ML22130A670

Trp 076 St. Lucie SLRA Breakout Questions Irr. Concr. Steel AP DD Jd (Combined)

ML22130A670
Person / Time
Site:	Saint Lucie
Issue date:	01/21/2022
From:	Vaughn Thomas NRC/NRR/DNRL/NLRP
To:
Thomas V, 301-415-5897
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ML22130A654	List: ML22130A670 ML22130A673
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Download: ML22130A670 (26)

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St. Lucie SLRA: Breakout Questions SLRA FE Sections 3.5.2.2.2.6 and 3.5.2.2.2.7:

TRP: 76 Note: Breakout Questions are provided to the applicant and will be incorporated into the publicly-available audit report.

Technical Reviewer A Prinaris 01/05/22 D Dijamco 01/06/22 J Dean 01/21/22 Technical Branch Chief J Colaccino Concurrence Date: 01/13/22 A Buford Concurrence Date: 01/10/22 R Lukes 01/21/22 (through Jeremy Dean)

Breakout Sessions Date/Time To be filled in by PM Applicant Staff NRC staff To be filled out by PM during breakout 76a: SLRA Section 3.5.2.2.2.6 Question SLRA SLRA Background/Issue Discussion Question/Request Outcome of Number Section Page (As applicable/needed) Discussion 1 3.5.2.2.2.6 3.5-35 The SLRA states that [t]he 7 ft. 3 in. thick [Primary By sharing a screen, review AP 3.5-39 Shield Wall] PSW and the mass concrete on which the aforementioned As Built it rests (elevation 7.5 to 18), which is identified as Drawings (as well as other the lower cavity concrete (LCC), surrounds the relevant drawings), discuss, RPV where potential radiation damage in the and clarify what are the concrete is maximum. Both the Unit 1 and Unit 2 compressive strengths of PSW/LCC have the same configuration. Table reactor cavity concretes.

3.5.2.2-1, PSL Primary Shield Wall/Lower Cavity Concrete compressive

Concrete Specifications, of the SLRA states that strengths and associated the compressive strengths of Units 1 and 2 PSW mechanical properties are and LCC concretes are 5,000 psi. Section 6.1 (pg. reduced when exposed to

39) of ePortal document 8770-16346, Revision 3, radiation. Concrete material EBASCO Backfit Engineering, St. Lucie Units 1 & properties are of importance 2 Civil Engineering Design Criteria, states that the particularly in areas where reactor internal structures concrete strengths are RPV support beams frame into 5,000 and 4,000 psi. ePortal As Built Drawing reactor cavity concrete wall(s).

8770-G-518, Revision 3, notes: Concrete shall be Class AA (4,000 psi) and indicates the reactor cavity concrete compressive strength to be 5,000 psi from approximate elevations of 18 feet to 36 feet. ePortal As Built Drawings 8770-G-529, Revision 2, and 2998-G-529, Revision 4, notes that concrete shall be class AAA (5,000 psi) and Class AA (4,000 psi) and indicate that only portion of the reactor cavity concrete strength to be 5000 psi.

It is not clear what are the actual/As Built concrete compressive strengths of the reactor cavity concretes (PSW/LCC).

2 3.5.2.2.2.6 3.5-35 The SLRA states that a 7 ft., 3 in. thick concrete a) By sharing a screen, show AP primary shield wall surrounds the reactor. [] The construction drawings of 7 ft. 3 in. thick PSW and the mass concrete on the Unit 1 and Unit 2 PSWs which it rests [] surrounds the RPV where to clarify what is the actual potential radiation damage in the concrete is (As Built) thickness of the maximum. It also states that [b]oth the Unit 1 concrete PSW cylinder at and Unit 2 PSW/LCC have the same each Unit.

configuration (emphasis added).

b) If the As Built PSW Section 3.8.3.4, Design and Analysis Procedures, cylinders are thicker than of the UFSAR states that the primary shield wall is that analyzed, particularly a 6 feet thick cylinder and is analyzed as a thick in the area where the RPV cylinder. Section B.2, Reactor Support support system (assembly)

Structure, of Unit 1 UFSAR Appendix 3H and Unit beams frame into the

2 UFSAR Section 3.8.3.1.1, Primary Shield Wall, cavity concrete, discuss however, confirm that the PSW concrete thickness the applicability of the 6-for both Units is 7 ft., 3 in as stated in the SLRA. feet thick concrete cylinder analysis to that of a 7 feet It is not clear whether the PSW concrete has 3 in. concrete cylinder variable thickness for Unit 1, whether the analysis construct.

for both Units were done for a 6 feet thick cylinder but the Units PSWs were built as 7 ft., 3 in.

thickness cylinders because of differing loading conditions. It is also not clear whether a 7 ft., 3 in.

thick cylinder has been used for conservatism in the PSL PSW Unit 1 and Unit 2 irradiated concrete safety evaluations (e.g., those in ePortal document NEESL00008-REPT-098 that reference the CLB analysis) although their CLB analysis may have varied.

