View source for 1CAN050401, Steam Generator C-3 Report for 1R18

{{Adams
| number = ML041340444
| issue date = 05/10/2004
| title = Steam Generator C-3 Report for 1R18
| author name = James D
| author affiliation = Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000313
| license number = DPR-051
| contact person = 
| case reference number = 1CAN050401
| document type = Inservice/Preservice Inspection and Test Report, Letter type:
| page count = 11
}}

=Text=
{{#Wiki_filter:Entergy Operations, Inc.
Ad En7ergy                                                                              1448 S.R. 333 Russellville, AR 72802 Tel 501858 5000 1CAN050401 May 10, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

==Subject:==
Steam Generator C-3 Report for 1R1 8 Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

==Dear Sir or Madam:==

Entergy is providing the attached Steam Generator (SG) inspection report to the Commission in accordance with ANO-1 Technical Specification 5.6.7.b. The report contains the results of SG tube inspections which fall into Category C-3. The inspection identified 234 tubes in "A" SG and 78 tubes in "B" SG which required plugging. Additionally, re-roll repairs were performed in the upper tubesheet resulting in 233 rolls installed in "A" SG and 110 rolls installed in "B" SG.
There are no new commitments contained in this correspondence. Should you have any questions regarding this submittal, please contact Fred Van Buskirk at (479) 858-3155.
Sincerely, Dale E. ames Mana ers Licensing DEJ/fpv Attachment
                                                                                                           @oL T)

I CAN050401 Page 2 cc:    Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Tom Alexion Mail Stop 0-7 D1 Washington, DC 20555-0001

Attachment I CAN050401 ANO-1 Steam Generator Inspection C-3 Report

ANO-1 1R18 C-3 Report The current steam generator inspection resulted in the steam generators being classified as C-3 (>10% of the tubes inspected are degraded or >1% are defective). ANO-1 began the outage with 14,536 in SG "A" and 15,133 SG "B" tubes in service. This report documents the C-3 report as required by Technical Specifications 5.6.7.b. There were 234 tubes plugged in SG 'A"and 78 plugged in SG "B". Additionally, re-roll repairs were performed in the upper tubesheet. There were 233 rolls installed in SG "A"and 110 in SG "B". The following is a summary of the historical repairs to date:
ANO-1 OTSG Repair Summary
                              # Plugged     Pugged7   j
                                                    'W#'I  insUi umulative                B   =                  GE D OUTAGE    DATE   EFPD EFPY    A     B    A     B       31  80"    31" 80"            A      B     A     B
'RESERV   Dec-74               0     2    0     2        0   0      0    0             0     2  0.00% 0.01%
1   Jan-77   500   1.37  0     0    0     2        0    0     0    0             0     2  0.00% 0.01%
2   Feb-78   770   2.11  5     0    5     2        0    0     0    0             5     2  0.03% 0.01%
3   Mar-79  1060   2.90  0     0    5     2        0   0      0    0             5     2  0.03% 0.01%
1L4A    Jul-80              3     0    8     2        0   0      0    0             8     2  0.05% 0.01%
1L4B   Sep-80               1     0    9     2        0    0     0    0             9     2  0.06% 0.01%
4   Jan-81  1389   3.81  7     0   16     2        0   0      0    0            16     2  0.10% 0.01%
I L5  May-82               10    0   26     2        0   0      0    0            26     2  0.17% 0.01%
5   Nov-82  1835   5.03 84    48   110    50       0   0      0    0           110     50 0.71% 0.32%
1L6    Ju1-83             43     0   153    50       0   0      0    0           153    50  0.99% 0.32%
1M6    Mar-84              43    37   196    87       0   0      0    0           196    87  1.26% 0.56%
6   Oct-84  2234   6.12 24     7   220    94       0   10     0    0           222    95  1.43% 0.61%
7   Sep-86  2678   7.34  1     4  221    98        0   50     0    0           229    103 1.47% 0.66%
1L8   Jan-87               1     0  222    98        0   50     0    0           230    103 1.48% 0.66%
8   Aug-88  3116   8.54  13    0  235     98       0  148     0   76           258    113 1.66% 0.73%
1F89-1  Feb-89               0     0  235    98        0  148     0    0           258    113 1.66% 0.73%
9   Oct-90  3511   9.62  0     9  235    107       0  176     0  147           263    125 1.69% 0.80%
10   Feb-92  3895  10.67 235   71  470    178      54  351    123 366           531    233 3.42% 1.50%
11   Sep-93  4371  11.98 22    12  492    190      54  427    123 366           565    253 3.64% 1.63%
12   Feb-95  4826  13.22  7     3  499    193      54  427    122 366           572    255 3.68% 1.65%
13   Sep-96  5334  14.61 68    48  567 241         54  422    121 363           639    303 4.12% 1.95%
14   Mar-98  5832  15.98 66    30  633 271         54  421    121 363 1963 1155 705    333 4.54% 2.14%
15   Sep-99  6333  17.35 213   63  846    334      54  414    121 363  78   33  917   404  5.90% 2.60%
16   Mar-01  6855  18.80 73    21  919    355      54  376    121 359 232 94    984    424 6.34% 2.73%
17   Oct-02  7384   20.2 76    43  994    398      53  374    121 358 187 154   1059   467 6.82% 3.01%
18   Apr-04  7805  21.37 234   78  1228 476        53  371    120 357 233 110   1292   545 8.3%   3.5%

