ML12033A179

From kanterella
Jump to navigation Jump to search
Calculation CA 11-004, Finite Element Stress Analysis of Monticello RPV Feedwater Nozzle, Enclosure 6
ML12033A179
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 02/17/2011
From: Mcgruder W
Xcel Energy
To:
Office of Nuclear Reactor Regulation
References
L -MT-12-002, SIA 1000847.301 CA 11-004, QF-0549 (FP-E-CAL-01), Rev 7
Download: ML12033A179 (32)
v • d • e


Text

ENCLOSURE6 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISE THE TECHNICAL SPECIFICATIONS TO INCLUDE A PRESSURE TEMPERATURE LIMITS REPORT CALCULATION CA 11-004 FINITE ELEMENT STRESS ANALYSIS OF MONTICELLO RPV FEEDWATER NOZZLE (SIA No. 1000847.302)

(31 pages follow)

Document Information NSPM Calculation (Doc) No: 11-004 Revision: 0

Title:

Finite Element Analysis of Monticello RPV Feedwater Nozzle Facility: 0 MT El PI Unit: 0 1 [ 2 Safety Class:

M SR LI Aug Q El Non SR Special Codes: [E Safeguards EL Proprietary Type: Calc Sub-Type:

(NOTE:

Print and sign name in signature blocks, as required.

Major Revisions E[

N/A EC Number: 17501 Vendor Calc Vendor Name or Code: Structural Integrity Vendor Doc No: 1000847.302 Integrity (SIA)

Description of Revision: New Calculation Issuance The following calculation and attachments have been reviewed and deemed acceptable as a legible QA recordI Prepared by: (sign)

IVtndOr

/(print) SIA Date: 1019/2010 Reviewed by:(sign) I. **

A/Z, (print) Wynter McGruder Date: 1/24/2011 Type of Review: [0 Design Ver~ication El Tech Review 0 Suitability Review Method Used (For DV Only): &_

ýe

[1 Alternate Calc EI Test Approved by: (sign)r I (print) Steve Kibler Date:-,9 / 7 b I I

Minor Revisions

[j N/A EC No:

El Vendor Caic:

Minor Rev. No:

Description of Change:

Pages Affected:

The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record Prepared by: (sign)

I (int)

Date:

Reviewed by: (sign)

/ (print)

Date:

Type of Review: [1 Design Verification [] Tech Review E] Suitability Review Method Used (For DV Only): *l Review [1 Alternate Calc [1 Test Approved by: (sign)

/ (print)

I Date:

Record Retention: Retain this form with the associated calculation for the life of the plant.

This reference table is used for data entry into the PassPort Controlled Documents Module reference tables (C012 Panel). It may NOTE:

also be used as the reference section of the calculation. The input documents, output documents and other references should all be listed here. Add additional lines as needed by using the "TAB" key and filling in the appropriate information in each column.

Reference Documents (PassPort C012 Panel from C020)

Controlled*

Document Name Document Doc Ref Type**

Doc?+ Type Number Rev INPUT OUTPUT I

U.S. NRC, Reactor Vessel Integrity Database, Version 2.0.1 N/A N/A X

2 x

SPEC GE Stress Report No. 22A7454, Revision 1, "Reactor Vessel MPS-841 1

X (System Cycling),"

3 ASME Boiler and Pressure Vessel Code,Section III including N/A N/A X

Appendices, 1977 Edition with Addenda through Summer 1978 4

Chicago Bridge & Iron No. 9, Revision 7, "Detail of 10" Stub Mk. #

x DRAW N4 AID for 17'2" L.D x 63'-2" Ins Heads Nuclear Reactor for 7

X General Electric Company for Northern States Power," NX-8290-73 5

X SPEC GE Design Specification No. 22A6996, Revision 0, "Reactor MPS-843 0

X Vessel System Cycling,"

6, ANSYS Mechanical and PrepPost, Release 11.0 (w/ Service Pack N/A N/A X

1), ANSYS, Inc. August 2007 7

x SPEC GE Design Specification No. 21A1112, Revision 6, "Reactor MPS-179 6

X Pressure Vessel" 8

9 10 11 12 14 15 Record Retention: Retain this form with the associated calculation for the life of the plant.

