Break Exclusion Region (BER) weld population.

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Clinton Power Station, Unit 1 Delta-CDF and Delta-LERF by System CDF LERF Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISI RI-ISI Criteria Criteria AAI 1.21E-10 1.00E-07 7.74E-12 1.00E-08 AAP -9.32E-11 1.00E-07 -2.80E-12 1.00E-08 PFW 8.41E-10 1.00E-07 1.41E-10 1.00E-08 PHP -6.87E-10 1.00E-07 -4.38E-11 1.00E-08 PLP -6.95E-10 1.00E-07 -4.42E-11 1.00E-08 PMS 9.26E-10 1.00E-07 4.40E-10 1.00E-08 PNB 7.11E-12 1.00E-07 4.27E-13 1.00E-08 PRH 8.24E-10 1.00E-07 1.04E-10 1.00E-08 PRI 6.06E-11 1.00E-07 -4.44E-11 1.00E-08 PRR 5.59E-09 1.00E-07 3.56E-10 1.00E-08 PRT 1.25E-09 1.00E-07 4.78E-10 1.00E-08 PSC -1.94E-10 1.00E-07 -2.92E-11 1.00E-08 PSD 1.00E-14 1.00E-07 1.94E-13 1.00E-08 Total 7.95E-09 1.00E-06 1.36E-09 1.00E-07

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Clinton Power Station, Unit 1 BER Weld Delta-CDF and Delta-LERF by System CDF LERF Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISI RI-ISI Criteria Criteria PFW 9.96E-11 1.00E-07 9.37E-11 1.00E-08 PHP 8.06E-12 1.00E-07 3.22E-13 1.00E-08 PLP 5.31E-13 1.00E-07 7.59E-14 1.00E-08 PMS 4.04E-10 1.00E-07 4.03E-10 1.00E-08 PRH 3.23E-11 1.00E-07 2.87E-11 1.00E-08 PRI 4.92E-11 1.00E-07 -4.69E-11 1.00E-08 PRT 5.93E-10 1.00E-07 4.36E-10 1.00E-08 Total 1.19E-09 1.00E-06 9.15E-10 1.00E-07 The actual "evaluation and ranking" procedure including the Consequence Evaluation and Degradation Mechanism Assessment processes of the currently approved (Reference 6) RI-ISI Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-112657, Revision B-A. These requirements of the RI-ISI Program have been and will continue to be reevaluated and revised as major revisions of the site Probabilistic Risk Assessment (PRA) occur and modifications to plant configuration are made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, Element Selection, and Risk Impact Assessment steps encompass the complete living program process applied under the CPS RI-ISI Program. All Risk Categories and Element Selections were validated as part of the new Fourth Interval ISI Program development, and the living program process will be maintained throughout the interval.

5. Proposed Alternative and Basis for Use The proposed alternative initially implemented in the CPS, "Risk Informed Inservice Inspection Evaluation," (Reference 3), along with the two enhancements noted below, provide an acceptable level of quality and safety as required by 10 CFR 50.55a(z)(1).

This initial program along with these enhancements was resubmitted and is currently approved for CPS's Third ISI Interval as documented in Reference 6.

The Fourth ISI Interval RI-ISI Program will be a continuation of the current application and will continue to be a living program as described in Section 4 of this relief request.

No changes to the evaluation methodology as currently implemented under EPRI TR-112657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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a. In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISI Selected Examinations" of EPRI TR-112657, CPS will utilize the requirements of Paragraph -2430, "Additional Examinations" contained in N-578-1 (Reference 4). The alternative criteria for additional examinations contained in N-578-1 provides a more refined methodology for implementing necessary additional examinations. The reason for this selection is that the guidance discussed in EPRI TR-112657 includes requirements for additional examinations at a high level, based on service conditions, degradation mechanisms, and the performance of evaluations to determine the scope of additional examinations, whereas N-578-1 provides more specific and clearer guidance regarding the requirements for additional examinations that is structured similar to the guidance provided in ASME Section XI, Paragraphs IWB-2430 and IWC-2430. Additionally, similar to the current requirements of ASME Section XI, CPS intends to perform additional examinations that are required due to the identification of flaws or relevant conditions exceeding the acceptance standards, during the outage the flaws are identified.

b. To supplement the requirements listed in EPRI TR-112657, Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods,"

CPS will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations" contained in N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 2013 Edition of ASME Section XI (i.e., CPS's Code of Record for the Fourth ISI Interval) will be utilized which parallel those referenced in the code case. Table 1 of N-578-1 will be used as it provides a detailed breakdown for "Examination Method" and "Categorization of Parts to be Examined." Based on these methods and categorization, the examination figures specified in EPRI TR-112657, Section 4 will then be used to determine the examination volume/area based on the degradation mechanism and component configuration. For piping structural elements not subject to a degradation mechanism, N-578-1, Table 1, Note 1 will be applied using the expanded examination volume.

The CPS RI-ISI Program, as developed in accordance with EPRI TR-112657, Revision B-A (Reference 1), requires that 25% of the piping structural elements that are categorized as "High" risk (i.e., Risk Category 1, 2, and 3) and 10% of the piping structural elements that are categorized as "Medium" risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI TR-112657 while the guidance for the examination method and categorization of parts to be examined are provided by the EPRI TR-112657 as supplemented by N-578-1.

For NRC consideration in the evaluation of this alternative RI-ISI Program, Enclosure CL-MISC-030, Revision 0 to this relief request contains a summary of the Regulatory Guide 1.200, Revision 2 (Reference 7) evaluation performed on CL-PRA-014, Revision 8 (Reference 8), and the impact of the identified gaps on the technical

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 8 of 9) adequacy of the CPS PRA Model to support this RI-ISI application (see Enclosure, Tables 3-1 and 3-2).

In addition to this risk informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to receive Code-required system pressure testing as part of the current ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the CPS System Pressure Testing Program, which remains unaffected by the RI-ISI Program.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-01, was authorized by NRC SE dated December 22, 2010. This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

Braidwood Station, Units 1 and 2, Relief Request I4R-01, was authorized by NRC SE dated January 17, 2019 (ADAMS Accession No. ML18318A272).

LaSalle County Station, Units 1 and 2 Relief Request I4R-01, was authorized by NRC SE dated January 8, 2018 (ADAMS Accession No. ML18003A247).

8. References

1. Electric Power Research Institute (EPRI) Topical Report (TR) 112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," Revision B-A, dated December 1999

2. Letter from W. H. Bateman (NRC) to G. L. Vine (EPRI) "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999)," dated October 28, 1999 (ADAMS Accession Nos. ML993190460 and ML993190474)

3. Initial Risk-Informed Inservice Inspection Evaluation, Revision 0 - Clinton Power Station, dated October 15, 2001. (Letter RS-01-219 from K. A. Ainger (AmerGen) to the NRC, Clinton Power Station Second Interval Inservice Inspection Program - Relief Request 4208, "Alternative to the ASME Boiler and Pressure Vessel Code Section XI Requirements for Class 1 and 2 Piping Welds Risk-Informed Inservice Inspection Program," dated October 15, 2001 (ADAMS Accession No. ML012950371))

10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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4. American Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1," dated March 28, 2000

5. Letter from A. J. Mendiola, (NRC) to J. L. Skolds (EGC) "Clinton Power Station, Unit 1 - Risk-Informed Inservice Inspection Program, Relief Request 4208 (TAC No. MB53211)," dated April 8, 2002 (ADAMS Accession No. ML020800820)

6. Letter from R. D. Carlson (NRC) to M. J. Pacilio (EGC), "Clinton Power Station, Unit No. 1 -Relief Requests I3R-01, I3R-02, I3R-03, I3R-04, and I3R-05 Associated with the Third Inservice Inspection Interval (TAC Nos. ME2987, ME2988, ME2989, ME2990, and ME2991)," dated December 22, 2010 (ADAMS Accession No. ML103360335)

7. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Revision 2, dated March 2009 (ADAMS Accession No. ML090410014)

8. Clinton Power Station Quantification Notebook, CL-PRA-014, Revision 8, August 2018. (PRA model CL117A)

9. Enclosure Clinton Power Station, PRA Technical Adequacy for Risk-Informed In-Service (RI-ISI)

Applications (CL-MISC-030), Revision 0, dated October 2019

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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CL-MISC-030, Revision 0, "Clinton Power Station PRA Capability Assessment for Risk-Informed In-Service Inspection (RI-ISI) Applications," dated October 2019 1.0 PURPOSE This notebook documents the technical adequacy assessment of the Clinton Full Power Internal Events (FPIE) PRA model (CL117A [6]) in order to support an In-Service Inspection (ISI)

Program relief request for continuation of the Clinton Risk-Informed ISI (RI-ISI) program for another 10-year interval.

The PRA technical adequacy assessment is in accordance with the requirements of Regulatory Guide 1.200 Revision 2 [2] and NRC approved EPRI guidance [3].

2.0 METHODOLOGY The approach used for this assessment is consistent with the approach used at other Exelon nuclear power sites (e.g., Byron & Braidwood, Dresden, and Quad Cities). Information regarding PRA technical gaps relative to the ASME / ANS PRA Standard [1], as documented in the Clinton Peer Review Report [4] and Clinton PRA Finding and Suggestion Level Fact and Observation Independent Assessment Report [5], was reviewed relative to the technical capability required to perform the RI-ISI analyses with input from the NRC approved EPRI report on technical adequacy for RI-ISI [3]. The impacts are discussed in subsequent sections of this notebook.

3.0 PRA TECHNICAL ADEQUACY ASSESSMENT Exelon Generation Company (Exelon) employs a multi-faceted approach to establishing and maintain the technical adequacy and plant fidelity of the PRA models for all operating Exelon nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews.

The following subsections describe the approach used for the Clinton PRA technical adequacy assessment.

3.1 PRA MAINTENANCE AND UPDATE The Exelon Risk Management (RM) process ensures that the applicable PRA model remains an accurate reflection of the "as-built, as operated" status of the plants. This process is defined in the Exelon Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation procedures.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Exelon procedure ER-AA-600-1015, "FPIE PRA Model Update" delineates the responsibilities and guidelines for updating the FPIE PRA models at all operating Exelon nuclear power plants.

The overall RM program, including ER-AA-600-1015, defines the following:

Process for implementing regularly scheduled and interim PRA model updates.

Process for tracking issues identified as potentially affecting the PRA models (e.g., plant changes, errors or limitations identified in the model, industry experience, etc.) through the use of the Updating Requirements Evaluation (URE) database.

Process for controlling the model and associated computer files.

To ensure that the current PRA model remains an accurate reflection of the "as-built, as-operated" configuration of the plants, the following activities are routinely performed:

Design changes and procedure changes are reviewed for their impact on the PRA model.

New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

Maintenance unavailabilities are captured and their impact on risk is trended.

Plant-specific initiating event frequencies, failure rates, and maintenance unavailabilities for equipment that can have a significant impact on the PRA model are updated approximately every four years.

o Longer intervals may be justified if it can be shown that the PRA continues to adequately represent the "as-built, as operated" configuration of the plant.

In addition to these activities, Exelon Risk Management procedures provide guidance for the following activities:

Documentation of the PRA model, PRA products, and bases documents.

The approach for controlling electronic storage of RM products including, but not limited to, PRA update information, PRA models, and PRA applications.

Guidance for updating the hazard-specific PRA model.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program risk evaluations for maintenance activities (e.g., corrective maintenance, preventative maintenance, minor maintenance, surveillance tests, and modifications) on systems, structures, and components (SSCs) with the scope of the Maintenance Rule (10 CFR 50.65(a)(4)).

The most recent update of the Clinton FPIE PRA model (CL117A) [6] was completed in August 2018 as a result of a regularly schedule update to the previous PRA model (CL114A)

[7]. The CL117A FPIE PRA model is the most recent evaluation of the risk profile at Clinton for internal event challenges (including internal flood). The Clinton PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the Clinton PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.

3.2 PRA MODEL HISTORY Several PRA technical capability assessments have been performed for the Clinton PRA and future assessments will continue with each update to the PRA model. A chronological list of the assessments performed include the following:

An independent PRA peer review was conducted under the auspices of the BWR Owners Group (BWROG) in 2000 using the industry PRA Peer Review process

[8]. This peer review included an assessment of the PRA model maintenance and update process.

A self-assessment analysis was previously performed against Addenda B of the ASME PRA Standard (ASME RA-Sb-2005 [9]) and a draft version of Regulatory Guide 1.200, Rev. 1 (DG-1161) to support scoping / planning of the Clinton 2006 PRA update.

From 2005-2006, the Clinton PRA model results were evaluated in the BWROG PRA cross-comparisons study performed in support of implementation of the Mitigating Systems Performance Indicator (MSPI) process.

The Clinton 2006 PRA self-assessment was revised in 3Q09 to address consistency with Regulatory Guide 1.200, Rev. 1 [10] in preparation of the Clinton 2009 PRA peer review.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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The Clinton 2006 PRA model was peer reviewed in October 2009 for internal events and internal floods [4] using the ASME / ANS PRA Standard (ASME /

ANS RA-Sa-2009 [1]) and any clarifications provided in Regulatory Guide 1.200 Revision 2 [2].

o Of the 316 Supporting Requirements (SRs) reviewed:

247 (~78%) were assessed as "Capability Category I/II" or greater 12 (~4%) were assessed as "Capability Category I" 57 (~18%) were assessed as "Not Met".

In 2011, the Clinton PRA model was updated as part of the periodic update and several of the Facts and Observations (F&Os) from the 2009 Peer Review were also resolved.

In 2014, a data-only update was performed for the Clinton PRA model and FLEX was incorporated into the model.

In 2017, the Clinton PRA model was updated as part of the periodic update. The 2017 model (CL117A) was subjected to an F&O Closure (December 2018) performed by an independent team to close out the F&Os linked to "Capability Category I" and "Not Met" assessments from the 2009 Peer Review [5].

o Of the 12 SRs assessed as "Met Capability Category I":

8 SRs (~67%) remain as "Met Capability Category I" o Of the 57 SRs assessed as "Not Met":

17 (~30%) remain as "Not Met"

In 2019, an update to the 2017 model (CL117A) is currently in-progress. The objective of the updated model (CL117B) is to resolve all open F&Os that were not assessed as "resolved" during the December 2018 F&O Closure. Once CL117B is completed, a subsequent F&O Closure will be scheduled.

3.3 GAP ASSESSMENT EPRI Report TR-1018427 [3] provides guidance on the PRA Standard Capability Category necessary to support RI-ISI applications. This report received a Safety Evaluation (SE) from the NRC in January 2012. Regulatory Guide 1.200 (Rev. 2) considers it a good practice to generally have the SRs meet at least a Capability Category II assessment for applications.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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However, according to this EPRI report, not all SRs require Capability Category II to adequately support RI-ISI applications.

Table 3-1 summarizes the minimum capability requirements for each SR as well as a relationship matrix between the SRs from PRA Standard at the time of the EPRI report's publication (i.e., ASME RA-Sb-2005 [9]) and the SRs from the current PRA Standard (i.e.,

ASME / ANS RA-Sa-2009 [1]) which was used during the Clinton Peer Review in 2009.

Using the minimum capability requirements documented in Table 3-1, a gap assessment was performed against the latest PRA model (CL117A) [6] and the F&O Closure results [5] in order to identify any gaps that could be risk-significant for the RI-ISI application.

Table 3-2 summarizes the gap assessment performed for this application. The table is organized by F&O that does not meet the minimum capability requirement specified in Table 3-1.

For the open F&Os identified with a potential impact on the RI-ISI application, a sensitivity analysis was performed to assess the potential impact of those F&Os on the RI-ISI application.

