ML101940366
| ML101940366 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/26/2008 |
| From: | Kandalepas C Automated Engineering Services Corp |
| To: | Florida Power & Light Co, Office of Nuclear Reactor Regulation |
| References | |
| GL-87-011 PBNP-994-21-05-P02, Rev. 0 | |
| Download: ML101940366 (42) | |
Text
ENCLOSURE I ATTACHMENT 4 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 261 EXTENDED POWER UPRATE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PBNP-994-21-05-P02, REVISION 0, MAIN STEAM PIPING GL 87-11 BREAK LOCATION DETERMINATION 41 pages follow
Automated Engineering CALCULATION SHEET Page: 4 of 13 Services Corp Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No E Date: 9/26/2008 TABLE OF CONTENTS 1.0 Purpose........................................................................................................................................................
5 2.0 Background...................................................................................................................................................
5 3.0 Assumptions and Analysis N otes............................................................................................................
6 4.0 M ethodology and Acceptance Criteria...................................................................................................
7 5.0 Referencees...................................................................................................................................................
9 6.0 GL 87-11 Break And Leak Location Calculations................................................................................
10 7.0 Results & Conclusions...............................................................................................................................
1I Attachments Pages A. Resultant Stress Calculations Tables...................................................................................................
AI-A2 B. SK -M S-FIG.3 - Postulated Line Breaks...............................................................................................
B I C. Calculation of Local Pipe Stresses at Integral Pipe Attachm ents......................................................
C1-C20 D. Calculation of Com bined Stresses at Tee N ode Point 395...................................................................
D1-D2 E. Com bined Support Loads and Review of N on-Seism ic Supports........................................................
El-E3 F. A ES Technical Position Paper for H ELB Program.................................................................
Fl-F3 Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering Services Corp Page: 5 of 13 Calc. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Light Station: Point Beach Nuclear Plant - Unit 2 Prepared 13y: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes lxI No El Date: 9/26/2008 1.0 PURPOSE The purpose of this calculation is to establish the locations of intermediate high energy large breaks and leakage cracks utilizing the criteria given in Generic Letter 87-11 and its attachment USNRC Mechanical Engineering Branch Technical Position, MEB 3-I, Revision 2 (Reference 3).
Combined stress tables for the Main Steam piping, between the anchors at containment penetrations P-1 and P-2 to HP Turbine Control Valves and to Condensers 2SC-IA and 2SC-IB in the Turbine Building (Ref. 6),
are developed for the sole purpose of determining the locations of intermediate large breaks and leakage cracks in accordance with the combined stress equations defined in Reference 3.
I
2.0 BACKGROUND
Point Beach Nuclear Plant's (PBNP) licensing basis for High-Energy Line Break (HELB) is documented in the Final Safety Analysis Report (FSAR) (Reference 2, Appendix A.2). Appendix A.2 of the FSAR defines a high-energy line as a line with design pressure greater than 275 psig and service temperature greater than 200°F. Both conditions have to be satisfied for a line to be designated high-energy. Additional background discussion regarding the BPNP HELB Program and details for establishing HELB break and leakage crack locations criteria (HELB Reconstitution Program) is provided in the AES technical position paper (see Attachment F).
Based on the above high energy line definition, Calculation PBNP-994-21-02 (Reference 8) identifies the Main Steam (MS) System Lines, Main Feedwater Piping, Steam Generator (SG) Blowdown Piping, and Sampling System Lines as high-energy lines (Reference 2, Appendix A.2).
The application of GL 87-11 methods to determine the new intermediate break and leakage crack locations is expected to be beneficial in addressing design concerns related to high energy line break effects. GL 87-11 still requires terminal end circumferential breaks t6 be postulated irrespective of the combined stress values at these locations.
This calculation determines break and crack locations in the high-energy lines outside containment, based on the combined stress criteria detailed in the GL 87-11 methodology. This calculation does not address the additional postulation of "a single crack, exclusive of stress, at the most severe location with respect to essential equipment" (TE Notice 2000-20, Reference 12), nor does this calculation address the consequences of or evaluate the impacts of breaks or cracks that are required to be postulated based on this criteria.
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering Services Corp Page:
6 of 13 Calc. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Liaht Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[J No D Date: 9/26/2008 3.0 ASSUMPTIONS AND ANALYSIS NOTES Assumptions Piping material, geometric data and stress analysis (computer results) given in the AOR (Reference 6) for Main Steam piping are used as input to develop the GL 87-11 combined stress tables that, in turn, are used to determine the intermediate break and leakage crack locations.
Analysis Notes The code of record for this plant is USAS B31.1 Power Piping Code, 1967 Edition (Reference 1). The Main Steam piping stress analysis documented in Reference 6 was performed using ASME B&PV Code,Section III, Subsection NC and ND, 1977 Edition up to and including 1978 Winter-Addenda (Reference 13). The use of this piping analysis Code is documented as acceptable in the Pipe Code Reconciliation Study performed by Impell (see Section 5.1.1 of Reference 10).
Application of the MEB 3.1, Rev. 2 methodology for Class 2 and 3 piping requires combined stresses to be calculated in accordance with the 1986 ASME Section III, Class 2 requirements (Reference 4). As such, for consistency with the GL 87-11 criteria, additional computer analysis based on the 1986 ASME Section HI Edition was performed (Reference 6). Stress results obtained from this analysis can be directly used and compared to the threshold limits as discussed in Section 4.0.
The piping between containment penetrations P-1 and P-2 and valves 2MS-2017A and 2MS-2018A respectively was evaluated as seismic category I piping (Reference 6). The piping downstream of the valves is non-seismic, however, two supports in each direction past the seismic category I piping were evaluated as seismic category I supports. This was done to obtain the effect of the non-seismic piping portion on the seismic category I piping. In order to apply the GL 87-11 criteria to the non-seismic region pipe stresses, it is necessary to show that the non-seismic supports can withstand the effects of the Operational Basis Earthquake (OBE). Therefore, the non-seismic supports are reviewed in Attachment E and shown to conform to Section 5.8 of DG-M09 (Reference 10).
The pipe stress analysis (Reference 6) calculated combined stresses at 30" tee Node Point 395 that were found to be slightly over the threshold limit for large break. Attachment D of the calculation removed some of the conservatism in the combined stresses by following ASME Section III Code, 1986 Edition and replacing the pressure portion of Equation 9 by the expression provided in Section NC-3651 (a).
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Autoinated E ngineering
/7 Services Corp Page: 7 of 13 CaIc. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Light Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes Ej No El Date: 9/26/2008 4.0 METHODOLOGY AND ACCEPTANCE CRITERIA This calculation uses the GL 87-11 (Reference 3) methodology to determine postulated pipe break and crack locations. The analysis in Reference 6 was performed using ASME B &PV Code,Section III, Subsection NC and ND, 1977 Edition up to and including 1978 Winter Addenda (Reference 13) as the piping code consistent with DG-G09 (Reference 10). For consistency with the GL 87-11 criteria and MEB 3-1, additional computer analysis based on the 1986 ASME Section III was performed (Reference 6). All stress components for the ASME Section III stress combination are obtained from the 1986 ASME Section III analysis.
The following is a discussion of the high-energy line break criteria used to establish the break locations using the GL 87-11 methodology.
4.1 Intermediate Large Breaks The GL 87-11 and MEB 3-1, Revision 2 criteria (Reference 3) for intermediate large breaks is based on the combined stress formula given by the sum of Equations 9 and 10 of ASME B&PV Code Section 1II, Class 2 and 3 as follows:
B, PDo/2t + B2 MDw/Z + B2 MoBE/Z + i MT11/Z Ž 0.8 (1.8 Sh + SA)
(1)
In above equation, the first term is the longitudinal pressure stress. The second and third terms represent the stresses due to deadweight and OBE loads cases respectively. The fourth term is the thermal expansion stress. A stress table which summarizes all Node Points that exceed the threshold limit for breaks and cracks was developed and is provided in Attachment A. Based on the results from the pipe stress analysis (Reference 6), stresses are shown for Equation 9 (Level B) and Equation 10.
Local stresses due to IWA's are also included and combined as applicable. The combined stresses are compared to the threshold limits for cracks and breaks.
Where:
P
= Design internal pressure, psi Do
= Outside diameter of the pipe, in t
= Nominal thickness of the pipe, in MOW
= Resultant moment due to dead weight, in-lbs Mo01
= Resultant moment due to operating basis earthquake, in-lbs MTHj
= Resultant moment due to thermal expansion, in-lbs Sh
= Material allowable stress at temperature, psi SA
= Material allowable stress range, psi Z
= Section modulus of pipe, in3 i
= stress intensification factor, as given in Figure NC-3 673.2(b)-I B1
= primary stress index for pressure stress as given in Table NB-3 681(a)-i B2
= primary stress index for bending stresses as given in Table NB-368 1(a)-I Form 3.1-3 Rev. 2
I Page: 8 of 13 Calc. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Light Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes N]
No El Date: 9/26/2008 Large intermediate breaks are to be postulated only in locations where the combined stress exceeds the threshold value of 0.8(1.8S,-+ SA) (Reference 3). The requirements for arbitrary intermediate large breaks are eliminated by Reference 3.
4.2 Circumferential Breaks at Terminal Ends of Main and Branch Lines Terminal Ends The GL 87-11 criteria state that circumferential breaks have to be postulated at terminal ends of the main run as well as the branch piping. Terminal ends of a piping run are defined as the ends terminating at components, or at other piping (run pipe), or at intermediate anchors. Footnote 3 of MEB 3-1, Rev. 2 provides a definition for the term "terminal ends" which was missing in the Giambusso letter (Reference 5). The footnote defines terminal ends as "Extremities of piping runs that connect to structures, components (e-g., vessels; pumps, valves), or pipe anchors that act as-rigid-constraints to the piping motion and thermal expansion. A branch connection to a main piping run is a terminal end of the branch run, except where the branch run is classified as part of the main run and is shown to have a significant effect on the main run behavior..."
