05000325/LER-2006-002, Regarding Cracking Found in B Loop Spray Header Piping
| ML061530280 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/18/2006 |
| From: | Waldrep B Progress Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 06-0055 LER 06-002-00 | |
| Download: ML061530280 (6) | |
| 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) 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 |
| 3252006002R00 - NRC Website | |
text
Progress Energy May 18, 2006 SERIAL: BSEP 06-0055 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Brunswick Steam Electric Plant, Unit No. 1 Docket No. 50-325/License No. DPR-71 Licensee Event Report 1-2006-002 Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power
& Light Company, now doing business as Progress Energy Carolinas, Inc., submits the enclosed Licensee Event Report.
Please refer any questions regarding this submittal to Mr. Randy C. Ivey, Manager - Support Services, at (910) 457-2447.
Sincerely, B. C.*
dr Plant General Manager Brunswick Steam Electric Plant MAT/mat
Enclosure:
Licensee Event Report Progress Energy Carolinas, Inc.
Brunswick Nuclear Plant PO Box 10429 Southport, NC 28461
Document Control Desk BSEP 06-0055 / Page 2 cc (with enclosure):
U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-051
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2007 (6-2004)
, the NRC may not conduct or sponsor, and a person is not requred to respond to the information collection.
- 3. PAGE Brunswick Steam Electric Plant (BSEP), Unit 1 05000325 1 OF 4
- 4. TITLE Cracking Found In B Loop Core Spray Header Piping
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILmES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL N
MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBI NO__________
05000 03 21 2006 2006 002 00 05 18 2006FACILITY NAME DOCKET NUMBER 1
_________________05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) 5 0l 20.2201(b)
[I 20.2203(a)(3)(i) 0l 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) ol 20.2201(d)
[E 20.2203(a)(3)(ii)
El 50.73(a)(2)(i4)(A) 0l 50.73(a)(2)(viii)(A) 0l 20.2203(a)(1)
[I 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[I 20.2203(a)(2)(i)
[I 50.36(c)(1)(i)(A) 0l 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
Dl 50.36(c)(1)(ii)(A) 0-50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 000 El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71 (a)(4)
El 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER Specify n Abstract below El 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D) or (If more space is required, use additional copies of NRC Form 3W6A) (17)being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as operation prohibited by the plant's Technical Specifications.
EVENT CAUSE
The root cause of this event is intergranular stress corrosion cracking (IGSCC) in the heat affected zone (HAZ) of weld P3c-270.
During B 116R1, invessel inspections were performed using a WesdynelWestinghouse automated Double-Up Inspection Tool with an ROS color camera calibrated to 0.0005-inch resolution. The use of the improved inspection equipment permitted increased coverage; thus locating the flaw.
The P3c-270 weld is unique to Unit 1, B loop CS, and was likely made to assist in fit-up of the piping during installation. Neither Unit 1 CS loop A or Unit 2 have a corresponding weld location.
SAFETY ASSESSMENT
The safety significance of this condition is considered minimal and was characterized as having very low safety significance in Brunswick Steam Electric Plant - NRC Integrated Inspection Report Nos.
05000325/2006002 and 05000324/2006002, dated April 30, 2006.
The as-found condition of the Unit 1 B loop CS pipe weld P3c-270 resulted in the Unit 1 B loop CS subsystem being inoperable for an indeterminate amount of time. Although the CS system mitigates several core damage sequences, B loop CS was still capable of mitigating all of the sequences with the exception of large break loss of coolant accident (LOCA). General Electric NEDO-20566A, "GE Analytical Model for Evaluation of LOCA Analysis in Accordance with 10 CFR 50 Appendix K," provides the basis for this conclusion. NEDO-20566A describes the design features required in various accident conditions, and concludes that for all pipe breaks, other than breaks in the recirculation system, the core would remain flooded over the entire fuel length, and injection from any one Emergency Core Cooling System pump is sufficient to maintain adequate core cooling. Therefore, for pipe breaks other than on the recirculation lines, it is not necessary to maintain a core spray pattern over the top of the fuel because it would not be uncovered.
Making the conservative assumption that the flawed weld would not remain intact, B loop CS was capable of delivering flow to the reactor vessel outside the shroud. Therefore, the vulnerable accidents can be limited to the design-basis double-ended break of a recirculation line, which is quantified as a Large LOCA accident sequence in the BSEP Probabilistic Safety Analysis (PSA). If the PSA model is quantified with (If more space Is required use additional copies of NRC Form 366A) (17)B loop CS unavailable for Large LOCA initiating events, the change in core damage frequency (CDF) would be sufficiently small to conclude that the safety significance of the CS piping crack is minimal. The estimated frequency of a severe seismic event is sufficiently small to conclude that consideration of a seismic event would have a negligible effect on CDF.
CORRECTIVE ACTIONS
On April 1, 2006, weld P3c-270 was repaired by installation of a clamp designed to structurally replace the weld. The clamp was designed in accordance with BWRVIP Report BWRVIP-19-A, "BWR Vessel and Internals Project, Internal Core Spray Piping and Sparger Repair Design Criteria," which includes BWRVIP-84, "BWR Vessel and Internals Project, Guidelines for Selection and Use of Materials for Repairs to BWR Internal Components."
PREVIOUS SIMILAR EVENTS
A review of events which have occurred within the past three years has not identified any previous similar occurrences.
COMMITMENTS
No regulatory commitments are contained in this report.