ML14170A401
| ML14170A401 | |
| Person / Time | |
|---|---|
| Site: | Robinson, Brunswick  |
| Issue date: | 08/20/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jackie Jones CAROLINA POWER & LIGHT CO. |
| References | |
| NUDOCS 7909070237 | |
| Download: ML14170A401 (8) | |
Text
0 j t
0U N I T E D S T A T E S M I S O NUCLEAR REGULATORY COMMISSION REGION 11 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 In Reply Refer To:1979 RII:JPOAU20l1 50-325 Carolina Power and Light Company ATTN:
Mr. J. A. Jones Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, NC 27602 Gentlemen:
Gentlmen:SupplmentNo 1, which clarifies Enclosed is IE Bulletin No. 79-02, Revision 1, Supeen o
1 h
c lifYies)
NRC positions on actions requested with regard to your power reactor facility(ies) with an operating license.
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Sincerely, James P.
O'Peil y Director
Enclosures:
- 1.
IE Bulletin No. 79-02, Revision No. 1 (Supplement No. 1)
- 2.
Listing of IE Bulletins Issued in Last Six Months 7 9 O.9O 702's7
Carolina Power and Light Company
-2 cc w/encl:
A. C. Tollison, Jr.
Plant Manager Box 458 Southport, North Carolina 28461 R. B. Starkey, Jr., Plant Manager Post Office Box 790 Hartsville, South Carolina 29550
SSINS:
6820 Accession No:
7908150164 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 August 20, 1979 IE Bulletin No. 79-02 Revision No. 1 (Supplement No. 1)
PIPE SUPPORT BASE PLATE DESIGNS USING CONCRETE EXPANSION ANCHOR BOLTS Description of Circumstances:
The supplement to IE Bulletin No. 79-02 is intended to establish criteria for the evaluation of interim acceptability of plant operation with less than the design factors of safety for piping supports due to as-built problems, under design, base plate flexibility, or anchor bolt deficiencies.
In the reviews for system operability of the Duane Arnold and Crystal River facilities, criteria have been developed by the NRC staff that defines pipe support operability. The criteria has been applied in lieu of other analysis or evalua tion. Specifically, the licensees identified problems with pipe supports in which the original design factors of safety were not met but some lesser margin was available.
The design margins of four or five are intended to be final design and installation objectives but systems may be classed as operable on an interim basis with some lesser margin providing a program of restoration to at least the Bulletin factors of safety has been developed. Facilities which fall outside the operability criteria are considered to probably require a Technical Specifica tion exception and will require review on a case by case basis.
Action to be Taken by Licensees:
For the following two cases, plant operation may continue or may begin:
- a. For the support as a unit, the factor of safety compared to ulti mate strengths is less than the original design but equal to or greater than two.
- b. For the anchor bolts the factor of safety is equal to or greater than two and for the support steel the original design factor of safety com pared to ultimate strengths is met.
The above criteria may be applied provided that the affected systems are up graded to design margins of safety expeditiously for normally accessibile supports and by the next refueling for nonaccessible supports. Accessibility is as de fined in Bulletin No. 79-14 where "normally accessible" refers to those areas of the plant which can be entered during reactor operation.
IE Bulletin No. 79-02, Revision No. 1 Page 2 of 2 (Supplement No. 1)
August 20, 1979
- 1.
Any support not satisfying the criteria should be classed as inoperable and the Technical Specification action statement met unless it can be shown that the system can function in a design basis seismic event without the support.
- 2.
Repairs to supports should result in return to the design factor of safety.
- 3.
Operations may be continued while repairs to upgrade the system from a factor of safety equal to or greater than two with respect to design loads are performed. Consideration must be given to the effect of the repair process on support function and system operability. In other words the time the support is not functional should be limited to T.S. action statement times or the support must be determined not to cause the system to be unable to perform its function in a seismic event. The licensee should also exercise care not to take several supports on a given system out of service at the same time or cause both trains of one safeguards system to be made inoper able at the same time. Control over workmen on safety related systems during plant operation requires a high degree of control by the licensee.
