ML20083C973
| ML20083C973 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/21/1983 |
| From: | Baxter T SHAW, PITTMAN, POTTS & TROWBRIDGE, UNION ELECTRIC CO. |
| To: | Chackes K CHACKES & HOARE |
| References | |
| ISSUANCES-OL, NUDOCS 8312270150 | |
| Download: ML20083C973 (11) | |
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weette e osatCT osak NvMete December 21, 1983 822-1090 Kenneth M. Chackes, Esquire Chackes and Hoare 314 N. Broadway St. Louis, Missouri 63102 In the Matter of Union E13ctric Company (Callaway Plant, Unit 1)
Docket No. STN 50-493 OL
Dear Mr. Chackes:
In my letter of December 6, 1983, to the Atomic Safety and Licensing Appeal Board, I reported that Applicant Union Electric Company would dist; rib'Ite its reply to the NRC Staff's letter of November 16, 1983, on the Integrated Design Inspection report for the Callaway Plant.
Please find enclosed a copy of that reply in the form of a letter (ULNRC-706) and attachment, dated December 19, 1983, from Union Electric Company to NRC Region III.
Sincerely,
/
IW Thomas A. Baxter Counsel for Applicant TAB:jah Enclosure 8312270150 831221 PDR ADOCK 05000483 C
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r SHAW, PITTMAN, POTTS & TROW; RIDGE A 342748 ASMe# OF #8BOrtteacceA4 COepomATiOng Kenneth M. Chackes, Esquire December 21, 1983 Page Two
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cc:
Alan S. Rosenthal, Esquire Gary J. Edles, Esquire Dr. Reginald L. Gotchy James P. Gleason, Esquire Mr. Glenn O. Bright Dr. Jerry R. Kline Robert G. Perlis, Esquire Docketing and Service Section Joseph E. Birk, Esquire A. Scott Cauger, Esquire Mr. John G. Reed Eric A. Eisen, Esquire
_____._______._______._______...._.______1
UNION ELECTRIC COM PANY 9 908 GR ATIOT STREET Sr. Louis, Missouri December 19, 1983
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Mr. C. E. Norelius, Director Division of Project and Resident Programs U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Norelius:
ULNRC-706 SECOND RESPONSE TO NRC INTEGRATED DESIGN INSPECTION (IDI) REPORT Ref:
Your letter to D. F. Schnell dated November 16, 1983 The referenced letter transmitted a request for additional information on seven items previously discussed in our June 15, 1983 letter to Region III.
The attachment to this letter repeats the NRC's concerns and provides the requested additional information.
The referenced letter suggested a meeting in the near future to discuss "all open IDI findings and unresolved items".
It is our hope that such a meeting will result in resolution of the entire IDI report.
We are willing to meet at any time.
Sincerely, C'
Donald F. Schnell ACP/lw cc:
John T. Collins, Region IV Richard C. DeYoung, Director, OIE Hg.
G.
L. Koester, KG&E D. T. McPhee, KCP&L NRC Resident Inspector, Callaway Plant H. M. Wescott, NRC Region III Missouri Public Service Commission T. A. Baxter, Shaw, Pittman, Potts & Trowbridge N. A. Petrick, SNUPPS B.
L. Meyers, Bechtel JL
' STATE OF MISSOURI )
)
Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By h
D6nald F. Schnell Vice President Nuclear SUBSCRIBED and sworn to before me this
/ 9/4 day of december, 1983
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[ VARY C. GUINN NOTARY PU'500-STATE OF MISSOURI ST. LOU:S CITY WY COMulS$10N EAPiRES JUNE 15.1936 i
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cc:
Glenn L. Koester Vice President Operations Kansas Gas 6 Electric P.O. Box 208 Wichita, Kansas 67201 Donald T. McPhee Vice President Kansas City Power and Light Company 1330 Baltimore Avenue Kansas City, Missouri 64141 Gerald Charnoff, Esq.
