ML20006A053
| ML20006A053 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/12/1990 |
| From: | Reeves E Office of Nuclear Reactor Regulation |
| To: | Hairston W ALABAMA POWER CO. |
| References | |
| IEB-88-009, IEB-88-9, TAC-72659, TAC-72660, NUDOCS 9001250143 | |
| Download: ML20006A053 (59) | |
Text
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January 12, 1990 g
i Docket Nos. 50-348 and 50 364 Mr. W. G. Hairston, 111 Senior Vice President
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Alabama Power Company 40 Inverness Center Parkway Post Office Box 1295 i
Birmingham, Alabama 35201
Dear Mr. Hairston:
SUBJECT:
AUDIT OF ACTIONS RELATING TO NRC BULLET 1H 38 09, " THIMBLE i
TUBE THINNING IN WESTINGHOUSE REACTORS," FOR JOSEPH M. FARLEY i
NUCLEAR PLANT, UNITS 1 AND 2, (TAC NOS, 72659 AND 72660)
By letter dated April 17, 1989, we advised you of completion of our review of your November 2,1988 response to the subject bulletin. As a followup action, we conducted a site audit on the issue of incore thimble tube wear on November 14 and 15 '989.
We appreciate your staff assistance during that audit.
A copy of the audit trip report is enclosed for your information. The eudit concludes that the inspection program is responsive to the Bulletin requirements, uses' acceptable. inspection methods with technically justifiable acceptance criteria, and you have a schedule for conducting inspections every refueling outage. These interim actions are acceptable.
In addition, your participation with the Westinghouse Owners Group working toward the final resolution of the issue is noteworthy.
Sincerely, Edward A. Reeves, Senior Project Manager
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Project Directorate 11 1 Division of Reactor Projects 1/11 Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page i
DISTRIBUTION:
,gM., im OGC (info only)
LMarsh 1
l NRC &.ocal PDRS EJordan BBuckley) jl PD21 r/f SVarga,14/E/4 ACRS (10 p01 Glainas,14/H/3 EAdensam EReeves PAnderson DVerrelli, Ril l l 0FC :LA:PM
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'NAME: pan 3ff
.EP s:s1 am i.....;...........:............:.............................................................
DATE:1//l/90
- 1/// /90
- 1/g/90
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l 9001250143 900112 PDR ADOCK 05000348 1
O PDC
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t Docket Nos. 50 348 I
and 50 364 i
Mr. W. G.
irs n, III Senior Vic Pr sident Alabama Powe Company Post Office x 2641 Birminghair A bama 35291 0400
Dear Mr. Hairst :
SUBJECT:
AUDIT OF ACTIONS RELATING TO NRC BULLETIN 88-09, " THIMBLE TUDE THINNING IN WESTINGHOUSE REACTORS " FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, (TAC NOS. 72659 AND 72660)
Dy letter dated April 17, 1989, we advised you of completion of our review of your November 2,1988 response to the subject bulletin.
As a followup action, we conducted a site audit on the issue of incore thimble tube wear on November 14 and 15, 1989. We appreciate your staff assistance during that audit.
A copy of the audit trip report is enclosed for your information. The audit concludes that the inspection arogram is responsive to the Bulletin requirements, uses acceptable inspection met 1ods with technically justifiable acceptance l
criteria, and you have a schedule for conducting inspections every refueling -
outage.
These interim actions are acceptable.
In addition, your participation with the Westinghouse Owners Group working toward the final resolution of the issue is noteworthy.
Sincerely, Edward A. Reeves, Senior Project Manager.
Project Directorate II.1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: See next page DISTRIBUTION:
Docket File OGC (info only)
LMarsh NRC & Local PDRS EJordan BBuckley l
PD21 r/f SVarsa, 14/E/4 ACRS(10)
Glainas,14/H/3 EAdensam EReeves PAnderson DVerrelli, RII I 0FC :LA:
1
- PM:
I
- D:PDII.1 NAME :PArp
- EREeve :s1 :EAdensam t.....:............:............:............:............:.........................:.........
DATE :1/lb/90
- 1//0/90
- 1/ /90 3
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Docket Nos. 50 348 and 50 364 Mr. W. G. Hairston, 111 Senior Vice President Alabama Power Company Post Office Eox 2641 Birmingham, Alabama 35291 0400
Dear Mr. Hairston:
SUBJECT:
AUDIT OF ACTIONS RELATING TO NRC BULLETIN 88 09, " THIMBLE TUBE THINNING IN WESTINGHOUSE REACTORS," FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, (TAC NOS. 72659 AND 72660)
By letter dated Aoril 17, 1989, we advised you of completion of our review of your November 2,1988, response to the subject bulletin. As a followup action, we conducted a site audit on the issue of incore thimble tube wear on November 14 and 15, 1989. We appreciate your staff assistance during that audit.
A copy of the audit trip report is enclosed for your information. The audit I
concludes tht the inspection program is responsive to bulletin requirements.
You use acceptable inspection methods with technically justifiable acceptance criteria. You have a schedule for conducting inspections every refueling outage. These interim actions are acceptable.
In addition, your participation with the Westinghouse Owners Group working toward the final resolution of the issue is noteworthy.
Sincerely, Edward A. Reeves, Senior Project Manager Project Directorate 11 1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
See next page DISTRIBUTION:
Docket File OGC ( nfo only)
LMarsh NRC & Local PDRS EJor an BBuckley PD21 r/f.
SVar a, 14/E/4 ACRS(10) f Glainas,14/H/3 EAd sam EReeves y
PAnderson DVe elli, RII OFC :LA:PDII.1
- PM:PDll-1
- D:PDil.1 NAME :PAnderson
- EReeves:s1
- EAdensam 1
DATE :1/ /90
- 1/ /90
- 1/ /90
8 Mr. W. G. Hairston, III L
Alabama Power Company Joseph M. Farley Nuclear Plant l
Cc:
Mr. R. P. Mcdonald Resident Inspector Executive Vice President U.S. Nuclear Regulatory Commission Nuclear Operations P. O. Box 24 - Route 2 Alabama Power Company Columbia, Alabama 36319 P. O. Box 1295 Birmingham, Alabama 35201 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Mr. B. L. Moore 101 Marietta Street, Suite 2900 Manager, Licensing Atlanta, Georgia 30323 Alabama Power Company P. O. Box 1295 Chairman Birmingham, Alabama 35201 Houston County Commission Dothan, Alabama 36301 Mr. Louis B. Long, General Manager Southern Company Services, Inc.
Claude Earl Fox, M.D.
Houston County Commission State Health Officer P. O. Box 2625 State Department of Public Health Birmingham, Alabama 35202 State Office Building Montgomergy, Alabama 36130 Mr. D. N. Morey General Manager - Farley Nuclear Plant P. O. Box 470 Ashford, Alabama 36312 Mr. J. D. Woodward
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Vice-President - Nuclear Farley Project Alabama Power Company P. O. Box 1295 I
Birmingham, Alabams 35201 l
i ENCLOSURE l
o AUDIT TRIP REPORT i
I PURPCSE:
Audit of Joseph M. Farley Units 1 & 2 on Bulletin 88-09 Issues (BMI Thimble Tube Wear)
LOCATION:
Joseph M. Farley Nuclear Plant, Dothan, Ala.
DATE:
November 14-15, 1989 NRC PERSONNEL:
S.N.
Hou (NRC),
G. DeGrassi (BNL)
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LICENSEE PERSONNEL:
A.E. Hammett (APCo),
C.M. Scrabis (Westinghouse) and others (See Attachment 1) i t
1.0 INTRODUCTION
The purpose of this audit was to review the Licensee's activities related to the Bulletin 88-09 issues on BMI thimble tubo This bulletin requires the eatablishment of an inspection wear.
program to monitor thimble tube wear.
The program should include the establishment, with technical justification, of appropriate l
acceptance criterion, inspection methodology and inspection frequency.
The program should be implemented-in accordance with the given schedule (next refueling outage for'most plants), and corrective actions should be taken for tubes which fail to meet the established acceptance criterion.
The Alabama Power Company's written response to the bulletin was included in a letter dated November 2, 1988 and is included as.
A copy of the audit agenda ir included as Attachment 2 and meeting attendance lists for the entrance and exit meetings are included in Attachment 1.
An entrance meeting was held on November 14.
The Licensee gave an overview of the plcnt's inspection program and results.