Terminal ends for the Main Steam piping are summarized on Table 7.1 (Section 7.0).
Branch Lines All branch lines to the Main Steam piping, except the 3" Main Steam piping to anchor EB-8-A 118, the 3" Main Steam piping to anchor EB-8-A121, and the Main Steam piping to Auxiliary Feedwater Pump 2-P29, have been analyzed integrally (Reference 6) along with the Main Steam header piping in a single computer model. The effects of the branch on the main run and vice versa have been automatically included in the overall stress analysis of the MS piping. Therefore, based on the Reference 3 criteria, branch points along the main run need not be considered as terminal ends of the branch line for the purpose of postulating a circumferential large break at these locations if the combined stresses at these locations do not exceed the large break threshold criterion.
4.3 High-Energy Line Leakage Cracks (Small Breaks)
The GL 87-11 and MEB 3-1, Revision 2 criterion for leakage cracks is based on the same combined stress formula given in equation (1) above, except the threshold stress value on the right side of the equation is reduced by one-half as follows:
B, PD0/2t + B2 MDw/Z + B2 MoB,/Z + i MTH/Z > 0.4 (1.8 Sh + SA)
(2)
Leakage cracks are to be postulated in locations where the combined stress exceeds the threshold value of 0.4(l.8Sh,+ SA).
Form 3.1-3 Rev. 2
Automated Engineering Services Corp Page:
9 of 13 Cale. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Light Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cafc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No 0i Date: 9/26/2008
5.0 REFERENCES
- 1. USAS B31.1.0, Power Piping Code, 1967 Edition.
- 2. PBNP FSAR, Appendix A.2, High Energy Pipe Failure Outside Containment.
- 3. Generic Letter 87 Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, June 19, 1987
- 4. ASME Section III, 1986 Editions.
-5. AEC-DOLs Letter to WPS of December 15, 1972 (Mr. Giambusso to Mr. James).
- 6. Shaw Calculation No. 129187-P-00 11, Rev. OA, "Point Beach Unit 2-Main Steam Outside Containment -
Piping Qualification for Extended Power Uprate Conditions".
- 7. PBNP Accession No. WE200098, Piping System Qualification Report for Main Steam Outside Containment to HP Turbine Control Valves, Unit 2, Rev. 0 including Addenda A and B.
- 8. Calculation No. PBNP-994-21-02, 1ELB Reconstitution Program-Task 2, High Energy System Selection, Rev. 0.
- 9. Not Used.
- 10. DG-M09, Rev. 2, Design Requirements for Piping Stress Analysis.
- 11. Drawing M-2201 Sh. 1, Revision 49, Piping Subsystems P&ID - Main & Reheat Steam System.
- 12. NRC Information Notice 2000-20, Potential Loss of Redundant Safety Related Equipment Because of the Lack of High Energy Line Break Barriers, 12/11/2000.
- 13. ASME Section III B&PV Code, Subsection NC and ND, 1977 Edition up to and including 1978 Winter Addenda.
- 14. Isometric Drawing: P-207, "Main Steam Outside Containment to H.P. Turbine Control Valves and to Condenser 30,24 EB-l, 24, 18, 16 EB-2 & 1OHB-12", Sheet I Rev. 1, Sheet 2 Rev. 0, & Sheet 3 Rev. 0.
- 15. ASME Code Case N-318-5, "Procedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 & 3 Piping," 4/28/94.
- 16. ASME Code Case N-392-1, "Procedure for Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 2 & 3 Piping," 12/11/89.
- 17. "Stresses in Elbows Created by Supported Lug Load," T. K. Emera and E. C. Rossow, 1979 PVP Spring Conference 79-PVP-51.
Form 3.1-3 Rev..72
Automated Engineering Services Corp Page:
10 of 13 Cale. No.: PBNP-994-21-05-P02 Revision: 0 Client:
Florida Power & Light Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-1 1 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[0 No El Date: 9/26/2008 6.0 GL 87-11 BREAK AND LEAK LOCATION CALCULATIONS As discussed in Section 4.0, the threshold stress limits used by GL 87-11 are determined in accordance with the requirements of ASME Section III Code, for Class 2 and 3 piping. This Section provides additional information on the GL 87-11 method to determine postulated break and crack locations.
6.1 Application of GL 87-11 Criteria The requirements of GL 87-11 (Reference 3) are applied to the Main Steam piping from containment penetrations P-i and P-2 to H.P. turbine control valves and condensers 2SC-1A and 2SC-IB. There are no intermediate anchors between the containment penetration anchors at Nodes 5 & 7010 and the Turbine Control Valves at Nodes 240 & 380.
A stress table which summarizes all Node Points that exceed the threshold limit for breaks and cracks was developed taking piping stress data from Reference 6 computer analyses. The stress table is provided in Attachment A. Based on the results from the pipe stress analysis (Reference 6), stresses are shown for Equation 9 (Level B) and Equation 10. Local stresses due to IWA's are also included and combined as applicable. This information was then used to calculate the combined stress value for comparison to threshold values established per equations (1) and (2) that were defined in Section 4.0.
6.2 Calculation of Local Pipe Stresses due to IWA's The localized stresses developed on the pipe due to integral welded attachments (IWA) are calculated in Attachment C and are superimposed on the pipe stresses calculated by the computer analysis. As discussed in the pipe stress analysis guidelines DG-M09 (Reference 10) the specific methodology used to calculate IWA stresses uses as a guide the applicable Code Cases (such as N-3 18 and 392), Welding Research Council Bulletins or Finite Element Analysis as appropriate.
The IWA calculations are not intended to evaluate the IWA's or the weld between the IWA and the pipe, but rather calculate the local stresses for use in evaluating HELB point locations in accordance with Section 4.0 of this calculation. The local stresses are added to the piping system stresses to obtain the combined stresses and compare them to the threshold stress limits (see pipe stress table in Attachment A).
Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page:
11 of 13 Services Corp Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Caic.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes W]
No El Date: 9/26/2008 7.0 RESULTS & CONCLUSIONS Results The Main Steam lines between the containment penetration anchors P-I and P-2 and the H.P. Turbine Control Valves and Condensers 2SC-1A & 2SC-1B including branch lines (Ref. 6) have been evaluated for break and crack locations following the requirements and criteria of Generic Letter GL 87-11.
Large Breaks Terminal end circumferential breaks are to be postulated at the terminal ends of the main steam lines at the penetration anchors inside containment and at.the-H.P.-Turbine Control Valves. The terminal ends of the 6" lines to Condensers 2SC-1A & 2SC-IB also require large breaks to be postulated. Locations where large breaks are required to be postulated are summarized in Table 7.1 below.
As seen from the stress tables in Attachment A, there is a single location where the combined stresses exceed the Intermediate Large Break Limits, and as such, this location requires large break to be postulated. This location is also summarized in Table 7.1 below.
Table 7.1 - Postulated Laree Breaks at Terminal Ends and Intermediate Locations Break Location Node Point Notes 30" MS header at Containment Penetrations P-1 & P-2 7010, 5 Terminal Ends Turbine Stop Valves 240, 380 Terminal Ends 8" & 6" branch lines from MS relief header (from 7070,8000, Terminal Ends 2HIX-IA) 8030,8060, 9055 8" & 6" branch lines from MS relief header (from 25016,25017, Terminal Ends 2HX-IB) 25018,25019, 23020 Inlet side of control valves to condenser 2SC-1A 630,725,825,925 Terminal Ends Inlet side of control valves to condenser 2SC-IB 2070,3065, Terminal Ends 4070,5065 3" MS to Auxiliary FW Pumps 16,7018 Terminal Ends 16" side of24x24x)16 Tee 405 Intermediate Break 6" Elbow upstream of control valve 2MS-2057 920 Intermediate Break All postulated large breaks at terminal ends and intermediate locations are shown on Drawing No. SK-MS-FIG. 3 (Attachment B)
Form 3.1-3 Rev. 2
CALCULATION SHEET Page:
12 of 13 Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Linht Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-1 1 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes 9
No ElI Date: 9/26/2008 Leakage Cracks (Small Breaks)
Leakage cracks need to be postulated at locations where the combined stress exceeds the threshold limits as shown in the stress table (Attachient A). Locations where leakage cracks are required to be postulated are summarized in Table 7.2 below.
Table 7.2 - Leakage Crack (Small Break) Locations Main Steam Lines Crack Location Node Point Notes 30" Pipe riser at supports EB-I-2H3 & EB-1-2141 5 120,6010 30"x24" eccentric reducers 165,305 24" Elbow north of 24" Tee 290
-24_'xl18" eccentric reducer 1453 6" Elbow (2 "d elbow upstream of 2MS-2054) 600 6" Elbows (3 elbows upstream of 2MS-2055) 701,702,703,710,720, 6" Elbow (4k" elbow upstream of 2MS-2056) 770 6" Elbows (2 elbows upstream of 2MS-2056) 802,810,820 6" Elbows (4 elbows upstream of 2MS-2057) 870,880,890,900,901, 902,903,910,920 6" Elbow (2 "d elbow upstream of 2MS-2050) 2035 6" Elbows (3 elbows upstream of 2MS-2051) 3015,3025,3030,3045 6" Elbow (2 "d elbow upstream of 2MS-2052) 4035 6" Elbow (2 elbows upstream of 2MS-2053) 5030,5035,5040,5045 5055 6"x3/4" branch conn. upstream of control valves 628,823,2068,3062, 4068,5062 6"Pipe upstream of 6"x3/4" branch connection 5060 30" Pipe riser (at support EB-1-2H202) 6004,6001 Both ends of check valve 2MS-2018A 390,385 24" Elbow by HP Turb. Stop Valve 2MS-2027 375 30" Tee near Penetration P-1 7000 30" Tees by check valves 2MS-2017A & 2MS-2018A 395,130 30x30x24 Tees downstream of check valves 2MS-300,150 2017A & 2MS-2018A 24" Tee 275 24x24x16 Tee 405 16x16x6 Tess at 6" lines to Condensers 550,555,560,565,1085, 1090 Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated T
Engineering CALCULATION SHEET Page:
13 of 13 Services Corp Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[Q No DI Date: 9/26/2008 Conclusions An evaluation of the MS lines from containment penetrations P-1 and P-2 to the respective H.P. Turbine Control Valves and Condensers 2SC-IA & 2SC-1B using the GL 87-11 and its associated USNRC Branch Technical Position MEB 3.1, Rev. 2 (Reference 3) is described in this calculation. The calculation shows that:
- No intermediate stress related and arbitrary large breaks are required to be postulated for the Main Steam header piping. An intermediate stress related large break is required to be postulated on the 6" line to condenser 2SC-1A (Node Point 920 just upstream of the terminal point at control valve 2MS;2057). Also, an intermediate stress related large break is required to be postulated on the 16" side of the 24x24x16 tee on the line to condenser 2SC-IB (Node Point 405).