Page 2 of 8 1.0      OTSG Description Functional Description ANO-1 is a Babcock & Wilcox Nuclear Steam System Supply design with two once-through design steam generators (OTSGs). The OTSG is a straight-tube, straight-shell, vertical, counter-flow; once-through heat exchanger with shell-side boiling to the extent that the steam produced is superheated. By nature of its design, the OTSG eliminates the need for steam separating equipment.
Primary fluid from the reactor enters through the OTSG inlet nozzle (hot leg) in the top head, flows down through the tubes, is collected in the bottom head and exits through two primary outlet nozzles (cold leg). The feedwater enters through a series of spray nozzles near the top of the annular feedwater heating chamber. Here, the feedwater is heated to saturation temperature by direct contact with high-quality or slightly superheated "bleed" steam. The resulting saturated feedwater enters the tube bundle through ports near the bottom of the tube bundle. Nucleate boiling starts immediately upon contact with the hot tubes. Steam quality increases as the secondary fluid flows upward between the tubes in counter-flow to the primary fluid inside the tubes. The departure from nucleate boiling occurs at about the 30-foot level at full power conditions. The mode of heat transfer then transitions from nucleate to film boiling. Steam quality continues to increase, but at a slower rate. After 100% quality is reached, the steam becomes superheated; flows radially outward; leaves the tube bundle at the upper tubesheet; flows down the steam annulus; and exits through two steam outlet nozzles.
Design Information An OTSG unit weighs approximately 570 tons, has an outer diameter of 12-1/2 feet and an overall height of 73 feet. Each steam generator has 15,531 triangularly pitched, alloy 600 tubes. The tube material is mill annealed alloy 600 that is in a sensitized condition as a result of undergoing a full-vessel stress relief at 1100 - 1150 degrees Fahrenheit for 12 to 18 hours. These tubes are 5/8-inch, outside diameter (OD) x 0.037-inch nominal wall x 52 ft. long excluding tubesheet lengths. They are partially roll-expanded (1 inch minimum from the cladding surface) and attached to the upper and lower tubesheets by fillet welds. The use of straight tubes results in almost pure counter-flow with resulting improved secondary flow distribution and primary-to-secondary temperature differentials. This design also has the benefit of placing the tubes in compression during normal operating conditions. This is due mainly to the fact that the alloy 600 tubes have a thermal coefficient of expansion slightly greater than that of the carbon steel shell. This compressive load tends to inhibit the initiation and propagation of stress related damage mechanisms. The OTSG tubes are subjected to tensile loads during cool down transients and postulated accident transients, such as a Main Steam Line Break (MSLB) or a small break loss of coolant accident condition, when the OTSG shell is at a higher temperature than the tubes.
The OTSG's have 15 tube support plates (TSPs) that are fabricated from 1-1/2 inch thick carbon steel plate, drilled and broached to provide surface contact and support along three axes for each tube at each tube support plate. An exception is the 15th TSP periphery tube locations, which are not broached, but only drilled. The support plates are non-uniformly

