IR 05000282/1988017
| ML20155G704 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/11/1988 |
| From: | Danielson D, James Gavula NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20155G694 | List: |
| References | |
| 50-282-88-17, 50-306-88-17, IEB-79-14, NUDOCS 8810170351 | |
| Download: ML20155G704 (6) | |
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{{#Wiki_filter:_ _ - -_________ - _____ ,. . V. S. NUCLEAR REGULATORY COMMISSION
REGION III
= ports No.: 50-282/88017(DRS);50-306/88017(DRS) socket Nos.: 50-282; 50-306 Licenses No.: DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Mall flinneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plant Units 1 and 2 Inspection At: Prairie Island Site Red Wing, MN Inspectiun Conducted: September 21-22, 1988 /C/7 M Inspector: u . Date~ Approved By: D. H. Danielson, Chief /o/N/// Materials and Processes Section Date ~ ' Inspection Summary inspection on September 21-22,1988 (Reports No. 50-282/88017(DRS); 50-306/88017(DRS)) Areas Inspected: Special safety inspection of facility modifications which resulted from the ongoing IEB 79-14 reviews (37701); and licensee action on previously identified concerns (92701).
Results: One violation was identified (inadequate bases for safety evaluation Paragraph 4.a). Ongoing reviews of IEB 79-14 records have identified discrepancies which exceeded allowable stress values specified in the Safety Analysis Report.
Modifications have been made to 20 supports to correct deficiencies resulting from Copes Vulcan center of gravity discrepancies and other IEB 79-14 related problems.
" Review process is approximately one-third completed, l 8810170351 681012 l , ' PDR O ADOCK 050002G2 PDC
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.. .- DETAILS 1.
Persons Contacted Northern States Power Company (NSP)
- +J. Goldsmith, Superintendent, Nuclear Technical Services
+G. Eckholt, Licensing Engineer
- +G. Rolfson, Engineer
- +J. Donate 11, Engineer
- +C. Baltos, Engineering Associate Fluor Daniel, Inc. (FDI)
- +B. Dickerson, Project Piping Engineer
+J. Khanna, Engineer Teledyne Engineering Services +M. McKeown, Project Engineer Nuclear Regulatory Comission (NRC) +J. Hard, Senior Resident Inspector
- M. Moser, Resident Inspector
+0. Dilanni, Licensing Project Manage'. NRR +S. Hou, Senior Mechanical Engineer, Nr R +B. Burgess, Chief, Projects Section 2A, RIII + Attended meeting at Prairie Island on September 21, 1988.
- Attended exit interview on September 22, 1988, 2.
Licensee Action on Previously Identified Items a.
(Closed) Open Item (282/87018-01; 306/87017-01): Comprehensive review program of all previous IEB 79-14 work was implemented by the licensee to identify and correct inconsistencies in applying and documenting reconciliation criteria.
Based on the potential significance of the discrepancies being found and the number of re-analyses required to resolve the identified discrepancies, the acceptability of the original IEB 79-14 effort requires further evaluation. On this basis, the above item is closed and an Unresolved Item will be opened.
(See Paragraph 4 of this report for details on this item (28?/88017-02; 306/88017-02)).
b.
(Closed) Open Item (282/88009-01; 306/88009-01): Criteria and methodology for establishing a justification for continued operation requires additional evaluation.
, Based on the meeting discussed in Paragraph 3 of this report, the final criteria submitted by NSP was determined to be acceptable.
This item is considered closed.
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3.
Meeting to Discuss Criteria for Justification for Continued Operation . A meeting was held between representatives of the licensee and the NRC - (denoted in Paragraph 1) on September 21, 1988.
The purpose of the meeting was to discuss details of NSP's recently submitted "Criteria for t Determining Justification for Continued Operation when Encountering "As-Built" Discrepancies Which Cause Large Increases in Seismic Stress".
. A draft copy was issued on June 8,1988, and a completely revised document was submitted for review during the September 21, 1988 meeting (See Enclosure 3).
The major points of discussion during the meeting pertained i .' mainly to administrative aspects of the criteria. Slight changes to some i of the wording in criteria were made during the meeting to resolve these
concerns.
From a technical perspective, the criteria was determined to l be acceptable for piping and pipe support operability stress limits.
, Based on the conclusions reached during the meeting, no additional l l evaluations of the criteria will be required. This issue is considered closed.
' 4.
