ML101940376
| ML101940376 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/03/2008 |
| 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-P06, Rev. 0 | |
| Download: ML101940376 (22) | |
Text
{{#Wiki_filter:ENCLOSURE 1 ATTACHMENT 8 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 LICENSE AMENDMENT REQUEST 261 EXTENDED POWER UPRATE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PBNP-994-21-05-P06, REVISION 0, STEAM SUPPLY PIPING TO AFW PUMP AND RADWASTE GL 87-11 BREAK LOCATION DETERMINATION 21 pages follow
Page: 4 of 17 Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
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Steam Supply Pipiing to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [] No El Date: 10/03/2008 TABLE OF CONTENTS Page No. 1.0 Purpose ....................................................................................................................................................... 5 2.0 Background ................................................................................................................................................... 5 3.0 Assum ptions and Analysis Notes ........................................................................................................ 6 4.0 M ethodology and Acceptance Criteria .................................................................................................. 7 5.0 References ................................................................................................................................................. 11 6,0 GL 87-11 Break And Leak Location Calculations ................................................................................. 12 7.0 Results & Conclusions ............................................................................................................................... 16 Attachments Pages A. Stress Indices, Resultant Stress Calculation Tables, Stress Tables ........................................ Al-A3 B. SK -FW -FIG.4 - Postulated Line Break Locations .................................................................. B1 C. A ES Technical Position Paper for H ELB Program ................................................................ C 1-C3 Form 3.1-3 Rev. 2
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes El No LI Date: 10/03/2008 1.0 PURPOSE The purpose of this calculation is to establish the locations of intermediate Iiigh 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 3-1, Revision 2 (Reference 3). Combined stress tables for the 3" steam supply piping from the Main Steam headers to Auxiliary Feedwater (AFW) Pump 2P-29 and Radwaste 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. 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" (IE Notice 2000-20, Reference 9), 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.
2.0 BACKGROUND
PBNP's 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 C). Based on the above high energy line definition, Calculation PBNP-994-21-02 (Reference 16) 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 to be postulated irrespective of the combined stress values at these locations. Since the purpose of this calculation is only to generate stress tables to determine GL 87-11 break and or crack locations, the pipe stress analyses for the steam supply piping (Reference 6) still remain valid. Rev. 2 Form 3.1-3 Rev. 2
Page: 6 of 17 Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Linht Revision: 0) Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes NI No ED Date: 10/03/2008 3.0 ASSUMPTIONS AND ANALYSIS NOTES Assumptions Piping material, geometric data and stress analysis (computer results) given in the Analysis of Record (AOR) (Reference 6) for the Steam Supply piping to the AFW Pump and Radwaste 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 Steam Supply 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 10). A pipe code reconciliation study was used to show that no significant differences exist between the two codes for piping analysis. 