16 17 Controlled Doc marked with an "X" means the reference can be entered on the C012 panel in black. Unmarked lines will be yellow. If marked with an "X', also list the Doc Type, e.g., CALC, DRAW, VTM, PROC, etc.

Mark with an "X" if the calculation provides inputs and/or outputs or both. If not, leave blank. (Corresponds to PassPort "Ref Type" codes: Inputs I Both =

"ICALC", Outputs = "OCALC", Other/ Unknown blank)

Other PassPort Data Associated System (PassPort C01 1, first three columns)

OR Equipment References (PassPort C025, all five columns):

Facility Unit System Equipment Type Equipment Number MT I

RPV Superseded Calculations (PassPort C019):

Facility Calc Document Number Title NIA N/A N/A Description Codes - Optional (PassPort C018):

Code Description (optional)

Code Description (optional)

Notes (Nts) - Optional (PassPort X293 from C020):

Topic Notes Text Record Retention: Retain this form with the associated calculation for the life of the plant.

E-Calc Introduction D Copy directly from the calculation Intro Paragraph or -- See write-up below IE (Specify)

Record Retention: Retain this form with the associated calculation for the life of the plant.

QF-0549 (FP-E-CAL-01), Rev. 7 Page 5 of 4 XcelEnergy Calculation Signature Sheet Monticello Specific Information Topic Code(s) (See MT Form 3805):

PLEX. RATE Structural Code(s) (See MT Form 3805):

[]YES 0]YES EN/A EN/A Does the Calculatio R YES Z No n:

Require Fire Protection Review? (Using MT Form 3765, "Fire Protection Program Checklist', determine if a Fire Protection Review is required.) If YES, document the engineering review in the EC. If NO, then attach completed MT Form 3765 to the associated EC.

Affect piping or supports? (If Yes, Attach MT Form 3544.)

Affect IST Program Valve or Pump Reference Values, and/or Acceptance Criteria? (If Yes, inform IST Coordinator and provide copy of calculation.)

E YES E] YES

~No

~No Record Retention: Retain this form with the associated calculation for the life of the plant.

V Structural Integrity Associates, Inc.

File No.: 1000847.302 CALCULATION PACKAGE Project No.: 1000847 Quality Program: 0 Nuclear El Commercial PROJECT NAME:

Monticello P-T Curves Revision According to the PTLR Methodology CONTRACT NO.:

1005, Release 18 CLIENT:

PLANT:

Xcel Energy, Inc.

Monticello Nuclear Generating Plant CALCULATION TITLE:

Finite Element Stress Analysis of Monticello RPV Feedwater Nozzle Document Affected Project Manager Preparer(s) &

Revision Pages Revision Description Approval Checker(s)

Signature & Date Signatures & Date 01

- 25 Initial Issue A-i Clark Oberembt Eric Houston CJO 10/8/10 EJH 10/9/10 Yu Chen YC 10/8/10 Page 1 of 25 F0306-O1R1

Structural Integrity Associates, Inc.

Table of Contents 1.0 OBJECTIVE..................................................................................................................

4 2.0 M ETH OD OLOGY.................................................................................................

4 3.0 DESIGN IN PUTS.....................................................................................................

4 4.0 A SSUM PTION S..........................................................................................................

.5 5.0 FINITE ELEM EN T M OD EL...................................................................................

6 6.0 AN ALY SIS....................................................................................................................

7 6.1 M echanical Load Stress A nalysis.................................................................

7 6.2 Therm al Transient Stress A nalyses...............................................................

7 6.2.1 Thermal Shock Transient.............................................................................

8 6.2.2 Cooldown Transient......................................................................................

8 7.0 RESULTS OF AN ALY SIS.....................................................................................