On November 15, more detailed discussions on these subjects were held primarily with Al Hammett and Chuck Scrabis.
Documentation including eddy current inspection reports, analyses and design drawings were made available for our review.
Additional informa-i tion was provided af ter the audit to resolve some open issues raised during our diccussions.
A brief exit meeting was held at the end of the audit to summarize our findings and recommendations.
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1 2.0 HIGHLIGHTS OF FARLEY EXPERIENCE AND LICENSEE ACTIONS o
APCo established an eddy current inspection program in 1986 with an inspection frequency of every refueling outage.
This frequency will be maintained until sufficient data is available to justify a longer interval between inspections, To date, no thimble tube leaks have occurred at Farley.
o o
To date, three eddy current inspections have been performed on Unit i thimble tubes and two inspections have been performed on Unit 2.
Significant wear has been detected in several tubes and corrective actions including tube repositioning and capping have been taken.
o A 65% wall loss acceptance criteria was established based on Westinghouse analysis.
o The thimble tubes have manual isolation valves which could isolate a tube leak if it should occur.
The Licensee has considered long term corrective actions o
but has made no definite commitment to date.
The Licensoe is participating in the Westinghouse owners o
Group (WOG) program on thimble tubes.
This program was scheduled for completion by the end of this year, but significant delays are expected.
3.0 AUDIT
SUMMARY
The following is a summary of the information obtained through our discussions and document reviews.
3.1 Thimble Tube Inspection Program The Licensee had established a thimble tube eddy current inspection program before the-issuance of NRC Bulletin 88-09.
Inspections were conducted during each refueling outage since 1986.
i Unit 1 was inspected during the cycle 7, 8 and 9 refueling outages i
in _ October 1986, March 1988 and September 1989.. Unit 2 was inspected during the cycle 5 and 6 outages-in November 1987, and April 1989.
The first three inspections were performed by Cramer and Lindell.
The last two 1989 inspections were performed by Echoram, a Westinghouse subsidiary.
Both vendors used similar eddy current test equipment and a multifrequency inspection procedure.
Measurement uncertainty was quoted as 10% by Echoram' and 15% by Cramer and Lindell.
The wear scar calibration standards were l
different.
Cramer and Lindell used a 180' crescent-shaped stan-dard.
Echoram used a 90' flat tapered wear scar.
Since eddy
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1 current inspection is based on volumetric measurement, the 90' wear scar should be more conservative.
i A summary of all the eddy current inspection results for both units is provided in Attachment 5.
It is difficult to compare exact changes in wall loss between cycles because of the uncer-tainties and the differences in the two vendors' reporting methods.
Cramer and Lindell reported a range of values while Echoram reported a single value but added a 10% uncertainty factor.
Nevertheless, it is clear that tube wear has been observed since the first inspection and tends to increase after each cycle.
Wear has been occurring mainly in the lower core plate area.
The core map figures summarize the latest inspection results.
In Unit 1, there are 36 out of 50 tubes with wear between 17% and 67%.
Three tubes with wall losses of 62%, 634 and 674 have been capped.
One tube was capped because it was blocked and could not be inspected.
Four tubes with wear between 45% and 51% have been repositioned.
In Unit 2, there are 10 out of 50 tubes with wear between 13% and 53%.
Two tubes with 53% wear were repositioned.
One tube (L9) had been repositioned after the first inspection although the 1989 inspection did not measure any detectable wear of the L9 tube.
No Unit 2 tubes have yet been capped.
The criteria for capping or repositioning thimbles was based on Westinghouse recommendations.
Westinghouse had performed a finite element analysis to demonstrate the structural adequacy of a thimble tube with a two inch long flat wear scar covering 90* of the tube circumference.
The corrective action criteria is to cap any tube with wall loss exceeding 65% and to reposition any tube that is predicted to exceed 65% wall loss before the next inspec-tion.
For Unit 1, Westinghouse assumed an average wear rate of 11.8% wall loss per cycle to predict total wear at the end of the next cycle.
Although wear rates exceeding that amount were ob'ierved between cycles for several tubes, Westinghouse feels that this rate is conservative in the longer term.
In reviewing the inspection reports,- it was noted that-the calibration tube wear scar used by Echoram was not in agreement with the wear scar used in the finite element model.
The model assumed a flat scar with a surface parallel to the tube axis.
The l
calibration standard used a flat wear scar with a surface inclined i
relative to tube axis.
Westinghouse was asked to justify the difference and agreed to provide additional information (see Section 3.4).
The Licensee stated that eddy current inspections will be performed during every refueling outage until there is sufficient data to justify longer inspection intervals.
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3.2 Design Parameters i
The Licensee provided design information and drawings on the thimble tube systems and reactor internals.
The thimble tubes are fabricated from SA213 Type 316-stainless steel cold drawn, heat treated tubing with
.300" OD and
.201" ID.
The lower reactor internals guide column dimensions were not available but Westinghouse agreed to provide the information (see Section 3.4).
The ID at the lower core plate is.545" in Unit i and.600" in Unit i
2.
The high pressure conduits which support the thimbles from the reactor vessel to the seal table have an ID of.400" in 42 tubes and.600" in 8 tubes.
It was noted that these dimensions are
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smaller than most other plants and would restrict the use of larger thimble tubes.
Both Farley Units are three loop plants with 12 i
foot cores.
Best estimate flow rates are 95900 gpm in Unit 1 and 95200 gpm in Unit 2.
Unit I went into commercial operation in j
December 1977. Unit 2 went into commercial operation in July 1981.
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Both Units have the same bottom mounted incore instrumentation system with thimble tubes that extend from the reactor core down through high pressure conduits to the seal table.
The system consists of drive units, 5 path rotary transfer devices, 10 path rotary transfer devices and manual isolation valves. High pressure seals form part of the reactor pressure boundary at the thimble tube to seal table interface.
In the event of a thimble tube leak, water would enter the ten path transfer devices.
'Each ten. path device has a drain line which feeds into a common drain header.
A level sensing switch is installed in the drain header.
In the event of a leak, the switch would sound an alarm on the flux mapping panel in the control room and open the drain valve which allows the water to drain to the containment sump.
In the event of a leak, plant personnel _would have to enter the containment, identify the leaking tube and close the manual isolation valve.
The Licensee provided a set of photographs of the seal table (see Attachment 4).
The photographs show the BMI transfer cart which is a frame structure that supports the isolation valves and 10 path transfer devices.
During refueling, the thimble tubes must be withdrawn.
This is accomplished by disconnecting the tube coupling at the seal table, jacking up the upper portion of the transfer cart and rolling the entire assembly to the side to provide vertical clearance for tube withdrawal (see photograph).
i When the transfer cart is in the normal operating position, the upper part of the cart is bolted to supports at each end.
- However, it was not clear that the lower part of the cart was restrained.
A concern was raised that during an earthquake the cart could roll and sever the thimble tubes.
The Licensee could not explain how the lower portion of the cart is restrained but agreed to provide additional information (see Section 3.4).
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3.3 Licensee Evaluation and Corrective Action Plans The eddy current inspections showed tube wear at scattered core locations.
There were no apparent trends suggesting that certain core locations are more susceptible to wear than others.
The most severe wear was generally observed at the lower core plate elevation.
More wear has been observed in Unit 1 than in Unit 2.
as may be expected since Unit.1 has been operating for a longer period of time.
The Licensee has been working closel thimble tube inspections and evaluations.y with Westinghouse in the The acceptance criteria for capping or repositioning tubes was developed by Westinghouse.
The use of a linear wear prediction technique based on a wear rate of 11.8%
wall loss per cycle appears nonconservative since increases in wear as high as 35% were observed between cycles 8 and 9 in Unit 1.
In the long term, the 11.8% rate appears conservative for predicting total wear of tubes over many cycles.
However, as seen in other plants, the increase in wear for any particular tubo in a given cycle is unpredictable and can vary significantly.
If a tube leak should occur, however, Westinghouse has demonstrated that maximum leakage of 35 gpm per tube for three tubes can be accommodated by the normal makeup capacity of the system.
Furthermore, there are manual valves available to isolate a leak.
Experience at-other plants has demonstrated the feasibility of isolating a leak with the manual valves. Actual leakage rates have also been well below the worst case 35 gpm predictions. Therefore, although the wear rate prediction method appears nonconservative, the overall program seems reasonable when consideration is given to the consequences of tube leaks, the tube isolation capability, and the planned inspection frequency.
