ML20151T329

Trip Rept for 880502-04 Meeting at Walnut Creek,Ca Re Seismic Upgrade of Yankee Nuclear Power Station

ML20151T329
Person / Time
Site:	Yankee Rowe
Issue date:	05/31/1988
From:	Russell M EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20151T316	List: ML20151T312 ML20151T326 ML20151T329
References
CON-FIN-A-6808 NUDOCS 8808160319
Download: ML20151T329 (62)

Text

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ENCLOSURE J TRIP REPORT FOR THE MAY 2-4, 1988 MEETING AT WALNUT CREEK, CA CONCERNING THE SEISMIC UPGRADE OF THE YANKEE NUCLEAR POWER STATION M. J. Russell May 1988 '-

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the United States Nuclear Regulatory Commission Washington, DC 20555 Under DOE Contract No. DE-AC07-76ID01570

, FIN No. A6808 egg 8QQg[g 69 P

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SUMMARY

A meeting was held May 2-4, 1388 between the NRC. staff and Yankee )

Atomic Electric Company, licensee for the Yankee Nuclear Power Station.

Also in attendance were representatives of Cygna Corp., consultant to tb5 licensea, and a representative of INEL, consultant to the NRC staff. The ,

topic of discussion was the ongoing work to seis;nically upgrade the station. One issue was discussed in detail, the method used by the licensee in calculating support loads for non seismic piping attached to <

gang hangers in the seismic scope. Audits of recent calculations were  !

also conducted. Two significant issues were generated. In general, the  !

calculations were being done according to criteria and methodology  !

approved by the NRC staff during the previous review cycle. Further, evidence was found that the licensee is taking actions necessary to meet )

his previous commitments.

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. i CONTENTS

SUMMARY

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1. INTRODUCTION ..................................................... 1

2. AUDIT RESULTS .................................................... 2 2.1 Ge r.e ral Aud i t F i nd i ng s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1.1 SRSS Combination of VC and RSS SAMs .................. 4 2.1.2 Anchor Bolt Interaction Equation ..................... 5 2.1.3 Fatigue Checks of Rod Hangers ........................ 6 2.2 Licensee Commitments ........................................ 6

3. DISCUSSION RESULTS: GANG HANGER LOADS ............................ 8

4. CONCLUSION ....................................................... 9 I

5. REFERENCES ...................................................... 10 f

TABLES 1 L i s t o f a t t e nd e e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

2. L i s t o f a c ro nym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

3. Li st o f audi ted documents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 APPENDIX A--MEETING HANDOUTS ........................................ A-1 l

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- TRIP REPORT l FOR THE MAY 2-4. 1988 MEETING AT WALNUT CREEK. CA l CONCERNING THE SEISMIC UPGRADE OF THE YANKEE NUCLEAR POWER STATION l

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l 1. INTRODUCTION In July of 1987, The Nuclear Regulatory Commission (12C) 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 comp'ating the latest review e.ycle pertaining to the seismic upgrade of Yankee e.;uipment and piping. They also list commitments made by the Yankee Atomic Electric Company (YAEC), licensee for Yankee, in a letter dateJ June 1,1987 (Reference 3). The licensee and his consultant, Cygna Energy Services of Walnut Creek, CA, are now working toward meeting those commitments.

A meeting was held in Walnut Creek during May 2-4, 1988 to discuss progress made on the work. A list of attendees is included in Table 1.

Audits of piping calculations were conducted to ensure that they utilized the criteria and methodology acceptable to the staff previously obtained.

The audits also allowed monitoring of the progress the licensee has made toward meeting his commitments. Discussions were held concerning efforts made independently of the calculations by the licensee in meeting the commitments. Since the licensee has yet to identify applications of alternate criteria requiring case-by-case review, no such reviews were conducted during the meeting. An audit of the latest reactor vessel ,

analysis was scheduled. This analysis was done to support the licensee's assertion that the vessel will not slide under NRC spectrum loadings. ,

This was a low priority item because the critical path for completing the I review of the reactor sliding issue is the review of the structural analysis which generated the loads used in the vessel analysis. This audit was not conducted due to lack of time.

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i The following sections of this report document the results of the l meeting. Section 2 contains the results of the audits. Section 3

contains the results of the discussions, primarily concerning the I calculation of loads for gang hangers applied by attached non seismic piping. A conclusion and list of references are provided in the following  !

sections. Table 2 provides a list of acronyms used in this report. All handouts received from the licensee during the meeting are included in Appendix A. I

2. AUDIT RESULTS Three piping calculations were audited: These are listed in Table 3.

The associated pipe support calculations were not sufficiently complete for full audit, so only a limited audit of support calculations was conducted. The general results of the audits are listed in the following subsection. This is followed by subsections which discuss specific issues I raised by the audits.

2.1 General Audit Findinas In general, the piping calculations audited met the requirements of the July 1987 SER and subsequent clarifications (References 1 and 2). A three dittwsional finite element analysis was performed in each case.

Response spectra for the analyses were generated for each of three orthogonal directions by enveloping spectra generated by structural analyses for all points of support. All of the spectra used were based on NRC ground spectrum. Results of the three separate seismic analyses done in each of three orthogonal directions were combined per the requirements of Regulatory Guide 1.92. Seismic anchor motions were taken from structural analyses which utilized the Yankee Composite Ground Spectrum.

Loads considered in the analyses included the effects of pressure, weight, thermal expansion, and seismic excitation, both inertial and anchor motion. Losd combinations were done in the standard fashion.

Stresses due to cmbined weight and pressure were limited to less than the 2

i allowable stress Sh . Stresses due to combined weight, pressure, seismic i inertia, and (optionally) seismic anchor motion were limited to less than j 1.8 times Sh for equivalent Class 1 piping, and 2.'4 times Sh fF e,quivalent Class 2/3 piping. Stresses due to thermal expansion, thermal anchor motion, and (optionally) seismic anchor motion were limited to less than Sa . Seismic anchor motion was included in one of the two places where it could be optionally included. If stresses including the thermal effects exceeded the allowable, a check was mad? with the stresses due to weight, pressure, thermal expansion, thermal anchor motion, and (optionally, as determined by the above choice) seismic anchor motion.

Allowable stress for the check was Sh plus Sa . In cases where stresses including seismic inertial effects exceeded the allowable stress, strain criteria were imposed. These criteria included checks for excessive strain, excessive fatigue damage, and 1ccal buckling potential.

The audits indicated that the licensee has taken steps to ensure that shortcomings in piping models found in the previous audits (discussed in Reference 1) were addressed. Calculations were done to ensure adequate masspoint spacing was used in the models. Missing mass calculations were done where appropriate. Valve weights and center of gravity locations appeared reasonable, and references were found to communications with valve vendors which corroborated the validity of some of this data.

A portion of the piping analyzed in one of the calculations (Problem No. 10) had been walked down by the author during a May 1987 inspection (see Reference 1). No discrepancies were found between the current model and the walkdown notes. l I

Although only one pipe support calculation was audited, it was found to conform to the criteria and methodology accepted by the NRC staff, except as noted below. Loads were correctly transferred from the piping calculation. A check was made to ensure that normal operating and seismic deflections were in the working range of the spring hanger. Bending and shear stresses were calculated for a beam supporting the hanger, and compared against allowables of 0.6 and 0.4 times the yleid strength of the material, respectively,. A linear interaction equation was used to  !

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, evaluate the beam's anchorage, with allowable loads taken as 1/4 the l ul tirtate.

Except for the cases discussed in the following subsections, the  !

piping calculations were found to be done according to the criteria and methodology found acceptable to the NRC staff in the SER and its clarification document (References 1 and 2). l 2.1.1 SRSS Combination of VC and RSS SAMs Audits of the piping calculations established that square root of the sum of the squares (SRSS) combination was used to calculate Vapor Container (VC) seismic anchor motions (SAMs) relative to the Reactor Support Structure (RSS). A check following the meeting of the DC-1 criteria document (Reference 4), the July 1987 SER, and of the author's audit, and confirmatory and example analysis records from the last review cycle established that this methodology had not been reviewed previously.

No precedent for its use could be found in a review of documentation for the Diablo Canyon or the San Onofre Plants. The general indication was that absolute summation was used. Specific guidance on this topic was found in Section 8.2 of NUREG/CR-0098, Reference 5. Absolute combination of, relative motion was identified as the appropriate combination method.

Since this method will produce larger SAMs than the SRSS method, and hence larger SAM stresses, it is more conservative than the SRSS method. Based on these findings, an acceptable basis for its use will be needed.  !

An informal discus: ion of the basis for SRSS combination was held with the licensee's consultants during the meeting. Acceptance of SRSS combination of VC SAMs and RSS SAMs was identified in an evaluation of the potential for impact between the RSS support columns and the VC at the i point of penetration of the columns into the VC during the structural portion of the previous review cycle. Sheets from previous meeting I summaries were provided in support of this (pages A-33 to A-35 of Appendix A). Review of these sheets established that the outcome of the impact issue was not dependent on the use of SRSS combination: use of absolute combination would have yielded the same result. A discussion with the NRC  !

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consultant involved in the previous cycle's structural review (N. C. Tsai)

I established that acceptance of SRSS combination was primarily based on a comparison of the natural frequencies of the VC and RSS. However, the consultant also stated that the consequences of impacting between the VC and the columns was not particularly severe, and that the consequences of an underestimation of the expected SAMs should be considered in reviewing the SRSS methodology.