8 8.0 REFEREN CES.....................................................................................................

9 A PPEND IX A : LIST OF SUPPORTIN G FILES...........................................................

A -i List of Tables Table 1: RPV M aterial Properties (SA -533 Grade B)......................................................

10 Table 2: N ozzle M aterial Properties (SA -508 Class II)..........................................................

10 Table 3: Cladding M aterial Properties (Type ER308 Stainless Steel)...................................

11 Table 4: Therm al Shock Transient D efinition........................................................................

11 Table 5: Cooldow n Transient D efinition................................................................................

-11 File No.: 1000847.302 Page 2 of 25 Revision: 0' F0306-OIR1

Structural Integrity Associates, Inc.

List of Figures Figure 1: Finite Element Model Geometry...................................

12 Figure 2: FEM O verview..................................................................................................

13 F igure 3: FE M D etail...............................................................................................................

14 Figure 4: Unit Pressure Loading with CYp Loads (psi)..................................................

15 Figure 5: FEM Boundary Conditions..... :........................................................................

16 Figure 6: Heat Transfer Coefficient Regions....................................................................

17 Figure 7: Thermal Shock Transient Convection Film Coefficients (Btu/sec-in2-OF)......

18 Figure 8: Cooldown Transient Convection Film Coefficients (Btu/sec-in2-OF)................ 19 Figure 9: Location of Maximum Hoop Stress due to Pressure (psi).................................

20 Figure 10: Path L ocation..................................................................................................

21 Figure 11: Thermal Shock Temperature Response (OF), Time = 100 Seconds................ 22 Figure 12: Thermal Shock Temperature Response (OF), Time = 5,100 Seconds.............. 23 Figure 13: Cooldown Temperature Response (OF), Time = 1,001 Seconds...................... 24 Figure 14: Cooldown Temperature Response (OF), Time = 4,501 Seconds...................... 25 File No.: 1000847.302 Revision: 0 Page 3 of 25 F0306-O1R1

Structural Integrity Associates, Inc.

1.0 OBJECTIVE The objective of this calculation is to perform finite element stress analysis on Monticello Nuclear Generating Plant's feedwater (N4) nozzles to support Pressure-Temperature (P-T) curve development.

The development of P-T curves for the N4 nozzles requires the consideration of a postulated /4 thickness flaw in the nozzle blend radius. The results of this analysis will be used to develop stress intensity factors at the postulated flaw tip.

2.0 METHODOLOGY A three-dimensional (3-D) finite element model (FEM) is constructed for the feedwater nozzle to be evaluated. For this analysis, the region of interest is the nozzle blend radius. The FEM also includes a portion of the reactor pressure vessel (RPV) as well as the internal cladding. Appropriate boundary conditions are applied to reflect symmetry in the model. A mechanical analysis with a unit internal pressure load and two thermal transient analyses are performed. One path through the nozzle blend radius is defined based on the maximum hoop stress at the inside surface of the nozzle, due to pressure, and the shortest path to the outside of the blend radius. The selection of a single path is most appropriate to support the P-T curve evaluation because RPV hoop stress is twice membrane stress and the thermal gradient is nearly uniform around the blend radius. Experience has shown that the hoop stress distribution due to pressure loading usually produces approximately a three times greater stress intensity factor than due to thermal stresses at the tip of the postulated flaw., Mapped hoop stresses are extracted along this path for the mechanical and thermal analyses. For the thermal analyses, the auto time stepping feature in ANSYS, along with a small initial step, is used to generate the time steps for temperature and stress solutions.

The assumptions made in order to define the evaluation approach and perform the analysis are summarized in Section 4.0. The application of these assumptions is indicated throughout the document using a set of braces containing the appropriate assumption number; for example, Assumption #1 would be indicated as { 1, §4.0}.