- 4.
There are no special reporting requirements for this supplement to the Bulletin; however, the reporting requirements as set forth in the regula tions and licenses must be met.
Enclosure IE Bulletin No. 79-02 Revision 1, Supplement No. 1 August 20, 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued To No.
79-18 Audibility Problems 8/7/79 All OL's for action Encountered on Evacua-All CP's for information tion of Personnel from High-Noise Areas 79-17 Pipe Cracks in Stagnant 7/26/79 opri licn Borated Water Systems at PWR Plants 79-16 Vital Area Access Controls 7/26/79 Alicants f Poe Reactor Operating Licenses who anticipate loading fuel prior to 1981 79-15 Deep Draft Pump 7/18/79 All Power Reactor (Supp. 1)
Deficiencies Licensees with a CP and/or OL 79-15 Deep Draft Pump 7/11/79 All Power Reactor Deficiencies Licensees with a CP and/or OL 79-14 Seismic Analyses for 7/27/79 Aclities wita (Correc-As-Built Safety-Related Ol or aC tion)
Piping System 79-14 Seismic Analyses for 7/18/79 All Power Reactor (Rev. 1)
As-Built Safety-Related Ol or aC Piping System 79-14 Seismic Analyses for 7/2/79 All Power Reactor As-Built Safety-Related Ol or aC Piping System 79-13 Cracking in Feedwater 6/25/79 All PWRs with an System Piping OL for action. All BWRs with a CP for information
Enclosure IE Bulletin No. 79-02 Page 2 of 4 Revision 1, Supplement No. 1 August 20, 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued issued To No.
79-12 Short Period Scrams at 5/31/79 All GE BWR Facilities BWR Facilities with an OL 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Fl or aC for Engineered Safety Systems 79-10 Requalification Training 5/11/79 All Power Reactor Progam SatisicsFacilities with an OL Program Statistics 79-09 Failures of GE Type AK-2 4/17/79 Aclites wita Circuit Breaker in Safety Ol or CP Related Systems 79-08 Events Relevant to BWR 4/14/79 AclitiesPwit antoL Reactors Identified During Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Fl or Cn 79-06C Nuclear Incident at Three 7/26/79 To all PWR Power Mile Island - Supplement wit a l Allh Cobsto Enine 79-06B Review of Operational 4/14/79 Errors and System Mis-ing Designed Pressurized Water Power Reactor alignments Identified During the Three Mile Island Incident 79-06A Review of Operational 4/18/79 Preactr Fates (Rev 1)
Errors and System Mis-o Wetou Deign alignments Identified During the Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water 79-0A Eror andSysem is-Power Reactor Facilities Errors and System Mis-o etnhueDsg alignments Identified During the Three Mile Island Incident
IE Bulletin No. 79-02 Page 3 of 4 Revision 1, Supplement No. 1 August 20, 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued To No.
79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with-an alignments Identified OL except B&W facilities During the Three Mile Island Incident 79-05C Nuclear Incident at Three 7/26/79 To all PWR Power Mile Island - Supplement Reactor Facilities with an OL 79-05B Nuclear Incident at 4/21/79 All B&W Power Reactor Three Mile island Facilities with an OL 79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island Facilities with an OL Nuclear Incident at 4/1/79 All Power Reactor 79-05A ula nieta Facilities with an O Three Mile Island OL and CP 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 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured by Youngstown Welding and Engineering Co.
79-02 Pipe Support Base Plate 6/21/79 All Power Reactor (Rev. 1)
Designs Using Concrete Ol or aP Expansion Anchor Bolts 79-02 Pipe Support Base Plate 3/8/79 All Power Reactor Designs Using Concrete Facilities with an Using Concrete Expansion OL or a CP Anchor Bolts
IE Bulletin No. 79-02 Enclosure Revision 1, Supplement No. 1 Page 4 of 4 August 20, 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued To No.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) 79-01 Environmental Qualification 2/28/79 All Power Reactor (Correc-of Class IE Equipment Facilities with an tion)
(Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves)78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP