05000456/LER-2011-001, Regarding Through Wall Crack on 1A Safety Injection Pump Discharge Line Due to Outside Diameter (Transgranular) Stress Corrosion Cracking Initiated at External Diameter of Pipe
| ML111160404 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/26/2011 |
| From: | Enright D Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW110041 LER 11-001-00 | |
| Download: ML111160404 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 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 |
| 4562011001R00 - NRC Website | |
text
10 CFR 50.73 April 26, 2011 BW110041 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456 Subject: Licensee Event Report 2011-001 Through Wall Crack on 1A Safety Injection Pump Discharge line Due to Outside Diameter (Transgranular) Stress Corrosion Cracking Initiated at External Diameter of Pipe The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "licensee event report system," paragraph (a)(2)(i)(B), any operation or condition which is prohibited by the plant's Technical Specifications. On February 25,2011, a through wall crack on the 1A safety injection pump discharge line was identified, and may have existed for a period longer than allowed by the Technical Specifications. 10 CFR 50.73(a) requires an LER to be submitted within 60 days following discovery of the event. Therefore, this report is being submitted by April 26, 2011.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Chris VanDenburgh, Regulatory Assurance Manager, at (815) 417-2800.
Respectfully,
~~~""'~
Daniel J. Enright Site Vice President Braidwood Station
Enclosure:
LER 2011-001-00 cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)
Illinois Emergency Management Agency - Braidwood Rep
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/3112013 (10-2010)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 13. PAGE Braidwood Station, Unit 1 05000456 1 of 4
- 4. TITLE Through Wall Crack on 1A Safety Injection Pump Discharge Line Due to Outside Diameter (Transgranular) Stress Corrosion Cracking Initiated at External Diameter of Pipe
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR ISEQUENTIALIREV MONTH DAY YEAR N/A N/A NUMBER NO.
FACILITY NAME DOCKET NUMBER 02 25 2011 2011. 001.
00 04 26 2011 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) o 20.2201(b) o 20.2203(a)(3)(i) o 50.73(a)(2)(i)(C) o 50.73(a)(2)(vii) 1 o 20.2201(d) o 20.2203(a)(3)(ii) o 50.73(a)(2)(ii)(A) o 50.73(a)(2)(viii)(A) o 20.2203(a)(1) o 20.2203(a)(4) o 50.73(a)(2)(ii)(B) o 50.73(a)(2)(viii)(B) o 20.2203(a)(2)(i) o 50.36(c)(1 )(i)(A) o 50.73(a)(2)(iii) o 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL o 20.2203(a)(2)(ii) o 50.36(c)(1 )(ii)(A) o 50.73(a)(2)(iv)(A) o 50.73(a)(2)(x) o 20.2203(a)(2)(iii) o 50.36(c)(2) o 50.73(a)(2)(v)(A) o 73.71(a)(4) 100 o 20.2203(a)(2)(iv) o 50.46(a)(3)(ii) o 50.73(a)(2)(v)(B) o 73.71 (a)(5) o 20.2203(a)(2)(v) o 50.73(a)(2)(i)(A) o 50.73(a)(2)(v)(C) o OTHER o 20.2203(a)(2)(vi)
[gI 50.73(a)(2)(i)(B) o 50.73(a)(2)(v)(D)
Specify in Abstract below or in 2011 001 00 In addition to the dried boric acid identified, evidence of a previous installed pipe clamp was noticed in the area where the dried boric acid was identified and it was later confirmed that a snubber had previously existed at this location. The snubber and pipe clamp were removed as part of the snubber reduction program in the early 1990s.
A failure analysis was performed on the section of pipe that was cut out of the system. An internal dye penetrant examination was performed and only one 0.010 inch long indication was detected on the pipe internal diameter.
