ML20140F036
| ML20140F036 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/21/1985 |
| From: | Le A CALSPAN CORP. |
| To: | Shaw H NRC |
| Shared Package | |
| ML20140F040 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130 TER-C5506-325, NUDOCS 8506250361 | |
| Download: ML20140F036 (49) | |
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l TECHNICAL EVALUATION REPORT l
NRC DOCKET NO. 50-254, 50-265 FRC PROJECT CSS 06 N RC TAC NO. --
FRC ASSIGNMENT 12 NRC CONTRACT NO. NRC-0341 130 FRC TASK 325 AUDIT FOR MARK I COffrAINMENT IDNG-TERM PROGRAM -
STRUCTURAL ANALYSIS FOR OPERATING REACTORS COMMotOfEALTH EDISON CCMPANY QUAD CITIES NUCLEAR GENERATING STATION UNITS 1 AND 2 TEE-C5506-325 Prepared for Nuclear Regulatory Commission FAC Group Leader: V. 21. Con Washington, o.C. 20555 NRC Lead Engmeer:
M..shaw June 21, 1985 This report was prepared as an account of work sponsored by an agency of the United States Go,ornment. Neither the United States Government nor any agency thereof, or any of trew empfcyees, makes any warranty, expressed or implied, or assumes any legal liabll:ty or responsibility for any third party's use, or the results of such use, of any informathn, appa.
retus, product or process disclosed in this report, or represents that its use by such third party erould not infringe privately owned rights.
Preparea by:
Reviewed by:
Approved by:
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Vu Yw h f h' W
Principal Author Group b ader os,.:s/niretor epartme o l
er os,.: a.zo-se os,.:
i->. n FRANKUN RESEARCH CENTER OtVtSION OF ARVIN/CALSMWs som a esce stseers.mentaastama.pn mes 85dDE636/.N+
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TECHNICAL EVALUATION REP' ORT l
NRC DOCMET NO. 50-254, 50-265 FRCPROJECT Casos NRC TAC NO.-
FRC.4SIGNMENT 12 I
NRC CONTRACT NO. NRC-0341 130 FRC TASK 325 i
A AUDIT FOR MARK I C00ffAINNENT IONG-TEppl PROC 31AM -
STRUCTURAL AMkLYSIS FOR OPE 3tATING lashC'! ORS, COISEDInlEALTSI EDISON COMPANY j
QUAD CITIES NUCLEAR GENERATDIG STATICEI mtITS 1 AIED 2 TER-05506-325 t-g Prepared for Nuclear Regulatory Commission FRC Group Leader: V. N. Con Washington, D.C. 20555 NRC Lead Engineer:, H. Shaw June 21, 1985 i
This report was prepared as an account of work sponsored by an agency of the United States Government. Nettner the United States Government nor any agency thereof, or any of their employees, makes any werranty, empreteed or implied, or assumes any legal liability or roeponeitHilty for any third party's use, or the results of such use, of any information, appa-retus, product of process disclosed in this report, or represents that its use by such third i
l party would not infringe privately owned rights.
P Prepared by:
Reviewed D)!
Approved by:
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- v. 4 k
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Principal Author Group b ader opartmerfDiretor
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Date: n -lo #f Date: J-28 fI Date:
l FRANKUN RESEARCH CENTER OfWISION OF ANViet/CALSfmN seen a eace stuusts pseuespina.pn mies
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TER-C5506-325 i
i CONTENTS f
I 8ection Title Pg 1
1 INTRODUCTION 3
2 2
AUDIT FINDINGS.
5 1
3 CONCLUSIONS.
1 6.
4 REFERENCES.
l APPENDIX A = AUDIT DETAILS APPENDIX B - ADDITIONAL INFolethTION REQUIRED i
APPENDIX C - TECHNICAL REPORT ON THE USE OF THE CMODF PROGRAM IN THE MARK I TORVS ATTACHED PIPING ANALYSIS 1
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8 TER-C5506-325 1
FOREWORD e
l This Technical Evaluation Report was prepared by Franklin Research Center i
under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
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TER-C5506-325 1.
INTRODUCTION The capability of the boiling water reactor (BWR) Mark I containment i
suppression chamber to withstand hydrodynamic loads was not considered in the original design of the structures. The resolution of this issue was divided into a short-term program and a long-term program.
Based on the results of the short-term program, which verified that each Mark I containment would maintain its integrity and functional capability when subjected to the loads induced by a design-basis loss-of-coolant accident (LOCA), the NRC staff granted an exemption relating to the structural factor of safety requirements of 10CFR50, 55(a).
The objective of the long-term program was to restore the margins of safety in the Mark I containment structures to the originally intended margins. The results of the long-term program are contained in NUREG-0661
[1), which describes the generic hydrodynamic load definition and structural acceptance criteria consistent with the requirements of the applicable codes and standards.
The objective of this report is to present the results of an audit of the Quad Cities Nuclear Generating Station Units 1 ant: 2 plant-unique analysis (PUA) report with regard to structural analysis. The audit was performed using a moderately detailed audit procedure developea earlier (2) and attached to this report as Appendix A.
The key items of the audit procedure are obtained from " Mark I Containment Program Structural Acceptance Criteria Plant f
Unique Analysis Application Guide" [3], which meets the criteria of Reference 1.
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I M?t-c5506-325 2.
AUDIT FINDINGS A detailed presentation of the audit for Quad Cities Units 1 and 2 provided in Appendix a, which contains information with regard to several key i
items outlined in the audit procedure (2]. Based on this detailed audit, it was concluded earlier that certain items in the Dresden Unita 2 and 3 PUA report (4] indicated noncompliance with the requirements of the criteria [3]
and that several aspects of the analysis required further information. Based on this conclusion, the Licensee was requested to provide information with regard to the items contained in Appendix B of this report.
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During the course of reviewing the analytical techniques for stress calculations of the torus attached piping systems, Franklin Research Center (FBC) staff raised concerns regarding the verification of the computer program CMDOF (Coupling of Multiple Degrees of Freedom), which was used by the NiffECH technical staff to qualify the Mark I torus attached piping systems in a i
number of nuclear power plants. Meetings were held with NUTECH technical staff and representatives of affected utilities to discuss and resolve concerns associated with this program.- In accordance with an FRC request for additional study to verify the program, the Monticello plant used some 4
in plant safety relief valve tests performed in 1980 for verification purposes, and the results of this study were found acceptable. This assessment is also applicable to Quad Cities Units 1 and 2, Appendix C of this I
report provides the background and assessments relating to this program. The Licensee has responded (5) to all the items contained in the request for additional information (Appendix B)r a brief review of each response is provided below.
4 i
j Request Ites' 1 In this response, the Licensee indicated that the wetwell-to-drywell vacuum breakers were modified and evaluated according to ASME Code Class 2 criteria and that an overview of this analysis has been submitted to the NRC.
Regarding safety relief valve (SRV) discharge line vacuum breakers, the Licensee indicated that they were replaced with valves qualified in accordance 2
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a TER-C5506-325 l
with the ASME Code Section III, Subsection NC, 1977. Since the criteria for vacutum breaker modifications are not addressed in Reference 3, the vacuum i
breaker evaluations are outside the scope of this technical evaluation report (TER). his issue will still be awamined as part of the Mark I Long-Tern 8
Program and will be addressed in a separate TER.
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i Request Item 2 In this response, the Licensee showed that the AISC specification was more conservative than the ASME Boiler and Pressure Vessel Code with respect to the analysis of the SRV discharge line supports by providing a comparison of allowable stresses derived from each. The comparison indicated that the j
ASME Subsection NF allowable streses were 40% to 68% higher than the AISC allowable stresses. The Licensee's response is satisfactory.