3 3.5.2.2.2.6 3.5-39 Portions of the PSL RPV beam supports are a) Clarify whether the loading AP embedded in concrete. PSL in ePortal document factor of 1.15 was NEESL00008-REPT-098, St. Lucie Units 1 and 2 considered in the faulted Subsequent License Renewal Primary Shield Wall loading condition for the Irradiation Evaluation, uses interaction ratios (IRs) PSW IR evaluation in to assess the capacity of the reactor cavity ePortal document concrete to carry imposed loads including those NEESL00008-REPT-098.

from the embedded RPV support beams. To this end the SLRA states, that the governing failure b) Discuss whether loads and mode at the reactor cavity wall is tensile failure of loading conditions in the vertical rebars at the inner face of the UFSAR Unit 1 and 2 PSW/LCC. It also states that based on the Tables 3.8-11 and 3.8-19, irradiation effects summarized in the section and respectively have been the original analysis of the PSW/LCC under CLB considered when loading conditions for both PSL Units 1 and 2, calculating PSW IR(s) in there will be minimal effect on the IR associated ePortal document with the governing failure mode of 0.77. It then NEESL00008-REPT-098.

states: If not considered, provide

this IR is based on a guillotine break of the an explanation justifying main primary loop piping; thus the actual IR their exclusion.

will be much lower considering both PSL Units 1 and 2 have implemented leak-before- c) Discuss the impact of an break of the primary loop piping as part of increase in exposure their CLBs. uncertainty by 20 percent on an IR of 0.77, the ePortal document NEESL00008-REPT-098, governing mode of tensile (hereinafter referred as the basis document) failure, and the RPV references Specifications 8770-16346, Revision 3, support beam embedment for St Lucie CLB reinforced concrete loading in concrete.

combinations and capacity reduction factors ().

Section 2.2, Current Licensing Basis Loading Condition, of the basis document references Section 4 of Specification 8770-16346, Revision 3 and recounting CLB calculations determines that the Pipe Break or Accident with Design Basis Earthquake is the governing loading combination for the reactor cavity concrete subject to exposure.

It then references UFSAR Unit 1 Appendix 3H page 20 for the definition of the SLRA listed IR of 0.77. For Unit 1, it is apparent that this IR is calculated based on the tensile stress of a vertical rebar of 27.6 ksi divided by the allowable yield strength of the rebar (strength reduction factor of rebar 0.9 X 40 ksi). However, the relevant loading condition in Appendix 3H of the UFSAR Unit 1 assigned for the definition of 0.77 IR includes a load factor of 15 percent increase for all loads (except for loads associated with temperature effects). Furthermore, when considering Unit 1 UFSAR Table 3.8-11, the governing PSW loading condition and its loading factors differ from that listed in NEESL00008-REPT-098. A similar

observation is made for Unit 2 PSW governing loading condition and its loading factors (see Unit 2 UFSAR Table 3.8-19). Furthermore, applied loads and resulting loading capacities for Units 1 and Unit 2 PSWs also differ (see corresponding UFSAR Unit 1 and Unit 2 Tables 3.8-11 and 3.8-19, respectively).

It is not clear whether the basis document considered all of the above governing loading conditions when discussing the IR of 0.77. If so, it is not clear whether the 0.77 IR considered also the 15 percent loading factor increase discussed above, and whether the 15 percent loading factor was used in the evaluation with the faulted loading condition. Furthermore, it is not clear whether the tensile mode of failure for the IR considered, accounted for effects of radiation (if any). If so, it is not clear what would be the IR value, particularly when radiation exposure uncertainty consistent with conclusions reached at Point Beach is maximized to 20 percent.

4 3.5.2.2.2.6 ePortal Drawings 8770-G-794 and 2998-G-794 a) Identify which type of AP indicate that RPV structural steel assembly column EMBECO 636 grout has base plates are supported at elevation 2.92 feet to been used on RPV concrete through a 3 inches thick concrete grout. structural system Section 4.0 Material Specification Data of ePortal (assembly) supports.

Document PSL-110389-001-M04, Summary &

Results Report for Reactor Head Drop Analysis, b) Discuss if there is any OE Attachment B (page B5) states that the grout used related to the performance in column anchorage is EMBECO 636. In a letter of the EMBECO 636 grout.

dated May 7, 2001(ADAMS Accession No. ML011310474) the supplier of EMBECO 636 c) Discuss measures to be notified NRC of a Possible Nonconformance taken to ensure adequacy

Related to the Performance of EMBECO 636 of grout performance (CMTR) Grout also noted in Event Notification during the subsequent Number 37685. It is not clear whether there is any period of extended Operating Experience (OE) at PSL regarding the operation, particularly if the performance of the EMBECO 636 grout to date. grout contains metallic and quartz aggregates and In addition, it is noted that the PLUS version (type) exposed to radiation and of the grout contains metallic and quartz corrosive (boric acid) aggregates (see https://marbri.net/embeco-636- environment.

plus-grout) which could be affected by boric acid corrosion and radiation effects.

Since the EMBECO 636 grout is identified as a nuclear safety related product, it is not clear, consistent with the reevaluation criteria of NUREG-1509, whether the environment in the reactor cavity has been examined as to the future performance of the EMBECO 636 grout during the subsequent period of extended operation.

