ML19241B793
| ML19241B793 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/21/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Andognini G BOSTON EDISON CO. |
| References | |
| NUDOCS 7907230642 | |
| Download: ML19241B793 (1) | |
Text
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2 1 J uii idl 3 Docht No. 50-293 Boston Edison Company M/C Nuclear ATTN:
Mr. G. Carl Andognini, Manager Nuclear Operations Department 800 Boylston Street Boston, Massachusetts 02199 Gentlemen:
Enclosed is IE Bulletin No. 79-02 Revision No. 1, which requires acti'on by you with regard to your power reactor facility (ies) with an operating license or a construction permit.
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Sincerely,
/
'n Boyce H. Grier Director
Enclosures:
1.
IE Bulletin No. 79-02 (Revision No. 1) 2.
List of IE Bulletins Issued in Lart Twelve Months cc w/encls:
P. J. McGuire, Pilgrim Station Manager A. Z. Roisman, Natural Resources Defense Council o
i 007 230ff2 '
468 011
ENCLOSURE 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICS OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 1 of 6 PIPE SUP"0RT BASE PLATE DESIGNS USING CONCRETE EXPANSION ANCHOR BOLTS Description of Circumstances:
Since the issuance of IE Bulletin 79-02 on March 8,1979, IE inspection h
experience and many inquiries from licensees indicate that additional informa-tion and clarification is needed.
This revision is intended'to serve that purpose.
None of the requirements of the original Bulletin have been deleted, and the due date for completion of the requested actions (July 6,1979) has not been changed.
The follcwing text supersedes the text of Bulletin 79-02.
Changes from the original text are identified by lines in the margin.
The purpose of this revision is to identify acceptable ways of satisfying the Bulletin requirements.
While performing inservica inspections during a March-April 1978 refueling outage at Millstone Unit 1, structural failures of piping supports for safety equipment were observed by the licensee.
Subsequent licensee inspections of undamaged supports showed a large percentage of the concrete ancnor bolts were not tightened properly.
Deficiency reports, in accordance with 10 CFR 50.55(e), filed by ong 7.sland Lighting Company on Shoreham Unit 1, indicate that design of base plates using rigid plate assumptions has resulted in underestimation of loads on scme anchor bolts.
Initial investigation indicated that nearly fifty pc cent of the base plates could not be assumed to behave as rigid plates.
In addition, licensee in;pection of anchor bolt installations at Shoreham has shewn over fifty percent of the bolt installations to be deficient.
Vendor Inspection Audits by NRC at Architect Engineering firms have shown a wide range of design practices and installation procedures which have been employed for the use of c 1 crete expansion anchors.
The current trends in the industry are toward more rigorous controls and verification of the instal-lation of the bolts.
The data available on dynamic testing of the concrete expansion anchors show fatigue failures can occur at loads substantially below the bolt static capa-cities due to material imperfections or notch type stress risers.
The data also show low cycle dynamic failures at loads below the bolt static capacities due to joint slippage.
- Lines indicate changes to previous editicn 7006290232 468 012
IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 2 of 6 Action to be Taken by Licensees and Permit Holders:
This Bulletin addresses those pipe support base plates that use concrete i
expansion anchor bolts in Seismic Category I systems as d Ifined by Regulatory Guide 1.29, " Seismic Design Classification" Revision 1, dated August 1973 or as defined in the applicable FSAR.
For older plants where Seismic Category I requirements did not exist d the time of licensing it must be shown that piping supoorts for safety related systems, as defined in the Final Safety Analysis Reprt, meet design requirements.
The revision is not intended to penalize licensees who have already completed some of the Bulletin requirements.
In those instances in which a licensee has completed action on a specific item and the Bulletin revision provides more conservative guidance, the licensee should explain the adequacy of the action already performed.
It should be reiterated that the purpose of the Bulletin actions are to assure opera,ility of Soismic Category I piping systems in the event of a seismic event.
1.
Verify that pipe support base plate flexibility was accounted for in the calculation of anchor bolt loads.
