ML20149J218
| ML20149J218 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 06/16/1997 |
| From: | Diane Jackson NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9707280120 | |
| Download: ML20149J218 (56) | |
Text
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t NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20665-0001 June 16, 1997
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APPLICANT: Westinghouse Electric Corporation FACILITY: AP600 ,
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS WESTINGHOUSE AP600 PIPING DESIGN ANALYSIS The subject meeting was held between the Nuclear Regulatory Commission (NRC) and Westinghouse Electric Corporation, the applicant, at the Westinghouse office in Monroeville, Pennsylvania, on April 16 through 18, 1997. The main purpose of this followup meeting and audit on the " leak-before-break" (LBB) application to piping was to discuss and resolve the remaining related issues identified in the NRC AP600 Draft Safety Evaluation Report (DSER). As a result of this meeting, significant progress was made in resolving the open items. However, in order to close these items, Westinghouse must provide revisions to the standard safety analysis report (SSAR) and a letter response to document some technical bases and resolutions with greater detail.
Attachment 1 is a list of meeting par'ticipants for the piping design review meeting. Attachment 2 is a summary of the issues reviewed at this meeting and draft SSAR markups provided by Westinghouse at the meeting. Attachment 3 is a l summary of the open items, up to SSAR Revision 11, discussed at the meeting. '
Attachment 4 is the input and results of independent verification on main 3 steam LBB bounding analysis. Attachment 5 is the Westinghouse presentation l on computer code verification. Attachment 6 is the summary of sample LBB ;
calcui;tions.
On April 17, 1997, Westinghouse presented a proposal for design acceptance criteria and inspection, testing, analysis, and acceptance criteria for piping. Attachment 7 is the list of meeting participants and Westinghouse handouts for the presentation.
You have requested that portions of the information submitted in the June 1992, application for design certification be exempt from mandatory public disclosure. While the staff has not completed its review of your i request in accordance with the requirements of 10 CFR 2.790, that portion of i the submitted information is being withheld from public disclosure pending the staff's final determination. The staff concludes that these questions and comments do not contain those portions of the information for which exemption is sought. However, the staff will withhold this meeting summary from public disclosure for 30 calendar days from the date of this meeting summary to allow Westinghouse the opportunity to verify the staff's conclusions. If, after that time, you do not request that all or portions of the information in the t attachments be withheld from public disclosure in accordance with 10 CFR '
! \
2.790, this meeting summary will be placed in the NRC Public Docum
~
[lR'*nMA M85Hga NRC FILE CENTER COPY
.. 4 l
June 16, 1997 ;
This meeting summary was prepared with input from the NRC consultants. If you ;
have any comments, please contact me at (301) 415-8548. l original signed by:
Diane T. Jackson, Project Manager l Standardization Project Directorate !
Division of Reactor Program Management Office Of Nuclear Reactor Regulation Docket No.52-003 l Attachments: As stated cc w/ attachments: See next page DISTRIBUTION _w/ attachments:
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SCollins/FMiraglia, 0-12 G18 TMartin, 0-12 G18 RZimmerman, 0-12 G18 MSlosson SWeiss TQuay WDean, 0-5 E23 ACRS (11) JMoore, 0-15 B18 DOCUMENT NAME: A:ECG4 16. MIN To receive e copy of this document, Indicate in the box: 'C' = Copy without ettechment/ enclosure "E" = Copy with attechment/ enclosure *N" = No copy 0FFICE PM:PDST:DRPM l, D:PDST:DRPM l l l NAME DTJackson:sg W TRQuay T4 DATE 06/(%/97 %J 06/ft,/97 0FFICIAL RECORD COPY
.~ . _ . . . . . _ _ - . - - . - -
}
, . .o June 16, 1997 4
This meeting summary was prepared with input from the NRC consultants. If you have any comments, please contact me at (301) 415-8548.
Q Diane T. Jacksen, Project Manager Standardization Project Directorate Division of Reactor Program Management Office Of Nuclear Reactor Regulation Docket No.52-003 ,
Attachments: As stated cc w/ attachments: See next page !
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Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Ms. Cindy L. Haag Advanced Plant Safety & Licensing Advanced Plant Safety & Licensing Westinghouse Electric Corporation Westinghouse Electric Corporation Energy Systems Business Unit Energy Systems Business Unit P.O. Box 355 Box 355 l Pittsburgh, PA 15230 Pittsburgh, PA 15230 1
Mr. S. M. Modro I Nuclear Systems Analysis Technologies Lockheed Idaho Technologies Company Post Office Box 1625 Idaho Falls, ID 83415 l
Enclosure to be distributed to the following addressees after the result of the proprietary evaluation is received from Westinghouse:
Mr. Ronald Simat Director Ms. Lynn Connor i Advanced Reactor Programs DOC-Search Associates Nuclear Energy Institute Post Office Box 34 1776 Eye Street, N.W. Cabin John, MD 20818 '
Suite 300 Washington, DC 20006-3706 Mr. Robert H. Buchholz GE Nuclear Energy
- Mr. James E. Quinn, Projects Manager 175 Curtner Avenue, MC-781 LMR and SBWR Programs San Jose, CA 95125 GE Nuclear Energy 175 Curtner Avenue, M/C 165 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq. 19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20874 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Charles Thompson, Nuclear Engineer AP600 Certification Mr. Frank A. Ross NE-50 U.S. Department of Energy, NE-42 19901 Germantown Road Office of LWR Safety and Technology Germantown, MD 20874 19901 Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design-Certification Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303
. . s NRC/ WESTINGHOUSE AP600 PIPING DESIGN MEETING APRIL-16 THROUGH 18, 1997 MEETING PARTICIPANTS NRiE ORGANIZATION Goutam Bagchi NRC/NRR/DE/ECGB Shou-nien Hou NRC/NRR/DE/ECGB Diane Jackson NRC/NRR/DRPM/PDST Matthew Mitchell (P/T) NRC/NRR/DE/EMCB Gulliano DeGrassi BNL (NRC Consultant)
Ron Gamble Sartrex (NRC Consultant)
Kevin Accornero Westinghouse-SMT Dulal Bhowmick Westinghouse-SMT Ed Cummins (E) Westinghouse Rolu Hundal CDI (Wesitnghouse Consultant)
Ed Johnson CDI (Westinghouse Consultant)
Don Lindgren ' Westinghouse-Licensing Moshe Mahlab Westinghouse Rao Mandava Westinghouse-Plant Engineer Seth Swamy Westinghouse Chunag-yeh Yaing Westinghouse-SMT (P/T) - Part-time participant (E) - Exit meeting participant i
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Attachment 1
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-MEETING TRIP REPORT j 1
1-
1.0 INTRODUCTION
I; On' April 16 through 18, 1997, the NRC staff and its consultants from the Brookhaven National Laboratory and the Scientech,_ Inc. conducted a meeting and audit on the " leak-before-break",(LBB) application to piping in AP600 standard plant at the Westinghouse's (the applicant) office in Monroeville, Pennsylvania. The purpose of the meeting was to resolve 1
- remaining open items and to perform audits on LBB calculations. This I report presents a summary of the meeting discussions, audit results, and followup actions agreed by the applicant for closing these items.
i The LBB-related open items were described in detail in Section 3.6.3,
" Leak-Before-Break Evaluation Procedure," of the NRC _ Draft Safety
, Evaluation ~ Report (DSER), which was issued by the staff ~in November
,L 1994. Since issuance of the DSER, there have been several additional i meetings-and correspondence between the staff-and the applicant to l resolve the issues, the applicant made a number of revisions to the
- SSAR to address the staff's concerns. Attachment 3 summarizes the L status of the LBB Open Items following the issuance of Revision'll of i-the SSAR and prior to this meeting. Items remaining open were the j subject of this meeting.
2.0 MEETING SUMMAR'Y As results of this meeting, the following is a summary of the discus-
- sions, findings, and agreements reached in each area.
