05000483/LER-2025-002, Unit 1, Reactor Vessel Bottom Head (Rvbh) Bottom Mounted Instrument (Bmi) Nozzle Pressure Boundary Leakage
| ML25183A213 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/02/2025 |
| From: | Ameren Missouri, Union Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25183A211 | List: |
| References | |
| ULNRC-06953 LER 2025-002-00 | |
| Download: ML25183A213 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4832025002R00 - NRC Website | |
text
Abstract
on May 6, 2025, with the Callaway plant in Mode 3 (Hot Standby), while performing a visual inspection of the outer diameter ofthe bottom of the reactor vessel, a boric acid leak was identified at 19:23 in the annulus area of bottom-mounted instrumentation (BMI) nozzle #48. The leak was determined to be a boric acid leak that constituted pressure boundary leakage. Consequently, the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.4.13, RCS Operational LEAKAGE, was declared not met at 22:20 on May 6, 2025, due to failure to meet the LCO requirement of No pressure boundary LEAKAGE. The plant was taken to Mode 5 in accordance with TS 3.4.13 after discovery of the leak.
Extent-of-condition inspections were performed for all 58 BMI nozzles, and it was found that three additional BMI nozzles had no apparent boric acid leaks but required repair based on indications identified from non-destructive examination. Corrective action taken to prevent recurrence consisted of installing a weld pad (half nozzle repair) on the outside of the reactor pressure vessel for each of all four degraded BMI nozzle junctions, thereby moving the reactor coolant pressure boundary from the internal part of the reactor vessel to the external part of the reactor vessel in these nozzle locations.
The most probable cause ofthe BMI nozzle #48 leak was the result of a stress riser from a welding/fabrication defect combined with PWSCC being the only possible in-service failure mechanism to develop a leak.
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET (See NUREG-1 022, R.3 for instruction and guidance for completing this form htto://www.nrc.pov/readinci-rm/doc-collections/nureps/staff/sr1022/r3/)
- 1. FACILITY NAME Callaway Plant, Unit No. 1
NARRATIVE
EXPIRES: 04/3012027
L_ I
- 3. LER NUMBER A primary function of the reactor coolant system (RCS) (EIIS: AB) is to serve as one of the three principal barriers (fuel cladding, RCS, and containment) against fission product release to the environment. The RCS is designed, constructed and maintained to minimize the potential for reactor coolant leakage.
Callaway Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.13 states:
RCS operational LEAKAGE shall be limited to:
a. No pressure boundary LEAKAGE; b. I gpm unidentified LEAKAGE; c. 10 gpm identified LEAKAGE; and d. I 50 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
TS LCO 3.4.13 is applicable in Modes 1, 2, 3, and 4. PerCondition B ofTS LCO 3.4.13, ifpressure boundary leakage exists, then the Required Action is to be in in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, per Required Actions B.1 and B.2, respectively.
The Callaway reactor vessel contains penetrations that include 58 BMI nozzle assemblies. The BMI nozzle assemblies allow the in-core instrumentation to enter the bottom of the reactor vessel. The BMI nozzle assemblies are fabricated from Alloy 600 and are internally connected to the vessel lower head with J-groove welds that are fabricated from Alloy 182.
Alloy 600 materials and Alloy 82/182 weld metals are known to be susceptible to primary water stress corrosion cracking (PWSCC). The nozzles and welds were water jet peened in 2017 to prevent the progression of PWSCC.
The leak at BMI nozzle #48, was from/through the associated Alloy 182 J-groove weld.
2. INITIAL PLANT CONDITIONS
The event described in this LER was identified during the current refueling outage (RFO 27) at Callaway, which began on 3/29/2025. On 5/3/2025, at 12:46, the plant was taken to Mode 4 in preparation for plant restart (as intended at the time).
The plant was then taken to Mode 3 on 5/5/2025 at 03:14. An under-vessel visual inspection (in accordance with plant procedures) was performed on 5/6/2025, during which the RCS leak (described below) was identified. This was in addition to an under-vessel examination that was performed earlier in the outage (also in accordance with plant procedures) just after plant shutdown for the outage.
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3. EVENT DESCRIPTION
On May 6, 2025, at I 9:23, while performing a visual inspection of the outer diameter of the reactor vessel bottom head (RVBH), a boric acid leak was identified in the annulus area of bottom mounted instrumentation (BMI) nozzle #48. The leak was determined to be a boric acid leak such that it constituted pressure boundary leakage. Consequently, the Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.4.1 3, RCS Operational LEAKAGE, was declared not met at 22:20 on May 6, 2025, due to failure to meet the LCO requirement of No pressure boundary LEAKAGE.
The plant was taken to MODE 5 on 5/7/2025 at 23:06 in accordance with TS 3.4.13, RCS Operational Leakage, after discovery of the leak.