Page 3 of 8 spaced axially at intervals of 35 to 46 inches to prevent resonant vibrations along the tube length, thus providing a high damping factor.
2.0      INSPECTION RESULTS The initial scope and expansions based on number of examinations are listed in Table 2.1:
Table 2.1 1R18 Inspection Summary Test Type                 # Planned   % Scope      Category     Expansion SG "A" General Examination                                          C-3
* Bobbin                             14536         100                       No
* RPC URT                            11862         100                       No
* RPC LRT                             5014         34                        No
* 80" 1600 B&W Sleeve                  101         100                       No
* UTS IGA                             328          100                       No
* RPC LTS Sludgepile/Crevice          659          20                        No
* Dented Locations                     128         20                        No
* Re-rolls                            2247         100                       No SG "B"                  #Planned    % Scope      Category     Expansion General Examination                                          C-3
* Bobbin                             15133         100                       No
* RPC URT                            13280         100                       No
* RPC LRT                             5257         34                        No
* 80" 1600 B&W Sleeves                  69         100                       No
* UTS IGA                              218         100                       No
* RPC LTS Sludgepile/Crevice          436          20                        No
* Dented Locations                      65         20                        Yes
* Dented Locations CA Exp              213         100                       Yes
* Re-rolls                            1374         100                       No There are two alternate repair criteria (ARC) approved for use in the ANO-1 SG's. The first is for outer diameter intergranular attack (IGA) limited to the upper tubesheet. This is documented under Technical Specification Amendment 213 (1CNA030106). The second ARC is for cracking in the tube ends (TEC). This is documented under Technical Specification Amendment 201 (1CNA099903).             In addition, Technical Specification Amendment 212 (1CNA030105) was approved to install re-rolls in the upper and lower tubesheets due to degradation associated with TEC and upper roll transition (URT) cracking.
3.0      EVALUATION OF INSPECTION RESULTS The inspections were performed with equipment and techniques qualified in accordance with Appendix H of the Electric Power Research Institute (EPRI) PWR Steam Generator Examination Guidelines, Rev. 6, or demonstrated equivalent. Table 3.1 lists the type and number of indications repaired during 1R18:

A -  Page 4 of 8 TABLE 3.1 1R18 Flaw Summary SG "A"                 SG 'VI
                                                    # Defects              # Defects Upper Tube End                               48                       6 Upper Roll Transition                        81                       91 Reroll Transitions                           16                       17 UTS Tubesheet Crevice                        63                        1 Upper Tubesheet IGA                           2                       3 Freespan294                                                          167 Sleeve Transitions                            0                       0 DentsDings                                    0                       8 Lower Tubesheet Crevice                       0                       0 Lower Roll Transition                         2                       0 Lower Tube End                                1                       4 As noted in Table 3.1, the number of defects in SG "A' and "B" were greater than 1% of the total tubes inspected in that SG. The tubesheet IGA and the tube end anomalies were covered under the ARCs and did not count toward the C-3 categorization. The mechanisms will be discussed individually.
Upper Tubesheet Intergranular Attack (IGA)
Intergranular Attack in the upper tubesheet has been present since the early 1980's with very little change or growth. Since the cause of the IGA (high sulfate due to resin intrusion in the late 1970's) has been eliminated, the IGA initiation and growth is essentially zero. An ARC has been implemented for the upper tubesheet IGA. The ARC allows the IGA detected during the 1R18 inspection to remain in service. The approach is based on comparing plus-point voltage and axial extent using a sign test and a paired t-test. As part of the acceptance criteria an overall population growth comparison was performed between the IGA flaws detected during 1R17 and 1R18. There were 374 IGA patches in 310 tubes in the "A" OTSG and 251 IGA patches in 159 tubes in the "B" OTSG that made up the population for the growth comparison.
The statistical tests to validate the "no growth" hypothesis were then performed for the voltages and axial extents. These tests were performed separately for each steam generator. All of the tests performed supported the "no growth" hypothesis.
Based on the conclusion of no population growth, the ARC was implemented. Additional measures per the management program require each indication to be assessed against a set of repair limits. Two indications in SG "A" and three in SG "B" exceeded the repair limit for voltage and/or the extreme value test. These five tubes were plugged.
Tube End Cracking (TEC)
During the 1R14 refueling outage a 100% plus-point upper roll transition (URT) area inspection was performed and as a result indications were detected in the Heat Affected Zone (HAZ) region of the tube end. A root cause analysis determined the flaws to be most likely Primary Water Stress Corrosion Cracking (PWSCC). PWSCC has been confirmed in this general area (URT) through tube pulls at ANO-1, Oconee and Davis Besse. The B&W

