RA-25-0264, Proposed Alternative for Acceptance of Through-wall Flaw in a Service Water System Valve Body
| ML25321A755 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/17/2025 |
| From: | Ellis K Duke Energy, Progress Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RA-25-0264 | |
| Download: ML25321A755 (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 Serial: RA-25-0264 10 CFR 50.55a November 17, 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:
Proposed Alternative for Acceptance of Through-wall Flaw in a Service Water System Valve Body Ladies and Gentlemen:
Pursuant to 10 CFR 50.55a(z)(2), Duke Energy Progress, LLC (Duke Energy) is proposing 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 is proposing to use alternative methods to calculate the stresses that are used in evaluating a Service Water (SW) system through-wall flaw. The proposed alternative is described in Enclosure 1.
A through-wall flaw has developed in control valve 1-E11-PDV-F068A, which connects the Unit 1 Residual Heat Removal (RHR) Heat Exchanger Service Water Discharge Header to the downstream Conventional and Vital Headers Drain. ASME Code Case N-869, Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 or 3 Piping Section XI, Division 1 is approved by the NRC for use and provides criteria to allow temporary acceptance of flaws, including through-wall flaws in Class 2 or 3 piping without performing repair or replacement activities in accordance with ASME Code Section XI, Article IWA-4000. Code Case N-869 includes guidance for evaluating flaws in piping; however, it does not address the evaluation of a flaw on a body of a valve. The proposed alternative applies ASME Section III 2007 Edition with 2008 Addenda, Design by Analysis approach in section NB-3200 for the evaluation.
Approval is requested prior to December 6, 2025 to prevent unit shutdown per the Technical Specifications until valve replacement can be made.
No new regulatory commitments have been made in this submittal.
If you have additional questions, please contact Ryan Treadway, Director - Nuclear Fleet Licensing, at 980-373-5873.
U.S. Nuclear Regulatory Commission RA-25-0264 Page2 General Manager - Nuclear Regulatory Affairs, Policy & Emergency Preparedness
Enclosures:
cc:
- 1. Proposed Alternative for Acceptance of Through-wall Flaw in Valve 1-E11-PDV-F068A : Digital Ultrasonic Thickness NOE Report dated November 6, 2025 : Liquid Penetrant Examination Report dated November 6, 2025 : Borescope Inspection Summary dated November 6, 2025 : SIA Calculation 2552884.301, "Brunswick Unit 1 1-E11-PDV-F068A Valve Evaluation" USNRC Region II Regional Administrator USN RC Senior Resident Inspector for BNP USN RC NRR Project Manager for BNP
RA-25-0264 Proposed Alternative for Acceptance of Through-wall Flaw in Valve 1-E11-PDV-F068A 9 Pages Follow
RA-25-0264 1 of 9 1.0 ASME CODE COMPONENT(S) AFFECTED:
A through-wall flaw has developed in a 16-inch x 20-inch automatic control valve (Valve No. 1-E11-PDV-F068A) which connects the Unit 1 Residual Heat Removal (RHR) Heat Exchanger Service Water Discharge Header (Line Number 1-E11-128-16-046) to the downstream Conventional and Vital Headers Drain (Line Number 1-SW-142-20-157A).
ASME Section XI Code Class: 3 Component
Description:
16-inch x 20-inch 90-degree angled, Valtek control valve on the Unit 1 Service Water System in between Line 1-E11-128-16-046 and Line Number 1-SW-142-20-157A.
Construction Code Year: USAS B31.1 -1967 Analysis Code Year: Section III 1980, including the Winter 1980 addenda Design Conditions:
Temperature: 215 °F Pressure: 400 psig (max)
Operating Conditions:
Temperature: 215 °F Pressure: 320 psig (max)
Material of Valve Body:
Aluminum Bronze, SB148 CD954 2.0 APPLICABLE CODE EDITION AND ADDENDA:
The current applicable Edition and Addenda of the ASME Code,Section XI is identified in Table 1.
Table 1 Plant/Unit ISI Interval ASME Section XI Code Edition/Addenda Interval Start Date Interval End Date1 Brunswick Steam Electric Plant (BNP), Unit 1 Fifth 2007 Edition, Through 2008 Addendum2 05/11/2018 05/10/2028 Note 1: Date listed is the currently planned interval end date. In accordance with IWA-2430(c)(1), this end date may be extended or shortened as necessary.
Note 2: Documented in Safety Evaluation Dated May 14, 2025, the NRC has approved use of Nonmandatory Appendix C of Section XI of the 2021 Edition of the ASME Code for all applicable pressure-retaining piping and components at Brunswick, Unit No. 1, for the remainder of the fifth 10-year inservice inspection (Agencywide Documents Access and Management System Accession No. ML25125A315).
RA-25-0264 2 of 9 3.0 APPLICABLE CODE REQUIREMENT:
In accordance with ASME Code Section XI, 2007 Edition with 2008 Addenda, subparagraph IWD-3120(b) requires that flaws in ASME Code Class components which do not meet the standards shall be subjected to supplemental examination, or to a repair/replacement activity. In regard to the flaw analysis, IWD-3500, "Acceptance Standards" for Class 3 components, states that the requirements of IWC-3500, "Acceptance Standards" for Class 2 components, may be used.
Additionally, IWA-4000 describes the repair/replacement activities to correct an unacceptable flaw.
Discovery of an area below the design minimum wall thickness in the structural portion of an ASME Code Class 1, 2, or 3 component is direct evidence of a flaw in the component.
The Code does not include analytical evaluation criteria for acceptance of through-wall flaws in pressure retaining base material of non-ferrous alloy pipe or fittings. Code Case N-869, "Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 or 3 Piping Section XI, Division 1," which has been conditionally approved by the U.S. Nuclear Regulatory Commission (NRC) in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 21, provides analytical evaluation rules for temporary acceptance of flaws in piping. Code Case N-869, however, does not apply to valve bodies.
4.0 REASON FOR REQUEST:
Duke Energy Progress, LLC (Duke Energy) is requesting a proposed alternative from the requirement to perform repair/replacement activities for the degraded valve body, which has a through-wall flaw. This alternative is requested due to the challenges of online code repair or replacement of valve 1-E11-PDV-F068A as described within this document.
On November 3, 2025, a through-wall flaw was discovered leaking in the Unit 1 Service Water (SW) system. The flaw is located on the 16-inch side of a 16 inch x 20 inch control valve body, connecting line 1-E11-128-16-046 to line 1-SW-142-20-157A (Figure 1). Currently there is approximately two drops per 30 minutes of external leakage from the piping system. ASME Code Case N-869 is approved by the NRC for use and provides criteria to allow temporary acceptance of flaws, including through-wall flaws in Class 2 or 3 piping without performing repair or replacement activities in accordance with ASME Code Section XI, Article IWA-4000. ASME Code Case N-869 includes guidance for evaluating flaws in piping, however, it does not address the evaluation of a flaw on a body of a valve. It is noted that Code Case N-513-5, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping and Gate ValvesSection XI, Division 1 latest NRC approved revision, is the sister code case that shares the same analytical equations from Section XI Appendix C, but with different pressure and augmented examination requirements. It also shares the same scope, with the addition of valve body ends.
