ML19344D745
| ML19344D745 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee, Beaver Valley |
| Issue date: | 01/18/1979 |
| From: | Keimig R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19344D724 | List: |
| References | |
| NUDOCS 8004280090 | |
| Download: ML19344D745 (95) | |
Text
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UNITED STATES
'If NUCLEAR REGULATORY COMMISSION f*( d
,h REGION I se s
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$31 PARK AVENUE h.4-af KING OF PRUSSIA. PENNSYl.VANI A 19404 18 JAN 1979 MEMORANDUM FOR:
E. L. Jordan, Assistant Director for Technical Programs, gROI,IE THRU:
Y" J. Brunner, Chief, Reactor Operations and Nuc ear k
Support Branch, RI i
FROM:
R. R. Keimig, Chief, Reactor Projects Section #1, RO&NS Branch, RI
SUBJECT:
BEAVER VALLEY POWER STATION UNIT 1 (DOCKET NO. 50-334)
SAFETY INJECTION PIPING STRESS ANALYSIS - REQUEST FOR.
ADDITIONAL TECHNICAL REVIEW (AITS NO. F14741H2)
The licensee submitted LER 78-53 to Region I on October 27, 1978 to report errors identified by the licensee's A/E in the piping stress analysis perfomed for safety injection piping % side the containment.
The LER was supplemented by additional reports on November 9 and Decem-ber 6,1978. The reported errors were discovered during an A/E review of the stress calculations after receipt of new information from the l
NSSS vendor which corrected the weights of check valves installed in the i
injection lines.
i During our review of these activities, we noted that the A/E had applied two essentially equivalent computer programs, PIPESTRESS and NUPIPE, to the reanalyses of three of the one hundred and three lines reviewed.
j The results of these analyses indicated an overstress condition in two lines requiring modification of the piping supports. Additionally, the results obtained from NUPIPE and PIPESTRESS for the same lines <tiffered significantly with each other and with the original design hand calcu-lations. Attempting to reconcile these differences, we discussed the analysis results with the A/E and infomally with NRR: DOR Engineering Branch (SteveHosford). Our discussion with NRR indicated that the magnitude of the differences in results could not be reconciled by only the differences in calculational methods used by the computer programs as was indicated by the A/E. While the two lines affected have been modified to accomodate the differences, we are concerned that, if the differences in supposedly equivalent analysis methods cannot be tech-nic=.lly reconciled, nonconservative results may have been preiously applied to pipe supports designed for safety related piping.
CONTACT:
D. Beckman (488-1268) 8 004280 MO
[
Memorandum for E. L. Jordan 2
18 JAN 1979 We request that the information contained in the attachment to this i
memorandum be forwarded to the NRR: DOR Engineering Branch for their i
review. We ' recommend that they assess the information provided and establish whether further review or action appears necessary. Our concerns include.
1.
Reconciliation of the differing analysis results to assure that the design methods used are neither incorrect nor unconservative; and, 2.
The need for further licensee review of piping potentially affected by any incorrect or nonconservative calculations.
Our discussions with the A/E, Stone and Webster, were conducted with Mr.
James Cumisky (Boston Office, 617-973-5685).
Mr. Cumisky'has indicated a willingness to provide additional information or answer questions re-garding this matter.
[
j
,g_ R. R. Keimig, Chief I"
Reactor Projects Section No.*1 RO&NS Branch
Attachment:
Chronology of A/E Reviews / Reanalyses bec w/attachmer.t:
R. Keimig D. Beckman v' Branch File e
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CllRON0 LOGY OF A/E REVIEWS / REANALYSIS The piping stress calculation errors reported in Beaver Valley Power Station Unit 1 LER No. 78-53 and its supplements were apparently the result of the misapplication of original hand calculations (described in the BVPS FSAR, Section 82.1.1.9).
In their analyses of these errors, the licensee's architect / engineer (Stone & Webster) applied two computer codes (PIPESTRESS and NUPIPE) to each of the three lines discussed below. While an additional 100 lines were reviewed in whole or part, using only the PIPESTRESS program and were found acceptable, these three lines were analyzed with both and displayed a significant difference between the NUPIP,E and PIPESTRESS results.
The stress results of the two computer analyses are provided below for the three lines in question.
Additionally, sketches showing piping arrarigement are attached and further discussed below.
Line 6"-SI-20:
NUPIPE Calculated Stress Allowable Stress 9 180 F 0
Oload + S1p 6193 16600 D10ad + 31p + OBET 31621 24900 Dioad + 31p + OBET + THERM 33433 49800 Dioad + SIP + DBET 32503 29880 i
PIPESTRESS Dioad
- Slp 7449 16700 Dload
- S1p DBET + TilERM 37000 44375 NOTE: All values ir psi.
