NL-18-004, Browns Station - American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section Xi, Inservice Inspection (ISI) Program, Unit 1 Third Ten-Year Inspection Interval, Unit 2 Fifth Ten-Year Inspection Interval, and Unit 3 Four

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Browns Station - American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section Xi, Inservice Inspection (ISI) Program, Unit 1 Third Ten-Year Inspection Interval, Unit 2 Fifth Ten-Year Inspection Interval, and Unit 3 Fourt
ML18135A357
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 05/11/2018
From: Shea J W
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-18-004
Download: ML18135A357 (12)
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Text

Tennessee Valley Authority , 1101 Market Street, Chattanooga, Tennessee 37402

CNL-18-004 May 11 , 2018 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR

-68 NRC Docket No. 50-259, 50-260, and 50

-296

Subject:

American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection (ISI) Program , Unit 1 Third Ten-Year Inspection Interval , Unit 2 Fifth Ten

-Year Inspection Interval, and Unit 3 Fourth Ten-Year Inspection Interval Request for Alternative for ISI-46

References:

1. NRC Letter to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Safety Evaluation for Relief Request ISI

-23 (TAC Nos. ME3396, ME3397, and ME3398)," dated October 28, 2010 (ML102440565) 2. EPRI Technical Report 1021005, BWRVIP

-241, "Probabilistic Fracture Mechanics Evaluation for the Boling Water Reactor Nozzle

-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2010 (ML11119A041) 3. NRC Letter to Chairman, EPRI BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)

-241 Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,' (TAC No. ME6328)," dated April 19, 2013 (ML13071A240)

In accordance with 10 CFR 50.55a(z)(1), Tennessee Valley Authority (TVA) is requesting an alternative, for Nuclear Regulatory Commission (NRC) review and approval, from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," Sub-Article IWB

-2500 , "Examination and Pressure Test Requirements." Enclosure 1 contains alternative request ISI

-46 that proposes an alternative to the requirements contained in ASME Section XI, Sub-Article IWB

-2500, Table IWB-2500-1, "Examination U.S. Nuclear Regulatory Commission CNL-18-004 Page 2 May 11 , 2018 Categories

," to allow reduced percentage requ i rements for nozzle-to-vessel weld and inner radius section examinations for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 based on Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI , Division 1." This alternative to the Section XI requirements is requested for the remainders of the BFN Unit 1 Third Ten-Year Inspection Interval , the BFN Unit 2 Fifth Ten-Year Inspection Interval , and the BFN Unit 3 Fourth Ten-Year Inspection Interval.

A similar alternative request was granted by the NRC for BFN Units 1, 2, and 3 in Reference

1. The spec i fic components affected by this request are provided in Enclosure 2. Enclosure 3 provides the TVA response to Boiling Water Reactor Vessel and Internals Project (BWRVIP)-241 (Reference
2) plant specific applicability criteria.

The NRC provided a Safety Evaluation (Reference

3) approving the generic technical basis and acceptability cr iteria for application of ASME Code Case N-702 , to wh i ch TVA conforms as described in Enclosure 1. The plant specific applicab i lity of the BWRVI P-241 report to BFN is demonstrated in Enclosure
3. TVA has determined that this alternative request provides an acceptable level of quality and safety in accordance with 10 CFR 50.55(a)(z)(1 ). TVA requests approval of these relief requests within one year from the date of this letter. There are no new regulatory commitments associated with this submittal.

Please address any questions regarding this request to Edward D. Schrull at 423-751-3850. Respectfully, J. W. Shea Vice President , Nuclear Regulatory Affairs & Support Services Enc l osures: 1. Request for Alternative, Use of Code Case N-702 in Lieu of Specific ASME Code Section XI Requirements ISl-46 2. BFN Units 1, 2, and 3 Request for Alternative ISl-46 -Table of Affected Components 3. Response to BWRVIP-241 Plant Specific Applicability Criteria cc (w/Enclosures)

NRC Regional Administrator

-Region II NRC Senior Resident Inspector

-Browns Ferry Nuclear Plant NRC Project Manager -Browns Ferry Nuclear Plant CNL-18-004 E1-1 of 5 Browns Ferry Nuclear Plant Units 1, 2, and 3 American Society of Mechanical Engineers, Section XI Inservice Inspection Program, Unit 1 Third

-Ten-Year Inspection Interval, Unit 2 Fifth

-Ten-Year Inspection Interval, and Unit 3 Fourth

-Ten-Year Inspection Interval Request for Alternative Use of Code Case N

-702 in Lieu of Specific ASME Code Section XI Requirements ISI-46 10 CFR 50.55a(z)(1)

