RS-17-072, Fourth 10-Year Interval Inservice Inspection Program Relief Requests

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Fourth 10-Year Interval Inservice Inspection Program Relief Requests
ML17150A449
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 05/30/2017
From: Gullott D
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-17-072
Download: ML17150A449 (98)
v • d • e


Text

4300 Winfield Road Warrenville. IL 60555 Exelon Generation@ 630 657 2000 Office RS-17-072 10 CFR 50.55a May 30, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

LaSalle County Station, Units 1 and 2, Fourth 10-Year Interval lnservice Inspection Program Relief Requests In accordance with 10 CFR 50.55a, "Codes and standards," paragraphs (z)(1) and (z)(2),

Exelon Generation Company, LLC (EGC) requests NRC approval of the attached relief requests associated with the fourth lnservice Inspection (ISi) interval for LaSalle County Station (LSCS),

Units 1 and 2. The fourth interval of the LSCS ISi Program is currently scheduled to begin on October 1, 2017, and end on S.eptember 30, 2027, and will comply with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda. EGC proposes the following relief requests for the LSCS Fourth 10-year ISi interval:

  • 14R-01 requests approval of alternative risk-informed ISi program and examination criteria for Examination Category 8-F, 8-J, C-F-1, and C-F-2 pressure retaining piping welds in accordance with ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1. 11
  • 14R-04 requests approval of alternative post-tensioning system inspection system inspection scheduling requirements for sites with two plants.
  • 14R-08 requests approval of alternative examination requirements for the hydrogen recombiner system piping.

May 30, 2017 U.S. Nuclear Regulatory Commission Page2

  • 14R-09 requests alternative examination requirements for the nozzle-to-vessel welds and inner radii sections.

The bases for these relief requests are provided in Attachments 1 through 8, respectively.

Relief Requests similar to 14R-01, 14R-02, 14R-03, 14R-04, 14R-06, 14R-07, 14R-08, and 14R-09 have previously been approved for use at LSCS for the Third 10-Year Interval. EGC is requesting NRC approval of the relief requests by January 19, 2018, to support implementation of the LSCS Fourth 10-year ISi interval prior to the LSCS Unit 1 spring 2018 refueling outage (L 1R17).

There are no regulatory commitments contained within in this letter. Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments:

1) 10 CFR 50.55a Relief Request 14R-01
2) 10 CFR 50.55a Relief Request 14R-02
3) 10 CFR 50.55a Relief Request 14R-03
4) 10 CFR 50.55a Relief Request 14R-04
5) 10 CFR 50.55a Relief Request 14R-06
6) 10 CFR 50.55a Relief Request 14R-07
7) 10 CFR 50.55a Relief Request 14R-08
8) 10 CFR 50.55a Relief Request 14R-09 cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, LaSalle County Station

ATTACHMENT 1 10 CFR 50.SSa Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

(Page 1 of 25)

1. ASME Code Component(s) Affected Code Class: 1and2

Reference:

Table IWB-2500-1, Table IWC-2500-1 Examination Category: 8-F, 8-J, C-F-1, and C-F-2 Item Number: 85.10, 89.11, 89.21, 89.31, 89.32, 89.40, C5.11, C5.21, C5.51, and C5.81

Description:

Alternate Risk-Informed Selection and Examination Criteria for Examination Category 8-F, 8-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Component Number: Units 1 and 2 Pressure Retaining Piping

2. Applicable Code Edition and Addenda The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.
3. Applicable Code Requirement Table IWB-2500-1, Examination Category 8-F, requires volumetric and surface examinations on all welds for Item Number 85.10.

Table IWB-2500-1, Examination Category 8-J, requires volumetric and surface examinations on a sample of welds for Item Numbers 89.11 and 89.31, and surface examinations on a sample of welds for Item Numbers 89.21, 89.32, and 89.40. 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 8-F.

ATTACHMENT 1 10 CFR 50.SSa Relief Request 14R*01 Proposed Alternative 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 (1) 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, C5.21, 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;

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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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 lnservice Inspection Evaluation Procedure,"

Revision B-A (Reference 1) along with two enhancements from ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1, 11 (Reference 4) will provide an acceptable level of quality and safety.

As stated in "Safety Evaluation Report Related to EPRI Risk-Informed lnservice 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 in accordance with 10 CFR 50.55a for the proposed alternative to the piping /SI requirements with regard to the number of locations, locations of inspections, and methods of inspection. "

The initial LaSalle County Station Risk Informed lnservice Inspection (RI-ISi) Program was submitted during the Second Period of the Second ISi Interval for Units 1 and 2.

This initial RI-ISi Program was developed in accordance with EPRI TR-112657, Revision B-A, as supplemented by ASME Code Case 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 December 27, 2001 (Reference 5).

The LaSalle County Station RI-ISi Program was resubmitted using the same approach during the Third ISi Interval for Units 1 and 2. The program was approved for use by the NRC via SE as transmitted to EGC on April 29, 2008 (Reference 6).

The transition from the 2001 Edition through the 2003 Addenda to the 2007 Edition with the 2008 Addenda of ASME Section XI for LaSalle County Station'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/Addenda will be implemented as detailed in the LaSalle County Station ISi Program Plan. Therefore, with the exception of specific weld locations that may have changed due to maintenance or modification activities (e.g., Fukushima FLEX modification), the proposed alternative

ATIACHMENT1 10 CFR 50.55a Relief Request 14R*01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

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RI-ISi Program for the 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 LaSalle County Station 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 Section XI Program to the Fourth Interval RI-ISi Program was within the 1.00E-06 and 1.00E-07 acceptance criteria for delta-core damage frequency (Delta-GDF) and delta-large early release frequency (Delta-LERF) as described in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. 11 The Delta-CDF and Delta-LERF values for LaSalle County Station, Units 1 and 2 are listed in the following table.

Change in Risk from LaSalle County Station Pre-RI-ISi Section XI Program to Fourth Interval RI-ISi Program Unit No. Delta-CDF Delta-LERF Unit 1 5.26E-09 4.34E-09 Unit 2 6.29E-09 4.94E-09 The following tables document the Delta-GDF and Delta-LERF for LaSalle County Station Units 1 and 2 over the initial ASME Section XI Program for the Fourth ISi Interval. The first two tables provide results for Unit 1. The results for Unit 1 are provided in the first table by system and the second table for only the Break Exclusion Region (BER) weld population. The next two tables provide the equivalent results for Unit 2.

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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LaSalle County Station Unit 1 Delta-GDF and Delta-LEAF by System ACDF ALE RF Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria CRD 4.16E-11 1.00E-07 1.17E-12 1.00E-08 ECCS -1.70E-10 1.00E-07 -6.09E-11 1.00E-08 FW 1.13E-09 1.00E-07 1.01 E-09 1.00E-08 HPCS 1.19E-10 1.00E-07 7.65E-11 1.00E-08 MS 1.08E-09 1.00E-07 1.07E-09 1.00E-08 RCIC 5.56E-10 1.00E-07 5.22E-10 1.00E-08 RCS 1.04E-09 1.00E-07 2.65E-10 1.00E-08 RWCU 1.46E-09 1.00E-07 1.46E-09 1.00E-08 Total 5.26E-09 <1.00E-06 4.34E-09 <1.00E-07 LaSalle County Station Unit 1 BER Weld Delta-GDF and DeIta- LEAF b1y S,ysem t

ACDF ALE RF Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria ECCS -2.80E-10 1.00E-07 -2.80E-10 1.00E-08 FW 9.54E-10 1.00E-07 9.55E-10 1.00E-08 HPCS 1.56E-12 1.00E-07 1.56E-14 1.00E-08 MS 1.01 E-09 1.00E-07 1.01 E-09 1.00E-08 RCIC 5.48E-10 1.00E-07 5.22E-10 1.00E-08 RWCU 1.46E-09 1.00E-07 1.46E-09 1.00E-08 Total 3.70E-09 <1.00E-06 3.66E-09 <1.00E-07

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

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LaSalle County Station Unit 2 Delta-CDF and I LEAF b1y Siystem Deta-ACDF ALE RF System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria CAD 7.07E-11 1.00E-07 1.46E-12 1.00E-08 ECCS 3.40E-10 1.00E-07 -1.42E-10 1.00E-08 FW 2.86E-09 1.00E-07 2.72E-09 1.00E-08 HPCS -1.36E-12 1.00E-07 2.41 E-11 1.00E-08 MS 4.30E-10 1.00E-07 4.38E-10 1.00E-08 RCIC 4.25E-10 1.00E-07 4.03E-10 1.00E-08 RCS 8.66E-10 1.00E-07 1.81 E-10 1.00E-08 RWCU 1.32E-09 1.00E-07 1.31 E-09 1.00E-08 Total 6.31 E-09 <1.00E-06 4.94E-09 <1.00E-07 LaSalle County Station Unit 2 BER Weld Delta-CDF and DeIta- LEAF b1y S1ysem t

ACDF ALE RF Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria ECCS -3.18E-10 1.00E-07 -3.18E-10 1.00E-08 FW 2.65E-09 1.00E-07 2.65E-09 1.00E-08 HPCS 2.09E-12 1.00E-07 1.04E-14 1.00E-08 MS 3.42E-10 1.00E-07 3.36E-10 1.00E-08 RCIC 4.25E-10 1.00E-07 4.03E-10 1.00E-08 RWCU 1.32E-09 1.00E-07 1.31 E-09 1.00E-08 Total 4.42E-09 <1.00E-06 4.39E-09 <1.00E-07 The actual 11 evaluation and ranking 11 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 portions of the RI-ISi Program have been and will continue to be reevaluated as major revisions of the site Probabilistic Risk Assessment (PAA) occur and modifications to plant configuration are made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, and Element Selection, and Risk Impact Assessment steps encompass the complete living program process applied under the LaSalle County Station RI-ISi Program.

ATIACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

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5. Proposed Alternative and Basis for Use The proposed alternative initially implemented in the LaSalle County Station, Units 1 and 2, "Risk Informed lnservice 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 LaSalle County Station'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 the 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.

a. In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISi Selected Examinations" of EPRI TR-112657, LaSalle County Station will utilize the requirements of Paragraph -2430, "Additional Examinations" contained in ASME Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in ASME Code Case 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 ASME Code Case 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, IWB-2430 and IWC-2430. Additionally, similar to the current requirements of ASME Section XI, LaSalle County Station 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, 11 LaSalle County Station will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations," contained in ASME Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 2007 Edition with the 2008 Addenda of ASME Section XI (i.e., LaSalle County Station's Code of Record for the Fourth ISi Interval) will be utilized which parallel those referenced in the code case. Table 1 of ASME Code Case 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 elements not subject to a degradation mechanism, ASME Code Case N-578-1, Table 1, Note 1 will be applied using the expanded examination volume.

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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The LaSalle County Station RI-ISi Program, as developed in accordance with EPRI TR-112657, Revision B-A (Reference 1), requires that 25% of the elements that are categorized as "High" risk (i.e., Risk Category 1, 2, and 3) and 10% of the 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 ASME Code Case N-578-1.

For NRC staff consideration in the evaluation of this alternative RI-ISi Program, Enclosure LS-LAR-007, Revision Oto this relief request contains a summary of the RG 1.200, Revision 2 (Reference 7) evaluation performed on LS-PSA-014, Revision 9 (Reference 8), and the impact of the identified gaps on the technical adequacy of the LaSalle County Station PAA Model to support this RI-ISi application (see Enclosure, Table 1).

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 LaSalle County Station 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 LaSalle County Station, Units 1 and 2.
7. Precedents
  • LaSalle County Station Third ISi Interval Relief Request 13R-01 was authorized per NRC SE dated April 29, 2008 (ADAMS Accession No. ML080940215). This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval, utilizes a similar RI-ISi methodology to the previously approved relief request.
  • Relief Request 14R-01 was authorized for Byron Station Units 1 and 2 by NRC SE dated December 20, 2016 (ADAMS Accession No. ML16327A396).
  • Relief Request 14R-01 was authorized for Limerick Generating Station Units 1 and 2 by NRC SE dated December 29, 2016 (ADAMS Accession No. ML16344A324).
  • Relief Request RR-7 was authorized for St. Lucie Plant, Unit 2 by NRC SE dated August 10, 2015 (ADAMS Accession No. ML15196A623).
  • Relief Request 4RR-01 was authorized for Susquehanna Steam Electric Station, Units 1 and 2 by NRC SE dated April 28, 2015 (ADAMS Accession No. ML15098A478).

ATTACHMENT 1 10 CFR 50.SSa Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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8. References
1. Electric Power Research Institute (EPRI) Topical Report (TR) 112657, "Revised Risk-Informed lnservice Inspection Evaluation Procedure," Revision 8-A, dated December 1999.
2. Letter from W. H. Bateman (NRC) to G. L. Vine (EPRI), "Safety Evaluation Report Related to EPRI Risk-Informed lnservice Inspection Evaluation Procedure (EPRI TR-112657, Revision 8, July 1999)," dated October 28, 1999 (ADAMS Accession Nos. ML993190460 and ML993190474).
3. Initial Risk-Informed lnservice Inspection Evaluation - LaSalle County Station, Units 1 and 2, dated February 2001.
4. American Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method 8,Section XI, Division 1," dated March 28, 2000.
5. Letter from A. J. Mendiola (NRC) to 0. D. Kingsley (EGC) "LaSalle County Station, Units 1 and 2 - Relief Request CR-35 (TAC Nos. MB1982 and MB1983)," dated December 27, 2001 (ADAMS Accession No. ML013610078).
6. Letter from R. Gibbs (NRC) to C. G. Pardee (EGC), "LaSalle County Station, Units 1 and 2 - Relief Request 13R-01, Associated with the Third 10-Year Interval for LaSalle County Station, Units 1 and 2 (TAC Nos. MD5457 and MD5458),"

dated April 29, 2008 (ADAMS Accession No. ML080940215).

7. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"
8. LS-PSA-014, Revision 9, "LaSalle Probabilistic Risk Assessment (PAA)

Quantification Notebook," November 2015.

9. Enclosure LS-LAR-007, Revision 0, "LaSalle Station, Units 1 and 2, PAA Capability Assessment for RI-ISi, (Summary: LS-LAR-007 PAA Capability Assessment for Risk-Informed lnservice Inspection Applications)," dated January 2017.

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a{z){1)

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ENCLOSURE LS-LAR-007, Revision 0, "LaSalle Station, Units 1 and 2, PRA Capability Assessment for RI-ISi, {Summary: LaSalle PRA Capability Assessment for Risk-Informed lnservice Inspection Applications)," dated January 2017 Introduction Exelon Generation Company, LLC (EGC) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PAA models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PAA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the LaSalle PAA.

PRA Maintenance and Update The EGC risk management process ensures that the applicable PAA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the EGC Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation procedures. EGC procedure ER-AA-600-1015, "FPIE PAA Model Update" delineates the responsibilities and guidelines for updating the full power internal events PAA models at all operating EGC nuclear generation sites. The overall EGC Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PAA model updates, for tracking issues identified as potentially affecting the PAA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files. To ensure that the current PAA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:

  • Design changes and procedure changes are reviewed for their impact on the PAA model.
  • New engineering calculations and revisions to existing calculations are reviewed for their impact on the PAA model.
  • Maintenance unavailabilities are captured, and their impact on core damage frequency (CDF) is trended.
  • Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities for equipment that can have a significant impact on the PAA model are updated approximately every four years.

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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In addition to these activities, EGC risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

  • Documentation of the PRA model, PRA products, and bases documents.
  • The approach for controlling electronic storage of Risk Management products including PRA update information, PRA models, and PRA applications.
  • Guidelines for updating the full power, internal events PRA models for EGC nuclear generation sites.
  • Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (1 O CFR 50.65{a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

The most recent update of the LaSalle PRA model (designated the LS2014A model) [6] was completed in November 2015 as a result of a regularly scheduled update to the previous LS2011 A model [8]. The LS2014A model is the most recent evaluation of the risk profile at LaSall~ for internal event challenges, including internal flooding. The LaSalle 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 LaSalle PRA is based on the event tree I fault tree methodology, which is a well-known methodology in the industry.

PRA Peer Review Several assessments of technical capability have been made, and continue to be planned for the LaSalle PRA model. A chronological list of the assessments performed includes the following:

  • An independent PRA peer review was conducted under the auspices of the BWR Owners' Group in July 2000, following the Industry PRA Peer Review process [4]. This peer review included an assessment of the PRA model maintenance and update process. All findings from this peer review were addressed and closed out.
  • A self-assessment analysis was performed using Addendum B of the ASME/ANS PRA Standard [9] and Regulatory Guide 1.200, Revision 1 [13]

as part of the periodic update of the LaSalle PRA. This was updated and finalized to represent the current status of the PRA model near the completion of the update in 2007.

ATIACHMENT1 10 CFR 50.SSa Relief Request 14R-01 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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  • In 2008, the 2006C version of the model was peer reviewed against the requirements of the ASME/ANS PRA standard [9] and any Clarifications and Qualifications provided in the Nuclear Regulatory Commission (NRC) endorsement of the Standard contained in Revision 0 to Regulatory Guide (RG) 1.200 [12].
  • Of the 331 internal events supporting requirements (SR) reviewed o 293 were considered Met
  • 286 at Capability Category I/II or greater
  • 7 at Capability Category I o 18 were considered N/A o 20 were considered Not Met According to EPRI TR-1021467-A [5], the 7 SRs that meet Capability Category I are sufficient for RISI; and 12 of 20 of the SRs that were not met are not required for the RISI application.

For the 8 Not-Met SRs that are required for RISI, the following provides a brief synopsis of each SR that was not met:

  • Success Criteria o Perform checks to determine the reasonableness and acceptability of thermal hydraulic, structural or other supporting engineering basis used to support the success criteria.
  • Human Reliability Analysis o For equipment modeled in the PRA, identify test and maintenance activities that require realignment of equipment outside of its normal operational or standby 'status, through a review of procedures and practices. This requirement is applicable to pre-initiator human failure events (HFEs).

o Identify calibration that if performed incorrectly can have an adverse impact on the automatic initiation of standby safety equipment. The SR states that this is accomplished through a review of procedures and practices. This requirement is also applicable to pre-initiator HFEs.

o Check the post-initator human error probability (HEP) quantifications for consistency.

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  • Quantification o Review of a sample of the significant accident sequences/cutsets sufficient to determine that the logic of the cutset or sequence is correct.

o Review of a sample of the non-significant accident sequences/cutsets sufficient to determine that the logic of the cutset or sequence is correct.

o Document the quantitative definition used for significant basic event, significant cutset, and significant accident sequence.

The Peer Review findings, including both those items not-met as well as those that meet Capability Category I, have negligible impact on the RISI analysis. Most of the findings relate to missing documentation rather than shortcomings in the PRA analysis.

A summary of the current Peer Review Not-Met and Capability Category I SRs relative to the RI-ISi relief request is provided in Table 1. Most of the Peer Review Not-Met and Capability Category I SRs have been resolved during subsequent PRA model updates since the Peer Review. Any unaddressed gaps are tracked by the PRA open issues list (i.e., Update Requirements Evaluation (URE) Database) and are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. These items are tracked and their potential impacts are accounted for in applications where appropriate. In addition, plant changes made since the last PRA update have been reviewed and determined to not have a significant PRA impact. These items are also documented in UREs for consideration in future PRA updates, as appropriate.

