05000265/LER-2012-002, For Quad Cities, Unit 2 Regarding Reactor Vessel Instrument Nozzle Leakage Due to Intergranular Stress Corrosion Cracking
| ML12165A245 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/04/2012 |
| From: | Hanley T Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVP-12-055 LER 12-002-00 | |
| Download: ML12165A245 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2652012002R00 - NRC Website | |
text
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Exelon Generation Company, LLC www.exeloncoTp.com Nuclear Quad Cities Nuclear Power Station 22710 2o6th Avenue North Cordova, I L 61242-9740 June 4, 2012 10 CFR 50.73 SVP-12-055 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-30 NRC Docket No. 50-265
Subject:
Licensee Event Report 265/2012-002-00, "Reactor Vessel Instrument Nozzle Leakage Due to Intergranular Stress Corrosion Cracking" Enclosed is Licensee Event Report (LER) 265/2012-002-00, "Reactor Vessel Instrument Nozzle Leakage Due to Intergranular Stress Corrosion Cracking," for Quad Cities Nuclear Power Station, Unit 2.
This report is submitted in accordance with 10 CFR 50.73 (a)(2)(ii)(A) which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this report, please contact Mr. W. J. Beck at (309) 227-2800.
Respectfully, Tim HanleyJ Site Vice President Quad Cities Nuclear Power Station cc:
Regional Administrator-NRC Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013
- 10-2010)
, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.
- 3. PAGE Quad Cities Nuclear Power Station Unit 2 05000265 1 OF 6
- 4. TITLE Reactor Vessel Instrument Nozzle Leakage Due to Intergranular Stress Corrosion Cracking
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A FACIL**Y NAME DOCKET NUMBER 04 04 2012 2012 002 -
00 06 04 2012 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
C3 20.2201(b)
El 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 4 El 20.2201(d)
El 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4) 0l 50.73(a)(2)(ii)(B)
[I 50.73(a)(2)(viii)(B)
_]
20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
[I 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii) 0l 50.36(c)(1)(ii)(A)
[I 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71 (a)(4) 000%
El 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii)
[I 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
[I 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[I 50.73(a)(2)(i)(B)
[I 50.73(a)(2)(v)(D)
Specify in Abstract below or in On April 4, 2012, ENS Notification 47806 was made in accordance with 10 CFR 50.72 (b)(3)(ii)(A). This event has been classified as a Maintenance Rule Functional Failure for the RPV.
Given the impact on the reactor vessel pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73 (a)(2)(ii)(A), which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
C.
CAUSE OF EVENT
The most probable root cause of the N-i 1 B instrument nozzle weld crack and leak was determined to be Intergranular Stress Corrosion Cracking (IGSCC). Contributing to crack development were higher residual stresses that remained in the nozzle assembly following nozzle replacement in 1970, prior to the initial start-up of Unit 2. A most probable cause was determined because the repair resulted in requiring the degraded nozzle remnant to remain installed in the RPV penetration, hence, the specific characterization of the flaw causing the leak could not be determined.
The root cause investigation identified that the N-11B was originally replaced during initial RPV construction in 1970.
Since the 1970 replacement nozzle was installed after the original RPV heat treatment, no additional post weld heat treatment was pursued at that time, and as such, both the N-i 1 B nozzle body and the attachment weld likely had higher residual stresses than other similar nozzles installed during initial RPV construction which were heat treated.
Based on the higher residual stresses that likely existed, and since the nozzle and nozzle attachment weld were both fabricated from susceptible materials (Inconel Alloy 600 for the nozzle, and Inconel Alloy 182 for the weld metal), it was concluded that the observed leak was most likely caused by IGSCC. This cause is supported by previous industry events (primarily Pressurized Water Reactors (PWRs)) where similar cracking has occurred in the same Inconel alloy materials.