r s Page5of8 Owners Group developed an alternate repair criterion for this specific degradation mechanism and location. The ARC was approved by the NRC for use at ANO-1 on September 14, 1999, allowing the indications to remain in service based on meeting specific criteria including radial position and associated flaw leakage.
During the 1 R1 8 inspection there were a total of 1776 tubes containing HAZ cracks in the
    'A" OTSG and 1038 tubes in the "B" OTSG. Of these flaws there were a total of 55 tubes that did not meet the ARC (e.g., circumferential orientation) and were repaired. Leakage was assigned to the flaws remaining in service based on the radial position and corresponding leakage as identified in the TEC ARC topical report which was the basis for Amendment 201. An adjustment of 15% is required to account for the Probability of Detection (POD) for flaws detected during 1 R1 8. Based on the implementation of the ARC, the operational assessment is essentially performed for Cycle 19 with the postulated end of cycle (EOC) accident induced MSLB leakage calculated to be 0.43 gpm for the worst case OTSG. This value will be added to the other contributing mechanisms to determine the total EOC leakage under accident conditions.
Upper Roll Transition (UR7)
The OTSG's were fabricated with a one-inch hard roll in each end of the tube and seal welds on the tube ends to the cladding. A 100% examination of the hot leg upper roll transitions was performed using the rotating plus-point/pancake coil probe. Axial and circumferential cracks and volumetric flaws were detected during the inspection resulting in the identification of 172 total indications (81 in "A" and 91 in "B"). All of these indications were re-rolled. The cracking is attributed to PWSCC in the roll transition regions of the tubes, while the volumetric indications were OD initiated. EPRI testing and B&W Owners Group OTSG tube integrity program testing has verified that axial indications constrained within the tubesheet will not burst, thus these flaws pose no structural challenge. The circumferential and volumetric flaw sizes met structural and leakage requirements under design basis accident conditions.
Lower Tubesheet (LTS) Sludgepile/Dented Region During 1R18, a 100% bobbin inspection of the LTS sludge pile/dented region and a 21%
plus-point examination was performed. This encompassed 659 and 436 tubes in "A" and "B" SGs, respectively. No indications of degradation were found during this inspection.
Re-roll Transition The 1R14 refueling outage was the first time that ANO applied the re-roll ARC for repairing defective tubes. The repair roll creates a new RCS pressure boundary beyond the degraded section of tubing (below the indication and original transition in the upper tubesheet). A second re-roll campaign was accomplished during1R15.              As part of the requirements for ARC implementation, a 100% eddy current exam was performed during 1R18 on the rolls installed prior to this outage. This exam consisted of using an RPC/plus-point technique for detection of degradation.           During this exam there were several indications detected mostly in the heel transition which were outside the pressure boundary, and there were some indications that propagated into the one-inch effective roll or initiated at the toe roll (16 in "A" and 17 in "B"). These indications were broken out into two categories (volumetric and axial/mixed mode). These indications were repaired by installing a second re-roll below the initial re-roll. The axial/mixed mode indications are likely PWSCC

