ML20052F129
| ML20052F129 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/15/1976 |
| From: | Millen C PUBLIC SERVICE CO. OF COLORADO |
| To: | Denise R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19308A354 | List: |
| References | |
| FOIA-82-149 P-76276, NUDOCS 8205120142 | |
| Download: ML20052F129 (1) | |
Text
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PUBLIC SERVICE COMPANY OF COLORADO v (
P. O. 80x 84O OCNVER. COLORAD0 eO2Ot C. E . >11LLEN atmon vict seatsiotwT December 15, 1976 Fort St. Vrain-Unit No. 1 P-76276 Mr. R. P. Denise Asst. Dir. for Advanced Reactors Nuclear Regulatory Commission Division of Reactor Licensing 7920 Norfolk Bethesda, MD 20034 Docket No. 50-267 Gentlemen:
We submitted, for your consideration, a request for a temporary change to Technical Specification LCO 4.2.2.d), on December 8, 1976. This request was subsequently granted by the Commission and consisted of a seven day period of relief from the conditions of the referenced portion of the Technical Specification.
Maintenance work on the s_ubject pump was completed on Saturday, December 11, and the pump was installed and testing began December 12. Testing has indicated that the pump performance is only marginally acceptable and at the present reactor primary system pressure of 6000, it has the capacity to supply makeup water to only two helium circulators. Internal inspection of the pump has revealed that the impellers are badly worn and require replacement. The supplier of the pump informs us it will be necessary to cast replacement impellers. Casting and machining vill take approximately 8 to 9 days. Upon receipt of these new parts it will take-approximately two days to reassemble the pump and another two days to reinstall and test it.
We are, therefore, requesting a 14 day extension for relief from the pro-visions of LCO 4.2.2.d). We also request that we be allowed to raise the power limitation from the 28% limit included in the 12/8/76 request to 35%
of rated reactor power to allow us to proceed with our rise to power testing.
Your earliest response to this request is appreciated.
Very truly yours, f f' 8205120142 820402 9,
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PDR FOIA LYNCH 82-149 PDR C. K. Millen Senior Vice President CR!:il
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(g November 17, 1977 Fort St. Vrain Unit No. 1 P-77229 Mr. Richard P. Denisc A9st. Director for Special Projects Division of Project Management Nuclear Regulatory Commission k'ashington, DC 20555 Docket No. 50-267 Gentlemen:
Subject:
Supports for 2" and Under Piping Ref: P-77199, 9-29-77 In correspondence P-77199, dated 9/29/77, I informed you of a deficiency identified in the supports for 2" and under piping classified as Seismic Class 1. I also indicated that the evaluation and corrective action would be completed in eight to ten weeks, during the period 12--l-77 to 12-15-77.
Based on our findings to date, the following revised schedule is submitted for your infornation:
- 1. Complete walkdown of 1200 Isometric Drawings not previously reviewed. 11-18-77
- 2. Complete physical work identified by walkdown in
- 1) above. 11-30-77
- 3. Complete re-walkdown of 400 Isometric Drawings -
previously reviewed. 12-15-77
- 4. Complete physical work identified by walkdown in
- 3) above. 12-31-77
- 5. Complete documentation update with all corrections identified in 1) and 3) above. 3-15-77 If you have any questions, please let me know.
Very truly yours, k'
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gb,/p ~ . K.-Fuller
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WASHINGTON, D.C. 20555 IE Bulletin No. 79-14 Date: July 2, 1979 Page 1 of 3 SEISMIC ANALYSES FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS Description of Circumstances:
Recently 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 orders for Beaver Valley, Fitzpatrick, Maine Yankee and Surry. It l
was also addressed in IE Bulletin 79-07 which was sent to all power reactor
. licensees.
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 79-04 (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 aubsequent 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-j sider the as-built configurat. ion.
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, Revision 1 Date: July 18, 1979 Page 2 of 3 Action to be taken by Licensees and Permit Holders:
All power reactor facility licensees and construction permit holders are requested to verify, unless verified to an equivalent degree withis,the 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, 1973 or as defined in the applicable FSAR. The action items that follow apply to all safety-related piping 2-1/2-inches in diameter and greater and to seismic Category I piping, regardless of size which was dynamically analyzed by computer. For older plants, where Seismic Category I requirements did not exist at the time of licensing, it must be shown that the actual configuration of /E//g safety-related systems, utilizing pip _ing 2-1/2 inches in diameter and greater, meets design require-ments.
