ML24109A178

Attachment 1 - VC Summer SLRA: Supplemental Audit Questions

ML24109A178
Person / Time
Site:	Summer
Issue date:	04/19/2024
From:	Lauren Gibson NRC/NRR/DNRL/NLRP
To:	Carr E Dominion Energy South Carolina
References
Download: ML24109A178 (4)

Text

1 ATTACHMENT 1 V.C. SUMMER SLRA: SUPPLEMENTAL AUDIT QUESTIONS SLRA Section: 3.5.2.2.2.6 Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation GAI-AOR = Structural Engineering Calculation No. 1.53. Reactor Building Interior Concrete Area. Reactor Vessel Support Anchor Assembly/Anchor Assembly under RV Support - Primary Shield Wall W-AOR = CGE-CA120-CN-SA-000001, Rev. 1, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment

1. It appears from GAI-AOR that the noted entries in the Subsequent License Renewal Application (SLRA) Table 3.5.2.2.2.6-2 Summary of SLR Demand to Capacity Ratios for PSW Concrete come from the single loading combination of accident, temperature, and pressure loads. Were there any other loading combinations that contained in addition to large loss-of-coolant accident (LOCA) loads, jet and seismic forces? Was the W-AOR evaluation considered the recalculated reinforcing steel at the same locations to match those of the GAI-AOR?

2. It is not clear what is the air gap of the Reactor Vessel (RV) annulus so that there is unrestricted air flow to the primary shield wall (PSW). Design Basis Document, Rev 9 states an air gap of 3, the site reports (reference: DC07020-002, Rev 3) 1.75 max and 1.5 min. Please clarify what is the current unrestricted air gap between the PSW and the RV with and without insulation.

3. With regard to gamma heating, discuss the concrete compressive strength for 10 or more portions of the PSW. Reference(s) indicate that potentially, it could vary from 0.9

• 5,381 to 0.9

• 6,083 psi or 4,843 to 5,475 psi (reference: Attachment 14 to Calcs DC0020-209, Rev 0, SP-201, ACI 318, and EPRI documentation).

4. It appears that the W-AOR calculations are based on a concrete clear cover of 2.5. The clear cover in construction drawings is shown to be 4.5 (vertical bars). What was the clear cover used in the legacy and W-AOR calculations? Is there enough margin in structural concrete calculations at Effective Full-Power Years (EFPY) 72 to obviate the concern of its more accurate representation?

5. The W-AOR states that given the conservatisms in the analyses quantification other considerations are not necessary. There is no description of what are these other omitted considerations from the analyses. Please discuss/describe these other considerations.

6. Confirmation needed that the PSW strength calculations are based on a 3,000 psi compressive strength concrete. Concrete construction drawings state that the compressive strength of concrete is 5,000 psi at 90 days. Was the 3,000 psi concrete compressive strength used for the corbel, the PSW, and other relevant concrete construct designs to loading conditions and combinations (including static and dynamic loads resulting from the mass/weight of the RV)?

2

7. The GAI-AOR states that the concrete carries vertical loads and the embedded steel caries lateral loads. The W-AOR states that concrete carries vertical loads, lateral loads, and torsional loads. Are both correct? Please clarify.

8. The grout rests on approximately 4 inches of zero compressive strength of PSW concrete. In addition, it interconnects the Westinghouse short columns to the Gilbert PSW/corbel embedded steel. Presented during the audit referenced material (e.g.,

CR-00-01234, CGE-CA120-TM-SA-000001) indicates that a portion of the grout that has been exposed to boric acid, may also be exposed to effects of radiation at 72 EFPY as well. Response to Audit Breakout Question Number 6 states that the GAI-AOR indicates the grout allowable compressive strength to be 6,453 psi at 72 EFPY versus that of 6,333 psi required @ 14 days. A further review of the GAI-AOR, however, shows the demand for the grout compressive strength to be as high as 7,104 psi @ 14 days. In addition, literature (NUREG/CR 7280) indicates that creep could also potentially affect the grout deformation and hence stability of the RV placement. Given that it is not clear what would be the compressive strength/load bearing capacity and state of the grout at 72 EFPY, its monitoring and management of applicable aging effects (e.g., loss of mechanical properties, reduction of strength, deformation, cracking) for all applicable environments (including radiation) consistent with the GALL-SLR and SRP-SLR Reports review principles is considered essential. Clarify steps are/to be taken to ensure that the grout fulfills its intended function and that stability of the RV under all loading combinations, conditions and environments remain consistent with the CLB to the end of the SPEO.

9. Locally the PSW concrete was exposed to boric acid. It is not clear whether the reinforced concrete structural steel and the embedded structural shapes and plates are/were affected by the boric acid. Hence monitoring of the PSW reinforced concrete is considered essential. Clarify steps are/to be taken to ensure that the effects of aging (e.g., loss of material, cracking, loss of strength) to all applicable environments (including that of radiation) for PSW reinforced concrete and embedded structural shapes and plates are managed consistent with the GALL-SLR and SRP-SLR Reports review principles to the end of the SPEO.