--Acceptable Level of Quality and Safety

-- 1.0 American Society of Mechanical Engineers (ASME) Code Component(s) Affected Code Class:

ASME Section XI Code Class 1 Component Numbers:

Various (See Table 1 and Enclosure 2 for detailed list of components)

Code

References:

ASME Section XI, 2007 Edition through 2008 Addenda Code Case N

-702 Examination Category:

B-D Item Number(s):

B3.90 and B3.100 Unit / Inspection Interval Applicability:

Browns Ferry Nuclear Plant (BFN) Unit 1 Third

-Ten-Year Inservice Inspection (ISI) Interval, BFN Unit 2 Fifth-Ten-Year ISI Interval, and BFN Unit 3 Fourth-Ten-Year ISI Interva l 2.0 Applicable Code Edition and Addenda ASME Section XI, 2007 Edition through the 2008 Addenda (Reference 1)

3.0 Applicable

ASME Code Requirements ASME Section XI, 2007 Edition through the 2008 Addenda, Sub-Article IWB

-2500, "Examination and Pressure Test Requirements," Table IWB

-2500-1, Examination Category B

-D, "Full Penetration Welded Nozzles in Vessels" requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 10

-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Mandatory Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems,"

is implemented.

CNL-18-004 E1-2 of 5 The plant, interval, ASME Section XI Code Edition and Addenda, interval start date, and interval end date are provided in Table 1. Table 1 Plant Interval Edition Start Date End Date BFN Unit 1 Third 2007 Edition through the 2008 Addenda February 1, 2016 January 31, 2026 BFN Unit 2 Fifth 2007 Edition through the 2008 Addenda February 1, 2016 January 31, 2026 BFN Unit 3 Fourth 2007 Edition through the 2008 Addenda February 1, 2016 January 31, 2026

4.0 Reason for Request

Regulatory Guide (RG) 1.147, Revision 18 (Reference 2), conditionally accepts the use of Code Case N-702 , "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle

-to-Shell WeldsSection XI, Division 1" (Reference 3)

. C ode Case N-702 provides an alternative to performing examination of 100 percent (%) of the nozzle-to-vessel welds and inner radii for examination category B-D nozzles with the exception of the feedwater nozzles 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.

As noted in RG 1.147 , the applicability of Code Case N

-702 must be shown by demonstrating that the criteria in Section 5.0 of the Nuclear Regulatory Commission (NRC) Safety Evaluation regarding Boiling Water Reactor Vessel Internals Project (BWRVIP) -108 (Reference 4) or Section 5.0 of the NRC Safety Evaluation regarding BWRVIP

-241 (Reference 5) are met. RG 1.147 also states, "The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case." The evaluation required by RG 1.147 regarding Code Case N

-702 is provided in Section 5 of this enclosure and Enclosure 3. 5.0 Proposed Alternative and Basis for Use

5.1 Proposed

Alternative

Pursuant to 10 CFR 50.55a(z)(1), BFN, Units 1, 2 and 3 requests approval to implement the alternative of Code Case N

-702 in lieu of the code

-required 100% examination of all nozzles and inner radii identified in Table 1. Enclosure 2 provides a detailed list of the nozzles and inner radii subject to this request. As an alternative, for the nozzle-to-shell welds and inner radii identified in Table 2 below , TVA proposes to examine a minimum of 25% of the nozzle

-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N

-702. TVA was previously granted an alternative request for the previous 10

-year intervals for BFN in Reference 6.

CNL-18-004 E1-3 of 5 The reactor pressure vessel (RPV) nozzle

-to-vessel welds and inner radii subject to this request are listed in Table

2. A detailed list of the affected components is provided in Enclosure 2.

Table 2 Identification Number Description Total Number Minimum Number to be examined N2 Recirculation Inlet 10 3 N3 Main Steam Outlet 4 1 N5 Core Spray 2 1 N6 Head Spray 2 1 N7 Head Vent 1 1 N8 Jet Pump Instrumentation 2 1 The exams in Table 2 will be scheduled in accordance with Table IWB-2411-1. As permitted by Code Case N

-702, a VT-1 visual examination of Item Nos. B3.20 1 and B3.100 may be performed in lieu of a volumetric examination.

