ML20098G884
| ML20098G884 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/02/1984 |
| From: | Faires W, Manz R COMMONWEALTH EDISON CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20098G869 | List: |
| References | |
| OL, NUDOCS 8410050652 | |
| Download: ML20098G884 (12) | |
Text
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. ATTACHMENT 12 l
. 'i ,,
i) ED UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 04 Cg7,y Al1:25 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD.
"9Cd[,[j,g!j a qj@Jf.' ry -
In.the Matter of )
) -
COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL
) 50-455 OL (Byron Nuclear Power Station, )
Units 1 & 2) )
JOINT STATEMENT OF ROBERT W. MANZ AND WADE FAIRES Introduction (Mr. Manz and Mr. Faires)
- 1. This statement addresses Findings 3-11 through 13-17 of the NRC Staff's Integrated Design Inspection (IDI)
-Report dated September 30, 1983, which was prepared by the NRC's. Office of Inspection and Enforcement. Those findings concern piping design work that was performed by Westinghouse Electric Corporation. The IDI Team found certain deficiencies in the sample of this work that it inspected. Westinghouse
' design engineers reviewed these portions of the IDI Report and submitted draft responses to Commonwealth Edison Company (CECO). CECO personnel reviewed these drafts and submitted
-revised responses to the NRC IDI Team on December 30, 1983.
This statement again addresses the specific deficiencies perceived by the IDI Team.
Finding 3-11: Pipe Damping Value (,Mr. Manz)
- 2. The IDI Team concluded that Westinghouse's use of 8410050652 841002 DRADOCK05000gg
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a 4' percent' damping value in the safe shutdown earthquake (SSE) analysis:of piping of 12 inch diameter and greater is
. appropriate for Westinghouse supplied equipment (i.e., the reactor coolant loop) but is inconsistent with the FSAR as it.is applied to the SSE analysis of the balance-of-plant systems.
O. The Westinghouse. criteria establish a 4 percent
. damping value for SSE analyses of the Reactor Coolant System
. Piping which includes connected piping 12 inches or greater in diameter, a 3 percent damping value for piping 12 inches and greater in diameter which is not connected the Reactor Coolant System,-and a 2 percent damping value for piping of less than~12' inches in diameter. In light of the concern Lexpressed in the IDI, the FSAR has been amended to clarify the use of the 4 percent damping factor. The table in the FSAR applicable to Westinghouse now prescribes the use of a 4 percent damping value for SSE analysis of the reactor coolant loop and piping subsystems of 12 inch diameter and:
greater which are attached directly to the reactor coolant loop. For other piping subsystems 12 inches and greater, the We stinghouse table now prescribes use of a 3 percent
. damping'value. To ensure that the damping values used were
,j consistent with the revised table, Westinghouse reviewed the r
SSE' analyses of all 521 subsystems within its scope of analysis. The SSE analyses of only 2 subsystems was found
, , to have used the inappropriate damping values. These two subsystems.were reevaluated and the existing design was found to be' acceptable.
Finding 3-12: Spray Additive Tank Calculation (Mr. Manz)
- 4. Finding 3-12 indicates that there is poor document control because a preliminary calculation was not noted as preliminary, voided or revised when final evaluations were performed.
- 5. As a general practice, Westinghouse routinely marks a calculation as revised or voided when a subsequent evaluation is performed. However, document control procedures of the Westinghouse Quality Assurance Program did not require any such action. To assure that our practice of marking calculations as revised or voided is consistently applied, the Westinghouse Procedure and Guideline. Manual for the Byron Project has been revised to direct such action when appropriate.
- 6. In the instance cited in-Finding 3-le, the final calculations appear in the same section of the analysis notebook as the initial calculations. Thus, the initial calculation was readily discernable as outdated. None-theless, that calculation has since been marked as superseded.
- 7. In any event, the failure to mark the initial i
calculation as preliminary would not have any effect on design because.of checks in the iterative design process.
.These steps include independent engineering review of calculation- internal audits against QA procedures, and review of analytical results at 70 percent and 100 percent
'as-built stages.
Finding 3-13: Pipe Movement in Relation to Gap Clearance (Mr. Manz)
- 8. Finding 3-13 asserts that the calculations that
'were performed to confirm the piping displacements at two wall. penetrations did not include the calculation of vector
' displacement. Therefore, the IDI Team concluded that the calculations failed to verify the modeling assumption of no
' interference between'the piping and the penetration walls.
.9. We disagree with the IDI Team's conclusion that
-the calculations were insufficient to verify the modeling assumption. The pipe displacements were all less than two-thirds of the~ gap size. Thus, the analyst who performed
.the piping' lateral displacement calculations was able to
. determine by inspection that the last step of calculating the vector. displacement was unnecessary since it was obvious
.'from the magnitude of the displacements that the vector sur
-would beusmaller.than the wall penetration gap. Therefore, the assumptions used in the model were verified adequately based on the calculation as originally performed. A sub-
m v sequent-calculation of the vector displacement confirmed the analyst's judgment that the vector sum were less than the wall penetration gap.
