ML051870299
| ML051870299 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/23/2005 |
| From: | Agrawal P, Hoang P, Verma V Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 1165-065 2005-06760 | |
| Download: ML051870299 (48) | |
Text
ENCLOSURE SARGENT & LUNDY CALCULATION, "ANALYSIS OF POSTULATED REACTOR HEAD LOAD DROP ONTO THE REACTOR VESSEL FLANGE," REVISION 3 DATED JUNE 23, 2005 (ABRIDGED VERSION EXCLUDING ATTACHMENTS A, B AND C INCLUDING REVISED PAGES OF ATTACHMENT D)
ISSUE
SUMMARY
Form SOP-0402-07, Revision 7 DESIGN CONTROL
SUMMARY
CLIENT: Nuclear Management Company UNIT NO.: 2 Page No.: 1 PROJECT NAME: Point BeachNuclear Generating Plant PROJECT NO.: 11165-065 0 NUCLEAR SAFETY- RELATED CALC. NO.: 2005-06760 0 NOT NUCLEAR SAFETY-RELATED TITLE: Analysis of Postulated Reactor Head Load Drop Onto the Reactor Vessel Flange EQUIPMENT NO.:
IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD First Issue: Page 1-20. B1-822. Cl-C12 and D1-D45 issued as Rev. 0 INPUTS/ ASSUMPTIONS 0 VERIFIED a UNVERIFIED REVIEW METHOD: Detailed Review REV. 0 STATUS: Approved DATE FOR REV.: 6/8/2005 PREPARER Phuong H. Hoang (Signature on File) DATE: 6/8/2005 REVIEWER Vijay K Verma (Signature on File) 6t8/2005 APPROVER P. K. Agrawal (Signature on File) DATE: 6/8/2005 IDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDEDNOIDED & REVIEW METHOD Attachment A: Page A1 -A30 Issued as a part of Rev. 0 INPUTS/ ASSUMPTIONS 0 VERIFIED o UNVERIFIED REVIEW METHOD: Detailed Review REV. 0 STATUS: Approved DATE FOR REV.: 6812005 PREPARER Mohammad Amin (Signature on File) DATE: 6/8/2005 REVIEWER Javad Moslemian (Signature on File) DATE: 6/8/2005 APPROVER P. K. Agrawal (Signature on File) DATE: 6/8/2005 IDENTIFICATION OF PAGES ADDED/REVISEDISUPERSEDEDNOIDED & REVIEW METHOD Rev 1 of the calculation supersedes Rev 0 of the calculation in Its entirety.
Prepared, Reviewed, and Approved Main Body of Calculation and Attachments 8, C, and D.
INPUTS/ ASSUMPTIONS 0 VERIFIED o UNVERIFIED REVIEW METHOD: Detailed Review REV. 1 STATUS: Approved DATE FOR REV.: 6/19/2005 PREPARER Phuong H. Hoang (Signature on File) DATE: 6/19/2005 REVIEWER Vijay K Verma (Signature on File) DATE: 6119/2005 APPROVER P. K. Agrawal (Signature on File) DATE: 6/1912005 NOTE: PRINT AND SIGN IN THE SIGNATURE AREAS SOP04023-REV7.DOC Page 1 of 40 Rev. Date: 04-08-2005
ISSUE
SUMMARY
Form SOP-0402-07. Revision 7 DESIGN CONTROL
SUMMARY
CLIENT: Nuclear Management Company UNIT NO.: 2 Page No.: 2 PROJECT NAME: Point BeachNuclear Generating Plant PROJECT NO.: 11165-065 0 NUCLEAR SAFETY- RELATED CALC. NO.: 2005-06760 0 NOT NUCLEAR SAFETY-RELATED TITLE: Analysis of Postulated Reactor Head Load Drop Onto the Reactor Vessel Flange EQUIPMENT NO.:
IDENTIFICATION OF PAGES ADDEDIREVISEDISUPERSEDEDNOIDED & REVIEW METHOD Rev 1 of the calculation supercedes Rev 0 of the calculation In Its entirety.
Prepared, Reviewed, and Approved Attachment A.
INPUTSI ASSUMPTIONS 0 VERIFIED o UNVERIFIED REVIEW METHOD: Detailed Review REV. I STATUS: Approved DATE FOR REV.: 6119/2005 PREPARER Mohammad Amin (Signature on File) DATE: 6119/2005 REVIEWER Javad Moslemian (Signature on File) 6/19/2005 APPROVER P. K. Agrawal (Signature on File) DATE: 6/19/2005 IDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDEDNOIDED & REVIEW METHOD Rev 2 of the calculation revised pages 3,7.36, 39,40 and added pages 36a, D102, D103, D104, 0105, D106. and D107 INPUTS ASSUMPTIONS 0 VERIFIED o UNVERIFIED REVIEW METHOD: Detailed Review REV. 2 STATUS: Approved - DATE FOR REV.:
PREPARER Phuong H. Hoang DATE: 4/' Z/O-REVIEWER Vijay K. Verma AdL DATE:
APPROVER P. K. Agrawal
- DATE: OrS IDENTIFICATION OF PAGES ADDEDIREVISEDISUPERSEDEDNOIDED & REVIEW METHOD INPUTS/ASSUMPTIONS o VERIFIED o UNVERIFIED REVIEW METHOD: REV.
STATUS: _ DATE FOR REV.:
PREPARER DATE:
REVIEWER DATE:
APPROVER DATE:
NOTE: PRINT AND SIGN IN THE SIGNATURE AREAS SOP04023-REV7.DOC Page 2 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 3 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop TABLE OF CONTENTS
1.0 Purpose and Scope
.................... Page 4 2.0 Design Input and Assumptions .................... Page 4 3.0 References .................... Page 13 4.0 Methodology .................... Page 14 5.0 Calculation .................... Page 22 6.0 Results / Conclusion .................... Page 35 ATTACHMENTS:
A Evaluation of Concrete for Column Loads ...... 11 Pages B DIT-1, DIT-2, DIT-3 and CMTRs ..... 19 Pages C ANSYS Model and Post Processing Input Files ..... 29 Pages D ANSYS Output Files ..... 107 Pages R2 SOP04023-REV7.DOC Page 3 of 40 Rev. Date: 04-08-2005
SVrgOAL LIUncly" Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 4 Project No.: 11165-065 Rev. No.1 Analysis of Postulated Reactor Head Load Drop 1.0 PURPOSE AND SCOPE The purpose of this calculation is to evaluate a postulated load drop of the Reactor Vessel (RV) head onto the RV flange for Unit 2, Point Beach Nuclear Generating Plant. Per the input provided by Nuclear Management Company (NMC), this analysis considers a flat vertical impact of the new RV head that weighs 194,000 pounds dropping from a height of 26.4 feet onto the RV flange. This impact analysis approach is consistent with the NUREG-0612, Appendix A (Ref.
3.11). This analysis also includes an evaluation of the structural integrity of supporting elements in the load path, and predicts the vertical displacement of the reactor vessel.
2.0 DESIGN INPUT & ASSUMPTIONS 2.1 General Impact Load Path
Description:
Refer to drawings 233-681, 233-682, and C-2320 (Ref. 3.2 (a), (b) and (c)). When reactor vessel (RV) flange is vertically impacted by the falling head, the impact effect is directly transmitted through the vessel to four RV nozzle supports and two RV bracket supports to a hexagonal framework of six horizontal box girders.
Each girder supports one reactor support at its center, and it is supported at its ends by 12" Schedule 120 pipe columns. Refer to drawing C-2320, Ref. 3.2(c). The top of girder elevation is 31 '-9 3/8". The bottom of girder is at elevation 30'-1 7/8". The columns are pinned at bottom, and anchor bolts anchor the base plate to concrete base mat at elevation 10'-0".
Refer to Ref. 3.2 (o), the RV Inlet and Outlet nozzles are connected to the Reactor Cooling System (RCS) Cold Leg and Hot Leg Piping. The RCS piping provides additional stiffliess to the RV nozzles under vertical impact loading.
Refer to Ref. 3.2(d, e, f. h, i and j). Each pipe column passes through a hole in a concrete shelf structure. The elevation of top of the shelf is 29'-10 1/2". The bottom of the shelf elevation is 14'-1 3". As seen on drawings C-326 and C-2326, Ref. 3.2(e and i), the gap between the column and the surrounding concrete in the hole of the shelf structure is 1 3/4". Therefore, in the event of column buckling after closure of this gap, the shelf structure can provide lateral support for the columns. The lateral force on the shelf structure is calculated to determine the adequacy of lateral support from the shelf structure.
Noting that the bottom of girder elevation is 30'-1 7/8", and top of shelf elevation is 29'-10 1/2",
it is seen that if impact causes the RV move downward less than 3 3/8", the girders will not SOP04023-REV7.DOC Page 4 of 40 Rev. Date: 04-08-2005
Svgrge W Lunchey s Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 5 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop impact the top of the shelf structure. (At the top of the support columns, the vertical gap to the shelf is smaller than 3 3/8" due to the column cap plate connection. However, the bearing area is small. Concrete will be locally crushed, without any gross damage to the shelf structure.)
In summary, the load path consists of the RV, RV supports at four nozzles and two brackets, six support girders, and six pipe columns and their pins and base plates and concrete foundation. The RCS piping will also transfer a portion of the impact load to the Steam Generator and the RC Pump support structures.
SOP04023-REV7.DOC Page 5 of 40 Rev. Date: 04-08-2005
S~Ererrm, LJunrV..