This is significant because the SAM loading resulting from VC motion relative to the RSS is the predominant loading contributing to stress in some of the piping spanning the VC penetration to the RSS. Stress levels in this area were sufficiently high to require application of strain criteria for at least three piping calculations (Reference 6). Further, the fatigue check associated with strain criteria established that the critical piping locations ranged from 79 to 94% of the maximum allowable seismic fatigue damage using the SRSS methodology. It is not likely that the piping support configurations shown acceptable in the calculations would remain acceptable if absolute combination were imposed. Thus the methodology used to combine VC and RSS SAMs will, at least in some cases, determine whether the pipe support configurations are acceptable.

The licensee needs to provide an acceptable basis for the use of SRSS combination if it is to be used.

2.1.2 Anchor Bolt Interaction Eauation The audits established that the pipe support anchorage interaction equation defined in the DC-1 criteria document was not being uniformly applied in the calculations. DC-1 defines a 5/3 power equation. In one case, a lincar equation was applied in a piping calculation, in another (support WCD-H-4 on the Loop 2 feedwater piping), a second order equation was used. Although not done to DC-1, boti the linear and second order equations meet the requirement established in the SEP guidelines. As discussed in the July 1987 SER, a second order equation, based on the use of such an equation in the IE Bulletin 79-02 work, is acceptable. The 5/3 power equation was found acceptable based on envelopment of it by the 5

second order equation. Both equations found in the calculations are also acceptable based on envelopment. Because of this, the failure to adhere to approved criteria and methodology does not adversely affect the audited calculations. However, it generates a concern about the calculations not audited. The licensee needs to assure that unaudited calculations contain anchor bolt equations which are the same as or more conservative than the second order equation found acceptable by the staff.

2.1.3 Fatiaue Checks of Rod Hanaers Conversations held with the licensee's consultants established that evaluation of rod hangers did not include a check for excessive fatigue damage. This is of concern because of the potential for fatigue damage to rod hangers as a result of significant lateral motion of the piping attached to them. Rod hangers are particularly sensitive to fatigue damage because of the threaded connections which exist at each end of the rod. Although the licensee has committed to evaluating all small bore vertical-only supports (which includes rod hangers) for excessive lateral motion, no commitment was made to inclade consideration of fatigue. The licensee is cautioned that fatigue damage is a plausible failure mechanism which should be considered in the evaluation of relatively short rod hangers which are subjected to significant lateral motions.

2.2 Lic*ensee Commitments On June 1, 1987, the licensee issued a letter defining all commitments l l

made during the previous review cycle (Reference 3). Where possible,  ;

checks were made during the audits of the licensee's progress in meeting

, these commitments. The following paragraphs contain the results of the checks.

Several of the comitments made by the licensee involved doubling  ;

Yankee Composite spectra (YCS) stresses and performing checks using SEP guidelines criteria. The licensee attempte'd to apply this methodology and found it too conservative. The alternative chosen was to reanalyze piping i

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, - using spectra based on the NRC ground response spectrum (NRC spectra) when the doubling methodology was required. Since NRC spectra are preferred over YCS by the staff, this alternative is ar.ceptable. All calculations audited were found to employ NRC spectra. This corroborates the licensee's stated intent to use NRC spectra exclusively.

One of the licensee's comitments was to ensure that the potential was considered for run piping to dynamically amplify floor motion in trans-ferring it to branch piping. The audit of problem No. 10 (SG Blowdown piping) revealed that floor spectra had been applied at the anchor point representing the attachment of the blowdown piping to 8" feedwater piping. When asked about this, the licensee stated that the feedwater piping was sufficiently rigid in the area of the branch that application of floor spectra was acceptable. Since this was not documented in the calculation, the licensee was asked to produce supporting evidence.

Spectra for the steam generator (input to the feedwater piping) and spectra for the branch point were produced (see the attachment). A comparison of these spectra established that the piping in near proximity to the branch point is supported rigidly enough that there was no significant difference in the two sets of spectra. Application of floor spectra at the anchor representing attachment of blowdown piping to feedwater piping is acceptable. The licensee has met his comitment in the one caso checked.

The licensee has made a comitment to regenerate spectra and SAMs for all points of attachment of piping to the VC, and to check the existing calculations to ensure the new values are enveloped by those used in the calculations. When questioned about this, the licensee stated that the new spectra and SAMs had been calculated, and were being used in the piping calculations. Sheets A-36 and A-37 of Appendix A identify the calculations where this was done. The audits have established that the licensee is using the new values in reanalyzing piping attached to the VC. This licensee is meeting this comitment in the current work effort.

Licensee comitments to use acceptable methodology for calculating support loads for non seismic piping attached to gang hangers supporting seismic piping are discussed in the following section.

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. 3. DISCUSSION RESVLTS: GANG HANGER LOADS l

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During the meeting, the licensee presented a description of the work done in calculating support loads for non seismic piping attached to gang hangers which also support seismic piping. Handouts from the presentation are included in pages A-38 through A-46 of Appendix A. A more detailed description of the effort is provided in Reference 7. During the previous review cycle, four approaches acceptable to the NRC staff were developed for calculating the support loads. These are: (1) the licensee could perform equivalent static analyses to calculate the support loads, using 1.5 times the spectral peak as the applied acceleration; (2) the licensee could perform confirmatory analysis of a few representative samples to show acceptability of the practice previously used; (3) the licensee could provide a similarity argument between the seismic and non seismic piping configurations to justify the previously used practice of defining non seismic support loads in terms of seismic piping loads; and (4) the licensee could perform standard response spectrum analyses of the non seismic piping to obtain the support loads, combining the results for the gang supports using the absolute summation combination method. Options 2 and 3 above required review by the NRC staff.

The approach taken by the licensee was a combination of items 3 and 4, above. The non seismic piping was reviewed, and two lines were chosen to represent the balance of the non seismic piping. A 3" line was chosen to represent the 3" piping, and a 2" line was chosen to represent 2" and I" piping. The lines were chosen to maximize free span between supports and pipe size, and to require spectra for the highest attachment points among the candidate lines. Drawings laying out the pipe routing were reviewed (96994P-40A, Revision 5, and 9699-FP-408, Revision 7). The licensee's judgement that the two lines chosen for analysis would provide conservative loads for all non seismic support loads was accepted.

Analyses of the the non seismic piping were done using the criteria and methodology defined in the DC-1 criteria document (Reference 4). This assured accurate loads were calculated, and that failure of the non seismic piping would not significantly change the support loads. The licensee reviewed the gang hanger designs with the new loads in hand and 8

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. decided to reduce the loadings by grouting non seismic piping penetrations of interior concrete walls. The analyses were redone with the new support configuration, and lower loads were obtained. The evaluations of the gang hangers using the new loads were not complete, and hence not available for audit.

During the presentation, the licensee asked for a statement concerning the need for future case-by-case reviews of this approach. Guidance in this is provided in the September 1985 Safety Evaluation Report for the San Onofre Nuclear Generating Station (Reference 8). In this SER, all applications of similarity were required to be reviewed on a case-by-case basis. The requirement was based on the wide variety of means available to establish similarity. Essentially, the use of similarity depends too heavily on engineering judgement to be allowed without the collegial process implied in a case by-case review. Since the approach under review here depends on similarity, case-by-case reviews will be needed for any future applications. Lacking a staff commitment to perform the review, the licensee would need to analyze all the non seismic piping attached to seismic gang hangers using the analytical procedures previously approved by the staff. As an alternative, the licensee could use the equivalent -

static analyses discussed under option 1, above.

4. CONCLUSION l

Four calculations were audited during the meeting. The audits '

established that the criteria and methodology accepted by the NRC staff during the previous review cycle are being correctly employed in the calculations. One issue was generated, with the following associated action item: The licensee needs to provide justification for using SRSS combination of VC and RSS SAMs to obtain VC SAMs relative to the RSS. In addition, the licensee is cautioned against ignoring the potential for rod hangers to fail in fatigue when those hangers are relatively short and subjected to large lateral displacements Dy the attached piping.

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In all cases where the audits touched on areas affected by the I licensee's commitments made during the previous teview cycle, the licensee was found to be following those commitments. l The methodology adopted by the licensee for calculating support loads for non seismic piping attached to seismic gang hangers is acceptable for the application reviewed during the meeting. Future applicatien of the methodology to piping not reviewed during the meeting will require  !

case-by-case reviews by the staff. Acceptable alternatives to this methodology which do not require case-by-case review are identified in Section 3 of this report.

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4. REFERENCES

1. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) dated July 16, 1988,

Subject:

NUREG-0825, Section 4.11 Seismic Design Considerations (TAC No. 51807).

2. Letter from H. B. Fairtile (NRC) to G. Papanic (YAEC) dated March 21, 1988,

Subject:

Clarifications to the July, 1987 Safety I Evaluation Report.

3. Letter from G. Papanic, Jr. (YAEC) to E. McKenna (NRC) dated i June 1, 1987,

Subject:

SEP Topic III-6 Commitment Summary.

4. Letter from M. B. Fairtile (NRC) to G. Papanic (YAEC) dated l April 30, 1987,

Subject:

SEP Topic III Technical Information.