Specifically, each licensee is requested to:
- 1. Identify inspection elements to be used in verifying that the seismic analysis input information 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 information 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 attachments 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 description 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.
TNormally accessible refers to those areas of the plant which can be entered during reactor operation.
IE Bulletin No. 79-14 Date: 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 nonconformance upon system operability under specifjed 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 descriptions 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 documents to reflect the as-built conditions in plant, and describe measures which are in effect which provide assurance that future modifications of piping systems, including their supports, will be reflected in a timely manner in design documents and the seismic analysis.
Facilities holding a construction permit shall inspect safety-relate?
systems in accordance with Items 2 and 3 and report the results within t
120 days.
Reports shall be submitted to the Regional Director with copies to the Director of the Of fice 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 (R0072); clearance expires 7/31/80. Approval was given ,
under a blanket clearance specifically for generic problems. l 1
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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 condition 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 as differences similar to built configuration l Surry. for reanalysis.
Pilgrim Snubber sizing wrong. Plan shutdown to restore Snubber pipe attachment original design condition.
welds and snubber support assembly nonconformances.
Brunswick I and 2 Pipe supports undersize. Both units shutdown to
- restore original design condition.
t 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.
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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-support construction built verification to be deviations. completed by July 1.
Davis-Besse Gussets missing from main Supports would be over-Steam Line Supports. stressed. Repairs will be completed prior to start-up.
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IE Bulletin No. 79-14 July 2, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTilS Bulletin Subject Date Issued Issued To No.
78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containment Torus Facilities for Welds action: Peach Bottom 2 and 3, Quad Cities 1 and 2, liatch 1, Monti-cello and Vermont Yankee 78-12 Atypical Weld Material 9/26/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP)78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP) 78-1211 Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP) 78-13 Failures in Source 10/27/78 All General and licads of Kay-Ray, Specific Licensees Inc., Gauges Models with Kay-Ray Gauges 7050, 7050B, 7051, 7051B, 7060, 7060B, .
7061 and 7061B Enclosure Page 1 of 4 7
IE Bulletin No. 79-14 July 2, 1979 78-14 Deterioration of
~ 12/19/78 All GE BWR facilities Buna-N Components with an Operating, in ASCO Solenoids License (OL) or Construction Permit e (CP) 79-01 Environmental Quali- 2/8/79 All Power Reactor fication of Class IE Facilities with an Equipment Operating License (OL) or Construction Permit (CP)79-01A Environmental Qualification 6/6/79 All Power Reactor of Class IE Equipment Facilities with an d
Operating License (OL) or Construction Permit (CP) 79-02 Pipe Support Base 3/8/79 All Power Reactor Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-02 Pipe Support Base 6/21/79 All Power Reactor (Rev. 1) Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 2 79-03 Longitudinal Weld 3/12/79 All Power Reactor Defects In ASME SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spools (OL) or Construction Manufactured by Permit (CP)
Youngstown Welding .
and Engineering Coupany 79-04 Incorrect Weights 3/30/79 All Power Reactor for Swing Check Facilities with an Valves Manufactured Operating License by Velan Engineering (OL) or Construction Corporation Permit (CP)
Enclosure Page 2 of 4 1
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IE Bulletin No. 79-14 July 2, 1979 79-05 Nuclear Incident at 4/1/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05A Nuclear Incident at 4/5/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction
, Permit (CP)79-05B Nuclear Incident at 4/21/79 All B&W Power Reactor Three Mile Island Facilities with an Operating License (OL) 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Power Reactor Facilities Misalignments Identified Except B&W Facilities During The Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Westinghouse PWR Errors and System Facilities with an Misalignments Identified Operating License During the Three Mile (OL)
Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev. 1) Errors and System Mis- Power Reactor Facilities align.nents Identified of Westinghouse Design During the Three Mile with an Operating License (OL)
Island Incident 79-06B Review of Operational 4/14/79 All Combust. ion Engineer-Errors and System ing PWR Facilities with Misalignments Identified an Operating License During The Three Mile (OL)
Island .