Guidance from EPRI Report on PRA Technical Adequacy for RI-ISi EPRI report TR-1021467-A [5] provides guidance on the PRA Standard Capability Category necessary to support RI-ISi. This report received a Safety Evaluation (SE) from the NRC in January 2012. Reg. Guide 1.200 considers it a good practice to have, in general, SRs meet Capability Category II for applications. However, according to the EPRI report not all SRs require Capability Category II to adequately support RI-ISi applications. According to the EPRI report [5] some of the LaSalle gaps listed in Table 1 do not require Capability Category II, but instead only require Capability Category I, the most basic level. Therefore, according to EPRI TR-1021467-A and the associated NRC SE, the LaSalle PRA model 2014A is adequate for use in the RI-ISi application.

General Conclusion Regarding PRA Capability The LaSalle PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in RI-ISi applications. As specific risk-informed PRA applications are performed, remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

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Conclusion Regarding PRA Capability for Risk-Informed ISi The LaSalle PRA model continues to be suitable for use in the risk-informed inservice inspection application. This conclusion is based on:

  • PRA maintenance and update processes in place,
  • PRA technical capability evaluations that have been performed and are being planned, and
  • RI-ISi process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RI-ISi process to PRA attribute capability beyond ASME PRA Standard Capability Category I.

In support of the PRA analyses for the LaSalle 10-year interval evaluation using the LS2014A model, the remaining gaps to the PRA standard have been reviewed to determine which, if any, would merit RI-ISi-specific sensitivity studies in the presentation of the application results. The result of this assessment concluded that no additional sensitivity studies are merited.

References

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

"Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications,"

ASME/ANS RA-S-2002, April 2002.

2. "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities," Regulatory Guide 1.200, U.S. Nuclear Regulatory Commission, March 2009, Revision 2.
3. U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, 11 Regulatory Guide 1.174, Revision 1, November 2002.
4. Boiling Water Reactor Owners' Group (BWROG), "BWROG PSA Peer Review Certification Implementation Guidelines," Revision 3, January 1997.
5. Electric Power Research Institute (EPRI), 'Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-informed lnservice Inspection Programs," TR-1021467-A, June 2012.
6. LS-PSA-014, Revision 9, "LaSalle Probabilistic Risk Assessment (PRA) Quantification Notebook," November 2015.
7. LaSalle Station PRA Peer Review Report Using ASME PRA Standard Requirements, July 2008.
8. LS-PSA-014, Revision 8, "LaSalle Probabilistic Risk Assessment (PRA) Quantification Notebook," March 2013.

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9. ASME/ANS, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addenda RA-Sb-2005 to ASME RA-S-2002, December 2005.
10. LS-PRA-004, Revision 7, LaSalle PRA Human Reliability Notebook, January 2013.
11. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sc-2007, August 2007.
12. "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities," Regulatory Guide 1.200, U.S. Nuclear Regulatory Commission, February 2004, Revision 0.
13. "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities," Regulatory Guide 1.200, U.S. Nuclear Regulatory Commission, January 2007, Revision 1.
14. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sa-2009, March 2009.
15. LS-PRA-013, Revision 7, LaSalle PRA Summary Notebook, March 2013.
16. LS-PSA-001, Revision 6, LaSalle PRA Initiating Events Notebook, January 2013.
17. U.S. Nuclear Regulatory Commission, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," NUREG-1855, Volume 1, Main Report, March 2009.

11

18. "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI, Palo Alto, CA: 2008, 1016737.

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement IE-A9 CCI CC I is sufficient for this RI-ISi application. CCI (IE-A7) This SR addresses review of plant-specific precursors to determine if potential initiating events are missing from the PRA. The requirement being evaluated here does not have an impact on the modeled initiating events. CC I is acceptable for RI-ISi applications.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0011 ). Appendix J was added to LS-PSA-001, LaSalle PRA Initiating Event Notebook [16], documenting a review of the LaSalle Licensee Event Reports (LERs) and other industry plant operating experience to determine if there are initiating events that have not been identified previously and modeled. No new initiating events were identified as a result of the review to resolve this open item.

IE-03 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met initiating events analysis are documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

AS-C3 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met accident sequence analysis are documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement SC-85 Not Met This SR requires that checks be performed to determine the reasonableness and cc 1/11/111 acceptability of thermal hydraulic, structural or other supporting engineering basis used to support the success criteria.

The 2008 Peer review team noted that while the PRA model documentation provided some selected comparison of previous LaSalle thermal hydraulic results to more recent calculations, there was no documented comparison of how the LaSalle success criteria compare to those used for sister plants or other similar comparisons. The team noted that the success criteria used for LaSalle appear to be consistent with those of other similar BWRs. The team documented in the peer review report [7] that "this is a documentation issue only."

To address this peer review open item, the success criteria and associated calculations were reviewed. Comparisons of the success criteria were reviewed and compared with other similar plants. No anomalies were identified in the LaSalle success criteria or supporting calculations. This peer review open item remains open to address only the documentation issue noted. This is tracked by URE LS2010-0030.

This issue has no impact on the RI-ISi application.

SC-C3 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met development of success criteria are documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement SY-A4 CCI CC I is sufficient for this RI-ISi application. CCI This SR addresses confirmation of the systems modeling in the PRA with plant operators and system engineers. CC I requires only interviews of plant personnel, while capability category 11/111 requires plant walkdowns as well as interviews to confirm that the systems analysis accurately reflects the as built, as operated plant.

This peer review open item remains open and is tracked by URE LS2010-0032. Plant walkdowns have been conducted for internal flooding, fire PRA development as well as seismic PRA development. No issues with system modeling have been identified during these walkdowns. Complete system walkdowns solely for the purpose of meeting CC 11/111 are a lower priority due to the maturity of the systems modeling in the PRA, the confirmation of modeling through interviews, and the desire to maintain the ALARA (as low as reasonablely achievable) principle with respect to radiation exposure.

SY-C3 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met systems analysis are documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

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TABLE 1 LaSalle Not Met and Capability Category {CC) I Supporting Requirements {SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement HR-A1 Not Met This SR requires, for equipment modeled in the PAA, identification of test and cc 1/11/111 maintenance activities that require realignment of equipment outside of its normal operational or standby status, through a review of procedures and practices. This SR is applicable to pre-initiator human failure events (HFEs).

With respect to this SR, the peer review team concluded that this SR was not met and noted the following: "While this is primarily a documentation issue, this documentation is necessary to meet the requirements of this SR. This requirement is not met strictly because there is no procedure list. It is expected that this review has been performed based on the pre-initiators provided."

This peer review open issue was addressed in the 2011 PAA update (URE LS2010-0042). While the peer review team desired to have a detailed documented list of procedures reviewed, this is not required by the SR. The action taken to resolve the open issue was to define a process to first identify, through systems analysis, potential misalignments or off-normal configurations, and then identify pre-initiator HFEs to include in the model based on the review of system procedures and maintenance practices for these potential off-normal configurations. This approach is documented in LS-PSA-004, LaSalle Human Reliability Analysis [1 O] and the results of the pre-initiator HFE identification process are documented in Appendix J of LS-PSA-004.

Additionally, the test and maintenance procedures used in the pre-initiator HRA are listed in Table B-5 of Appendix B of LS-PSA-004.

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement HR-A2 Not Met This SR requires identification of system calibrations that if performed incorrectly can cc 1/11/111 have an adverse impact on the automatic initiation of standby safety equipment. The SR states that this is accomplished through a review of procedures and practices. This SR is applicable to pre-initiator HFEs.

With respect to this SR, the peer review team concluded that this SR was not met and noted the following: "This requirement is not met because documentation does not provide evidence of the procedures reviewed. It just says procedures were reviewed."

Further, the peer review report stated that "while this is primarily a documentation issue, this documentation is necessary to meet the requirements of this SR."

This peer review open issue was addressed in the 2011 PRA update (URE LS2010-0042). The action taken to resolve the open issue was to define a process to first identify, through systems analysis, potential miscalibrations, and then identify pre-initiator HFEs to include the model based on the review of system procedures and maintenance practices for these potential off-normal configurations. This approach is documented in LS-PSA-004, LaSalle Human Reliability Analysis [1 O] and the results of the pre-initiator HFE identification process are documented in Appendix J of LS-PSA-004. Additionally, the test and maintenance procedures used in the pre-initiator HRA are listed in Table 8-5 of Appendix 8 of LS-PSA-004.

HR-81 CCI CC I is sufficient for this RI-ISi application. CCI This SR requires that if screening of activities is used, rules are established for screening. To meet CC I, classes of activities can be screened. To meet CC 11/111, individual activities must be screened. This SR refers to pre-initiator HFEs.

This peer review open issue was addressed in the 2011 PRA update (URE LS2010-0043). A screening methodology was established and individual activities were screened. This screening approach is documented in LS-PSA-004, LaSalle Human Reliability Analysis [1 O] and the results of the pre-initiator HFE identification process are documented in Appendix J of LS-PSA-004.

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TABLE 1 LaSalle Not Met and Capability Category {CC) I Supporting Requirements {SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement HR-G6 Not Met This SR requires that the post-initator human error probability (HEP) quantifications be cc 1111/111 checked for consistency. Specifically, this SR requires a review of the HFEs and their final HEPs relative to each other to check their reasonableness given the scenario context, plant history, procedures, operational practices, and experience.

This SR was addressed in the 2011 PAA update (URE LS2010-0044). LS-PSA-004, LaSalle Human Reliability Analysis [1 O] was updated and provides a comparison and a reasonableness check of the final post-initiator HEPs.

HR-13 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met human reliability analysis are documented.

This peer review open item was addressed in the 2011 PAA update (URE LS2010-0018). The LaSalle PAA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PAA Summary Notebook [15].

DA-CB CCI CC I is sufficient for this RI-ISi application. CCI For CC I, this SR requires estimating the time that components were configured in their standby status. To meet CC 11/111, plant specific operational records must be used to determine the time that components were configured in their standby staus.

This peer review open item is tracked by URE LS2010-052 and remains open. As noted in the peer review report, some standby times used in the LaSalle PAA are plant specific while others are estimated. This remains open to investigate the standby time of all components that are not normally running.

DA-C10 CCI CC I is sufficient for this RI-ISi application. CCI For CC I, this SR requires a review of test procedures to detemine whether a test should be credited for each possible failure mode and requires that only completed tests or unplanned operational demands be counted for component operation.

This peer review open item is tracked by URE LS2010-0051 and remains open. As noted in the peer review report [7], this issue relates primarily to documentation and how surveillance tests were used to estimate demands on various components.

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement DA-E3 Not Met This SR requires that sources of uncertainty and assumptions associated with the data Not Required to be Met analysis are documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NU REG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

IFSN-A8 CCI CC I is sufficient for this RI-ISi application. CCI (IF-C3b) For CC I, this SR does not require an analysis of inter-area propagation given that flood areas are independent.

The resolution of this CC I peer review assessment open item is tracked by URE LS2010-0061 and remains open. The resolution of this open item is estimated to have a negligible impact on the results of the internal flooding evaluation because the results of the internal flooding PRA are dominated by large flood events that bypass installed barriers such as water tight doors.

IFPP-83 Not Met This SR requires that sources of uncertainty and assumptions associated with the Not Required to be Met internal flooding analysis are documented.

lFS0-83 This peer review open item was addressed in the 2011 PRA update (URE IFSN-83 LS2010-0018). The LaSalle PRA model uncertainty analysis follows the industry lFEV-83 guidance documented in NUREG-1855 [17] and the associated EPRl report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

lFQU-83 (1F-F3)

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement QU-01 Not Met This SR requires a review of a sample of the significant accident sequences/cutsets cc 1/11/111 sufficient to determine that the logic of the cutset or sequence is correct.

(QU-01a)

It was determined based on interviews with the PAA model developers that significant accident sequences and cutsets were reviewed. This was a documentation issue and was tracked by URE LS2010-0069.

This peer review open item was resolved during the 2011 PAA update. LS-PSA-014, LaSalle PAA Quantification Notebook, Revision 8, [8] includes a discussion of the review and well as documentation of a sample of significant cutsets/accident sequences.

QU-05 Not Met This SR requires a review of a sample of the non-significant accident cc 1/11/111 sequences/cutsets sufficient to determine that the logic of the cutset or sequence is (QU-04) correct.

It was determined based on interviews with the PAA model developers that non-significant accident sequences and cutsets were reviewed. This was a documentation issue and was tracked by URE LS2010-0070.

This peer review open item was resolved during the 2011 PAA update. LS-PSA-014, LaSalle PAA Quantification Notebook, Revision 8, [8] includes a discussion of the review of significant cutsets/accident sequences.

QU-E2 Not Met This SR requires that assumptions made in the development of the PAA are identified. Not Required to be Met This peer review open item was addressed in the 2011 PAA update (URE LS2010-0018). The LaSalle PAA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17) and the associated EPRI report [18); and is documented in LS-PSA-013, LaSalle PAA Summary Notebook [15).

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting- Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement QU-E4 Not Met This SR requires that an evaluation be done to evaluate the impact of model Not Required to be Met uncertainities on the results of the PRA model.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0071 ). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

QU-F3 CCI CC I is sufficient for this RI-ISi application. CCI For CC I, this SR requires that the significant contributors to CDF are documented in the PRA results summary. For CC 11/111, this SR additionally requires a detailed description of significant accident sequences or functional failure groups.

This peer review open item (URE LS2010-0073) was resolved during the 2011 PRA update. LS-PSA-014, LaSalle PRA Quantification Notebook, Revision 8, [8] includes a detailed description of significant accident sequences.

QU-F4 Not Met This SR requires that sources of uncertainty and assumptions associated with the PRA Not Required to be Met model be documented.

This peer review open item was addressed in the 2011 PRA update (URE LS2010-0074). The LaSalle PRA model uncertainty analysis follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PRA Summary Notebook [15].

QU-F6 Not Met This SR relates to documenting the quantitative definition used for significant basic cc 1/11/111 event, significant cutset, and significant accident sequence.

It was noted by the peer review team that other than in the HRA notebook, the documentation did not include the applied definition of "significant". This peer review open item (URE LS2010-0076) was resolved during the 2011 PRA update. LS-PSA-014, LaSalle PRA Quantification Notebook, Revision 8, [8] includes a definition of "significant."

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TABLE 1 LaSalle Not Met and Capability Category (CC) I Supporting Requirements (SRs)

Supporting Capability Requirement Category Risk-Informed ISi Evaluation Impact EPRI TR-1021467 (Note 1) Requirement LE-F3 Not Met This SR requires the identification of contributors to LEAF and characterization of Not Required to be Met LEAF uncertainties.

This peer review open item was addressed in the 2011 PAA update (URE LS2010-0080). The LaSalle PAA model uncertainty analysis, including an analysis of LEAF, follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PAA Summary Notebook [15].

LE-G4 Not Met This SR requires the documentation of assumptions and sources of uncertainty Not Required to be Met associated with the LEAF analysis.

This peer review open item was addressed in the 2011 PAA update (URE LS2010-0080). The LaSalle PAA model uncertainty analysis, including an analysis of LEAF, follows the industry guidance documented in NUREG-1855 [17] and the associated EPRI report [18]; and is documented in LS-PSA-013, LaSalle PAA Summary Notebook [15].

LE-G6 Not Met This SR requires documentation of the quantifitative definition used for significant cc 1/11/111 accident. This was a documentation issue with no impact on PAA model results.

This peer review open item (URE LS2010-0081) was resolved during the 2011 PAA update. LS-PSA-014, LaSalle PAA Quantification Notebook, Revision 8, [8] includes a definition of "significant."

Notes:

1. The LaSalle PRA peer review was against the ASME/ANS 2005 PRA standard [9]. The current 2009 ASME/ANS standard [14] SRs are provided with the equivalent peer reviewed SR listed in parentheses where applicable.

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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: 813.10, 813.20, 813.30, and 813.40

Description:

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

2. Applicable Code Edition and Addenda The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.
3. Applicable Code Requirements ASME Section XI requires the examination of components within the Reactor Pressure Vessel. These examinations are included in Table IWB-2500-1 Categories B-N-1 and B-N-2 and identified with the following item numbers:

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

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

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

813.40 Examine surfaces of the welded core support structure each interval by the VT-3 visual examination method.

These examinations are performed to assess the structural integrity of components within the boiling water reactor pressure vessel.

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

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BWRVI P guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVIP Inspection and Evaluation (l&E) guidelines have recommended aggressive specific inspection by BWR operators to completely identify material condition issues with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. l&E guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanisms, and require re-examination at conservative intervals. In contrast, ASME Section XI inspection requirements were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience.

Use of this proposed alternative will maintain an adequate level of quality and safety and avoid unnecessary inspections.

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

LaSalle County Station, Units 1 and 2 will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 1 in accordance with the latest Nuclear Regulatory Commission (NRC) approved BWRVIP guideline requirements. This relief request proposes to utilize the identified BWRVI P guidelines in lieu of the associated ASME Section XI requirements, including 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 guidelines are applicable to this relief request:

- BWRVIP-03, 11 BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines"

- BWRVIP-18, Revision 2-A, 11 BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines"

- BWRVIP-25, 11 BWR Core Plate Inspection and Flaw Evaluation Guidelines"

- BWRVIP-26-A, 11 BWR Top Guide Inspection and Flaw Evaluation Guidelines"

- BWRVIP-27-A, 11 BWR Standby Liquid Control System/Core Plate ,1P Inspection and Flaw Evaluation Guidelines"

- BWRVIP-38, 11 BWR Shroud Support Inspection and Flaw Evaluation Guidelines"

- BWRVI P-41, Revision 3, 11 BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines"

- BWRVIP-42, Revision 1, "Low Pressure Coolant Injection (LPCI) Coupling Inspection and Flaw Evaluation Guidelines"

- BWRVIP-47-A, 11 BWR Lower Plenum Inspection and Flaw Evaluation Guidelines"

- BWRVIP-48-A, "Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines 11

- BWRVIP-49-A, 11 lnstrument Penetration Inspection and Flaw Evaluation Guidelines"

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- BWRVIP-76, Revision 1-A, BWR Core Shroud Inspection and Flaw Evaluation Guidelines"

- BWRVIP-94NP, Revision 2, "Program Implementation Guide"

- BWRVI P-138, Revision 1-A 11 Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines"

- BWRVI P-180, "Access Hole Cover Inspection and Flaw Evaluation Guidelines" 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. Current LaSalle County Station deviations from the subject guidelines above are summarized in Table 2.

Inspection services, by an Authorized Inspection Agency, will be applied to the proposed alternative actions of this relief request.

BWRs examine reactor internals in accordance with BWRVIP guidelines. These guidelines have been written to address the safety significant vessel internal components and to examine and evaluate the examination results for these components using appropriate methods and reexamination frequencies. The BWRVI P has established a reporting protocol for examination results and deviations. Enclosures 2 and 3 contain the "Reactor Internals Inspection History" for LSCS Units 1 and 2. This summary provides, 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. This table also contains the identified corrective actions. The information provided reflects the compilation of the BWRVI P 120-day reports. Corrective actions and examinations performed prior to the BWRVI P were implemented to the requirements of ASME Section XI, as applicable. The NRC has agreed with the BWRVI P approach in principal and has issued Safety Evaluations for these guidelines (see References 1 through 14 below). Therefore, use of these guidelines, as an alternative to the subject ASME Section XI requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

As additional justification, Table 1 of 14R-02, 11 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVI P Guidance Requirements, 11 provides specific examples which compare the inspection requirements of ASME Section XI Item Numbers 813.10, 813.20, 813.30, and 813.40 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. These comparisons demonstrate that use of these guidelines, as an alternative to the subject ASME Section XI requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

Table 1 compares present ASME Section XI Examination Category B-N-1 and B-N-2 requirements with the above current BWRVI P guideline requirements, as applicable, to LaSalle County Station. Therefore, Table 1 only represents a current comparison. Any deviations from the BWRVI P guidelines referenced within this relief request for the

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duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process.