Facility Modifications Resulting from IEB 79-14 Reviews j a.
Background As documented in NRC Inspection Reports No. 50-282/87018; 50-306/87017, t a review program is currently underway at FDI to reassess the original
IEB 79-14 work.
This effort resulted from FDI's recognition that >
' some inconsistencies potentially existed in applying and documenting [ the original reconciliation criteria at Prairie Island.
During the ' implementation of this program, the discrepancies on the Copes Vulcan ' valve weights and centers of gravity were discovered (See NRC ' Inspection Reports No. 50-282/88009; 50-306/88009 for additional i
details). As a result of these discrepancies, several piping systems were found to exceed code stress allowables and in one instance, the l stresses were calculated to be approximately 160 ksi for the design basis earthquake (DBE) load.
This was almost six times the allowable , stress limit.
As a result of this situation, Safety Evaluation Report No. 259-1, ( Revision 1 "Justification for Continued Operation Due to Increased i Valve Weights of Small Bore Copes Vulcan Control Valves" was issued on May 6, 1988.
This evaluation provided the bases and criteria l used to justify continued operation until the analyses could be ' completed and any resulting required modifications implemented.
During the review of this report, the NRC inspector noted the following problems:
- The original piping system was designed and constructed in
, accordance with ANSI B31.1, Power Piping, 1967, but allowable stress limits for ASME Section III, Class I piping were used.
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Although the report states that Class I piping requires a
fatigue evaluation and also takes into account the plastic behavior of the piping material, no justification was given , for differences between ASME Class I material and construction requirements and ANSI B31.1 macerial requirements.
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The report states that, "it is assumed for purposes of this evaluation that the seismic stresses calculated using non-linear techniques is less than 3 times the allowable stress intensity".
No quantitative justification is provided for this assumption.
- A Class I fatigue evaluation and associated stress criteria was
. , used to determine the operability of the piping system. No > bases were given as to why primary plus secondary stress
intensity ranges would be applicable to a strictly primary stress situation. The limitations for Level D service limits for faulted load conditions were exceeded by over a factor of l three.
- Because of the concerns stated above, it cannot be concluded
, that the accumulator pressure boundary would survive a DBE.
Furthermore, the bases for stating that the charging pumps for
the safety injection system would keep up with a leak, if a ' , ' portion of the piping system in question ruptured, was not , provided in the evaluation.
As a result of the above, it was determined the bases for the above
safety evaluation were not adequately established. This is an f < example of a violation of 10 CFR 50.59 (282/88017-01; 306/88017-01).
f b.
Modification Implementation
' Even though NSP had completed the above justification for continued
operation, imediate steps were taken to inodify the pipe supports
to brina a 1 stresses within the code allowables.
In addition, t i " all of *. 5er discrepancies which were identified prior to the Unit 1 dli
- T were corrected before the unit was restarted.
For Unit 1, the following modifications have recently been made: t ' Stress Report No. PI-216-II, Line 9.
- Support No. 1-RHRRH-21:
Reinforced existing support.
< Support No. 1-RHRRH-24: Reinforced existing support.
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Stress Report No. PI-233-28, Line 39.
l Support No. 1.CCRH-41: Added U-bolt to existing support.
- Stress Report No. PI-206-1 & !!!, Line 6.
New anchor installed (never installed originally).
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Stress leport No. PI-206-43, Lines 38B & 38C.
Support No. 1-RCVCH-857: Reinforced existing support.
New support added to valve CV-31334.
, Pipe Whip No. 14CVCS-2: Modified into a 2-way guide.
- Stress Report No. PI-234-XIV, Lines 113/114.
! New support added to valve CV-31329.
j Support No. RCRH-14: Rod replaced to accomodate uplif t.
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- - . . _ _ i .* .* , i Support No. RCRH-18: Rod replaced to accommodate uplift.
, Support No. RCRH-19: Kicker added to existing support.
i Support No. RPCH-2: Vertical modified into 2-way guide.
Support No. RCVCH-1451: Increase weld size.
- Stress Report No. PI-205-22, Line 14.
New support added to valve CV31445.
Support No. RSIH-32: U-bolt sti ffened.
l S+ress Report No. PI-205-VI, Line 16.
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New support added to valve CV-31442.
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Stress Report No. PI-205-14, Line 108.
New support added to valve CV-31447.
New support added to valve CV-31448.