5 3.5.2.2.2.6 3.5-35 SLRA Table 3.5.2.2-1, PSL Primary Shield a) Clarify the origin of the AP 3.5.2.2.2.1 3.5-28 Wall/Lower Cavity Concrete Specifications, states aggregates and discuss if 3.5-38 that concrete mix [f]ine aggregate consists of there is a material natural and/or manufactured sand [and c]oarse certification indicating aggregate consists of hard, durable crushed rock pit/quarry origin or provide or natural gravel. Although the SLRA discusses evidence that the qualification of aggregates (e.g., SLRA Section aggregates are not quartz, 3.5.2.2.2.1) consistent with ASTM and ACI and if they are, discuss that standards, it does not identify whether the concrete their potential RIVE mix had quartz or limestone aggregates. swelling is of no concern.

The SLRA also states that the inner surface of the b) Discuss condition of PSW concrete does not have a steel liner. PSW/LCC inner surface.

However, construction drawings (e.g., 8770-G-794 State if there are areas of SH2, Revision 3 and 2998-G-794 SH2, Revision 5) concrete spalling. Provide RPV cavity concrete

indicate the inclusion of a liner in areas where RPV (PSW/LCC) pictures steel beam supports frame into the PSW concrete. indicating that no spalling or other aging effects are It is not clear what is the origin of the manufactured occurring.

sand. Typically, manufactured sand is silica based and mostly derived from quartz or granite. Given c) Discuss and clarify the the uncertainty in radiation exposure, RIVE maybe purpose of the liner a factor contributing to spalling of PSW concrete installed at location of the surface particularly if the concrete aggregates RPV beams framing into were of quartz origin. the PSW concrete. Was the liner installed to mitigate potential concrete spalling at location of RPV beam support framing into the PSW concrete?

6 3.5.2.2.2.6 3.5-35 FPL letter (page 3) to NRC (ADAMS accession No a) By sharing screen show AP ML18108A562) states that [w]ith the exception of relevant Unit 1 and Unit 2 the reactor cavity wall, the Unit 2 reactor coolant construction drawings system support design duplicates the Unit 1 illustrating differences in design. With regard to cavity wall design, the Unit areas of said modifications 2 design has been modified [] to provide in the PSW/LCCs.

additional margin above that shown to be acceptable for Unit 1. The SLRA states that b) Clarify the cause, need,

[b]oth the Unit 1 and Unit 2 PSW/LCC have the and importance for such same configuration, which is indicative that there modifications. Discuss to are no differences in PSW/LCC and/or reactor what extent (if any) such cavities for the two PSL Units. modifications may have helped improve radiation It is not clear what are the reactor cavity shielding, maintenance, or modifications at Unit 2 PSW/LCC. It is also not aging effects management clear what precipitated these modifications and of PSW/LCC concrete and whether their examination is addressed in RPV structural steel inspection procedures (current and those proposed system support assembly for the subsequent period of extended operation) (e.g., of the RPV steel support components

particularly if they are in areas of increased encased in concrete) and radiation. further addressed in inspection procedures.

c) Discuss whether such modifications would not alter the previously discussed IR(s) for Unit 2.

7 3.5.2.2.2.6 ePortal document NEESL00008-REPT-098, St. Clarify whether the CLB AP Lucie Units 1 and 2 Subsequent License Renewal analysis referenced in the Primary Shield Wall Irradiation Evaluation, (aka basis document is summarized basis document) references (page 23) the original in the aforementioned FPL CLB analysis for safety evaluation of the letters to NRC. If not PSW/LCC in the subsequent period of extended summarized and not discussed operation. It is not clear whether the referenced in Appendix 3H of the UFSAR, CLB analysis regarding the PSW has been summarize the relevant summarized in FPL letters to NRC (ADAMS sections of the CLB analysis, Accession Nos.: ML18114A219 and particularly in areas of high ML18108A562) and in UFSAR Appendix 3H. exposure and where the RPV support beams (assembly) frame into the PSW concrete.

8 3.5.2.2.2.6 3.5-39 SLRA Table 3.5.2.2-2, End of Subsequent Period Clarify the below:

JD and of Extended Operation (72 EFPY) Exposures for 3.5.2.2.2.7 PSL Concrete, provides fluence and gamma dose What is the estimate of the values for PSL Units 1 and 2. The SLRA then uncertainty associated with the states: neutron fluence and gamma dose or displacements per Based on these results, the projected end of atom (dpa) results for:

SPEO gamma doses for PSL Unit 1 and PSL Unit 2 fall below the NUREG-2191 and

• Reactor Pressure Vessel NUREG-2192 concrete irradiation damage (RV) beltline threshold for gamma radiation (1.0 x 1010

• Biological Shield Wall rads). Accordingly, no further evaluation of (PSW/LCC)

the PSL Units 1 and 2 PSW/LCCs for

• RV support steel (ring gamma irradiation effects is required girder and support columns)

Uncertainties in radiation exposure are the issue at hand for the RV beltline, RV steel support assembly, and BSW 9 3.5.2.2.2.6 3.5-38 The SLRA states: Clarify the below:

JD and 3.5.2.2.2.7 Unlike what was done for initial license a) Based on the projected renewal, future projections for PSL SLR margin in the SLRA, are included a 10% positive bias on the there any fuel loading peripheral and re-entrant corner assemblies restrictions needed? This on the projection fuel cycle. Peripheral should include any shield assemblies have one or more faces exposed assemblies and their to the core baffle plates and re-entrant designs if needed.

corner assemblies have one corner exposed b) If so, please state the the core baffle plates. loading restrictions and how they are being The 10% positive bias applied to the implemented/managed?

projection cycle peripheral and re-entrant corner assembly relative powers is intended to account for normal cycle-to-cycle variations that have been observed in past PSL core designs and are expected to occur in future ones as well.