Also, the reactor vessel internals inspection program at LaSalle County Station has been developed and implemented to satisfy the requirements of BWRVI P-94. It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to include enhancements in inspection techniques and flaw evaluation methodologies.

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 an NRG-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.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units 1 and 2.
7. Precedents
  • The Exelon Generation Company/AmerGen fleet-wide Relief Request for BWRVIP was authorized conditionally by NRC Safety Evaluation (SE) dated April 30, 2008 (ADAMS Accession No. ML080980311) (i.e., LaSalle County Station Third ISi Interval Relief Request 13R-02.} This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval utilizes a similar approach to the previously approved relief request. (Reference 15}
  • Relief Request was authorized for Limerick Generating Station, Units 1 and 2, by NRC SE dated November 21, 2016. (Reference 16)
  • Relief Request was authorized for Perry Nuclear Power Plant Unit 1 by NRC SE dated January 31, 2012. (Reference 17)
8. References
1. Letter from K. Hsueh (NRC) to BWRVIP, "U.S. Nuclear Regulatory Commission Approval Letter for Electric Power Research Institute Topical Report, BWRVIP-18, Revision 2-A, BWR [Boiling Water Reactor] Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines"

{TAC No. MF8415}, 11 dated December 21, 2016 (ADAMS Accession No. ML16273A083).

2. Letter from NRC to BWRVIP, "Final Safety Evaluation of BWRVIP Vessel and Internals Project, 'BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25),"' EPRI Report TR-107284 (TAC No. M97802), dated December 19, 1999.

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3. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-26-A, 'BWR Vessel and Internals Project, Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines,"' dated September 9, 2005.
4. Letter from NRC to BWRVIP, Proprietary Version of NRC Staff Review of BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate LiP Inspection and Flaw Evaluation Guidelines, 11 dated June 10, 2004.
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), 11 dated July 24, 2000.
6. BWRVIP-41, Revision 3, BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1021000, dated September, 2010.
7. "BWRVIP-42, Revision 1: BWR Vessel and Internals Project, LPCI Coupling Inspection and Flaw Evaluation Guidelines," dated June, 2010.
8. Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-47-A, 'BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines, 111 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-49-A: BWR Vessel and Internals Project, Instrument Penetration Inspection and Flaw Evaluation Guidelines," dated March, 2002.
11. Letter from NRC to BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel and Internals Project 76, Revision 1-A Topical Report, "Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines" (TAC No. ME8317)," dated November 12, 2014.
12. Letter from Chairman, BWR Vessel and Internals Project to NRC, "Project No.

704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2), 11 dated September 22, 2011 (ADAMS Accession No. ML11271A058).

13. Letter from NRC to BWRVIP, "Electric Power Research Institute Final Safety Evaluation for Technical Report 1016574 "BWRVIP-138, Revision 1-A: BWR

[Boiling Water Reactor] Vessel and Internals Project 'Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines'" (TAC No. ME2191)," dated May 14, 2012.

14. "BWRVIP-180: BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines," dated November 2007.

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15. Letter from R. Gibbs (NRC) to C. G. Pardee (Exelon Generation Company/AmerGen), "Clinton Power Station Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief Request to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements {TAC Nos. MD5352 through MD5363)," dated April 30, 2008 (ADAMS Accession No. ML080980311 ).
16. Letter from S. S. Koenick (NRC) to B. C. Hanson (Exelon Nuclear), "Safety Evaluation of Relief Requests 14R-02 and 14R-1 O for the Fourth 10-Year Interval of the lnservice Inspection Program for Limerick Generating Station, Units 1 and 2 (CAC Nos. MF7587 and MF7588)," dated November 21, 2016 (ADAMS Accession No. ML16301A401 ).
17. Letter from J. I. Zimmerman (NRC) to V. A. Kaminskas (FirstEnergy Nuclear Operating Company), "Perry Nuclear Power Plant, Unit No. 1, Re: Safety Evaluation In Support of 10 CFR 50.55a Requests for the Third 10-Year In-Service Inspection Interval {TAC Nos. ME5373, ME5376, ME5377, ME5379, and ME5380), 11 dated January 31, 2012 (ADAMS Accession No. ML120180372).
9. Enclosures
1. Comparison of Code Examination Requirements to BWRVIP Examination Requirements
2. LaSalle Unit 1 Reactor Internals Inspection History, dated April 6, 2016
3. LaSalle Unit 2 Reactor Internals Inspection History, dated March 16, 2017

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TABLE 1 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Section ASME ASME ASME BWRVIP XI Item Section XI Authorized BWRVIP BWRVIP Component Section Section XI Exam Number, Table Exam Alternative Exam Frequency Scope XI Exam Frequency Scope IWB-2500-1 B13.10 Reactor Vessel Accessible VT-3 Each BWRVIP-18-R2- Overview examinations of components during Interior Areas period A, 25, 26-A, 27- BWRVIP examinations satisfy ASME Section XI VT-A, 38, 41-R3, 42- 3 visual examination requirements.

R1, 47-A, 48-A, 76-R1-A, 138-R1-A, and 180 B13.20 Interior Attachments Accessible VT-1 Each BWRVIP-48-A, Riser Brace EVT-1 100% in first 12 Within Beltline Region Welds 10-year Table 3-2 Attachment years; 25% during

- Jet Pump Riser Interval each subsequent 6 Braces years.

Lower Surveillance BWRVI P-48-A, Bracket VT-1 Each 10-year Specimen Holder Table 3-2 Attachment Interval.

Brackets B13.30 Interior Attachments Accessible VT-3 Each BWRVIP-48-A, Bracket VT-3 Each 10-year Beyond Beltline - Welds 10-year Table 3-2 Attachment Interval.

Steam Dryer Hold- Interval down Brackets Guide Rod Brackets BWRVI P-48-A, Bracket VT-3 Each 10-year Table 3-2 Attachment Interval.

Steam Dryer Support BWRVIP-48-A, Bracket EVT-1 Each 10-year Brackets Table 3-2 Attachment Interval.

Feedwater Sparger BWRVIP-48-A, Bracket EVT-1 Each 10-year Brackets Table 3-2 Attachment Interval.

Core Spray Piping BWRVIP-48-A, Bracket EVT-1 Every 4 Refueling Brackets Table 3-2 Attachment Cycles.

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TABLE 1 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Section ASME XI Item ASME ASME BWRVIP Section XI Authorized BWRVIP BWRVIP Component Section Section XI Exam Number, Table Exam Alternative Exam Frequency Scope XI Exam Frequency Scope IWB-2500-1 Upper and Middle BWRVI P-48-A, Bracket VT-3 Each 10-year Surveillance Table 3-2 Attachment Interval.

Specimen Holder Brackets Shroud Support (Weld 2 BWRVIP-38, Weld H9 EVT-1 or UT Based on as-found H9) including gussets 3.1.3.2, including conditions, to a where applicable Figures 3-2 and gussets maximum of 6 years 3-5 (where for one sided EVT-applicable) 1, 1O years for UT.

Shroud Support Legs (Rarely BWRVIP-38, Weld H12 Per When accessible.

(Weld H12) Accessible) 3.2.3 BWRVIP-38 NRC SER (07/24/00),

inspect with appropriate 3

method B13.40 Welded Core Support Accessible VT-3 Each BWRVIP-38, Shroud EVT-1 or UT Based on as found Structure - Shroud Surfaces 10-year 3.1.3.2, Support and conditions, to a Support Interval Figures 3-2 and Leg Welds maximum of 6 years 3-5 including for one sided EVT-gussets as 1, 1O years for UT applicable where accessible.

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TABLE 1 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Section ASME ASME ASME BWRVIP XI Item Section XI Authorized BWRVIP BWRVIP Component Section Section XI Exam Number, Table Exam Alternative Exam Frequency Scope XI Exam Frequency Scope IWB-2500-1 Shroud Horizontal BWRVIP-76-R1- Welds H1- EVT-1 or UT Based on as found Welds A, H7 as conditions, to a 2.2.1 applicable maximum 6 years for one sided EVT-1, 1O years for UT where accessible.

Shroud Vertical Welds BWRVIP-76-R1- Vertical and EVT-1 or UT Maximum of 6 years A, Ring for one-sided 2.3, Segment EVT-1, 1O years for Figure 3-3 Welds as UT.

applicable Shroud Repairs BW RVI P-76-R 1- Tie-Rod VT-3 Per designer A, Repair recommendations Section 3.5 per BWRVIP 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 BWRVI P document.
2. In accordance with Appendix A of BWRVI P-38, a site specific evaluation will determine the minimum required weld length to be examined.
3. When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection of these welds.

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TABLE 2 BWRVIP Deviations PLANT BWRVIP LETTER DATE TO NRC DEVIATION APPLICABILITY DOCUMENT LaSalle County BWRVIP-25 Letter from S. E. Kuczynski Postponement of Core Plate Bolt inspections on both Unit This Deviation does Station Unit 1 (EGC) to NRC dated 1 and Unit 2 because meaningful inspections are not not impact the basis and Unit 2 March 31, 2011. currently possible using existing methods. for the use of this relief request.

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ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS The following discussion provides a comparison of the examination requirements provided in ASME Section XI Item Numbers 813.10, 813.20, 813.30, and 813.40 in Table IW8-2500-1, to the examination requirements in the 8WRVI P guidelines. Specific 8WRVI P guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. ASME Section XI Requirement - 813.10 - 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 3 years, during the First ISi Interval, and each period during each successive 10-Year ISi Interval. Typically, these examinations are performed every other refueling outage of the inspection 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 8WRVIP Guidelines require access to accessible areas of the reactor vessel during each refueling outage.

Examination of core spray piping and spargers (8WRVI P-18-R2-A), top guide (8WRVIP-26-A), jet pump welds and components (8WRVIP-41-R3), interior attachments (8WRVIP-48-A), core shroud welds (8WRVIP-76-R1-A), shroud support (8WRVIP-38),

LPCI couplings (8WRVIP-42-R1), and lower plenum components (8WRVIP-47-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 8WRVI P examination requirements. Therefore, the specified 8WRVI P Guideline requirements meet or exceed the subject ASME Section XI requirements for examination method and frequency of the interior of the reactor vessel.

Accordingly, these 8WRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Section XI requirements.

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ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS

2. ASME Section XI Requirement - 813.20 - Interior Attachments Within the Beltline (B-N-2)

ASME Section XI requires a VT-1 visual examination of accessible reactor interior surface attachment welds within the beltline each 10-year interval. In the BWR/5 model, this includes the jet pump riser brace welds-to-vessel wall and the lower surveillance specimen support bracket welds-to-vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the lower surveillance specimen support bracket welds, and requires an EVT-1 visual examination on 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 BWRVIP examination requirements.

Comparison to BWRVIP Requirements - Jet Pump Riser Braces (BWRVIP-41-R3 and BWRVI P-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.
  • The BWRVIP requires an EVT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds the first 12 years and then 25% during each subsequent 6 years.
  • BWRVI P-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 enhanced VT-1 (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 inter-granular 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 (depending on applicable Edition) that at a maximum distance of 2 feet or a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires resolution of 0.044 inch on the examination surface. BWRVIP-48-A includes a diagram for the configuration and prescribes examination for this plant.

The calibration standards used for BWRVIP EVT-1 visual examinations utilize ASME Section XI characters, thus assuring at least equivalent resolution compared to ASME

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ENCLOSURE1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Section XI. 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 interior surface attachment welds beyond the beltline each 10-year interval. In the BWR/5 model, this includes the core spray piping primary and supplemental support bracket welds-to-vessel wall, the upper surveillance specimen support bracket welds-to-vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support and hold down bracket welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, the shroud support plate-to-vessel wall, and the shroud support gussets. 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 enhanced visual examination technique 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 ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVI P requirements are equivalent to that provide by ASME Section XI.

For the core spray primary and secondary support bracket attachment welds, the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-vessel welds, as applicable, the BWRVIP Guidelines require an EVT-1 visual examination at the same frequency as ASME Section XI. Therefore, the BWRVI P requirements provide the same level of quality and safety to that provided by ASME Section XI.

The core spray piping bracket-to-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 - Core Spray Piping Bracket Welds (BWRVIP-48-A)

  • ASME Section XI examination requirement is a VT-3 visual examination of each weld every 1O years.
  • The BWRVIP visual examination requirement is an EVT-1 for the Core Spray piping bracket attachment welds with each weld examined every four cycles (8 years for units with a two year fuel cycle).

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ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS The BWRVIP visual examination method EVT-1 has superior flaw detection and sizing capability, the examination frequency is greater than 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 enhanced 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, the relevant degradation mechanisms for BWR internal attachments.

Therefore, with the EVT-1 visual examination method, the same examination scope (accessible welds), an increased examination frequency (8 years instead of 1O years) in some cases, the same flaw evaluation criteria (ASME Section XI), the level of quality and safety required by the BWRVIP criteria is superior to than that required by ASME Section XI.

4. ASME Section XI Requirement - 813.40 - Welded Core Support Structures {B-N-2)

ASME Section XI requires a VT-3 visual examination of accessible surfaces of the welded core support structure each 10-year interval. In the BWR/5 model, the welded core support structure has primarily been considered the shroud support structure, including the shroud support plate (annulus floor), the shroud support ring, the shroud support welds, the shroud support gussets, and the shroud support legs (if accessible).

In later designs, the shroud itself is considered part of the welded core support structure.

Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examination replaces this ASME Section XI requirement with specific BWRVI P guidelines that examine susceptible locations for known relevant degradation mechanisms.

Comparison to BWRVI P Requirements - Shroud Supports (BWRVI P-38)

  • ASME Section XI requires a VT-3 visual examination of accessible surfaces each 10-year interval.
  • The BWRVIP requires either an enhanced visual examination technique (EVT-1) or volumetric examination {UT) every 1O years as compared to ASME Section XI requirement {VT-3 visual examination). (Only 10% of the weld is required to be examined.)

BWRVI P recommended examinations of welded 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 volumetric examination of the susceptible welds at a frequency identical to ASME Section XI requirement.

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ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS For other welded core support structure components, the BWRVI P requires an EVT-1 visual examination or UT of core support structures. The core shroud is used as an example for comparison between ASME Section XI and BWRVIP examination requirements as shown below.

Comparison to BWRVI P Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline (BWRVI P-76, Revision 1-A)

  • ASME Section XI requires a VT-3 visual examination of accessible surfaces each every 10-year interval.
  • The BWRVIP requires an EVT-1 visual examination from the inside and outside surface where accessible or ultrasonic examination of each core shroud circumferential weld that has not been structurally replaced with a shroud repair at a calculated 11 end of interval" (EOI) that will vary depending upon the amount of flaws present, but not to exceed ten years.

The BWRVI P 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 ASME Section XI required VT-3 visual examination for flaw detection and characterization. The superior flaw detection and characterization capability and the same flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that required by ASME Section XI requirements.

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

ENCLOSURE 2 LaSalle Unit 1 Reactor Internals Inspection History 15 pages follow

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection Core Spray Piping L1R16 (2016) EVT-1 Visual examination of 34 core spray piping welds. No indications.

EVT-1 I VT-1 Visual examination of 5 core spray piping brackets and attachment welds. No attachment weld indications. Slight wear identified at interface between one bracket and the piping.

UT Re-sized existing flaw BP4a; no significant change in length.

Re-exam scheduled in 2 cycles.

L1R15 (2014) EVT-1 Visual examination of 33 core spray piping welds (implemented the sampling of P4 welds). No indications.

L2R14 (2012) EVT-1 Visual examination of 46 core spray welds, including two LaSalle 1-unique welds, and the BP4a welds that was examined by UT. No indications.

UT Re-sized existing flaw BP4a; no significant change in length.

Re-exam scheduled in 2 cycles.

L1 R13 (2010) EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated. No indications. Visual examination of four piping brackets. No indications.

L1R12 (2008) UT Ultrasonic examination of 38 welds for which the UT technique is now demonstrated. Re-sized flaws on BP4a, DPS, and DP6 and due to new Demonstration, the flaws on DP5 and DP6 have been re-characterized as geometry-related; no flaws exist. Flaw evaluation performed on BP4a and weld scheduled for examination again in L1R14.

EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated or where access is limited.

No indications. Visual examination of five piping brackets.

No indications.

L1 R11 (2006) UT Re-sized flaws on BP4a, DP5, and DP6. Flaw evaluation performed and welds scheduled for examination in L1R12.

EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated. No indications.

L1R10 (2004) UT Ultrasonic examination of 34 welds for which the UT technique is demonstrated. Re-sized flaws on BP4a, DPS, and DP6. Flaw evaluation performed and welds scheduled for examination in L1R11.

EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated. No indications.

L1R09 (2002) EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection L1A08 (1999) UT Ultrasonic examination of the welds for which the UT technique is demonstrated. Re-sized flaws on BP4a, DP5, and DP6. Flaw evaluation performed and welds scheduled for examination in L1 A10.

EVT-1 Visual examination of those core spray piping welds for which UT technique is not demonstrated. No indications.

Visual examination of 50% of the core spray sparger welds.

No indications.

Core Spray Sparger L1A16 (2016) VT-1 Visual examination of 50% of the core spray sparger S3 welds. No indications.

Visual examination of seven sparger brackets. New indication noted on sparger bracket af 225°.

L1R15 (2014) EVT-1NT-1 Visual examination of 25% of the core spray sparger welds.

No indications.

Visual examination of six sparger brackets. New indications noted on sparger bracket at 225°.

L1R14 (2012) VT-1 Visual examination of 25% of the core spray sparger welds.

No indications.

Visual examination of six sparger brackets.

No rew indications L1R13 (2010) EVT-1NT-1 Visual examination of 50% of the core spray sparger welds.

No indications.

Visual examination of eight sparger brackets. No indications.

L1R12 (2008) EVT-1 Visual examination of 25% of the core spray sparger welds.

No indications. Visual examination of four sparger brackets.

No indications.

L1R11 (2006) EVT-1 Visual examination of 50% of the core spray sparger welds.

No indications.

L1A10 (2004) EVT-1 Visual examination of 50% of the core spray sparger welds.

No indications.

L1A09 (2002) EVT-1 Visual examination of 50% of the core spray sparger welds.

No indications.

L1ROS (1999) EVT-1 Visual examination of 50% of the core spray sparger welds.

No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection .Results, Frequency of Scope Method Used Repairs, Replacements, Relnspections Inspection Attachment Welds L1R16 (2016) EVT-1 / VT-3 Visual examination of one steam dryer support lug. Wear on top and gouge on side are unchanged. No indications on weld.

Visual examination of 12 feedwater sparger attachment welds and associated brackets. No indications on welds. No change in previously identified wear on bracket pins.

Visual examination of both guide rod attachment welds. No indications.

L1A15 (2014} VT-3 Visual examination of one steam dryer support lug. Wear on top and gouge on side are unchanged. Weld was not examined.

L1R14 (2012) EVT-1 Visual examination of one steam dryer support lug. Wear on top and gouge on side are unchanged. No indications on weld.

L1R13 (201 O) EVT-1 (see core spray section for those attachment welds) Visual examination of one steam dryer support lug attachment weld (185°). No change in the wear.