Request Item 3 l
In response to this item, the Licensee confirmed that all large bore and small bore piping systems were classified as essential. Also, all active pumps and valves were evaluated for operability and are considered operable.
The Licensee's response has resolved this concern.
I Request Item 4 In this response, the Licensee provided a summary of the method for applying the 10% rule that exempted some small bore pipes from analysist the i
summary is listed below.
o At the small bort piping attachment point, the stresses in the large e
bore piping due to combined Mark I loads,were calculated.
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he large bore piping stress combinations for Imvels B, C, and D were compared against 10% of the respective allowables. Stress
{
intensification factor values were also included where applicable.
o Any small bore piping connected to large bore piping that met the 10%
rule at the attachment point was then exempted from further Mark I evaluation.
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TER-C5506-325 The Licensee has also provided a table showing the distance from the j
torus along each large bore line to the point at which the 10% rule cosas into effect. The Licensee's response indicates that sufficient calculations have i
been made to ensure compliance with the 10% rule of section 6.2d of the criteria [3].
i g est Item 5 t
In this response, the Licensee indicated that some equipment at the Quad -
Cities plant was qualified by the 10% rule of Section 6.2d of the criteria
[3]. A susmary of the method for applying the 10% rule at equipment nozzles was also provided: the susmary is presented belows At the pipe-to-nozzle junction, the piping stress due to combined o
Mark I loads was calculated.
I Stress combinations for Levels B, C, and D were compared against 10%
o of the respective allowable. Stress intensification factor values were also included in the stress combinations where applicable.
q.
The Licensee's response has resolved this concern.
Request Item 6 In response to this item, the Licensee stated that the results of the suppression chamber analysis for lateral asynsietric loads used in the Quad 8
Cities plant-unique analysis report envelop those that would have been obtained using a 180* model of the torus. Bounding values of the lateral loads were developed using the maximum spectral acceleration and maximum dynamic load factors. The resulting loads were added absolutely and were assumed to be transferred by two of the four seimliic restraints. Stresses in the suppression chamber shell and column / saddle desembly caused by asymmetric lateral loads are small compared with those caused by other major torus l
loads. The Licensee's response is satisfactory.
Request Item 7 In this response, the Licensee asserted that, despite the proximity of certain stress results to allowable limits, the margins of safety of the _ _ _
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i TER-C5506-325 original design have been restored or increased. The following reasons were given: the-code allowable limits provide adequate factors of safetys stress results represent peak values which occur over a tiny area of the structure; loads are conservatively defined based on test results; and conservative load combinations are used, in which peak responses are aesumed to occur simultaneously. This response is satisfactory.
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3.
CONCLUSIONS Based on the audit of the Quad Cities Units 1 and 2 plant-unique analysis report, it was concluded earlier that certain aspects required additional information. Based on the Licensee's responses [5] to the request for additional information, it is concluded that the Licensee's structural analyses with regard to major plant modifications and the torus-attached piping conform to the criteria requirements. With reference to the verification of the computer program CMODF used to quilify the torus attached piping systems, the results of a verification study (based on U Monticello in-plant safety relief valve tests) perforleed by NUTECH tech'dcal staff were found acceptable as documented in Appendix C of this report. The Licensee's approach to the evaluation of piping fatigue conforms to the approach reccannended by the Mark I Owner's Group, which has been accepted by the NRC.
I The evaluation criteria of the containment vacuum breaker modifications are not addressed in Reference 3 and are therefore outside the scope of this TER; however, this issc* will still be examined as part of the Mark I Long-Term Program.
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4.
REFERENCES j
1.
" Safety Evaluation Report, Mark I Containment Iong-Term Program nasolution of Generic Technical Activity A-7" i
Office of Nuclear Reactor Regulation USNRC 1
July 1980 2.
Technical Evaluation Repose Audit Procedure for Ms.x I Containment Long-Term Program - Structural Analysis Franklin Research Center, Philadelphia, PA June 1982, TER-C5506-308 f
3.
" Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide" l
General Electric Co., San Jose, CA October 1979 l
4.
Quad Cities Nuclear Generating Station Units 1 and 2 3
Plant Unique Analysis Report, Revision 0 Commonwealth Edision Company Nutech Engineers, Inc.
May 1983 5.
B. Rybak I
Letter to H. R. Denton (NRC) with Attachment
Subject:
Response.to Questions Ccncerning Mark I Containment Plant-Unique Analysis l
Commonwealth Edision Company l
l March 1984 I
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APPENDIX A
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AUDIT DETAILS I
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o FRANKUN RESEARCH CENTER DM90N OP ARVIN/CALSPAN 20th & RACE STREETS, PHILADELPHIA.PA 19103 I
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I TER-C5506-325 1.
INTRODUCTION l
The key items used to evaluate the Licensee's general compliance with the requirements of NUREG-0661 [1] and specific compliance with the requirements of " Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysia Application Guide" (2) are contained in Table 2-1.
This audit procedure is applicable to all Mark I containments, except the Brunswick containments, which have a concrete torus.
i For each requirement listed in Table 2-1, several options are possible.
Ideally, the requirement is met by the Licensee, but if the requirement is not met, an alternative approach could have been used. This alternative approach will be reviewed and compared with the audit requirement. An explanation of why the approach was found conservative or unconservative will be provided. A column indicacing ' Additional Information Required
- will be used when the information provided by the Licensee is inadequate to make an assessment.
A few remarks concerning Tables 2-1 and 2-2 will facilitate their future uses o A summary of the audit as detailed in Table 2-1 is provided in Table 2-2, highlighting major concerns. When deviations are identified, reference to appropriate notes are listed in Table 2-1.
o Notes will be used extensively in both tables under the various columns when the actual audits are conducted, to provide a reference I
that explains the reasons behind the decision. Where the criterion is satisfied, a check mark will be used to indicate compliance.
o When a particular requirement is not met, the specific reasons for noncompliance will be given.
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1 _-
NRC Contract No.NIC4541c130 j
l A 'lin Raesarch Center 100llFrenii PRC Project No.C8008 Page PRC W g./ 1 g
I A onessa es The Frenideiinemus.
PRCTask No. 3 am and name seese. Phas.. Pa.19103 (2156 4401000 PsentName %o Cmes Uunis I d2.
Table 21. Audit Procedure for Structural Acceptance Critoria of Mark iContainment 1.ong Term Program Ucensee Usee Section Keyitems Considered N
Addtl.
Alternate Approach No. [2]
In the Audit Not Info.
NA Remarks Conser-Unooneer-l gg g
utive wellve I
1.2 All structural slee.nts of I
the vent system and suppres-sion damber must be considered in the review.
l Se following pressure r6taining elements (and their supports) must be considered in the reviews o trus shell with associ-ated penetrations, reinforcing rings, and support attae ments o trus shell supports to
[
the containment structure o Vents between the drywell
_y and the vent ring header 7
(including penetrationa therein) o Begica of drywell local
[
to vent penetrations I
l 0 Bellows between vents and
/
l torus shell (internal or l
external to torus) l o hat ring header and the downconers attached to it o vent ring header supports to the torus f
grF LICAM3fE'S I
l o vacuum breaker valves Magr gr3Ao45r HAS l
attached to vent penetra-j p W D D ti.5 tions within the torus gggy (where applicable)
N#### ####A~##
o meuum breaker piping
-)(- W4VEN04//r;dfTm#
systems, including eacuum AscM/4E W breaker valves attached 8@8 # I to torus shell penetra-re l.
