RA-25-0117, Response to Request for Additional Information (RAI) Regarding Proposed Alternative for Acceptance of Through-wall Flaw in a Reducer Small End Transition Zone
| ML25126A250 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/06/2025 |
| From: | Ellis K Duke Energy, Duke Energy Progress |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RA-25-0117 | |
| Download: ML25126A250 (1) | |
Text
Kevin M. Ellis General Manager Nuclear Regulatory Affairs, Policy &
Emergency Preparedness Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 843-951-1329 Kevin.Ellis@duke-energy.com 10 CFR 50.55a Serial: RA-25-0117 May 6, 2025 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Brunswick Steam Electric Plant, Unit No. 1 Renewed Facility Operating License No. DPR-71 Docket No. 50-325
SUBJECT:
Response to Request for Additional Information (RAI) Regarding Proposed Alternative for Acceptance of Through-wall Flaw in a Reducer Small End Transition Zone
REFERENCES:
- 1. Duke Energy letter, Proposed Alternative for Acceptance of Through-wall Flaw in a Reducer Small End Transition Zone, dated March 21, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25080A235)
- 2. NRC email, FINAL Brunswick U1 RAI Reducer Alternative, dated April 17, 2025 Ladies and Gentlemen:
In Reference 1, pursuant to 10 CFR 50.55a(z)(2), Duke Energy Progress, LLC (Duke Energy) proposed an alternative to the requirements of American Society of Mechanical Engineers (ASME) Code,Section XI for the Brunswick Steam Electric Plant (BSEP), Unit No. 1.
Specifically, Duke Energy proposed to use alternative methods to calculate the stresses that are used in evaluating a Service Water (SW) system through-wall flaw. In Reference 2, the NRC staff requested additional information regarding Reference 1. Enclosure 1 provides Duke Energys response to the Reference 2 RAI.
No new regulatory commitments have been made in this submittal.
U.S. Nuclear Regulatory Commission RA-25-0117 Page 2 If you have additional questions, please contact Ryan Treadway, Director - Nuclear Fleet Licensing, at 980-373-5873.
Kevin M. Ellis General Manager - Nuclear Regulatory Affairs, Policy & Emergency Preparedness
Enclosures:
cc:
- 1. Response to Request for Additional Information USNRC Region II Regional Administrator USN RC NRR Project Manager for BSEP USN RC Senior Resident Inspector for BSEP
RA-25-0117 Response to Request for Additional Information 2 Pages Follow
RA-25-0117 Page 1 of 2 NRC RAI-1 Confirm that a flooding analysis to evaluate the impact of the leakage at the allowable leak rate on the safety-related structures, systems, and components in the room and/or vicinity of the leaking pipe has been completed and summarize its conclusions. Provide a leak rate limit (or corresponding hole size), based on the flooding analysis, beyond which corrective action (e.g.,
repair or replacement) must be taken or confirm that this limit would correspond to a hole size larger than the 10-inch by 10-inch hole considered in the structural integrity evaluation.
Duke Energy Response to NRC RAI-1 Based on a review of design basis calculations, this section of piping will remain under vacuum with the design basis accident flow of 8000 gpm through the A Loop Residual Heat Removal (RHR) Heat Exchangers. With the piping under vacuum conditions, leakage from the system is not postulated. In the unlikely event external leakage were to occur, the following provides a description of the available flood protection:
The Service Water line containing the leak is located in a vertical pipe chase in the Northwest corner of the reactor building. Since there is no internal positive pressure in the piping, any postulated leakage would be considered to be contained within the pipe chase and collect on the floor of the North Core Spray Pump Room. The North Core Spray pump room contains a 600 gallon capacity sump and a sump pump with a capacity of 50 gpm. The sump is equipped with a high-level switch connected to an annunciator to alert operations of high level in the sump which would indicate leakage exceeding the capacity of the sump pump. Additionally, the North Core Spray pump room is equipped with an additional level switch and annunciator to alert operations if the water level in the area is above the top of the sump for flooding.
For external flooding events, evaluations postulate the potential for a total of 17 gpm leakage into the North Core Spray pump area through doors and other sources. This provides 33 gpm margin in pump capacity for other leakage sources.
Based on the above, limiting the leakage from the through-wall flaw to less than 10 gpm is considered reasonable for protecting the North Core Spray Pump room from internal flooding.
This limit is within the capacity of the area sump pump and provides margin to accommodate other leakage. The daily monitoring described in Section 5.0 of the proposed alternative will identify any increasing trend in leakage that would approach this limit. Annunciator alarms will alert operations of any unexpected leakage increases.
The piping containing the flaw is on the discharge side of all Safety Related and Non-Safety Related heat exchangers. External leakage at this location will not impact the heat removal capability of any Safety Related or Non-Safety Related equipment.
If the identified leakage exceeds the 10 gpm leak rate, repair/replacement activities will be performed to restore the integrity of the piping.
RA-25-0117 Page 2 of 2 NRC RAI-2 Section 1.0 of Attachment 3 to the proposed alternative states that the interior surface of the degraded pipe has a cement lining.
- a. Discuss whether the cement lining provides any structural support of pipe loads.
- b. Discuss whether the cement lining is modeled in the finite element model or provide justification for not modeling.
- c. Discuss whether the weight of cement lining is included in the deadweight of pipe as part of the stress analysis or provide justification for not including.
- d. Provide the average thickness of the cement lining.
Duke Energy Response to NRC RAI-2
- a. The cement lining is unreinforced and does not possess the strength in tension necessary to be considered a structural component. The cement lining is not credited for providing any structural support in the pipe stress calculation or the finite element analysis in the proposed alternative.
- b. The cement lining is considered non-structural and was not modeled in the finite element model because the purpose of the cement liner is to prevent corrosion/tuberculation within the piping. The cement is unreinforced and does not possess the strength in tension necessary to be considered a structural component. Neglecting the stiffness of the cement liner is conservative because doing so increases the stress in the steel pipe.
- c. The pipe stress calculation (SA-SW-761/762/763) includes the weight of the cement lining in the weight per foot input value for the piping material plus contents. The forces and moments developed in the pipe stress analysis are used as input into the finite element analysis.
- d. Plant drawings show the minimum thickness of the cement lining is 5/16 for pipe sizes 12 to 20 and 3/8 thick for pipe sizes 24 to 36.