The extent of condition of this leakage event is limited to the N-11B nozzle. Attached to each Unit RPV are 36 nozzles, 177 Control Rod Drive (CRD) [AA] penetrations, and 53 flux monitor penetrations. Each of the RPV nozzles and penetrations are visually inspected for leakage during the RPV Class 1 system pressure test that occurs each refuel outage. Based on the most recent system pressure test on both units (Q1 R21 and Q2R21) there were no other RPV penetrations identified exhibiting evidence of similar through-wall failures. Additionally, following the N-11 B leak during Q2R21, the insulation was removed from similar N-11 and N-12 nozzles, and no leakage was observed.
The extent of cause is IGSCC, and is likely limited to the N-11B nozzle, since the N-11B nozzle is unique in that following the replacement of the nozzle in 1970 no post weld heat treatment was performed.
All other nozzles installed in the Unit 1 and Unit 2 RPVs were heat treated along with the vessel, and as such would likely have lower residual stresses. This assumption is supported by industry experience given there have been no similar RPV nozzle failures at other Boiling Water Reactors (BWRs).
Previous industry events are either specific to PWRs, are associated with external weld failures, or are non-IGSCC related. In addition, it is possible that as a result of the N-11 B nozzle replacement in 1970, IGSCC was able to initiate earlier, or propagate more quickly, due to the shallower than normal J-groove weld depth along the nozzle to weld interface. Based on the possibility of IGSCC as a potential cause, the extent of cause will be expanded to include inspections of other nozzles of similar design/material on Units 1 and 2.
D.
SAFETY ANALYSIS
System Design
The purpose of the RPV and its appurtenances, such as the N-11B nozzle, is to retain the reactor core coolant-moderator within the RPV and to serve as a high integrity barrier against leakage steam used for power production and leakage of radioactive materials to the drywell [NH] during all modes of plant operation. The N-i 1 B nozzle serves as the connection point for the 2-inch diameter "reference" leg associated with the "B" train of RVLIS. The purpose of RVLIS is to provide initiation signals to the Emergency Core Cooling System (ECCS) [JE] based on reactor water level, and also provide initiation signals to trip functions in the Anticipated Transient Without Scram (ATWS) system.
The N-i 1 B nozzle is located on the RPV, 35 inches above normal reactor water level (+30 inches), and during plant operation is exposed to the steam section of the RPV. The N-12B nozzle, which is of similar construction as the original N-11 B nozzle, is located on the RPV below the normal reactor water level, which serves as the connection point for the 2-inch "variable" leg for the B-train of RVLIS, and during plant operation is exposed to the water section of the RPV.
Safety Impact The safety significance of this event was minimal given the leakage was very small, was found while the reactor was shutdown for refueling, and if leaked during plant operation, did not exceed Technical Specification (TS) leakage limits for unidentified drywell leakage.
During normal power operation, this nozzle is exposed to a reactor steam environment, and not the water solid conditions of the Class 1 pressure boundary system leakage test that is performed each refueling outage in accordance with ASME Section XI, IWB-2500.
Had a worst case scenario occurred, in which the 2-inch diameter N-i 1 B connection line to RVLIS failed completely such that there was a 2-inch diameter opening in the RPV, the consequences would have been minimal. The 2-inch diameter line break (0.02 sq ft in area) would be bounded by the line breaks of up to 0.12 sq ft in area, as discussed in Chapter 15 of the Updated Failure Safety Analysis Report (UFSAR) which provides that the High Pressure Coolant Injection (HPCI) [BJ] system could supply sufficient coolant to depressurize the vessel and cool the core for line breaks of up to 0.12 sq ft in area.
Additionally, the leakage through the N-11 B nozzle was small and had no impact on the ability of the RVLIS system to monitor RPV level. Since the leak was located on the reference leg of the RVLIS system, where the reference leg is maintained full by backfilling from the CRD system, the leak at the nozzle would not impact the RVLIS reference leg unless a complete failure of the 2-inch line had occurred. Furthermore, even if a complete line failure had occurred, the redundant "A" train of RVLIS would have provided the necessary level indication for safety system actuations.