Shaw, Pittman, Potts 6 Trowbridge 1800 M. Street, N.W.
Washington, D.C.
20036 Nicholas A. Petrick Executive Director SNUPPS 5 Choke Cherry Road Rockville, Maryland 20850 John H. Neisler Callaway Resident Of fice U.S. Nuclear Regulatory Commission RR$1 Steedman, Missouri 65077 Jim Konklin Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region II 799 Roosevelt Road Glen Ellyn, Illinois 60137 n
Pcga 1 of 6 NRC COMMENT ON THE RESPONSE TO FINDING 3-1 (11-18-83)
Your response does not appear to address the concern.
The concern involved the vector decom of the seismic building response. position of a single direction Based on the discussion in the ME 101 Users Manual for skewed supports, the computer program multiplies the building displacement by the support cosine vector to determine the movement along the support direction for input to the static displacement perpendicular to the support direction.
It should be noted that, since this is vector decomposition of a single component of the building displacement, these components are perfectly correlated.
The specific concern involves the method of solution used by the computer program to determine forces and moments in the piping system.
If the program resolves the imposed displacement along the support axis back into the global system using the cosine vectors, the imposed global displacement will be less than the original building displacement and a ficticious displacement will be added to the perpendicular global direction.
This could occur if the original resolution of the displacement perpendicular to the support axis was lost as stated above.
You are requested to provide a detailed description of the method used by the computer to formulate the forces and moments and, if the method involves simplifying assumptions such as disregarding a displacement component, you are requested to describe the conservatism or lack of conservatism in the approach.
Alternatively, an acceptable approach to resolution of the concern would be to run simplified test cases to demonstrate that displacement output at the support point in the global directions matches the building input motion at the support point.
SECOND RESPONSE (12-16-83)
In order to demonstrate that skewed supports are properly treated in the seismic anchor movement analysis of Bechtel's ME 101 program, a sample problem was run to compare input and output values of seismic anchor movements - this method is in accordance
' with the alternative approach suggested by the NRC.
The sample problem contained 3 skewed supports which addressed the range of l
conditions necessary to insure that no data has been misplaced by l
ME 101.
l The results of this investigation are documented in Bechtel Study PDE-80-08.
An exact correlation of input and output displacement values was demonstrated.
The displacement output at the support point in the global directions matched the building input motion at the support point.
1 On this basis, it was concluded that the ME 101 non-conservatism postulated by the NRC does not exist.
Page 2 of 6 NRC COMMENT ON THE RESPONSE TO UNRESOLVED ITEM 3-5 (11-18-83)
Your response does not adequately resolve the item.
Test data on components as cited in NUREG-0307 have demonstrated that some components such as welding tees have moment capacities equal to or greater than the attached straight pipe.
The reduction procedure in TB-Oli is not conservative for all components based on actual test results cited in NUREG-0307.
Based on your response that the procedure was only used at elbows for Callaway, the design of those anchors should be adequate.
However, the procedure should be modified to reflect the actual data for other types of components such as welding tees.
You are requested to describe your plans for modifying the procedure.
SECOND RESONSE (12-16-83)
Bechtel Procedure TB-Oll will be revised to limit the application of stress intensification factor (SIF) to reduce the collapse piping loads at elbows only.
A survey of all Bechtel projects using TB-Oll has confirmed that the SIF reduction provision has been used for elbows only.
The revised procedure will be issued and distributed. no later than December 16, 1983.
NRC COMMENT ON RESPONSE TO UNRESOLVED ITEM 3-6 (11-18-83)
Your response does not adequately resolve the item.
The response addressed the question of the general stiffness of major structural elements such as concrete shear walls.
However, you did not address the specific question of the stiffness at the support change 2FC-1191-MH or the specific concern of I-beam members loaded in torsion.
The issue of loading structural I-beams in torsion with pipe supports has been found to be a problem at some facilities because the structural designers were not aware of support requirements.