Appendix A summarizes the model changes associated with each sensitivity analysis performed in support of this application.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

AS-A7 AS-A7 Met CC I-II Met CC I-II 1-19 CC I-II Yes AS-A8 AS-A8 Met CC I-III Met CC I-III 1-19 CC I-III Yes AS-A8 AS-A8 Met CC I-III Met CC I-III 3-8 CC I-III Yes AS-A9 AS-A9 Met CC II Met CC II 1-21 CC I Yes AS-A9 AS-A9 Met CC II Met CC II 6-8 CC I Yes AS-B3 AS-B3 Not Met Not Met 1-21 CC I-III No AS-B3 AS-B3 Not Met Met CC I-III 1-22 CC I-III Yes AS-C3 AS-C3 Not Met Met CC I-III 1-13 None Yes DA-B1 DA-B1 Met CC I Met CC II 4-10 CC I Yes DA-B1 DA-B1 Met CC I Met CC II 4-11 CC I Yes DA-B2 DA-B2 Met CC I-II Met CC I-II 4-12 CC I-II Yes DA-C10 DA-C10 Not Met Not Met 5-2 CC I No DA-C11 DA-C11 Not Met Met CC I-III 4-2 CC I-III Yes DA-C14 DA-C13 Not Met Met CC I-III 4-18 CC I-III Yes DA-C14 DA-C13 Not Met Met CC I-III 6-7 CC I-III Yes DA-C15 DA-C14 Not Met Not Met 1-27 CC I-III No DA-C6 DA-C6 Not Met Not Met 5-2 CC I-III No DA-C7 DA-C7 Met CC I Met CC I 5-3 CC I Yes DA-C8 DA-C8 Met CC I - 5-4 CC I Yes DA-D6 DA-D6 Met CC I Met CC I 1-42 CC I Yes DA-D6 DA-D6 Met CC I Met CC II 4-21 CC I Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

DA-E1 DA-E1 Met CC I-III Met CC I-III 1-42 CC I-III Yes DA-E1 DA-E1 Met CC I-III Met CC I-III 4-11 CC I-III Yes DA-E1 DA-E1 Met CC I-III Met CC I-III 4-12 CC I-III Yes DA-E1 DA-E1 Met CC I-III Met CC I-III 4-18 CC I-III Yes DA-E1 DA-E1 Met CC I-III Met CC I-III 4-2 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 1-42 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 4-11 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 4-12 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 4-18 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 4-2 CC I-III Yes DA-E2 DA-E2 Met CC I-III Met CC I-III 4-9 CC I-III Yes DA-E3 DA-E3 Not Met Met CC I-III 1-13 None Yes HR-A1 HR-A1 Not Met Met CC I-III 2-15 CC I-III Yes HR-A1 HR-A1 Not Met Not Met 3-13 CC I-III No HR-A2 HR-A2 Not Met Met CC I-III 2-15 CC I-III Yes HR-A2 HR-A2 Not Met Not Met 3-13 CC I-III No HR-A3 HR-A3 Not Met Met CC I-III 2-14 CC I-III Yes HR-B2 HR-B2 Not Met Met CC I-III 2-16 CC I-III Yes HR-C2 HR-C2 Met CC I Met CC I 6-9 CC I Yes HR-D2 HR-D2 Met CC I Met CC I 5-11 CC I Yes HR-D3 HR-D3 Met CC I Met CC I-III 5-10 CC I Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

HR-D4 HR-D4 Not Met Met CC I-III 5-10 CC I-III Yes HR-D5 HR-D5 Not Met Met CC I-III 5-10 CC I-III Yes HR-D6 HR-D6 Not Met Met CC I-III 5-7 CC I-III Yes HR-G6 HR-G6 Not Met Met CC I-III 1-31 CC I-III Yes HR-G7 HR-G7 Not Met Met CC I-III 1-32 CC I-III Yes HR-G7 HR-G7 Not Met Met CC I-III 1-33 CC I-III Yes HR-G7 HR-G7 Not Met Met CC I-III 1-34 CC I-III Yes HR-H3 HR-H3 Not Met Met CC I-III 1-32 CC I-III Yes HR-H3 HR-H3 Not Met Met CC I-III 1-33 CC I-III Yes HR-H3 HR-H3 Not Met Met CC I-III 1-34 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 1-31 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 2-14 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 2-16 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 5-10 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 5-11 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 5-7 CC I-III Yes HR-I1 HR-I1 Met CC I-III Met CC I-III 6-9 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 1-31 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 2-14 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 2-16 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 5-10 CC I-III Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 9 of 37)

Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

HR-I2 HR-I2 Met CC I-III Met CC I-III 5-11 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 5-7 CC I-III Yes HR-I2 HR-I2 Met CC I-III Met CC I-III 6-9 CC I-III Yes HR-I3 HR-I3 Not Met Met CC I-III 1-13 None Yes IE-A2 IE-A2 Met CC I-III Met CC I-III 1-28 CC I-III Yes IE-A5 IE-A4 Not Met Met CC II 1-2 CC I Yes IE-A5 IE-A4 Not Met Met CC II 1-3 CC I Yes IE-A6 IE-A4a Not Met Met CC II 1-4 CC I Yes IE-A8 IE-A6 Met CC I Met CC II 1-6 CC I Yes IE-A9 IE-A7 Met CC I Met CC III 1-7 CC I Yes IE-C14 IE-C12 Not Met Met CC I-II 2-10 CC I-II Yes IE-C14 IE-C12 Not Met Not Met 2-7 CC I-II No IE-C14 IE-C12 Not Met Met CC I-II 5-12 CC I-II Yes IE-C2 IE-C1a Met CC I-III Met CC I-III 1-10 None Yes IE-C3 IE-C1b Not Met Met CC I-III 1-17 CC I-III Yes IE-C5 IE-C3 Met CC I-II Met CC I-II 1-10 None Yes IE-C5 IE-C3 Met CC I-II Met CC I-II 1-11 None Yes IE-D3 IE-D3 Not Met Met CC I-III 1-13 None Yes IFEV-B3 IF-F3 Not Met Met CC I-III 1-13 CC I-III Yes IFPP-B3 IF-F3 Not Met Met CC I-III 1-13 CC I-III Yes IFQU-B3 IF-F3 Not Met Met CC I-III 1-13 CC I-III Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 10 of 37)

Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

IFSN-A6 IF-C3 Not Met Met CC I-II 1-14 CC II Yes IFSN-A8 IF-C3b Met CC II Met CC II 1-36 CC I Yes IFSN-B3 IF-F3 Not Met Met CC I-III 1-13 CC I-III Yes IFSO-B3 IF-F3 Not Met Met CC I-III 1-13 CC I-III Yes LE-C10 LE-C8b Met CC I Met CC I 1-21 CC I Yes LE-C10 LE-C8b Met CC I Met CC I 1-47 CC I Yes LE-C10 LE-C8b Met CC I Met CC I 5-12 CC I Yes LE-C10 LE-C8b Met CC I Met CC II 5-13 CC I Yes LE-C11 LE-C9a Met CC II-III Met CC II-III 1-21 CC I Yes LE-C12 LE-C9b Met CC I Met CC I 1-47 CC I Yes LE-C3 LE-C2b Met CC I Met CC I 1-43 CC I Yes LE-C3 LE-C2b Met CC I Met CC I 3-18 CC I Yes LE-C5 LE-C4 Met CC II Met CC II 1-48 CC I Yes LE-C7 LE-C6 Met CC I-III Met CC I-III 3-17 CC I-III Yes LE-C9 LE-C8a Met CC II-III Met CC II-III 1-21 CC I Yes LE-C9 LE-C8a Met CC II-III Met CC II-III 5-12 CC I Yes LE-D4 LE-D3 Not Met Met CC II 2-10 CC I Yes LE-D4 LE-D3 Not Met Not Met 2-7 CC I No LE-D4 LE-D3 Not Met Not Met 5-12 CC I No LE-E1 LE-E1 Not Met Met CC I-III 1-13 CC I-III Yes LE-E1 LE-E1 Not Met Not Met 1-27 CC I-III No

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 11 of 37)

Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

LE-E1 LE-E1 Not Met Met CC I-III 1-32 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-33 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-34 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-35 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-37 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-38 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-41 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 1-42 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 2-11 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 2-14 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 2-15 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 2-16 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 2-18 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 3-11 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 3-13 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-10 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-11 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-12 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-15 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-16 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-18 CC I-III Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

LE-E1 LE-E1 Not Met Met CC I-III 4-19 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-2 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-20 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-21 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 4-3 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 5-10 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 5-11 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 5-2 CC I-III Yes LE-E1 LE-E1 Not Met Not Met 5-3 CC I-III No LE-E1 LE-E1 Not Met -- 5-4 CC I-III No LE-E1 LE-E1 Not Met Met CC I-III 5-7 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 6-7 CC I-III Yes LE-E1 LE-E1 Not Met Met CC I-III 6-9 CC I-III Yes LE-E4 LE-E4 Met CC I-III Met CC I-III 1-32 CC I-III Yes LE-E4 LE-E4 Met CC I-III Met CC I-III 1-33 CC I-III Yes LE-E4 LE-E4 Met CC I-III Met CC I-III 1-34 CC I-III Yes LE-E4 LE-E4 Met CC I-III Met CC I-III 1-43 CC I-III Yes LE-E4 LE-E4 Met CC I-III Met CC I-III 2-3 CC I-III Yes LE-F2 LE-F1b Not Met Met CC I-III 6-10 CC I-III Yes LE-F3 LE-F3 Not Met Met CC I-III 1-13 None Yes LE-G4 LE-G4 Not Met Met CC I-III 1-13 None Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

LE-G5 LE-G5 Not Met Not Met 3-22 CC I-III No LE-G6 LE-G6 Not Met Met CC II 1-48 CC I-III Yes QU-A5 QU-A4 Not Met Met CC I-III 1-32 CC I-III Yes QU-A5 QU-A4 Not Met Met CC I-III 1-33 CC I-III Yes QU-A5 QU-A4 Not Met Met CC I-III 1-34 CC I-III Yes QU-A5 QU-A4 Not Met Not Met 2-18 CC I-III No QU-B2 QU-B2 Not Met Met CC I-III 2-3 CC I-III Yes QU-B3 QU-B3 Not Met Not Met 1-49 CC I-III No QU-C1 QU-C1 Not Met Met CC I-III 1-32 CC I-III Yes QU-C1 QU-C1 Not Met Met CC I-III 1-33 CC I-III Yes QU-C1 QU-C1 Not Met Met CC I-III 1-34 CC I-III Yes QU-C2 QU-C2 Not Met Met CC I-III 1-32 CC I-III Yes QU-C2 QU-C2 Not Met Met CC I-III 1-33 CC I-III Yes QU-C2 QU-C2 Not Met Met CC I-III 1-34 CC I-III Yes QU-D1 QU-D1a Not Met Met CC I-III 1-43 CC I-III Yes QU-D4 QU-D3 Met CC I Met CC II-III 1-45 CC I Yes QU-D5 QU-D4 Not Met Met CC I-III 1-46 CC I-III Yes QU-E1 QU-E1 Not Met Met CC I-III 1-13 None Yes QU-F3 QU-F3 Not Met Met CC II-III 4-23 CC I Yes QU-F4 QU-F4 Not Met Met CC I-III 1-13 None Yes QU-F6 QU-F6 Not Met Met CC I-III 4-23 CC I-III Yes

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-1 MINIMUM CAPABILITY REQUIREMENTS FOR EACH SUPPORTING REQUIREMENT NOT MET WITH AT LEAST A CAPABILITY CATEGORY II(1)

Current Previous Supporting Supporting Peer Review Ind. Review Team Minimum CC Requirement(2) Requirement(3) Assessment(4) Assessment(5) F&O Requirement(6) Minimum Met?(7)

SC-B4 SC-B4 Met CC I-III Met CC I-III 3-10 CC I-III Yes SC-B5 SC-B5 Not Met Met CC I-III 4-5 CC I-III Yes SC-C3 SC-C3 Not Met Met CC I-III 1-13 None Yes SY-A13 SY-A12b Not Met Not Met 1-30 CC I-III No SY-A18 SY-A17 Not Met Not Met 1-24 CC I-III No SY-A20 SY-A18a Not Met Met CC I-III 6-7 CC I-III Yes SY-A22 SY-A20 Not Met Not Met 1-26 CC I No SY-A24 SY-A22 Not Met Not Met 1-27 CC I-III No SY-A5 SY-A5 Met CC I-III -- 1-25 CC I-III No SY-A9 SY-A10 Met CC I-III Met CC I-III 3-12 CC I-III Yes SY-B14 SY-B15 Met CC I-III Met CC I-III 1-21 CC I-III Yes SY-B14 SY-B15 Met CC I-III Met CC I-III 1-22 CC I-III Yes SY-B6 SY-B6 Not Met Not Met 6-8 CC I-III No SY-C3 SY-C3 Not Met Met CC I-III 1-13 None Yes Notes to Table 3-1:

(1)

The highlighted rows indicate the Supporting Requirements (SRs) that do not meet the minimum capability requirements specified in EPRI 1018427. These SRs and associated F&Os are further discussed in Table 3-2.

(2)

The "current" Supporting Requirement designations are based on the current ASME / ANS PRA Standard (RA-Sa-2009 [1]).

(3)

The "previous" Supporting Requirement designations are based on the previous ASME PRA Standard (RA-Sb-2005 [9]), which is referenced in EPRI 1018427.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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(4)

The 2009 Peer Review results are documented in Reference [4].

(5)

The 2018 F&O Closure results are documented in Reference [5].

(6)

The minimum capability requirements are specified in EPRI 1018427.

(7)

If the minimum capability requirement is met for the Clinton model, then no further actions are required for this application. If the minimum capability requirement is not met, then the identified gap is further discussed in Table 3-2.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-21 Nearly all event trees model success of ECCS pump LE-C10; The new basic event that captures the failure of ECCS Non-Significant Impact operation following a loss of suppression pool cooling if LE-C11; equipment after venting is not in the PRA model of record. A sensitivity analysis was containment venting is successful. No evaluation of the SY-B14; The staff states that it was inadvertently removed during other performed with the inadvertently ability of ECCS pumps to operate at post-containment- AS-B3; AS- maintenance activities. removed basic events (1SY--

venting temperatures was provided. If ECCS pump A9; LE-C9 STEAMBOUND- and operation for such conditions cannot be supported, then RECOMMENDATION - Re-insert the necessary event(s) into 1CVOPVENTCTRLH--)

additional core damage sequences could result. the model. reinserted into the model.

(This F&O originated from SR AS-B3)

Agree that this is Maintenance. Based on the results of the sensitivity analysis, the increase AS-B3 remains NOT MET (see also F&O 1-22) in risk is less than 2%, so this LE-C10 no impact (see also F&Os 1-47, 5-12, 5-1) model correction does not LE-C11 remains MET CC-II-III significantly impact the RI-ISI SY-B14 remains MET CC I-III application.

AS-A9 remains MET CC-II LE-C9 remains MET CC II-III See Appendix A for further details regarding the sensitivity analysis.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 17 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-24 From a review of system notebooks, it appears that no SY-A18 F&O 1-24 is Partly Resolved, with a Recommendation. The No Impact concerted attempt was made to identify and model RCIC and MS/CCW system notebooks were reviewed and the A systematic review of the accident conditions that could cause system failures. For errors noted in the F&O were found to be corrected. assumptions documented in the example, in the RCIC notebook, Assumption 6 states the Assumption #6 in RCIC was the one updated, however system notebooks is not water leg pump is only needed if AC power is available. Section 2.1 still states "The water leg pump must operate expected to identify any risk-Assumption 17 says that the water leg pump is not continuously to assure RI operability" and confirmed modeling significant discrepancies. Since needed and the fault tree gate UGATE06 included water of spurious temp signal for RCIC, e.g., event 1RITS--N604E-F- this is primarily a documentation leg pump failure as a failure of the RCIC system. Also, -. While the notebooks are reviewed as a part of each model issue, there is no impact on the the high steam flow isolation is not addressed in the RCIC update, no systematic review of the notebooks was performed RI-ISI application.

notebook nor is isolation based on area temperature to address this concern.

sensors. The RCIC gland seal compressor is included as a failure mode for RCIC, but it is not clear from the RECOMMNEDATION - It is recommended that a systematic documentation why that failure mode is included. review of the system notebooks should be performed to Another example is the circulating water system. No ensure this issue has been completely resolved.

mention is made of the circulating water pump trip in high level in the turbine building. Because of the inconsistent Agree that this is model maintenance.

documentation and above mentioned omissions from the documentation, and further because no evidence of a SY-18 is NOT MET Cat I-III review for such conditions was identified in the documentation, it appears that this condition is not an isolated error but a complete omission from the analysis.

Because of these issues this SR is assessed as not met.

(This F&O originated from SR SY-A18)

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 18 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-25 The potential for water hammer-induced failure of ECCS SY-A5 N/A - Not reviewed. Potentially Significant Impact systems when aligned for testing and coincident with a A sensitivity analysis was loss of offsite power is not addressed. If an ECCS performed with the inclusion of system is aligned for test and a LOSP occurs, then water hammer scenarios.

portions of the system will drain in the period between the loss of power and repowering the buses. While the Based on the results of the potential for a LOSP following an initiating event is sensitivity analysis, the increase addressed and maintenance alignments are included in in risk is approximately 5% for the models, the water hammer that could result are not CDF and 0.6% for LERF, so this assessed. model refinement represents a potentially significant impact to the RI-ISI application.

See Appendix A for further details regarding the sensitivity analysis.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 19 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-26 No calculations supporting room cooling requirements are SY-A22 The Dependency Notebook provides details and references for Potentially Significant Impact documented. No evaluation of control room HVAC was the room cooling calculations where applicable. However, Regarding Item #1 of the identified. Better documentation of analysis to several issues were identified in the implementation, so the Independent Review Team's demonstrate that rated or design capabilities are not F&O remains only partially resolved. assessment for this F&O, a exceeded is needed within the system notebooks. Identified Issues: sensitivity analysis was (This F&O originated from SR SY-A22) 1. The RHR room cooling logic under gate RHRA-PMP- performed with the common RMCLG does not appear to handle the CCF fan terms cause failures of the room consistently with the independent terms, which are modeled to cooling fans treated consistently prevent the door opening action from creating success. This with the independent failures.

should be corrected and checked in other systems that use a similar approach. Based on the results of the

2. The SX System Notebook and the Dependency Notebook sensitivity analysis, the increase were not properly updated regarding the room cooling in risk is approximately 6% for requirements for SX. Specifically, CL-PRA-006 discusses CDF and approximately 1% for pump A/B/C with B&C requiring cooling, while CL-PRA-005.12 LERF, so this model correction discusses division 1/2/3 with 1&2 requiring room cooling, while represents a potentially the model appears to only require cooling for division 2. This significant impact to the RI-ISI issue is only a documentation issue. application.