This was the first time a reactor vessel BMI nozzle leaked at Callaway. All the BMI nozzles and J-groove welds were water jet peened in 201 7 to reduce the probability of PWSCC in components containing Alloy 600 and Alloy I 82 which are known to be susceptible to PWSCC.
4. ASSESSMENT OF SAFETY CONSEQUENCES
In response to operating experience, the nuclear industry has addressed the potential for reactor vessel BMI nozzle assembly leakage with requirements to perform periodic visual examinations to detect leakage before safety significant failure can occur. At Callaway, visual examinations of BMI nozzle assemblies for evidence of leakage and corrosion on adjacent ferritic items have been scheduled every other refueling outage to meet the requirements of I 0 CFR 50.55a(g)(6)
(ii)(E), Augmented 151 requirements: Reactor coolant pressure boundary visual inspections, (based on Electric Power Research Institute (EPRI) MRP-206, Materials Reliability Program: Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants, and ASME Code Case N-722-I The probabilistic risk assessment performed for the EPRI MRP-206 study concluded that, for plants with 1 8-and 24-month fuel cycles (Callaway has an 18-month fuel cycle), a bare metal visual examination every other refueling outage is sufficient to ensure the calculated change in incremental core damage frequency (ICDF) resulting from degraded BMI nozzle assemblies remains at or below 1.OE-06. The risk evaluations summarized in EPRI MRP-206 show that periodic inspections provide reasonable assurance against nozzle ejection and significant head wastage. During the investigation of this event, only axial indications were found in the J-groove weld (at BMI nozzle #48) and no loss of reactor vessel lower head shell material was found. Therefore, the RCS leak is considered to be of low nuclear safety significance.
The BMI nozzle #48 leakage rate was very small with little boron residue accumulation on the lower reactor vessel head and no appreciable boron residue accumulation on the structures beneath the vessel. The event did not result in the release of radioactive materials to the environment, and the event did not adversely affect the safe operation of the plant or health and safety of the public. The event occurred and was found during shutdown conditions, and it was repaired during shutdown conditions.
The event did not result in a potential for a transient more severe than those analyzed in the Final Safety Analysis Report (FSAR) Chapters 6, Engineered Safety Features, and I 5, Accident Analysis. The condition would not have prevented the fulfillment of a safety function, and the condition did not result in a safety system functional failure as defined by 10 CFR 50.73 (a)(2)(v).
5. REPORTING REQUIREMENTS
This LER is submitted pursuant to 10 CFR 50.73(a)(2)(ii)(A) to report an event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded (in this case, degradation of the reactor coolant pressure boundary).
This LER is also submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) to report an operation or condition that was prohibited by the plants Technical Specifications (in this case, TS 3.4.13, RCS Operational LEAKAGE) due to non-zero reactor coolant system pressure boundary leakage. Specifically, the TS violation occurred due to the unintended delay in identifying the existence of pressure boundary leakage. As noted previously, no such leakage was identified during the under-vessel inspection performed early in the outage. However, after identification of the leakage on 5/6/2025, a past operability evaluation was performed. The evaluation concluded that the leakage began during Mode 4 (i.e., during heat-up and pressurization). As noted previously, the plant was taken to Mode 4 at 12:46 on 5/3/2025 and then to Mode 3 at 03:14 on 5/5/2025. The leak was identified at 19:23 on 5/6/2025.
Per TS 3.4.13, with the LCO not met due to pressure boundary leakage during Mode 4, the plant is required to be in Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> per Required Action B.2. Thus, even if it is assumed thatthe leak occurred at the end of Mode 4, at 03:14 on 5/5/2025, the plant was not taken to Mode 5 until 23:06 on 5/7/2025, when it should have been taken to Mode 5 at or by 15:14 on 5/6/2025. Thus, due to the unintended delay in identifying the pressure boundary leakage, a condition prohibited by the plants Technical Specifications occurred.
This condition was reported to the NRC pursuant to I 0 CFR 50.72(b)(3)(ii)(A) via Event Notification 57695 on 5/7/2025.
6. CAUSE OF THE EVENT
The most probable cause of the BMI nozzle #48 leak was the result of a stress riser from a welding/fabrication defect combined with PWSCC being the only possible in-service failure mechanism to develop a leak.
7. CORRECTIVE ACTIONS
The entire population of 58 nozzles was non-destructively inspected by ultrasonic testing (UT), and the results were independently reviewed by EPRI. The corrective actions were to repair the four BMI nozzles with indications with a half nozzle repair which moves the RCS pressure boundary from the J-groove weld inside the reactor vessel to a new J groove weld outside the reactor vessel, i.e., the repair method relocated the structural weld reactor coolant pressure boundary (RCPB) from the reactor vessel bottom head inside diameter to the reactor vessel bottom head outside diameter.
YEAR NUMBER 2025H1 002 HPage 5
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