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 111, Class 2 requirements (Reference 4). Because the way stresses are factored in the two codes is different, the individual load components (moments) for the different load cases (weight, pressure, thermal, and OBE seismic) from the AOR are extracted and adjusted for the stress indices and stress intensification factors, and used as input to calculate the new combined stresses in accordance with Reference 4. Stress Intensification Factors (SIFs) that are calculated per Reference 10 are identical to those calculated per Reference 4. As such, SlFs friom the analysis in Reference 6 are used in the calculations of secoudary stresses per Reference 4. These SIFs appear in the stress tables. Note that SIF is not specified at every point in the existing piping analysis. SIFs are equal to 1.0 for those locations without SIF specifically identified. Factored stresses, calculated based on Reference 4, uses SIPF and stress indices, B, and B 2. These stress indices are determined /calculated in accordance with Table NB-3681 (a)-1 and associated sections of Reference 4. Stress indices for straight pipe runs are obtained from Table NB-3681 (a)-I (Reference 4) and they are Bt 0.5 and B 2 = 1.0. Form 3.1-3 Rev. 2
Automated Engineering CALCULATION SHEET Page: 7 of 17 Services Corp Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [ No El Date: 10/03/2008 4.0 METHODOLOGY AND ACCEPTANCE CRITERIA This calculation uses the GL 87-I 1 (Reference 3) methodology to determine postulated pipe break and crack locations. The analysis in Reference 6 was performed using ASME Section III, Subsection NC and ND, 1977 Edition (Reference 10) as the piping code. The results from this analysis are used to calculate the combined stresses using the methodology in ASME Section III, 1986 Edition (Reference 4) consistent with GL 87-11 criteria. All stress components for the ASME Section IllI stress combination, except longitudinal pressure stress, are obtained from the AOR. The resultant stresses along with appropriate stress indices, "B1 " and "B2" and the stress intensification factor (SIF) 'i' values used in the calculation of the stress components in equations (1) and (2) are given in Sections_4.1 and 4.3 of this calculation. 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 ofASME B&PV Code Section III, Class 2 and 3 as follows: B, PDo/2t + B2 Mnw/Z + B 2 MoBE/Z + i M'rjj/Z > 0.8 (1.8 Sh + SA) (1) In above equation, the first term is the longitudinal pressure stress. The resultant stresses (M/Z) for each load case (deadweight, thermal expansion, and OBE from simultaneous X and Y directions or OBE fi'om simultaneous Y and Z directions) are calculated on the resultant stress calculation Table in Attachment A based on the moments in X, Y and Z (MA, MB, Mc) directions obtained from the AOR (Reference 6). S = [(M^2 + MB2 + Mc 2) 1 2] / Z To obtain the corresponding ASME Section III factored stresses for the deadweight and seismic load cases, the stresses obtained fr'om the resultant stress calculation Table in Attachment A, are multiplied by the corresponding B2 values. Note that thermal expansion stresses are multiplied by the SIF (i). Replacing the M/Z term by S, the modified combined stress equation is as follows; B1 PD 0/2t + B2 SDW + B2 SOBE + i STn Ž 0.8 (1.8 Sh + SA) (1A) Where: P = Design internal pressure, psi Do = Outside diameter of the pipe, in Rev. 2 3;1-3 Form 3:1 Form Rev. 2
Page: 8 of 17 Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [] No [I Date: 10/03/2008 t = Nominal thickness of the pipe, in MDW = Resultant moment due to dead weight, in-lbs MoBE = Resultant moment due to operating basis earthquake, in-lbs MTH = Resultant moment due to thermal expansion, in-lbs SDW = Resultant stress due to dead weight, psi, from Resultant Stress Table in Attachment A SoBE = Resultant stress due to OBE (Max of X & Y or Y & Z earthquake), psi, from Resultant Stress Table in Attachment A STH = Resultant stress due to thermal expansion, psi, from Resultant Stress Table in Attachment A Sh, = Material allowable stress at temperature, psi SA = Material allowable stressrange, psi Z = Section modulus of pipe, in3 i = stress intensification factor, as given in Figure NC-3673.2(b)-1 B, = primary stress index for pressure stress as given in Table NB-368 1(a)-I and further defined in Section 4.4 B2 = primary stress index for bending stresses as given in Table NB-3681(a)-I and further defined in Section 4.4 Large intermediate breaks are to be postulated only at locations where the combined stress exceeds the threshold value of 0. 8(l. 8 Sh+ 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 Steam Supply piping to AFW Pump 2P-29 are summarized on Table 7.1 (Section 7.0). Rev. 2 Form 3.1-3 Form Rev. 2