3. The 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months assumption for RH pumps needs more justification (or modification) since the EQ temp is only 3F See Appendix A for further above the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months temp. details regarding the sensitivity

4. The basis for using RH B room cooling calculations as the analysis.

basis for ECCS room cooling requirements requires additional justification (or modification). Regarding Items #2-4 of the Independent Review Team's Agree that this in Maintenance. assessment for this F&O, the issues are primarily SY-A22 is NOT MET CC I-II documentation issues, so there is no impact on the RI-ISI application.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 20 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-27 The CPS PRA models one instance where credit is given LE-E1; DA- Reviewed the FPIE Component Data Notebook, CL-PRA-010 Non-Significant Impact for repair of hardware failures, repair of RHR. Per the C15; SY- REV. 4, Appendix G Additional Data Assessments, Section A sensitivity analysis was requirements of RG 1.200, plant-specific data to quantify A24 G.14, and the Fire PRA Plant Response Model Notebook, and performed using a plant-specific the probability of repair should be collected and analyzed CL-PRA-021.05 REV. 1, Appendix F - Generic Recovery RHR recovery failure probability.

to support any credit given. The analysis for CPS uses Justification.

generic industry data. Therefore, this SR is assessed as Based on the results of the not met. Since the finding specified collecting plant-specific data for the sensitivity analysis, the increase (This F&O originated from SR SY-A24) RHR repair time and that is now covered in the Fire PRA PRM in risk is less than 2%, so this notebook Appendix F, the FPIE Component Data Notebook model correction does not should utilize that MTTR analysis and include it as part of significantly impact the RI-ISI Appendix. application.

It is agreed that this is model maintenance. See Appendix A for further details regarding the sensitivity DA-C15 remains NOT MET analysis.

LE-E1 remains NOT MET CC I-III SY-A24 remains NOT MET CC I-III

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 21 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-30 Flow diversion criterion or discussion otherwise is SY-A13 The comparison of flow diversions to the MSO results does not No Impact provided in the "Assumption" sections of the individual provide assurance unless the MSO analysis uses required For each system notebook, the system notebooks. All flow diversion pathways that divert flow criteria. The basis for the MSO analysis is not provided, assumptions were reviewed to less than ten-percent of system flow are categorically so it is not a sufficient justification. identify the systems that utilize excluded without consideration of the flow actually the "less than 10% flow diverted" required. RECOMMENDATION - Instead of the current criteria, evaluate exclusion criterion. This review (This F&O originated from SR SY-A13) each potential flow diversion using required flow criteria and identified the following systems:

document each decision. - RHR

- LPCS Agree that this is model maintenance. - RCIC

- HPCS SY-A13 remains NOT MET CC I-III - SX (Shutdown Service Water)

Upon further review of these systems, the flow diversion pathways that were excluded using the "less than 10% flow diverted" exclusion criterion can be further justified by comparing the required flowrate for system success (as specified in the applicable plant-specific MAAP cases) and the diverted flowrate through the pathway of concern.

Therefore, this issue is primarily a documentation issue, so there is no impact on the RI-ISI application.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 22 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 1-49 As described in the PRA standard, convergence can be QU-B3 Reviewed Quantification Notebook and Appendix D of CL- No Impact considered sufficient when successive reductions in MISC-015 CPS F&O Closure Summary Notebook Rev 0, the Although the "less than 5%

truncation value of one decade result in decreasing Clinton Summary Report. It is recognized that computer change per decade" example changes in CDF or LERF, and the final change is less limitations prevent solution at a lower truncation level. criterion is not met for the CDF than 5%. The truncation study in Section 3.1.2 of CPS- However, the graphs presented in CL-PRA-013 Summary and LERF results, convergence PSA-013 clearly demonstrates that this convergence has 13264 FINAL as Tables 3.7-1 and 3.7-2 are only just starting can be demonstrated by showing not occurred. For example, in Table 3.1-4 of CPS-PSA- to flatten out. This suggests that at the currently solvable that no significant accident 013, there is a 20.5% difference in CDF between 1E-11 cutoff CDF and LERF could be underestimated. sequences are inadvertently and 1E-12 truncation limits. Similarly, convergence of RECOMMENDATION - It is recommended that an alternative eliminated during the LERF was not demonstrated with a 29% increase from a method could be used to estimate convergence or that the quantification process. A review truncation of 1.E-11 to 1E-12. additional frequency of CDF and LERF be estimated. of the CL117A CDF and LERF (This F&O originated from SR QU-B3) importances demonstrate that no It is agreed that this is model maintenance. risk-significant accident sequences are inadvertently QU-B3 remains NOT MET eliminated between the truncation limits of 5E-13/yr (which is used in the base model) and 5E-14/yr.

Also, the CDF and LERF results at the 5E-13/yr truncation are consistent with the NEI PRA Peer Review Guidelines which indicate that a truncation of four orders of magnitude below the risk metric is adequate for a high-quality PRA. Since Clinton's CDF and LERF values are approximately six orders of magnitude higher than the truncation limit, the results are appropriate.

Since this is primarily a documentation issue, there is no impact on the RI-ISI application.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 23 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 2-7 The initiating event frequency values appear to be much IE-C14; LE- The NSAC-154 guidance is of similar vintage (1991) as the Non-Significant Impact lower than others in the industry and are developed D4 NUREGs mentioned, so it appears to be generally up-to-date. A sensitivity analysis was based on old data. Furthermore, the frequency While recovery actions are still discussed in the performed to account for "state-calculations do not appear to consider correlated data documentation, the ISLOCA frequencies used in the model, of-knowledge" uncertainties which is recommended in recent industry documents which are only the 'Rupture' cases, do not credit any recovery associated with the ISLOCA such as NUREG/CR-5744 and NUREG/CR-5124. Also, actions. They consist of Interface Failure x Piping Rupture, analysis, as recommended by the recovery check valve failures for the valves that have with the Early Isolation Failure set to 1.0. Independent Review Team's failed to cause the initiating event is credited in the However, the issue of addressing correlated failures is a recommendation.

analysis. No basis for the credit is provided. newer issue that is generally expected in all ISLOCA analyses (This F&O originated from SR IE-C14) under the PRA Standard. The uncertainty distributions applied to the various component RECOMMENDATION - One way to address the correlation of failure modes modeled in the failures is to use UNCERT (ensuring that the basic events overpressurization fault trees are have correlated uncertainties) to identify the mean frequency evaluated using the UNCERT of each ISLOCA event for input into the mode. software to determine the final ISLOCA initiating event It is agreed that this is model maintenance. frequencies used in the PRA.

IE-C14 remains NOT MET Based on the results of the LE-D4 remains NOT MET sensitivity analysis, the increase in risk is less than 1%, so this model correction does not significantly impact the RI-ISI application.

See Appendix A for further details regarding the sensitivity analysis.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 24 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 2-18 It is unclear how operator recovery actions are included in HR-H1; HR- The FPIE HRA Notebook CL-PRA-004 REV. 5 was reviewed. No Impact the model. Where is the documentation for evidentiary I1; LE-E1; As stated in the Finding, Section 2.3.4 Recovery Actions A table summarizing the recovery confirmation that EOPs and APs, etc. were reviewed for HR-H2; HR- Added to Cut Sets states that "No 'recovery actions' other than events credited in the PRA is a potential recovery actions. I2; LE-E4; offsite power recovery...were included after the model cut sets documentation issue only.

QU-A5 were generated. Rather, successes and failures of actions as Therefore, there is no impact on The CPS HRA documentation states that successes and directed by the Clinton EOPs or auxiliary procedures were the RI-ISI application.

failures of actions as directed by the Clinton EOPs or evaluated for inclusion in the event tree and fault tree models auxiliary procedures were evaluated for inclusion in the directly."

event tree and fault tree models directly. If the actions Generally recovery actions are identified during cutset reviews are in the model as stated, the events are very difficult to as additional actions that could assist in mitigating accident trace from within the documentation. scenarios and therefore risk. At that point, procedures are (This F&O originated from SR HR-H1) reviewed to determine whether such actions are already proceduralized at the plant and can be credited. The discussion in Section 2.3.4 seems to indicate that such recovery actions were added to the event trees and fault trees but there is no listing in the HRA Notebook as to what these actions were and what HFEs were credited (hence the confusion on the part of the original reviewer who made the Finding).

Add a table to Section 2.3.4 listing the recovery actions that were added to the Event Trees or as HFEs in the Fault Trees as a result of cutset reviews.

It is agreed that this is model maintenance.

HR-H1 remains MET CC II HR-H2 remains MET CC II HR-I1 remains MET CC I-III HR-I2 remains MET CC I-III LE-E1 is MET CC I-III LE-E4 remains MET CC I-III QU-A5 remains NOT MET

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 25 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 3-13 Per Section 2.3.2.1.1 of CPS-PSA-004, the pre-initiators HR-A1; LE- The FPIE HRA Notebook CL-PRA-004 REV. 5 was reviewed. No Impact were identified by ' it was judged that the best use of E1; HR-A2 Since system and procedure familiarization were performed This is primarily a documentation resources is to identify the leading candidates for risk and operating experience was reviewed, it is believed that the issue associated with the Human significant pre initiators and not to expend substantial pre-initiator identification performed as discussed in Appendix Reliability Analysis Notebook.

effort to quantify other preinitiator HEPs for which no data J meets the requirements of SR HR-A1. However, Section Therefore, there is no impact on are available and for which the contribution can be 2.3.2.1.1 on Random Pre-Initiator Human Interactions casts the RI-ISI application.

expected to be small.' The section states a methodology doubt on the analysis with the subsection identified in the

'when resource become available', the identification Finding that begins "When resources become available, should be performed in a way that would meet the SR. activities that support this examination process should (This F&O originated from SR HR-A1) include:" Recommend deleting this subsection since it introduces confusion rather than clarity.

Agree that this is model maintenance.

HR-A1 remains NOT MET CC I-III HR-A2 remains NOT MET CC I-III LE-E1 is MET CC I-III 3-22 Limitations in the LERF analysis that would impact LE-G5 This F&O remains open. CL-PRA-015 was reviewed. The No Impact applications were not identified. new Section 7.4 added to the Level 2 Notebook (CL-PRA-015) This is primarily a documentation (This F&O originated from SR LE-G5) provides a place to document model limitations but no actual issue associated with the Level 2 model limitations were identified. The items discussed as Notebook. A systematic review limitations are actually boundary conditions defining the model of the assumptions associated scope. Review of the complete report identified a few with the development of the Level scattered assumptions which could potentially result in model 2 model was performed an no limitations (no Assumptions Section was found in the limitations were identified for use notebook). in future applications. Therefore, RECOMMENDATION - Add an Assumptions section to the there is no impact on the RI-ISI report. Identify model limitations based on the modeling application.

assumptions made in the analysis.

Agree that this is model maintenance.

LE-G5 remains NOT MET CC I-III

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 26 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 5-2 Demand and run-time data is documented in Table C.2-3. LE-E1; DA- Reviewed the demand and run-time data in Appendix C of the No Impact The data sources used were provided in the table. A C10; DA-C6 Component Data Notebook (CL-PRA-010, Vol. 1, Rev. 4) to This is primarily a documentation footnote in the table indicates that plant-specific data check that it is based on actual plant experience (when issue associated with the were not used. Some of the data is based on the MSPI available) and includes surveillance tests and maintenance Component Data Notebook.

basis document which should provide a basis for the events. The Component Data Notebook identified the source Therefore, there is no impact on period of the basis document. However, data was of data as the Maintenance Rule program, but the the RI-ISI application.

extrapolated beyond the period of time considered in the Maintenance Rule program data provided (a spreadsheet) did basis document. No consideration of surveillance tests, not show that each of the failure modes listed in the SR has maintenance acts, etc., was documented. There was no been addressed.

accounting for the exclusion of post maintenance demands. RECOMMENDATION - Add documentation describing how each of The failure modes listed in the SR are addressed by the empirical data and provide a reference to the specific MR report used.

Agree that this is model maintenance.

DA-C6 remains NOT MET DA-C10 remains NOT MET LE-E1 is MET CC I-III

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 27 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 5-3 Actual plant experience, practices and plant maintenance LE-E1; DA- Reviewed the test and maintenance data portion of Appendix No Impact plans were not used to base the number of surveillance C7 F of the Component Data Notebook (CL-PRA-010, Vol. 1, Rev. This is primarily a documentation tests and maintenance acts. 4), specifically the Train Availability data portion of several issue associated with the (This F&O originated from SR DA-C7) components. The Component Data Notebook identified the Component Data Notebook.

source of data as the Maintenance Rule program, but the Therefore, there is no impact on Maintenance Rule program data provided (a spreadsheet) did the RI-ISI application.

not show that the test and maintenance surveillance frequencies were included.

RECOMMENDATION - Add documentation describing how the Maintenance Rule data addresses the plant surveillance requirements and unplanned maintenance, and provide a reference to the specific MR report used.

Agree that this is model maintenance.

DA-C7 remains MET CC I LE-E1 remains NOT MET CC I-III 5-4 Plant-specific records were not used to determine the LE-E1; DA- N/A - Not reviewed. Non-Significant Impact time that components were configured in their standby C8 System Manager interviews were status. used to determine the time that (This F&O originated from SR DA-C8) components were configured in their standby status. Review of plant-specific data for the standby configurations is not expected to change probabilities used in the PRA model. Any changes to the standby probabilities would slightly change the ranking of the different trains of the system.

Therefore, there any change in the PRA results will likely be non-significant and would not impact the RI-ISI application.

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 28 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 5-12 Credit has been taken for the capability to close valves LE-C10; IE- The FPIE ISLOCA Notebook CL-PRA-011 REV. 3 and the No Impact against ISLOCA dynamic loads. The evaluation of C14; LE- Initiating Events Notebook (CL-PRA-001) were reviewed. This is primarily a documentation closure capability only evaluated normal loads and delta D4; LE-C9 The IE Notebook section 2.4.9 states that: "Failure of the issue. The ISLOCA initiating pressures. isolation interface (check valves and MOVs) between LPCS or events only model the rupture (This F&O originated from SR IE-C14) LPCI and the RCS during power has the potential to result in failure mode, which never failure of the RCS boundary and bypass of the primary credited the operator action for containment; this issue is addressed under the ISLOCA isolating the MOV. As mentioned initiator category." in the Independent Review The ISLOCA Notebook, Appendix E - ISLOCA Early Isolation, Team's assessment, there is no section E.1 "Analysis" states that "MOVs that failed due to quantitative impact on the PRA leakage or rupture (including dynamic failure to hold) may not results. Therefore, there is no be used to isolate the ISLOCA." impact on the RI-ISI application.

However, the Event Trees E-1 and E-2 contain a probability for the MOV failure based on the fraction of the overpressurization frequency involving a mispositioned or spuriously opened MOV, which in turn is based on a percent of cutsets containing interface leakage or rupture of the MOV.

Response from PRA team: A summary was provided of how the ISLOCA frequencies are calculated, as summarized in Section 5.3 of the ISLOCA Notebook [shows that the MOV is no longer credited].

Although Event Trees E-1 & E-2 are no longer used quantitatively in the FPIE PRA, the figures and calculations should be updated to reflect that operator action of the MOVs is not feasible during an ISLOCA incident.

Agree that this is model maintenance.

IE-C14 is ET CC I-II LE-C9 remains MET CC II-III LE-C10 remains MET CC I LE-D4 remains NOT MET

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 29 of 37)

Table 3-2 CLINTON GAP ASSESSMENT FOR RI-ISI APPLICATION Applicable F&O F&O Description SR(s) Independent Review Assessment from F&O Closure RI-ISI Impact 6-8 ECCS Room Cooling applications document the need for SY-B6; AS- The Dependency Notebook provides details and references for No Impact room cooling. Credit is provided for the opening of doors A9 the room cooling calculations where applicable. However, the This is primarily a documentation (LPCS pump room for example). No formal engineering basis for using RHR B room cooling calculations as the basis issue associated with the analyses were identified that supports this application. for ECCS room cooling requirements requires additional Dependency Notebook.

(This F&O originated from SR SY-B6) justification (or modification). Therefore, there is no impact on the RI-ISI application.

DOCUMENTATION - The Dependency Notebook requires additional justification of the bases for the room cooling requirements for the other ECCS pump rooms Agree that this is model maintenance.

SY-B6 remains NOT MET AS-A9 remains MET CC-II

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 30 of 37)

4.0 CONCLUSION

S The Clinton PRA maintenance and update processes and technical capability evaluations described in the previous section of this notebook provide a robust basis for concluding that the PRA is suitable for use in Risk-Informed In-Service Inspection (RI-ISI) applications.

This conclusion is based upon the following considerations:

Existing PRA Model Maintenance and Update procedures present robust, yet flexible guidance for maintaining the "as-built, as-operated" configuration of Clinton.

The Gap Assessment performed for this application demonstrates that Clinton's PRA model is technically adequate.