In lieu of supporting analysis jus-tifying the assumption of rigidity, the base plates should be considered flexible if the unstiffened distance between the member welded to the plate and the edge of the base plate is greater than twice the thickness of the plate.
It is recognized that this criterion is conservative.
Less conservative acceptance criteria mus be justified and the ju,:tification submitted as part of the response to the Bulletin.
If the base plate is determined to be flexible, then recalculate the bolt loads using an appropriate analysis ypfgp piJJ/.acpppp,t fpr Jpg gffggy pf Qear 7/Kep,5fpp fp,teragtfop,/m/nfmpm/edgp pistance/apd/ proper po)t/spacipg/
If possible, this is to be done prior to testing of anchor bolts.
These calculated bolt loads are referred to hereafter as the bolt design loads.
A descrip-tion of the analytical model used to verify that pipe support base plate flexibility is accounted for in the calculation of anchor bolt loads is to be submitted with your response to the Bulletin.
It has been noted that the schedule for analytical work on base plate flexibility for some facilities extends beyond the Bulletin reporting time frame of July 6,1979.
For those facilities for which an anchor bolt testing program is required (i.e., sufficient QC documentation does not exist), the anchor bolt testing program should not be delayed.
2.
Verify that the concrete expansion anchor bolts have the following minimum factor of safety between the bolt design load and the bolt ultimate capa-city determined from static load tests (e.g. anchor bolt manufacturer's) which simulate the actual conditions of installation (i.e., type of con-crete and its strength properties):
468 013
IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 3 of G a.
Four - For wedge and sleeve type anchor bolts, b.
Five - For shell type ancher bolts.
The bolt ultimate capacity rhould account for the effects of shear-tension interaction, minimum edge distance and proper bolt spacing.
If the minimum factor of safety of four for wedge type anchor bolts and five for shell type anchors can not be shown then justification must be -
provided.
3.
Describe the design requiremelts if applicable for anchor bolts to with-stand cyclic loads (e.g. seismic loads and high cycle operating loads).
4.
Verify from existing QC documentation that design requirements have been met for each anchor bolt in the following areas:
(a)
Cyclic loads have been considered (e.g. anchor bolt preload is equal to or greater than bolt design load).
In the case of the shell type, assure that it is not in contact with the back of the support plate prior to preload testing.
(b) Specified design size and type is correctly installed (e.g. proper embedment depth).
If sufficient documentation does not exist, then initiate a testing program that will assure that minimum design requirements have been met with respect to sub-items (a) and (b) above.
A samoling technique is accept-able.
One acceptable technique is to randomly select and test one anchor bolt in each base plate (i.e. some supports may have more than one base plate).
The test should provide verification of sub-items (a) and (b) above.
If the test fails, all other bolts on that 1se plate should be similarly tested.
In any event, the' test program should assure that each Seismic Category 1 system will perform its intended function.
The preferred test method to demonstrate that bolt preload has been acccmplished is using a direct pull (tensile test) equal to or greater than design load.
Recognizing this method may be df#ficult due to acces-sibility in some areas an alternative test method such as torque testing may be used.
If torque testing is useo it must be shown and substantiated that a correlation between torque and tension exists.
If manufacturer's data for the specific bolt used is not available, or is not used, then site specific data must be developed by qualification tests.
Bolt test values of one-fourth (wedge type) or one-fifth (shell type) of bolt ultimate capacity may be used in lieu of individually calculated bolt design loads where the test value can be shown to be conservative.
468 014
IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 4 of 6 The purpose of Bulletin 79-02 and this revision is to assure the operability 'of each seismic Category I piping system.
In all cases an evaluation to confirm system operability must be performed.
If a base plate or anchor bolt failure rate is identified at one unit of a multi-unit site which threatens operability of safety related piping systems of that unit, continued operation of the remaining units at~that site must be immediately evaluated and reported to ite NRC.
The evaluation must consider the generic applicability of the
+,nti fied failures.
Ap7endix A describes two sampling methods fr..- testing that can be used.
Otaer sampling methods may be used but must be justified.
Those options may be selected on a system by system basis.