- 10. - At the time the IDI Team made Finding 3-13, Westinghouse was already in the process of developing a
, systematic review of all penetrations. For this review, we developed a standardized calculation form which includes a calculation of the vector sum. This systematic review has been implemented and has been completed.
Finding 3-14: Branch Line Analysis (Mr. Manz)
- 11. The IDI team reviewed the calculations that evaluated two small branch lines off the main piping system which were not included in the mathematical model. The team apparently did not feel that the methodology used in these calculations.was sufficiently detailed to support seismic qualification of the lines.
- 12. We disagree with the conclusion that the seismic qualification of these lines is not adequate.- . In this case the analyst chose to use a simplified, conservative method for evaluating.these lines instead of a more detailed calculation similar to the IDI recommendations. Use of such simplified methods is a typical engineering practice. If such a conservative calculation yields unacceptable results, a more detailed calculation is performed. To demonstrate L:
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's- 4 the conservatism of the initial. analysis, a more detailed evaluation was performed as recommended by the IDI Team.
This resulted'in calculated stresses approximately a factor of 3 lower than the simplified results.
- 13. . Typically, these branch lines are included as part of the main piping models. Separate hand calculations are performed in cases when~they are not included with a main piping model. As demonstrated above, the hand calculations
-which.were performed prior to the IDI result in conservative values of pipe stress. Furthermore, to ensure unformity in this type of evaluation, standard methods similar to those <
recommended by the IDI are now incorporated into the West-inghouse Procedure and Guideline Manual for the Byron Project.
Finding 3-15: Hanger Dimension LMr. Faires)
- 14. The.IDI~ Team found that.there was a discrepancy between the piping isometric drawings and the support drawings involving the location of two pipe supports.
- 15. Discrepancies between Hunter piping isometric drawings and Westinghouse support designs do occur in the-design process. However, these are expected and are resolved systematically through the 100% as-built program.
- 16. -When the construction work is 100% complete, CECO Lu
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- i o issues 100% as-built packages to' Westinghouse, containing the isometric drawings with the as-built support locations identified. Westinghouse site personnel review all 100 percent as-built packages to verify the correctness of all documentation. Information on support drawings showing locations of the supports is compared to the. corresponding information on the as-built isometric drawings. CECO is notified of any deviations or discrepancies in the as-built dimensions and proceeds with further verification and, if necessary, corrective action. This process had not been completed at the time of the IDI for the hangers in question, but has since been completed.
- 17. For the specific instance cited in Finding 3-15, field personnel verified the supports' location. In both cases,.the information on.the Westinghouse design documents was correct. There were no violations in documentation nor any design deficiencies. Commonwealth Edison has.been notified.of the discrepancy on the Hunter Isometric drawing and the isometric drawing has been reissued.
Finding 3-16: U-Bolt Analysis (Mr. Faires)
- 18. The IDI team found that the calculations for a particular support revealed that a cantilever angle section and a.U-bolt were overstressed. A letter to the Westinghouse site personnel indicated: that a modification of the angle J
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- was required, but not a corresponding modification of the U-bolt. A decision not-to modify the U-bolt was not supported by calculations at the time.of the IDI.
-19. The support referred to in the finding required a modification because it was overstressed. For all supports that require modifications, support calculations are com-
- _pletely verifie'd in the field before the modification is implemented. In addition, after the modification is com-pleted, CECO issues a revised 100 percent as-built package to Westinghouse and the calculations are completely reverified again by the design engineer. In the event the field Engineering Change Notice (ECN) does not resolve the design deficiencies the process is repeated. This verification cycle in the design process with regard to the particular
- support identified had not been completed at the time of the IDI.
- 20. At the time of the inspection, support verification Z
had indicated an overstressed condition of both the~ angle and the U-bolt. As the IDI Team pointed out, the letter informing the field personnel of the need for support modifi-cations reflected the overstressing of the angle but not the coverstressing of the U-bolt. However, for supports indicated las needing_ modifications, a complete calculation reverification
'is done and an ECN is then issued. In this instance, this
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, included the U-bolt as well as the angle. Once the calculation review is complete, the field engineer decides whether it is more feasible to modify the existing support configuration or to design a new.one. At the time of the Staff's inspection it was believed that the design of the U-bolt was appropriately conservative. However, at that time the calculation verification
-had not yet been performed and the ECN had not yet been issued. After'the calculation was performed, the field engineer determined that the most feasible modification
.would be to replace the angle with a tube steel section and to replace the U-bolt with a frame of angle sections around the pipe. A field ECN reflected these modifications. After the modification.was completed, the design engineer confirmed that-it incorporated all necessary changes to confirm structural adequacy. Finally, a review of the U-bolt calculation has been completed and documented since the IDI, and no changes
..in the U-bolt would have been required.