The sensitivity analysis (Appendix A) incorporating model changes addressing the open F&Os show that the base model remains adequate for this application.

In support of the PRA analyses for the Clinton 10-year interval evaluation using the CL117A PRA model, a gap assessment was performed and sensitivity analyses were developed for potentially risk-significant gaps that could impact the RI-ISI application.

As documented in Appendix A, resolving each individual gap, when assessed independently from one another, does not increase the overall risk results substantially (i.e., largest increase is associated with F&O 1-26, which increases CDF approximately 6%). When the cumulative impacts of all identified gaps are assessed (i.e., all F&O model changes incorporated at once),

the overall impact on CDF is approximately 16% and LERF is approximately 2%. Although this suggests a significant impact on CDF, the sensitivities performed for this application present the conservative / bounding assessments that would be refined as part of a normal PRA update process. Therefore, use of the CL117A model for the RI-ISI application is appropriate.

5.0 REFERENCES

[1] American Society of Mechanical Engineers / American Nuclear Society (ASME / ANS),

"Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application", ASME /

ANS RA-Sa-2009, March 2009

[2] U.S. Regulatory Commission (USNRC), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities",

Regulatory Guide 1.200, Revision 2, March 2009

[3] Electric Power Research Institute (EPRI), "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs", EPRI 1018427, December 2008

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 31 of 37)

[4] Boiling Water Reactor Owners' Group (BWROG), "Clinton Power Station 2009 PRA Peer Review Report Using ASME / ANS PRA Standard Requirement", April 2010

[5] Boiling Water Reactor Owners' Group (BWROG), "Clinton PRA Finding and Suggestion Level Fact and Observation Independent Assessment", 032362-RPT-12, Revision 0, January 2019

[6] Exelon, "Clinton Power Station Probabilistic Risk Assessment Quantification Notebook",

CL-PRA-014, Revision 8, August 2018

[7] Exelon, "Clinton Power Station Probabilistic Risk Assessment Quantification Notebook",

CPS-PSA-014, Revision 7, March 2014

[8] Boiling Water Reactor Owners' Group (BWROG), "BWROG PSA Peer Review Certification Implementation Guidelines", Revision 3, January 1997

[9] American Society of Mechanical Engineers (ASME), "Addenda to ASME RA-S-2002:

Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", ASME RA-Sb-2005, December 2005

[10] U.S. Regulatory Commission (USNRC), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities",

Regulatory Guide 1.200, Revision 1, January 2007

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 32 of 37)

APPENDIX A RI-ISI SENSITIVITY

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 33 of 37)

A.1 Purpose This appendix documents the sensitivity analysis performed in support of the RI-ISI application.

The appendix is organized by F&O and an overall / cumulative assessment is performed for all F&Os.

A.2 F&O 1 Basis for ECCS Pump Operation Following Loss of DHR For CL117A, HFE 1CVOPVENTCTRLH-- was removed from the model and subsumed into the other venting HFEs (i.e., 1CVOPSECT25-6H-- & 1CVOP-SECT27-H--) as an execution step for controlling containment pressure. By removing this action, the event tree logic becomes complicated and resetting the logic to the previous structure is required.

The following model changes, as recommended by the Independent Review Team's recommendation during the F&O Closure, were made for the sensitivity analysis supporting this F&O:

Gate ZZ-VNT-C (AND gate of 1CVOPVENTCTRLH-- & 1SY--STEAMBOUND-)

from the previous PRA model (CL114A) was reinserted into the current PRA model (CL117A) under the following parent gates:

o ZZ-VNT-CS o ZZ-VNT-CS-ML o ZZ-VNT-CS-SL o ZZ-VNT-HP o ZZ-VNT-LPI o ZZ-VNT-LPI-ML o ZZ-VNT-LPI-SL

The following HEP values changed as a result of removing the execution step associated with HFE 1CVOPVENTCTRLH-- from the other venting HFEs:

o 1CVOPSECT25-6H-- 9.10E-03 o 1CVOP-SECT27-H-- 6.10E-03

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 34 of 37)

A.3 F&O 1 Water Hammer The potential for water hammer-induced failure of ECCS systems when aligned for testing and coincident with a loss of offsite power is not addressed in CL117A. If an ECCS system is aligned for test and a loss of offsite power (LOOP) occurs, then portions of the system will drain in the period between the LOOP and repowering the buses. Restarting the pumps could result in a water hammer event that fails the ECCS system.

The following model changes, as recommended by the Independent Review Team's recommendation during the F&O Closure, were made for the sensitivity analysis supporting this F&O:

For the following systems, water hammer scenarios are considered given a LOOP initiating event (including transient-induced LOOPs and LOCA-induced LOOPs) or a failure of the water-leg pump when the system is in standby:

o LPCS o RHR A o RHR B o RHR C o HPCS o RCIC - No LOOP pre-condition o SX A - No water-leg pump o SX B - No water-leg pump o SX C - No water-leg pump A.4 F&O 1 RHR Room Cooling For CL117A, common cause failures of the RHR A & B room cooling fans were inconsistently modeled with the independent failures. Failure of the RHR room cooling fans would prevent successful room cooling even if operators open the room doors.

The following model changes, as recommended by the Independent Review Team's recommendation during the F&O Closure, were made for the sensitivity analysis supporting this F&O:

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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For the following RHR room cooling parent gates, the common cause gate is included in the logic:

o RHRA-PMPRM-FAN Added RGATE107 o RHRA-PMPRM-FAN-RSDP Added RGATE107 o RHRA-RMCLG-FAN Added RGATE107A o RHRB-PMPRM-FAN Added RGATE107 o RHRB-RMCLG-FAN Added RGATE107A A.5 F&O 1 Plant-Specific RHR Recovery Using the Clinton Maintenance Rule Out of Service data for RHR SPC and RHR A & B pumps for the time period of 2011 through 2016 (i.e., 6 years), the mean time to repair (MTTR) was calculated to be 28.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . For Class IIA sequences (without upper pool dump), the allowed time for repair (ATFR) is 23.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . For sequences where upper pool dump is successful, the allowed time for repair increases to 39.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Using these times, the plant-specific failure to recover RHR can be calculated using the following equation:

=

The following model changes, as recommended by the Independent Review Team's recommendation during the F&O Closure, were made for the sensitivity analysis supporting this F&O:

Using the data and equation previously discussed, the plant-specific failure to recover RHR probabilities were updated to the following values:

o 1RHRXDHRRECLTH-- 4.33E-01 o 1RHRX-REC-UPDH-- 2.43E-01 A.6 F&O 2 ISLOCA IE Frequencies The ISLOCA initiating event frequencies were updated to account for State of Knowledge Correlation (SOKC) uncertainties. Using the ISLOCA initiating event fault trees and associated uncertainty distributions, a parametric uncertainty evaluation was performed, and the results of that evaluation are used in this sensitivity analysis.

The following model changes, as recommended by the Independent Review Team's recommendation during the F&O Closure, were made for the sensitivity analysis supporting this F&O:

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 36 of 37)

The following ISLOCA initiating event frequencies were updated to account for SOKC uncertainties:

o %ISLOCA-CS 8.02E-10/yr o %ISLOCA-LPA 8.02E-10/yr o %ISLOCA-LPB 8.02E-10/yr o %ISLOCA-LPC 8.02E-10/yr o %ISLOCA-SDCA 8.02E-10/yr o %ISLOCA-SDCB 8.02E-10/yr o %ISLOCA-SDC 5.85E-09/yr A.7 Cumulative Results A sensitivity analysis was performed to assess the cumulative impact of all of the F&Os with potential impacts on the RI-ISI application previously discussed in this appendix.

Table A-1 summarizes the cumulative results of this sensitivity. As shown in Table A-1, CDF increases approximately 16% and LERF increases approximately 3%.

Table A-1 SENSITIVITY RESULTS FOR CUMULATIVE Metric Base Sensitivity %Delta CDF 2.19E-06 2.54E-06 16%

LERF 1.25E-07 1.29E-07 3%

A.8 Impact on the RI-ISI Weld Selections All of the consequence calculations from the Clinton RI-ISI program were recalculated using the cumulative model (i.e., 134 cases in total were rerun for both CDF and LERF to arrive at revised CCDPs and CLERPs). Only 2 consequence calculations increased the PRA Consequence Rank as a result of the F&O model changes documented in this appendix. Table A-2 summarizes the change in results between CL117A and the sensitivity model (i.e., revised CL117A).

Enclosure 10 CFR 50.55a Relief Request I4R-01 Alternate Risk Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds in Accordance with 10 CFR 50.55a(z)(1)

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

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Table A-2 Consequence Segments Affected by Sensitivity Model Change CL117A Sensitivity Consequence Description CCDP CLERP CCDP CLERP CCDP CLERP LPCI B injection between PRHB-C-03 inside MOV F042B and 186% 0% 7E-07 1E-07 2E-06 1E-07 inside check valve F041B LPCI B line between PRHB-C-04 containment and inside 186% 0% 7E-07 7E-08 2E-06 7E-08 MOV F042B For both consequence groups, the PRA Consequence Rank went from Low to Medium with the sensitivity model. There are no welds susceptible to a damage mechanism among the 74 welds associated with these two segments. Since they went from Low PRA Consequence Rank to Medium Consequence Rank, the segments now move from Low Risk Category into the Medium Risk Category which requires 10% be inspected (i.e., which is equivalent to approximately 7 welds).

A.9 Conclusion A sensitivity model was developed to address the open items from the F&O Closure. That model incorporated six of the open items that are relevant to RI-ISI based on the EPRI guidance. The RI-ISI consequence groups were then rerun using the revised model to assess the impact on the weld selections. The sensitivity model shows a 16% increase in base CDF while LERF only went up slightly with a 3% increase. As a result, only 2 consequence segments changed Consequence Rank, going from Low Consequence to Medium Consequence. There are 74 welds in total among the two groups and none of them are susceptible to a damage mechanism. Since the 74 welds were originally Low Risk Rank, they are now Medium Risk Rank which requires approximately 7 welds be added for inspection. This is a small impact from the remaining open items from the F&O Closure.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 1 of 8)

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

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90 and B3.100

Description:

Alternative Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section (IWB-2500, Table IWB-2500 Inspection Program)

Component Number: Nozzles N1, N2, N3, N5, N6, N7, N8, N9, and N16 (See the Enclosure for a complete list of nozzle identification numbers)

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2013 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2013 Edition is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xiv) and (xviii).

3. Applicable Code Requirement

The applicable requirement is contained in Subsection IWB, Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzle in Vessels - Inspection Program," Class 1 Reactor Vessel nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Numbers B3.90, "Nozzle-to-Vessel Welds," and B3.100, "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval.

All of the nozzles identified in the Enclosure are full penetration welds.

4. Reason for Request

NRC Regulatory Guide (RG) 1.147, Revision 19 conditionally accepts the use of ASME Code Case N-702 (N-702) (Reference 3). This code case provides an alternative to performing examination of 100% of the nozzle-to-vessel welds and inner radii for Examination Category B-D nozzles with the exception of the Feedwater and Control Rod Drive Return Line (CRDRL) nozzles. The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 2 of 8)

N-702 has been approved for use in RG 1.147, Revision 19 with conditions as noted below:

The applicability of Code Case N-702 for the first 40 years of operation must be demonstrated by satisfying the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240).

The use of Code Case N-702 in the period of extended operation is prohibited. If VT-1 is used, it shall utilize ASME Code Case N-648-2, "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1," with associated required conditions specified in Regulatory Guide 1.147.

Note: This code case was previously approved with conditions, the conditions have been revised for Revision 19 of Reg. Guide 1.147.

The analyses in BWRVIP-108NP and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth.

The proposed alternative provides an acceptable level of quality and safety based on the technical content of BWRVIP-108 and BWRVIP-241, as endorsed by the NRC SEs, and the reduction in examination scope could provide a dose savings of as much as 25 Rem for the entire Fourth ISI Interval.

5. Proposed Alternative and Basis for Use In accordance with 10 CFR 50.55a(z)(1), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Tables 5-1 and 5-2 below (see the Enclosure for a list of RPV Examination Category B-D nozzles applicable to this relief request). As an alternative, for all welds and inner radii identified in Table 5-1, CPS proposes to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with N-702. For the nozzle assemblies identified in the Enclosure, this would mean 25 percent from each of the groups identified in Table 5-1 during the 120-month interval.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 5-1 Clinton Power Station, Unit 1 RPV Examination Category B-D Nozzle Summary Minimum Total Comments Group Number to be Number Results1 Examined 20" Recirculation Outlet 2 1 One (1) nozzle was examined in Nozzles the Third ISI Interval.

(N1) No rejectable indications.

10" Recirculation Inlet 10 3 Three (3) nozzles were Nozzles examined in the Third ISI (N2) Interval.

No rejectable indications.

24" Main Steam 4 1 One (1) nozzle is scheduled to Nozzles be examined in the Third ISI (N3) Interval.

No rejectable indications.

12" Core Spray 2 1 One (1) nozzle was examined in Nozzles the Third ISI Interval.

(N5) No rejectable indications.

10" Low Pressure 3 1 One (1) nozzle was examined in Coolant Injection the Third ISI Interval.

Nozzles No rejectable indications.

(N6) 6" Head Spray Nozzles 2 1 Two (2) nozzles were examined (N7 and N8) in the Third ISI Interval.

No rejectable indications.

4" Jet Pump 2 1 One (1) nozzle was examined in Instrumentation the Third ISI Interval.

nozzles No rejectable indications.

(N9)

Vibration 1 1 One (1) nozzle was examined in Instrumentation Nozzle the Third ISI Interval.

(N16) No rejectable indications.

Note:

1. The nozzle-to-vessel weld and inner radius examinations are performed together.

The examinations in Table 5-1 will be scheduled in accordance with ASME Section XI, IWB-2411, "Inspection Program."

N-702 stipulates that a VT-1 visual examination may be used in lieu of the volumetric examination for the nozzle inner radii (i.e., Item Number B3.100, "Nozzle Inside Radius Section"). This VT-1 visual examination is outlined in Code Case N-648-2 (N-648-2)

("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1"). CPS will perform either volumetric examination or VT-1

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 4 of 8) visual examination of the inner radius as required by N-702. (Note, however, that CPS is not currently using N-648-2 and is planning to continue to perform volumetric examinations of all required nozzle inner radii.)

The Electric Power Research Institute (EPRI) Technical Report 1003557 (Reference 1),

"BWRVIP-108 Boiling Water Reactor Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," (Reference 1) found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure (LTOP) event are very low (i.e., < 1 x 10-6 for 40 years) with or without any inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified.

EPRI Report BWRVIP-241 received a final NRC Safety Evaluation Report on April 19, 2013 (ML13071A240). In the NRC Safety Evaluation Report, Section 5.0, "Conditions and Limitations," indicates that each licensee who plans to request relief from ASME Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections may reference the BWRVIP-241 report as the technical basis for the use of N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating that the following general and nozzle-specific criteria are satisfied:

In the case of CPS, the single set of values (e.g., nozzle radii, nozzle thicknesses, etc.)

used in the following equations are correct and applicable to CPS. These values are minimum design values.

Responses to NRC Plant Specific Applicability

1. The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour.

This criterion is met by adherence to CPS Technical Specification (TS) 3.4.11, "Reactor Coolant System Pressure/Temperature Limits," Surveillance Requirement 3.4.11.1 which requires verification that the Reactor Coolant System (RCS) heatup and cooldown rates are limited to less than or equal to 100°F in any one hour period and, less than or equal to 20°F in any one hour period during RPV pressure testing.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

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2. For the Reactor Recirculation Inlet Nozzles (N2), the following criteria must be met:

a. (pr/t)/CRPV 1.15; p=RPV Normal Operating Pressure 1025 psig r=RPV inner radius 110.19 inches t=RPV wall thickness 6.1 inches CRPV= 19332 Result: (pr/t)/CRPV = 0.96 1.15 The calculation for the CPS N2 Nozzle results in a maximum value of 0.96, which is less than 1.15 and satisfies this criteria.

b. [p(ro2 +ri2)/(ro2-ri2)]/CNOZZLE 1.15; p=RPV Normal Operating Pressure 1025 psig ro=nozzle outer radius 11.69 inches ri=nozzle inner radius 5.81 inches CNOZZLE 1637 Result: [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE = 1.04 1.15 The calculation for the CPS N2 Nozzle results in a maximum value of 1.04, which is less than 1.15 and satisfies this criteria.

3. For the Reactor Recirculation Outlet Nozzles (N1), the following criteria must be met:

a. (pr/t)/CRPV 1.15; p=RPV Normal Operating Pressure 1025 psig r=RPV inner radius 110.19 inches t=RPV wall thickness 6.1 inches CRPV= 16171 Result: (pr/t)/CRPV = 1.14 1.15 The calculation for the CPS N1 Nozzle results in a value of 1.14, which is less than 1.15 and satisfies the criteria.

b. [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE 1.15; p=RPV Normal Operating Pressure 1025 psig ro=nozzle outer radius 16.3125 inches ri=nozzle inner radius 12.97 inches CNOZZLE 1977 Result: [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE = 0.97 1.15

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 6 of 8)

The calculation for the CPS N1 Nozzle results in 0.97, which is less than 1.15 and satisfies this criteria.