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 6 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 2.2 Design Input:
The following verified design input from Ref. 3.10 will be used for the Reactor Vessel (RV)
Head drop analysis for Units 2 of Point Beach:
2.2.1 Drop Parameters:
Maximum Drop Height = 26.4 feet Weight of RV Head = 194,000 lbs Total Weight of RV (Including internals, fuel and water) = 707,600 lbs 2.2.2 Material Properties:
a) Concrete:
In-place concrete compressive strength is 5200 psi.
b) Reinforcing Bars:
All reinforcing bars are deformed bars conforming to ASTM Al5 with yield strength of 40 ksi and ultimate strength of 75 ksi.
c) Liner:
Liner plate material is ASTM A167 type 304L with yield strength of 25 ksi and tensile strength of 70 ksi.
d) RV Support Girders:
Girder plate material is ASTM A517 Grade F with yield strength of 90,000 psi and tensile strength of 105,000 psi.
e) RV Support Columns:
The 12.75" OD schedule 120 pipe columns are ASTM A53 Grade B. Per CMTR (Heat No. 1573) the yield strength is 42,800 psi and tensile strength is 78,900 psi.
SOP04023-REV7.DOC Page 6 of 40 Rev. Date: 04-08-2005
Onrgmet Luncfdy' Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 7 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop f) Pins for RV Support Columns:
The 4" diameter pins are ASTM A490. Per CMTR (Heat No. AC3P2489) the yield strength is 148,800 psi and the tensile strength is 164,900 psi.
g) RCS Piping RCS piping material is ASTM A376 TP 316 per Ref. 3.2 (q) h) RV Support Bracket:
The bracket material is ASTM A-516 Grade 70 per Ref. 3.2(g) i) RV Nozzles:
RV nozzles are per ASTM A508-64, Class 2. See the CMTRs in Attachment B.
Table 2.2.3: Inlet/Outlet Nozzle Material Properties from Material Test Certificates (ASTM A-508-64 Class 2), Ref. 3.10 Test Heat Sy ( si) u (ksi) % Elong Notes X-1 9-5742 71 93.5 26 Inlet Nozzle X-2 9-5742 68 93.5 26 Inlet Nozzle X-1 9-5414 68.5 92 24 Inlet Nozzle X-2 9-5414 72.5 95 24.5 Inlet Nozzle X-a 9-5691 64 98.5 24.2 Outlet Nozzle X-b 9-5691 64 98.0 27.2 Outlet Nozzle IR2 X-a 9-5716 69.5 92.5 25 Outlet Nozzle X-b 9-5716 65.5 87.5 26 Outlet Nozzle Median 68 93.5 -25 :
SOP04023-REV7.DOC Page 7 of 40 Rev. Date: 04-08-2005
rgrsl Landy"5 Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 8 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 2.2.3 Dimensions and Configurations:
Dimensions and configuration are based on the information given in Ref. 3.2:
Ref. 3.2(a and k) show the dimensions of the Reactor Vessel, the Inlet nozzle and Outlet nozzles. The Inlet nozzle and Outlet nozzle dimensions are similar. The dimensions of the Inlet nozzle were used in the analysis. The dimensions used to construct the model are shown in Figure 2.1.
ELEMENTS -0 d J4- 14.47" AN JUN 1 2005 REAL NUo 20.8 16:08:28 66" -i
-0" 53.76"
__ _ Center Line of Inlet Nozzle iD
_42.66" I 8.217" 836156_ Pad 24" wide 7-Center Line of RV 195.56" 10_ .5" A-
-4.125" RV Nozzle (Point Beach)
Figure 2.1: RPV and Inlet Nozzle (Dimensions are in inches)
SOP04023-REV7.DOC Page 8 of 40 Rev. Date: 04-08-2005 r'io
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ftff" IRV Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 9 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop AN JUN 3 2005 16:33.49 Figure 2.2: A 60-degree Sector FEA Model of the RPV Support Bracket ELDER AN REL 2- 17i48:24 1 5" Thick \ 02'
- ~rtical Ax 0 El. 31'- 9 318' El. 29'- 10 El." 3' 1 7/8 Top of Concrete Shelf s lE.3'178 2 3" Thick Top and Bottom El 17'-1 3/.'
Bottom of Concrete Shelf 12' Sch. 120 Pipe El. 10'- 0" Figure 2.3: A 60-degree Sector FEA Model of the Hexagonal Girder Box Frame SOP04023-REV7.DOC Page 9 of 40 Rev. Date: 04-08-2005 001
StergeLA nclyl-Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 10 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop The RV support bracket was analyzed with dimensions that were initially measured and scaled from Ref. 3.2 (1, in). A detailed drawing of the RV support bracket obtained later, Ref. 3.2 (g) shows approximately the same bracket dimensions as the analyzed dimensions. The deviations are minor and conservative for the bracket stress calculation and do not significantly affect the combined stiffness of the RV support structure. The dimensions used to construct the bracket model are shown in Figure 2.2.
Ref. 3.2(c) and 3.2(n) show the dimension of the girder box frame and the support columns. The dimensions used to construct the girder box frame and support column model are shown in Figure 2.3.
Ref. 3.2(o) shows the general layout of the RCS piping. The detailed dimensions are provided in Ref. 3.2(p) for Loop A and Loop B piping. Piping material is shown in Ref. 3.2(q). The dimensions used to construct the RCS piping models are shown in Figure 2.4 and Figure 2.5 for the hot leg and the cold leg respectively.
SOP04023-REV7.DOC Page 10 of 40 Rev. Date: 04-08-2005
15 0M.
Mew!O45 .. ltw- L.Ur-lciv---
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 11 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop JHM 17 2005 08527:20 SG RV deg LC 155.88H eech)6 RCS Hot Leg Piping Model (Point Be-ch)
Figure 2 .4: Hot Leg Piping Layout - Elevation View 17 WINN 2005 09:34:39 11 114.5' 36.29" 1<- >e RCP 26 deg RV RCS Cold Leg Piping Model (Pjoit Beach)
Figure 2 .5: Cold Leg Piping Layout - Plan View SOP04023-REV7.DOC Page 11 of 40 Rev. Date: 04-08-2005
SrgOOAC. Luncdyr Client: Nuclear Management Company Caic. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 12 Project No.: 11165-065 Rev. No.1 Analysis of Postulated Reactor Head Load Drop
2.3 Assumptions
2.3.1 The RV, nozzles and support bracket material is modeled using the true-stress true-strain curve of a SA-106 Grade B specimen as shown in Figure 5.3. The actual RV nozzle material properties listed in Table 2.2.3 are higher than the strength of SA-106 Grade B material, and irradiated material also has higher yield strength. Per Ref. 3.2 (g), the RV support bracket is ASTM-A516 Grade 70, which has a slightly higher strength than SA-106 Grade B. However, since the RV flange, shell and nozzle are much stiffer than the girder box support columns, a higher material strength will not significantly affect the analysis result. Therefore, this assumption does not require verification.
2.3.2 Contact stiffness at the RV flange and the head on impact is taken as 100 times higher than the stiffness of the target RV and its supports such that it is nearly rigid contact.
2.3.3 In this analysis, only impact damping is considered. An impact damping of 5% of the critical damping is used in the analysis. This damping consideration is judged to be reasonable for this application because of the following:
- Energy loss due to plastic damage at the impact surface between the reactor head and the reactor vessel flange.
- Energy loss due to imparted damage to six lateral supports for the hexagonal girder frame. The damage will be due to shearing of the 1 3/4" diameter studs and bending of the partially embedded lateral supports girder.
- Energy loss due to local damage to the liner and concrete crushing at the top of the six support columns.
The 5% damping is the same as that used in the Prairie Island calculation on the same subject (Ref. 3.1).
SOP04023-REV7.DOC Page 12 of 40 Rev. Date: 04408-2005
'.S 52rg~nt Amp LunclMjri Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 13 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop
3.0 REFERENCES
3.1 Sargent & Lundy Calculation 2005-05621, " Analysis of Postulated Reactor Head Load Drop Onto the Reactor Vessel Flange", for Prairie Island Nuclear Generating Plant, Revision 1.
3.2 (a) 233-68 1, "General Arrangement (Elevation)", Rev. 1 (b) 233-682, "General Arrangement (Plan)", Rev. 0 (c) C-2320, "Reactor Steel Supports", Rev. 5.
(d) C-325, "Unit 1, Containment Structure Biological Shield Liner Plate Penetrations",
Rev.7 (e) C-326, "Unit 1, Containment Structure Biological Shield Liner Plate", Rev.3 (f) C-135, "Unit 1, Containment Structure Internals, Reinforcing Sections Sheet2", Rev. 11.
(g) 233-692, "Reactor Vessel Miscellaneous Details", Rev. 1 (h) C-2325, " Containment Structure Biological Shield Liner Plate Penetration", Rev 3.
(i) C-2326, " Containment Structure Biological Shield Liner", Rev 2.
(j) C-2135, " Containment Structure Interior Reinforcing Section", Sheet 2, Rev 9.
(k) 233-686, "Nozzle Details", Rev. 2.
(1) 233-685, "Pressure Vessel Final Machining for Westinghouse Corp.", Rev. 6.
(m)233-683, ", "Pressure Vessel Final Machining for Westinghouse Corp.", Rev. 4.
(n) Miscellaneous Nooter Corporation Drawings: 34534, Rev. 0, 34535, Rev. 0, 34536, Rev.
0, 34537, Rev. 0, 34528, Rev. 3, 34529, Rev. 4, 34530, Rev. 4.
(o) P-258 Sh. 1 Rev. 0, Bechtel Drawing, " Piping Isometric Primary Coolant System".