" Circumferential large breaks are required to be postulated at the terminal ends as shown in Table 7.1.
Leakage cracks (break size = '/2 times the pipe wall thickness x Y2 the pipe internal diameter) are required to be postulated at the locations summarized in Table 7.2.
This calculation does not address the postulation of a single crack, exclusive of stress, at the most severe location with respect to essential equipment (iE Notice 2000-20, Reference 12), nor does this calculation address the consequences or evaluate the impacts of breaks or cracks that are required to be postulated based on this criteria.
Form 3.1-3 Rev. 2
Caic. No. PBNP-994-21-05-P02, Rev. 0 Attachment A, Page Al of A2 Prepared by: Chris Kandalepas Checked by: Dan Quijano Stress Table for Main Steam System - Unit 2 MS from Containment Penetration P-1 & P-2 to H.P. Turbine Control Valves and to Condenser (Based on analysis results from Shaw Group Pipe Analysis, Reference 6)
Input Data from Analysis Eq. 9B Eq. 10 Ratio Ratio Outside Pipe Pipe IWA Pipe IWA Comb.
Limit Limit
=
=
COMMENTS NODES dia.
thickness Stress Stress Stress Stress Stress fotr for Comb. S$
Comb. St Do (in)
- t. (in) psi psi psi psi psi Crack Break Crack Limit Break Limit 120 30 0.908 10743 7721 1343 3089 22896 19800 39600 1.1564 0.5782 IWA (EB-1-2H3) 155 30 0.908 20342 6665 27007 19800 39600 1.3640 0.6820 30x24 Ecc. Reducer 160 24 0.968 20351 6612 26963 19800 39600 1.3618 0.6809 30x24 Ecc. Reducer 165 24 0.968 16389 9915 26304 19800 39600 1.3285 0.6642 30x24 Ecc. Reducer 280 24 0.968 8158 12343 20501 19800 39600 1.0354 0.5177 24" Elbow 290 24 0.968 9424 12958 12817 0
35199 19800 39600 1.7777 0.8889 24" Elbow, IWA(EB-1-2H14) 295 24 0.968 6550 13552 20102 19800 39600 1.0153 0.5076 24" Elbow 453 24 0.968 15615 8849 24464 19800 39600 1.2356 0.6178 24x18 Ecc. Reducer 455 18 0.750 15540 8468 24008 19800 39600 1.2125 0.6063 24x18 Ecc. Reducer 565 16 0.656 7364 2972 744 1745 12825 19800 39600 0.6477 0.3239
]WA (EB-2-219) 600 6.625 0.280 7747 5774 5439 3219 22179 19800 39600 1.1202 0.5601 IWA (EB-2-2H101 628 6.625 0.280 13820 6646 20466 19800 39600 1.0336 0.5168 6x3/4 Branch Conn.
630 6.625 0.280 14382 7187 21569 19800 39600 1.0893 0.5447 Valve 2MS-2054 end 701 6.625 0.280 20700 7645 28345 19800 39600 1.4316 0.7158 6" Elbow End 702 6.625 0.280 20166 5774 7734 3219 36893 19800 39600 1.8633 0.9316 6" Elbow, IWA(EB-2-2H11) 703 6.625 0.280 19277 7182 26459 19800 39600 1.3363 0.6682 6" Elbow End 710 6.625 0.280 18493 10830 29323 19800 39600 1.4810 0.7405 6" Elbow End 720 6.625 0.280 22954 11303 34257 19800 39600 1.7302 0.8651 6" Elbow End 725 6.625 0.280 15116 8321 23437 19800 39600 1.1837 0.5918 Valve 2MS-2055 end 770 6.625 0.280 5952 14331 20283 19800 39600 1.0244 0.5122 6" Elbow End 802 6.625 0.280 10113 5774 6672 3219 25778 19800 39600 1.3019 0.6510 6" Elbow, IWA(EB-2-2H12) 810 6.625 0.280 11093 10932 22025 1980 39600 1.1124 0.5562 6" Elbow End 820 6.625 0.280 9916 11286 21202 19800 39600 1.0708 0.5354 6" Elbow End 823 6.625 0.280 15484 5548 21032 19800 39600 1.0622 0.5311 6x3/4 Branch Conn.
825 6.625 0.280 16168 6048 22216 19800 39600 1.1220 0.5610 Valve 2MS-2056 end 870 6.625 0.280 9901 18779 28680 19800 39600 1.4485 0.7242 6" Elbow End 880 6.625 0.280 6468 15947 22415 19800 39600 1.1321 0.5660 6" Elbow End 890 6.625 0.280 20984 4927 25911 19800 39600 1.3086 0.6543 6" Elbow End 900 6.625 0.280 20359 4979 25338 19800 39600 1.2797 0.6398 6" Elbow End 901 6.625 0.280 19445 7722 27167 19800 39600 1.3721 0.6860 6" Elbow End 902 6.625 0.280 20181 5774 7955 3219 37129 19800 39600 1.8752 0.9376 6" Elbow, JWA(EB-2-2H13) 903 6.625 0.280 18458 7183 25641 19800 39600 1.2950 0.6475 6" Elbow End 910 6.625 0.280 25292 13615 38907 19800 39600 1.9650 0.9825 6" Elbow End 920 6.625 0.280 29513 14291 43804 19800 39600 2.2123 1.1062 6" Elbow End 925 6.625 0.280 16978 10580 27558 19800 39600 1.3918 0.6959 Valve 2MS-2057 end 2000 16 0.656 6992 3272 0
4469 14733 19800 39600 0.7441 0.3720 IWA (EB-2-2H18) 2035 6.625 0.280 12708 3239 7509 1136 24592 19800 39600 1.2420 0.6210 6" Elbow, IWA(EB-2-2H19) 2068 6.625 0.280 10482 10720 21202 19800 39600 1.0708 0.5354 6x3/4 Branch Conn.
2070 6.625 0.280 10573 11314 21887 19800 39600 1.1054 0.5527 Valve 2MS-2050 end 3015 6.625 0.280 12100 8634 20734 19800 39600 1.0472 0.5236 6" Elbow End 3025 6.625 0.280 13631 9939 23570 19800 39600 1.1904 0.5952 6" Elbow End
Calc. No. PBNP-994-21-05-P02, Rev. 0 Attachment A, Page A2 of A2 Prepared by: Chris Kandalepas Checked by: Dan Quijano Stress Table for Main Steam System - Unit 2 MS from Containment Penetration P-1 & P-2 to H.P. Turbine Control Valves and to Condenser (Based on analysis results from Shaw Group Pipe Analysis, Reference 6)
Input Data from Analysis Eq. 9B Eq. 10 Ratio Ratio Outside Pipe Pipe IWA Pipe IWA Comb.
Limit Limit
=
=
COMMENTS NODES dia.
thickness Stress Stress Stress Stress Stress for for Comb. St Comb. St.
Do (in) tn (in) psi psi psi psi psi Crack Break Crack Limit Break Limit 3030 6.625 0.280 10439 3239 7963 1136 22777 19800 39600 1.1504 0.5752 6" Elbow, IWA(EB-2-2H20) 3045 6.625 0.280 12938 7047 19985 19800 39600 1.0093 0.5047 6" Elbow End 3062 6.625 0.280 21051 8690 29741 19800 39600 1.5021 0.7510 6x3/4 Branch Conn.
3065 6.625 0.280 22977 10114 33091 19800 39600 1.6713 0.8356 Valve 2MS-2051 end 4035 6.625 0.280 8608 3239 9010 1136 21993 19800 39600 1.1108 0.5554 6" Elbow, IWA(EB-2-2H21) 4068 6.625 0.280 9371 12902 22273 19800 39600 1.1249 0.5624 6x3/4 Branch Conn.
4070 6.625 0.280 9556 13596 23152 19800 39600 1.1693 0.5846 Valve 2MS-2052 end 5030 6.625 0.280 11334 10830 22164 19800 39600 1.1194 0.5597 6" Elbow End 5035 6.625 0.280 11949 3239 11514 1136 27838 19800 39600 1.4060 0.7030 6" Elbow, IWA(EB-2-2H22) 5040 6.625 0.280 13096 11093 24189 19800 39600 1.2217 0.6108 6" Elbow End 5045 6.625 0.280 15016 9935 24951 19800 39600 1.2602 0.6301 6" Elbow End 5055 6.625 0.280 11305 9496 20801 19800 39600 1.0506 0.5253 6" Elbow End 5060 6.625 0.280 15806 4545 20351 19800 39600 1.0278 0.5139 6" Pipe 5062 6.625 0.280 22481 8998 31479 19800 39600 1.5898 0.7949 6x3/4 Branch Conn.