Based upon the above information, all CPS RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections meet the general and nozzle-specific criteria in BWRVIP-241.

The analyses for the nozzles in BWRVIP-108NP and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address an extended operating period. Pressure-Temperature Limits reports applicable to CPS, concluded that peak fluence over the period of extended operation (54 effective full power years) is expected to be less than the fluence criteria of 1.0E17 n/cm2, as contained in 10 CFR 50, Appendix H for all nozzles and welds for which this relief request is applied. Therefore, the fluence criteria is satisfied and use of BWRVIP-108 and BWRVIP-241 remain applicable to the CPS nozzles contained in this relief request.

Therefore, use of N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1) for all applicable full penetration RPV nozzle-to-vessel shell welds and nozzle inner radii sections for the Fourth ISI Interval.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS, or until the NRC approves N-702, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-02 was authorized by NRC SE dated December 22, 2010 (ADAMS Accession No. ML103360335) (Reference 7). This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

Peach Bottom Atomic Power Station, Units 2 and 3 Relief Request I5R-04 was authorized by NRC SE dated December 21, 2018 (ADAMS Accession No. ML18331A216) (Reference 8).

James A. FitzPatrick Nuclear Power Plant Relief Request I5R-05 was authorized by NRC SE dated September 10, 2018 (ADAMS Accession No. ML18239A010)

(Reference 9).

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

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8. References

1. EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,"

dated October 2002

2. ASME Boiler and Pressure Vessel Code, Code Case N-648-2, "Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI, Division 1," September 4, 2014

3. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," dated February 20, 2004

4. NRC Regulatory Guide RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 19

5. Letter from Matthew A. Mitchell (NRC) to Rick Libra (BWRVIP Chairman), "Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108),'" dated December 19, 2007 (ADAMS Accession No. ML073600374)

6. Letter from Kevin Hsueh (NRC) to Tim Hanley (BWRVIP Chairman), "Revised Final Safety Evaluation for the License Renewal Appendix A for "BWRVIP-241-A:

BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and "BWRVIP-108NP-A: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend" (TAC NO. MF4638), April 26, 2017(ADAMS Accession No. ML17114A096)

7. Letter from R. D. Carlson (NRC) to M. J. Pacilio (EGC), "Clinton Power Station, Unit No. 1 -Relief Requests I3R-01, I3R-02, I3R-03, I3R-04, and I3R-05 Associated with the Third Inservice Inspection Interval (TAC Nos. ME2987, ME2988, ME2989, ME2990, and ME2991)," dated December 22, 2010 (ADAMS Accession No. ML103360335)

8. Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Alternative Requests Related to the Fifth Inservice Inspection Interval (EPID L-2018-LLR-0055, EPID L-2018-LLR-0057, EPID L-2018-LLR-0058, and EPID L-2018-LLR-0059)," dated December 21, 2018 (ADAMS Accession No. ML18331A216)

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

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

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9. Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "James A. FitzPatrick Nuclear Power Plant - Issuance of Relief from the Requirements of the ASME Code N-702 for Plant Nozzle-to-Vessel Welds and Inner Radii Examinations (EPID L-2017-LLR-0093)," dated September 10, 2018 (ADAMS Accession No. ML18239A010)

9. Enclosure Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 1 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N1A 20" Recirculation Outlet Nozzle N1A to Vessel B-D B3.90 Weld N1A-IRS 20" Recirculation Outlet Nozzle N1A Inner B-D B3.100 Radius N1B 20" Recirculation Outlet Nozzle N1B to Vessel B-D B3.90 Weld N1B-IRS 20" Recirculation Outlet Nozzle N1B Inner B-D B3.100 Radius N2A 10" Recirculation Inlet Nozzle N2A to Vessel B-D B3.90 Weld N2A-IRS 10" Recirculation Inlet Nozzle N2A Inner Radius B-D B3.100 N2B 10" Recirculation Inlet Nozzle N2B to Vessel B-D B3.90 Weld N2B-IRS 10" Recirculation Inlet Nozzle N2B Inner Radius B-D B3.100 N2C 10" Recirculation Inlet Nozzle N2C to Vessel B-D B3.90 Weld N2C-IRS 10" Recirculation Inlet Nozzle N2C Inner B-D B3.100 Radius N2D 10" Recirculation Inlet Nozzle N2D to Vessel B-D B3.90 Weld N2D-IRS 10" Recirculation Inlet Nozzle N2D Inner B-D B3.100 Radius N2E 10" Recirculation Inlet Nozzle N2E to Vessel B-D B3.90 Weld N2E-IRS 10" Recirculation Inlet Nozzle N2E Inner Radius B-D B3.100 N2F 10" Recirculation Inlet Nozzle N2F to Vessel B-D B3.90 Weld N2F-IRS 10" Recirculation Inlet Nozzle N2F Inner Radius B-D B3.100 N2G 10" Recirculation Inlet Nozzle N2G to Vessel B-D B3.90 Weld N2G-IRS 10" Recirculation Inlet Nozzle N2G Inner B-D B3.100 Radius N2H 10" Recirculation Inlet Nozzle N2H to Vessel B-D B3.90 Weld N2H-IRS 10" Recirculation Inlet Nozzle N2H Inner B-D B3.100 Radius N2J 10" Recirculation Inlet Nozzle N2J to Vessel B-D B3.90 Weld N2J-IRS 10" Recirculation Inlet Nozzle N2J Inner Radius B-D B3.100

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 2 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N2K 10" Recirculation Inlet Nozzle N2K to Vessel B-D B3.90 Weld N2K-IRS 10" Recirculation Inlet Nozzle N2K Inner Radius B-D B3.100 N3A 24" Main Steam Nozzle N3A to Vessel Weld B-D B3.90 N3A-IRS 24" Main Steam Nozzle N3A Inner Radius B-D B3.100 N3B 24" Main Steam Nozzle N3B to Vessel Weld B-D B3.90 N3B-IRS 24" Main Steam Nozzle N3B Inner Radius B-D B3.100 N3C 24" Main Steam Nozzle N3C to Vessel Weld B-D B3.90 N3C-IRS 24" Main Steam Nozzle N3C Inner Radius B-D B3.100 N3D 24" Main Steam Nozzle N3D to Vessel Weld B-D B3.90 N3D-IRS 24" Main Steam Nozzle N3D Inner Radius B-D B3.100 N5A 12" Core Spray Nozzle N5A to Vessel Weld B-D B3.90 N5A-IRS 12" Core Spray Nozzle N5A Inner Radius B-D B3.100 N5B 12" Core Spray Nozzle N5B to Vessel Weld B-D B3.90 N5B-IRS 12" Core Spray Nozzle N5B Inner Radius B-D B3.100 N6A 10" Low Pressure Core Injection Nozzle N6A to B-D B3.90 Vessel Weld N6A-IRS 10" Low Pressure Core Injection Nozzle N6A B-D B3.100 Inner Radius N6B 10" Low Pressure Core Injection Nozzle N6B to B-D B3.90 Vessel Weld N6B-IRS 10" Low Pressure Core Injection Nozzle N6B B-D B3.100 Inner Radius N6C 10" Low Pressure Core Injection Nozzle N6C to B-D B3.90 Vessel Weld N6C-IRS 10" Low Pressure Core Injection Nozzle N6C B-D B3.100 Inner Radius N7 6" Top Head Spray Nozzle N7 to Vessel Weld B-D B3.90 N7-IRS 6" Top Head Spray Nozzle N7 Inner Radius B-D B3.100 N8 6" Top Head Spare Nozzle N8 to Vessel Weld B-D B3.90 N8-IRS 6" Top Head Spare Nozzle N8 Inner Radius B-D B3.100 N9A 4" Jet Pump Instrumentation Nozzle N9A to B-D B3.90 Vessel Weld N9A-IRS 4" Jet Pump Instrumentation Nozzle N9A Inner B-D B3.100 Radius N9B 4" Jet Pump Instrumentation Nozzle N9B to B-D B3.90 Vessel Weld N9B-IRS 4" Jet Pump Instrumentation Nozzle N9B Inner B-D B3.100 Radius

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 3 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N16 Vibration Instrumentation Nozzle to Vessel B-D B3.90 Weld N16-IRS Vibration Instrumentation Nozzle Inner Radius B-D B3.100

10 CFR 50.55a Relief Request I4R-03 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 1 of 5)

1. ASME Code Component(s) Affected Code Class: 2 and 3

Reference:

Table IWC-2500-1, IWC-5200 Table IWD-2500-1, IWD-5200 Examination Category: C-H, D-B Item Number: C7.10, D2.10

Description:

Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All ISI Class 2 Instrument Air (IA) Piping and the ISI Class 3 IA Piping Supplying, all Safety Relief Valves (SRVs), and both Feedwater Containment Outboard Isolation Check Valves Component Number: Multiple lines (See Note below)

Drawing Number: M05-1040, Sheet 7 M10-9002, Sheet 1 M10-9002, Sheet 2 M10-9004, Sheet 8 Note: A more detailed description of the pressure testing boundary is identified below.

The following ISI Class 2 IA piping and components between containment isolation valves 1IA012A/B and 1IA013A/B and check valves 1IA042A/B require examination.

This includes the following lines, valves, and components shown on CPS Piping and Instrumentation Diagram (P&ID) M05-1040, Sheet 7 not listed above.

Lines 1IA71BA/BB-1, 1IA14GA/GB-1, 1IA95A/B-1, 1IA93AA/BA-3/4, and 1IA96AA/BA-3/4

Valves 1IA131A/B, 1IA129A/B, and the blind flanges on lines IIA95A/B-1 The following ISI Class 3 IA system piping and components require examination. This includes the following IA lines and valves supplying all 16 SRVs and both Feedwater containment outboard isolation check valves.

P&ID M05-1040, Sheet 7 lines - 1IA79CA/CB-1, 1IA92AA/BA-3/4, 1IA102BA-1/2, 1IA103BA-1/2, 1IA71AA/AB-1, 1IA87A/B-1/2, 1IA125A/B-1/2, 1IA122A/B-1, 1IA88A/B-1/2, 1IA71CA/CB-1, 1IA71DA/EA/FA/GA-1/2, and 1IA71DB/EB/FB/GB/FC-1/2

P&ID M05-1040, Sheet 7 valves - 1IA075A/B, 1IA076A/B, 1IA130A/B, 1IA1170A/B, 0IA18MA/B, 1IA044A/B, 1IA1171A/B, 1IA1172A/B, 1IA096C/D, and 1IA097A/B.

NOTE - Strainers 1IA26FA/FB are not Code components

10 CFR 50.55a Relief Request I4R-03 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 2 of 5)

P&ID M10-9002, Sheet 1 lines - 1IA71DA/DB/EA/EB/FA/FB/FC-1/2, 1IA85A/B/C/D/E/F/G-1/2, 1MS71CE/DE-1/2, 1MS72AE/BE-1/2, 1MS73BE/CE-1/2, 1MS74CE-1/2, 1MS71CG/DG-3/4, 1MS72AG/BG-3/4, 1MS73BG/CG-3/4, 1MS74CG-3/4, 1MS71CH/DH-1/2, 1MS72AH/BH-1/2, 1MS73BH/CH-1/2, 1MS74CH-1/2, 1MS71CF/DF-3/4, 1MS72AF/BF-3/4, 1MS73BF/CF-3/4, 1MS74CF-3/4, 1MS71CC/DC-2, 1MS72AC/BC-2, 1MS73BC/CC-2, 1MS74CC-2, 1MS71CJ/CK/DJ/DK-1 1/4, 1MS72AJ/AK/BJ/BK-1 1/4, 1MS73BJ/BK/CJ/CK-1 1/4, and 1MS74CJ/CK-1 1/4

P&ID M10-9002, Sheet 1 valves - 1IA094A/B/C/D/E/F/G, 1B21-F039B/C/D/E/H/K/S, 1B21-F331C/D, 1B21-F332A/B, 1B21-F333B/C, 1B21-F334, valves 'G' on M10-9002, Sheet 1 and 1B21-F082B/C/D/E/H/K/S

P&ID M10-9002, Sheet 1 accumulators - 1B21-A003B/C/D/E/H/K/S

P&ID M10-9002, Sheet 2 lines - 1IA71GA/GB-1/2, 1IA86C/E-1/2, 1MS75AE/BE-1/2, 1MS76CE/DE-1/2, 1MS77AE/CE/DE-1/2, 1MS78BE/CE-1/2, 1MS75AC/BC-2, 1MS76CC/DC-2, 1MS77AC/CC/DC-2, 1MS78BC/CC-2, 1MS75AG/AH/BG/BH-1 1/4, 1MS76CG/CH/DG/DH-1 1/4, 1MS77AG/AH/CG/CH/DG/DH-1 1/4, and 1MS78BG/BH/CG/CH-1 1/4

P&ID M10-9002, Sheet 2 valves - 1IA095C/E, 1B21-F036A/F/G/J/L/M/N/P/R and 1B21-F081A/F/G/J/L/M/N/P/R

P&ID M10-9002, Sheet 2 accumulators - 1B21-A004A/F/G/J/L/M/N/P/R

P&ID M10-9004 Sheet 8 lines - 1FW26BA/BB-1/2, 1FW27BA/BB-1/2, 1FW26CA/CB-2, and 1FW28AA/AB-3/4

P&ID M10-9004, Sheet 8 valves - 1B21-F433A/B and 1B21-F492A/B

P&ID M10-9004, Sheet 8 accumulators - 1B21-A300A/B

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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 Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all ISI Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWC-5220. This pressure test is to be conducted once each inspection period.

Table IWD-2500-1, Examination Category D-B, Item Number D2.10, requires all ISI Class 3 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWD-5220. This pressure test is to be conducted once each inspection period.

10 CFR 50.55a Relief Request I4R-03 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 3 of 5)

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.

Performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, many of which are in elevated dose rate areas with limited access. VT-2 visual examinations would result in additional radiation exposure (estimated to be 2 rem) and industrial safety challenges without any added benefit in the level of quality and safety. These examinations would not be consistent with radiation exposure practices of As Low As Reasonably Achievable (ALARA).

Relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Tables IWC-2500-1 and IWD-2500-1 for all ISI Class 2 IA piping and the ISI Class 3 IA piping supplying all SRVs and both Feedwater containment outboard isolation check valves.

5. Proposed Alternative and Basis for Use As an alternative to the examination requirements of Tables IWC-2500-1 and IWD-2500-1, CPS will perform pressure decay testing on the ISI Class 2 and 3 IA piping supplying all 16 SRVs and both Feedwater containment outboard isolation check valves as required in surveillance procedure CPS 9061.11, "Instrument Air Check Valve Operability and Pipe Pressure Test."

Surveillance procedure CPS 9061.11, verifies the operability of SRV actuation capability and check valves in the IA supply lines to all 16 SRVs and both Feedwater containment outboard isolation check valves. This surveillance test is performed for each individual SRV and both Feedwater containment outboard isolation check valves as a requirement of the CPS Inservice Testing (IST) Program. One specific test this surveillance performs, is a pressure decay test of the SRV and Feedwater containment outboard isolation check valve accumulators, as well as associated piping and valves. The pressure decay test is performed by isolating and pressurizing these accumulators and associated piping to the nominal operating pressure. The decay in pressure is then monitored through calibrated pressure measuring instrumentation. If any pressure decay acceptance criterion is exceeded in Table 1, the surveillance procedure identifies appropriate troubleshooting steps to perform, including soap-bubble application to locate leakage.

10 CFR 50.55a Relief Request I4R-03 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 4 of 5)

Table 1: Acceptance Criteria From Procedure CPS 9061.11 (For Information Only)

Clinton Power Station, Unit 1 Pressure Leakage Drop Test Component Criterion Comments Duration (psig)

(minutes)

Accumulator Headers for all SRVs except 1B21- 1.5 psig 108 minutes F051C and D Accumulator Headers for Smaller volume than 1.5 psig 31 minutes 1B21-F051C and D other SRVs.

Accumulator Headers for Smaller volume than 1.5 psig 26 minutes Feedwater Check Valve SRVs.

This examination covers ADS Supply Header to 22 psig 60 minutes over 200 feet of piping Accumulator Headers and components.

The pressure decay test performed as part of CPS 9061.11 identifies degradation of the ISI Class 2 and 3 Automatic Depressurization System (ADS) supply piping and the SRV and Feedwater containment outboard isolation check valve accumulators and associated piping. The volume tested by this surveillance encompasses all piping and components requiring testing under ASME Section XI for these portions of the IA system. This surveillance is performed on a greater frequency than that required in Tables IWC-2500-1 or IWD-2500-1 and the test pressure is consistent with the pressure requirements of both tables. Thus, the testing performed during this surveillance will provide the same level of quality and safety as the pressure testing and VT-2 visual examination requirements of Tables IWC-2500-1 and IWD-2500-1.

The VT-2 visual examination described in Tables IWC-2500-1 and IWD-2500-1 and performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, many of which are in elevated dose rate areas with limited access. VT-2 visual examinations would result in additional radiation exposure (estimated 2 rem) and industrial safety challenges without any added benefit in the level of quality and safety.