The flaw in the valve at BNP is within the region identified in Code Case N-513-5, where it would be acceptable to use the straight pipe hand calculations to solve for an allowable flaw size. However, due to the complexity of the valve and to account for future degradation, a finite element analysis under limit load criteria from Section III is utilized in this proposed alternative instead of the hand calculation values in Section XI Appendix C.
Flaw Characterization:
During ultrasonic thickness (UT) examinations (Attachment 1) of the valve, the wall thickness of the valve body near the through-wall flaw and around circumference of the valve body at the inlet flange
RA-25-0264 3 of 9 show measured values to be at nominal wall thickness. The flaw is characterized as a pinhole with no additional cracking based on the Dye-Penetrant Testing Report dated November 6, 2025 (Attachment 2).
A borescope inspection (Attachment 3) was performed to examine the internal valve condition and to further characterize the degradation. Based on a review of borescope pictures and given the known size and contour of the inlet of the valve body, an area of material degradation in the area of the through wall flaw was conservatively estimated to be approximately 2-inches by 2-inches.
Code Case N-869, supplemented with finite element analysis, was then invoked to establish that the component remains structurally stable and not subject to catastrophic failure. Section 5.0 Augmented Examinations of the N-869 Code Case specifies that augmented examinations of the most susceptible locations on the affected system shall be performed, which are to be completed within 30 days of the initial detection. These inspections will include the sister valve on Unit 1, and the two comparable valves on Unit 2, as these are the only locations to check with similar conditions. The flaw on valve 1-E11-PDV-F068A will be re-inspected to monitor the flaw for growth and any leakage using the guidance provided by Code Case N-869 as modified by this proposed alternative.
The location of the flaw is the inlet side of valve 1-E11-PDV-F068A (Figure 1), which is located at the top of a vertical riser, approximately 45 feet from the bottom of the drop. From this low point the piping exits into a pipe tunnel and discharges into the top of the Circulating water discharge at near atmospheric conditions. Based on the large elevation difference and the small pressure drop in the discharge piping, the pressure conditions on the downstream side of this valve are at vacuum. The flaw is on the upstream side of the valve body which is currently filled with Well Water for layup, and is experiencing approximately two drops per 30 minutes of external leakage. The portion of the piping system containing the flaw is laid up with Well Water while not in service to help prevent brackish water corrosion. The Well Water Pressure is significantly lower than normal operating pressure. The flaw location would see normal operating pressure and temperature conditions when the RHR Heat Exchangers are in service for shutdown cooling or quarterly testing. When quarterly testing occurs, the daily leakage monitoring of this proposed alternative will take place at that time to ensure monitoring at normal system pressure and temperature.
Figure 1. Flaw Location
RA-25-0264 4 of 9 Hardship of Implementing Repairs The flaw within the affected valve is to be repaired by replacing the degraded valve body. Based on site specific and industry operating experience, valve replacement is the prudent repair method over weld repairs on cast aluminum bronze valves. Internal Brunswick OE shows a previous attempt to perform a weld repair on an aluminum bronze valve body was unsuccessful and rendered the valve unusable.
Additionally, valve repair vendors have indicated that successful weld repairs have been made on similar valves to remove casting defects prior to placing in service, however no OE has been found of repairs being made on valves that have been in service for a significant period of time. If such a repair were possible, it would likely require the valve body to be removed, and the following discussion on valve replacement confirms that valve body removal cannot be safely performed online.
Valve body replacement cannot be accomplished safely with the Unit online. The location of the affected valve is at the top of a vertical riser, approximately 45 feet above the bottom of the downcomer. From this point, the piping exits into a pipe tunnel and discharges into the top of the Circulating Water discharge at near atmospheric conditions. Based on the large elevation difference, and small pressure drop in the discharge piping, the pressure conditions where valve replacement work would be conducted are in a vacuum. No record of valve body replacement work at this location during online conditions could be found at Brunswick. While it is noted that with the valve removed, the vacuum conditions in the piping would most likely be reduced, working at this location with Reactor Building Closed Cooling Water (RBCCW) Service Water discharge flow entering the piping approximately 4 feet below the bottom flange of the valve (Figure 1), presents the risk of water exiting the piping and providing very little distance between the open work area and this service water discharge (up to 7300 gpm).
Additionally, in the final steps of valve reassembly, vacuum conditions at this location will be re-established and could present a personnel safety and equipment concern, particularly during lifting and rigging evolutions to set the valve plug in the body.
Additionally, the exfiltration from the Reactor Building that is caused by removing 1-E11-PDV-F068A from the system causes Secondary Containment to be inoperable unless the control room and site boundary dose consequence of the opening is found to be acceptable. Plant calculations show that the exfiltration from a nominal 20 diameter pipe opening contribute to dose consequences that would exceed regulatory limits. Technical Specification Limiting Condition for Operation (LCO) 3.6.4.1 states that for an inoperable Secondary Containment in Modes 1, 2, and 3, Secondary Containment must be restored within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. To replace 1-E11-PDV-F068A online from a Secondary Containment operability standpoint, several iterations of entry and exit from this LCO would be required via installation of blank flanges during disassembly and reassembly evolutions. These iterations are required as it would not be possible to perform the entire valve removal and reinstallation within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> LCO timeframe. The vacuum on the downstream piping would again introduce the safety hazard of repeatedly installing and removing blind flanges against the vacuum to restore Secondary Containment between disassembly and assembly of the major components of the valve.
Given the aforementioned concerns, completion of a valve body replacement on this component will require isolation of the Conventional Service Water system discharge header. Although the time period allowed by the associated Technical Specification Required Actions for the Service Water system may provide sufficient time to complete a body replacement, isolation of the affected line to perform this work would require shutdown of the Unit due to the loss of cooling water to the supported equipment. The Conventional Service Water system discharge header is the flow path for Service Water discharge from the Reactor Building Closed Cooling Water (RBCCW) system heat exchangers. The RBCCW system provides cooling for reactor building auxiliary equipment which includes Drywell Coolers, Spent Fuel Pool Cooling, Reactor Water Clean-up, Penetration Cooling and Recirc Pump Cooling. Without Drywell Cooling, atmospheric temperatures in the drywell would increase and exceed Technical Specification
RA-25-0264 5 of 9 limits. For this reason, replacement of the valve body cannot be performed with the Unit online. Because the affected repair would require isolating the 1A Loop of Residual Heat Removal Service Water (RHRSW) and RBCCW Service Water, this outage would require the Unit to enter Mode 5. For this shutdown, the reactor pressure vessel head would need to be removed and the reactor cavity flooded to allow placing the 1B Loop of RHR in Shutdown Cooling and Augmented Spent Fuel Pool Cooling (since RBCCW is out of service).
