ML20151T326
| ML20151T326 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/31/1988 |
| From: | Russell M EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20151T316 | List: |
| References | |
| CON-FIN-A-6808 NUDOCS 8808160317 | |
| Download: ML20151T326 (36) | |
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TOPICS REQUIRING FURTHER REVIEW RELATED TO THE SEISMIC UPGRADE OF THE YANKEE PLANT H. J. RUSSELL JULY 1988 Prepared for the United Sta'.es Nuclear Regulatory Comission Washington, DC 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6808
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SUMMARY
The Yankee Atomic Electric Company, licensee for the Yankee Nuclear Power Station, has recently indicated an interest in performing the seismic upgrade of the Yankee Plant in a fashion which will minimize the amount of review work required of the NRC staff. This document identifies all the criteria, methodology, and issue resolution strategies which require staff reviews. It also identifies the available alternatives which do not require such reviews.
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CONTENTS !
SUMMARY
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- 1. INTRODUCTION .................................................... 1
- 2. TOPICS REQUIRING NRC STAFF REVIEW, AND ALTERNATIVES ............. 1 2.1 MS/ FW An al y s i s (4. 3.13, 2. 2. 5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.2 Response Ratio Method (4.4.1.1, 2.2.2) ..................... 2 2.3 Bracketing of Piping Models (4.4.1.1, 2.2.2) ............... 2
2.4 Nozzle Anchor Flexibility using WRC-297 (4.4.1.1, 2.2.2) ... 3 2.5 Strain Criteria Application Exceeding 1%(4.4.1.1,2.2.2)... 3 2.6 SRSS Combination of SI and SAM for Pipe Supports (4.4.2. 2.3.1) ........................... 4
- 2.7 Gang Hanger Out of-Scope Piping Loads (4.4.2, 3.2.1) ....... 4 2.8 U-Bolts Used as Axial Restraints (4.4.2, 3.2.2) . . . . . . . . . . . . 5 2.9 Equipment Nozzle Analysis Using WRC-297 (4.5.1, 3.2.2) . . . . . 5 2.10 Review of Time Histories Used in Reactor Sliding Issue Resolution ........................... 5 2.11 SRSS Combination of VC and RSS SAMs ........................ 6 ;
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- 3. CONCLUSION ...................................................... 7 REFERENCES ......................................................
- 4. 7 TABLES
- 1. LIST OF ACRONYMS ................................................ 8 1
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TOPICS RE0VIRING FURTHER REVIEW l RELATED TO THE SEISMIC UPGRADE OF THE YANKEE PLAST
- 1. INTRODUCTION j In July of 1987, the Nuclear Regulatory Commission (NRC) staff issued a Safety Evaluation Report (SER) concerning the seismic upgrade of the Yankee Nuclear power Station (Yankee). This SER is identified in Reference 1. In March of 1988, the NRC staff issued a document clarifying their position on several issues discussed in the July 1987 SER (Reference 2). These documents benchmark the progress made in completing the latest ;
review cycle pertaining to the seismic upgrade of Yankee equipment and l
piping. Criteria and methodology acceptable to the staff for implementing i
the upgrade are identified. Of interest here are the criteria and I methodology, the application of which require case-by-case review by the staff before acceptance. Also of interest are resolution strategies for unresolved issues identified in the July 1987 SrR which require staff review. All topics in these documents which involve staff review are discussed in the following section.
I Two meetings i' ave occurred between the licensee and the staff since j the July 19?? SER was issued concerning the upgrade. The first meeting l (held in Bethesda, MD December 21,1987) concerned the integrity of tne i reactor support under seismic loading. The second (held in Walnut Creek, !
CA, May 2-4, 1988) dealt primarily with the method used by the licensee in '
calculating support loads for nonseismic piping attached to gang hangers 1 in the seismic scope. Any potential staff review work related to these
, meetings is also discussed in the following section. 1
- 2. TOPICS REQUIRING NRC STAFF REVIEW, AND ALTERNATIVES In each of the following subsections, a topic which involves the potential for future NRC staff review is identified. Specific areas requiring review are identified, as well as alternatives which do not require review. Two section numbers are provided in parentheses in the 1
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title of each subsection below. The first is the section number in the July 1987 SER where the topic is discussed. The second is the section number in Attachment 3 to the SER, where a much more detailed discussion of the issue is typically provided. References to other pertinent.
l documents are explicitly made in the text.