10. Potential loss of grout support function/compressive strength at approximate elevation 426-8 may also result to liner overload and reduction to its barrier/retaining wall capacity to further contain portions of the grout and concrete that are considered ineffective to support the Westinghouse short columns in their entirety at 72 EFPY. Potential liner bulging or deformation could affect the RV short column leveling (as each of these is tangent to the liner), RV balance, and potentially the Nuclear Steam Supply System (NSSS) configuration at 72 EFPY. Hence monitoring of the liner for potential deformation is considered essential. Clarify steps are/to be taken to ensure that the effects of aging (e.g., liner deformation, bulging) for the PSW liner, grout, and reinforced concrete are managed consistent with the GALL-SLR and SRP-SLR Reports review principles to the end of the SPEO.

11. Attachment 14 to Calcs DC0020-209, Rev 0, states that the PSW concrete coarse aggregates are siliceous. A number of publications (e.g., see references in Section 3.5.2.2.2.6 of PBN SLRA SE - ML22140A127) show that such aggregates when irradiated could produce as much as a 20 percent radiation induced volumetric expansion (RIVE). As a result, at the fuel mid-plane elevation potentially the expanding

3 concrete/aggregates and irradiated strained/cracked liner, at 72 EFPY could affect the stability of the RV. Hence monitoring of the liner for deformation and cracking is considered essential. Clarify steps are/to be taken to ensure that the effects of aging (e.g., loss of material, cracking, RIVE) to all applicable environments (including that of radiation) for the PSW liner and reinforced concrete are managed consistent with the GALL-SLR and SRP-SLR Reports review principles during the SPEO.

12. The applicant defined radiation screening rules in proprietary W-AOR, V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Primary Shield Wall Concrete Assessment, Revision 1, appear to reduce the effects of radiation/fluence on the RV support components (steel, cementitious materials) at 72 EFPY. State where these screening rules were applied and confirm that their use did not materially affect results of calculations described in SLRA Section 3.5.2.2.2.6 including those in Demand to Capacity (D/C) ratios Tables.

13. GAI-AOR states that the embedded steel assembly design did not consider any composite action of the surrounding concrete, especially in calculating stresses in the wide-flange (WF) sections. This is conservative, hence the overall adequacy of the anchor assembly to transfer the Westinghouse reactor vessel loads to the main concrete of the primary shield wall is satisfactory. However, it does say that the headed concrete anchors engage the concrete. Clarify the two statements.

14. GAI-AOR also states the following loads for each WF anchor support: Vertical Seismic Design-basis event (DBE) = 700 kips, Vertical Accident = 2950 kips, Horizontal Shear Seismic DBE = 500 kips, Horizontal Shear Accident = 3000 kips. Discuss what accident load, e.g., large break LOCA, was analyzed in the GAI-AOR. (Can be discussed under the GAI-AOR/W-AOR clarification question.)

15. SLRA 3.5.2.2.2.6 states: the primary vertical, hoop, and shear reinforcement which is considered in the analysis of record is located outside the zones of neutron and gamma threshold exceedances. Therefore, there is no reduction in the ability of the primary reinforcement to carry load, and no reduction in PSW capacity as a result of rebar capacity reduction. Details of the capacity reduction calculations are in the proprietary W-AOR. Clarify whether reinforcement in the capacity reduction calculations in the W-AOR includes only the ((

)) (This one can also be discussed under the GAI-AOR/W-AOR clarification question.)

16. W-AOR states: ((

)) It is not clear in the response to question 17a of the breakout audit (ML24085A701) whether the 3/4 x 7-3/16 headed concrete anchors are the same as the (( )) described in the W-AOR. The response to question 17a further states that the headed concrete anchors engage the concrete, but composite action is not credited in the design. This response seems to be inconsistent with the W-AOR.

Clarify/discuss this inconsistency.

4

17. SLRA Table 4.3.4-1 reports cumulative usage factors due to cyclic fatigue loads for the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. The RV outlet and inlet nozzles are supported by the support shoes of the RV steel supports.

Therefore, it is not clear whether the support shoes experience cyclic fatigue loads.

Proprietary document CGE-MC000-TR-CF-000005 provides some details of the RV Outlet Nozzles and Support Pads and RV Inlet Nozzle and Support Pads. CR 01577 shows a block of metal referred as Weld Build-Up between the RV nozzle and the support shoe and states that the Weld Build-Up is part of the nozzle and is therefore considered part of the pressure boundary. Clarify whether this Weld Build-Up is the support pad.

18. The bases for the 20 percent and 25 percent uncertainties applied to the fluence estimates to the PSW and RV support structure, respectively, are not documented in the application or in a document that can be cited in the NRC staffs safety evaluation.