79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating License (OL) or Construction Permit (CP)
Enclosure Page 3 of 4
IE Bulletin No. 79-14 July 2, 1979 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors Identified Facilities with an During Three Mile Island Operating License Incident (OL) or Construction Permit (CP) 79-09 Failures of GE Type AK-2 5/11/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems Operating License (OL) or Construction Permit (CP) 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an Operating License (OL) 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety Operating License (OL) or Systems a Construction Permit (CP) 79-12 Short Period Scrams at 5/31/79 All Power Reactor Facilities BWR Facilities with an Operating License (OL)
- l or a Construction Permit (CP) 79-13 Cracking in Feedwater 6/25/79 All PWR with an Operating System Piping License (OL) for action.
All BWR with a Construction Permit (CP) for information C
Enclosure Page 4 of 4 4
CUORA1. I ll.ES PDR:HQ
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Docket No. 50-267 Public Service Company of Colorado ATTN: Mr. C. K. Millen Senior Vice President P. O. Box 840 Denver, Colorado 80201 Centlemen:
IE Bulletin 79-14 is revised to limit the scope of work. required. The changes are indicated on the enclosed replacement page for the bulletin. If you desire additional information regarding this matter, please contact this oEiIce.
, Sincerely,
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Enclosure:
IE Bulletin No. 79-14 Revision 1 cc: D. W. Warembourg, Nuclear Production Manager Fort St. Vrain Nuclear Station P. O. Box 368 Platteville, Colorado 80651 L. Inrcy, Manager, Quality Assurance
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WASHINGTON, D.C. 20555 l Supplement IE Bulletin No. 79-14 Date: August 15, 1979 Page 1 of 2 SEISMIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS Description of Circumstances:
IE Bulletin No. 79-14 was issued on July 2, 1979, and revised on July 18, 1979.
The Bulletin requested licensees to take certain actions to verify that seismic analyses are applicable to as-built plants. This supplement to the Bulletin provides additional guidance and definition of Action Items 2, 3, and 4.
To comply with the requests in IE Bulletin 79-14, it will be necessary for licensees to do the following:
- 2. Inspect Part. of the Accessible Piping For each system selected by the licensee in accordance with Item 2
, of the Bulletin, the licensee is expected to verify by physical t'
inspection, to the extent practicable, that the inspection elements .
meet the acceptance criteria. In performing these inspections, the licensee is expected to use measuring techniques of sufficient accuracy to demonstrate that acceptance criteria are met. Where inspection elements important to the seismic analysis cannot be viewed because of thermal insulation or location of the piping, the licensee is expected to remove thermal insulation or provide access. Where physical inspection is not practicable, e.g., for valve weights and materials of construction, the license is expected to verify conformance by inspection of quality assurance records. If a nonconformance is found, the licensee is expected in accordance with Item 4 of the Bulletin to perform an evaluation of the significance of the nonconformance as rapidly as possible to determine whether or not the operability of the system might be jeopardized during a safe shutdown earthquake as defined in the Regulations. This evaluation is expected to be done in two phases involving an initial engineering judgement (within 2 days), followed by an analytical engineering evaluation (within 30 days). Where either phase of the evaluation shows that system operability is in jeopardy, the licensee is expected to meet the applicable technical specification action statement and complete the inspections required by Item 2 and 3 of the Bulletin as soon as possible. The licensee must report the results of these inspections in accordance with the require-ments for content and schedule as given in Item 2 and 3 of the Bulletin.
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Supplement IE Bulletin No. 79-14 Date: August 15, 1979 Page 2 of 2
- 3. Inspect Remaining Piping The liceesee is expected to inspect, as in Item 2 above, the remaining safety-related piping systems which were seismically analyzed and to report the results in accordance with the requirements for content and schedule as given in Item 3 of the Bulletin.
4A. Evaluate Noncomformances With regard to Item 3A for the Bulletin, the licensee is expected to include in the initial engineering judgement his justification for continued reactor operation. For the analytical engineering evaluation, the licensee is expected to perform the evaluation by using the same analytical technique used in the seismic analysis or by an alternate, less compicx technique provided that the licensee can show that it is conservative.
If either part of the evaluation shows that the system may not perform its intended function during a design basi earthquake, the licensee must promptly comply with applicable action statements and reporting requirements in the Technical Specifications.
