05000346/LER-2010-002-01, Regarding Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
 
Line 1: Line 1:
{{Adams
#REDIRECT [[05000346/LER-2010-002, Regarding Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage]]
| number = ML102800416
| issue date = 09/30/2010
| title = Regarding Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage
| author name = Allen B
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000346
| license number = NPF-003
| contact person =
| case reference number = L-10-258
| document report number = LER 10-002-01
| document type = Letter, Licensee Event Report (LER)
| page count = 5
}}
{{LER
| Title = Regarding Control Rod Drive Nozzle Primary Water Stress Corrosion Cracking and Pressure Boundary Leakage
| Plant =
| Reporting criterion = 10 CFR 50.73(a)(2)(i)(B), 10 CFR 50.73(a)(2)(ii)(A), 10 CFR 50.73(a)(2)(i), 10 CFR 50.73(a)(2)(vii), 10 CFR 50.73(a)(2)(viii)(A), 10 CFR 50.73(a)(2)(ii)(B), 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), 10 CFR 50.73(a)(2)(x), 10 CFR 50.73(a)(2)(v)(A), 10 CFR 50.73(a)(2)(v)(B), 10 CFR 50.73(a)(2)(i)(A), 10 CFR 50.73(a)(2)(v)
| Power level =
| Mode =
| Docket = 05000346
| LER year = 2010
| LER number = 2
| LER revision = 1
| Event date =
| Report date =
| ENS =
| abstract =
}}
 
=text=
{{#Wiki_filter:FENOCF5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Barry S. Allen 7..t
(
419-321-7676 Vice President-Nucle*,r.,
Fax: 419-321-7582 September 30, 2010 L-1 0-258 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 Revision 01 Enclosed is Revision 01 to.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.
This LER is being revised to provide results of the completed Root Cause evaluation.
There are no new 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. Patrick J. McCloskey, Manager, Site Regulatory Compliance, at (419) 321-7274.
Sincerely, Barry S. Allen GMW Enclosure: LER 2010-002-01 cc:
NRC Reg*rwi III Administrator
.. NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board
 
;n#'
, )_
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 4
: 4. TITLE Control Rod Drive Nozzle PrimaryWater Stress Corrosion Cracking and Pressure Boundary Leakage
: 5. EVENT DATE
: 6. LER NUMBER
: 7. REPORT DATE
: 8. OTHER FACILITIES INVOLVED SUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER MONTH DAY YEAR 05000 03 12 2010 2010 002 01 09 30 2010 FACILITY NAME DOCKET NUMBER 05000
: 9. OPERATING MODE
: 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
E-20.2201(b)
[L 20.2203(a)(3)(i) ri 50.73(a)(2)(i)(C)
E]
50.73(a)(2)(vii) 6
[
20.2201(d)
E] 20.2203(a)(3)(ii)
[
50.73(a)(2)(ii)(A)
Ii 50.73(a)(2)(viii)(A)
E] 20.2203(a)(1)
[
20.2203(a)(4)
E]
50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[] 20.2203(a)(2)(i)
LI 50.36(c)(1)(i)(A)
.E]
50.73(a)(2)(iii)
EL 50.73(a)(2)(ix)(A)
: 10. POWER LEVEL 0
20.2203(a)(2)(ii)
E) 50.36(c)(1)(ii)(A)
[E 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
EI 20.2203(a)(2)(iii) 0 50.36(c)(2)
[] 50.73(a)(2)(v)(A)
[]
73.71 (a)(4).
000 E]
20.2203(a)(2)(iv)
[]
50.46(a)(3)(ii)
[] 50.73(a)(2)(v)(B)
LI 73.71 (a)(5)
LI 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
LI OTHER Specify in Abstract below
_0 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)
E]
50.73(a)(2)(v)(D) or in
 
==CAUSE OF EVENT==The DBNPS reactor vessel closure head was replaced due to the discovery of degradation as documented in DBNPS Licensee Event Report 2002-002. The current DBNPS reactor vessel closure head, which was purchased from the cancelled Midland Unit #2 and placed in service in,6eptember 2003, has been in service for three operating cycles. No. RCS pressure boundary leakage had been identified during previous bare metal visual examinations of this RVCH.
Operating temperatures of the DBNPS RVCH were previously considered to be equal to the highest RCS hot leg temperature at approximately 606.4°F. Analysis performed following discovery' of RCS pressure boundary leakage on March 13, 2010, determined the temperature of the water circulating
,through the upper region of the RVCH is approximately seven to eight degrees warmer than previously understood. This is due to channeling of warmer water leaving fuel assemblies located below the control rod guide tubes to the RVCH. The core fuel design determines the fuel assemblies' power and subsequent water exit temperature. The RVCH temperature is used to calculate Effective Degradation Years and Re-Inspection Years, which are used in the calculation to determine when the next inspection is required to be performed per regulations.
The Root Cause of the CRDM nozzle cracking and J-groove weld flaws with RCS pressure boundary leakage was a less than adequate perception of the risk of PWSCC susceptibility with the replacement RVCH resulting in inadequate identification, development, and implementation of interim actions to mitigate degradation prior to replacement with a PWSCC resistant Alloy 690 RVCH.
 
==ANALYSIS OF EVENT==
The bare metal visual inspection did not identify any discernable areas of wastage of the RVCH. The majority of the indications/flaws identified on the nozzles from UT examinations were primarily axial in direction, and the few identified with circumferential aspects were no more than 15 degrees in length.
Based on industry evaluation of CRDM nozzle indications, the critical size of a circumferential flaw that could result in a CRDM nozzle ejection is on the order of 330 degrees in length, and axial nozzle cracking is not a credible mechanism for CRDM nozzle ejection. Since the CRDM nozzle flaws observed at the DBNPS were identified by a planned inspection designed to detect those types of flaws, and since the flaws that were detected were well below flaw sizes required for nozzle ejection.
and there was no discernable head wastage, this issue was of very low safety significance.
Reportability Discussion:
Most of 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 XI 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 hours 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 hours (also reference Event #45764).
 
==ANALYSIS OF EVENT==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 were modified utilizing the inside diameter temper bead (IDTB) welding method to restore the pressure boundary of the degraded nozzles. These activities were 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, and authorized by the NRC verbally on June 4, 2010 and via letter dated September 20, 2010.
In accordance with the NRC Confirmatory Action Letter issued to the DBNPS on June 23, 2010, the DBNPS will be shut down no later than October 1, 2011, to replace the RVCH. Until replacement of the RVCH, upon reaching Action Level 3 of procedure EN-DP-01171, "Engineering Implementation of the RCS Integrated Leakage Program," the DBNPS shall be shutdown in 30 days if RVCH leakage cannot be ruled out. During subsequent shutdown as part of the containment inspection for RCS leakage, if RVCH leakage cannot be ruled out a bare metal visual examination of the RVCH will be performed per the applicable ASME Code Case and 10 CFR 50.55a(g)(6)(ii)(D).
Lessons learned training on the cause of this event including the less than adequate perception of risk related to the event will be provided to Engineering Support Personnel and Site Supervisors.
 
==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.
.NRC FORM 366A (9-2007)
}}
 
{{LER-Nav}}

Latest revision as of 04:57, 14 January 2025

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=05000346/LER-2010-002-01,_Regarding_Control_Rod_Drive_Nozzle_Primary_Water_Stress_Corrosion_Cracking_and_Pressure_Boundary_Leakage&oldid=5299121"