L 2.1 Boundina Analysis Procedures (OITS# 608)
Westinghouse explained their procedure for developing the LBB bounding
! analysis curves given in SSAR Appendix 38. The curves are based on a minimum of two points. One point corresponds to a low normal stress and i the other point to a high normal stress. The first point includes the
- axial force for normal pressure combined with ne bending moment
- corresponding to a very low bending stress. These loads are used to determine the corresponding leakage flaw size at normal 100-percent
- power condition for 10 times the leak detection capability (5 gpm). A d
stability analysis is performed to determine the maximum additional 1 moment for a critical flaw size of twice the leakage flaw size. This defines the maximum stress corresponding to the low normal stress
, condition. The second point on the curve includes the same axial . force l for normal pressure combined with a high bending moment corresponding to J a higher than expected bending stress. The'same calculational procedure is followed to define the leakage flaw size at the high normal stress condition and the maximum stress for a critical flaw size of twice the leakage flaw size. The bounding analysis curve is generated by plotting these two points on a normal versus maximum stress plot and joining the two points by a straight line. These bounding curves are generated Attachment 2
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o.
for each piping system-to be qualified for LBB. More than one bounding
- curve may_be needed for each system with consideration given to pipe size, pipe schedule, operating temperature, and operating pressure. The calculations- are based on minimum wall thickness at the weld counterbore, minimum yield and ultimate strengths of piping material' from the ASME Code and modulus of elasticity from the ASME Code.
In the meeting discussions, the audit team expressed the following concerns:
- 1. The use of minimum pipe wall thickness' and material properties is conservative for the stability analysis at maximum stress but not for defining the detectable flaw size at normal stress._ It was suggested that maximum thickness and material properties' be used to
. define the leakage flaw' size while minimum thickness and material properties be used in the stability analysis. The applicant felt that this was too conservative and stated that minimum thickness and material properties had consistently been used in the past and their effects on bounding curves were shown to be insignificant.
The applicant agreed to either perform additional calculations or provide a SSAR revision to confirm this, based on previous study resul ts'.
- 2. The calculations assume that axial force ~ is due to normal pressure alone and that axial force does not change between normal and ,
maximum stress conditions. The audit team questioned whether this. '
is a valid assumption. The applicant indicated that.their experi-ence has shown that pressure produces the dominant axial stress.
The applicant agreed to provide a SSAR revision to verify that the axial load due to pressure at full power operation is the dominant load in producing the axial stress.
- 3. The beunding curves are based on a straight line between two points.
The audit team questioned why a straight line is acceptable. The applicant explained that for many cases, additional intermediate points were determined and those points fell above the straight line ;
curve. Therefore a straight line is conservative. The explanation appears reasonable and acceptable.
- 4. Draft standard review plan (SRP) Section 3.6.3 requires the maximum stress load combination for-demonstrating crack stability to include
~
pressure, deadweight, thermal and SSE loads. In Section 3B.3.3 of the SSAR, the applicant includes this combination'plus two other load combinations, one of which includes valve thrust in place of SSE and the other which includes thermal maximum in place of thermal normal and SSE. The audit team questioned why these additional load combinations were included. In addition, if valve thrust must be considered, it seems'that this type of force may significantly
- l
.. , 1 increase axial force. If this is the case, the bounding curves which assume constant axial. force only induced by pressure as
- discussed above may not be valid.
. The applicant. agreed to reconsider the load combinations and revise the SSAR if necessary or provide additional justification.
2.2 Restraint of Pressure-Induced Bendina on LBB (0ITS# 608) I Recent work published by Battelle in NUREG/CR-6443 discussed the potential effects of restraint to pressure-induced bending in a piping
, system on LBB calculations. This restraint effect may increase both the )
- leakage crack length and the critical crack length. In their study, j i - Battelle performed calculations to determine net effect of restraint
- on a 28-inch pipe and a 4-inch pipe. IL showed that the~ net
, effect was negligible for the large pipe but % very significant for
' the small pipe. Thus for small pipe applications, the LBB methodology which generally does not consider this effect may be unacceptable for cracks near restrained. locations such as nozzles.
The. applicant stated that the studies were not directly applicable to AP600 because they were based on thinner pipes typical of boiling water l reactors (BWRs)~rather than the thicker pipes in pressurized water reactors (PWRs). On the other hand, the pressures were more representa-tive of the higher pressures in PWRs. The leak' rate flaw sizes in the study were calculated based on crack surface roughness of intergranular stress corrosion cracking associated with BWRs which is not appropriate ,
for PWRs. The applicant argued that because of the differences, the net effect on LBB calculations for AP600 may be small or even positive.
However, to support their position, the applicant agreed to perform a detailed three-dimensional finite element analysis of a six-inch pipe
~
(the smallest diameter LBB pipe in AP600 plant) to investigate this effect. ~The intent was to assess the most limiting case in AP600 LBB applications for a summary disposition of the staff concern. They proposed a model of a section of pipe and nozzle with a circumferential crack at a distance of one pipe diameter from the nozzle. The~ nozzle end would be fixed-and the applied loads would correspond to the high normal end of the bounding curve. The audit team agreed with this :
approach but questioned whether the appropriate load was being applied.
This was discussed in detail. The applicant felt that the high loads were more critical. The audit team pointed out that this was not clear from the Battelle results and suggested that a range of loads be considered. It was finally agreed that the analysis would be performed using a single load corresponding to the actual design load for the line.
The results of the analysis were made available at the end of the meeting. For an assumed 3.2-inch.long circumferential crack, a leakage rate of 6.25 gpm was calculated. Thus the margin of 10 on leak rates was satisfied. Crack stability analysis was performed for a crack e, e , , . - - - ,..-.-.w ,.
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6.6. inches long (greater than twice the leakage crack size) and the crack was shown to be stable. Based on these results, the bounding curve was shown conservative when restraint effects on both leak rate I and load carrying capacity were considered.
Thus the issue of restraint effects on pressure-induced bending to AP600 i LBB analyses was. technically resolved, and its closure is pending the i following confirmatory actions by the applicant: 1 (1) Provide a request for additional information (RAI) response to the staff describingL the analysis and conclusions as presented during the meeting. !
(2) Provide- the calculated value of the applied J-Integral .for the flaw with restrained bending, and confirmation' that the' material J-R curve used in the applicant's analysis is consistent with the data obtained by Battelle in the NRC Degraded Piping and the IPIRG..
Programs.
2.3 Weld Inspection Reauirements (0ITS# 611) l The applicant described the fabrication and in-service inspection requirements for Class 2 and 3 LBB lines. The primary staff concern was that one of the LBB lines, the Accumulator Discharge Line which is part of the emergency core cooling system (ECCS) is classified as an ASME Code' Class 3 line. One of significant differences between Class 3 piping and Class ?. piping is in the radiographic inspection requirements for fabrication. Class 2 piping requires radiographic examination of the welds and Class 3 piping does not. In the SSAR, the applicant states that for Class 3 lines required for emergency core cooling functions, radiography will be conducted on a' random sample of welds.
The audit team asked for full radiographic inapection of all welds in these lines if LBB is applied.
The applicant agreed to incorporate this requirement into the SSAR.
2.4. Verification of Computer Codes (0ITS# 618) !
The applicant gave a' presentation on computer code verification. A copy of their presentation material is included in Attachment.5. The audit -
. team' reviewed the procedure used to benchmark the leak rate computer- >
code. . The procedure was described as including laboratory data from
-Battelle. leak rate experiments, and in-service leak rate data from the Duane Arnold Nuclear Plant. These are the same data previously used to benchmark other leak rate codes, such as the EPRI developed software PICEP and the NRC developed software SQUIRT. The NRC previously reviewed the benchmark procedure for the Westinghouse leak rate software and found it acceptable. Based on the information provided by the applicant and reported'by the NRC in the South Texas SAR (NUREG-0781, Supplement #4, July 1987), it is concluded that the software has been
. o adequately benchmarked. The applicant also stated that their methods -
~have not changed since their codes were developed or since the South Texas methodology was. approved by the staff. This issue is technically resolved.