3.0 DESIGN INPUTS The following design inputs are used in this calculation:

  • RPV material = SA-533 Gr. B [1]

N4 material = SA-508 Cl1II [2]

" Cladding material = Type ER308 stainless steel [7]

  • N4 Design Code of Record [2, Sheet No. 2]= ASMIE Section III, 1977 Editionwith Addenda through Summer 1978 [3]

" Temperature dependent material properties are obtained from Reference [3] and are listed in Tables 1 through 3

" RPV and N4 geometry are taken from References [4, 5] and reproduced in Figure 1 { 1, §4.0}

o Reference [4] shows the original N4 nozzle configuration o Reference [5] shows the modified N4 nozzle configuration

" Maximum fluid temperature under normal operating condition = 546°F [5, Section 4.6.7.2.2]

File No.: 1000847.302 Page 4 of 25 Revision: 0 F0306-OIRI

V Structural Integrity Associates, Inc.

" Normal operating thermal transients applicable to the N4 nozzle are defined in Reference [5, Section 4.6.7 ] {2, §4.0}

" The effect of the thermal sleeve is approximated by calculating the fluid temperature in the annulus when there is feedwater flow using the methods described in Reference [5, Appendix 20]

  • Ambient air temperature outside of the vessel is 100°F [5, Section 4.3.2.1]
  • Heat transfer coefficients (see Figures 6, 7, and 8 and Section 6.0 for region definitions and application areas on the model) o Exterior surface heat transfer coefficient (Btu/hr-fý-°F) = 0.2 [5, Section 4.3.2.1]

o Vessel region internal heat transfer coefficient (Btu/hr-ft2-OF) = 500 [5, Appendix 20]

4.0 ASSUMPTIONS The assumptions made in order to define the evaluation approach and perform the analysis are summarized in the following list. The application of these assumptions is indicated throughout the document using a set of braces containing the appropriate assumption number; for example, Assumption

  1. 1 would be indicated as {1, §4.0}.
1. The following assumptions are made in regards to the modeled geometry:
a. Some features remote from the area of interest are not modeled. This includes the safe end, safe end-to-nozzle weld, thermal sleeve and the nozzle-to-vessel weld. This simplification of the model is remote from the area of interest and the impact of this assumption is minimal.
b. Some features remote from the area of interest are simplified. Some fillet radii are ignored and modeled as points, the nozzle ID is assumed constant at 10.88 in. upstream of the blend radius modification, and the location of the nozzle-to-vessel material transition is approximated at the centerline of the nozzle-to-vessel weld. This simplification of the model is remote from the area of interest and the impact of this assumption is minimal.
c. For the pressure load analysis, the cladding is assumed not to contribute to the stress solution and the associated cladding volumes are removed from the analysis (cladding is not removed for the thermal transient analyses) consistent with Reference [3].
2. Reference [5] defines the thermal transients applicable to the N4 nozzle, the following assumptions are made in regard to the bounding transients:
a. The N4 nozzle thermal shock transient from 546°F to 100°F, when feedwater flow is initiated, is assumed to occur over 1 second. The effects of the thermal sleeve are approximated using the methods described in Reference [5, Appendix 20] to calculate the annulus fluid temperature. Annulus fluid temperature variation in the nozzle axial direction is conservatively ignored.
b. The initial vessel heatup from 100F to 546°F at 100°F/hr is defined in Reference [5], and a cooldown transient is not defined. However, a cooldown transient typically produces more severe hoop stress at the inner surface of the nozzle and vessel. Since the analysis File No.: 1000847.302 Page 5 of 25 Revision: 0 F0306-OIR1

V Structural Integrity Associates, Inc.

will be used in generating P-T limit curves, the locationý of interest is near the inner surface of the nozzle blend radius. Therefore, the defined heatup transient is conservatively modeled as a cooldown transient from 546°F to 100°F at -100°F/hr.