You are requested to describe your plans for addressing the stiffness of this specific example and the general concern of I-beams loaded in torsion.
SECOND RESPONSE (12-16-83)
As noted previously, Bechtel's standard practice is not to consider the contribution of Building Structural members in overall pipe support stiffness calculations.
Therefore, no plans are currently in place to evaluate individual cases of building steel being loaded in torsion for stiffness.
Nevertheless, Field Change Request 2FC-1191-MH has been re-evaluated to consider the contribution of torsional rotation on pipe support 0-FC01-R004/1350 and subsequently on stress problem No. 68.
Pcgo 3 of 6 The overall stiffness of pipe support 0-FC01-R004/1350 was calculated and provided as input for a reanalysis of stress problem No. 68.
The resulting stress levels were well within allowable and indicate only minor changes from the previous analysis.
Based on this example and the factors noted in our original response, this is not considered to be a generic Concern.
The issue concerning the torsional loading of structural I-beams i
is not a problem at the Callaway Plant because the effects of torsional loading are considered by the civil group in its review of pipe supports.
1 The loads received by civil, from plant design engineering, for each hanger review include all loads at the civil / plant design interface point, including torsional moment.
Civil then reviews the adequacy of the structural steel to handle these loads and approves the hanger, or makes modification, as necessary.
NRC COMMENT ON RESPONSE TO FINDING 3-8 (11-18-831 Your response does not adequately resolve the item.
You stated that standard support components (such as snubbers) are norme.lly loaded axially and are therefore significantly stiffer than the associated structural support members and you cited a single example problem.
However, the finding provided a specific example where the snubber was so flexible that the support assembly could not meet the stiffness requirements of Specification M-217, regardless of the stiffness of the associated structural support members.
Furthermore, it is doubtful that the standard support components are generally so stiff that valid results will be provided by only considering the associated structural support stiffness.
You are requested to describe your plans to resolve the question about the specific support cited in the finding, which does not meet the stiffness requirements of Specification M-217 and, therefore, could affect the validity of the piping analysis.
In addition, you are requested to describe your plans for identifying other instances where supports do not meet the stiffness assumptions used in the piping analysis and determining whether or not they have a significant effect on the analyses.
SECOND RESPONSE (12-16-83)
A study was performed on the support in question (0-FC01-R0 20/1350) to determine its overall support stiffness.
The revised stiffness value was then input into stress problem No. 60 to evaluate the effect.
The study shows a negligible increase in piping stress levels.
= ____*_ _ _ _ _ _.
Pcgo 4 of 6 We suggest the results of this study be reviewed with the cognizant NRC inspector.
We believe any examination of the approach and methodology used in the referenced study, when considered in conjunction will the rationale provided to NRC in the initial response, will sufficiently support our contention that further consideration of this item is not required.
NRC COMMENT ON RESPONSE TO FINDING 6-2 (11-18-83)
Your response does not adequately resolve the item.
Tha error identified in logic diagram J-2AL01 should have been detected and corrected by the review process prior to issuance of the drawing.
The fact that it was detected and corrected during the review of a subsequent revision to the drawing (issued during the NRC inspection) does not attest to the effectiveness of the original design review.
You are requested to describe your plans for review of other logic diagrams against the applicable schematics to determine whether or not a systematic problem is present.
SECOND RESPONSE (12-16-83)
We do not believe there exists a systematic problem involving logic diagrams.
The discrepancy noted between logic diagram J-02AL01 and electrical schematic E-03AL01A&B was solely the result of a misinterpretation of load sequencer interlocks.
To ensure that similar misinterpretations have not occurred, all logic diagrams and electrical schematics involving the load sequencer will be reviewed.
This review will be completed by February 15, 1984.
NRC COMMENT ON THE RESPONSE TO FINDING 6-3 No further information is required regarding this specific error, which appears to have been corrected by revision of the PSAR.
