ML20147D814
| ML20147D814 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/02/1988 |
| From: | Michael Ray TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8803040170 | |
| Download: ML20147D814 (14) | |
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TENNESSEE VALLEY AUTHORITY CHATTANOOGA TENNESSEE 374ol SN 105B Lookout Place MAR 2 1988 U.S. Nuclear Regulatory Comissior.
ATTN: Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
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Docket Nos.
50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SQN) - INTEGRATED DESIGN INSPECTION (IDI) - NRC INSPECTION REPORT NOS. 50-327/07-48 AND 50-328/87-48 provides TVA's supplemental response to IDI item D4.6-1.
- provides a list of comitments being made by TVA in this submittal.
It is our understanding that the information provided herein completes TVA's actions on this item for SQN unit 2 restart.
If you have any questions, please telephone D. L. Williams at (615) 632-7170.
Very truly yours, l
l TENNESSEE VALLEY AUTHORITY M. J.
y, Deputy Director I
Nuclear Licensing and Regulatory Affairs Enclosures cc:
See page 2 s
i 8803040170 880302 PDR ADOCK 05000327
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' I An Equal Opportunity Employer
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.. i U.S. Nuclear Regulatory Commission p
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-Enclosures cc (Enclosures):-
Mr. K. P. Barr, Acting Assistant Director.
for ' inspection. Programs TVA Projects Division U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-Mr. G. G. Zech, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission one White Flint, North 11555 Rockville Pike.
Rockville, Maryland 20852 Sequoyah Resident Inspector Sequeyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 L
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ENCLOSURE 1 SEQUCYAH NUCLEAR PLANT (SQN)
ITEM NUMBER:
D4.6-1 TITLE:
Equipment Drawing Versus Calculation Discrepancy
SUMMARY
OF ITEM:
The design drawings for equipment foundations and supports are not in agreement with the calculations.
CLASSIFICATION:
Minor Calculation Error (Items 1 and 2 as listed below)
Design Deficiency (Item 3 as listed below)
SUPPLEMENTAL RESPONSL':
The foundation embedment designs for the following equipment were includud in the IDI review:
1.
Component Cooling Water Heat Exchanger (CCWHx)
TVA agrees that the baseplate thickness of 1/2-inch shown on the drawing (48N1269 R11) at the top of the concrete pedestal for the CCWHx is less than the 3/4-inch thickness required by the original calculations. A Condition Adverse to Quality Report (CAQR) SQP871450IDI (reference 1) has been written to address this discrepancy.
This issue has been resolved in conjunction with IDI item D3.4-3.
The CCWHx has been reanalyzed, and the loads from this reanalysis were used to qualify the embedded anchorages (reference 2) to the load combinations and allowable stresses of SQN-DC-V-1.3.3.1 except that the maximum allowable stress for abnormal / extreme load combinations was limited to 0.9 Fy as provided in SQN-DC-V-1.3.2.
The 1/2-inch embedded plate meets these criteria.
2.
Component Cooling Water Surge Tank (CCWST)
TVA agrees that a drawing discrepancy exists between the original CCWST support calculations and the design drawing (48N1271).
Specifically, the drawing detail is unclear with respect to number and spacing of studs required.
The CCWST anchorage has been reviewed in response to IDI item D4.6-2 (tank No. 12 as listed in the attachment to response for IDI item D4.6-3).
Based on a conservative interpretation of the anchorage details on the design drawing, the details have been evaluated and the anchorage I
meets design criteria SQN-DC-V-1.3.2 and DS-C1.7.1 requirements (reference 3).
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3.
Containment Spray Heat Exchanger (CSHx)
TVA agrees that the diameter of the concrete anchors and the plate thickness shown on drawings 48N1266 and 48N1267 for the CSHx anchorage are less than required by the original calculations. This discrepancy had been previously identified by TVA undue the "Detailed Technical Review of Miscellaneous Steel Structures" (reference 4).