VT-1NT-3 Visual examination of the upper and lower surveillance capsule attachment welds. No indications.

L1 A12 (2008) EVT-1 Visual examination of 12 feedwater sparger attachment welds, both the upper and lower surveillance capsule welds at three locations. No indications.

EVT-1 Visual examination of four steam dryer support lug attachment welds. No change in the wear on the steam dryer support lugs at 5° and 185° where previous wear was observed.

L1 R11 (2006) EVT-1/VT-1/ (see jet pump and core spray sections for those attachment VT-3 welds.) Visual examination of 2 guide rod attachment welds, 12 feedwater sparger attachment welds, and both the upper and lower surveillance capsule welds at three locations. No indications EVT-1 Visual examination of the steam dryer support lug at 185° where wear was observed last outage. No change in the wear.

L 1R10 (2004) EVT-1NT-1/ (see jet pump and core spray sections for those attachment VT-3 welds.) Visual examination of 4-steam dryer support lug welds, 2 feed water sparger attachment welds, and both the upper and lower surveillance capsule welds at three locations. The steam dryer support lug at 185° showed signs of wear and was accepted for one cycle.

L 1ROB (1999} EVT-1NT-1 (see jet pump and core spray sections for those attachment welds.) Visual examination of 4 steam dryer support lug welds. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection Core Shroud L1R15 (2014) UT All accessible areas of core shroud weld H4 were (Note: LaSalle has two ultrasonically examined. On the upper side of the weld, beltline horizontal welds 89.8% of the weld length was examined, and 3.6% of the and thereby unique examined weld length was flawed. On the lower side of the designation) weld, 100% of the weld length was examined, and 2.6% of the examined weld length was flawed. Due to the high fluence on H4, a site specific evaluation was performed, supporting re-inspection of weld H4 in 1O years.

L1R14 (2012) UT UT of welds H2 (lower only), H3, HS, H6, and HB (LaSalle-specific numbering). Welds H2 and HS were not due for examination but were partially examined due to tooling availability. 100% of the accessible areas of H3, H6, and H8 were examined, and indications were less than 10% of each weld. Due to the high stresses on HS, a site specific evaluation was performed for this weld. Re-inspection of welds H3, H6, and HS is required in 1O years.

L1R11 (2006) UT UT of welds H3, H4, H6, and H8 (LaSalle-specific numbering). Coverage on H6 and HS was less than 50%,

and a site-specific flaw evaluation was performed and re-inspection is in 6 years. Note that 100% of the accessible areas were not examined, and a Deviation Disposition was submitted. Indications were less than 10% on each weld.

L1R07 (1996) UT UT of welds H3, H4, HS, H6, and H8 (LaSalle-specific numbering). No indications noted except on H4, where indications were 3.0%. Next inspection in 2006.

Shroud Support L1R16 (2016) EVT-1 Visual examination of approximately 25% of H8a(BWRVI P Weld HS). No indications.

Visual examination of six shroud support plate gussets, No indications.

L1R14 (2012) UT Ultrasonic examination of 100% of the H9 weld from the vessel outside diameter. No indications.

EVT-1 Visual examination of 100% of both access hole covers. No indications.

EVT-1 Visual examination of 2 shroud support plate gusset welds.

No indications.

L1R13 (2010) EVT-1 Visual examination of 7 shroud support plate gusset welds.

No indications.

EVT-1 Visual examination of approximately 12.5% of H8a. No indications.

L1R12 (2008) EVT-1 Visual examination of both access hole covers. No indications.

EVT-1 Visual examination of 7 shroud support plate gusset welds.

No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection L1 R11 (2006) EVT-1 Visual examination of 8 shroud support plate gusset welds.

No indications.

VT-3 Visual exam of 100% of the accessible portion of the tope of H9 and both access hole covers. No indications.

VT-3 Visual examination of the accessible portions of the bottom of H9 beneath jet pumps 5, 6, 9, and 10 due to the removal of the inlet mixers. NRI.

L1R10 (2004) EVT-1 Visual examination of 11 shroud support plate gusset welds.

No indications.

EVT-1 Visual examination of approximately 20% of H8a(BWRVIP weld HS). No indications.

VT-3 Visual examination of the accessible portions of the bottom of H9 beneath all jet pumps due to the replacement of the inlet mixers. NRI.

L1R09 (2002) UT Ultrasonic examination of 100% of the H9 weld from the vessel outside diameter. No indications.

L1ROS (1999) EVT-1 Visual examination of 6 shroud support plate gusset welds.

No indications.

EVT-1 Visual examination of approximately 2% of H8a, 23% of the top of H9, and both access hole covers. No indications.

L1 R07 ( 1996) VT-1 Visual examination of both access hole covers. No indications.

Standby Liquid Control L1R16 (2016) VT-2 Visual examination during the system leak test. No indications.

L1R15 (2014) VT-2 Visual examination during the system leak test. No indications.

L1R14 (2012) VT-2 Visual examination during the system leak test. No indications.

L1R13 (2010) VT-2 Visual examination during the system leak test. No indications.

L1R12 (2008) VT-2 Visual examination during the system leak test. No indications.

PT Surface examination. No indications.

L1R11 (2006) VT-2 Visual examination during the system leak test. No indications.

L1R10 (2004) VT-2 Visual examination during the system leak test. No indications.

PT Surface examination. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components In BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection L 1R09 (2002) VT-2 Visual examination during the system leak test. No indications.

L1 ROS ( 1999) VT-2 Visual examination during the system leak test. No indications.

Jet Pump Assembly L1R16 (2016) EVT-1 Visual examination of IN-1 at 5 locations. No indications.

Visual examination of IN-2 at 5 locations. No indications.

Visual examination of RB-1 a,b,c,d on 5 pumps. No indications.

Visual examination of RS-1 at 3 locations. No indications.

Visual examination of RS-2 at 3 locations. No indications.

Visual examination of RS-3 at 5 locations. No indications.

Visual examination of RS-6 at 2 locations. No indications.

Visual examination of RS- 7 at 3 locations. No indications.

Visual examination of RS-8 at 5 locations. No indications.

Visual examination of RS-9 at 12 locations. No change in existing indication on Jet Pump 1/2 riser.

Visual examination of strain relief welds RS-AW on three risers. No indications.

VT-1 Visual examination of all twenty WD-1 main wedges. No indications on two wedges. No change to previously identified indications on 14 wedges. New indications of wear on four wedges.

Visual examination of fourteen auxiliary wedges. New indications of wear on wedges for Jet Pumps 4 and 7 were identified. The vessel-side aux wedge for Jet Pump 16 had reached the bottom of its available travel, and required removal. Inspection of the associated set screw AS-1 and AS-2 locations were satisfactory with no indications.

VT-3 Visual examination of Slip Joint Clamps at four locations.

New indication on one clamp, and no change to previously identified indications on the remaining three clamps.

UT Ultrasonic examination of RS-9 at four locations. Confirmed presence of three flaws previously identified visually.

Evaluated as acceptable for one cycle.

UT Ultrasonic examination of AD-1 and AD-2 at eighteen locations. No indications.

L1 R15 {2014) UT Ultrasonic examination of thirteen Group 2 beams. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection EVT-1 Visual examination of RS-2 at 3 locations. No indications.

Visual examination of IN-2 at 5 locations. No indications.

Visual examination of RB-2a,b,c,d on 5 pumps. No indications.

Visual examination of RS-6 on 2 pumps. No indications.

Visual examination of RS-7 on 2 pumps. No indications.

Visual examination of RS-9 on 1O pumps. No change in existing indications on riser 1/2, 3/4, 5/6, 9/10, 11/12, and no indications on other 5 risers.

VT-3 Visual examination of all 20 slip joint clamps. Existing wear at contact point with middle vane unchanged on jet pumps 7, 13, and 14. New wear identified at contact point with middle vane on jet pump 12. No contact observed at middle vane on jet pump 10, and a review of video indicates that the clamp was not in contact after original installation and has not changed since original installation.

VT-1 Visual examination of all 20 main wedges. No change in the wear on 14 pumps, and the other 6 pumps had no indications. All 20 main wedge rods were examined in response to BWRVIP Letter 2014-019. Existing wear was unchanged on 14 of the rods, and new wear was identified on one rod. The other 5 rods had no wear.

VT-1 Visual examination of 5 auxiliary wedges. Existing wear on 3 auxiliary wedges showed no change, and the other 2 auxiliary wedges had no wear.

L1A14(2012) UT Ultrasonic examination of diffuser welds DF-1 (bottom only),

DF-2, and DF-3 (top only) on all twenty pumps. No indications. (Note that the bottom of DF-3 is not accessible due to the presence of curved adaptor)

EVT-1 Visual examination of RB-1 welds at 18 locations. No indications.

Visual examination of RS-1 at 3 locations. No indications.

Visual examination of RS-8 at 3 locations. No indications.

Visual examination of RS-6 at 1 location. No indications.

Visual examination of RS-7 at 1 location. No indications.

Visual examination of RS-9 at 10 locations. Existing flaws at two locations unchanged; indications noted on the edges at four locations.

VT-1NT-3 Visual examination of twenty slip joint clamps. Recordable indications on three clamps, and the other 17 had no indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection VT-1 Visual examination of WD-1 on all twenty pumps. No recordable indications on 5 wedges, unchanged wear on 14 wedges, and new wear on one wedge.

Visual examination of WD-2a and WD-2b on two pumps, no indications.

Visual examination of set screw to inlet mixer contact on four pumps. No indications.

Visual examination of 10 auxiliary wedges; recordable indications on three, and no recordable indications on 7 locations.

VT-3 Visual external examination of the jet pump 9 assembly, including the nozzles and sensing line. No indications.

Visual external examination of the jet pump 10 assembly, no indications.

L1 R13 (2010) Performed an access study on 4 pumps to assist in tooling development for UT examination of unique welds AD-1, AD-2, and DF-3.

EVT-1 Visual examination of RS-1 on 4 pumps. No indications.

EVT-1 Visual examination of RS-3 on 5 pumps. No indications.

EVT-1 Visual examination of RS-8 on 1Opumps. No indications.

(Due to Laguna Verde)

EVT-1 Visual examination of RS-9 on 1Opumps. No new indications, no apparent change in three existing indications.

(Due to Laguna Verde)

EVT-1 Visual examination of IN-1 on 5 pumps. No indications.

VT-1 Visual examination of WD-1 on 20 pumps. No new indications, no apparent change in wear on 14 wedges. (Due to Laguna Verde)

VT-1 Visual examination of vessel side auxiliary wedges on 9 pumps. No new indications, no apparent change in wear on 1 wedge. (Due to Laguna Verde)

VT-1 Visual examination of shroud side auxiliary wedges on 8 pumps. No new indications. (Due to Laguna Verde)

EVT-1 Visual examination of strain relief welds RS-RW on the 9 risers that contain the welds. No new indications. (Due to Laguna Verde)

VT-3 Visual examination of 20 jet pump sensing lines due to SIL 420 Revision 1. No indications.

L1R12 (2008) UT UT of 14 hold down beams at BB-1, BB-2, and BB-3.

Indication found at 88-3 on Jet Pump 18 and beam replaced.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components In BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Relnspections Inspection VT-1 Visual examination of 9 auxiliary wedges. One indication on Jet Pump 16; accepted as is. No other indications.

VT-1 Visual examination of WD-1 on 1O pumps. New indications noted on jet pumps 8 (an auxiliary wedge was installed) and on jet pump 11 (accepted as-is).

EVT-1 Visual examination of 8 DF-2 welds. No indications.

VT-3 Visual examination of 5 slip joint clamps. No indications.

VT-1NT-3 Visual examination of 2 riser brace clamps installed in L1R11.

No indications.

VT-3 Visual examination of the inside of the diffuser on jet pumps 19 and 20. No indications.

L1R11 (2006) The hold-down beams on jet pumps 5, 6 9, and 1O were 1

proactively replaced with low stress beams.

EVT-1 Visual examination of RB-2 welds on 6 pumps. NRI.

Installation of riser brace clamps on the risers for jet pumps 5/6 and 9/10 to repair the RS-9 flaws identified in L1R10.

The slip joint clamps on jet pumps 5, 6, 9, and 1O were upgraded to a new style.

VT-3 Visual examination of the 16 old style slip joint clamps installed in the previous outage. No indications.

EVT-1 Visual examination of RB-1 on 12 jet pumps and RB-2 on 6 jet pumps. No indications.

VT-1 Visual examination of WD-1 on 20 jet pumps. No change in the wear identified in L1R10.

EVT-1 Visual examination of RS-3 on 5 pumps. No indications L1 R10 (2004) UT BB-1, BB-2, and BB-3 areas of all 20 hold-down beams.

Indications at BB-1 on Jet Pump 15 resulted in replacement of this beam with a low stress beam. When the inlet mixer for Jet Pump 19 was replaced, the beam was proactively replaced.

EVT-1 Visual examination of RS-3 on 5 risers. No indications.

VT-3 Best effort examination of the inaccessible welds AD-1, AD-2 and DF-3 on all 20 jet pumps. No indications.

EVT-1 Visual examination of DC-3 on 8 pumps. No indications.

EVT-1 Visual examination of DF-1 on 11 Jet Pumps. No indications.

EVT-1 Visual examination of DF-2 on 2 Jet Pumps. No indications.

EVT-1 Visual examination of RS-1 welds on all 1O risers. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection EVT-1 Visual examination of RS-2 welds on 5 risers. No indications.

EVT-1 Visual examination of RS-3 on 5 risers. No indications.

EVT-1 Visual examination of RS-6 and RS-7 on 1Ojet pumps. No indications.

EVT-1 Visual examination of RS-8 on all 20 jet pumps. No indications.

EVT-1 Visual examination of RS-9 on all 20 jet pumps. Indications found on 3 jet pumps (5, 6, and 9). Flaw evaluation performed and required the installation of a repair in L1R11.

EVT-1 Visual examination of IN-1 on 11 jet pumps. No indications.

EVT-1 Visual examination of IN-2 on 11 jet pumps. No indications.

EVT-1 Visual examination of MX-2 on 11 jet pumps. No indications.

EVT-1 Visual examination of RB-1 on 19 of the jet pumps. No indications.

EVT-1 Visual examination of RB-2 of 18 jet pumps. No indications.

VT-1 Visual examination of WD-1 on 20 jet pumps. Wear identified on 10 jet pumps. Wear accepted as-is on 9 jet pumps; inlet mixer for jet pump 19 replaced with a different inlet mixer.

VT-1 Visual examinations of 1O auxiliary wedges installed in previous outages. No indications.

Installed auxiliary wedges at the following vessel side locations: jet pumps 4, 12, 13, 141 15, 16, and 19. Installed auxiliary wedges at the following shroud side locations: jet pumps 1, 3, 4, 12, 14, and 16.

EVT-1 Visual examination of the strain relief welds on the 10 risers.

No indications.

Slip joint clamps were installed on all 20 jet pump inlet mixers.

L1R09 (2002) VT-3 Visual examination of WD-1 on 4 jet pumps. No indications.

Installed auxiliary wedges at the fallowing vessel side location: jet pump 6. Installed auxiliary wedge at the following shroud side location: 11.

VT-1 Visual examination of 2 auxiliary wedges installed in previous outages. No indications.

L1ROS (1999) UT UT of 10 jet pump beams at the BB-1 and BB-2 locations. No indications.

EVT-1 Visual examination of DF-1 on 1OJet Pumps. No indications.

EVT-1 Visual examination of DF-2 on 1OJet Pumps. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Relnspectlons Inspection EVT-1 Visual examination of RS-1 welds on 5 risers. No indications.

EVT-1 Visual examination of RS-2 welds on 5 risers. No indications.

EVT-1 Visual examination of RS-3 on 5 risers. No indications.

EVT-1 Visual examination of RS-6 and RS-7 on 10 jet pumps. No indications.

EVT-1 Visual examination of RS-8 on 1Ojet pumps. No indications.

EVT-1 Visual examination of RS-9 on 10 jet pumps. No indications.

EVT-1 Visual examination of IN-1 on 10 jet pumps. No indications.

EVT-1 Visual examination of IN-2 on 10 jet pumps. No indications.

EVT-1 Visual examination of MX-2 on 10 jet pumps. No indications.

EVT-1 Visual examination of R8-1 on 1Ojet pumps. No indications.

EVT-1 Visual examination of RB-2 on 1Ojet pumps. No indications.

VT-3 Visual examination of WD-1 on 20 jet pumps. Due to wear observed in L1R07. the inlet mixer on jet pump 9 was replaced and the wedge was oversized, and the restrainer bracket was machined to accommodate the larger wedge. To prevent flow imbalance the inlet mixer on jet pump 1O was proactively replaced.

Auxiliary wedges installed at the following vessel side locations: jet pumps 1, 5, 7, 8, and 10. Auxiliary wedge installed at the following shroud side location: jet pump 6.

VT-1 Gaps at the vessel side set screw were identified on 1 pump and accepted without installation of an auxiliary wedge for one cycle. Gaps at the shroud side set screw were identified on 1 pump and accepted without installation of an auxiliary wedge for one cycle.

The temporary auxiliary wedges installed on the vessel and shroud side of jet pump 9 were replaced with permanent auxiliary wedges. The wear on WD-1 was accepted for another cycle.

L1R07 (1996) VT-3 Visual examination of WD-1 on 2 jet pumps with wear observed on jet pump 9. Flaw evaluation determined acceptable for one cycle.

UT UT of all 20 jet pump holddown beams at 88-1; one indication on #9 beam; beam replaced.

VT-1 A gap was identified on the vessel side set screw of jet pump 9, and temporary wedges were installed at both set screws on jet pump 9.

LPCI Couplings L1R16 (2016) EVT-1 / VT-3 Visual examination of four locations on one coupling (315°).

I VT-1 No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components In BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Relnspectlons Inspection L 1R14 (2012) EVT-1 / VT-3 Visual examination of four locations on one coupling (45°).

I VT-1 No indications.

L1R13 (2010) EVT-1 Visual examination of one location (45-12) on one coupling (135°). No indications.

L 1R12 (2008) EVT-1 / VT-3 Visual examination of four locations on one coupling (135°).

I VT-1 No indications.

L1 R10 (2004) EVT-1 / VT-3 Visual examination of four locations on all three couplings.

I VT-1 No indications.

L1ROB (1999) EVT-1 / VT-3 Visual examination of four locations on all three couplings.

I VT-1 No indications.

Lower Plenum L1R14 (2012) VT-3 Areas below the core plate made accessible due to inspection of the bottom head drain line. No indications.

ICH RPV-1 at four locations.

ICHGT ICH-1 at four locations.

ICHS ICGT-1 at four locations.

ICHS-1 at four locations.

CRDH ST at eight locations.

CRDH-1 at eight locations.

ST RPV-1 at eight locations.

L1 R11 (2006) VT-3 Areas below the core plate made accessible due to the removal of the inlet mixers for jet pumps 5, 6, 9, and 1 o.

Areas include CRD/ST-1, bottom of H9, and ICH/RPV-1. No indications.

L 1R10 (2004) VT-3 Areas below the core plate made accessible due to the removal of the inlet mixer for jet pump 19. Areas include CRD/ST-1, bottom of H9, and ICH/RPV-1. No indications.

L 1R09 (2002) VT-3 / EVT-1 Visual examination of the fuel support guide tube pins (FS/GT-ARPIN-1) at 20 locations, CRGT-1 at 20 locations, CRGT-2 at 21 locations, and CRGT-3 at 21 locations. No indications.

