ML20206M548
| ML20206M548 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/14/1986 |
| From: | Joseph Holonich Office of Nuclear Reactor Regulation |
| To: | Ballard R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8608210250 | |
| Download: ML20206M548 (55) | |
Text
,. .
AUG 2 4 jgg3 MEMORANDUM FOR: Ronald Ballard, Chief Engineering Branch d dd Division of PWR Licensing-A FROM: Joseph J. Holonich, Project Manager Project Directorate #4 Division of PWR Licensing-A
SUBJECT:
REQUEST FOR REVIEW OF INTERIM CRITERIA Enclosed is an advanced copy of the Tennessee Valley Authority (TVA) submittal on the interim acceptance criteria being used at Sequoyah. TVA will make an official submittal of this information within the next few days. However, because of the compressed review schedule which is presently in place, TVA has provided an advanced copy so that the staff can begin its evaluation.
Therefore, the purpose of this memorandum is to request that the Engineering Branch (EB) begin'to review the enclosed information. Once the fomal submittal is made, Project Directorate #4 will provide EB with a copy. The TVA responses to the EB questions, contained in your memorandum dated August 7,1986, are not included in the enclosure nor will they be included in the fomal transmittal.
TVA will provide its responses by August 29, 1986. Once these are received, EB should provide a safety evaluation report by September 12, 1986.
If you have any questions, please contact me at extension 27270.
6\
Joseph J. Holonich, Project Manager Project Directorate #4 Division of PWR Licensing-A
Enclosure:
As stated DISTRIBUTION: See Other Page l
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% ,,,,,+ AUG 141986 MEMORANDUM FOR: Ronald Ballard, Chief Engineering Branch Division of PWR Licensing-A FROM: Joseph J. Holonich, Project Manager Project Directorate #4 Division of PWR Licensing-A
SUBJECT:
REQUEST FOR REVIEW 0F INTERIM CRITERIA
. Enclosed is an advanced copy of the Tennessee Valley Authority (TVA) submittal
- on the interim acceptance criteria being used at Sequoyah. TVA will make an l-~ official submittal of this infomation within the next few days. However, i because of the compressed review schedule which is presently in place, TVA has provided an advanced copy so that the staff can begin its evaluation.
Therefore, the purpose of this memorandum is to request that the Engineering Branch (EB) begin to review the enclosed infomation. Once the fomal submittal is made Project Directorate #4 will provide EB with a copy. TLe TVA responses to the EB questions, contained in your memorandum dated August 67,1986, are not included in the enclosure nor will they be included in the fomal transmittal.
TVA will provide its responses by August 29, 1986. Once these are received, EB should provide a safety evaluation report by September 12, 1986.
If you have any questions, please contact me at extension 27270.
i d Yo*1 Joseph J. Holonich, Poject Manager t' Project Directorate #4 Division of PWR Licensing-A
Enclosure:
As stated
- e. 7 TENNESSEE VALLEY AUTHORITY I?
SN 157B Lookout Place Director of Nuclear Reactor Regulation Attention: Mr. B. Youngblood, Project Director i
FWR Project Directorate No. 4 Division of Pressurized Water Reactors (PWR)
Licensing A ^
U. S. Nuclear Regulatory Commission Washington, D.C. 20555 -
Dear Mr. Youngblood:
In the matter of ) Docket No. 50-327 Tennessee Valley Authority ) 50-328 Interim Acceptance Criteria has been developed for Sequoyah to be used for the temporary resolution of several critical engineering issues. These criteria involve minor deviations to the original FSAR licensing commitments, yet assure that the system functions required for safe power plant operation and accident mitigation are not compromised. Overall standards for public safety therefore are not lessened. Rach deviation can be justified by current industey practice, other plant precedents, or recent code developments.
TVA proposes to utilize these acceptance guidelines on a limited basis. Given that public safety is assured, application of these criteria will serve as the basis for timely restart and operation of Sequoyah units 1 and 2.
Subsequently, full compliance to the original design bases and FSAR commitments will be reestablished, in accordance with a defined post-restart program schedule.
Application of these interim acceptance criteria will be limited to the
- following disciplines
c
- 1. Small Bore Piping and Supports
- 2. Safety-Related Cable Tray Suports All remaining areas of plant design will continue to satisfy the commitments previously established within the original Sequoyah design basis documentation.
- t Proposed interim acceptance criteria are presented as an enclosure to this letter. The engineering basis and justification accompanies each deviation taken from the original FSAR commitments.
- . o- p 2-Director of Nuclear Reactor Regulation Additionally, twelve technical information requests (TIR) which resulted from our presentation on July 18, 1986, are addressed with this submittal. Please review the enclosed. If any further clarification or details are required, please call N. R. Harding at (615) 870-6422.
Very truly yours, TENNESSEE VALLEY AUTHORITY
--~
R. Gridley, Director Nuclear Safety and Licensing Enclosure cc: U.S. Nuclear Regulatory Commission Region II Attn: Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Nr. Carl Stahle Sequoyah Project Manager U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue i, Bethesda, Maryland 20814 9
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SEQUOYAH NUCLEAR PLANT '
. NRC TECHNICAL INFORMATION REQUEST -
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- INTERIM ACCEPTANCE CRITERIA T
CIVIL ENGINEERING PROGRAMS B
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., .t TABLE OF CONTENTS I. Introduction II. Purpose / Scope III Technical Information Requests s
IV. Cable Tray Supports Interim Acceptance Criteria 4
A. Comparison with FSAR Commitments and Justifications B. Cable Tray Supports on the Steel Containment Vessel
- 3. ) C. Cable Tray Supports on Reactor Building Shield Wall
, D. All, Other Cable Tray Supports
_ , , V. Small Bore Piping Analysis and Support Design Interim Acceptance Criteria A. Comparison with FSAR Commitments and Justification - Piping Analysis B. Comparison with PSAR Commitments and Justification - Support Design
.Y I.
- Anchor Bolt Load Capacities .
VII. Cable Tray Design and Construction - QA/QC Procedures Attachments i
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., . e 1. Introduction As a result of TVA-NRC meetings held in Bethesda, Maryland, on July 17-18,1986 to discuss the provisions and use of Civil Engineering Interim Acceptance Criteria as part of the SQN Restart Program, a list of twelve Technical Information Requests (TIRs) was developed. TVA committed at the meeting to expeditiously respond to the requests.
II. Purnose and Scone F
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The purpose of this transmittal is to provide the information NRC
. . _ , requested and expedite the approval of the interim acceptance criteria.
The twelve TIRs are presented and cross-referenced to the pertinent responses.
III. Technical Information Reauests (TIRa)
- 1. Provide a description of the analysis procedures bei~ng used for cable tray supports interim criteria. Compare these to the original analyses. Include the technical basis for the interim criteria analysis.
Response in Sections IV. B.2. C.2. and D.2.
- 2. NRC to provide a list of generic questions on cable trays. TVA should be prepared to address them.
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., .e TVA is prepared to respond to these upon receipt.
- 3. Describe how the D3A response spectra on the SC7 have been developed. How many cable tray supports are affected by the DBA ,
spectra?
Response in Sections IV. B.1, B.3, IV. C.I . , and IV.D.I .
, , 4. Provide description of how anchor bolt load caps' cities are determined by TVA.
Response in Section VI.
- 5. How will employee concerns be integrated and resolved for the cable tray and piping programs?
TVA's Employee Concern Program is defined in detail in the Nuclear Performance Plan (revised) for Sequoyah. This plan was transmitted to NRC in a letter dated July 17,1986, from S. A. White to the Honorable Lando W. Zech, Jr.