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Line 6"-SI-/3:
NUPIPE NUPIPE Run PIPESTRESS Run Allowable Run with New with New Stress As Restraint PIPESTRESS Run Restraint on 9 1800F Built on 6"-SI-72 As Built 6"-SI-72 Dload + Slp 16,600 9,583 9,945 10,385 10,231 Dload + S1p + OBET--
24,900
- 206,238 18,268
~*29,899-19,488 D10ad + Slp + OBET + THERM 49,800
- 235,994 40,949
- 36,439 46,755 D10ad
- Slp + DBET 29,880
- 218 689 21,463
- 32,884 20,776 Line 6"-SI-72 NUPIPE NUPIPE Run PIPESTRESS Run Allowable Run with New with New Stress As Restraint PIPESTRESS Run Restraint on 9 180 F Built on 6"-SI-72 As Built 6*-SI-72 Dioad + S1p 16,600 11,798 11,780 11,782 11,722 Dioad + Slp + OBET' 24,900
- 109,652 15,616
- 167,881 15,769 Dioad + Slp + OBET + TilERM 49,300
- 164,778 40,834
- 202,341 40,240 Dload + S1p + DDET 29,880
- 113,609 17,306
- 177,982 17,283 The stress values indicated by asterisk (*) in the above tablulation show the considerable variation between the PIPESTRESS and NUPIPE results.
In particular, the stresses annotated for Line 6"-SI-73 for the "as built" condition show differences of a factor of 6.5-6.8 with the PIPESTRESS program appearin9 to be the less conservative.
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Although modifications were made,to line SI-72 which reduced the stress to allowable values in both it and SI-72, the difference in the respec-tive as built results raised the question of the accuracy arn tor conser-vatism of the PIPESTRESS program as applied to all the 103 lines analyzed.
The following information, extracted from the S&W report to the licensee l
on this matter discusses the three lines involved and the evaluation of each line's stress analysis results.
The figures mentioned are attached.
"The model run in NUPIPE and in PIPESTRESS are geometrically similar, i
however, the mass distribution and support stiffness are different.
Further, the method of force sumation (intramodal) is different l
between PIPESTRESS and NUPIPE.
NUPIPE utilizes more conservative techniques for intramodal combinations of generalized loadings.
I These newer techniques arose following establishment of BV1 Design Criteria....
The PIPESTRESS methods used were accepted dynamic analysis tech-j niques for Beaver Valley 1 generation plants and are the basis for all computerized Category 1 pipe stress analysis done for Beaver Valley Unit 1.
6 Figure IA gives the hanger location and peak local stress vs.
allowable from the NUPIPE model for line 6"-SI-20.
Figure IB is a i
sketch of the hangers-pipe (SI-A-60/6"-SI-20) interface showing the l
overstressed area. The table attached to this sketch identifies
)
the differences in the hanger attachment loads resulting from the two different computer models.
Based on the above, the Safety Injection line 6"-SI-20 is acceptable as designed."
{
i Discussions with S&W indicated that this acceptance was based upon ob-taining satisfactory stress values using the techniques comitted to for j
the original design. The differences in NUPIPE and PIPESTRESS results were attributed to NUPIPE using root mean square values for individual and total loads where the PIPESTRESS program algebraically combines individual loads.
The S&W report continues:
" Figures II A&B give the most highly stressed hanger, location and stresses resulting from the NUPIPE and PIPESTRESS runs on 6"-SI-73.
Based on this data, 6"-SI-73 is acceptable as designed.
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Figures IIIA and B give the most highly stressed hanger, location and stresses resulting from the NUPIPE and PIPESTRESS runs for line 6"-SI-72.
Based on this data, modification to hanger VC-LC-H306A j
(Figure IV) and the addition of hanger LSS-H is required to be added to line 6"-SI-72".
i The reasons given by S&W for the differences between the NUPIPE and PIPESTRESS results were the same as those for 6"-SI-20.
In these cases, the addition of the supports on 6"-SI-72 also affected the response of 6"-SI-73 (as noted in the far right hand column of the previous tabulation) resulting in both the NUPIPE and PIPESTRESS values being within code allowable values.
In conclusion, the S&W report states:
"We believe that the remainder of the containment annulus piping is acceptable based on the fact that the pipe stress analysis section has completed a review of seismic piping shown on the BP-3 series drawings (annulus piping). The review was limited to piping 2 1/2" OD to 6" OD because of the possibility that these sizes may have been analyzed by the ' chart' method.
The attached tabulation i
(Table I) contains all the seismic lines falling between 21/2" and l
6".. This tabulation contains 103 seismic lines of which 55 were l
reviewed and found acceptable.
A large portion of this piping was analyzed during the "as built review" (in 1974) using computer program PIPESTRESS.
PIPESTRESS results are available for all or portions of 45 of the (remaining) tabulated lines and are acceptable".