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5. NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected !

Nuclear Power Plants," May, 1978. '

6. Letter from N. H. Williams (Cygna) to P. Y. Chen (NRC) dated April 19, 1988,

Subject:

Alternate Criteria Usage YNPS SEP Piping Job No. 87150.

7. Letter from N. H. Williams (Cygna) to P. Y. Chen (NRC) dated May 9, 1988,

Subject:

Gang Support Load Generation YNPS SEP Piping Job l No. 87150. i

8. Letter from H. Thompson, Jr. (NRC) to K. P. Baskin (SCE) dated September 19, 1985,

Subject:

Long Term Service (LTS) Seismic Criteria '

and Methodology - San Onofre Nuclear Generating Station, Unit 1.

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i TABLE 1. LIST OF ATTENDEES NAME AFFILIATION l

PEI-YING CHEN USNRC/EMEB MORT FAIRTILE USNRC BRUCE HOLMGREN YANKEE DONALD LEFRANC0lS YANKEE DARLENE LEONG Cygna GEORGE PAPANIC, Jr. YANKEE HARK RUSSELL INEL T. NEAL WATTS 1 Cygna NANCY WILL Cygna -

T.Y.WONG{AMS Cygna i

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1. These individuals attended only limited portions of the meeting.

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- TABLE 2. LIST OF ACRONYMS ACRONYM DESCRIPTION IE INSPECTION AND ENFORCEMENT DIVISION OF THE NRC I NRC NUCLEAR REGULATORY COMMISSION

! SCE SOUTHERN CALIFORNIA EDIS0N CO.

! SER SAFETY EVALUATION REPORT SRSS SQUARE ROOT OF THE SUM OF THE SQUARES COMBINATION YAEC YANKEE ATOMIC ELECTRIC COMPANY YANKEE THE YANKEE NUCLEAR POWER STATION YCS YANKEE COMPOSITE SPECTRA i

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o TABLE 3. LIST OF AUDITED DOCUMENTS TOPIC CALC NO. REVISION REMARKS

$orkInstruction 1 0 Procedure for piping analyses.

Boiler Feed Piping 87150/9/F 0 Strain criteria Inside VC were applied.

Calculation for support WCD-H-4 was audited Main Coolant and 87150/30/F 0 Strain criteria Pressurizer Urains were used. .

Steam Generator 87150/32/F 0 Blowdown Piping i

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E APPENDIX A HEETING HANDOUTS l

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YANKEE ATOMIC ELECTRIC COMPANY su\s# '

US NUCLEAR REGULATORY COMMISSION CYGNA ENERGY SERVICES NAY 2 - 4, 1988 TECHNICAL REVIEW MEETING AGENDA  !

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l o INTRODUCTION OF ATTENDEES I

o MEETING OBJECTIVES )

l o YNPS PIPING DESIGN CRITERIA I l

o PIPING ANALYSES IN PROGRESS o GANG SUPPORTS o DESIGN REVIEW (CASE-BY"CASE) i o RPV SLIDING EVALUATION l

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YNPS PIPING DESIGN CRITERIA I. CRITERIA DEVELOPMENT Deceaber 1977 - NRC issues gencrio letter describingthe Systeimtic Evaluation Program (SEP)

March 1978 - NRC issues generic letter presenting the detailed description of all SEP topics October 1980 - NRC orders Yankee to inititate a program "to demonstrate the seisnio design adequacy of the facility" January 1981 - Yankee letter to NRC documents information provided to the staff on the preliminary seismic evaluations for YNPS February 1981 - Yankee letter to NRC submits Retrofit Criteria Document (DC-1), Rev. O.

April 1981 - Yankee submits the Yankee Composite Spectrum (YCS) to NRC June 1981 - NRC finalizes their site specific spectrum

• December 1981 - Yankee letter to NRC defines scope of the seismic evaluation as those reactor systems and structures necessary for maintaining the plant in safe hot shutdown.

May 1982 - Yankee, NRC, Cygna and EG&G meet to discuss the results of the seismic evaluatiens of the hot shutdown system piping July 1982 - NRC issues SED Guidelines, Rev. O August 1982 - Yankee, NRC, Cygna IG&C meet to discuss results of the seismic evaluations 7f hot shutdown system structures and piping September 1982 - NRC issues SEP Guidelines, Rev.1 Yankee begins installation of 76 large bore pipin6 seismic support February 1983 - NRC issues safety evaluation on SEP Topic III-6 July 1983 - NRC issues Integrated Plant Safety Assessment for )

YNPS (NUREG-0825) l February 1986 - April 1987 - Yanxee, NRC, Cygna. EG&G and LLNL meet l to discuss evaluations preformed by Yankee /Cygna for Topic III-6 July 1987 - NRC issues safety evaluation for SEP Topic III-6  !

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August 1987 - Yankee requests clarification of the staff's position for 10 items in the Topic III-6 safety evaluation March 1988 - NRC issues safety evaluation for the 10 items from the SEP Topic III-6 SE for which Yankee requested clarification.

II. SAFE SF"TDOWN SYS*D1 SCOPE A. Rh or Coolant Pressure Boundary (RCPB - to insure primary sysatem integrity B. Secondary Coolant Pressure Boundary (SCPB) - to insure secondary system integrity.

C. Dedicated Safe Shutdown System (DSSS) - provides primary and secondary makeup and provides for decay heat removal. '

III. DESIGN DOCUMENTS A. Seismic Reevaluation and Retrofit Criteria (DC-1), Revision 4: -

Sections 8 (Piping) and 9 (Supports)

B. Safety Evaluation for 10 items requiring clarification, (Letter, NRC to TAEC, dated March, 1988) - Items 4 and 6 C. Cygna Pipe Stress and Pipe Support Work Instruction for YNPS SEP Piping D. Alternate Criteria Worksheets IV. ALTDLNATE CRITTRIA A. Strain Criteria - applied in accordance with Section 8 3 1 of the Seismic Reevaluation and Retrofit Criteria (DC-1), Rev. 4, SEP

. Topic III-6 Safety Evaluation, and alternate criteria worksheets.

B. SRSS combination of seismic inertia and seismic anchor motion loads on supports - applied in accordance with March 21, 1988 Safety Evaluation and alternate criteria worksheet=.

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l] 8.0 PIPING ANALYSIS CRITERIA

,a This section describes the criteria to be used in the stress analysis of the piping systems which are part of the Safe Shutdown System or are otherwise included in the seismic scope.

8.1 Load Description The following load cases shall be considered for the piping stress analysis.

In addition, local stress concentration due to integral supports shall be evaluated.

8.1.1 Thermal Load Loads due to steady state temperature effects, including thermal anchor mo vements. -

8.,1.2 Weight Load Leads due to the weight of the pipe, its contents, and its insulation.

8.1.3 Pressure load i Loads due to the steady state internal design pressure. ,

8.1.4 Seismic load i Loads due to earthquake excitations, including both seismic inertia effects and seismic anchor movements. Pipe stresses due to these loadings are limited to the values in Section 8.3.

i Yankee Nuclear Power Station 51 Seismic Reevaluation Criteria '

80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 kill!!Ill liittilfililll$1ll-0439B A

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(a) Each problem shall be considered from anchor to anchor. If an k./ . anchor to anchor prslem exceeds program limitations, it may be broken up into smaller problems with adequate overlap areas. The overlap areas will be chosen to properly represent the truncated portion of piping. The results of the multiple computer runs in the overlap areas will be enveloped to assess boundary or loading conditions. All usage of overlap criteria will be justified on a case by-case basis.

(b) The geometry and restraint conditions shall be modeled in accordance with as built isometrics.

(c) The piping analysis shall be performed using Yankee Piping Specifications (YS-497 .and YS-4652), YAEC flow diagrams, Yankee Insulation Specifications (YS-2304), and vendor catalog data, pipe i and pipe support material properties for the specified analysis conditions are stained from Appendices A, B, and C of Reference 3(a).

, ,o (d) Branch lines may be decoupled from run piping if the following three criteria are met:

3 o The ratio of m ments of inertia (run/b ranch) is 25:1 or ,

greater. '

I o There are no anchors or supports on the branch line close to the branch point which serve to restrain the run pipe. j o There are no nozzles on the branch line close to the branch I point which restrain the run pipe.

When analyzing the branch line, the point of decoupling from the l run line shall be considered an anchor. The deflections of the run piping at the decoupling point shall be input as anchor motions for

, the branch line analysis.

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• r Yankee Nuclear Power Station 53 g [ ;, ,

Seismic Reevaluation Criteria ,

tillitiltilillit!Illilr!!Illil 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 )

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! Seismic accelerations of the valves shall be sumarized. The I l

allowable valve accelerations shall be [ Reference 4(h)]: 1 Resultant horizontal acceleration < 4.25 g f

Vertical accelerations < 3.0 g (g) Flanges shall be considered as additional luged weights. Flange thicknesses shall be assumed to be the same as that of the pipe for purposes of codeling stiffness. Additional flange information may be obtained from ANSI B16.51977.

(h) Stress intensification factors for tees, reducers, flanges, elbows and couplings (half and. full) shall be considered as per Appendix 0 of Ref erence 3(a).