. Support No. 1-RSIH-423: Modified existing 1-way into 2-way.
Support No. 1-RS!H-424: Modified existing 1-way into 2-way, i Drawings for the above modifications were briefly reviewed by the
NRC inspector. Additional inspections of the design analyses associated with these modifications are planned in the future, , The majority of the above modifications were walked down by the ' NRC inspector while at the site.
Fo adverse comments were made i relative to the general installation configuration or workmanship.
The discrepancies associated with the above modifications were caused by several factors. As previously mentioned, changes in the weights and centers of gravity for certain Copes Vulcan valves required re-analysis and subsequent modifications to be made.
This was the case in most of the above modifications; however, several modifications were required which were not associated with the Copes Vulcan issue.
In one case, an anchor was not installed as originally designed.
Although the IEB 79-14 field walkdown indicated it was not installed, the review of the analyses did not identify this deficiency.
another instance, the wrong response spectra was used in the seismic analysis, c.
Scope of Future Modifications The current review program has identified 91 subsystems which will require re-analysis to reconcile the as-built configuration to the design analysis. To date, approximately one-third of this re-analysis is completed. Since the worst cases are being addressed first, it is assumed that very few additional modifications will be required.
There are currently four subsystems in Unit 2 which will require some degree of modifications but other than those, no additional modifications are anticipated.
Based on the current scope of the required re-analysis and the modifications required to date, the acceptability of the original IEB 79-14 effort will require additional evaluation by the NRC.
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.. . Pending the determination of the extent and significance of the as-built discrepancies, this will be considered an Unresolved Item (282/88017-02; 306/88017-02).
5.
Unresolved items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item, a deviation, or a violation.
Unresolved items disclosed during this inspection are discussed in Paragraph 4.c.
6.
Exit Interview The Region 111 inspector met with the licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on September 22, 1988.
The inspector sumnarized the purpose and findings of the inspection. The licensee representatives acknowledged this information.
The inspector also df tcussed the likely informational content of the inspection report with regard to dacuments or processes reviewed during the inspection.
The licensee representatives did not identify any such documents / processes as proprietary.
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i .- ENCLOSURE 3 . . Northern States Power Company ' 414 Ngoliet Man Minneapoks. Minnesota 55401 Telephone (612) 330 5500 September 26, 1988 . Director of Nuclear Reactor Regulation U S Nuclear Regul' tory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE IS1AND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR 42 50-306 DPR 60 Generie Criteria For Justification Of Continued Ooeration on September 21, 1988, the NRC Staff met with NSP representatives at Prairie Island to discuss generic criteria to be utilized for the justification of continued operation when major discrepancies in "as-built" safety related piping are encountered. Attached is a copy of the generic criteria agreed to during the meeting.
Please contact us if you have any questions regarding the information provided.
[[2)c.1-)Y)Y'J5 David Musolf j Manager Nuclear Support Se rvices
c: Regional Administrator III, NRC NRR Project Manager, NRC j Senior Resident Inspector, NRC G Charnoff Attachment, Criteria For Deteruining Justification For Continued Operation When Encountering Major Discrepancies in "As Built" Safety Related Piping, September 21, 1988 .
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', . . , . . . CRITERIA FOR DESERMINING JUSTIFICATION FOR CONTINUED OPERATION WHEN ENCOUNTERING MAJOR DISCREPANCIES IN "AS-BUILT" SAFETY RELATED PIPING . SEPTEMBER 21, 1988 .
NORTHERN STATES POWER CO.
' PRAIRIE ISLAND NUCLEAR GENERATING PLANT
1717 WAKONADE DRIVE EAST , f WELCH, MN 55089
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. _ _ _ ___ . . . TABLE OF CONTENTS
PAGE 1.0 INTRODUCTION & SCOPE.................................
3 2.0 CRITERIA............................................
3.0 CONCLUSION........................ .................
4.0 REFERENCES.......................................... . l l l
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. .. . . . 1.0 INTRODUCTION & SCOPE These criteria are in' tended to assure the operability require- - ments of safety related piping and associated supports if it is determined that stresses exceed allowables presented in the Prairie Island USAR.
These criteria permit operation for an interim period only.
Modifications will be made which return the system to within USAR allowables by the nexu re-fueling outage or sooner if operation permits.
These criteria are intended to expeditiously perform necessary evaluations to determine interim operability and not to delay appropriate actions.