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NRC Contract No. NRC641-1.J Frenidin Research Center FRC W No.2008 PeGe l
A Dauemon ed The Frankkiinasuse PRC W d
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- m. m.s r,*..r. me ms,wiam p
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i Tame 21. Audit Procedure for Structurel Acceptance Cetterte of Mark I Contaitunent Long-Term Program I
r Ucensee Usee e Section Keyitems Considered N
Addtl. Alternese W No. [2l In the Audit Not info.
NA Romerks
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Wet Met Rogd.
wellwe wellwe 1.2 (Cont.)
tions and to vest penetrations external to i
the torus (where l
applicable) o Piping systems, including pumps and valves. internal to the torus, aEtacized to the torus shell and/or l
vent penetrations o All main steam system
[
safety relief valve (SRV) piping o Applicable portions of
/
l the following piping systems:
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- Active containment system piping systans I
(e.g., emergency core cooling system (EG) and other piping required to j
maintain core cooling af ter losnf-coolant accident (IDCA))
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- piping systems whicts provide a drywell-to-wetwell pressure dif-forential (to alleviate pool swell effecta) l t
j
- other piping systems, I
including vent drains o supports of piping systems [
mentioned in previous item
[
o vent header deflectors including associated hardware 1
,1 A.
NRC Convest No.NRC.m.ei.ia UUUU Franklin Research Center PRC Project No.CAIDS Page N Assignment No. IE PRCTeek No. 3 z 5 4
A Dewomen of The Fsenhisiinusuas PlantName OVAD CtT7M5 UM/f5 [N 20th and Race Seeses. PMs.. Pa. 19103(2156 448 1000 l
I Table 21. Audit Procedure for Structural Acceptance Crtteria of Mark l Containment Lont-Te:m Program i
Liconese uses Section Keyitems Considered N
Adleid-AlterneseApproach No. [2]
in the Audit Not kh.
NA Romerks Consor. Unooneer-Met Met Resid.
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1.2 (Cont.)
o Internal structural elements (e.g., monorails, catwalks, their supports)
I whose failure might impair the containment function L 1 CGM 6GE'S RI!3PtA35E HAS 1.3
- a. The structural,,.
EtE504 VED 1hiS acceptance criterla g
Il for existing alark I cowApJ containment systems are contained in the I
American Society of I
lesct,anical Bigineers (ASid!) Boiler and Pressure 9essel (3677) Code,Section I
III, Division 1 (1977 Edition), bith addenda through the l
Summer 1977 Addenda
[3] to be referred I
herein as the Code. The alternatives to this i
criteria provided in Drference 2 are also l
.coe,tah1e.
l
- h. i.,.a o 1ete a,pu-g cation of the criteria b
(ites 1.3a) results in hardships or unusual difficultiea without a compense-ting increase in level of quality and safety, other structural acceptance critaria any be used after approval by the Baclear angulatory Commission.
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,I NRC ConWect No.NRC43411J W Franlun Research Center FRC 6 No.C0008 Page lI A Dhamma elThe Freddwiinesmas M
No. S FRCTask No. 32 5 5
aedi end Race s
, Phs.. Pa. 19:03 (215e 44410o0 MontName hD 0T78:3 (ANTS f / 2-g Table 21. Audit Procedure for Structural Acceptance Criterls of MarkiContainment WTerm Program useneeeUsee n
A.ed.
A,or,.,e _
No.
h the Audit Not info.
NA Romerks I,
[fl Conser-unooneer-met uet Reed.
l 2.1 a.
Identify the code i
or other classification of tne structural element b.
Prepare specific di==naional boundary I
definition for the specific Mark I contain-ment systems (mtes Wolds connecting piping to a nossle are piping welds, not class MC welds) 2.2 (kaidelinee for classification of structural elements and I
tr=>adary definition are as follows:
(Refer to hble 2-3 and j
Sable 2-4 for non-piping and l
piping structural elements, l
respectively, and to item 5 l
in this table for row designations used for I
defining limits of hr=trwta rio s) a.
tras shell (tw 1)
I
.e torus.e.hr.e in ocabination with reinforcing rings, g
penetration elements within the M-3334 [3]
limit of reinforce-ment normal to the I
torus shell, and attactuneet welds to the inner or outer surface of the above membera but not to nossles, is a Class NC [3] vessel.
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PRCTWNo. 3M h
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Table 31. Audit Procedure for Structural Acceptance Crisorts of Mark i Containment Long-Term Program l
1 Ucensee Usee Secton Mey Nome Considered N
Addtl.
ANornate Approach No. [2]
in Wie Audit Not info.
NA Romerks Consor-Unooneer-yg g
veuve votive 2.2 (Cont.)
[
b.
meus shell supports (kna 1) - subeection NF (3) support structures i
i between the torus shell l
and the building strueture, esclusive of the attactunent welds i
I to the torus shells welded or med anical attaements to the l
building structures s
(escluding hts);
and seismic constraints l
between the torus shell l
and the building structure are Class NC
[3] supports.
c.
kternal vents and
[
vent-to-torus houows (Epw 1) - Me esternal l
vents (between the attaement weld to the I
drywell and the attaement weld to the bellows) including:
vent penetrations within the EB-3334 [3]
limit of reinforcement normal to the vent, internal or external attdmaat welds to the esternal vent but act to nossles, and the I
vent-to-torus bellows (including attaement welds to the torus shell and to the external vents) are Class HC [3] vessels.
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NRC Convect No. NRCH2 81-1L) a UUR Frankhn Research Center PRC Project No.C3008 Page I
FRC Assignment No.12.
A h W The Fem h l
FRC Teek No.31"a Plant Neme 6 (,MD df 7"/ES UMr7% ffb som and nm. s Phd... Pa. 19103 (2154 MS 1000 g
i Table 31. Audit Procedure for Structural Acceptance Crtterte of MarklContainment Long-Term Program I
ucenee use.
Secton Keyitems Considered Criteria Addtl.
ANornato Appromoh No. [f]
In the Audit Not info.
NA Romerks Consor. Unooneer-gg g,
l 4
I 2.2 (cont.)
d.
Drywell-vent connection region (Bow 1) - Vent l
welded connections to I
the drywell (the drywell and the drywell region of interest for this l
program is up to the MB-3334 (3) li5it of reinforcement on the drywell shell) are Class MC (3) vessels.
e.
Internal vents (kars 2
[
I and 3) - Are the continuation of the vents internal to the torus shell from the vent-bellows welds and includes the cylindrical shell, the closure head, penetrations in the cylindrical shell or closure head within the m-3334 [3] limit of
(
l reinforcement normal to the vent, and attaement 4
welds to inner or outer l
surface of the vent but I
not to nossles.
f.
Dent ring header (Rows
[
1 l
4 and 5) and downoomers (Bow 6) - Vent ring header including the l
downoceers and internal l
i or esterna1 att.e.or,t welds to the ring l
l header and the attamment welds to the i
l downcomers are Class NC
[3] vessels.
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,1 gg NRC Contract No.MRC43 41413 si uu Franbri Research Conse, PRC Project No.C3005 b;s !
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IE A Denman of The Freihtn inamuse PRC
.5 6
i 3th and Rose h Phen.. Pa 19103 (2156 440-1cc0 PRC Teek No.
j PientName Ocan40 Qr.M WHT$ e h C.-
I Table 31. Audit Procedure for Structural Acceptance Cetteria of werk 1 Catninment Long-Term Prograrn
,,",,",,,",,"O seemn xer name Coneu. red cram addv.
l No.[2]
in the Audit Not info.
NA Memerks conser unoonese.
uet met Reet.