Page: 9 of 17 Cale. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-1 1 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [ No [I Date: 10/03/2008 Branch Lines The 3" steam supply piping to AFW pump 2P-29 and Radwaste is modeled and analyzed in a separate analytical part (Ref. 6) from the Main Steam header piping. These lines are affected by and in turn affect the main line, and as such, the branch connections to the Main Steam header (Node Points 16 & G20) are considered terminal ends. 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, PD 0/2t + B 2 MDw/Z + B2 MOBE/Z + i M.m/Z ->0.4 (1.8 Sh + SA) (2) In equation 2, the term M/Z is also replaced by the resultant stress S term similar to the modification to equation 1. The modified equation is as follows: B1 PD0/2t + B 2 Sow + B 2 SOBE + i STH > 0.4 (1.8 Sh + SA) (2A) Leakage cracks are to be postulated in locations where the combined stress exceeds the threshold value of 0.4 (l.8SI,+ SA). 4.4 Calculation of Stress Indices Calculated values for various pipe sizes and schedules associated with the piping in the stress tables are provided on Table - I in Attachment A. Stress Indices are calculated in accordance with ASME Section III, Subsection NB, Table NB-3681 (a)-I found in Reference 4, and SIFs are calculated in accordance with ASME Section III, Subsection NC, Figure NC-3673 ;2(b)-l. The following equations are used to calculate the stress indices (B1 and B2) and SIF (i) for welding elbows and welding tees in accordance with ASME Section I1I, NC and NB rules (Reference 4). Form 3.1-3 Rev. 2
Page: 10 of 17 CALCULATION SHEET Cale. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [] No El Date: 10/03/2008 0 Curved Pipe or Butt Welding Elbows (NB-3683.7): B, =-0.1 + 0A-h but not <0 nor >0.5 where: 1.30 R= nominal bend radius of curved pipe B2-but not < 1.0 or elbow, in 112) r,, mean pipe radius, in, = (Do - t)/2 t = nominal wall thickness of pipe, in Do outside diameter of pipe, in and from Figure NC-3673.2(b)- I h = flexibility characteristic t-R 0.9 2 h13 ) ril 0 Butt Welding Tees (NB-3683.9): 13B= 0.5 (Table NB-3681(a)-I) where: 2 R1 = mean radius of run pipe, in, = (E6 - t)/2 Tr= nominal thickness of run pipe, in R.3 but not < 1.0 Do = outside diameter of pipe, in 132b = 0-4 2-B2b B2 stress index for branch pipe 2 B2r B2 stress index for run pipe 132 r =05 but not < 1.0 and fi'om Figure NC-3673.2(b)- 1 0.9 h4.4-T, k~r h(13 ) Stress Indices for straight pipe run are obtained firom Table NB-3681(a)-I (Ref. 4), and they are: B, =0.5, and B2 = 1.0. Rev. 2 Form 3.1-3 Form 3.1-3 Rev. 2
Page: II of 17 Cale. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [ No D Date: 10/03/2008
5.0 REFERENCES
- 1. USAS B31.1.0 - 1967, Power Piping Code.
- 2. PBNP FSAR (08/04), Appendix A.2.
- 3. Generic Letter 87 Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, June 19, 1987.
- 4. ASME Section I1I, 1986 Editions.
- 5. AEC-DOLs Letter to WPS of December 15, 1972 (Mr. Giambusso to Mr. James).
- 6. PBNP Pipe Stress Analysis Reports.
a) PBNP Accession No.-WE-200044, 2EB8A/B-3" Auxiliary Steam to Auxiliary FW Pump 2P-29, Revision 0 including Addendum A. b) PBNP Accession No. WE-200101, Piping System Qualification Report for Main Steam Piping to Anchor EB-8-A1 18 (Outside Containment), Unit 2, Revision OA. c) PBNP Accession No. WE7200102, Piping System Qualification Report for Main Steam Piping to Anchor EB-8-A 121 (Outside Containment), Unit 2, Revision 0. d) PBNP Accession No. WE-300042, Piping System Qualification Report for Radwaste Steam, Units 1
& 2, Revision 2 including Addenda A, B, and C.
e) PBNP Accession No. WE-300004, Auxiliary Steam Line 3" SA-3 from Anchors SA-3, S-32, and S-29 to Anchor S-408, Units 1 & 2, Revision 1.
- 7. DG-M09, Rev. 2, Design Requirements for Piping Stress Analysis.
- 8. Drawing M-2201 Sh. 1, Revision 45, Piping Subsystems P&ID - Main & Reheat Steam System.
- 9. NRC Information Notice 2000-20, Potential Loss of Redundant Safety Related Equipment Because of the Lack of High Energy Line Break Bairriers, 12/11/2000.
- 10. ASME Section III B&PV Code, Subsection NC and ND, 1977 Edition up to and including 1978 Winter Addenda.
- 11. Isometric Drawing P-206, Revision 7, "Main Steam to Aux. Feedwater Pump 2P-29, Y3-EB-8".
- 12. Isometric Drawing PBA-2044 Sheet 1, "Aux Steam to Aux FW Pump 2P-29", Revision 0.