(p) Westinghouse Drawings: F-126, Rev. 0, F-127, Rev. 0, F-130, Rev. 0, F-131, Rev. 0.
(q) ISI Isometric Drawings: ISI-2120, Rev. 9, ISI-2121, Rev. 7.
3.3 Sargent & Lundy SVVR 03.7.596-8.1, "ANSYS/Mechanical/FLOTRAN, Version 8.1",
Rev.0, dated 3/28/05.
3.4 Point Beach DIT-2 and DIT-3, dated 6/17/05. (Attachment B).
3.5 Proceedings of Second ASCE Conference on Nuclear Power, Vol.5, "Report of Committee on Impactive and Impulsive Loads", Knoxville, Tennessee, September 15-17, 1980.
3.6 ACI 209R-92, "Prediction of Creep, Shrinkage, and Temperature Effects in Concrete Structures-Report by ACI Committee 209", Approved 1997.
3.7 Thomson, W.T.," Theory of Vibration with Application", 4h Edition, Prentice Hall, New Jersey.
3.8 Battelle Pipe Fracture Encyclopedia, U.S. Nuclear Regulatory Commission, 1997.
3.9 ASTM A-516/A 516M-01 3.10 Point Beach DIT-1, dated 6/7/05. (Attachment B).
3.11 NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants", July 1980.
3.12 AC1349-01/349R-01, "Code Requirements for Nuclear Safety Related Concrete Structures (AC1349-01) and Commentary (ACI349R-01)", American Concrete Institute, 2001.
SOP04023-REV7.DOC Page 13 of 40 Rev. Date: 04-08-2005
S%;rgero t & L-uncdty Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 14 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop
- 4. METHODOLOGY This section describes the analysis models, solution process and acceptance criteria.
4.1 Analvsis Model Analysis models consist of static analysis models for stiffness calculations and a dynamic impact model. These finite element analysis (FEA) models are constructed by using the ANSYS computer program (Ref. 3.3).
The static analysis models include:
- A detailed model of RV flange and RV shell below the flange, including a nozzle resting on a supporting shoe.
- A similar detailed model of RV flange and RV shell below the flange with a support bracket resting on a supporting shoe.
- A detailed model of the hexagonal girder box frame supported by 6 pipe columns at the vertices.
- Piping models for the RCS hot legs and cold legs.
These models are used to construct static load-displacement diagrams for all steel components, which are on the impact load path. Section 4.1.1 describes the element types, and properties of this model.
The dynamic transient model consists of a two-mass model with springs and dash-pot in a vertical configuration. The top mass represents the falling head, and the bottom mass represents the target RV model supported by various springs. which represent the stiffness of the nozzlelbracket support, the girder box frame/column supports and the RCS piping.
Section 4.1.2 describes the parameters of this impact analysis model.
4.1.1 Static Analvsis Models ANSYS SHELL181, a four node finite strain element is used in FEA models for static load-deflection analyses. Non-linear material properties are modeled with a strength increase factor of 10% to account for the strain rate effects due to the dynamic impact.
Large deformation analysis option was selected to account for potential buckling and yielding in the structural components along the impact load path. ANSYS plastic pipe element PIPE20 and plastic elbow PIPE60 are used to model the RCS piping.
SOP04023-REV7.DOC Page 14 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 15 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 4.1.1.1 RV Nozzle Stiffness:
Figure 4-1 shows the FEA model for one of the RV nozzles. The Inlet and Outlet nozzle are similar, therefore, only one nozzle (Inlet nozzle) is modeled. Due to symmetry, a 60-degree sector of the RV is modeled with symmetric boundary conditions., The RV material is modeled using the true-stress true-strain curve of a SA-106 Grade B test specimen (Figure 5.3) from Battelle Pipe Fracture Encyclopedia, U.S. Nuclear Regulatory Commission, 1997, Ref. 3.8. It should be noted that, the actual RV material in Section 2.2.3(g) has a higher strength than SA-1 06 Grade B material, and irradiated material also has higher yield strength. However, since the RV flange, shell, nozzles and bracket are much stiffer than the support columns of the hexagonal girder box frame, a higher material strength will not significantly affect the analysis result. A strength increase factor of 1.1 is applied to the stress strain curve to account for the effect of impact loading.
Static vertical displacement is applied to the RV flange uniformly and a reaction force is calculated to construct the force-displacement curve of the RV nozzles.
ELNENTS AN REAJt NEo J1UN 1 2D00 16:07:43 RV Nozzle (Point Beech)
Figure 4.1 Reactor InletlOutlet Nozzle 60-degree Sector FEA Model SOP04023-REV7.DOC Page 15 of 40 Rev. Date: 04-08-2005 coAj2
L drcy e Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 16 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 4.1.1.2 RV Bracket Support Stiffness:
Figure 4-2 shows the FEA model for the RV bracket support with a 60-degree sector of the RV shell. The material properties of the bracket were assumed to be the same as that of the nozzles (See Assumption 2.3.1).
AN JUN 1 2005 15:57:28 Figure 4.2 Reactor Support Bracket 60-degree Sector FEA Model SOP04023-REV7.DOC Page 16 of 40 Rev. Date: 04-08-2005
rgaroerrt L~uncdtp-Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 17 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 4.1.1.3 Girder Box Frame and Support Column Stiffness:
Figure 4.3 shows a 60-degree sector FEA model for the girder box frame and a column support.
Load is applied at the mid-span of the girder box beam (the symmetric boundaries of the model).
The support column base is a pin connection with a base plate that is anchored to concrete base mat at elevation 10'-0. The column top plate is bolted to the girder box beam connection. The support columns pass through the vertical holes in the concrete shelf structure'. In the event the column buckled under compressive load it will come in contact with the surrounding concrete shelf. The gap between column and surrounding concrete is 1.75". However, there are channels and openings in the concrete shelf inner surface at elevations above 17'-1 3/4", therefore, only the outer half of surrounding concrete is modeled. Figure 4-3 shows the lateral gap elements along the support column.
The girder box material is A-5 17 Type F, a high strength steel with Sy =90 ksi and Su =110 Ksi.
A true-stress true-strain curve from a similar material strength from Ref. 3.8 was modified to match with the Sy and Su of A-517 Type F material. The true Stress Strain Curve for this material is shown in Figure 5.1. The true-stress true-strain of the support column material (A-53 Grade B) is available from the material database as shown in Figure 5.2. A strength factor of 1.1 is applied to the stress strain curve to account for the effect of impact loading. Static vertical displacement is applied uniformly to the support shoe located at the mid-span of the girder box beam, the reaction force is calculated to construct the force-displacement curve of a girder box frame support column.
SOP04023-REv7.DOC Page 17 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Caic. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 18 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop
.1 ELN s AN T!PI MlM JUR 8 2005 16:30:35 Figure 4.3 Hexagonal Girder Frame 60-degree Sector FEA Model 4.1.1.4 RCS Hot Leg Piping Stiffness:
The RCS hot leg piping includes a 12'-1 1 7/8" long straight pipe, a 50-degree vertical elbow and a short spool piece at the connection to the SG. The pipe OD is 34" and the pipe wall thickness is 2.5". The hot leg connection to the RV Outlet nozzle is modeled as a vertical roller that allows ,
vertical movement but restrains pipe rotation. To obtain the stiffness two bounding cases are analyzed. In Case 1, fixed boundary condition is used at the Steam Generator (SG) location and in -
Case 2 pinned boundary condition is used. In both cases, the pipe axial movement is ,
conservatively released to account for a potential horizontal movement of the SG. .-
The hot leg material is ASTM A-376 TP3 16. A true-stress true-strain curve for A-376 TP3 16 pipe ,.
test specimens from Ref. 3.8 is used for the RCS piping. The true Stress Strain Curve for this material is shown in Figure 5.4. A strength factor of 1.1 is applied to the stress strain curve to account for the effect of impact loading. Static vertical displacement is applied at the connection -
to the RV nozzle and the reaction force is calculated to construct the force-displacement curve for ,
the hot leg piping. Note that, the RCS Loop A and Loop B are identical, therefore the force-SOP04023-REV7.DOC Page 18 of 40 -
Rev. Date: 04-08-2005 CoC '
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 19 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop displacement curve is applicable for both Loop A and Loop B.
4.1.1.5 RCS Cold Leg Piping Stiffness:
The RCS cold leg piping includes a 9'-6 1/2" long straight pipe, a 26-degree horizontal elbow and a short spool piece at the connection to the Reactor Vessel (RV). The pipe OD is 32.25" OD and the pipe wall thickness is 2.375". The cold leg connection to the RV Inlet nozzle is modeled as a vertical roller that allows vertical movement but restraints pipe rotation. To obtain the stiffness two bounding cases are analyzed. In Case 1, fixed boundary condition is used at the Reactor Coolant Pump (RCP) location and in Case 2 pinned boundary condition is used. In both cases, the pipe axial movement is conservatively released to account for a potential horizontal movement of the RCP.
The cold leg material is the same as the hot leg material. Similar to the analysis for the hot leg, static vertical displacement is applied at the connection to the RV nozzle and the reaction force is calculated to construct the force-displacement curve for the cold leg piping.
4.1.2 Impact Analysis Model The impact analysis model is a vertical model consisting of a single mass representing the falling head, and a second single mass represents the RV target assembly. Figure 4.4 depicts this model; MI represents the 194,000 lb falling mass of the vessel head and M2 represents the 707,600-lb reactor vessel target including fuel, internals and water.
