2CAN079604, Discusses Evaluation of Steam Generator Integrity for Cycle 12 Operation & Forwards Steam Generator Operational Assessment
| ML20116F282 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/30/1996 |
| From: | Mims D ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20116F285 | List: |
| References | |
| 2CAN079604, 2CAN79604, NUDOCS 9608060303 | |
| Download: ML20116F282 (2) | |
Text
q ceim o-.ue.. iec.
-:-:- ENTERGY
-m RuzcMe. AR 72301 Te 501858 fMO July 30,1996 2CAN079604 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station Pl-137 Washington, DC 20555
Subject:
Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No. NPF-6 Evaluation of Steam Generator Integrity for Cycle 12 Operation Gentlemen:
1 l
l The ANO-2 steam generators (SGs) were last inspected during the eleventh refueling outage l
(2R11) which ended November 20,1995. A report detailing the results of that inspection was l
provided on February 29,1996 (2CAN029608). The 2R11 inspection was the fifth extensive l
(100% of the tubes) examination of the susceptible hot leg expansion transition region in both SGs since circumferential cracking was discovered in the spring of 1992. During 2Rll an
)
initial 20% sample of the cold leg expansion transitions, concentrated in the sludge pile region, was also inspected. Additionally, a 100% bobbin coil examination was performed similar to that completed in the previous two refueling outages.
As was the case in previous outages, circumferential cracking was observed in both the A and B SG hot legs at the expansion transition where the tube exits the tube sheet. Additionally, during 2R11 circumferential cracks were detected in the A SG cold leg at the expansion transition. The cold leg examination was expanded to 100% in the A SG which resulted in the identification of 36 circumferential flaws. While there were more circumferential cracks detected during 2R11 (487 in A and 215 in B) than in previous outages, the average size of the flaws and the largest flaws were substantially smaller than those detected in previous outages. The increased number of flaws can partially be attributed to the enhanced capability of the Plus Point eddy current probe and other advanced instrumentation in detecting sniall circumferential cracks in the expansion transition region of the ANO-2 SG tubing.
Axial cracking at the eggerate supports was observed during 2R11 in both the A and B SGs, as has been the case in previous outages. The bobbin coil signal for the axial cracks at the 1
eggerates continues to be small in voltage. Comparisons to previous eddy current indications, n e n n'7c;
,f'/pl/
- l 9608060303 960730 PDR ADOCK 0500036.8 4
O PDR
l U. S. NRC July 30,1996 l
2CAN079604 Page 2 on average, show little to no growth. During the 2R11 bobbin inspection, several freespan indications were detected which were further characterized by Plus Point to be axial cracks.
Additional bounding Plus Point inspections were performed. A total of 18 freespan axial crack indications were detected in 16 tubes. The largest flaw measured 1.9 volts with a pancake coil and was approximately 1.85 inches in length. All 16 tubes were removed from service.
To funher assess the structural integrity of the steam generators, in-situ pressure tests were performed on four of the largest circumferential cracks, the largest freespan axial crack and the largest eggerate crack. The largest flaws were easily determined based on eddy current testing parameters (depth, length, and amplitude). All of these tests showed the tubes capable of withstanding the limiting Regulatory Guide 1.121 requirement of 3AP (4050 psig). No leakage was detected during the in-situ pressure tests of the circumferentially cracked tubes when pressurized up to 6800 psig.
The freespan axial crack also did not leak when pressurized to 4750 psig. The largest eggerate axial crack leaked at approximately 0.4 gpm at 2950 psig.
Since the 2Rll outage, extensive evaluations have been conducted to assess the integrity of the steam generators during the current cycle of operation. Included in these evaluations was a detailed review of the eddy current results from the 2RI1 outage, as well as previous outages, and an extensive review ofindustry data from other licensees with similar damage mechanisms.
Deterministic as well as probabilistic safety assessments were performed utilizing acceptance criterion found in Generic Letter 95-05 " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," Regulatory Guide 1.121, and the ANO-2 licensing basis. These analyses included conservative assumptions with respect to the flaw morphology (i.e., assumed the flaws to be planner without crediting ligament strengthening), material propenies, non-destructive examination uncertainty, tube structural limits, and flaw growth rates. The results of these evaluations demonstrate ANO-2's operation through the current cycle does not present any undue risk to public health and safety and that compliance with the ANO-2 licensing basis is assured.
Deterministic evaluations show that all ANO-2 SG tubing will maintain compliance with the structural requirements of Regulatory Guide 1.121 through the end of cycle 12. Probabilistic analysis assessed the probability of a tube burst following a main steam line break during cycle 4
12, considering all known damage mechanisms, to be 2.40x10 This is within the acceptance 4
criterion of lx10 specified in Generic Letter 95-05. The worst case leakage following a main steam line break was calculated to be less than 8.7 gpm, well within the plant makeup system's capability. An offsite dose calculation was performed with this leakrate utilizing Standard Review Plan methods, with the exception of a probabilistic iodine spiking model for the coincident iodine spike case. In addition, ICRP 30 dose conversion factors were used l
with the resulting calculated offsite doses falling below 10CFR100 limits. A complete report l
summarizing the result of the evaluations performed to date is attached.
U. S. NRC July 30,1996 2CAN079604 Page 3 i-Following the NRC's review of this report, Entergy Operations would be pleased to answer any questions you may have. Please. contact me if you have any specific questions or if you believe a meeting to discuss the results of our evaluations is necessary.
Very truly yours, hl) p 4
wi
.Mims d
Dire to Nuclear Safety i
DCM/dej attachment i
cc:
Mr. Leonard J. Callan Regional Administrator U. S. Nuclear Regulatory Commission t
Region IV l
611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 l
Mr. George Kalman l
NRR Project Manager Region IV/ANO-1 & 2 l
U. S. Nuclear Regulatory Commission l
NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 l
f i
l l
I
,