Employee concerns in the cable tray support area and the small bore piping area have been identified. The subject of several of these employee concerns are being addressed in the cable tray and small bore piping programs. All of the employee concerns involving cable ' tray supports and small bore piping that af fect safety vill be addressed and resolved prior to restart of the plant.
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6.
Provide TVA's best estimate of our schedule for Phase II (after plant restart) Laplementation.-
Phase II will consist of development of designs, analyses, and technical positions to finalize and document the plant's design basis consistent with FSAR commitments. This process will i include additional plant walkdowns which will necessitate staging data gathering with outage schedules.
We, estimate that, based on presently identified scope-of-effort
_, for the two programs discussed in this transmittal, that 2 to 3 scheduled outages will be required to obtain data and make the necessary modifications. Phase II scope and schedules will be identified prior to restart.
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7.
, Provide a list of exceptions to FSAR commitments being taken during the interim period (Phase I) and complete justification for interim acceptance criteria.
l, Responses in Section IV.A. for Cable Tray Supports and Sections
- Y. A. and V.B. for Small Bore Piping and Pipe Supports.
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- 8. What QA/QC procedures were applied to the original cable tray design and construction? For the Phase I and Phase Il programs?
}
Responses in Section VII.
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. , '. e 9. Make available to the NRC reviewers copies of all documents pertinent to cable tray design and construction as well as
. cookbook analysis techniques. Material is to be available j
July 21, 1986. l l
The following material has been provided to NRC reviewers.
TVA drawings 48N1501 through 1506 (supports on SCV)
. - Walkdown Criteria SMI-0-317-36 -
TVA drawings for supports in auxiliary _ and control building
_. CEB Report 80-5 SQN Alternate Criteria for Piping Analysis and Support CEB Report 76-5 WBN Alternate Criteria for Piping Analysis and Support SQN Alternate Analysis Review Progran Procedures s .
~ 10. Describe the walkdown procedures f or data gathering to ensure that the interim cable tray qualification program addresses as-built conditions. -
Responses in Section IV. B.4. , C.3. , and D.3.
- 11. Provide justification for separating secondary stresses or provide alternate technical approaches for SAM / TAM issues.
Responses in Section V. A.
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1: , .e 12. Provide additional justification for 2Sy stress level used in f aulted primary stresses.
Response in Section V. A.
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., .e IV. CABLE TRAY SUPPORTS A. Interim Accentsoce Criteria: Comparison with FSAR Conusitaents and Justification The interim acceptance criteria for cable tray supports is limited in its application and is consistent with current knowledge and Practice. It is compared to the FSAR commitments (Attachments A j , , and B) and justification provided for each exception below.
Those supports that utilize the interim acceptance criteria vill be reevaluated for the plant design basis criteria and modified as required during Phase II (after plant restart).
- 1. Dampina Compare to FSAR Commitment
[ The interim acceptance criteria preacribes 7 percent damping for the safe shutdown earthquake (SSE) and for the Design Basis Accident.(DBA). Table 3.7.1-3 Attachment A herein, of the SQN FSAR allows a maximum of 5-percent damping for bolted structures for the SSE vben the stresses are at or near i
i yield. For welded structures, it allows a maximum of 2-percent damping. For the 1/2 SSE, the FSAR allows 1- and 2-percent damping f or velded and bolted structures, res pectively.
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Justification In recent years, seismic testing of various types of electrical raceway systems (both cable trays and conduits) has demonstrated their inherent ruggedness and ability to absorb energy while continuing to perf orm their function, when subjected to seismic loads.
. 7 TVA v'as an active participant in the raceway test program conducted by Bechtel Power C'orporation and ANCO Engineers, Incorporated. Reference the report, " Cable Tray and Conduit i
Raceway Seismic Test Program Release 4 " Report 1053-21.1-4, ANCO Engineers Incorporated, December 1978. These tests demonstrated high damping (up to 50 percent) for flexible unistrut supports, as well as for trays mounted directly to the .
shake table. The latter condition is equivalent to_an infinitely rigid support. Therefore, the tests encompassed a
, wide range of system frequencies. .
i' Tests conducted by URS/ John A. Blume and Associates. Engineers (URS/Blume) for the Systematic Evaluation Progran (SEP) Owners Group found similar results for the specific rod-hung and l unistrut support configurations tested. The test report is
" Shaking-Table Testing for Seismic Evaluation of Eeletrical Raceway Systems " April,1983 URS/Blume. The evaluation of the supports is contained in the report, " Analytical Techniques Models, and Seismic Evaluation of Electrical Raceway Systems," August 1983, URS/Blume.
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,, .e Although specific SQN tray and tray supports were not among those tested, the test results are applicable to SQN on the following basis:
- 1. The type of tray and tray supports utilized in the tests did not significantly influence system desping.
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- 2. The stiffnesses of the SQN cable tray support systems fall
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, within the range of stiffnesses ef the flexible supports and the higher bound range of rigid suppprts in the tests.
- 3. The flexible type support systems tested had a range of .
system fundamental frequency in the transverse, vertical, and. longitudinal direction.s of '2-6 Hertz. The rigid type supports had a frequency range of 9-25 Bertz. SQN support system frequencies f all within the range of these frequencies. .
Although the type of tray and tray supports used in the SQN systems have not been specifically tested, they have dynamic characteristics similar to those employed in the test program.
On this basis, the use of higher damping is acceptable.
The 7-percent, damping prescribed for the SSE by the interim acceptance criteria is, therefore, a conservative application of the data considering the much higher values found from the tests.
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- Of the numerous plants approved for cable tray damping in excess of the values allowed by the USNRC Regulatory Guide (RC) 1.61, the Vogtle plant is the one with cable tray supports most like those at SQN. Both plants use welded structural tube supports, with the trays bolted to them. Vogtle was approved for up to 15 percent damping. Due to the similarities with SQN supports, the Vogtle plant provides a strong basis for the much lower 7 percent SSE damping in the interim acceptance crit e.ria. .
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[ -- - Finally, the 7 percent SQN damping is consistent with the 7
- percent SSE damping allowed by RC 1.61 for bolted structures.
Most of the cable tray system mass and flexibility is in the cables and trays. The SQN trays are bolted to the supports and bolted connections are used in the splice plates joining
, straight ' tray section's to each other an'd -to tray fittings.
These connection details plus the nonlinearities and energy h absorption provided by the cables will cause the system to experience increased damping. Compared to damping currently
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allowed on bolted structural systems, the use of 7-percent SSE damping for the SQN cable tray supports is reasonable.
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.. ,, 2. DBA - SSE Load Combination Connare to FSAR Commitment The interim acceptance criteria requires these two dynamic l loads to be combined by the square root of the sum of the l squares (SRSS) method. The design criteria requires their effects to be added by the absolute sua method. The FSAR does not explicitly define the combination method "for these loads. -
Consequently, the application of the SRSS method is not an
[- exception to the FSAR. It is addressed herein since it is an element of the interim acceptance criteria which differs from the design criteria.
Justification
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The technical justification ~ for this combination nethod is based on the fact that the DBA and SSE are low probability.
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independent (one does not initiate the other), dynamic events.
NUREG-0484 Revision 1 documents NRC acceptance of this combination method and references reports by General Electric, Westinghouse, Brookhaven National Laboratories, and a number of consultants.