Plant shutdown activities result in additional 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 Duke Energy time for safe and orderly long term repair actions.
5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:
Duke Energy is requesting approval to apply ASME Section III 2007 Edition with 2008 Addenda, Design by Analysis approach in section NB-3200 to evaluate the non-uniform wall thickness and local thinning of the valve body. As a result, a detailed three-dimensional (3-D) finite element model (FEM) was utilized in the evaluation to calculate the stress field associated with the non-uniform wall thickness (Attachment 4). In accordance with 10 CFR 50.55a(z)(2), the proposed alternative 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.
Given that there is at least one reading below tmin identified, which is inferred by the through wall leak, the primary stress design criteria for Service Level A (Normal) and Level B (Upset) conditions are based on the limit load design criteria of Subparagraph NB-3228.1, Limit Analysis. NB-3228.1 states:
The limits on General Membrane Stress Intensity (NB-3221.1), Local Membrane Stress Intensity (NB-3221.2), and Primary Membrane Plus Primary Bending Stress Intensity (NB-3221.3) need not be satisfied at a specific location if it can be shown by limit analysis that the specified loadings do not exceed two-thirds of the lower bound collapse load. The yield strength to be used in these calculations is 1.5Sm.
The allowable Stress, S, is used instead of the Class 1 Design Stress Intensity, Sm. As a result, the S value is substituted for the Sm, in regard to establishing the yield stress defined above. This is consistent with the design rules, as the yield stress defined here is the equivalent of the Local Primary Membrane Stress, which uses an allowable stress of 1.5Sm for Class 1 components, or 1.5S for Class 2/3 and B31.1 components.
Subparagraph NB-3213.28, Limit Analysis Collapse Load states (note that references from the code refer to pipe, while the component being evaluated is a valve body):
The methods of limit analysis are used to compute the maximum load that a structure assumed to be made of ideally plastic material can carry. At this load, which is termed the collapse load, the deformations of the structure increase without bound.
A total load of 1/(2/3) = 150% of the applied maximum loads on the degraded section of pipe must be sustained without any portion of the pipe plastically collapsing (i.e., the membrane stress across an entire section for a given location does not exceed the defined yield stress of 1.5S).
The primary stress design criteria Service Level D (Faulted) conditions will be based on the limit load design criteria of Appendix F, F-1200(a), and will be based on the ASME Code,Section III,
RA-25-0264 6 of 9 Appendix F, Subparagraph F-1341.3. F-1341.3, which states:
Static or equivalent static loads shall not exceed 90% of the limit analysis collapse load using a yield stress which is the lesser of 2.3Sm and 0.7Su A total load of 1/0.9 = 111.1% of the applied faulted loads on the degraded section of pipe must be sustained without any portion of the pipe plastically collapsing (i.e., the membrane stress across an entire section for a given location does not exceed the defined yield stress of 2.3S or 0.7Su).
To that end, the 3-D FEM of the degraded piping and the appropriate maximum Normal/Upset and Faulted loads (internal pressure and piping moments) was evaluated. The total load, as a percentage of nominal load has been compared to the required 150% of the Service Level A/B (Normal/Upset) loads and 111.1% of the Service Level D (Faulted) loads as defined in the criteria identified above.
Based on the above criteria, the valve meets the limit load design criteria outlined in Subparagraph NB-3228-1, Limit Analysis with a conservatively extended flaw size of 6.2 inches axially by 5 inches circumferentially, with the addition of 20% wall loss (Attachment 4).
Consistent with the requirements of ASME Code Case N-869 Section 2, the following compensatory actions shall be performed until such time that the flaw is repaired.
(a) Frequent periodic examinations of no more than 30 day intervals shall be used to determine if flaws are growing and to establish the time, at which the detected flaw will reach the allowable size.
(b)
For through-wall leaking flaws, leakage shall be monitored daily to confirm the analysis conditions used in the evaluation remain valid. When the RHR Heat Exchangers are in service at normal operating temperature and pressure for quarterly testing, the daily leakage check will be performed during this time.
(c)
If engineering evaluation of examinations reveal flaw growth rate to be unacceptable, a repair/ replacement activity shall be performed.
(d)
Repair/replacement activities shall be performed no later than when the predicted flaw size from either periodic Inspection or by flaw growth analysis exceeds the acceptance criteria (6.2 inches axially or 5 inches circumferentially), or during the Spring 2028 scheduled refueling outage, whichever occurs first. Repair /replacement activities shall be in accordance with IWA-4000.
Flooding Considerations:
The Service Water valve containing the leak is located in a vertical pipe chase in the Northwest corner of the Reactor Building. The orientation of the leak is on the bottom side of the valve such that any leakage is 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.
RA-25-0264 7 of 9 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 a maximum of 20 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. It is acknowledged that the flaw described in the BNP alternative approved via safety evaluation dated May 16, 2025 (ADAMS Accession No. ML25135A417) is a hole in the reducer on line 1-SW-142-30-157, which if it leaked, water would flow into the same sump. The alternative conservatively limited leakage from the reducer hole to 10 gpm. The cumulative allowed leakage from both flaws of 30 gpm is within the 33 gpm margin for sump pump capacity. Additional confidence in the margin is provided by the fact that the reducer flaw is under vacuum and is not expected to leak externally. The daily monitoring described in this 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 valve containing the flaw is in piping 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 20 gpm leak rate limit, repair/replacement activities will be performed to restore the integrity of the valve.
Repair/Replacement Timeline:
Exhaustive searches have been performed to locate an existing replacement valve within the industry that could be installed in the next Spring 2026 refueling outage, with no success. In addition, current procurement lead times for a new valve of 30-50 weeks eliminate the possibility of replacing this valve in the next refueling outage, currently scheduled to begin on February 28, 2026.
6.0 DURATION OF PROPOSED ALTERNATIVE:
The proposed alternative is requested to be authorized to remain in place until a new valve can be procured, and then installed during the Spring 2028 Refueling Outage which is currently scheduled to start on March 4, 2028. Valve replacement in the Spring 2028 Refueling Outage versus the Spring 2026 Refueling Outage is acceptable considering that the analysis and basis used for this proposed alternative are not time dependent. Duke Energy will continue to monitor the flaw and ensure it stays within the acceptance criteria of the evaluation until the valve can be replaced.