In the event TVA elects to apply the alternative VT

-1 examination for BFN , TVA will meet the NRC conditions specified for Code Case N-648-1 , "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1

," in R G 1.147, Revision 18, Table 2, as follows. "In lieu of a UT examination, licensees may perform a VT

-1 examination in accordance with the code of record for the inservice Inspection Program utilizing the allowable flaw length criteria of Table IWB

-3512-1 with limiting assumptions on flaw aspect ratio."

Based on the above and the evaluation in Enclosure 3, TVA conforms to the conditions of RG 1.147 and the criteria specified in Code Case N

-702. 5.2 Basis for Request As noted above, the conditions in RG 1.147 specify that the applicability of Code Case N

-702 must be shown by demonstrating that the criteria in Section 5.0 of Reference 7 or Section 5.0 of Reference 8 are met. Enclosure 3 provides the evaluation of the criteria in Reference 8.

There have not been any recordable indications of the components for which an alternative is requested since NRC's approval of the previous similar alternative request for BFN Units 1, 2, and 3 (ISI-23) for previous ten

-year intervals (Reference 6). The examination history for examinations performed prior to NRC approval of ISI

-23 is contained in Reference 6. 6.0 Duration of Proposed Alternative The duration of this alternative request is for the remainder of the current ten

-year intervals for BFN Units 1, 2, and 3 as shown in Table 1.

1 Note that Item No. B3.20 was changed to B3.100 in the 2007 Edition of Section XI.

CNL-18-004 E1-4 of 5 7.0 Conclusion Based on the information in Enclosures 1, 2, and 3, the affected components for which an alternative is requested, meet the conditions of R G 1.147, Revision 18. Therefore, the proposed alternative to use the ASME Code Case N

-702 in lieu of the ASME Section XI requirements in the 2007 Edition with the 2008 Addenda is acceptable, and provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

8.0 Precedents

In addition to Reference 6 , this alternative request is similar to the following NRC Safety Evaluations that approved the use of ASME Code Case N

-702: "Peach Bottom Atomic Power Station , Units 2 and 3 - Requests for Relief 14R-51 and 14R-52 (TAC N os. ME5392, ME5393, ME5394 AND ME5395)," dated January 24, 2012 (ML112770217) "James A. Fitzpatrick Nuclear Power Plant - Safety Evaluation for Request for Relief Using the Requirements of the ASME Code Case N

-702 and BWRVIP

-241 for Plant Nozzle-to-Vessel Welds and Nozzle Inner Radii (CAC N o. MF8301)," dated December 6, 2016 (ML16334A440)

"Dresden Nuclear Power Station , Units 2 and 3 - Issuance of Safety Evaluation for Request Re: Inservice Inspection Interval Proposed Alternative (15R-08) (CAC N os. MF8090 and MF8091)," dated June 28, 2017 (ML17073A121)

"Hope Creek Generating Station - Relief from the Requirements of the ASME Code for Alternative to Nozzle

-to-Vessel Weld and Inner Radius Examinations (CAC No. MF9554)," dated August 17, 2017 (ML17223A483)

"Quad Cities Nuclear Power Station, Units 1 And 2 - Alternative to the Requirements of the ASME Code Regarding Reactor Pressure Vessel Nozzle Assemblies- Relief Request 15R-07 (C AC Nos. MF8989 and MF8990) (RS 256)," dated August 25, 2017 (ML17221A264)

CNL-18-004 E1-5 of 5 9.0 References

1. ASME Code,Section XI, 2007 Edition with the 2008 Addenda
2. Regulatory Guide 1.147 Revision 18, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," dated March 2017 (ML16321A336)
3. ASME 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 letter to 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 (ML073600374) 5. NRC Letter to Chairman, EPRI BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)

-241 Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle

-to-Vessel Shell Welds and Nozzle Blend Radii,' (TAC No. ME6328)

," dated April 19, 2013 (ML13071A240)

6. NRC Letter to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Safety Evaluation for Relief Request ISI

-23 (TAC Nos. ME3396, ME3397, and ME3398)," dated October 28, 2010 (ML102440565)

7. EPRI Technical Report 1003557, BWRVIP-108, "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 (ML023330203) 8. EPRI Technical Report 1021005, BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boling Water Reactor Nozzle

-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2010 (ML11119A041)

CNL-18-004 E2-1 of 2 BFN Units 1, 2, and 3 Request for Alternative ISI

-46 Table of Affected Components Component No. Category Item Number

  • Description N2A-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle-to-Vessel Weld N2A-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2B-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2B-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2C-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2C-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2D-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2D-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2E-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2E-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2F-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2F-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2G-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2G-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2H-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2H-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2J-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2J-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N2K-NV B-D B3.90 11.56" Reactor Recirculation Inlet Nozzle