The Licensee has considered some~1ong term corrective actions but has made no definite plans or commitments at this time.
The use of sleeves in the lower internals guide column have been considered.
The Unit 2 guide column ID could be reduced from.600" to.468".
Unit 1, however, has a unique problem in that the ID is only
.545".
Installation of sleeves would require boring out a larger diameter in the core plate and upper guide column to make sleeve installation possible.
The use of larger diameter thimble tubes is not feasible due to the use of some small diameter (.400" ID) high pressure conduits between the seal table and lower reactor vessel.
Westinghouse has been wc. ding on wear resistant coatings for thimble tubes but they are not presently being considered by the Licensee.
Although there is not commitment, the Licensee has allocated funds for tube replacement, if that should be necessary in the future.
The Licensee is participating in the Westinghouse owners Group (WOG) Program on thimble tube wear.
This program will develop more accurate wear scar standards and refine the acceptance criteria for wear based on testing of tube samples from operating plants.
The 5
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program is behind schedule because of the difficulty in obtaining tube samples.
At this time, Westinghouse has four sample tubes available from Diablo Canyon.
Eddy current inspections indicated 90% wear on one tube.
Westinghouse plans to re peat the eddy current inspections and compare the results agatnst hot cell examinations of the sample.
Burst testing of the tube samples is also planned.
Westinghouse expects to receive additional tube samples from Kewaunee and other plants. A firm completion date for the program could not be obtained at this time.
3.4 Closeout of Open Items
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By the end of the audit, there were three open items for which the Licensee agreed to provide additional information.
This was provided in a letter dated November 29, 1989 which is included as.
The information was reviewed and found acceptable.
The following is a summary description of the open items and their resolution.
Item 1 The wear scar geometry described in the Westinghouse finite element analysis report, RPVSA-89-1351, to justify the 65 percent allowable wall loss criteria is different from that described for the wear scar calibration standard in the Echoram ECT Report.
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Resolution 1 The wear scar geometries in the two reports are
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different. The geometry used by Westinghouse in the finite element analysis was judged to be conserva-tive for predicting maximum stress values for givon wear scar depths.
The geometry used by Echoram as a
calibration standard for measuring wear is considered to be a more reprer.entative approximation of actual scars and would be more conservative, for estimating wear depth for a given volume of material removed, than that defined by the finite element analysis model. ECT inspections determine wall loss by measuring the volume of material removed.
A new finite element analysis has been performed by Westinghouse, modeling the geometry of the Echoram calibration standard.
The results of this new analysis shows that the predicted stress values for 65 percent wall loss are less than originally predicted.
This confirms the original analysis assumption that the scar geometry was conservative.
A summat) stress report of this new analysis was included in the November 29th transmittal (Attachment 6).
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Item 2 It is not clear how the bottom portion of the BMI transfer cart, which contains the track wheels for I
rolling the transfer cart away from the seal table, is restrained from rolling or against seismic excitation after the upper portion of the transfer cart is bolted down to its front and back anchor post.
Resolution 2 The upper and bottom portions of the BMI transfer cart are never separated from each other.
The 4 bottle jacks that raise and lower the upper portion relative to the bottom portion are permanently fixed at the top with 4 bolts each to the top portion and at the bottom with 4 bolts each to the bottom-i portica of the transfer cart.
Once the upper t
portion is bolted to its anchor posts above the seal table the entire transfer cart is restrained from moving in a seismic event.
The seismic qualification reports for the Farley transfer carts and copies of drawings showing the areas of concerns were included in the November 29th transmittal (Attachment 6).
Item 2
-Design drawings of the instrument columns in.the lower core area through which the thimble tubes pass showing the variation in internal diameter were not 1
available for review.
Resolution 3 Westinghouse prepared sketches of the Farley Unit 1 and 2 lower internals area showing.3 instrument columns with the changes in internal diameter along the lengths.
These sketches were included-in the November 29th submittal (Attachment 6).
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4.0 CONCLUSION
S AND RECOMMENDATIONS Based on the information obtained during and after the audit, l
our conclusions and recommendations-are as follows:
The Licensee has defined and implemented an adequate program o
j which is responsive to Bulletin 88-09 requirements.
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The program applied acceptable state of the art inspection r
i methods with technically justifiable wear acceptance criteria.
The inspection frequency of every refueling outage is acceptable.
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o The Licensee has taken appropriate short term corrective i
actions (capping and repositioning) to minimize the potential i
for leaks in tubes with significant wear.
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o The Licensee has considered long term corrective actions and is prepared to replace tuber, if necessary.
Although no long term commitments have been made, we understand that the Licensee will continue participating in industry programs (such as WOG) and follow new developments related to this issue.
o Concerns raised at other plants regardirig the seismic l
restraint of the seal table frame assembly do not apply to Farley.
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i ATTACHMENT 2 i
Audit Agenda c
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i AGENDA NRC Audit of Bulletin 8P.-09 Issues Farlev Units 1 &2 4
I.
Audit Discussion Items A.
Thimble Tube Insoection Procram Inspection Methods Description including assumptions o
and uncertainties o
Inspection Frequency and basis o
Wear Acceptance criteria and basis o
Corrective Action o
Inspection Results B.
Review of Parameters Affectina Tube Wear o
Hardware Design - Thimble tubes, Tube supporting-structures internal and external to Reactor 1
o Flow rates o
Isolation capability o
-- Operating History C.
Licensee Evaluation of Wear o
Evaluation of inspection results/significant findings Westinghouse owners Group findings / recommendations o
o Root cause analysis i
Assessment of safety significance o
D.
Lono Tgrm Corrective Action Procram Status o
Addition of sleeves o
Addition of isolation valves i
o Hot cell examination o
Other Long Term Plans i
II.
Document Review o
Inspection Reports /Results o
Design drawings Analyses supporting acceptance criteria and inspection o
frequency other relevant Licensee or Westinghouse reports o
III.
Hardware Inspection i
o Seal table room inspection (if accessible) t l
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ATTACHMENT 3 j
Licensee Response to Bulletin 88-09 4
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Alabama 6we'CDmcany 600 fvo9m 18th $1rtet d
PostOmce Bc 264' B,emingham A;atama 357914400 Teiop%oo+ 20$ 2504637 W o Newsten ill Senior vice Pres,ctet Nuciear operpens AlabamaPower Docket Nos. 50-340 50-364 ene s:wiern,i enc ege, November 2, 1958 U.S. Nuclear Regulatory Commission Attn Document Control Desk Vashington, DC 20555 Centlement Joseph M. Parley Nuclear Plant - Units 1 & 2 thimble Tube Thinning in Vestinghouse Reactors NRC Bulletin No. 88-09 NRC Bulletin No. 68-09 requests that each addressee establish and implement an inspiction program to monitor thimble tube performance and take appropriate corrective actions should the thimble tube fail to meet the established acceptance criterion.
This program should include the establishment and technical justification of an appropriate thimble tube acceptance criterion and inspection frequency and the establishment of an inspection methodology. Holders of operating licenses that already had an established inspection program to monitor thimble tube integrity consistent with that requested by this bulletin and, based upon the results of the last inspection, took appropriate corrective actions for the thimble tubes that failed to satisfy the established acceptance criterion, are requested to implement the inspection pro accordance vith their established inspection frequency. gram in Alabama Pover Company began to utilize the services of an eddy current vendor to perform incore flux measuring system thimble tube eddy current testing (ECT) at Farley Nuclear Plant in 1986. In order to be able to identify a vide range of defects, the ECT vendor developed a calibration standard to include ASME' Boiler and Pressure Vessel Code standard defects, typical vear patterns, and service defects.
includes performing ECT at aach refueling outage until adequateThe current program confidence is established in vear rate projections. Thimble tubes that do not meet the current acceptance criteria are either slightly withdravn, in order to align the vear scar to a nev location and provide an undamaged thimble tube vear surface at locations where the degradation had been previously identified, or capped, depending upon the percentage of vall loss.
During the Unit I seventh and eighth refueling outages and the Unit 2 fifth refueling outage, all thimble tubes (except those blocked or capped) were inspected full length and appropriate corrective actions vere taken.
In the future, thimble tubes that cannot be eddy current inspected due to blockages vill be preventively capped.