As a result of this review, CAQR SQp870188 (reference 5) was written to document deficiencies in the heat exchanger support structure.
Additional evaluations have determined that a modification was required to meet the design criteria requirements. The modification has been completed.
Calculations confirming the adequacy of the modified structure are contained in reference 6.
Generic Review TVA has conducted a generic review to resolve the IDI finding that design drawings for equipment foundations and supports n?v not be in agreement with the calculations and that the foundations and supports may be structurally inadequate. To evaluate this issue generically, TVA has established three categories of equipment: tanks, heat exchangers, and other equipment.
The twelve Category I tanks identified by the Design Baseline and Verification program as required for safe shutdown have been reviewed for calculation and drawing compatibility.
Supplemental calculations have been developed where required in the resolution of IDI item D4.6-2.
These calculations demonstrate that the tank anchorages meet the design criteria requirements of SQN-DC-V-1.3.2 and DS-C1.7.1.
Eleven heat exchangers have been similarly reviewed in response to IDI items D3.4-3 and D3.4-4.
The results of this review are discussed in the response to those items.
The adequacy of the equipment supports has been further evaluated by reviewing a sample of 60 supports.
The following attributes were considered in selecting the supports:
higher building elevations, closely spaced anchor bolt patterns, small diameter anchor bolts, equipment with intervening steel structures, and equipment with high center of gravity and high mass. The evaluation of the 60 support structures demonstrated that 59 of the structures meet design criteria requirements.
The reactor coolant pump (RCP) anchorage meets all design criteria requirements except for the load combination involving a pipe rupture in the reactor coolant loop and a concurrent SSE.
However, there is an adequate margin of safety such that the RCp will perform its design function. After restart, the anchorage will be further evaluated to encure that the RCP anchorage meets design criteria requirements for the extreme load combinations.
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Based upon the evaluations that have been completed. TVA has determined that calculation / drawing discrepancies do exist. However, when the features are evaluated for verified loads and as-built configurations, there is a high degree of confidence that the supports will meet design criteria requirements and, thus, are acceptable.
As previously noted in TVA's January 19, 1988 letter to the NRC, in TVA's response to NRC Observation CEB-15 the miscellaneous steel calculations, which include equipment supports, will be reviewed post-restart and revised as required te verify the adequacy of the miscellaneous steel features. provides TVA's program plan for completing the generic evaluation of equipment supports.
The root causes for the calculation / drawing discrepancies have been determined to be inadequate verification and design change control of the design calculations and drawings. TVA has updated its Nuclear Engineering Procedures (NEPs) to further clarify the requirements for independent design verification of the calculations and drawings, and to enhance the design change control program.
The implementation of increased design verification and the enhanced design change control program provides assurance that future design calculations and drawings will be compatible.
In addition to the review described above, during the NRC's inspection in Knoxville, Tennessee, the week of February 15, 1988, responses to questions and/or commitments for additional ac.tions from TVA were requested on several equipment supports evaluated during this review.
Issues to be addressed before and after SQN unit 2 restart were identified. The following providos TVA's response / commitments on these issues for each TVA item number where additional questions were raised:
PRE-RESTART ITEMS ITEM 8 - REACTOR COOI. ANT PUMP ANCHORAGE The NRC requested clarification on the anchorage loads supplied by the NSSS vendor that identified higher OBE loads than DBE loads. The vendor (Westinghouse) has supplied a technical justification (attachment 2) for the higher OBE loads.
It should be noted that the higher OBE loads occurred only on some of the reactor coolant pump support columns, i
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ITEM 62 - COMPRESSOR Calculation SCG-4M-00174 identified a missing nut on an anchor bolt for the subject equipment. The NRC questioned whether the nut had been replaced.
TVA r
initiated a maintenance request on October 28, 1987, to install the missing nut. The nut was installed on January 15, 1988.
The subject calculation has been revised to document the process by which the nut was replaced.