Contact stiffness Kl at the RV flange and the head on impact is taken as 100 times higher than the stiffness of the target RV and its supports such that it is nearly a rigid contact. The spring KI and the dashpot Cl are modeled using ANSYS COMBIN40 elements with an initial gap equal to 26.4 feet and the damping value is Cl = 244(K1 M2 ) per Ref. 3.7, where the damping ratio is 4 =
0.05 of the critical damping is considered in the analysis. This damping ratio is the same as the damping factor used in the Prairie Island Nuclear Station head drop analysis, Ref. 3.1, (See Assumption 2.3.3).
SOP04023-REV7.DOC Page 19 of 40 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 20 Project No.: 11165-065 Rev. No.1 Analysis of Postulated Reactor Head Load Drop H =26.4' KI H C, K4N- K2B K6 G K2HL ' Kay Figure 4.4 Dynamic Impact Model There are two parallel non-linear spring elements connected to M2. Spring K4N represents four RV nozzles, and spring K2B represents two RV bracket supports. These springs are connected in series to springs K6G, K2HL and K2CL which represents the girder box frame support columns, the RCS hot legs and cold legs. These non-linear springs are modeled with the ANSYS COMBIN39 based on the nonlinear load-displacement diagrams obtained from analysis model described in Sections 4.1.1.1,4.1.1.2,4.1.1.3,4.1.1.4 and 4.1.1.5.
4.2 Solution Process The impact analysis model is a dynamic transient analysis started at time t = 0 when the mass Ml is released. For a 26.4-foot drop, the impact occurs at time --4(2*26.4/g)= 1.276 sec. The time step at the impact is taken as smaller than 1/15 of the period of the contact spring. The frequency of the contact spring is fn = 4(KjfM2) / (2*3.14) = 190 hz. Therefore, the time step should be smaller than 1/(15* 190) - 0.00035 sec.
The analysis was performed for a time duration of 6 sec. This time duration is adequate to capture the rebounds of the head mass and the subsequent impacts until the head mass essentially stops and rests on top of the RV flange.
SOP04023-REV7.DOC Page 20 of 40 Rev. Date: 04-08-2005
rS-roo Am Lncl 'I Client: Nuclear Management Company Caic. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 21 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 4.3 Acceptance Criteria In addition to the requirement that the supporting structure should be capable of resisting the impact load, the following specific acceptance criteria shall be met:
- 1. After its contact with the flange, the dropped head should undergo an upward velocity.
Subsequent strikes and rebounds of the head must be associated with noticeably decreased rebound amplitudes approaching nearly zero rebound.
- 2. In order for the concrete shelf to provide lateral support for the supporting columns, the lateral force must not exceed the concrete shelf lateral load capacity.
SOP04023-REV7.DOC Page 21 of 40 Rev. Date: 04-08-2005
Bneroorr/. LunctV-Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 22 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop 5.0 CALCULATION 5.1 Stress-Strain Curve for the Girder Box Per the design input 2.2.2 (d), the girder box material Sy=90 ksi, which is about 10%
higher than the highest yield strength of all materials in Ref. 3.8, listed in Table 5.1.
Therefore, the stress strain curve No. 1 is adjusted by a factor of 10 % to match with the material properties of the girder box.
Table 5.1: Material Stress-Strain Data from PIPE FRACTURE ENCYCLOPEDIA, Ref. 3.8 Yield Ult. Percent Area Ramberg-Osgood Coefficients NO: MATERIAL Stren. Stren. Elong. Reduct. Sigop Epsqo lalpha n ksi ksi cksi 1 15NiCuMoNb5 81.2 98.6 22.0 55.0 84.2 0.002784 5.61 7.79 2 20MnMoNi55 74.4 89.9 25.0 73.0 74.4 0.002509 6.02 6.18 3 2_MnMoNi55 74.4 89.9 25.0 73.0 74.4 0.002509 6.06 6.2 4 20MnMoNi55 74.4 89.9 25.0 73.0 74.4 0.002509 6.02 6.18 5 SAW 73.5 97.3 53.4 58.9 73.5 0.002720 1.767 9.023 6 TIG Weld 74.8 100.5 36.5 ND 69.6 0.002531 5.363 5.5741 7 SAW 64.2 104.5 31.0 43.3 69.3 0.002100 0.781 7.020 8 20MnMoNi55 68.2 89.8 23.0 70.0 68.2 0.002397 1.69 2.07 9 API 5LX65 63.0 82.4 25.1 66.9 64.4 0.002420 1.210 10.840 10 Plate SAW 64.2 104.5 31.0 43.3 64.2 0.002289 0.662 9.260 11 STS410 Weld 61.3 76.4 61.1 0.002190 8.08 8.726 12 NiMoCr-Melt 60.2 91.8 19.0 49.0 60.2 0.002116 0.77 8.09 13 SAW 60.2 83.4 18.5 51.5 59.8 0.002136 1.325 9.7714 14 SAW 60.2 83.4 18.5 51.5 59.8 0.002136 1.325 4.0377 15 Inconel 54.0 86.9 NA NA 54.0 0.002086 3.11 3.37 16 STS410 Weld 54.2 85.6 53.5 0.002070 2.1 6.341 17 SAW 53.1 72.9 14.6 58.5 53.1 0.002060 1.547 11.221 18 SAW 51.7 80.7 20.4 NA 51.7 0.001920 1.770 7.882 19 STS42 Weld 47.1 69.6 38.4 47.1 0.001570 5.754 9.197 20 SAW WELD 47.1 67.6 31.5 44.2 47.1 0.001830 2.598 4.863 21 E7018 SMAW 45.7 97.0 26.0 46.4 45.7 0.001693 1.134 5.343 22 STS410 47.2 71.1 45.7 0.001640 2.9 8.822 23 TIG Weld 43.3 64.8 31.5 73.6 43.3 0.001680 2.680 9.033 24 A53 Gr B 44.0 66.2 30.5 ND 43.2 0.001442 6.018 6.287 25 Plate SAW 53.1 72.9 14.6 s8.5 53.1 0.002058 1.547 11.221 26 TP316 33.2 72.7 41.9 67.3 33.2 0.001250 0.565 8.280 SOP04023-REV7.DOC Page 22 of 40 Rev. Date: 04-08-2005
ergern; e.LunctyV Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 23 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop The stress-strain curve generated from the Ramberg-Osgood parameters for this test specimen is shown in Figure 5.1 using the Ramberg-Osgood stress-strain relationship, Ref 3.8.
Where, co = Referenced yield strain 0.002784, given in Table 5.1 ao = Referenced yield strength 84,200 psi, given in Table 5.1 a=5.61 n = 7.79 120000-100000-
- 80000 -
'a EI Un 40000-0 0.01 0.02 0.03 0.04 0.05 0.06 Strain (inWin)
Figure 5.1: Stress-Strain curve of a test specimen (psi)
Note that the above stress-strain curve does not include the strength adjustment factor of 1.1 SOP04023-REV7.DOC Page 23 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 24 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 5.2. Stress-Strain Curve for The Support Column The material properties for the support columns (ASTM A-53 Gr B, Ref. 3.10) are available in Ref. 3.8 and are listed as curve No. 24 in Table 5.1. The Ramberg-Osgood parameters for curve No. 24 are so = Referenced yield strain 0.001442 00 = Referenced yield strength 43,200 psi a = 6.018 n= 6.287 90000 80000 70000 -
0 60000 c 50000-c} 40000-30000-20000 10000 -
0 0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 Strain (inrin)
Figure 5.2: Stress-Strain curve for ASTM A-53 GrB (psi)
Note that the above stress-strain curve does not include the strength factor of 1.1 SOP04023-REv7.DOC Page 24 of 40 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Caic. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 25 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 5.3 Stress-Strain Curve for RV and Nozzle Available test data in digitized form from Ref. 3.8 for SA-106 Grade B is used. Note that the stress-strain curve in Figure 5.3 does not include the strength factor of 1.1 TRUE STRESS-STRAIN CURVE 120000
- 100000 0.
en 80000 7 I 60000 W 40000 F 20000 0 0.1 0.2 0.3 0.4 0.5 0.6 TRUE STRAIN Figure 5.3: Material Stress Strain Curve for SA-106 Gr B 5.4 Stress-Strain Curve for RCS Piping.
The material properties for the RCS hot legs and cold legs (ASTM A-376 TP3 16, Ref.
3.2(q)) are available in Ref. 3.8. The Ramberg-Osgood parameters for the stress strain curve are:
co = Referenced yield strain 0.00125 c0 = Referenced yield strength 33,200 psi c = 0.565 n=8.28 SOP04023-REV7.DOC Page 25 of 40 Rev. Date: 04-08-2005
0ugtrovm Lucancdy Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 26 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop 80000 70000 60000 a 50000 FI mm 40000 en 30000 20000 10000 0
0 0.1 0.2 0.3 0.4 0.5 Strain (inlin)
Figure 5.4: Stress-Strain curve for ASTM A-376 TP316 (psi)
Note that the above stress-strain curve does not include the strength factor of 1.1 5.5 Load-Deflection Curve Analysis is performed using the static models as described in Section 4.1.
Displacement is applied gradually on the RV flange for the RV nozzle and bracket models, and on the mid-span of the girder box beam with support column. Figures 5.5.1, 5.5.2 and 5.5.3 show the load-deflection curve for the RV nozzle, the support bracket and hexagonal girder box frarne. Figures 5.5.4 through 5.5.7 show the load deflection curves for Hot and Cold legs. See Attachment D for ANSYS Output.
SOP04023-REV7.DOC Page 26 of 40 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 27 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop POSTZ6 AN Y 27 2005 17.30 48 Dl-i (2h
.. .; .5 .7 .
.a 4ir.b a/
131 ap..