3. The forced convection heat transfer coefficients applied to the nozzle blend radius and the nozzle region (upstream of the blend radius) during the thermal shock analysis are taken from Reference

[5, Appendix 20] and are modified as follows:

a. The nozzle blend radius heat transfer coefficient is assumed to be constant at the maximum value given from "c to f' of 1500 Btu/hr-ft2-°F. A higher HTC will produce a larger thermal gradient, which results in higher thermal stress and is conservative.
b. The nozzle region heat transfer coefficient is assumed to be constant at the maximum value given from "a to c" of 750 Btu/hr-ft2 -F.

A higher HTC will produce a larger thermal gradient, which produces higher stress and is conservative.

4. Although Reference [2] specifies lower natural convection heat transfer coefficients for 0%

feedwater flow, the vessel region forced convection heat transfer coefficient of 500 Btu/hr-ft2 -F is applied to the entire inner surface of the model for the cooldown transient analysis. Higher heat transfer coefficients typically produce higher thermal stresses for this type of transient.

5. The material Poisson's ratio and weight density are assumed to be temperature independent at 0.3 and 0.283 lbf/in3, respectively.
6. Unspecified material properties at -100°F are assumed equal to properties at 70'F.
7. All analyses are performed with a stress free reference temperature of 700F.
8. All materials are assumed to follow linear-elastic isotropic material behavior.

5.0 FINITE ELEMENT MODEL The three-dimensional (3-D) finite element model (FEM) is developed for the N4 nozzle using the ANSYS finite element software package [6] using 8-node SOLID45 3-D linear structural elements. The SOLID45 elements are converted to thermal solid elements (SOLID70) for the thermal transient analyses. The dimensions used for the nozzle and RPV shell are provided in References [4, 5] as modified by the assumptions in { 1, §4.0}. The dimensions used in the model are presented in Figure 1.

The one-quarter 3-D model extends the end of the machined portion of the nozzle a total of 5 inches, while the RPV shell extends 450 circumferentially and 60 inches axially from the nozzle axis. These dimensions are large enough to eliminate non-representative boundary effects on the nozzle blend radius region in the finite element analysis (FEA) solution. The FEM is shown in Figures 2 and 3. The ANSYS input files used to generate the FEM are included with the electronic supporting files listed in Appendix A.

File No.: 1000847.302 Page 6 of 25 Revision: 0 F0306-O1R1

Structural Integrity Associates, Inc.

6.0 ANALYSIS A static mechanical loading stress analysis, using a unit pressure, and two thermal transient stress analyses are performed in the FEA. A mesh density study is completed using the unit pressure load case to verify the quality of the mesh. Three different mesh densities are evaluated: Mesh 1 = 5,473 nodes, Mesh 2 = 30,775 nodes, and Mesh 3 = 178,507 nodes. The results for maximum hoop stress at the blend radius converged within 0.1% from Mesh 2 to Mesh 3. Mesh 2 is used in this analysis.

6.1 Mechanical Load Stress Analysis For the unit pressure case, a uniform pressure of 1,000 psig is applied to the inside surface of the RPV and nozzle bore. The mechanical load stress results are intended to be linearly scaled during subsequent analyses as required. Note that the cladding is not included in the unit pressure analysis {lc, §4.0}. A cap load, Pcjj, is applied to the upper horizontal cut plane in the model, which is calculated as:

P, P.JR,,

p,-oR

_ -IRo,,

where:

Punit = unit pressure, psig IRves = inside radius of RPV excluding cladding, in.

ORves = outside radius of RPV, in.

A cap load, P,12, is also applied to the free end of the nozzle, which is calculated as:

untnoz OR,

- OR-JR nRz where:

IRoz = inside radius of nozzle, in.

ORnoz = outside radius of nozzle free end, in.

Symmetry boundary conditions are applied on both of the vertical cut planes and the lower horizontal cut plane in the 3-D model. The nodes at the free end of the nozzle are coupled in the axial direction, as are the RPV nodes on the upper horizontal cut plane. The unit pressure loads and boundary conditions are shown in Figures 4 and 5, respectively. The ANSYS input file for the unit pressure load is included with the electronic supporting files listed in Appendix A.

