RA17-086, County Station, Units I & 2 - 10CFR50.55a Relief Request 14R-01

County Station, Units I & 2 - 10CFR50.55a Relief Request 14R-01

ML17314A009
Person / Time
Site:	LaSalle
Issue date:	11/10/2017
From:	Exelon Generation Co, True North Consulting
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17317A006	List: ML17314A008 RA17-086, County Station, Units I & 2 - 10CFR50.55a Relief Request 14R-01
References
RA17-086
Download: ML17314A009 (139)

Text

/SI Program Plan LaSalle County Station Units I & 2, Fourth fllterval 10CFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page I of 29)

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

Reference:

Table IWB-2500-1, Table IWC-2500-1 Examination Category: B-F, B-J, C-F-1, and C-F-2 Item Number: 85.10, 89.11, 89.21, 89.31, 89.32, 89.40, CS.I I, C5 .21, CS.SI, and CS.81

Description:

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

2. Applicable Code Edition and Addenda The Fourth I0-Y car Interval of the LaSalle County Station, Units I and 2 lnscrvicc Inspection (ISi) Program is hascd on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Reuuirement Tahle IWB-2SOO- I, Examination Category B-F. requires volumetric and surface examinations on all welds for Item Number BS. I0.

Tahle IWB-2SOO- I, Examination Category B-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:

(I) primary plus secondary stre)\s intensity rnnge of 2.4S,,, for fcrritic steel "ml uustenitic steel.

(2) cumuh1tive usuge foctor U of 0.4.

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/SI Program Pfau LaSalle Co1111ty Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 29)

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

(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 docs not include welds exempted by IWB-1220. These additional welds may be located as follows (I) For BWR plants (a) One reactor coolant recirculation loop (where a loop or run branches, only one branch)

(h) One branch run representative of an essentially symmetric piping configuration among each group of branch runs that arc 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 hrnnches, only one hranch)

(e) Each piping and hranch exclusive of the categories of loops and runs that ure part of the system piping of (a) through (d) ahove Tahle IWC-2500-1. Examination Categories C-F- 1 and C-F-2 re<1uire volumetric and surface examinations on a sample of welds for Item Numhers C5. I l, C5.2 I. and C5.5 I.

and surface examinations on a sample of welds for Item Numher C5.K I. The weld population selected for inspection is spccilied in Note (2) for hoth Exmnination Cutegories.

Note (2) The welds selected for cxumination shall include 7.5'~* . hut not less than 28 welds, of nil dissimilar metal. austenitic stainless steel or high alloy weld)\

<Examination Category C-F- 1) or of all curhon und low alloy steel welds (Examimllion Cutegory C-F-2) not exempted hy IWC-1220. (Some weld)\ not exempted hy IWC-1220 arc not re<tuired to he nondestrm:tivcly examined per

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/SJ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.5Sa(z)(l)

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

Revision 0 (Page 3 of 29)

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; (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 Rcuucst In accordance with IOCFR50.55a(z)( I). relief is requested on the basis that the proposed alternative utilizing Electric Power Reseurch Institute (EPRI) Topical Report (TR) 112657, "Revised Risk-Informed lnservice Inspection Evaluation Procedure,"

Revision B-A (Reference I) along with two enhancements from ASME Code Case N-578-1. "Risk-Informed Requirements for Class I. 2. or 3 Piping. Method B.Section XI. Division I," (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):

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The initial LuSalle County Station Risk -Informed lnservicc Inspection (RI-ISi) Prognun was submitted during the second period of the Second ISi Interval for Units I amt 2.

This initial RI-ISi Pmgrnm was developed in ucconlancc with EPl~I TR-112657.

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/SI Program Plan LaSalle Co1111ty Station Units I & 2, Fourth /11terval 10CFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 4 of 29)

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 l 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 200 I 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 he 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 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, hut rather, the previously approved RI-ISi methodology and evaluation will he 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 clement selection for the Fourth ISi Interval RI-IS I Progrnm. This same process lrns hccn nrnintained in each revision to the LaSalle County Station RI-ISi assessment that has heen performed to date.

Based on the hn1rth 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 I .OOE-06 and I .OOE-07 acceptance criteria for delta-core damage frequency (Dclta-CDF) und delta-large early release frequency (Dclta-LERF) as described in Regulatory Guide 1.174, "An Approach for Using Prohahilistic Risk Assessment in Ri!>k -

lnformed Decisions on Plant-Specilk Changes to the Licensing Basis ." The Delta-CDF and Delta-LERF values for LaSalle County Station, Unit!> I and 2 arc listed in the following tahle.

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/SI Program Plan LaSalle Co1111ty Station U11its I & 2, Fourth lllterva/

10CFRSO.SSa Relief Request 14R-01 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Page 5 of 29)

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 I 5.26E-09 4.34E-09 Unit 2 6.29E-09 4.94E-09 The following tables document the Delta-CDF and Delta-LERF for LaSalle County Station Units I and 2 over the initial ASME Section XI Program for the Fourth ISi Interval. The first two tables provide results for Unit I. The results for Unit I arc 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.

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IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 6 of 29)

LaSalle County Station Unit I Delta-CDF and Delta-LERF by System ACDF ALE RF Events/Reactor-Year Events/Reactor-Year System Acceptance Accepta-nce RI-ISi RI-ISi Criteria Criteria CRD 4.16E-l I l.OOE-07 I. I7E-l 2 I .OOE-08 ECCS -l.70E-I 0 l.OOE-07 -6.09E-l I l.OOE-08 FW I. I3E-09 l.OOE-07 l.OIE-09 l.OOE-08 HPCS l.19E-IO l.OOE-07 7.65E-1 I 1.00E-08 MS l.08E-09 l.OOE-07 l.07E-09 l.OOE-08 RCIC 5.56E-IO l.OOE-07 5.22E-IO 1.00E-08 RCS l.04E-09 l.OOE-07 2.65E-IO 1.00E-08 RWCU I .46E-09 l.OOE-07 1.46E-09 1.00E-08 Total 5.26E-09 <l.OOE-06 4.34E-09 <l.OOE-07 LaSalle County Station Unit I BER Wc Id Dc Ita- CDF*an d Dc Ita- LERF. h1y S,ystcm ACDI" ALERI" Events/Reactor-Year Events/Reactor-Year System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria ECCS -2.80E-IO l .OOE-07 -2.80E- IO l.OOE-08 FW 9.54E-IO l .OOE-07 9.55E-10 l.OOE-08 llPCS 1.56E-l 2 I .<X>E-07 l .56E-14 I .<X>E-08 MS 1.0IE-09 I .<X>E-07 1.01 E-09 1.<X>E-08 RCIC' 5.48E-IO I .<X>E-07 5.22E-IO I .<X>E-08 RWC'U 1.46E-09 I .OOl~-07 I .46E-09 1.<X>E-08 Totul J.70E-09 <I .OOE-06 J.66E-09 <I .OOE-07 l 'riu* i\'ortll <'1111.mlt/111:. /./.(' I S H7./.fllflfl.fl4

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 7 of 29)

LaSalle County Station Unit 2 De Ita- CDF an d De Ita- LERF b'Y S ystem ACDF ALE RF System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria CRD 7.07E-1 I l.OOE-07 J.46E-12 l.OOE-08 ECCS 3.40E-10 I .OOE-07 -l.42E-IO l.OOE-08 FW 2.86E-09 l.OOE-07 2.72E-09 l.OOE-08 HPCS -l.36E-12 l.OOE-07 2.41 E-11 l.OOE-08 MS 4.30E-IO l.OOE-07 4.38E-10 I .OOE-08 RCIC 4.25E-IO l.OOE-07 4.03E-10 l.OOE-08 RCS 8.66E-IO l.OOE-07 1.81 E-IO l.OOE-08 RWCU I .32E-09 l.OOE-07 1.31 E-09 l.OOE-08 Total 6.31 E-09 <l.OOE-06 4.94E-09 <l.OOE-07 LaSalle County Station Unit 2 BER Wc Id Dc Ita- CDF an d Dc Ila- LERF h y S vstcm ACDI<' ALE RF Evcnts/Rcactor-Y car Events/Reactor-Year System Acceptance Acceptance RI-ISi RI-ISi Criteria Criteria ECCS -3. I 8E-IO l.OOE-07 -3.18E-IO l.OOE-08 FW 2.65E-09 I .OOE-07 2.65E-09 l.OOE-08 llPCS 2.09E-12 1.00E-07 1.04E-14 l.OOE-08 MS 3.42E-IO I .OOE-07 3.36E-IO l.OOE-08 RCIC' 4.25E- IO I .OOE-07 4.03E-IO l.OOE-08 RWCU I .32E-09 1.00E-07 1.31 E-09 I .OOE-08 Total 4.42E-09 <l.OOE-06 4.39E-09 <1.00E-07 l'rur ,\'11rth ('1111.rnlti11Jl, /./.('

ISi Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 8 of 29)

The actual "evaluation and ranking" procedure including the Consequence Evaluation and Degradation Mechanism Assessment processes of the currently approved (Reference

6) RI-ISi Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-112657, Revision B-A.

These portions of the RI-ISi Program have been and will continue to be reevaluated as major revisions of the site Probabilistic Risk Assessment (PRA) occur and modifications to plant configuration are made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, Element Selection, and Risk Impact Assessment steps encompass the complete living program process applied under the LaSalle County Station RI-ISi Program.

5. Proposed Alternative and Basis for Use The proposed alternative initially implemented in the LaSalle County Station, Units I and 2, "Risk-Informed lnservice Inspection Evaluation," (Reference 3), along with the two enhancements noted helow, provide an acceptable level of quality and safety as required hy IOCFR50.55a(z)( I). This initial program along with these enhancements, was resuhmitted und is currently approved for LaSalle County Station's Third ISi Interval us documented in Reference 6.

The Fourth ISi Interval RI-ISi Program will he a continuation of the current application and will continue to he a living program as dcscrihcd in Scc1ion 4of1his relief request.

No changes to the evaluation methodology as currently implemented under EPRI TR-112657, Revision B-A, arc required as part of this interval update. The following two cnhanccmcnls will continue to he implcmcn1cd.

a. In lieu of the evalualion and sample expansion requirements in Section 3.6.6.2, "RI-ISi Selecled Examinalions" of EPRI TR-112657, LaSalle Counly Stalion will u1ilize the requiremcnls of Paragraph -2430, "Additional Examinations" conlaincd in ASME Code Case N-578-1(Rel'crence4). The alternalive criteria for addilional exami11&11ions conlained in ASME Code Case N-578-1 provides a more relined 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, hased on service condilions. degradation mechanisms, and the performance of evalualion" 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. Addi1ionally. simihir to the current rc<1uircmcnts of ASME S"*ction XI. LaSalle Coun1y Slation intends 10 perform additional examinations that are rel1uircd due to the identification of llaws or 1'r11r Nnrtll ('1111.mltillJl, /./.('

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-01 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Page 9 of 29) 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,"

LaSalle County Station will utilize the provisions listed in Table I, Examination Category R-A, "Risk-Informed Piping Examinations," contained in ASME Code Case N-578-1 (Reference 4 ). To implement Note I0 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 I of ASME Code Case N-578-l will be used as it provides a detailed breakdown for "Examination Method" and "Categorization of Parts to be Examined." Based on these methods and categorization, the examination figures specified in EPRI TR-112657, Section 4 will then be used to determine the examination volume/area based on the degradation mechanism and component configuration . For piping structural clements not subject to a degrudation mechanism, ASME Code Case N-578-1, Table I, Note I will he applied using the expanded examination volume.

The LaSalle County Station RI-ISi Program, as developed in accon.lancc with EPRI TR-112657, Revision B-A (Reference I), requires that 25"/ri of the piping structural clements that arc categorized as "High" risk (i.e., Risk Category I, 2, and 3) and 10% of the piping structural clements that arc categorized as "Medium" risk (i.e., Risk Categories 4 and 5) he selected for inspection. For this application, the guidance for lhc examinalion volume for a given degrndalion mechanism is provided hy lhc EPRI TR-112657 while the guidance for the cxaminalion mc1hod and ca1cgoriza1ion of parts lo he examined arc provided hy lhc EPRI TR -112657 as supplcmelllcd hy ASME Code Case N-578-1.

For NRC' slaff considernlion in lhe evalualion of lhis ahcrnalivc RI-ISi Progrnm, Enclosure LS-LAR -007, Revision 0 lo this relief rcquesl contains a summary of the Regulatory Guide 1.200, Revision 2 (Reference 7) cvalualion performed on LS-PSA-014, Revision 9 (Rcl'crcnce 8), and lhc impacl of lhc identified gaps on the lechnical adequacy of lhc LaSalle C'ounly Slalion PRA Model to supporl lhis RI-ISi applicalion (sec Enclosure, Table I ).

In addilion to thi-; risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, rcgardlc~s of risk clas~ilicalion, will l*onlinue lo receive \ode-required system prcssme tesling as parl of the CUITl.'llt ASME Section XI Program. VT-2 vi~ual examinations m*e scheduled in al.'cordance wilh the LaSalle County Station System Pressure Testing Program, which remains unaffected hy lhe RI

• IS I Pmgrnm.

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page I0 of 29)

6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units I and 2.

7. Precedents

• LaSalle County Station, Units I and 2, Third ISi Interval Relief Request 13R-O I was authorized per NRC SE dated April 29, 2008 (ADAMS Accession No. ML0809402 I5). This relief request for the LaSalle County Station, Units I and 2, Fourth ISi Interval, utilizes a similar RI-ISi methodology to the previously approved relief request.

• Relief Request 14R-O 1 was authorized for Byron Station Units I and 2 by NRC SE dated December 20, 2016 (ADAMS Accession No. MLl6327A396).

• Relief Request 14R-O I was authorized for Limerick Station Units 1 and 2 hy NRC SE dated December 29, 2016 (ADAMS Accession No. ML16344A324).

• Relief Request RR-7 was authorized for St. Lucic Plant, Unit 2 hy NRC SE dated August 10, 2015 (ADAMS Accession No. MLl5196A623).

• Relief Request 4RR-O I was authorized for Susquehanna Steam Electric Station, Units I and 2 by NRC SE dated April 28, 2015 (ADAMS Accession No. MLI 5098A478).

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

2. Letter from W. 11. Batcnwn (NRC) lo G. L. Vine (EPRI). "Safety Evaluation Report Related to EPRI Risk -Informed lnservice Inspection Evaluation Procedure (EPRI TR-112657. Revision B. July 1999)," dated October 28, 1999 (ADAMS Accession Nos. Ml..993190460 and Ml.993190474 ).

3. lniti<1l l~b.k-lnl'ormed 111'ervicc Inspection Evalm11ion - LaSalle County Station, Units I and 2. dated Pebruary 2001. (Letter from C. G. Pardee (Commonwe<1lth Edison Comp*my) to the NRC. "LuSalle County St<1tion Interval 2 lnservice Inspection Pro~rnm: Relief Request CR -35. Alternative to the ASME Boiler und 1'r1w N11rtl1 l'm1.rnlti111t. I.I.(' l ..'iH74.llWfi*fl./

IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-01 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Page 11 of 29)

Pressure Vessel Code Section XI Requirements for Class I and Class 2 Piping Welds, Risk-Informed Inservice Inspection Program," dated May 18, 200 I)

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

5. Letter from A. J. Mendiola (NRC) to 0. D. Kingsley (EGC), "LaSalle County Station, Units I and 2 - Relief Request CR-35 (TAC Nos. MB 1982 and MBl983)," dated December 27, 2001 (ADAMS Accession No. MLOl3610078).

6. Letter from R. Gibbs (NRC) to C. G. Pardee (EGC), "LaSalle County Station, Units I and 2 - Relief Request 13R-O I, Associated with the Third 10-Ycar Interval for LaSalle County Station, Units I and 2 (TAC Nos. MD5457 and MD5458),"

dated April 29, 2008 (ADAMS Accession No. ML0809402 I 5).

7. NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009 (ADAMS Accession No. ML090410014).

8. LS-PSA-014, Revision 9, "LaSalle Prohabilistic Risk Assessment (PRA)

Quantification Notehook," dated November 2015.

9. Enclosure LS-LAR-007, Revision 0, "LaSalle Station, Units I and 2. PRA Capahility Assessment for RI-ISi. (Sununary: LS-LAR-007 PRA Capahility Assessment for Risk -Informed lnservice Inspection Applications)," dated January 2017.

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ISi Program Plan LaSalle County Station Units 1 & 2, Fourth lllterval 10CFRSO.SSa Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 12 of 29)

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 PRA models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the LaSalle PRA.

PRA Maintenance and Update The EGC risk management process ensures that the applicable PRA 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 PRA Moc.lei Update" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGC nuclear generation sites. The overall EGC Risk Management progrnm, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g .. due to changes in the plant, errors or limitations iden1ilicd in the model, industry operating experience), and for controlling the model and associated computer tiles. To ensure that the current PRA model remains an accurate reflection of the as-built. as-operated ph111ts, the following activities arc routinely performed:

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

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

• Maintenance unavaih1bilities arc captured, and their impact on core dmrn1ge frc4ucncy (CDF) is trended.

• Plant specific initiating event frequencies, failure rates. and nmintenancc unavailabilities for equipment that can have u !-.ignilkant impact on the PRA model arc updated approximately every four years.

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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-Linc 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 Ruic (IOCFR50.65(a)(4)).

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

The most recent update of the LaSalle PRA model (designated the LS2014A model) [61 was completed in November 2015 as a result of a regularly scheduled update to the previous LS2011 A model 181. The LS20 I4A model is the most recent evaluation of the risk prolile at LaSalle for internal event challenges, including internal llooding. 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 hased on the event tree I fault tree methodology, which is a well-known methodology in the industry.

PRA Peer Review Several a~se!'lsments of technical capability have heen made, and continue to he planned for the LaSalle PR/\ model. I\ chronological list of the as!'lessments performed includes the following:

• An independent PRA peer review was conducted under the au!'lpiccs of the BWR Owners' Group in July 2000, following the Industry PR/\ Peer Review process 141. This peer review included ml assessment of the PRA model maintenance and update pnll*ess. All lindings from thi~ peer review were

<1ddressed and dosed out.

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• A self-assessment analysis was performed using Addendum B of the AS ME/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.

• 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 1.200 [ 12].

• Of the 331 internal events supporting requirements (SR) reviewed o 293 were considered Met

• 286 at Capahility Category I/II or greater

• 7 at Capahility Category I o 18 were considered N/ A o 20 were considered Not Met According to EPRI TR-1021467-A (5J, thc 7 SRs that meet Capahility Category I arc sufticicnt for RISI: and 12 of 20 of the SRs that were not met arc not required for the RISI application.

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

• Swxess Criteria o Perform checks to determine the reasonableness and accepiability of thermal hydrnulic, structurnl or other supporting engineering basis used lo support the success criteria.

• I luman Rcliahility Analysis o For el1uipment modeled in the PRA. identify tc'l and m.aintcnance

.ictivitics tlml rel1uirc rc.ilignmcnt of equipment outside of its normal operational or standby status. through a review of procedures and 1'ruC' N11rtll C'lm.rnlti1111, I.I.<'

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Revision 0 (Page 15 of 29) 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.

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

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

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

• Large Early Release Frequency (LERF) o Document the quantitative definition used for significant accident.

The Peer Review findings, including hoth those items not-met as well as those that meet Capahility 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-Mel and Capability C.11cgory I SRs relative lo the RI-IS I relief request is provided in Table I. 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 unaddrcsscd gaps me tracked by the PRA open issues list (i.e .* Update RclJllircmcnts Evaluation (URE> Database) and arc judged to have low impact on the PRA model or its ahility to support u full range of PRA applications. These items arc tracked and their potential impacts arc accounted for in applications where appropriate. In addition, plant changes mmlc since the last PRA update have hccn reviewed *md determined to not have a significant PRA impact. These items arc also documented in UR Es for consideration in future PRA updates. us upproprh11c.

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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 lo 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 I 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 arc performed, remaining gaps to specific requirements in the PRA standard will he reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

Conclusion Regarding PRA Capability for Risk-Informed ISi The LaSalle PRA model continues to he 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 tlmt lmvc been performed and arc being planned. and

• Rl*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 analy!'IC!'I for the LaSalle IO-year interval evaluation using the LS2014A model, the remaining gaps to the PRA standanl have been reviewed to determine which, if any.

would merit Rl

• ISl -specilk scn~itivity studies in the prc~cnt11tion of the application results. The result of this 11ssessment concluded that no 11dditional sensitivity studies arc merited.

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References I. 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," Regulatory Guide l.174, Revision I, 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: Prohahilistic Risk Assessment Technical Adequacy Guidance for Risk-informed lnservice Inspection Programs," TR-1021467-A. June 2012.

6. LS-PSA-014. Revision 9, "LaSalle Prohahilistic Risk Assessment (PRA) Quantification Notehook," Novemher 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.

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 lluman Reliahility Notehook, January 2013.

11. ASME/ANS, "Standard for Level I/Large Early Release Frequency Prohahili:-.tic Risk Assessment for Nucle;ir Power Plant Applications," ASME/ANS RA*Sc-2007, August 2007.

12. "An Approach for Determining the Technical Adequacy of Pmhahili:-.tic Risk As:-.es:-.mcnt results for Ri:-.k-lnformed Activitic:-.," Regulatory Guide 1.200, U.S. Nudcm Regulatory Commission. Fchruary 2004, Revision 0.

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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 I.

14. ASME/ANS, "Standard for Level I/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-00 I, 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 I, Main Report, March 2009.

18. "Treatment of Parameter and Model Uncertainty for Prohahilistic Risk Assessments,"

EPRI, Palo Alto, CA: 2008, 1016737.

1'r11c* .\'ort/1 Cm1mlti111:. /./.(' l ..'iH 74.f Hflfl.fl./

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TABLE 1 La."311e ~ol Met and Capability Cal~ory (CC) I Supporting Requirements (SRs)

Sapportiai Capability I Reqaimnrat

(~el)

Cat~* Risk-Informed ISi E\*aluation Impact EPRI TR-1021467 Requirement I

i IE-A9 11E-A71 CCI CC I is sufficient for 1his RI-ISi applicalion.

Thi~ SR addn."SSC!S rt:\ ie\\ of plan1-~p..-cific pn.-cursors 10 delermine if polential initiating e\'ents CCI II are missing from 1he PRA. The n.-quirement heing ernluated here does not ha\'e an impac1 on 1he modekd ini1ia1ing e\en1s. CC I is acceplable for RI-ISi applica1ions.

I I

This p.."t."f rt:\ ie\\ op.:n ilcm was addrcsSL>d in lhc 2011 PRA updale (URE LS20I0-0011 ).

I App..-ndi"< J \\as add..>d 10 LS-PSA-001 . LaSalle PRA lni1ia1ing faenl No1ebook [ 16).

docullll."llling a rt:\ie\\ of1he LaSalle Licensee E\ent Repons (LERs) and other industry plant I op.."r.lling C"<J><.'fiL"llCC 10 dL"lerminc if 1hcre are ini1ia1ing even1s 1ha1 have not been identified i pre\iousl~ and modeled. So new iniliating evenls were identified as a resuh of1he review to I n."Whe 1his open i1em.

i I IE-Ol  :"ot ~k"l This SR n.-quin.-s !hat sourcL"S of unccnainty and assump1ions associaied wilh 1he initialing Not Required to be Met I e\L"lllS anal~sis are documented.