Dofl.ction (inch)
Figure 5.5.1 Vertical Load-Deflection Curve for the RV Nozzle KS?26 AN JUN 6 2005 14:07:21 PD!r ND. 1 i .
. .4 . . .
Defleation (inch) iRV aaket Foirce-apDmq ont Qave (Point 9mb)
Figure 5.5.2 Vertical Load-Deflection Curve for the RV Support Bracket -
2 2
S SOP04023-REV7.DOC Page 27 of 40 -
Rev. Date: 04-08-2005 A
3 Cog '
amSa Luri.V.n Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 28 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 26 AN 3iN 8 200S 09 4:31 POT NO. 2
'4
,--7 A /""It I
0 1 2 4 1fltioi(inch)'.
Gi,¢, ,-, Fr6 C , , _, av ,(Poit UF N, , , , ,,, , , I Figure 5.5.3 Vertical Load-Deflection Curve for the Girder Box Frame POT26 AN JUN 16 2005 10:12:48 154.-93 3II "3
.52. 5. . .
0 1 2
.5 1.S 2.5 3. 5 4.5 DefleCti.. (inch)
RCS HOt Leg FO.Ce-Di.place...t Cur.e (Point Be-ch)
Figure 5.5.4 Vertical Load-Deflection Curve for the Hot Leg Case 1 SOP04023-REV7.DOC Page 28 of 40 Rev. Date: 04-08-2005
W *nods 1ucV Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 29 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop P04T26 A~N JUN 16 2005 10:24:19 117 6.60-G
.29b14
.0 S 0 2 3 4 5
.0 0.0 2.5 3.0 4.5 Deflection (inch)
RCS Cold Force-Dieplacee..t Curve (Point Beach)
Figure 5.5.5 Vertical Load-Deflection Curve for the Cold Leg Case 1 POT826 AN J~e 17 2005 0 1 2 3 4 4
.4 1.5 2.5 5.5 4.5 Deflection (inch)
RC$ H400 Leg F-ce-Displ ..... nt C-rv (Point DBeach)
Figure 5.5.6 Vertical Load-Deflection Curve for the Hot Leg Case 2 SOP0402Z3-REV7.DOC Page 29 of 40 Rev. Date: 04-08-2005 coc
L ajrudcI Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 30 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop POT26 uAN
,UN 17 2005 13 34 53
. 0 2.5 3. .5 Defleotio. (i.coh)
RCS Cold Fore-DiepI ...... t Curve (Poist Be..h)
Figure 5.5.7 Vertical Load-Deflection Curve for the Cold Leg Case 2 SOP04023-REV7.DOC Page 30 of 40 Rev. Date: 04-08-2005 010
7 Sr5;1rondf LndyszI Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 31 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop The above load deflection curves are used to develop the non-linear springs for the impact model described in Section 4.1.2. The digitized data of the curve is provided in ANSYS Post26 output (Attachment D); with variable 4 (4 PROD) as displacement, and with variable 5 (5 PROD) as force. The load of the nozzle load-deflection curve (Figure 5.5.1) is multiplied by 4 to construct the non-linear spring K4N of the dynamic impact model. Similarly, the load deflection curve for the bracket is multiplied by 2 to construct spring K2B and the load deflection curve for the girder box frame is multiplied by 6 to construct spring K6G. Similarly springs K2HL and K2CL are obtained for the Hot Leg and the Cold leg. The following load-deflection data shown in Tables 5.5.1 through 5.5.7 is the input for the dynamic impact model.
Table 5.5.1 Non-Linear Spring K4N for 4 RV Nozzles Deflection (in) Force (Ibf) Deflection Force (Ibf)
__ _ __ _(in) 0.05 17169200 (*) -0.45 -79291600 0.00 0 -0.50 -81948800
-0.05 -17169200 -0.55 -84337600
-0.10 -33971800 -0.60 -86482800
-0.15 -47442800 -0.65 -88423600
-0.20 -56676400 -0.70 -90225600
-0.25 -63456400 -0.75 -91951600
-0.30 -68632800 -0.80 -93592800
-0.35 -72697600 -0.85 -95137600
-0.40 -76234800 -0.90 -96643200 Table 5.5.2 Non-Linear Spring K2B for 2 RV Support Bracket Deflection (in) Force (Ibf) Deflection Force (lbf)
__ _ _ __ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _(in ) _ _ _ _ _ _ _
0.05 4040980 -0.45 -8733200 0.00 0 -0.50 -9086580
-0.05 -4040980 -0.55 -9423600
-0.10 -5183280 -0.60 -9732820
-0.15 -5851220 -0.65 -10041080
-0.20 -6437560 -0.70 -10276160
-0.25 -6966620 -0.75 -10514540
-0.30 -7458740 -0.80 -10775800
-0.35 -7919800 -0.85 -10984960
-0.40 -8342720 -0.90 -11178660 SOP04023-REV7.DOC Page 31 of 40 Rev. Date: 04.08-2005
- g. nd Client: Nuclear Management Company Caic. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 32 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop Table 5.5.3 Non-Linear Spring K6G for the Girder Box Frame Deflection (in) Force (Ibf) Deflection Force (Ibf)
(in) 0.046 168999*6 -2.250 -10080420 0.000000 0 -2.500 -10341720
-0.046 6*168999 -2.750 -10479120
-0.433 -6* 0.155248E+07 -3.000 -10621620
-0.621 -6* 0.136627E+07 -3.250 -10766340
-0.750000 -6* 0.152094E+07 -3.500 -10919100
-1.000000 -9643320 -3.750 -11039940
-1.500000 -9660360 -4.000 -11150760
-1.750000 -9579840 -4.250 -11295960
-2.000000 -9528900 -4.500 -11455560 Note (*): The slope at the original is extended to the tensile side as required for the nonlinear spring element COMBIN39.
Table 5.5.4 Non-Linear Spring K2HL for the RCS Hot Leg Case 1 Deflection (in) Force (Ibf) Deflection Force (Ibf)
(in) 0.249 303048 -2.250 -1865454 0.000 0 -2.500 -1928182
-0.249 -303048 -2.750 -1976648
-0.500 -606148 -3.000 -2019480
-0.750 -909306 -3.250 -2063260
-1.000 -1195336 -3.500 -2095340
-1.250 -1414618 -3.750 -2129240
-1.500 -1575728 -4.000 -2160480
-1.750 -1695666 -4.250 -2184760
-2.000 -1788510 -4.500 -2210940 SOP04023-REV7.DOC Page 32 of 40 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 33 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop Table 5.5.5 Non-Linear Spring K2CL for the RCS Cold Leg Case 1 Deflection (in) Force (Ibf) Deflection Force (Ibf)
(in) 0.249 672590 -2.250 -2587180 0.000 0 -2.500 -2649620
-0.249 -672590 -2.750 -2691480
-0.500 -1341114 -3.000 -2726440
-0.750 -1824580 -3.250 -2770160
-1.000 -2093840 -3.500 -2805040
-1.250 -2255660 -3.750 -2839020
-1.500 -2373100 -4.000 -2864840
-1.750 -2458260 -4.250 -2903900
-2.000 -2532760 -4.500 -2931600 Table 5.5.6 Non-Linear Spring K2 L for the RCS Hot Leg Case 2 Deflection (in) Force (lbf) Deflection Force (Ibf) l_ (in) 0.248 123650 -2.250 -918874 0.000 0 -2.500 -953504
-0.248 -123650 -2.750 -987832
-0.500 -247456 -3.000 -1008254
-0.750 -371424 -3.250 -1030556
-1.000 -495554 -3.500 -1050996
-1.250 -619846 -3.750 -1065946
-1.500 -730672 -4.000 -1082836
-1.750 -820494 -4.250 -1098132
-2.000 -875780 -4.500 -1106792 SOP04023-REV7.DOC Page 33 of 40 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 34 Project No.: 11165-065 Rev. No. I Analysis of Postulated Reactor Head Load Drop Table 5.5.7 Non-Linear Spring K2C L for the RCS Cold Leg Case 2 Deflection (in) Force (lbf) Deflection Force (Ibf)
(in) 0.249 157962 -2.250 -932728 0.000 0 -2.500 -958898
-0.249 -157962 -2.750 -980356
-0.500 -315926 -3.000 -999902
-0.750 -473894 -3.250 -1016450
-1.000 -621848 -3.500 -1029442
-1.250 -731802 -3.750 -1049196
-1.500 -806628 -4.000 -1056698
-1.750 -860988 -4.250 -1070554
-2.000 -900302 -4.500 -1087462 5.6 Column Pin Connection Shear Load Capacity.
Pin material: ASTM A490, Pin diameter: 4 inch Tensile Strength: a tp:= 164900 psi Yield Strength: o yp := 148800 psi With a strength increase factor of 10% under impact load a tp:= I.I .o t a yp:=I1.l.0 2
Shear Area: A 5:=2 I in2 A sp= 12.566.in Faulted Load 0.42-c tp 4
Shear Allowable: 0 allow.p:= mii 0 c allow.p= 7.618-10 Ipsi 0.60La 2-Load capacity of the pin: Fmax:= 2-iAsp'Gallow.p Fmax= 1.915-10 6 Ibf SOP04023-REV7.DOC Page 34 of 40 Rev. Date: 04-08-2005
- ~- 0 rm rk" LU nv9-g'1t Lnc" p Client
- Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 35 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop 6.0 RESULTS / CONCLUSION The load deflection curve is used in a dynamic transient analysis as described in Section 4.1.2.
The output files for the dynamic transient analysis are documented in Attachment D. The output variable definitions and the extreme values are listed on Tables 6.1a and 6.1b.