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A final consideration is the recent NRC position on General Design Criteria 4 published in the Federal Register Volume 51, No. 70 April 11,1986. It defines a basis for eliminating the DBA dynamic events as a design load (excluding the containment l DE02;036211.02
vessel and some essential piping systems), providing that certain fracture mechanics assumptions can be verified on a plant-specific basis.' It should be noted, however, that ITA has taken no advantage of this new position.
3.0 1/2 SSE SSE Connare to FSAR Commitment The interim acceptance criteria waives the requirement fcr considering the 1/2 SSE load case. The FSAR states in section 3.10.2 (Attachment B) that: "For dead loads combined with live loads and for dead loads combined with the 1/2 safe shutdown earthquake loads, the designs are based on the allowable '
stresses of the AISC Specification on Structural Steel for Buildings." '
j Justification ,
i
- i, The SSE will always produce greater structural response than '
the 1/2 SSE and will, therefore, be the controlling load case in terms of maintaining functionality following a seismic event. Consequently, evaluation for SSE under the interim acceptance criteria ensures the ability to bring the plant to a safe shutdown for any seismic event less that or equal to the SSE.
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- In addition, the USNRC has identified the need for additional 9-i seismic evaluation of equipment and supports in older plants.
An example is the establishment of the Systematic Evaluation Program (SEP) for eleven older operating plants. There is also an Unreviewed Saf ety Issue (USI) A-46 and associated draf t NUREG-1030 documenting outstanding seismic issues on equipment and supports f or operating plants.
, The majority of plants covered by these progr~ans have continued t
to operate while the issues are resolved. In many instances, l' - - - the basis for continued operation is the demonstration of safe functionality following the SSE. For example, a February 8,1984 letter from NRC to Southern California Edison (docket number 50-206) documents NRC acceptance of a '
" functionality criteria" f or piping and pipe supports for San Onofre Unit'l as the basis for plant restart Only the SSE is j
. considered in Chat criteria.
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lip The URS/Blume report on Analysing Technical Raceway Systems l .[ addresses the seismic response of raceway systems (cable trays and conduits) .for various SEP plants. Again, the purpose is to denonstrate functionality of the safe shutdown systems following the SSE.
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The application of the SQN interim acceptance criteria for SSE
. evaluation provides the necessary assurance f or plant safety I
and functionality.
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. .' 4. Allowable Stresses Compare to FSAR Commitment The interim acceptance criteria allowable stresses are identical to those in section 3.8.4 of the USNRC Standard Review Plan ('SRP), NUREG-0800. Section 3.10.2 of the SQN FSAR (see Attachment 2), states: "For dead load combined with safe shutdown earthquake loads the stresses are limited to 90
~
percent of yield stress for the material involved." There is no
_, explicit FSAR commitment for cable tray support stresses for load combinations including the DBA.
r The SQN cable tray supports are generally fabricated from square structural tubes. Under the provisions of the AISC specification, these' members usually qualify as compact sections and have an allowable in bending of 0.66 times yield stress. Applying the SRP allowables to these members for a
- load combination including DBA and SSE gives an allowable of 1.7 times the AISC allowable, or 1.12 times the yield stress.
For the SSE locd without. DBA, the SRP allows 1.6 times the AISC i
allowable, or 1.06 times the yield stress. Thus, the above allowables are seen to be an increase over the 0.9 times the yield stress committed in the FSAR.
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. ,, . . Justification-The SRP stress allowables have been used extensively in evaluations of plants under construction as well as for evaluation of older plants. The use of these allowable stresses for evaluations under the interim acceptance criteria vill ensure that stresses are maintained within current NRC criteria and acceptable industry standards.
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- IV.B.
CABLE TRAY SUPPORTS ATTACHED TO STEEL (3mframmenarn ygg3EE
- 1. Scope The reevaluation of the supports attached ttno tethe meta =41 - l- -
vessel is required as a part of the resolutna= af NIDR SN i
All 560 of the cable tray supports attachee as talte aumen aanttainnmann.
- ! vessel were evaluated. 551 of these mappactra imumme ====t=r dt hyj f&is,ec
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[ typical designs. The remaining nine (9) uniignum magamartta warm-
- _, individually analysed. The analysis identifmad Himmt (diD asraan, requiring modification of 3 existing suppcata sans alta =hniini nni i of 112 new supports. The modified supports and the meur anaggestin ====a dheiguadi to meet original design criteria requiranerets 'Jhus (CID) af tdhs, modifications involved reinforcing the existiing maggmuitas tam =or.u-mi--H the effects of th'e design loads. One (1) modtilf=n=dnimn mar n=gi An-H tte prevent overstressing the vertical stiffeness =nen=rit=H hm Elle EstEE1 3 containment vessel. The twelve (12) additi==m11 anagguurtte meure auqpsame in areas where the unsupported spans of trays were feas 7/ fEemet troo ll@
feet in length. These spans were 75 percent its 1551 punzene grammaar j than spans elsewhere on the containment. The seemdl maggnats inasrae
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predominantly cantilevers supported from .franig mumlbans letbast og=an-H l
the vertical stif feners on the steel contalment vusmeL 2.a. Analysis Procedures for Interim Acceptance' Design Dynamic Response: The dynamic response of the ann == nun 1u0 syscana composed of the cable trays and their supports mer mia'llyaredl wstdie tihe mct.,4mrenn.0122
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, .,, GTSTRUDL computar ced2 usitt ths 01cetic response spectra modal analysis technique. Both SSE and DBA dynamic design events were evaluated using 10 percent broadened, 7 percent damped spectra.
Model: The cable tray and its supports were modeled using elastic beam elemsents. The vertical mass of the cables was lumped on the tray while the horizontal mass was lumped on the tray side rails. If the support failed to meet interim acceptance criteria when analysed using the design basis fully loaded weight of 45d/ft, the actual tray i
l loading determined by field walkdown was modeled. The flexibility of the model support points was modeled using spring constants determined by a finite element analysis of the containment vessel and c its stiffeners.
Thermal and Pressur's Loadings: The thermal gradient between the
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cable tray supports and the containment vessel was evaluated using ,
the thermal loading capabilities of the CTSTRUDL computer code. The effect on the support of liner expansion due to internal pressure was evaluated using an equivalent temperature gradient. -
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Reactions on Containment Vessel: The reactions from the cable tray support analysis were applied to the contalmsent vessel model to insure that the vessel and its stiffeners were within allowable stresses.
The analysis procedures as described above represent a comprehensive 4
analysis of the supports and the vessel and reflects current industry analysis techniques.
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- 17.b. Analysis Pr:ccdurce for Original Cable Trcy Support Designs The analysis of the cable tray supports is described in section 3.10.2 of the FSAR (Attachment B). In addition, the dynamic response of the cable trays and.their supports was originally analyzed by hadd calculations using the elastic response spectra
, modal analysis technique applied to a single degree of freedom model. Unbroadened spectra for 1 percent damping were used. A 4
structural model that lumped the mass load of the cables (80 lb/f t) on each support was evaluated.
- 3. DBA Response Spectra:
The DBA analyses are described in section 3.8.2.4.2 of the SQN FSAR.
'4. Walkdown Procedures A walkdown of the cable tray system was perf ormed to establish actual tray loading. Cable profiles were recot ded at all points where significant numbers of cables entered or ezited the trays attached to the SCV. Nessuranents to determine cable cross sectional area were taken on both sides of where the cables entered or exited the trays.
In addition, measuranents of the cross section of the exiting / entering cables were also taken. Actual tray loading was then calculated from the profile measurenents.