I This p.."t."f re\ ie\\ op.:n i1em was addn."Ssed in lhe 2011 PRA upda1e (URE LS2010-0018). The I LaSalle PRA model uncenaim~ analysis follows lhe indus1ry guidance documented in NUREG-I I 1855 [ 17) and !he associalL>d EPRI repon [I 8): and is documented in LS-PSA-013. LaSalle I PRA Sumlllaf) Sotebook [ 15).

I I AS-C~  :"lll ~l<.'1 This SR n.-quin."S 1ha1 sourc<."S of uncenaint~ and as~umptions associaied wilh 1he accident Not Required to be Met I

~-qucncc anal~sis arc documenled.

I This p.."t."f rt:\ie\\ op.:n i1cm was addresSL>d in lhe 201 I PRA update (URE LS2010-0018). The II LaSalk PRA model uncenain1y analysis follows 1he industry guidance documented in NUREG-I 1855 [ 171and1he associa1ed EPRI repon [I 8): and is documen1ed in LS-PSA-013 . LaSalle I PRA Sumlllaf) So1<.>book [I 5 ].

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TABLE I La.'\alle ~ol '.\lei and Capability Cat~ory (CC) I Supporting Requirements (SRs)

~ C.pabilily I I lleqairemelll cSoce J)

Ca1~- Risk-lnfonned ISi fa*aluation Impact EPRI TR-1021467 Requirement SC-85 ~Ol ~kt This SR n."quin."S 1ha1 cht."Cks be p.:rformed lo delennine lhe reasonableness and acceplabilily of CC I/II/III I

I' 1ht.-nnal h~draulic. suuclural or ocher supporting engineering basis used 10 support lhe success l.Tilt.Tia.

I ~ 2008 Pet.'f re\ iew learn noted 1ha1 while lhe PRA model documenlacion provided some

!>c!lt."\.-lt.d comparison of pre\ ious LaSalle 1hermal hydraulic resuhs lo more recenl calculalions.

lht.'fC \\-;is no documenlt."li comparison of how 1he LaSalle success criceria compare 10 chose used for sislt.-r planes or other similar comparisons. The learn noled 1ha1 lhe success criceria used for LaSalle appt."ar 10 be consislent wich chose of ocher similar BWRs. The learn documenled in lhe I pt."\.'f re\ie\\ repon (7) lhal "chis is a documenlalion issue only."

!I I To addn."Ss chis pt.-.:r re\ie" op.:n icem. 1he success crileria and associaled calculalions were re\ie\\t.-d. Comparisons oflhe succt."SS crileria were reviewed and compared wich ocher similar II planes. ~o anomalit.-s were identifit."li in 1he LaSalle success criteria or supporting calculacions.

I This pt."\.'f re\iew open item remains open 10 address only lhe documentacion issue noced. This I is 1rackt.-d h~* l"RE LS::!Ol0-0030.

I This issue has no impacl on lhc RI-ISi applicacion.

II SC-0 ~Ol ~k"t This SR n."quin.-s 1ha1 wurct."S of uncertainty and assumplions associaled wilh lhe development Not Required to be Met II j

I of sucet."Ss crilt.-ria are documenled.

I This pt."\.'f re\iew open ilem \\as addn.-sSt."li in lhe 2011 PRA updace (URE LS20I0-0018). The LaSalle PRA model unct.'flainty analysis follows 1he induslry guidance documcnled in NUREG-1855 [ 17) and 1he associalt."li EPRI report [ 18): and is documcnled in LS-PSA-013. LaSalle I PRA Summary ~oteliook [ 15).

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TABLE I La.'ialle ~ot ~let and Capability Category (CC) I Supporting Requirements (SRs)

Sapporti~ Capability Reqairrmml Cat~* Risk-Informed ISi E\*alualion Impact EPRI TR-1021467 i *~**

Requirement SY-A-' CCI CC I is sufficient for this RI-ISi application. CCI I

This SR addn.-ss..-s confirmation of the systems modeling in the PRA with plant operators and II ~~* stem engilk."\.'TS. CC I n."quin.-s only inten*iews of plant personnel. while capability category 11/111 n."quin:s plant walkdowns as well as inten*iews to confirm that the systems analysis acrnratc:l,Jo n:tlt."Cls the a.~ built. a.~ operated plant.

I This pet.'T n:\ ie\\ open item n:mains open and is tracked by URE LS2010-0032. Plant I \\aU..do\\ns ha\e bt.~n conductt.-d for internal flooding. fire PRA de\'elopment as well as seismic PRA dc\c:lopment. :'\o issues with system modeling ha\'e been identified during these I walkdo\\TIS. Compk'le system walkdO\ms solely for the purpose of meeting CC 11/111 are a lo\\t.'T priority due to the maturity of the systems modeling in the PRA. the confirmation of modeling through inteniews. and the desin: to maintain the ALARA (as low as reasonablely achie\ablel principle \\i th n.-spt.-ct to radiation exposure.

II SY-Cl '.'Oot !\.k1 This SR n."quin.-s that sotm.'"CS of uncenainty and assumptions associated with the systems Not Required to be Met anal~ sh; an: dOl."Umt.'fltt.-d.

I I This pt."\.'T n:\ie\\ open item was addn.-ssed in the 2011 PRA update (URE LS2010-0018). The LaSalle PRA model uncenainty analysis follows the industry guidance documented in NUREG-1855 (I 7) and the associatt.-d EPRI n:pon [ 18): and is documented in LS-PSA-013. LaSalle I PRA Sum~* :'\otehook (15 ).

Tn1e .V< addresSL-d in the 2011 PRA update !URE LS2010-00-l2).

\\"hile the pt."\.'1" re\ iew team dc:sin.-d to ha\'e a dc:tailed documented list of procedures reviewed.

II this is not requin.-d by the SR. The action takc:n to resol\'e the open issue was to define a I pnx-.:ss to first itk"fltify. through systems analysis. potential misalignments or off-normal I configurations. and then identify pre-initiator HFEs to include in the model based on the review I of sys1c:m procL-dun."S and maintenance practices for these potential off-normal configurations.

I I This approach is documenlL-d in LS-PSA-00-l. LaSalle Human Reliability Analysis [ 10) and the I

n."Sults of the pre-initiator HFE identification process are documented in Appendix J of LS-I PSA-OQ.l. Additional!~-. the IL"St and maintenance procedures used in the pre-initiator HRA are I

li!>tL-d in Tahlc: B-5 of Appendi'\ 8 of LS-PSA-OQ.J.

I Tn1e .\"onh Consulting. LLC LS8743R06-04

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TABLE I LaSalle ~ot ~let and Capability Cat~ory (CC) I Supporlin~ Requirements (SRs)

Sapporti111: Capability I ReqaiJ emrat

• Sole ..

I Cat~* Risk-Informed ISi Evaluation Impact EPRI TR-1021467 Requirement II HR-A:!  :'\oc ~kt This SR n.-quin."S identification or system calibrations that if perfonned incorrectly can have an cc 1/11/111 I

I I

I ad\<."f"Sc! impact on the automatic initiation or standby safety equipment. The SR states that this b accomplislk.-d through a re\ ie" of proc<.-dures and practices. This SR is applicable to pre-initiator HFEs.

I With n."Sp<."CI to this SR. the peer re\ iew team conclud<.-d that this SR was not met and noted the I

follo1Aing: 9This n.-quirement is not met because documentation does not provide evidence of I the proct.-dun."S re\ ielA<.-d. It just says proc<.-dures were re,*iewed." Further. the peer review I n.-pon stat<.'\J that - IAhile this is primarily a documentation issue. this documentation is necessary I to m<."l.'1 the n.-quirements or this SR.-

I I

I This pt."l."f re\ic:"' open issue: was addn.-ss..-d in the 2011 PRA update (URE LS2010-0Cl.i2). The action taken to n.-sohc: the open issue was to dc:finc a process to first identify. through systems I anal ~sis. pot<.'11tial miscalibrations. and thc:n identify pre-initiator HFEs to include the model

~~"don the re\ icw or system procedun."S and maintc:nance practices for these potential off-II normal configurations. This approach is documented in LS-PSA-00-l. LaSalle Human Rdiabilit~ Anal~sis ( 10) and the n.-sults of the pre-initiator HFE identification process arc Jocum..'11tt."d in Appendi:"( J of LS-PSA-00-l. Additionally. the test and maintenance procedures I us.."d in the pre-initiator HRA are listed in Table B-5 of Appendix B of LS-PSA-00-l.

I I HR-81 CCI CC I is sufficient for this RI-ISi application. CCI I This SR n.-quin."S tha1 if scr<."l!ning or activities is used. rules arc es1ablished for screening. To m<."l.'1 CC I. cla.~S<."S or acth iti<.-s can he screened. To meel CC 111111. indi\'idual acti\'ities must he I s..-n."l!ll<."d. This SR rekrs to pre-ini1iator HFEs.

I This pt."l!r re\ ii!"' open issue: was addressed in the 2011 PRA update (URE LS2010-0Cl.i3). A I S<.'"f\."l.'11ing m.."thodolog~ was c:s1ablished and individual activities were screened. This screening I approach is documenl<.-d in LS-PSA-00-l. LaSalle Human Reliahilily Analysis ( 10) and the n."Sults of 1he pre-initiator HFE identification process arc documented in Appendix J of LS-

!I PSA~ .

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TABLE I La.'\alle ~ot  :\let and Capability Category ICC) I Supporting Requirements (SRs)

~ Capability I I Rrqaiftmeat

• ~It Cat~* Risk-Informed ISi Evaluation Impact EPRI TR-1021467 Requirement HR-G6 '."O( \kt This SR n.'quin."S that th<.' post-initator human error prohahility (HEP) quantifications he cc 1/11/111 Ii cht.-ck1."l.I for consistenc) . Sp..-cifically. this SR requires a re\'iew of the HFEs and their final HEPs relathe to 1.-ach oth<.'r to ch1.-ck their reasonahleness gi\'en the scenario context. plant hi~o~. pnx:1."l.lun.-s. operational practices. and experience.

I I This SR "'as addn.-ss..'d in th<.' 2011 PRA update (URE LS2010-00.W). LS-PSA-00-t LaSalle Human Reliahilit) Anal) sis [ 10) was updated and pro\'ides a comparison and a reasonableness I cht.-ck of th<.' final post-initiator HEPs.

I l

HR-I.~

I '."01 \let This SR n.-quin.-s that sourc1.-s of uncenainty and assumptions associated with the human reliahilit) analysis are documented.

Not Required to be Met I This p..>t.-r re' iew open item was address..-d in the 2011 PRA update (URE LS20I0-0018 ). The I LaSalle PRA model unc1."ftainty analysis follows the industry guidance documented in NUREG-1855 [ 17) and th<.' associat..-d EPRI repon [ 18 ): and is documented in LS-PSA-013. LaSalle PRA Sumlll3f) '."otehook (15).

I I DA-C8 CCI CC I is sufficknt for this RI-ISi application. CCI I

I I For CC I. this SR n.-quin.-s estimating the time that components were configured in their standby I status. To m.."t."I CC 11/111. plant sp..-cific operational r1.-cords must be used to determine the time that compont.-nts \l.1.-re configun.-d in th<.'ir standby staus.

II This p..>t.-r re'ie"' open item is track1.-d hy URE L.52010-052 and remains open. As noted in the I

p..>t.-r re'ie"' n.-pon. some standhy times used in the LaSalle PRA are plant specific while others II are 1.-stimat1."l.I. This remains open to imestigate the standby time of all components that are not I

normall) running.

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TABLE 1 LaSaUe ='ot :\let and Capability Cat~ory (CC) I Supporting Requirements (SRs)

I ~

Rftlairnmol (Sokl1 Capabili~*

Ca~* Risk-Informed ISi E,*aJuation Impact EPRI TR-1021467 Requirement OA-CIO I CCI CC I is sufliciL"fll for this RI-ISi application. CCI I

I For CC I. this SR rt."quin.-s a re\ iew of IL-st procL-<lures to detemine whether a test should be l.-n.-ditL-d for L"3Ch possible failure mode and rl."quires that only completed tests or unplanned I oPLT.llional demands be counted for component operation.

I I This pt!LT re\ ie" open item is trackL-<l by LIRE LS20 I 0-005 I and remains open. As noted in the I PL"'-T re\ ie" repon (7). this issue relalL-s primarily to documentation and how surveillance tests

\\LTe us..-d to L~timate demands on \arious components.

I I

DA-E.' Sot \kt This SR n."quin.-s that !>Otlrces ofuncenainty and assumptions associated with the data analysis Not Required to be Met I an: documentL-d.

This PL"'-T re\ie" open item was addresSL-d in the 201 I PRA update (URE LS20I0-0018). The II I

LaSalle PRA model uncenainty analysis follows the industry guidance documented in NUREG-1855 [ 17) and the associalL-d EPRI repon [ 18): and is documented in LS-PSA-013. LaSalle PRA Summary Sotebook [I 5).

I II IFSS-AR CCI CC I is suOicient for this RI-ISi application. CCI II clF-C.~ti1 For CC I. this SR dOL-s not rt."quire an analysis of inter-area propagation given that llood areas an: indL"J><."Tldent.

I I 1llC n.-solution of this CC I PL"'-T re\*iew asSL-ssment open item is tracked by URE LS2010-0061 j and remains OPL"fl. 1llC n.-solution of this open item is estimated to have a negligible impact on the n.~hs of the intL-rnal llooding e\ aluation because the results of the internal llooding PRA i

I I an: dominalL-d b~ large llood e\ents that bypass installed barriers such as water tight doors.

True .\"CNtlt Consulling. UL LS8743R06-04

/SI Program Plan LaSalle Count)* Station Units I & 2, Fourtlr Interval 10CFRS0.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 26 of 29)

TABLE I LaSalle Sol :\lei and Capabilily Cat~o11* (CC) I Supporting Requirements (SRs)

I I

SapportiJll:

ltaqWn meol (Sok.,

Capabilily Ca1~* Risk-Informed ISi Evaluation Impact EPRI TR-1021467 Requirement I

IFPP-R~ ~Cl( ~kt This SR n..-quin.."S that soun:t."S of uncenaim~* and assump1ions assodaied wilh 1he internal Nol Required to be Met I flooding anal~sis are documentt.'ti.

IFS0-8.l I IFS:"'-8~

This pt.'t.'f rc\ie" open i1em "as addn.."SSt.-d in 1he 2011 PRA upda1e (URE LS2010-0018). The I IFEV-H~

LaSalle PRA model uncenainty analysis follows lhe induslr)' guidance documenled in NUREG-1855 ( 171and1he associalt."d EPRI rcpon ( 18): and is documented in LS-PSA-013. LaSalle I PRA Sum~ :"'ot~k ( 15 (.

IFQL'-8~

I 1IF-Fl1 II Ql'-DI  :"oc ~kt Thi~ SR n..-quin.."S a re\ ie" of a sample of lhe significant accident sequences/cu1se1s sufficienl 10 CC 11111111 I dt.'1t."flltine 1hat Ille logic of the cutsel or St."quence is correcl.

1Ql'-Dla1 I It was dt.'1t.-nnint.-d ha.<;(."d on interviews with the PRA model developers thai significant accident I SL"qU<.'1lC\."S and cutSt.'1s Wt.'fC re\ iewed. This was a documentation issue and was tracked by URE L~'.?010-0069.

I Thi~ pt.'t.'f rc\ie" open i1em "~ rt."SOh<."d during the 1011 PRA updale. LS-PSA-01-l. LaSalle I

PRA Quantification :-.:otehook. Re,*i!oion 8. (81 includes a discussion of the re\'iew and well as iI documt.-ntation of a sample of significant cutsels/acddent sequences.

I Ql'-05  :"ot ~l ..'1 Thii. SR n..-quin.."S a re\ ie" of a !oample of 1he non-significant accident sequences/cu1sets CC 1/11/111 I sufficient 10 dt.'1t.-nnine lhat 1he logic of 1he cutset or St."quence is correcl.

1Qt:-o.i 1 I It wa.<; dt.'1t."flltint."d haSt."d on intef\ iews with the PRA model developers that non-significant accidt.-nt SL"qU<."TICt."S and l'Ulsels \\ere reviewed. This was a documentation issue and was I 1rad,..-d hy l'RE L~:?O 10-0070.

I Thi~ pt.'t.'f rc\k\\ open ilem \\a.o; rt."SOht."d during lhe 1011 PRA update. LS-PSA-01-l. LaSalle I PRA Quantification :-.:otebool... Re\ ision 8. (81 includes a discussion of the review of significant cutSt.'1..Vaccident SL-quenCt."S.

Trae .\*orth Consullillg. UC LS8743R06-04

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval IOCFR50.55a Relief Request 14R-01 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 27 of 29)

TABLE I La."alle ~ot ~let and Capability Cat~ory (CC) I Supporting Requirements (SRs)

II

~

R.tqaiftmnll fSole It Capabili~*

Cat~ Risk-Informed ISi E,*aluation Impact EPRI TR-1021467 Requirement 1

~sumptions I Qt:-E! ~(I( '.\kt This SR n..-qutn."S thal made in the de\ elopment of the PRA arc idcmilic:d. Not Required to be Met I This ~'\.-r re\iC" open item \\as addressed in the 2011 PRA update (URE LS20I0-0018). The LaSalle PRA model unccnaimy analysis follo\\s !he industry guidance documented in NUREG-I 1855 f I 7) and the a..'\socia!t."d EPRI repon [ 18): and is documented in LS-PSA-013. LaSalle PRA Sumlllar) ~Olcbook f I :'i ).

I I Ql'-E.! ~ot '.\kt This SR n.-quin."S thal an C\ alualion be done 10 C\ aluale !he impact of model unccnainities on

!he rt."SUhs of !he PRA model.

Not Required to be Met I

Thi'\ ~'\.-r re\iC\\ open item was addrt."SS(."d in the ::?011 PRA update (URE LS2010-0071 ). The II I

LaSalle PRA model unct.'ftainl) analysis follows !he industry guidance documemed in NUREG-1855 f 17) and !he associa!t."d EPRI repon f 18): and is documented in LS-PSA-013. LaSalle PRA Sumlllar) ~Olebook f 15 ).

I I

I I Ql'-F~ CCI CC I is sufficit.-m for this RI-ISi application. CCI I

For CC I. this SR n.-quin.-s !hat the significant conlrihulors lo CDF are documented in !he PRA I n."Sults sumlllar)*. For CC 111111. this SR additionally rt.'quires a detailed description of I' significant accident ~ucnct."S or functional failure groups.

I This ~'\.-r re\icw o~-n ilcm ll'RE L'i2010-0073) was resol\"ed during !he 2011 PRA update.

L'i-PSA-01-'. LaSalle PRA Quamificalion :"lotebook. Re\"ision 8. f8) includes a detailed I dt.~"ription of significant accident !>t."qUenccs.

Tnte .\ '"orth Con$ulling. UC LS8743R06-04

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-Ol Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 28 of 29)

TABLE 1 l..a..~e :oiiot ~let and Capability Category (CC) I Supporting Requirements (SRs)

~ Capabiliay

~ Cat~* Risk-Informed ISi E,*aJualion Impact EPRI TR-1021467 fSok ., Requirement Ql*-F~  :"lll ~kt Thi~ SR n.-quin."S that sources of uncenainty and assumptions associated with the PRA model be Not Required to be Met dOl.""Umenh..-d.

This P.,."t."f re\ie" open item was addresse!d in the 2011 PRA update (URE LS:!OI0-007..f). The LaSalle PRA model uncenaint)* analysis follows the industry guidance documented in NUREG-1855 (I 7) and the aswciatt.-d EPRI repon [ 18): and is documented in LS-PSA-01 J. LaSalle I PRA Summary :-.;01ehool.. [ 15).

Ql.-Ff>  :"lll ~k1 This SR relatt."S 10 documenting the quantitati\"e definition use!d for significant hasic event. cc 1/11/111 I ~*gnificant cuts..'l. and significant accident sequence.

It "as nlllc."\J h) the pc.-er re\ ie" team that other than in the HRA notebook. the documentation I

did nlll include the applic.-d definition of "significant". This peer review open item (URE I I Lli2010-00761 "as n."SOhc:d during the 2011 PRA update. LS-PSA-01..f. LaSalle PRA I' I Quantification :"otebook. Re\ ision 8. (8) includes a definition of "significant."

II LE-F.~  :"lll ~kt This SR n.-quin."S the idc.-ntification of contrihutors to LERF and characterization of LERF Unct."flaintic."S.

Not Required to be Met i

I This pt."t."f re'i.:" opc.-n item "as addn.-ss..-d in the 2011 PRA update (URE LS20I0-0080J. The I LaSalle PRA model uncenainty analysis. including an analysis of LERF. follows the industry guidance docurnc.-ntc:d in :'\l'REG-1855 (I 7) and the associatc:d EPRI repon [ 18): and is I dOl.""Umentc."\J in LS-PSA-013. LaSalle PRA Summary Notebook ( 15).

I I

True .\"onlr Consulling, UC LS8743R06-04

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval IOCFR50.55a Relief Request 14R-Ol Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 29 of 29)

TABLE I La.'\alle Sol ~lel and Capability Cat~ory (CC) I Supporting Requirements (SRs)

I Sapportini Capmbilily lbqainmeot Calq:ory Risk-Informed ISi Evaluation Impact EPRI TR-1021467 I fSoce II Requirement I LE-G.l ="01 \lc..'t This SR n."quin.>s the docullll!ntation of assumptions and sources of uncertainty associated with Not Required to be Met I the LERF anal) sis.

I This Jl<."t."f re\ie\\ OJl<."11 item was addn.>ssed in the 2011 PRA update (URE LS2010-0080). The I

I LaSalle PRA model unct.-naimy analysis. including an analysis of LERF. follows the industry guidance d<x.-ullll!ntt."d in ="L"REG-1855 ( 17) and the associated EPRI report I 18): and is documr.."lltt."d in L'i-PSA-OD. LaSalle PRA Summary Notebook (15).

I  !

t I LE-Gt> ="°' \lt.'t Thi' SR n."quin.>s docullll!ntation of the quantifitathe definition used for significant accident.

Thb \\as a docullll!ntation issue \\ith no impact on PRA model results.

CC I/II/Ill Thb Jl<."t."f f'C\ie\\ open item IURE LS2010-0081) was resolved during the 2011 PRA update.

II' I LS-PSA-01~. LaSalle PRA Quantification Notebook. Revision 8. (8) includes a definition of

- ~i gni ficant. -

Note'>:

I. The LaSalle PRA peer review wa'> against the ASME/ANS 2005 PRA standard [9]. The current 2009 ASME/ANS standard [ 1-t) SRs are provided with the equivalent peer reviewed SR listed in parentheses where applicable.

Tn1e Sonlr Consulling. UC LS8743R06-04

IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

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

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 Dcscri pt ion: 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 Bcltlinc Region, Interior Attachments beyond Beltlinc Region, and Core Support Structure

2. Applicable Code Edition and Addenda The Fourth I0-Y car Interval of the LaSalle County Station, Units I and 2 lnservice Inspection (ISi) Program is hascd on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Reuuirements ASME Section XI requires the examination of components within the Reactor Pressure Vessel. These examinations arc included in Tahlc 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 hy the YT-3 visual examination method (B-N-1 ).