1 2.1 MS\FW Analysis (4.3.13. 2.2.5)
The analysis of the main steam (MS) and feedwater (FW) piping outside the Vapor Container (VC) is a special case. The steel framing supporting this piping is comparable in mass and stiffness to that of the piping.
The licensee has proposed special analytical techniques to analyze the MS/FW piping and supports. Approval was given to the proposal in Reference 1, with the understanding that the analyses to be done will be subject to a detailed review to determine final acceptance by the NRC staff. The only acceptable alternative to the use of special analytical techniques requiring case-by case review is for the licensee to modify the steel frame to the point where decoupling the piping from the frame is possible. This is not considered a feasible alternative.
2.2 Resoonse Ratio Method (4.4.1.1. 2.2.2)
During the course of evaluating piping inside the Vapor Container, the licensee regenerated spectra for branch piping connected to the reactor coolant loop. This regeneration made use of a refinement of the reactor support portion of the structural model which substantially reduced and broadened the spectra generated. Rather than reanalyzing the piping already analyzed, the licensee developed a response ratio method for calculating stresses for the new spectra based on the original stresses.
This method was reviewed as a result of an audit question, and found acceptable on a case by-case basis. The alternative to the response ratio method is a reanalysis using the new spectra.
2.3 Bracketino of Pioina Models (4.4.1.1. 2.2.2)
In most cases, piping models extended from anchor to anchor. Two exceptions to this oce.urred. In these cases, bracketing criteria were 2
used. With bra:keting criteria, the coundary between piping analysis models was not located at an anchor. Interaction effects between the models were included by extending each model past the arbitrary boundary a sufficient distance to include such effects. The amount of extension was per the engineering judgement of the analyst. Confirmatory analysi,s established the acceptability of these two applications. The -
acceptability of future applications of bracketing criteria will need to be established on a case-by-case basis because of the engineering judgement exercised in their application. There are two alternatives to the use of bracketing criteria: (1) create a large model with all piping between anchors included, if possible; and (2) install an anchor at a convenient intermediate location, with piping models terminating at the anchor.
2.4 Nozzle Anchor Flexibility Usino WRC-297 (4.4.1.1. 2.2.21 The licensee proposed defining nozzle terminations of the piping as anchors in the models in most cases. This is industry practice acceptable to the NRC staff. In a limited number of cases, the use of Welding Research Council Bulletin No. 297 (WRC-297) was proposed for defining more realistic boundary conditions. This was judged a reasonable approach, but it lacked sufficient basis for blanket acceptance. Therefore, the use of WRC 297 in defining nozzle flexibility was accepted with the restriction that every application of it be subject to a case-by-case review by the staff. The alternative is to use anchors at the terminatien points.
2.5 Strain Criteria Acolications Exceedino 1% (4.4.1.1. 2.2.2)
For NP.C spectrum loading, the licensee proposed functional capability criteria based on the SEP Guidelines. As a first check, stresses due to occasional loads are compared to allowable stresses of 1.8S h for equivaleni Class 1 piping, and 2.4Sh for equivalent Class 2 and 3 piping. If these stresses are exceeded, strain criteria were proposed.
With these criteria, a strain is calculated based on the ASME Plastic, Simplified Elastic methodology. This strain is compared to a limiting strain of 1% for carbon steel and 2% for stainless steel, in addition, a wrinkling check is performed for large bore, Sch 40 piping; a maximum 3
stress based on low cycle fatigue limits is imposed; and factored increases on flanged joint, nozzle, and support loads are imposed in the area of strains in excess of 1% (this being applicable only to stainless steel piping due to the 1% limit for carbon steel). The clearance check performed in the plant will use displacements increased by a factor of a 3.33 in areas of strain criteria application. Applications of these criteria where strains exceed 1% are subject to a case-by case review by the NRC staff. The alternative to a case by-case review is to reconfigure the supports to reduce strain levels in the pipe to below 1%.