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... t. IV. C. Cable Trav Suncorts on the Reactor Buildian Shield Wall
- 1. Scope The reevaluation of the supports attached to the shield wall is required as's part of the resolution of SCRSQNCD8622.
The 'approximately 400 supports attached to the interior of the Reactor Building Shield Wall were evaluated through usage of enveloping techniques to identify worst-case /most critically loaded q, supports. The enveloping technique consisted of placing supports i into groups of similar configurations. Calculations were then performed for the controlling support in each group. The supports consisted primarily of cantilever and braced frame type steel structures that are welded to plates attached to the concrete wall with expansion anchor bolts. A small number of the supports were welded to plates embedded in the shield wall. All of the supports ~
were found to meet interim acceptance criteria and, thus, no support f
modifications were required. -
4 2a. Analysis Procedure f or Interim Acceptance Criteria:
Dynamic Response:
The dynamic response of the structural system composed of the cable trays and their supports was calculated by multiplying the supported mass by the peak of the applicable response spectra. A frequency analysis of the support was performed to ensure predominantly singic DE02;036211.02 1
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The cable tray supports were modeled using elastic bean elements.
- w The mass of the cables and tray was lumped on the supports. The i
flexibility of the support baseplate was modeled using spring constants determined by the stiffness of the . anchor bolts.
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ja } - Baseplate and Concrete Anchorage:
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7 The support reae.tions were applied to baseplate / anchorage models (using the BASEPLATE II option of the ANSYS computer code. Field measured anchor locations were evaluated co determine the worst -
i condition for analysis. -
The analysis ^ procedure as described represents a comprehensive ,
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evaluation of the supports and reflects current industry analysis l techniques. -
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2b. Analysis Procedure for Original Support Design:
The analysis procedure utilized for the original design of the cable tray supports on the shield wall is the, same as tbst previously provided in IV.B.2.b. of this report for the supports on the SCV.
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All cantilevered and braced free type supports were walked down and evaluated for any occurrences of the following: potential cable tray overloading, insulation on supports, undocumented attachments to supports,'C-32 anchor bolt violations, and any other noticeable deviations. In addition, the worst-case cantilever supports determined from a design standpoint were also walked down and avaluated for anchor bolt and attachment locations- on the plates.
The supports attached directly to plates embedded in the wall were also visually inspected but no deviations were identified that required consideration in the evaluation of the supports.
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IV. D.
, , , ALL OTHER CABLE TRAY SUPPORTS
, 1. Scope The reevaluation of other cable tray supports is required to resolve ECR SQNCEB8622.
There are approximately 2900 cable tray supports in the Category I structures (excluding the Reactor Building annulus area which has been previously defined). A majority of them are in the Auxiliary Building and Control Building, approximately d
1700 and 850, respectively. Thirty worst-case configurations were identified for evaluation from a design perspective. Some worst case supports were identified by a drawing review and -
utilized enveloping techniques to determine the supports that would transmit the largest loads to the baseplates. Other worst '
. case supports were chosen because they are unique and appear to be heavily loaded. There were 10 worst-case configurations
-i selected in the Auxiliary Building representing 148 supports,10
, in the control Building representing 26 supports, 5 in the Raactor Building, representing all supports, and 5 in the Diesel
~
Cenerator Building, repretenting all supports. These worst-cast \
I supports envelope the design of all cable tray supports.
The supports consist primarily of cantilever and braced frame type steel structures that are welded to embedded plates in concrete or surface mounted plates attached with expansion anchors. The worst-case support evaluation is in progress using "as-constructed" information.
DE02;036211.02
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_ l i
- 20. Analysis Procsd:rcs (Interin Acesptcaca Critsria)
- v Dynamic response: The dynamic response of the structural system l composed of the cable tray systems are being analyzed on the GTSTRUDL program using the elastic response spectra modal analysis technique. The SSE is being evaluated using the 10 percent broad.ened and a 5 percent or 7 percent damped spectra.
Computer Model: The cable tray supports are modeled
. - using elastic beam elements. The tributary mass, of the cable
. tray is lumped r.t the centerline of the cable tray on the
^
-~
bracket.
Baseplate model: The baseplates are modeled using the finite element prograu BASEPLII which processes the ANSYS computer program.
The analysis procedure as described above utilises a-comprehensive approach which reflects usage of current industry analysis techniques.
2b. Analysis Procedure for Original Design The analysis procedure for the original design of the supports is generally described in the FSAR as previously identified.
The dynamic response of the cable trays and their supports were originally analyzed by various methods such as hand calculations and response spectra modal analysis. The unbroadened response DE02;036211.02
r e .
spectro for th2 structura fcr cithcr 1 parcszt er 2 percent damping were used. A structural model that lum;ed the mass of the cables. (ranging from 45 lb/f t to 80 lb/f t) en each support was evaluated.
- 3. Walkdown Procedures:
The 30 worst-case supports were walked down to determine the "as constructed" configuration for evaluation. The "as constructed" configuration included identifying any occurrences of th.e following:
overloaded cable trays, insulation on supports, undocumented
~~
attachments to supports, location of support on baseplate, other attachments on baseplate, G-32 violations, span lengths, and any other noticeable deviations. Evaluation of this information is in progress.
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9
, , Y. Small Dianotcr Pipirs (4 Inch cad Less) Qualified by " Cookbook" s'
- Procedures - Analysis and Support Design I
A.
Comparison with FSAR Commitments and Justification - Piping Analysis w
- 1. FSAR Commitment - Pioinn Analysis The code of record ior analysis of Sequoyah Nuclear Plant is ANSI 531.1,1967. Load combinations and stress limits u. sed to qualify category I piping systems are shown in FSAR table 3.9.2-5. Table 1 i -- is a tabulation retaining only the terus in FSAR table 3.9.2-5 that p
are normally applicable to small diameter piping qualified by cookbook (alternate analysis).
- 2. Exception / Justification TVA proposes three exceptions to the FSAR commitment to reevaluate small diameter piping for interim acceptance criteria. The FSAR commitment and interim acceptance criteria are indicated in .
,' Table 2. Each exception is justified as follows:
- a. Secondary Stress Evaluations Excention - Secondary stresses due to seismic anchor movements (SAM) and thermal (Th) plus thermal anchor movements (TAM) will be evaluated separately to equation 10 limits.
Dm2;036211.02 4
______n_.____.__ - - - - - - - - -
~
i, ._. --
se TABLE 1 1
CATEGORY I PIPING SYSTEMS LOADING CONDITIONS AND' STRESS LIMITS POR SMALL DIAMETER PIPING OUALIPIED BY COOKBOOK ANALYSIS NC-3652(1)
Plant Condition ' Load Combinations Stress Limits Equations Normal SLp+ SDL Sh 8 SST 8A 10
- - - - - - - - - - - - og - - - - - - - - - - - -
SLP+8DL+8ST 8A+8h II Upset SsL+ Sgt+ SI/2SSE 1.2 Sh 9(
Paulted SLy+SpL+ SSSE 2.4 Sh '9 where SyL = Longitudinal pressure stress.
S = Longitudinal bending stress due to dead load.
DL S1 /2SSE = Maxime bending stress due to inertial loadings of the 1/2SSE.
SSSE = Maximum bending stress due to inertial loadings of the SSE.
SST = Secondary stress due to thermal expansion and anchor movement stress associated with the 1/2SSE.
Sh = Basic material allowable stress at hot temperature.
, SA = Allowable stress range for expansion stress.
L Notes: (1) The code of record did not provide guidance or allowable stress for
, many of the loading combinations presently required by NRC.