5065 6.625 0.280 24660 10575 35235 19800 39600 1.7795 0.8898 Valve 2MS-2053 end 6010 30 0.908 11030 6419 4185 4218 25852 19800 39600 1.3057 0.6528 IWA(EB-1-2H15) 6004 30 0.908 12486 8196 20682 19800 39600 1.0445 0.5223 30" Pipe 6001 30 0.908 12769 9258 22027 19800 39600 1.1125 0.5562 30" Pipe 6000 30 0.908 12785 9631 22416 19800 39600 1.1321 0.5661 30" Pipe 397 30 0.908 12798 9818 22616 19800 39600 1.1422 0.5711 30" Pipe 390 30 0.908 12317 9325 21642 19800 39600 1.0930 0.5465 30" Valve (2MS-2018A) 385 30 0.908 11726 11517 23243 19800 39600 1.1739 0.5869 30" Valve (2MS-2018A) 305 30 0.908 21792 8469 30261 19800 39600 1.5283 0.7642 30x24 Ecc. Reducer 310 24 0.968 18374 12300 30674 19800 39600 1.5492 0.7746 30x24 Ecc. Reducer 375 24 0.968 9791 12590 22381 19800 39600 1.1304 0.5652 24" Elbow 380 24 0.968 12134 13024 25158 19800 39600 1.2706 0.6353 H P Turb. Stop Valve 7000 30 0.908 24700 1349 26049 19800 39600 1.3156 0.6578 30" Tee 395 30 0.908 See Note 39094 19800 39600 1.9744 0.9872 30" Tee 300 24 0.968 24688 8891 33579 19800 39600 1.6959 0.8480 30x3Ox24 Tee 130 30 0.908 18197 7042 25239 19800 39600 1.2747 0.6373 30" Tee 150 24 0,968 20361 8097 28458 19800 39600 1.4373 0.7186 30x30x24 Tee 275 24 0.968 17968 4954 22922 19800 39600 1.1577 0.5788 24" Tee 405 16 0.656 29240 13406 42646 19800 39600 2.1538 1.0769 24x24x16 Tee 550 6.625 0.280 18592 6327 24919 19800 39600 1.2585 0.6293 16x16x6 Tee 555 6.625 0.280 24356 7148 31504 19800 39600 1.5911 0.7956 16x16x6 Tee 560 6.625 0.280 15632 7452 23084 19800 39600 1.1659 0.5829 16x16x6 Tee 565 6.625 0.280 16409 8713 25122 19800 39600 1.2688 0.6344 16x16x6 Tee 1085 6.625 0.280 23729 3271 27000 19800 39600 1.3636 0.6818 16x1 6x6 Tee 1090 6.625 0.280 19161 3016 22177 19800 39600 1.1201 -
0.5600 116x 6 Tee Note. Combined stresses at Node Point 395 are as shown in Attachment D.
8 I
/
4 s
4 1
I-I Y
Z (SOUGIJTH)
> ~
Coo Nj (0 - ;
x~
H E
F E
D c
B B
ACCESSION NO. WE-200098 SUBSYSTEM 30'-EB-I POINT BEACH NUCLEAR PLANT
- LEGEND, ENOTES POSTULATED PIPE LINE BREAK S-DENOTES LINE BREAK NUMBER, A
AUAUTOMATED ENGINEERING SERVICES CORP.
40 SWU.4M80 0100. SUIWE 220. 64P8EFLLE.
I. 60563 0600:
MAIN STEAM SYSTEM, UNIT 2 OUTSIDE CONTAINMENT TO HP TURBINE CONTROL VALVES & TO CONDENSERS LINE BREAKS ISOMETRIC O
Mli NC 00,: O7-Og-OB sr.W08]
NONE siT. ROY.
cw0o:
Io-mo 6 0
,SK-MS-FIG.3 I I
{
I 4
9 2
a 7
1 6
1 5
43 K
Automated
/
Engineering CALCULATION SHEET Page: C1 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Revieived By: Dan Quijano Safety Related Yes N
No ED Date: 9/26/2008 CALCULATION OF LOCAL PIPE STRESSES AT INTEGRAL PIPE ATTACHMENTS The localized stresses developed on the pipe due to integral welded attachments (IWA) are calculated below and shall be superimposed on the pipe stresses calculated by the computer analysis.
As discussed in the pipe stress analysis guidelines (DG-M09, Ref. 10) the specific methodology used to calculate IWA stresses uses as a guide the applicable Code Cases (such as N-318 and 392),
Welding Research Council Bulletins or Finite Element Analysis as appropriate. ASME Code Cases N-318-5 (Ref. 15) and N-392-1 (Ref. 16) are used to calculate the local stresses due to shear lugs and hollow circular attachments respectively since these are the latest versions of the applicable Code Cases. Since the Code does not provide a criteria for analyzing pipe elbows with integral lugs, the method for calculating local stresses due to elbow lugs is in accordance with Reference 17.
The calculations below are not intended to evaluate the IWA's or the weld between the IWA and the pipe, but rather calculate the local stresses for use in evaluating HELB point locations in accordance with Section 4.0 of this calculation. The local stresses from the IWA analysis are added to the piping system stresses (Reference 6) using ASME B&PV Code Section III, Class 2 and 3, Equations 9 plus 10 to obtain the combined stresses and compare them to the threshold stress limits as given in Equations I and 2 in Section 4 (see pipe stress table in Attachment A).
Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Services Corp Attachment C Page: C2 of C20 Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By; Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[1 No D
Date: 9/26/2008 Support No. EB-1-2H3 (Drawing P-407 Sh. 57A-57C):
Geometry Input DO:= 30.0 in t:= 0.908 in L,
0.5.3.Oin
,2 0.5,3.0in KL := 3.6 w:= 0.375n Lo := min(Lt,L 2)
Ld := mar(Li, *2) j:=1..6 R, := 0.5D, Do-t 2
Ll = 1.5in l-2 = 1.5in La := inin(t, f 2)
Lb := min(t,L,)
L, = 1.Sin Ld = 1.5in R.= 5in r = 14.546in L' = 0.908inl 4= 0.908in Mt Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 10 Peak**
W
.2 1
I I 6...
lb 01 o
01 C)
F 01 01 Q2 I11 147921 2
3.
21033 27273 I6 I lb 01 46751 01 ML=
1 1
23 II1 01 0
0 5
1 Ilbi 01 01 el Note: Loads are based on two lugs active (conservative). Shear lugs are welded to the clamp.
MA/Z (MA+MBIZ)
(MA+MBIZ)
(MA+MB/Z)
Mc/Z Psiressj>
1 0d ps23 12 I
I 101 0101 0
P := O.psi Pmax:= 0.psi (Note: Stress due to pressure is included in above pipe stresses)
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C3 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes W]
No El Date: 9/26/2008 Limitations r
Y:=
y = 16.0 2 50 t
1
!L
=
31=0.103 0.5 r
A'ITACHMENT PROPERTIES A 1 := 4.Lr.L2 A 1 = 9in2 OK OK 1-2 f32 0. 103 0.5 OK 131132-0.011 0.075 OK ZIL.,=(ý)(LIL22) 3 2
ZI N := (1)(L I. L,2) 3 ZIL =4.5in 3
ZIN =4.5111 3
STRESS INDICES i:= 1..3 2.2 40 A):=
2.0 0 := 50 -deg 11.8 140)
(0*0 '
0*05 Xo:= -0.45)
Y0 := -055!
Y,-0.75) t,-0.60J X, := X 0 + Iog(131)
-0.991 Xl
'-1.44J k.174)
Y,: O + iog(13 2)
(-0.94)
(-1.591 2i Ilii:= -(XI,COS (0~) +t Yi, sin(oi)) -
A
.,isin(oi) -
1,00j)
Cr := ma{3.82yI6..131.P 2.(,l 1)1.54, 1]
CL:
mnato. 26y 7. P1.1 2.(112)44.1]
CN= Mat0-38y19.01 l. 5 2,6f3)3, 1]
CT = 6.13 CL = 1.18 CN = 1.47 Br: rma{3.Cr, Ij BL: ma{-3. CL, 1)i Bw :=i rn{..CNIi)
Br = 4.087 BL = 1.00 BN = 1.00 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C4 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[N1 No El Date: 9/26/2008 LOCAL STRESSES j:= 1.. 4 B-rW 1,i BLMLI1j Q21,j Stan+
- =
j
+
+
1d AI ZIL 2-.L 2 Lb CT'W 1,5 CL'MLI, Q21.
S+
1,
+
1,5 A,
ZIL 2.1.2 1,b Sp, KL-Sni Sniee:' GrWW 6 + CLML1,6 Q21,6 Al ZIL 2.L.2.Lb Eq. 8 Eq. 9B Eq. 9C Eq. 9D Sm1 =
2 3
4 psi III1 54301 77211 100121 0o Sni = 1716.226psi Range S= 6178.41psi Range Sn1 ee = 0 psi Peak**
MODIFIED CODE EOUATIONS B1 := 0.5 EQ, 8 B2 := 1.0 for straight pipe Bi*. P****+ Pstress
+
EQ. (9B)
B1.+
PstresS
+
2-t 1,2 SInl I = 5430psi Sm112= 7721 psi EQ. (9C)
P. DO B1 3"
.- + Pstress,
+ Sml,3 = 10012psi EQ. (10)
PP Pstress
+ -
= 3089psi 1,5 2
Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C5 of C20 Services Corp Attachment C Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes EZ No EL Date: 9/26/2008 Support No. EB-2-2H9 (Drawing P-407 Sh. 12A-12B):
Geometry Input DO := 16.Gin t := 0.656in L, := 0.5.0.in L2:= 0.5.3.0in KL := 3.6 w:= 0.25in L.:= mir(Lj,L 2)
Ld :=,na,(L,, Q*
R0 := 0.5-Do Do - t 2
L,=0.25in L2= 1.5 in L,=ini(t,L2 Lb=mir(tl,L:)
L, = 0.25 in Ld = 1.5 in k = 8in r = 7.672in La,= 0.656in Lb = 0.25in j:=1-.6 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 10 Peak**
w
=
12
- -]
3:
4 56
'lb 111 18511 26901 35291 ol 5851 Q 2 = 1 1
1 2
4
.3 4
1 5
6:-.
1 0
0 0
0 0
0 lb
?, =I :ýý I
- , I % 2 1I: ~Tý.l4I [ý:
1,ý:: 6
ý I I,.
L,,
111 0
0o o 0o 0o 0o LU" ILL MA/Z (MA+MS/Z)
(MA+MB/Z)
(MA+MB/Z)
Mc/Z I.