It is not clear whether fuel loading restrictions are accounted in the above.

10 3.5.2.2.2.6 3.5-34 There are no discussions of insulation in SLRA Clarify the types, locations, JD and through Sections 3.5.2.2.2.6 and 3.5.2.2.2.7 relevant to and results of and inspections 3.5.2.2.2.7 3.5-47 fluence calculations. of any insulation materials relevant to fluence calculations. Areas of focus are reactor coolant piping, Primary Shield Wall (PSW),

and BSW. This should include:

• Reflective Metal Insulation (RMI) - reference (IN) 2007-21, Supplement 1

• Any other flexible, rigid, or metallic insulations 11 3.5.2.2.2.6 3.5-34 Table 3.5.2.2-2 of SLRA Section 3.5.2.2.2.6 Discuss how any radiation-JD through indicates exceedance of fluence limits (see SRP- induced volumetric expansion 3.5-40 SLR 3.5.2.2.2.6) on concrete for PSL Units 1 and 2 (RIVE) effects are evaluated in for 72 EFPY. The section however does not combination with other include a discussion of radiation-induced structural stresses that may be volumetric expansion (RIVE) calculation and its present for all the aging effects on concrete other than the statement: management systems under evaluation.

Localized cracking and spalling of the concrete at the peak areas of neutron fluence due to radiation-induced volumetric expansion are not expected.

12 3.5.2.2.2.6 3.5-34 The SLRA section states: Clarify whether there are any JD through components of the PSW/LCC 3.5-40 Although the PSL PSW/LCCs do not have a that require any additional or liner plate on their outside surfaces, localized augmented inspections/actions cracking and spalling of the concrete at the over the subsequent period of peak areas of neutron fluence due to extended operation based on radiation-induced volumetric expansion are the results of OE and not expected. The reactor cavity areas for information provided in this both units will continue to be inspected as SLRA.

part of the Structures Monitoring AMP (Section B.2.3.33).

Since there is no liner to inhibit any concrete spalling that could affect safety related SSCs, the issue at hand is adequacy of inspections for the PSL PSW/LCCs.

76b: SLRA Section 3.5.2.2.2.7 Questio SLRA SLRA Background / Issue Discussion Question / Request Outcome of n Section Page (As applicable/needed) Discussion Number 1 3.5.2.2.2.7 3.5-44 The SLRA states that Westinghouse performed a Discuss/explain/justify that AP 3.5-46 qualitative assessment of the PSL Units 1 and 2 there was no need to consider structural steel RPV supports in References the loading conditions listed in 3.5.4.7 and 3.5.4.8. SLRA Reference 3.5.4.7 Section 3.8.3.3.2.3 and Table (WCAP LTR-SDA-21-021-NP, Revision 1) 3.8-12 of Unit 1 and Unit 2 addresses the specifics of the assessment and UFSARs, respectively in loads states that in order to perform the comparison development and that the PSL assessment with PBN RPV support evaluation in Unit 1 UFSAR Appendix 3H WCAP-18554 the branch line pipe break (BLPB) faulted loading condition(s) for PSL are required and calculated. It also states is/are adequate for a that the PSL BLPB load development is not conservative estimate of loads complete as the reactor coolant system models to predict the most stressed from the analysis of record need to be updated. location of the RPV steel Hence, conservatively estimated loads are used support system (assembly).

instead based on faulted loading condition (including e.g., thrust/reactor vessel internals (RVI) blowdown pressure waves loads) to predict the most stressed location of the RPV steel support system (assembly).

FPL letters to NRC (ADAMS Accession Nos.:

ML18114A219 and ML18108A562) include analyses to determine PSL RPV support system (assembly) loads and evaluation of supports restraint capability based on a faulted loading condition. They conclude that the support system (assembly) safety margins are adequate.

Subsequent analyses summarized in PSL UFSAR Unit 1 Appendix 3H (e.g., pages 3H-3, 3H-19, etc.)

with augmented thermal loads contributing to vertical thermal growth of the RPV steel support

assembly with displacement to limit its loading capacity (displacement control).

Although Section 3.8.3.3.2.3 of the Unit 1 UFSAR and Table 3.8-12 of Unit 2 UFSAR list both service load and factored loading conditions for the RPV steel support system (assembly), it is not clear whether PSL in developing the unitless ratios in SLRA Table 3.5.2.2-5 considered these as well. If so, which one? Without updating the PSL RCS models, it is not clear how the plant specific loading conditions used for calculating flaw tolerance for each of the PSL Units would align with those of PBN to afford such a comparison (e.g., see PSL Unit 1 Appendix 3H thermal loads, LOCA break times discussed in ePortal document Book C142, North Anna Syndrome Analysis, etc.)

in an irradiated environment.

1.

References:

WCAP LTR-SDA-21-021-NP/P, Revision 1, St.

Lucie Units 1&2 Subsequent License Renewal:

Reactor Pressure Vessel Supports Assessment, June 24, 2021.