2 1. Page6of8 at the roll transition and, in some cases, extended into the one-inch effective roll. The volumetric and mixed mode indications have been classified as ODIGA and IGSCC.
Freespan IGSCC This form of degradation is classified as axial aligned IGA and believed to be associated with grooves created by the broached lands when the tubes were inserted during assembly.
The degradation is located in the freespan between the support structures, mainly in the upper bundle region (elevated temperature). This degradation was identified in the ANO-1 OTSG tubing for the first time in 1R13. The number of tubes with indications detected exceeded the estimate in the operational assessment. The original number was based on historical experience at ANO during previous refueling outages. The increase in 1R18 is primarily attributed to the training provided to the eddy current analysts on broadening criteria for identifying freespan indications based on industry experience. One other operating OTSG plant identified in a previous inspection that indications that had higher than expected phase angles (resulting in very low to zero percent through wall values) should be identified as potential repairable indications. This practice was put into place during the ANO-1 1R18 inspection and resulted in a significant number of additional indications. Repairing these indications increases the confidence level that structurally significant flaws will not develop during the next operating cycle.
Super-Heated Dent Inspection The superheated region is defined as the region above the 8th TSP. The initial examination consisted of a 20% Motorized Rotating Pancake Coil (MRPC) dent examination at the 8th TSP and above including the freespan and UTS areas. Defining this region as a critical area is based on past ANO-1 MRPC inspection results and a sharp temperature gradient below the superheated region between the 8th and 9th TSPs. The tube temperature drops slightly (approximately 50 F) as the reactor coolant flows from the UTS to the 9th TSP.
Beyond the 9 th TSP the gradient increases. Selecting the 8th TSP provides reasonable assurance that the tube temperature in the defined critical area is measurably higher than the remaining areas of the OTSG's.
Initially a 20% MRPC exam of dents >2.0 volts was scheduled for both "A" and "B" OTSGs.
During the 20% exam in the "B" OTSG one indication was found: A critical area was then developed that contained all dents >2.0 volts between the 15th TSP and 3" above the secondary face of the UTS. A buffer zone was created between the 14th support plate and 15th span. A second indication was identified and the critical area and buffer zone was redefined to be the 10th support to the UTS + 3 inches and 7th support to the 10th support respectively. A 100% MRPC exam was performed on dents >2.0 volts from the 7t support plate to 3" above the secondary face of the UTS in the "B" OTSG. There were a total of eight single axial indications (SAI) reported at dents in the "B" SG. All indications were located above or near the 12S support plate. No additional indications were identified in the buffer zone. 20% of the critical area in SGA was acquired in the initial sample. No further expansions were required.
Sleeves A 100% (170 sleeves) inspection of all roll transitions in the 1600 sleeves in the "A" and "B" SGs were MRPC inspected during 1R18. No flaws were found in the pressure boundary of any of the roll transitions.