L 1ROS (1999) VT-3 Visual examination of the fuel support guide tube pins (FS/GT-ARPIN-1) at 19 locations, the CRGT-1 at 19 locations. No indications.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components In BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection Steam Dryer L1R16 (2016) VT-1 Visual examination of one Drain Channel Vertical Weld. No indications.

Visual examination of four horizontal welds. No indications.

Visual examination of 29 vertical welds. No change to previously identified indications on V09-090, V04a-090, V04c-090, V04c-270, and V13-270. New indication on V04b-090.

Visual examination of nine Tie Bars with no change in one previously identified indication.

Visual examination of four Tie Rods with no change in one previously identified indication.

Visual examination of one lifting lug bracket location with no change in a previously identified indication.

L1R15 (2014) VT-1 Visual examination of lifting lug brackets at two locations, and the one flawed bracket was unchanged from last outage.

Visual examination of one tie rod with no change in degradation.

Visual inspection of two vertical welds, and no change in the indications.

Visual inspection of portions of the Upper Support Ring with no change in the indications.

L 1R14 (2012) VT-1 Visual examination of upper guide bracket with no indications Visual examination of existing flaws; Vertical welds in three locations with no changes noted; Tie Rods at two locations with no changes noted; Lifting lug welds at four locations with no changes noted; Upper support ring for 360 degrees with three new indications noted. All were evaluated and accepted without repair.

L1R13 (2010) VT-1 Examination of the dryer included 21 tie bars, 23 vertical welds, 5 horizontal welds, and 5 tie rods. The upper support ring was examined for 360°. The lug and four brackets on two lifting assemblies (225° and 315°) were examined.

Indications identified previously were examined and there were no changes in any indications. New indications were noted on the lifting lug #2, #3, and #4 brackets at 315°, tie rod 17-90°, V01-270°, and the USA from 180-360°. All indications were evaluated and accepted without repair.

Page 13

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspectlons Inspection L 1R12 (2008) VT-1 All welds on the half of the dryer between 0° and 180°,

including drain channels, tie bars, vertical welds, horizontal welds, and tie rods on both sides of the dryer. New indications were identified on TB-03, TB-08, TR-05-270, TR-05-90, TR-06-270, TR-06-90, TR-09-270, TR-09-90, TR 270, TR-10-90, TR-13-270, TR-13-90, TR-14-270, TR-14-90, TR-16-90, TR-17-270, TR-17-90, TR-18-270, TR-18-90, V04a-90, V04c-90, V05-90, V06-90, V09-90, V10-90, V13-90, V14-90, V15-90, V17-90, and upper support ring between 90-180. All were evaluated and accepted without repair.

VT-3 General inspection of half of the dryer between 180° and 360° above the waterline. No indications.

L1 R11 (2006) EVT-1 Re-inspection of lower guide bracket at 180° and hood A plate 5 where previous indications existed and were stop drilled. No new indications.

VT-1 All welds on the half of the dryer between 180° and 360°:

access hole cover, drain channels, vertical welds and horizontal welds. No new indications. Indications at V13-270 and V14-270 were re-examined and there was no growth.

L1R10 (2004) VT-3 Visual exams on the end panels and welds; one indication on bank 8, bank 2 which was stop drilled, and one previous indication on bank.. D bank 4 and there was no growth. All four lifting lugs and their brackets (previous indications at five locations with no growth), 100% of tie rods (1 O previous indications unchanged), 100% of tie bars VT-1 Visual examination of upper and lower guide brackets with an indication on the lower guide at 180° which was stop drilled, all horizontal welds, all horizontal plates (hood A plate 5 indication was stop drilled), hood F plate 1 (previous indication did not grow), 100% of the tie bars Top Guide L1R14 (2012) EVT-1 Visual examination of ten grid cells; two metal slivers Identified.

VT-3 Visual examination of one c-clamp; no indications.

L1R13 (2010) EVT-1 Visual examination of two grid cells; no indications.

VT-3 Visual examination of one c-clamp; no indications.

L 1R12 (2008) VT-3 Visual examination of two c-clamps; no indications.

L1R10 (2004) VT-3 Visual examination of two c-clamps; no indications.

L 1Roa {1999) VT-3 Visual examination of four c-clamps; no indications.

Vessel L1R12 (2008) VT-3 Inspection of the general condition of the RPV interior surface from the RPV closure flange elevation to the Steam Dam, 360° around the RPV interior. NRI.

Page 14

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Method Used Repairs, Replacements, Reinspections Inspection Inspection of the general condition of the RPV interior surface at the shroud support elevation above the gussets, 360° around the RPV interior. NRI.

L1R10 (2004) VT-3 Inspection of the general condition of the RPV interior surface from the RPV closure flange elevation to the Steam Dam, 360° around the RPV interior. NAI.

Inspection of the general condition of the cladding at the Steam Dam elevation, 360° around the RPV interior. NRI.

Inspection of the general condition of the RPV interior surface from below the core plate to the shroud support plate. NRI.

L1R09 (2002) VT-3 Inspection of the general condition of the RPV interior surface from the RPV closure flange elevation to the Steam Dam, 360° around the RPV interior. NRI.

Inspection of the general condition of the cladding at the Steam Dam elevation, 360° around the RPV interior. NRI.

DM Welds- L1A13 (2010) UT Inspection of 16 Category C OM welds; 10 automated and 6 BWRVIP-75-A manual. No indications Two Category D OM welds were identified on a flow venturi in the drywell in 2009, and the flow venturi was removed and replaced with a venturi that does not contain any welds.

Details will be provided to the BWRVIP and NRC under a separate letter.

L1R12 (2008) UT There were no dissimilar metal welds examined this outage.

Integrated Surveillance L1R13 (201 O) Removed the surveillance capsule at 120° to support analysis Program of the contents under the ISP.

Moisture Separator L1R16 (2016) UT Ultrasonic examination of all 24 shroud head bolts. No indications.

Other L1R16 (2016) VT-1 Visual examination of four IRM dry tubes (upper two feet and verification of plunger engagement per SIL 409). RI for one plunger not fully engaged with top guide, accepted as-is for one cycle.

Page 15

ATIACHMENT2 1o CFR 50.55a Relief Request 14R-02 Proposed Alternative In Accordance with 10 CFR 50.55a{z){1)

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

ENCLOSURE 3 LaSalle Unit 2 Reactor Internals Inspection History 11 pages follow

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Core Spray Piping L2R1S (201S) EVT-1 Welds APl, AP2, AP3, AP4a, AP4b, AP4c, AP4c-1, AP4d, APS, AP6,AP7,AP8a,AP8b,BP3,BP4a, BP4b,BP4c,BP4c-1,BP4d, BPS,BP6,BP7,BP8a,BP8b,CP1,CP2,CP3,CP4a,CP4b,CP4c, CP4c-~CP4d,CPS,CP6,CP7,CP8a,CP8~DP3,DP4~DP4~

DP4c, DP4c-1, DP4c-2, DP4d, DPS, DPG, DP7, DP8a, and DP8b.

Piping Brackets PB-068, PB-09S, PB-26S, and PB-292.

All NRI.

L2R14 (2013) EVT-1 Welds APl, AP2, AP3, AP4a, AP4b, AP4c, AP4c-l, AP4d, APS, AP6, AP7, AP8a, AP8b, BP3, BP4a, BP4b, BP4c, BP4c-1, BP4d, BPS,BP6,BP7,BP8a,BP8b,CP1,CP2,CP3,CP4a,CP4b,CP4c, CP4c-1,CP4d,CPS,CP6,CP7,CP8a,CP8b,DP3,DP4a,DP4b, DP4c, DP4c-1, DP4c-2, DP4d, DPS, DPG, DP7, DP8a, and DP8b.

Piping Brackets PB-015, PB-16S, PB-195, and PB 3S3.

All NRI.

L2R13 (2011) EVT-1 AP1,AP2,AP3,AP4a,AP4b,AP4~AP4d,AP8a,BP3,BP4a, BP4b,BP4c,BP4d,BP8a,CP1,CP2,CP3,CP4a,CP4b,CP4c, CP4d,CP8a,CP8b,DP3,DP4a,DP4b,DP4c,DP4d,DP8a,and DP8b. NRI. Three new longitudinal welds identified located adjacent to CP4a at 187°. All examined by EVT-1 with NRI.

L2R09 (2009) UT UT of 34 welds; APl, AP2, AP3, AP4a, AP4b, APS, AP6, AP7, AP4c, AP8b, BP3, BP4a, BP4b, BPS, BPG, BP7, BP4c, BP8b, CPl, CP2,CP3,CP4a, BP4b,CPS,CP6,CP7,CP4c,DP3,DP4a,DP4b, DPS, DP6, DP7, and DP4c. Existing flaw on BPS determined to be geometry. All others NRI. Four new welds identified, (AP4c-1, BP4c-l, CP4c-1, and DP4c-1) located between P4c and P7 (OE28372). All examined by UT with NRI.

EVT-1 APl, AP2, AP3, AP4a, AP4b, AP4c, AP4c-1, AP4d, AP8a, AP8b, BP3, BP4a, BP4b BP4c, BP4c-l, BP4d, BPS, BP8a, BP8b, CPl, 1

CP2,CP3,CP4a,CP4b,CP4c,CP4c-1,CP4d,CP8a,CP8b,DP3, DP4a, DP4b, DP4c, DP4c-l, DP4c-2, DP4d, DPS, DP8a, and DP8b. NRI. Five new welds identified, (AP4c-1, BP4c-1, CP4c-1, DP4c-1 and DP4c-2) located between P4c and P7. All examined by EVT-1 with NRI.

L2Rll (2007) EVT-1 Visual examinations of core spray piping welds for which the UT is not demonstrated. No indications (NRI).

8 piping brackets; NRI.

L2R10 (2005) UT UT of 34 welds; APl, AP2, AP3, AP4a, AP4b, APS, AP6, AP7, AP4c,AP8b,BP3,BP4a,BP4b,BPS,BP6,BP7, BP4c,BP8b,CP1, CP2,CP3,CP4a,CP4b,CPS,CP6,CP7,CP4c,DP3, DP4a,DP4b, DPS, DP6, DP7, and DP4c. Existing flaw on BPS re-sized with no growth. All others NRI.

EVT-1 Welds for which UT is not demonstrated: APl, AP4d, AP8a, BPS,BP4d,BP8a,CP1,CP4d,CP8a,CP8b,DP4d,DP8a,and DP8b. NRI.

2 core spray piping brackets; NRI.

Page 1

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Core Spray Piping L2R09 (2003) EVT-1 Visual examination of those core spray piping welds for which (continued) the UT technique is not demonstrated. No indications L2R08 {2000) UT UT for those welds for which the UT tool is qualified.

EVT-1 8 piping brackets; NRI.

Core Spray Sparger L2R15 (2015) EVT-1 SlA, 52A (left and right), S4A (left and right),

SlB, S2B (left and right), 54B (left and right),

SlC, S2C (left and right), 54C (left and right),

510, 520 (left and right), 540 (left and right).

All NRI.

VT-1 53A-a, 53A-b, and S3A-c from 007 to 088° 1 530-a and 530-b from 352 to 088°.

Bent sparger nozzle deflector identified in L2Rll unchanged; All others NRI.

6 sparger brackets; NRI.

L2R14 (2013) VT-1 6 sparger brackets; NRI.

S3A-a, 53A-b, and S3A-c from 272 to 007°.

S3C-a, S3C-b, and S3C-c from 092 to 187°.

530-a, and 530-b from 272 to 352°.

All NRI L2R13 (2011) VT-1 6 sparger brackets; NRI.

S3B-a, 53B-b and S3B-c from 172.5 to 268°, 53C-a and 53C-b from 187 to 268°; bent sparger nozzle deflector identified in L2R11 unchanged. All others NRI.

EVT-1 51A, S2A (left and right), S4A, 51B, 528 (left and right), 548, SlC, S2C (left and right), S4C, 510, 520 (left and right), 540.

NRI.

L2R12 (2009} VT-1 6 sparger brackets; NRI.

53A-a, 53A-b and S3A-c from 268 to 7.5°, S3A-a, 53A-b and 53A-c from 7.5 to 88°, 536-a, 538-b, and 538-c from 172.5 to 268° and 538-a, S3B-b, and S3B-c from 88 to 172.5°; bent sparger nozzle deflector identified in L2R11 unchanged. All others NRI.

L2Rll (2007) VT-1 530-a, 530-b, 530-c from 352 to 88° and 53C-a, S3C-b and S3C-c from 7.5 to 88°; one bent sparger nozzle deflector; all others NRI. Bent nozzle accepted for one cycle.

EVT-1 SlA, S2A (left and right), 54A, 518, 528 (left and right), S4B, SlC, S2C (left and right), S4C, SlD, 520 (left and right), 540.

NRI.

VT-1 6 sparger brackets; NRI.

L2R10 (2005) VT-1 S3A-a and 53A-b from 268 to 008° and 530-a and 530-b from 352 to 268°; NRI.

6 sparger brackets; NRI.

L2R09 (2003) EVT-1 Visual inspection of half of the core spray sparger welds. NRI.

6 sparger brackets; NRI.

Page 2

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Core Spray Sparger l2R08 (2000) VT-1 100% of all sparger welds. NRI.

(continued) EVT-1 12 sparger brackets; NRI.

Attachment Welds l2R16 (2017) VT-3 2 Guide Rod vessel attachment welds; NRI.

One feedwater sparger bracket pin, no change to previously identified minor wear.

All four steam dryer hold-down lugs; NRI.

(see jet pump section of this report) l2R15 (2015) VT-1 lower surveillance capsule bracket at 030°; NRI.

VT-3 Upper surveillance capsule bracket at 030°; NRI.

EVT-1 Twelve Feedwater sparger bracket to vessel welds; NRI.

VT-3 I VT-1 Eight Feedwater sparger bracket pins; minor wear at pin/bracket interface at six locations. Accepted as-is.

(see core spray section of this report)

L2R14 (2013) VT-1 Surveillance capsule holder lower bracket; NRI.

VT-3 Surveillance capsule holder upper bracket; NRI.

EVT-1 Four steam dryer support lug attachment welds. NRI.

Four Feedwater sparger bracket to vessel welds; NRI. Minor wear at pin/bracket interface on three of the spargers.

L2R13 (2011) VT-1 Lower surveillance capsule bracket at 120°; NRI.

VT-3 Upper surveillance capsule bracket at 120°; NRI.

L2Rll (2007) (see jet pump and core spray sections of this report)

L2R10 (2005) EVT-1 Steam dryer attachment welds, four locations; NRI.

VT-3 Upper bracket attachment welds for surveillance baskets at three locations, NRI.

VT-1 Lower bracket attachment welds for surveillance baskets at three locations. Basket disengaged at 120° location and accepted for one cycle. All others NRI.

EVT-1 All Feedwater sparger attachment welds; NRI.

L2R09 (2003) EVT-1 All Feedwater sparger attachment welds; NRI.

L2 R08 (2000) VT-1 Steam dryer attachment welds, four locations, NRI.

VT-3 Guide Rod attachments at 0° and 180°; NRI.

Upper surveillance capsule brackets at three locations; NRI.

VT-1 Lower surveillance capsule brackets at three locations; NRI.

Core Shroud L2R16 {2017) EVT-1 Visual inspection of H4 for off-axis cracking in response to EPRI Letter 2016-030; NRI.

L2R15 (2015) UT UT of welds Hl, H2, H3, H4, HS, H6, H7, and H8. Flaws identified on Hl, H3, and H4. Accepted as-is with 10 year EOI.

EVT-1 Visual inspection (from shroud ID and OD) of vertical welds V12, V13, V14, and VlS. NRI.

Visual inspection (from shroud OD) of horizontal welds Hl and H2; NRI.

L2R11 (2007) VT-3 Surfaces of the shroud for ASME Section XI. NRI.

Page 3

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Core Shroud L2R10 (200S) UT UT of welds H3, HS, HG, and H8 (continued) All welds are .NRI.

L2R07 (199G) UT UT of HS, HG, and HS. NRI.

L2ROG (199S) UT UT of H3, H4, HS, HG, and HS. NRI.

Shroud Support L2R1G (2017) EVT-1 Top of H8a weld at the 0° and 180° locations. NRI.

L2R1S (201S) EVT-1 Access hole cover at 0°; NRI.

L2R14 (2013) EVT-1 Top of H9 weld at the 0° and 180° locations. NRI.

Access Hole Cover at 180°. N RI.

L2R13 (2011) EVT-1 H8a welds (BWRVIP weld HS) for> 10%-- NRI.

L2R12 (2009) EVT-1 Access Hole Covers at O and 180°-- NRI.

L2Rll (2007) VT-3 Access Hole Covers at 0 and 180° for ASME Section XI. NRI.

Accessible portions of the top of the shroud support plate for ASME Section XI. NRI.

Top of H9 weld (accessible locations) for ASME Section XI. NRI.

L2R10 (200S) VT-1 Access Hole Covers at O and 180°-- NRI.

VT-3 Inspections of the general condition of the RPV interior surface from the RPV closure flange elevation to the steam dam, 3G0° around the RPV interior. NRI.

Inspection of the general condition of the cladding at the steam dam elevation, 360° around the RPV interior. NRI.

Examined RPV cladding from below core plate to shroud support plate due to removal of the inlet mixers. NRI.

EVT-1 H8a weld (BWRVIP weld HS) for> 10%-- NRI.

VT-3 Inspection of the general condition of weld H9 from below the shroud support plate due to removal of all jet pump inlet mixers. NRI.

L2R09 (2003) UT UT of 100% of H9 from the RPV OD. NRI.

VT-3 Inspection of the general condition of the RPV interior surface from the RPV closure flange elevation to the Steam Dam, 360° around the RPV interior. NRI.

Inspection of the general condition of the cladding at the steam dam elevation, 360° around the RPV interior. NRI.

Top Guide L2R16 (2017) VT-3 C-clamp at 0°. NRI.

L2R14 (2013) VT-3 C-clamp at 270°. NRI L2R13 {2011) VT-3 C-clamp at 090 and 180° locations - NRI.

EVT-1 Lower Beam and Slot intersections of 19 cells - NRI.

L2Rll (2007) VT-3 C-clamp at 0°. NRI.

Accessible portions of the top guide for ASME Section XI. NRI.

L2R10 (2005) VT-3 C-clamps at 4 locations - NRI.

Standby Liquid L2R16 (2017} UT UT of the partial penetration weld and HAZ. NRI.

Control VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

Page4

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Standby Liquid L2R15 (2015) VT-2 Visual inspection of the partial penetration weld to the bottom Control head during the Section XI system leak test. NRI.

(continued) L2R14 (2013) VT-2 Visual inspection of the partial. penetration weld to the bottom head during the Section XI system leak test. NRI.

L2R13 (2011) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

L2R12 (2009) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

L2Rll (2007) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

UT UT of the partial penetration weld and heat affected zone.

NRI.

L2R10 (2005) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

L2R09 (2003) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test. NRI.

PT Surface examination. NRI.

L2R08 (2000) VT-2 Visual inspection of the partial penetration weld to the bottom head during the Section XI system leak test NRI.

Jet Pump Assembly L2R16 (2017) VT-1 Main wedge WD-1 on all 20 pumps. Minor new wear on JP 3 prompted inspection of Jet Pump 3 AS-1, AS-2, and WD-2 locations. Gaps seen at AS-1 were closed by adjusting main wedge. Indication on one of two VS AS-2 tack welds accepted for one cycle.

Four auxiliary wedges; no change in existing wear on JP 15 VS, all others NRI.

L2R15 (2015) VT-1 Main wedge WD-1 on all 20 pumps; minor new rod wear on 2 pumps. Accepted as-is, all others were either NRI or wear was unchanged from previous exam.

Auxiliary wedge on Vessel Side of JP 15, no change in existing wear.

L2R14 (2013} VT-1 Sensing line brackets on eight pumps; NRI.

Main wedge WD-1 on all 20 pumps; minor new rod wear on 3 pumps; all others were either NRI or wear was unchanged from last exam.

Auxiliary wedge on Vessel Side of JP 15; no change in existing wear.

VT-3 Visual of the slip joint area on all 20 pumps in response to GEH SC 12-12 and 12-14. NRI.

Pages

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Jet Pump Assembly L2R13 (2011) VT-1 WD-1 on all 20 pumps; three wedges showed movement with (continued) no change in wear from previous examinations; five rods showed new wear; accepted as-is.

Four auxiliary wedges examined, two vessel side and two shroud side. One vessel side auxiliary wedge had new minor wear at the contact point with the belly band and movement accepted-as-is. Other auxiliary wedges were NRI.

L2R12 {2009) VT-1 WD-1 on 9 pumps; all showed minor wear with most unchanged from previous examinations.

Three auxiliary wedges examined, two vessel side and one shroud side. One vessel side auxiliary wedge had minor wear.

Set screw on same pump confirmed to be in contact with the belly band and tack welds intact. Other auxiliary wedges were NRI.

L2R11 (2007) VT-1 WD-1 wedges on all 20 pumps; 7 wedges/rods showed minor wear; accepted-as-is. Auxiliary wedges installed at four locations on 3 pumps to compensate for observed gaps.

VT-3 Examination of ratchet teeth engagement on 13 jet pump hold down beams due to fitup issues in the previous outage. NRI.

L2R10 {2005) VT-1 All 20 inlet mixers were replaced with new inlet mixers with labyrinth seals in the slip joint area, and with new non-stellite main wedges. New hold down beams were installed on 17 pumps. After replacement, three point contact verified at all locations (AS-1 shroud side, AS-1 vessel side, and WD-1). NRI.

L2R09 (2003) Replacement of 3 beams. After a review of material certification paperwork that identified them as Group 1 beams, three holddown beams were replaced with low stress beams.

VT-1 WD-1 on 3 pumps; NRI.

Installed 3 Aux. Wedges to ensure three point contact for three pumps.

L2R08 {2000) UT UT exams of 10 beams at the BB-1 and BB-2 locations. NRI.

VT-3 Exam of WD-1 on all 20 pumps; NRI. Exam of all set screw to belly band contact points; installed 7 auxiliary wedges to maintain three point contact; all others NRI.

Jet Pump Diffuser L2R16 (2017) EVT-1 AD-2 on 5 pumps; NRI.

DF-1 on 5 pumps; NRI.

DF-2 on 5 pumps; NRI.

DF-3 on 5 pumps; NRI.

IN-1 on 5 pumps; NRI.

IN-2 on 5 pumps; NRI.

Page 6

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Jet Pump Diffuser L2R15 {2015} EVT-1 AD-2 on 5 pumps; NRI.

(continued) DF-1 on 5 pumps; NRI.

DF-2 on 5 pumps; NRI.

DF-3 on 5 pumps; NRI.

L2R13 (2011) EVT-1 IN-1on5 pumps; NRI.

IN-2 on 5 pumps; NRI.

L2Rll (2007) EVT-1 AD-2 on 6 pumps; NRI.

DF-1 on 10 pumps; NRI.

L2R10 (2005) EVT-1 AD-2 on 4 pumps; NRI.

DC-3 on 10 pumps; NRI.

DF-2 on 10 pumps; NRI.

DF-3 on 10 pumps; NRI.

L2R09 (2003) EVT-1 AD-2 on 4 pumps; NRI.

DF-1 on 4 pumps; NRI.

DF-2 on 4 pumps; NRI.

DF-3 on 4 pumps; NRI.

IN-1 on 10 pumps; NRI.

IN-2 on 10 pumps; NRI.

L2R08 {2000) UT AD-2 on 6 pumps; NRI.

DF-1 on 6 pumps; NRI.

DF-2 on 6 pumps; NRI.

DF-3 on 6 pumps; NRI.

MX-2 on 6 pumps; NRI.

Jet Pump Riser L2R16 (2017) EVT-1 RS-2 on 3 risers; NRI.

RS-3 on 5 risers; NRI.

RS-6 on 2 risers; NRI.

RS-7 on 3 risers; NRI.

RS-8 on 3 risers; N RI.

Corners of RS-9 on all 10 risers; NRI.

RB-1 on 5 jet pumps; NRI.

RB-2 on 5 jet pumps; NRI.

All RS-1 welds (including pup piece welds) on 4 risers. Re-sized flaw on RS-le on 19/20 with no change in length, accepted for four cycles. All others NRI.

Strain relief welds on four risers; NRI.

L2R15 (2015) EVT-1 Corners of RS-9 on all 10 risers; NRI.

RS-6 on 1 riser; NRI.

RS-7 on 1 riser; NRI.

L2R14 (2013) EVT-1 RS-9 on 9 risers; NRI.

RS-la on 19/20; NRI.

Re-sized flaw on RS-le on 19/20; no change in length.

Page 7

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Jet Pump Riser L2Rl3 (2011) EVT-1 All RS-1 welds (including pup piece welds) on 2 risers; NRI.

(continued) RS-2 on 3 risers; NRI.

RS-3 on 5 risers; NRI.

RS-6 on 2 risers; NRI.

RS-7 on 1 riser; NRI.

RS-8 on all 10 risers; NRI.

RS-9 on all 10 risers; NRI.

RB-1 on 5 jet pumps; NRI.

RB-2 on 5 jet pumps; NRI.

Strain relief welds on four risers; NRI.

L2Rl2 (2009) EVT-1 All RS-1 welds (including pup piece welds) on 8 risers; Re-sized flaw on RS-le on 19/20; no change in length. All others NRI.

L2Rll (2007) EVT-1 Re-sized flaw on RS-le on 19/20; no change in length.

RS-1 on 2 risers; NRI.

RB-1 on 12 jet pumps; NRI.

L2R10 (2005) EVT-1 Examined strain relief welds on all 10 risers. NRI.

Re-sized flaw on RS-le on 19/20; no change in length.

RS-2 on 3 risers; NRI.

RS-3 on 5 risers; NRI.

RS-6/7 on 10 jet pumps; NRI.

RS-8/9 on all 20 jet pumps; NRI.

RB-1 on 10 jet pumps; NRI.

RB-2 on all 20 jet pumps; NRI.

L2R09 (2003) EVT-1 Re-examined flaws on two RS-1 welds; that on the 1/2 riser was determined to be non-relevant; those on the 19/20 riser were re-sized, with no change since L2R07 (1996).

MX-2 on 4 pumps; NRI.

RS-6/7 on ten pumps; NRI.

L2R08 (2000) UT UT exam of MX-2 on 6 pumps; NRI.

L2R07 (1996) VT-1 RS-1 on all ten risers; two indications; one on the Yz riser and the second on the 19/20 riser; both accepted for two cycles.

RS-2 on all ten risers; NRI.

RS-3 on all ten risers; NRI.

Steam Dryer L2R16 {2017) VT-1 Ten tie bars; NRI.

One tie rod; no change in existing indication One access hole cover at 270° Two vertical welds on the 90° side and ten vertical welds on the 270° side; NRI Lifting Lug Bracket welds at 45°, NRI.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Steam Dryer L2R15 (2015) VT-1 Fifteen tie bars; NRI.

(continued) Two tie rods; NRI for TR-01-090, no change in existing indications on TR-21-090.

One drain channel vertical weld at 240°, eleven vertical welds 0

on the 270° side, seven vertical welds on the go side, four horizontal welds on the 270° side. All NRI.

L2R14 (2013) VT-1 Two access hole covers associated with the seismic blocks at 5° and 185°. NRI.

Re-examined tie bars TB-05 and TB-28; no change in existing indications.

Re-examined tie rod TR-21-090; no change in existing indications.

  • L2R13 (2011) VT-1 One vertical weld on the 270° side; NRI.

0 One horizontal weld on the go side; NRI.

Upper and lower guide rod bracket at 180° and lower guide bracket at 0°. Previous indications showed no change.

Two tie bars. Previous indications on tie bars reviewed and unchanged.

Four tie rods; one previous indication re-confirmed with no change. All others NRI.

Upper support ring from 0-360°; previous indication re-confirmed with no change.

0 L2R12 (20Qg) VT-1 11 vertical welds on the go side; all NRI.

6 horizontal welds; all NRI.

Upper and lower guide bracket at 180° and lower guide bracket at 0°. All RI as was the top of the guide rod at 180°.

Conditions accepted as-is.

Upper guide bracket at 0°; NRI.

15 tie bars. Previous indications on tie bars reviewed and unchanged. All others NRI.

Two tie rods; one previous indication re-confirmed with no change. Second tie rod was NRI.

Upper support ring from 0-360°; top and side RI; bottom NRI.

All indications accepted as-is.

L2Rll (2007) EVT-1 All welds recommended by BWRVIP-13g and SIL 644 Revision 2 0

for a curved hood dryer on the go side of the dryer, tie rods on both sides, upper support ring external surfaces, upper and lower guide at 180°, (indication on the lower guide bracket accepted for one cycle), lifting lugs and lifting assembly brackets at 45 and 135°, and 18 tie bars. Previous indications on tie bars reviewed and no change in sizes. All other welds NRI.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Steam Dryer L2R10 (2005) Vf-1 All welds recommended by SIL 644 Revision 1 for a curved (continued) hood dryer on the 270° side of the dryer, horizontal bank welds on both the 90° and 270° sides, all four lifting lugs, lifting assembly brackets at 225° and 315° locations, and all tie bars.

Indications found on three tie bars and accepted for one cycle.

Upper strap on lifting assembly at 215° found broken and was removed. All other welds NRI.

Vessel L2R10 {2005) VT-3 Inspection of the general condition of the RPV interior surface from the RPV closure flange elevation to the Steam Dam, 360° around the RPV interior. NRI.

L2R09 (2003) VT-3 Inspection of the general condition of the cladding at the Steam Dam elevation, 360° around the RPV interior. NRI.

LPCI L2R15 (2015) EVT-1 Examined welds 45-12 and 45-3b at the 045° coupling; NRI.

Vf-1 Examined weld 45-Sa,b,c,d at the 045° coupling; NRI.

VT-3 Examined bolts 45-6a,b,c,d at the 045° coupling; NRI.

L2R14 (2013) EVT-1 Examined weld 45-12 at the 315° coupling. NRI.

L2R13 (2011) EVT-1 Examined welds 45-12 and 45-3b at 135°; NRI.

VT-1 Examined weld 45-Sa,b,c,d at 135°. NRI.

VT-3 Examined bolts 45-Ga,b,c,d at 135°; NRI.

L2Rll (2007} EVT-1 Examined welds45-12a,b,c,d and 45-03a-d coupling at 315°;

NRI.

VT-1 Examined weld 45-08a,b,c,d at 315°; NRI.

VT-3 Examined bolts45-06a,b,c,d coupling at 315°; NRI.

L2R10 (2005) EVT-1 Examined welds45-12a,b,c,d and 45-03a,b,c,d at 45, 135, and 315°; NRI.

VT-1 Examined weld 45-08a,b,c,d at 45, 135, and 315°; NRI.

VT-3 Examined bolts45-06a,b,c,d at 45, 135, and 315°; NRI.

L2R08 {2000) EVT-1 BWRVIP-42, visual examination of all three LPCI couplings, No indications detected.

L2R07 (1996) VT-3 (every other outage) VT-3 of all three couplings; NRI.

Lower Plenum L2R16 {2017) VT-3 Best-effort examination of all areas below the core plate made accessible by removal of two guide tubes for access to bottom head drain. Areas examined included bottom head drain, four ICH RPV-1, four ICHGT ICH-1, four ICHS-ICGT-1, four ICHS-1 welds, ten CRDH ST, ten CRDH-1, and ten ST RPV-1 welds. All NRI.

L2R14 (2013) VT-1 Examined all areas below the core plate made accessible by FME inspection of bottom head area and inside of bottom head drain line. Areas examined included four ICH RPV-1 welds, four ICHS ICGT-1 welds, four ICHS-1 welds, eight CRDH ST welds, eight CRDH-1 welds, and eight RPV-1 welds. All NRI.

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REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVIP Scope Method Used Repairs, Replacements, Reinspections Inspection Lower Plenum L2R10 (2005) VT-1 Examined all areas below the core plate made accessible by (continued) disassembly of 20 jet pumps. Areas examined included CRD/ST-1, ST/RPV-1, H8a, H9, HlO, Hll, H12, ICH/RPV-1, and bottom head cladding. NRI for all twenty locations.

L2R09 (2003) VT-3 / EVT-1 Visual examination of the fuel support guide tube pins (FS/GT-ARPIN-1) at 4 locations, CRGT-1at4 locations, CRGT-2 at 21 locations, and CRGT-3 at 21 locations. No indications.

L2R08 (2000) VT-3 Visual examination of the fuel support guide tube pins (FS/GT-ARPIN-1) at 14 locations, CRGT-1at15 locations. No indications.

DM Welds- L2R16 (2017) UT One weld examined by automated UT; no unacceptable flaws BWRVIP-75-A detected.

Category B L2R15 (2015) UT Five welds examined by automated UT; no unacceptable flaws detected.

L2Rl3 (2011) UT Eight welds examined by manual UT; no flaws.

DM Welds- L2R16 (2017) UT One weld examined by automated UT; no unacceptable flaws BWRVIP-75-A detected.

Category C DM Welds- L2R13 (2011) UT Two welds examined by manual UT; no flaws.

BWRVIP-75-A L2R12 (2009) Two previous un-identified Category D welds were identified Category D and the spoolpiece on which the welds were located was replaced with a spoolpiece with non-IGSCC susceptible welds.

Other L2R16 (2017) VT-1 Visual examination of 5 IRM/SRM dry tubes (upper two feet and verification of plunger engagement per SIL 409). RI for 3 plungers not fully engaged, accepted as-is for one cycle.

Replacement of four original equipment IRM dry tubes FME Search Removed FME from inside the bottom head drain line. Minor particulate noted and vacuumed from bottom head area.

VT-3 Verified engagement of surveillance capsule basket at 030°.

VT-3 Guide Rod Cap at 000° 1 tack weld indications accepted as-is.

L2R15 (2015) VT-1 Visual examination of 2 SRM dry tubes and 4 IRM dry tubes (upper two feet and verification of plunger engagement per SIL 409 Rev 4). RI for 4 plungers not fully engaged with top guide, accepted as-is for one cycle.

Replacement of 1 SRM dry tube, due to detector replacement.

L2R14 (2013) VT-3 Verified engagement of surveillance capsule basket at 030°.

FME Search Removed FME from inside the bottom head drain line. Minor particulate noted on bottom head area.

VT-1 Visual examination of the upper two feet of four IRM dry tubes and 1 SRM dry tube. All NRI.

L2R13 (2011) VT-3 Verified engagement of surveillance capsule basket at 030°.

L2R11 (2007) VT-3 Removal of the surveillance capsule basket from 120° due to a broken spring.

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ATTACHMENT 3 10 CFR 50.55a Relief Request 14R-03 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)

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

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-K Item Number: B10.10

Description:

Examination of the Reactor Pressure Vessel (RPV)

Stabilizer Bracket Welds on Shell Course Component Number: RPV Stabilizer Bracket Welds Drawing Number: GEL-1109 (Unit 1)

GEL-1119 (Unit 2)

2. Applicable Code Edition and Addenda The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Requirement

Per Table IWB-2500-1, Examination Category B-K, Item Number B10.10, a surface examination is required over essentially 100% of the length of the weld for Pressure Vessel welded attachments. Examination Category B-K, Note 4 states that for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination.

4. Impracticality of Compliance In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical.

Due to other plant structures and components, accessibility of 100% of the examination areas for these welds was not provided in the original plant design which occurred prior to the issuance of ASME Section XI lnservice Inspection requirements. As indicated by Figure 14R-03.1, surface examination of the vessel stabilizer bracket weld is not possible due to the proximity of the bioshield wall, the stabilizer bar, and vessel insulation.

The stabilizer bracket lugs are approximately 1411 long by 7" wide. From past examination data and configuration details, LaSalle County Station has identified that the entire lower length of the weld is inaccessible for surface or visual examination due to the biological shield wall and the stabilizer bar. The two side lengths and the entire top length are also not accessible for surface examination due to the Reactor Pressure Vessel insulation; however, limited visual access can be obtained. LaSalle County Station has calculated the best effort examination coverage to be between 52% and 66% of the entire weld surface. This would account for the entire top length of the weld and essentially the full length of the side welds.

ATTACHMENT 3 10 CFR 50.55a Relief Request 14R-03 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)

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Due to the vessel insulation being panel type linked by several small screws which are inaccessible for the subject panels due to the stabilizer bars and brackets, the ability to remove the insulation is not practical due to the tight clearances and potential to damage the fasteners and insulation joints.

Based on the limited geometrical configuration described above and detailed in Figure 14R-03.1, LaSalle County Station requests relief from the ASME Section XI surface examination coverage area requirements for the six stabilizer bracket lugs per unit.

5. Burden Caused by Compliance Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating the subject Reactor Pressure Vessel. Based on this, the Code requirements are deemed impractical in accordance with 10 CFR 50.55a(g)(5)(iii).
6. Proposed Alternative and Basis for Use As an alternate examination to the Code required surface examination, LaSalle County Station will perform a VT-1 visual examination on the accessible portions of the stabilizer bracket welds (approximately 52% of the weld length) when one of these lugs is scheduled for examination in accordance with ASME Section XI.
7. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units 1 and 2.
8. Precedent LaSalle County Station Third ISi Interval Relief Request 13R-04 was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated January 30, 2008 (ADAMS Accession No. ML073610587). This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.

ATTACHMENT 3 1o CFR 50.55a Relief Request 14R-03 Relief Requested In Accordance with 10 CFR 50.55a(g)(5)(iii)

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Figure 14R-03.1 REACTOR PRESSURE VESSEL STABILIZER BRACKET WELDS r

14.0"

~-7.0" _...__...,

17.25"

.., . ' .I

~-~___. ~

    • "'~hieldWall

'\.i I

' '"... I Reactor Vessel O.D.

Insulation 34.0"

ATTACHMENT 4 10 CFR 50.55a Relief Request 14R-04 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

Reference:

IWL-2421 Examination Category: L-B Item Number: L2.10, L2.20

Description:

Post-Tensioning System Inspection Scheduling Requirements for Sites with Two Plants Component Number: Tendons and Wire Strands for Class CC Concrete Containment

2. Applicable Code Edition and Addenda

The Third 10-Year Interval of the LaSalle County Station, Units 1 and 2 Containment lnservice Inspection (CISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Requirement

IWL-2421 (a) of ASME Section XI allows the test schedule for concrete containment post-tensioning systems for sites with two plants to be modified if the following are met:

  • Both primary containments utilize the same prestressing system and are essentially identical in design;
  • Post-tensioning operations for the two primary containments were completed not more than two years apart, and;
  • Both containments are similarly exposed to or protected from the outside environment.