POST-RESTART ITEMS ITEM 8 - REACTOR COOLANT PUMP ANCHORAGE TVA will demonstrate that the affected anchorage structure meets design criteria requirements for the extreme load combination.
ITEM 15 - LOWER COMPARTMENT COOLING UNIT C-A Calculation SCG-4M-00177 R1 was based on the lower and upper portions of the subject equipment being dynamically uncoupled.
In response to an inquiry from an NRC reviewer, a preliminary calculation to justify the decoupling was developed and presented to the reviewer on February 18, 1988. That calculation will be incorporated into SCG-4M-00177 R2.
POST-RESTART ITEM 3 k
ITEM 82 - RHR PUMP l
The NRC requested confirmation that the nozzle load forces on the RHR pump were properly applied and that the different coordinate systems used in the load generation and anchorage calculations were reconciled.
Calculation SCG-4M-00210 will be revised to provide better documentation of the nozzle load forces used.
In addition, calculation scc 1S173X-082 will be revised to provide better documentetion of the relationship between the coordinate systems used in the load calculations and those used in the anchorage calculations.
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ITEM 122 - 480-VOLT SHUTDOWN BOARD 2B1-B NRC requested documentation of qualification of the 3/4-inch holddown bolts on the 480-volt shutdown boards. The 3/4-inch diameter holddown bolts are j
included in the Westinghouse scope of supply (as ubown by drawing 7057D25).
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Appendix II of calculation SCC-AM-00195 documents the seismic qualification test of the subject transformer.
By accepted industry practice, this test i
would have included the holddown bolts.
Consequently, no additional i
evaluation is required.
However, calculation SCG1S173X122 will be revised to provide a calculation qualifying the subject bolts, l,
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REFERENCES:
1.
Condition Adverse to Quality Report (CAQR) SQP871450!DI 2.
SWEC Calculations 17341-11-CS(B)-4 (B25 880222 312) 3.
DNE Calculations (B25 871104 455) 4.
Detailed Technical Review of Miscellaneous Steel Structures (SQW-CEB-87-02) 5.
CAQR SQP870188 6.
SWEC Calculations 17341-CS(B)-13. (B25 880222 314) and 17341-31-NM(B)001 (B25 880223 310) 1 o
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.. l ATTACHMENT'1 IDI D4.6-1 PROCRAM PLAN FOR CENERIC EVALUATION OF EQUIPMENT SUPPORTS
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o Resolve specific IDI finding (i.e., discrepancy between calculations and drawings).
o Evaluate CCW Heat Exchanger and Support Structure o
Ivaluate CCW Surge Tank Anchorage o
Evaluate Containment Spray Heat Exchanger and Support Structures Complete generic evaluations.
o Review 11 Category I heat exchangers and support structures o
Review tank anchorage design and support structures for all (12) 4 Category I tanks identified by the Design Baseline and Verification
,y Program as required for safe shutdown 3
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o Review calculations versus drawings for 60 equipment supports l
Evaluate supports through the development of supplemental l
calculations where calculation / drawing discrepancies are identified
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In addition to the defined actions for the resolution of the IDI item, TVA. ' ' '
has fully evaluated, through its CAQR program, additional equipment i
supports for a total of 60 supports. These evaluations are being conducted utilizing reverified loads and "as-constructed" configuration.
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Acceptability of the equipment supports is demonstrated if supplemental
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j calculations that are developed to reso.'.ve the calculation / drawing h
discrepancies or to evaluate the technical adequacy of the feature demonstrate that all features meet design criteria requirements and compliance with FSAR commitments.
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o If a feature does not meet design criteria requirements. TVA will develop
'i a Condition Adverse to Quality Report (CAQR) and evaluate against failure i
evaluation criteria.
If the feature meets the failure evaluation criteria, then it is acceptable for restart.