- 13. Isometric Drawing PBA-2044 Sheet 2, "Aux Steam to Aux FW Pump 2P-29", Revision 0.
- 14. Isometric Drawing PBA-3004, "Aux Steam to W.D. Blowdown Area Elevation 44'-0, Aux. Building Central", Revision 4.
- 15. Isometric Drawing PBA-3042, "Radwaste Steam Supply", Revision 2.
- 16. Calc. No. PBNP-994-21-02, HELB Reconstitution Program-Task 2, High Energy System Selection, Rev. 0.
Form 3.1-3 Rev. 2
Page: 12 of 17 Cale. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cakl.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: 'Dan Quijano Location Determination Safety Related Yes [ No El Date: 10/03/2008 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 I1Code, 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 87-1 I (Reference 3) are applied to the Steam Supply piping from the branch connection points at the 30" Main Steam headers to AFW Pump 2P-29 and Radwaste. Part of the requirements to implementing GL 87-11 is that the stress evaluation is to be in accordance with ASME Section III, 1986 edition. Table I in Attachment A calculates the B1 and B 2 stress indices and SIF (i) for the steam supply piping. These indices are calculated in accordance with Paragraphs NB-3683.7, NB-3683.8, NB-3683.9, Table NB-368 1(a)-I and Figure NC-3673.2(b)-l. 6.2 Manipulation of Stress Equations in EXCEL Spreadsheets A spreadsheet or stress table for each line segment of concern was developed taking piping stress data from References 6 computer analyses. Appropriate stress indices were included for each analytical point under consideration. This information was then used to calculate the combined stress value for comparison to threshold values established per equations (1A) and (2A) that were defined in section 4.0. A detailed explanation of the different columns used in the stress tables (spreadsheets) is as follows: Column Description Remarks A &X Node or Analysis Point Number B Outside Diameter of pipe segment, in C Nominal Thickness of pipe segment, in Taken from Ref. 6. D Internal pressure = design pressure, psi E Resultant Dead Weight Stress, Sow, psi Calculated in Resultant F Resultant Thermal Expansion Stress, STH, psi Stress Table based on G Resultant Seismic Stress (Due to X & Y earthquake), SOBEXY, Psi moments from computer H Resultant Seismic Stress (Due to Y & Z eamthquake), SoBEYZ, Psi output of Ref. 6 J Stress Intensification Factor (S]F), (i) K B 1 stress index Calculated as shown in L B2 stress index Section 4.4 and Table - 1 M Longitudhial Pressure Stress, psi. This is the first component in = BI*P*D/2*t equations (1) and (2) shown in sections 4.1 and 4.3 N Deadweight stress, psi. This is the second component in = B 2*(SDw) equations (1A) and (2A) shown in sections 4.1 and 4.3 Form 3.1-3 Rev. 2
Page: 13 of 17 Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes IZ No El Date: 10/03/2008 Column Description Remarks P Thermal expansion stress, psi. This is the fourth component in = i *(STH) equations (1A) and (2A) shown in sections 4.1 and 4.3 Q Operating basis earthquake (OBE) stress, psi. This is the third = B2*[SoBE (max of SoBExy component in equations (IA) and (2A) shown in sections 4.1 or SoBEYz)] and 4.3 R Combined Stress, psi = Sum of columns M, N, P, and Q S Threshold Limit for Leakage Crack Postulation = 0.4 (1.8 Sh + SA) T Threshold Limit for Large Break Postulation = 0.8 (1.8 Sh + SA) V Ratio of combined stress / threshold crack limit Crack postulated if> 1 W Ratio of combined stress/Ithreshold break limit -Break postulated if> I The Steam Supply piping material is A106 Gr. B (Reference 6). At temperatures < 650 'F, the allowable hot condition stress, S1, and allowable stress range, SA, per Reference 4 are as follows: S,= SI = 15000 psi from Reference 4, Table 1.7.1 SA= 22500 psi, where SA = f (1.25S, + 0.25 St) and f= 1.0, per NC-3611.2(e) Threshold Limit for Intermediate Large Circumferential or Longitudinal Breaks: 0.8(1.8*S1 , + SA) = 39600 psi Threshold Limit for Leakage Cracks: 0.4(l.8*Sh, + SA) = 19800 psi Rev. 2 Form 3.1-3 Form- 3.1-3 . Rev. 2