I I
Pstrcss=
[ ::*! *:i!.
- 2 1
2 3:ý!-:
I I
I I
i1 4
- psi 00 oj 01 0o P:= 0. psi Pmax:= 0.psi (Note: Stress due to pressure is included in above pipe stresses)
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C6 of C20 Services Corp Attachment C Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes No No El Date: 9/26/2008 Limitations r
y:=~
y= 11.7
[31 :=--
[31 =0,033 A
- =( 4Lr 1
L 2 AI 50 OK 0.5 OK 1.5in 2
= 0.75in 3
= 0.125in 3
i.
Pr1P[2 = 0.006 02 = 0.196 0.5 OK
< 0.075 OK STRESS INDICES i:= L..3 2.2 (40 A,:=
2.0 0:= 50.deg 1.8
.40) 0.0 r 0.05 Xo:=
-0.45 Y :=
-0.55
-0.75)
-0.60)
"-1.49' X1 := Xo + log(P 1)
X1 L 1.94 Y := Y + log(0 2)
( -2.2 4, CT:= n-3.82.y
.64 0 2+11)1.54,]
CL := maTO,26y' 74. *1j0q2 ()474,1 CN:=maO.39yl9. 31.p 2 2(,
3).4 1]
(-0.66' Y=
-1.26
- -1.31)
BT ma
.cr, )
BT = 1.657 BL = 1,00 BN = 1.00 Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Services Corp Attachment C Page: C7 of C20 Calc. No.: PBNP-994-21-05-P02 i
Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes El No 0
Date: 9/26/2008 LOCAL STRESSES j:= 1..4
ýBT'.W I~
BL.MLI Q21,j Smlj A1 n,j 2.LLbo Snl
+ C
+MJ A l ZIL 2"L2"Lb Spi := Kf; SnI CT'W 1, 6 CL.ML 1,6 Q21,6 Snlee~
A-
+
+
A f ZIL 2.I_2.1 MODIFIED CODE EQUATIONS Eq. 8 Eq. 9B Eq. 9C Eq. 9D I 1 20451 2972 3
0 psi 29721 38991 0
Sj=969.539psi S,=3490,34psi 5n1 e 0 psi Range Range Peak**
B, := 0.5 EQ. 8 B2 := 1.0 for straight pipe P.Do 131,-
+ Pstress
+ 5in1
= 2045psi EQ. (9B) 1--
+ PStresS2 + S,2=
2972psi 2.t 1,2 Sfl1,2 EQ. (9C)
'1 3-
+ Pstress
+ Stl,
= 3899psi 2E t 1,3 1,3 EQ, (10)
Pstress 1
+
I = 745psi 15 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C8 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-1 1 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[R]
No El Date: 9/26/2008 Support No. EB-I-2H15 (Drawing P-407 Sh. 58A-58C):
Geometly Input Do := 30.0in t:= 0.908 in L, := 0,5.3.0in L2:= 0.5.4.Oin KL :=3,6 w := 0.375n Ld:= main(LIL 2)
Ld := rna4L,, L,)
R,,:=0.5-13, 2
L2= 2in Lb := mi:(t,L12)
Ld 2in Ro = 15i r = 14,546in l" 0.908inl Lb = 0.908in j:= 1..6 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 10 Peak**
W=111 1121314,3
.5:
k@! 6
(
- 1111 lb Q* =
LiI 14 01 o1 of 0
0 fj 1 2'
3 45 6VII
- ii2lb
-= 1 170021 233141 296261 01 8510 L 0
ML =
1 2..
6 L
-I 1
o
° 01 0o 1
lb-in Note: Loads are based on two lugs active (conservative). Shear lugs are welded to the clamp.
MA/Z (MA+MB/Z)
(MA+MB/Z)
(MA+MB/Z)
Mc/Z PýStress 4
4 ps I2i I ;Ii 0I o0 0
Pmax:= 0 psi (Note: Stress due to pressure is included in above pipe stresses)
P :=.psi Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Services Corp Attachment C Page: C9 of C20 Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[Z No M
Date: 9/26/2008 Limitations r
y-t L.
y = 16.02
< 50 OK 31 = 0.103 0.5 OK I.1 I1i.32 = 0.014 02 = 0.137 0.5 OK 0.075 OK A'TTACHMNT PROPERTIES A, := 4,LIL 2 A,
-12in 2 ZIL:= (4 )(LI'L22)
ZIL = 8in3 ZIN :=
?J(L12.L)
ZIN = 6in 3 STRESS INDICES i:= L..3 2.2' 40",
Ao:=
2.0 0:= 50 deg S1.8,
- 401 (0.0 (0.05'"
X. := -0.45 Yo:= -0.55
- -0.759
-0.60)
X 0
=X + iog(p 1)
(-0,99"1 Xi =' -1.44
\\-1,74)I Y, := Yo + log(3 2)
-0.81 Y1
-1.41
[-1.461 2
12 11(x=.,cos(0i) + Y,.sin(0i)) - (-).}(xi.sin(e~) - Y, COS(0i)) 2 maT
=i43.82y 1.64. I3IP2. (,n,)1.54,,]
CL := naf O.-26y
. 1.PI-02.,2)4.74 CN := MaT0.38y ~.f 01.P2'(113)3.4,,
Gr ý 7.467 CL = 1.64 CN = 1.75 Br := ma{-3.C. I)
BL := M4{-3CL, 1)
BN := ma{ -3cN.I 1
Br = 4.978 BL = 1.09 BN= 1.17 Form 3.1-3 Rev. 2 Form 3. I-3 Rev. 2
/Im Automated Engineering CALCULATION SHEET Page: C10 of C20 Services Corp Attachment C Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes El No 11 Date: 9/26/2008 LOCAL STRESSES j:= 1..4 BT'WI,
+ B+ML
,j Sm1lj A, --
+
+
Ij A 1 ZIL 2'L2*L*
Eq. 8 Eq. 9B Eq.9C Eq. 9D Smt 1
2
- 4.
ps
-1 4681 6419 81 0
CTrW 1, 5 CL'ML1,5 Q21,s An1
+
+
A, ZIL 2"L2"Lb Spl:= KL.Snl S
crW1,6
+
1'MLI,6
+
1,6 AI ZIL 2"L2'Lb S,,, = 2343.062psi p,= 8435.02psi Sni ee = O Psi Range Range Peak**
MODIFIED CODE EQUATIONS B, :=0.5 EQ, 8 B2 := 1.0 for straight pipe P. Do BI'-
+ Pstress
+ Smnil 4681 psi EQ. (9B)
BI1. -"D + Pstrcss
+ Sm,
= 6419psi 2-t 1,2 1,2 EQ. (9C)
P' D,
+ pstress
+ Sml
= 8157psi 2-t 1,3 1,3 EQ. (10)
Pstress 1
+ L-
= 42l8psi Form 3,1-3 Rev. 2
/%M Automated Engineering CALCULATION SHEET Page: C11 of C20 Services Corp Attachment C Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes IX]
No El Date: 9/26/2008 Suimort No. EB-2-2H18 (Drawinix P-407 Sh. 17A-17B):
Geometry Input D,:= 16.0in t := 0.656 in L,
0.5.0,5in L2 0.5.3.Oin KL 3.6 w := 0.25in L,:= min( 1,-,L2)
Ld:= ma(Li,L2)
R, : 0.5-D.
2 L, = O.25in L2 15in L4,:= Miri(t,L1)
L,,= 0.25ini Ldj = 1.5in R. = 8in r =7,672in La = 0.656in Lb = 0.25in
_Pý,e Q2 j:= 1..6 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 9D Peak**
4 5
6 W
1 2
3 lb Ii.
18971 29611 40251 01 14981 01 1
2 3
1 4:"::
5 b
Q2 = [II 1 o1 011 o1 01 01 ML= I :I 1 2
3 4 4 516 I Ih-in 111 0o 0o 1
00 0
0 Ih.;n MA/Z (MA+MB/Z)
(MA+MB/Z)
(MA+MB/Z)
Mc/Z pstress = I *i:!l*I!':
+ -.
ý2 3
I I1 0
0 0
0 01 0o si11 P := 0.psi Pmax:= 0-psi (Note: Stress due to pressure is included in above pipe stresses)
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C12 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes RZ No El Date: 9/26/2008 Limitations yl:= r y
11.7 50 0
t L,
31 := -
131=0.033 0.5
(
r AY1'ACHMENT PROPERTIES A1 := 4.LL, 1 2 A, = 1.5in2 K
)K P2=-
132 =0.196 0.5 OK 31132 =0.006 0.075 OK ZIL:=(
(L.
,2 ZL =0.75in3 ZIN := (1)(1,212)
ZIN = 0.125in 3 STRESS INDICES i:= 1.. 3 2.2]
¢=40]
0.0 "
=
0.05 A:= 2.0
- =
50 *deg Xo:= -0.45]
-0.55 L.8 40
-0.
-0.60
-1.49' x
0
=X,+ iog(i3 1) 1 1i.94J Y, := Y, +i Iog(r32) q:=- (X1.*Cos (0 ) + Yi
- sit(00) -. L}X, isii(o~) - Yi, Cos(()i))2 CT: M4 382y 1.64, V
2,(11 1)1.54,1]
Cl rnain 26y
. 01-P 12.(,2)4.74, i]
CN: Mafa,13&y 19.P132.3 2 '(113)34 1]
0T = 2.5 CL = 1,00 CN = 1.00
-0.66)
Yj =
-1.26
-1.31) q=2 2/
Br := ma{.-3 C'r. t)i BL := ma{-.PCL, I 1N=ma(_3 C14 1)
BT = 1.657 BL = 1.00 BN = 1.00 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Services Corp Attachment C Page: C13 of C20 Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[*]
No 0l Date: 9/26/2008 LOCAL STRESSES j:= 1.4 3T'W 1.