For cases involving components classified as ASME Code Class I where USAR allowables are exceeded, NSP shall be notified upon discovery and NSP chall evaluate reportability require-ments per 10CFR50.
. 2.0 CRITERIA 2.1 Piping Operability Criteria ' The piping analysis shall be in accordance with ASME,
Section III NC-3600 service level D limits (Ref. 1).
' The design loading conditions to be applied in tha analysis shall include the DBE earthquake.
Following is the pipe stress criteria for justifying continued operation of the plant: (Fgp + SWT * DBE $ 2.0 Sy] (Ref. 1 equation 9) Where: S = Longitudinal Pressure Stress gp S = Dead Weight Stress WT SDBE = Stresses Resulting From Design Basis Earthquake S = Material Yield Stress (Reference 1 y Appendices) Code Case N-411 allows for increased damping values, independent of pipe diameter, for seismic analysis.
Therefore, increased damping values, in accordance with reference 2, will be acceptable when performing these analyses to meet operability.
Should the piping a stress analysis exceed the value of 2.0 Sy, or pipe supports do not meet their operable limits (see Sect.
2.2), then additional iterstive analysis of the piping Page 3 - _ - . -. _ - -.. .--. . - - - - .. - - - - - - _ . .
. . ... .. . may be required.
The iterative analysis may uma ths knowledge that a support is not capable of withstand-f . ing the loads, and can be removed from the analysis.
- Where feasible, the actual support stiffness may be included in the iterative analysis, along with other
refinements.
' For cases where piping secondary stresses are determined to exceed USAR allowables, a specific case by case approach will be used to determine interim operability.
2.2 Pipe Support & Hanger operability Criteria As a first step in evaluating the support, a linear elastic analysis method will be used to determine the stress in the support members.
In addition to the load 3ng in Section 2.1, the support loads must include pipe thermal loads and results from free end displace-ment and anchor motion.
Supports will be analyzed . using the allowables listed below to meet operability requirements.
Structural Steel 1.20 Sy but Tension F = t < 0.70 Su 1.20 Sy but Bending F = b < 0.70 Su 0.72 Sy but Shoar F = y < 0.42 Su < F but not to exceed 2/3 Per Compression F, t Combined Stress For axial compression and bending or axial tension and bending, use AISC 1.6., (Ref. 6) 1.0 Sy Web Crippling = 0.42 Su (of weld material) Weld Stress F = y Anchor Bolts Use Factor of Safety of 2 against ultimate tension and shear values.
Snubbers Hydraulic: Load < manufacturers one time load capacity.
Movement < total travel Springs Load within catalog range without bottoming out Struts FS = 2 and < 2/3 Per Page 4
. .- .- o All remaining Usa manufactursrs publishsd fcultsd Catalog Items load rating.
Where level D allow-ables are not given, and the factor . of safety is specified in the catalog, use design allowables but with FS = 2.
(Typical catalog FS = 5, therefore use 2.5 x catalog capacity).
Where: F = Al wa e ensile Stress t F = Allowable Bending Stress b F = Allowable Shear Stress y F, = Allowable Axial Compressive Stress F = Allowable Weld Stress y . Per = Maximum Strength of Axially Loaded Compression Member Sy = Specified Minimum Field Strength at Temperature (See Note 1) 't Su = Specified Minimum Tensile Strength Temperature FS = Factor of Safety < NOTE 1: Actual yield strength may be used where CMTR's are available for the material.
It: a support fails using the linear elastic method, then , a more refined analysis may be performed using plastic analysis techniques.
The plastic analysis will follow l ' the design rules of ASME Section III, Appendix F, (Ref. 1).
3.0 CONCLUSION If the above criteria cannot be met, reportability per 10 CFR l 50 must be evaluated and system operability requirements per Plant Technical Specifications must be evaluated and appro-priate actions taken.
4.0 REFERENCES 1.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Codes, Section III, 1983 Edition, through Winter 1985 Addenda.
2.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Codes, Case N-411, Dated 9/17/84.
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NRC-IE Bulletin 79-02, "Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts," Revision No. 1 ' (Supplement No. 1), Dated 8/20/79.
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USAS B31.1.0-1967, Power Piping Code.
5.
Updated Safety Evaluation Report for PINOP.
6.
"Manual of Steel Construction," American Institute of Steel Construction, Inc., Eighth Edition, 1980.
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