4B. Submit Nonconformance Evaluations The licensee is expected to submit evaluations of all nonconformances and, where the licensee concludes that the seismic analysis may not be conservative, submit schedules for reanalysis in accordance with Item 4B of the Bulletin or correct the noncomformances.
4C. Correct Nonconformances If the licensee elects to correct nonconformances, the licensee is expected to submit schedules and work descriptions in accordance with Item 4C of the Bulletin.
4D. Improve Qualtiy Assurance If noncomformances are identified, the licensee is expected to evaluate and improve quality assurance procedures to assure that future modifica-tions are handled efficiently. In accordance with Item 4D of the Bulletin, the licensee is expected to revise design documents and seismic analyses in a timely manner.
The schedule for the action and reporting requirements given in the Bulletin as originally issued remains unchanged.
Approved by GAO (R0072); clearance expires 7/31/80. Approval was given under a blanket cicarance specifically for generic problems.
l Supplement IE Bulletin No. 79-14 l
. August 15, 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 IE Equipment Facilities with an Operating License (OL) or Construction Permit (CP) 79-02 Pipe Support Base 3/8/79 All Power Reactor Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-02 Pipe Support Base 6/21/79 All Power Reactor (Rev. 1) Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-03 Longitudinal Weld 3/12/79 All Power Reactor Defects In ASME SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spools (OL) or Construction Manufactured by Permit (CP)
Youngstown Welding and Engineering Company 79-04 Incorrect Weights 3/30/79 All Power Reactor for Swing Check Facilities with an Valves Manufactured Operating License by Velan Engineering (OL) or Construction Corporation Permit (CP) 79-05 Nuclear Incident at 4/1/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)
Enclosure Page 1 of 4
r Supplement IE Bulletin No. 79-14 August 15, 1979 t
79-05A Nuclear Incident at 4/5/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05B Nuclear Incident at 4/21/79 All B&W Power Reactor Three Mile Island Facilities with an Operating License (OL) 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Power Reactor Facilities Misalignments Identified Except B&W Facilities During The Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Westinghouse PWR Errors and System Facilities with an Misalignments Identified Operating License During the Three Mile (OL)
Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev. 1) Errors and System Mis- Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an Operating License (OL)
Island Incident 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System ing PWR Facilities with Misalignments Identified an Operating License During The Three Mile (OL)
Island 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating License (OL) or Construction Permit (CP)
Enclosure Page 2 of 4
Supplement IE Bulletin No. 79-14 August 15, 1979 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors Identified Facilities with an During Three Mile Island Operating License Incident (OL) or Construction Permit (CP) 79-09 Failures of GE Type AK-2 5/11/79 All Power Reactor '
Circuit Breaker in Safety Facilities with an Related Systems Operating License (OL) or Construction Permit (CP) 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an Operating License (OL) 79-11 Faulty overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety Operating License (OL) or Systems a Construction Permit (CP) 79-12 Short Period Scrams at 5/31/79 All Power Reactor Facilities BWR Facilities with an Operating License (OL) or a Construction Permit (CP) 1 79-13 Cracking In Feedwater 6/25/79 All PWRs with an Operating
\ System Piping License (OL) for action.
All BWR with a Construction Permit (CP) for information ,
l 79-14 Seismic Analyses for 7/2/79 All Power Reactor facilities As-Built Safety-Related with an Operting License Piping System (OL) or a Construction Permit (CP) 79-15 Deep Draft Pump 7/11/79 All Power Reactor Facilities Deficiencies with a Construction Permit and/or Operating License (OL)
'79-10 Vital Area Access Controls 7/26/79 Al'1 Power Reactors with an Operating License (OL) or anticipating fuel loading prior to January 1981.