2.6 SRSS Combination of SI and SAM for Pioe Suocorts (4.4.1.3. 2.3.11 A limited application of square root of the sum of the squares (SRSS) combination of seismic inertia and anchor motion loads was proposed for evaluation of supports under NRC spectrum loading. This was approved, with the understanding that all such applications were subject to case by case reviews. The alternative is absolute summation.
2,7 Gana Hanaer Out-of-Scone Pioina loads (4.4.2. 3.2.1)
Use of in-scope piping loads to estimate the loads associated with l out-of-scope piping on gang hangers is not acceptable. This issue was i
discussed at the April 1987 meeting. Four alternatives acceptable to the NRC staff were identified to resolve it: (1) The licensee could perform static analyses to calculate the non-safety related piping loads. Such analyses would include use of peak spectral accelerations with a factor of l 1.5 as recommended in SRP section 3.7 (Reference 3). (2)Thelicensee j could perform confirmatory analysis of a few representative samples to 2
show acceptability of the practice. (3) The licensee could provide a similarity argument between the safety related and non-safety related piping configurations to justify using the loads for one to estimate loads for the other. (4) The licensee could perform standard response spectrum analysis of the non safety related piping and combine the resulting loads using absolute summation. Options (2) and (3) would reouire review by the NRC staff for acceptability. Options (1) and (4) are alternatives that I
would not require review.
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All current gang hanger loads ha* e been generated by the licensee, and I reviewed and approved by the ataff during the May 2 4, 1988 meeting !
(Reference 4). The particular methodology used would need review in the .
l future if it were to be useo with piping other than that involved in the application already approved. If such a need arises in the future, ;
options (1) and (4), described above, could be used without further review l by the staff. ,
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2.8 U-Bolts used as Axial Restraints (4.4.2. 3.2JL),
Use of bullets as axial restraints without a testing basis for the allowable loads used is not useptable. Several alternatives acceptable I to the NRC staff were developed to resolve this issue: (1) The licensee could use available test results to validate the allowable loads used; (2) l The licensee could perform the necessary testing t.a validate the allowable loads used; and (3) The licensee could replace the subject supports with l conventional supports. If one of the alternatives that involve validating the allowable loads is chosen, review by the NRC staff will be required. l The remaining alternative would not require review. !
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2.9 Eouiement Nozzle Analysis Usino WRC-297 (4.5.1. 2.1.2) i The methods contained in Welding Research Council (WRC) Bulletin 107 were proposed for calculating stresses in the vicinity of nozzles. In some cases, the methods of WRC bulletin 297 were proposed as an l alternative. In either case, the geometrical parameters of the nozzle under evaluation would be checked to ensure they fall in the range of applicability of the Bulletin being applied. These methods are acceptable, with the understanding that all applications of WRC Bulletin 297 are subject to a case-by-case review by the NRC staff. If WRC Bulletin 107 is used, no staff review is required.
2.10 Review of Time Histories Used in Reactor Slidina Issue Resolution During the December 21, 1987 meeting, the licensee presented the results of an analysis intended to demonstrate that the reactor vessel 5
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will not siide during an NRC spectrum earthquake. The analysis involved time histories generated at the attachment points of the reactor vessel supports to the Reactor Support Structure (RSS) by a nonlinear analysis of the RSS. The analysis which generated these time histories was not reviewed previously by the staff, and a need for such a review was identified during the meeting. A review by the Structural Engineering Branch of the NRC has been planned, but has not yet been completed.
Although the licensee has withdrawn in original proposal to use floor spectra generated uring the new time histories in piping analysis, these time histories still support the licensee's conclusion that the reactor vessel will not slide under an NRC spectrum earthquake. A review i still needed. As an alternative, the licensee can reanalyze using time histories previously reviewed and approved by the staff. No further review would be required, provided that the reactor vessel modal used in the analysis has been previously reviewed and approved by the staff. This may not be possible, be:tuse the lack of sliding predicted by the analysis resulted primarily from reductions in excess margin obtained in the new time histories.