C (1) Penetrations shall be analyzed as follows:

Grouted penetrations: A bilateral restraint condition snall be 3 considered to eFiSt on either side of the penetration for all load cases. Generally, axial restraint of the pipe shall not be

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considered unless the pipe has a valded collar which is enbedded in the penetration. However, for piping exerting relatively low axial O loads on the penetration, the grout may still afford axial restraint. In these cases, axial restraint may be taken if the bond between the pipe and the grout remains intact. This check will be made if credit is taken for the axial restraint.

Ungrouted penetrations: At ungrouted penetrations, deflection of the pipe less than 1/4" shall be considered acceptat,le. Where deflections exceed 1/4", further review of actual penetration j clearances shall be initiated. Deflections shall be based on the conbined thermal, deadweight, and seismic conditions.

M3 IHHHulHHunnuunHI Yankee Nuclear Power Station Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 e

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. l piping shall be analyzed. The thermal anchor movement (TAM) stress due to normal operating temperature shall be added to the thermal expansion stress to cbtain the total thermal stress. Zero deflection criteria may be used based on the recommendations of WRC-300.

8.2.4 Seismic Analysis l

1

, (a) The basic dynamic analysis technique will be the response spectrum  !

modal superposition method using lumped mass models. Sufficient )

mass puints shall be used in the computer mdel to adequately l represent the mass distribution. When available, the ' Automatic Mass Point Spacing' option shall be used. When not available, the maximum span length between mass points shall not exceed: )

I L < f {h} h} (Eq. 8.2.4)

Where: f = Cutoff frequency, Hz l

E = Young's modulus, psi i 1 = Pipe moment of inertia, in 4 w = Weight / unit length of pipe + contents, Ib/in L = Maximum allowable span between mass points, in For rod hanger supports, when the uplift due to seismic load (include thermal load if it is upward) is larger than the weight load, the effect of the rod hanger support on the system shall be considered as follows: Two seismic analyses shall be considered.

In the first analysis, the rod hanger shall be considered ef fective. In the second analysis, the particular rod hanger support will not be included in the model. The results of the two

analyses will be enveloped.

Both seismic inertia analysis eM seismic anchor movement analysis shall be performed.

Yankee & clear Power Station 57 i Seismic Reevaluation Criteria tillllliittiillillilill!!illi 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 0439B ,, e-7

(c) Cut-off frequency and minimum nunber of trodes:

A cut-off frequency of 33 cps, with no less than 10 modes, shall be considered in the analysis. An equivalent static seismic analysis, using a constant spectral acceleration at the 33 cps cut-off frequency, shall be performed when the contributions of higher modes (>33 cps) are significant. The results of the static analysis shall be conbined by SRSS with the dynamic results.

(d) Damping values _:

For either the YCS or RC seismic event, damping as specified in USRC Regulatory Guide 1.61 or ASME Code Case N-411 [ Reference 3(m)] shall be used. The two different dampings (Regulatory Guide 1.61 or N-411) may not be used concurrently on the same piping pr& lem, f

8.2.5 Seismic Anchor Movement Analysis (SAM)

The seismic anchor movement load condition shall be considered for both stress and support load evaluations. SAM will generally be conbined with seismic inertia loads by sunning absolutely. However, the SRSS method .

may also be used with RC spectra loadings.

8.2.6 Pressure Effect i.

The effect of internal pressure shall be con 1)ered in computiag longi-tudinal stresses.

8.2.7 Effects Due to Relief Valve Blow-Off and Other Occasional Loads The effects due to relief valve blow-off and other occasional loads shall be calculdted and applied as external forces to the piping.

Yankee Nuclear Power Station

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Note: The pressure stress term in Equations 8.3.1-A, -B, and -D may be replaced by the following: l Pg 0,2. d 2 i

where d = inside diameter of the pipe.

(b ) The effects of pressure, weight, other sustained loads, and  !

earthquake est meet the following requirements:  !

I PD I O + 0.751 g + '

(Eq. 8.3.1-B) 3 4t n Z A

Z b " I*"h i for the Yankee Composite Spectra.

\

where MB = Resultant mment loading on the' cross section due to earthquake loads, in.-lbs. All other terms are the same as in Equation 8.3.1-A.

For the 5C spectra, the piping systems must be shown to remain U

functional under the effects of pressure, weight, other sustained loads, and earthquake. As a first check, the SEP allowables will be used in Equation 8.3.1-B as follows: 1. 8 S h for Class 1 and 2.4Sh for Class 2 and 3 piping. In cases where these allowables are not satisfied, the equivalent strain at any point in the system exceot equipment and valve nozzles, threaded connections, the  !

specially fab ricated tee and fork components in the Mah Steam /Feedwater piping outside the VC, shall be limited to 1%.

Stress at those nozzles and threaded connections shall meet the SEP

allowab les. The strain calculations will be used and justified on a case by-case basis.

' Yankee Nuclear Power Station 61 g( , Seismic Reevaluation Criteria .

80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 IIlittlitttitittlltitttttlitti 04398 -

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Elastically calculated displacements in the area of inelastic

(

b eha vior shall be multiplied by 3.33 to account for added deformation. The effect of the added displacements shall be reviewed.

For stainless steel piping with strains greater than 0.01, elastica 11y calculated flanged joint, nozzle and support loads -

shall be multiplied by the following factor, X:

X = 1.0 + 170.0 ( c - 0. 01)

(c) Thermal Expansion Stress (SE )

1M SE* z (Eq. 8.3.1-C) where:

s, MC = The range of resultant moments due to thermal

/ expansion, i n. -lb s. Also include moment effects of anchor displacement due to earthquake if anchor disp.lacement effects were omitted from Eq. 8.3.1-B

,' Fohhe d *T.dse, since Equation 8.3. da ,

3

Level A/B str comparison, and since PRC Spectra l l represent noch grea load case an an OBE, 60% of {

/ the SAM effects will b d in Mc. This 60% is !

intended to limit SAM result o the same levels I as an OBE ce e analysis would. Howev if the 60%

S ination is used at any point, th. train,/

,, 4 eria shal not be used at the samegnt.__. -

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(f) Although only the response spectrum analysis method is considered in this criteria, it does not preclude the possibility of using the time history analysis method, if the situation warrants its application. Specific criteria for time history analysis will be provided when the need arises.

8.3.? Allowable Stresses  !

Allowable stress values to be used for safe shutdown piping systems are given in Appendix A of Reference 3(a). Those values shall be used for piping stress analyses.

For' material allow 2 1e stress values not available in Appendix A of Reference 3(a) or for material yield stress values at design I

tegeratures, Reference 3(g) shall be used. The appropriate allowable stress values shall be taken from tables contained in Appendix I of the reference.

8.3.3 Allowable Deformations .

Deformations will be limited to existing clearances to prevent igact of adjacent components. Also sr.e Section 8.2.1(1).

8. 4 Small Bore Pipe Stress Analysis l This section applies to piping with a nominal outside diameter of 2" or smaller. The stress qualification for small bore piping shall be performed using one of the methods outlined in Sections 8.4.1 and 8.4.2.

8.4.1 Detailed Stress Analysis  ;

For detailed stress analysis, the same procedures and methods as those i

for large pipe stress analysis shall be folloved (Sections 8.1 through ,

I Yankee MJclear Power Station 65 Seismic Reevaluation Criteria i 061/86064 Nc. %. L1; Rev. 4 litililill lillillfilllilill!

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p distributions will be postulated where necessary, to cbtain less

\ severe thermal expansion stress distributions.

(c) Seismic Stress - Seismic pipe stresses shall be avaluated by the equivalent static mithod. Span stresses are calculated assuming simply-supported spans b etween supports. For the initial evaluation, the peak of the response spectrum is used to accelerate the sigly-supported span. Since 100% mass participation is assumed for this static "first mode" which maximizes span stress. a dynamic aglification is not used.

If the piping is overstressed using the peak acceleration from the response spectra, an evaluation is performed to determine if

. frequency testing or detailed analysis should be the next course of action. Zero period acceleration (the acceleration at 33 cps, in this case) is used to accelerate the pipe span. If the span S qualified by this method and appears to be relatively rigid (high fundamental frequency), frequency testing is warranted. If the span is not qualified by this method, detailed analysis or addition of supports is necessary.  ;

For some piping systems, frequency testing will be performed for three orthogonal directions to ob tain less conservative

accelerations for the equivalent static analysis. The testing methods are describ ed in Section 8.4.3. The acceleration

~

corresponding to the tested fundamental frequency will be used in l the piping analysis, if the fundamental frequency is greater than the frequency at the peak acceleration. Again, the dynamic l amplification factor for these cases is taken as 1.0. If the fundamental frequency of the piping is below the frequency at the j peak acceleration, the peak acceleration is retained in the analysis. '

1 i

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excitation will be used for relatively flexible lines by displacing the pipe and releasing it to vibrate freely. At least four individual tests for each response direction and location will oe m de.

To obtain the final frequency response functions to be used in the stress analyses', the individual frequency response functions will be avecaged and plotted. The frequency of the first significant response peak will -

be used as the fundamental frequency of the span for the tested direction.

M Buried Pipe Stress Analysis Buried piping shall be analyzed per Section 6.2 of Appendix F.