Table 6.la Post26 Summary O Variable Extreme Values and Definitions for Case 1 VARI TYPE IDENTIFI NAME DEFINITION MINIMUM AT TIME MAXIMUM AT TIME ABLE _ _ERS 2 NSOL 2 UY RPV RPV Displacement -2.716 2.396 -0.2912E-01 0.5120 3 NSOL 3 UY UY RV Head Displacement -319.4 2.896 0.000 0.10E-07 4 OPER 4 FILL 4 Drop Height 316.8 0.10E-07 316.8 0.10E-07 5 OPER 5 ADD RV Head RV Head Position -2.633 2.896 316.8 0.10E-07 6 RFOR I FY TotalF Total Load on Supports -0.1386E+08 1.336 0.1038E+08 1.309 7 ESOL I SMIS 1 f2b Load on Support Brackets -0.2854E+07 1.309 0.2584E+07 1.336 8 OPER 8 DERI Y Velo Impact Velocity -494.6 1.281 157.2 1.283 9 RFOR 4 FY girder Load on Girder Columns -0.1386E+08 1.336 0.1038E+08 1.309 11 ESOL 2 SMIS I F4n Load on Nozzles -0.1212E+08 1.309 0.1098E+08 1.336 Table 6. lb Post26 Summary f Variable Extreme Values and Definitions for Case 2 VARI TYPE IDENTIFI NAME DEFINITION MINIMUM AT TIME MAXIMUM AT TIME ABLE ERS 2 NSOL 2 UY RPV RPV Displacement -3.200 2.122 -0.3259E-01 0.6120 3 NSOL 3 UY UY RV Head Displacement -319.9 2.496 0.000 0.1000E-4 XOPER 4 FILL 4 Drop Height 316.8 0.10E-07 316.8 0.10E-07 5 OPER 5 ADD RV Head RV Head Position -3.100 2.496 316.8 0.IOE-07 6 RFOR I FY TotalF Total Load on Supports -0.1215E+08 1.342 0.1070E+08 1.314 7 ESOL I SMIS I f2b Load on Support Brackets -0.2425E+07 1.314 0.2152E+07 1.342 8 OPER 8 DERI Y Velo Impact Velocity -494.6 1.281 171.5 1.283 9 RFOR 4 FY girder Load on Girder Columns -0.1215E+08 1.342 0.1070E+08 1.314 H1 ESOL 2 SMIS I F4n Load on Nozzles -0.1030E+08 1.3 14 0.9143E+07 11.342 Figures 6-1 through 6-4 depict the results of Case 1 and Case 2 analyses showing the RV Head and vessel flange time-displacement histories, respectively. The first impact occurs at time 1.281 sec with an impact velocity of 494.6 in/sec. The exact impact time is q1(2*26.4f32.2) = 1.281 sec and the exact impact velocity is q(2*26.4*32.2) = 41.233 Wl/s or 494.796 in/s. This indicates that ANSYS accurately calculates the input impact energy.
SOP04023-REV7.DOC Page 35 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 36 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop The maximum dynamic displacement of the RV is -2.72 inches for Case 1 and -3.20 inches for Case 2. These displacements are less than 3.375". Therefore, the girder box will not come in contact with the concrete shelf.
The maximum impact load on the column foundation (Case 2) is 10.70E+06 lbf, or 1.783E+6 lbf on each column, which is smaller than the maximum load of the load-deflection curve shown in Figure 5.5.3. The load per support column was evaluated and found acceptable for the pin connection in Section 5.6 and for the concrete basemat (See Attachment A).
The concrete shelf is only required to provide lateral support for the stability of the support columns located within the shelf. Between elevations 17'- 1 3" and 29'- 10 112", the concrete shelf will be subjected to lateral stability loads directed away from the reactor centerline towards the biological shield wall as shown in Tables 6.2. Thus, the concrete shelf will simply transfer these loads to the biological shield wall in bearing. By engineering judgment, the biological shield wall is considered adequate for these minor loads without any detailed evaluations. At elevation 14'-1 3/4", the concrete shelf will be subjected to lateral loads from each support column which are directed towards the centerline of the reactor (See Table 6.2). The lateral load between the elevation 14'-1I"3 and Elev. 17'- 1 3 h"is evaluated in the Attachment A and is found to be acceptable.
Table 6.2: Result Summary (bounding values from Casel and Case 2) R2 RV Head Drop Analysis Results Results Notes Vertical Load per Support Column (Ibf) 1.783E+6 < 1.909E+6 Lbf Capacity Vertical Load per Nozzle (lbf) 3.030 E+6 <0.241608E+08 Lbf Capacity Vertical Load per Support Bracket (Ibf) 1.427E+06 < 0.55893E+07 Lbf Capacity Lateral Load between Elev. 14'-1 4"and .13435E+6 Radial Force, toward RV Elev. 17'- 1 3/4" per Support Column .02635E+6 Tangential Force Lateral Load between Elevationl7'- 1 3/4" .09101E+6 Radial Force, toward Shield Wall and 29'- 10 'A" (lbf) per Support Column .04353E+6 Tangential Force Maximum RV Dynamic Deflection (in) 3.20 < 3.375" SOP04023-REV7.DOC Page 36 of 40 Rev. Date: 04-08-2005
Urgr mcl.- LUnd.y.g Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 36a Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop For the RCS nozzle stress evaluation, the reaction forces and moments from the RCS piping at the RV nozzles are extracted for a deflection of 3.2" from the force-deflection analyses. The RCS piping reaction force is 1.377E+6 lbf and the RCS piping reaction bending moment is 94.76E+6 in-lbf, which are from the bounding Cold Leg fixed-fixed model at 3.2" deflection (See page D105). These loads are used in the RV nozzle stress evaluation as follows:
The sum of moments at the RV due to nozzle reaction of 3.03E+6 lbf (refer to Table 6.2),
moment due to RCS piping reaction force of 1.377E+6 lbf and the RCS piping reaction bending moment of 94.76E+6 in-lbf is:
M = 3.03E+6 lbf
- 16.156" + 1.377E+6 lbf
- 27.5" + 94.76E+6 in-lbf = 181.580E+6 in-lbf.
where, 16.156" is the distance from the nozzle support saddle to the middle of the RV wall (Ref.
3.2(a)) and 27.5" is the distance from the RCS nozzle ends to the middle of the RV wall (Ref.
3.2(o)).
An equivalent reaction force, FEQ at the nozzle support saddle, which produces the same total moment M at the RV nozzle junction is:
FEQ= M / 16.156" = 181.580E+6 in-lbf/ 16.156" = 11.239E+6 lbf.
The Von Mises stress in the RV nozzles extracted from the nozzle load-deflection analysis for 11.239E+6 lbf load is shown in Figure 6.5. The maximum Von Mises stress of 41.234 ksi (membrane plus bending) is less than the ASME Section III, Appendix F allowable stress for membrane stress intensity of 0.7 Su (0.7*80 ksi = 56 ksi). Where Su = 80 ksi for A-508 Class 2 material per ASME Section III, Appendices, 1977 Edition.
Similarly, the Von Mises stress in the RV support brackets extracted from the bracket load-deflection analysis for 1.427E+06 lbf load is shown in Figure 6.6. The maximum Von Mises stress of 39.2 ksi (membrane plus bending) is less than the ASME Section III, Appendix F allowable stress for membrane stress intensity of 0.7 Su (0.7*70 ksi = 49 ksi). Where Su = 70 ksi for A-516 Grade 70 per ASME Section III, Appendices, 1977 Edition.
Figures 6.7 and 6.8 provide the stress and lateral deflection plots, respectively, for the girder box frame support columns at 3.2 inches RV deflection.
SOP04023-REV7.DOC Page 36a of 40 Rev. Date: 04408-2005
S;EWr0 n LALnclyt Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 37 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop PT26 AN JUN 17 2005 PR 19:49:02 RV Had PV=Ib NO. 1 240 INH 160 40 II 41 1.2 2.4 3.6 4.0 6 11 3 4.2 5.4 TIM RV Had 26.4 root DV Tio Hi pSy-Ca-. 1 Figure 6.1: RV Head Drop Time History Case 1 AN RW JUNJ17 2001 19:49:02 PLO NO. 2 1.2 i 2 4 3 3 .6 4.. 4.8 1.
- Tf_l.
Dkt9aut- Tim, I~so RV w Scn Figure 6.2: RV Response Time History Case 1 SOP04023-REV7.DOC Page 37 of 40 Rev. Date: 04-08-2005 Cif
Lncrity L
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 38 Project No.: 11165-065 Rev. No. 1 Analysis of Postulated Reactor Head Load Drop gST26 AN Eons RNV rmIJ~a4 PLOP . I L.2 2.4
.e6 1.8 3 4.2 5,4 TDI RVHead 26.4 Jot p Tip m sto N.C. 2..
Figure 6.3: RV Head Drop Time History Case 2 pi 26 AN I RWV JUN 17 2005 19:59:14 PLOTN. 2
.6 1. 1.68 2.4 3.6 4.3 6 3 4.2 2.4
_. Ti- .- _..... 2 -. _.. -- . .. ...........