(2) Seismic anchor movements (SAM) may be included in equation 9 in the upset condition. They are not included in the emergency or faulted condition.
DE02;036211.02 '
- * - -o l
I .
I t t 4 *
- TABLE 2 '.' ,
SEQUOYAR NUCLEAR FLANT
! INTERIM ACCEPTANCE CRITERIA Safety-Related Small Diameter Pinine Syst- - DIN SON-DC-V-13.4 1
1 Cperating Condition
} (Lona Tern /Interia). Loadina Combinations Stress Limits Code Eaustions i C2twal Long Ters - If SAM effects omitted from equation 9U.
- ] ________________,,_________________
l P + W + Th + TAM + SAM SA+8h II i ,
1 .
j Restart - Thermal loads and seismic anchor point movement loads can be evaluated separately.
- - - - - - - - - - - - - - - -or- - - - - - - - - - - - -----
P + W + Th + TAM ,
SA+8h II i S
P + W + SAM S +S 11 !
i A h .
Upset OBE load combinations will not be considered for restart.
1 i Fculted
- Lcng Teru - P + W + SSE 2.4Sh ( ) -
9F !-
Restart - Faulted allowable stress increased to Present code allowable.
) (1)
P + W + SSE 3Sh :$ 2S y 9F j Nstes: (1) Stress intensification factors (i) will be used for restart evaluations.
(2) Present code allowables may be used with primary stress indices.
01h191.02
, , Justification;
- r. 1.
- 1. The ASME code pemits the evaluation of Th plus TAM only to the full allowable of equation 10 because it is permissible to include SAM in equations 9 or 11.
- 2. For this vintage plant, it was general industrial practice to field route small diameter pipe. SAM was evaluated to i
equation 10 and thermal effects were considered negligible. Further, deadiced and pressure stress for small diameter pipe qualified by alternate analysis is
~
generally less than 0.25 Sh . This would result in a secondary stress allowable in equation 11 of 2.25 Sh.
- 3. It is reasonable to evaluate thermal and SAM separately
, because secondary stress allowables are based on 7000 331.1 permits a fatigue usage evaluation based on cycles.
the number of cycles of each stress.
- 4. Equation (10a) permits an allowable stress of 3Sc for any single nonrepeated anchor movement. In effect. this allowable could be invoked for the first earthquake occurrence since imposed thermal stresses would shake down on the first seismic cycle.
i l
I 1
l l
0 DID2;036211.02 i
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w w e s - a o vme -~ ,w-m -
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- b. OBE Evaluation Exception - OBE or upset inertial loads will not be evaluated for interim operation of Sequoyah.
Justification:
Sequoyah OBE piping analysis is based on 1/2-percent damping and SSE piping analysis is based on 1-percent damping.
Evaluation of this piping to' these conservative SSE limits will ensure the plant can be shut down saf ely.
- c. Faulted allowable . pipe stress Exception:
current code allowables stress will be used with stress
, intensification factor (i). .
Justification
- 1. This allouable stress is currently permitted by the ASME code, but used with primary stress indices (B). For straight pipe, primary stress indices (B) and stress intensification f actors (i) are equal. For elbows and tees, the ratio of B to i can approach 1.5. Screening to the f aulted limits proposed will identify any significant l
DE02 g036211.02
l
_, . - . -- * * * * * ' ' ~'
r.
strces pr:blems. Pipe passing this scrscaing critsrin has a high probability of being qualified by a acre rigorous analysis. '
- 2. A cookbook or hand calculation piping analysis has been shown to be inherently conservative by comparing the results to computer analysis. More supports are required by a cookbook analysis.
. . 3.
The criteria is based on conservative damping by comparison to variable damping presently permitted in the industry.
, 4. Tests presently beirg sponsored jointly by NRC and EPRI Indicate loads 20 times greater than ASME level D limits
, do not fail pipe. These test indicate seismic inertial loads act more like secondary loads. .They do not act in ~
one direction long enough to collapse a pipe. Fatigue stress intensification factors may be more appropriate than Primary stress indices.
5 1
- 5. Earthquake experience data reported in NUREG-1061 indicates industrial grade pipe does not fail during an earthquake if large anchor point movements are prevented.
- 6. A 2.4 Sh allowable permits stainless steel to. exceed 2.0 Sy at high temperature and carbon steel to exceed 1.4 Sy at high temperature. These allowables would apply 4
to the most important pipe. This limit is much more DE02;036211.02
.- f. r:strictivo f or low taperettro, Icss irportcat pipe thca
~
it is to high temperature pipe. The proposed limits are more realistic. (See paper published in the Journal of Pressure Vessel Technology November 1982, Vol.104/35/ by I 1
S. E. Moore and E. C. Rodabaugh). l
- 7. A2S ylimit was permitted for restart of the San Onofre l Nuclear Station Unit 1. This limit was used for computer l analysis of large pipe. Evaluation of small bore pipe (2-1/2 inches and less) by analysis was not , required for
. restart. (See NRC Docket, No. 50-206LS05-84-02-021.)
DED2;036211.02
- - ---. - . - . ... . . - - - . - - . _ _ - - - - - - . - _ - _- a.-... - - _ - _-
, - .,e**----#*'M*4 -**h - e-Y.B. Comparissa cith FSAR Commitacite cid T;chnicci Justificatics for Small Bore Fipe Support Design 1.0 Allowable Stresses The FSAR does not explicitly address allowable stresses for l
small bore pi. ping supports. The FSAR sections 3.9.2.5.2, and 3.7.3.14, Attachments C & E. require the use of the ANSI B31.1 1967 Fower Piping Code. The B31.1 code, section 120.2.4, prescribes the use of American Institute of Stee.1 Construction or equivalent. TVA equivalent is " Guidelines for Design of
[ -- Component Supports for TVA Classes A through D," CEB Report 2
80-75. In addition, FSAR Section 3.8.4.5.2, Attachment ]
which TVA uses for design of miscellaneous steel allows 0.9 Fy for the SSE load case. TVA has issued a new design criteria SQN-DC-V-24.1, for new designs and modifications entitled, " Location and Design of Piping Supports and Supplemental Steel in Category I Structures," which limits the stress to a maximum of 1.6 AISC. for the f aulted condition.
Excention i
The interim acceptance criteria (for small bore piping alternately analyzed) provides an allowable stress of 1.7 x AISC allowables for faulted conditions. In addition, designs are being evaluated for the SSE load case only.
i i
DID2;036211.02
,, _,.,__._ m_--.- --a.-
Jus tification ;
l This increased allowable and the consideration of the SSE load case only are justified for the following reasons:
- 1. The SSE condition will always produce greater structural s
response than the 1/2 SSE and will, therefore, be the controlling load case in terms of maintaining functionality
' ~
fol' lowing' a seismic event. Consequently, evaluation for SSE under the interim acceptan'ce criteria ensures the ability to bring the plant to a safe shutdown for any seismic event less than or equal to the SSE. .
4
- 2. The SRP 3.8.4 stress allowables (maximum of 1.7 times AISC allowables) have been used extensively in evaluations of plants under construction as well as for evaluation of older L plants. The use of these allowable atresses for evaluation under the interim acceptance criteria will ensure that stresses are maintained within acceptable industry I5 standards.
l
. 3. ASME Section III, Subsection NF, Appendix F,1974 edition, allows the use of a maximum stress of 1.88 x Appendix IVII allowables f or the f aulted loading condition, but not greater than 0.7 Su in tension. Stresses must also be less than 2/3 critical buckling in compression.
t 5
DID2 3 036211.02
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, , 4. The ANSI /AISC N690 Nucicer Coda c11ove th2 us2 cf a maximum
- r. s.
stress of either 1.7 x AISC or 0.7Fu, whichever is less for the f aulted condition.