6.2 Thermal Transient Stress Analyses For the transient case, the model exterior and interior surfaces are broken into different regions as shown in Figure 6. The regions are defined as follows:

o Region 1 - Exterior surfaces

  • Region 2 - Vessel wall (interior surfaces) e Region 3 - Blend radius (interior surfaces) a Region 4 - Nozzle (interior surfaces, upstream of the blend radius)

File No.: 1000847.302 Page 7 of 25 Revision: 0 F0306-OIRI

Structural Integrity Associates, Inc.

The FEM is set to steady state with appropriate initial conditions before each transient is analyzed.

6.2.1 Thermal Shock Transient The heat transfer coefficients described in Section 3.0 with {3,, §4.0} are applied to the respective regions as shown in Figure 7. The air temperature for Region 1 is set at 100'F [5]. The fluid temperature for Region 2 is equal to the vessel temperature of 546°F [5]. The fluid temperature is calculated for Regions 3 and 4 using the methods described in Reference [5, Appendix 20]. Table 4 provides a summary of the thermal shock transient definition as analyzed. TheANSYS input files used to generate the thermal stresses are included with the electronic supporting files listed in Appendix A.

6.2.2 Cooldown Transient The heat transfer coefficients as described in Section 3.0 with {4, §4.0} are applied to the respective regions as shown in Figure 8. The air temperature for Region 1 is set at 100'F [5]. The fluid temperature for Regions 2, 3, and 4 are equal at 546'F and decrease to 100°F at -l00F/hr. Table 5 provides a summary of the cooldown transient definition as analyzed. The ANSYS input files used to generate the thermal stresses are included with the electronic supporting files listed in Appendix A.

7.0 RESULTS OF ANALYSIS After running the unit pressure load case, the nodes along the blend radius at both cut planes are queried for the location of maximum hoop stress. This confirms that the maximum hoop stress is along the RPV axial cut plane as expected. Figure 9 shows the location of maximum hoop stress due to pressure. A stress extraction path is defined for the cut plane that includes this maximum location. The path starts at the nodal location of maximum hoop stress and extends through the blend radius along the shortest path length. Figure 10 shows the path location. The mapped hoop stresses are extracted along this path for the static unit pressure analysis. For the thermal transient load case, the mapped hoop stresses are extracted along the same path for all time steps. The stress extraction results from the FEA are contained in the ".OUT" output files listed in Appendix A. Comparing the results from the output files, verifies that the use of auto time stepping, along with a small initial step, is sufficient in capturing the peak stresses within acceptable limits.

Examples of the temperature responses due to the thermal shock transient load case are shown in Figures 11 and 12 for selected times showing that the model is behaving as expected. Examples of the temperature responses due to the cooldown transient load case are shown-in Figures 13 and 14 for selected times showing that the model is behaving as expected.

File No.: 1000847.302 Page 8 of 25 Revision: 0 F0306-O1R1

8.0 1.

2.

3.

4.

5.

6.

7.

Structural Integrity Associates, Inc.

REFERENCES US Nuclear Regulatory Commission, Reactor Vessel Integrity Database, Version 2.0.1.

GE Stress Report No. 22A7454, Revision 1, "Reactor Vessel (System Cycling)," SI File No.

NSP-36Q-202.

ASME Boiler and Pressure Vessel Code,Section III including Appendices, 1977 Edition with Addenda through Summer 1978.

CB&I Drawing No. 9, Revision 7, "Detail of 10"0 Stub Mk.# N4 A/D for 17'-2" I.D. x 63'-2" Ins. Heads Nuclear Reactor for General Electric Company for Northern States Power," NX-8290-73, SI File No. NSP-21Q-235.

GE Design Specification No. 22A6996, Revision 0, "Reactor Vessel System Cycling," SI File No. 1000847.201.

ANSYS Mechanical and PrepPost, Release 11.0 (w/ Service Pack 1), ANSYS, Inc., August 2007.

GE Specification No. 21A1 112, Revision 6, "Reactor Pressure Vessel," SI File No. MONT-14Q-204.