BL'MLI+
Q2 tj
+
+
I,j A1 ZIL
+
+
Al ZIL 2"L2.Lb Spi := KL.Snl CT'W 1,6 CL"MLl,6 Q21,6 Sne.-
,6 1, +
Al
+
ZIL 2-L2"Lb MODIFIED CODE EQUATIONS BI:= 0.5 B2 := 1.0 for straight pipe EQ. 8 P.Do Br--
+ Pstress
+ Sin
= 2096psi 2.t 1,1 Srl,I EQ. (9B)
P-130 B1 ---
+ Pslress
+ Sm,2 = 3272psi 2-t 1,2 f,
EQ. (9C)
P. D0 B--
- + Pstres
+ SMI
= 4447psi 2-t 1,3 1,3 EQ. (10)
Sp1 Pstress,5+ --
= 4469psi Eq. 8 Eq. 9B Eq. 9C Eq. 9D 11 1 20961 32721 44471 01 s
S,, 2482.683psi pI8937.66psi 5fl1 ee O psi Range Range Peak**
Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATI Services Corp Attachn Client-Florida Power & Light ON SHEET ient C Page: C14 of C20 Calc. No.: PBNP-994-21-05-P02 Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-1 1 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes IJ No D Date: 9/26/2008 Suppolt No. EB-1-2H14 (DrawingP-407 Sb. 37A-37B):
Geonietry Input D,:= 24.0 in t :=0.968 in L, : 0.5-0.75in L.2:
0.5 7.Oin KI,:= 3.6 w: 0.25in Ld rna(LI, L2) j:= 1.. 6 Ro := 0,.SDo, D, - t 2
L1 = 0.375in L2 = 3.5 in L,,:= min~t, L2)
Lb:= min(t,L 1)
L¢ = 0.375in Ld = 3.5in
%0 =l2in r = 1I.516in L,
0.968in Lb = 0.375in t
Ru!n -jpip Q2 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. '10 Peak**
1 31914 31914 31914 0
0 0
1 2 3 5
6 b
I.1:1 01 cii 01 01 01 0
M,-=I :I.*LI
ý :ý:
I ý:
5 I
6 l Ill cii 01 01 01 0,
0 MA/Z (MA+MB/Z)
(MA+sMB/Z)
(MA+MB/Z)
Mc/Z pstress I I
,IIpsi I 1V 01 01 01 0i P := 0.psi Pmax:= 0. psi (Nbte: Stress due to pressure are included with the pipe system stresses)
Form 3.1-3 Rev. 2
aV Automated Engineering CALCULATION SHEET Setvices Corp Attachment C Page: CI5 of C20 Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No 0
Date: 9/26/2008 t
PI =L Y= 11.9 50 OK P2
[3r-P2 =0.01 P32 = 0.304 0.5 OK 31 = 0.033 0.5 OK 0.075 OK ATTACHiMENT PROPERTIES A : 4,LI.L 2 A1 = 5.25in 2
ZIL := ('(IL)
Z11, = 6.12in 3 ZIN= (4 (LI2.L,2)
ZN= 0,656in 3
STRESS INDICES i:= 1.. 3 2.2 40")
Ao:=
2.0 0:= 50.deg
.1.8 40) 0.0 (0.05 xo:= -0.45
-Y:=
_0.55 r-0.75)
-0.60)
_1.49' Xl:=X, + log( p1)
= -1.94
- = Y + og(13 2)
- -2.24)
Ili: -(xi.COS(of) + Y,.sin(o i)) - (-L}.X (x 1.s(0
- 1) - Ycos (a.))
-0.47 YJ
-1.07
-1.12) 1.277" 1.744 2.244, CL:=ma4jO26y I0 l.71 132()4.74 1 ]
CNq-.
01140.39y19. N 12 PT (,,3~. 4, CT = 3.198 cL = 1.00.
CN = 1.00 BI-:= ni4
{ 3. 01,I )
BL:m ra{.
CLi BN=Ma{-3CN, I)
BT = 2.132 BL = 1.00 BN = 1.00 Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
/l IAutomated Engineering CALCULATION SHEET Page: C16 of C20 Services Corp Attachment C Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steamn Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No ED Date: 9/26/2008 LOCAL STRESSES j:= 1..4
=BT'W,j+
BL'Ml,+
Q21,j
.-1
++
1,j A1 ZIL 2"L2',Lb CT'WI, 5 CL'MLI,5 Q21,5 AI ZIL 2'1L2"Lb Spl:= KLSnl CT'W 1,6 CL'ML,6 Q21,6 Snl ee:-
+
+ I AI ZIL 2-L2,Lb MODIFIDT) CODP EOIFJATIONS Eq. 8 Eq. 9B Eq. 9C Eq. 9D 1
2 1
12958 12958
- -3 7; ':'.. :4 3
psi 12958 0
Snl = Opsi Sp, = Opsi Snle = 0 psi Range Range Peak**
B2 := 1.0 for straight pipe BI := 0.5 EQ. 8 P-Do BI-
+ Pslress
+ Sia
= 12958psi EQ. (9B)
BD,
+ Sm 12958psi B4-
+ Ps1ress1,2 1,2 =
EQ. (9C)
Bj
+ Pstress
+ SW
= 12958psi 4.t 1,3 1,3 EQ. (10)
Pstressl 5 + SP' = Opsi 2
Form 3.1-3 Rev.2 Form 3.1-3 Rev. 2
/Automnated Engineering CALCULATION SHEET Page: C17 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes Q]
No
[L Date: 9/26/2008 SSuonort No. EB-2-21-I1 0 2H111. 21-112- & 21-113 (Drawirng P-407 Sh-I IA-1 3M, Geomnett, Input Do:= 6.625in R := 9.000 in D := 3.000in t := 0.280 in R.:= 0.5-D.
0:= acos ( R D
L:= sin(0)
R,= 3.3l2iii r:= 3.173 in 0
43.Odeg L =4.396in Wieg : 0.25 in KL := 3.600 j:= 1.. 6 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 10 Peak**
W=1 23011 31141 39271 01 6431 o0 Ilb MA/Z (MAi+M B/Z)
(MA +MB/Z)
(MA+MB/Z)
MCZ Pstress =
2 3
psi I1 0
0 0
0 0
P := 0- psi Prax:= 0 psi (Note: Stress due to pressure are included with the pipe system stresses)
Local Stresses L05.r 1.25 K:=
t 1.75 K = 82
> 15
& < 280 O.K.
0.97-W-K 0.9 a
1,r Lr1 +0o.2i!~
Eq. 8 Eq. 9B Eq. 9C Eq. 9D Range Peak UL.,nb =1 5325.91 7207.71 9089.41 0 1 1488.31 0 P si 1.1&W.K0.77. r + 0.47(r't)0.5 Eq. 8 Eq. 9B Eq. 9C Eq. 9
- C.mnb :-
I I
CFC.,nb = 1 6399.61 8660.8 1 10921.9 1 D
Range Peak 0 1 1788.31 o-- psi j," I Pornh 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
/U Automated Engineering CALCULATION SHEET Services Corp Attachment C Page: C18 of C20 Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[N No El Date: 9/26/2008 SMj 1 2 ma G.L.mb t, I¢Cmb I i aa(Lmb1,a C'mb 1,5)
Sp, KL-Snt S
ln
- = maý*(UL'mb 1,6' 9C.mb 1,6)
MODIFIED CODE EOUATIONS Eq. 8 Eq. 9B Eq. 9C Eq. 9D Sint ý 42661 57741 72811 0 1 psi Sý,= 178 Spsi Sp, = 6438psi S",.OO0psi Range Range Peak B2 := 1.0 for straight pipe B1 := 0.5 EQ. 8 B 1-- -
+ Pstress
+ S1l1
= 4266psi 2-t 1,1 11 EQ. (9B)
P-Do Bj.-
+ Pstrcss
+ Smn,
= 5774psi 2-t 1,2 1,2 EQ. (9C)
P.Do B1. -,
+ psress
+ 5n,1
= 7281 psi 2-t 1,3 1,3 EQ. (10)
- Psressl,
+-
= 3219psi 15 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: C19 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes.
Q No EL Date: 9/26/2008 Support No. EB-2-21HI9 2H20, 2H21, & 21-122 (Drawing P-407 Sh. 18, 19A-19B):
Geometry Input DO := 6.625 in R := 9.000 in D := 3.000in t:= 0.280in KL := 3.600 Ro := 0.5.)D 0:=acos (
R)
R+ R, D
sin(o)
R, = 3,312in r:= 3.173in 0 = 43.Odeg L = 4.396in Wleg := 0.25 in j:= 1..6 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Eq. 10 Peak**
W=1 12321 17471 22621 01 2271 0 1 Ibt MA/Z (MA+MB/Z)
(MA+MBIZ)
(MA+MB/Z)
Mc/Z estress
- 5,
psi S
0p 0N 0
0d 0
P := 0-psi Pmax:= 0.psi (Note: Stress due to pressure are Included with the pipe system stresses)
Local Stresses L0.51.25 K:=
1.75 t
K = 82
>15
& < 280 O.K.
0.97.W.K 0.8 L.~m I.r (1 +0.2;1 Eq. 8 Eq. 9B Eq. 9C Eq. 9D Range Peak l-mb = 1 2851.61 4043.6 I 5235.6 I ol 525.41 ol psi 1.ISW-K0.77. I + 0.47.(r't)051 Eq. 8 Eq. 9B Eq. 9C GC.mb :=
I.
I aCrmb = I 3426.51 4858.8 1 6291.2 1 Eq. 9D Range Peak 01 631.31 01 psi r-- "
Form 3.1-3 Rev. 2 Formn 3.1-3 Rev. 2
Automated 7,
Engineering CALCULATION SHEET Page: C20 of C20 Services Corp Attachment C Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[X]
No ED Date: 9/26/2008 i:= 1.4 Eq. 8 Eq. 9B Eq. 9C Eq. 9D 2
i SmI 3 ma4(CL.mb
,C.mbI S, ma4 CrLflb 1 5 ~Cymb 15 sK,.