"8h*
M8h*
l 1
2.2 (Cont.)
l
[
- The portion of the dowacomer within the
,1 IEE-3334 [3] limit of reinforcement normal to
{
the vent ring header and portion of the vent I
ring header vi, thin 35-3334 limit of reinforcement arc l
ocasidered under hw 5.
g.
Dent ring header l
supports (Bow 7) -
l g
Subsection IEF {3]
l supports, exclusive of lI the attar *==at welds to the vent ring header and to the torus shell, are Class sec [3]
I supports.
1 LJtl*SVSEE'S t
h.
Essential (hws
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10 and 11) and M
ggggg g non-essential (hws 3
O ll 12 and 13) pi.)ing MOAM M d l
systems - A p ping system or a portion Il of it is essential if the system is necessary to a. u e the integrity of the reactor coolant i
pressure boundary, the capability to shut down the reactor and maintain it in a abutdown ocedition, or the l
capability to i
prevent or mitigate the consequences of I
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PRC Assignment
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FRCTeek No. 3 2
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Wh and Rene Seese. h. Pa. 19103 (21 S 448 1000 Totte F1. Audit Procedure for Structural Acceptance Cettetts of MarkiCentainment Long Term Program
'g t
M KeyitemeGeneidered AddW.
..m i,,,,e -
,.ot o
NA Rome,te conser-unooneer.
uet met Reed.
votive vanve 2.2 (Cont.)
I accidents witicks could result in i
potential off site esposures comparable to the guideline exposure of 10CFRKO [4]. Piping should be considered essential if if performs a safety-related role at a later l
time during the event I
combination being considered or during any subsequent event combination.
I L.Ic.trA/SEE 5 L.
Active and inactive
- 4 57tpA/5 d" #4 5 l
p vgD rMS component (Bows 3
.1>13) - Active ggw oosqponent is a puq l
or valve in an essential piping system wttick is requized to perform I
a anchanical motion during the course of accompliaking a system safety function.
'S RL8'SN
- j. matainment vacuum g
Edu5ct.v50 7MS breakers (ant 2) -
f cgpAgg4M Vacuum breakers valves mounted on the vent internal to the torus or on piping associated l
with the torus are Class 2 (3) oceponents.
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---e.-,
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l NRCConwect No.MAC4341-11]
Frer'ddn Research Cenw PRC Pnpot No.C0008 Page N W No. G i
A Dheen elThe Fva -:i anamuse PRCTeek No. 3 M plantName Ot. Ao Cires VAni;s WI.-
Sh and Aans Seese. Phen. Pa. 19103(2156 448 1000 TatWe St. Audit Precedure for Structural Acceptance Criterte of leerk 1 Containment Lent Term Program Licensee Uses Seatten Keyitems Considered Criterte Addtl. AlternaleW
..wi R.eAudi e.
i,de.
- 4 nome,te i
Conser-Uneeneer-gg g,
velho veelve 2.2 (Cont.)
k.
Baternal piping and supports (aows 10-13):
}
- No Class 1 piping 3
- piping external to and penetrating the l
torus or the external vents, includisg the j
attaement veld to the torus or vent nossle is i
Class 2 (3) piping. The i
other terminal end of I.
such esternal piping should be determined based on its function and isolation capability.
- maheection NF [3]
support for sue i
esternal piping including welded or ameanical attactument to l
structurer escluding any attamaant welds to the piping or other pressure retaining component are class 2 (3) supports.
LJC.EM5 SM RESR5AJEF W 1.
Internal piping and N
Aract.vtro rMrs l
supports (nows WCMOJ 10-13) - Are class 2 or i
l Class 3 piping and l
Class 2 or Class 3
- t.upports.
m.
Internal structures
[
l,
<=,w e) - mon.af.er-related elements which ll are not pressure I
retaining, esclusive of attachunent welds to any pressure retalaing lI I
_ _ _ _. _ ~.. _ _.. _ _. _. _ _ _ _ _ _ _ _ _ _ _ _ _. -. _, _, _. _, _ - _,
T
,1 NHC Om Me.NMC#41-13B Frenien Reneerch Center PRC Prolset No. QM08 Pa08 MMM Ib A Om af The Freder,wasmee l
na w neu s pm.p. mas am ees.iam N T**h*
3 2 1J Mentfoeme otA4p Csrigg ban?? _ /M d f.
8 Totne 31. Audit Presseure for Struchsted Acceptanee Cetterte of hiertiContainment Long Term Program g
t useneseUsee I
i l
CHearts
- Aggqi,
.i soonen Keyitems Considered Not W u_
N4 femarks No.{2) in the AudM m. Wet 1Inood.
conser-u-mewe vaghe 2.2 (Cost.)
I
.em.er (e....
menoraLis laddera, i
[
catwalks, and their i
supports).
a.
Pont deflectors (mov 9) l
- Dent header f)ow deflectors and
- associated hardware (not l
including attachsent j
welds to Class NC vessels) are internal l
s tructures.
3.2 Imad terminology used
[
abould be based on Final
]
I Safety Analysis Esport a
(FSAR) for the unit or the i
i Imad Definition kaport (125) [5]. Za case of oo.fu.t. the I u,.d.
aball be used.
3.3 Consideration of all load
~
comoLastions defino4 in Section 3 of the IDR (5]
+
ohall be provided.
i ll4.3 l
l p
a.
Iso reevaluation for
, /
limits set for design i
pressure and desip temperature values is f
needed for present structural elements.
)
l b.
Desip limit
/
f requirements used for initial. oonstruction I
following moraal
{
practice with respect
[
to load definition and j
allowable stress shall be used for systems or I
I i
i i
l l
[
i i
I
1 c'c cm. :: ;..u t 1-a i
,1 4,,,
4 hWJ Funkbn Rassach Centar F*C N ic W. N Page rac Aeese.mentNo. IE il a %., m y,m m PMCTammo.325
/2 n+ w r c.w m...F. me ms mom praetu.,ame Guao C.ines uwirs N 2.
Teb662-1. Audit Proreuure for 5tnntuiW Accepww:e Caterte d MortiConthimneett Long Term Program g
i
-I a Lcensee Deee x, _
A,,e e.,,,,r_,,
NA plemarks 1
I Medf]
tr@s Audtt Not lefo.
C'"**r.
unooneer 1 j
w uet need.
www m
I i
4,3 (Cunt.)
I portions ci systems
(
that are replaced and for cow systead.
Ses definition l
4.4 Sersloe Lisits and
]'
Design Proceduret shall for Se:vice f
$(JV Cbda, Section 'tII,
'I-Limits in he based o2 thy section 4 of Divisico 1 1 cluding Beierence 2.
l
.eddenda up to Sument 1$71 Mdeoda (3), specifica11yt I
l v/
l A.
Clasa PC g
contalmesst vessels: Article EE-30!!0 [3]
l
. Lf c r.USEEE'S m
M5 b.
LfAdar-type p
C'*T"* P L (C14*A 3 RESCL vEO 7"44:5 arid 2) suppo:2-c,opsegafa with three I
modificat10c4 to j
.e ede, i
4
- For bolted
)
-ti..,.e requirements of Servloe Limita A
{[,
(
and 3 abil. be l
I applied ts Service Limits C and D I
j wi'2eut increase in I
.e.u.ee.ie.
i
~ ~e applicable to l
Service Invals A end as l
l
- EF-3231.1 (a)
(3) is for primary j
plus mar = Aary
. tr.,e,.e, l
a t
l
[g NRC Centreet No. NRC48 0113 I
WW M Rennesch Center PRC ProtestNo.CAAIS Page 1
PRC Asessaamntme. /g 4 a
.e n. w i, es.