813.20 Examine interior attachment welds within the hcltlinc region each interval by the YT-I visual examination method (B-N-2).

Bl3.30 Examine interior attachment welds beyond the hcltlinc region each interval hy the YT-3 visual examination method (B-N-2).

B 13.40 Examine surl'ilccs of the welded core support structure each interval hy the VT-3 visual cxmnination method (B-N-2).

These examinations arc performed to asscs!'o. the !'o.tl'llctural integrity of componcnt!'o. within the boiling water reactor pressure vessel.

1'r11r ,\/11rtl1 Cm1.rnlti111:. I.I.< '

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

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4. Reason for Request In accordance with IOCFR50.55a(z)(l ), relief is requested for the proposed alternative to ASME Section XI requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVIP Inspection and Evaluation (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 Tahlc I for Examination Category B-N-1 and B-N-2.

LaSalle County Station, Units I and 2 will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table I in accordance with the latest Nuclear Regulatory Commission (NRC') approved BWRVIP guideline requirements. This relief request proposes to utilize the identilied BWRVIP guidelines in lieu of the associated ASME Section XI requirements, including examination method, examination volume.

frequency. trnining. successive and additional examinations. llaw evaluations. and reporting.

Not all the components addressed hy these guidelines arc ASME Section XI components.

The following guidelines arc applicable to this relict' request:

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

• BWRVIP-18, Revision 2-A. "BWR Core Sprny lntcnmls Inspection and Flaw Evaluation Guidelines"

• BWRVIP-25. "BWR Core Plate Inspection and Flaw Evaluation Guideline!\"

• BWRVIP-26-A. "RWR Top Guide Inspection and Fhlw Evaluation Guidelines" 1'ml' ,\'ortlr ('1111mlti111t. /./,(' /SH74.fNt16*fl4

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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

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

- BWRVIP-41, Revision 3, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines"

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

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

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

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

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

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

- BWRVIP-138, Revision 1-A "Updated Jct Pump Beam Inspection and Flaw Evaluation Guidelines"

- BWRVIP-180, "Access Hole Cover Inspection and Flaw Evaluation Guidelines" Any deviations from the referenced BWRVIP Guidelines for the durntion of the proposed alternative will he appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process. Current LaSalle County Station deviations from the suhject guidelines ahovc arc sumnrnrizcd in Tahlc 2.

Inspection services, hy an Authorized Inspection Agency, will he 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 <tnd reexamination fre<.1uencies . The BWRVIP has established a reporting protocol for examination results mHl deviations. Enclosures 2 mHI 3 contain the "Reactor Internals Inspection I listory" for LaSalle County Station Units 1 and 2. This summary provides. on a component-by-component basis, the examination methods utili1.ed. the examination frequency to date, and the results of the examinations during the previous interval. This tahle also contains the identified corrective actions. The information provided reflects the compilation of the BWRVIP 120-day report!\.

Corrective actions und exumim1tions performed prior to the BWRV IP were implemented to the requirements of ASME Section XI, as applicahle. The Nl~C has agreed with the BWRVIJ> appro*u:h as documented in References I through 14. Therefore, Ul\e of these guideline!\, as an altenrntive to the subject ASME Section XI rc<.1uirementl\, provides an 1'ruc* N11rtll ('1111.rnlti111t. I./.(' u;H7./JHfl6-fl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Page 4 of 17) acceptable level of quality and safety and will not adversely impact the health and safety of the public.

As additional justification, Table I of 14R-02, "Comparison of ASME Section XI Examination Category 8-N-I and 8-N-2 Requirements with 8WRVIP Guidance Requirements," 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 IW8-2500-l, to the inspection requirements in the 8WRVIP documents. Specific 8WRVIP documents arc 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 I compares present ASME Section XI Examination Category 8-N-I and 8-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to LaSalle County Station. Therefore, Table I only represents a current comparison. Any deviations from the 8WRVIP guidelines referenced within this relief request for the 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 al LaSalle County Station has been developed and implemented lo satisfy the requirements of BWRVIP-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 NRC-authorized proposed alternative to the requirements of IOCFR50.55a, it may he implemented. Olherwise, lhe revised guidelines will only he implemented afler NRC approval of the revised BWRVIP guidelines or a plant-sped lie request for rclicl' has been approved.

(,, Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle Counly Slalion, Unils I and 2.

7. Precedents

• The Exelon Genern1ion C'omp*my/AmerGcn tleet-widc relief requcsl for BWRVIP was authorized conditionally by NRC' Safety Evalualion (SE) dated April JO. 200K (ADAMS Acccl\,ion No. Ml.OK09K03 I I) (i.e .* LaSalk Counly 1'r111* .\'11rtlr C'm1mlti111t. 1.1.C'

/SI Program Plan lASalle County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRS0.55a(z)(l)

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

Revision 0 (Page 5 of 17)

Station Third ISi Interval Relief Request 13R-02.) This relief request for the LaSalle County Station, Units I and 2, Fourth ISi Interval utilizes a similar approach to the previously approved relief request. (Reference 15)

• Relief Request 14R-02 was authorized for Limerick Station, Units I and 2, by NRC SE dated November 21, 2016. (Reference 16)

• Relief Request IR-056, Rev. I was authorized for Perry Nuclear Power Plant Unit I by NRC SE dated January 31, 2012. (Reference 17)

8. References I. 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)," dated Decemher 21, 2016 (ADAMS Accession No. ML I6273A083 ).

2. Letter from NRC to BWRVIP, "Final Safety Evaluation or 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 Decemhcr 19, 1999.

3. Letter from NRC to BWRVIP, "NRC Approval Letter or BWRVIP-26-A. 'BWR Vessel and Internals Project, Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines,"' dated Septemher 9, 2005 .

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

5. Letter from NRC' 10 BWRVIP. "Final Safety *~valuation of the 'BWR Vessel and Internals Project, BWR Shroud Support ln.,peclion and Flaw Evaluation Guidelines (BWRVIP-38),' EPRI Reporl TR- 10882J (TAC' No. M99618)," dated July 24, 2000.

6. BWRVIP-41. Revision 3, BWR Vessel and Internal~ Project. BWR Jct Pump A~~emhly ln~pcction and Fh1w Evaluation Guidelines, EPKI Tcdmical Report 1021000, dated Scptemhcr. 2010.

1'r111* ,\ '11rt/1 <'1111.rnlti111:, /./.(' I ~4'H7./.JNfJf>.fJ./

/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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7. "BWRVIP-42, Revision l: BWR Vessel and Internals Project, LPCI Coupling Inspection and Flaw Evaluation Guidelines," dated June, 2010.

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

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

10. "BWRVIP-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. Lcllcr from Chairman, BWR Vessel and Internals Project to NRC. "Project No.

704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2),"

dated September 22, 2011 (ADAMS Accession No. MLI 1271 A058).

13. Letter from NRC to BWRVIP, "Electric Power Research Institute Final Safety Evaluation for Technical Report 1016574 "BWRVIP-138, Revision I-A: BWR I Boiling Water Reactor I Vessel and Internals Project 'Updated Jct Pump Beam Inspection and Flaw Evaluation Guidelines'" (TAC No. ME2 l 9 I)," dated May 14.

2012.

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

15. Leller from R. Gihhs (NRC) to C. G. P<mlee (Exelon Generation C'ompany/AmerGen). "Clinton Power Station Unit No. I: Dresden Nuclear Power Station, Units 2 and 3: LaSalle County Station. Unit!> I and 2; Limerick Station.

Unit!> I and 2: Oyster Creek Nuclear Station: Peach Bottom Atomic Power Station, Units 2 und 3: and Quad C'itie1' Nuclear Power Station, Units I and 2 -

Relief Request to Use Boiling Water Reactor Vessel and lntcnrnls Project Ciuidclines in Lieu of Specilic ASME Code Requirements <TAC No\. MD5352 through MD~U63 )," dated April 30. 2008 (ADAMS Accession No.

Ml.080980311 ).

1'rt11* "'"" <'1111.rnlt/111:. /./.('

/SI Program P/a11 LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 7 of 17)

16. Letter from S. S. Koenick (NRC) to B. C. Hanson (Exelon Nuclear), "Safety Evaluation of Relief Requests 14R-02and14R-IO for the Fourth IO-Year Interval of the Inservice Inspection Program for Limerick Station, Units I and 2 (CAC Nos. MF7587 and MF7588)," dated November 21, 2016 (ADAMS Accession No. MLl6301A401 ).

17. Letter from J. I. Zimmerman (NRC) to V. A. Kaminskas (FirstEnergy Nuclear Operating Company), "Perry Nuclear Power Plant, Unit No. I, Re: Safety Evaluation In Support of IOCFR50.55a Requests for the Third I0-Y car In-Service Inspection Interval (TAC Nos. ME5373, ME5376, ME5377, ME5379, and ME5380)," dated January 31, 2012 (ADAMS Accession No. ML120180372).

9. Enclosures I. Comparison of Code Examination Requirements to BWRYIP Examination Requirements

2. LaSalle County Station Unit I Reactor Internals Inspection History. dated April 6, 2016

3. LaSalle County Station Unit 2 Reactor Internals Inspection History. dated March 16,2017

'/'rnr N11rtli C1111.rnlti11Jt, 1.1.C U iH74.fHIJf>.fl./

ISi Program Plan IASal/e County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 ENCLOSURE I Comparison of Code Examination Requirements to HWRVIP Examination Requirements I0 pa~es follow 1"r11t' .\'11r1/i ('t111,rnl1i11R. I .I.<*

/SI Program Plan IASal/e County Station Units I & 2, Fourtlr Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 8 of 17)

I TABLE I Comparison of AS:\IE Section XI E.umination Category B-N-1 and B-N-2 i Requirements with BWRVIP Guidance Requirements*

I AS~IE Stttioa XI AS:\IE AS~IE ASME BWRVIP l Item !l'amber.

, Table m"B-i.s.1 Componmt Section XI ExamSc:ope Section XI Exam Section XI FreQuency Authorized Alternative Exam Scope BWRVIP Exam BWRVIP Frequency I 8HIO Accc."Ssihlc VT-J Each period BWRVIP-18-R2-A. 0\ erview examinations of components during BWRVIP I

I Ri:actor \' c."S!d An:~ 25. 26-A. 27-A. JS. examinations satisfy ASME Section XI VT-J visual I

I Ink-nor -l 1-RJ. -l2-R I. -l7- examination requirements.

' A. -l8-A. 76-R I-A.

t I J8-R I -A. and 180 I

I 81.~ .~ I kt Pump R1~ Bracc."S Acet!Ssihlc Wdds VT-I Each BWRVIP--l8-A. Riser Brace EVT-1 100% in first 12 I

I ln1c:nor 10-year Tahlc J-2 Attachment years; 25% during

, Atuchrnt:'llb lntcnal each subsequent 6 I \\ utun lkltlinc years .

• I I Rep on Lo\loc."f Sunc."lllam"\: BWRVIP--l8-A. Bracket VT-I Each 10-year i Spc....;m..-n Hl1IJ..-r Table J-2 Attachment Interval.

I I I Brad:c."b

!I 81~ 30 St~ l~c."f Hold- Accc."S~ihle VT-J Each BWRVIP--l8-A. Bracket VT-J Each 10-year

Interior dl~\loll Brad.c."ts Wdds IO-year Table J-2 Attachment Interval.

I ~hm.."TIU Guid.: Rod Hrad:c."ts lntenal BWRVIP--l8-A. Bracket VT-J Each 10-year I

' Hc~onJ lkhlux Table J-2 Attachment Interval.

l St12111 ~c."f Suppon BWRVIP--l8-A. Bracket EVT-1 Each 10-year I

I 8rxl..c."t!> Table J-2 Attachment Interval.

I fc.'\."\JWalc."f Spar~ BWRVIP--l8-A. Bracket EVT-1 Each 10-year I

Brad.ct~ Table J-2 Attachment Interval.

I II I Core Spra~ Pipmg BWRVIP--l8-A. Bracket EVT-1 Every 4 Refueling I Hrad.c."t!i> Table .3-2 Attachment Cycles.

True .\"o111r ConsalhnK, UC LS8743R06-04

/SI Program Plan LaSalle County Station Units I & 2, Fourtlr Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Altemath*e Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 9 of 17)

I TABLE I

!I Comparison or ASME Section XI EHmination Category B-N-1 and B-N-2 t Requirements with BWRVIP Guidance Requirements' I

I AS~IE Sedioa XI I l*ftll s llJllbft'.

T..- m"B-.z..qo.1 Component AS~IE Section XI ExamSc:ope ASME Section XI Enm ASME Section XI Frequency Authorized Alternative BWRVIP Exam Scope BWRVIP Exam BWRVIP Frequency I

I l"ppcr and ~hJdk BWRVIP-48-A. Bracket VT-3 Each 10-year I Suodllance S~*dlTk.'fl Table 3-2 Attachment Interval.

Hokk.-r Hrad.L"ts

't I ShrouJ Support 1WelJ BWRVIP-38. Weld H9~ EVT-1 or UT Based on as-found H91 mdudmg gtJ!'5L"ts 3. 1.3.2. including conditions, to a

~h.:n: apphL-ahk Figure~ 3-2 and 3-5 gussets maximum of 6 years (where for one sided EVT-applicable) 1, 10 years for UT.

Shroud Suppon U.-gs 1Rarel~ BWRVIP-J8. Weld Hl2 Per BWRVIP- When accessible.

tWcld Hl2 1 ACCL'Ssiblc l J .2.J 38 NRC SER (07/24/00).

I inspect with I

I appropriate I

I I method' I'

BP.W j ShrouJ Support t WelJ r\CCL'SSib)C VT-J Each BWRVIP-38. Shroud EVT-1 or UT Based on as found

\\'ddcd Ll"Jre HIOI Surf~ JO-year J . l .J .2. Support conditions, to a I

I J

Suppon S~"ture lnterYal Figures 3-2 and 3-5 (Weld HIOl maximum of 6 years I and Leg for one sided EVT-I I

Welds including 1, 1 0 years for UT where accessible.

I gussets as I

I applicable Tnu .\'o'1h Co1Uulling. UC LS8743R06-04

IS/ Program Plan LaSalle County StaJion Units I & 2, Fourth Interval 10CFR50.5Sa Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 10 of 17)

I TABLE I Comparison or ASl\IE Section XI E."lamination Category B-N-1 and B-N-2 II Requirements with BWRVIP Guidance Requirements 1 j A..~IE S<<tioa XI AS~IE ASl\IE ASl\IE BWRVIP ITalM ltna Stllllbft-.

J\\"B-?Se8-I Component smionXJ E.umScope Section XIE.um Section XI Frequency Authorized Alternath*e Exam Scope BWRVIP Exam BWRVIP Frequency I

I i Shroud Hori1ontal BWRVIP-76-Rl-A. Welds HI- EVT-1 or UT Based on as found i Wdd.~ 2.2.1 H7 as conditions, to a I

I applicable maximum 6 years I

I for one sided EVT-

1, 10 years for UT I

I where accessible.

j BWRVIP-76-Rl-A. Vertical and EVT-1 orUT Maximum of 6 years I

Shroud \-'f'lk"al Welds I 2.3. Ring for one-sided 1

Figure 3-3 Segment EVT-1, 10 years for I Welds as UT.

I applicable II Shroud R'--pai~ BWRVIP-76-Rl-A. Tie-Rod VT-3 Per designer Section 3.5 Repair recommendations I

I per BWRVIP I I R1-A.

~Oles:

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

2. In accordance with Appendix A of BWRVIP-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.

True .\"onh Consulting, U.C LS8743R06-04

/SI Program Plan LaSall~ County Station Units I & 2, Fourth Interval IOCFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 11 of 16)

I TABLE2 BWRVIP Deviations I

Pl.A.,,.

II La..~k cowu ..

BWR\"IP DOCL~IE.'7 BWR\'IP-~'i LETTER DA TE TO ~RC l..."th.'T from S. E. Kucz~nski DEVIATION Postponement of Core Plate Bolt inspections on both Unit I and APPLICABILITY This Deviation docs Sl3llon l"m~ I cEGO to :'\RC dat..-d Unit 2 Ix-cause meaningful inspections are not currently possible not impact the basis I md:? ~larch ~I.  :?O 11 using existing methods. for the use of this I

I relief reauest.

TTrl~ .\'ortlr Consulting, UC LS8743R06-04

IS/ Program Pla11 LaSalle Cou11ty Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request I4R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 12 of 17)

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 B 13.10, B 13.20, B 13.30, and B 13.40 in Table IWB-2500-1, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines arc provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. ASME Section XI Requirement - B 13.10 - 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 IO-Year ISi Interval. Typically, these examinations arc 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 lleel. 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 BWRYIP guidelines require access to accessible areas of the reactor vessel during each refueling outage.

Examination of core spray piping and spargers (BWRVIP-l 8-R2-A). top guide (BWRVIP-26-A). jct pump welds and components (BWRVIP-41-RJ). interior attachments ( BWRVIP-48-A). core shroud welds (BWRVIP-76-R I-A). shroud support (BWRVIP-38), LPCI couplings (BWRVIP-42-R I). and lower plenum components (BWRVIP-47-A) provides such ucccss. Locating and examining specific welds and components within the reactor vessel ureas above. hclow (if accessible). and surrounding the core (annulus area) entails access hy remote camera systems that essentially perform equivalent YT-3 visual examination of these areas or spaces as the specific weld or component examinations urc performed. This provides un equivalent method of visu11l cxmnination on a more fn!lJUCnl hasis than that required hy ASME Section XI. Evidence of wear, structural degradation, loose, missing, or disph1ced parts. foreign material~. and rn1To,ion product buildup can he, und has been oh~crved during the cour~c of 1'r11r N11rtli l'1111.mlti111t. /./.(' ISH74.flW6*tl4

ISi Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 13 of 17)

ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS implementing these BWRVlP examination requirements. Therefore, the specified BWRVIP 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 BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Section XI requirements.

2. ASME Section XI Requirement - B 13.20 - Interior Attachments Within the Bcltlinc

{B-N-2>

ASME Section XI requires a VT-I visual examination of accessible reactor interior surface attachment welds within the belt line each I0-ycar interval. In the BWR/5 model, this includes the jct 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 EYT-1 visual examination on the remaining attachment welds in the heltline region in the first 12 years, and then 25% during each subsequent 6 years.

The jct pump riser hracc examination requirements arc provided hclow to show a comparison between ASME Section XI and BWRYIP examination requirements.

Comparison to BWRVIP Requirements - Jct Pump Riser Braces (BWRVIP-4 l-R3 and BWRVIP-48-A>

• ASME Section XI requires a I00'~* VT- I visual examination of the jct pump riser hracc-to-rcactor vessel wall pad welds each I0-ycar interval.

• The BWRVIP requires an EYT-1 visual examination of the jct pump riser hracc -

to-reactor vessel wall pad welds the lirst 12 ycms and then 25% dming each subsequent 6 years.

• BWRYIP-48-A specilically delincs the susceptible regions of the attachment that arc to he cxumined.

ASME Section XI VT-I visual examination is l'onductcd to detect discontinuities and imperfections on the surfaces of components. including such l'omlitions us cracks, wear.

l'ormsion, or cro!>>ion. The BWRVIP enhanced YT-I (EVT-1) visual examination is conducted to detcl*t discontinuities und imperfections on the surface of component!'> and is additionally !>>pecilied to detect potentially very tight cracks charnctcristic of fatigue and 1"r111* ,\'11rtlt Cmn11ltit1Jl, /./.(' UiH7./.fHfJfl.fJ./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRS0.55a Relief Request I4R-02 Proposed Alternative in Accordance with 10CFRS0.55a(z)(l)

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

Revision 0 (Page 14 of 17)

ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS 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-I visual examination method requires 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 characters on the examination surface. BWRVIP-48-A includes a diagram and prescribes examination for the configuration of 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 Section XI. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced llaw 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 hy ASME Section XI.

3. ASME Section XI Requirement - B 13.30 - Interior Attachment Be yond the Belt line Region <B-N-2 )

ASME Section XI requires a VT-3 visual examination of accessihle reactor interior surface attachment welds beyond the heltline 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 weld. and the shroud support gussets. BWRVIP-48-A requires as a minimum the same YT-3 visual examination method as ASME Section XI for some of the interim attachment welds beyond the bcltline region. aml in some cases spccilics an enlmnced visual examination technique EYT-1 for these welds. For those interior attachment welds that have the same YT-3 method of visual examination. the same scope of examination (accessible welds).

the same examination frel1uency (each 10-ycar interval) and ASME Section XI tlaw evaluation criteria. the level of quality and safety provided hy the BWRYIP requirements arc CllUivalent to that provided by ASME Section XI.

.,.,,,,. "'"' ( '1111.ml1i111:. /,/.(. UiH74.IHfl6-fl./

/SJ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request I4R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 15 of 17)

ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS 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 BWRVIP 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 10 years.

• The BWRVIP visual examination requirement is an EVT-1 for the Core Sprny piping brm:kct attachment welds with each weld examined every four cycles (8 years for units with a two year fuel cycle).

The BWRVIP visual examination method EVT-1 has superior flaw detection and sizing capability, and the same flaw evaluation criteria arc 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 cxamin;ition is conducted to detect discontinuities and imperfections on the examination surfoccs. including such conditions as tight cracks caused hy IGSC'C' or fatigue. the relevant degradation mechanisms for BWR internal attachments.

Therefore, with the EVT- 1 visual exami1rn1ion method. the same examination scope (accessible welds). the s*11ne examination frequency, the same flaw evaluation criteria (ASME Section XI). the level of quality and safety required by the BWRVIP criteria i" superihr to than tlmt rel1uired by ASME Section XI.

1'r11c* .\'ortll ( *,,,, rnltl111t. /./.(

• UiH7./.fHtJf>.fl./

IS/ Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFR50.55a Relief Request I4R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 16 of 17)

ENCLOSURE I COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS

4. ASME Section XI Requirement - B 13.40 - 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 I0-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 BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

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

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

• The BWRVIP requires either an enhanced visual examination technique (EVT-1) every 6 years or volumetric examination (UT) every 10 years as compared to ASME Section XI requirement (VT-3 visual examination). (Only 10'7'* of the weld is required to he examined.)

BWRVIP recommended examinations of welded core support structures arc focused on the known susceptihle areas of this structure, including the welds and associated weld heat affected zones. Jn many locations. the BWRYIP guidelines require a volumetric examination of the susceptihle welds at a frequency identical to ASME Section XI rclJUiremcnt.

For other welded core support structure components. the BWRVIP requires an EVT-1 visual examination or UT of core support structures. The core shroud is used as an example for comparison hetween ASME Section XI and BWRVIP examination requirements us shown hclow.

1'rur .Vorth ('1111rnlti111:. /./.(' UiH7".fHfl6.fl-I

ISi Program Pla11 LaSalle Co1111ty Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request I4R-02 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 17 of 17)

ENCLOSURE 1 COMPARISON OF CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Comparison to BWRVIP Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline <BWRYIP-76, Revision I-A)

• ASME Section XI requires a VT-3 visual examination of accessible surfaces each I0-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 "end of interval" (EOI) that will vary depending upon the amount of flaws present, hut not to exceed ten years.