Further testing confirmed the suspected through wall crack. Based on the conclusions provided under the test report, the through wall crack was caused by ODSCC that initiated from the external surface of the pipe caused by chloride exposure. Qualitative analysis of the exposed crack surface indicated the cracking was caused by an aqueous chloride exposure. Additional qualitative analysis was performed on the tape residue removed from the flawed pipe revealing a high chloride peak, which confirmed the probable source of chloride. The tape residue was located in the area where the pipe clamp was installed. Additionally, the test report indicated that industry literature reports show pipe clamps as preferential locations for externally initiated stress corrosion cracking. As a result, the previously removed snubber clamp contributed to the leak for the following reasons: 1) The clamp generated crevice conditions that could locally retain moisture for long periods of time, and 2) The clamp to pipe crevice would concentrate detrimental chlorides by a crevice corrosion mechanism. The inadequate removal of tape residue occurred during initial plant construction.
p.
Safety Consequences
There were no actual safety consequences impacting plant or public safety as a result of this event. Following identification of the dried boric acid, the 1A SI train was declared inoperable and the appropriate TS required actions were taken.
In a design basis accident, there is reasonable assurance the SI system would have performed its design function, based on:
The size of the defect, 0.010 inches, is very small as indicated in the PowerLabs failure analysis report (report number BRW-36383). The report also stated the piping wall thickness ranged from 0.325" to 0.343" near the leak area, which was within the 12.5% tolerance for 4" NPS, Schedule 80S piping that was fabricated to ASTM/ASME A312-Type 304. Based on the relatively small indication size and the 45 degree angled orientation for the cracking, there was little concern for a catastrophic pipe failure.
Austenitic stainless steel piping is not prone to brittle fracture due to the inherent ductility at all relevant temperatures of the material. The annealed austenitic stainless steel used for piping applications has a high resistance to brittle fracture and this resistance does not diminish over time, because the material is not expected to be embrittled as a result of neutron radiation nor are the wrought forms of austenitic stainless steels susceptible to thermal embrittlement since they have low <<1 percent) delta ferrite. For these reasons, catastrophic failure of austenitic stainless steel pipe due to sec is rare.
The failure mechanism involved with this failure, ODSCC, is a slow progressing mechanism. Braidwood's probabilistic risk assessment (PRA) model credits the SI pump operation for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. Therefore, there would be a very limited time frame for flaw growth in an accident scenario.
No failure of the discharge piping occurred during the most recent SI pump runs, with flow in recirculation; Prior to the discovery of the through wall crack on February 25,2011, on December 3,2010 an ASME Section XI Inspection (VT-2) was completed on this pipe. The system was pressurized to approximately 1575 psig for approximately two hours. No indication of leakage was identified.
Additionally, between this time period (December 3, 2010 and February 25, 2011), the 1A and 1B SI pumps were operated in support of testing on February 1, 2011 and February 23, 2011, respectively. During pump operation, the discharge piping of both trains of SI was pressurized to approximately 1565 psig and 1500 psig, respectively. The system was pressurized for approximately one half hour in both instances. No adverse effect on the discharge piping was evident.
The discharge pressure of the Sl pumps operating with flow in recirculation to the RWST would bound the pressure expected in an accident condition.
Based on the information provided above, this event did not result in a safety system functional failure.
E.
Corrective Actions
Completed corrective actions include:
Replaced the portion of the SI line containing the flaw with new stainless steel pipe.
Performed extent of condition walkdowns of portions of the SI discharge piping. No other anomalies were identified.
Additional corrective actions include:
Perform a flaw evaluation to determine the allowable flaw size to demonstrate that the structural integrity of the Sl discharge line was maintained.
Review the sticker labels used by site or by suppliers for chloride content for conformance to concentrations listed in the Chemical Control Program procedure.
Review tape adhesive currently available at the station for chloride content for conformance to concentrations listed in the Chemical Control Program procedure.
Previous Occurrences
There has been one previous, similar Licensee Event Report identified at the Braidwood Station in the past three years:
Licensee Event Report 2010-003 Unit 1 Through-weld Leak of the Line from the 1B Seal Injection Filter to the Vent Valve G.
Component Failure Data
This event has been reported to EPIX as Failure Report NO.1 013.