RA-25-0264 8 of 9
7.0 PRECEDENTS
ML25135A417, May 16, 2025 - Brunswick Steam Electric Plant, Unit 1 - Proposed Alternative for Acceptance of Through-Wall Flaw in Reducer (EPID L-2025-LLR-0038)
This proposed alternative was submitted to support the expansion of Code Case N-513-5 to the small end transition region of a reducer for continued operation at BNP.
Flaw evaluations of Section III limit load were applied to predict allowable flaw size via finite element analysis. Similarities are: flaws occurred in the same system,Section III limit load criteria, finite element analysis was used, and both asked for expansion of the code case scope of components. The current proposed alternative was largely structured after this precedent.
ML080580577, March 26, 2008 - McGuire Nuclear Station, Unit 1, Request for Relief 06-MN-001, for Third 10-Year Interval Inservice Inspection Program Plan Regarding Repair of Chemical Volume And Control System Valve 1NV-240 (TAC No. MD6274)
This request supported an evaluation performed in 2007 on a Class 2 isolation valve for a pinhole leak in the Charging Supply System for continued operation at MNS. In the provided evaluation, the allowable size of the flaw was determined by the use of limit load and flaw criteria from Section XI, Appendix C. This evaluation was requested for relief before Code Case N-869 was created and lead to the creation of Code Case N-869 and push for more components to be accepted in Code Case N-513.
Similarities are: high pressure valve, flaw on the valve, and limit load criteria.
ML19003A188, September 18, 2019 - Virgil C. Summer Nuclear Station, Unit No. 1 - Relief Request (RR-4-16) for use of ASME Code Case N-513-4 and NRC Generic Letter 90-05 (EPID L-2018-LLR-0100)
This request supported the expanded use of Code Case N-513-4. At the time (2019) N-513-4 was not yet approved, and the new revision of N-513 supported additional components from the previous revision. These additional components are now approved and incorporate flaws in valve bodys in the newest N-513 approved revision. N-513 is the sister Code Case of N-869, and the code cases use the same analytical approaches from Section XI Nonmandatory Appendix C.
ML12201A256, August 16, 2012 - South Texas Project, Unit 2 - Request for Relief RR-ENG-3-08 from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Requirements for Repairs to Class 3 Valves in Essential Cooling Water System Piping (TAC No. ME8159)
This request submitted in March 2012 requested relief from the requirements of IWA-5250 of Section XI of the ASME BPV for repairs of flaws found in bodies of two Class 3 cast aluminum-bronze valves in Essential Cooling Water system piping.
Specifically, the licensee requested to defer code repairs of the flaws until after STP Unit 2 was returned to service scheduled for mid-April 2012 at the time of the request.
The relief was needed due to nonavailability of necessary parts until after the scheduled restart date.
RA-25-0264 9 of 9 ML14217A203, September 4, 2014 - Millstone Power Station, Unit No. 3 - Relief from the Requirements of the ASME Code Section XI, Requirements For Repair/Replacement of Class 3 Service Water Valves (TAC No. MF1314)
This request supported a deferral of repair to the next refueling outage. Similarities are: Class 3 line, aluminum bronze valve.
8.0 REFERENCES
8.1 ASME Boiler and Pressure Vessel Code, Code Case N-869, Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 or 3 Piping Section XI, Division 1, July 5th, 2018.
8.2 ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda 8.3 ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, 1980 Edition with 1980 Winter Addenda 8.4 ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda
RA-25-0264, Attachment 1
, Attachment 1 Digital Ultrasonic Thickness NDE Report dated November 6, 2025 2 Pages Follow
RA-25-0264, Attachment 1
([_-, DUKE
~ ENERGY.
Site/Unit: BNP
/ _1 _____ _
Summary No.: 1-Ell-PDV-F068A Workscope: BOP Code:
N/A Drawing No.:
D-25037 SH0001 UT Calibration/Examination Procedure: NDE-NE-ALL-6403 Procedure Rev.: _3 ______________ _
Work Order No.: 20756874-01 Cat/Item:
_N~/_A~/_N,_/A _________ _
Location:
REACTOR BLDG
==
Description:==
2045 RHRSW Outage No.:.!:N!l./!:!A~--------
Int / Per:.!:N!l./!:!A~--------
Report No.: UT-25-1006 Page: 1 of 2 System ID:
_2_0_4_5 ______________________________________
Material: Aluminium - Bronze Component ID: 1-E11-PDV-F068A Size/Length:
N/A Thickness/Diameter: ~1~6:...." ___________ _
Limitations*
Yes - Gusset location Start n 1me:
0930 lnlS I
Fi
- h Tme 1045 Instrument Settings Search Unit Cal.
Axial Oriented Search Unit Checks Time Date Serial No.:
20COOGKD Serial No.:
13G012C6 Calibration Signal Sweep Sound Path Manufacturer:
GEIT Initial Calib 0700 11/06/2025 Reflector Amplitude%
Division Manufacturer:
GE Model :
USN 60 NDA18 0.375" BDH BO 1.9 0.375" Size: 0.25" Model: Gamma Inter. Calib.
N/A Linearity:
LN-25-1001 1.125" BDH 80 5.6 1.125" Freq.:
2.25 MHz Center Freq: N/A Inter. Calib.
0900 11/06/2025 1.500" Step 80 8.2 1.500" Delay:
2.1 275 Range:
2.000" Exam Angle: 0 Squint Angle: N/A Inter. Calib.
N/A 2.000" Step 80 10.0 2.000*
M'tl Cal/Vel:
0.2228 Reject:
0%
Measured Angle: N/A Mode:
Long.
Final Calib.
1150 11/06/2025 Damping:
500 Ohms Inst. Freq.:
2.25MHz Exit Point: N/A
- of Elements: 1 Couplant PRF:
Auto High Rectify:
Full Wave Circumferential Oriented Search Unit Config.:
Single Focus:
N/A Cal. Batch:
21C032 Voltage:
450V Pulse Width: 220 ns Calibration Signal Sweep Sound Path Shape:
Round Contour: Flat Type:
ULTRAGELII Reflector Amplitude %
Division Pulser Type:
Sguare N/A See Axial Wedge Style:
Mfg:
MAGNAFLUX Exam Batch: 21C032 Ax. Gain (dB): 33.1 Circ. Gain (dB): 33.1 Search Unit Cable Type:
ULTRAGELII 1
Screen Div= 0.2 of Sound Path Type: RG-174 Length : 6' No. Conn. : _o __
MAGNAFLUX calibration Block Mfg:
Scan Coverage Reference/Simulator Block Cal Block No.:
AL - Bronze Cal Blk Upstream~
Downstream ~
Reference Block Gain Signal Sweep Scan dB:
- Reflector Sound Path Material:
AL-Bronze cw~
ccw ~
dB Amplitude%
Division Scan dB:
- Serial No.:
N/A Thickness:
1.5
- 2.5 Dia.: Flat N/A Exam Surface:
O.D.