Spl :
KL-Snl Sni ma*(OLjnb 1 6 ' CrC.mb 1,6)
MODIFIED CODE EOUAT[ONS SmIj 22841 32391 41941 01 psi S,- 63 ]psi s =2273psi S",= O.OO0psi Range Range Peak B1 := 0.5 EQ. 8 B2 := 1.0 for straight pipe P*D.
Bl.
+ Pstress
+
+ S
=
1 2284psi 2.t tall,!
EQ. (9B)
P-Do BI.
1-
+ Pstress
+ StM
= 3239psi 2-t 1,2 1,2 EQ. (9C)
BI-
-- + Pstress
+S,3
= 4194psi 2-t 1,3 1,3 EQ. (10)
Pstress
+I, -
= l136psi IS 2 Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: DI of D2 Services Corp Attachment D Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes NI No El Date: 9/26/2008 CALCULATION OF COMBINED STRESSES AT TEE NODE POINT 395
Purpose:
Calculate the combined stresses at tee Node Point 395 for use in evaluating HELB point locations in accordance with Section 4.0 of this calculation.
Method:
The stresses at tee Node Point 395 are recalculated-to remove some of the conservatism in the combined stress of Reference 6 by following Section NC-3651 (a) of the ASME Section III Code, 1986 Edition. The pressureportion of Equation 9 is replaced by the expression providedin the calculations below. All inputs are obtained from the pipe stress analysis (Reference 6).
Calculations:
Deadweight and thermal moments for Tee Node Point 395 are from (Reference 6), The OBE stress is obtained by subtracting the equation 8 stress from equation 9 stress provided in the stress summary of Reference 6.
D:= 30. in t:= 0.908 in D, := D t D, = 28.184in P:= 1085psi Section Modulus:
. - D14) 32-D i:=2.131 B1 := 0.5 B2:= 3.18 Z = 585.87in 3 (Ref. 6)
M~d,~:= 5059ft~lbf Mt,=154360ft~lbf U
c,:
aq9 -
ucq8 Mxd,:= 14021ft~lbf Mt,=472980ft.lbf GeqS :=~ 1023 Spsi Myd, := 2496ft. IV yt,=435S ft. lbf crcq9:= 18484psi Gobe = 8249psi Form 3.1-3 Rev, 2
/U Automated Engineering CALCULATION SHEET Page: D2 of D2 Services Corp Attachment D Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No El Date: 9/26/2008 P.(D - 2.t) 2 D2 - (D - 2.t)2 MA:=
Mxdw + MydW 2+
Mzd,2 MA adw:
132-M z
Mc=
2 + Myt2 + M2 alp = 8157psi MA = 14908ftlbf cadw = 971psi Mc = 497550RbIbf Ot Mc Z
at[) = 21717psi The combined Equation 9 and 10 stresses at Node 395 for use on the stress table of Attachment A are:
Gcjh=ap+ adw, + Gib2 + ab (7(,b= 39094psi Form 3.1-3 Rev. 2 Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: El of E3 Services Corp Attachment E Caic. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes 1Z No
[]
Date: 9/26/2008 Review of Non-Seismic Supports Loads:
Support No. Load Global Forces (lb)
(Point #)
Combination Fx Fy Fz Comments Deadweight
-3433 Thermal
-1290 Analyzed Load 6000 lbs OBE 820 Maximum Upset Load = 5543 lbs < 6000 lbs O.K.
EB H5 (480)
Normal Min
-4723. '-.
Normal Max
-3433 Upset Min
-5543 Upset Max
-2613 Deadweight
-2551 Thermal 341 Analyzed Load = 2100 lbs OBE 525 Maximum Upset Load = 3076 lbs > 2100 lbs EB-2-2H6 Normal Min
-2551 Minimum hanger item capacity = 4300 # O.K.
(490)
Normal Max
-2210 Aux. steel check: fb*=7.34x1,46= 10.75 ksi < 21.6 ksi Upset Min
-3076 O.K.
Upset Max
-1685 Deadweight
-3277 Analyzed Load = 3600 lbs Thermal
-52 Maximum Upset Load = 3954 lbs > 3600 lbs OBE 625 Minimum hanger item capacity = 2385 # > 2350 # O.K.
EB-2-2H7 Normal Min
-3329
[3954# x (3.66716.167)]= 2350 #
(525) ~
~
~
3277.,.
Normal Max 3277 Channel check: fb =13.96xl.10= 15.33 ksi"< 20.62 ksi Upset Mmn
- ii!!
-3954
.:::::::.-O.K.
Upset Max
-2652 Deadweight
-5937 Thermal 633 Analyzed Load = 5600 lbs OBE 1202 Maximum Upset Load = 7139 ibs > 5600 lbs EB-2-2H-8....
Normal Min
-5937 Minimum hanger item capacity
-3770 3570 # O.K.
Normal Max
-5304 Trapeze hanger (7139/2 = 3570#)
Upset Min
-7139 Upset Max
-4102 Deadweight
-1029 Thermal 353 Analyzed Load = 1200 lbs EB22H0OBE 315 Maximum Upset Load = 1344 lbs > 1200 lbs (6005)
Normal Min
-1029 Minimum hanger item capacity = 2710 # O.K.
Normal Max
-676 See Note 2 Upset Min
-1344 Upset Max
-361 Deadweight
-2206 Analyzed Load = 2300 lbs Thermal 377
!:___:: Maximum Upset Load = 2806 lbs > 2300 lbs OBE 600 Minimum hanger item capacity = 2710 #
EB-2-2H 111 (702)
Normal Min
-2206 Considering factor of safety (FS) = 2 and typical Normal Max
-1829 catalog FS = 5, therefore, use 2.5 x catolog capacity Upset Min
-2806 6775 # > 2806 #
O.K.
Upset Max
-1229
___:___ See Note 2 Form 3.1-3 Rev. 2
CALCULATION SHEET Page: E2 of E3 Attachment E Calc. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
Title:
Main Steam Piping GL 87-1 1 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes El No
[
Date: 9/26/2008 Support No. Load Global Forces b)
(Point #,).Fx Fy Fz Comments Deadweight
-1132 Thermal 509 Analyzed Load 1200 lbs OBEE13-2-2 418
__:__=:
Maximum Upset Load = 1550 lbs > 1200 lbs (02)H Normal Min 1132 Minimum hanger item capacity= 2710 # OK.
Normal Max
-623 See Note 2 Upset Min i
-1550 Upset Max 205:::.::..:::::.:.::_
-205_:_:_:_:_:_
Deadweight
, -2301
== Analyzed Load = 2400 lbs Thermal 643 Maximum Upset Load = 3114 lbs > 2400 lbs EB-2-2H13 OBE.....___
813 Minimum hanger item capacity = 2710 #
Normal M3
-2301
- Considering factor of safety (FS) = 2 and typical (902)
Normal Max
-1658 catalog-FS = 5, therefore, use 2.5 x catolog capacity Upset Min
-3114 6775 # > 3114 #
O.K.
Upset Max
-845 See Note 2 Deadweight
-3227 Analyzed Load = 12000 lbs Thermal
-8736 Maximum Upset Load = 13336 lbs > 12000 lbs EB-2-2H15 OBE 1373 Minimum hanger item capacity = 11630 #
025)
Normal Mn
-11963 Considering factor of safety (FS) = 2 and typical (1
Normal Max
-3227 catalog FS = 5, therefore, use 2.5 x catolog capacity Upset Min
-13336
= 29075 # > 13336#
O.K.
Upset....
-1854 See Note 3 Deadweight 0
Thermal 1
.....:. k..+.
EB-2-2H16 OBE Normal Min 0
Analyzed Load = 2500 lbs O.K.
(1060)
Normal Max 1
Upset M in Upset Max 2
Deadweight
-3492 Thermal 5468 Analyzed Load 6200 lbs OBE 3085
- Maximum Upset Load = 6577 lbs > 6200 lbs E13-2-2H17 (1075)
Normal Min 3492 Minimum hanger item capacity 6840 # O.K.
Normal Max 1976 "See Note 5 UpsetMin 6577 Upset Max 5061 Deadweight
-1897 Thermal
-1498 Analyzed Load = 3000 lbs EB-2-2H18 OBE 1064 Maximum Upset Load = 4459 lbs > 3000 lbs (2000)
Normal Min
-3395 Minimum hanger item capacity = 4960 # O.K.
Normal Max 1897 Aux. Steel and connections have significant margins Upset Min
-4459 Hanger is adequate for maximum upset load.
Upset Max
-833 I_
Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: E3 of E3 Services CotrP Attachment E Cale. No.: PBNP-994-21-05-P02 Client:
Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
Title:
Main Steam Piping GL 87-11 Break Location Determination Reviewed By: Dan Quijano Safety Related Yes
[]
No E Date: 9/26/2008 Support No. Load (Point #)
Combination Global Forces (IbL Fx Fy Fz Comments Deadweight
-417 Thermal 186 Analyzed Load = 650 lbs OBE 463 Maximum Upset Load = 880 lbs > 650 lbs
(.2035)
Normal Mi
-417 Minimum hanger item capacity = 2710 # (lug) O.K.
Normal Max T2311 Support designed for uplift O.K.
Upset Min
-880 Shell anchors are designed for 1400# > 880 # O.K.
U pset M ax 232 Deadweight
-1232 Thermal
-227 Analyzed Load 1700 lbs EB-2-2H20 OBE 515 Maximum Upset Load = 1974 lbs > 1700 lbs Normal Min
-1459 Minimum hanger item capacity 2710 # (lug) O.K.
Normal Max
-1232 See Note 4 Upset Min
-1974 Upset Max
-717 Deadweight
-355
_=___..
Thermal
-51 AalyzedLa=90lb OBE 277
"_:__:_ Maximum Upset Load = 683 lbs < 900 lbs O.K.
Normal Min
-406
- -:.:..: See Note 4 (4035)
Normal Max
-355 Upset Mi
-683 Upset Max
-78 Deadweight
-1190 Thermal
-270 Analyzed Load 1700 lbs OBE 503 Maximum Upset Load = 1963 lbs > 1700 lbs (5035)
Normal MI
..... -1460 Minimum hanger item capacity = 2710 # (lug) O.K.