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U.S. Nuclear Regulatory Commission tage 2 i
Alabama Power Company vill continue to monitor thimble tube year by i
periodic testing and vill participate in Vestinghouse Ovners Group (V0G) activities to establish recommended testing options, acceptance criteria. and recommended corrective actions.
When issued, the VOC recommended actions vill be reviewed and the Alabama Power Company pregram modified, as appropriate.
In the interim period prior to issuance of the V0G recommendations, Alabama Power Company vill continue with its cur,rently ettablished program which is consistent with the i
requirements of NRC Bulletin 88-09.
t 2
If there are any questions, please advise, Respectfully submitted, ir,)
lYn W
V. G. Hairston III VGH/AEH a
$vorn to and subscribed before me th day of Nuenbed,1988
. L.
nt OriEM D'I~
Mr. E. A. Reeves Hr. G. F. Maxwell NOTARY PUBLIC commission expires O
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4 ATTACHMENT 4 Seal Table Photographs Photo # 1 -
seal Table with BMI Transfer cart in normal operating position.
Photo # 2 -
Seal Table with BMI transfer cart rolled to side and thimble tubes withdrawn.
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1 ATTACHMENT 5.
Eddy Current Testing Inspection Results 4
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Page 1-Unit 1 Thimble Tube ECT results During The 7th and 8th and 9th Refueling Outages PERCENT YEAR TUBE #
VALL LOSS LOCATION
- 1. 1986 01-J7 10-20%
LCP 1988 01-J7 17-20%
LCP 1989 J7 0%
1989 J7 0%
- 2. 1986 02-G7 10-20%
LCP 1988 02-G7 16-21%
LCP 1988 02-G7 17-19%
TP 1989 G7 26%
- 3. 1986 03-G9 BLOCKED 83' 1988 03-G9 BLOCKED Considered permanently blocked.
1989 G9 BLOCKED Capped and isol. 51v. shut.
- 4. 1986 04-H6 10-20%
LCP usfg> 3n 1988 04-H6 18-26%
LCP 1988 04-H6 16-19%
TP 1989 H6 20%
- 5. 1986 05-F8 0%
1988 05-F8 0%
l 1989 F8 0%
- 6. 1986 06-J10.-
0%
1986 06-J10 0%
LCP 1988 06-J10 24-35%
TP 1989 J10 47%
V/D 1"
- 7. 1986 07-F9 0%
1988 07-F9 0%
LCP 1988 07-F9 18-33%
TP 1989 F9 20%
- 8. 1986 08-F6 10-20%
LCP 1988 08-F6 17-38%
LCP 1989 F6 27%
- 9. 1986 09-H11 0%
1988 09-H11 0%
LCP 1988 09-H11 10-19%
TP 1989 H11
- 10. 1986 10-L8-0%
0%
i 1988 10-L8 0%
1989 LB 0%-
- 11. 1986 11-L9 0%
-1988 11-L9 0%
LCP -
1988 11-L9 17-21%
TP 1989 L9 17%
- 12. 1986 12-J5 0%
1988 12-J5 0%
1989 J5 0%
- 13. 1986 13-L6 25%
LCP i
1988 13-L6 19-33%
LCP 1988 13-L6 17-25%
CSF 1989 L6 29%
1989 L6 35%
i
- 14. 1986 14-F11 25-34%-
LCP 1988 14-F11 17-24%
LCP 1989 Fil 38%
4 b
,A l
l' P:ge 2
'15. 1986 15-H4 10-20%
LCP 1988 15-H4 17-26%
LCP 1989 H4 19%
- 16. 1986 16-J12 0%
1988 16-J12 0%
LCP 1988 16-J12 19-35%
TP 1989 J12 31%
- 17. 1986 17-D7 0%
1988 17-D7 0%
LCP-1988 17-D7 9-21%
TP 1989 D7 18%
)
- 18. 1986 18-L11 20%
LCP 1988 18-L11 17-27%
LCP L
1988 18-L11 24-42%
TP l
1989 L11 20%
- 19. 1986 19-L5 25%
LCP 1988 19-L5 18-39%
LCP l
1988 19-L5 17-20%
TP l
1988 19-L5 16-19%
DP l
1989 L5 20%
I-
- 20. 1986 20-E5 10-20%
LCP 1
1988 20-E5 0%
1989 E5 0%
- 21. 1986 21-E11 25%
LCP
~~
1988 21-E11 8-28%
LCP
~
1989 Ell 29%.
- 22, 1986 22-F4 50-77%
V/D 3", capped & Iso.
viv. shut.
1988 22-F4 CAPPED V/D an additional 1/2".
1989 F4 23%
1989 F4 30%
1989 F4 62%
Replaced cap.
l
- 23. 1986 23-D10 34-35%
LCP 1988 23-D10 11-39%
LCP-1988 23-D10 20-39%
TP 1988 23-D10 16-31%
CSF 1989 D10 27%
1989 D10 28%
24, 1986 24-H13 25-30%
LCP 1988 24-H13 24-36%
LCP 1989 H13 36%
- 25. 1986 25-N8 0%
1988 25-N8 0%
LCP-1988 25-N8 16-17%
TP 1989 N8 0%
- 26. 1986 26-L4 25%
LCP 1988 26-L4 19-35%
LCP 1989 L4 30%
27, 1986 27-H3 0%
1988 27-H3 0%
LCP I
1988 27-H3 0%
1988 27-H3 16-19%
TP 1989 H3 36%
- 28. 1986 28-D5 0%
1988 28-D5 0%
LCP 1988 28-D5 16-18%
TP 1989,D5 18%
i
+
h 9
y Paga 3 P-
- 29. 1986 29-C8 0%
1988 29-C8 0%
1989 C8 0%
- 30. 1986 30-N7 0%
1988 30-N7 0%
1989 N7 0%
- 31. 1986 31-J3 0%
1988 31-J3 0%
1988 31-J3 15-26%.
LCP TP 1988 31-J3 16-33%
CSF 1989 J3 38%
- 32. 1986 32-N10 10-20%
LCP 1988 32-N10 0%
LCP 1988 32-N10 0%
i 1989 N10 0%
- 33. 1986 33-F13 10-20%
LCP 1988 33-F13-0%
LCP i
1988 33-F13 16-28%
TP 1989 F13 24%
l
- 34. 1986 34-D12 25%
LCP l
1988 34-D12 9-38%
LCP 1988 34-D12 18-38%
CSF 1989 D12 24%
1989 D12 33%
- 35. 1986 35-N5 25%
LCP 1988 35-N5 19-31%
LCP 1988 35-N5 18-30%
TP l
1989 N5
'33%
l 36, 1986 36-B8 0%
1988 36-B8 0%
1989 B8 0%
- 37. 1986 37-B7 10-20%
LCP i'
1986 37-B7 0%
CSF 1988 17-B7 19-25%
LCP l
1988 37-B7 16-18' CSF 1989 B7 _
24%
- 38. 1986 38-G14 10%
LCP 1988 38-G14 17-26%
LCP 1989 G14 38
- 39. 1986 39-F2 10-20%
LCP 1988 39-F2 16-19%
LCP 1989 F2 22%:
- 40. 1986 40-B10 25-27%
LCP
-1988 40-B10 20-38%
LCP 1988 40-B10 16-24%
TP 1989 B10 2B%
- 41. 1986 41-N12 0%
1988 41-N12 0%
LCP 1988
'41-N12 15-18%
TP 1989 N12 18%
Paga 4 42.-1986 42-M3 40%
LCP 1988 42-H3 26-43%
LCP 1988 42-M3 19-36%
TP 1988 42-H3 22-36%
CSF (top)
-1988 42-H3 22-33%
V/D 1/2".
1989 M3 40%
1989 M3 46%
V/D an additional 1".
- 43. 1986 43-D3 0%
1988 43-D3 0%
1989 D3 0%
- 44. 19C6 44-C12 25%
LCP-1988 44-C12 17-39%
LCP 1988 44-C12-19-35%
CSF 1989 C12 20%
1989 C12 4$4 V/D 1"
- 45. 1986 45-L14 0%
1986 45-L14 0%
I 1988 45-L14 0%
1988 45-L14 0%
LCP 1988 45-L14 28-35%
CSF 1989 L14-67%
V/D 1", capped & iso.
v1v. shut.
