ML19224D644
| ML19224D644 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Byron, Braidwood, Quad Cities, Zion, LaSalle  |
| Issue date: | 07/02/1979 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Brian Lee COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 7907130236 | |
| Download: ML19224D644 (1) | |
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'o UNITED STATES
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c NUCLEAR REGULATORY COMMISSION
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REGION 111 0,
8 799 ROOSEVELT ROAD
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GLEN ELLYN, ILLINOIS 60137 JUL 2 bis Docket Nos. 50-10, 50-237, 50-249, 50-254, 50-265, 50-295. 50-304, 50-373, 50-374, 50-454, 50-455, 50-456 and 50-457 Commonwealth Edison Company ATTN:
Mr. Byron Lee, Jr.
Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:
The enclosed IE Bulletin No. 79-14 is forwarded to you for action.
Written responses are required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, James G. Keppler Director
Enclosure:
IE bulletin No. 79-14 cc w/ enc].
Mr. B. B. Stephenson, Mr. R. Cosaco, Project Station Superintendent Superintende.,t Mr. N. Kalivianakis, Central Files Station Superintendent Direc'.or, NRR/DPM Mr. N. Vandke, Station E2re_ tor, NRR/ DOR Superintendent PDR Mr. L. J. Burke, Site Local PDR Project Superintendt..
NSIC Hr. T. E. Quaka, Quality TIC Assuranc Supervisor Anthony Roicman, Esq.,
Mi. R. H. Holyoak, Station Attorney Superintendent Mr. Dean Hansell, Office Mr. Gunner Sorensen, Site of Assistant Attorney Project Superintendent General
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U.S. NUCLEAR REGULATORY LOMMISSION i
0FFICE OF INSPECTION AND ENFORCEMENT REGION III July 2, 1979 IE Bulletin No. 79-14 SEISMIC ANALYSE 3 FOR AS-BUILT SAFETY-RELATED PIPlNG SYSTEMS Description of Circumstances:
Recenty two issues were identified which can cause seismic analysis of safety-related piping systems to yield nonconservative results.
One issue involved algebraic summation of loads in some seismic analyses.
This was addressed in show cause orderu for Beaver Valley, Fitzpatrick, Maine Yankee and Surry.
It was also addressed in IE Bulletin 79-07 which was sent to all power reactor lice sces.
The other issue involves the accuracy of the information input for seismic analyses.
In this regard, several potentially unconservative factors were discovered and subsequently addressed in IE Bulletin 79-02 (pipe supports) and 75-04 (valve weights). During resolution of these concerns, inspection by IE and by licensees of the as-built configuration of several piping systems revealed a number of nonconformances to design documents which could potentially affect the validity of seismic analyses. Nonconformances are identified in Appendix A to this bulletin.
Because apparently significant non-conformances to design documents have occurred in a number of plants, this issue is generic.
The staff has determined, where design specifications and drawings are used to obtain input information for seismic analysis of safety-related piping systems, that it is essential for these documents to reflect as-built con-figurations. Where subsequent use, damage or modifications affect the con-dition or configuration of safety-related piping systems as described in documents from which seismic analysis input information was obtained, the licensee must consider the need to re-evaluate the seismic analyses to con-sider the as-built configuration.
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l IE Bulletin No. 79-14 July 2, 1979 Page 2 of 3 Action to be taken by Licensees sad Permit IIolders:
All power reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree within che last 12 months, that the seismic analysis applies to the actual configura-tion of safety-related piping systems. The safety related piping includes Seismic Category I systems as defined by Regulatory Guide 1.29, " Seismic Design Classification" Revision 1, dated August 1, 1975 or as defined in the applicable FSAR. For older plants, where Seismic Category 7 require-ments did not exist at the time of licensing, it must be shown that the actual configuration of these safety-related systens meets cesign require-ments.
Specifically, each licensee is requested to:
1.
Identify inspection elements to be used in verifying that the seismic analysis input inf emation conforms to the actual configuration of safety-related systems. For each safety-related system, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic analyses. Also submit a description of the seismic analysis input informatica which is contained in each document.
Identify systems or portions of systems which are planned to be inspected during each sequential inspection identified in Items 2 and 3. Submit all of this information within 30 days of the date of this bulletin.
2.
For portions of systems which are normally accessible *, inspect one system in each set of redundant systems and all nonredundant systems for con-formance to the seismic analysis input information set forth in design documents.
Include in the inspection: pipe run geometry; support and restraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachements; and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).
Within 60 days of the date of this bulletin, submit a de 'ription of the results of this inspection.
Where nonconformances are found which affect operability of any system, the licensee will expedite completion of the inspection described in Item 3.
- Normally accessible refers to those areas of the plar.t which can be entered during reactor operation.
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IE Bulletin No. 79-14 July 2, 1979 Page 3 of 3 3.
In accordance with Item 2, inspect all other normally accessible safety-related systems and all normally inaccessible safety-related systems.
Within 120 days of the date of this bulletin, submit a description of the results of this inspection.
4.
If nonconformances are identified:
A.
Evaluate the effect of the nonconfornance upon system operability under specifien earthquake loadings and comply with applicable action statements in your technical specifications including prompt report-ing.