The BWRYIP recommended examinations specify locations that arc known to he vulnerable to BWR relevant degradation mechanisms rather than accessible surfaces. The BWRVIP exmnination methods (EYT-1 or UT) arc 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 sufety equivalent to or superior to that required hy ASME Section XI requirements.

1'r11c* N11rtll ('cm.rnltillJl, /./.(

• IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRS0.55a(z)(l)

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

Revision 0 ENCLOSURE2 LaSalle County Station Unit l Reactor Internals Inspection History 15 pages follow

'/'rw* ,\'11rtll ( '1111.rn//i1111, I.I.('

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components in BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Inspection Method Used Repairs, Replacements, Reinspections 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 slgnif icant change in length.

Re-exam scheduled In 2 cycles.

L1R13 (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, DP5, and DP6 and due to new Demonstration, the flaws on DPS 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 llmlted.

No indications. Visual examination of five piping brackets.

No Indications.

L1R11 (2006) UT Re-sized flaws on BP4a, DPS, 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 L1A11.

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.

Page 1

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components In BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Method Used Repairs, Replacements, Reinspectlons Scope Inspection L1R08 (1999) UT Ultrasonic examination of the 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 L1A10.

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 L1R16 (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 at225°.

L1A15 (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.

l 1ROB (1999) EVT* 1 Visual examination of 50% of the core spray sparger welds.

No indications.

Page2

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components In BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Method Used Repairs, Replacements, Reinspections Scope 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.

L1R15 (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 (2010) 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.

L1R12 (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.

L1R11 (2006) EVT*1NT*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 obsorvod last outage. No change In the wear.

L1R10 (2004) EVT*1NT*1/ (see jet pump and coro 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.

L1R08 (1999) EVT*1NT*1 (see jet pump and coro spray sections for those attachment welds.) Visual examination of 4 steam dryer support lug welds. No Indications.

Page 3

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 S9.S% 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 10 years.

L1R14 (2012) UT UT of welds H2 (lower only), H3, H5, H6, and HS (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, HS, and HS 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, HS, and HS Is required in 1Oyears.

L1A11 (2006) UT UT of welds H3, H4, H6, and Ha (LaSalle-specific numbering). Coverage on HS and HS was less than 50%,

and a site-specific flaw evaluation was performed and re*

Inspection Is In S 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, H5, HS, and HS (LaSalle-specific numbering). No indications noted except on H4, where indications were 3.0%. Next Inspection In 200S.

Shroud Support L1R1S (2016) EVT-1 Visual examination of approximately 25% of HSa (BWRVIP 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 (200S) EVT*1 Visual examination of both access hole covers . No Indications.

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

No Indications.

Page4

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components in BWAVIP Date or Inspection Summarize the Following Information: Inspection Results, Scope Frequency of Method Used Repairs, Replacements, Reinspections Inspection L1R 11 (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 1Odue 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 HBa (BWAVIP 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. NAI.

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

L1ROB (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.

L1R07 ( 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.

L1A14(2012) VT-2 Visual examination during the system leak test. No Indications.

L1R13 (2010) VT-2 Visual examination during the system leak lest. No Indications.

L1A12 (2008) VT-2 Visual examination during lhe system leak test. No Indications.

PT Surf ace examination. No Indications.

L1R11 (2006) VT*2 Visual examination during lhe system leak lest. No indicalions.

L 1A10 (2004) VT*2 Visual examination during lhe system leak lest. No Indications.

PT Surface examination. No indications.

Page 5

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components In BWRVIP Date or Inspection Frequency of Summarize the Followlng Information: Inspection Results, Scope Inspection Method Used Repairs, Replacements, Reinspectlons L1R09 (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 AB-1a,b,c,d on 5 pumps. No indications.

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

Visual examination of AS-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 AS-7 at 3 locatlons. No indications.

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

Visual examination of AS-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 ldentif led. 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 lour locations.

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

UT Ultrasonic examination of RS-9 at lour locations. Confirmed presence of three llaws previously identlliod visually.

Evaluated as acceptable for one cycle.

UT Uilrasonlc examination of A0-1 and A0-2 at eighteen locations. No indications.

L1A15 (2014) UT Ultrasonic examination of thirteen Group 2 beams. No Indications.

Page6

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components in BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Scope Frequency of Method Used Repairs, Replacements, Relnspections 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 10 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 mlddle 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.

L1R14 (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 ls 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*1/VT*3 Visual examination of twenty slip joint clamps. Recordable Indications on three clamps, and tho othor 17 had no Indications.

Page7

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components in BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Scope Inspection Method Used Repairs, Replacements, Reinspections 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.

L1R13 (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 10 pumps. 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 eKamlnation ol strain relief welds RS-AW on the 9 risers that contain the welds. No new Indications. (Due to Laguna Verde)

VT*3 Visual examination ol 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 BB*3 on Jet Pump 18 and beam replaced.

Pages

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components In BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Method Used Repairs, Replacements, Reinspections Scope 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 10 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 1Owere proactlvely 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 9/1 Oto repair the RS-9 flaws Identified In L1A1 O.

The slip joint clamps on jet pumps 5, 6, 9, and 10 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 L1A10.

EVT-1 Visual examination of RS-3 on 5 pumps. No indications L1R10 (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 AS-3 on 5 risers. No Indications.

VT-3 Bost effor1 examination of the Inaccessible welds AD* 1, AD-2 and OF*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 OF*1 on 11 Jet Pumps. No Indications.

EVT-1 Visual examination of OF*2 on 2 Jet Pumps. No Indications.

EVT*1 Visual examination of AS*1 welds on all 10 risers. No Indications.

Page 9

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components In BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Frequency of Method Used Repairs, Replacements, Reinspections Scope 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 10 jet 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 AB* 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-1on20 jet pumps. Wear Identified on 1Ojet 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 1Oauxiliary wedges Installed In previous outages. No Indications.

Installed auxiliary wedges at the following vessel side locations: jet pumps 4, 12, 13, 14, 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 auxlliary wedges at the followlng vessel side location: jet pump 6. Installed auxiliary wedge at the followlng shroud side location: 11.

VT*1 Visual examination of 2 auxiliary wedges installed In previous outages. No Indications.

L1ROS (1999) UT UT of 1OJet pump beams at the BB* 1 and BB*2 locations. No Indications.

EVT*1 Vlsual examination of DF*1 on 10 Jot Pumps. No Indications.

EVT*1 Visual examination of DF*2 on 10 Jet Pumps. No Indications.

Page 10

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Components in BWRVIP Date or Inspection Summarize the Following Information: Inspection Results, Scope Frequency of Method Used Repairs, Replacements, Reinspections 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 10 jet pumps. No indications.

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

EVT-1 Vlsual 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 RB-1 on 10 jet 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 p~event flow Imbalance the Inlet mixer on jet pump 1owas proactlvely replaced.

Auxiliary wedges Installed at the following vessel side locations: jet pumps 1, 5, 7, e, and 1o. 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 W0-1 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 cyclo.

UT UT of all 20 jet pump holddown beams at BB* 1: ono Indication on #9 beam: beam replaced.

VT-1 A gap was ldenllf led 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 L1A16 (2016) EVT*1 / VT*3 Visual examination of four locations on one coupling (315°).

/VT*1 No Indications.

Page 11

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Frequency of Inspection Summarize the Following Information: Inspection Results, Scope Inspection Method Used Repairs, Replacements, Reinspections L1R14 (2012) EVT-1 I 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.

L1R 12 (2008) EVT*1 I VT-3 Visual examination of four locations on one coupling (135°).

I VT*1 No indications.

L1R10 (2004) EVT-1 I VT*3 Visual examination of four locations on all three couplings.

I VT*1 No indications.

L1R08 (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.

L1R11 (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 10.

Areas Include CRD/ST*1, bottom of H9, and ICH/RPV*1. No Indications.

L1R10 (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.

L1R09 (2002) VT-3 I 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 locallons, and CRGT-3 at 21 locations. No Indications.

L1R08 (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.

Page 12

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 1 Date or Components in BWRVIP Frequency of Inspection Summarize the Following Information: Inspection Results, Scope Method Used Repairs, Replacements, Relnspections 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.

L1R14 (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.

lndlcalions Identified previously were examined and there were no changes In any lndicalions. New Indications were noted on the lifting lug #2, #3, and #4 brackets at 315°, lie rod

. 17-90°, vo1-210°, and the USA from 180*360°. All Indications were evaluatod 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, Reinspections Inspection L1 A12 (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.

L1R11 (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.

L1A10 (2004) VT-3 Visual exams on the end panels and welds; one Indication on bank B, 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 (10 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 Gulde L1A14 (2012) EVT-1 Visual examination of ten grid cells: two metal slivers Identified.

VT-3 Visual examination of one c-clamp: no Indications.

L1A13 (2010) EVT-1 Visual examination of two grid cells; no Indications.

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

L1R 12 (2008) VT-3 Visual examination of two c-clamps; no Indications.

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

L1ROB (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 sur1ace 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 Inspection Summarize the Following Information: Inspection Results, Components in BWRVIP 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. NRI.

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.

OM 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.

L1R 12 (2008) UT There were no dissimilar metal welds examined this outage.

Integrated Surveillance L1R13 (2010) 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 bolls. 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-ls for one cycle.

Page 15

/SI Program Pla11 LaSalle Cou11ty Statio11 U11its I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-02 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 ENCLOSURE3 LaSalle County Station Unit 2 Reactor Internals Inspection History 11 pages follow 1'rt1r ,\ '11rtl1 ( 'mu11/ti11i:. / ./,('

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 L2RlS (201S) EVT-1 Welds APl, AP2, AP3, AP4a, AP4b, AP4c, AP4c-l, AP4d, APS, AP6, AP7, AP8a, AP8b, BP3, BP4a, BP4b, BP4c, BP4c-l, BP4d, BP5,BP6,BP7,BP8a,BP8b,CP1,CP2,CP3,CP4a,CP4b,CP4c, CP4c-l, CP4d, CPS, CP6, CP7, CP8a, CP8b, DP3, DP4a, DP4b, DP4c, DP4c-l, DP4c-2, DP4d, DPS, DP6, 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-1, AP4d, AP5, AP6,AP7,AP8a,AP8b, BP3,BP4a,BP4b,BP4c,BP4c-l,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, DP6, DP7, DP8a, and DP8b.

Piping Brackets PB-015, PB-165, PB-19S, and PB 353.

All NRI.

L2R13 (2011) EVT-1 AP1,AP2,AP3,AP4a,AP4b,AP4c,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,BP5,BP6,BP7,BP4c,BP8b,CP1, CP2,CP3,CP4a,BP4b,CPS,CP6,CP7,CP4c,DP3,DP4a,DP4b, DPS, DPG, DP7, and DP4c. Existing flaw on BPS determined to be geometry. All others NRI. Four new welds identified, (AP4c-1, BP4c-1, CP4c-1, and DP4c*l) located between P4c and P7 (OE28372). All examined by UT with NRI.

EVT*l APl, AP2, AP3, AP4a, AP4b, AP4c, AP4c-l, AP4d, AP8a, APBb, BP3,BP4a,BP4b,BP4c,BP4c-1,BP4d, BPS,BP8a,BP8b,CP1, CP2,CP3,CP4a,CP4b,CP4c,CP4c*l,CP4d,CP8a,CP8b,DP3, DP4a, DP4b, DP4c, DP4c*l, DP4c*2, DP4d, DPS, DP8a, and DP8b. NRI. Five new welds identified, (AP4c-l, BP4c-1, CP4c-1, DP4c-1 and DP4c-2) located between P4c and P7. All examined by EVT-1 with NRI.

L2Rll (2007) EVT*l Visual examinations of core spray piping welds for which the UT ls not demonstrated. No indications (NRI).

- - - - - - -*------ _!.P_!pln~_bra~~!s; NRI. -*-*---**---* ----.-

L2Rl0 (200S) 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*l Welds for which UT ls not demonstrated: APl, AP4d, AP8a, BPS,BP4d,BP8a,CP1,CP4d,CP8a,CP8b,0P4d,DP8a,and OP8b. 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, S2A (left and right), S4A (left and right),

SlB, S2B (left and right), S4B (left and right),

SlC, S2C (left and right), S4C (left and right),

SlD, S2D (left and right), S4D (left and right).

All NRI.

VT-1 S3A-a, S3A-b, and S3A-c from 007 to 088",

S3D-a and S3D-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.

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

53C-a, 53C-b, and 53C-c from 092 to 187".

530-a, and 530-b from 272 to 352".

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

S3B-a, 538-b and 538-c from 172.5 to 268", 53C-a and 53C-b from 187 to 268*; bent sparger nozzle deflector Identified In L2Rll unchanged. All others NRI.

EVT-1 51A, 52A (left and right), 54A, 518, 528 (left and right), 548, 51C, 52C (left and right), 54C, 510, 520 (left and right), 540.

NRI.

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

53A-a, 53A-b and 53A-c from 268 to 7.5", 53A-a, 53A*b and 53A-c from 7.5 to as*, 538-a, 538-b, and 538-c from 172.5 to 268* and 538-a, 538-b, and 538-c from 88 to 172.5"; bent sparger nozzle deflector Identified In L2Rll unchanged. All others NRI.

L2R11 (2007) VT*l 530-a, 530-b, 530-c from 352 to as* and 53C-a, 53C*b and 53C-c from 7.5 to as*; one bent sparger nozzle deflector; all others NRI. Bent nozzle accepted for one cycle.

EVT*l 51A, 52A (left and right), 54A, 518, 528 (left and right), 548, 51C, 52C (left and right), 54C, 510, 520 (left and right), 540.

NRI.

VT*l 6 sp~~~.-~~ack.~m; N~-*- *---- -*

L2R10 (2005) VT*l 53A*a and 53A-b from 268 to ooa* and 530-a and 530-b from 352 to 268*; NRI.

6 spa~~~!_!lrac~ets.!_~~:- * - -

L2R09 (2003) EVT*l 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)

L2Rl5 (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 lnterface 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.

L2R08 (2000) VT-1 Steam dryer attachment welds, four locations, NRI.

VT-3 Gulde Rod attachments at o* and 180"; NRI.

Upper surveillance capsule brackets at three locations; NRI.

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

Core Shroud L2R 16 (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, HG, H7, and HS. Flaws Identified on Hl, H3, and H4. Accepted as-ls with 10 year EOI.

EVT* l Vlsual Inspection (from shroud ID and OD) of vertical welds V12, V13, V14, and VlS. NRI.

Vlsual Inspection (from shroud OD) of horizontal welds Hl and H2; NRI.

L2Rll (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 HS (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 HSa weld at the o* and 1so* locations. NRI.

L2R1S (201S) EVT-1 Access hole cover at O"; NRI.

L2R14 (2013) EVT-1 Top of H9 weld at the o* and 1so* locations. NRI.

Access Hole Cover at lSO". NRI.

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

L2R12 (2009) EVT-1 Access Hole Covers at o and 1so*-- NRI.

L2Rll (2007) VT-3 Access Hole Covers at 0 and 1so* 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 (2005) VT-1 Access Hole Covers at o and 1so*-- NRI.

VT-3 Inspections of the general condition of the RPV Interior surface from the RPV closure flange elevation to the steam dam, 3GO" around the RPV Interior. NRI.

Inspection of the general condition of the cladding at the steam dam elevation, 3GO" 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 HBa 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, 3GO" around the RPV Interior. NRI.

Top Gulde L2R1G (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" locatlons - NRI.

EVT*l Lower Beam and Slot Intersections of 19 cells - NRI.

L2Rll (2007) VT*3 C*clamp at o*. 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 sys~em leak test. NRI.

Page 4

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.

L2Rl3 (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-ls, 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.

L2Rl4 (2013) VT-1 Sensing line brackets on eight pumps; NRI.

Main wedge WD*l 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.

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.

l2Rll (2007) VT-1 WD-1 wedges on all 20 pumps; 7 wedges/rods showed minor wear; accepted-as-ls. 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 fltup 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-1on3 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 locatlons. 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*l AD *2 on 5 pumps; NRI.

DF-1 on S pumps; NRI.

OF -2 on S pumps; NRI.

DF -3 on S pumps; NRI.

IN-1 on S pumps; NRI.

IN* 2 on S pumps; NRI.

Pagc6

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-1 on 5 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-1on10 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; NRI.

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

RB-1 on 5 jct pumps; NRI.

RB-2 on 5 Jet pumps; NRI.

All RS*l welds (Including pup piece welds) on 4 risers. Re-sized flaw on RS*lc on 19/20 with no change In length, accepted for f~!_cyclcs. _Al!_<?th~!.~ NR!:_ ______

Strain relief welds on four risers; NRI.

L2R1S (2015) EVT*l Corners of RS*9 on all 10 risers; NRI.

RS*6 on 1 riser; NRI.

RS* 7 on 1 riser; NRI.

L2R14 (2013) EVT*l RS*9 on 9 risers; NRI.

RS* la on 19/20; NRI.

Re*slzed flaw on RS* le on 19/20; no change In length .

REACTOR INTERNALS INSPECTION HISTORY Plant: LaSalle 2 Date or Components in Inspection Summarize the Following Information: Inspection Results, Frequency of BWRVI P Scope Method Used Repairs, Replacements, Reinspections Inspection Jet Pump Riser L2R13 (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; NRt.

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 Yi 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 L2Rl6 (2017) VT-1 . Ten tie bars; NRI.

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

Pases

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 on the 270" side, seven vertical welds on the 90° 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.

One horizontal weld on the 90" side; NRI.

Upper and lower guide rod bracket at 180° and lower guide bracket at o*. 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.

L2R12 (2009) VT-1 11 vertical welds on the 90" side; all NRI.

6 horizontal welds; all NRI.

Upper and lower guide bracket at 180" and lower guide bracket at o*. All RI as was the top of the guide rod at 180".

Conditions accepted as-ls.

Upper guide bracket at O"; 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-ls.

L2Rl 1 (2007) EVT-1 All welds recommended by BWRVIP-139 and SIL 644 Revision 2 for a curved hood dryer on the 90" 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.

Page9

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 L2Rl0 (2005) VT-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.

VT-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 45-6a,b,c,d at 135"; NRI.

L2R11 (2007) EVT-1 Examined welds45-12a,b,c,d and 45-03a-d coupling at 315";

NRI.

VT-1 Examined weld 45-0Sa,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 45-0Sa,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*l 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, elRht CRDH*l welds, and el5ht RPV* l welds . All NRI.

Page 10

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, HBa, H9, HlO, Hll, Hl2, ICH/RPV-1, and bottom head cladding. NRI for all twenty locations.

L2R09 (2003) VT-3 I 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 locatlons. 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.

L2R13 (2011) UT Eight welds examined by manual UT; no flaws.

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

Category C DM Welds- L2Rl3 (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 spoolplece on which the welds were located was replaced with a spoolplece 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-ls 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 Gulde Rod Cap at ooo*, tack weld Indications accepted as-ls.

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-ls for one cycle.

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

. ~R14 (1:_013) VT-3 Verified ensagcment of surveillance ca ~ sule 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 en11agement of surveillance capsule basket at 030".

L2Rll (2007) VT-3 RcmovJI of the survclllJnce capsule basket from 120* due to a brok<?n spring.

P;igc 11

/SI Program Pla11 LaSalle Co1111ty Statio11 U11its 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-04 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page l of 4)

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

Reference:

IWL-2421 Examination Category: L-B Item Number: L2. I0, 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 JO-Year Interval of the LaSalle County Station, Units I and 2 Containment Inservice Inspection (CISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Re<1uirement IWL-2421 (a) of ASME Section XI allows the test schedule for concrete containment post-tensioning systems for sites with two plants to he modified if the following arc met:

• Both primary containments utilize the same prcstrcssing system and arc essentially identical in design;

• Post-tensioning operations for the two primary containments were completed not more than two years apart, and;

• Both containments arc similarly exposed to or protected from the outside environment.

IWL-242 l(h) of ASME Section XI specifics the modilicd test schedule when the conditions of IWL-2421 (a) arc mer.

4. Reason for Re<1uest In uccordancc with IOCFR50.55a(1.)( I), relief is rcquc1'tcd on the hasis that the propmcd alternative will provide un ucccptahlc level of quality und safety.

On Augu1't K I996, the Nuclear Rcguh1tory Commission (NRC) published u linal rule in the Fede ml Register (FR) (i.e .. 61 FR 41 ~O~) to mncnd IOCFl~50 . 55a. "Codc1' and

~tandard1', to inl'orporntc hy reference Suh..,cction IWL of ASME Section XI.

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/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-04 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 4)

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 I and 2 primary containments are Class CC components.

The amended I OCFR50.55a required incorporation of Subsection JWL into inspection programs by September 9, 200 I. The initial lWL examinations were allowed to be based on the existing (prior to September 9, 1996) post-tensioning system program schedules per IOCFR50.55a(g)(6)(ii)(B)(4) at that time. After establishing this initial Subsection IWL inspection date, subsequent 5-ycar inspections arc 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 (I) both containments utilize the same prestrcssing system and arc essentially identical in design, (2) post-tensioning operations for the two containments were completed not more than two years apart, and (3) hoth containments arc similarly exposed to or protected from the outside environment. LaSalle County Station, Units I and 2, primary containments utilize the same prestrcssing system, arc essentially identical in design. and hoth primary containments urc similarly exposed to or protected from the mllsidc environment.

Regarding the final condition. LaSalle County Station, Unit I. 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 IOCFR50.55a, hy NRC letter dated June 3.

1994 (Reference I), the NRC in Amendment No. 100 for Unit I and Amendment No. 84 for Unit 2 approved the use of the guidance contained in Regulatory Guide 1.35.

"lnscrvicc Inspection of Ungroutcd Tendons in Prcstrcssscd Concrete Containments,"

Revision 3, and the use of the provisions of Survcillm1cc Requirement (SR) 3.0.2.

Additionally. the NRC reviewed LaSalle County Station's rct1uest al that lime 10 treat the Units I and 2 primary containments a~ "twin containme111s" even though the initial Structural In1cgri1y Tests (SITs) were nol within two years of each other as described in Regulatory Guide 1.35. (Note tlml the 2-year period in Regulatory Guide 1.35 wa~ based on the SIT dates whereas the two year period in IWL-24 IO(a) is hased on Post-Tcnsioning Operntion .) The LaSalle County Station. Unit I. initial SIT was performed in December 1978. and the Unit 2 initil1I SIT was performed in June 1983 ( 55 months upmt).

A~ d<K'Umented in RcferenL*c I. the NRC approved this approach balied on a detailed review of dat11 from ~ix llnil I and four Unit 2 inscrvicc in~pcctinn'\. The NRC reviewer~

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IS/ Program Plan IASalle County Station Units I & 2, Fourth lllterval 10CFRSO.SSa Relief Request 14R-04 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Page 3 of 4) 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 I 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 or the LaSalle County Station, Units I and 2, primary containments as "twin containments." Post-tensioning system inspections completed to date have been performed for the Isi, 3n1, 5' 11, I0111, 15'\ 20'11, 25' 11, 30111 and 35 111 years for Unit I, and the Ist, 3rll, 5th, IOt 11, 15 111 , 20111 , 25' 11, and 30111 years for Unit 2, with the tendon and wire/strand tests being completed every other 5-year period. These Post-Tensioning 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 lo acceptable condition. The results or these inspections completed to date demonstrate that the performance or the Unit 2 post-tensioning relative to the Unit I 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 modi lied test schedule of IWL-2421 (h) only if post-tensioning operations for the two primary containments were completed not more than two years apart. Based on the ahove discussion, the proposed alternative will provide an acceptable level of quality and safety in accordance with IOCFR50.55a(z)( I) for the Fourth ISi Interval, as well as the remaining term of the renewed facility operating licenses for LaSalle County Station, Units I and 2.