79-17 Pipe Cracks in Stagnant 7/26/79 All PWRs with Operating Borated Water Systems at License (OL)
PWR Plants
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Enclosure Page 3 of 4
Supplement IE Bulletin No. 79-14 August 15, 1979 79-18 Audibility Problems 8/6/79 All Power Reactor Encountered on Evacuation Facilities with an of Personnel from High- Operating License (OL)
Noise Areas 79-19 Packaging Low-Level 8/10/79 All Power and Research Radioactive Waste for Reactors with Operating Transport and Burial Licenses (OLs), fuel facilities except uranium mills, and certain
' materials licensees 79-20 Packaging Low-Level 8/10/79 Materials Licensees Radioactive Waste for who did not receive Transport and Burial Bulletin No. 79-19 79-21 Temperature Effects on 8/13/79 All PWR's with an Level Measurements Operating License (OL) for action. All BWR's with a Construction Permit (CP) for information i
Enclosure Page 4 of 4
l)f UNITED STATES NUCLEAR REGULATORY COMMISSION q OFFICE OF INSPECTION AND ENFORCEMENT j WASHINGTON, D.C. 20555 9 IE Bulletin No. 79-14 Supplement 2 Date: Septenber 7, 1979 Page 1 of 2 SEISHIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS Description of Circumstances:
IE Bulletin No. 79-14 was issued on July 2, revised on July 18, and first supplemented on August 15, 1979. The bulletin requested licensees to take certain actions to verify that seismic analyses are applicable to as-built plants. Supplement 2 provides the following additional guidance with regard to implementation of the bulletin requirements:
Nonconformances One way of satisfying the requirements of the bulletin is to inspect safety-related piping systems against the specific revisions of drawings which were used as input to the seismic analysis. Some architect-engineers (A-E) however, are recommending that their customers inspect these systems against the latest revisions of the drawings and mark them as necessary to define the as-built configuration of the systems. These drawings are then returned to the A-E's offices for comparison by the analyst to the seismic analysis input. For licensees taking this approach, the seismic analyst will be the person who will identify nonconformances.
The first supplement to the bulletin provided guidance with regard to evaluation of nonconformances. That guidance is appropriate for licensees inspecting against later drawings. The licensee should assure that he is promptly notified when the A-E identifies a nonconformance, that the initial engineering judgment is completed in two days, and that the analytical engineering evaluation is completed in 30 days. If either the engineering judgement or the analytical engineering evaluation i'ndicates that system operability is in jeopardy, the licensee is expected to meet the applicable tech ical specification action statement.
Visual Approximations Some licensees are visually estimating pipe lengths and other inspection elements, and have not documented which data have been obtained in that way.
Visual estimation of dimensions is not encouraged for most measurements; however, where visual estimates are used, the accuracy of estimation must be within toler-ance requirements. Further, in documenting the data, the licensee must specif-ically identify those data that were visually estimated.
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IE Bulletin No. 79-14 Supplement 2 Date: September 7, 1979
. Page 2 of 2 Thermal Insulation In many areas, thermal insulation interferes with inspection of pipe support details, i.e. , attachment welds, saddles, support configuration, etc. In some areas, the presence of thermal insulation may result in unacceptably large uncertainties for determination of the location of pipe supports.
Where thermal insulation obstructs inspection of support details, the insulation should be removed for inspection of a minimum of 10% of the obstructed pipe supports in both Items 2 and 3 inspections. In the Item 3 response, the licensee should include a schedule for inspecting the remaining supports.
Where necessary to determine the location of pipe supports to an accuracy within design tolerances, thermal insulation must be removed.
Clearances For exposed attachments and penetrations, licensees are expected to measure or estimate clearances between piping and supports, integral piping attachments (e.g., lugs and gussets) and supports, and piping and penetrations. Licensees are not expected to do any disassembly to measure clearances.
Loose Bolts Loose anchor bolts are not covered by this bulletin, but are covered by IE Bulletin No. 79-02. Any loose anchor bolts identified during actions taken for this bulletin should be dispositioned under the requirements of IE Bulletin No. 79-02.
Other loose bolts are to be treated as nonconformances if they invalidate the seismic analysis; however, torquing of bolts is not required.
Difficult Access Areas where inspections are required by the IE Bulletin, but are considered impractical even with the reactor shutdown, should be addressed on a case-by-case basis. Information concerning the burden of performing the inspection and the safety consequence of not performing the inspection should be documented by the licensee and forwarded for staff review.
Schedule The schedule for the action and reporting requirements given in tLe IE Bulletin as originally issued remains unchanged.
Approved by GAO (R0072); cicarance expires 7/31/80. Approval was given under a blanket clearance specifically for generic problems.