After unit 2 restart, TVA j
will re-evaluate equipment support features and modify, as required, to
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meet design criteria requirements.
(Refer to response to NRC's l
Observation CEB-15.)
i NOTE: A previous commitment, Corporate Commitment Tracking Item NCO-87-0361-052, stated that equipment support features not meeting design criteria requirements would be evaluated against an interim acceptance criteria. However, as discussed with the NRC reviewer, it was not necessary I
to utL11 e an interim acceptance criteria; the previous commitment can be i
replaced by the above commitment stating that items failing'to meet design j
I criteria requirements will be evaluated against a failure evaluation criteria.
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- Restart activities are complete.
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ATTACHMENT 2 WESTINGHOUSE TECHNICAL JUSTIFICATION S
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FEB.26 'E817:00 R&D 701 BLDG 412-256-6743 P,02 c
Westinghouse PowerSystems Electik Corporation P0 8t"" 8 "" *"*
EE!ce monaism em Mr. J. B. Hosmer Sequoyah Project Engineer TVA-88-553 Tennessee Valley Authority February 26. 1988 Sequoyah Nuclear Power Plant, DSC-A P.O. Box 2000 Soddy Daisy TN 37379 Tennessee Valley Authority Sequoyah Nuclear Power Plant Unit'Humber 1 and 2 Reactor coolant Pumo Seismic column lq1 tit
Dear Mr. Hosmer:
Thesequoyahprimaryloopstressreport(50119)hasbeenreviewed.
at the request of DNE, to determine the reasons for the seismic loads on the Reactor Coolant Pump colu:nns being CBE condition than for the DBE condition. generally higher for the This inquiry was made to assist DNE ir, its response to the NRC questions during their Sequ IDI reviews.
Enclosed is the Westinghouse response to your inquiry.
Should there be any questions concerning our response, please a~ dv e
Very truly yours.
WESTINGHOUSE ELECTRIC CORPORATION dn
. A. Lordi. Manager TVA sequoyah Project cc K. $. Siedle K. Spates J. Rochelle W. R. Manglante
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FEB.26 '6G 17:01" RtD 701 BLDG 482-256-6743 c.
P.03 TVA Sequoyah l
Reactor Coolant Pop Support Columns Response to NRC Questions from the laauovah IDI Review
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Executive sumary i
Westinghouse provides the following in response'to the NRC's questi concerning the difference of the TVA sequoyah Reactor Coolant Pum column se'smic loads.
The difference in RC Pump support column OBE and DBE seismic leads is due to the difference in response spctra used in the ReactorCoolant1. cop (bym)ultiplyingth RCL seismic analysis. The individual RC Pump column i
loads were developed RCL seismic analysis by support member influence coefficients.e seism s
individual member loads ta4ulat The i
are the result of thi,s process.ed in Appendix B of the 50-119 stress report n.
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4 TVA sequoyah Reactor Coolant pts!p Support Columns Response to NRC questions NRC Question 1
Reactor Coolant System RCSDuring the NRC's IDI reviews column loads. question c(once)rning the Reactor C resulted in a In reviewing West'nghouse RC5 Equipment support stress report 8011g, the NRC questioned the RC pump support colum OBE and Dbt seismic co transient are larger, nditions.
in magnitude than the column loads for the seismic transient.
These loads wer,e questioned since the Design Basis Earth uakes Earth uakes ( Bare larger in magnitude than the Operational Basis expla ation o( I Consequently the NRC has asked TVA to pr ese column suppo,rt seismic load differences. ovide an Response Spectra Comparison-s.