Page: 14 of 17 I Cale. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [] No El Date: 10/03/2008 6.3 Sample Calculations Sample of Resultant Stress Calculations (Table -2. Attachment A2): For Example Nodes 485 and 530 (Ref. 6a) b):= 3.5Gin t := 0.216in D - 2.t D, = 3.068in 7D - D 4). 3 Section Modulus S S = 1.72in 32:D DW TH OBE(X+Y) OBE(Y+Z) DW TH OBE(X+Y) OBE(Y+Z) (104 54 262 262. my:=(177 350 306 306)'ft'lbf5 K 104 2 262 262) 3 429 156 1561If b DW TH OBE(X+Y) OBE(Y+Z) (345 240 248 248).ft-lbf K 82 703 205 205) DW TH OBE(X+Y) OBE(Y+Z) Resultant Stress
- MX2 + MY2 + MZ2 C(2794 2978 3293 3293") 485 S 922 5732 2557 2557)psi 530 Rev. 2 Form Form 3.1-3 3.1-3 Rev. 2
Automated Page: 15 of 17 Engineering Alzkm Services Corp Cale. No.: PBNP-994-21-O5-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-1 1 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [Z No El Date: 10/03/2008 Sample of Stress Table Calculation (Table - 3, Attachment A3): For exam pie row of Node Point 485 is considered COLUMN A Choose Node 485 from the Stress Table for Steam Supply to AFW Pump 2P-29 B: Outside Diamter of pipe, Do := 3.50in C: Pipe Thickness, tn.:= 0.216in D: Internal pressure, P := I t45psi E: Resultant Deadweight stress, SDW := 2794:psi: F: Resultant Thermal Expansion stress, Sm := 2978 psi G & H: Resultant Seismic (max OBE) stress, 8o]3c:= ma,(3293,3293).psi J: Stress intensification factor, i:= 1.777 K: Stress index, B1 := 0.044 L: Stress index, B2 =2.566 BI.P.D0 M: Longitudinal pressure stress, ColM .- DColM =408psi N: Deadweight stress, CoIN := 132.SDw CotN = 7169'psi P: Thermal Expansion stress, CoIP := i'STH ColP = 5292 psi Q: OBE stress, CoIQ:= B3SoBE2 Col_Q 8450psi Combined stress, Col_R := ColM + ColN + ColP + Co_Q Co_R =21319psi Limit for Crack, ColS := 19800psi Limit for Break, ColT := 39600psi V: Ratio of combined stress to limit for crack, CoV:= Col-R CotlV= 1.077 Col_S Col R W: Ratio of combined stress to limit for break, ColW := ColW = 0.538 ColT Form 3.1-3 Rev. 2
I Page: 16 of 17 Calc. No.: PBNP-994-21-05-P06 Client: Florida Power & Light Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Calc.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes [I No El Date: 10/03/2008 7.0 RESULTS & CONCLUSIONS Results The 3" Steam Supply lines from the 30" Main Steam headers (Node Points 16 & G20) to AFW Pump 2P-29 and Radwaste 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 steam supply lines at the branch connection (Node Points 16 & G20) to the Main Steam headersin the Auxiliaty Building, at intermediate anchors EB-8-A118 & EB-8-A121, and at the inlet side of Motor Operated Valves 2MOV-2019 and 2MOV-2020 upstream of the AFW Pump. Locations where large breaks are required to be postulated are summarized in Table 7.1 below. As seen from the stress table in Attachment A, all steam supply piping combined stresses are well below the Intermediate Large Break Threshold Limit. Therefore, no intermediate large breaks need be postulated for the 3" Steam Supply Lines from the MS headers to the AFW Pump and Radwaste. Table 7.1 - Postulated Large Breaks at Terminal Ends and Intermediate Locations Break Location Node Point Notes 3" Steam Supply lines from MS Header Branch 16, G20 Terminal End Connections to AFW Pump 3" Steam Supply line at anchor EB-8-A1 18 5 Terminal End 3" Steam Supply line at anchor EB-8-A 121 1190 Terminal End 3" Steam Supply at inlet of 2MOV-2019 550 Terminal End 3" Steam Supply at inlet of 2MOV-2020 785 Terminal End 3" Steam supply line at anchor SA-3-$32 (Ref. 13) 1535 Terminal End 3" Auxiliary steam supply line at anchor SA-1-$408 400 Terminal End (Ref. 14) 6" Radwaste steam supply line at orifice flange 130 Terminal End downstream of valve SA-I (Ref. 15) Form 3.1-3 Rev. 2
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-I-ClicuP Florida Power & TITht Revision: 0 Station: Point Beach Nuclear Plant - Unit 2 Prepared By: Chris Kandalepas Cale.