IWL-2421 (b) of ASME Section XI specifies the modified test schedule when the conditions of IWL-2421 (a) are met.

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.

On August 8 1996, the Nuclear Regulatory Commission (NRC) published a final rule in the Federal Register (FR) (i.e., 61 FR 41303) to amend 10 CFR 50.55a, "Codes and standards," to incorporate by reference Subsection IWL of ASME Section XI.

Subsection IWL of ASME Section XI, provides rules for the containment inservice inspection and repair/replacement activities of the reinforced concrete and post tensioning systems of Class CC components. LaSalle County Station, Units 1 and 2 primary containments are Class CC components.

ATIACHMENT4 10 CFR 50.55a Relief Request 14R-04 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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The amended 10 CFR 50.55a required incorporation of Subsection IWL into inspection programs by September 9, 2001. The initial IWL examinations were allowed to be based on the existing (prior to September 9, 1996) post-tensioning system program schedules per 10 CFR 50.55a{g)(6)(ii)(B)(4) at that time. After establishing this initial Subsection IWL inspection date, subsequent 5-year inspections are based on IWL-2400.

LaSalle County Station maintains an inspection program to implement Subsection IWL requirements.

IWL-2421 (a) allows the test schedule for post-tensioning systems of the concrete containments for sites with two plants to be modified if (1) both containments utilize the same prestressing system and are essentially identical in design, (2) post-tensioning operations for the two containments were completed not more than two years apart, and (3) both containments are similarly exposed to or protected from the outside environment. LaSalle County Station, Units 1 and 2, primary containments utilize the same prestressing system, are essentially identical in design, and both primary containments are similarly exposed to or protected from the outside environment.

Regarding the final condition, LaSalle County Station, Unit 1, post-tensioning operation was performed in July 1978, and Unit 2 post-tensioning operation was performed in December 1980 {29 months apart).

Prior to the endorsement of the IWL rules in 10 CFR 50.55a, by NRC letter dated June 3, 1994 (Reference 1), the NRC in Amendment No. 100 for Unit 1 and Amendment No. 84 for Unit 2 approved the use of the guidance contained in Regulatory Guide (RG) 1.35, 11 lnservice Inspection of Ungrouted Tendons in Prestresssed Concrete 11 Containments, Revision 3, and the use of the provisions of Surveillance Requirement (SR) 3.0.2. Additionally, the NRC reviewed LaSalle County Station's request at that time to treat the Units 1 and 2 primary containments as "twin containments" even though the initial Structural Integrity Tests (SITs) were not within two years of each other as described in RG 1.35. (Note that the 2-year period in Regulatory Guide 1.35 was based on the SIT dates whereas the two year period in IWL-241 O(a) is based on Post-Tensioning Operation.) The LaSalle County Station, Unit 1, initial SIT was performed in December 1978, and the Unit 2 initial SIT was performed in June 1983 {55 months apart) .

As documented in Reference 1, the NRC approved this approach based on a detailed review of data from six Unit 1 and four Unit 2 inservice inspections. The NRC reviewers noted that for the lift-off forces, the difference between the two unit's construction dates was of little significance. The NRC review of this data concluded that there is reasonable agreement in the deflection values obtained during the SITs at comparable locations of the primary containments and that the treatment of the LaSalle County Station, Units 1 and 2, primary containments as "twin containments" was acceptable.

The current Subsection IWL Program for tendons and wires/strands is based on the continued treatment of the LaSalle County Station, Units 1 and 2, primary containments as "twin containments!' Post-tensioning system inspections completed to date have been performed for the 1st, 3rd, 5th, 1oth, 15th, 20th, 25th, 30th and 35th years for Unit 1, and the 1st, 3rd, 5th, 10th, 15th, 20th, 25th, and 30th years for Unit 2, with the tendon and wire/strand tests being completed every other 5-year period. These Post-Tensioning

ATTACHMENT 4 10 CFR 50.55a Relief Request 14R-04 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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System tests and examinations for both units have all met the applicable acceptance criteria, or ASME Section XI repair and replacement activities have been completed to return them to acceptable condition. The results of these inspections completed to date demonstrate that the performance of the Unit 2 post-tensioning relative to the Unit 1 post-tensioning (i.e., 29 months apart) is not a factor contributing to any unique condition that may subject either primary containment to a different potential for structural or tendon deterioration.

Relief is requested from the IWL-2421 (a) requirement to apply the modified test schedule of IWL-2421 (b) only if post-tensioning operations for the two primary containments were completed not more than two years apart. Based on the above discussion, the proposed alternative will provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) for the Fourth ISi Interval, as well as the remaining term of the renewed facility operating licenses for LaSalle County Station, Units 1 and 2.

5. Proposed Alternative and Basis for Use The modified test schedule of IWL-2421 (b) will continue to be used for LaSalle County Station, Units 1 and 2 tendon tests (L2.10) and wire/strand examinations (L2.20). The initiation of the IWL-2400 rolling 5-year schedule was based on the previous inspection dates under the Station Tendon Surveillance Program prior to Subsection IWL being endorsed and will continue throughout each 120-month interval.
6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval, as well as and the remaining term of the renewed facility operating licenses (NPF-11 for Unit 1 and NPF-18 for Unit 2), which currently expire at midnight on April 17, 2042, and at midnight on December 16, 2043, for LaSalle County Station, Units 1 and 2, respectively. The Fourth ISi Interval, as well as the remaining term of the renewed facility operating licenses refers to the LaSalle County Station, Units 1 and 2 current fourth and upcoming fifth and sixth 120-month ISi Program intervals (Reference 2).
7. Precedent LaSalle County Station Second CISI Interval Relief Request 13R-05 was authorized per NRC Safety Evaluation (SE) dated January 15, 2008 (ADAMS Accession No. ML073521568). This relief request for the LaSalle County Station, Units 1 and 2 Third CISI Interval, utilizes a similar approach to the previously approved relief request.
8. References
1. Letter from Anthony T. Gody (NRC) to D. L. Farrar (Commonwealth Edison Company), 11 lssuance of Amendments (TAC Nos. M87305 and M87306), 11 dated June 3, 1994 (ADAMS Accession No. ML021130097)

ATTACHMENT 4 10 CFR 50.SSa Relief Request 14R-04 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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2. Letter from Jeffrey S. Mitchell (U.S. Nuclear Regulatory Commission) to M. P. Gallagher (EGC), "Issuance of Renewed Facility Operating Licenses for LaSalle County Station, Units 1 and 2 (TAC Nos. MF5347 and MF5346)," dated October 19, 2016 (ADAMS Accession No. ML16202A075).

ATIACHMENT5 10 CFR 50.55a Relief Request 14R-06 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

Reference:

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

Description:

Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators Component Number: CRD Accumulators and Associated Piping

2. Applicable Code Edition and Addenda

The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Requirement

Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with IWC-5220. This pressure test is to be conducted once each inspection period.

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.

LaSalle County Station, Units 1 and 2, Technical Specification (TS} Surveillance Requirement (SR) 3.1.5.1 requires each control rod scram accumulator pressure to be equal to or greater than 940 psig for the control rod scram accumulator to be considered operable. The TS SR is required to be met whenever the unit is operating in Modes 1 and 2. The accumulator pressure is continuously monitored by system instrumentation and surveillance is performed on a weekly basis that requires a physical walkdown of all CRD accumulators. The walkdown is intended to identify any system air leaks and negative trending in system pressure. The accumulators are isolated from the source of makeup nitrogen, thus the continuous monitoring of the CRD accumulators currently functions as a pressure decay type test. The accumulators are maintained at a pressure of approximately 1100 psig during operation. Should accumulator pressure fall below 1000 psig (-15 psig), an alarm is received in the control room. The pressure drop for the associated accumulator is then recorded in the control room log, and the accumulator is recharged by station procedure LOP-RD-20, 11 Control Rod Accumulator Recharging/Water Removal. 11 Other corrective actions, including soap bubble application to locate leakage or equipment repair are performed, as required, in accordance with the Corrective Action Program.

ATTACHMENT 5 1o CFR 50.55a Relief Request 14R-06 Proposed Alternative In Accordance with 10 CFR 50.55a{z){1)

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Since the monitoring of the nitrogen side of the accumulator at pressures consistent with the requirements of Table IWC-2500-1 is continuous, any degradation of the accumulator and associated piping would be detected by normal system instrumentation. The accumulators are normally passive components and are susceptible to slow developing failure modes. Corrosion and tubing connection integrity are the primary modes of failure. Continuous monitoring will detect degrading conditions of individual accumulators due to these failure modes before similar detection by the code required examination. The continuous monitoring of the CRD accumulators and associated piping exceeds the code requirement of inspecting the system once per inspection period. The additional VT-2 visual examination 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 would require applying a leak detection solution to 185 accumulators per unit in an elevated dose rate area. This results in radiation exposure (estimated 108 mrem three times per interval for a total of 324 mrem) without any added benefit in the level of quality and safety. This inspection would not be consistent with as low as reasonably achievable (ALARA) practices.

Relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Table IWC-2500-1 for the nitrogen side of the CRD system accumulators and associated piping on the basis that the requirements of TS SR 3.1.5.1 exceeds the code required examinations.

5. Proposed Alternative and Basis for Use As an alternative to the VT-2 visual examination requirements of Table IWC-2500-1, LaSalle County Station will perform continuous pressure decay monitoring for the nitrogen side of the CRD Accumulators and associated piping and a weekly surveillance in accordance with TS SR 3.1.5.1 that requires a physical walkdown of all CRD accumulators.
6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units 1 and 2.
7. Precedents
  • LaSalle County Station Third ISi Interval Relief Request 13R-09 was authorized per Nuclear Regulatory Commission (NRG) Safety Evaluation (SE) dated January 30, 2008 (ADAMS Accession No. ML073610587). This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.
  • Susquehanna Steam Electric Station Fourth ISi Interval Relief Request 4RR-08 was authorized per SE dated June 9, 2014 (ADAMS Accession No. ML14141A073).

ATTACHMENT 6 10 CFR 50.55a Relief Request 14R-07 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

Reference:

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

Description:

Alternative Pressure Testing of the Safety Relief Valve (SRV) Automatic Depressurization System (ADS)

Accumulators Component Number: SRV ADS Accumulators and Associated Piping

2. Applicable Code Edition and Addenda

The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Requirement

Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with IWC-5220. This pressure test is to be conducted once each inspection period.

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.

LaSalle County Station operating surveillance LOS-MS-R7 11 Main Steam Safety Relief Valve Operability11 performs operability testing of the Main Steam safety relief valves including the seven relief valves and accumulators per unit that are required to provide automatic depressurization. These surveillances are performed on a refueling outage frequency as a requirement of LaSalle County Station*s lnservice Testing (IST} Program.

One specific test that these surveillances perform is a pressure decay test of the ADS accumulators, associated piping and valves. The pressure decay test is performed by isolating and pressurizing the ADS accumulators and associated piping to the nominal operating pressure (i.e., 100 pounds per square inch, gauge). The decay in pressure is then monitored through calibrated pressure measuring instrumentation. If the acceptable pressure decay criteria are exceeded, the surveillances identify appropriate troubleshooting steps to perform, including soap-bubble application to locate leakage.

The pressure decay test performed as part of LOS-MS-R7 will identify any degradation of the ADS accumulators and associated piping. The volume tested by these surveillances encompasses the entire ASME Section XI Code boundary. These surveillances are performed on a greater frequency than the required period frequency of Table IWC-2500-1 and the test pressure is consistent with the pressure requirements

ATTACHMENT 6 10 CFR 50.55a Relief Request 14R-07 Proposed Alternative In Accordance with 10 CFR 50.55a{z){1)

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

(Page 2 of 2) of Table IWC-2500-1. Thus, the testing performed during these surveillances will provide the same level of quality and safety as the pressure testing and the VT-2 visual examination requirements of Table IWC-2500-1. The additional VT-2 visual examination 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 seven accumulators per unit and associated piping in an elevated dose rate area with limited access. This results in radiation exposure (estimated 1440 mrem three times per interval for a total of 4320 mrem) without any added benefit in the level of quality and safety. This inspection would not be consistent with As Low As Reasonably Achievable (ALARA) practices.

Relief is requested from the performance of system pressure tests and the VT-2 visual examination requirements specified in Table IWC-2500-1 for the SRV ADS Accumulators and associated piping on the basis that the existing LaSalle County Station surveillances provide an acceptable level of quality and safety.

5. Proposed Alternative and Basis for Use As an alternate to the examination requirements of Table IWC-2500-1, LaSalle County Station will perform pressure decay testing on the ADS Accumulators and associated piping every refueling outage in accordance with surveillance procedure LOS-MS-R7 for Units 1 and 2.
6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units 1 and 2.
7. Precedent LaSalle County Station Third ISi Interval Relief Request 13R-1 O was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated January 30, 2008 (ADAMS Accession No. ML073610587). This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.

ATTACHMENT 7 10 CFR 50.55a Relief Request 14R-08 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

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

Reference:

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

Description:

Alternate Examination Requirements for the Hydrogen Recombiner System Piping Component Number: HG Unit Cross-Tie Piping From check valve 1HG007 to check valve 2HG016 From check valve 1HG016 to check valve 2HG007 From check valve 1HG009 to check valve 2HG006B From check valve 1HG006B to check valve 2HG009

2. Applicable Code Edition and Addenda The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda.
3. Applicable Code Requirement Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with IWC-5220. This pressure test is to be conducted once each inspection period.

IWC-521 O(b){2) requires test procedure to include methods for detection and location of through-wall leakage from the components of the system tested when the pressurizing medium is gas.

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.

Relief is requested from the system pressure test requirements of IWC-5221 and the periodicity requirements of Table IWC-2500-1, as well as the requirements of IWC-521 O{b)(2) as applied to the cross-tie piping of the Hydrogen Recombiner System, as depicted in Figure 14R-08.1 and as defined in above Component Numbers. Air is used as the pressurizing medium for the Hydrogen Recombiner System because the system contains air during normal operation. The application of a leak detection solution (e.g., soap bubble solution) to the surface of the piping would be necessary per IWC-521 O{b)(2) in order to allow for the detection and location of potential through-wall air leakage. To access the surface of the cross-tie piping, scaffolding will be required

ATTACHMENT 7 10 CFR 50.55a Relief Request 14R-08 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

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(Page 2 of 4) because there are long runs of piping located approximately 30 feet overhead. An accumulated dose of 480 mrem three times per interval for total of 1440 mrem would be required to perform a leakage test of cross tie piping. Furthermore, this total dose does not include the hours required for a significant amount of scaffolding that would have to be erected around several sensitive instrument racks and systems on both units that, if jarred, could result in a unit trip or other challenges to the operators.

Alternatively, LaSalle County Station will challenge the unit cross-tie piping to provide assurance of its structural integrity by performing pressure test at peak accident pressure and applying a soap bubble solution to all pipe welds once per inspection interval. Necessary scaffolding will be erected and leak detection solution will be applied to the surface of the unit cross-tie piping to the extent required by IWC-521 O(b)(2) if through wall leakage is detected during pressure testing of accessible components and associated piping, which is performed once every inspection period, or if through wall leakage is detected during pressure testing unit cross tie piping welds. The condition of the accessible components as determined by pressure testing of the accessible components once every inspection period in accordance with ASME Section XI rules would be indicative of that of the inaccessible components. Both the accessible and inaccessible components are designed/constructed to the same requirements and subject to similar operating conditions. Additionally, the Hydrogen Recombiners, including the unit cross-tie piping, are functionally tested every refuel outage to verify system temperature, pressure, and flow requirements to further insure system operability and structural integrity.

Based on the above discussion, reasonable assurance of the unit cross-tie piping structural integrity is achieved by the performance of the alternate pressure test of piping welds once every inspection interval.

5. Proposed Alternative and Basis for Use A pressure test will be performed on the unit cross-tie piping welds, at peak accident pressure, once each inspection interval.

Necessary scaffolding will be erected and leak detection solution will be applied to the surface of the unit cross-tie piping to the extent required by IWC-521 O(b)(2) if:

  • Through wall leakage is detected during pressure testing of accessible components and associated piping. (Remainder of system for which no relief is requested)

OR

  • Through wall leakage is detected during pressure testing of unit cross tie piping welds.
6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units 1 and 2.

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7. Precedent LaSalle County Station Third ISi Interval Relief Request 13R-11 was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated May 28, 2008 (ADAMS Accession No. ML081190470). This relief request for the LaSalle County Station, Units 1 and 2, Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.

ATTACHMENT 7 10 CFR 50.55a Relief Request 14R-08 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

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Figure 14R-08.1 HYDROGEN RECOMBINER SYSTEM CROSS-TIE PIPING T 2HG007 1HG007 I

I I

I Ui*i t2

~~

C:>l'ltain:xltn:

Stnr:hn I

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

Reference:

ASME Section XI, Table IWB-2500-1 Examination Category: B-D (Inspection Program)

Item Number: 83.90 and 83.100

Description:

Alternate Examination Requirements for the Nozzle-to-Vessel Welds and Inner Radii Sections Component Numbers: Reactor Vessel Nozzles: N1, N2, N3, NS, N6, N7, NB, N9, N16, and N18 (See Tables 1 and 2 for complete list of nozzle identifications)

2. Applicable Code Edition and Addenda

The Fourth 10-Year Interval of the LaSalle County Station, Units 1 and 2 lnservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2007 Edition with the 2008 Addenda. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2007 Edition with the 2008 Addenda is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv).

3. Applicable Code Requirements 11 The applicable requirement is contained in 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 11 delineated in Item Number 83.90, "Nozzle-to-Vessel Welds, and 83.100, "Nozzle Inside Radius Section. The required method of examination is volumetric. All nozzles with full 11 penetration welds to the reactor vessel shell (or head) and integrally cast nozzles are examined each interval.

All of the nozzle assemblies identified in Enclosures 1 and 2 are full penetration welds.

4. Reason for Request

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 Enclosures 1 and 2 for a list of RPV Examination Category B-D Nozzles for which this relief request is applicable).

The Federal Register Notice (FRN) published November 5, 2014, contains the rulemaking that amends 10 CFR 50.55a to incorporate by reference Regulatory Guide

( RG) 1.147, Revision 17, Inservice Inspection Code Case Acceptability, ASM E 11 Section XI, Division 1. As stated in the FRN, licensees may use the Code Cases listed 11

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(Page 2 of 13) in RG 1.147 as alternatives to engineering standards for the construction, inservice inspection, and inservice testing of nuclear power plant components. ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1,"(Reference 1) is listed in RG 1.147, Table 2, "Conditionally Acceptable Section XI Code Cases." In addition, the Nuclear Regulatory Commission (NRC) staff has also conditionally approved ASME Code Case N-702 for incorporation into Draft RG 1.147, Rev 18 (Reference 2). Draft RG 1.147, Rev 18, will be incorporated by reference into the next final rulemaking of 10 CFR 50.55a, which is scheduled to be issued in the Spring of 2017. The Condition associated with ASME Code Case N-702 is as follows:

The applicability of ASME Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation (SE) regarding BWRVIP-108 dated December 19, 2007 (ADAMS Accession No. ML073600374),

or Section 5.0 of NRC Safety Evaluation regarding BWRVI P-241 dated April 19, 2013 (ADAMS Accession No. 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 Code Case.