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IE Bulletin No. 79-14 Supplement 2 September 7, 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTilS Bulletin Subject Date Issued Issued To No.79-01A Environmental Qualification 6/6/79 All Power Reactor of Class IE Equipment Facilities with an Operating License (OL) or Construction Permit (CP) 79-02 Pipe Support Base 3/8/79 All Power Reactor Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-02 Pipe Support Base 6/21/79 All Power Reactor (Rev. 1) Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction
[ Permit (CP) 79-03 Iongitudinal Weld 3/12/79 All Power Reactor Defects In ASME SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spools (OL) or Construction Manufactured by Permit (CP)
Youngstown Welding and Engineering Company 79-04 Incorrect Weights 3/30/79 All Power Reactor for Swing Check Facilities with an Valves Manufactured Operating License by Velan Engineering (OL) or Construction Corporation Pe rmit (CP) 79-05 Nuclear Incident at 4/1/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)
Enclosure Page 1 of 4
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IE Bulletin No. 79-14 Supplement 2 September 7, 1979 79-05A Nuclear Incident at 4/5/79 All Power Reactor Three Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05B Nuclear Incident et 4/21/79 All B&W Power Reactor Three Mile Island Facilities with an Operating License (OL) 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Power Reactor Facilities Misalignments Identified Except B&W Facilities During The Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Westinghouse PWR Errors and System Facilities with an Misalignments Identified Operating License During the Three Mile (OL)
Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev. 1) Errors and System Mis- Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an Operating License (OL)
Island Incident 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System ing PWR Facilities with Misalignments Identified an Operating License During The Three Mile (OL)
Island 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating License (OL) or Construction Permit (CP)
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IE Bulletin No. 79-14 Supplement 2 September 7, 1979 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors identified Facilities with an During Three tille Island Operating License Incident (OL) or Construction Permit (CP) 79-09 Failures of GE Type AK-2 5/11/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems Operating License (0L) or Construction Permit (CP) 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an Operating License (OL) 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety Operating License (OL) or Systems a Construction Permit (CP) 79-12 Short Period Scrams at 5/31/79 All Power Reactor Facilities BWR Facilities with an Operating License (OL) or a Construction Permit (CP) 79-13 Cracking In Feedwater 6/25/79 All PWRs with an Operating System Piping License (OL) for action.
All BWR with a Construction Permit (CP) for information 79-14 Seismic Analyses for 7/2/79 All Power Reactor facilities As-Built Safety-Related with an Operting License Piping System (OL) or a Construction Permit (CP) 79-15 Deep Draft Pump 7/11/79 All Power Reactor Facilities Deficiencies with a Construction Permit and/or Operating License (OL) 79-16 Vital Area Access Controls 7/26/79 A51PowerReactorswithan Operating License (OL) or anticipating fuel loading prior to January 1981.
79-17 Pipe Cracks in Stagnant 7/26/79 All PWRs with Operating Borated Water Systems at License (OL) or a PWR Plants Construction Permit (CP) i Enclosure Page 3 of 4
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IE Bulletin No. 79-14 Supplement 2 September 7, 1979 79-18 Audibility Problems 8/6/79 All Power Reactor Encountered on Evacuation Facilities with an of Personnel from High- Operating License (OL)
Noise Areas or a Construction Permit (CP) 79-19 Packaging Low-Level 8/10/79 All Power and Research Radioactive Waste for Reactors with Operating Transport and Burial Lisenses (OLs), fuel facilities except uranium mills, and certain materials licensees 79-20 Packaging Low-Level 8/10/79 Materials Licensees Radioactive Waste for who did not receive Transport and Burial Bulletin No. 79-19 79-21 T mperature Effects on 8/13/79 All PWR's with an Level Measurements Operating License (OL) or Construction Permit (CP)
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g Eo? r Eji.E eL,di.H 4S senice company of Cobrado T C '-s
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Fort St. Vrain Unit No. 1 P-79238 Mr. Karl V. Seyfrit, Director riuclear Regulatory Commission Region IV Office of Inspection and Enforcement 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76012
Subject:
Safety Related Piping System Audit
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Reference:
PSC Letter P-79161 dated Aug. 1, 1979 PSC Letter P-79198 dated Sept.4, 1979
Dear Mr. Seyfrit:
On September 24, 1979, Public Service Company of Colorado notified the t1RC Resident Inspector that the inconsistencies noted in the 12 pipe hangers of the random sample audit conducted on the safety related piping systems and reported to the flRC in PSC Letter P-79198 dated September 4, 1979 had been resolved and that the Fort St. Vrain plant would resume power operation effective September 25, 1979.