In other comparable industry practices, San Onofre Nuclear Generating Station Unit I utilized similar structural steel allowables and consideration of only the SSE load case to demonstrate safe plant functionality.
- 2. Rigidity Raquirements .
^
Connarison to FSAR Commitment The FSAR, Section 3.7.3.14.6 (Attachment F) describes the need for a minimum moment of inertia requirement. This requirement is specified in CEB Report 80-75, as 1/16" deflection under ,
normal load and a 1/8" deflection under design load. This' requirement is addressed in SQN-DC-V-24.1.
t Exception d
. For interim acceptance criteria. TVA requires that supports which carry load primarily in bending meet deflection requirements as defined above for the first two support locations in either direction from rotating equipment.
DID2;036211.02
.' , Justification
. s.
Alternately analyzed pipe was inetslied using typical supports designed to the stiffness requirenents specified in CEB 80-75.
However, to insure that rotating equipment is protected from seismic forces induced by adjacent pipe, the first two supports adjacent to rotating equipment will be verified to meet the specified deflection requirements in SQN-DC-V-24.1.
- 3. Friction Loads
_. Comoarison to FcAn Ca-itment Friction is not explicitly defined in the FSAR, but in section 3.9.2.5.2 and 3.7.3.14.6 of the FSAR, reference is made to ANSI 1
, 331.1-1967 Fower Piping Code. TVA did not require the
~
consideration of friction loads in that the ef fect of friction loads due to taperature on small bore pipe supports is insignificant and thus can be neglected.
The interim acceptance criteria states that the of facts of friction loads do not have to be considered in the reevaluation of existing supports.
DID2;036211.02
e,_%~a,, -,-.. - - = - - - - + ~ *** *- * * ~
, , Justification
's e, =e TVA has done a study of WBN pipe support designs for the ef fects of friction loads. (
Reference:
From J. C. Standifer to C.
Wadewitz, " Watts Bar Nuclear Plant Unit 1 - Pipe Supports -
, Significance of Friction Force in Controlling Design -Resolution of Unresolved Item 5-390, 391/84-05-05", R26 860106 003.) The results of this study confirmed that the friction loads did not generally govern the design of the supports. TVA will' do a similar study for the SQN pipe support designs af ter restart.
~
i
, , The results of this study will, be utilized to confirm that the
-~
ef fects of friction loads on support designs are negligible.
. It should also be noted that the pipe supports have successfully experienced the friction loads due to thermal growth of the piping during plant operation and this load condition will not effect the integrity of pipe supports or piping during the faulted condition.
l j DE02;036211.02 -
---,-e.,~~ --
- . , - , - - - , , - , - - - . , , - . - ,y ,,__,.-,_,,.-,-y, ,-- - -- - ,- .----,--m,-,-,-w- , - , . - , ,,,-
vm~,,,c,.,-..--.-.,-,,,-,---r--.
, , 4. Componert Stad:rd Supports f *
, .as 1 Comoarison to FSAR Commitment Variable Sprinns Suncort The FSAR section 3.9.2.3.2 references ASME Section III ar.J ANSI R31.1 Codes, both of which recommend i 25 percent variability.
^
Exception to Sorinas For springs, we are not concerned with variability under normal conditions, but will check to insure the spring will not bottom 1
t :
out for the f aulted condition.
t
'Justifleation
(
j SQN has had no probless with springs near equipment or in runs of piping for the past five years of operation. Since springs only support dead load, they will not jeopardise the safety of
' the plant under the faulted condition.
Other Comoonent Standard Succorts l
l 1
The FSAR does not specifically state what load factor to use,
]
I but CEB Report 80-75 defines the load f actor for components. l l
1 DE02;036211.02 I *
)
.-- - . - - - -__s-.-.-.._.~.-. - - . - - - . - . - . . -
Exception 1
For all component standard supports except snubbers,1.7 times the normal load rating or the NF faulted load rating which ever is greater will be used for faulted conditions. -
Justification
- 1. MSS-SP-58, (1967 edition) uses a factor-of-safety equal to 5
. - against ultimate load. TVA by using a load factor of 1.7,
, would still have a f actor-of-saf ety of 2.94 against ultimate.
, 2. ASME Section III, Subsection NF (1974) allows a factor of 1.88 times normal allowable as defined in Appendix F
. (Section F1370) for the f aulted loading condition. This factor is applied to the load rating procedure in NF-3262 i for standard components.
I 3.
~
The NRC Standard Review Flan Section 3.8.4 for structures
- allows a maximum of 1.7 x A13C allowables for the f actored load condition which is similar to the 1.88 factor allowed by NF for standard components and steel.
l l
l
)
DIl023 036211.02 O 8
e .
, VI. ANCEOR BOLT LOAD CAPACITIES
,+. t.
TVA General Construction Specification G-32 specified minimum ultimate tensile capacities for self-drilling and wedge bolt anchors (revision 5 of G-32 was submitted to MRC OIE as a part of the response to Bulletin 79-02 in a letter to James P. O'Reilly dated July 5,1979). The.
specification requires qualification testing to assure that the anchors achieve the ultimate capacities in G-32. The requirements for qualification testing were added in 1977. The current allowable design loads for expansion anchors are equal to the required ultimate tensile capacity given in G-32 divided by a f actor-of-safety (4 for wedge bolts
~ --
and 5 for self-drilling anchors).
The allowable expansion anchor loads used for design at SQN were originally baed on the test results for Phillips Self-Drilling Expansion Anchors. The tests published by,the manufacturer and some tests perf ormed by TVA in 1972 were used to develop allowable design loads. A f actor-of-safety of 4 was applied to obtain working load i i
allowables. The tensile capacities from the TYA tests were 10 percent l
less than the manufacturer's t.ests. This reduction was included in the determination of the allowables.
A construction specification requiring use of Phillips or equal was issued on 1972. This specification also included requirements for improcess inspection and proof load testing of the self-drilling anchors. Onsite qualification tests were performed on several brands in 1976 to determine equivalency to the Phillips anchors.
DE023 036211.02
_ _ - , - . . - ,,.,.7-- . . - , , . . ,w- m , ,,.,..-.,,.-_p_..-,,, ,.,.,y, ,.m,. . - - , - - . _ . ,
, ,. Substitution cf redgo bsit cach:rs for s21f-drilling orchsrs ecs Permitted in 1976. The allowable loads for these anchors were based primarily on test results for Hilti Kwik-Bolts. A factor-of-saf ety of 4 was applied for working loads. A list of acceptable wedge bolt anchors was provided. Onsite qualification tests were performed on Thunderstud wedge bolts in 1976 since they were not included in the original list of approved wedge bolts.
As the result of NRC-01E Bulletin 79-02, the design standard for anchors was revised to provide a factor-of-safety of 5 for self-drilling' anchors and to require this.f actor-of-saf ety to be applied for
. ~'
all loading conditions. Revision of ultimate tensile capacities in G-32 was not necessary.
During NRC review of TVA's response to Bulletin 79-02, a response was prepared for an NRC question relating to the basis for anchor capacities.
This response was provided in a memorandum to L. S. Rubenstein. NRC-NRR. from L. M. Mills TVA dated February 1,1980.