5. Proposed Alternative and Hasls for Use The modi lied test schedule of IWL-2421 (h) will continue to he used for LaSalle County Station, Units I and 2 tendon tests (L2.10) and wire/strand exmnination!'I (L2.20). The initiation of the IWL-2400 rolling 5-year schedule was hased on the previous inspection dates under the Station Tendon Surveillance Program prior to Subsection IWL heing endorsed and will continue throughout each 120-month interval.

(,, Duration of Proposed Alternative Relief is requested for the rourth ISi Interval. as well us nnd the remaining term of the renewed fm:ility operating licenses (NPF-11 for Unit I and NPr- tX for Unit 2). which currently expire ut midnight on April 17. 2042. und at midnight on Decemher 16. 2043.

for LaSalle County Station, Unit~ I und 2. The Fourth ISi lntervul. us well as the remaining term of the renewed facility operating license~ rcfer!'I to the LaSalle County 1'r11c* i\'11rtlr Cm1.mltin1t. /./.('

IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-04 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 4 of 4)

Station, Units I and 2 current fourth and upcoming fifth and sixth 120-month ISi Program intervals (Reference 2).

7. Precedent LaSalle County Station, Units I and 2, Second CISI Interval Relief Request BR-05 was authorized per NRC Safety Evaluation (SE) dated January 15, 2008 (ADAMS Accession No. ML07352I568). This relief request for the LaSalle County Station, Units I and 2, Third CISl Interval, utilizes a similar approach to the previously approved relief request.

8. References I. Letter from Anthony T. Gody (NRC) to D. L. Farrar (Commonwealth Edison Company), "Issuance of Amendments (TAC Nos. M87305 and M87306)," dated June 3, 1994 (ADAMS Accession No. ML021130097)

2. Letter from Jeffrey S. Mitchell (U.S. Nuclear Regulatory Commission) lo M. P. Gallagher (EGC), "Issuance of Renewed Facility Operating Licenses for LaSalle County Station, Units I and 2 (TAC Nos. MF5347 and MF5346)," dated Octohcr 19, 2016 (ADAMS Accession No. MLl6202A075).

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/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 1 of 9)

• • • NOTE***

LaSalle County Station Fourth ISi Interval Relief Request 14R-05, Revision 0 is simply an administrative placeholder. This relief request was previously submitted and approved under the Second ISi Interval ISi Program Plan as Relief Request CR-38, Revision 0. The approval authorized under Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER) dated January 28, 2004, ADAMS Accession No. ML033300147, for LaSalle County Station, Units 1 and 2, Exelon Generation Company, LLC (EGC), was for permanent relief. This approval applies to the remaining term of the initial operating licenses, which covers only a portion of the Fourth IS I Interval.

Formatting for Relief Request CR-38, Revision 0 varied from the standard ISi Program Plan format due to the fact that it also requested relief from the Augmented Vessel examination contained in IOCFR50a(g)(6)(ii)(A)(2).

The relief request is carried here and renumhered as 14R-05, Revision 0 purely for administrative purposes. All ASME Code references were made in accordance with the 1989 Edition with No Addenda of ASME Section XI. No changes to the actual approved relief request have heen made and no further or revised authorization is required.

(Note: Additional relief to complete the Fot111h ISi Interval will require resuhmittal and approval of this relief request prior to entering the Period of Extended Operation under License Renewal, Anril 17. 2022 for Unit I and Decemher 16, 2023 for Unit 2.)

I. ASME Code Component(s) Affected Code Class: I Examination Category: B-A Item Number: BI. I I

Description:

Alternative Volumetric Examination of RPV Circumferential Shell Welds Component Number: Class I Pressure Retaining RPV Shell Circumferential Welds.

2. Appllcuhle Code Edition and Addenda In accordance with the provisions of IOCFR50.55a(a)(3). LaSalle County Station requests permanent relief for the remaining term of the opernting licenses for Units I and 2 from the following requirement!-.:

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• IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 2 of 9)

a. Volumetric examination of all RPV shell circumferential welds in the Reactor Pressure Vessel in accordance with the requirements of the ASME Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 1989 Edition, Examination Category B-A, Item Number B 1.11.

b. Successive Inspections for RPV shell circumferential welds in accordance with the requirements of the ASME Section XI, 1989 Edition, IWB-2420.

c. Additional Examinations for RPV shell circumferential welds in accordance with the requirements of ASME Section XI, 1989 Edition, IWB-2430.

3. Applicable Code Rec1uirement ASME Section XI 1989 Edition with No Addenda is applicable to the ISi Program for the current interval.

4. Reason for Rec1uest LaSalle County Station requests this relief to reduce the number of circumferential welds requiring inspection as endorsed by Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the Requirements on Reactor Pressure Vessel Circumferential Shell Welds."

5. Proposed Alternative and Hasis for Use I. Alternative Provisions:

In accordance with IOCFR50.55a(a)(3)(i). LaSalle County Station will implement the following alternate provisions for the subject weld examinations. Unless stated otherwise. all references to the ASME C'odc arc to the 1989 Edition with No Addenda of ASME Section XI.

u. lnservke Inspection Scope The failure frequency for ASME Section XI. Table IWB-2500-1, Exmnination Category B-A. Item Number B 1.11. "Reactor Pressure Vessel Shell Circumferential Welds," is sufficiently low to justify their elimination from the ISi requirement of IOCFR50.55a(g) hased on the NRC Safety Evaluation <Reference 2).

The ISi requirements of the ASME Section XI. Table IWB -2500-1.

Examination Ca1egory B-A. Item Number B 1. 12. "Reaclor Pre!>!>lll'e

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 3 of 9)

Vessel Shell Longitudinal," shall be performed, to the extent possible, and shall include inspection of the circumferential approximately 2 to 3% of the RPV shell circumferential welds. EGC believes that when this examination is performed, an automated ultrasonic inspection system will provide the best possible examination of the RPV shell longitudinal welds.

Automatic examinations will be implemented where practical and supplemented by manual examinations to maximize volumetric coverage when necessary. The procedures for these examinations shall be qualified such that flaws relevant to the RPV integrity can be reliably detected and sized, and the personnel implementing these procedures shall be qualified in the use of these procedures. Qualification and examination will be completed in accordance with the 1995 Edition with the 1996 Addenda of ASME Section XI Appendix VIII as modified by the Performance Demonstration Initiative (PDI) and 10CFR50.55(a), "Codes and Standards."

b. Successive Examination of Flaws For ASME Section XI. Tahle IWB-2500-1, Examination Category B-A.

Item Numher B 1.11. "Reactor Pressure Vessel Shell Circumferential Welds," at intersections with longitudinal welds. successive examinations per IWB-2420 "Successive Inspections," arc not required for non-threatening llaws (i.e .. original vessel material or fabrication llaws such as inclusions which exhibit negligible or no growth dming the life of the vessel), provided that the following conditions arc met:

I. The llaw is characteri1.ed as subsurface in accordance with BWRVIP-05 <Reference 1).

2. The NDE technique and evaluation that detected and characteri1.cd the llaw as originating from material manufacture or vessel fabrication is documented in a llaw evaluation report. and

3. The vessel containing the llaw is acceptable for continued service in uccordance with ASME Section XI. IWB-36<X>. "Analytical Evaluation of Flaws," and the llaw is demonstrated acceptable for the intended ~crvice life of the vessel.

For ASME Section XI, Table IWB-25<X>- I. Examination Category B-A, Item Number B 1.12. "Real'tor Pressure Vessel Shell Longitudinal Welds,"

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• UiH74.fHfl6-tl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFRSO.SSa Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l) (10CFRS0.55a(a)(3)(i))

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

Revision 0 (Page 4 of 9) all flaws shall be reinspected at successive intervals consistent with ASME Section and regulatory requirements.

c. Additional Examinations of Flaws For ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item Number B 1.11, "Reactor Pressure Vessel Shell Circumferential Welds," at the intersection with longitudinal welds, additional requirements per ASME Section XI, IWB-2430, "Additional Examinations," arc not required for flaws provided the following conditions arc met:

I. If the flaw is characterized as subsurface in accordance with BWRVIP-05 (Reference I) then no additional examinations arc required.

2. If the flaw is not characterized as subsurface in accordance with BWRVIP-05 (Reference I) then an engineering evaluation shall he performed, addressing the following as a minimum:

• A determination of the root cause of the llaw,

• An evaluation of any potential failure mechanisms,

• An evaluation of service conditions which could cause subsequent failure,

• An evaluation per ASME Section XI. IWB-3600 demonstrating that the vessel is acceptable for continued service.

3. If the llaw meets the criteria of ASME Section XI. IWB-3600 for intended service life of the vessel, then additimrnl examinations may he limited to those welds subject to the root cause conditions and failure mechanisms, up to the number of cxmnination-;

required hy ASME Section XI. IWB -2430(a). If the engineering evaluation determine~ that there arc no additional welds subject to the same root cuusc conditions or no failure mechanism exists.

then no additional examinations arc required.

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/SI Program Plan LaSalle Cottnty Station Units I & 2, Fottrth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 5 of 9)

For ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item Number B 1.12 additional examination for flaws shall be in accordance with ASME Section XI, IWB-2430, "Additional Examinations." All flaws in RPV shell longitudinal welds shall require additional weld examinations consistent with ASME Section XI and regulatory requirements. Examinations of the RPV shell circumferential welds shall be performed if RPV longitudinal welds reveal an active, mechanistic mode of degradation.

II. Basis for Relief The technical basis providing justification for relief from the examination requirements of RPV shell circumferential welds is contained in a report submitted from the BWRVIP to the NRC (i.e., Reference I). The NRC evaluated this report and responses to Requests for Additional Information, and issued Safety Evaluations to the BWRVIP (i.e., References 2 & 5). Additionally, NRC Generic Letter (GL) 98-05 (i.e., Reference 6) permits BWR licensees to request permanent relief from the inservice inspection requirements of IOCFR50.55a(g) for the volumetric examination of RPV shell circumferential welds (ASME Section XI, Tahle IWB-2500-1, Examination Category B-A. Item Numher BI. I I, Circumferential Shell Welds). This relief can he granted hy demonstrating two criteria: (I) at the expiration of their license, the RPV shell circumferential welds will continue to satisfy the limiting conditional failure prohahility for RPV shell circumferential welds that is estahlished in the NRC Safety Evaluation (Reference 2), and (2) licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC Safety Evaluation (Reference 2).

Criterion I, Demonstrute thut ut the explrutlon of the license, the RPV shell clrcumferentlul welds will continue to sutlsfy the limiting condltlonul fullure prohnhllity for RPV shell clrcumferentlul welds thut Is estuhllshed In the .July 30, 1998 Sufety Evuluutlon.

The NRC evaluation of BWRYIP-05 utili1.cd a prohahilistic l'rl1l*turc mechanics (PFM) analysis to estimate the RPV shell weld failure prohahilities. Three key assumptions of the PFM analy~is arc : (I) the neutron llucncc used was the estimated end-of-life mean lluem:c: (2) the chemistry values arc mean values hascd on vcl-.scl types: and (3) the potential for hcyond-<.lcsign-h.isis events is considered.

Tahle I provides u comparison of the limiting RPV circumferential weld parameters for each RPV to tho~c found in Tahlc 2.6-4 of the NRC Final Sakty

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/SI Program Plan lASalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 6 of 9)

Evaluation of BWRVIP-05. LaSalle County Station, Unit I, RPV was manufactured by Combustion Engineering (CE) and Unit 2 RPV was manufactured by Chicago Bridge and Iron Company (CB&l). The CE vessel chemistry limits included in the "NRC Limiting Plant Specific Analysis (32 EFPY)" column arc those from the NRC Safety Evaluation (Reference 2) as reported in the Combustion Engineering Owner's Group (CEOG) Report, Reference 22 of Reference 2. The basis for the LaSalle County Station, Units 1 and 2, copper and nickel values in Table I is the LaSalle County Station response to the NRC Request for Additional Information associated with Generic Letter 92-01, "Reactor Vessel Structural Integrity (Reference 7)."

LaSalle County Station requested a change to the Reactor Coolant System Pressure and Temperature (Pff) Limits Technical Specifications of Units 1 and 2 (Reference 4). The 32 Effective Full Power Year (EFPY) nucnce was calculated using the methodology of NEDC-32983P, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation," which was approved hy the NRC in Reference 3. Additionally, as a conservative assumption, the 32 EFPY llucnces were calculated at the inside surface of the reactor vessel wall using the current uprated power and twenty-four month fuel management designs throughout the entire operational life of LaSalle County Station. The Reference 4 submittal is currently under review hy the NRC, and the lluence results of these calculations were used in this submillal.

The summary of the evaluation of the Unit 1 RPV is shown on Tahlc 1. The copper content for Unit I is higher than the value utilized in the NRC analysis, however. the Unit I nickel content and chemistry factor arc considcrahly lower than the values utilized in the NRC' analysis. The unirradiatcd reference temperature is lower than that used in the NRC analysis. The calculated 32 EFPY lluence for Unit I is considernhly lower than the NRC estimated values. Overnll, the relatively high copper content on Unit I compared with the content used in the NRC' analysis is more than compensated for hy the lower nickel content, unirradiated temperature and lluence: resulting in a lower calculated mean reference temperature than the NRC mean analysis values.

The summary of the evaluation of the Unit 2 RPV is also ~hown on Tahle I.

From the tahle, the Unit 2 chcmbtry compo~ition *md chemistry factor arc lower tlmn the values u~ed in the NRC analysis. The unirradh1ted reference temperature i~ higher than that u~cd in the NRC analysis for Unit 2. The calculated 12 EFPY lluence for Unit 2 is considcrnhly lower th*m the NRC cstim.ited values. For Unit

2. the relatively higher unirradiated reference temperature compared with the NIU' limit ii\ more than compensated for hy lower l'oppcr mul nickel content uml 1'r11e* ,\'11rtl1 C'1111.m/11t11:. /./,(' / ..'iH 7./. f RIJft.IJ./

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFR50.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 7 of 9) fluence, resulting in a lower calculated mean reference temperature than the NRC mean analysis values.

The RPV shell circumferential weld RTNDT due to fluence is calculated to be less for each Unit than the NRC's limiting case and therefore each Unit's RPV shell circumferential weld failure probabilities arc bounded by the conditional failure probability, P(FIE), in Table 2.6-4 of the NRC Safety Evaluation (Reference 2) through the initial end of license.

Table l, Effects of Irradiation on RPV Circumferential Weld Properties LaSalle County Station, Units 1 and 2 Parameter LaSalle Unit l LaSalle Unit 2 NRC Limiting Description Parameters at 32 Parameters at 32 Plant Specific EFPY EFPY Analysis (32 EFPY)

CERPV CB&IRPV CE CB&I RPV RPV Copper, wt. % 0.205 0.04 0.183 0.10 Nickel, wt% 0.105 0.94 0.704 0.99 Chemistry Factor 98 54 172.2 134.9*

End of Life Inside 0.102 0.109 0.20 0.51 Diameter Fluence, X 10 1" n/cm 2

~RTN1rr. "F 41.2 23.5 98.I 109.5 RTNl>T11J1. "F -50 -34 0 -65 Mean RTN11r. "F -8.8 -10.5 98.I 44.5

"' Revised value from the Reference 5 leller.

Criterion 2. Licensees hnve implemented operator trnininit nnd estnhllshed procedures thnt limit the fret1uency of cold overpressure events to the nmount specified in the .July 30. 1998 Snfety Evnluutlon.

EGC has procedures in place for LaSalle County Station, Units I and 2 that guide operntors in controlling and monitoring reactor pre!\sure during all phases of operation, including cold shutdown. U!\e of the!\e procedures will prevent a Low Temperature Over-Pressuri.1.ation (LTOP) event. and ure reinforccJ through operator training. Operating Procedures have !\Uflicient guidance to prevent a LTOP event. A real*tor ves-;cl pres-;ure test i' performed prior to each re!\tart ufter 1'rw* .\'11rtli Ctm.rnltln1:, /./.(

• l.SH7./.1Hflfi.IJ4

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFRS0.55a(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 8 of 9) a refueling outage. This procedure requires an Operations briefing prior to test commencement with all involved personnel. Vessel temperature and pressure are required to be monitored and controlled to within the Technical Specification pressure-temperature (P-T) curve during all portions of testing. The normal and contingency methods to enact pressure control are specified.

A Senior Reactor Operator, who is designated as a Test Coordinator during cold pressure testing, is responsible for the coordination of the test from initiation to conclusion and maintains cognizance or test status. A controlled rate of pressure increase is administratively limited in the test procedure to no greater than 50 psi per minute. If the rate of pressurization exceeds this limit, contingencies arc included to reduce the rate of pressure increase by depressurizing through the Reactor Water Clean Up system and by securing the Control Rod Drive (CRD) pump.

Other than the CRD system, the high pressure coolant sources that could inadvertently initiate and result in a LTOP event arc the Feed water, Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Spray (HPCS) systems.

During a normal RPV till sequence prior to pressure testing, the condensate system is used to fill the reactor. The Motor Driven Reactor Feedpump (MDRFP) is prevented from starting hy the high level Feedwatcr pump trip signal, which is present due to the high reactor water levels required during pressure testing.

During pressure testing, the reactor is in cold shutdown, and as a result, there is no steam availahlc to drive the turhinc driven RCIC and Turhine Driven Reactor Fcedpumps (TDRFPs). The l-IPCS pump control switch is placed in pull-to-lock to ensure an inadvertent initiation will not occur.

Low pressure coolant sources include the Emergency Core Cooling Systems (ECCS) (i.e., Low Pressure Core Spray and Low Pressure Coolant Injection (LPCI) systems) and the Condensate system. The shut-off heads of the ECCS pumps and condensate pumps arc sufllciently low to preclude a LTOP event that would exceed the P-T curve limits due lo an inadvertent low pressure ECCS injection.

During cold shutdown when the reactor head is tensioned. a LTOP event is prevented hy the opernting shutdown procedure, which requires the operator to place the RPV head vent valves in an open position when reactor coolant tcmpcrntmcs arc hclow 2 I 2"F.

In addition to the prnccdurul harrier'\, licensed operutors arc provided spcdfic training on the P-T curve!'> and the u~!'>ociatcd rclluircmcnt!'> of the Technical Trur N11rtl1 C'1111.rn/tit1Jl, /./.(' I ..'\H7-l.f Rfl6*fl4

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-05 (CR-38)

Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l) (10CFR50.55a(a)(3)(i))

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

Revision 0 (Page 9 of 9)

Specifications. Simulator sessions are conducted which focus on plant heat-up and cool-down and equipment surveillance where adherence to these examinations is required. Additionally, in response to industry operating experience and events, the Operations Training instructors and staff routinely evaluate and develop operating training programs to reduce the possibility of events such as LTOP.

In summary, EGC has reviewed the methodology used in BWRVIP-05 (Reference I), and considering LaSalle County Station plant specific materials properties, llucncc and operational practices, and the provisions of the NRC Safety Evaluation Report (Reference 2), EGC believes the criteria established in Generic Letter 98-05 arc satisfied. Therefore, permanent relief is requested from the examination requirements of IOCFR50.55a for Reactor Pressure Vessel circumferential shell welds since the proposed aherm1tivc provides an acceptable level of quality and safety.

6. Duration of Proposed Alternative Permanent relief is requested for the remaining term of the initial operating licenses of Units I and 2.

'/'r111* .\'11rtll C1mrnlti111:. I ./.('

ISi Program Pla11 LaSalle Cou11ty Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-06 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 0 (Pagel of3)

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

Reference:

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

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 I0-Y car Interval of the LaSalle County Station, Units l and 2 lnscrvicc Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition with the 2008 Addenda.

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

4. Reason for Reuuest In accordance with IOCFR50.55a(z)( I). relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

LaSalle County Station, Units I and 2, Technical Specification (TS) Surveillance Requirement (SR) 3.1.5.1 requires each control rod scram accumulator pressure to he equal to or greater than 940 psig for the control rod scram accumulator to he considered operable. The TS SR is required to he met whenever the unit is operating in Modes I and

2. The accumulator pressure is continuously monitored hy system instrumentation and surveillance i), performed on a weekly basis that requires a physical walkdown of all CRD uccumuh1tors. The wulkdown is intended to identify any sy),tem air leaks und negative trending in system pressure. The accumulators arc isolated from the !-.ource of nmkeup nitrogen, thus the continuous monitoring of the CRD accumulators currently functions as u pressure decay type test. The <U:cumulators ure maintuined at a pres!-.tll'c of approximately I IOO psig during operation. Should accumulator pressure fall below IOOO p!-.ig (*I~ P'ig), un alarm is rccl!ivcd in the t*ontrol room. The prC!.'illl'C drop for the 1'rw* i\'11rtli ('1111.mlti111:. /./.(' I ..\'H7./J Rf16.fJ.I

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-06 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 3) associated accumulator is then recorded in the control room log, and the accumulator is recharged by station procedure LOP-RD-20, "Control Rod Accumulator Recharging/Water Removal." Other corrective actions, including soap bubble application to locate leakage or equipment repair are performed, as required, in accordance with the Corrective Action Program.

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 arc normally passive components and arc susceptible to slow developing failure modes. Corrosion and tubing connection integrity arc 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 YT-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 YT-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 mid associated piping on the basis that the requirements of TS SR 3.1.5.1 exceeds the Code-required examinations .

5. Proposed Alternative and Husls for Use As an alternative to the YT-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 *u.:cordance with TS SR 3.1.5.1 that re,1uires a physical walkdown of all CRD accumulators.

6. Duration of Proposed Alternative Relief i~ requested for the Fom1h ISi Interval I'm LuSalle County Station. Units I nnd 2.

1'r111* "'"" ('1111.mltillJ:, /./.(. UiH7./JHtl6°tl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-06 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 3 of 3)

7. Precedents

• LaSalle County Station, Units I and 2, Third ISi Interval Relief Request 13R-09 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 I 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. MLl4141A073).

1'rll<<' ,\'11rtl1 <'1111.mlti111:, /./.( *

/SI Program Plan LaSalle County Statio11 Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-07 Proposed Alternative in Accordance with 10CFRS0.55a(z)(l)

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

Revision 0 (Page l of 3)

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

Reference:

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

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 IO-Year Interval of the LaSalle County Station, Units I and 2 lnservicc Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition with the 2008 Addenda.

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

4. Reason for Reuucst In accordance with IOCFR50.55a(z)( I). 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 "Main Steam Safely Relief Valve Opcrahility" performs opcrahility testing of the Main Steam sat'cty relief valves including the seven relief valves and accumulators per unit that me required to provide automatic dcprcssurization. These surveillances arc performed on a refueling outage frequency as a requirement of LaSalle County Station\ lnservicc Testing ( IST) Progrnm.