- 46. 1986 46-B5 30%
LCP
{
1988 46-B5--
14-37%-
LCP 1988 46-B5 15-17%
CSF 1989 B5 31%
- 47. 1986 47-R8 0%
1986 47-R8 0%
92' 1988 47-R8 0%
1988 47-R8 22-37%
I 1988 47-R8 34-37%
CSF-(BASE 1 1989 R8 52%
7r tj g 1989 R8 63%
V/DJL 9",
capped &
-iso. viv. shut.
48.~1986 48-H1 40%
LCP 1988 48-H1 30-38%
LCP 1988 48-H1 16-18%
V/D 1/2".
1989 El 26%
1989 B1 51%
V/D an additional 1".
- 49. 1986 49-J15 0%
1988 49-J15 0%
1988-49-J15 0%
60'-65' 1989 J15 0%
- 50. 1986 50-A9 0%
1988 50-A9 17-23%
LCP 1988 50-A9 20-36%
CSF 1989
-A9 20%
1989 A9 40%
6 PULLED BACK 4 CAPPED NOTE: *= A THIMBLE TUBE 1 BLOCKED (G-9 blocked & capped)
VITHDRAVN OR VORKED DURING THAT OUTAGE i
=-
s e
Page 5 Unit 2 Thimble Tube ECT Results During The 5th & 6th Refueling Outage PERCENT TUBE #
'VALL LOSS LOCATION.
COMMENTS
- 1. 1987 01-J7 0%
GOOD LCP 1989 01-J7 NDD
- 2. 1987 02 c7 0%
SLIGHT DIST @ LCP 1989 02-07.
NDD
- 3. 1987 03-G9-0%
SLIGHT DIST @ LCP=-
1989 03-G9 NDD 4, 1987 04-H6 0%
GOOD LCP 1989 04-H6 25%
- 5. 1987 05-F8 0%
SLIGHT DIST @ LCP 1989 05-F8 NDD
- 6. 1987 06-J10 0%
-SLIGHT DIST @ LCP 1989 06-J10 NDD
- 7. 1987 07-F9 0%
GOOD LCP 1989 07-F9 ~
NDD
~'
- 8. 1987 08-F6 0%
GOOD LCP-1987 08-F6 0%
45' DEPOSIT 1989 08-F6 NDD 1
- 9. 1987 09-H11 0%
SLIGHT DIST @ LCP 1987 09-H11 0%
CSF DEPOSIT l
1969 09-H11 NDD j
- 10. 1987 10-L8 0%
GOOD LCP El 1989 10-L8 27%
- 11. 1987 11-L9 22-48%
90' V/D 411.m 1 ys 1
2 3
1989 11-L9 NDD-
- 12. 1987 12-J5 0%
GOOD LCP l
1989 12-J5 NDD j
- 13. 1987
- 13-L6 0%
GOOD LCP-1 1989 13-L6 NDD t
14, 1987 14-F11 0%
- SLIGHT'DIST @ LCP 1989 14-F11 NDD
- 15. 1987 15-H4 0%
. GOOD LCP 1
1989 15-H4 NDD
- 16. 1987 16-J12 0%
GOOD LCP 1987 16-J12 0%
TP-RC DEPOSITS 1989 16-J12 NDD j
- 17. 1987 17-D7 0%
SLIGHT DIST @ LCP
.{
1987 17-D7 0%
TP DEPOSITS a
1989 17-D7 NDD i
- 18. 1987 18-L11 26-45%
105' ID DEFECT i
1987 18-L11 0%
GOOD LCP-1989 18-L11 NDD
- 19. 1987 19-L5 0%
GOOD LCP l
1987 19-L5 0%
DEPOSITS ALONG COND 1989 19-L5 NDD i
20, 1987 20-E5 0%
GOOD LCP.
l 1989 20-E5 NDD j
Pag 2 6-l
- 21. 1987 21-E11 0%
GOOD LCP 1989 21-E11 NDD
- 22. 1987 22-F4 0%
GOOD LCP 1989 22-F4' NDD
- 23. 1987 23-D10 0%
GOOD LCP 1989 23-D10 NDD
- 24. 1987 24-H13' 0%
GOOD LCP 1989 24-H13 NDD
- 25. 1987 25-N8 27-44%
84' VEAR @ TP l
1987 25-N8 0%-
GOOD LCP u
1989 25-N8 53%
V/D 2. 2" -- ;. ;;
I,jI
- 26. 1987 26-L4 0%
GOOD LCP 1989 26-L4 NDD
- 27. 1987 27-H3 0%-
GOOD LCP-1989 27-H3 NDD
- 28. 1987 28-D5 0%
GOOD LCP 1989 28-D5 NDD
- 29. 1987 29-C8 0%
GOOD LCP 1989 29-C8 NDD
- 30. 1987 30-N7 0%
GOOD LCP 1989 30-N7 NDD 4
- 31. 1987 31-J3 0%
GOOD LCP i
1987 31-J3 0%
SLIGHT DIST @ CSF l
1989 31-J 3 --
NDD
- 32. 1987 32-N10 0%
GOOD.LCP l
1989 32-N10 NDD
- 33. 1987 33-F13 0%
GOOD LCP 1989-33-F13 NDD l
- 34. 1987 34-D12 0%
GOOD LCP i
1989 34-D12 NDD
- 35. 1987 35-N5-0%
GOOD LCP i
1989 35-N5 NDD
- 36. 1987 36-B8 0%
GOOD LCP l
1989 36-B8 16%
- 37. 1987 37-B7 0%
GOOD LCP j
1989 37-B7 13%
- 38. 1987 38-G14 0%
GOOD LCP i
t 1989 38-G14 NDD
- 39. 1987 39-F2 22-43%
87' VEAR 9 LCP-i 1989 39-F2 29%
1
- 40. 1987 40-B10 22-34%
92' VEAR @ CSF i
1987 40-B10 0%
GOOD LCP 1989 40-B10 28%
- 41. 1987 41-N12 20-40%
94' VEAR 0 TP l
1987 41-N12 0%
GOOD LCP 1987 41-N12 0%
108' DEPOSIT 1989 41-N12 26%
- 42. 1987 42-H3 0%
GOOD LCP 1989 42-H3 NDD
- 43. 1987 43-D3 0%
GOOD LCP i
1989-43-D3 NDD
- 44. 1987 44-C12 0%
GOOD LCP 1989 44-C12 NDD i
~I i
I Page 7 h
45.-1987 45-L14 0%
GOOD LCP 1981 45-L14 0%
DEPOSIT @ TOP DP 19B9 45-L14 NDD
- 46. 1987 46-B5 20-26%
85' VEAR @ CSP 1987 46-B5 0%
GOOD LCP u
1989 46-B5 53%
V/D 2.:"
- 0.27 Ef
- 47. 1987' 47-R8 0%
GOOD LCP W
1989 47-R8 14%
cw +. ( F 9)jg ". f fk/m h k e+
- 48. 1987 48-H1 0%
GOOD LCP
' k 5 1.
F.' v =j 1989 48-H1 NDD VJ
- 49. 1987 49-J15 0%
GOOD LCP 1989 49-J15 NDD
- 50. 1987 50-A9 0%
DIST @ LCP 1989 50-A9 NDD k#PULLEDBACK T CAPPED T aLOCKED e
?
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!'i, STATUS OF FARLEY NUCLEAR PLAhT.
INCORE THD:BLES U5IT yL,,,,
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i STATUS OT T..fa.ET ;rCLEAR PL.CT INC;EE 3.:.s tES C;17 'Two,
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ATTACHMENT 6 November 29, 1989 Submittal i
ease 1
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4 en
. AlabaillaPOWtr
~v November 29, 1989 a
Mr. Giuliano DeGrassi Structural Analysis Division Department of Nuclear Energy, Building 129 Brookhaven National Laboratory Upton, New York 1197*
J. H. Farley Nuclear Plant - Units 1 & 2 NRC Audit of bulletin 88-09 Issues
Dear Mr. DeGrassi:
Enclosed are copies of the following documents to resolve all open items from your recent audit of Bulletin 88-09 issues:-
(1) Vestinghouse letter report HED-RPV-2574,-
Evaluation of the Echoram Vear Scar-(2) Vestinghouse letter report ALA-87-608:
Unit 1 Flux Happing System Seismic Analysis-(3) Vestinghouse letter report ALA-86-741:
Unit 2 Flux Happing System Seismic Analysis (4) Sketch shoving Unit 1 BMI instrument column thimble tube dimensions (5) Sketch shoving Unit 2 BMI instrument column thimble tube dimensions (6) 26350 (4 sheets) EANCO'Inc. drawings:
Control System - Flux Happing (7) 26353 (1 sheet) EANCO Inc. drawing:
Carriage Assembly If there are any questions please advise.