PRCTaskNo n5 jj
/3 l
mm e n s he..n mas mis en ione PlantName 6uAO C"'97165 UAfr N ! 7 L
+
TatWe 2-1. Audit Procedure ter aan ctural Acceptense Cetterte of Mark 1 Contoirweent Lane Term Pre 0 rem I
usensee uses Section KeyitemsConeidered Criterts Aggg-Alternate Approach No. [2]
in the Audit Not inte.
NA Romerks i
Met Met Road.
veeve venue
- All increases in i
allowable strees permitted by sabeectica NF [3] are limited by
.I Appendix IVII-2110(b) l
[3] trhen buckling is a consideration.
t
]
c.
Class 2 and 3 piping, pumps, valves,"and internal structures (also class IIC) 1 5.3 The components, ocuponent
/
l l
loadings, and service level I
assigrusents for Class IC l
[3] oogtments and laternal structures shall be as l
defined in Sohle 5-1 of i
n
,ef.r 1.
)l 5.4
' /
She comyceents, Pat 1*ings, and servios level I
assigmeents for Class 2 and l
Class 3 piping systema j
shall be defined in 1hkle 5 2 of meference 2.
l 5.5-the definition of operability is the ability to perform required l
sechanical motion and functionality.is the l
ahtLity to pass rated flow.
l uceums 1
m.
Active componesta m
gM H8t'!>
t l
l shall be proven g
p T>N S.
l l
operable. Active mymrw
-te saan w 4
i acasidered operable
.if dervice Limits
+
l A or 3 or more ccuservative limits g
(if the turiginal l
1 desip criteria required it) are aart..
t
\\
l NRC Contect No. NRCHIS 811U r,.nkin w ce,,,,,
rRC erwooiNo.Come N W No. O 1l A Dhnamn of The h buses Shh and Rame Seesm. Phen.. Pa. 19103 (2th 410o0 FRCTeek No. 32.5 plantName auso Cme =s LAnrs j42-
/p t
Tatne 2-1. Audit Procedure for Structurel Acceptance Cettoris of Mark l Containment Lont Term Pro 0 rem r
Keyitems Considered Criterts Addtl. m lSecten No. l2l In me Audit Not Info.
NA Remarka Conser-Unooneer-Met Met Rogd.
vedve velhe 1
5.5 (Cont.)
/
i b.
Piping components shall be proven functional in a manner consistent i
with the original decip criteria.
i l
6.1 Analysis guidelines provided herois shall apply to all structural l
elements identified in i
item 1.2 of this table.
a.
All loadings defined in See section 3.3 subsection 3.2 of of this table.
Reference 2 shall be
- id****-
l
,icesses's
/
=
arsw
- o.
As - to.nio.1 report on the analysis Rllr3cL.VE*. O TNIS i
ahall be submitted to f
{
l c gp A M,4m. R A.J tse unc.
l i
6.2 the following general guidelines shall be applied i
I to all structural elements analysed:
I 5.
Perform analysis according to guideline fl defined herein for all lands defined in IAR 3
(5].
(m r loads considered in original e
decip, but not redefined by LDR, previous analyses or new analyses may be esed.)
b.
skly limiting load ocabination events need 3
be consideged.
j4 e.Rc contest No. NRc43-411J litik Fratshn *1eser n Center PRc PreMot No.C8806 Page PRc Aeolenment No. / 2.
4l PRcTW No.525
/5 A fMmon elThe Fem buses ment namec um Co r ics f.hirs // 2.
ahh and name seese. Musa, Pa.19 03 aise eeMua, y
Tense 21. Aust Procencre for structural Accessence cinerte of Mest Icontainment t.one Tenn Proetem useneee uese crnerm Ades.
,,,,,,, ~. _ _ _ _,
seceen nor noms conessered NA Romerks No.lat inom Aumi mot mee.
con or unooneer-
,,, u men new
'l 6.2 (Cbnt.)
J/crNuu"5 A/MCACWC M S c.
Fatigue effects of all 70 7WE AleC operational cycles M#
8 shall be considered.
l ucsuser s RESPtsv5F M45 d.
iho further evaluation
/
g RF.SOWE) W of structural elements J
COWCK4 AJ S l
for which combioed effoct of loads defined in Ist [5] produces I
stresses 1ess than 10%
of allowable is required. calculations demons trating conformance with the I
10% rule shall be provided.
I I
e.
Dsging values used in dynamic analyses shall l
be in accordance with i c gu1 story oa.e Il6.3 1.61 (61 Structural responses for l
loads resulting from the l
combination of two dynamic I
phenomena shall be cotained in the folloeing manners a.
Abeolute sum of atross
[
c components, or i
b.
Cumulative distribution
/
(
l function method if absolute sum of stress ocuponents does not satisfy the acceptance criteria.
6.4 Sorus analysis shall consist of:
1 i
l Il
1 l
NRC Contreet No.MAC4H11J Free 4Ln Research Center PRC W No.C3006 Page i
4 % g.,g % %
PRC Assignment No. I*Z.
PRC7 No O
j k 30m and Race Sesos. Phas.. Pa 19103 (21W 4e01000 Table 21. Audit Procedure for Structurel Acceptance Cettetts of wertl Conteinment Long Term Proerem Ucensee Usee Section Keyitems Considered N_
Addtl.
Afternese Approach hewAuet Not Info.
No. [tl Conser Unconsor-u t met need.
I 1
6.4 (Cont.)
LicEMSEE 'S l
SEK RESPCAJSC M5 a.
Finite element analysis Nyyy g g g rWi$
for Igdrodynamic loads egg (tias history analysis) t and norinal and other loads (static analysis) making up the load l
combinations =hmt1 be l
Performed for he most highly loaded segment l
of the torus, including the shell, ring, l
girders, and support.
U b
b.
Evaluation of overall
/
SEE effects of seismic and W
ResoLAMC) 'D+ S other nonsymmetric l
loads shall be provided CI 1
I using been models (of at least 180* of the l
torus including colisens and meismic restraints) by use of either dynmaic load factors or g
time history analysis.
l c.
Provide a non-linear time history analysis, l
using a spring mass model of torus and support if not tensile forces are produced in columns due to upeard phase of loading.
r i
d.
Bijlaard fo-ilms shall l
be used in analysing l
ese torus nossle for I
effect of reactions produced by attached piping. If Bijlaard j
i formulas are not 1,.
i
.I NRC Centract No. NRC 43 011J 0 rr.okan ne e,es ceoier rRCPres iN.. Cases p
A Dnamen af The Franham inensee ERO W NO-
'l PRCTask No. 325
/7 1l PsentName QtAQD QTirs UAnts llE.
20m and Race Seeses. Phda. Fa 19103 f2154 M1000 l
] Table 2-1. Audit Procedure for Structural Acceptance Crnerts of Mark lContainmeert Lane Term Program Uceneep Usee Section Key hems Coneklered CrMerte AddII.
Altomate Approach No. [2]
In the Audit Not info.
NA Remarks
,y Met Met Road.
6.4 (cont.)
I applicable for any nossle, finite element I
analysis shall be performed.
6.5 In analysis of the vent I
system (including vent i
penetration in drywell, vent pipes, ring header, Il 6cemoceers and their intersections, vent column supports, vent-torus i
l bellonra, vacuum breaker penetration, and the vent I
deflectors), the following I
guidelines shall be followed:
a.
Finite element model shall represent the 1
most hi$1y loaded portion of ring header
.i
)
l shell in the 'non-vent' l
bay with the daemooners l
I attached.
b.