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Steam Supply Piping to AFW Pump and Radwaste GL 87-11 Break Reviewed By: Dan Quijano Location Determination Safety Related Yes E] No E] Date: 10/03/2008 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 (Attachment A). Locations where leakage cracks are required to be postulated are summarized in the Table 7.2 below. Table 7.2 - Leakage Crack (Small Break) Locations for Steam Supply Lines Crack Location Node Point Notes Riser elbows upstream of 3"tee conn. to relief valve 485 2MS-252 Elbow upstream of MOV 2N4S-2020 800 F 2"" Elbow (top of riser) upstream of MOV 2MS-2020 835, 845, 848B 3" tee downstream of 2MS-252 1275 3 rd Elbow downstream of relief valve 2MS-251 1285 1 1/2/" line downstream of 3" tee by anchor EB-8-A118 400, 585 3" Radwaste steam supply line at inlet of valve RS-3 55 Conclusions An evaluation of the Steam Supply lines from the 30" MS headers to AFW Pump 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:
- Intermediate stress related and arbitrary large breaks need not be postulated anywhere along the entire Steam Supply lines to AFW Pump and Radwaste.
- Circumferential large breaks are required to be postulated at the terminal ends as shown in Table 7.1.
- Leakage cracks (break size = /2times the pipe wall thickness x 2the 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 (11 Notice 2000-20, Reference 9), 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. Rev. 2 3.1-3 Form 3.1-3 Rev. 2
Calc. No. PBNP-994-21-05-P06, Rev. 0 Attachment A, Page Al of A3 Prepared by: Chris Kandalepas Checked by: Dan Quijano Date: 8/19/2008 TABLE - 1: STRESS INDICES (FOR USE IN STRESS TABLES) Outside Nominal Mean Radius Curved pipe or Butt Weldinq Elbows Welding Tees Nominal Diameter Pipe Thickness of pipe ( rm) Bend Flex. Char. SIF (i) Index B Index B2 Flex. SIF (I) Index B2 Note 1 Pipe Size of Pipe Schedule of pipe (t) Note I Radius (R) (h) Note I Note Char.(h) I (Do) Note 2 Note 1 Note I Note 1 Tee-Run Tee-Br Note 1 3 3.50 40 0.216 1.6420 4.5 0.3605 1.7768 0.0442 2.5664 0.5788 1.2958 1.933 1.547 3 3.50 40 0.216 1.6420 3.0 0.2403 2.3282 0.0000 3.3630 0.5788 1.2958 1.933 1.547 Notes:
- 1) For applicable equations for calculating rm, Rm, r, h, I, B, and B2, see Section 4.4 of main calculation.
- 2) Bend Radius is long radius elbows (1.5 x diameter) or special radius given in pipe stress analysis input (Ref. 6).