Risk Insights Prior to and during this event, the capability of RVLIS was not lost since there were no identified RVLIS leaks during the prior operating cycle. This condition did not create any actual plant or safety consequences since the Unit was not in an accident or transient condition requiring use of RVLIS initiation signals to ECCS or ATWS during this period of time or prior operating cycle. The nozzle leak was discovered during system pressure testing which is designed to identify any RPV leakage before the unit is placed into power operation.
There is no makeup rate required to mitigate an RPV leakage of 60 dpm. N-11B is a 2 inch penetration. Even a complete failure of the penetration would only result in Small LOCA (SLOCA). The break diameter for a SLOCA is between 0.5 and 2 inches and would be well within the capability of HPCI or one RHR pump. 60 dpm is negligible compared to any SLOCA.
The N-i 1 B instrument nozzle, which is attached to the wall of the pressure vessel, provides the connection point for the reference leg of the "B" train of the RVLIS. Steam from the RPV travels through piping to the condensing pot and maintains a constant level in the reference leg. The 60 dpm when converted to steam would become a very small flow of leakage steam, and would not result in diverting any measurable steam flow needed in the piping for the reference leg condensing pot.
Based on the above, the impact on the RPV and RVLIS of a 60 dpm leak is negligible, therefore, considering the impact of this condition on the Plant Probabilistic Risk Assessment (PRA), the change in Core Damage Frequency (CDF) due to the observed leakage will be less than 1.OE-06/yr. In conclusion, the overall safety significance and impact on risk of this event was minimal.
E.
CORRECTIVE ACTIONS
Immediate:
- 1. A Relief Request was submitted to the NRC since the defect in N-i 1 B nozzle would not be removed, and since a qualified technique to perform volumetric non-destructive examination (NDE) of the partial penetration weld for characterizing the flaw and determining flaw growth in the specific configuration does not exist. The NRC subsequently approved the Relief Request prior to Unit startup.
- 2. The degraded N-i 1 B nozzle was partially removed to allow a half-nozzle repair to be installed. The half-nozzle design was welded to the outer vessel wall, effectively moving the pressure boundary from the degraded nozzle to the new externally attached nozzle. The new half-nozzle assembly was fabricated from Inconel Alloy 690 and attached to the RPV with Inconel Alloy 52M weld metal. Both materials are considered IGSCC resistant.
- 3. A failure assessment and flaw evaluation were completed prior to startup to demonstrate the acceptability of leaving the original partial penetration attachment weld, with a maximum postulated flaw, in place for one operating cycle.
Follow-up:
- 1. N-11A/B and N-12A/B nozzles will be inspected from inside the RPV using enhanced VT-1 methods during the next Ul and U2 refuel outage. Additional corrective actions will be identified as necessary as a result of these inspections.
- 2. Perform an Elastic Plastic Fracture Mechanics (EPFM) flaw evaluation for the N-11 B remnant J-groove weld to support continued operation beyond the next refuel outage for Unit 2.
F.
PREVIOUS OCCURRENCES
The station events database, LERs, EPIX, and NPRDS were reviewed for similar events at Quad Cities Nuclear Power Station. This event was a RPV nozzle attachment weld leak associated with IGSCC. There were no previous similar occurrences identified at Quad Cities Nuclear Power Station that involved an event of this type.
I
G.
COMPONENT FAILURE DATA
The failed component was the N-i 1 B instrument nozzle which was fabricated by Chicago Bridge and Iron in 1970 after the original nozzle was replaced due to damage that occurred after initial construction of the RPV. The original construction nozzle, along with the other nozzles installed in Unit 2, were fabricated by Babcock and Wilcox.
The N-i 1 B nozzle is a 2 inch diameter RPV nozzle located in the steam section of the vessel approximately 35 inches above normal reactor water level. Materials of construction are, Inconel Alloy 600 for the nozzle, and Inconel Alloy 182 for the weld metal.
This event has been reported to EPIX as Failure Report No. 1152.