On the basis of discussions with NRR: DOR Engineering Branch personnel, we consider additional technical review is necessary to reconcile the differences in the PIPESTRESS and NUPIPE results.
These discussions in-dicated that the differences appeared to be incongruous in light of their knowledge of the two programs.
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3 a cc - 251-151-43 3 " cc - 252-151-0 3 a c3 - 106-153v-42 3 " cc - 255151-0 3 a cs - 106-152-42 3 " cc W151-0 3 a CH - 125-1503 A " cc - 164-151-43
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NRCscF+56 U.S. NUCLEAR REGULATCRY COMMIS$1CN
't ACTION ITEM CONTROL FORM INITIATING CFFICE f
TRACK NUMBER
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ACTION OFFICE PERSON ASSIGNED ESTA8USHED COMPLETION DATE e4ri REGIONAL l
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ACTION TAKEN ACTION DATE S UT ACION TRANS.
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RETURNED TO INITIATING OFFICE UPON COMPLETION
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Duquesne Ugat
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.ss sam Avenue I
uron..'ennsylvania DUQUESNE LIGHT COMPANY
.9 Beaver Valley Power Station Post Office Box 4
(
Shippingport, PA 15077 i
November 9, 1978 BVPS: JAW:609 l
Beaver Valley Power Station, Unit No. 1
(
Docket No. 50-334, License No. DPR-66 l
LER 78-53/OlT l
Mr. B. H. Grier, Director of Regulation United States Nuclear Regulatory Commission
[
Region 1 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Mr. Grier:
In accordance with Technical Specification 6.9.1.8.1, the following interim report is submitted.
Further investigation by the Beaver Valley architect-engineer into I
the discovery of apparent errors in the stress analysis for safety injection piping inside containment has determined that the original report was erroneous. The line stresses were thought to have been hand calculated only, when in fact they were subsequently computer calculated and found acceptable.
A full report on the situation is being prepared by the architect-engineer for Duquesn'e Light. A follow-up report will be submitted after a review of the architect-engineer report by Duquesne Light engineers
)
and the Onsite Safety Committee.
~
Very truly yours, bM I
hJ.A.Werling Superintendent l
781111 0.2./7
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- 3. H. Grier ember 9, 1978
$VPS: JAW:609 Page 2 i
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cc: Director, Office of Management Information & Program Control United States Nuclear Regulatory Commission Washington, D. C.
20555 E. A. Reeves, BVPS Licensing Project Manager United States Nuclear Regulatory Commission Washington, D. C.
20555 f
W. J. Raymond, Nuclear Regulatory Commission, King of Prussia, PA l
-i G. A. Olson, Secretary, Prime Movers Committee - EEI l
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[.7,urgh Pennspransa DUQUESNE LIGHT COMPANY Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077 December 6, 1978 BVPS: JAW:615 Beaver Valley Power Statien, Unit No. 1 Docket No. 50-334, License No. DPR-66 LER 78-53/01T-0 Mr. B. H. Grier, Director of Regulation United States Nuclear Regulatory Commission Region 1 631 Park Avenue King of Pruscia, Pennsylvania 19406
Dear Mr. Grier:
In accordance with Appendix A, Beaver Valley Technical Specifications, the following Licensee Event Report is submitted:
LER 78-53/01T-0, Technical Specification 6.9.1.8.1, Error In Safety Injection System Piping Stress Analysis.
Very truly yours, J. A. Werling Superintendent At'.achment a
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Mr. 3. H. Grier l
December 6, 1978
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BVPS: JAW:615
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I, f;l cc: Director, Office of Management Information & Program Control United States Nuclear Regulatory Commission Washington, D. C.
20555 E. A. Reeves, BVPS Licensing Project Manager /
i United States Nuclear Regulatory Commission
- i. '
Washington, D. C. 20555 l
W. J. Raymond, Nuclear Regulatory Commission, King of Prussia, PA 3i i.
- l G. A. Olson, Secretary, Prime Movers Committee - EEI l!
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75 REPCMT QATE So EVENT DESCRIPTICN AND PRC5ABLE CONSECUENCES h l
lo i2l l As a result of an error in vendor-suoplied data for the veicht of check valves in i
1 A I the SI System, a review of SI piping stress calculations was oerforned. Uoon I
1 lo6 tl completion of-the review, an error in the original einine stress analvsis for one I o t s i I SI line had been discovered. A further review of containment annulus nipine was I o $ s I l conducted with the review limited to oicine 2 1/2 to 6 inch OD beenuse these s4-es I
[TTT} } may have been ans17:ed by the chart nethed. This piping is secentable as inctslied.!
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.o CAUSE CESCRIPTICN AND CORRECTIVE ACTIONS 27 l i i o l l The error has been attributed to the misapplication of the s1=olified analvsis I
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