In summary, the required ultimate tensile capacities for concrete expansion anchcrs were originally based on manufacturer's and TVA tests. The allowable anchor loads in the TVA design standard for anchors are equal to the required ultimate tensile capacity divided by the required factor-of-safety.
-3 8- DE02 036211.02
. -~, -~~"~~*~~#~' ~~
e .
. . . VII. QA/qc Proccdsres
,** *t.
A. Desima and Construction 0A/0C Procedures for Interim Accentance Criteria i
Desian .
Division of Nuclear Engineering Procedures Manual I
Cable Tray, Support Design - Sequoyah Nuclear Plant'- Design Criteria ,
, f or Category 1 Cable Tray Support ,Systens - Design Criteria l
SQN-DC-V-1.3.4 including Design input Memorandums 1. 2. . and 3.
, Steel Containment Vessel Design - Tennessee Valley Authority Design Specification SNP-DS-1705-9803 Sequoyah Nuclear Plant Units 1 and 2 Structural Steel Containment Yessels for the Reactor Buildings. 1968 ASME Boiler and Pressure Yessel Code.Section III.
Winter Addendum.
Civil Design Standard DS-CI.7.1 - General Anchorage to ' Concrete and Q11 CD 86-006-001 Supplanent 4.
Construction Construction Quality Assurance / Quality Control Sequoyah Nuclear Plant - Specification N2C-877 for Identification of Structures.
Systems, and Components Covered by the Sequoyah Nuclear Plant Quality Assurance Progras.
DID2 g 036211.02
. . . _ ~
i ..
,. 4
! Fabrication and Erections. American Institute of Steel Construction Specification.
Welding: TVA General Construction Specification G-29C - Process '
Specifications for Weldihg, Heat Treatment, No'ndestructive Oxamination, and Allied Field Fabrication Operations.
Bolting: TVA General Construction Specification G ASME Section
,, III and Non-ASME.,Section III (Including AISC, ANSI /ASME B31.1, and ANSI B31.5) Bolting Material.
Bolt Anchors: TVA General Construction Specification G-32 Bolt Anchors Set in Hardened Concrete The Phase II program (long term) will be implemented underuthe above QA/QC procedures, including any revisions and/or additions that are in effect at that time.
4 B. Design and Construction QA/QC Procedures for Original Design and
, Construction Design Division of Engineering Design Engineering Procedures Cable Tray Support Design: Sequoyah Nuclear Plant - Design Criteria for Miscellaneous Steel Components for Class I Structures - SQN-DC-V- -
1.3.2.
. , . Construetion m a* v..
Construction Quality Assurance / Quality Control: Drawings required certification that material conform to appropriate ASTM apecification and that f abrication and erection be performed in accordance with AISC.
Welding: TVA Ceneral Construction Specification G Process Specifications f or Welding, Best Treatment. Nondestructive Examination, and Allied Field Fabrication Operations.'
4 i
i Dt023036211.02
- --..~ - - - -. - - .
f.'..'.,
.s . .s TABLE 3.7.1-3 DAMP 1NC RATIOS USED IN ANALYSIS OF CATECORY !
STRUCTURES. SYP EMS, COMPOWENTS. AND SOIL FOR STRUCTURES LISTED IN TABLE 3.7.1-1 Damping Ratio. Percent of Critical Viscous Damping 1/2 Safe Shutdown Safe Shutdown item Earthquake Earthquake Steel Containment Vessel 1 1 1*
Concrete Shield Building 2 5 7 and Internal Concrete St ruc ture
\l Other Welded Steel 1 1 2
"~~
Stevetures
. Bolted Steel Structures 2 2 5
+ - Other Reinforced Concrete 5 5 7 St ruc tures Bolted or Nailed Wooden 5 5 5 St ruc tures -
t Damping for Determining 10 10 10 Amplification through Soils for Soll Supported Structures Vital Piping System ** 0.5 0.5 1 l3 aDamping values used when stress levels are at or near yield. All other damping values are for lower stress levels, i
- Response spectra were also computed for frequency variable damping of 5 percent to 10 hertz, decreasing linearly to 2 percent at 20 hertz, and remaining at 2 percent to 33 hertz as described in ASME Code Case N-411.
Variably damped spectra used in piping analyses were not modified to include higher damping permitted by U.S. KRC Regulatory Culde 1.61.
Variable damping was used in new analyses as well as reanalyses (reconcillation work) using the response spectrum method but not in time 3 history analyses.
To account for possible increased flexibility when Code case N-411 damping was used, piping system displacements were checked for adequate clearance with adjacent structures, components, and equipment. Also, involved equipment was checked to withstand increased motion.
Revised by Amendment 3 A 77AC&f ENT A
i
'." [ ', )*. . SNP ,
t2.- Analyses E For dead loads combined with live loads and for dead loads combined with 1/2 safe shutdown earthquake loads the designs are based on the allowable stresses of the AISC Specification on Structural Steel for Buildings. For desd loads combined with safe shutdown earthquake loads the stresses are.11mited to 90 percent of yield stressed for the material involved.
For the seismic loads the actual astural frequency response of each support is calculated and the appropriate seismic fro-quency amplication factor is selected from the floor response spectrum covered in Section 3.7.
Conduit Banks and Suonorts
- 1. Restraint Neasures
- -. The Category I electrical exposed conduit supports, and electrinal sonduit box supports have been designed to provide vertical and horizontal support for the spacing recommended la TVA Construction Specif!sa sion.
- 2. Analyses The Category I underground electrical coaduit banks which run from the auxiliary building to the diesel-generator bs!! ding, to the auxiliary cooling towers, and to.the pumping station j were analyzed by two methods.
l The conduit banks were first analyzed as a beam with 4 unconstrained ends, on sa elastic fosadation and were found to have the same motion as the soll deposit la which they
, were buried. ,
Secondly, the soil deposit was them a ssumed to be an infinitely long uniform soll deposit resting on a rigid ,
foundation which responds to earthquake motion by moving a !
continuous sinusoidst plane wave. The displacement of the soll is l
i l
i l
- A 7 TA G HP12 N T B \
i l l
3.10-7
. . . 4. Beeldu31 heat removal system 7~ ~ *. t: . . l3
- 9. Component cooling system V
- 10. Essentini raw cooling water
- 11. Auxiliary boiler piping - '
l2. Upper head injection piping
- 13. parts of other systems which require rigorous analysis.
l3 3.9.2.5.2 Analytl,tal Methods Leadinz Conditions and Stress Limits The design loading combinations and the allowable stress limits considered in the design of TVA piping systeau: within the scope of subparagraph 3.9.2.5.1 are shown La Table 3.9.2-5. Design loading combinations are categorized with respect to normal, upset, and faulted conditions. Piping components have been
~~ designed to allowable stress intensity levels given by the ANSI 331.1-1967 Power Piping Code.
While the referenced code did not define stress levels for the loading combinations considered in Table 3.9.2-5, the allowable stress intensity levels are in agreesset with subsection WC3000
- of the ASME,Section III. Winter 1972 Addenda. The referenced i
subsection is considered to be equivalent to ANSI 331.1 with appropriate consideration to the modifications where they exist. .
Analyses
- 1. Stress evaluations due to loadings such as deadweight, thermal expansion, and anchor movements are perforar.d using static analysis techniques, while stress avsluations due to earthquake loadings are performed using dynamic analysis .
techniques. The computer programaans for applicatism of both techniques is described in subparagraph 3.9.2.5.3.