The applicant agreed to revise the SSAR to describe the benchmarking of' the LBB computer code.-
The applicant also presented some possible reasons for the differences between their results and the results obtained by the staff using the PICEP code. There are differences in the way the codes calculate section modulus and a correction factor-needs to.be applied to account for this difference. Other differences may be in the assumed shape and smoothness of the postulated crack surface. The explanation appears reasonable.
2.5 Water Hammer loads.-(0ITS# 620)
The staff had-a concern over preliminary results from small-break loss-of-coolant-accident (LOCA) tests performed at Oregon State University which indicated that rapid condensation events have the potential to cause unanticipated dynamic-loads in the AP600 reactor coolant system (RCS). The applicant was requested to address whether these water hammer type loads should be considered in the LBB analysis. The applicant indicated that they had reviewed and evaluated the final results of the tests and concluded that the condensation events had resulted in a lot of noise but no damage. In order to measure the effects, faster instrumentation was installed which indicated that the pressure peaks'from these events were very low (only a few psi) and would not produce high stresses in the piping.
The applicant also discussed the water hammer screening criteria proposed by Professor Griffith of MIT. The criteria are divided into two basic categories which include configuration and thermohydraulics.
Pipe configuration is first reviewed to determine slope and length of pipe. If the pipe has a slope less than 0.5 inches /ft and its length exceeds 24 pipe diameters, it meets the configuration criteria for possible steam bubble collapse / water hammer potential and must be evaluated against the thermohydraulic criteria. If the Froude number is less than one, the subcooled temperature is greater than 38*F,. and the pressure is greater than 150 psia, water hammer may occur in that line.
The applicant reviewed all of the proposed LBB lines against these criteria and concluded that steam bubble collapse / water hammer will not occur in any of the lines.
The audit team found the information provided by the applicant accept-able but asked for a documentation to summarize their evaluations. The
appl::"' agreed to include the screening criteria in the SSAR and will provide a detailed discussion of both the Oregon test evaluations and the water hammer screeaing reviews as part of their RAI response.
2.6 Failure Mechanisms for Main Steam Pipino (OITS# 620)
During. discussions, the audit team raised a quest.iun regarding potential for erosion-corrosion (EC);in the main steam piping. The information provided in Section 38.2.1 of the SSAR on EC degradation was reviewed.
Based on this review and service experience of operating PWRs, the audit team concluded that additional information was needed to assess the potential for EC of main steam line. The team suggested that EC screening criteria similar to those used in operating plants be used to assess the potential for EC. The applicant subsequently indicated that in addition to the reasons given'in SSAR Section 32.2.1, the AP600 main steam line has a low. flow rate' (less than 150 ft/sec at- full power) which is less than currently used in industry EC screening criteria.
Based on this information, the team concluded that there is a low potential for EC in the main steam piping. The low potential for EC. is consistent with the guidelines in SRP 3.6.3 for application of LBB.
The applicant agreed to revise the SSAR to include this additional information.
?. 6 Audit of LBB Calculations (0ITS#.608)
In order..to verify conformance with bounding analysis procedures and LBB acceptance criteria, a sample of five LBB calculations were reviewed.
The LBB calculations are included in the piping analysis reports.- The applicant provided the following piping analysis reports for audit:
e Main Steam Line (32-inch pipe): SGS-PLR-030 Rev. A, May 1996, by UTE/-Initec e Pressurizer Surge Line (18-inch pipe): RCS-PLR-040 Rev. 8, July 8 1994, by Ansaldo e PRHR Return Line (10-inch pipe): PXS-PLR-040 Rev 1 (Draft),
April 15, 1997, by Westinghouse e CMT-26 Supply Line (8-inch pipe): PXR-PLR-060 Rev F, June 1995, by UTE/Initec e Direct Vessel Injection Line A (6-inch line): PXS-PLR-010 Rev A, April 4, 1996, by Ansaldo The selected analyses covered a range of pipe sizes and materials.
The calculations were performed by the applicant, Ansaldo, and UTE/Initec as noted above. Each calculation package was~ reviewed by the audit team. The piping analysis reports contained a section
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[ t i
I which sunnarized the_LBB' stress calculations and comparisons _to
! bounding curves. The pipe stresses were calculated from the axial
! loads and . bending moments which were taken from the. results of the i appropriate piping analysis load cases. The algebraic sum method
- was used to determine the normal stress from the pressure, dead-
! weight, and thermal load case results. The absolute summation method was used to calculate the maximum stresses from the required load cases. A post processor program was used to read the PS+CAEPIPE piping analysis results and calculate the LBB stresses in i accordance with the required load combinations. The maximum stress point and its corresponding normal stress was then compared against '
the appropriate bounding analysis curve. A summary of these LBB evaluations for the five selected piping systems is included in Attachment 6. Two curves are shown for each piping system corre-sponding to different maximum stress load combinations. One is for the normal plus SSE load combination and the other is for the normal plus maximum thermal load combination. It was ncted that for the PRHR Return Line, two normal stress points were checked. They correspond to normal stratified and unstratified conditions. '
i The audit team reviewed the LBB calculations and found them to be in I conformance with the bounding analysis procedure. The piping -
isometrics were reviewed to identify the maximum stress point locations which appeared to be reasonable. The lines which were '
audited met the LBB acceptance criteria. The audit team questioned
/ -the verification of the post processor program used in the evalua-tions. The applicant stated that.the post processor is a small program which does not require formal verification. The results-of the analysis are contained in a formal checked calculation and verification of the post processor results is the responsibility of the checker. To satisfy this concern, the applicant performed their own additional verification of the high stress points. The audit team also randomly spot checked the calculated stresses at some locations and did not find any errors. ,
The audit team checked the calculations to verify that SSE loads considered all applicable soil conditions. All of the selected calculations were performed using the enveloped response spectrum analysis method. From the piping calculation references, it was determined that the response spectra used in the piping analyses were taken from Bechtel report GW-S2R-002 which was made available for review. The spectra represent an envelop of the four soil design cases and includes 15 percent broadening.
l The audit team also verified that pipe breaks at branch pipe connec-tions 2re postulated unless LBB is applicable to the branch line. i The audit team reviewed calculation GW-POC-004 Rev 0, " Break Loca-tions for High Energy Pipe Break Hazard Evaluation," dated March 21, 4 1997, by Westinghouse. Four terminal end pipe breaks were i
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i . l i I
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l postulated at LBB line branch connections. They include two four- !
-inch spray .line connections to the cold leg and two four-inch t automatic depressurization system (ADS) stage 1. piping connections j l to the 14-inch ADS stage 2/3.line.
There'were'some additional discussions on the completion status of !
the calculations. The audit team noted that the calculations were identified as. preliminary and had letter revisions versus number revisions. In addition, the pipe stress analyses did not include Class 1 fatigue evaluations nor did they consider all loads such as valve opening and closing loads. The applicant stated that the calculations are considered preliminary because design data for the valves, equipment, modules and pipe supports was not final. The applicant believes that the analysis performed is sufficient for .
design certification and that fatigue evaluations and analysis of '
other load cases will be completed in the final design stage.
However, since this did not have a direct impact on the LBB review, the applicant suggested that further discussions on this subject be deferred to a later piping. design review meeting on inspection, testing, analysis, and acceptance' criteria.
2.7 Independent Verification of leak Rate and Boundina Analysis (OITS# 618)
During-the audit, the applicant was asked to provide information that would allow independent calculation by the team to confirm the calcula-tions by the applicant on leak rc a and bounding curves. The piping selected was the 32-inch main steam, which is ferritic steel piping.
This request included identifying the type of weld rod used for welding this ferritic piping. .The' applicant reported that the weld rod is 7018. .