These examinations would not be consistent with radiation exposure practices of ALARA.

In summary, relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Tables IWC-2500-1 and IWD-2500-1 for the ISI Class 2 and 3 IA system piping and components identified in this relief request on

10 CFR 50.55a Relief Request I4R-03 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for All Class 2 Instrument Air (IA) Piping and the Class 3 IA Piping Supplying All Safety Relief Valves (SRVs) and both Feedwater Containment Outboard Isolation Check Valves in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 5 of 5) 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. The alternative approach outlined above is supported by IWA-5120(e) of the 2017 Edition of ASME Section XI.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request was authorized by NRC Safety Evaluation Report (SER) dated December 13, 2007 (ADAMS Accession No. ML103360335). This relief request for the Clinton Power Station, Unit 1 Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

8. References None

10 CFR 50.55a Relief Request I4R-04 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs) in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 1 of 4)

1. ASME Code Component(s) Affected Code Class: 3

Reference:

Table IWD-2500-1, IWD-5200 Examination Category: D-B Item Number: D2.10

Description:

Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs)

Component Number: Multiple lines (See Note below)

Drawing Number: M10-9002, Sheet 5 Note: A more detailed description of the pressure testing boundary is identified below.

The following ISI Class 3 Instrument Air (IA) system piping and components require examination. This includes the following IA lines and valves supplying all eight (8)

MSIVs (i.e., four (4) inboard and four (4) outboard).

Drawing M10-9002, Sheet 5 lines - 1MS79AA/BA/CA/DA, 1MS79AC/BC/CC/DC, 1MS79AB/BB/CB/DB, 1MS80AA/BA/CA/DA, 1MS80AC/BC/CC/DC, and 1MS80AB/BB/CB/DB.

Drawing M10-9002, Sheet 5 valves - 1B21-F083A/B/C/D, 1B21F024A/B/C/D, 1B21-F084A/B/C/D and 1B21-F029A/B/C/D.

Drawing M10-9002, Sheet 5 accumulators - 1B21-A001A/B/C/D and 1B21-A002A/B/C/D

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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 Table IWD-2500-1, Examination Category D-B, Item Number D2.10, requires all ISI Class 3 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWD-5220. This pressure test is to be conducted once each inspection period.

10 CFR 50.55a Relief Request I4R-04 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs) in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 2 of 4)

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.

Performance of a VT-2 visual examination would require applying a leak detection solution to this piping and components, all of which are in elevated dose rate areas with limited access. VT-2 visual examinations would result in an estimated additional radiation exposure of 0.5 rem, and industrial safety challenges without any added benefit in the level of quality and safety. These examinations would not be consistent with As Low As Reasonably Achievable (ALARA) radiation exposure practices.

Relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Table IWD-2500-1 for all ISI Class 3 IA piping, valves, and components supplying all eight (8) MSIVs at CPS.

5. Proposed Alternative and Basis for Use As an alternative to the examination requirements of Table IWD-2500-1, CPS will perform pressure decay testing on the ISI Class 3 IA piping supplying all eight (8) MSIVs as required in surveillance procedure CPS 9061.11, "Instrument Air Check Valve Operability and Pipe Pressure Test."

The diameter of the majority of the instrument air piping supplying the inboard and outboard MSIVs is 1/2 inch, but also includes some 2 inch piping containing a few 3/4 inch drain taps. This piping is seamless austenitic stainless steel (i.e., ASME SA-312 or SA-376, Grade TP304).

This relief request is applicable to approximately 360 feet of piping based on field walkdowns. This estimate is based on lengths of piping in the order of 40 feet for each of the four inboard MSIVs and 50 feet for each of the four outboard MSIVs.

Surveillance procedure CPS 9061.11, verifies the operability of MSIV closure capability and check valve repositioning in the IA supply lines to all eight (8) MSIVs. This surveillance is performed for each individual MSIV as a requirement of the CPS Inservice Testing (IST) Program. One specific test this surveillance performs is a pressure decay test of the MSIV air supply components. The pressure decay test is performed by pressurizing and isolating these accumulators and associated piping at nominal operating pressure. The decay in pressure is then monitored through calibrated pressure measuring instrumentation. If any pressure decay acceptance criterion as shown in Table 1 is exceeded, the surveillance procedure identifies appropriate troubleshooting steps to perform, including soap-bubble application to locate leakage.

10 CFR 50.55a Relief Request I4R-04 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs) in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 3 of 4)

Design calculations determined that each MSIV would need 35 gallons of air to close an MSIV following the failure of the normal air supply. The sizing of MSIV air system piping is based on continuous leakage of one standard cubic foot per hour (SCFH) at the minimum air pressure of 100 psig. Each MSIV has an air accumulator that is slightly oversized at 39 gallons to provide additional margin. The pressure decay test acceptance criterion is based upon leakage equaling one-half of the assumed one SCFH continuous leakage, or 0.5 SCFH. This equates to a pressure drop of 0.0242 psi per minute. For a 63 minute test, the total pressure drop would be 1.5 psig. The 63 minute test duration is based on the resolution of the instrument used for the test and the time necessary to obtain conclusive and reliable results. If the pressure drop measured during the test exceeds 1.5 psig, the affected piping will be inspected using a soapy liquid leak detecting solution to locate leakage. Leakage that is located will be corrected until satisfactory results are obtained. If through-wall leakage from a safety related component is discovered, the component will be repaired or replaced in accordance with the applicable ASME Section XI requirements.

Table 1: Acceptance Criteria from CPS Procedure CPS 9061.11 (For Information Only)

Clinton Power Station, Unit 1 Component Leakage Criterion Pressure Drop Comments (psig) Test Duration (minutes)

Performed under Accumulator Sections 8.11 and Headers for all 1.5 psig 63 minutes 8.14 of Procedure MSIVs CPS 9061.11.

The pressure decay test performed as part of CPS 9061.11 identifies degradation of the ISI Class 3 IA supply piping to the MSIVs and associated isolation check valves, accumulators, and valves. The volume tested by this surveillance encompasses all piping and components requiring testing under ASME Section XI for these portions of the IA system. This surveillance is performed each 24-month refueling cycle, which is a greater frequency than that required in Table IWD-2500-1 and the test pressure is consistent with the pressure requirements of this table. Thus, the testing performed during this surveillance will provide the same level of quality and safety as the pressure testing and VT-2 visual examination requirements of Table IWD-2500-1. The normal operating pressure of the Instrument Air (IA) system supplying the MSIV accumulators is 110 psig.

The VT-2 visual examination described in Table IWD-2500-1, which is performed once per inspection period, would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, all of which are in elevated dose rate areas with limited access. VT-2 visual examinations

10 CFR 50.55a Relief Request I4R-04 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all Inservice Inspection (ISI) Class 3 Instrument Air (IA) Piping Supplying Eight (8) Main Steam Isolation Valves (MSIVs) in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 4 of 4) would result in estimated additional radiation exposure of 0.5 rem and industrial safety challenges without any added benefit in the level of quality and safety. These examinations would not be consistent with ALARA radiation exposure practices.

In summary, relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Table IWD-2500-1 for the ISI Class 3 IA system piping and components identified in this request 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. The alternative approach outlined above is supported by IWA-5120(e) of the 2017 Edition of ASME Section XI.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-07 was authorized for use by the NRC in a Safety Evaluation (SE) dated November 21, 2011 (ADAMS Accession No. ML112900885). This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

LaSalle County Station, Units 1 and 2 Relief Requests I4R-07 was authorized per SE dated November 17, 2017 (ADAMS Accession No. ML17354A854).

8. References None

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 1 of 6)

1. ASME Code Component(s) Affected Code Class: 2

Reference:

Table IWC-2500-1, IWC-5200 Examination Category: C-H Item Number: C7.10

Description:

Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB.

Component Number: Multiple lines (see Note below)

Drawing Number: M05-1063, Sheet 1 M05-2063, Sheet 1 Note: A more detailed description of the system function and pressure testing boundary is identified below.

The Combustible Gas Control (HG) system maintains the hydrogen concentration in the Drywell and Containment below the ignition level. This relief request involves a portion of the HG system designed to control hydrogen generated during a Loss of Coolant Accident (LOCA) when the core is not degraded.

The ISI Class 2 HG system piping and components requiring examination are listed below. These components and piping control and route flow from the containment to Hydrogen Recombiners 0HG01SA and 0HG01SB and back to the containment.

All piping is seamless carbon steel, ASTM A106, Gr. B and/or ASME SA-106, Gr. B.

The 2 inch and smaller diameter piping is schedule 80 thickness and the 2 1/2 inch diameter piping is schedule 40 thickness.

Drawing M05-1063, Sheet 1 - Lines associated with hydrogen recombiner 0HG01SA in flow path order starting inside the containment just prior to containment penetration 1MC-71 and ending inside containment just past containment penetration 1MC-72 are: 1HG01A-2 inch, 1HG07A-3/4 inch, 1HG01B-2 1/2 inch, 1HG11A-3/4 inch, 0HG03A-2 inch, 0HG03B-2 inch, 0HG04A-2 inch, 0HG04B-2 inch, 0HG02A-2 1/2 inch, 1HG12A-3/4 inch, 1HG02B-2 inch, and 1HG08A-3/4 inch.

Drawing M05-1063, Sheet 1 - Valves associated with hydrogen recombiner 0HG01SA in flow path order starting inside the containment just prior to containment penetration 1MC-71 and ending inside containment just past containment penetration 1MC-72 are: - 1HG012-2 inch, 1HG016-3/4 inch, 1HG001-2 inch, 1HG020-3/4 inch, 1HG002-2 inch, 2HG002-2 inch, 2HG003-2 inch, 1HG003-2 inch, 1HG021-3/4 inch, 1HG004-2 inch, 1HG017-3/4 inch, and 1HG013-2 inch.

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

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Drawing M05-1063, Sheet 1 - Hydrogen Recombiner 0HG01SA.

Drawings M05-1063, Sheet 1 and M05-2063, Sheet 1 - Lines associated with hydrogen recombiner 0HG01SB in flow path order starting inside the containment just prior to containment penetration 1MC-166 and ending inside containment just past containment penetration 1MC-62 are:

1HG03A-2 inch, 1HG09A-3/4 inch, 1HG03B-3 inch, 0HG01B-2 inch, 0HG01A-2 inch, 0HG02A-2 inch, 0HG02B-2 inch, 1HG04A-2 1/2 inch, 1HG04B-2 inch, and 1HG010A-3/4 inch.

Drawings M05-1063, Sheet 1 and M05-2063, Sheet 1 - Valves associated with hydrogen recombiner 0HG01SB in flow path order starting inside the containment just prior to containment penetration 1MC-166 and ending inside containment just past containment penetration 1MC-62 are:

1HG014-2 inch, 1HG018-3/4 inch, 1HG005-2 inch, 1HG006-2 inch, 2HG006-2 inch, 2HG007-2 inch, 1HG007-2 inch, 1HG008-2 inch, 1HG019-3/4 inch, and 1HG015-2 inch.

Drawing M05-2063, Sheet 1 - Hydrogen Recombiner 0HG01SB.

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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 Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all ISI Class 2 pressure retaining components to be subjected to a system leakage test via a VT-2 visual examination in accordance with Paragraph IWC-5220. This pressure test is to be conducted once each inspection period. IWC-5000 permits the use of a pneumatic test for ISI Class 2 components in accordance with Paragraph IWA-5211(c).

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 or safety.

Performance of the ASME Section XI prescribed VT-2 visual examination of this piping would require applying a leak detection solution to the safety-related piping and components to locate evidence of leakage. Performing this examination would pose significant safety hazards, as most of this piping requires working at heights. Either a ladder or scaffold will be needed to reach this piping, and all

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

Revision 0 (Page 3 of 6) piping is insulated. Based upon field walkdowns, the piping associated with Hydrogen Recombiner 0HG01SA is conservatively estimated at 550 linear feet and the piping associated with Hydrogen Recombiner 0HG01SB is conservatively estimated at 1050 feet in length.

Examining the piping associated with 0HG01SB will also result in an estimated radiation exposure of 150 to 200 mrem. Performing this VT-2 visual examination would not be consistent with As Low As Reasonably Achievable (ALARA) radiation exposure practices.

Relief is requested from the performance of system leakage tests and VT-2 visual examination requirements specified in Table IWC-2500-1 for the ISI Class 2 HG piping, valves and components associated with Hydrogen Recombiners 0HG01SA and 0HG01SB at CPS.

5. Proposed Alternative and Basis for Use As an alternative to the examination requirements of Table IWC-2500-1, Exelon Generation Company, LLC (EGC) will perform leak rate tests on the ISI Class 2 HG piping associated with Hydrogen Recombiners 0HG01SA and 0HG01SB as required in surveillance procedure CPS 9861.02 Datasheet D041, Local Leak Rate Testing (LLRT) Data Sheet for 1MC071/1MC072- H2 Recombiner 0HG01SA and Datasheet D033, LLRT Data Sheet for 1MC062/1MC166 - Hydrogen Recombiner 0HG01SB.

LLRT procedure CPS 9861.02 Datasheets D033 and D041 determine the leakage rate from the components and piping associated with this relief request as well as leakage through affected boundary valve internals and mechanical connections.

These datasheets perform three separate leak rate tests (i.e., Test Sets A, B, and C) on each hydrogen recombiner train. The test pressure for each test is maintained between 9.1 psig and 9.9 psig and the elapsed test (i.e., hold) time is a minimum of 15 minutes from stabilization of test pressure. Stabilization can be declared when the measured flow rate is stable for a period of approximately five minutes. According to the piping design tables, the maximum operating pressure of this piping is nine psig and the maximum design pressure is 15 psig. Test Set

'A' tests the piping and components between the HG return header opening inside containment and containment isolation valve 1HG008 or 1HG004. Test Set 'B' tests the piping and components between the HG inlet header opening inside containment and containment isolation valve 1HG005 or 1HG001. Test Set 'C' tests all of the piping and components between the HG inlet header opening inside containment and the HG return header inside containment. Administrative limits have been established for each train to prompt further examinations and/or evaluations. Train 'B' has an administrative limit of 500 standard cubic centimeters per minute (sccm) and Train 'A' has an administrative limit 1000 sccm.

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

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Normal system flow is 50 standard cubic feet per minute (scfm). The accuracy of the pressure measuring instrumentation is +- 0.15 psig and the accuracy of the flow measuring instrumentation is +-2% of full scale.

The lines and components being addressed in this relief request are a possible secondary containment bypass pathway. As such, any leakage detected is included in the overall secondary containment bypass pathway limit of 28,882 sccm, or 0.08 of the maximum allowable containment atmospheric leak rate at the calculated peak accident pressure (La). The administrative limits established in LLRT procedure CPS 9861.02 Datasheets D033 and D041 were selected to ensure evaluation and trending of the test results is performed well before the overall secondary containment bypass pathway limit is challenged. If the administrative limit is exceeded as shown in Table 1 by any of the three test sets, the ISI Pressure Test Program Manager or their Designee will be contacted to evaluate the test results and determine if the leakage is due to through-wall leakage in the pressure boundary. Troubleshooting will be performed to quantify all valve seat leakage and to perform VT-2 visual examinations of all mechanical joints associated with the train in question with a leak detection solution. Any leakage identified will be corrected and the test set run again. If troubleshooting determines that leakage above the administrative limit may be attributed to a through-wall pressure boundary leak, the ISI Pressure Test Program Manager or their Designee will perform a VT-2 visual examination of the entire piping run with a leak detection solution. If through-wall pressure boundary leakage from a safety related component is discovered, the component or piping will be repaired or replaced in accordance with the applicable ASME Section XI requirements. If evidence of safety-related pressure boundary leakage is not located, the examination will be considered acceptable for ISI Pressure Testing Program requirements for that inspection period.

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

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Table 1: Acceptance Criteria From Procedure CPS 9861.02 Datasheets D033 and D041 (For Information Only)

Clinton Power Station, Unit 1 Pressure Leakage Drop Test Component Criterion Duration (sccm)

(minutes)

HG piping and components associated 1000 15 minutes with Hydrogen Recombiners 0HG01SA HG piping and components associated 500 15 minutes with Hydrogen Recombiners 0HG01SB The leakage test performed as part of CPS 9861.02 Datasheets D033 and D041 identifies degradation of the ISI Class 2 HG piping and components associated with the Hydrogen Recombiners. The volume tested by these surveillances encompasses all piping and components requiring testing under ASME Section XI for these portions of the HG system. These surveillances are performed every inspection period to comply with the frequency required in Table IWC-2500-1. The test pressure used is consistent with the pressure requirements of this table.

Thus, as a minimum, the testing performed during this surveillance will provide the same level of quality and safety as the pressure testing and VT-2 visual examination requirements of Table IWC-2500-1.

This test is a flow test and, as described in CPS 9861.02, there are three 'test sets' for each train of the hydrogen recombiner subsystems (i.e., train 'A' and train 'B') in the HG system. The three test sets for each of the respective subsystems are as follows:

Test set A, for the 'A' Hydrogen Recombiner train, tests piping and components on the inlet line from the open-ended piping in the containment to motor-operated containment isolation valve 1HG001.