Cale. No. PBNP-664-21-05-P06, Rev. 0 Attachment A, Page A2 of A3 Prepared by: Chris KanJalsas Checked by: DanOcilana Date; 8119/2008 TABLE - 2 Resultant Stress Calculation Table for Steam Supply Piping from Anchors EB78-A1 18 & -Al 21 to AFW Pump 2P-29 (Based on analysis results from Accession No. WE-209944 Rev. 0, Dated 6-21-87) Input Data from A n yels Moments From Anal sie Outside Pipe Section alF Dead Wealht Moments Thermal Expansion Moments MaxX + Y Seismic Moments MaxY I Z Seismic Moment NODES die, thickness Modulus Cola. X Y Z X Y Z X Y Z X Y Do (In) t. (in) S an') ft-lbs ft-lbs ft-ibs ft-ibs ft-lbs ft-lbs ft-lbs ft-lbs ft-lbs ft-lbs ft-hbs f 485 3.5 0.216 1.72 1.777 104 177 349 54 300 240 292 306 248 262 306 530 3.5 0,216 1.72 1.399 104 3 92 2 429 703 252 159 205 262 156 800 15 0.219 1.72 1.777 12 144 126 124 92 571 258 370 194 298 370 810 3.5 0.216 1.72 1.300 12 94 55 227 348 398 298 299 153 258 269 635 325 0.216 1.72 1.777 7 119 57 227 139 1268 52 35 102 52 35 845 3.5 0.216 1.72 1.777 37 49 94 1109 891 97 90 77 33 90 77 8488 3.5 0.218 1.72 1.777 30 12 27 5 945 1007 33 70 84 33 70 1190 2.0 0,216 1.72 2,100 27 221 256 437 349 76 1000 59 78 1000 126 32,9 0.216 1.72 1.777 49 17 177 250 197 739 212 82 166 212 62 1275 3.5 0,216 1.72 1.200 195 193 196 376 1404 824 192 teo 258 182 166 1285 3.5 0.219 172 1.777 195 127 11i 376 552 447 172 157 144 182 157 1325 3.5 0,216 1.72 1.777 20 94 16 519 928 150 100 124 64 too 124 Resultant Stress Calculation Table for Steam Supply Piping from 30" MS Header to Anchor EB-8-A118 (Based on analysis results from Accession No.WE-200101 Rav.0, Dated 12-29-95) Input Data from Anaysise NODES die, au~d thickness Pip ato~in Madulus S5P Calc. DO (in) t, tIn) tt,,e0 Resultant Stress Calculation Table for Steam Supply Piping from 30" MS Header to Anchor EB-8-Al 21 (Based on analysts results from Accession No, WE-200102 Rev. 0, Dated 12-18-95) Input Data from Analyses Momenta From Analysis Calculated Resultant Stress Otside] Pipe Section Sir lF Dead Weliht Moments Thermal Expansion Moments MaxX Y Seismic Moments NODES( MaxY + Z Seismic Moments DW THR S1es X+Y es v+Z die, I thickne.s Modulus Caelc. X V I z x I . 2 x y 2 1 V Z IRes. Stress Res. Stress Res. Stress Rae. tress Do (in) t. (in) S (In') .b I ft-lbs I ft-the ft-lbs h-the I ft-ike I I ft-lbs ft-lbs I h-lbs ft-ibe ftls jI t-lbs I psi psi I psi I pal 225 3.5 0.21 t 197 1.? 8 153 I 2 I 20 213 i 32 I 307 1 48 I 53 I 70 1 9I5 60 I 1 1 i 2221 1 2610 1 695 i 747 235 3.5 0.21 i 1.72 1.879 145 I 2 I 282 326 i 32 I 277 1 163 1 53 i 195 i 210 I 69 I 185 1 2207 1 2986 1 1807 i 2006 Resultant Stress Calculation Table for Auxiliary Steam line 3" SA-3 from Anchors SA-3-$32 and -329 to Anchor SA-3-$408 (Based on analysis results from Accession No. WE-300004 Rev. 1, Dated 03-08.86) Input Data from Analysis Moments From Analysis Outsaid Pipe Sectian 9IF NODES die. tckness Madalus Calc. 0, lint ,t^int S lin='t ; Resultant Stress Calculation Table for Radwaste Steam (Based on analysis results from Accession No. WE-300042 Rev. 2. Dated 01-18-02) mpu, aatamAayi Outsida Pipe Section SiF NODES I dis. 0 thickness Modutas 1 a (in) t. (in) a fin'i Cab. psi I psi I psi I psi I