2.11 SRSS Combination of VC and RSS SAMs Audits performed during the May 2 4 meeting (Reference 4) established that seismic anchor motions of the Vapor Container (VC) relative to the Reactor Support Structure (RSS) used in analyzing piping in the VC were generated using the SRSS combination methodology. Justification for this is required, which implies further staff review. As an alternative, absolute sua combination can be used in place of SRSS combination without further review.
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- 3. CONCLUSION Although many of the approaches requiring staff review have viable alternatives which do not require such review, some may not. The following are likely to require further review:
- 1. MS/FW analysis - Modifications to simplify the analytical requirements are probably prohibitive.
- 2. Use of unreviewed time history in reactor sliding issue resolution -
Under the assumption that the licensee would not have generated the new time histories if the reactor vessel could be shown not to slide using existing, previously reviewed time histories, a review will be needed of the analysis which generated the new time histories.
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- 4. REFERENCES
- 1. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) dated July 17, 1987,
Subject:
NUBEG 0825, Section 4.11 Seismic Design Considerations (TAC No. 51807).
- 2. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) dated March 21, 1988.
- 3. NUREG 0800, "Standard Review Plan for the Review of Safety Analysis :
Reports for Nuclear Power Plants, LWR Edition," July,1981.
- 4. Letter from C. F. Obenchain (INEL) to P. Y. Chen (NRC) dated June 8, 1988,
Subject:
Trip Report for the May 2-4, 1988 Meeting
, Concerning the Seismic Upgrade of the Yankee Plant (A6808) -
Oben 71-88. (Included in the document package which contains this j docE* ant)
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o Table 1. List of Acronyms ACRONYM DESCRIPTION FW FEED WATER PIPING SYSTEM INEL IDAHO NATIONAL ENGINEERING L/BORATORY tis MAIN STEAM PIPING SYSTEM NRC NUCLEAR REGULATORY COMMISSION RSS REACTOR SUPPORT STRUCTURE
, SAM SEISMIC ANCHOR MOTION SER SAFETY EVALUATION REPORT SI SEISMIC INERTIA SRSS SQUARE ROOT OF THE SUM 0F THE SQUARES 4
VC VAPOR CONTAINER WRC WELDING RESEARCH COUNCIL YAEC YANKEE ATOMIC ELECTRIC COMPANY YANKEE YANKEE NUCLEAR GENERATING STATION l
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sa Eb c L O.S t QE 2 cc: R. G. Rahl 8MA.
p Wh40 AM File A6808 g EGsG Idaho l
lNTEROFFICE CORRESPONDENCE Date: December 28, 1987 To: R. C. Guenzler Frem: M.J. Russell */7 l
subject MEETING
SUMMARY
- YANKEE R0WE REACTOR SU/ PORT INTEGRITY - MJR-13-87 l
Refs: (a) M. 8. Fairtile (NRC) ltr to G. Papanic (Yankee Atomic l Electric Company), NUREG-0825, Section 4.11 Seismic Design I Consideraticns (TAC No. 51807), July 16, 1987 (b) Meeting summary fru. T. Cheng to YAEC, Meeting Sumary on .
Systematic Evaluation Program (SEP) Seismic Review (Topic III-6), July 25, 1986 A meeting was held in Bethesda, MD on December 21, 1987 to discuss the results of a licensee evaluation of the potential for the Yankee Rowe reactor vessel to slide under NRC spectrum loadings. This was identified as an unresolved issue in the last $afety Evaluation Report covering the seismic upgrade of the Yankee plantt a). Other issues related to the upgrade were also discussed: (1) the performance of case-by-case reviews '
soecified in Reference (a); and (2) use of spectra generated using the rasults of the reactor vessel sliding evaluation in the evaluation of piping and supports mounted on the reactor support structure (RSS). The l meeting was attended by representatives of the NRC staff (EMEB), l representatives of the Yankee Atomic Elsetric Company (the licensee for Yankee), representatives of the licensee's consultant (Cygna Corp.), and myself. A list of attendees is presented in ths first attachment. A copy ,
l of the licensee's meeting handouts is included in the second attachment. '
In evaluating the central issue, the licensee regenerated free field ground motion time histories for the NRC ground spectrum, performed a nonlinear analysis of the RSS to obtain time histories for the reactor support area, and used the resulting time histories in a linear analysis of the reactor vessel, neutron shield support tank and associated supports. The analysis results were used to calculate a coefficient of friction needed to keep the reactor frvm slidirg. This coefficient was shown to be less than those defined for the steel-concrete interface under question in both the PCI and ACI standatds.