Test set A, for the 'B' Hydrogen Recombiner train, tests piping and components on the outlet line from the open-ended piping in the containment to motor-operated containment isolation valve 1HG008.

10 CFR 50.55a Relief Request I4R-05 Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for the Inservice Inspection (ISI) Class 2 Combustible Gas Control (HG) piping to and from Hydrogen Recombiners 0HG01SA and 0HG01SB in Accordance with 10 CFR 50.55a(z)(2)

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

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Test set B, for the 'A' train, tests piping and components on the outlet line from motor-operated containment isolation valve 1HG004 to the open-ended piping in the containment.

Test set B, for the 'B' train, tests piping and components on the inlet line from motor-operated containment isolation valve 1HG005 to the open-ended piping in the containment.

Test set C for both trains tests piping and components on the inlet and outlet lines from the open-ended inlet piping in the containment, through the respective hydrogen recombiner (i.e., 0HG01SA or 0HG01SB) to the open-ended outlet piping in the containment, respectively.

In summary, relief is requested from the performance of the system pneumatic test and VT-2 visual examination requirements specified in Table IWC-2500-1 for the ISI Class 2 HG system piping and components identified in this request 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.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-09 was authorized for use by the NRC in a Safety Evaluation (SE) dated April 18, 2013 (ADAMS Accession No. ML13107A099). This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

LaSalle County Station, Units 1 and 2 Relief Request I4R-08 was authorized per SE dated January 3, 2018 (ADAMS Accession No. ML17354A854).

8. References None

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-N-1 and B-N-2 Item Number: B13.10, B13.20, B13.30, and B13.40

Description:

Use of BWRVIP Guidelines in Lieu of Specific ASME Section XI Requirements on the Reactor Pressure Vessel Internals and Components Inspection Component Name: Vessel Interior, Interior Attachments within Beltline Region, Interior Attachments beyond Beltline Region, and Core Support Structure

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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 ASME Section XI requires the examination of components within the reactor pressure vessel. These examinations are included in Table IWB-2500-1, Examination Categories B-N-1 and B-N-2 and identified with the following item numbers:

B13.10 Examine accessible areas of the reactor vessel interior each period by the VT-3 visual examination method (B-N-1).

B13.20 Examine interior attachment welds within the beltline region each interval by the VT-1 visual examination method (B-N-2).

B13.30 Examine interior attachment welds beyond the beltline region each interval by the VT-3 visual examination method (B-N-2).

B13.40 Examine accessible surfaces of the core support structure each interval by the VT-3 visual examination method (B-N-2).

These examinations are performed to assess the structural integrity of the reactor vessel interior, its welded attachments, and the core support structure within the boiling water reactor (BWR) pressure vessel.

The components/welds listed in Table 2 are subject to this request for alternative.

Table 2 provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1 and the appropriate Boiling Water Reactor Vessel and Internals Project (BWRVIP) document.

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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4. Reason for Request In accordance with 10 CFR 50.55a(z)(1), relief is requested for the proposed alternative to ASME Section XI requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVIP Inspection and Evaluation (I&E) guidelines recommend specific inspections by BWR owners to identify material degradation with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. The BWRVIP guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying known or potential degradation mechanisms, and require re-examination at appropriate intervals. The scope of the BWRVIP guidelines meet or exceed that of ASME Section XI and in many instances include components that are not part of the ASME Section XI jurisdiction.

As an alternative to ASME Section XI requirements, use of BWRVIP guidelines will avoid duplicate or unnecessary inspections, while conserving radiological dose.

5. Proposed Alternative and Basis for Use In lieu of the requirements of ASME Section XI, the proposed alternative is detailed in Table 2 for CPS for Examination Categories B-N-1 and B-N-2.

CPS will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 2 in accordance with the latest Nuclear Regulatory Commission (NRC) approved BWRVIP guideline requirements. This relief request proposes to utilize the identified BWRVIP guidelines in lieu of the associated ASME Section XI requirements, including the examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting.

Not all the components addressed by these guidelines are ASME Section XI components. The following BWRVIP guidelines are applicable to this relief request:

o BWRVIP-03, "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines" o BWRVIP-18, Revision 2-A, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" o BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines" o BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines" o BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines" o BWRVIP-41, Revision 4-A, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines" o BWRVIP-42, Revision 1-A, "Low Pressure Coolant Injection System (LPCI)

Coupling Inspection and Flaw Evaluation Guidelines"

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 3 of 10) o BWRVIP-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines" o BWRVIP-48-A, "Vessel ID [Internal Diameter] Attachment Weld Inspection and Flaw Evaluation Guidelines" o BWRVIP-76, Revision 1-A, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" o BWRVIP-94, "BWR Vessel and Internals Project, Program Implementation Guide" o BWRVIP-138, Revision 1-A, "Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines" o BWRVIP-180, "Access Hole Cover Inspection and Flaw Evaluation Guidelines" o BWRVIP-183-A, "Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines" Inspection Services, by an Authorized Inspection Agency, will be applied to the proposed actions of this relief request.

BWRs examine reactor internals in accordance with BWRVIP guidelines. These guidelines are written for the safety significant vessel internal components and provide appropriate examination and evaluation criteria with using appropriate methods and re-examination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVIP approach in principal and is expected to issue Safety Evaluations for many of these BWRVIP guidelines. Therefore, use of these BWRVIP guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

As additional justification, Enclosure ("Comparison of ASME Section XI Examination Requirements to BWRVIP Examination Requirements"), provides specific examples that compare the inspection requirements of ASME Section XI Item Numbers B13.10, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are provided as examples. This comparison also includes a discussion of the inspection methods.

The BWRVIP provides BWR Vessel and Internals Inspection Summaries to the NRC periodically. Table 1 contains the BWR Vessel and Internals Inspection Summaries transmitted to the NRC that includes CPS. These summaries provide, on a component-by-component basis, the examination methods utilized, the examination frequency to date, and the results of the examinations during the previous interval. These summaries also contain the identified corrective actions. This information reflects the compilation of the BWRVIP outage reports. Corrective actions and examinations performed prior to the BWRVIP were implemented to the requirements of ASME Section XI, as applicable.

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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Table 1 BWR Vessel and Internals Inspection Summaries Unit Accession Number Document Title Document Date Project No. 704 - BWR Vessel and Internals Clinton Power February 7, ML18040A464 Inspection Summaries for Station, Unit 1 2018 Spring 2017 Outages (Reference 18)

When a BWRVIP guideline refers to ASME Section XI, the technical requirements of ASME Section XI as described by the BWRVIP guideline will be met, but the examination is under the auspices of the BWRVIP program as defined by BWRVIP-94, "BWR Vessel and Internals Project, Program Implementation Guide." The reactor vessel internals inspection program at CPS has been developed and implemented to satisfy the requirements of BWRVIP-94. It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to address industry operating experience, include enhancements to inspection techniques, and add or adjust flaw evaluation methodologies. BWRVIP-94 states that where guidance in existing BWRVIP documents has been supplemented or revised by subsequent correspondence approved by the BWRVIP Executive Committee, the vessel and internals program shall be modified to reflect the new requirements and implement the guidance within two refueling outages, unless a different schedule is specified by the BWRVIP. However, if new guidance approved by the Executive Committee includes changes to NRC approved BWRVIP inspection guidance that are less conservative than those approved by the NRC, the less conservative guidance shall be implemented only after the NRC approves the changes, which generally means publication of a "-A" document or equivalent.

Where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC-authorized proposed alternative to the requirements of 10 CFR 50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for relief has been approved.

CPS is a BWR/6 design. Table 2 compares present ASME Section XI Examination Category B-N-1 and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to CPS. Therefore, Table 2 only represents the most current comparison. Any deviations from the referenced BWRVIP guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process.

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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Note that other requirements (e.g., NUREG-0619, IGSCC) are still implemented separately as Augmented Examination Programs.

In the event that conditions are identified that require repair or replacement and the component is within the jurisdiction of ASME Section XI (welded attachments to the reactor vessel or welded core support structure), the repair or replacement activities will be performed in accordance with ASME Section XI, Article IWA-4000 or a separate relief request will be submitted. Subsequent examinations will be in accordance with the applicable BWRVIP guideline.

As part of the BWRVIP initiative, the BWR reactor internals and attachments were subjected to a safety assessment to identify those components that provide a safety function and to determine if long-term actions were necessary to ensure continued safe operation. The safety functions considered are those associated with (1) maintaining a coolable geometry, (2) maintaining control rod insertion times, (3) maintaining reactivity control, (4) assuring core cooling, and (5) assuring instrumentation availability.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-10 was authorized by NRC Safety Evaluation (SE) dated April 29, 2016 (ADAMS Accession No. ML16071A233) (Reference 15). This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

Nine Mile Point Nuclear Power Station, Units 1 and 2 Relief Request I5R-03/I4R-03 was authorized conditionally by NRC SE dated December 13, 2018 (ADAMS Accession No. ML18318A275) (Reference 16).

Peach Bottom Atomic Power Station, Units 2 and 3 Relief Request I5R-03 was authorized conditionally by NRC SE dated July 18, 2018 (ADAMS Accession No. ML18179A394) (Reference 17).

8. References

1. "BWRVIP-03: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines"

2. "BWRVIP-18, Revision 2-A: BWR Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002008089, dated August 2016

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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3. Letter from NRC to BWRVIP, "Propriety Version of NRC Staff Review of BWRVIP-27-A, 'BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines,'" dated June 10, 2004

4. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-26-A, 'BWR Vessel and Internals Project, BWR Top Guide Inspection and Flaw Evaluation Guidelines,'" dated September 9, 2005

5. Letter from NRC to BWRVIP, "Final Safety Evaluation of the 'BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38),' EPRI Report TR-108823 (TAC No. M99638)," dated July 24, 2000

6. "BWRVIP-41, Revision 4-A: BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002014254, dated December 2018

7. "BWRVIP-42, Revision 1-A: BWR Vessel and Internals Project, Low Pressure Coolant Injection System (LPCI) Coupling Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002010548, dated November 2017

8. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-47-A, 'BWR Vessel and Internals Project, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines,'" dated September 9, 2005

9. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-48-A, 'BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guideline,'" dated July 25, 2005

10. "BWRVIP-76, Revision 1-A: BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002005566, dated April 2015

11. "BWRVIP-94: BWR Vessel and Internals Project, Program Implementation Guide"

12. BWRVIP-138, Revision 1-A: BWR Vessel and Internals Project, Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1025136, dated October 2012

13. "BWRVIP-180: BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1013402, dated November 2007

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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14. "BWRVIP-183-A: BWR Vessel and Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002010551, dated November 2017

15. Letter from T. Tate (NRC) to B. Hanson (EGC), "Clinton Power Station, Unit 1 and Nine Mile Point Nuclear Station, Units 1 and 2 - Relief Request Alternative RE: Use of Boling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (CAC Nos. MF6116 and MF6117)," dated April 29, 2016 (ADAMS Accession No. ML16071A233)

16. Nine Mile Point Nuclear Station, Units 1 and 2 - Issuance of Relief Requests Re:

Use of Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (EPID L-2018-LLR-0085), dated December 13, 2018 (ADAMS Accession No. ML18318A275)

17. Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Safety Evaluation of Relief Request I5R-03 Regarding the Fifth 10-Year Interval of the Inservice Inspection Program (EPID No. L-2018-LLR-0056)," dated July 18, 2018 (ADAMS Accession No. ML18179A394)

18. Letter 2018-015 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Spring 2017 Outages," dated February 7, 2018 (ADAMS Accession No. ML18040A464)

9. Enclosure Comparison of ASME Section XI Examination Requirements to BWRVIP Examination Requirements

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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TABLE 2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Item ASME ASME ASME Section XI Section XI Authorized BWRVIP Component Section Section XI BWRVIP Exam BWRVIP Frequency No. Table Exam Alternative Exam Scope Scope XI Exam Frequency IWB-2500-1 B13.10 Reactor Vessel Accessible VT-3 Each BWRVIP-18-R2- Overview examinations of components during BWRVIP Reactor Interior Areas period A, 26-A, 27-A, 38, examinations are performed to satisfy ASME Section XI VT-3 Vessel 41-R4-A, 42-R1- visual examination requirements.

Interior A, 47-A, 48-A, 76-R1-A, 138-R1-A, 180, and 183-A B13.20 Jet Pump Riser Accessible VT-1 Each 10- BWRVIP-48-A, Riser Brace EVT-1 25% during each 6 years Interior Braces Welds year Table 3-2 Attachment Attachments Lower Surveillance Interval BWRVIP-48-A, Bracket VT-1 Each 10-year Interval Within Specimen Holder Table 3-2 Attachment Beltline Brackets Region B13.30 Steam Dryer Hold- Accessible VT-3 Each 10- BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Interior Down Brackets Welds year Table 3-2 Attachment Attachments Guide Rod Brackets Interval BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Beyond Table 3-2 Attachment Beltline Steam Dryer BWRVIP-48-A, Bracket EVT-1 Each 10-year Interval Support Brackets Table 3-2 Attachment Feedwater Sparger BWRVIP-48-A, Bracket EVT-1 Each 10-year Interval Brackets Table 3-2 Attachment Core Spray Piping BWRVIP-48-A, Bracket EVT-1 100% every 4 refueling Brackets Table 3-2 Attachment cycles Upper Surveillance BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Specimen Holder Table 3-2 Attachment Brackets

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 9 of 10)

TABLE 2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Item ASME ASME ASME Section XI Section XI Authorized BWRVIP Component Section Section XI BWRVIP Exam BWRVIP Frequency No. Table Exam Alternative Exam Scope Scope XI Exam Frequency IWB-2500-1 Shroud Support BWRVIP-38, Weld H9 EVT-1 or UT Based on as-found Welds 3.3, Figure 3-5 conditions, to a maximum of 6 years for EVT-1, 10 years for UT Shroud Support Leg Accessible VT-3 Each 10- BWRVIP-38, Weld H12 Per BWRVIP-38 When accessible (Weld H12) Welds year 3.2.3 NRC SE (Beneath Interval (07/24/00),

core plate, rarely examine with accessible) appropriate method2 B13.40 Shroud Support Accessible VT-3 Each BWRVIP-38, Welds H8 and EVT-1 or UT Based on as-found Core Support Surfaces 10-year 3.3, H9 conditions, to a maximum Structure Interval Appendix A of 6 years for EVT-1, 10 Figures 3-4 and years for UT 3-5 Shroud Support Accessible BWRVIP-38, Shroud Per BWRVIP-38 When accessible Legs Surfaces 3.2.3 support leg NRC SE (beneath welds (07/24/00),

core plate, rarely examine with accessible) appropriate method2 Shroud Vertical Accessible BWRVIP-76-R1- Vertical Welds EVT-1 or UT Maximum of 6 years for Welds Surfaces A, one-sided EVT-1, 10 3.3, Figure 3-2 years for UT

10 CFR 50.55a Relief Request I4R-06 Alternative for the Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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TABLE 2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Item ASME ASME ASME Section XI Section XI Authorized BWRVIP Component Section Section XI BWRVIP Exam BWRVIP Frequency No. Table Exam Alternative Exam Scope Scope XI Exam Frequency IWB-2500-1 Shroud Repairs3 Accessible BWRVIP-76-R1- Tie-Rod VT-3 Per repair designer Surfaces A, Repair recommendations, per 3.5 BWRVIP-76-R1-A NOTES:

1) This table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1, and the appropriate BWRVIP document.

2) When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection of these welds.

3) CPS has a tie-rod shroud repair.

Enclosure 10 CFR 50.55a Relief Request I4R-06 Alternative to Use BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 1 of 5)

Comparison of ASME Section XI Examination Requirements to BWRVIP Examination Requirements The following paragraphs provide a comparison of the examination requirements in ASME Section XI, Table IWB-2500-1, Item Numbers B13.10, B13.20, B13.30, and B13.40, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. ASME Section XI Requirement - B13.10 - Reactor Vessel Interior Accessible Areas (B-N-1)

ASME Section XI requires a VT-3 visual examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately three years, during the First ISI Interval, and each inspection period during each successive 10-year ISI Interval.

Typically, these examinations are performed every inspection period during the 10-year ISI Interval. This examination requirement is a non-specific requirement that is a departure from the traditional ASME Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts; loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products, wear, and structural degradation.

Portions of the various examinations required by the applicable BWRVIP guidelines require examination of accessible areas of the reactor vessel during refueling outages.

Examination of core spray piping and spargers (BWRVIP-18-R2-A), top guide (BWRVIP-26-A), shroud support (BWRVIP-38), jet pump welds and components (BWRVIP-41-R4-A), LPCI couplings (BWRVIP-42-R1-A), lower plenum components (BWRVIP-47-A) interior attachments (BWRVIP-48-A), core shroud welds, (BWRVIP-76-R1-A), access hole cover (BWRVIP-180), and top guide grid beams (BWRVIP-183-A) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 visual examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by ASME Section XI. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements.

Therefore, the specified BWRVIP guideline requirements meet or exceed the subject ASME Section XI requirements (including method and frequency requirements) for

Enclosure 10 CFR 50.55a Relief Request I4R-06 Alternative to Use BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 2 of 5) examination of the interior of the reactor vessel. Accordingly, these BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Section XI requirements.