Type:
N/A Cal. Blk. Temp.: 72 Temp. Tool: 6500048 Surface Condition: Smooth Comp. Temp.:
68 Temp. Tool: 6500048 Material: N/A Recordable Indication{s):
Yes D No ~
Reject 0 (If Yes, Ref. Attached Ultrasonic Indication Report.)
Comments:
Range adJusted m field.
- Gain adjusted as necessary.
Results:
Accept D Info ~ _______________
Reviewed Previous Data:
..:..;N:<./.:..:A'--------
Examiner Level II-N.
Exam Date RKvi~er Ransom Gre l.
11/06/2025
. egan Jr NOE LIii Examiner
- Leve7, Site Review le.~-
10 0
N/A N/A Examiner Level Signature Date ANII Review Signature Date N/A N/A
RA-25-0264, Attachment 1
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~ ENERGY.
Summary No.: 1-E11-PDV-F068A Supplemental Report Sketch or Photo:
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- J Report No.: UT-25-1006 Page:..;;:2:__ __
of 2
RA-25-0264, Attachment 2
, Attachment 2 Liquid Penetrant Examination Report dated November 6, 2025 1 Page Follows
NOE Report#
100282 DUKE ENERGY Unit: 0 [!11 02 03 LIQUID PENETRANT EXAMINATION REPORT Page WO:
of 20756874-07 Plant: [!)BNP HNP RNP MNS CNS DONS Keowee Other:
N/A Time:
0800 Date:
111512s ID/Weld No.:
Pin hole area 2045 Line/Component:
1-E11-PDV-F068A System:
Code:
[!) ASME 111 0 ASME VIII ASME XI B31.1 AWS D1.1 Other:
N/A Code Class:
3 Material: Des ss [!}other AL-BRONZE Stage of Mfg.: Root Intermediate Final Repair [!) Base Metal N/A Thickness: _._81_0*_* _ OD/Length: __ 3_"_X_4'_' _
Surface Finish: As-Welded Ground Machined [!]Other:
BM Drawing No.:
FSP-2609SSH.101 Joint Design: Fillet open-butt socket BMR cavity[!) NIA BM Procedure No: -N-D-E--N-E-A_L_L--4-,o-,--R-e_v_: __
s Accept Criteria or Attach:,r-J.li) C/~f-:? Technique Attach:
White Light Meter S/N: ___
N_IA ___ Calibration Due:
N/A Illumination Level Verified:
> 1oofc Black Light Meter S/N:
N/A Calibration Due:
N/A Illumination Level Verified:
D >1000 µW/cm2 Character Card S/N:
020150002*
Calibration Due Not Applicable Visible Characters Resolved:
0 VT-1 Thermometer S/N:
Gsoo48 Calibration Due:
1_11_s_12_0_26 __ Temperature:
OF 68 Examination Type:
0 Visible D Fluorescent [!) Solvent Removable D WaterWashable Cleaner:
SPOTCHECK SKC-S Batch No.:
24G01C Dry Time:
5 Minutes Penetrant:
SPOTCHECK SKL-SP2 Batch No.:
21F04K Dwell Time:
10 Minutes Cleaner Remover: SPOTCHECK SKC-5 Batch No.:
24G01C Dry Time:
5 Minutes Developer:
SPOTCHECK SKD-52 Batch No.:
21M11C Evaluation Time:
10 Minutes Area of Coverage Achieved:
0 100% D Other:
N/A If< 100%, Reason:
N/A NCR#
N/A Sketch and Remarks
- EPRI CARD VERIFIED PRE/POST EXAM
/ ;./ t) ~T7bµS..C1'://
Due to casting defects rounded indications are expected in the base materiatlNo cracks were noted with the exam. Below are rounded indications and number 3 is the location of the through-wall pin hole as noted on the NCR engineering generated 025746211.
Engineering acceptance criteria looking for cracks. No cracks noted.
I Indication No:
Indication Descri tion: Rounded or Linear/ Size I Remarks 5 Rounded in line separated by <1 /16" in transition area 2
roun ed )116" <1 /8" above transition area 3
rounded> 1 /16" <1 /8" through wall location in transition area 4
rounded > l /16" <1 /8" in transition area NRI
[!] Examination Results Acceptable Level: ffoate: 11/6/25 Level: N/A Date:
N/A Reviewer: (print/sign) tJOO'ceb 16o.l +ez.c (K'"
Level:% Date: 11 /6/25 ANII Review (as applica
):
~
n nl.;2 Date:
J',1\\A NDE-NE-ALL-4101-F1 Rev.008 0 Accept Reject 0 Accept Reject 0 Accept Reject 0 Accept Reject 0 Accept Reject Reject NCR#:
N/A N/A RA-25-0264, Attachment 2
RA-25-0264, Attachment 3
, Attachment 3 Borescope Inspection Summary dated November 6, 2025 4 Pages Follow
Borescope Inspection Summary (11/06/2025):
Engineering (Sexton), witnessed the borescope inspection of the inlet side of the valve, performed by Trident. Trident was able to access the valve and examine the entire circumference of the inlet area of the valve. The inspection did note a buildup of material at the bottom-center (6 oclock position) of the valve inlet and two small "aws near this buildup. The Trident team attempted to retrieve a sample of the buildup, but the material was too hard to collect remotely with a grabber. No other degradation was noted around the inlet of the valve body. Condition of the rest of the valve was consistent with conditions noted during the 2018 internal inspection. Based on todays inspection, the "aw appears to be localized in the bottom center of the inlet of the valve, at the bore radius as it enters the valve cavity towards the cage trim. Photos below.
Figure 1. Material Buildup on bottom center of inlet of valve (looking downstream).
RA-25-0264, Attachment 3
Figure 2. Material Buildup on bottom center of inlet of valve (looking downstream).
RA-25-0264, Attachment 3
Figure 3. Material Buildup on bottom center of inlet of valve (looking downstream).
RA-25-0264, Attachment 3
Figure 4. Small "aw / hole at inlet of valve (located at 0530 position if looking downstream).