[
shall be performed to l
evaluate local effects in the ring header i
shell and doernoomer l
l intersections. One I
time history analysis for pool swell transient and equivalent static analysis for doimoceer lateral loads.
l l
m s
NRC Congrect No.NRC SM111)
I
=
Freh Research Center PRC W No. M PRC No. I L A Chuman of The Frenhhn inenww PRCT No
/6 shh and Rees Seese. Phda.. Pa. 19103(215e estono y
s
'l Table 2-1. Audit Procedure for Structural Acceptance Generis of Mark lConselnment Long Term Program e
usensee uses N
Addtl.
Alternese Approach Section Keyitems Considered No. [2]
In the Audit Not Irdo.
NA Romerks Conser-une.smer-ua votive wedve I
6.5 (Cont.)
c.
Evaluation of overall effects of seismic and
^l o.or nonsymmetrical I
loads shall be provided using beam models (of at least 180* of the l
I vent system including vent pipes, ring header and column supports) !sy ll oe use of ei.e, dynamic load factors or time history analysis.
d.
Use beam andels in analysis of vent deflectors.
l v1p/T" CNFFLAt'.TD At.
I e.
Consider appropriate j
g,moggy su superposition of gg gg I
reactions from e vent I
deflectors and ring headers in evaluating the vent support columns for pool swell.
6.6 a.
Analysis of torus internals shall incimie the catwalks with supports, annorails, and miscellaneous internal piping.
b.
It shall be based on
/
hand calculations or simple beam models and dynamic load factors and equivalent static analysis.
I 1
7
_. _ _ ~ _ _., _ _. _. _. _. _ _... _ _ _ _ _ _ _ _ _. _.., _. _ _.. _. _ _ _
.m.,-._,
\\
l NRC Contract No. NRC4241-1D I
FreeUdin Research Censar FRC 6 No. M PRC Assignment No. 1 A Dnsessa d h h lasense l
T h
8 mn and Rose seese. Phan. Pe 19103 (2156 m1000 Table 21. Audit Procedure for Structural Acceptance Catteria of Mark 1 Containment Long Term Program I
ucensee uses Section Keyitema Considered N
Addtl. ANornese W No. [fl in the AudM Not 184 0.
Conser Unconser-Met Met Regd.
l l
s.s (cont.)
/
\\
c.
It shall consider Service Intel D or E
'l when specified by the i
I structural acceptance criteria using a
(
simplified nonlineae analysis technique i
l (e.g., Rigg 's als thod).
6.7 Analysis of the torus r
attached piping shall be performed as follows:
Lt C E'A.!$CE'S a.
Designate in the g
M 8F M l
summary technical AlFsouMe TwS 3
" ? 3LA#
'l report submitted all I
piping systems as essential or non-essential for each i
l load combination.
b.
Analytical andel shall represent piping and siqpports from torus to I-first rigid andor (or where effect of torus action is j
insignificant).
c.
One response spectrum l
or time history analysis for dynamic effect of torus motion at the attadiment point, emoept for piping systems less than 6' in diameter, for whie equivalent static analysis (using appropriate l
amplifloation factor) any be performed.
l N-
\\
s l
E Contract No.NRC434118 0 FrerUdmResearchCenter PRC W No.C0000 Page PRC Assignment No. /E A Dheen of The Frankhninsasues PRCTW No M l
PlantName m C.rrres Durs jd2 Shh and Aara Seeen. Phen. Pa 19:03(21544 4 1000 Table 2-1. Audit Procedure for Structural Acceptance Cetterta of werk i Containment Lont Term Program Ucensee Usee N
Adott. Anernate W Secean Key items Considered No.[2]
in the Audit Not info.
NA Remarks Conser-unooneer-I uet met Reed.
weeve wellve 6.7 (Cont.)
l
/
d.
Ef fect of anchor disf ar===rtt due to tocus motion may be neglected from aquation i
9 of EC or ND-3652.2 [3]
if considered in Equations 10 and 11 of M: or MD-3652.3 [3).
l 6.8 Safety relief valve ofEE disearge piping shall be A'OTS LX WM'.5 r
RESPt:>NSF MAS analyzed as follows:
g REi5cLVED rNi5
^
I a.
Analyze each disearge COAJc:lldDlitAJ j
line.
b.
mdel shall represent piping and supports, from nossle at main steam line to dircharge in suppression pool, and include discharge device and its supports.
c.
For discharge thrust
/
4 loads, use time history
,I analysis.
).
d.
Use spectrum analysis or dynamic load factors I
for other dynamic loads.
l l
l l
I I
s NMC Centract No. NRC6411D A
FRC ProMot No.C0000 j
0000 Frenian Research Center FRC Aaelgn.ent No. / 2.
PRC,ask No. 325 7/
a w,,n, % %
mm-dam s nn...r.. masmis = imo f
**tNe**Oudo Grws /%wrs //2.
C-.-
i 1111 111J!t 11101 I
Memarks
~
a.
Meus shell with asacciated [
penetrations, reinforeing rings, and support attachments i
b.
trus shell supports to
/ [/
[ !
the building atructure c.
Dents between the drywell g/
[ /'//'
and the vent ring header (including penetrations therein) d.
Region of drywell local to
/
vent penetrations e.
Bellows between vents and
/
/
[ [ /
l torus shell (internal or external to torus) f.
Vent ring header and the
/
downoomers attaeed to it g.
Vent ring header supports
[
[
to the torus shell l
h.
Vacuum breaker valves M M M M
Y ALVES ARE attached to vent penetra-tions within the torus I
(where applicable) g I
1.
Vacuum breaker piping M8t MA MA WA MA MA MA MA systems, including vacuum breaker valves attamed to torus shell penetrations and to vent penetrations external to the torus (where applicable)
- j. Piping systems, including
[ [ [
peps and valves internal to the torus, attaeed to the torus shell and/or vent penetrations
. ~...
s NRC Contract No. NRCKlH1-130 Frarian Reesarch Center N W No.CSWIS A Dhesen of The Fienhan insame
[g mT No
)
sa anm== s====. Pha... r 1,ios als ses.iano Table 2-2. Audit Summary for Structural Acceptance CrNorts of Mark 1 Containment Long Term Program l
,,,u,,emen,s
-s u-,s
_.em._
_ ts dl l fi lJ 1J1 1 1
k.
All main steam system safety relief valve (SRV) piping 1.
Applicable portions of the
/ [ l // / / l l
following piping systems:
(1) Active ccatainment system piping systems (e.g., emergency core I
cooling system (aces) suction piping and other piping required to maintain core cooling after loss-of-coolan t accident (IDCA))
(2) Piping systems which provide a drywell-tcr-wetwell pressure dif-forential (to alleviate pool swell effects)
(3) Other piping systems, including vent drains I
a.
Supports of piping systems
/
[ /
/
/
/
aentioned in previous iten u.
Dent header deflectors
/
/
/
/
/
including associated hardware f
/ /
l l
o.
Internal atructura1 elements (e.g., monorails, catwalks, their supports) l whose failure might impair the containment function i
t I
TER-C5506-325 Table 2-3.
Non-Piping Structural Elements
~
STRUC1 URAL ELEMENT ROW External Class NC Tbrus, Bellows, 1
External Vent Pipe, Drywell (at Vent),
Attachment Welds, Tbrus Supports, Seismic Restraints l
Internals Vent Pipe General and 2
Attachsent Welds Ar. Penetration 3
(e.g., Header)
Vent Ring Header General and 4
Attachment Welds At Penetrations 5
(e.g., Downconers)
Downconers General and 6
Attachment Welds Internals Supports 7
Internals Structures General 8
l 3
Vent Deflector 9
l l
l l _
._~
~
l TER-C5506-325 Table 2-4.
