ML21153A052

Psl SLRA Preapplication Meeting 06-03-21

ML21153A052
Person / Time
Site:	Saint Lucie
Issue date:	06/02/2021
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
Rodriguez-Luccioni H
References
EPID L-2021-RNW-0002
Download: ML21153A052 (21)

Text

Saint Lucie Units 1 and 2 Subsequent License Renewal Application Pre-Application Meeting #2 June 3rd, 2021

Draft Application Materials Subject to Change 2

Agenda

• Introduction - Topics of Interest

• Surveillance Capsule Removal Schedule

• Irradiation of Concrete - Design and Discussion

• Irradiation of Reactor Pressure Vessel (RPV)

Supports - Design and Discussion

• Questions

• Action Items 3

Surveillance Capsule Removal Schedule

• The next capsule removal for St. Lucie (PSL) Unit 1 is coming up:

- Table 5.4-3 of the Unit 1 FSAR has the Approximate Removal Schedule as 38 Effective Full Power Years (EFPY) for capsule 263

- However, as part of the SLRA, FPL plans to modify the capsule removal schedule which would push the next Unit 1capsule removal out to 47 EFPY

• Unit 1 will only be at 37.6 EFPY or less for the next outage (Fall, 2022)

• FPLs interpretation is the capsule is to be removed at the first refueling outage that meets or exceeds 38 EFPY which would mean the capsule would be removed in the Spring of 2024

- A similar clarification was made to the Turkey Point and Point Beach surveillance capsule schedules

• With the capsule 263 removal planned for the Spring 2024 outage, there is time to modify the removal schedule via the SLRA 4

Surveillance Capsule Removal Schedule

• Similar to Turkey Point & Point Beach, an adjustment to the approved withdrawal schedule will allow sufficient material data and dosimetry for the end of the subsequent period of extended operation (SPEO)

• As an example, see the excerpt from Table 4.4-2, Turkey Point UFSAR 5

Irradiation of Concrete - Design Primary Shield Wall (PSW) Design configuration

• The PSW is a 7 ft 3 in thick cylindrical reinforced concrete wall and surrounds the Reactor Pressure Vessel (RPV)

• Both Unit 1 and Unit 2 PSWs have the same dimensions and the same concrete properties

• Slight differences in reinforcement steel Unit 2 PSW has slightly less vertical reinforcement Unit 2 PSW has Grade 60 reinforcement steel vs Grade 40 for Unit 1

• 5000 psi concrete strength

• No liner plate 6

Irradiation of Concrete - Design PSW Elevation View of the Reactor Building 7

Irradiation of Concrete - Design PSW Plan View of the Reactor Building at Elevation 18 ft 8

Irradiation of Concrete - Design 9

Plan View of RPV Supports

Irradiation of Concrete - Discussion

• Maximum exposures on the inner surface of PSW at the end of the SPEO (72 EFPY) based on Westinghouse calculations (Unit 2 bounding)

- Gamma dose - 6.62 x 109 rads (less than the NUREG-2192 threshold of 1.0 x 1010 rads )

- Neutron fluence E > 0.1 MeV - 1.42 x 1019 n/cm2 (greater than the NUREG-2192 threshold of 1.0 x 1019 n/cm2)

• PSW exposures result in minimal impact to the concrete

- No impact due to gamma dose

- Neutron fluence reaches threshold at ~0.8 inches into PSW

- Neutron fluence effects including radiation-induced volumetric expansion (RIVE) will have minimal impact on interaction ratio (IR) 10

Irradiation of Concrete - Discussion Concrete strain at ultimate strength per ACI 318-69 is 0.003

- Maximum strain due to RIVE is less than 0.003 thus RIVE depth is 0 11

Irradiation of Concrete - Discussion

• IR in UFSAR (applicable to both units) = 0.77

- Based on 42 guillotine break

- Loads considering leak-before-break (LBB) of main loop piping consistent with CLB will result in a much lower IR

• Evaluation demonstrates PSW maintains its structural integrity under current licensing basis (CLB) loading when considering 80-year irradiation effects 12

Irradiation of RPV Supports - Design Design configuration

• RPV Supports

- The RPV is supported at three points on three T-shape steel beam-column assemblies within the reactor cavity

- The horizontal support beams are embedded in the PSW approximately 6 ft on each end

- The column is bolted to the underside of the horizontal support beam at mid-span and to the reactor cavity floor

- Load transfer between the RPV system and the RPV support occur between the nozzle pad which is welded to the reactor nozzle and steel bearing plates designed into the top of the steel horizontal support beam 13

Irradiation of RPV Supports - Design 14

Irradiation of RPV Supports - Design 15 Plan View: RPV Supports

Irradiation of RPV Supports - Design RPV Inlet/Outlet Nozzle Nozzle Pad Elevation View:

RPV Load-Path 16

Irradiation of RPV Supports - Discussion

• FPL is performing a qualitative assessment of the PSL Units 1 & 2 RPV supports, as it pertains to the irradiation aging effects for the SPEO

• This assessment will provide the technical justification to support an inspection-based approach per NUREG/CR-1509

• The assessment has two (2) elements:

Qualitative comparison of the technical attributes Compare PSL RPV Supports to Point Beach Nuclear (PBN)

Inspection-based attributes 17

Irradiation of RPV Supports - Discussion Assessment topic areas include:

• Compare Geometry & Materials Identify analogous components between PBN and PSL Components reduce to plates and round bars Compile all material types for downstream evaluation

• Calculate Fracture Toughness (KIC) using Conservative Fluence & Material Properties Consider CMTRs and industry guidance as applicable

• Calculate Faulted Stresses () from Plant-Specific Model using Conservative Loads Loads are consistent with LBB auxiliary line breaks in the CLB

• Affirm Current Inspection Program Review inspection capabilities

• Assessment Compare ratios between PSL and PBN involving KIc and at critical locations Favorable ratios (i.e., > 1) imply PSL postulated critical flaw sizes would be greater than those of PBN; therefore, PSL would have more margin than PBN, which FPL has evaluated as acceptable for an inspection-based approach Ratios ranging from 2.5 - 19.6 have been calculated Ample margins exist resulting in a favorable comparison supporting above premise Utilize above comparison to affirm validity of current RPV Supports In-service Inspection Program 18

Irradiation of RPV Supports - Conclusion FPL concludes the assessment will support that:

• A plant specific Aging Management Program (AMP) or an enhancement to an existing AMP will not be required (current inspections performed by ASME Section XI, Subsection IWF are sufficient)

• The RPV supports will continue to perform their intended functions consistent with the CLB when considering irradiation effects through the SPEO 19

Questions 20

Action Items 21