8. 6 Main Steam /Feedwater Piping Outside the Vapor Container 8.6.1 Geonetry and Computer Modeling

>O The main steam /feedwater piping outside the VC is supported on a series of light structural frames. The mass and stiff.iess of the support j structure and piping runs are of similar magnitude; therefore, the rigid I l

support assumption generally used in the majority of the piping analyses

) is not valid here. Additionally, a nunber of pipes may be supported by the same frame, promoting interaction between the pipes. For the above reasons, the main steam /feedwater piping and support structure shall be analyzed together in one problem. Non-seismic piping shall be modeled and analyzed if it affects the response of the seitmic piping and the support structure.

The conbined main steam /feedwater piping and support structure system is large and complex, making it prohibitively costly and time-consuming to analyze a detailed mode' of the entire system. To minimize the size of the actual seismic analy h model, s* structuring methods may be used to Yankee Nuclear Power Station 69 Seismic Reevaluation Criteria Illlllilli lillll1111111 til 0439B M -/#/

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9. 0 PIPE SUPP0lT DESIGN CRITERIA 9.1 Introductio2 The following criteria shall be used to evaluate or redesign pipe supports at l the Yankee Nuclear Power Station.  !

9. 2 Codes. Standards and References, The following codes shall be used for the design of pipe supports:

i 9.2.1 ASME, Reference 3(g),Section III, Stbsection NF,1977 edition. I 9.2.2 ANSI, Reference 3(a).

9.2.3 AISC, Reference 3(c).

l 9.2.4 ITT, Reference 4(ap).

9.2.5 Hilti, Reference 4(ad). '

9.2.6 Other manufacturer's ptblished catalogs.

9. 3 Loading Description All loadings obtained from piping stress analysis shall be used for support desi gn.

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Yankee Nelear Power Station Seismic Reevaluation Criteria 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 76 04393 .3 (. /) /)

,. indi vidual responses of. each of the three components of

i. earthquake load shall be used to review the support. Seismic loads shall be considered to act in both the positi ve and negative restraining directions.

9.3.2.1.1 Seismic loads considered are (a) the Yankee Composite Spectra (YCS) or (b) the RC Spectra (RC).

9.3.2.1.2 Seismic Anchor Movement, SAM . Loads cbtained from seismic anchor movement of either YCS or EC (SAMYCS and SAMSC) shall be considered. For supports, the full SAM loads must be used, not 60% as allowed in the pipe stress criteria (Section 8.3.1-c) 9.4 Loadino Conbinations lead Case load 1 D + TH + TAM + F4.

2 0 + TH + TAM + YCS + SAM YCS 3 0 + TH + TAM + RC + S AM EC Load Case 3 will be used for the confirmatory arialyses and selected example analyses. Thermal and TAM loads will only be added to the load conbinations if they increase the magnitude 'of tbs design load. The larger of lDj and lD + TH + TAM l will be used as the deadweight + thermal contribution to the design load. ' ,

9. 5 Frequency For supports which are assumed to be rigid in the large bore piping analyses, the natural frequency of a seismic restraint with its tributary pipe mass must be greater than 33 Hertz in the pipe's restrained direction. The mass used to calculate the natural frequency shall include the weight of the restraint, Yankee Nuclear Power Station 78 Seismic Reevaluation Criteria .

JL A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 liitttttltitt!!!!!Ititittititt 0439B , f)-/6

requirements of Appendix A of USMC Standard Review Plan 3.9.3. Limiting the

() emergency / faulted allowables to 0.9F y ensures support functionality.

For non-symetric structural steel sections such as angles, principal axis properties shall be used for all evaluations.

l For anchor bolts and other catalog items, vendor-specified allowables shall be used with appropriate factors of safety. Table 9.6-1 sumarizes the component allowables used for the qualification of pipe supports at YNPS. For some catalog items, the allowables are given at elevated temperatures; however, the  ;

conponents may be used in supports stbjected to much lower design i temperatures. In these cases, the catalog allowables may be increased by a i ratio of material allowables at the design temperature versus the catalog I temperature. l When evaluations to the MC spectra are performed, they will be first done to the allowables discussed herein. However, when these are not satisfied, additional refinements will be considered and reevaluations performed. Tiese

. O refinements wiii inciude:

a) Increased anchor bolt allowables. A safety factor of less than 4.0 but greater than 2.0 shall be permitted for interim operation. ,

b) Increased anchor bolt allowables. A safety factor of less.than 4.0 but greater than 2.0 shall be permitted for long tem operation provided that -

1) the base plate in question has at leasrt four anchor bolts, and not more than half are simultaneously stbjected to tensile loads, f

2) loads greater than that associated with a factor of safety of 4.0 can be redistributed to adjacent supports.

Yankee Nuclear Power Station 80 Seismic Reevaluation Criteria L. LJk A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lililillllllilllll!!!!!Illill!

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Table 9-1 Allow &1e Stresses for Pipe Supports SERVICE LEVEL STRESS -

LOAD CASE 1 LOAD CASES 2 AND 3 Value KS!(1) Value KSI(l)

Tension 0. 6 Fy 19.8 0.9 Fy 29.7 Shear 0. 4 Fy 13.2 0.6 F y 19.8 W& Crippling 0.75 F y 24.8 0.9 F y 29.7 Conpression F 5maller of 1.5 Fg or a

2/3 F ee O se#4'#9 o e e, 19 8 o9e y z9 7 Bearing 0. 9 F y 29.7 N/A NA Bolts Tension Allowele Tension 1.5 X (Alloweie Tension per AISC per AISC)

Shear Allowe le Shear 1.5 X (Allowable Shear per AISC per AISC)

Anchor Bolts Hilti (2) (2)

Star Slug-in (3) (3)

Through Bolts . (4) (4)

Toggle Bolts (5) (5)

Yankee Nuclear Power Station 82 Seismic Reevaluation Criteria .

i L k A 80023/81060/81061/86064 Doc. No. DC-1; Rev. 4 lililllilllilillllllllllllllll 0439B , .

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)S/3 ( YA )5/3 < 1. 0 Additionally, the applied shear, V, shall not exceed 40% of the tensile allowab le. If the applied shear exceeds the 40%, the exponent shall be reduced from 5/3 to 1. Bolt spacing and edge distance shall be considered per vendor requirements.

(3) Allowable loads for Star Slug-in anchors shall be as given in Table 9-2.

These values are based on information provided in Ref.4(aq).

(4) Pipe supports shall be attached to masonry walls by either through bolts, toggle bolts or other types of connectors developed specifically for use with hollow masonry. Through bolts are preferrable and will be used where practicable. Allowables for through bolts shall be per the AISC Code, 8th 1 Edition.

(5) Pipe supports shall be attached to masonry walls using toggle bolts if through bolts are not practical. Allowables for toggle bolts shall be limited to manufacturer values.

(6) If catalog allowables are given for emergency / faulted loading conditions, they will be used instead of the allowables calculated here.

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l Yankee Nuclear Power Station 84 Seismic Reevaluation Criteria L l L A 80023/81060/81061/86064 Doc. %. DC-1; Rev 4 I.'Jilllililllllllllilllilllll 0439B , ,

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! \' of all calculations which contain features which could not be accepted I in general by the staff, so that acceptance on a case specific basis was required. Audits of YAEC's piping and support calculations were also conducted to assure that the criteria and methods accepted by the staff were correctly executed. Thesearedetailedrevjewsofa set of calculations. Detatis of the audits and case-by-case reviews can be found in Attachment 3.

4.4.1 Criteria and Analysis Method for Pipino and Supports The bases for acceptance of criteria and methodology by the staff in- l cluded references to regulatory documents defining practices acceptable i to the staff, such as the SEP guidelines, the Regulatory Guides, the '

' Standard Review Plan (NUREG-0800); Codes and Standards acceptable to ,

the staff such as the ASME Boiler and pressure Vessel Code, and the AISC Specification; and previous Safety Evaluation Reports issued for other SEP plants, particularly the SER issued for the SEP review of l San Onofre Nuclear Generating Station Unit 1 (SONGS 1) (Ref. 39). l a- Bases demonstrating a favorable comparison to such practices were '

also acceptab,le.

Based on (1) the staff's review of YAEC's criteria and methods, (2) the audits and case-by-case reviews of the calculations, and (3) the confirmatory and example analyses, the staff has identified accep- '

I table criteria and methods for use in YCS analysis of piping and supports, and criteria and methods &cceptable for use in NRC spectrum analysis of piping and supports. These are summarized separately for large-bore piping, small-bore piping, the main steam /feedwater piping and support structure out of containment, and pipe supports. They are discussed in greater detail in Attachment 3, where the bases for their acceptance is discussed at lengt.h.

4.4.1.1 Laroe-Bore Pipino Large-bore piping was analyzed for the effects of weight, pressure; thsrsal expansion, thermal anchor motion, seismic loads (including seismic inertia and anchor motion), and other loads (including reitef valve blow-off, snow loads, and wind loads). Two levels of seismic loads were defined, one related to YCS, one related to the NRC spectra. These are the two seismic loads found acceptable in the IPSAR. Different criteria and methodology were used for the two seismic loadings, as discussed below.

Finite element analysis techniques were defined, including three 1 dimensional lintar e?artic response spectrum analysis for seismic inertia loads. Tht three directions of simultaneous excitation were combined per Reguistory Guide 1.92.