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The confirmatory analysis used the methodology described in NUREG/
CR-6281. Based on the use of weld rod 7081, the team employed the factors listed in Table 2-4 in NUREG/CR-6281. A summary of the input and results is enclosed as Attachment 4.
Results from the confirmatory analysis indicated that the team calculated maximum allowable axial plus bending loads are greater than the loads determined ey the applicant for the bounding analysis of the 32-inch main steam piping. Based on these results, the confirmatory analysis verified that the bounding analysis performed by the applicant is conservative and acceptable.
3.0 CONCLUSION
S l
As discussed above, significant progress was made in resolving the remaining LBB open items. Items that remained open after the meeting were technically resolved and agreements on the following Westinghouse actions to close these items were reached during the meeting: l l
l
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. o j e For OITS# 608:
- a. The applicant will provide a SSAR revision to address the i effects of using minimum wall thickness, minimum material properties, and axial load in LBB bounding analyses. I
- b. The applicant will provide a SSAR revision to clarify load I combinations for calculating maximum stress in the LBB bounding analyses. j
- c. The applicant will provide a RAI response describing procedures and results of a sensitivity study conducted by the applicant on a six-inch pipe, the calculated value of applied J-Integral, and the material J-R curve used in the sensitivity study, l
e For 0ITS# 611:
- a. The applicant will provide a SSAR revision to incorporate augmented weld inspection requirements for the ASME Code Class 3 LBB lines.
e For 0ITS# 618:
- a. The applicant will provide a SSAR revision describing the appli-cant efforts on LBB computer code verification.
-e For 0ITS# 620:
i
- a. The applics.nt will provide a RAI response with a detailed l description of the applicant's efforts-and conclusions on the !
water hammer screening criteria reviews and Oregon State Univer-sity rapid condensation loads evaluation. ;
- b. The applicant will provide a SSAR revision providing additional information on potential erosion-corrosion in AP600 main steam line.
,3. Desigi cf Structures, Ccmponents, Equipment, end Systems introduction of a large volume of cold water sufficient to result in a bubble collapse water hammer.
He design and component selection of reactor coolant branch lines and other lines evaluated i
for mechanistic pipe break follow design guidelines intended to minimize the potential for I
water hammer. Comparison of the AP600 piping to the screening criteria in Subsection 5.29 I
of NUREG/CR-6519 (Reference 13) demonstrates that there is not a significant potential for
- I water hammer in the leak-before-break piping.
Thermal stratification of water in stagnant or slowly flowing lines can result in thermal fatigue in a pipe. He piping and system design requirements for AP600 address the potential for thermal stratification. For additional information of thermal stratification, see subsections 3.9.3, 5.4.3, and 5.4.5.
The water chemistry and flow velocities in the main steam lines are controlled to minimize I
the potential for erosion and corrosion. At full power the flow rate in the main stream line l I is less than the nuclear industry criteria for Aeam velocity in advanced light water reactors of
- I~ 150 ftdsec. The main steam lines are not subject to water hammer or thermal stratification by the nature of the fluid transported.
The steam line is protected from being filled with water due to steam generator overfill by implement:. tion of operating instructions or isolation requirements included in the protection system logic or both. See Section 7.3 for information on the protection system design to prevent overfill.
In addition to requirements on the design, fabrication, and inspection of the piping systems, the application of mechanistic pipe break requires a qualified leak detection capability. Leak detection systems inside containment meet the guidelines of Regulatory Guide 1.45. See 4
subsection 5.2.5 for a discussion of the leak detection system for the reactor coolant system and connected piping.
3.6.3.2 Design Criteria for Leak before Break The methods and criteria to evaluate leak-before-break in the AP600 are consistent with
] the guidance in NUREG-1061 (Reference 11) and Draft Standard Review Plan 3.6.3 (Reference 12). The application of the mechanistic pipe, break in AP600 requires that the following design requirements are met.
Pre-service inspection of welds is required.
l e For ASME Code Class 1. Class 2, and Class 3 systems for which leak-before break is demonstrated, the ASME Section XI preservice and insenice inspection will provide i
for the integrity of each system. He weld and welder qualification, and weld I inspection requirements for ASME Code, Section 111, Class 3 leak-before break lines I
are equivalent to the requirements for Class 2. The inservice inspection requirement for 4
oksarrvlTO306n.Rl244 t ?97 Revision: 12 W Westiligh0US8 3.6-33 Draft,1997 J
- . -- --- .- - . - - _ . ., ... . . - . ~ - .
?" - "
j I 5
- 3. Design cf Structures, Ccmponents, Equipm:nt, tnd Systems l
' each Class 3 leak-befom-break line includes a volumetric inspection equivalent to the I
requirements for Class 2 for the weld at or closest to the high stress location.
1 Inservice inspection and testing of snubbers (if used) are performed to provide for a low j snubber failure rate.
l For the maximum stress due to steady-state vibration refer to subsection 3.9.2.
De leak-before-break bounding analysis curves are developed for each applicable piping system. He bounding analysis methods are described in Appendix 3B. These i curves give the design guidance to satisfy the stress limits and leak-before-break acceptance criteria. The highest stressed point (critical location) detennined from the piping stress analysis is compared to the bounding analysis curve and has to fall on or under the curve. He points on or under the bounding analysis curve satisfy the j requirements for leak.before-break.
i The analyzed normal stress and muimum stress are not required to construct the
- bounding analysis curve. The analyzed stresses are calculated by the equation; i-F j c=2+M._1
, A Z 1
where: ,
1 l l c is the stress !
Fx is the axial force Mb is the applied bending moment
- . A is the piping cross-sectional area Z is the piping section modulus.
The normal stress is calculated by the algebraic summation of load combination method a=
'and the maximum stress is calculated by the absolute summation of load combination method.
0 3-
- 5 The co:Tosion-resistant piping materials, including base metal and welds, have an appropriate toughness. The piping materials containing primary coolant are wrought stainless steel. The welds in stainless steel pipe are made using the gas tungsten arc
- (GTAW) process. These matnials are very resistant to crack extension. The tensile properties for the leak-before-break evaluation are those found in the Section III Appendices of the ASME Code. During the design stage, the material properties used are based on the ASME Code minimum values. During the as-built reconciliation stage, l certified material test report values are reviewed to verify that ASME Code ;
requirements are satisfied,
- For those lines fabricated using non. stainless ferritic materials, the materials used and j the associated welds have. adequate toughness to demonstrate that leak-before-break Revision
- 12 ssarrvi2o30sn.Ri24 i?97 i Draft,1997 3.6-34 W W85tilighouse
)- J t
, .3 Design cf Structures, C:mponents, Eqdpment, tnd Syst;ms criteria are satisfied. The welds are made using the gas tungsten arc (GTAW) process.
He tensile propenies for the leak-before-break evaluation are obtained from actual material tests. During the design stage, the material properties are based on test results.
During the as-built reconciliation stage, cenified material test report values are reviewed to verify that the toughness and strength requirements of the ASME Code,Section III 4
are satisfied.
=
Potential degradation by erosion, erosion / corrosion and erosion cavitation is examined to provide low probability of pipe failure.
Wall thicknesses in elbows and other fittings are evaluated to confirm that ASME Code, Section Ill piping requirements are met as a minimum.
The as-built condition of the piping and support system is evaluated based on the guidelines in EPRI NP-5630 (Reference 10) and reconciled to the analysis of the leak-before-break criteria based on the design information. The locations and characteristics of the supports, including any gaps between the supports and piping, or other configurations that result in a nonlinear response are included in the as-built evaluation.
=
Adjacent structures and components are designed for the safe shutdown eanhquake event to provide low probability of indirect pipe failure.
The piping supports are anchored to reinforced concrete structures, to concrete-filled steel plate structures, or to steel ',tructures anchored to these types of saucturcs. Piping is not supported by masonry block walls.