2. ASME Section XI Requirement - B13.20 - Interior Attachments Within the Beltline (B-N-2)

ASME Section XI requires a VT-1 visual examination of accessible reactor vessel interior surface attachment welds within the beltline each 10-year interval. In the General Electric (GE) Company BWR/6 design, this includes the jet pump riser brace welds-to-reactor vessel wall and the lower surveillance specimen support bracket welds-to-reactor vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the lower surveillance specimen support bracket welds, and requires an enhanced VT-1 (EVT-1) visual examination of the remaining attachment welds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years.

The jet pump riser brace examination requirements are provided below to show a comparison between ASME Section XI and the BWRVIP examination requirements.

Comparison to BWRVIP Requirements - Jet Pump Riser Braces BWRVIP-48-A

ASME Section XI requires a 100% VT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds each 10-year interval.

BWRVIP-48-A requires an EVT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds and heat affected zones 25% each subsequent 6 years.

BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined.

ASME Section XI VT-1 visual examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP EVT-1 visual examination is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and intergranular stress corrosion cracking (IGSCC), the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT-1 visual examination.

ASME Section XI VT-1 visual examination method requires that at a maximum distance of 2 feet or a letter character with a maximum height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires resolution of 0.044 inch characters on the examination surface and additionally the performance of a cleaning assessment and cleaning as necessary. BWRVIP-48-A includes a diagram for the configuration and prescribes examination for each configuration including CPS.

Enclosure 10 CFR 50.55a Relief Request I4R-06 Alternative to Use BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 3 of 5)

The calibration standards used for BWRVIP EVT-1 visual examinations utilize the ASME Section XI characters, thus assuring at least equivalent resolution compared to the ASME Section XI requirements. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced flaw detection capability of an EVT-1 visual examination with a less frequent examination schedule provides an acceptable level of quality and safety to that provided by ASME Section XI.

3. ASME Section XI Requirement - B13.30 - Interior Attachment Beyond the Beltline Region (B-N-2)

ASME Section XI requires a VT-3 visual examination of accessible reactor vessel interior surface attachment welds beyond the beltline each 10-year interval. In the BWR/6 design, this includes the core spray piping support bracket welds-to-reactor vessel wall, the upper surveillance specimen support bracket welds-to-reactor vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, and the shroud support plate-to-reactor vessel weld. BWRVIP-48-A requires as a minimum the same VT-3 visual examination method as ASME Section XI for some of the interior attachment welds beyond the beltline region, and in some cases specifies an EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of visual examination, the same scope of examination (accessible welds), the same examination frequency (each 10-year interval), and the same ASME Section XI flaw evaluation criteria are used. Therefore, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by ASME Section XI.

For the core spray support bracket attachment welds, the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-reactor vessel welds, the BWRVIP guidelines require an EVT-1 visual examination at the same frequency as ASME Section XI, or at a more frequent rate. Therefore, the BWRVIP enhanced examination requirements provide the same level of quality and safety compared to that provided by ASME Section XI.

The feedwater sparger bracket-to-reactor vessel attachment weld is used as an example for comparison between ASME Section XI and BWRVIP examination requirements as discussed below.

Comparison to BWRVIP Requirements - Feedwater Sparger Bracket Welds (BWRVIP-48-A)

The ASME Section XI examination requirement is a VT-3 visual examination of each weld every 10 years.

The BWRVIP-48-A visual examination requirement is an EVT-1 for the feedwater sparger bracket attachment welds with each weld examined every 10 years.

Enclosure 10 CFR 50.55a Relief Request I4R-06 Alternative to Use BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

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The BWRVIP-48-A visual examination method EVT-1 has superior flaw detection and sizing capability, the examination frequency is the same as the ASME Section XI requirements, and the same flaw evaluation criteria are used.

ASME Section XI VT-3 visual examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An EVT-1 visual examination is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, and the relevant degradation mechanisms for BWR internal attachments.

Therefore, because the EVT-1 visual examination method provides the same examination scope (accessible welds), the same examination frequency in most cases, and the same flaw evaluation criteria as ASME Section XI, the level of quality and safety provided by the BWRVIP criteria meets or exceeds that provided by the ASME Section XI requirements.

4. ASME Section XI Requirement - B13.40 - Core Support Structures (B-N-2)

ASME Section XI requires a VT-3 visual examination of accessible surfaces of the reactor vessel core support structure each 10-year interval. In the BWR/6 design, the core support structure has primarily been considered the shroud support structure, including the shroud. Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examinations replace this ASME Section XI requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

Comparison to BWRVIP Requirements - Shroud Supports (BWRVIP-38)

ASME Section XI requires a VT-3 visual examination of accessible surfaces each 10-year interval.

The BWRVIP-38 requires an EVT-1 visual examination every 6 years or ultrasonic examination (UT) every 10 years.

BWRVIP recommended examinations of core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. In many locations, the BWRVIP guidelines require a UT of the susceptible welds at a frequency identical to the ASME Section XI requirement.

The BWRVIP guidelines require an EVT-1 or UT of core support structures. The core shroud is used as an example for comparison between the ASME Section XI and BWRVIP examination requirements as shown below.

Enclosure 10 CFR 50.55a Relief Request I4R-06 Alternative to Use BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Enclosure Page 5 of 5)

Comparison to BWRVIP Requirements - Core Shroud (BWRVIP-76-R1-A)

Shroud repair Tie-Rods have been installed at CPS. Therefore, BWRVIP-76-R1-A requires inspection of the vertical shroud welds and the Tie-Rod repair hardware.

ASME Section XI requires a VT-3 visual examination of accessible surfaces each 10-year interval.

The BWRVIP requires an EVT-1 visual examination or UT of shroud vertical welds every 10-years minimum, as compared to the ASME Section XI requirement (VT-3).

For Shroud Repairs, the BWRVIP requires a VT-3 visual examination and other appropriate techniques to examine the Tie-Rod repair hardware every ten years.

Therefore, the BWRVIP referenced examinations are the same or superior to ASME Section XI requirements. Shroud vertical welds and repair Tie-Rod examinations are recommended in BWRVIP-76-R1-A and have the same basic VT-3 method of visual examination or better, the same examination frequency (each 10-year interval) and comparable flaw evaluation criteria. Therefore, the BWRVIP requirements provide a level of quality and safety equivalent to that provided by ASME Section XI.

For other core support structure components, the BWRVIP requires an EVT-1 visual examination or UT of core support structures.

Summary The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than accessible surfaces.

The BWRVIP examination methods (EVT-1 or UT) are superior to the ASME Section XI required VT-3 visual examination for flaw detection and characterization. In most cases, the BWRVIP examination frequency is equivalent to or more frequent than the examination frequency required by ASME Section XI. In cases where the BWRVIP examination frequency is less frequent than required by ASME Section XI, the BWRVIP examinations are performed in a more comprehensive manner and focus on the areas most vulnerable. Therefore, the superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency, or with a less frequent examination frequency but with those examinations being performed in a more comprehensive manner, and using comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that required by the ASME Section XI requirements.

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

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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: 64 RPV Threads in Flange

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) 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)(1), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety. Exelon Generation Company, LLC (EGC) 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 CPS.

EGC has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in the United States (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 evaluates potential degradation mechanisms and includes a stress analysis / flaw tolerance evaluation, and includes a survey of inspection results from over 168 units, a review of operating experience (OE) related to RPV flange/bolting. 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.

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

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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 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 / indications),

then subsequent inservice 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 preservice examinations and over 10,000 inservice 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 ASME 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 threads in flange component as input to a flaw tolerance evaluation.

Sixteen nuclear plant units (10 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 CPS as compared to the values used in the evaluation of the bounding preload stress are shown in Table 1. Specifically, the Reference 1 preload stress is 42,338 psi, whereas the preload stress is 26,567 psi at CPS. As shown in Table 1, the preload stress used in the analysis is also bounding compared to that at CPS. As can be seen from this table, the diameter of the stud used in the analysis is

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 3 of 14) smaller and has fewer studs compared to those at CPS. The smaller stud diameter and the fewer number of studs results in higher preload per bolt. Dimensions of the analyzed geometry are shown in Figure I4R-07-1. The CPS preload stress is bounded by the Reference 1 report which demonstrates that the report remains applicable to this relief request.

For comparison purposes, the global force per flange stud can be estimated by the pressure force on the flange (p**r2, where p is 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 741 kips per stud for the CPS configuration, indicating that the configuration used in the analysis bounds that at CPS.

The specifications for the threads and thread geometry for CPS as compared to that used in the analysis in Reference 1 is shown in Table 2. As this table shows, the flange hole diameter used in the analysis is slightly larger than those at CPS. The larger hole diameter results in a smaller remaining ligament between holes, and is therefore conservative. As can be seen from Table 2, the pitch of the threads used in the analysis is identical to the pitch of the threads for CPS. For CPS, the depth of the thread is slightly larger than that used in the analysis. However, considering the margins in the analysis for this plant, the minor difference is considered negligible. Hence the thread geometry used in the analysis is representative of the thread geometry for CPS.

Dimensions of the analyzed geometry are shown in Figure I4R-07-1.

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

Installed Evaluated (inches) Hole (inches)

(inches)

Clinton Power 64 64 6.25 218.5 12.87 1264.7 26,567 Station, Unit 1 Values Used in 54 54 6.0 173 16 2500 42,338 Bounding Analysis Table 2: RPV Flange Thread Geometry Nominal Bolt Thread Thread Plant Hole Diameter in Pitch Depth Specification Flange (inches) (inches)

Clinton Power Station, Unit 1 6.25"-8N-2B 6.25 8 0.06765 Analysis Geometry per EPRI 7"-8N-2B 7.00 8 0.06500 Report 3002007672

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

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The analytical model is shown in Figures I4R-07-2 and I4R-07-3. The loads considered in the analysis consisted of:

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

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

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 threads in flange component for the three loads described above.

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 Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Section XI, IWB-3600 was performed.

Stress intensity factors 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 (a/t) 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 I4R-07-4 for the flaw model with a/t = 0.77 a/t 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 3 for the four crack depths. From Table 3, 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.

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

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Table 3: Maximum K vs. a/t K at Crack Depth (ksiin)

Load 0.02 a/t 0.29 a/t 0.55 a/t 0.77 a/t 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:

KI < KIc/10 = 69.6 ksiin

Where, KI = Allowable stress intensity factor (ksiin)

KIc = Lower bound fracture toughness at operating temperature (220 ksiin)

As can be seen from Table 3, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of a/t = 0.77. Hence the allowable flaw depth of the 360o circumferential flaw is at least 77% of the thickness of the flange. The 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 a/t = 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 K and 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 4 below provides the RPV flange RTNDT values and the bolt-up temperatures for CPS. These were determined using the RTNDT value from plant records. As can be seen from this table, the minimum (T-RTNDT) is 66°F for CPS. From the equations in paragraph A-4200 of ASME Section XI, Appendix A, the corresponding value of KIc is 111 ksiin. Using a structural factor of 10, the allowable KIc value is 35 ksiin. This value is more than the maximum stress intensity factor (KI) for the preload condition of 17.4 ksiin shown in Table 3, thus the report evaluation is bounding for CPS.

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 6 of 14)

Table 4: RPV Flange RTNDT and Bolt-Up Temperature Preload Temp Minimum Plant Name Flange RTNDT (°F)

(°F) T-RTNDT (°F)

Clinton Power Station, 0 70 66 Unit 1 The stress analysis / flaw tolerance evaluation presented above shows that the threads in flange component at CPS is very flaw tolerant and can operate for 80 years without violating ASME Section XI safety margins. This clearly demonstrates that the threads in flange examinations can be eliminated without affecting the safety of the RPV.

Operating Experience Review Summary As discussed above, the results of the survey, which includes results from CPS, 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 5 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 5: 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

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 7 of 14) 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.

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 examination, CPS proposes that the industry report (Reference 1) provides an acceptable technical basis for eliminating the requirement for the RPV threads in flange examination.

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. The CPS specific parameters (e.g., vessel diameter, number of studs, inservice inspection findings) have been confirmed to be consistent with or bounded by Reference 1.

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

To protect against non-service related degradation, CPS 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

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 8 of 14) removed, and the procedures document each step. These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service.

Procedure MA-CL-716-102, "Reactor Disassembly" (Reference 7) provides steps necessary to remove the drywell head, reactor pressure vessel (RPV) head piping and insulation, RPV head (EIN: 1B13D003) and RPV head studs, steam dryer (EIN: 1B13D005) assembly, shroud head & steam separator (EIN: 1B13D004) assembly, and install main steam line plugs. This procedure also provides instructions for removal of associated pipe supports and closing of drywell bulkhead penetrations.

The procedure requires verification that the RPV stud threads and RPV head flange, including around nuts and washers, are free of debris prior to de-tensioning and cleaned.

Additionally, the procedure requires an inspection of the RPV studs and RPV flange for debris, and cleaning; and requires cleaning as necessary, and inspection of nut and stud threads for nicks, marks, or signs of galling.

Procedure MA-CL-716-103, "Reactor Assembly" (Reference 8) provides instructions for installation of steam dryer (EIN: 1B13D005) assembly, shroud head/steam separator (EIN: 1B13D004) assembly and for removal of main steam line plugs. This procedure also provides steps necessary to install the reactor pressure vessel (RPV) head (EIN: 1B13D003), to tension RPV head studs, to install the reactor head insulation and piping, and to install the drywell head. MA-CL-716-103 provides guidance for the cleaning, inspection, and lubrication the RPV head flange and components (studs, nuts, and washers) of all dirt and granular type material by brushing, blasting with air (except O-ring grooves), wiping with lint free wipes, or by using either demineralized water or approved cleaning fluid. Additionally, the procedure requires cleaning RPV stud holes and lubricating threads with approved thread lubricant.

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 requirements in this relief request are based upon ASME Section XI Code Case N-864 (N-864) (Reference 4) 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 1," and thus, is not available for application at nuclear power plants without specific NRC approval.

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 9 of 14)

6. Duration of Proposed Alternative Relief is requested for the Fourth ISI Interval for CPS, or until the NRC approves N-864, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents

Clinton Power Station, Unit 1, Third ISI Interval relief request was authorized by NRC Safety Evaluation (SE) dated June 26, 2017 (Reference 3). This Clinton Power Station, Unit 1 relief request was part of an EGC 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 the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.

Peach Bottom Atomic Power Station, Units 2 and 3, Relief Request I5R-06 was authorized by NRC SE dated December 21, 2018 (ADAMS Accession No. ML18331A216) (Reference 5).

Nine Mile Point Nuclear Station, Units 1 and 2, Relief Request I5R-05/I4R-05 was authorized by NRC SE dated December 21, 2018 (ADAMS Accession No. ML18334A236) (Reference 6).

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 (EGC), 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. ML17170A013)

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

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 10 of 14)

5) Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Alternative Requests Related to the Fifth Inservice Inspection Interval (EPID L-2018-LLR-0055, EPID L-2018-LLR-0057, EPID L-2018-LLR-0058, and EPID L-2018-LLR-0059)," dated December 21, 2018 (ADAMS Accession No. ML18331A216)

6) Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "Nine Mile Point Nuclear Station, Units 1 and 2 - Alternative to the Requirements of the ASME Code (EPID L-2018-LLR-0087)," dated December 21, 2018 (ADAMS Accession No. ML18334A236)

7) Clinton Power Station, Unit 1 Procedure MA-CL-716-102, "Reactor Disassembly," Revision 14

8) Clinton Power Station, Unit 1 Procedure MA-CL-716-103, "Reactor Assembly,"

Revision 14

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 11 of 14)

Figure I4R-07-1 Modeled Dimensions

10 CFR 50.SSa Relief Request 14R-07 Alternative to Use ASME Code Case N-864, Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 12 of 14)

Figure 14R-07-2 Finite Element Model Showing Bolt and Flange Connection 1

ELEMENTS REAL NOM ANO_Vessel_Flange

10 CFR 50.55a Relief Request I4R-07 Alternative to Use ASME Code Case N-864, 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)(1)

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

Revision 0 (Page 13 of 14)

Figure I4R-07-3 Finite Element Model Mesh with Detail at Thread Location

10 CFR 50.SSa Relief Request 14R-07 Alternative to Use ASME Code Case N-864, Examination of ASME Section XI, Examination Category B-G-1, Item Number 86.40, Threads in Flange in Accordance with 10 CFR 50.55a(z)(1)

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

Revision 0 (Page 14 of 14)

Figure 14R-07-4 Cross Section of Circumferential Flaw with Crack Tip Elements Inserted After 10th Thread from Top of Flange

v t e Letter Sequence Request
EPID:L-2019-LLR-0112, Relief Requests Associated with the Fourth Inservice Inspection Interval (Open)
Initiation Request Acceptance... Administration Questions Answers 13 May 2020 Results Approval... Other: ML20218A660
MONTHYEAR2020-05-13 [Table View]

RS-19-116, Relief Requests Associated with the Fourth Inservice Inspection Interval

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RS-19-116, Relief Requests Associated with the Fourth Inservice Inspection Interval

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