RA-25-0264, Attachment 3
RA-25-0264, Attachment 4
, Attachment 4 SIA Calculation 2552884.301, Brunswick Unit 1 1-E11-PDV-F068A Valve Evaluation 22 Pages Follow
CALCULATION PACKAGE File No.: 2552884.301 Project No.: 2552884 Quality Program Type:
Nuclear Commercial PROJECT NAME:
U1 BNP Finite Element Analysis and N-869 CONTRACT NO.:
03021365 00124 CLIENT:
Duke Energy PLANT:
Brunswick Nuclear Plant, Unit 1 CALCULATION TITLE:
Brunswick Unit 1 1-E11-PDV-F068A Valve Evaluation Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) &
Checker(s)
Signatures & Date 0
1 - 20 A A-2 Initial Issue Yanni Patten 11/12/25 Grant Wu 11/12/25 Minghao Qin 11/12/25 1
Page 6 and 17 Editorial Changes Yanni Patten 11/16/25 Grant Wu 11/16/25 Minghao Qin 11/16/25 RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 2 of 20 F0306-01R4 Table of Contents 1.0 OBJECTIVE.............................................................................................................. 4 2.0 METHODOLOGY...................................................................................................... 4 3.0 DESIGN INPUTS....................................................................................................... 6 3.1 Geometry and Material................................................................................... 6 3.2 Piping Loads.................................................................................................. 6 3.3 Wall Thickness Data and Borescope.............................................................. 7 4.0 ASSUMPTIONS........................................................................................................ 7 5.0 FINITE ELEMENT MODEL........................................................................................ 8 5.1 Geometry and Element Selection................................................................... 8 5.2 Boundary Conditions...................................................................................... 8 5.3 Beam Elements.............................................................................................. 8 5.4 Internal Design Pressure................................................................................ 8 5.5 Piping Moment Loads.................................................................................... 8 6.0 LIMIT LOAD COLLAPSE EVALUATIONS................................................................. 9 7.0 RESULTS OF ANALYSIS.......................................................................................... 9
8.0 CONCLUSION
S........................................................................................................ 9
9.0 REFERENCES
........................................................................................................ 10 COMPUTER FILES..................................................................................... A-1 RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 3 of 20 F0306-01R4 List of Tables Table 1: Piping Loads...................................................................................................................... 7 Table 2: Results............................................................................................................................... 9 List of Figures Figure 1. Leak Location.................................................................................................................... 4 Figure 2: Through - Wall Leak Picture............................................................................................ 11 Figure 3: Borescope Inspection..................................................................................................... 12 Figure 4: PT Inspection Report....................................................................................................... 13 Figure 5: UT Inspection Report....................................................................................................... 13 Figure 6: Finite Element Model Geometry, Global View.................................................................. 14 Figure 7: Finite Element Model Geometry, Local View.................................................................... 14 Figure 8: Mechanical Boundary Conditions..................................................................................... 15 Figure 9: Applied Pressure Loads................................................................................................... 15 Figure 10: Example of Applied Piping Loads (+Y Direction)............................................................ 16 Figure 11: von Mises Stress Plot with Hole, SRSS +Y Moment at 179.84%.................................... 16 Figure 12: von Mises Stress Plot with Hole, SRSS +X Moment at 179.84%.................................... 17 Figure 13: von Mises Stress Plot with Hole, SRSS -X Moment at 179.84%..................................... 17 Figure 14: von Mises Stress Plot with Hole, SRSS +Z Moment at 175.98%.................................... 18 Figure 15:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +Y Moment at 179.84%
18 Figure 16:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +X Moment at 179.84%
19 Figure 17:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS -X Moment at 179.84%
19 Figure 18:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +Z Moment at 175.98%
20 RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 4 of 20 F0306-01R4 1.0 OBJECTIVE A through-wall leak was discovered in the Unit 1 Service Water (SW) system at Brunswick Nuclear Plant (BNP). The leak is located on the 16-inch side of a 16 inch x 20 inch control valve, with a valve ID of 1-E11-PDV-F068A. See Figure 1 for location of the identified leak on BNP provided drawing [10].
The 16-inch x 20-inch valve is cast aluminum bronze, SB-148 CDA 954, and is safety-related Class 3 piping [5].
Structural Integrity Associates, Inc. (SI) has been contracted to perform a detailed finite element analysis incorporating wall thinning and through-wall wall loss determined from the UT thickness data and borescope inspections to justify continued operation utilizing ASME Code,Section III design by analysis criteria. The objective of this calculation is to develop the finite element model, define the loads, and to perform limit analysis - collapse load evaluations utilizing the ASME Code,Section III design by analysis rules [6].
The final of this calculation is to provide BNP with an allowable hole size based on ASME Code structural stability acceptance limits.
Figure 1. Leak Location 2.0 METHODOLOGY The original Code of Construction is B31.1, 1967 Edition [7]. Per Reference [5], this valve was designed to Section III 1980 edition with Winter 1980 addenda [8].
To evaluate non-uniform wall thickness/local thinning, or complex or combined stresses, detailed design qualification rules are taken from the ASME Code Design by Analysis approach (Section III, NB-3200 [6]). As a result, a detailed three-dimensional (3-D) finite element model (FEM) will be utilized RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 5 of 20 F0306-01R4 in subsequent evaluations to calculate the stress field associated with localized thinning, which is consistent with the Design by Analysis approach of NB-3200.
Given that there is at least one reading is through wall, the primary stress design criteria for Service Level A (Normal) and Level B (Upset) conditions are based on the limit load design criteria of Subparagraph NB-3228.1, Limit Analysis [6]. NB-3228.1 states:
The limits on General Membrane Stress Intensity (NB-3221.1), Local Membrane Stress Intensity (NB-3221.2), and Primary Membrane Plus Primary Bending Stress Intensity (NB-3221.3) need not be satisfied at a specific location if it can be shown by limit analysis that the specified loadings do not exceed two-thirds of the lower bound collapse load. The yield strength to be used in these calculations is 1.5Sm.
The allowable Stress, S, is used instead of the Class 1 Design Stress Intensity, Sm. As a result, the S value is substituted for the Sm, in regard to establishing the yield stress defined above. This is consistent with the design rules, as the yield stress defined here is the equivalent of the Local Primary Membrane Stress, which uses an allowable stress of 1.5Sm for Class 1 components, or 1.5S for Class 2/3 and B31.1 components.
Subparagraph NB-3213.28, Limit Analysis Collapse Load [6] states:
The methods of limit analysis are used to compute the maximum load that a structure assumed to be made of ideally plastic material can carry. At this load, which is termed the collapse load, the deformations of the structure increase without bound.
Thus, a total load of 1/(2/3) = 150% of the applied maximum loads on the degraded section of pipe must be sustained without any portion of the pipe plastically collapsing (i.e., the membrane stress across an entire section for a given location does not exceed the defined yield stress of 1.5S).
The primary stress design criteria Service Level D (Faulted) conditions will be based on the limit load design criteria of Appendix F, F-1200(a), and will be based on the ASME Code,Section III, Appendix F, Subparagraph F-1341.3 [8]. F-1341.3 states:
Static or equivalent static loads shall not exceed 90% of the limit analysis collapse load using a yield stress which is the lesser of 2.3Sm and 0.7Su Thus, a total load of 1/0.9 = 111.1% of the applied faulted loads on the degraded section of pipe must be sustained without any portion of the pipe plastically collapsing (i.e., the membrane stress across an entire section for a given location does not exceed the defined yield stress of 2.3S or 0.7Su).