. . .J. . . ...... I -,,-
Figure 6.4: RV Response Time History Case 2 SUP04023-REV7.DOC Page 38 of 40 Rev. Date: 04-08-2005
Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 39 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop NODAL SOLUTION AN JUN 22 2005 SUa =4 11:37:58 S3EV (AVG)
DNX =.140415 . 131568 8MN =.131568 mu43=41234 4582 9163 13745 18326 22908 27488 32071 36653 41234 Figure 6.5: Nozzle Von Misses Stress Under the Maximum RV Deflection R2 -
JUN822 2005 18 20:06 1749 69 4356 8711 13067 17422 21778 26133 30489 34844 39200 Figure 6.6: Bracket Von Misses Stress Under the Maximum RV Deflection SOP04023-REV7.DOC Page 39 of 40 Rev. Date: 04-08-2005 CI
LeuncdtV Client: Nuclear Management Company Calc. No 2005-06760 Project: Point Beach Nuclear Generating Plant Page No. 40 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop NODAL SOLUTION AN 8TEP-1 a~~3 JUN 1:222 205 41:3 8US =10 17 4131 T01812 .83 8EQV (AVG)
OMS =3.211 i 311X-07 amCc-. 311-07 SMX =71153 7906 15812 23718 31624 39530 47435 55341 63247 71153 R2 Figure 6.7: Support Column Von Misses Stress NODAL SOLUTION AN 8TEP=13 JUN 22 2003 SUB =10 17:45:09 TIML-12 .83 UZ (AVG)
R5YS= X -1.769 Di =3 .211 SMN =-1.769 -1.371 SMS =1.806
-.97407
-. 576847
-. 179624
.217599
.614822 1.012
- 1. 409 1.806 Figure 6.8: Support Column Lateral Deflection SOP04023-REV7.DOC Page 40 of 40 Rev. Date: 04-08-2005 C /y-
SUaromrm LunclfV Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D102 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop Reaction Force and Moment for Cold Leg Piping Moment.out
- ANSYS POST26 VARIABLE LISTING ****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 0.20000E-01 424395. -6725.89 -0.500000E-02 0.40000E-01 848791. -13451.8 -0.1000OOE-01 0.70000E-01 0.148538E+07 -23540.6 -0.175000E-01 0.11500 0.244027E+07 -38673.9 -0.287500E-01 0.18250 0.387261E+07 -61373.8 -0.456250E-01 0.28250 0.599459E+07 -95003.2 -0.706250E-01 0.38250 0.811656E+07 -128633. -0.956250E-01 0.48250 0.102385E+08 -162262. -0.120625 0.58250 0.123605E+08 -195892. -0.145625 0.68250 0.144825E+08 -229521. -0.170625 0.78250 0.166045E+08 -263151. -0.195625 0.88250 0.187265E+08 -296780. -0.220625 0.94125 0.199731E+08 -316538. -0.235312 1.0000 0.212198E+08 -336295. -0.250000 1.0200 0.216442E+08 -343021. -0.255000 1.0400 0.220686E+08 -349747. -0.260000 1.0700 0.227052E+08 -359836. -0.267500 1.1150 0.236601E+08 -374969. -0.278750 1.1825 0.250924E+08 -397669. -0.295625 1.2825 0.272144E+08 -431299. -0.320625 ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY uy 1.3825 0.293364E+08 -464929. -0.345625 1.4825 0.314584E+08 -498559. -0.370625 1.5825 0.335804E+08 -532189. -0.395625 1.6825 0.357023E+08 -565819. -0.420625 1.7825 0.378243E+08 -599449. -0.445625 1.8825 0.399420E+08 -632954. -0.470625
- 1. 9413 0.411557E+08 -651759. -0.485313 2.0000 0.423689E+08 -670557. -0.500000 2.0200 0.427933E+08 -677282. -0.505000 2.0400 0.431957E+08 -683373. -0.510000 2.0700 0.438148E+08 -692962. -0.517500 2.1150 0.447476E+08 -707460. -0.528750 2.1825 0.461127E+08 -728193. -0.545625 2.2825 0.480960E+08 -757677. -0.570625 2.3825 0.500690E+08 -786789. -0.595625 2.4 825 0.519341E+08 -814047. -0.620625 2.5825 0.535165E+08 -837043. -0.645625 2.6825 0.549715E+08 -858330. -0.67062S 2.7825 0.564396E+08 -878407. -0.695625 2.8825 0.576300E+08 -893080. -0.720625
- ANSYS POST26 VARIABLE LISTING *****
SOP04023-REV7.DOC Page D102 of D107 Rev. Date: 04-08-2005
e £roo Lundy ...
Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D103 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop TIME 12 MZ 12 FY 12 UY MZ FY Uy 2.9413 0.583192E+08 -902204. -0.735313 3.0000 0.590733E+08 -912290. -0. 750000 3.0200 0.593889E+08 -918207. -0.755000 3.0400 0. 595787E+08 -919713. -0.760000 3.0700 0.599678E+08 -924699. -0.767500 3.1150 0.606170E+08 -933586. -0.778750 3.1825 0.615507E+08 -945639. -0.795625 3.2825 0.626478E+08 -958235. -0.820625 3.3825 0.637320E+08 -971382. -0.845625 3.4825 0.648381E+08 -985968. -0.870625 3.5825 0.659303E+08 -0.100083E+07 -0.895625 3.6825 0.669762E+08 -0.101469E+07 -0.920625 3.7825 0.678061E+08 -0.102524E+07 -0.945625 3.8825 0.685155E+08 -0.103322E+07 -0.970625 3.9413 0.689897E+08 -0.104037E+07 -0.985313 4.0000 0.694378E+08 -0.104692E+07 -1.00000 4.0200 0.696646E+08 -0.105190E+07 -1.00500 4.0400 0.696416E+08 -0.105115E+07 -1.01000 4.0700 0.698729E+08 -0.105430E+07 -1.01750 4.1150 0.701028E+08 -0.105615E+07 -1.02875
- ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 4.1825 0.706338E+08 -0.106293E+07 -1.04562 4.2825 0.715184E+08 -0.107407E+07 -1.07062 4.3825 0.721805E+08 -0.108201E+07 -1.09562 4 .4 825 0.728246E+08 -0.109011E+07 -1.12062 4.5825 0.734803E+08 -0.109884E+07 -1.14562 4.6825 0.739848E+08 -0.110576E+07 -1.17062 4.7825 0.746159E+08 -0.111384E+07 -1.19562 4.8825 0.750941E+08 -0.112029E+07 -1.22062 4.9412 0.754325E+08 -0.112499E+07 -1.23531 5.0000 0.756971E+08 -0.112783E+07 -1.25000 5.0200 0.758787E+08 -0.113239E+07 -1.25500 5.0400 0.757702E+08 -0.113115E+07 -1.26000 5.0700 0.759871E+08 -0.113357E+07 -1.26750 5.1150 0.760643E+08 -0.113308E+07 -1.27875 5.1825 0.764735E+08 -0.113816E+07 -1.29562 5.2825 0.771922E+08 -0.114746E+07 -1.32062 5.3825 0.776058E+08 -0.115332E+07 -1.34562 5.4825 0.780422E+08 -0.115906E+07 -1.37062 5.5825 0.784186E+08 -0.116252E+07 -1.39562 5.6825 0.788220E+08 -0.116577E+07 -1.42062
- ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 5.7825 0.791788E+08 -0.117156E+07 -1.44562 5.8825 0.796850E+08 -0.117865E+07 -1.47062 5.9412 0.798673E+08 -0.118296E+07 -1. 48531 6.0000 0.801206E+08 -0.118655E+07 -1.50000 6.0200 0.802668E+08 -0.119071E+07 -1.50500 6.0400 0.799163E+08 -0.118437E+07 -1.51000 SOP04023-REV7.DOC Page D103 of D107 Rev. Date: 04-08-2005
53rger LUncly'V-'
7l Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D104 Project No.: 11165-065 Rev. No.2 Analysis of Postulated Reactor Head Load Drop 6.0700 0.798194E+08 -0.118209E+07 -1.51750 6.1150 0.799854E+08 -0.118327E+07 -1.52875 6.1825 0.803767E+08 -0.118800E+07 -1.54562 6.2825 0.807110E+08 -0.118993E+07 -1.57062 6.3825 0.813433E+08 -0.119952E+07 -1.59562 6.4825 0.817101E+08 -0.120757E+07 -1.62062 6.5825 0.820007E+08 -0.121133E+07 -1.64562 6.6825 0.823751E+08 -0.121508E+07 -1.67062 6.7825 0.826950E+08 -0.122054E+07 -1. 69562 6.8825 0.830071E+08 -0.122503E+07 -1.72062 6.9412 0.831869E+08 -0.122663E+07 -1.73531 7.0000 0.833753E+08 -0.122913E+07 -1.75000 7.0200 0.833458E+08 -0.123062E+07 -1.75500 7.0400 0.833301E+08 -0.123070E+07 -1.76000
- ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 7.0700 0.834657E+08 -0.123225E+07 -1.76750 7.1150 0.837525E+08 -0.123558E+07 -1.77875 7.1825 0.838292E+08 -0.123518E+07 -1.79562 7.2825 0.839699E+08 -0.123183E+07 -1.82062 7.3825 0.843209E+08 -0.123869E+07 -1.84562 7.4 825 0.847753E+08 -0.124835E+07 -1.87062 7.5825 0.849487E+08 -0.124948E+07 -1.89562 7.6825 0.851295E+08 -0.125055E+07 -1. 92062 7.7825 0.854792E+08 -0.125654E+07 -1. 94562 7.8825 0.857903E+08 -0.126190E+07 -1. 970 62 7.9412 0.859026E+08 -0.126436E+07 -1.98531 8.0000 0.860246E+08 -0.126638E+07 -2.00000 8.0200 0.861638E+08 -0.127056E+07 -2.00500 8.0400 0.857695E+08 -0.126321E+07 -2.01000 8.0700 0.856208E+08 -0.125989E+07 -2.01750 8.1150 0.857586E+08 -0.126036E+07 -2.02875 8.1825 0.861671E+08 -0.126481E+07 -2.04562 8.2825 0.865852E+08 -0.127076E+07 -2.07062 8.3825 0.868848E+08 -0.127177E+07 -2.09562 8.4825 0.871421E+08 -0.127707E+07 -2.12062
- ANSYS POST26 VARIABLE LISTING ***
TIME 1 2 MZ 12 FY 12 UY MZ FY Uy 8.5825 0 .875007E+08 -0.128457E+07 -2.14562 8.6825 0 .876199E+08 -0.128738E+07 -2.17062 8.7825 0 .877267E+08 -0.128432E+07 -2.19562 8.8825 0 .878697E+08 -0.128497E+07 -2.22062 8.9412 0 .880417E+08 -0.128872E+07 -2.23531 9.0000 0 .882377E+08 -0.129359E+07 -2.25000 9.0200 0 .883649E+08 -0.129766E+07 -2.25500 9.0400 0 .883415E+08 -0.129822E+07 -2.26000 9.0700 0 .884293E+08 -0.129873E+07 -2.26750 9.1150 0 .884590E+08 -0.129831E+07 -2.27875 9.1825 0 0.887275E+08 -0.130141E+07 -2.29562 9.2825 0 0.888712E+08 -0.130291E+07 -2.32062 9.3825 0 .892329E+08 -0.130855E+07 -2.34562 9.4825 0 .893893E+08 -0.131109E+07 -2.37062 SOP04023-REV7.DOC Page D104 of D107 Rev. Date: 04-08-2005
1.