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Component flexibilities, stress intensification factors, and i material properties at temperature were defined using the  ;

i Appendices to the ANSI B31.1-1977 Code. If not available '

there, the Appendices to the 1977 ASME Boiler and Pressure Vessel Code were used. This is acceptable to the staff based 1 on a review of a favorable comparison between these sources of data and the Code of record defined in the SEP Guide- l lines, the 1980 ASME Code, Winter 1980 Addenda.

! l

! The following criteria were accepted for decoupling branch piping from run piping in the run piping analyses: (1) The

! coment of inertia ratio of run pipe to branch pipe must be I 25:1 or greater; (2) there are no anchors or supports on the .

~

i branch line in close proximity to the branch point; (3) there .

are no nozzles on the branch line close to the branch point. l

' Acceptance was based on previous acceptance of this meth-  !

odology per the SONGS 1 SER. For decouplin l the branch piping in a branch piping model,g run piping additional from con- I i *- sideration of the suitability of application of floor response l _

spectra at the branch points was required because of the expected additional amplification of the floor response spectra from the ,

anchor point of run pipe to the branch point. This requirement .

I was based on the finding of the example analyses reported

' in Attachment 3. The licensee reviewed all affected piping j l calculations and identified calculations where application of 4

floor response spectra was not appropriate. To correct this deficiency, the Itcensee committed to generating response 1

soectra at the branch points usin of the run piping in some cases. gInaother time history analysis cases, piping models were to be consolidated or extended to remove the branch point terminations. Some branch piping models were acceptable as-is, based on the rigidity of the attached run piping. The 1

licensee's efforts in this area were reviewed in detail by the j staff and its consultants, and were found acceptable. ,

! Terminations of the piping at equipment nozzles wert either 1 defined as anchors, or defined according to the provisions of 4

Welding Research Council Bulletin No. 297 (WRC-297). All applications of WRC-297 were reviewed by the staff and its

] consultants on a case-by-case basis and found ecceptable.

Acceptance of this methodology and the restriction on its application was based on the SONGS 1 SER.

The masses of manually o at the pipe centerlines.perated Theand check center valves were of gravities modeled (CG) of air- i i

or motor-operated valves were modeled an appropriate distance off of the pipe centerlines. When possible, the CG offset was as specified by the valve manufacturer. When this*in-i l formation was not available, the total mass was concentrated 1

at a point 1/3 the stem length from the valve centerline, i l

The second method was not acceptable for defining the CG l locations of small-bore motor operated valves. YAEC provided i a basis for this methodology, which was reviewed and found acceptable by the staff. I

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' (C' excess of 1% (this being applicable only to stainless steel piping j due to the 1% limit for carbon steel). The clearance check 4 performed in the plant uses displacements increased by a factor

of 3.33 in areas of strain criteria application. Based on the SONGS 1 SERs, these criteria are acceptable to the staff, provided I that applications nf strain criteria in excess of 1% are submitted &

to the staff for case-by-case review. Although YAEC proposed the use of time history analysis in place of response spectrum analysis,

this analysis technique was never used. Therefore, the staff did not formulate a position as to its acceptability.

l , 4.4.1.2 Small-Bore Pioina .

YAEC applied the same criteria and methodology as was approved .

i for large bore piping to a number ofremall-bore piping analyses.

This is acceptable to the staff based on acceptance of the criteria and methodology for large-bore piping.

Alternatively, YAEC used a simplified approach to analyzing

! ~ small-bore piping based on simplified equivalent static analysis techniques. Simple span calculations were perfomed to obtain i

2 stresses due to weight, thermal, thermal anchor motion, seismic, and seismic anchor motion loads. The acceleration corresponding i to the peak of the' associated spectra was used in the seismic

! analysis. The calculated stresses were combined in the same l i

4 fashion as for the large-bore stresser, and compared to the same l allowable stresses. One exception to \his was the pressurizer  !

] spray piping problem Nos. 2 and 3, whic.h vere analyzed using the 1 l 1983 ASME Code. This exception was reviewei and found acceptable

by the staff.

YAEC reduced the acceleration used in seismic analysis in some

!" cases by utilizing frequency test results. In these c.ases, the acceleration used in the calculation was the maximum of the , ,

] associated spectra at or above the fundamental natural frequency i i established by test. Test procedures and sample test results were  !

reviewed and found to be an acceptable means of establishing the 1 fundamental natural frequency of the piping. l lO l l The criteria and analysis methods proposed for analyzing small bore piping are acceptable to the staff based on similarity to j simplified anal l'

review of YAEC'ysis methodology s benchmark found acceptable calculations, in NUREG-0800, and review of the test j procedures and sample test results.

1 4.4.1.3 Pipe Support I Loads defined for the analysis of pipe supports includ'ed the 1

effects of weight (D), thermal expansion (TH). thermal anchor i motions (TAM), friction (FL), and earthquakes (both YCS and NRC j , spectrum loads (NRC)). Both seismic inertia and seismic anchor I

1 o-22 m

Allowable stresses for load case No. 3 (D + TH + TAM + NRC) were generally the same as those for load case No. 2, and were accepted based on a favorable comparison to the pipe support criteria found in the SEP Guidelines. The staff approved one supplement to the SEP Guideline, criteria. Anchor bolt factors of safety were allowed to decrease to a factor of safety of 2 (versus the normal factor of 4) for a limited interim period. In addition, the staff accepted the combination of seismic inertia and anchor motion loads using the square root of the sum of the squares (SRSS) combination methodology.

A 5/3 power interaction equation in conjunction with factors of safety of 4.0 for expansion anchors and 5.0 for shell type anchors was proposed for evaluating concrete expansion anchor bolts under both YCS and NRC spectrum loading. This was found unacceptable for YCS .

analysis because it failed to provide the margin necessary to ensure that SEP Guideline requirements are met under NRC spectrum loadings.

Consequently, the licensee made a comitment to evaluate anchor bolti in YCS analysis by doubling the YCS seismic loads and treating the resulting loads as NRC spectrum loads. For NRC spectrum loads, the 5/3 power interaction equation was accepted by the staff based on a favorable comparison to the methodology of IE Bulletin 79-02.

Audits of YAEC's support calculations established a short-l coming in the criteria used. No check was performed to assure that one-way supports (rod hangers, pipe rests etc.) would not be C/ subjectedtounacceptableloadsindirectIonsforwhichsupport actbn Vas trot credited. Criteria acceptable to the staff for such a check were developed, and YAEC committed to checking all affected pipe support calculations using it.

The review established that YAEC was calculating loads on gang hangers for non-safety related piping based on the loads for safety related piping attached to the hangers. An acceptable basis for this was not developed. Although this issue could not be resolved during the review, resolution strategies acceptable to the staff were developed, as discussed in the section below defining the status of piping and supports.

Information Notice (IN) 86-94, which reduced allowable loads for some sizes of Hilti Kwik-Bolts, was released during the course of the review.

This was discussed with YAEC who stated that a group was main-tained which reviewed all such notices and assessed their impact on the plant. The staff agreed that the licensee could address the concerns 4 about anchor bolt adequacy raised by the IN using such a review.

This remains to be done.

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8. A basis must be provided for use of U-bolts as axial restraints, or the subject supports must be replaced with conventional supports. Several alternatives acceptable to the staff were develcped to resolve this issue: (a) YAEC could use available test results to validate the allowable loads used, (b) YAEC could perform the necessary testing to validate the allowable loads used, and (c) YAEC could replace the subject supports with conventional supports. If one of the alternatives that involve validating the allowable loads is chosen, review by the staff will be required.

4.5 Mechanical Eaufpment In the 1983 IPSAR, two sets of loadings and criteria were defined for the .

evaluation of major mechanical equipment and associated supports. As discussed in Section 4.3, the use of the floor response spectra generated from the NRC site-specific ground response spectrum and criteria and A methods based on the SEP guidelines (NRC spectrum analysis) is acceptable to the staff. As an alternative the staff also accepted analysis using Yankee Composite Spectra and Code Design Allowables (YCS analysis). Review

• of the criteria and methods used by the licensee in the analysis of major mechanical equipment and associated supports established that the work completed by the licensee had not met the requirements specified in the IPSAR. The criteria and methods proposed for YCS analysis were identical O to those allowed for NRC spectrum analysis. Since they correspond to the criteria and methods defined in the SEP guidelines, they are acceptable for NRC spectrum analysis. Being identical to the SEP Guidelines criteria and methods, the triterie proposed for YCS analysis will not assure the SEP guidelitm reqWement$ will be met under NRC spectrum loadings, and are therefore not enc ptabls *oe YCS analysis.

Although YAEC's YCS Aalculations were performed using criteria and methods which do no% assure adquaM margin, it was possible that the designs exhibited sufficient rrtin to a.ake i.$ for the shortcomings in criteria and methodology, used. The *Wf d6cided to salvage what value it could from the YCS calcu-lations, and subsequently rarformed case-by-case reviews of all of them.

Since the NRC spectrum is dout a factor of 2 higher than YCS as discussed in Section 3.1, YCS seismic inertia and anchor motion stresses were doubled to approximate NRC spectrum stresses, and evaluated with the other stresses (weight, pressure, thermal, etc.) using the SEP guidelines criteria. The reviews indicated that the equipment designs are acceptable, and identified the critical components in the population: the pressurizer supports. These components had minimum margin for a support member in bending, minimum margin for a support member under compressive loading, and minimum margin for concrete expansion anchor colts. This made the pressurizer supports an ideal candidate for confirmatory analysis to verify the review findings.