Yours truly, X ca. e. X A. E. Hammett Nuclear Haintenance Support AEH Attachments Distribution:
Mr. Giuliano DeGrassi - v/1 Mr. A. E. Hammett
- v/1 File: c-56
- v/0 t
i 4
MED RPV 2574 nn MECHANICAL EQUIPMENT DESIGN
- wu 236-6366 c3.,
November 27, 1989 g,3 Evaluation of Echoram Wear Scar 3
J. A. Knochel/EC-West 232A L. F. Dougherty/EC-West 232 cc:
C. H. Boyd/STC 701-306 D. E. Boyle/STC 701-306 C. M. Scrabis/STC 701-303 D. Merkovsky/STC 701-303
REFERENCE:
Calc. Note RPVSA 89-1505 We have completed an evaluation of the Echoram wear scar identified as the " check-mark" configuration.
Please transmit the following information to the ALA customer.
As a follow-up to conversations between Primary Component Engineering personnel and the ALA customer we have completed, and are hereby documenting, the evaluation of the Echoram wear scar with the
" check-mark" configuration. The results for the " check marked" configuration indicate a_ maximum stress intensity of 22,109 psi which is less than the previously report value of 23,318 psi.
This analysis was based on wear scars idegtified in.Echoram drawing-WS-A-007-89, and can be described as a 90 wear scar, with.65% wall loss, and a length of 0.97 inches.
All other: features of the previous report remain'as presented.
The only purpose of this report is for a comparison to the previous analysis.
Attached with this letter are color plots of.the stress' intensity-contours from this evaluation.
U1 A. J. Kuenzel Reactor Pressure Vessel System Analysis
/kls Attachment w
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_ CENTPM ENG' NEE:.:NG FILE ALA-87-608
_ A L A - l J J - F u 3
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ENGINEER:6 /> Sh/~.
April 21,1987 -
1 Mr. W. G. Hairston, lil, General Manager Nuclear Support l
Alaba9a Power Company 600 North Eighteenth' Street Birmingham, AL 35291-0400 Attn:
J. A. Ripple
~
Joseph M. Farley Nuclear Plant Unit No. 1 FMS ORIVE SYSTEM SElSMIC ANALYSIS
Dear Mr. Hairston:
Attached for your information is the subject report. As discussed in-the
' report, the four bolts in the Bechtel restraints which are used to hold the FMS cart in position over the seal table should be replaced at the next available opportunity, while the installed bolts will not yiel'd or fail when subjected to the Farley seismic levels, the analysis showed they could be.
stressed beyond AISC allowable levels.
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if you have any questions, please contact this office.
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Us, ',+a i~ z Very truly yours, F<C 7"ec -
-WESTINGHOUSE ELECTRIC CORPORATION g g g,,~. v-r..
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a./r/rs-C. Eicheidinger, Manager Ap JAK/fIc Attachment
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ALA.87-608 Ref:
C.0, BH-65051 FAR 90067 J. A. Ripple April 21, 1987 cc: R. P. M: Donald 1L
- n. G. Hairston ill 1L, 1A J. D. Aoodard 1L,1 A K. C. Gandhi 1L, 1A L. B. Lorg 1L J. R. Crane 1L R. H. - Baul ig 1L,1 A R. W. Wise 1L J. A. Ripple.1L,1 A
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Attict.rer.t tc SE6C-ECT-127 Pale 1 of C SE:SM:C EVALUA!!S CF TFE FAF.:.EY, UM7 1, FLUX P.APF:N3 SYS3M l
Nw-4 Prepared by :
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A.J. Harte. ann Equi;Nr.t Cualificatien Techno1cgy Reviewed by: C 4
J.E. Drexler Eculpment Qi lification Technology ~.
) $ 3[
I Approved by:
b L.f. Walker, ManaE,er Equipment Qualification Technology i
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Attachrer t te SE K-EOT-;E?
Page 2 cf &
SE S*:C EVAL'JAT:C.'. CF TEE FAF. LEY, US 1,
FLU.\\ PAFF:N3 SYSTEM 1.0 Irt rocu:t i:n A seisnie analysis ef the Far:ey Unit 1 Flux Mar;ir.g Systet (FP.S ) har teen perferred to address tr e ccncern that it may interact with anc fecparci:e the seal table pressure teundary during a seistic ever,t.
The ana;ysis was_ performed in accordance~ w;th the recor.iended practi:es cf IEEE 3;L-1975 (Eeference :) to deterr:ne :f the FMS would maints;r its stru:tural it.tegrity durir.E a seistic event, i
2.0 Ecufr-ert Insivred J
The equipment ar.aly ed is the reveatle FMS cerined fr Eance :nc.
i drawing nutters 263E0 Rev. C (Unit 1) and 26975 Fev. A (Unit 2). - A i
walkdcwr. cf the equiptent installed in Unit I was performed on October, 19E6 to cor.firr. the drawings. ar.d gather the necessary deta:Is to perform the ar.alysis. The FMS is mounted to rails and is located abcVe the seal tatie durinE operatier. at televation 129'-0" in the contair.rer.t l
buildir.g. The FMS is composed of structural steel which suppertt.'Cr ive i
i assemblies, transfer devices, ar.d thittle tubing.
During the walkdown, the FMS equipment was inspected to deterr.ine what pertions of the FMS could interact with the seal table and therefore require analysis. It was determined that only the moveable cart required analysis. The five (5) path transfer devices and their supporting structures are locatec off the moveable cart, at least 4 feet from the seal table. It was judgec that these components could i
not interact with the seal table since they are located so far froc. the seal table. Furthermore, steel grating, as well as the plate on top of i
the F13 moveable cart, is located between these components. and the seal j
table, and it is highly unlikely that these components: or their I
supporting structures would fail durir.g a seismic event. Therefore,-
only the moveable cart was judged to be able to interact with the seal 1
table and was the only piece of FMS equipment analyzed.
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n Attachter t to SE60-E07-127 Page 3 of 6-3.C if 3: u. s -
Finite e:erer t ::rputer r. ce; were deve:c;ed usits the WECA!: corruter ccce. Thret ci.er.siena; team, cars, en: the:1 e:erents were used tt t:ce; the F"S.
Fo:e; toundary.ccr,c:t:cns were chcsen to result :r the i
rest rea;is:ic yet cctservat:ve rese:ts for each type of analysis.
Sc;te: c Tre: :cn peir.t within.the tc4e; were constrained cr2y in the ap;re;r: ate trar.s:sti:nal and/or rctaticna; directions for the type cf cent.e::::n. Tr.ese belted ccnnecticn were eva2uated with hand ca:cu:et icr.: us:rg the contined ;eads derived. from the cetputer ana;ys:s.
L' sing the avai;sb;e Farley Safe Shutdown Earthquake (SSE) Requirec I
Fespcnse S;e:trut (RES), si darping EES curves were developed for the centair.rer.: building at elevatten 129'. A moda15 ERS analysis using these RES Vis ;erferred to' derive Icads and stresses in each principal.
axis of the FMS. The results fret ea:h rcde were combined by the-
~
square-roc -ef-the-sun-cf-the-squares (SRSS) method,'except for c1cse;y space: teces (within 10% of each other) which'were combined-l absc;ute;y. Static analyses were also performed Eto derive stresses and i
loads due :: structure deadweight and Zero Period Acceleratior, (EFA)
[
1evels 'in esch cf the equipterrt principal axes as defined by the Farley
[:
RRS. The IPA ar.alyses were perforred to include the. effects ef hitter l
frequency c: des in the analysis.
The RRS analysis results were i
ccmbined atselutely with the ZPA results for each principal directicn.
The two hori:enta) principal direction ERS plus ZPA results and the vertical directicn ERS plus ZFA results were combined by the SRSS method. Finally, an absolute sum method was used to combined these results with the deadweight results The final-.results were considered in determining the acceptability of the structure when subjected to seismic loading. These resulting member stresses were evaluated for acceptability based on American Institute of Steel Construction (AISC) specifications as defined jn Reference 2.
4.0 Fesults Results of the stress evaluation revealed that the only' elements which were stressed beyond AISC allowables (assuming A-307 bolts were used) were the two bolts connecting the FP.S C-Channel restraint to the Bechtel designed restraints.
However if. the present bolts are replaced by 1/2 in diameter A325 (or equivalent) bolts, the connection will be l
acceptable. All other eierents were found to be. stressed below AISC allowable levels, including connections which were evaluated with hand calculations using loads derived from the corputer analysis.
l l
t Atta;r.rer.t tc SE6;-ECT-;27 Page 4 cf 6
?
~he.4:r.it; ces:Ered restra:nts were ine;uced ir our ana:ysis. Tat 2e 2.
prev:tes the ::a:s.a: tre :r.terface tetween the :-).2x;/t ar.g;e and the Ex3 tute stee; fer tr.e "C-:har.r.e;" restra:rt.
Tat;e 2 fr0v::es the
! cats at the.irit* Tate betweer the ext >1/k tute stet: an: the Ex3 tule stee:. f tr tr.e " tut e stee." res tra:nt.
5.0 Cone:ur!cro F-: Fe:c-e dF'fers A seisric reduired res;;r.se spe:tr.:r analysis was Ferfor ed en the
.Farley FMS usir.g the i:ECAt:-firite elerent computer code. A13 seist fc r.erter stresses were fcund to te within AISC allewab3e 2 eve;s, except fcr the t::ts used in cre of the Sechte! des 15ned restraints. The bolts present;y used te attact the C-channel to the Bechte: designed i
restraint cust te upgrade: to ASTM-A-325 boits' (cr equivalent).
As deter-ined in a prev:cus ana;ys:s (Reference 3), the bo;ts' are not stressed beyonc their yie;d stress litits. Therefore, the restraint will not yts.;d or fail when ex;csed to the Farley seistic levels.
However the existing 1/2 inch disreter ASTM-A307 bolts should be replaced with 1/2 inch dia eter ASW-A325 (SAE Grade 5) belts at the next avai;st;e c;;cr:Un:ty.
This wil; ensure that the desi5n bolt stresses w;;l remain within allowable. limits.
1 6.C Feferences I-
.1.
"IEEE Reconnended Fractices for Seistric Qualification of Class IE Equipment for Nuclear Fower Generating Stations", IEEE 344-1975, The Institute of Electrical and Electronics Engineers,'Inc., New York,- New York, January -33,1975.
2.
Specificatien for the Design, Fabrication and Erecticn of Strvetural Steel for Buildings, Effective November 1978, AISC.
3 E Project Letter ALA-86-741, dated September 2,1986, subject: "FMS Drive Syste Seismic Analysis".
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b0nd! hTe ITCVIOed Et the IT,*.erfhte LCt'eer, a 2x;/t ar.E e and the Ex tute stee; the x
i 1
1 EX : ICTCe ir. A direttiCT. : E2 1DE.
Fy : Fcree :n y directi:n : 326~;ts.
Pz : Force ir. z directicn : 630 lte.
Mx : Morent about-x direction : 6968 in.- bs.
My = Mor.ent about y directicn = 325 in.-1bs.
Mz : Morent about z direction : 171 in.-;bs.
Where: - y directicn is paralle]' to rails- (horizontal).
y direction is perpendicular to rails (horizontal).
1 z directicn is vertical.
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Attxt.rer.t to SELC-E07-1E?
Page 6 of 6 TAELE 2 "TCEE STEEL" F.LS7F.Ai!.T LCIIS L:s2 are previced at tre f r.te,Tsce t.et' ter.
tr.e kxtx1/4 ant h ar.d tt.e Ex3 tute stet.
Fx : Fcree it. x directien : 592 its.
Py : Fcree in y direction : 391 lbs.
F: : Fcree ir. : directicn : 610 lbs.'
Mx : Merent abcut x direction : 10666 in.-Ibs.
I i
My : Merer.l. about y direction : 7960 in.-Ibs.
M: : Mcrer.t about : direction : 1228 in.-lbs.
k*here:
> direct:c.n is paralle] to rails (hort:enta]).
y direction is perpendicular to rails (horizontal).
z direction is vertical.
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- Westinghouse Power Systems
- D *5"5 fleCific C0fpDf8110n 5"" D**
- sci 355 Pmswp Pennsevama it233 C2tt
-i ALA-46-741 Ref: G.O. BH-44284' FAR 91366 September 2,1986 Mr.
W.' G. Hairston, til, General Manager Nuclear Support' Alabama Power Company 600 North Eighteenth Street Birmingham, AL 35291 Attn:-J. A. Ripple Joseph M. Farley Nuclear Plant' Unit No. 2 FMS DRIVE SYSTEM SEISMIC ANALYSIS
Dear Mr. Hairston:
Attached for your information is the subject report.- As discussed in the report,.the four bolts in the Bechtel restraints which are used.to hold tho' FMS cart in position over the seal table should be replaced at the next i
l available opportunity. While the bolts currently will not yield or fall when subjected to the Farley seismic levels, the analysis'showed.they could be stressed beyond AISC allowable-levels.
A Safety Evaluation Checklist for the report is also attached..
Very truly yours.
WESTINGHOUSE ELECTRIC CORPORATION p C. ElcheIdinger, Manager-
. Alabama' Project JAK/pmh Attachment l
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cc: R. P. Mcdonald 1L, 1A W. G. Nairston lil 1L,1A J. D. Woodard 1L, 1A K. C. Gandhi 1L, 1A L. 8. Long 1L, 1A J. R. Crane 1L, 1A i
R. H. Baulig it,1 A R. W. Wise IL, 1A J. A. Ripple 1L,1 A l
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EQ&T-EQT-3864 Equipment Qualifiestion Technology n
236-6287
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gn, August 21, 1966 we Tarley Flum Mapping Systes Seistic Evaluation Summary Report l
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t to J. A. Knochel - MNC 237 e
cet J.J. McInerney - MNC 409 W.T. Cuerin - MNC 409 J.M. Ludwierak - R&D 701/307
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" ?.1.
iMC 157 "
K.G. Luna - IMC 264 W.W. Wassel - 177C 264
'A.J. Bartsann - R&D 701/307 Filet ALA-144-TMS/2 i
In res case to the customer's request, attached is a samary report for i
the seismic evaluation of the Tarley Flus Mapping System (TMS).
i Using a conservative load combination procedure, the analysis revealed that the only members which were stressed beyond allowable limits when subjected to Tarley seismic levels were the bolts used in both techtel designed restraints.
Further detailed analysis however, revealed that the bolts were not stressed beyond their yield stress limit.
Therefore, although these bolts are stressed beyond allowable levels, they will not yield or fail when subjected to seismic loads.
j these 1/2 inch diameter ASTM-A307 bolta should be replaced with 1/2Revever, inch diameter ASTM-A325 (sAE Crede 5) bolts or equivalent at the nest available opportunity.
t The detailed analysis package for this effort will be maintained at E GTSD central file under central file number A1A-144-TMS and is available for audit. Should further details on the analysis be required, or if there are any questions, please,j contact the undersigned.
1.\\'s, I
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Ab P.T. $sith Equipment Qualification Technology L 1. Walker, Manager Equipment Qualification Technology
(
33 attachment i
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s At t ac hment t o EQ&T-EQT-3864 Page 1 of 6
$EISMIC ETALUATICK OF TEE FARI.EY FLUK MAPPING SYSTEM 4
~
Prepared by -
N P.T. $sith Equipment Qualification Technology h./.
Reviewed by:
A.J. Wartmann Equipment Qualification Technology
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l Approved by:
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'IJ L.f. Walker,' Manager Equipment Qualification Technology I
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EQ67-EQT-3864 Page 2 of 6 i
SEISMIC ETAIUATION OF TEE yARLET FtrI MAPPIWG SYSTEM 1.0 Introduction A seienic analysis of the Farley Flun Mapping System (FMS) has been performed to address the concern that it may interact with and jeopardise the seal table pressure boundary during a seienic event.
he analysis was performed in accordance with the recommended practices of IEEE 344-1975 (Reference 1) to determine if.the FMS would maintain its structural integrity during a solumic event.
i s
2.0 Enulement Analvred The equipment analysed is the moveable FMS defined in Eanco Inc.
drawing-sunbers 26350 Rev. C (Unit 1) and 26975 Rev. A (Ut.it 2).