U4-tronrm Lurlcjv.&.
Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D105 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop 9.5825 0.895933E+08 -0.131369E+07 -2.39562
- 9. 6825 0.897010E+08 -0.131215E+07 -2.42062 9.7825 0.897888E+08 -0.130928E+07 -2.44562 9.8825 0.900510E+08 -0.131436E+07 -2.47062 9.9412 0.902479E+08 -0.132011E+07 -2.48531 10.000 0.904231E+08 -0.132481E+07 -2.50000
- -** ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 10.040 0.902520E+08 -0.132192E+07 -2.51000 10.080 0.902799E+08 -0.132311E+07 -2.52000 10.140 0.906014E+08 -0.132812E+07 -2.53500 10.230 0.907252E+08 -0.132617E+07 -2.55750 10.330 0.908982E+08 -0.132946E+07 -2.58250 10.430 0.911629E+08 -0.133460E+07 -2.60750 10.530 0.913661E+08 -0.133773E+07 -2.63250 10.630 0.913528E+08 -0.133288E+07 -2.65750 10.730 0.914582E+08 -0.133326E+07 -2.68250 10.830 0.916669E+08 -0.133682E+07 -2.70750 10.930 0.919494E+08 -0.134231E+07 -2.73250 11.000 0.920545E+08 -0.134574E+07 -2.75000 11.040 0.921470E+08 -0.134822E+07 -2.76000 11.080 0.919905E+08 -0.134627E+07 -2.77000 11.140 0.920952E+08 -0.134472E+07 -2.78500 11.230 0.924421E+08 -0.134962E+07 -2.80750 11.320 0.926777E+08 -0.135435E+07 -2.83000 11.420 0.928752E+08 -0.135552E+07 -2.85500 11.520 0.929216E+08 -0.135567E+07 -2.88000 11.620 0.931726E+08 -0.136044E+07 -2.90500 ANSYS POST26 VARIABLE LISTING ****
TIME 12 MZ 12 FY 12 UY MZ FY uy 11.720 0.933561E+08 -0.136344E+07 -2.930. DO 11.820 0.933894E+08 -0.136030E+07 -2.955 00 11.920 0.935499E+08 -0.136105E+07 -2.980 00 12.000 0.936139E+08 -0.136322E+07 -3.000' 00 12.040 0.938356E+08 -0.137002E+07 -3.010' 00 12.080 0.937638E+08 -0.136920E+07. -3.020i 00 12.140 0.938510E+08 -0.136800E+07 -3.035, DO 12.230 0.941096E+08 -0.137234E+07 -3.057! 50 12.320 0.943009E+08 -0.137590E+07 -3.080 00 12.420 0.942690E+08 -0.136999E+07 -3.1051DO 12.520 0.943028E+08 -0.136894E+07 -3.130. DO 12.620 0.944457E+08 -0.137212E+07 -3.155' jO 12.720 0.947139E+08 -0.137826E+07 -3.180 00 12.820 0.947631E+08 -0.137698E+07 -3.205 Do 12.920 0.949303E+08 -0.138154E+07 -3.230 00 13.000 0.951043E+08 -0.138508E+07 -3.250' DO 13.040 0.952325E+08 -0.138940E+07 -3.260' DO 13.080 0.951251E+08 -0.138789E+07 -3.270' DO 13.140 0.952758E+08 -0.138922E+07 -3.2851 DO 13.230 0.953967E+08 -0.138973E+07 -3.307! 50
- ANSYS POST26 VARIABLE LISTING *****
SOP04023-REV7.DOC Pacge D105 of D107 Rev. Date: 04-08-2005
Srgenr!f & L-uncty"t Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D106 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop TIME 12 MZ 12 FY 12 UY MZ FY UY 13.320 0.955812E+08 -0.139301E+07 -3.33000 13.420 0.955373E+08 -0.138800E+07 -3.35500 13.520 0.957080E+08 -0.139090E+07 -3.38000 13.620 0.959258E+08 -0.139551E+07 -3.40500 13.720 0.961592E+08 -0.140114E+07 -3.43000 13.820 0.962107E+08 -0.140061E+07 -3.45500 13.920 0.962162E+08 -0.139859E+07 -3.48000 14.000 0.963877E+08 -0.140252E+07 -3.50000 14.040 0.963452E+08 -0.140394E+07 -3.51000 14.080 0.965241E+08 -0.140788E+07 -3.52000 14.140 0.963586E+08 -0.139974E+07 -3.53500 14.230 0.964579E+08 -0.139429E+07 -3.55750 14.330 0.966906E+08 -0.139513E+07 -3.58250
- 14. 430 0.967743E+08 -0.139888E+07 -3.60750 14.530 0.970210E+08 -0.140672E+07 -3.63250
- 14. 630 0.972026E+08 -0.141409E+07 -3.65750 14.730 0.973193E+08 -0.141421E+07 -3.68250 14.830 0.974625E+08 -0.141570E+07 -3.70750
- 14. 930 0.975246E+08 -0.141857E+07 -3.73250 15.000 0.975500E+08 -0.141951E+07 -3.75000
- -* ANSYS POST26 VARIABLE LISTING *****
TIME 12 MZ 12 FY 12 UY MZ FY Uy 15.040 0.973286E+08 -0.141547E+07 -3.76000 15.080 0.973074E+08 -0.141560E+07 -3.77000 15.140 0.975689E+08 -0.141931E+07 -3.78500 15.230 0.979661E+08 -0.142489E+07 -3.80750 15.330 0.978152E+08 -0.141801E+07 -3.83250
- 15. 430 0.980224E+08 -0.142217E+07 -3.85750 15.530 0.982355E+08 -0.142739E+07 -3.88250
- 15. 630 0.984213E+08 -0.143199E+07 -3.90750 15.730 0.983181E+08 -0.142610E+07 -3.93250 15.830 0.983964E+08 -0.142542E+07 -3. 95750 15.930 0.984917E+08 -0.142761E+07 -3.98250 16.000 0.986953E+08 -0.143242E+07 -4.00000 16.04 0 0.986805E+08 -0.143535E+07 -4.01000 16.080 0.985047E+08 -0.143397E+07 -4.02000 16.140 0.986829E+08 -0.143609E+07 -4.03500 16.230 0.988492E+08 -0.143074E+07 -4.05750 16.330 0.990010E+08 -0.143356E+07 -4.08250
- 16. 430 0.990983E+08 -0.143800E+07 -4.10750 16.530 0.993080E+08 -0.144347E+07 -4.13250 16.630 0.993716E+08 -0.144392E+07 -4.15750
- ANSYS POST26 VARIABLE LISTING *****
TIME 1 2 MZ 12 FY 12 UY MZ FY Uy 16.730 0 .995191E+08 -0.144394E+07 -4.18250 16.830 0 .996544E+08 -0.144755E+07 -4.20750 16.930 0 .998164E+08 -0.145111E+07 -4.23250 17.000 0 .997917E+08 -0.145195E+07 -4.25000 SOP04023-REV7.DOC Page D106 of D107 Rev. Date: 04-08-2005
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Client: Nuclear Management Company Calc. No 2005-6760 Project: Point Beach Nuclear Generating Plant Page No. D107 Project No.: 11165-065 Rev. No. 2 Analysis of Postulated Reactor Head Load Drop 17.040 0.994502E+08 -0.144457E+07 -4.26000
- 17. 080 0.993765E+08 -0.144309E+07 -4.27000 17.140 0.996078E+08 -0.144575E+07 -4.28500 17.230 0.999940E+08 -0.145099E+07 -4.30750 17.330 0.100143E+09 -0.145456E+07 -4.33250 17.430 0.100106Ef 09 -0.144970E+07 -4.35750 17.530 0.100237E+09 -0.145233E+07 -4.38250 17.630 0.100400E+09 -0.145596E+07 -4.40750 17.730 0.100608E+09 -0.146156E+07 -4.43250 17.830 0.100624E+09 -0.146058E+07 -4.45750
- 17. 930 0.100725E+09 -0.146287E+07 -4.48250 18.000 0.100861E+09 -0.146580E+07 -4.50000 SOP04023-REV7.DOC Page D107 of D107 Rev. Date: 04-08-2005