NRC spectrum analyses were performed by both the licensee and the staff's consultants, and the pressurizer supports were found acceptable In both analyses. Details of the confirmatory analysis are contained in Appendix B to Attachment 3

_A d

En Calculation

[4 % 'fd lillllilllllllittilllillllllli Sheet

1. "'waru e r- cate YNPS SEP Piping Pipe Stress Evaluation l system JooNo File No 87150 /F Analysis No Rev No sheet No o A-STRAIN CRITERIA WORKSHEET - CARBON STEEL i 1. Node number:

l l 2. Is the node located at a valve or equipment nozzle or at a threaded connection?

Yu _ No ,

If yes, strain criteria are not applicable at this point.

3. Equation 12 stress calculation 1

Deadweight

f. Pressure NRC Inertia YCS SAM (if applicable)

Total '

4. Equivalent strain calculation ,

l E= 5JZ=5( ) s 0.01 E ( ) 3

5. Fatigue Check cr seismic = s 44 ksi l 6. Stability Check l

Does the application of strain criteria at this point alone, or at this point and at an adjacent location simultaneously, cause any instability in the system?

Yes No l

l

/) -W 1

1

. Calculation llllilllllllllllllllllllllllll I .

""'YNPS SEP Piping # " " 8 ' 0*

5**'"' chute er caie Pipe Stress Evaluation SyUem Jo0 No. "*"*

87150 /r Ansiys.: No Rev No 8""'"

0 A-

7. Displacement calculation: predicted maximum displacement in span near '

strain criteria application point.

Liit Node:

Max. Disp!. = 333 ( )=

.Y:dit Node:

Max. Displ. = 333 ( )=

Z:dit Node:

Max. Displ. =.333 ( )=

N 4

l 4

1

)

l,

'1 A-24

Calculation g(

, i Sheet llilllllllllllllllllllllllllll Project Prepared Br oate

$weiect Cnecaec By Date Pipe Stress Evaluation System J00 No rate No R71%n /r Analysis No Rev. No $heet No STRAIN CRITERIA WORKSHEET - STAINLESS STEEL L Node number:

2. Is the node located at a valve or equipment nozzle or at a threaded connection?

Yes . No If yes, strain criteria are not applicable at this point.

3. Equation 12 stress calculation Deadweight

( Pressure NRC Inertia YCS SAM (if applicable)

Total

4. Equivalent strain calculation i E = ff7g = 6 67 ( 1 s 0.02 '

E ( ) -.

5. Wrinkling Check Is the pipe large bore Sch 407 Yes No If no, go to 6.

E= s 0.4 in = 0.4 ( 1 =

Do - tn ( ).( )

6. Fatigue Check oseismic = s 44 ktl

7. Stability Check Does the application of strain criteria at this polat alone, or at this point and at an adjacent location simultaneously, cause any instability in the system?

Yes No A-27

- Calculation c4 L t 4 Sheet lillllllilllllllllllillllllill ,

Preparec By oste

"' YNPS SEP Piping Ch"uc ey

• • • ' Pipe Stress Evaluation o,i, sysiem  ;* y '"

87150 /F Analysis No Rev. No sheet No.

I

8. Displacement calculation: predicted maximum displacement in span near strain criteria application point.

X <ii.t Node:

Max. DispL = 333 ( )=

.Y.:dit Node:

Max. Displ. = 333 ( )=

7.cdit Node:

( Max. Displ. = 333 ( )=

9. Increased loads at flanged joints, nozzles, and supports in the vicinity of the strain application point (0.01 s E $ 0.02) l Factor = 1 + 170 (E - 0.01) fed 2 Elastic Load Factor Etetored Load l

i i

I l

l l

l l

A -25 l

m .-

.. . ymiPE@ . Calculation  !

did t If Sheet lillllllllllllilllllllllllllli i

PrearM Sy Date 6o"' Y.1PS SEP Piping Sweiect Checked By Cate Pipe Support Evaluation _ ,

System JoDNo File No 87150 /F Analysis No Rev, No, Smeet No.

O B-SRSS OF SEIS5f!C INERTIA AND SAh! LOADS WORKSliELT l t Node numben Support ID.

2.

3. Support Loads Deadweight  !

Dermal & TAhi Inertia SAh! __

4. Support results for the most highly stres:,ed components using absolute sum of inertia and SAh! loads Anchor bolt factor of safety l Structural steel ratio to allowable Weld ratio to allowable l Component ratio to allowable l

l l S. Support results for the most highly stressed components using SRSS combination i of inertia and SAh! loadt 1

l Anchor bolt factor of safety h 4.0 (wedge) or 5.0 (shell) l Structural steel ratio to allowable s LO Weld ratio to allowab!: $ LO i l Component ratio to allowable s LO

6. Results for the most highly stressed components for adjacent supports

a. Support ID-Anchor bolt factor of safety _ 2 4.0 (wedge) or 5.0 'shell) l Structural steel ratio to allowable $ 1.0 ,

Wcld ratio to allowable __ $ !.0 l Component ratio to allowable - __ s LO g_g 1

l

Calculation 4L i i Sheet lillllilllllllllllil!!!!!!!Ill P"P8'u ey caie l '#YNPS SEP Piping i sweiect Checked 8y cate l Pipe Supriort Eveluation syltem JoD No Fite No 87150 /P Analysis No Rev. No. sheet No.

O s-

b. Support ID Anchor bolt factor of safety 2 4.0 (wedge) or 5.0 (shell) i Structural steel ratio to allowable $ LO Weld ratio to allowable $ 1.0 Component ratio to allowable s LO

7. Evaluation of surrounding area l

a. Have strain criteria been applied at or adjacent to this support?

Yes No

b. Does the piping on the spans on either side of the support have flanged or threaded connections or any other features which may exhibit nonductile ,

behavior?

Yu No l

c. Will the adjacent supports exhibit ductile behavior?

Yes No

d. Is the piping capable of isolation upstream of this support?

Yes No .

c. Does the piping and support layout represent good design practice?

Yes No

8. Justification of appropriateness of using the SRSS combination of inertia and SAM i loadr l

l l

l l-) - fj) l

1 1

l

. PIPING ANALYSIS IN PROGRESS l

l I. 1988 SCOPE j I

Total scope of piping inside the Vapor Container to be evaluated for I seismic loads is listed in Yankee letter to NRC, dated June 1, 1987 Approximately 1/2 of the total scope will be evaluated in 1988. Support modifications as a result of those evaluations will be installed in the November, 1988 outage. The remainder of the evaluations will be performed in 1989 with any required modifications install 3d in the 1990 refueling outage.

The piping systems to be evaluated / modified in 1988 are: , b

1. Main Steam Piping (SCPB, DSSS)/ " l 2.

3 Feedwater piping (SCPB)

%Ab ([

Shutdown Cooling Piping (PCPB),.f_ ,

g t

Pk d I

4. Steam Generator Blowdown and Eme fency Boiler Feed (DSSS SCPB L j.I l S.

6.

Main Coolant and Pressurizer Drr. ins (PCPB)- @ jg~b/g9 ,

I Pressurizer Sample and Vent (PCPB) 6 4) g ed,%.

ghi :

^ -

l 7 Charg2ng thru the Drain Box (DSSS,'PCPB 4 c.f/ *" Q Leen II. PIPING DESCRIPTIONS

1. Main Steam piping - 14 in, sch. 80, carbon steel

- 4 loops, from Steam Generator to Vapor Container (VC) penetration

2. Feedwater - 8", sch. 80, carbon steel d )

- from VC penetration to Steam Generatdors (4 loops) 3 Shutdown Cooling - 6", sch.160 and sch 40, stainless steel

- from VC penetration to tap into Loop 4 hot leg; (Prob. 121) and from VC penetration to tap into Loop 4 cold leg (Prob. 122).

l

, 4. Steam Generator Blow-down and Emergency l

. l Boiler Feed - 2' , sch. 80, carbon steel

{

- Blowdown (4 loops) - from Steam Generators '

to VC penetration

- EBF (4 loops) - from Feedwater piping to Blowdown tie-in 5 Main Coolant and Pressurizer Drains - 1" (Pressurizer Drain), 11" (Main Coolant Drains), and 2" (Drain Header), sch. 80 and ,

160 stainless steel l

- from Hot and Cold legs (4 loops) and Press-urizer to 2" Drain Header to VC penetraion l

A -1/ l l

1

4

6. Prersurizer Sample &

Vent - 3/4", sch. 80, stainless steel

- from Capillary Tubes to Vent Header to VC penetration 7 Charging thru Drain Box

- 3/4" sch. 80 and 2" sch. 160, stainless i steel

- charging from VC penetration to HX-E-8-1, Safety Injection drain, HX drain, charging l drain, and lower HX cross connect piping.  ;

III. SCHEDULE FOR 1988 EVALUATIONS / MODIFICATIONS i l

Piping evaluations and designs for support modifications are scheduled j to be completed by June 1, 1988. '

1 Support modifications will be installed during the 1988 refueling outage, which will start in mid-November.

i i

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C/lf/86 PAGE $ l

. 1 I

Yale (EE II1CLIAR PorER STATION - SUMARY OF SEP TOPIC 111-6 WC 00EST!DNS  ! l DUES 110N RESOLUTION  ! l ITEM NO. ISSUE SUMARY SOURC(o RISOLUT!DN REFEREuCES SCHECULE

• j i

816 For building-founded tants, E666 Case by case revies. Criteria . 9/23/86 l Justify use of AISC and ACI Docueent E -1, l

l criteria proposed in Section Rev. 3. i l 5.4.2.2. Shos comparison

• l l sith SEP guidelines, j j i

C STRUCTURES ,

l 1 l

Cl VAPOR CONTAINER (VC)  !  !