3.6.3.3 Analysis Methods and Criteria The methods used to develop the bounding analysis curves are described in Appendix 3B.
Development of the bounding analysis curves provides an evaluation method that is consistent I
with NRC requirements and guidance. The calculation method and computer codes used for l
AP600 are benchmarked to test data and has been previously accepted by the NRC for leak-I before break evaluations in operating nuclear power plants.
Analyzable sections run from one terminal end or anchor to another terminal end or anchor. !
A terminal end is typically a connection to a larger pipe or a component. For the structural analysis, a normally closed valve between pressurized and unpressurized portions of a line is i
not considered a terminal end. Figure 3.6-3 is a schematic of a portion of a piping system j that illustrates the meaning of analyzable segments. In the figure the analyzable portion of the pipe runs from point A to point D.
The leak-before-break evaluation is based on a fracture mechanics stability analysis companng l the selected leakage crack to the critical crack size. The following discussion outlines the ar']ysis method.
1 l
l m Re* m 12 ,
T Westinghouse 3.6-35 Draft,1997 1
,, , 3. Desiga of Structures, Components, Equipmert cod Systems .
l 3B.3.3 Evaluation of Piping System Using Bcunding Analysis Curves To evaluate the applicability of leak-before-break, the results of the pipe stress analysis are compared to the bounding analysis curve. The critical location is the location of highest maximum stress as determined by the pipe stress results. A comparison is made with the applicable bounding analysis curves for the analyzable piping systems. As outlined in 3B.3.1.1 and 3B.3.2.1, bounding analysis curves are calculated for different combinations of pipe size, pipe schedule, operating pressures, operating temperatures.
The bounding analysis curves are useu during the layout and design of the piping systems to provide a design that satisfies leak-before-break criteria. In addition, the Combined License holder compares the results of the as-built piping analysis reconciliation to the bounding analysis curves to verify that the fabricated piping systems satisfies leak-before-break criteria.
See subsection 3.6.4.2 for the Combined License information item associated with this verification.
l l At the critical location, the load combinations for the maximum stress calculation uses the I
absolute sum method. The load comt:tation is as follows: s-includeahe--fellowing
, combinations--
l l l (1) l Pressure l + l Deadweight l + l Thermal (100% Power)*l + l Safe Shutdown l Earthquake l (2)-4Pressuredf--r-Maximune fDeadweightf ' 4Therwal--(100 i--Power)f-4--l%Ive (MPressured-4-4Deadweightf+fThermal-Maximum *f
- -Level-A-end4eveleef-ASMEC-eddoad-conditions--%1ve thrust maximmn-includes anticipated-water-hammer events resulting4 rom 4apid-valve <!osure or openingrincluding pressuricer-safety-valve-opening-(Level-G)--Thermal-maximum-inc4udes-applicable stratification-loads:
!
- Includes applicable stratification loads.
}
l The normal stress is calculated using the algebraic sum method at critical location and the following load combination.
(1) Pressure + Deadweight + Thermal (100% Power)
Revision: 12
[ Westiligh0US$ 3B-17 Draft,1997
, . 3. D. 'gn of Struct:res, Cemponents, Equipment end Syst:ms examination requiremerits for Class 2 pipe require radiographic examination of the welds and normally Class 3 pipe does not. As noted in subsection 3.2.2.5, for Class 3 lines required for emergency core cooling functions, radiography will be conducted on a random sample of welds. The Class 3 leak-before-break lines are included in the lines that are radiographed.
I In addition see subsection 3.6.3.2 for augmented inspection of Class 3 leak-before-break lines.
For the fabrication of welds in the Cass 1. Class 2 and Class 3 pipes there is no significant differences.
The differences in fabrication and nondestructive examination requirements do not affect the leak before-break analyses assumptions, criteria, or methods.
38.7 References
- 1. Standard Review Plan 3.6.3, " Leak Before Break Evaluation Procedures," Federal Register, Volume 52, Number 167 Friday, August 28, 1987: Notice (Public Comment Solicited),
pp. 32626-32633.
- 2. NUREG-1061, " Evaluation of Potential for Pipe Breaks, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," Volume 3,(prepared by the Pipe Break Task Group),
November 1984.
- 3. Deleted .
- 4. Deleted
- 5. Deleted
- 6. Deleted
- 7. Deleted
- 8. Deleted
- 9. ASME Boiler and Pressure Vessel Code, Section 111. " Rules for Construction of Nuclear Power Plant Components; Division ! - Appendices," 1989 Edition, July 1,1989.
i i
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i Revision: 12
[ W85tingh00S8 3B-21 Draft,1997 1
1
{' !
i
- l
! Status of AP600 Open Items in SSAR Section 3.6.3: Leak-before-break j l
(Up to SSAR Revision 11)
, 3/7/97 1
IT m No. & STATUS DFSCRIPTION i
' D5tRf 3.6.3.4-1 LBB Bounding Analysis (0!TS 608) i Action W/N.;
- Add description in SSAR 38.3.1.3 and 38.3.2.3 on i bounding curve construction procedures, or in'3.6.3.3 3
on bounding Analysis to explain how bounding curves l meet LBB acceptance criteria.
1 j
- NRC will audit calculations to ensure that the j bounding curves satisfy LBB acceptance criteria.
!
- Uncertainties in applying LBB to small lines (see NUREG/CR-6443, Section 3.5 on pressure-induced bending l effects to leakage flaw size and max stress) needs to i
- be discussed. M should perform sensitivity studies. !
n On February 7, 1997, E presented a markup on SSAR l
4 Section 3.6.3 and Appendix 3B, which was subsequently
{ incorporated in Revision 11. LBB application to 4-i inch diameter pipe was deleted. Thus this issue is j closed.
- Applying LBB methodology to the feedwater line is unacceptable.due to lack of operating experience and i potential waterhammer load which can not be defined.
! Revisions to the SSAR to delete the main feedwater i line from LBB consideration will be tracked under DSER# 3.6.3.5-5 (0ITS 614). For further discussion of j this issue, see NRC letter dated January 24, 1997.
! On February 7, 1997, H presented a markup on SSAR Section 3.6.3 and Appendix 35, which was. subsequently l incorporated in Revision 11. LBB application to i feedwater line was deleted. Thus t11s issue is 3- closed.
i
- Results from PICEP computer code do not agree with
[ Westinghouse LBB analyses. Need an explanation.
i i
f Attachment 3 l
i i DSERf 3.6.3.4-2 COL applicant to verify L88 bounding analyses (same as COL (01TS 609) Action Item 3.6.3.4-1 or 01T51883) on materials, as-built 4 Closed analyses, and acceptance parameters.
Newly revised SSAR Section 3.6.4.2 (Revision 10) is
- . acceptabe. .Thus DSER Open Item 3.6.3.4-2 and OSER COL
< Action Item 3.6.3.4-1 (OITS 1883) are closed.
!, DSERf 3.6.3.5-1 Leak detection methods (RAI 252.8)
(0!TS 610)
Closed Changes in sump SSARlevel Section 5.2.5.3.1, is under Rev.10 separateregarding 'f containment monitor staf review. Thus in review areas of Mechanical Engineering Branch, this issue is considered closed.
l 4
) DSERf 3.6.3.5-2 Class 1 vs. Class 2 differences in analysis, fabrication, l (OITS 611) and inspection - RAI 252.5 l l Action M ;
- Due to concern on vibrations induced by normal operation l
- transients, fatigue evaluation equivalent to ASME Class 1
! piping and augmented ISI appear necessary at the main steam 4
connection weld to the steam generator nozzles. Revision of j SSAR Sections 38.2.4, and 39.8 is needed. i
) DSERf 3.6.3.5-3 Location of Main Steam (MS) and Feedwater (FW) anchors (0ITS 612)
- Closed Changes in SSAR Revision 4, Section 3.6.3 paragraph 7 and
- Appendix 3E identify the scope of LBB analysis for the main i steam line, as well as anchor locations. The anchors are to
. remain at the exterior wall of the auxiliary building. This
- is acceptable. In Revision 11 to the SSAR, LBB application i *o main feedwa.ter lines was deleted. Thus OSERf 3.6.3.5-3 is cl o s et. ..