Calc. No. PBNP-994-21-05-P08, Rev. 0 Attachment A, Page A3 of A3 Prepared by: Dan OQulno Checked by: Chis Kandalepas Date: 8/1l/2008 TABLE - 3 Stress Table for Steam Supply Piping from Anchors E-B--A118 & -A121 to AFW Pump 2P-29 (Based on analysis results from Accession No. WE-200044 Rnv. 0,Dated 9.21-87) F 0 1 H J1 K L I M I N P I Q R S T V W X Long. Dead Thermal aBE Comb. Ratio Ratio Outside Pipe internal Dead Thermal Mex X+Z Max Y+Z SIP INDEX INDEX Pr. Weight Expansion Stress Stress Limit Limit NODES dea. thickness Preee.re Weight ExpansIon Seismic Seismic Calt. Bt B2 Stress Stress Stress for for Comb. St. Comb. St. NODES Do (In) ItI (in) P (psi) psi psi psi psi psi 13i psi nai Dai Crack Crock LimIt Break Limit 8291 19(00 39600 7452 7316 5022 4 10025 9795 17408 17080 2325 8882 10490 28755 1 19800 1 39800 1.4523 1 0.7261 11901 Stress Table for Steam Supply Piping from 30" MS Header to Anchor EB-8-A118 (Based on analysis results from Accession No. WE-200101 Rev. 0, Dated 12-29-98) 382 I : I 470 I A0O 8542 I I 955 I 2454 1 1 1713 1 : 1 958 I Stress Table for Steam Supply Piping from 30" MS Header to Anchor EB-8-A121 (Based on analysis results from Accession No. WE-200102 Rev. 0. Dated 12-18-95) SAeS I I I 1807 Li I 4 Stress Table for Auxiliary Steam line 3" SA-3 from Anchors SA-3-$32 and -S29 to Anchor SA-3-S408 (Based on analysis results from Accession No. WE-300004 Rev. 1, Dated 03-08-86) 105 I 3.5 t 0.216 I I n.800 I 1.0 1 4196 1: 7 I10,3108 I 1ot I 110 I 3.5 i N 0.2I r I 0.044 I 2.0 1 8259 I 1 4 I 0,3002 I 110 1 290 1 3.5 1 0.216 F 1 0.044 I 2.0 1 3842 I 1 S 1 0.3779 I 2990 Stress Table for Radwaste Steam (Based on anaivsis result& from Accession No. WE-300O42 Rev. 0. Dated 01M1A-2882l I 1143 I 0.3684 1 30 1 14960 14587 I1 1988o0 19800 1 39600 39600 11 0 11 208566 1 9800 16823 19B00 1 3993022 1 M90 II 2 221 1 1.870 0.4248 1 66 1
Y z* X (SOUTH) z I 0ý C15-9' LEGENDi POINT BEACH NUCLEAR PLANT MAIN STEAM SYSTEM, UNIT 2
- DENOTES POSTULATED PIPE LINE BREAK TO AUX, FEED WATER PUMP 2P-29
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- DENOTES LINE BREAK NUMBER. Av-AAUTOMATED
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TECHNICAL POSITION PAPER FOR ESTABLISHING HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Rev. 0, December 4, 2006 Rev. 1, August 7,2008 CaO~tD PBNP-1?-z7.-oI-05PO6,fev 1 1.0 Introduction ACzt ° N e C6
-54 V 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 environment outside Containment, 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 Sh +SA) to establish the EQ parameters (Reference 4, 5). It is noted that Revision 1 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 I 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. Automated EngineerngServices Corp 42'ý,vk,.
HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Page 2 Ca& I'JO. 2.0 Line Characterization Criteria and Break Selection Rules /\-{ C PCJ e C 2 6.Uocj 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-I, 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-I, 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 PROC 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 of criteria 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 3-1, 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 (132 indices) and longitudinal pressure (B1 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. 1.c(2)(b)(ii)) and not at each pipe fitting (Section B. 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-
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HELB BREAK & LEAKAGE CRACK LOCATION SELECTION CRITERIA Page 3 3653 of the ASME Code Section III) and local stresses at the integral welded attachments(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. C&Cc. No. P'60-Y94-21-o5-Po6,,Rev. o 4.0 Regulatory & Licensing Issues A&&LL. C1 P03e C3 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 Containment", 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
- 8. AES Corp. PBNP HELB Reconstitution Program Task 1 Report, Rev. 0.
- 9. PBNP Design Guide DG-M09, Rev. 2, DesignRequirements for Pipe Stress Analysis.
- 10. ASME B & PV Code Section III, Subsections NC and ND, 1986 Edition.
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