Piping Structural Elements STRUCTURAL ELEMENT ROW Essential Pipine Systems With IBA/DBA 10 With SBA 11 Nonessential Piping Systems With IBA/DBA 12 With SBA 13 I
I I
l I
I I,
t
]
TER-C5506-325 NOTES RELATED TO TABLES 2-1 AND 2-2 Note 1: The Licensee has not provided a sussary of the analysis of the vacuum breaker valves and has not indicated that they are Class 2 components
-as required by the criteria [2].
(The Licensee's response has resolved this concern.)
Note 2: Nith respect to Sections 5-3.3 and 5-4.3 of the PUA report [7), the Licensee has used the AISC code in place of ASME,Section III, Division 1, Subsection NF for Class 2 or 3 SRVDL vent line supports.
(The Licensee's response has resolved this concern.)
4 I
j l
Note 3: The Licensee has not designated any torus attached piping systems as i
I essential or nonessential and has not indicated whether active pumps or valves are considered operable.
(The Licensee's response has resolved this concern.)
i Note 4: Sections 6-3.1 and 7-3.1 of the PUA report [7] state that some sasil bore piping was excluded from the analysis on the basis of the 10%
l rules however, no calculations demonstrating conformance to this rule have been provided as required by Section 6.2d of the PUAAG [2].
(The Licensee's response has resolved this concern.)
Note 5: Sections 6-5.3.1 and 7-5.3.1 of the PUA report [7) state that the 10%
rule was a criterion for the qualification of equipments however, no calculations demonstrating conformance to this rule have been provided as required by Section 6.2d of the PUAAG-[2). -(The Licensee's response has resolved this concern.)
I Note 6: The Licensee should justify the reasons for not considering a 180*
beam model of the torus including columns, saddles, and seismic restraints in order to determine the effects of nonsymmetric loads j
such as SRV and chugging for Quad Cities Units 1 and 2.
~(The 1
Licensee's response has resolved this concern.)
{
l Note 7: According to Table 2-2.5-3 of the PUA report [7], certain suppression chamber stresses are close to the allowables. The Licensee should indicate conservatisms in the analysis to show that these calculated stresses would not be exceeded if a different analytical approach were to be used.
(The Licensee's response has resolved this concern.)
i l
f i
A 4
4 TER-C5506-325 3.
REFERENCES FOR APPENDIX A 1.
l
" Safety Evaluation Report, Mark I Containment Iong-Term Program Resolution of Generic Technical A::tivity A-7" Office of Nuclear Reactor Regulation USNRC July 1980 2.
" Mark I Containment PS;ogram Structural Acceptance Criteria Plant Unique Analysis Application Guide" General Electric Co., San Jose, CA October 1979 g
3.
American Society of Mechanical Engineers l
Boiler and Pressure Vessel Code,Section III, Division 1
" Nuclear Power Plant Components" New York: 1977 Edition and Addenda up to Summer 1977 I
4.
Title 10 of the Code of Federal Regulations 5.
NEDO-21888 Revision 2
" Mark I Containment Program Load Definition Report" General Electric Co., San Jose, CA November 1981 6.
NRC
" Damping Values for Seismic Design of Nuclear Power Plants" l
Regulatory Guide 1.61 October 1973 1
7.
Quad Cities Nuclear Generating Station Units 1 and 2 Plant Unique Analysis Report i
Revision 0 Commonwealth Edison Company Nutech Engineers, Inc.
May 1983 i
t l
l I
i l
=
I l
\\
l
~
l APPENDIX B l
o i
ADDITIONAL INFORMATION REQUIRED
\\
1 I
l l
l I
l l
l i
l FRANKLIN RESEARCH CENTER l
DM90N OF ARVIN/CALSPAN 20tn & RACE STREETS.PHH.ADELPHIA,PA 19103 l
l TER-C5506 'J25 REQUEST FOR INFORMATION I
Iten 1: Provide a summary of the analysis of the vacuum breaker valves and indicate whether they are Class 2 components as required by the criteria [1]. Also indicate whether any vacuum breaker valves are attached to torus shell penetrations.
i Item 2: With respect to Sections 5-3.3 and 5-4.3 of the PUA report (2), show that SRVDL support stresses due to extreme environmental and emergency conditions do not exceed the Service Level C and D Limits specified in the ASME B&PV Code,Section III, Division 1, Subsection NF for Class 2 or 3 linear supports.
Item 3: Designate which torus attached piping systems are essential and which are nonessential as required by the PUAAG [1], Section 6.7a.
Also indicate whether all active pumps or valves associated with the piping are considered operable.
Item 4: With respect to Sections 6-3.1 and 7-3.1 of the PUA report [2),
I provide calculations demonstrating conformance to the 10% rule of Section 6.2d [1] that exempted piping systems at Quad Cities Units 1 and 2 from analysis.
Item 5: With respect to Sections 6-5.3.1 and 7-5.3.1 of the PUA report (2],
indicate whether any equipment was qualified by the lot rule of Section 6.2d [1] and, if so, provide calculations demonstrating conformance to this rule.
Item 6: Provide and justify the reasons for not considering a 180* beam model I
of the torus including columns, saddles, and seismic restraints in order to determine the effects of nonsymmetric loads such as SRV and chugging for Quad Cities Units 1 and 2.
Item 7: Table 2-2.5-3 of the PUA report [2] indicates that the calculated values of certain stresses are close to respective allowables.
Indicate conservatisms in the analysis to show that these calculated values would not be exceeded if a different analytical approach were to be used.
I I l
6 i
TER-C5506-325 REFERENCES POR APPENDIX B
" Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide" General Electric Co., San Jose, CA October 1979 I
2.
Quad Cities Nuclear Generating Station Units 1 and 2 Plant Unique Analysis Report Revision 0 Commonwealth Edison Company Nutech Engineers, Inc.
May 1983 3.
" Safe'ty Evaluation Report, Mark I Containment Iong-Term Program Resolution of Generic Technical Activity A-7" l
Office of Nuclear Reactor Regulation July 1980 4.
NEDO-21888 Revision 2
" Mark I Containment Program Load Definition Report" General Electric Co., San Jose, CA November 1981 l
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l APPENDIX C TECHNICAL REPORT ON THE USE OF THE CMDOF PROGRAM IN THE MARK I TORUS A'ITACHED PIPING ANALYSIS I
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PRANKUN RESEARCH CENTER DMSION OF ARVIN/CALSPAN 20tti & RACE ff1EETS.PMMADELPHIA.PA 19103 I
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1 TECabi1 CAL REPORT ON THE USE OF THE CMDOF PROGAMt IN THE MARK I TORUS ATTACHED PIPING ANALYSIS 4
l BY i
V. N. Con A. A. Okaily 1
l FRANKLIN RESEARCH CENTER l
PHILADELPHIA, PA.
19103 l
i April 1985 I
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Background Information The purpose of this report is to provide assessments and to document-f l
activities associated with the computer program CMDOF (Coupling of Multiple Degrees of Freedca) which was used by the NUTECE technical staff to qualify 1
j the Mark I torus attached piping systems in a number of nuclear power plants.