J DSERf 3.6.3.5-4 MS and FW definitions for LBB (Audit issue) l (0!TS 613)
! Closed Scope of LEB analysis for main steam line is consistent due
! to changes in SSAR Revision 4, Section 3.6.3 paragraph 7 and Appendix 3E. In addition, in Revision 11 to the SSAR, LBS l application to main feedwater lines was deleted. Thus OSERf
- 3.6.3.5-3 is closed, i
! DSERf 3.6.3.5-5 Justification of LBB for MS and FW - RAI 252.13 i (0ITS 614 &
- _ 2422 to 2428) The Westinghouse proposal to apply LBB methodology to the
- Closed feedwater line is unacceptable due to lack of complete configurations, lack of operating experience and high degree l
i
i .
! of uncertainties to define the water hammer load for piping '
- design. Further discussion of this issue is stated in NRC letter dated January 24, 1997. '
l On February 7, 1997, M presented a markup on SSAR Section 3.6.3 and Appendix 3B, which was subsequently incorporated in Revision 11 to the SSAR. LBB application to feedwater' line was delstad. Thus DSERf 3.6.3.5-5 is closed.
DSERf 3.8.3.6-1 Soil conditions for LBB analyses - RAI 210.10 "
(0!TS 615)
Action N The NRC staff will audit calculations of LBB bounding analysis for verifying that the seismic loads used in the analysis have adequately considered all applicable soil conditions, and that the critical flaw stability in all LBB applications are based on N+SSE loading. This is also related to the staff effort stated in DSERf3.6.3.4-1 for verifying acceptability of bounding curves, specific' ally under limiting loading conditions.
DSERf 3.6.3.6-2 LBB applicability for branch connections - RAI 252.11 (017S 616)
Action N According to Section V in SRP 3.5.3 (Draft), pipe breaks at branch connections to the main piping run applicable to LBB should still be postulated, unless the branch linestare also qualified for LBC. Westinghouse agreed with SRP position during a meeting on February 14-15, 1995. However, LBB analyses were not competed at that time. For verification purpose, the staff will audit LBB calculations.
Thus DSERf 3.6.3.6-2 remain open for NRC action.
DSERf 3.6.3.6-3 0.5 gpm vs. 1.0 gpm leakage rate (0ITS 617)
Closed The following criteria were incorporated into the SSAR: 1) capability to monitor 0.5 gpm leak rate is demonstrated, ((2) using absolute sum combination of normal and SSE load with load factor of 1.0, and (3) using a factor of two between critical crack length under normal plus SSE and the leakage crack length under normal load. DSER# 3.6.3.6-3 is closed.
DSER# 3.6.3.6-4 Lea 6. age rate evaluation methodology (OITS 618)
Action M This issue is also a part of DSER Open Item 3.6.3.4-1.
The closure of DSER Dpen Item 3.6.3.6-4 is pending Westinghouse to show their benchmarks during a staff audit for verifying leak rate methodology used in their LBB analyses.
I DSERf 3.6.3.5-5 Justification of the part-through flaw criterion (0!T3 619) '
closed In S$AR Section 3.6.3.3 Revision 4, the part-through flaw j criterion was deleted. ,This is acceptable.
i
- DSERf 3.4.3.5-8 Waterhammer type loads in LB8 analyses (Test results issue)
(0175 620)
Action M Preliminary results from small-break LOCA tests performed at i
i Oregon State University indicate that rapid condensation i
events have the potential to cause unanticipated dynamic loads to occur in the AP600 RCS. These water hammer type i loads have not been considered in the piping design loads to
' justify a LB8 approach for the AP600 main coolant loop and i
attached piping., M was requested to address whether these water hammer type loads from condensation events need to be i
considered in its L88 an:1yses or, if not, justify why these
' loads can be excluded and incorporate relevant discussions in the SSAR.
. 1 i
I l
l i
I i
1 j Summary of input and Results for independent Vertfication of Bounding Analysis for 32' AP-800 Main Steam Piping
- (See NURECWOR-8281, Table 24 and Section 3)
}
i OD = L2.0000 inch t= 1.5400 inch R= 15.2300 inch R/t = 9.8896 4
R1 = 14.4600 loch j i Yield = 32.0000 kal Ultimate = 66.3400 kai
- Flow Stress (Sigma f) = 49.4700 kai
} B= 0.7583 l A= 0.9965
- Zeta (see Table 2-4) = 0.0268 ;
Leakage Flaw = 6.8900 inch l Critical Flaw Length = 13.3800 inch i Crttical Theta /Pi = 0.1398 Pressure = 0.8183 kH
, Axial Force - 537.5250 k$a i
! Sa = 3.6475 kai 4
Zeta
- Sa = 0.0977 kol
- Sigma m = 4.8345 kal I
Sigma b = 45.2595 ksi (Sigma b + Sigma m)/ Sigma f = 1.0126 (Sa + Sb) for Critical Flaw Size = 37.7950 kai i
i Computed (Sa + Sb) = 37.8 kal, and is larger than 2
allowable load from the Westinghouse boundirig analysis 1 .
i s
Attachment 4
MEETING WITH .
i THE UNITED STATES-NUCLEAR REGULATORY COMMISSION ON "AP600 LEAK-BEFORE-BREAK AND RELATED ISSUES" PITTSBURGH APRIL 16-18,1997 4
33 e
s I i e
. - . - _ . . - - - . - . . . - . - - _ . . . _ . - . - . - . . - . . ~ . . - - - - - - . - . . - . - . - . . - .
6 ITEM # IV: Verification of computer codes used in calculation of-leak rate, crack length and crack stability. t
. Westinghouse describes verification process used and show documentation, including benchmarked test data. l t
I RESPONSE: The NRC has approved the Westinghouse LBB analysis methodology and criteria for various applications. Therefore we believe our methodology is currently accepted by the staff. Westinghouse has l l successfully demonstrated leak-before-break for the primary coolant loops l and auxiliary lines in over 40 plants and obtained NRC approval. ;
i Wi :
<W !
1
I At the request. of the NRC, benchmark evaluations were performed to validate Westinghouse methodology. These leak rate and stability calculation methods were accepted by the NRC as described below:
t Verification of Westinghouse Codes:
The Westinghouse computer code to calculate leak rates was developed in the latter part of the 1970's. The Code was verified by comparison with experimental data and the results were presented in WCAP-9558, Rev. 2 (Reference 1).
In 1986 the NRC staff requested to perform a comprehensive review of the benchmarking of all the available and newer data.
In response to that Westinghouse did the followings:
a Bench mark calculations were performed and compared with the leakage observed in the recirculation pipe at Duane Arnold Nuclear Power Plant (Reference 2, WCAP--11256, Supplement 1).
e e
i i
o A comparison between the experimental data performed by Battelle Columbus Laboratories (Reference 3 ) and Westinghouse code was shown in Reference 2. The Westinghouse code shows a good correlation.
a Recognizing the need for detailed and accurate leak rate measurement data, Westinghouse, Framatome, Electricite de France (EDF) and Commissariat a l'Energie Atomique (CEA) performed research on the subject. The actual leak rates were conducted by CEA at Aquitane ll (Reference 4). A comparison of the experimental ,
results and the analytical predictions was shown in Table 2.4-2 of reference 2. The Westinghouse code shows good agreement between experimental data and analytical r predictions.
o For the fracture mechanics calculations, analytical predictions were compared with the experimental results and shown in reference 5. Westinghouse code shows a good correlation. :
i 4
f
=
l W>
i
.a . w4 i
~
i After reviewing the information as stated above the NRC reconfirmed the acceptability of the Westinghouse leak-before-break methodology and the benchmark data for the leak-before-break applications as documented in reference 6. ,
The NRC staff concluded (reference 6):
"The staff has reviewed the benchmark information supplied by the applicant and concludes that the software used to estimate leak rate is adequately verified and has acceptable accuracy for application to LBB analysis when~used with appropriate margin on predicted leak rate".