-to-Vessel Weld N2K-IR B-D B3.100 11.56" Reactor Recirculation Inlet Nozzle Inside Radius Section N3A-NV B-D B3.90 23.75" Main Steam Nozzle-to-Vessel Weld N3A-IR B-D B3.100 23.75" Main Steam Nozzle Inside Radius Section CNL-18-004 E2-2 of 2 BFN Units 1, 2, and 3 Request for Alternative ISI

-46 Table of Affected Components Component No. Category Item Number

  • Description N3B-NV B-D B3.90 23.75" Main Steam Nozzle-to-Vessel Weld N3B-IR B-D B3.100 23.75" Main Steam Nozzle Inside Radius Section N3C-NV B-D B3.90 23.75" Main Steam Nozzle-to-Vessel Weld N3C-IR B-D B3.100 23.75" Main Steam Nozzle Inside Radius Section N3D-NV B-D B3.90 23.75" Main Steam Nozzle-to-Vessel Weld N3D-IR B-D B3.100 23.75" Main Steam Nozzle Inside Radius Section N5A-NV B-D B3.90 8.78" Core Spray Nozzle-to-Vessel Weld N5A-IR B-D B3.100 8.78" Core Spray Nozzle Inside Radius Section N5B-NV B-D B3.90 8.78" Core Spray Nozzle-to-Vessel Weld N5B-IR B-D B3.100 8.78" Core Spray Nozzle Inside Radius Section N6A-NV B-D B3.90 6.22" Head Vent Nozzle-to-Vessel Weld N6A-IR B-D B3.100 6.22" Head Vent Nozzle Inside Radius Section N6B-NV B-D B3.90 6.22" Head Vent Nozzle-to-Vessel Weld N6B-IR B-D B3.100 6.22" Head Vent Nozzle Inside Radius Section N7-NV B-D B3.90 4.25"Head Vent Nozzle-to-Vessel Weld N7-IR B-D B3.100 4.25" Head Vent Nozzle Inside Radius Section N8A-NV B-D B3.90 3.81" Jet Pump Instrumentation Nozzle-to-Vessel Weld N8A-IR B-D B3.100 3.81" Jet Pump Instrumentation Inside Radius Section N8B-NV B-D B3.90 3.81" Jet Pump Instrumentation Nozzle-to-Vessel Weld N8B-IR B-D B3.100 3.81" Jet Pump Instrumentation Inside Radius Section
  • All nozzle sizes are nominal.

CNL-18-004 E3-1 of 3 RESPONSE TO BWRVIP

-108 PLANT SPECIFIC APPLICABILITY CRITERIA Regulatory Guide (RG) 1.147, Revision 18 (Reference 1) conditionally accepts the use of ASME Code Case N

-702 (Reference 2) with the following condition: The applicability of Code Case N-702 must be shown by demonstrating that 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) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the application of the Code Case.

BWRVIP-108 (Reference3) contains the technical basis supporting ASME Code Case N

-702 for reducing the inspection of RPV nozzle

-to-vessel shell welds and nozzle inner radius areas from 100 percent (%) to 25% of the nozzles for each nozzle type during each 10

-year interval at the Browns Ferry Nuclear Plant (BFN). As noted in Reference 3, the failure probability of the nozzles due to a low temperature overpressure (LTOP) event at the nozzle blend radius region and nozzle

-to-vessel shell weld is very low (i.e., less than 1E

-6 for 40 years) with or without inservice inspection. Reference 3 also states that inspection of 25% of each nozzle type is technically justified.

Additionally, BWRVIP

-241 (Reference 4) provides supplemental analyses to BWRVIP

-108 for BWR reactor pressure vessel (RPV) recirculation inlet and outlet nozzle

-to-shell welds and nozzle inner radii that also apply to BFN.