In the section of the FRN associated with NRG Responses to Public Comments on Draft Regulatory Guides, the NRC responses to comments specific to ASME Code Case N-702 start on page 9 of 40 {79FR65783}. An excerpt from the FRN is included as follows:

Licensees who plan to request relief from the ASME Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVI P-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, licensees should demonstrate the plant-specific applicability of the BWRVI P-241 report to their units in the relief request by addressing the conditions and limitations specified in Section 5.0 of the NRC SE for BWRVI P-241.

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.

5. Proposed Alternative and Basis for Use As an alternative for all welds and inner radii identified in Tables 5-1 and 5-2, EGC proposes to examine a minimum of 25 percent of the LaSalle County Station, Units 1 and 2, nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with ASME Code Case N-702. For the nozzle assemblies identified in Enclosures 1 and 2, this would mean 25 percent from each of the groups identified in Tables 5-1 and 5-2 during each 120-month interval.

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TABLE 5-1: LaSalle County Station, Unit 1 RPV Examination Category B-D Nozzle Summary Total Minimum Number Group Comments*

Number to be Examined Two (2) nozzles were Reactor Recirculation inspected in the Third ISi 2 1 Outlet (N1) Interval; No rejectable indications Three (3) nozzles were Reactor Recirculation inspected in the Third ISi 10 3 Inlet (N2) Interval; No rejectable indications Four (4) nozzles Main Steam inspected in the Third ISi 4 1 (N3) Interval; No rejectable indications Two (2) nozzles were Core Spray inspected in the Third ISi 2 1 (NS and N16) Interval; No rejectable indications One ( 1) nozzle on Top Nozzles On Top Head Head (N7) was inspected 3 1 (N7, NB and N18) in the Third ISi Interval; No rejectable indications Two (2) nozzles were Jet Pump Instrument inspected in the Third ISi 2 1 (N9) Interval; No rejectable indications Three (3) nozzles were Residual Heat Removal inspected in the Third ISi 3 1.

(N6) Interval; No rejectable indications

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

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TABLE 5-2: LaSalle County Station, Unit 2 RPV Examination Category B*D Nozzle Summary Total Minimum Number Group Comments*

Number to be Examined Two (2) nozzles were Reactor Recirculation inspected in the Third 2 1 Outlet (N1) ISi Interval; No rejectable indications Three (3) nozzles were Reactor Recirculation inspected in the Third 10 3 Inlet (N2) ISi Interval; No rejectable indications Four (4) nozzles were Main Steam inspected in the Third 4 1 (N3) ISi Interval; No reiectable indications One ( 1) Core Spray nozzle (N5) was Core Spray 2 1 inspected in the Third (N5 and N16)

ISi Interval; No rejectable indications One ( 1) nozzle on Top Nozzles On Top Head Head (N7) was inspected 3 1 (N7, NB, and N18) in the Third ISi Interval; No rejectable indications One ( 1) nozzle was Jet Pump Instrument inspected in the Third 2 1 (N9) ISi Interval; No rejectable indications Three (3) nozzles were Residual Heat Removal inspected in the Third 3 1 (N6) ISi Interval; No rejectable indications

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

The examinations in Tables 5-1 and 5-2 will be scheduled in accordance with ASME Section XI, IWB-2411, Inspection Program.

ASME Code Case N-702 stipulates that a VT-1 visual examination may be used in lieu 11 of the volumetric examination for the inner radii (i.e., Item Number 83.100, Nozzle Inside Radius Section ).

  • EGC may utilize ASME Code Case N-648-1 with associated 11 RG 1.147 conditions for the nozzles selected for examination. Volumetric examinations of the inside radius section of those reactor vessel nozzles selected for examination will be completed if ASME Code Case N-648-1 is not applied.

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Electric Power Research Institute (EPRI) Technical Report (TR} 1003557, "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, 11 provides the technical basis for ASME Code Case N-702.

BWRVI P-108 determined that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure (LTOP) event are 6

very low (i.e., <1 x 10- for 40 years) with or without lnservice Inspection. The report concluded that inspection of 25 percent of each nozzle type is technically justified. The BWRVIP-108 report was approved by the NRC in a SER dated December 19, 2007 (ADAMS Accession No. ML073600374) and requires additional criteria to be met in order to apply the technical basis of BWRVI P-108 for the reduction of inspection coverage of the RPV nozzles and nozzle-to-vessel shell welds.

11 BWRVIP-108 was supplemented by EPRI TR 1021005, BWRVIP-241: Boiling Water Reactor Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and was approved by the NRC in a SE dated April 19, 2013 (ADAMS Accession No. ML13071A240}. This report revised the acceptance criteria associated with the NRC additional criteria.

11 As stated in the BWRVIP-241 NRC SE, Section 5.0, Conditions and Limitations," 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 ASME Code Case N-702 as an alternative.

However, each licensee should demonstrate the plant-specific applicability of the BWRVI P-241 report to its units in the relief request by demonstrating that the following general and nozzle-specific criteria are satisfied:

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

LaSalle County Station, Units 1 and 2, Technical Specifications (TS}

3.4.11, "Reactor Coolant System (RCS) Pressure and Temperature (PIT)

Limits, 11 provides a Surveillance Requirement limiting heatup and cooldown rates to s 100 °F in any one-hour period. This heatup/cooldown rate is also described in the LaSalle County Station Updated Final Safety Analysis Report (UFSAR), Section 5.2.3.3.1.7, 11 "Operating Limits During Heatup, Cooldown and Core Operations.

For the recirculation inlet nozzles (N2}, the following criteria must be met:

(2) (pr/t)/CRPV S 1.15 p = RPV normal operating pressure (psi),

r = RPV inner radius (inch),

t = RPV wall thickness (inch), and CRPV = 19332;

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The calculation for the LaSalle County Station, Units 1 and 2, N2 Nozzle results in a maximum value of 1.064, which satisfies this criteria.

2 2 2 2 (3) [p(ro + ri )/(ro - ri )]/CNoZZLE S 1.47 p = RPV normal operating pressure (psi),

ro = nozzle outer radius (inch),

ri =nozzle inner radius (inch), and CNOZZLE = 1637; The calculation for the LaSalle County Station, Units 1 and 2, N2 Nozzle results in a maximum value of 1 .134, which satisfies this criteria.

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

(4) (pr/t)/CRPV S 1.15 p = RPV normal operating pressure (psi),

r = RPV inner radius (inch),

t = RPV wall thickness (inch), and CRPV = 16171; The calculation for the LaSalle County Station, Units 1 and 2, N1 Nozzle results in a value of 1.025 for Unit 1 and a value of 1.272 for Unit 2. The Unit 1 results satisfy the criteria; however, the Unit 2 results are greater than 1.15.

p = RPV normal operating pressure (psi),

ro =nozzle outer radius (inch),

ri =nozzle inner radius (inch), and CNOZZLE = 1977.

The calculation for the LaSalle County Station, Units 1 and 2, N1 Nozzle results in a maximum value of 1.114, which satisfies the criteria.

Based upon the above information, all LaSalle County Station RPV nozzle-to-vessel shell or head full penetration welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles on Unit 2, meet the general and nozzle-specific criteria in BWRVIP-241. BWRVIP-241 Section 6.0 notes that for plants having recirculation outlet nozzles with Condition 4 greater than 1.15, such as for LaSalle County Station, Unit 2, a plant-specific analysis following the approach described in this report may be able to justify values greater than 1.15.

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Because the Unit 2 N1 nozzles did not meet the BWRVIP-241 criteria, a bounding analysis was performed to qualify all the Unit 1 and Unit 2 nozzles. The Probability of Failure (PoF) was calculated based on operation for 60 years and assumes no inspections were performed in the initial 40 years of operation.

To address the elevated fluence issue of certain nozzles in the belt-line region of the reactor vessel, the fluence associated with the Unit 2 N6 nozzle (Low Pressure Coolant Injection nozzle) at the end of 60 years of operation was used as input. This beltline nozzle has the highest fluence of all the Unit 1 or Unit 2 nozzles, and although the current licenses for LaSalle County Station, Units 1 and 2, expire after 40 years, the use of the 60-year fluence provides additional margin in the analysis.

The VIPERNOZ computer program, as used in BWRVIP-108 and BWRVIP-241, was used in the LaSalle County Station analysis. The same assumptions used in BWRVI P-108 and BWRVI P-241 were used in the LaSalle County Station analysis, such as the assumed number of stress corrosion initiation and fabrication flaws, the flaw size distribution, etc.

The bounding load cases analyzed included the following:

1. Unit pressure
2. Turbine Generator Trip-SCRAM
3. Loss of Feedwater Pumps/Isolation Valves Close The number of thermal cycles used in the analysis was based on the LaSalle County Station reactor pressure vessel thermal cycle diagrams.

The results of the analysis are shown in the following table.

PoF per year from LTOP PoF per year from Normal events for 25% lnservice Operating Condition for Maximum Inspection for period of 25% lnservice Inspection PoF Extended Operation (Zero for period of Extended per year**

Inspection for initial 40 Operation (Zero Inspection years)* for initial 40 years)

Nozzle Blend 1.4E-9 4.2E-7 5.0E-6 Radii Nozzle-to-shell <<2.0E-10 3.3E-9 5.0E-6 weld

  • Values include 1E-3 probability of LTOP occurrence.
    • Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61 ), NUREG-1806, Volume 1, August 2007.

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Since the bounding nozzle, the N1 on Unit 2, has been shown to meet the NRC safety goal of 5E-6 per year, and all other nozzles meet the plant specific applicability criteria from the BWRVIP-241 report and are bounded by the Unit 2 N1 nozzle analysis, the application of ASME Code Case N-702 to all the Unit 1 and Unit 2 nozzles listed in Enclosures 1 and 2 is acceptable.

Therefore, use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 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, as well as the remaining term of the renewed facility operating licenses for LaSalle County Station, Units 1 and 2.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval, as well as the remaining term of the LaSalle County Station, Units 1 and 2, renewed facility operating licenses (NPF-11 for Unit 1 and NPF-18 for Unit 2), which currently expire at midnight on April 17, 2042, and at midnight on December 16, 2043, for LaSalle County Station, Units 1 and 2, respectively. The Fourth ISi Interval, as well as the remaining term of the renewed facility operating licenses refers to the LaSalle County Station, Units 1 and 2 current fourth and upcoming fifth and sixth 120-month ISi Program intervals.
7. Precedents
  • LaSalle County Station Third ISi Interval Relief Request 13R-14 was authorized per NRC SE dated October 30, 2015 (ADAMS Accession No. ML15226A412).

This relief request for the LaSalle County Station, Units 1 and 2 Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.

  • Letter from R. Guzman (U.S. Nuclear Regulatory Commission) to Site Vice President (Entergy), "James A Fitzpatrick Nuclear Power Plant - Relief from the Requirements of the ASME Code Case N-702 and BWRVIP-241 for Plant Nozzle-To-Vessel Welds and Nozzle Inner Radii (CAC No. MF8301 ), 11 dated December 6, 2016 (ADAMS Accession No. ML16334A440).
  • Letter from S. S. Koenick (U.S. Nuclear Regulatory Commission) to 8. C. Hanson (EGC), "Safety Evaluation of Relief Requests 14R-02 and 14R-10 for the Fourth 10-Year Interval of the lnservice Inspection Program for Limerick Generating Station, Units 1 and 2 (CAC Nos. MF7587 and MF7588)," dated November 21, 2016 (ADAMS Accession No. ML16301A401).
  • Letter from R. J. Pascarelli (U.S. Nuclear Regulatory Commission) to M. E.

Reddemann (Energy Northwest), "Columbia Generating Station - Relief Request for Alternative 41Sl-04 Applicable to the Fourth 10-Year lnservice Inspection 11 Program Interval (CAC No. MF7331 ), dated October 5 2016 (ADAMS Accession No. ML16263A233).

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8. References
1. ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor 11 (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1.

Draft Regulatory Guide DG-1296, lnservice Inspection Code Case Acceptability, 11 2.

ASME Section XI, Division 1, dated March 2016 (ADAMS Accession No.

11 ML15027A202).

3. Letter from Jeffrey S. Mitchell (NRC) to M. P. Gallagher (EGC), "Issuance of Renewed Facility Operating Licenses for LaSalle County Station, Units 1 and 2 (TAC Nos. MF5347 and MF5346), dated October 19, 2016 (ADAMS Accession 11 No. ML16202A075).

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ENCLOSURE 1: Applicable LaSalle Countu Station, Unit 1 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N1A Nozzle 8-D 83.90 Recirc Outlet 24 11 N1A IR* 8-D 83.100 Recirc Outlet 24 11 N18 Nozzle 8-D 83.90 Recirc Outlet 24 11 N181R 8-D 83.100 Recirc Outlet 24 11 N2A Nozzle 8-D 83.90 Recirc Inlet 12 11 N2AIR 8-D 83.100 Recirc Inlet 12 11 N28 Nozzle 8-D 83.90 Recirc Inlet 12 11 N281R 8-D 83.100 Recirc Inlet 12 11 N2C Nozzle 8-D 83.90 Recirc Inlet 12 11 N2CIR 8-D 83.100 Recirc Inlet 12 11 N2D Nozzle 8-D 83.90 Recirc Inlet 12 11 N2DIR 8-D 83.100 Recirc Inlet 12 11 N2E Nozzle 8-D 83.90 Recirc Inlet 12 11 N2EIR 8-D 83.100 Recirc Inlet 12 11 N2F Nozzle 8-D 83.90 Recirc Inlet 12 11 N2FIR 8-D 83.100 Recirc Inlet 1211 N2G Nozzle 8-D 83.90 Recirc Inlet 1211 N2GIR 8-D 83.100 Recirc Inlet 12 11 N2H Nozzle 8-D 83.90 Recirc Inlet 12" N2HIR 8-D 83.100 Recirc Inlet 12 11 N2J Nozzle 8-D 83.90 Recirc Inlet 12 11 N2JIR 8-D 83.100 Recirc Inlet 12" 11 N2K Nozzle 8-D 83.90 Recirc Inlet 12 N2KIR 8-D 83.100 Recirc Inlet 12 11 N3A Nozzle 8-D 83.90 Main Steam 26 11 N3AIR 8-D 83.100 Main Steam 26" N38 Nozzle 8-D 83.90 Main Steam 26 11 N381R 8-D 83.100 Main Steam 26" N3C Nozzle 8-D 83.90 Main Steam 26" N3CIR 8-D 83.100 Main Steam 26 11 N3D Nozzle 8-D 83.90 Main Steam 26 11 N3DIR 8-D 83.100 Main Steam 26 11 N5 Nozzle 8-D 83.90 Core Spray 12" 11 N51R 8-D 83.100 Core Spray 12 N6A Nozzle 8-D 83.90 LPCI** 12 11 N6AIR 8-D 83.100 LPCI 12" N68 Nozzle 8-D 83.90 LPCI 12" N681R 8-D 83.100 LPCI 12 11

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

(Page 11 of 13)

ENCLOSURE 1: Applicable LaSalle CountH Station, Unit 1 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N6C Nozzle 8-D 83.90 LPCI 1211 N6CIR 8-D 83.100 LPCI 12 11 Reactor Core N7 Nozzle 8-D 83.90 6" Isolation Coolinq Reactor Core N71R 8-D 83.100 6" Isolation Coolinq 8-D 83.90 Head Vent 411 NB Nozzle 8-D 83.100 Head Vent 411 NB IR Jet Pump N9A Nozzle 8-D 83.90 6" Instrumentation Jet Pump N9AIR 8-D 83.100 6" Instrumentation Jet Pump N98 Nozzle 8-D 83.90 6" Instrumentation Jet Pump N981R 8-D 83.100 6" Instrumentation N16 Nozzle 8-D 83.90 Core Spray 1211 N161R 8-D 83.100 Core Spray 12 11 N18 Nozzle 8-D 83.90 Spare 6" N181R 8-D 83.100 Spare 6"

  • IR - Inner Radius

ATTACHMENT 8 10 CFR 50.55a Relief Request 14R-09 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

(Page 12 of 13)

ENCLOSURE 2: Applicable LaSalle Countv Station, Unit 2 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N1A Nozzle 8-D 83.90 Recirc Outlet 24 11 N1A IR* 8-D 83.100 Recirc Outlet 24 11 N18 Nozzle 8-D 83.90 Recirc Outlet 24 11 N181R 8-D 83.100 Recirc Outlet 24 11 N2A Nozzle 8-D 83.90 Recirc Inlet 12 11 N2AIR 8-D 83.100 Recirc Inlet 12 11 N28 Nozzle 8-D 83.90 Recirc Inlet 12 11 N281R 8-D 83.100 Recirc Inlet 12 11 N2C Nozzle 8-D 83.90 Recirc Inlet 12 11 N2CIR 8-D 83.100 Recirc Inlet 1211 N2D Nozzle 8-D 83.90 Recirc Inlet 12 11 N2DIR 8-D 83.100 Recirc Inlet 1211 N2E Nozzle 8-D 83.90 Recirc Inlet 12 11 N2EIR 8-D 83.100 Recirc Inlet 1211 N2F Nozzle 8-D 83.90 Recirc Inlet 12 11 N2FIR 8-D 83.100 Recirc Inlet 12 11 N2G Nozzle 8-D 83.90 Recirc Inlet 12 11 N2GIR 8-D 83.100 Recirc Inlet 12 11 N2H Nozzle 8-D 83.90 Recirc Inlet 12 11 N2HIR 8-D 83.100 Recirc Inlet 12 11 N2J Nozzle 8-D 83.90 Recirc Inlet 12 11 N2JIR 8-D 83.100 Recirc Inlet 12 11 N2K Nozzle 8-D 83.90 Recirc Inlet 12 11 N2KIR 8-D 83.100 Recirc Inlet 12 11 N3A Nozzle 8-D 83.90 Main Steam 26 11 N3AIR 8-D 83.100 Main Steam 26 11 N38 Nozzle 8-D 83.90 Main Steam 26 11 N381R 8-D 83.100 Main Steam 26 11 N3C Nozzle 8-D 83.90 Main Steam 26 11 N3CIR 8-D 83.100 Main Steam 26 11 N3D Nozzle 8-D 83.90 Main Steam 26 11 N3DIR 8-D 83.100 Main Steam 26 11 N5 Nozzle 8-D 83.90 Core Spray 12 11 N51R 8-D 83.100 Core Spray 12 11 N6A.Nozzle 8-D 83.90 LPCI** 12 11 N6AIR 8-D 83.100 LPCI 12 11 N68 Nozzle 8-D 83.90 LPCI 12 11 N681R 8-D 83.100 LPCI 12 11

ATTACHMEN T 8 10 CFR 50.SSa Relief Request 14R-09 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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

( Page 13 of 13)

ENCLOSURE 2: Applicable LaSalle Countv Station, Unit 2 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N6C Nozzle 8-D 83.90 LPCI 12 11 N6CIR 8-D 83.100 LPCI 12 11 Reactor Core N7 Nozzle 8-D 83.90 6" Isolation Coolinq Reactor Core N71R 8-D 83.100 6" Isolation Coolinq 8-D 83.90 Head Vent 411 NS Nozzle N81R 8-D 83.100 Head Vent 4" Jet Pump N9A Nozzle 8-D 83.90 6" Instrumentation Jet Pump N9AIR 8-D 83.100 6" Instrumentation Jet Pump N98 Nozzle 8-D 83.90 6" Instrumentation Jet Pump N981R 8-D 83.100 6" Instrumentation N16 Nozzle 8-D 83.90 Core Spray 12 11 N161R 8-D 83.100 Core Spray 12 11 N18 Nozzle 8-D 83.90 Spare 6" N181R 8-D 83.100 Spare 6"

  • IR - Inner Radius
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