The difference in 08E and Ott Rb pump column seismic load
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the difference in the containment building response spectra
. values used to develop the response spectra used for the ACL 1
s The desping analysis were 2 percent for the building and 1
'Olt, and 5 percent for the building and 1 perc/2 percent for the RCL for 1
exceeding the DIE accelerat'ons for most frequenc Using the actual. frequency response values.stakom frontihe i
analysis, the: relative,difforec43etivesiCh6r.12ontal sputu Wes determined. Tho'.me~stistiMffotnt Sequetth' ipr 1r1FR I
at 4.94 10
.L these re,spon.6r-10st, ~4nd.16J Jer,tG' Tho'hoHi6Tita'listiceTestib6 ~ pi Olt to DRE acceler;q'compH6Hy'tahulating the ratio'of"tfie co setaar J. t.
1 ations.
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.... 2 Frequency (ha
..................).......OBEAccel./DBEAccel.
4.g4 0.g67
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10.6 10.8 1.42 16.9 1.47 1.33 i
envelopes the DBE spectrum.From this table it can be concluded 3
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, in general, horizontal spectra.In addition, the vertical CBE and DIE respo Therefore e
vertical DBE spectrum by the sa,me magnitudes.the vertical OBE spectra w
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FEB.26 '89 17:02 R&D 701 BLDG 412-256-6743 P,05 e
RC Pump Support Plane Leads The RCL seismic analysis provides RC Pump loads in the fo plane loads. Support plane loads degrees of freedom at the RC pump c(enterline.SPL) consist of loadings in the resolved into individual support member loads by multiThe support plane load RC Pump support column mester i~nfluence coefficients. plying the SPLs by used to develop the RC Pump column _ loads of stress report 501 This process was in the magnitude of the six components of suIn addition differences in RC Pump support column loads.pport plano loads also produce SPLs are distributed to the force components and the three moment components. s resultant moment vector.three moment components of SPLs pro vectors determine the amount of load distribute example As an The resu,ltant moment vector of these two tems is directed at an angle of 45 degrees with respect to the orthogona
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The resultant moment vector is distributed to the support colum forces from the 45. degree position ial loads would be different for a resu. The RC Pump support column axial ltant moment vector distributed from a 75 degree position.
The mode shapes for the RCL seismic analysis are the same and 08E condittens. However variations in the ratio of 08E and 08E the magnitude and direction of the resultant fo i
the SPLs. Therefore, the distribution of the CBE and DIE resultan i
RC Pump support produces ths differences in the column load o the stress report 30 11g.n n
Review 1
During the review of the stress reurt 3D-11g RC pump support c 4
loads,atranscriptionerrorfortheCBEmoment(M)ofcolumn6was identified.
This transcription e;'ror was considered conserva
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embedsent evaluations and was not corrected.
Conclusion stress report SD 119 are applicable for the CB j
variation in column loads for the CBE and DBE co i
higher Olt spectra than DBE soectra and the resulting differenc i
magnitude and direction of the SPL vectors.
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,a ENCLOSURE 2 LIST OF COMMITHENTS
,IDI D4.6-1 1.
After SQN unit 2 restart, the anchors of the reactor coolant pump will be further evaluated to demonstrate that it meets design criteria requirements for the extreme load combinations (Item 8).
2.
As noted in response to NRC Observation CEB-15 (January 19, 1988 submittal), the miscellaneous steel calculations, which include equipment L
supports, will be reviewed post-restart and revised as required to verify J
the adequacy of the miscellaneous steel features.
3.
After SQN unit 2 restart, the preliminary calculation to justify the i
decoupling for the dynamic analysis of the lower and upper portions of the lower compartment cooling unit C-A will be incorporated into calculation SCG-4M-00177 R2 (Item 15).
4.
After SQN unit 2 restart, calculation SCG-4M-00210 will be revised to provide better documentation of the nozzle load forces used (Item 82).
S.
After SQN unit 2 restart, calculation SCG1S173X-082 will be revised to provide better documentation of the relationship between the coordinate systems used in the load calculations and those used in the anchorage calculation (Item 82).
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After SQN unit 2 restart, calculation SCG13173X122 will be revised to provide a calculation qualifying the 3/4-inch holddown bolts on the 480-V shutdown boards (Item 122).
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