Normal Max
-1190 See Note 4 Upset Min
-1963 I Upset Max
-687 Where: Normal Min = MIN(DW, DW+TH)
Normal Max = MAX(DW, DW+TH)
Upset Min = Normal Min - OBE Upset Max = Normal Max + OBE Notes:
- 1) Loads obtained from Shaw Group pipe analysis (Ref, 6), considering EPU conditions, are used in the load combinations.
The S&L support qualification (Ref. 7) is used as the basis for the review of these non-seismic supports.
- 2) Per Ref. 7 supports EB-2-2H10, -2H11, -2H12, and -2H13 are ganged to a single frame. Frame stresses, fb= 9.0 ksi Considering worst increase of 30% for hanger EB-2-H13 (3114/2400 = 1.30), the frame bending stresses are:
fb = 9.0 ksi x 1.3 = 11.7 ksi < 16.0 ksi O.K. Also, connection welds and 6" channel inserts are adequate.
- 3) Per Ref. 7 support EB-2-H15 is attached to auxiliary steel. The aux. steel stresses increase by 11% to 25.1 ksi.
This is still well below the operability allowable of 40.6 ksi. Connection welds are also adequate.
- 4) Per Ref. 7 supports EB-2-2H20, -2H21, and -2H22 are ganged to the same frame. A single frame is installed to take the uplift forces for all supports. The down load is supported separately by each hanger. Since there is no uplift, the supports are evaluated separately as shown above.
- 5) Support EB-2-2H17 is a non-seismic support (rod hanger) and cannot withstand uplift forces. The upward load of 5061#
will be distributed to adjacent supports EB-2-2H1 5 and EB-2-2H1 8 which are loaded downwards and can accommodate the subject uplift load.
Form 3.1-3 Rev. 2
TECHNICAL POSITION PAPER FOR ESTABLISHING HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Rev. 0, December 4, 2006 Rev. 1, August 7, 2008 Co.,k, NO. P'BNP-1174 -2l-1 P02,ReWv 1.0 Introduction CA.
-tJo-.
F', PoiOe f 2,
Point Beach currently utilizes different Pipe Break Outside Containment (PBOC) location selection criteria in the HELB Program and EQ Program in regards to environmental parameters. The HELB Reconstitution Program (Program), as currently envisioned, will prepare documented calculations for the pressure, temperature and humidity time histories for a variety of HELB scenarios. Since the Program will reconstitute the design basis for PBOC and the resultant event enviromnent outside Contaimnent, these environmental parameters would be equally applicable and used as the input to the EQ Program. With this approach PBNP will have a single unified HELB approach to address impacts on EQ and structural effects including compartment pressures and temperatures, jet impingement, pipe whip among others Before proceeding with the Program, a major consideration needs to be addressed and agreed upon by PBNP. This involves the adoption of Generic Letter 87-11 and its associated NRC Mechanical Engineering Branch Technical Position, MEB-3-1, Revision 2. Considerable discussions have taken place in the past on the extent and use of GL 87-11 and its associated MEB 3-1, Rev. 2. Currently, PBNP EQ Program uses a variation of the MEB 3-1 document involving the use of the combined stress threshold for break location of 0.8(1.2 S1, +SA) to establish the EQ parameters (Reference 4, 5). It is noted that Revision I of MEB 3-1 stipulates the above break location threshold limit.
The PBNP FSAR Appendix A.2 (Reference 2) states "Break locations are selected in accordance with Reference 1. Consideration of arbitrary intermediate pipe ruptures is no longer required per NRC Generic Letter 87-11." The Reference 1 stated in the foregoing quotation is the Giambusso Letter of December 19, 1972. The Giambusso criteria included the threshold limit of 0.8(Sh+SA) and other requirements.
PBNP HELB DBD T-47 (Reference 6) provides a detailed discussion of the background history for the break location criteria. Without repeating these details, it is appropriate to state that the HELB location criteria have evolved over the years and there is a realization that these sets of criteria are "non-mechanistic" in nature. In other words, even though the pipe is designed to all design and analysis rules, additional precautions were imposed to provide added assurance for designing the plant SSCs against postulated pipe breaks. To provide a basis for establishing break locations, the AEC and NRC staff promulgated rules that tied these location selections to the stresses in the piping system. As the ASME Section III Piping Code equations (specifically Equations 9 and 10) (Reference 10) evolved so has the break location stress threshold limits. These changes in the break location criteria have led to the numerous discussions cited in the HELB DBD and the differences in the criteria used in the EQ Program and the FSAR citation.
K,,. Automated Engineering Services Corp
HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Page 2 (oJgcýto. PBPJP-9q7-V1-05-P02 Re 1 0 2.0 Line Characterization Criteria and Break Selection Rules A,-ot,4 V )Po* e*r.Z It is noted that the criteria for the identification of HE lines outside containment (Design pressure>275 psi and service temperature > 200'F) and the fact that the current licensing bases of most vintage plants, including PBNP, do not recognize moderate energy lines, are separate and distinct criteria that should not be linked to the break location selection. In other words, changes to the HE break location selection criteria do not automatically require the re-visitation of the criteria for high and moderate energy line characterization. In fact, SRP 3.6.2, GL87-11 and MEB 3-1, Rev. 2 do not address the line characterization criteria, which is reviewed in SRP 3.6.1.
Since the line characterization for line breaks remains the same as stated in the FSAR, the section of MEB 3-1, Rev. 2 pertaining to moderate energy lines do not apply since the PBNP licensing basis does not characterize lines in this category. Similarly, the HE line definition for PBNP remains unchanged and only the lines that satisfy the "and" criteria and the "normally depressurized" rule need to be included in the HELB Program.
3.0 Proposed Unified P13OC Criteria for the PBNP HELB Reconstitution Program The following criteria for the Pipe Break Outside Containment (PBOC) are proposed for the HELB Reconstitution Program. Adoption and use of this set ofcriteria will be across all PBNP Programs (EQ, HELB and others).
3.1 Retain the definition that all lines outside containment are designated as ASME Section III Class 2 and 3 as stated in the FSAR, Appendix A.2 and DG-M09 (Reference 9).
3.2 Retain the current definitions for HE lines, which does not require the characterization of lines for moderate energy.
3.3 Adopt the use of GL 87-11 and MEB 3-I, Rev. 2 rules for HE lines only including the rules for break and leakage crack location selection in their entirety. These rules utilize the 1986 ASME Code Equations 9 and 10 with the use of stress indices for dead weight and OBE resultant moments (B2 indices) and longitudinal pressure (BI indices) and stress intensification factors (i) for thermal expansion only. It is noted that the pipe stress analyses compute the resultant moments for the load cases. These resultant moments are independent of which Design Code is used, The code equations or in this case of establishing the break locations, the combined stress equation are computed from the stress resultants based on the specific formulations.
3.4 In addition, IN 2000-20 (Reference 7) clarifies the requirement of postulating a single open crack at the location most damaging to those essential structures and systems.
3.5 Types of breaks and cracks should be in accordance with the MEB 3-1 Section B.3.
3.6 When break criteria are based on stress calculation, it is recommended that breaks and cracks be based on the calculated stresses (Section B. I.c(2)(b)(ii)) and not at each pipe fitting (Section 13.1.c(2)(b)(i)) of MEB 3-1, Rev. 2. The stress requirement of Section B. 1.c(2)(b)(ii) should be based on the primary piping stress evaluation (Section NC/ND-
.,Qi.,* AutoinatedEnglneering Services Carp
HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Page 3 3653 of the ASME Code Section III) and local stresses at the integral welded attaclhments(IWA), where applicable.
3.7 Where breaks locations are selected without the benefit of stress calculations, it is recommended that breaks be postulated at the piping welds to each fitting, valve, or welded attachment.
cec, wo. PBO P-I994 O05PO21 Rev. o 4.0 Regulatory & Licensing Issues PopcLCL.
F; I0.X e F3 Use of the MEB 3-1 equations to determine break and crack locations does not require prior NRC approval. The 50.59 process and changes to the FSAR would be required.
In order to be compatible with the activities previously performed for the EQ Program, a 50.59 Screening/Evaluation should be performed to accept the use of the of the proposed PBOC criteria for determining break and crack locations.
The proposed PBOC criteria has the potential of eliminating all intermediate large breaks and almost all small breaks (leakage cracks), except the one (single) mandatory crack at the most adverse location. The 50.59 Screening/Evaluation should also address the elimination of the longitudinal crack at the terminal ends required by the Giambusso letter, but eliminated by MEB 3-1.
5.0 Conclusion The above approach would result in a single unified set of HELB/EQ criteria that would be applicable to all HELB related design parameters for the evaluation plant SSCs. The possible elimination of large breaks should result in lower environmental loads (compartment pressure and temperature) that would result in increasing design margins for the plant SSCs.
The HELB Reconstitution Program will utilize the Proposed PBOC Criteria and systematical address and documents the analysis and results in the various tasks outlined in the Task 1 Report (Reference 8) 6.0 References
- 1. "General Information Required for Consideration of the Effects of a Piping System Break Outside of Contaimnent", AEC December 19, 1972 (Giambusso Letter)
- 2. PBNP FSAR Appendix A.2, High Energy Pipe Failure Outside Containment
- 3. USNRC Generic Letter GL 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements and associated Revised MEB 3-1 of SRP 3.6.2
- 4. PBNP Calculation M-09334-357-HE.1
- 5. PBNP Calculation M-09334-357-HE.2, Rev. 01, High Energy Line Breaks in Selected Piping Systems
- 6. PBNP HELB DBD T-47, High Energy Line Break Design Basis Document, Rev. 0
- 7. USNRC Information Notice 2000-20, Potential Loss of Redundant Safety-Related Equipment Because of the Lack of High-Energy Line Break Barriers
- 9. PBNP Design Guide DG-M09, Rev. 2, Design. Requirements for Pipe Stress Analysis.
Z/01"Ji, Automated Engineering Services Corp