Cla Provide calculations for 18CT(1), Calculate the required stresses. 2 Resolved. l secar stresses in coluans MC and tie besas.

t Clb Provide calculations for NCI(5), Evaluatethesubjectbolts. 2 Resolved. l  !

pull-out of coluen base NRC anchor belts.

Cic See ites Il above, nCT U), Sase as ites il dove. -

8/15/86 '

NRC l

l <

Cid P.5/lFD pipe anchor leads NCT(3) Re-evaluation of YC penetrations 5 . '

used to evalify VC penet- sill be perforced when the MS/Fu l

i rations ser no longer be analysis is cosoleted. l valid because of recent I reanalysis of those lines. I l

Cle Provide the evaluation NCfH) Perfore the sdject evaluation. 2 Resolved. j results for the clevises l

and turnbuckles on the I diagonal tir rods. I Clf Evaluatetheadequacyof NCil6) Perfore the subject evaluations. 2 Resolved. m I' l '

the clearance netsten the l VC and the RSS, and the VC and the radioactive pipetunnel.

Cig Generate ARS under kPC NC10) Ni!! be resolved via Cid. -

Resolved.

spectrue loads.

j i

1 4.p/

' 06/18/86 PAGE 41 YMEE FJCLEAR POWER STATION - SUMARr 0F SEP 10Pic !!!-6 NRC 90EST!0NS GUESTION ITEM E0. RESOLUT!0W ISSVE SUMARY SOURCte RES0Wi!04 REFERENCES SCHEDAE e Muserals in parentheses following NCT correspond to fostnote numbers la Table 1 of the N:7 Revie Repert. A 10) appears for iteet such as 3C2, C4f, and '

C5e because ther do not appear in Table i but the issues are raised in the te:t.

Nueerals in parantheses fo!!ceing E6t3 corresped to E616 question numbers. ,

4 l

RITERINCES:

1. Presentation to hRC, docusent dated February 24. 1996.

2. Presentation to HEC, docusent dated A;ral 8,1936. )

3. Presentatien to N*r.C, docusent dated May 19, 1986.

4. Presentation to ARC, docueent dated June 24,1986.

5. Presentation to N5C, docusent dated August 6,1986.

5 5

I I

I I .

I I

l /) - b-

e Calculation j)6,

=

[*hP. J in Sheet #p4 lillllilllillifillllllllllill! [

Project g Prepared By. Date gg

~

Subject Checked By' Oate System Job No. File No.

Analysis No. Rev. No, Sheet No.

COMPARI GON OF YCS SAM OF- VC g YC 6 SAM ("')

~

elf::\/ATIOd plRECfl0M otp (f) l NE W (2)

N-S 0.47 S45 t '

g- u/ 0.43 003 ELsVs Io'7I'-O" Vggnat, o, o 1I 'O.007 "

ELEV. l140'- 6 (CLD HDOGL)

~

1. ' #' W E-W 0.47 0.01 GL&. ll52'- O' (%) 40DE L.) qqqq;(gt, 9, p ; y, n,9 g ; y

( 0. 4 m yog d-S 0.9 @

G-W 0. + l 0 79 V6RTICAL 0. 01 2. 0,0f3 j,d )

't El.nl. to l3'.9 (ctD HDDil) N'S 9O 4 03 ,;,, '

ELsv. tot 4'-4." (hjE+l H0 DEL) '

V6R'ficAL O.Oli 0,0Il BOTTO4 OF VC, N-6 0.46 0,63 F'W #' A I 0' O '

ELEV 10 4 /- o * -V6RTicAL O.0/0 ,, 0 p f a_

NoT66 :

(t) Old model h6S $$ra(Juyc] domping p{ 5%, SAM was ca/cula/ed LtsIng inCdal Sqeffosidien me fh 0d. References : (I C\.) " Vapor lon$$iner W RCNYC i YAf.S , S]rutAttral Ana sis feprt ", Rev.8 , l'd W >

Arcii IA84 . (I.b.) O g n a. C d cul s.h on 90023/4/p . %c syn 31ubc.

he hulories )

( (2) Neu model has &mcAurcJ ktmping of 3%, SAM NOS u k<dabed Ming direct integraHon 4 e - his+0rg we{kod. k'eferences t (1.c< ) Cyg na. la)cd^ 40n 97 I SC /* l F , Set E . ( model y Sue - his10ry analpis) (2,b) (git ( (dc. 97150/1/F, Se.A C . UpkHc fime- hidorW)

/)--yf.

1 Calculation jM t 5

[.4 D h'fd Sheet ehid i llillllllllillllllllilllllllll b g)!

Protect 4gp pjg Prepared By Date p Suciect Checked By Date System JoD No File No Analysis No. Rev. No. Sheet No CDMfAKl40td 0F YCs GAH OF RSS 1

YC G 6A M Cin ) l OLO AllAL. (') HLu AuAl.. (2)

ELW.toO7 L b(otD M AL.) (#- 5) 0.745" 1.1 x HOR 1204Mt- )

i i G-w) 0745" =

0.%" l ELW.10 79 - 6. (EN M AL. ) l (VERT) 0.0341 (ygeT ) 0,07" I l

'b ) 'b "

El.6v. I127L2" (CLD ANAL. ) l'l X h8D28dTAb O'

(E-W) o. M " e. I, O

('

ELEV. u 10 '- 7 (CE4 AMAL. ) (V6ET) 0.0 3#  %) o, O "

i WOTcs :

(o Tor #e cid snstpls , de ciructural ds.,qing is 7% ,

The sAas aece. uJudated asing modnl saperposikon me Mod l da h>o kovibonfal and cee verHeal 9aand Hee-hisicvies t>ppheA sowlktuac)y. Referentes :

(1. A)

• kdoy fivppoyt (fruc}vre , YdfS, !?lrucktrAl /ndysis )1port',,

~

hv. 3, Qt s , March 1993 ( hr RSS inodel.) l (I.b.) n MS/ SAM Hagpong 4r gglp5 ", Doc. No. HS-I, fu;f , W. ML7. (Ar%H) .

O c) hya G.LaJakon eco23/4/p, (for Sy,%gc_ ne-Hiskries )

(1) Tor k. wr udyds, Du shuclaral daryhg is 57. 71u S/\Ms utte (AlWahA sq Cred* integrafion me kod wHk I.I v horQordal a-A verRC4I N - Wsforg opphed heuw.ous/g . Tk u*r ar ALpis wA %. sm Rs;s linc.r nodd c.c ra. cid s w s y is.

Refuenus : (2.N Gyps GA. 97 iso /I/F, C<A D.((or SAM Olsdnksn)

(2.b) Gg p Gk. 01 so/ /r1 sea c . (hr Cynkuta 7-Hs)

==

g.p

9 f' {

Y' /,,d l

YAEC GANG SUPPORT LOAD GENERATION )

o SELECTION OF MODELS 2" Piping j l

o Similarity 1 o Branch Connections o Mass of Decoupled Branch Connections 3" Piping j i

j o Similarity l o Mass of System o Unsupportet. Length o MCP Connection Elevation i l

i o MODELING l 1

o Program ADLPIPE o DC-1 Design Criteria b

'~

o ANALYSIE yW [L' t .

o Leading Conditions ,

/

je.r* g . l i

M

- Peessure (Design Press. =l2fgi) f

- Deadweight )

- Thermal (Design Temp = 150 deg. F)

- Thermal Anchor Movements

- Seismic

- Seismic Anchor Movements o Loading Combinations o Gang Support Load Combinations l .

n o l

_ _ _ . - . . . - - , , - .~

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( -.- . 'S . e N ueoATr - @) 17 '/2 j f A  ; i GEU. UPDATE P y/Jn l m

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TIO 4212 ORGNAL coca 34-3 3 7de,or, 155,0E too a .v. cescazerios av cxxo A'eg BILL OF MATERIALS ITEM QTY MAT'L' DESCRIPTION I 4 A-36 R. '/4.* x 2.* x 0 - (o 8/ 2 " LG ,

2 7 A-36 R. I/4*w2"x0' T

• LG 4 i C.S. 1/2" ROO 5 I C.S. GRIff311 FIG. 290, 1/2" OIA 4ELDLESS EYE PAJT , OR EQUtv4 LENT 6 i C.S. GRIffELL FIG 212, 2" OIA PIPE CLALP , OP EQUtVALENT 7 Z i% ZH 8/Z' 9 HEX NUT S i

YANKEE NUCLEAR POWER STATION DWC7. NO. %')9 -MP-12Co AM- 4 A . 4//7

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