Page 7 of 8 Insitu Pressure Testing Insitu pressure testing was performed on I tube in the VB"OTSG. The indication was an axial crack in a dent in the super heat region of the tubing. The reason for testing was to determine the as-found condition of the tubing and provide information for assessment of the upcoming operating interval. Selection of the tube was in accordance with the EPRI Insitu Pressure Testing Guidelines, Rev. 2. The tube is listed in Table 3.2. The tube was tested     in stages at          500 psi increments             at   pressurization     rates of
< 200 psi/sec. which included normal operating, MSLB, and three times normal operating AP. No leakage was observed at normal operating AP, MSLB, nor 3AP. Based on the result of this test and previous insitu tests performed at ANO-1, all tubes met the condition monitoring criteria for accident-induced leakage and structural integrity.
TABLE 3.2 ANO-1 IN-SITU TEST RESULTS FOR 1R18 SG      Row       Tube         Flaw Type          NOP           MSLB          3AP       Leakage B        88        40     SAI - Freespan         1600          3050          4500          0 SAI - Single Axial Indication 4.0      Root Cause and Corrective Actions The tubing degradation identified above has been previously reported to the NRC in earlier reports. The majority of the affected tubes that are classified as C-3 were attributed to freespan axial cracking. Additionally, cracking at the original roll transition and axial cracking in dents were also identified.
The roll transition cracks are attributed to PWSCC. A 100% inspection of the upper tubesheet was performed with the plus point. All indications were repaired by re-rolling or plugging the tube. Based on the age and material properties, it is anticipated that this degradation will continue, but is manageable such that the tube integrity will be maintained until steam generator replacement in 2005.
The freespan cracks are identified as outside diameter IGA/SCC and are generally located in the area of grooves/scratches in the tubing. The increase in freespan groove IGA is the result of a more focused inspection. Industry experience has identified that indications which read essentially zero percent through-wall can still be a crack. Additionally, recent results at other OTSG plants have identified that higher phase values (outside the calibration) should also be reported. The indications are first identified by bobbin as a non-quantifiable indication (NQI). Analysts were trained to focus on this area of degradation and conservatively call indications. These are then tested with the plus point for confirmation.
Approximately twice the number of NQIs were called by bobbin and the confirmation rate was increased from approximately 3 percent to 22 percent. This is indicative of an increased POD. The combination of industry experience and analyst sensitivity contributed to the increase in the number of indications. The extreme values of the population were consistent with past results. Similar to the roll transition cracking, this mechanism is manageable until replacement in 2005 based on the 100% bobbin inspection with plus point confirmation.

Page 8 of 8 5.0     Condition Monitoring Results from the 1R18 inspection were compared to the 1R17 operational assessment results. The actual number of detected indications was within the frequency of occurrence predicted by the operational assessment model with the exception of freespan axial indications in "A" SG that were discussed previously. Thus, the flaw distribution predicted by the model was reasonable with respect to the actual flaw projections.
In-situ pressure test screening was performed on the most limiting flaws. The one flaw that was in-situ pressure tested passed all pressure target values including 3AP and exhibited no leakage.
The sum total of the leakage attributed to all identified indications was less than the limit of 1.0 gpm at MSLB conditions. The majority of the leakage is attributed to the ARC for tube end cracking and upper tubesheet intergranular attack. These values are calculated based on the identified population. There were no tubes identified with actual leakage.
Based on meeting both burst and leakage criteria, condition monitoring was met.
Repairs were made to the indications that were not covered by the ARCs or support plate wear which is sized and left in service below 40% through-wall. Preventative repairs were made to reduce the total leakage with added margin.
6.0     Operational Assessment Based on the results of the condition monitoring assessment, the 1R18 eddy current testing, and in-situ pressure test results, a preliminary operational assessment has concluded that the OTSG's are acceptable to operate until the final Cycle 19 operational assessment is completed.
7.0     Conclusions In summary, a comprehensive eddy current examination was performed in accordance with NEI 97-06. Both SGs were tested 100% full length with the bobbin coil and 100% at the hot leg roll transition region with plus-point. An in-situ pressure test was performed in accordance with EPRI guidance, which demonstrated that the condition monitoring criteria were satisfied.
Based upon the comprehensive actions performed during 1R18, which included the plugging/repair of 655 tubes, in conjunction with the ability to rapidly detect and respond to any primary-to-secondary leakage, ANO-1 is safe to resume plant operation. Therefore, considering the results of the 1R18 inspection, it is believed that the Cycle 19 operational assessment will demonstrate that it is safe to operate ANO-1 for the entire duration of the next fuel cycle. The next planned outage is in the fall of 2005, which will include replacement of the currently installed OTSG's.}}