' Finding 3-17: Piping Response for Changed Span Length (Mr. Manz)
- 21. The IDI Team concluded that the use of hand calculations, rather than computer reanalysis, to evaluate the' effects on the piping system of the relocation of a 4
particular support was not adequate for seismic qualifica-
-tion.
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- 2. The method used to evaluate the effect of the support relocation is fully acceptable. The assessment included load increases of 40 pe.rcent on the moved vertical support (004) as well as the two adjacent vertical supports (001 and 005). Although a precise prediction of the effect of this relocation would have required a computer reanalysis,
- the use of a hand calculation was deemed appropriate in this case in view of the limited changes from the analyzed configuration, the large margins in the pipe stress, tank nozzle loads, and valve accelerations, the conservative nature of the hand calculations, and engineering judgment based upon experience with such analyses.
- 23. Nonetheless, to verify the adequacy of the assessment made in this reconciliation process, an SSE computer reanalysis was conducted with the support relocation included. This resulted in SSE support load changes of -15 percent, +20 percent, and +33 percent from the original analysis for supports 001, 004, and 005, respectively. Therefore, the
+40 percent increase which was estimated for all three
- upports was indeed conservative. Reanalysis of the piping subsystem, with the support location deviation included, demonstrated that the estimated support load increases were conservative, and justified the simplified analysis used in the original evaluation.
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( 5ibfD UNITED STATES OF AMERICA SI ~S NUCLEAR REGULATORY COMMISSION 47 *D C ,_
BEFORETHEATOMIC-SAFETYANDLICENSINGBb II -t In.the Matter of )
)
COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454 OL
) 50-455 OL (Byron Nuclear Power Station, )
Units 1 & 2) }
CERTIFICATE OF SERVICE The undersigned, one of the attorneys for Common-wedith Edison Company, certifies that he filed the original and two copies of the attached " COMMONWEALTH EDISON COMPANY'S ANSWER TO INTERVENORS' MOTION TO REOPEN THE RECORD" with the Secretary of the Nuclear Regulatory Commission and served copies on the persons and at the cddresses shown on the attached service list. Unless otherwise noted on the Service List, service on the Secretary and all parties was made by deposit in the U.S. Mail, first-class postage prepaid, this 2nd day of October, 1984..
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,One of/the Attorney,s for Commonwealth Edison Company ISHAM, LINCOLN & BEALE Three First National Plaza Chicago, Illinois 60602 (312) 558-7500 L
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SERVICS LISm
-COMMONWEALTH EDISON COMPANY -- BYRON STATION Docket Nos. 50-454 and 50-455 CIvan W. Smith, Chairman Mrs. Patricia Morrison Administrative Judge 5568 Thunderidge Drive Atomic Safety and Licensing Board Rockford, Illinois 61107 U.S. Nuclear Regulatory Commission 4350 East West Highway *Mr. Stephen Lewis West Tower - Room 428 Office of the Executive Bethesda, Maryland 20814 Legal Director CDr. A. Dixon Callihan U.S. Nuclear Regulatory Commission 7735 Old Georgetown Road Administrative Judge Room 9604 102 Oak Lane Bethesda, Maryland Oak Ridge, Tennessee 20814 37830 CDr. Richard F. Cole Atomic Safety and Licensing Board Panel
. Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Washington, D.C. 20555 U.S. Nuclear Regulatory Commission 4350 East West Highway Atomic Safety and Licensing West _ Tower - Room 428 Appeal Board Panel Lethesda,-Maryland -20814 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i
Joseph-Gallo, Esq.
.Icham, Lincoln & Beale Docketing and Service Section Suite 840' 1120 Connecticut Avenue, N.W. Office of the Secretary Washington, D.C. U.5. Nuclear Regulatory Commission 20036 Washington, D.C. 20555 R;gion III Howard A. Wilbur U.S. Nuclear Regulatory Commission Administrative Judge s Office of Inspection & Enforcement Atomic Safety and Licensing Board 799 Roosevelt Road U.S. Nuclear Regulatory Commission Glen Ellyn, Illinois 60137 Washington, D.C. 20555 Ms. Betty Johnson Dr. Reginald L. Gotchy 1907 Stratford Lane Administrative Judge.
Rockford, Illinois 61107 Atomic Safety and Licensing M3. Diane Chavez*
- 3. Appeal Board S. Nuclear e
Regulatory Commission SAFE 528 Gregory Street 4350 East West Highway Rockford, Illinois 61108 3rd Floor - call 492-7663 Bethesda, Maryland 20814 Dr. Bruce von'Zellen Alan S. Rosenthal, Chairman Department of Biological Sciences Atomic Safety and Licensing i
Northern Illinois University Appeal Board D Kalb, Illinois 60115
' U.S. Nuclear Regulatory Commission
_ CCDouglas W. Cassel, Jr. 4350 East West Highway BPI 3rd Floor - call 492-7663 109 N. Dearborn St.; Suite 1300 Bethesda, Maryland 20814 Chicago, Illinois 60602
- Served by Federal Express
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