Technical documents BWRVIP

-108 (Reference 3) and BWRVIP

-241 (Reference 4) provide the basis for the code case, but only consider 40

-year plant operation. To extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PFM) evaluation, consistent with the methods of BWRVIP

-108 and BWRVIP

-241, was performed to ensure that the probability of failure (PoF) remains acceptable. Considering the additional thermal cycles and fluence a t the end of the period of extended operation (i.e., December 20, 2033, for BFN Unit 1, June 28, 2034, for BFN Unit 2, and July 2, 2036, for BFN Unit 3), the limiting PoF due to an LTOP event is 1.53E

-6 per year for the nozzle blend radius, and 8.33E

-12 per year for the nozzle-to-shell weld. These values are less than the Nuclear Regulatory Commission (NRC) limit of 5.0E

-6 failures per year in accordance with NUREG

-1806, Volume 1 (Reference 5). Therefore, the PFM evaluation concluded that, after consideration of the additional thermal cycles and fluence at the end of the period of extended operation, the N1 nozzles (including other applicable nozzles that are bounded by the N1 nozzles) are qualified for reduced inspection using ASME Code Case N

-702 through the end of the period of extended operation. Because the period of extended operation for BFN Units 1, 2, and 3 is beyond the end of the current ten

-year intervals for BFN Units 1, 2, and 3 (see Table 1 to Enclosure 1), the PFM evaluation is conservative for this alternative request.

The NRC safety evaluation (SE) for BWRVIP

-241 (Reference 6) states:

"Licensees who plan to request relief from the ASME Code Section XI requirements for RPV nozzle

-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP 241 report as the technical basis for use of ASME Code Case N

-702 as an alternative."

CNL-18-004 E3-2 of 3 Section 5.0 of Reference 6 list five criteria that licensees need to address to demonstrate the plant-specific applicability of the BWRVIP

-241 report to their units. Each criterion is addressed below: Criterion 1 states, "The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour

." BFN Units 1, 2 and 3 Technical Specification (TS) Surveillance Requirement (SR) 3.4.9.1.b verifies that the reactor coolant system (RCS), heatup and cooldown rates are 100°F in any one-hour period. Therefore, Criterion 1 is satisfied. Evaluation of Criteria 2 through 5 of Reference 6 is provided below in Table 1. Table 1 Evaluation of Criteri a 2, 3, 4, and 5 Recirculation Inlet Nozzles (N2)

Recirculation Outlet Nozzles (N1)

Criterion 2:

(pr/t)/CRPV 1.15 Criterion 4:

(pr/t)/CRPV 1.15 p = RPV normal operating pressure (psig) 1035 p = RPV normal operating pressure (psig) 1035 r = RPV inner radius (inch) 125.688 r = RPV inner radius (inch) 125.688 t = RPV wall thickness (inch) 6.125 t = RPV wall thickness (inch) 6.125 C RP V 19332 CRPV 16171 (pr/t)/CRPV ~1.10 (pr/t)/CRPV ~1.31 ~ Criteri on is met ~1.31 1.15 Criteri on Not Satisfied Criterion 3:

[p(r o 2 + r i 2) / (r o 2 - r i 2)]/CNOZZLE 1.47 Criterion 5:

[p(r o 2 + r i 2) / (r o 2 - r i 2)]/CNOZZLE 1.59 p = RPV normal operating pressure (psig) 1035 p = RPV normal operating pressure (psig) 1035 r o = nozzle outer radius (inch) 12.5 r o = nozzle outer radius (inch) 26.5 r i = nozzle inner radius (inch) 5.941 r i = nozzle inner radius (inch) 15.566 CNOZZLE 1637 CNOZZLE 1977 [p(r o 2 + r i 2) / (r o 2 - r i 2)]/CNOZZLE ~1.00 [p(r o 2 + r i 2) / (r o 2 - r i 2)]/CNOZZLE ~1.08 ~1.047 Criteri on is met ~1.08 59 Criteri on is met As shown above in Table 1, the BFN Units 1, 2 and 3 N1 nozzles do not meet Criterion 4 of Reference 6, and are not in the scope of this alternative request. Based upon the above information, the RPV nozzle

-to-vessel shell welds and nozzle inner radii sections, with the exception of the N1 nozzles, meet the BWRVIP

-241 report criteria. Therefore , Code Case N-702 is applicable.

CNL-18-004 E3-3 of 3 References

1. Regulatory Guide (RG) 1.147 Revision 18, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," dated March 2017 (ML16321A336)
2. ASME 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

3. EPRI Technical Report 1003557, BWRVIP-108, "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 (ML023330203) 4. EPRI Technical Report 1021005, BWRVIP-241, "Probabilistic Fracture Mechanics Evaluation for the Boling Water Reactor Nozzle

-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2010 (ML11119A041) 5. NUREG 1806 Volume 1 , "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)," dated August 2007 (ML072830076)

6. NRC Letter to Chairman, EPRI BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP)

-241 Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle

-to-Vessel Shell Welds and Nozzle Ble nd Radii,' (TAC No. ME6328)

," dated April 19, 2013 (ML13071A240)

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