One spccitic test that these surveillances perform is a pressure decay test of the ADS accumulator..;, associated piping and valvei.. The pressure decay test is performed hy isolating and pressurizing the ADS accumuh1tors and ussocialed piping to the nominal operating prei.i.ure (i.e .* UX> pounds per square inch. gauge). The decay in prei.i.ure is then monitored through calihrnted pres!-.urc mcusuring instrumentation. If the acceptable prei.,ure dcl*ay criteria arc exceeded. the surveillances identify uppmpri;1tc 1'r111* i\'11rtll C'm1mlti111:. I.I.('

IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-07 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 3) 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 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 YT-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 YT-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 rndiation exposure (estimated 1440 mrem three times per interval for a total of 4320 mrem) without any added henefit in the level of quality and safety.

This inspection would not he consistent with As Low As Reasonably Achievable (ALARA) practices.

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

S. Proposed Alternative and Hasis for Use As an alternate to the examination requirements of Tahle 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 I and 2.

6. Duration of Proposed Alternutlve Relief is requested for the Fourth ISi Interval for LaSalle County Station. Unit:-. I and 2.

7. Precedent LaSalle County Station . Units I and 2, Third ISi Interval Relief Request BR-10 was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated fomrnry :m. 2008 (ADAMS Accession No. ML073610~M7). This relief rcl1ucst for the LaSalle County Station, Units I nnd 2, r:ourth ISi Interval, utilizes a similar approach to the prcviou,ly approved rclicl' l'CllUCsl.

1'r111* .\ '11rtl1 C'm1.mlti111:. /./.( '

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-07 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 3) 1'r11r .\ 'ort/1 ( 'm1mltl111:. /./.(

• l ..'\H 7./.1Hfl6*fl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-08 Proposed Alternative in Accordance with 10CFR50.55a(z)(2)

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

Revision 0 (Page I of 4)

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

Reference:

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

Description:

Alternate Examination Requirements for the Hydrogen Recombincr System Piping Component Number: HG Unit Cross-Tic Piping From check valve I HG007 lo check valve 2HGO 16 From check valve I HGOl6 to check valve 2HG007 From check valve I HG009 to check valve 2HG006B From check valve I HG006B to check valve 2HG009

2. Applicable Code Edition and Addenda The Fourth I0-Y car Interval or the LaSalle County Station, Units I anc.1 2 lnscrvicc Inspection (ISi) Program is hascc.1 on the American Society or Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Coe.le,Section XI. 2007 Edition with the 2008 Addenda.

3. Applicable Code Reuuirement Tahle IWC-2500-1, Examination Category C-H. Item Number C7. I0, requires all Class 2 pressure retaining components he suhject to a system leakage test with a YT-2 visual examination in accordance with IWC-5220. This pressme test is to he conducted once each inspection period.

IWC-5210(h)(2) requires test prncedme to include methods for detection and location of through-wall leakage from the components of the system tested when the pressmizing medium is gas.

4. Reason for Reuuest In accordance with IOCFR50.55a(z)(2). relief is requested on the hasis that compliance with the specitied requirements would result in hardship or unusual diflkulty without .a compensating increase in the level or quality and safety.

Relief is re,1uested from the system pressure test rel1uirements or IWC-5221 and the periodicity requirements of Table IWC-2500-1. as well as the requirement' of 1'r111* N11rtll ('1111.rnlti111:. 1.1.C'

/SI Program Plan l.ASalle County Station Units I & 2, Fourth lllterval 10CFRSO.SSa Relief Request 14R-08 Proposed Alternative in Accordance with 10CFRS0.55a(z)(2)

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

Revision 0 (Page 2 of 4)

IWC-52 IO(b)(2) as applied to the cross-tie piping of the Hydrogen Recombiner System, as depicted in Figure 14R-08. I 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-tic piping, scaffolding will be required because there are long runs of piping located approximately 30 feet overhead. An accumulated dose of 480 mrcm three times per interval for total of 1440 mrcm would be required to perform a leakage test of cross tie piping. Furthermore, this total dose docs 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-tic piping to provide assurance of its structural integrity hy performing pressure test at peak accident pressure and applying a soap huhhlc solution to all pipe welds once per inspection interval.

Necessary scaffolding will he erected and leak detection solution will he applied to the surface of the unit cross-tic piping to the extent required hy IWC-52 IO(h)(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 tic piping welds. The condition of the accessible components as determined hy pressure testing of the acccssihle components once every inspection period in accordance with ASME Section XI rules would he indicative of that of the inaccessible components. Both the accessible and inaccessible components arc designed/constructed to the same requirements and subject to similar operating conditions. Additionally. the I lydrngcn Recomhincrs. including the unit cmss-tic piping. arc functionally tested every refuel outage to verify system temperature, pressure. and llow requirements to further insure system opcrnhility and structural integrity.

Based on the ahovc discussion. reasonable assurance of the unit cross-tic piping structural integrity is achieved hy the performance of the alternate pressure h!sl of piping welds once every inspection interval.

'l'r111* .Vort/1 C'1111rnlti111:. 1.1.l

• IS/ Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-08 Proposed Alternative in Accordance with 10CFR50.55a(z)(2)

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

Revision 0 (Page 3 of 4)

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-tic 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 tic piping welds.

6. Duration of Proposed Alternative Relief is requested for the Fourth ISi Interval for LaSalle County Station. Units I and 2.

7. Precedent LaSalle County Station. Units I and 2, Third ISi Interval Relief Request IJR-11 was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated May 28, 2008 (ADAMS Accession No. ML08 I I90470). This relief request for the LaSalle County Station. Units I and 2. Fourth ISi Interval, utilizes a similar approach to the previously approved relief request.

l'ruc* ,\ '11rtl1 l'1111rnltin1:. I.I.(' l ..'iH 7-1.f Nflfl.fl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-08 Proposed Alternative in Accordance with 10CFRS0.55a(z)(2)

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

Revision 0 (Page 4 of 4)

Figure 14R-08.l HYDROGEN RECOMBINER SYSTEM CROSS-TIE PIPING I

Uld 2+-i __.,.Uni: l 1HG007

~1:--~..........o1o1--1"-~1--~~~........._ I .,,.-~~~--i,..,.._.,.....,..~- 2HG01e -........_ ..............

_.....~, .................

U1*it2 CJrtaimlln:

~tn1~m 1ttGOlllB 1'rt1f' .\'11rth C'm1mltin1:. I .I.( '

IS/ Program Plan LaSalle County Statio11 Units I & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 1 of 15)

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

Reference:

ASME Section XI, Table IW8-2500- J Examination Category: 8-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: NI, N2, N3, N5, N6, N7, N8, N9, N 16, and N 18 (Sec Tables I and 2 for complete list of nozzle identilications)

2. Applicable Code Edition and Addenda The Fourth 10-Year Interval of the LaSalle County Station, Units I and 2 lnservice Inspection (ISi) Program is based on the American Society of Meclmnical Engineers (ASME) Boiler and Pressure Vessel (8PV) 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 modilied hy IOCFR50.55a(h)(2)(x v).

3. Applicuble Code Requirements The applicable requirement is contained in Table IWB-2500-1, "Exmnination Category B-D, Full Penetration Welded Nozzle in Vessels - Inspection Program." Class I Reactor Vessel nozzle-to-vessel weld and noale inner radii examination requirements arc delineated in Item Number 83.90, "Nozzle-to-Vessel Welds," and B3. IOO, "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full pcnclralion welds lo the reactor vessel shell (or head) and integrally cast nou.lcs arc examined each interval.

All of the 1101.1.lc assemblies idcntilicd in Enclosures I and 2 arc full penetration welds.

4. Reuson for Recmcst In accordmll'C with IOC'FR50.55a(1.)( I). relief is rcl1ucstcd from performing the required examinations on 100 percent of the nonle assemblies identified in Tables 5-1 and 5-2 helow {sec Enclosure~ I mul 2 for u list of RPV Exm11ination Category B-D No1.1.les fnr which this relief rcquc~t is upplicahle).

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IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 2 of 15)

The Federal Register Notice (FRN) published November 5, 2014, contains the rulemaking that amends IOCFR50.55a to incorporate by reference Regulatory Guide 1.147, Revision 17, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division I." As stated in the FRN, licensees may use the code cases listed in Regulatory Guide 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 I ,"(Reference I) is listed in Regulatory Guide 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 Regulatory Guide 1.147, Rev 18 (Reference 2). Draft Regulatory Guide 1.147, Rev 18. will be incorporated by reference into the next linal rulemaking of IOCFR50.55a, 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 BWRVJP-108 dated December 19, 2007 (ADAMS Accession No. ML073600374). or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 {ADAMS Accession No. MLl3071A240) arc met. The evaluation dcmonstrnting 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 NRC Re.\'fWll.H'.\' to />uh/it' Co111111e111.\* 011 Drc{/i Rt*gulatory Guide.\', 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 nol.Zle-to-vesscl shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. 1lowever. licensees should demonstrate the plant -

specilic applicability of the BWRVIP-24 I repot1 to their units in the relief reque!'>t by addressing the conditions and limitations specified in Section 5.0 of the NRC SE for BWRVIP-241.

The proposed altenrntive provides un acccptuhle level of 'luality and 1'<1fety ba!>ed on the technical content of BWRVIP-IOX *ind BWRVIP-24 I, a!> endorsed by the NRC' SEs.

'/'mr N11rtlr ('m1rn/1i111:, I.I.< *

/SI Program Pla11 LaSalle Co1111ty Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 3 of 15)

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 I 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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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 4 of 15)

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

Number Examined Two (2) nozzles were Reactor Recirculation inspected in the Third ISi 2 1 Outlet (N l) Interval; No rejcctable indications Three (3) nozzles were Reactor Recirculation inspected in the Third ISi IO 3 Inlet (N2) Interval; No reiectable indications Four (4) nozzles Main Steam inspected in the Third ISi 4 I (N3) Interval; No rejectable indications Two (2) nozzles were Core Spray inspected in the Third ISi 2 1 (NS and N 16) Interval; No re iectahle indications One ( I ) nozzle on Top Nozzles On Top I lead Head (N7) was inspected 3 I (N7. N8 andN18) in the Third ISi Interval:

No rejectahle indications Two (2) nozzles were Jct Pump Instrument inspected in the Third ISi 2 1 (N9) Interval; No rejectahle indications Three (3) nozzles were Residual I !cat Removal inspected in the Third ISi (N 6) 3 I Interval; No rejectahle indications

"'The no1.zle-to-vesscl weld and inner radius examinations arc performed together.

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/SI Program Pla11 lASal/e County Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

Revision 0 (Page 5 of 15)

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

Number Examined Two (2) nozzles were Reactor Recirculation inspected in the Third 2 I Outlet (N l) ISi Interval; No reiectable indications Three (3) nozzles were Reactor Recirculation inspected in the Third 10 3 Inlet (N2) ISi Interval; No reiectable indications Four (4) nozzles were Main Steam inspected in the Third 4 l (N3) IS I Interval; No reiectahle indications One (I) Core Spray nozzle (N5) was Core Spray 2 I inspected in the Third (N5andN16)

IS I Interval; No rejectahlc indications One ( I ) nozzle on Top Nozzles On Top I lead Head (N7) was inspected 3 I (N7. N8. and N 18) in the Third ISi Interval; No reiectahlc indications One ( I ) nm.i'.lc was Jct Pump Instrument inspected in the Third 2 I (N9) ISi Interval:

No rejectahle indications Three (3) non.Jes were Resillual I lcat Removal inspected in the Third (N6) 3 I ISi Interval:

No rejectahle indications

"'The nozzle -to-vessel weld and inner radius examinations arc performed together.

The examination!'. in Tables 5* I aml 5-2 will he schedu1cd in 11Cl.'01'lla111.:e with ASME Section XI. IWB-2411. ln!-.pcction Progrnm.

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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ASME Code Case N-702 stipulates that a VT-I visual examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item Number B3. I00, "Nozzle Inside Radius Section"). EGC may utilize ASME Code Case N-648-1 with associated Regulatory Guide 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.

Electric Power Research Institute (EPRI) Technical Report (TR) I003557, "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," provides the technical basis for ASME Code Case N-702.

BWRVIP-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 arc very low (i.e.,< I x 10-<i 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 hy the NRC in a SER dated December 19, 2007 (ADAMS Accession No. ML073600374) and requires additional criteria to he met in order to apply the technical basis of BWRVIP-108 for the reduction of inspection coverage of the RPV nozzles and nozzle-lo-vessel shell welds.

BWRVIP-108 was supplemented hy EPRI TR 1021005, "BWRVIP-241: Boiling Water Reactor Vessel and Internals Project Prohahilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and was approved hy 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.

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-2..J I report as the technical hasis for the use of ASME Code Case N-702 as an alternative. However. each licensee should demonstrate the plant-specific applicahility of the BWRVIP-241 report to its units in the relief request hy demonstrating that the following general and nozzle-specific criteria arc satisfied:

(I) The maximum RPV heatup/rnoldown rate is limited to less than 115 "f-/hour.

.,.,,,,. N11rtlr <'1111.rnlti111:. /./.('

/SI Program Pla11 LaSalle Cou11ty Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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LaSalle County Station, Units l and 2, Technical Specifications (TS) 3.4. l l, "Reactor Coolant System (RCS) Pressure and Temperature (Pff)

Limits," provides a Surveillance Requirement limiting heatup and cooldown rates lo :S l 00 °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. l .7, "Operating Limits During Healup, Cooldown and Core Operations."

For the recirculation inlet nozzles (N2), the following criteria must be met:

(2) (pr/t)/CRl'V :S 1.15 p = RPV normal operating pressure (psi),

r = RPV inner radius (inch),

l = RPV wall thickness (inch), and CRl'V = 19332; The calculation for the LaSalle County Station, Units 1 and 2, N2 Nozzle results in a maximum value of 1.064, which satislies this criteria.

(3) (p(r} + 1})/(ru2 - rhJICNo/.1.1.E :S 1.47 p = RPV normal operating pressure (psi),

ru =nozzle outer radius (inch),

r, = no1.zle inner radius (inch). and CN0/./1.1'. = 1637; The calculation for the LaSalle County Station, Units 1 and 2. N2 Nozzle results in a maximum value of 1.134, which satislies this criteria.

For the Recirculation Outlet Nou.les (N 1), the following criteria must he met:

(4) (pr/t )/C'1u*v _ 1.15 p = RPV nornml operating pre)\sure (psi),

r = RPV inner radius (inch),

t = RPV wall thickness (inch). and C'1<1'V : 16171; The calculation for the LaSalle County Station. Unit)\ 1 and 2. N 1 No1.1.h:

result)\ in a value of 1.025 for Unit 1 and a value of 1.272 for Unit 2. The

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• l ..'iH 74.f HtJfl.tJ-I

/SI Program Pla11 LaSalle County Station U11its I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-09 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

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Unit I 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),

n = nozzle inner radius (inch), and CNoZZLE = 1977.

The calculation for the LaSalle County Station, Units I and 2, NI Nozzle results in a maximum value of 1.114, which satisfies the criteria.

Based upon the ahove 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 descrihed in this report may he ahle to justify values greater than 1.15.

Because the Unit 2 NI nozzles did not meet the BWRVIP-241 criteria, a hounding analysis was performed to qualify all the Unit 1 and Unit 2 nozzles. The Prohahility 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 lluence issue of certain nozzles in the heh-line region of the reactor vessel. the lluence 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 heltline nozzle has the highest lluence of all the Unit I or Unit 2 nozzles, and although the current licenses for LaSalle County Station, Units I and 2, expire after 40 years, the use of the 60-year tluence provides additiom1l 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 BWRVIP- 108 and RWRVIP-241 were u~ed in the LaSalle County Station analysis. such as the a~sumed number of stress corro~ion initiation and fabrication llaws, the llaw si1.e dist rihut ion, l'tl.'.

The hounding load cases analyzed included the following:

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IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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I. 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 arc shown in the following table.

PoF per year from LTOP PoF per year from Normal events for 25 % Inservice Operating Condition for Maximum Inspection for period of 25 % Inservice 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 I E-3 probability of LTOP occurrence.

• • Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Ruic ( IOCFR50.6 I). NUREG- 1806, Volume I. August 2007.

Since the hounding nozzle, the NI 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 arc hounded hy the Unit 2 NI nozzle analysis. the application of ASME Code Case N-702 to all the Unit 1 and Unit 2 no1.1.lcs listed in Enclosures I and 2 is m:ceptahle.

Therefore. u~e of ASME Code Case N-702 provides an acceptable level of quality and safety in accon.lance with IOCFR50.55a(1.)( I) for all applicable full penetration RPV no1.zle-to-vcssel shell welds and no1.:r.lc 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 I and 2.

<*. Duration of Proposed Altcrnntlvc 1'r11r ,\ '11rtl1 C1111rn/1i111:. /./.('

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-09 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

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Relief is requested for the Fourth ISi Interval, as well as the remaining term of the LaSalle County Station, Units I and 2, renewed facility operating licenses (NPF-11 for Unit land 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 I and 2. The Fourth ISi Interval, as well as the remaining term of the renewed facility operating licenses refers to the LaSalle County Station, Units l and 2 current fourth and upcoming fifth and sixth 120-month ISi Program intervals.

7. Precedents

• LaSalle County Station, Units I and 2, Third ISi Interval Relief Request 13R-14 was authorized per NRC SE dated October 30, 2015 (ADAMS Accession No. ML I5226A4 I 2). This relief request for the LaSalle County Station, Units I 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 ),"dated December 6, 2016 (ADAMS Accession No. MLl6334A440).

• Letter from S.S. Kocnick (U.S. Nuclear Regulatory Commission) to B. C.

Hanson (EGC), "Safety Evaluation of Relief' Requests 14R-02 and 14R-I 0 for the Fourth I0- Ycar Interval of the lnscrvicc Inspection Program for Limerick Station, Units I and 2 (CAC Nos. MF7587 and MF7588)," dated November 21. 2016 (ADAMS Accession No. MLl6301A401).

• Letter from R. J. Pascarelli (U.S. Nuclear Regulatory Commission) to M. E.

Reddemann (Energy Northwest). "Columbia Station - Relief Request for Alternative 41Sl-04 Applicable to the Fourth IO-Year lnservice Inspection Program Interval (CAC No. MF7.B I)," dated October 5 2016 (ADAMS Accession No. MLl6263A233).

'/'rm* :V11rtli ('1111rnltin1:. /./.( ' I ..'\H74. f Hflfl.tJ./

IS/ Program Plan LaSalle County Station Units I & 2, Fourth fllterval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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8. References I. ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division I."

2. Draft Regulatory Guide DG-1296, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division I," dated March 2016 (ADAMS Accession No. MLI 5027 A202).

3. Letter from Jeffrey S. Mitchell (NRC) to M. P. Gallagher (EGC), "Issuance of Renewed Facility Operating Licenses for LaSalle County Station, Units I and 2 (TAC Nos. MF5347 and MF5346)," dated October 19, 2016 (ADAMS Accession No. MLI 6202A075).

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• l ..'\H 74.'1W6-tJ./

IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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ENCLOSURE 1: Aoolicable LaSalle County Station, Unit 1 Nozzles Category Item Nominal Component ID System Number Number Pipe Size NIA Nozzle B-D B3.90 Recirc Outlet 24" NIAIR* B-D 83.100 Recirc Outlet 24" NIB Nozzle B-D B3.90 Recirc Outlet 24" NIBIR B-D 83.100 Recirc Outlet 24" N2A Nozzle B-D B3.90 Recirc Inlet 12" N2AIR B-D 83.100 Recirc Inlet 12" N2B Nozzle B-D B3.90 Recirc Inlet 12" N2BIR B-D 83.100 Rccirc Inlet 12" N2C Nozzle B-D B3.90 Rccirc Inlet 12" N2CIR B-D 83.100 Recirc Inlet 12" N2D Nozzle B-D B3.90 Rccirc Inlet 12" N2DIR B-D B3.IOO Rccirc Inlet 12" N2E Nozzle B-D B3.90 Recirc Inlet 2" N2EIR 8-D 83.100 Recirc Inlet 2" N2F Nozzle 8-D 83.90 Recirc Inlet 2" N2FIR 8-D 83.100 Recirc Inlet 2" N2G Nozzle 8-D 83.90 Recirc Inlet 2" N2GIR 8-D 83.100 Recirc Inlet 2" N2H Nozzle 8-D 83.90 Recirc Inlet 2" N2HIR B-D B3.IOO Recirc Inlet 2" N2J Nozzle B-D B3.90 Recirc Inlet 2" N2J IR B-D B3. IOO Recirc Inlet 2" N2K Nozzle B-D B3.90 Recirc Inlet 2" N2KIR B-D B3.IOO Recirc Inlet 2" N3A Nozzle B-D B3.90 Main Steam 26" N3AIR B-D B3.IOO Main Steam 26" N3B Nozzle B-D B3.90 Main Steam 26" N3BIR B-D B3.IOO Main Steam 26" N3C Noule B-D B3.90 Main Steam 26" N3CIR B-D B3.IOO Main Steam 26" N3D Nozzle B-D B3 .90 - Main Stemn 26"

- 26" N3DIR B-D B3 . IOO--- Main Stemn 12" N5 Nozzle B-D B.l 90 Core Snrn)'.