However, we are concerned that this error may be indicative of other similar errors; inconsistencies between the FSAR commitments and the actual design.
That concern is strengthened by other similar findings during the IDI review.
Accordingly, please provide additional information concerning the actions you have taken to assure that other FSAR deficiencies of this type do not exist, and that further errors of this type will be precluded or identified.
SECOND RESPONSE (12-16-83)
Our original responses to findings 2-1, 2-7 and 6-3 provided sufficient information to conclude that, 1) the actual configuration of the plant meets the safety design bases, 2) procedures that control FSAR preparation and revision processes are adequate to minimize the introduction of inconsistencies, and
t Pcg2 5 of 6 3) the FSAR inconsistencies noted did not form the bases for a safety concern.
In the case of findings 2-7 and 6-3, it was determined that FSAR change control procedures were not executed properly.
As a consequence, inconsistencies between the PSAR and actual plant configuration were not corrected.
However, we do not believe these two examples are an indication of a breakdown in this area.
As indicated in our previous comment with respect to finding 2-1, we believe existing design to be consistent with FSAR commitments.
Project Engineering procedures provide measures for control of the FSAR and are used to preclude the introduction of inconsistencies between evolving design detail and licensing commitments.
Where changes in licensing detail are anticipated due to design evolution, the procedures provide a mechanism for identification, tracking and implementation of required changes.
The primary procedures in place for this purpose at Bechtel are EDPI 4.22- 01 " Preparation and Control of SAR," and EDPI 4.23-01 "SAR Change Control."
There are several additional mechanisms available for identifying FSAR material which may be inconsistent with current derign configuration.
Primarily, these are by audit and licenuing reviews conducted internally within Bechtel or initiated through SNUPPS Staff / Utility review.
In addition, final system design descriptions are updated at time of design completion and system turnover.
This updating process is undertaken to assure compatibility and consistency between final system design and FSAR commitments.
An added check of consistency with FSAR commitments is provided through preparation of preoperational test procedures which focus on demonstrating capability of the existing, as-built design to satisfy FSAR commitments.
In conclusion, although two cases were noted where the FSAR was not properly brought up to date to reflect final design configuration, the process of FSAR change is adequately controlled and has not resulted in any unresolved safety concerns.
Consequently, no action beyond that currently indicated is proposed.
NRC COMMENT ON THE RESPONSE TO FINDING 6-4 Please describe your assessment of the causes of this error, and the bases for your determination that this was an isolated incident.
l SECOND RESPONSE (12-16-83)
Bechtel acknowledges a procedural violation in that the J-435 l
(orifice plate) specification was issued prior to signoff approval of a formal calculation.
The applicable Bechtel i
i i
Pcga 6 of 6 procedure, EDPI 4.37-01, allows such issue with permission of the Project Engineer.
Although the issuance of the purchase specification was a sound engineering decision on the basis of completed computer calculations, permission to release the specification was not obtained.
The technical adequacy of the specification was verified through a subsequent review of the calculations at which time the Bechtel design was confirmed to be Correct.
Since the time of the initial NRC inspection, further reviews were. conducted in this area.
While no additional examples have been found of purchase specification release in advance of calculatics checking and approval, instances were identified in which desig.
' 9 wings had been released in advance of approval of supporting ceae >ations.
This finding has led to further technical revi ws and resulted in issuance of a directive by the Bechtel Division Engineering Manager reaffirming Division policy that release of design documents will be supported by checked and approved calculations.
Instructions have since been issued by Bechtel to obligate the responsible engineering discipline to indicate whether issuance of a new (or revised) design document is supported by calculation and to require verification that such calculations have been checked and approved in accordance with existing procedure.
Bechtel compliance with these procedural controls will be subject to increased SNUPPS/ Utility monitoring and audit.
It should be noted that these technical reviews have not resulted in any design document changes.
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