To that end, the 3-D FEM of the degraded piping and the appropriate maximum Normal/Upset and Faulted loads (internal pressure and piping moments) will be evaluated. Additional uniform wall thinning will be applied until the structure meets the limit load criteria, or plastically collapses (in terms of FEM this will be at the point of numerical instability). The total load, as a percentage of nominal load will be compared to the required 150% of the Service Level A/B (Normal/Upset) loads and 111.1% of the Service Level D (Faulted) loads as defined in the criteria identified above.
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 6 of 20 F0306-01R4 3.0 DESIGN INPUTS 3.1 Geometry and Material Valve dimensions were scaled from reference [4.a]. The 16" OD by 0.375" thick pipe length was taken from reference [4.b]. The 16" welded flange dimensions were taken from reference [11].
The following are the inputs for the valve and piping used in this analysis. The 1980 ASME Code only provides values for Maximum Allowable Stress, therefore the Young's Modulus will be taken from the 1986 ASME Code [14] and the Poisson's Ratio and density will be taken from the 2007 ASME Code
[13].
Valve Body o
SB-148 CDA954 [2]
Maximum Allowable Stress, S at 215 °F: 18,700 psi, interpolated from [8, Table I-8.4]
Young's Modulus at 215 °F: 14,570 ksi, interpolated from [14, Table I-6.0]
Poisson's Ratio: 0.33 [13, Table NF-1]
Density: 0.274 lbm/in3 [13, Table NF-2]
Piping and Flange o
C71500 [4.b]
Maximum Allowable Stress, S at 215 °F: 7,840 psi, interpolated from [8, Table 1-8.4]
Young's Modulus at 215 °F: 21,440 ksi, interpolated from [14, Table I-6.0]
Poisson's Ratio: 0.33 [13, Table NF-1]
Density: 0.323 lbm/in3 [13, Table NF-2]
Flange Bolts [4.c]
o SB-164 No. 4400 or 4405 (Hot Work) [4.c]
Maximum Allowable Stress, S at 215 °F: 12,500 psi, interpolated from [8, Table I-7.3]
Young's Modulus at 215 °F: 25,340 ksi, interpolated from [14, Table I-6.0]
Poisson's Ratio: 0.32 [13, Table NF-1]
Density: 0.319 lbm/in3 [13, Table NF-2]
3.2 Piping Loads The piping max operating conditions are as follows [2 and 3]:
Max Operating Pressure is 320 psig Max Operating Temperature is 215 °F Piping moments are provided from Reference [9] and are listed in Table 1.
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 7 of 20 F0306-01R4 Table 1: Piping Loads Loads Mx (ft-lb)
My (ft-lb)
Mz (ft-lb)
DW 2,049
-833 1,023 T1 1,468 3,081
-621 T2
-7,939
-11,455 2,246 T3
-8,632
-11,027 2,143 T4 696
-431 104 T5
-559
-927 324 T6
-7,884
-11,457 2,287 T7
-8,013
-11,388 2,256 T8
-8,335
-11,239 2,152 T9 894 959
-337 T10 1,425 3,084
-656 T11 1,531 3,028
-630 T12 1,804 2,901
-543 OBE 2,536 1,176 1,467 DBE 3,795 1,755 2,300 Tmax range 10,436 14,541 2,943 Per stress report [9] and stress analysis isometric drawing [10], Y direction in Table 1 is the axial direction of the 20" pipe, which is aligned with the Y-axis in the finite element model (FEM).
In calculating the piping loads to apply to the model, the moment in each axis was SRSSed to determine the total moment. To bound the limiting directions that the moment could occur on the piping, the following loading directions were considered:
- 1. Apply the SRSS of the calculated moment loading as a FEM +Y moment.
- 2. Apply the SRSS of the calculated moment loading as a FEM +X moment.
- 3. Apply the SRSS of the calculated moment loading as a FEM -X moment.
- 4. Apply the SRSS of the calculated moment loading as a FEM +Z moment.
+Z was the most bounding direction.
3.3 Wall Thickness Data and Borescope The nominal wall thickness used for this evaluation conservatively takes the minimum wall thickness from the UT report [11] of 1". A 6.2" axial by 5" circumferential hole was modeled at the flaw location to conservatively represent material degradation. The hole size was determined based off the borescope photos [12], to eliminate any area withing the area of wastage.
An additional case was run where the entire portion was uniformly thinned by 20% wall thickness, therefore the maximum wall thickness at any location within the valve body, fully circumferentially around the valve where the flaw was found is 0.8" thick. This is to account for any additional degradation that may occur to the wall during the remainder of the determined operating period.
4.0 ASSUMPTIONS The following assumptions are used in this evaluation:
The valve body dimensions were scaled from references [4.a and 11], with dimensions rounded to the nearest 1/16" where deemed reasonable.
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 8 of 20 F0306-01R4 The gusset plate underneath the flaw location was not modeled, as the model was configured with a hole in this area and the gusset would not have an area to attach to the valve flange. The gusset would add more stiffness and wall thickness, therefore removing it would be more conservative.
The transition from the 16" end of the valve to the valve body was not modeled. This is conservative as the additional material would help reduce the stresses and strains in the area.
5.0 FINITE ELEMENT MODEL A three dimensional (3-D) FEM of the valve and piping is developed with the ANSYS finite element analysis software [15].
5.1 Geometry and Element Selection The geometry consists of the valve body, the 16" raised neck flange, and the 16"x0.375" piping. The FEM is constructed using ANSYS 20-node SOLID186 structural solid elements. The valve body and piping have at least 4 elements through the thickness. Figure 6 shows the global view of the geometry used to the develop the FEM, while Figure 7 shows a local view of the geometry.
5.2 Boundary Conditions The flange face of the 20" end of the valve is restrained in the axial and circumferential directions. The radial direction is left unconstrained to allow for expansion due to internal pressure. The free end of the piping is left free in order to apply the bending moment and torque loads. The applied boundary conditions are shown in Figure 8.
5.3 Beam Elements Beam188 elements were used to connect the flange to the valve body. Beam elements constrain the faces of the holes on either side of the connection.
5.4 Internal Design Pressure A pressure of 320 psig [3] and is applied to the interior surfaces of the model. In order to properly model the longitudinal stresses caused by pressure on the interior surface of the piping, an induced end-cap load is applied to the unconstrainted free end of the valve and piping, calculated as follows:
Pendcap = P ID2 4
- where, Pend-cap = End cap force on free end (lbf.)
P
= Internal pressure (psi)
ID
= Inside diameter of the free end (in.)
The end-cap force is applied to the pilot node on free end surface of the valve and piping. The pilot node is described in Section 5.5. The internal pressure and end-cap pressures are shown in Figure 9.