N5 IR ___ _ _ _B * ~-- - B.l l Q<J__ Core ~m1 y 12" 1'r111* .\'11rtll C'm1rnlti11Jl, /./.( *

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-09 Proposed Alternative in Accordance with 10CFRSO.SSa(z)(l)

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

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ENCLOSURE 1: A 1mlicable LaSalle County Station, Unit 1 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N6A Nozzle 8-D 83.90 LPCI** 12" N6AIR 8-D 83.100 LPCI 12" N68 Nozzle 8-D 83.90 LPCI 12" N68IR 8-D 83.100 LPCI 12" N6C Nozzle 8-D 83.90 LPCI 12" N6CIR 8-D 83.100 LPCI 12" Reactor Core N7 Nozzle 8-D 83.90 6" Isolation Cooling Reactor Core N7 IR 8-D 83.100 6" Isolation Cooling N8 Nozzle 8-D 83.90 Head Vent 4" N8 IR 8-D 83.100 Head Vent 4" Jct Pump N9A Nozzle 8-D 83.90 6" Instrumentation Jct Pump N9AIR 8-D 83.100 6" Instrumentation Jct Pump N98 Nozzle 8-D 83.90 6" Instrumentation Jct Pump N981R 8-D 83.100 6" Instrumentation Nl6 Nozzle 8-D 83.90 Core Spray 12" Nl61R B-D 83.100 Core Spra y 12" N 18 Nozzle B-D 83.90 Spare 6" Nl81R B-D 133.100 Snare 6"

"' IR - Inner Radius

"'* LPCI - Low Pressure Coolant Injection

'/'rm* .\'11rtli C1111.m/1i11R. /./.(' U\H74.f Hflfi.fl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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ENCLOSURE 2: A 1mlicable LaSalle County Station, Unit 2 Nozzles Category Item Nominal Component ID System Number Number Pipe Size NIA Nozzle B-D 83.90 Recirc Outlet 24" NIAIR* B-D 83.100 Recirc Outlet 24" NIB Nozzle B-D 83.90 Recirc Outlet 24" NIBIR B-D 83.100 Recirc Outlet 24" N2A Nozzle B-D 83.90 Recirc Inlet 12" N2AIR B-D 83.100 Recirc Inlet 12" N2B Nozzle B-D 83.90 Recirc Inlet 12" N2BIR B-D 83.100 Recirc Inlet 12" N2C Nozzle B-D 83.90 Recirc Inlet 12" N2CIR B-D 83.100 Recirc Inlet 12" N2D Nozzle B-D 83.90 Recirc Inlet 12" N2DIR B-D 83.100 Recirc Inlet 12" N2E Nozzle B-D 83.90 Recirc Inlet 12" N2EIR B-D 83.100 Recirc Inlet 12" N2F Nozzle B-D 83.90 Recirc Inlet 12" N2FIR B-D 83.100 Recirc Inlet 12" N2G Nozzle B-D 83.90 Recirc Inlet 12" N2GIR B-D 83.100 Recirc Inlet 12" N2H Nozzle B-D 83.90 Recirc Inlet 12" N2HIR 8-D 83.100 Rccirc Inlet 12" N2J Nozzle 8-D B3.90 Rccirc Inlet 12" N2J IR 8-D 83.100 Rccirc Inlet 12" N2K No1.zlc 8-D B3.90 Rccirc Inlet 12" N2KIR 8-D 83.100 Rccirc Inlet 12" N3A Noulc 8-D 83.90 Main Steam 26" N3AIR B-D 83.100 Main Steam 26" N3B No1.1.lc B-D B3.90 Main Steam 26" N3BIR B-D B3.IOO Main Steam 26" N3C Nol.Zic B-D 83.90 Main Steam 26" N3CIR B-D B3.100 Main Steam 26" N3D Nozzle B-D B3 .90 Main Steam 26" N3DIR B-D B3 . IOO Main Steam 26"

---

-- *------- ----- C~reS~--- ----

N5 Nozzle B-D B3 .90 12" N5 IR B-D B3. IOO Core S ~--- -- 12" -

--- --~--------- -

1'mc* ,\'11rtl1 C1111mltin1:. /./.(' l ..'tH7./J Rflfi.fl./

/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-09 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

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

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ENCLOSURE 2: A pplicable LaSalle County Station, Unit 2 Nozzles Category Item Nominal Component ID System Number Number Pipe Size N6A Nozzle 8-D 83.90 LPCI*

• 12" N6AIR 8-D 83.100 LPCI 12" N68 Nozzle 8-D 83.90 LPCI 12" N68IR 8-D 83.100 LPCI 12" N6C Nozzle 8-D 83.90 LPCI 12" N6CIR 8-D 83.100 LPCI 12" Reactor Core N7 Nozzle 8-D 83 .90 6" Isolation Cooling Reactor Core N7 IR 8-D 83.100 6" Isolation Cooling N8 Nozzle 8-D 83.90 Head Vent 4" N8 IR 8-D 83.100 Head Vent 4" Jct Pump N9A Nozzle 8-D 83.90 6" Instrumentation Jct Pump N9AIR 8-D 83.100 6" Instrumentation Jct Pump N9B Nozzle B-D 83.90 6" Instrumentation Jct Pump N9BIR B-D 83.100 6" Instrumentation Nl6 No:t.zle 8-D 83.90 Core Spray 12" Nl61R 8-D 83.100 Core Snra v 12" Nl8 Nozzle 8-D 83.90 Snare 6" Nl81R 8-D 83.100 Spare 6"

• IR - Inner Radius

"'* LPC'I - Low Pressure Coolant Injection 1'r111* Nortli l'1111.rnlti111l. /./.('

/SI Program Plan LaSalle County Station Units I & 2, Fourth /11terval 10CFR50.55a Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFR50.55a(z)(2)

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

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1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel (BPY),Section XI, Class 2 and 3 components that meet the operational and configuration limitations of ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division I,"

paragraphs l(a), l(b), l(c), and l(d).

2. Applicable Code Edition and Addenda The Fourth Ten-Year Interval of the LaSalle County Station, Units I and 2, lnservice Inspection (ISi) Program is based on the ASME BPV Code,Section XI, 2007 Edition with the 2008 Addenda.

The Fourth interval of' the LaSalle County Station ISi Program is currently scheduled to begin on October I, 2017, and end on September 30, 2027, and will comply with the ASME BPV Code,Section XI, 2007 Edition with the 2008 Addenda.

3. Applicable Code Reuuirement IWC-3120 and IWC-3130 of' ASME Section XI. require that flaws exceeding the defined acceptance criteria he corrected hy repair/replacement activities or evaluated and accepted hy analytical evaluation. IWD-3 I20(h) of' ASME Section XI. requires that components exceeding the acceptance standards of' IWD-3400 he subject to supplemental examination or to a repair/replacement activity.

4. Reason for Reuucst In accordance with IOCFR50.55a(z)(2). rclicl' is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Exelon Generation Company. I.LC ( EGC) is requesting a proposed alternative from the requirement to perform repair/replacement .activities for degraded Class 2 uml 3 piping whose maximum operating temperature docs not exceed 200"F and whose maximum operating pre:-.surc docs not exceed 275 psig for LaSalle County Station. Moderntcly dcgrndcd piping could require a plant shutdown within the required action :-.tatcmcnt timcframe:-. to repair observed tk*gradation . Plant shutdown activitie ... result in additional 1'r11t* N11rtl1 C'1111.rnlti1111. 1.1.l ' / ..-;H74.fHll6-ll./

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFR50.55a(z)(2)

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

Revision 0 (Page 2 of 6) dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow EGC to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current Code requirements results in a hardship without a compensating increase in the level of quality and safety.

ASME Code Case N-513-3 docs not allow evaluation of flaws located away from attaching circumfcrcntial piping welds that arc in cl hows, bent pipe, reducers, expanders, and hnmch tees. ASME Code Case N-513-3 also docs not allow evaluation of flaws located in heat exchanger external tuhing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations.

5. Proposed Alternative and Hasis for Use EGC is requesting approval to apply the evaluation methods of ASME Code Case N-5 I3A to Class 2 and 3 components that meet the operational and configuration limitations of ASME Code Case N-513-4, paragraphs I(a), I(h), I(c). and 1(d) for LaSalle County Station in order to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements.

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

"Guidance for Performing Temporary Non-Code Repair of Class I, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptahility of limited degradation in moderate energy piping. The generic leller defines l'onditions that would he al*ceptahle to utilize temporary non-Code repairs with NRC approval. The ASME recognilcd that relatively

!\mall llaws could remain in service withmll risk to the structural integrity of a piping sy!\tcm and developed ASME Code Case N-513 . NRC approval of ASME Code Case N-513 versions in Regulatory Guide 1.14 7, "lnscrvicc Inspection Code c;a!\e Acl'Cplahility. ASME Section XI. Division I," Revision 17 (Rcferenl*e 4). allows acl*eptancc of partial through -wall or thmugh-wall lc:aks for un opl*rating cycle provided all conditions of the code ca-.c and NRC l'ondition-. arc met. The code ca!'.c ulso requires the Owner to demonstrate system opcrahility due to leakage.

'/'r11c* N11rtl1 l'c111rnlti111:, /./.('

/SJ Program Pla11 LaSalle County Statio11 U11its I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFR50.55a(z)(2)

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

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

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

2) Added applicahility to piping elbows, hent pipe, reducers, expanders, and hranch tees where the llaw is located more than (Rut) 112 from the centerline of the attaching circumferential piping weld.

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

4) Revised to limit the use to liquid systems.

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

6) Revised to address treatment of !laws in austenitic pipe llux welds.

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

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

Detailed dis1.*ussion of significant changes in ASME Code Case N-513-4 when compared to NRC approved Code Case N-513-4 is provided in Attachment 4 of the Reference 2 letter.

The de!'iign hasb is conl-iidcred for e<1ch lc<1k mul evalu*1ted using the EGC Opcrahility Evaluation pmccl-i .... The evaluation proccsloi mu ... t consider n.*~1uir1.'ml*n1s or l'Olllmitmcntl-i established for the system, continued degradation und potenti<1l consequences. operating expcricnn'. and l'nginccring judgement. /\!'a rl*quircd hy the code case, the evaluation 1'r11 ,\'11rtll ('m1.rnl1i111t. 1.1.C'

/SI Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFRS0.55a Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFR50.55a(z)(2)

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

Revision 0 (Page 4 of 6) process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding.

Leakage rate is not typically a good indicator of overall structural stability in moderate energy systems, where the allowable through-wall flaw sizes arc often on the order of inches. The periodic inspection interval defined using paragraph 2(e) of ASME Code Case N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size.

The effects of leakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate hy a safety factor of four (4). The critical leakage rate is determined as the lowest leakage rate that can he tolerated and may he based on the allowable loss of inventory or the maximum leakage than can he tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon ASME Code Case N-705 (Reference 3), which is accepted without condition in Regulatory Guide 1.147, Revision 17. Paragraph 2.2(e) of ASME Code Case N-705 requires a safety factor of two (2) on flaw size when estinrnting the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for nonplanar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable hased upon the precedent of ASME Code Case N-705. Note that the alternative herein docs not propose to use any portion of ASME Code Case N-705 and that citation of N-705 is intended only to provide technical hasis for the safety frn.:tor on leakage.

During the temporary acceptance period, leaking flaws will he monitored daily as required hy paragraph 2( f) of ASME Code Case N-513-4 to con Ii rm the analysis conditions used in the evuluation remain valid. Signilicant change in the leakage rate is reason to question that the analysi)\ condition.; remain valid, und would require re-inspeetion per paragraph 2( f) of the code cu)\e. Any rc-inspel*tion mu)\l he performed in accordance with paragrnph 2(a) of the code case.

The leakage limit provide)\ <JUantitative mcasurnhlc limits which cnMirc the opcrnhility of the )\ystcm mtd cmly idcntilkation of i'suc.; that could erode ddcn,c-in-depth and lead to udvcrSl' COllSClllleOCe)\.

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/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFRSO.SSa Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFRS0.55a(z)(2)

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

Revision 0 (Page 5 of 6)

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

6. Duration of Proposed Alternative The proposed alternative is for use of ASME Code Case N-513-4 for Class 2 and Class 3 components within the scope of the code case. An ASME Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first. Relief is requested for the Fourth ISi Interval for LaSalle County Station, Units I and 2, or such time as the NRC approves ASME Code Case N-513-4 in Regulatory Guide 1.147 or other document. If a flaw is evaluated near the end of the interval for LaSalle County Station an<l the next rcf'ucling outage is in the subsequent interval, the llaw may remain in service under this relief request until the next refueling outage.

7. Precedent LaSalle County Station, Units I an<l 2, Third IS I Interval Relief Request 13R- I 9 was authorized by NRC Safety Evaluation (SE) dated September 6, 2016 (Reference 5).

LaSalle County Station Relief Request 13R- I 9 was part of an EGC lleet-wide suhmitt*ll.

and the use of ASME Code Case N-513-4 was authori1.ed for vurious stations whose ISi Program was based on ASME Section XI. 2007 Edition with the 2008 Addenda. This relief request for the LaSalle County Station, Units I and 2. Fourth ISi Interval. utilizes a similar approach to the previously approved relief request.

8. References I) NRC Generic Letter 90-05, "Guidam:e for Performing Temporary Non-Code Repair of ASME Code \hiss I. 2. and 3 Piping (Generic Letter 90-05 ),"dated June 15. 1990 1'r111* .\'11rtll C'1111mltill1t. 1.1.C

• UiH7".fRtl6°tJ.I

IS/ Program Plan LaSalle County Station Units I & 2, Fourth /11terval 10CFRSO.SSa Relief Request 14R-10 Proposed Alternative for Use of ASME Code Case N-513-4, Evaluation Criteria for Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping in Accordance with 10CFRS0.55a(z)(2)

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

Revision 0 (Page 6 of 6)

2) Letter from D. T. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Proposed Alternative to Utilize Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division I," dated January 28, 2016

3) ASME Boiler and Pressure Vessel Code, Code Case N-705, "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and TanksSection XI, Division l ,"dated October 12, 2006

4) NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division l," Revision 17, dated August 2014

5) Letter from G. E. Miller (U.S. Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2; Byron Station Unit Nos. I and 2; Calvert Cliffs Nuclear Power Plant, Units I and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units I and 2; Limerick Station, Units I and 2; Nine Mile Point Nuclear Station, Units I and 2; Oyster Creek Nuclear Station; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2: R. E. Ginna Nuclear Power Plant; and Three Mile Island Nuclear Station, Unit 1 - Proposed Alternative to Use ASME Code Case N-513-4 (CAC Nos. MF7301-MF7322)," dated Septemher6, 2016 (ADAMS Accession No. MLl6230A237>

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision I (Page I of 15)

• • • NOTE***

LaSalle County Station Fourth ISi Interval Relief Request 14R-l l, Revision I is simply an administrative placeholder. This relief request was submitted as an Exelon Generation Company, LLC (EGC) Fleet-wide request under cover letter RS-16-215 on November 2, 2016 (Revision 0). Revision 1 was resubmitted under cover letter RS-17-044 (RAI) on March 13, 2017 (ADAMS Accession No. ML! 7072A385) and was approved for the Fourth ISi Interval as Relief Request 14R- l l, Revision 1. This approval was authorized under Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated June 5, 2017, which for LaSalle County Station covers the Fourth ISi Interval as specified in Table I of the NRC SE (ADAMS Accession No. ML7150A091.)

No changes to the actual approved relief request have been made in the Fourth Interval ISi Program Plan and no further or revised authorization is required.

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

2. Applicable Code Edition and Addenda

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision l (Page 2 of 15)

PLANT INTERVAL EDITION START END Nine Mile Point Nuclear Station.

Third 2004 Edition April 5. 2008 June 15. 2018 Unit 2 Peach Bollom Atomic Power 2001 Edition thmugh 2003 Fourth Novcmhcr 5, 2008 Dcccmhcr 31. 2018 Station, Units 2 and 3 Addenda Quad Citic~ Nuclear Power 2007 Edition thmugh 2008 Fifth April 2. 2013 April I. 2023 Station. Units I and 2 Addenda R. E. Gi1111a Nuclear Power Firth 2004 Edition January I. 20 I0 Dcccmhcr 31. 20 I lJ Plant Three Mile Island Nuclear Fourth 20!14 Edition April 20. 2011 April I lJ, 2022 Station. Unit I

3. Applicable Code Re<1uirement IOCFR50.55a(b)(2)(xx)(B) requires that "The NOE provision in IWA-4540(a)(2) of the 2002 Addenda of Section XI must he applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)( I )(ii) of this section." IWA-4540(a)(2) of the 2002 Addenda of Section XI requires that the nondestructive examination method and acceptance criteria of the 1992 Edition or later of Section Ill he met prior to return to service in order to perform a system leakage test in lieu of a system hydrostatic test. The examination requirements for ASME Section Ill. circumferential butt welds arc contained in ASME Section Ill. Suharticlcs NB-5200, NC-5200, and ND-5200. The acceptance standards for radiographic examination arc spccilicd in ASME Section Ill. Suharticlcs NB-5300, NC-5300, and ND-5300.

IW A-4221 requires that items used for repair/replacement activities meet the applicable Owner's Rct1uircmcnts and Construction Code requirements when performing repair/replacement activities. IW A-4520 requires that welded joints made for installation of items he examined in accordance with the Construction Code identilied in the Repair/Replacement Plan.

4. Reason for Reuuest Replacement of piping is periodically performed in support of the Flow Accelerated Corrosion (FAC) program as well as other repair and replacement activities. The U!'le of enl*oded Phased Array Ultrasonic Examination Techniltlle!'I (PAUT) in lil*u of radiography (RT) to perform the reltllired examinations of the replaced welds would eliminate the !'lafcty ri"ik a"isodatl*d with performing RT, which include~ the planned exposure and the potl'ntial for acddent:il 1x*rsonncl l'Xposure. PAUT minimizes the

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IS/ Program Plan LaSalle County Station Units 1 & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFRSO.SSa(z)(l)

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

Revision I (Page 3 of 15) impact on other outage activities normally involved with performing RT such as limited access to work locations and the need to control system fill status because RT would require a line to remain tluid empty in order to obtain adequate examination sensitivity and resolution. In addition, encoded PAUT has been demonstrated to be adequate for detecting and sizing critical tlaws.

Exelon Generation Company, LLC (EGC) requests approval of this proposed alternative to support anticipated piping repair and replacement activities sturting in the fall 2017 outage season. The duration of the proposed alternative request is for the remainder of the ISi Interval for the plants defined in Section 2 of this relief request.

5. Proposed Alternative and Uasis for Use EGC is proposing the use of encoded PAUT in lieu of the Code-required RT examinations for ASME fcrritic piping repair/replacement welds. Similar techniques arc heing used throughout the nuclear industry for examination of dissimilar metal welds and overlaid welds, as well as other applications including ASME 831.1 piping replacements.

This proposed alternative request includes requirements that provide an acceptahle level of quality and safety that satisfy the requirements of IOCFR50.55a(z)( I). The capahility of the alternative technique is comparnhle to the examination methods documented in the ASME Sections Ill, VIII, and IX, and associated code cases (References I, 3, 5, 6, 8, 9, I0, 11, 12 and 13) related to using ultrnsonic examination techniques for weld acceptance. The examinations will he performed using personnel and procedures qualilied with the requirements of Section 5.1 helow.

The electronic data files for the PAUT examinations will he stored as parl of the archival -

quality records. In addition, hard copy prints of the data will also he included as part of the PAUT examination records lo allow viewing wi1hou11he use of hardware or software.

5.1 Proposed Alternative EGC is proposing to perform encoded PAUT exmnination techniques using demonstrated procedures, equipment und personnel in accordance with the process documented hclow:

(I) The welds lo he examined shall meet the surface conditioning requirements of the demonstrated ultrusonic procedure.

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ISi Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFRS0.55a(z)(l)

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

Revision I (Page 4 of 15)

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

(0.8 mm) gap between the search unit and the scanning surface.

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

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

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

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

(c) A supplemental straight beam examination shall he performed on the volume or base metal through which the angle beams will travel to locate any reflectors that can limit the ability or the angle beam to examine the weld. Detected reflectors that may limit the angle beam examination shall he recorded and evaluated for impact on examination coverage. The straight hcam examination procedure. or portion or the procedure. is required to he qualified in accordance with ASME Section V. Article 4 ;md may he performed using non-encoded techniques.

(5) All detected flaw indications from (4)(a) and (4)(h) ahove shall he considered plmrnr flaws and compared to the preservicc acceptance standards for volumetric examination in accordance with IWB-3000.

IWC-3000 or IWD-3000. Preservice acceptance l\tandards shall he applied. Analytical evaluation for acceptance of flaws in accordance with IWB-3600, IWC-3600 or IWD-3600 i~ permitted for flaws that exceed the applicable acceptance standards and arc confirmed hy surface or volumetric examination to he non -surfol'C connected.

(6) Fh1ws CXl'Ceding the applil*ahlc acccpt<llll'e ~tandards and when analytkal evaluation ha~ not hccn perfnrml*d for lll'Ceptanl*e. shall he reduced to an 1'r11** ,\'11rtlr C'm1.rnlti111:, I./,(' I .SX 74.IHIJfi.fl./

ISi Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision I (Page 5 of 15) acceptable size or removed and repaired, and the location of the repair shall be reexamined using the same ultrasonic examination procedure that detected the flaw.

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

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

(8) A written ultrasonic examination procedure qualified hy performance demonstration shall he used. The qualification shall he applicable to the scope of the procedure, e.g .* flaw detection and/or sizing (length or through-wall height), encoded or non-encoded, single and/or dual side access. etc. The procedure shall:

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

(h) specify which parameters arc considered essential variables, *md a single value, a range of values or criteria for selecting each of the essential variables:

(c) list the cxmnination equipment, including nrnnufocturer and model or series; (d) define the scanning requirements; suc:h as beam angles, Sl'an patterns, hcam direction, maximum scan speed, extent of scanning.

and access; (e) contain a description of th..: calibration method (i ...: .. uctiono; required to ensure that the sensitivity mul accuracy of the signal mnplitude and time output" of the examination !o.Y!-itcm, whcthL*r 1'r11c* ,\ ',,rt/1 C'1111rnltit11t. /./.('

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision l (Page 6 of 15) displayed, recorded, or automatically processed, are repeated from examination to examination);

(f) describe the method and criteria for discrimination of indications (e.g., geometric indications versus indications of flaws and surface versus subsurface indications); and (g) describe the surface preparation requirements.

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

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

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

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

(C) The demonstration set shall include specimens not thicker than 0.1 in. (2.5 mm) more than the minimum thickness, nor thinner than 0.5 in. (I J mm) less than the maximum thickness for which the examination procedure is applicahle. The demonstration l'ICI slrnll include the minimum. within 1/1 inch of th c nominal pipe si1.e ( NPS ), and maximum pipe diameters for which the exmnination prol*cdurc is applic11hlc. If the procedure is applicahle to outside diameter (O.D.) piping of 24 in . (600 111111) or larger, the specimen set must include at lcul'll one specimen 24 in. O.D. (600 111m) or h1rger hut need not include the maximum diameter.

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IS/ Program Pla11 LaSalle Co1111ty Statio11 U11its 1 & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision 1 (Page 7 of 15)

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

(e) The demonstration specimen set shall include geometric conditions that require discrimination from flaws (e.g.,

counterbore, weld root conditions, or weld crowns) and limited scanning surface conditions for single-side access, when applicable.

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

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

(i) Flawed grading units shall he the actual flaw length, plus a minimum of 0.25 in. (6 mm) on each end of the flaw .

Unllawed grading units shall he at least I in. (25 mm).

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

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

(i) For thickness greater than 0.50 in. ( 13 mm); at least 20'h*

of the llaws shall he distributed in the outer third of the specimen wall thickness, at least 20% of the llaws shall he distributed in the middle third of the specimen wall thickness and at least 40% of the flaws slrnll he di'\trihuted in the inner third of the specimen wall thickness. For thickness 0.50 in. (I 3mm) and less, at least 20% of the llaws shall he distributed in the outer half of the spe'-*imen wall thickness and ut least 40'h* of the llaws shall he distributed in the inner half of the specimen wall thickness.

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JS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFRS0.55a(z)(l)

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

Revision 1 (Page 8 of 15)

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

of the flaws shall be classified as subsurface planar flaws in accordance with IW A-3320.

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

(iv) The flaw through-wall heights shall be based on the applicable acceptance standards for volumetric examination in accordance with lWB-3400, IWC-3400 or IWD-3400. At least 30% of the flaws shall be classified as acceptable planar flaws, with the smallest flaws being at least 50% of the maximum allowable size based on the applicable a/I aspect ratio for the flaw.

Additional smaller flaws may he included in the specimens to assist in establishing a detection threshold, hut shall not he counted as a missed detection if not detected. At least 30'.:i' of the flaws shall he classified as unacceptable in accordance with the applicable acceptance standards. Welding fabrication flaws arc typically confined to a heigl1 of a single weld pass. Flaw through- wall height distribution shall range from approximately one to four weld pass thicknesses, based on the welding process used.

(v) If applicahlc. at least two flaws, hut no more than 30'fi, of the flaws, shall he oriented perpendicular to the weld fusion line and the remaining flaws shall he circumferentially oriented.

(vi) For demonstrnl ion of si nglc-side-access capahi Iit ies, at least 30'.:i* of the flaws slrnll he located on the far side of the weld centerline and at least 30'it* of the planar flaw!'. shall he located on the near side of the weld l'enterline. The remaining flaws shall he distributed on eithl'r 'iidl' of the weld.