Justification for omitting certain bolts from sample testing which are in high radiation areas during an outage must be based on other testing or analysis which substantiates operability of the affected system.
Solts which are found during the testing program not to be preloaded to a load equal to or greater than bolt design load must be properly pre-loaded or it must be shown that the lack of preloading is not detrimental to cyclic loading capability.
If it can be established that a tension load on any of the bolts does not exist for all loading cases then no pre-load or testing of the bolts is required.
If anchor bolt testing is done prior to completion of the analytical work on base plate flexibility, the bolt testing must be performed to at last the original calculated bolt load.
For testing purposes factors may be used to conservatively estimate the potential increase in the calculated bolt load due to base plate flexibility.
After completion of the analytical work on the base plates the conservatism of these factors must be verified.
For base plate supports using expansion anchors, but raised from the supporting surface with grout placed under the base plate, for testing purposes it must be verified that leveling nuts were not used.
If leveling nuts wer: used, then they must be backed off such that they are not in con-tact with the base plate before applying tension or torque testing.
Bulletin No. 79-02 requires verification by inspection that bolts are properly installed and are of the specified size and type.
Parameters which should be included are embedment depth, thread engagement, plate bolt hole size, bolt spacing, edge distance to the side of a concrete member and full expansion of the shell for shell type anchor bolts.
If piping systems 2 1/2-inch in diameter or less were computer analyzed then they must be treated the same as the larger piping.
If a chart analysis method was used and t.is method can be shown to be highly con-servative, then the proper installation of the base plate and anchor bolts should be verified by a sampling inspection.
The parameters inspected should include those described in the preceding paragraph.
If small diameter piping is not inspected, then justification of system oper-ability must be provided.
468 015
IE Bulletin No. 79-0J (Revision No. 1)
Date:
June 21, 1979 Page 5 of 6 5.
All holders of operating licenses for power reactor facilities are requested to complete items 1 through 4 within 120 days of date of issuance of the Bulletin.
No extension of time to complete action requested in Bulletin 79-02 is granted by issuince of this revision of the Bulletin. (Due Dace - July 6, 1979) A reactor shutdown is not required to be initiated solely for purposes of chis inspection above.
However, it is expectad that testing of otherwise inaccessible supports will be performed during the earliest extended outage following Bulletin issuance.
It is also expected that testing of anchor bolts in acces-sible areas in operating plants will be performed within the reporting interval.
In the event the required testing is not completed at the time of the initial report, on or about July 6,1979, the licensee should justify system operability and therefore montinued plant operation based upon the results of testing completed.
Maintain documentation of any sampling inspection of anchor bolts required by item 4 on site and available for NRC inspection.
Report in writing within 120 days of date of Bulletin issuance, to the Director of the appro-priate NRC Regional Office, completion of your verification and describe any discrepancies in meeting items 1 through 4 and, if necessary, your plans and schedule for resolution.
For planned action, a final report is to be submitted upon completion of your action.
A copy of your report (s) should be sent to the United States Nuclear Regulatory Commission, Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555.
These reporting requirements do not preclude nor substitute for the applicable requirements to report as set forth in the regulations and license.
6.
All holders of construction permits for power reactor facilities are requested to complete items 1 though 4 for installed pipe sucport base plates with concrete anchor bolts within 120 days of date of issuance of the Bulletin.
No extension of time to complete action requested in Bulletin 79-02 is granted by issuance of this revision of the Bulletin.
For pipe support base plates which have not yet been installed, document your actions to assure that items 1 though 4 will be satisfied.
Maintain documentation of these actions on site and available for NRC inspection.
Report in writing within 120 days of date of Bulletin issuance, to the Director of the appropriate NRC Regional Office, completion of your review and describe any discrepancies in meeting items 1 though 4 and, if neces-sary, your plans and schedule for resolution.
A copy of your report should be sent to the United States Nuclear Regulatory Commission, Office of Inspection and Enforcement, Division of Reactor Construction Inspection, Washington, D.C.
20555.
Approved by GAO B180225 (R0072); clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
Attachment:
l 1.