The resolutions of the discrepancies in the 12 hangers involved in the sample audit were discussed with the t1RC Resident Inspector during the week of September 24, 1979. The lines associated with the hangers in question required the utilization of sophisticated analytical techniques to ascertain system operability which could not be completed in the alloted time thereby necessitating a plart shutdown until the anlaysis work could be completed. The lines and hangers have since been analyzed using sophisticated analytical techniques when necessary and, as a result, has allowed PSC Engineering to conclude that there was no system impairment in eleven of the twelve cases. The twelf th case involved a very complex hanger (llA-II-46130) containing numerous Class I and Class II lines that required model simplification techniques to analyze without exceeding computer core capability on the available computers. PSC Engineering continued the analysis to the point where it was concluded that the hanger would meet seismic design requirements if aight fillet welds were increased from 3/16" to 5/16" as originally specified on the hanger drawing. A craft action was written and the additional weldment was applied to the hanger. The analysis required to assess the acceptability of the as-found .
condition of. the hanger was both time consuming and costly. For these reasons, Y
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the analysis to determine operability was discontinued at the point of craft action completion. At this point in time system operability was arbitrarily ass"med to be in jeopardy when, in fact, such might not have been the case.
As a result of the sample audit program PSC concludes: (1) the use of simplified analysis techniques is conservative in that system impairment may be indicated, when in reality, system impairment would not be indicated when analyzed by sophisticated analytical techniques, (2) the simplified tech-niques would have been satisfied had corrective action been taken to correct the discrepancies to return the hangers to it's originally intended configuration and (3). even though discrepancies exist, system operability is not in jeopardy.
Based on the above, PSC intends to implement the requirement of I & E Bulletin 79-14 as follows:
- 1. A 100% audit of all accessible 21/2" and larger and 2" and under computer analyzed Class I piping. The piping audit is being conducted at this time. The systems or portions of systems required to facilitate reactor primary coolant depressurization and decay heat removal through PCRV liner cooling are being audited on a first priority basis.
- 2. Discrepancies found during the field audit shall be resolved by expediently returning the discrepant item to it's originally intended configuration. Any discrepant item that can not be resolved in this manner will be evaluated per item 3.
- 3. An evaluation will be performed based on engineering judgement using the as-built drawings. The evaluation will be completed within two days. If this evaluation indicates that the affected system operability is in jeopardy, the applicable Technical Specification action will be taken.
- 4. If the initial engineering judgement indicates that the system operability-is not in jeopardy, an analytical engineering evaluation will be conducted and completed within 30 days. The technique used in the analytical evaluation will be based upon simplified analytical techniques. If the simplified analyti9 al technique indicates that the piping or hanger does not meet the design requirements of the simplified technique, corrective action to bring the piping or hanger into compliance with the simplified analytical techniques will be completed within tha 30 day period.
If the piping system or hanger analysis and craft action (if required) is not completed within this 30 day period, the applicable Technical Specification action will be taken. In addition, approximately five per cent of the discrepancies will be analyzed using a sophisticated analytical technique to verify the judgemental evaluations and the conservatism of the simplified analytical techniques. At the completion of the program, all affected documents will be revised to reflect the as-built conditions in the plant. PSC's current m
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engineering and design procedures will be modified, if required, to provide assurance that modifications to systems include updating of piping supports. The current PSC Engineering procedures are adequate but may be further refined as we proceed through the as-built programs.
All items requiring Technical Specification reporting will be reported in accordance with the applicable directives contained therein. All
" discrepancies" will be reported to the Commission on 60-day intervals.
The 60 day report shall include a short summary of all discrepancies and l corrective actions. PSC intends to correct all discrepancies in a timely nenner throughout the duration of the hanger pro' gram. l Based upon the scope and available manpower, we anticipate a projected project completion date of December,1980 with a pessimistic completion ~
date of March,1981. If there are any questions concerning this program, please contact this office.
Very truly yours, W
Frederic E. Swart Nuclear Project Manager FES/LMM/scp
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