ML13316B797
| ML13316B797 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/05/1980 |
| From: | Haynes J SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| IEB-79-02, IEB-79-2, NUDOCS 8010300682 | |
| Download: ML13316B797 (4) | |
Text
Southern California Edison Company P O0BOX 800 22
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WAN UT GROV5 AV'ENVUE N1 ROSEMEAD.
CAL FORNIA 91 77 0 September 5, 1980 Cp U. S. Nuclear Reculatorv Commission Office of Inspection and Enforcement Reoion V 1990 North California Boulevaro Suite 202, Walnut Creek Plaza Vialnut Creek, California 94596 Attention:
Mr. R. H. Encelken, Director DCCKET No. 50-206 SAN ONOFRE -
UNIT 1
Dear Sir:
IE BULLETIN No.
79-02 REVISION 2, PIPE SUPPORT BASE PLATE DESIGN USING CONCRETE EXPANSION ANCHOR BOLTS
Reference:
(1) SCE (A. Arenai) to NRC (R. H. Engelken), letter dateo Aucust 15, 1979. Subject.
Testing of Concrete Expansion nchors per IE Bulletin 79-02 Rev. i.
-)rE J. T. Heau, Jr.) to NRC (R. H. Engelken), letter dated acember 7, 1979.
Subject:
Response to IE Bulletin 79-02 (3)
SCE (Robert N. Coe) to NRC (R. H. Engelken) letter dated rFebruary 8, 1974.
Subject:
Regulatory Operations Information Request No. 74-1.
In accorcance with Reference 1 above, a continuation of our concrete expansion anchor testing and inspection program was performeo at San Onofre Unit I during the 1980 refueling outage. In addition, a walkdown of all safety-relatec piping greater than 2-1/2 inches in diameter was conducted to ensure that all pipe supports were included in our program. A total of 60 pipe supports involving 78 base plates located insiae containment and 37 additional pipe supports involving 37 base plates located outside containment were iuentifieo to be within the scope of the program.
s....so b.
U. S. Nuclear Regulatory Commission IE Bulletin 79-02,Rev. 2 Page #2 The prcram inside containment incluced the test of all anchor bolts in each base plate. This ensured that retesting in an area of limited accessi bility would not be required if a high failure'rate were observed. For supports located outsioe containment only one bolt per base plate was tested provicing that the test was successful. Among the items included in our program were torque testing expansion anchors with torque values corresponding to a pullout of at least one.fifth the bolt ultimate 'capacity, and inspecting for proper thread engagement, anchor expansion and imbedment depth. The details of the testing and inspection proceoure have been discussed and reviewed by your staff and portions of the test program observed by an NRC staff member. The results of the program are summarized below:
INSIDE OUTSIDE CONTAINMENT CONTAINMENT Total Base Plates 78 37 Total Anchors Tested 211 75 Successful Torque Test 205 73 Failea Anchors (rotated in hole) 6 2
Not Tested (Damaged during removal) 19 10 Not Testec (Bolt removal not possible) 23 0
For piping supports inside containment it should be notec that 5 of the 6 faileo anchors were located on one support involving a feedwater line.
It appears that the anchors on this support were installed properly but were subsequently subjectec to a force which caused the anchor sleeves to loosen in the concrete.
A water hammer -event which could have caused such a force has been previously reported to you in Reference 3. We are presently conducting a program to evaluate the adequacy of our feedwater support design in the event of a water hammer.
The one remaining failed anchor appeared to be a result of original installation.
For piping supports outside containment the 2 failed anchors were located on one support involving the feedwater pump recirculation line.
The failures appear to have occurrea during original installation as a result of limited accessibility due to surrounding installations.
The remaining two anchors on this support were tested successfully.
U. S. Nuclear Regulatory Commission IE Bulletin 79-02 Rev. 2 Page #3 During the insoection program insice containment 8 inoications of inadequate imbedment depth were observed where steel reinforcement bar imbeced in the concrete prevented proper anchor imbedment during installation. These anchors were subsequently replaced. No such indications were observed for supports located outside containment.
Durinc our inspections for proper anchor to-bolt thread engagement, there were 6 instances inside containment and 1-7 instances outside containment where threao enoacements were less than the minimum required by our inspection procedure.
The majority of the thread engagement deficiencies were of such a nature that they would not be a primary failure mechanism durino load ing conditions.
In.all cases where threac encacement deficiencies were observed the remaining bolts in the plate were inspectec, the bolts replaced as necessary to achieve the reauireo threac engagement, and the new bolt torque tested.
The anchors reported as not testec were camagec curing bolt removal or removal was not possible due to geometrical and/or safety considerations.
(e. g. pipe operability required).
in acdition to the above there were 3 supports involving the residual heat removal (RHR) system which could not be testec ue tc high raciation levels.
Since only 2 of the 60 total pipe supports insice containment were found in a failed condition, expansion of the sampling program to include these three RHR system supports was not consicereo warrantec.
In all cases each support inclucec in our program was repaired as necessary to ensure a safety factor of five for existing shell type expansion anchors anu a safety f'actor of four for instances where cefective or camaged anchors were replaced with wedge type anchor bolts.
Based on the results of, our recent testing prograni and the results previously reported in Reference 1, we consider that the operability of all safety-related piping is assured in the event of a cesicn basis earthquake.
in accorcance with item 4 of Reference 2 we are providing the following information concerning our investigation into the effects of preload on the ultimate capacity of anchor bolts under dynamic loading:
- 1.
Regarcirg shear strength, Report No.
CEB 75-32 by the Tennessee Valley Authority Division of Engineering Design Thermal Power Engineering states, "There was no difference in the ultimate strength between preloading bolts or tightening nuts finger tight, however, under service 10aG concitions the preloaded bolt connections were much stiffer."
This report states further, "Preloading of expansion anchors to any degree of
U. S. Nuclear Regulatory Commission IE Bulletin 79-02 Rev. 2 Page #4 certainty coes not appear to be practical because of the slip characteristics of these anchorages."
In adoition, preload is lost on exp&nsion anchors at a fast rate, further obscuring the degree of preload with the passage of time. Based upon this latter fact, it would be difficult to justify taking credit for preload in the support design.
- 2. Secuoyah Nuclear Plant performed tests on wedge bnchors and embedded anchors in order to determine the effects of preload. The test considered combined shear and'tension-loads. A summary of test results and the interpretation of the test data can be found in "Sequoyah Nuclear Plant, Information on Anchorage Analysis."
The report concludes that, "As seen by these tests,, installation torque has a significant impact on the stiffness characteristics of the anchorage. It has no effect, however, on ultimate capacities."
- 3. In.order to account for the loss of stiffness as a result of oecreasihg preload over time both of the aforementioned references suggest the use of larger safety factors. by utilizing the safety factors suggesteo in IE Bulletin 79-02 it is our feeling that this recommendation is fulfilled.
Should you have any further questions regarding this matter, please do not hesitate to contact me.
Sincerely, J. G. Haynes Manager of Nuclear Operations cc:
Director, Office of Inspection and Enforcement, Division ot Reactor Operations lnspection