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JS/ Program Plan LaSalle County Statio11 Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision I (Page 9 of 15)

(I 0) Ultrasonic examination procedures shall be qualified by performance demonstration in accordance with the following requirements.

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

(b) The non-blind performance demonstration is used to assist in optimizing the examination procedure. When applying the non-blind performance demonstration process, personnel have access to limited knowledge of specimen naw information during the demonstration process. The non-blind performance demonstration process consists of an initial demonstration without any flaw information, an assessment of the results and feedback on the performance provided to the qualifying candidate. After an assessment of the initial demonstration results, limited flaw information may be shared with the candidate as part of the feedback process to assist in enhancing the examination procedure to improve the procedure performance. In order to maintain the integrity of the specimens for hi ind personnel demonstrations. only generalities or the flaw information may he provided to the candidate. Procedure modilications or enhancements made to the procedure. hased on the feedback process, shall he applied to all applicable specimens based on the scope of the changes.

(c) Objective evidence of a flaw's detection, length and through-wall height sizing. in accordance with the procedure requirements. shall he provided to the organi1.atinn administering the performance dcmlmstral ilm.

(d) The procedure demonstration specimen sci shall he representative of the procedure scope and limitations (e.g .. thkkness range.

dimncter rnnge. nmtcrial, access, surface condition).

(c) The demonstration set shall include specimens to repre'ient the minimum and nmximum dimneter ;md thickness covered hy the procedure. If the procedure spmt'i u rnnge of diameters and thicknesses, mldition;il specimens shall he included in the set 10 demonstrate the effectiveness of the procedure throughout the entire range.

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/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision I (Page IO of 15)

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

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

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

(DN 150) and snmller, and 0. 75 in. ( 18 mm) for diameters greater than NPS 6.0 in. (DN 150).

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

(iv) RMS error shall he calculated as follows:

RMS=

n where:

111, =mea~ured tlaw ~i1.e n = number of flaws measured t, = true tfaw ~i1e 1'r11e* ,\'11rt/1 l'1111.rnlti111l. I.I.<* I.SH 7./.f Rflfi.fl./

/SI Program Pla11 LaSalle County Station Units I & 2, Fourth Interval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision I (Page 11 of 15)

(h) Essential variables may be changed during successive personnel performance demonstrations. Each examiner need not demonstrate qualification over the entire range of every essential variable.

( 11) Ultrasonic examination personnel shall be qualified in accordance with IW A-2300. In addition, examination personnel shall demonstrate their capability to detect and size llaws by performance demonstration using the qualified procedure in accordance with the following requirements:

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

(i) The demonstration specimen set shall contain at least 10 flaws and shall meet the llaw distribution requirements or (9)(h) above, with the exception of (9)(h)(v). When applicable, at least one llaw, hut no more than 20% of the llaws, shall he oriented perpendicular to the weld fusion line and the remaining llaws shall he circumf'crentially oriented.

(h) Personnel performance demonstration acceptance criteria:

(i) To he qualified for flaw detection, personnel performance demonstration shall meet the requirements of the following table for hoth detection <md false calls.

Any non-flaw condition (e.g .* geometry) reported as a flaw shall he considered a false call.

1'r111* i\'11rtl1 C'rm.rnltl1111, /,/.('

ISi Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFRSO.SSa(z)(l)

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

Revision I (Page 12 of 15)

Performance Demonstration Detection Test Acceptance Criteria Detection Test Acceptance Criteria False Call Test Acceptance Criteria No. of Flawed Minimum No. of Unflawed Maximum Number Grading Units Detection Criteria Grading Units of False Calls 10 8 15 2 11 9 17 3 12 9 18 3 13 10 20 3 14 10 21 3 15 11 23 3 16 12 24 4 17 12 26 4 18 13 27 4 19 13 29 4 20 14 30 5 Note I: Flaws ~ 50% of the maximum allowahle size. hased on the

• 1pplkahle al( aspect ratio for the llaw.

(ii) To he qualified for flaw length sizing. the RMS error of the llaw lengths estimated hy ullrasonics. as compared with the true lengths, shall not exceed 0.25 in. (6 mm) for NPS 6.0 in. (DN 150) and smaller. and 0.75 in. ( 18 nun) for diameters larger than NPS 6.0 in. (DN 150).

(iii) To he qualified for flaw through-wall height sizing. the RMS error of the llaw through-wall heights estimated by ultrasonks. as compared with the trne through-wall heights, shall not cxl.*eed 0.125 in. (3 nun).

( 12) Documentation of the qualificution~ of prm:edures and per~onnel shall he maintained. DocumenlHtion shall include identilication of personnel, NDE procedures. equipment und spcl.'imcns u~ed during lllmlilicution. und results of the performance demonstration.

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/SI Program Plan LaSalle County Station Units 1 & 2, Fourth /11terval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision l (Page 13 of 15)

( 13) The pre-service examinations will be performed per ASME Section XI (Reference 4).

5.2 Basis for use The overall basis for this proposed alternative is that encoded PAUT is equivalent or superior to RT for detecting and sizing critical (planar) flaws. In this regard, the basis for the proposed alternative was developed from numerous codes, code cases, associated industry experience, articles, and the results of RT and encoded PAUT examinations. The examination procedure and personnel performing examinations arc qualified using representative piping conditions and flaws that demonstrate the ability to detect and size flaws that arc both acceptable and unacceptable to the defined acceptance standards. The demonstrated ability of the examination procedure and personnel to appropriately detect and size flaws provides an acceptable level of quality and safety alternative as allowed by IOCFR50.55a(z)( I).

6. Duration of Proposed Alternative This relief request will be applied for the duration of the inservicc inspection intervals defined in Section 2 of this relicf request.

7. Precedents

• Oconee Request for Rel icf No. 2006-0N -0 I, dated February 2, 2006, requested an alternative for examination of hull welds between the Pressurizer Level and Sample Tap nozzles and their respective Safe Ends. The reason for the request was hased on the diniculty to perform the Code-required radiography.

The alternative was to perform ultrasonic examination per similar requirements to ASME Code C'ase N-659-1. (ADAMS Accession No. Ml.060450464).

• Wolf Creek IOCFR50.55a Request ET 06-0029, dated Septemher I. 2006, requested an alternative for examination of Main Steam and Feedwater piping welds being replaced due to llow assisted corrosion. The reason for the request was hased on the ucccptahility of the proposed ultrasonk examination alternative process. radiation exposure reduction. outage costs and duration. and radiogrnphycxpo'\ure ri'\k . (ADAMS Accession No. ML06250009-'>.

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IS/ Program Plan LaSalle County Station Units I & 2, Fourth Interval 10CFRSO.SSa Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFRSO.SSa(z)(l)

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

Revision 1 (Page 14of 15)

• Palo Verde Nuclear Station Relief Request 48, dated August 1, 2012, ADAMS Accession No. ML12229A046. NRC approval dated April 12, 2013, ADAMS Accession No. ML13091Al77.

• Millstone Power Station Unit 2 Alternative Request RR-04-16, dated August 1, 2013, ADAMS Accession No. ML13220A019. NRC approval dated April 4, 2014, ADAMS Accession No. ML14091A973.

• Millstone Power Station Units 2 and 3 Alternative Requests RR-04-21 and IR 25, dated October 6, 2014, ADAMS Accession No. ML l 4283A 128. NRC approval dated September 21, 2015, ADAMS Accession No. ML15257A005.

8. References

1. ASME Section Ill Code Case N-659-2, "Use of Ultrasonic Examination in Lieu of Radiography for Weld Examination Section Ill, Divisions 1 and 3," dated June 9, 2008.

2. Pacific Northwest National Laboratory Report PNNL-19086, "Replacement of Radiography with Ultrasonics for the Nondestructive Inspection of Welds -

Evaluation of Technical Gaps -An Interim Report," dated April 2010.

3. ASME 831. I Case 168, "Use or Ultrasonic Examination in Lieu of Radiography for 831.1 Application," dated June 1997.

4. ASME Section XI Editions and Addenda applicable to every site.

5. ASME Section Ill Code Case N-818, "Use of Analytical Evaluation approach for Acceptance of Pull Penetration Butt Welds in Lieu of Weld Repair," dated December 6. 2011.

6. ASME Code Case 2235-9. "Use of Ultrasonic Examination in Lieu of Radiography Section I, Sel*tion VIII. Division!>. I and 2, and Sel*tion XII," dated October 11, 2005.

7. Jourrrnl of Pressure Vessel Technology, "Technical Basis for ASME Section VIII Code Case 2235 on llltrnsonic Exumination or Welds in Lieu of Radiogrnphy:"

Runa. l lcdden. Cow fer. und Boycl'. Volume 12.'. <lated August 200 I.

'/'rue* ,\ '11rtl1 Ccm.rnltitlllo I ./.('

/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 10CFR50.55a Relief Request 14R-11 Proposed Alternative for Use of Encoded Phased Array Ultrasonic Examination Techniques In Lieu of Radiography in Accordance with 10CFR50.55a(z)(l)

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

Revision l (Page 15of 15)

8. ASME Code Case 2326, "Ultrasonic Examination in Lieu of Radiographic Examination for Welder Qualification Test CouponsSection IX," dated January 20, 2000.

9. ASME Code Case 2541, "Use of Manual Phased Array Ultrasonic Examination Sect ion V," dated January 19, 2006.

I0. ASME Code Case 2558, "Use of Manual Phased Array E-Scan Ultrasonic Examination per Article 4 Section V," dated December 30, 2006.

11. ASME Code Case 2599, "Use of Linear Phased Array E-Scan Ultrasonic Examination per Article 4 Section V," dated January 29, 2008.

12. ASME Code Case 2600, "Use of Linear Phased Array S-Scan Ultrasonic Examination per Article 4 Section V," dated January 29, 2008.

13. ASME Code Case N-713, "Ultrasonic Examination in Lieu of Radiography Section XI, Division I," dated November I 0, 2008.

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/SJ Program Plan LaSalle County Station Units I & 2, Fourth Interval

9.0 REFERENCES

The references used to develop this ISi Program Plan include:

9.1 NRC References 9.1.1 Code of Federal Regulations, Title I0, Part 50, "Energy."

a. Paragraph 50.55a, "Codes and Standards."

b. Paragraph 2, "Definitions," the definition of "Reactor Coolant Pressure Boundary."

c. Appendix J, "Primary Reactor Containment Testing for Water Cooled Power Reactors."

9.1.2 NRC Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division I" (Sec NRC.gov Reading Room for the most current revision.)

9.1.3 NRC Regulatory Guide 1.150, Revision I, "Ultrasonic Testing of Reactor Vessel Welds During Prcservicc and lnscrvicc Examination."

9.1.4 NRC Regulatory Guide 1.193, "ASME Code Cases Not Approved for Use" (Sec NRC.gov Reading Room for the most current revision.)

9.1.5 NRC Regulatory Guide 1.26, Revision 3, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive Waste- Containing Components of Nuclear Power Plants."

9.1.6 NRC NUREG-0313, Revision 2, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,"

dated January 1988.

9.1.7 NRC NUREG-0519, "Safety Evaluation Report related to the operation of LaSalle County Station Units I and 2," dated March 1981.

9.1.8 NRC NUREG-0578, "TMl-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," dated July 1979.

9.1.9 NRC NUREG-0619, "BWR Fecdwatcr Nozzle and Control Rod Drive Return Linc Nozzle Cracking," dated Novemher 1980.

9.1.10 NRC NUREG-0737, "Clarification of TMI Action Plan Requirements,"

dated Novemher 1980.

9.1.11 NRC NUREG-180 I. "Generic Aging Lessons Learned."

9.1.12 NRC Mechanical Engineering Branch ( MEB) Technical Position 3-1 (MEB 3-1 ), "lligh Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," dated November 24, 1975.

9.1.13 NRC Final SER related to "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-0 I Inspection Schedules BWRVIP-75), EPRI Report TR -113932, October 1999," (TAC NO.

MA5012), dated May 14. 2002.

9.1.14 NRC Final SER related h~ "BWR Vessel und Internals Project, Technical Basi~ for Revisions to Generic Lcller 88-0 I Inspection Schedules (BWRVIP-75-A). EPRI Report TR-1012621. October 2005," dated March 16, 2006.

1'r11c* ,\ 111rtl1 C'1111.rnltl111:. I .I.< ' 9./ UiH7./JRIJ6.IJ./

H,.1*i.d1111 II

ISi Program Plan 1.ASalle County Station Units I & 2, Fourth Interval 9.1.15 NRC Final SER related to the "Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A71-0594, 'Alternate BWR Feedwater Nozzle Inspection Requirements, August 1999,' (TAC No. M94090),"

dated June 5, 1998.

9.1.16 NRC Final SER related to the "Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A 71-0594-A, Revision I, 'Alternate BWR Feedwater Nozzle Inspection Requirements, May 2000,' (TAC No.

MA6787)," dated March IO, 2000.

9.1.17 NRC Final SER related to the "BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05), EPRI Report TR-105697, September, 1995," dated July 28, 1998.

9.1.18 NRC Final SER related to EPRI Topical Report TR-112657, Rev. B, Final Report, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 1999," dated October 28, 1999.

9.1.19 NRC Final SER related to EPRI Topical Report TR-1006937, Rev. 0, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISi)

Methodology to Break Exclusion Region (BER) Programs," dated June 27, 2002.

9.1.20 SECY-96-080, "Issuance of Final Amendment to l OCFR50.55a to Incorporate by Reference the ASME Boiler and Pressure Vessel Code,Section XI, Division I," Subsections IWE and IWL.

9.2 Industry References 9.2.1 ASME Boiler and Pressure Vessel Code,Section XI. Division I.

"lnservice Inspection of Nuclear Power Plant Components."

a. 2007 Edition with the 2008 Addenda (including Appendix VIII)

(Subsections IWE and IWL for CISI) (4th ISi Interval and 3rd CISI Interval).

b. 200 I Edition through the 2003 Addenda (Subsections IWE and IWL for CISI) (3rd ISi Interval and 2nd CISI Interval).

c. 2001 Edition with No Addenda.

d. 1998 Edition with No Addenda (Subsections IWE and IWL for CISI) (1st CISI Interval Final).

e. 1995 Edition through the 1997 Addenda.

r. 1995 Edition with the 1996 Addenda (4th ISi Interval).

g. 1992 Edition with the 1992 Addenda (Subsections IWE and IWL for CISI) (Isl CISI Interval Original).

h. 1989 Edition with No Addcmla (2nd ISi lluerval ).

1. 1980 Edition with Addenda through the Winter 1980 (80W80) (Isl ISi Interval).

9.2.2 ASME Boiler und Pres~urc Ve~scl Code,Section V. "Nondestructive Exmuirrntion,"2007 Edition with the 2008 Addend*** !The Edition <llld Addend** for ASME Section V ure the same a" the Edition mu.I Addenda of ASME Section XI used for the inspl*ction interval for both ISi and Non-ISi NDE examination~. Reference ASME lntl*rpretation Xl 89-021 .

.,.,,, ** "'"" C'c111.rnlti111:. /,/,(' 9-1 UiH 7".fRIJfi.IJ./

Rf'1*/.fim1 fl

/SI Program Plan LaSalle County Station Units I & 2, Fourth Interval 9.2.3 ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants,2004 Edition through the 2006 Addenda (Subsections ISTA and ISTD) (4th Snubber Interval).

9.2.4 Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A 71-0594-A, Revision I, "Alternate BWR Feedwater Nozzle Inspection Requirements," dated May 2000.

9.2.5 Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A 71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements," dated August 1999.

9.2.6 BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-0 I Inspection Schedules (BWRVIP-75-A), EPRI Report TR-I 012621, dated October 2005.

9.2.7 BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), EPRI Report TR-113932, dated October 1999.

9.2.8 Generic Letter 88-0 I, Revision 2, "NRC Position on lntergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping,"

dated January 25, 1988.

9.2.9 Generic Letter 88-0 I, Supplement I, "NRC Position on lntergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," dated February 4, 1992.

9.2.10 Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds,"

dated Novemher I 0, 1998.

9.2.11 BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05), EPRI Report TR-105697, dated Septemher, 1995.

9.2.12 BWR Vessel and Internals Project Program Implementation Guide (BWRVIP-94). EPRI Report TR-1011702, dated Decemher 2005.

9.2.13 EPRI Topical Report TR-112657. Rev. B-A. Final Report, "Revised Risk-Informed lnservice Inspection Evaluation Procedure, dated Decemher 1999.

9.2.14 EPRI Topical Report TR-1006937. Rev. 0-A. "Extension of the EPRI Risk -Informed Inservice Inspection (RI-ISi) Methodology to Break Exclusion Region (BER) Programs," dated August 2002.

9.2.15 EPRI Containment Inspection Program Guide (TR-110698-R I).

9.2.16 INPO Engineering Program Guide EPG-11. "lnservice Inspection Program."

9.3 Licensee Rclcrences 9.3. I LaS.ilk County Station Units I and 2. Final Safety Analysis Report (FSAR) 9.3.2 LaSalle County Station Unit!-1 I <llld 2. Updaf\.*d Fimtl S.ifcty Analysis Report (UFSAR).

9.J .." LaSalk County Station Units I and 2. Technical Spccilicatinn!-1 (TS).

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/SI Program Plan LaSalle County Station Units I & 2, Fourth lllterval 9.3.4 LaSalle County Station Units 1 and 2, Technical Requirements Manual (TRM).

9.3.5 LaSalle County Station Units I and 2, ISi Classification Bases Document (SL-4829), Second Ten-Year Inspection Interval.

9.3.6 LaSalle County Station Units I and 2, ISi Classification Basis Document (LS8743R06-03 ), Fourth Ten-Year Inspection Interval.

9.3.7 LaSalle County Station Units I and 2, ISi Selection Document (LS8743R06-05), Fourth Ten-Year Inspection Interval.

9.3.8 LaSalle County Station Units l and 2, EGC Risk-Informed Inservice Inspection Evaluation (Final Report) (LS8743R06-06) for LaSalle County Station Units I and 2.

9.3.9 LaSalle County Station Units l and 2, Snubber Program Document (LS8743R06-07), Fourth Ten-Year Inspection Interval.

9.3.10 Sargent & Lundy External Design Information Transmittal (EDIT)

DIT-LS-EXT-0045-1. This EDIT transmitted Calculation ATD-0204, Revision 4, "Makeup Calculation in Conjunction with Paragraph IWB-1220, ASME Section XI."

9.3.11 LaSalle County Station Letter RA06-062 from Susan R. Landahl (Site Vice President) to NRC (Document Control Desk), "Inservice Inspection ISi Intervals," dated September 22, 2006.

9.3.12 Structural Integrity Associates, Inc. File No. 1400187.30 I, Revision l, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the Unit 2 NI Nozzle," dated February 6, 2015.

9.3.13 Structurnl Integrity Associates, Inc. File No. 1400187.302, Revision 2, "Probability of Failure for LaSalle Unit 2 NI Nozzle-to-Shell-Welds and Nozzle Blend Radii Regions," dated May 6, 2015.

9.3.14 LS-LAR-007, Revision 0, "LaSalle Station, Units I am.I 2, PRA Capability Assessment for RI-ISi, (Summary: LS-LAR-007 PRA Capability Assessment for Risk-Informed lnservice Inspection Applications)," dated fonuary 2017.

9.3.15 LaSalle County Station Procedures LAP-100-14, "Leak Reduction Program," LTS-300-7, "Leakage Reduct ion and Control Progrnm," LOP-RD-20, "Control Rod Accumulator Recharging/Water Removal," LTS-500-18, "Unit I Main Steam Safety Relief Valve Opernhility," and LTS-500-19, "Unit 2 Main Steam Safety Relief Valve Opernhility."

9.3.16 Procedures ER-AA-330, "Conduct of lnservice Inspection Activities,"

ER-AA-330-<X>I. "Section XI Pressure Testing," ER-AA-330-002, "lnservice Inspection of Section XI Welds and Components," ER-AA-330-<XH. "lnservice Inspection of Section XI Component Supports," ER-AA-330-<K>S. "Visual Examination of Section XI Class CC Concrete Containment Structures," ER-AA-330-<K>6. "lnservice Inspection und Testing of The Pre-Stressed Concrete Containment Post Tcmioning Systems," ER-AA-330-007, "Vism1I Examination of Section XI Cla~s MC Smfoccs and Class CC Liner!'.," ER-AA-.H0-009, "ASME Section XI Repair/Replacement Progrnm," ER-AB-331. "BWR Internals Pmgrnm 1'rm* .\'11rtl1 ( '1111.rnlti111t. /./.( * /~-;H74.flW6-114 Hr1*i.ti1111 II

/SI Program Plan LaSalle County Station Units I & 2, Fourth fllterval Management," and LTS-600-8, "Reactor Vessel Internals Inservice Inspection During Reactor Refueling."

9.4 License Renewal References/Commitments 9.4.1 LaSalle County Station Units I and 2, "Operating License Renewal Application," December 9, 2014.

9.4.2 NRC Final SER "Safety Evaluation Report - Related to the License Renewal of the LaSalle County Station, Units I and 2 - Docket Nos. 50-373 and 50-374 - Exelon Generation Company, LLC (NUREG-2205),"

dated September 30, 2016.

9.4.3 CM-1, Action Tracking AR 1603204-01, LSCS License Renewal Commitment, ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD Aging Management Program.

9.4.4 CM-2, Action Tracking AR 1603204-03, LSCS License Renewal Commitment, Reactor Head Closure Stud Bolting Aging Management Program.

9.4.5 CM-3, Action Tracking AR 1603204-05, LSCS License Renewal Commitment, BWR Fcedwater Nozzle Aging Management Program.

9.4.6 CM-4, Action Tracking AR 1603204-06, LSCS License Renewal Commitment, BWR Control Rod Drive Return Linc Nozzle Program.

9.4.7 CM-5, Action Tracking AR 1603204-07, LSCS License Renewal Commitment, BWR Stress Corrosion Cracking Aging Management Program.

9.4.8 CM-6, Action Tracking AR 1603204-08, LSCS License Renewal Commitment, BWR Penetrations Aging Management Program.

9.4.9 CM-7, Action Tracking AR 1603204-23, LSCS License Renewal Commitment, Unit I One-Time Inspection of ASME Code Class I Small-Bore Piping Aging Management Program.

9.4.10 CM-8. Action Tracking AR 1603204-48, LSCS License Renewal Commitment. Unit 2 One-Time Inspection of ASME Code Class I Small -

Bore Piping Aging Management Program.

9.4.11 CM-9, Action Tracking AR 1603204-29. LSCS License Renewal Commitment, ASME Section XI. Subsection IWE Aging Management Program (Steps 1.1. 6.0. 6.1. 6.2. and 6.3).

9.4.12 CM-JO, Action Tracking AR 1603204-30. LSCS License Renewal Commitment, ASME Section XI. Subsection IWL Aging Management Program (Steps 1.1. 6.0. 6.1, 6.2, and 6.3 ).

9.4.13 CM-11. Al*tion Tracking AR 160320..t-3 I. LSCS License Renewal Commitment. ASME Section XI. Subsection 1w1: Aging Mairngement Progrnm (Steps I. I, 4.0 (entire section). and Tahles 7.1-1 and 7.1 -2).

1'rm* ,\'11rtl1 Cm1.rnlti111:. /./.( ' UiH74.flltJfl.tJ4 H,.1*i.flm1 tJ