File No.: 1000847.302 Revision: 0 Page 9 of 25 F0306-OIRI

V Structural Integrity Associates, Inc.

Table 1: RPV Material Properties (SA-533 Grade B)

Meanf Coefficient

_Tempeature

-Thermal Thermal Specific E

oTermal Elastic iF);

  • Conduc/tiviy' Diffusiv'ity' Heat' 3 -.a-6 Mo UlUS,X 10
  • O F )

E o a c v

:f f 2 i i ),

il x p a n s i o n X 1 0.....

I hr-ft-0 F)

(ft2/hrF)(Btu/lb-0 F),

(psi) 1~003150.62.113 2O/F)6 30.4

-100 31.5 0.5692 0.1132 6.07 30.4 70 31.5 0.5692 0.1132 6.07 29.9 200 30.0 0.5246 0.1169 6.38 29.5 300 29.1 0.4928 0.1208 6.60 29.0 400 28.1 0.4616 0.1245 6.82 28.6 500 27.2 0.4338 0.1282 7.02 28.0 600 26.2 0.4061 0.1319 7.23 27.4 Notes:

1. Reference [3], Table 1-4.0 {6, § 4.0)
2. Calculated assuming constant density of 0.283 lb/in3 {5, § 4.0}
3. Reference [3], Table 1-5.0 {6, § 4.0)
4. Reference [3], Table 1-6.0 Table 2: Nozzle Material Properties (SA-508 Class II)

Thermal

'Mean (?befficient Easi

'... =Thermal

ý.A I,

herm al, S1

  • S
cific, T::I:'.>

. i:

,
.,:, >,e sm,!{i:

Temperature, Tha t

pcii ansoy3im

(,r)

Conductivity

,,Di'ffivit Hat.*

-on Te rml

.. 1

,.du us'

,,,(Btu/hr-ft-0 F)

(/hr)

(BtuAl2-F) i.o 1

(psi),

-100 31.5 0.5692 0.1132 6.07 30.4 70 31.5 0.5692 0.1132 6.07 29.9 200 30.0 0.5246 0.1169 6.38 29.5 300 29.1 0.4928 0.1208 6.60 29.0 400 28.1 0.4616 0.1245 6.82 28.6 500 27.2

-0.4338 0.1282 7.02 28.0 600 26.2 0.4061 0.1319 7.23 27.4 Notes:

1. Reference [3], Table 1-4.0 (6, § 4.0)
2. Calculated assuming constant density of 0.283 lb/in 3 (5, § 4.0)
3. Reference [3], Table 1-5.0 {6, § 4.0)
4. Reference [3], Table 1-6.0 File No.: 1000847.302 Revision: 0 Page 10 of 25 F0306-OIRI

V Structural Integrity Associates, Inc.

Table 3: Cladding Material Properties (Type ER308 Stainless Steel)

= ~ ~~~~~~~

Mean Coefficient Temeraure Thermal Thermal Specific

'Mean ice" Elastic Tem peHrature

,..l1S, Heat2 P

of Thermald T~fe 2,

Mou 4 x

.. 0L_

C onductivity'

ý,DI1I.fUslVliy' H at*-

[-

3,Mocli6:",

F)

('Expansion X 1(0Bl*fF(psi7 (n/in/0F).-i

-100 8.35 0.1498 0.1140 9.11 29.4 70 8.35 0.1498 0.1140 9.11 28.3 200 8.90 0.1548 0.1176 9.34 27.7 300 9.35 0.1589.