This program was originally developed by Dr. R. F. Eennedy [1] of Structural Mechanics Associates and modified by IRFFECE technical staff to establish the l
i stress level of the torus attached piping under various hydrodynamic loading conditions associated with the Mark I structural evaluation program. In the l
course of reviewing the analytical procedures for stress calculations of the torus attached piping systems, Franklin Research Center (FRC) staff raised a
concerns associated with the verification of this program, which will be summarized in the next section of this report. A meeting was held with the 1
NUTECH technical sta";f and a number of affected utilities on August 9 and 10, f
1984 to discuss a number of technical issues related to this program. As a l
result of this meeting, a number of action items were requested from the af fected utilities, to which the NUTECH technical staf f responded [2]. The reviews of NUTECH responses indicated that the main corecern, which is the validation of the program, remained unresolved. A report was then prepared l
and submitted to the NRC by FRC [3] to provide the review status of this i
l program and highlight areas of concern associated with the use of this l
program.
A subsequent meeting was held on January 4,1985 with the NUTECH technical staff, Dr. R. P. Kennedy of Structural Mechanics Associates, and
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representatives of the Mark I owner group and a number of utility companies.
l In this meeting, Dr. Kennedy provided an overview of the technical background l
1 of this program. It was also learned that the Bechtel Power Corporation attempted to verify the program by comparing the results obtained by the f
program with those obtained f rom a combined torus / piping model. However, due l
tonumericalinstabilitiesofthecombinedtorus/pipingmodel,thisattempt l
was not successful. At the end of this meeting, it was obvious that FRC's concerns were not resolved and the affected licensees expressed their opposition to perform further investigations regarding the progras l
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verification. However, it was learned later that the Monticello plant selected some in-plant test data (SRV in-plant test data) to verify the program. The results of this study were submitted for review (4). FAC review of this latest document is given in Section 4 of this report.
I 2.
Technical Beckeround of the QWOF Program j
l The standard practice for performing dynamic analysis of the torus and 4
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attached piping systems is to perform independent uncoupled dynamic analysis of the torus and of the attached piping. First, the torus model is developed and a dynamic analysis of the torus subjected to the postulated hydrodynamic 4
load is performed using this uncoupled model. The response time history at a
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the penetration point of the attached piping is obtained. Then this response -
time history is used in conjunction with the uncoupled dynamic model of the j
g attached piping to calculate piping responses. This approach is known as an uncoupled analysis because the dynamic model of the torus and the attached l
piping are never directly coupled. It has been recognized that this approach I
results in a conservative estimate of the piping responses.
4 The other acceptable approach is to carry out a coupled analysis in which i
the torus and associated piping are combined in a single coupled model. The model is fairly complicated and also results in high computational cost, especially when a significant number of loading time histories have to be 4
considered. Therefore, this coupled analysis does not represent an attractive alternative. In fact, none of the Mark I facility resorts to this approach.
1 The CMDOF program was developed to take into account the coupling effects without carrying out the coupled analysis described above. Essentially, this program is used to modify the response time history obtained from the uncoupled torus model at the penetration point of the attached piping and this modified time history is then used to obtain the piping response of the 4
i ut. coupled piping model. In order to use this program, the modal response characteristics of the torus and attached piping have to be established first by applying an unit force at the attachment location. These modal response characteristics along with the uncoupled response time history of the torus at the penetration point will be input into the QOOF program, which will produce i
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f a modified response time history to be used in obtaining the piping response.
l This program, in principle, is supposed to remove the conservatism associated with the uncoupled analysis.
1 3.
Concerns Associated with the CMDOF Program i
Based on the review of pipe stresses obtained via this program and other information relating to this program, FRC staff raised a number of questions j
in connection with the validation of this program (3). A program of this nature requires a substantial validation effort in order to use it in a production mode. Also, this program is relatively new and the originator of
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g the program cautioned:
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"It has been carefully programmed and checked against a number of test j
cases by comparing its results for coupled response with those obtained l
from coupled structure and equipment analyses. However, it has not been l
used to date (April,1980) by other than the authors. It is not a production program which can be used as a " black box".
Users should l
independently verify their own use of the program and understand its u
basis and applicability before using it in a production mode." (1) 4 j
FIC's concerns are briefly summarised below l
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o The verification problems provided were extremely simple compared i
with the problem of the torus and attached piping. Basically, the l
verification problem consists of a spring-mess system with a few l
g degrees of freedom.
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The parameters (esse and stiffness) given in the verification l
l problems did not resemble a wide range of values (mass and stiffness)
I encountered in the actual problem.
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o Based on some study by NUTECH [4], it was observed that the CMDOF l
could reduce the input loading to the attached piping by as much as 3 or 4 times when compared with a standard uncoupled analysis.
o calculated stresses of the affected piping systems in a number of plants in some cases were closed or equal to the stress allowables.
4.
Review of CISOF Verification In-plant SRV tests performed at the Monticello plant in 1980 were used as a basis for verification of the CMDOF program. Test data f rom five tests were selected for comparison. Specifically, data from strain gauges located on the i
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e RCIC turbine exhaust line (RS3-8 in-HE) approximately 1 foot and 20 feet f rom the torus penetration, as shown in Figure 1, were used for comparison. The tests were conducted by actuating one safety relief valve under cold pipe and normal water leg conditions with a reactor power level of 80%. Plots of strain time histories were recorded during each test and were compared directly with the predicted values obtained by the CM00F program.
With regard to load development, two programs (GE computer codes WFORO4 and QBUBSO3) were used to develop the SRV torus shell pressure time histories corresponding to the test case conditions (i.e., cold pipe, normal water leg, reactor at 804 rated power). With respect to the torus and piping structural models, the Licensee indicated that these models were developed to reflect the as-tested condition.
The C300F program was used in conjunction with the modal characteristics of the torus and attached piping to obtain the modified responses at the attachment location to the test SRV loadings. Displacement, velocity, and acceleration responses were developed at all pipirq degrees of f reedom coupled to the torus. From these responses, a modal superposition was employed in I
conjunction with transfer junction methodology to obtain stress time histories at the strain gauge locations of interest for comparison with the test results.
The Monticello SRV test strain gauge data (converted to stress) were compared with the predicted stresses obtained by the CMDO' program. The responses on the time domain and f requency domain (by Fourier transformation) at strain gauge locations were compared with those obtained by the analysis.
In addition, the maximum stress values were used in the comparison. The results indicated that a factor of conservatism is excess of 3 was observed in the analysis.
Based on FRC's review of various stress time histories and the maximum stress level of the test data and analysis, it is' observed that there is conservatism associated with the analytical procedures. This conservatism could be attributed to the following sources: methodology by which loads were generated, low damping values used in the analysis, possible nonlinearity resulting from pipe supports. The comparison between the test and predicted l
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values indicated that there is conservatism associated with the analytical procedures, which provides a basis for alleviating the concerns related to some calculated stress values presented in the Licensee's original submittals.
REFERENCES 1.
Kennedy, R. P. and Kincaid, R. H., "CMDOF - A Computer Program to Couple the Response of Structures and Supported Equipment for Multiple Degrees of Coupling Using the Results f rom Uncoupled Structure and Equipment Analysis," SNA 12101.03, Structural Mechanics Associates, Inc., Newport Beach, California, November 1980 2.
R. W. McGaughy (Iowa Electric Light and Power Company)
Letter with Attachments to H. Denton (NIC)
Subject:
Clarifications Regarding the Duane Arnold Energy Center Plant Unique Analysis Report, Mark I Containment Program, NG-84-3937 i
September 17, 1984 3.
Con, V. N., " Review of the Cosputer Code CMDOF (Coupling of Multiple Degrees of Freedom," Franklin Research Center, October 1984
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4.
D. Musolf (Northern States Power Company)
Letter with Attachments to H. Denton (NRC)
Subject:
Additional Information Related to Computer Program CMDOF February 25, 1985 t
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