- 2. Loads on equipment nossles are crabined and evaluated against allowables as follows
Fog, + FST + F1/2SSE i Alloweble
- 3. Seismic valve accelerations are maintained below 2 g vertical, and 3 s horizontal for the SSE condition.
3.9.2.5.3 procrams Used for Category T Ploins Analyses i
Dynaale analyses on TVA C1' ass A, 8. and C piping are perforved j using plSOL1A, TPIPE, or DACS. Static analyses en TVA Class A,
' B, and C pipings are performed using FISOL3A. TP!PE, PFA, or .
3.9-28 k' l
r- c1. c. m t. C g .
- ms~ --s--------------- .-e-ew,--mmn,-n..
' u - *? s.
.w...e The stresse: is sheer walls parallel to the direettom of the lateral earthquake forces in the Assiliary Bs:1 ding are as follows:
N : ssa Calculated Stress Allowable Elevation Safe Shutdove Earthenske Stresses (1bs/in s) horth-South East-Veat 669-690 146 112 250 690-714 104 120 250 714-734 98 to 250 734-749 90 96 250 749-763 92 62 250 763-778 ' 176 114 250 Above 778 234 60 250 r_, Stresses for 1/2 SSE are one-half of those tabsistad.
F.a r t h q u a ke shear stresses were insignificant la all other structures.
All Category I structures are esssatially low profile box structures with height to base ratios less than 1.0 and a high factor of safety against sliding or overturning ander the most severe loading conditions. In addition, all structures are auf.ficiently heavy that there is no flotation problem under
. maximum flood c ondi ti on s .-
3.8.4.5.2 Structural Steel Streeteral steel and welds are designed in accordance with AISC "Nassal of Stect Constraction." Seventh Edities, for Case I
, leading condition so that the stress la the members and connections do act exceed the allowable stress criteria as set forth in the February 1969. AISC 'Specifica tions for Design.
Fabrication. and Erec tion of Structural Steel f or Buil ding s . "
For the factor of safety of these allowable stresses with respect to specified minissa yield point of the material used, see Section 1.5 of "Commen t a ry on the Spe cifica tion f or the Design.
Fabrication, and Erection of Structural S t e st f or Building s .
- Both specification and c omm e n t a ry are included in the AISC "Nanual of Steel C o n s t r u c t i o n .
For Case 11 loading condition the actual stresses do not exceed the allowable s t re s se s as set forth in Table 3.8.4-2. The a l l ow a b l e stresses for Case II loading have a sinissa factor of safety of 1.11 based on the spe cif ie d minissa yield poin t of the asterial ased.
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o..o oNP-2 3.7.3.13 Interaction of Other Pleism with Cstenorv I Pinina i:
The seismic-indoced ef f e ct s of non-Ca te gory I pipin g sy s t em s on Category I piping is accounted for by faclading in the a naly si s of the Ca t o g o ry I pi pin g a length of the non-Ca t e g o ry I system equal to at least the,first se i sm ic r e st r aint or anchor beyond the point of change'la cl a s si f ica tion. Norms 11y, a v alve serves as a s ei smic-mon s si smic bo tada ry la a fi nid sy st em. Th e valve ca pa bil i ty to maintsia a pressure bo un da ry in the event of a l sei smic event is assured by sei smically de signing piping on the I moncl as sified side through the first seismis restraint or anchor beyond th e valve.
. - 3.7.3.14 Field Location of Sannorts and Restrsists Criterin have. been developed f or field use in loca ting supports for Category I, TV A Cl a s s B, C, and U proce ss and instrument piping. The applicability of the criteria according to line size, schedule, temperature, prs sure, and loca tion is described la Se c tion 3.9.2.6. 2 1
The criteria was based upon a de tailed smalytical st udy th a t !
ev alua ted the important pareaetors which included dead and live l weights, seismic, and th e rm al con sidera tions f or a range of f,
operstlag temperatures and a giv en n aziman ope r a tin g p r e s sur e.
1 In order to maintala stresses in the piping compone nt s to well w i th in the ANSI B31.1.0-1967 Pow er Piping Lode limits, support r equir eme nt's w ere determined for a r a'a g e of line sizes and m a te ri al s that are sted in the pl ant.
L 1
Data was generated for each pipe size by assasing a straight ran ,
of pipe re s traine d f rom mov eme nt by la teral support s. The pipe p length a s s um e d w a s four horisontal spans. A vertical support was 3 assumed at each la teral support and a t eq u al spaces between l a te ral supports when the vertical spacing was excee de d. The period, ma xim aa seismic displacement, m a z is am stress, and reactions were de t e rn ise d for the range of lateral support spacings considered si gnifi ca n t. Th e m a x im an par ticipa tion by the dynam ic model was conservatively assamed to occur in the first mode. An envelope spectra was used as the horizontal spectra, and the v ertical spectr a was a s s um e d t o be a constant vaine equal to the highe st attainable acceleration corresponding to th e mo st a c t iv e natural frequency all owed by the ver tical support spacing.
The m aximum re sul ting l oa ds were extr apolated t o the peak of the acceleration spectra and used a s sei smic p13: de a dw ei gh t l oa ds .
Prim a ry stre sse s were con sidered by combialms th e ma ximum re ss1 ting seismic, pl us' de a dw ei gh t stresses, with the pressure l st res s r e sul ting from maximum allow ed pr e ssures, and comparing to l
l r the code allowables for all as t e ri al s considered.
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- 6. The support shall be designed so that the inertial loading from the system is not amplified by seismic response of the support. This is accomplished by e stablishing minimum soment af inertis requirements.
To ensure that the criteria is sorrectly app 11ed and that limits are not exceeded, a step-by-step procedure is developed in the criteria skish the field personnel f ollow. The procedure consists of three maj or sections:
- 1. Support toes tlos guide with all de sign rules f or the correct support procedures, accompanied by a is11 illustration in the use of the rules. Daly spring hangers, rod hangers, and rigid supports ar's seasidered.
- 2. Flexibility analysis guide with all applisable cases considered in tabular f orm, secompanied by a full illustration in the use' of all rules.
- 3. Support selection guide with all applicable sase geometries fully illustrated sad the structural desism requirements incorporated into the tabulated values. A load table and l selection procedste is presented which acconats for dead i load, seismie, and thermal loads induced on a support by l various a dj ac e n t pipe geometries. Final selection record of f
- supports is kept in the form.of a support workshee t whleh laclades s up'po r t. geometry, loads used, and seometry of piping supported.
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h 3.7.3.15 _Seisnie Analysis for Fuel Elements. Control Rod q Assemblies, and Control Rod Drives 5
Fuel assembly component stresses induced by horizontal seismic disturbances are analysed through the use of finite element computer modellas. The time-history floor response based on a standard seismis time history moraalised to SSE levels is used as the seismic input. The reactor internals sad the f uel a s sembl i e s are modeled as spring and lamped ma ss systems. The seismic response of the f uel assemblies is smalysed to determine design adequacy. A deta iled discussion of the analyses perf ormed f or .
typical fuel assemblies is sontained in Reference 16.
The Control tod Drive Nechanisms (CIDM) are seismically analyzed to confirm that system st re sse s mader seismic conditions do not
, exceed allowable levels as defined by the ASME Boiler and j Fressure Yessel Code, Se c t i on III f or Up se t ' and "Faul t ed 8 l conditions. Based on these stress criteria, the allowable seismic stresses in terms of bending acaents la the structure are determined. The CRDM is ma thema tic ally modeled a s a system of lumped and distributed masses. The model is analyzed mader ,
appropriate seismic excitation and the ressitant seismic bending
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3.7-45
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