B.
Submit an evaluation of identified nonconformances on the validity of piping and support analyses as described in the Final Safety Analysis Report (FSAR) or other NRC approved documents. Where you determine that reanalysis is necessary, submit your schedule for: (1) completing the reanalysis, (ii) comparisons of the results to FSAR or other NRC approved acceptance criteria and (iii) submitting descrip-tions of the results of reanalysis.
C.
In lieu of B, submit a schedule for correcting nonconforming systems so that they conform to the design documents. Also submit a descrip-tion of the work required to establish conformance.
D.
Revise documenti io reflect the as-built conditions in plant, and describe measures which are in effect which provide assurance that future modit'ications of piping systems, including their supports, will be reflected in a timely manner in design documents and the r,eismic analysis.
Facilities holding a construction permit shall inspect safety-related systems in accordance with Items 2 and 3 and report the results within 120 days.
Reports shall be submitted to the Regional Director with copies to the Director of the Office of Inspection and Enforcement and the Director of the Division of Operating Reactors, Office of Nuclear Reactor Regulation, Washington, D.C.
20555.
Approved by GAO (RG072); clearance expire. 7/31/80. Approval was given under a blanket clearance specifically for generic problems.
Enclosures:
1.
Appendix A 2.
Listing of IE Bulletins Issued in Last Twelve
., r g Months
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APPENDIX A PLANTS WITH SIGNIFICANT DIFFERENCES BETWEEN ORIGINAL DESIGN AND AS-BUILT CONDITION OF PIPING SYSTEMS Plant Difference Remarks Surry 1 Mislocated supports.
As built cendition Wrong Support Type, caused majority of pipe Different Pipe Run overstress problems, not Geometry.
algebraic summation.
Beaver Valley Not specifically identified.
As built condition resulted Licensee reported "as-built in both pipe and support conditions differ signifi-overstress.
cantly from orginal design."
Fitzpatrick IE inspection identified Licensee is using sa differences similar to built configuration Surry.
for reanalysis.
Pilgrim Snubber sizing wrong.
Plant shutdown to restore Snubber pipe attachment original design condition, welds and snubber support assembly nonconformances.
Brunswick 1 and 2 Pipe supports undersize.
Both units shutdown to restore original design condition.
Ginna Pipe supports not built Supports were repaired to original design, during refueling outage.
St. Lucie Missing seismic supports.
Install corrected Supports on wrong piping.
supports before start up from refueling.
Page 2 APPENDIX A Plant Difference Remarks Nine Mile Point Missing seismic supports.
Installed supports before startup from refueling.
Indian Point 3 Support location and Licensee performing as aupport construction built verification to be deviations.
completed by July 1.
Davis-Besse Gussets missing from main Se7 ports would be over-Steam Line Supports.
stressed. Repairs will be completed prior to start-up.
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IE Bulletin No. 79-14 Enclosure July 2, 1979 Page 1 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-13 Cracking In Feedwater 6/25/79 All PWRs with an System Piping CL for action. All BWRs with a CP for information.
79-02 Pipe Support Base Plate 6/21/79 A' ' Dower Reactor (Rev. 1)
Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or a CP 79-12 Short Period Scrams at 5/31/79 All GE BWR Facilitien BWR 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 a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Prograu Statistics Facilities with an OL 79-09 Failures of GE Tyre AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactols Identified During Facilities with an OL Three Mile Island Incident 79-07 Seismic Stress Analysis 4/'4/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 767 366 J
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IE Bulletin No. 79-14 Enclosure July 2, 1979 Page 2 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subj ect Date Issued Issued To No.79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System Mis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating Licensee 79-06A Review of Operational 4/10/79 All Pressurized Water (Rev 1)
Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Inciden'.79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors end System Mis-Power Reactors with an alignments Identified OL except B&W facilities During the Three Mile Island Incident 79-05A Nuclear Incident at 4/5/79 All D&W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Pcwer Reactor Three Mile Island Facilitiee with an OL and CP 7h'.b 3blb J
IE Bulletin No. 79-14 Enclosure July 2, 1979 Page 3 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE,M01TTHS Bulletin Su 'ect Date Issued Issued To No.
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 78-12B Atypical Weld Material 3/19/79 All Power R actor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reacter 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 3/2/70 All Powcr Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-01A Environmental Qualification 6/6/79 All Power Reactor of Class 1E Equipment Faciliti.'s with an (Deficiencies in the Envi-OL or CP roniental Qualification of ASCO Solenoid Valves) 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equip ent Facilities with an OL or CP 78-14 Deterioration of Buna-N 12/19/78 All CE EWR facilities Component In ASCO with an OL or CP Solenoids
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IE Bulletin No. 79.4 Enclosure July 2, 1979 Page 4 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
78-13 Failures in Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges 11fic licensees Models 7050, 7050B, 7051, i the subject 7051B, 7060, 7060B, 7061
-Ray, Inc.
and 7061B 3 <ges78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12 Atypical Weld Material 9/29/78 All l'ower Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containwent Torus Welds Facilities for action:
Peach Bottom 2 and 3 Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee
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