2. WCAP-18554-NP/P, Rev. 1, "Fracture Mechanics Assessment of Reactor Pressure Vessel Structural Steel Supports for Point Beach Units 1 and 2," September 2020 2 3.5.2.2.2.7 3.5-42 SLRA Figure 3.5.2.2-3 shows that the RPVs for a) Provide history of ASME AP 3.5-44 Units 1 and 2 are supported by 3 beam-column Section XI, subsection IWF 3.5-46 assemblies. The SLRA states that the guidance of inspections for the Unit 1 NUREG-1509 is used for the embrittlement and Unit 2 RPV structural evaluation of the PSL RPV supports. In its OE support system (assembly) segment of the Section, the SLRA states that to date.

consistent with Section 4.3.1.1 of NUREG-1509, physical examination of the RPV supports is b) Clarify whether any of the essential to the evaluation. The NUREG also past inspections were states that the assessment includes a mandatory, consistent with the systematic reevaluation of the current condition of requirements of ASME the supports with an emphasis placed on Section XI, Subsection IWF components supports in tension, as well as or deviated/augmented as observations and predictions of their possible noted in the FPL letter to future weaknesses. NRC. If so, state what prompted such deviations Such examinations in subsequent license renewals and whether PSL plans to are performed through the ASME Section XI, continue implementing Subsection IWF Inservice Inspections (ISIs) these along with the programs with additional guidance of GALL-SLR guidance provided in GALL-AMP XI.S3. Consistent with ASME Section XI, SLR AMP XI.S3 so as to Subsection IWF Table IWF-2500-1, they include avoid possible future 100 percent (visual) VT-3 examination of supports. weaknesses during the However, for multiple components, within a system subsequent period of of similar design, function, and service, the extended operation.

supports of only one of the multiple components are required to be examined. c) Discuss whether differences in PSL ASME Section XI Table IWF-2500-1 Category F-A, Supports of Programs for Subsection ePortal PSL 4th ISI Interval Program Plan for Unit 2 IWF of Units 1 and 2 RPV states that only one of multiple components, within structural support systems a system of similar design, function, and service (assemblies) inspections are to be examined. The ePortal 5th ISI Program are just limited to sampling.

Plan for Unit 1, however, does not provide such clarification. The SLRA indicates that existing OE for Uni1 included VT-3 of hot and cold leg accessible portions of support, while for Unit 2, it states that VT-3 inspections were performed on all accessible areas of hot and cold leg supports to the extent possible.

Furthermore, FPL letter to NRC (ADAMS Accession No. ML18114A219) (page 3) indicates that FPL considered augmented NDEs within the cavity beyond those of Section XI inspection interval requirements. Past OE discussed in the SLRA states that the Unit 2 RPV supports were not being examined. ePortal AR 01716657 (page

65) dated 12/15/2011 discusses Unit 1 RPV supports nonconformance to ASME Code Section XI, Subsection IWF and states that the ISI Program of Unit 1 did not include examination of the RPV Class 1 components supports as required.

NUREG-1779, dated July 2003, however, indicates that the applicants in its initial LRA (page 3.5-36) did include AMRs for ASME Section XI Subsection IWF and Boric Acid inspections of the RPV supports.

It is not clear when the first ASME Section XI inspection of Unit 1 and Unit 2 RPV structural support assemblies took place and whether any of the past inspections were consistent with the requirements of ASME Section XI, Subsection IWF or augmented as noted in the FPL letter. It is also not clear whether current Unit 1 ISI Program follows the sampling requirements of ASME Code Section XI, Subsection IWF Table-2500-1 Category F-A, Supports or has augmented NDE as noted above.

3 3.5.2.2.2.7 3.5-46 ASME Section XI, Subsection IWF-1300, Support a) Discuss whether IWF AP B-247 Examination Boundaries, in its Figure IWF-1300-1 and/or the SMP aging 2.3-57 identifies the extent of support examination management programs (or boundaries and states that the boundary of an both) inspect the RPV integral support (C) connected to a building support beams and structure (E) is the surface of the building columns and whether such

structure. In addition to the ASME Section XI, inspections include Subsection IWF inspection requirements, the examination of the concrete GALL-SLR AMP S6, Structures Monitoring, surface where the RPV provides guidance for inspection of structures steel support components including those within the reactor containment. frame into the cavity concrete/basemat. For the SLRA AMP B.2.3.33 states that the Structures RPV support beams that Monitoring Program (SMP) includes inspection of would include inspections of containment internal structures. SLRA Section the PSW surface at about 2.3.3.12, Ventilation, describes the necessity to 18 feet of elevation and for ventilate the reactor annulus so that concrete is not the column at the base dehydrated and cracks and thermal growth of the plate resting on the 3 inches reactor vessel supporting steelwork is limited to thick EMBECO 636 grout.

3/16-inch. ePortal IWF Supports includes folders for Units 1 and 2 supports that include inspection b) Provide relevant OE and reports and numerous photos. discuss cracking of concrete surface/grout (if It is not clear whether the aforementioned any) witnessed during IWF concrete-steel interfaces at the reactor cavity are ISI/SMP inspections and routinely examined and inspected for cracking due examinations. Present to potential thermal effects (including those that pictures detailing condition could develop from potentially added gamma dose of concrete surface where exposure). It is also not clear whether there is OE the RPV support beam(s)

(e.g., ARs, CRs, and WO) associated with such frame into the PSW and of inspections and examinations. the grout at column base plate(s).

4 3.5.2.2.2.6 3.5-45 The SLRA states that Westinghouse assessed the a) Discuss what is the AP 2.3-57 RPV slide plate lubricant for degraded conditions lubricant used in the sliding such as a decrease in viscosity due to radiation plate and how it is applied effects. In addition to irradiation, the lubricant is to the plate.

exposed to an elevated temperature discussed in Section 2.3.3.12 of the SLRA which states that the b) Clarify whether the sliding temperature at the bottom of the lubrication plate shoe assembly has been evaluated for potential

between the reactor and support leg to be as high increased frictional forces as 300°F (degrees Fahrenheit). (and to this end the RPV structural steel beams),

Although, neutron flux was identified as the key excessive wear, and loss of parameter in irradiation aging effects of the dry film its intended mechanical lubricant it is not clear what the lubricant is and function due to whether the combined effect of radiation and ineffectiveness of the elevated temperatures could lead to increased lubricant caused by its frictional forces, excessive wear, and potentially exposure to radiation and loss of sliding shoe intended mechanical function. elevated temperatures.

In addition, it is not clear whether the support shoe assembly components are exposed to an overall c) Clarify whether the support environment conducive to stress corrosion shoe assembly components cracking (SCC). are exposed to an overall environment conducive to SCC.

5 3.5.2.2.2.7 3.5-44 The SLRA states: Discuss whether the SLRA AP proposed evaluation A comparison of the key inputs to ASME methodology for the Section XI critical flaw size calculations was embrittlement of the RPV made between PSL and Point Beach beam-column support system Nuclear (PBN) in order to ascertain the (assembly) has accounted for acceptability of the PSL RPV steel supports all of the prescribed for the subsequent period of extended reevaluation criteria of NUREG-operation (SPEO) with consideration of 1509 (see examples provided) irradiation aging effects. These key inputs for the subsequent period of consist of the fracture toughness and extended operation.

stresses of the RPV support components and were combined into a comparative ratio In particular, in order for the term based on the general form of stress staff to determine that the intensity factor. This comparative ratio comparative ratios in SLRA effectively normalizes the fracture Table 3.5.2.2-5 would hold true toughness and stress relative to PBN as it for the PSL RPV steel supports pertains to the calculation of critical flaw and provide the same level of sizes [] therefore, the conclusions assurance of protection against

contained within the detailed PBN fracture cracking due to loss of fracture mechanics evaluation (References 3.5.4.9 toughness as the PBN RPV and ML21111A155) can be applied to PSL. supports, assumptions and conservatisms for the PSL RPV LTR-SDA-21-021-NP/P, Revision 1 details the steel support evaluation similar formula used to evaluate PSL RPV structural to the assumptions and steel system supports to radiation effects. conservatisms in the PBN RPV Both the numerator as well as the denominator steel support evaluation, as of the formula are ratios. The numerator is a discussed in the PBN SLRA ratio of mode I critical fracture toughness (Kic, Supplement 1 (ADAMS a measure of flawed steel materials ability to Accession No. ML21111A155, resist fracture to applied loads/stresses) of Attachment 21, pages 12 irradiated material used for the PSL and PBN through 16), need to be steel support systems, while the denominator discussed. Therefore, discuss is a ratio of max stresses/tractions presumably such assumptions and applied normal to potential PSL and PBN conservatisms for the PSL RPV cracks. steel support evaluation. As was done for the PBN SLRA, It is not clear whether the proposed the discussion of these methodology of mixing irradiated material assumptions and states with stresses fulfills the guidance of conservatisms would need to NUREG-1509 for plant specific evaluation of be brought upfront into the the RPV steel supports for each plant, which SLRA.

should also consider; for example:

1. Assessment of the existing condition of the supports at the time of reevaluation;

2. Comparison with the initial construction conditions, fabrication procedures (i.e.,

pre-service examinations of the plates and welds that comprise the RPV steel support assembly);

3. Degree of degradation predicted by the end of plant life;

4. Original design and safety margins;

5. Original design methodology, load combinations for which the supports were designed, allowable stresses and their margins with respect to the actual stresses in the members, and codes governing the original design.

It is not clear whether the above noted criteria for reevaluation have been considered (implicitly or otherwise) in the implementation of the proposed methodology for evaluation of embrittlement of the PSL RPV beam-column supports.

6 3.5.2.2.2.7 3.5-43 The Westinghouse analysis in LTR-SDA In view of the information AP 3.5-39 021-NP/P, Revision 1 assumes full fixity at the contained in the UFSAR Unit 1 RPV beam supports framing into the PSW Appendix 3H, clarify the RPV concrete for calculating its critical stresses and beam support fixity into the its Mode I fracture toughness. Essentially. the PSW concrete and discuss the full fixity at the supports results in a typical conservatism of performed short stubby support beam (see SLRA Figure structural analysis of the RPV 3.5.2.2.-4) with max tensile stresses beam-column support system developing on the exposed portion of the (assembly). Also discuss the support beam. potential for cracking of the RPV steel support beams The concrete analysis of the PSW Unit 1 within the concrete.

discussed in its UFSAR Unit 1 Appendix 3H, however, indicates that tensile cracking is the potential governing mode of failure (see also SLRA pg. 3.5-39) of the PSW. Summarized Finite Element Analyses in Appendix 3H indicate also potential cracking within the concrete where the RPV support beams are anchored into the concrete is also possible. It is not clear how fixity for the RPV support beam can be assumed under these conditions

to evaluate embrittlement of the exposed support beam presented in LTRSDA-21-021-P/NP, Revision 1.

7 3.5.2.2.2.7 3.5-44 The first paragraph in Section 5.0 of LTR-SDA Clarify the source(s) for the DD page 021-NP, Revision 1 (ADAMS Accession No. fluence values. Are they from 10 of ML21215A320, Enclosure 4, Attachment 3 to the the appropriate tables in LTR-Ref SLRA) states that the PSL unit-specific fluence REA-21-1-NP and NP, 3.5.4.8 values are taken into consideration for the Revision 1 (ADAMS Accession embrittlement using Figure 3-1 of NUREG-1509 [2] No. ML21215A320, Enclosure upper bound curve. 4, Attachments 1 and 2 to the SLRA)?

8 3.5.2.2.2.7 3.5-44 Section 5.0 of LTR-SDA-21-021-NP/P, Revision 1 a) Clarify the step-by-step DD page discusses the fracture toughness determination for determination of fracture 10 of the plant Saint Lucie (PSL) Units 1 and 2 reactor toughness for PSL Unit 1 as Ref pressure vessel (RPV) steel supports. The staff described in the enclosed 3.5.4.8 needs clarification on the step-by-step brackets in Section 5.0 of LTR-determination of fracture toughness for each unit SDA-21-021-NP, Revision 1.

as described in Section 5.0 of LTR-SDA-21-021-NP/P, Revision 1. Note that Sections 5.1.1.1 and b) Clarify the step-by-step 5.1.1.2 of WCAP-18554-P (Ref. 4 of LTR-SDA determination of fracture 021-NP/P, Revision 1) discusses the step-by-step toughness for PSL Unit 2 as determination of fracture toughness of the RPV described in the enclosed steel supports for the Point Beach nuclear plant brackets in Section 5.0 of LTR-(PBN). SDA-21-021-NP, Revision 1.

Discuss why the referenced

• • • Because a majority of the discussion in the publicly available documents in referred section is proprietary, the discussion this discussion of PSL Unit 2 during the breakout session would very likely are redacted cover proprietary information. ***

c) Show sample fracture toughness calculations shown in Table 5-1 of LTR-SDA 021-NP, Revision 1 for PSL Unit 1 and PSL Unit 1.

d) Discuss how the fracture toughness values for the welds in Table 5-1 of LTR-SDA 021-NP, Revision 1 were obtained.

9 3.5.2.2.2.7 3.5-44 In the referenced section and page, the SLRA To confirm the results of the DD states the following about the key inputs for the comparative ratio approach, did comparative ratio approach: you attempt to solve for the critical flaw depth iteratively These key inputs consist of the fracture using the equation on page 18 toughness and stresses of the RPV support of LTR-SDA-21-021-NP, components and were combined into a Revision 1 for PSL and for PBN comparative ratio term based on the general (for at least one of the cases form of stress intensity factor. This comparative reported in SLRA Table 3.5.2.2-ratio effectively normalizes the fracture toughness and stress relative to PBN as it 5)? Note that in the equation, pertains to the calculation of critical flaw sizes. fracture toughness and stress are known (Tables 5-1 and 6-1 Details of the comparative ratio approach are of LTR-SDA-21-021-NP, discussed in Section 7.1 of LTR-SDA-21-021-NP, Revision 1, respectively), but Revision 1 (details redacted in NP version). the shape factor F must be iteratively determined in order to solve for the critical flaw depth.

10 3.5.2.2.2.7 3.5-44 Figure 6-3 of LTR-SDA-21-021-NP, Revision 1 a) Does the FEA stress contour DD page shows stress contour plot (redacted as shown) of plot in Figure 6-3 of LTR-SDA-16 of only one T-shaped RPV steel support assembly. 21-021-NP, Revision 1 Ref represent the highest stressed 3.5.4.8 T-shaped support for each unit?

b) Do the other T-shaped assemblies have similar stress plots as in Figure 6-3 or are the stress contour completely different?

11 3.5.2.2.2.7 3.5-44 The staff couldnt verify from the references a) For PWHT of PSL Unit 1 DD page provided on post-weld heat treatment (PWHT) RPV steel supports, page 18 of 18 of discussion on page 18 of LTR-SDA-21-021-NP, LTR-SDA-21-021-NP, Revision Ref Revision 1 that PWHT was performed for the RPV 1 states that reference [9.a]

3.5.4.8 steel supports of both PSL units. has no indication or notes of any field welding. Clarify in the drawing in Ref 9.a that there is no field welding at the high stress locations shown in Figure 6-4 of LTR-SDA-21-021-NP, Revision 1.

b) For the PSL Unit 2 RPV steel supports, page 18 of LTR-SDA-21-021-NP, Revision 1 states that Ref. 14 has information on PWHT, but the staff couldnt verify from Ref. 14 (PSLWEC-21-0066 dated June 15, 2021) that PWHT was performed for the RPV steel supports. Clarify from Ref. 14 (or other references) that PWHT was performed for the PSL Unit 2 RPV steel supports.