5.5 Piping Moment Loads The piping loads are applied by making use of a pilot node to transfer the loading. The TARGE170 target element type from the ANSYS element library is used to create the pilot node. The CONTA174 contact element is used to create the contact surfaces that are aligned with the pilot node. The pilot node and surfaces are bonded together such that the loads applied to the pilot node are transferred to RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 9 of 20 F0306-01R4 the respective surfaces on the pipe. Figure 10 shows the applied torque and bending moments in the model.
The load combination for Service Level A/B is DW+OBE+ThermalMax, and for Service Level D is DW+DBE+ ThermalMax.
6.0 LIMIT LOAD COLLAPSE EVALUATIONS Per Section 2.0, the static or equivalent static loads shall not exceed 2/3 (for Normal/Upset) or 90% (for Faulted) of the limit analysis collapse load. Thus, a total load of 1/(2/3) = 150% (Normal/Upset) or 1/0.9
= 111.1% (Faulted) of the applied load must be sustained without plastic collapse. The ratio of the required load between the Normal/Upset and Faulted combinations is 150%/111.1%=35%.This ratio is higher than the Service Level D to Service Level A/B loads (max. of 15%). Therefore, only Service Level A/B loading is required to be analyzed as it controls the results of the analysis.
7.0 RESULTS OF ANALYSIS The models with a hole and with uniform thinning successfully meet the criteria specified in Section 6.0 for both Normal/Upset and Faulted conditions. Table 2 tabulates the results for the evaluations.
Table 2: Results Model Percentage of Service Level A/B Load Applied (1)
Model with 6.2"x5" hole in the valve flange 175.98%
Model with 6.2"x5" hole and 20%
uniform thinning around the entire valve flange 175.98%
Notes:
1.
Using Service Level A/B loads, a passing score of greater than 150% is needed. Only Service Level A/B loads need to be considered as discussed in Section 6.0.
Von Mises stress plots are shown for the last converged step for each model in Figure 11 through Figure 17.
8.0 CONCLUSION
S Based on the application of UT data and Borescope inspections provided to SI [11 and 12] for the valve, the components meet the limit load design criteria outlined in Subparagraph NB-3228.1, Limit Analysis [6] with a conservatively extended hole size of 6.2 axially x 5 circumferentially with 20%
additional uniform wall loss, fully circumferentially around the hole size.
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 10 of 20 F0306-01R4
9.0 REFERENCES
- 1. ASME B16.5, "Pipe Flanges and Flanged Fittings: NPS 1/2 through NPS 24, Metric/Inch Standard," 1953.
- 2. E-mail from J. Goelz (BNP) to Y. Patten (SI) "RE: [EXTERNAL] Re: Brunswick Service Water Leak," 11/4/2025, SI File No. 2552884.208.
- 3. E-mail from J. Goelz (BNP) to Y. Patten (SI) "RE: [EXTERNAL] Proposal for work of AL-Bronze Valve," 11/7/2025, SI File No. 2552884.208.
- 4. Drawings
- a. Carolina Power & Light Company Dwg. Dwg. O-FP-81463 Sheet 1 of 2, Revision A, "16" x 20" Pressure Differential Valve (Motor Operated), SI File No. 2552884.205.
- b. Duke Energy Dwg. FSP-26095 Sheet 101, Revision 1, "Reactor Building RHR Heat Exchanger "1A" Service Water Discharge Elev. 50'-0" Piping Line Isometric," SI File No.
2552884.205.
c.
Duke Energy Dwg. FSP-26095 Sheet 102, Revision 3, "Reactor Building Service Water Discharge from RHR Heat Exchanger "1A" Elevation 50'-0" Piping Isometric,", SI File No. 2552884.205.
- 5. Duke Calculation, 0RHR-0033, Seismic Qualification for Valtek Automatic Control Valves Mark No. 1-E11-PDV-F068A, 1-E11-PDV-F068B, 2-E11-PDV-F068A, 2-E11-PDV-F068B (Design Report No. 35411-1, 2, 9, Revision 001, April 2025, SI File No. 2552884.206.
- 6. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components, 2007 Edition with Addenda through 2008.
- 7. USAS B31.1.0, Power Piping, 1967 Edition.
- 8. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components, 1980 Edition with Winter 1980.
- 9. Duke Stress Report Calculation, SA-SW-761/762/763, June 1994, SI File No. 2552884.201.
- 10. Drawing D-28046, SH. 763, Reactor Building - Unit 1, Service Water System, Revision 1, March 1995, SI File No. 2552884.205.
- 13. ASME Boiler and Pressure Vessel Code,Section II Part D, Materials, 2007 Edition.
- 14. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, Division I Appendices, 1986 Edition.
- 15. Ansys Workbench, Release 24.2, ANSYS, Inc., May 2024.
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 11 of 20 F0306-01R4 Figure 2: Through - Wall Leak Picture RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 12 of 20 F0306-01R4 Figure 3: Borescope Inspection RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 13 of 20 F0306-01R4 Figure 4: PT Inspection Report Figure 5: UT Inspection Report RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 14 of 20 F0306-01R4 Figure 6: Finite Element Model Geometry, Global View Figure 7: Finite Element Model Geometry, Local View RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 15 of 20 F0306-01R4 Figure 8: Mechanical Boundary Conditions Figure 9: Applied Pressure Loads (In this figure, the pressures shown are 200% of Design Pressure)
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 16 of 20 F0306-01R4 Figure 10: Example of Applied Piping Loads (+Y Direction)
(In this figure, the piping loads shown are 200% of Service Level A/B (Normal/Upset) Case.)
Figure 11: von Mises Stress Plot with Hole, SRSS +Y Moment at 179.84%
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 17 of 20 F0306-01R4 Figure 12: von Mises Stress Plot with Hole, SRSS +X Moment at 179.84%
Figure 13: von Mises Stress Plot with Hole, SRSS -X Moment at 179.84%
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 18 of 20 F0306-01R4 Figure 14: von Mises Stress Plot with Hole, SRSS +Z Moment at 175.98%
Figure 15:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +Y Moment at 179.84%
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 19 of 20 F0306-01R4 Figure 16:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +X Moment at 179.84%
Figure 17:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS -X Moment at 179.84%
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page 20 of 20 F0306-01R4 Figure 18:. von Mises Stress Plot with Hole and 20% Uniform Thinning, SRSS +Z Moment at 175.98%
RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page A-1 of A-2 F0306-01R4 COMPUTER FILES RA-25-0264, Attachment 4
File No.: 2552884.301 Revision: 1 Page A-2 of A-2 F0306-01R4 File Name Description 2552884.301.R0.wbpz ANSYS Workbench archive file containing both models and all moment loading runs loads.xlsx Calculation of piping loads applied to the model MaterialProps.xlsx Linear interpolation of material properties used in the model RA-25-0264, Attachment 4