Since the first analysis of the evaluation, that of the RSS, is beyond my area of experi!se, no judgement as to its general validity will be made here. This includes the proposal to generate response spectra for the analysis of piping supported from the RSS using the results of the RSS analysis. Based on my attendance at many of the meetings concerning building analysis conducted during the last review round. I questioned the "Providing research and development services to the govemment" -
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l R. C. Guenzler December 28, 1987 ,
MJR-13-87 i Page 3 of 3 . l I
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l and methodology. The portion of the SER discussing strain criteria stated i that strain criteria would be used only if allowable stresses were exceeded. The portion discussing square-root-of-the-sum-of-the-squares combination of seismic inertia and anchor motion loads for pipe support i analysis stated that such application would be limited. The discussion ia i the SER of the third case, the response ratio method, indicates our i understanding at the conclusion of the review that all applications of this '
method had already been identified and reviewed on a case-by-case basis.
i'uture applications of this methodology were not expected. Considering the additional cost and effort involved in case-by-case reviews, across the board application of criteria and methodology requiring such reviews is not sv.tsable. ;
At the end of the meeting, the licensee had agreed to submit to the staff the following: (1) a document defining the differences between the previous i and current reactor vessel sliding analyses and providing justification for l the differences, with the document specifically to include the methodology used in obtaining the 10% damping response spectra from the 5% spectra previously reviewed and approved, the correlation coefficient between the 1 horizontal and vertical time histories used in the analyses, and a plot of damping versus frequency with the RSS natural frequencies identified on it; 1 j
and (2) a document clarifying the licensee's application of methodology {
requiring hase-by-case-review.
mm1 Attachments:
As Stated (2) t l
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Attachment 1 December 28, 1987 MJR-13-87 :
Page 1 .
LIST OF ATTENDEES NAME AFFILIATION P. Chen NRC/NRR/ EMES J. Haseltine Yankee Atomic Electric Co. (YAEC)/ Project Manager B. Holmgren YAEC/ Lead Mechanical Engineer P. Kuo* NRC/NRR/EMEB V. Nuses NRC/NRR/ Project Manager G. Philley YAEC/ Consultant M. Russell INEL T. Wang Cygna/ Technical Consultant
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R. Wessman* NRC/NRR/EMEB N. Williems Cygna/ Project Manager
- Attended only a portion of the meeting.
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1 Attachment 2 December 28, 1987 MJR-13-87 Page 1 l
MEETING HANDOUTS l
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REACTORPRESSUREVESSEL/NEUTR0!!SHIELDTAliK -
SEISMICRES?0liSIAllDARSGENERATI0li FOR YANKEEATOMICELECTRICCOMPA!iY l
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ISSUISMARY:
Perfor. a stability analysis of ths Reacter Pressurs Visssi (RPV)/ Neutron Shield Tank (NST) to deconstrate that the R?V/NST
, WillnotupliftorslideunderthecombineddeadandNRCseistic spectra 10 ads.
CEnsratsthea:plifiedresponssspectra(ARS)atsignificant ReactorSupportStructure(RSS)locationsandattheMainCoolant Loop (MCW nozziss.
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1.3 Potential For Uplift / Sliding j Becauseofthetangentialrestraint,theRPVcannot slide against the RSR. The RSR, however, is sitting en theconcretecorbelandisunrestrainedhorizontally !
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CONCLUSION: TheR?V/NSTwillneitherupliftnorslideunderthecombineddead andh10seismicspectraloads. Therefore,thelinearRPVanalysis and the ARS generated using the time-histories generated by this analysisarevalidandtheARSareacceptableforuseinthe pipinganalysis. I
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