~
A valkdown of the equipment installed in Unit 2 was performed on April 8, 1986 to confirm the drawings and gather the necessary details to i
The FMS is mounted to rails and is located above perform the analysis.
the oest table during operation at elevation 129'-0* in the containment building.
The FMS is composed of structural steel' which supporte drive assemblies, transfer devices, and thimble tubing.
During the walkdown, the FMS equipment was inspected to deternise what i
portions of the FMS could interact with the seal table and therefore require analysis.
It was determined that only the moveable eart-required analysis. The five (5) five path transfer devices and their supporting structures are located off of the moveable tatt, at least 4 feet from the seal table.
It was judged that these components could not interact with the seal table since they are located so far from the seal table.
Furthersore, steel grating, as well as the plate on top of the FMS moveable cart, is located between these components and the seal table, and it is highly unlikely that these components or their supporting structures would fait during a seismic event. Therefore, only the moveable cart was judged to be able to interact with the seal table and was the only piece of FMS equipment analyaed.
r i
1 3.0 Ansivste l
Finite element computer models were developed using the WECAN computer i
code.
Three dimensional been, mass, and shell elements were. used to medel the FMS.
Model boundary conditions were chosen to result in the most realistic yet conservative results for each type of analysis.
Bolted connection points within the model were constrained only in the appropriate translational and/or rotational directions for the type of
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At t ac haset t o BQET-EQT-3664
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Fate 3 of 6 l
connection. These belted connections were evaluated with hand l
calculations using the combined leads derived from the computer analysis.
i tising the available Farley Safe $batdown Earthquake ($$E) Requir ed Response Spectrus (115), 41 damping RAS curves were developed for the containment building et elevation 129'. A model 118 analysis using i
i this Ris was performed to derive loads and stresses in each principal amis of the FMS.
The results from each mode were combined by the square-root-of-the-sum-of-the-squares ($RSS) method, sacept for closely spaced modes (within 10% of each other) which were combined i
absolutely. Static analyses were also performed to derive stresses and loads due to structure deadweight and Zero Period Acceleration (IPA)
+
levels in each of the equipment principal ames as defined by the Farley E15 The IFA analyses were performed to include the effects of higher frequency modes in the analysis.
The RAS analysis results were combined absolutely with the ZPA results for each principal direction. ~
i The two horizontal principal direction RAS plus EPA results and the vertical direction RRS p1ve IPA results were combined by the SRSS method.
Finally, an absolute sua method was used to combine these results with the deadweight results.
The fical results were considered i
in determining the acceptability of the structure when subjected to seismic loading.
These resulting member stresses were evaluated for acceptability based on American Institute of steel Construction (AISC) specifications as defined in Reference 2.
i A.0 Results Results of the stress evaluation revealed that the only elements which vere stressed beyond the AISC allevable levels were the bolts used in both Bechtel designed restraints. Further detailed analysis was performed with a less conservative load combination method which combined each 118 result with each EPA result by $185 rather than by absolate sum.
This analysis of the 1/2" dissieter ASTM-A307 bolts revealed that although they were stressed beyond the AISC shear allevable levels, they were not stressed beyond the material shear yield strength and therefore would not yield or fait during a seismic i
All other elements were found to be stressed below AISC event.
allevable lesels including connections which were evaluated with hand calculations using loads derived from the computer analysis.
The Bechtel designed restraints shown in Bechtet drawing D-206116 were included in the analysis. Table 1 provides the loads at the interface between the 4 a 2 a 1/4 angle and the 8 a 3 tube steel for the r
"C-channel" res traint. Table-2 provides the loads at the interface between the 4 x 4 a 1/4 tube steel and the 8 a 3 tube steel for the "tabe steel" restraint.
o f
O At t ac hsa tt t o EQ67-EQT-3E64 Page 4 ef 6 5.0 Cene1usiens and teeparendations i seismic required response spectrum analysis was performed on the Farely FMS using the b'ECAN finite element cesputer code.
All seismic araber stresses were found to be within AISC allevable levels, encept for the botto used in both Bechtel designed restraints.
Justification for interia operation is based on a further detailed analysis which showed that the restraint bolts were not stressed beyond their yield stress limits.
Therefore, the restraints will not yield or fail when asposed to the Farley seisait levels.
Bowever, the saisting 1/2 itch disseter A87H-A307 bolts should be repisced with 1/2 inch diameter ASTM-A325 (SAE Crade 5) or equivalent bolts at the seat available opportunity.
This will ensure that the bolt stresses will fall-within allpesble limits.
6.0 lefere6ees
~
- !EEE Recommended Practices for Seistic Qualification of Class !!
1.
Equipment for Rutlear Power Generating stations " 1EEE-344-1975 The Institute of Electrical and Electronics Engineers, Inc., Few York, New York, January 31, 1975.
2.
Estus) of Steel Cetstruction, Eighth Edition, American Institute of Steel Construction, Chicago, IL,1980.
1 i
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At t ac hme nt t o l
BQ&T-EQT-3864 Page 5 of 6 t
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i TABLE 1 l
[
"C-CEANNEL" RESTEAINT LOADS I
Loads are provided at the interface between the 4 a 2 a 1/4 angle and the 8 a 3 tube steel j
l Pa = Terce in a direction = 76 lbs.
Py = Force in y direction = 321 lbs.
i Fa = Force in a direction = 616 lbs.
Ma = Moment-about a direction = 6913 in.-lbs.
i
~
My = Moment about y direction = 305 in.-lbs.
Ma = Moment about a direction = 87 in.-!bs.
I t
Vberet a direction is parallel to rails (borizontal).
y direction is perpendicular to rails (borizontal).
a direction is vertical.
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t EQ47-EQT-3 864 Page 6 of 6 1ABLE 2 "TME STEEL" REETRAINT 14 ADS Leads are provided at the interface between the 4 a 4 x 1/4 tube steel and the 8 a 3 tube steel Pa = Torce in a direction = $35 lbs.
Py = Force in y direction = 349 lbs.
Pa = Terce in a direction = 553 lbs.
Ma = Moment about a direction = 9595 in.-lbs.
My = Moment about y direction.= 7077 in.-lbs.
~
MW = Moment about a direction = 466 in.-lbs.
Where a direction is parallel to rails (borizontal).
y direction is perpendicular to rails (borizontal).
a direction is vertical.
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i WESTINGHOUSE l
NUCLEAR SAFETY EVALUATION CHECK LIST
}) NUCLEAR PLANT (5
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- 2) CHECK LIST APPLICABLE TO:
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(Swbject af Change)
- 3) The written safety evaluation of the revised procedure, design change. or l
modification required by 10CFR50.59 has been prepared to the entent reautres 5 and is attached.
If a saf ety evaluation is not required or is incomplete i
f or any roauton, explain on Page 2.
Parts A and 5 of this Saf ety Evaluation Check Liut are to be completed only on the basis of the safety evaluation performed.
i CHECK LIST - PART A
~
(3.1*
Yes,,,,,, No.[ A change to the plant as described in the F5 ART (3.20 Yes.. No A change to procedures as' described,in the FSaR7 (3.3)
Yes...No A test or' emperiment not described in the FEART l
(3.4)
Yes... No
,, A change to the plant technical specifications 3
(Appendia A to the Operating License)?
- 4) CHECK LIST - PART 3 (Justification f or Part B answers must be included on Page 2.)
(4.1)
Y e s,,,,,,, N o,,,,,,,,
Will the probability of an accleent previously evaluated in the FSAR be increased?
(4.2)
Yes,,,,,, No Will the consequences of an accident previously evaluated in the FSAM be increased?
(4.3)
Ye s,,,,,,, No d,,,
May t he p os si b i l i t y of. an ac c i dent eehi ch i s dif f erent ther) any already evaluated in the j
FSAR be created?
(4.4)
Yes... No y..
Will the probability of a malfunction of equipment important to safety previously
/
ovaluated in the FSAR be increased?
(4.5)
Yes..
No,,V,,
Will the consequences of a malfunction of equipment important to saf ety previously ovaluated in the FSAR be increased?
(4. 0 Ye s,,,,,,,, N o..,,,
May the possibility of a malfunction of eouipment important to saf ety dif f erent than any already evaluated in the FSAR be created?
44.7)
Yes,,,,, No,,,,, Will the margin of saf ety as defined in the bases j
to any technical specification be reduced?
t
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