12 3.5.2.2.2.7 3.5-44 The staff couldnt determine in the discussion of Discuss cyclic loadings (from DD pgs 13- loads and stress in Section 6.0 of LTR-SDA thermal or other loads), if any, 14 of 021-NP, Revision 1 if there are any cyclic loadings that could cause postulated Ref (from thermal or other loads) that affect the RPV cracks to grow in the RPV steel 3.5.4.8 steel supports of PSL Units 1 and 2. supports of PSL Units 1 and 2.

13 3.5.2.2.2.7 3.5-45 The staff needs clarification regarding SLRA Are these two components DD Tables 3.5.2.2-3 and 3.5.2.2-4. listed in the referenced tables, Support Column and Bottom of horizontal support (top of column) (56 below top of 6 plate), coincident locations?

14 3.5.2.2.2.7 3.5-37 The staff needs clarifications on the support shoes a) Go over the support shoe DD 3.5-41 and socket/slide assembly (how they work and socket and slide assembly 3.5-44 whether its components are susceptible to stress configuration and how it works.

3.5-45 corrosion cracking due to presence of lubricant) See Figure 4-4 of LTR-SDA pages described in the referenced pages. 021-NP, Revision 1, for the 10, 20 configuration (the SLRA says to 21 of the support shoe is welded to Ref the underside of the RPV 3.5.4.8 nozzle); if any photographs of the configuration is available, please show during the discussion.

b) Page 3.5-46 of the SLRA states that magnetic particle examination was performed for the nozzle support. Discuss the latest examination results for both units.

c) Go over the evaluation of effects of lubricant, as discussed in the Section 7.3 of LTR-SDA-21-021-NP/P, Rev. 1.

• • • Discussion would likely include proprietary information. ***

d) How is the above (part b) tied to the discussion on page 3.5-45 of the SLRA on the RPV slide plate lubricant degradation (decrease in viscosity). Also, page 3.5-45 of the SLRA mentions a flux threshold: what

is the threshold value and document/specs for this value?

15 3.5.2.2.2.7 Drawin Drawings of the PSL RPV steel supports a) Discuss the impact of partial DD gs in assemblies in the ePortal (EBASCO Drawing penetration welds on the ePortal 8770-G-794, Rev. 5 for Unit 1 and EBASCO stresses used in the evaluation Drawing 2998-G-794, Rev. 3 for Unit 2) indicate of the PSL RPV steel supports.

that the plates of the assemblies are joined by fillet Were there sensitivity studies or groove welds. If the size of the fillet or groove performed on the finite element weld is small compared to the thickness of the analyses (FEA) with the welds joined plates, only a partial penetration welded joint joining the plates modeled as can result instead of a full penetration welded joint. partial penetration welds versus the welds modeled as full penetration welds (i.e., as fully bonded joint)?

b) Even though a weld joint is shown as full penetration in the drawings (for example, Sector B-12 of EBASCO Drawing 2998-G-794, Rev. 3 for Unit 2),

the welding symbol in the drawing shows what appears to be a 1/4-inch double-bevel groove weld for the 4-inch thick plate, which would result in a partial penetration weld. (Hard to understand why a 4-inch plate would be joined with a 1/4-inch weld, so could the drawing be incorrect? What are we missing?) If the FEA is based on fully-bonded joints, is there a physical way to confirm that the groove (or fillet) welded joints are full penetration, especially the locations of high stress

(Figure 6-4 of LTR-SDA 021-NP, Revision 1)?

16 3.5.2.2.2.7 Drawin Some drawings in the ePortal (such as those with Clarify these MT and UT DD gs in filenames 8770-4306_001_2 and 2998- examinations of the welds of ePortal 5376_001_2) seem to indicate some welds of the the plates comprising the PSL plates comprising the PSL RPV steel support RPV steel support assemblies:

assemblies have received magnetic particle testing (MT) and ultrasonic testing (UT) a) Which welds received MT examinations. and UT examinations? Did the welds at the locations of high stress (Figure 6-4 of LTR-SDA-21-021-NP, Revision 1) receive them?

b) Are these MT and UT examinations pre-service examinations?

c) Confirm that the design specifications for the PSL RPV steel supports that specify these MT and UT examinations are:

• FLO-8770-761, Rev 2 for PSL Unit 1

• FLO-2998-761, Rev 6 for PSL Unit 2 17 3.5.2.2.2.7 3.5-44 Section 7.6 of LTR-SDA-21-021-NP, Revision 1 What does relatively low DD page discusses the anchor bolts at the base plates and means in terms of a fluence 21 of that the fluence at the base plate is relatively low. value (compare with the Ref exposure levels in SLRA 3.5.4.8 Tables 3.5.2.2-3 and 3.5.2.2-4).

18 3.5.2.2.2.7 3.5-40 The SLRA provides Interaction Ratios (IRs) for Provide a summary of the IRs JD through concrete, however no IRs and details on their and material strength values for 3.5-47 each component/system being

calculation are stated for the RV steel support evaluated. Also include the assembly specific components/systems. method/reference for these values any whether they include radiation effects.