Appendix A 468 016
IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 6 of 6 APPENDIX A SAMPLING METHODS Item 4 of this Bulletin states that for anchor bolt testing purposes a sampling program is acceptable.
Two samoli1g methods are discussed below, but other methods may be used if justified.
a.
Test one bolt on each plate as origirally recommended in Bulletin No. 79-02.
If the test fails, all other bolts on that base plate should be similarly tested.
A high failure rate should be the basis for-increased testing.
b.
'andomly select and test a statistical sample of the bolts to provide a 95 percent confidence level that less than 5 percent defective anchors are installed in any one seismic Category I sy'..em.
The sampling pro-gram should be done on a systen oy system basis.
468 017
IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21,1979 Page 1 of 4 ENCLOSURE 2 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued issued To No.
78-10 Bergen-Paterson 6/27/78 All BWR Power Reactor Hydraulic Shock Facilities with Suppressor Accumulator an OL or CP Spring Coils 78-11 Examination of Mark I 7/24/78 BWR Power Reactor Containment Torus Facilities with an OL Welds for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee.
All other BWF Power Reactor Facilitius with an OL for information 78-12 Atypica? deld Materiai 9/29/78 All Powe.- Reactoc in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-128 Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-13 Failures In Source Heads 10/27/78 All General and of Kay-Ray, Inc., Gauges Specific Licensees Models 7050, 70508, 7051, with the subject 70518, 7060, 70608, 7061 Kay-Ray, Inc.
and 70618 Gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR Faci-Components In ASCO lities with an OL Solenoids (for action), and all other Power Reactor Facilities with an OL or CP (for information) 468 018 IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 2 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued To No.
79-01 Environmental Qualif-2/8/79 All Power Reactor ication of Class IE Facilities with an OL, Equipment except the 11 Systematic Evaluation Program Plants (for action), and all other Power Reactor Facilities with an OL or CP (For Information)79-01A Environmental Qualification 6/6/79 All Power-Reactor of Class 1E Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) 79-02 Pipe Support Base Plate 3/8/79 All Power Reactor Design Using Concrete Facilities with an OL Expansion Anchor Bolts or CP 79-03 Longitudinal Weld Defects 3/12/79 All Power Reactor in ASME SA-31' T: ce Facilities with 304 Stainless Steel Pipe an OL or CP Spools Manufactured by Youngstown Welding and Engineering Company 79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-05 Nuclear Incident at 4/1/79 All Babcock and Three Mile Island Wilcox Power Reactor Facilities with an OL, Except Three Mile Island 1 and 2 (For Action),
and All Other Power Reactor Facilities With an OL or CP (For Information) 468 0i?
IT Bulletin No. 79-02 Date:
June i 1079 Page 3 of 4 LISTING 0F IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Q illetin Subject Date Issued Issued to No.79-05A Nuclear Incident at 4/5/79 Same as 79-05 Three Mile Island -
Supplement 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-
. Power Reactor Facil-alignments Identified ities with an OL Except During the Three Mile B&W Facilities (For Incident Action), All Other Power Reactor Facil-ities with an OL or CP (For Information)79-06A Same Title as 79-06 4/14/79 All Westinghouse Designed Pressurized Power Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-06A Same Title as 79-06 4/18/79 All Westinghouse (Revision 1)
Designed Pressurized Pcwer Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-068 Same Title as 79-06 4/14/79 All Combustion Engineering Designed Pressurized Power Reactor Facilities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information) 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 463 020 IE Bulletin No. 79-02 (Revision No. 1)
Date:
June 21, 1979 Page 4 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.
79-08 Events Relevant to 4/14/79 All BWR Power Boiling Water Fower Reactor Facilities Reactors Identified with an OL (For Ouring Three Mile Action), All Other Island Incident Power Reactor Facil-ities with an OL or CP (For Information) 79 Failures of GE Type 4/17/79 All Power Reactor AK-2 Type Circuit Facilities with an Breaker in Safety OL or CP Related Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety OL or CP Systems 79-12 Short Period Scrams at 5/31/79 All GE BWR Facilities BWR Facilities with an OL 463 02i