Also the NRC staff concluded (reference 6):
The computer code used for fracture mechanics calculations has been benchmarked against data from the bend test of a. cracked pipe. The computer code also has been benchmarked against an established fracture mechanics code used previously by Westinghouse. The staff reviewed the benchmark information supplied by the applicant and concluded that the computer code has acceptable accuracy ..........................". j l
i e ,
m .
REFERENCES:
- 1. Palusamy, S. S. and Hartmann, A. J., " Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack," WCAP-9558, Rev. 2, Class 2, May 1981 (Westinghouse Proprietary Class 2).
2, Swamy, S. A., Witt, F. J. and Bamford, W. H., " Technical Bases for Eliminating .l Pressurizer Surge Line Ruptures as the Structural Design Basis for South Texas Project Units 1 and 2," WCAP-11256, Supplement 1, November 1986 (Westinghouse Proprietary Class 2).
3m Collier, R. P. , J. S. K. Liu, M. E. Mayfield, F. B. Stulen, " Study of Critical Two-Phase Flow Through Simulated Cracks", Battelle Columbus Laboratories, Interim Report BCL-EPRI-80-1, November 1980.
t h t
. . . . . i t
r
l VI. Audit of LBB Calculations 9 Sample of LBB Calculations Main Steam A (32" pipe)
System : SGS-PLA-030 Pzr. Surge Line (18" pipe) System : RCS-PLA-040 1
PRHR Retum Line (10" pipe) System : PXS-PLA-040 CMT-B (8" pipe) System : PXS-PLA-060 DV)-A (6" pipe) System : PXS-PLA-010 Verify conformance with bounding analysis procedures LBB Screening evaluations are performed based the criteria and Bounding Analysis Curvers included in Section 3B of the '
AP600 Standard Safety Analysis Report (SSAR).
Verify meeting LBB Acceptance Criteria Results from LBB screening evaluations for the above mentioned piping systems are summarized on the following pages. Calculated LBB stresses are ploted against the appropriate Bounding Analysis Curve for each applicable pipe size.
j l
l i
j
\
i Attachment 6 o
i I
O- . Verify the SSE loads have considered all applicable soil conditions Seismic analysis of all seismic Category I / II auxiliary piping systems is performed based on enve!oped response spectra analysis.
The response spectra is generated from the data provided in GW-S2R-002, " Nuclear Island Stmetures - Safe Shutdown Earthquake Seismic Response Report" The Amplified Response spectra provided in GW-S2R-002 represent an envelope of the following four soil design cases and includes.a 15% broadening to account for variations in structural frequencies :
soft rock site soft-to-medium soil site soft-to-medium soil (upper bound) site hard rock site l
G Verify that pipe break at branch line connection is postulated unless t
LBB is also applicable to the branch line.
This applies only to 2 piping systems, the Primary Coolant Loop and the Automatic Depressurization System, Stage 2 / 3.
Terminal end breaks are postulated for the 4" spray line attached to the Primary Loop, and for the 4" ADS stage 1 piping attached to ADS stage 2 / 3.
4 m
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SGS-030 BOUNDING ANALYSIS CURVE 28 35 i
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0 5 10 15 20 25 30 OD = 32.00 Tmin = 1.54 MAIN STEAM
( s s e)
SGS-030 BOUNDING ANALYSIS CURVE 45 30 l I i
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- M s tu susn (Ynnuna n g x ,,,, u n )
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RCS-040 l 50 BOUNDING ANALYSIS CURVE 36 I
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- 50 BOUNDING ANALYSIS CURVE 37 I
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l PXS-040 BOUNDING ANALYSIS CURVE 11 i
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BOUNDING ANALYSIS CURVE 22
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PXS-060 BOUNDING ANALYSIS CURVE 16 50 I I
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- PXS-060 BOUNDING ANALYSIS CURVE 23
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NRC/WESTINGiidUSE AP600 MEETING
, DESIGN ACCEPTANCE CRITERIA & INSPECTION, TESTING, ANALYSIS AND ACCEPTANCE CRITERIA FOR PIPING APRIL 17, 1997 MEETING PARTICIPANTS NEE ORGANIZATION Goutam Bagchi NRC/NRR/DE/ECGB Thomas Cheng NRC/NRR/DE/ECGB i Matthew Mitchell NRC/NRR/DE/EMCB David Terao NRC/NRR/DE/EMCB Joseph Sebrosky (T) NRC/NRR/DRPM/PDST Jerry Wilson (T) NRC/NRR/DRPM/PDST Ed Cummins Westinghouse Don Lindgren Westinghouse-Licensing Roger Schreiber Westinghouse Jeff Willis Westinghouse (T) - on telephone in Rockville Attachment 7
.-.l 5
AP600 Certified Design Material Proposed Piping & Structural ITAAC t
i i
Roger Schreiber Jeff Willis 4/17/97 c
I
Overview i
e In a telecon on May 31 with the Civil and Structural Branch, we proposed an alternate approach to piping and structural requirements in the CDM i
e Our request is based on the fact that the AP600 design is far more complete in ;
these areas than were the evolutionary plants at their Tier 1 submittals e We believe credit should be given for the level of design detail available at Design Certification rather than requiring NRC review and approval at each ;
COL application -
1 e It should be possible to replace the DAC in the evolutionary plant submittals with more definitive ITAAC e The staff requested that we submit a written description of our proposed approach for consideration i
e Following are our proposals for: Piping (Roger Schreiber)
Structural (Jeff Willis)'
i
. t
e Piping ITAAC -
. Based on RAI 640.19, transmitted in letter of March 4
= That RAI listed 8 specific commitments that should be added to Tier 1 in the form of a piping DAC
= Following gives our proposed resolution of eacI(of hese 8 points individually es;
- Illustrates how each one can be addressed ty inspection, test or analysis of the as-built piping, rather than by a DAC ,
e Format: '
- DAC Commitment requested by staff
- Discussion
- Proposed ITAAC commitment
' ~
Commitment #1: .
Designing the piping to the ASME Boiler and Pressure Vessel Code, Section IH to ensure pressure boundary integrity.
Disctission: .
- Design and analysis of ASME piping for the AP600 is substantially ,
completed e This analysis is being reviewed and approved by the NRC now e Thus, the only inspection that need be specified in Tier 1 is to verify that the as-built piping is reconciled with the design and the completed ASME analysis
-e We propose to. add an ITAAC commitment as follows to each system that contains ASME piping
ASME Code Piping ITAAC:
Table 2.2.3-4: Inspections, Analyses,'and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria 2.b) The piping identified in Figure Inspection of the ASME design The ASME Code Section III design 2.2.3-1 as ASME Code Section Ill is reports will be conducted. repons exist for the as-built piping designed and constructed in identified as ASME Code Section III accordance with ASME Code . in Figure 2.2.3-1.
Section 111 requirements.
3.b) Pressure boundary welds in Non-destructive examination of the A reports exists and concludes that i piping identified in Figure 2.2.3-1 as as-built pressure boundary welds the ASME Code Section III ASME Code Section Ill meet ASME will be perfonned in accordance requirements am met for the quality Code Section 111 requirements. with the ASME Code Section III. of pressure 'oundary welds.
- 4) The piping identified in Figure A hydrostatic test will be performed A report exists and concludes that 2.2.3-1 as ASME Code Section 111 on the piping required to be the results of the hydrostatic test of retains pressure boundary integrity at hydrostatically tested by the ASME the piping identified in Figure 2.2.3-its design pressure. Code Section III. I as ASME Code Section III conform with the requirements of the ASME Code Section III. ;
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= _ = - - . - _ _ -- _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ .
1 Commitment #2:
Designing the piping to ensure itsfimctional capability.
Disctission: l i f
. SSAR Table 3.9-11 gives the functional capability requirements for ASME piping systems that must maintain an adequate fluid flow path to mitigate a Level C or Level D plant event i i
e Since these analyses are also substantially completed for AP600, the most that is needed of Tier 1 is to.
a) Identify the lines to which these criteria apply, and I b) Verify that the as-built piping is reconciled with the completed design -
and analysis, which is being reviewed and approved now.
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. To address this point, we propose adding the following to our existing Tier 1 !
piping entries which are contained within the individual system ITAAC i
k Functional Capability ITAAC:
Table 2.2.3-2 Funetional Erosion-ASME Section Leak-Hefore- Capability Corrosion Line Name Line Numbers ill Hreak Required Allowance IRWST injection PXS-L125 A, yes yes yes n/a
, line A to DVI line A PXS-L127A 1
Table 2.2.3-4: Inspections, Analyses and Acceptance Criieria Design Commitment inspections. Tests, Analyses Acceptance Criteria 6.b) Each of the lines identified in Inspection will be perfomied for the -
A report exists and concludes that Table 2.2.3-2, for which functional existence of a report concluding that each of the as-built lines identified capability is required, meets the the as-built piping meets the in Table 2.2.3-2, for which requirements for functional requirements for functional capability. functional capability is required, capability. meets the requirements for functional capability.
k
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i Commitment #3: .
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Minimizing the effects of erosion-corrosion.
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Disctission: '
j e For the small amount of ASME piping that.is subject to erosion-corrosion, we will include an entry in the system ITAAC. '
j e This entry will commit to a report verifying the allowance in the pipe wall thickness for erosion-corrosion.
r e Following is an examples of the proposed ITAAC.
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I-Erosion-Corrosion ITAAC: .
Table 2.2.4-2 Functional Erosion-ASME Section Leak-Hefore- Capability Corrosion Line Name Line Numbers 111 Break Required Allowance (inches)
Main Steam Lines SGS-LOO 6A yes yes no .08 Within Containmen+ SGS-LOO 6B t
Table 2.2.4-4: Inspections, Analyses and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria 1 6.a) For the ASME Code Section III Inspection will be performed for the For ASME Code Section III piping ;
piping identified in Table 2.2.4-2 as existence of a report confirming identified in Table 2.2.4-2 as having ;
having an allowance for erosion- that the pipe wall thickness includes an allowance for erosion-corrosion, corrosion, a wall thickness at least the allowance for erosion- a report exists and concludes that allowance is provided for the corrosion specified in Table 2.2.4-2. the as-built pipe wall thickness ,
potential effects of erosion- includes at least the specified i corrosion. allowance. <
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Commitment #4: ,
Ensuring that equipment nozzle loads are met.
Discussion:
. The completed ASME piping analysis on AP600 includes verification of equipment nozzle loads.
. The piping analysis is being reviewed and approved by the NRC now.
. Therefore, the only inspection that need be specified in Tier 1 is to verify that the as-built piping is reconciled with the design and the completed ASME analysis. '
- Since the ITAAC shown above under commitment I already accomplish that, t no additional ITAAC are needed. !
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O Commitment #5:
Benchmarking the piping computer code. ,
t Disctission:
f
. Since the NRC has a eady accepted the benchmarked codes used for AP600 piping analysis, no additional ITAAC are needed. !
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Commitment #6:
Ensuring that high-energy line breaks and environmental effects are adequately considered. .
1 Discussion: .
. This item encompasses a number of different concerns related to the dynamic '
and environmental effects of pipe rupture.
- Environmental effects on safety-related equipment.
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- Dynamic effects on equipment
- Effects on structures 1
= We propose to address cach of these concerns individually as follows.
i e
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Environmental effects on equipment: .
. The system ITAAC already include a commitment to environmental qualification of IE equipment.
Table 2.2.3-4: Inspections, Analyses and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria The Class 1E equipment identified in Type tests, analyses, or a - -
A report exists and concludes that Table 2.2.3-1 as being qualified for a combination of type tests and the as-installed Class IE equipment harsh environment can withstand the analyses will be performed on Class identified in Table 2.2.3-1 as being environmental conditions that would IE equipment located in a harsh qualified for a harsh environment exist before, during and following a environment. can withstand the environmental design basis accident without loss of conditions that would exist before, safety function for the time required during and following a design basis to perfomi the safety function. accident without loss of safety function for the time required to perform the safety function.
. No additional ITAAC are necessary to address this issue ,
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Dynamic effects on equipment: .
. We have completed the analysis necessary to identify specific break locations and essential targets that must be protected by pipe whip restraints.
. The NRC is reviewing and approving this analysis now.
. We propose to include this information in Tier 1 Section 3.3 Nuclear Island Buildings, and inspect the as-built plant to ensure that the necessary protection is provided.
Effects on structures: .
. This will also be addressed by revising Section 3.3, Nuclear Island Buildings.
. Jeff will show you the proposed' changes
Commitment #7:
Ensuring that proper materials are used to prevent brittlefracture and reduce the possibility of cracking during service. -
a Disctission: -
t
- Pipe material specifications are completed for AP600. They call for low carbon stainless and other crack-resistant raaterials. .
. The most that should be required in Tier 1 is: j
- A commitment to meet Code requirements for brittle fracture !
- Verification that the as-built pipe is constructed of the material specified t i
e We will include entries as follows in the system ITAAC. ,
e w Material Fracture Toughness / Cracking ITAAC:
Table 2.2.3-2 _
Funetional Erosion-ASME Section Leak-liefore- Capability Corrosion Line Name Line Numbers III Break Required Allowance IRWST injection PXS-LI25A, yes yes yes n/a line A to DVIline A PXS-LI27A Table 2.2.3-4: Inspections, Analyses and Acceptance Criteria Design Commitment ~ Inspections, Tests, Analyses Acceptance Criteria -
The ASME Code Section III piping Inspection will be performed for the For ASME Code Section III piping .
identified in Table 2.2.3-2 is existence of a report concluding identified in Table 2.2.3-2, a report constmeted of a material that meets that ASME Code Section III exists and concludes that the as-the ASME Code Section III fracture fracture toughness requirements are built material of construction meets toughness requirements. met. the ASME Code Section III fracture i toughness requirements. ,
The ASME Code Section III piping Inspection will be perfonned for the For ASME Code Section III piping identified in Table 2.2.3-2 is existence of a mport concluding identified in Table 2.2.3-2, A report constructed of a material that is that the as-built pipe is constructed exists and concludes that the as-resistant to environmentally assisted of the material specified. built pipe is constructed of the cracking during service. material specified.
1 Commitment #8:
Ensuring that adequate clearances are provided during construction.
Discussion: -
1
. We propose to perform a walkdown of each piping system during hot functional testing or at another time when the piping will be filled with fluid '
at operating temperature.
e The walkdown will verify that the piping is unrestrained from thermal movement.
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= Following is the proposed Tier i entry.
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Piping clearance ITAAC: ..
Table 2.2.3-4: Inspections, Analyses and Acceptance Criteria Design Commitm _ . _ _ . Inspections, Tests, Analyses Acceptance Criteria Design of the ASME Code Section Inspection of the as-built piping A report exists and concludes that III piping shown on Figure 2.2.3-1 will be performed, with the piping the as-built ASME Code Section III provides for cleamnce between filled with fluid at operating . piping shown on Figure 2.2.3-1 has adjacent piping, components and temperature, to ensure that clearance from adjacent piping, other stnictures when the piping clearance exists. components and other stmetures. '
moves due to design static and thermal loadings.
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Summary e Each of the issues raised in the March 4 letter can be addressed by inspection, test or analysis of the as-built piping.
. No piping DAC is necessary
. The ITAAC changes and additions just described will be incorporated in the next submittal 1
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