05000346/LER-2010-002, Regarding Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage
| ML101340128 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/11/2010 |
| From: | Allen B FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-10-119 LER 10-002-00 | |
| Download: ML101340128 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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 |
| 3462010002R00 - NRC Website | |
text
FENOC 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Barry S. Allen 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 May 11,2010 L-10-119 10 CFR 50.73 ATTN: Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station Docket Number 50-346, License Number NPF-3 Licensee Event Report 2010-002 Enclosed is Licensee Event Report (LER) 2010-002, "Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage." This LER is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), operation in a condition prohibited by the Technical Specifications, and 10 CFR 50.73(a)(2)(ii)(A), condition of the plant, including its principal safety barriers, being seriously degraded.
There are no regulatory commitments contained in this letter or its enclosure. The actions described represent intended or planned actions and are described for information only. If there are any questions or if additional information is required, please contact Mr. Brian T. Hennessy, Acting Manager, Site Regulatory Compliance, at (419) 321-8592.
Sincerely, BarrS. A Zen GMW Enclosure: LER 2010-002-00 cc:
NRC Region III Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 8/31/2010 (6-2004)
, the NRC may (See reverse for required number of not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
- 3. PAGE Davis-Besse Nuclear Power Station 05000346 1 OF 3
- 4. TITLE Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED ISEQUENTIAL I
REV MFACILITY NAME DOCKET NUMBER MONTH DAY YEAR EAR NUMBER NO.
ONTH DAY YEAR 05000 03 FACILITY NAME DOCKET NUMBER 03 12 2010 2010 002 00 05 11 2010 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
E] 20.2203(a)(3)(i)
[5 50.73(a)(2)(i)(C)
E]
50.73(a)(2)(vii) 6 E5 20.2201 (d)
E]
20.2203(a)(3)(ii)
Z 50.73(a)(2)(ii)(A) 5] 50.73(a)(2)(viii)(A) 5 20.2203(a)(1)
E] 20.2203(a)(4) 5j 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[] 20.2203(a)(2)(i)
[] 50.36(c)(1)(i)(A)
E]
50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii) 5j 50.36(c)(1)(ii)(A)
E]
50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
E]
20.2203(a)(2)(iii) 5 50.36(c)(2)
[] 50.73(a)(2)(v)(A)
[1 73.71(a)(4) 000 5] 20.2203(a)(2)(iv) 5 50.46(a)(3)(ii)
E5 50.73(a)(2)(v)(B)
El 73.71(a)(5)
[] 20.2203(a)(2)(v)
[] 50.73(a)(2)(i)(A)
[] 50.73(a)(2)(v)(C)
S OTHER Specify in Abstract below
_20.2203(a)(2)(vi)_
50.73(a)(2)(i)(B) 5 50.73(a)(2)(v)(D) or in
==ANALYSIS OF EVENT==Reportability Discussion:
The indications found in the CRDM nozzles were determined to be unacceptable and require repair.
.. Section 3.2.4 of NUREG-1022, Event Reporting Guidelines, identifies that defects in the RCS pressure boundary that cannot be dispositioned as acceptable per ASME Section Xl represent a condition that results in the nuclear power plant, including its principal safety barriers, being seriously degraded.
These conditions were initially reported to the Nuclear Regulatory Commission per 10 CFR 50.72(b)(3)(ii)(A) on March 13, 2010 at 0445 hours0.00515 days <br />0.124 hours <br />7.357804e-4 weeks <br />1.693225e-4 months <br /> of this condition (reference Event #45764).
As a result of the bare metal visual inspection, evidence of leakage from a CRDM nozzle penetration was observed, which also indicates serious degradation of a principle safety barrier. These conditions were also reported to the Nuclear Regulatory Commission per 10 CFR 50.72(b)(3)(ii)(A) on March 13, 2010 at 1903 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.240915e-4 months <br /> (also reference Event #45764).
Technical Specification Limiting Condition for Operation 3.4.13 states that RCS operational Leakage shall be limited to no pressure boundary leakage. The evidence of leakage from the CRDM Nozzles indicates the plant operated in a condition prohibited by the Technical Specifications. Therefore, this issue is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B), operation in a condition prohibited by the Technical Specifications, and 10 CFR 50.73(a)(2)(ii)(A), serious degradation of a principle safety barrier. This event does not meet the definition of a Safety System Functional Failure.
CORRECTIVE ACTIONS
The CRDM nozzles containing indications or evidence of leakage are being modified utilizing the inside diameter temper bead (IDTB) welding method to restore the pressure boundary of the degraded nozzles. These activities are being conducted in accordance with the 1995 Edition through the 1996 Addenda of ASME Code Section XI, Code Case N-638-1, Code Case N-729-1, and alternative requirements as requested via separate correspondence (letter L-1 0-099 dated April 1, 2010, and others) from the FirstEnergy Nuclear Operating Company (FENOC) to the NRC.
Any additional Corrective Actions identified following completion of the Root Cause Evaluation for this event will be provided in a revision to this Licensee Event Report.
PREVIOUS SIMILAR EVENTS
DBNPS LER 2002-002 documented a previous event where RCS pressure boundary leakage occurred due to primary water stress corrosion cracking of CRDM nozzles, which resulted in wastage and degradation of the RVCH. The root cause of the 2002 RVCH degradation was boric acid corrosion due to an inadequate Boric Acid Corrosion Control Program. No discernable wastage of the RVCH was identified for the current event. The RVCH was replaced as a result of the 2002 event with an unused RVCH from the cancelled Midland Unit #2.