0.1203 9.47 27.1 400 9.80 0.1630 0.1229 9.59 26.6 500 10.23 0.1659 0.1261 9.70 26.1 600 10.70 0.1707 0.1282 9.82 25.4 Notes:

1. Reference [3], Table 1-4.0 f6, § 4.0}
2. Calculated assuming constant density of 0.283 lb/in 3 {5, § 4.0}
3. Reference [3], Table 1-5.0 {6, § 4.0}
4. Reference [3], Table 1-6.0 Table 4: Thermal Shock Transient Definition Tim s)

Temperatuires (OF)l Heat Transfer Coeffidents (Btu/hr-ft2.-- F)

Time *(s),

Region1$ Region.2,.Region 3 Reio4Region:I1 Region.2Region3 Relgion 4

..,...... R d'R e g i o 21' RR egoo h3_

k ' g o ni oke in

, 4:4 0

100 546 546 546 1

100 546 528 510 0.2 500 1500 750 5100 100 546 528 510 Notes:

1. Temperatures for Regions 3 & 4 are calculated using the methods from Reference [5, Appendix 20], based on a feedwater flow temperature of 100°F at 25% flow Table 5: Cooldown Transient Definition Tirn~

(s) i Temp~eratures '(*F)

_____HaTrnfer Cbeff icibnts-(Bu/hrý-ft2,F)

ý. im e* (.s)

Regibn I*

Region 2', Region 3 Region 4 Region,,

Region 2. 'Region 3 Region 4 0

100 546 546 546 3600 100 446 446 446 0.2 500 500 500

.16056 100 100 100 100 18156 100 100 100 100 File No.: 1000847.302 Revision: 0 Page 11 of 25 F0306-O1R1

V Structural Integrity Associates, Inc.

R103.000 188 Notes:

1. Dimensions from References [4, 5].
2.

Units in inches.

Figure 1: Finite Element Model Geometry File No.: 1000847.302 Revision: 0 Page 12 of 25 F0306-O1RI

V Structural Integrity Associates, Inc.

ELEHENTS ANSYS 11.OSP1 MAT NMN ANSYS II.OSPI FEEDIAITER NOZZLE

'Aý x Figure 2: FEM Overview File No.: 1000847.302 Revision: 0 Page 13 of 25 F0306-OIRI

Structural Integrity Associates, Inc.

Figure 3: FEM Detail File No.: 1000847.302 Revision: 0 Page 14 of 25 F0306-O1RI

V Structural Integrity Associates, Inc.

ELE=ENTS ANSYS 11.0SP1 MAT NUM PPE S-NORM

-9245

-6968

-4692

-2415

-138.325

-8107

-5830

-3553

-1277 1000 FEEDWATER NOZZLE Figure 4: Unit Pressure Loading with Cap Loads (psi)

File No.: 1000847.302 Revision: 0 Page 15 of 25 F0306-OIRI

V Structural Integrity Associates, Inc.

Figure 5: FEM Boundary Conditions File No.: 1000847.302 Revision: 0 Page 16 of 25 F0306-OIR1

V Structural Integrity Associates, Inc.

Region I Region 2 Region 4 Figure 6: Heat Transfer Coefficient Regions File No.: 1000847.302 Revision: 0

  • Page 17 of 25 F0306-OIR1

V Structural Integrity Associates, Inc.

ELEMEMt HAT NUN

':Q T-H

.386E-06

.322E-03 EVENT: Shock Figure 7: Thermal Shock Transient Convection Film Coefficients (Btu/sec-in 2-OF)

File No.: 1000847.302 Revision: 0 Page 18 of 25 F0306-O1R1

Structural Integrity Associates, Inc.

1 E LEBENTS MIAT WNM i:ON17-HCOE X

. 386E-06

.108E-03 EVENT:

COOL DOW~N Figure 8: Cooldown Transient Convection Film Coefficients (Btu/sec-in 2-°F)

File No.: 1000847.302 Revision: 0 Page 19 of 25 F0306-OIR1

V Structural Integrity Associates, Inc.

I MODAL SOLUTION STEP=1 RSYS~l OHIX ý. 066, SWN

  1. 583.

SHX q,5O025

-5583 31639 4AU4 b 620. 455 FEEDVATER NOZZLE 6824 19231 3 il;39 44046 50250 13028 25435 37842 Figure 9: Location of Maximum Hoop Stress due to Pressure (psi)

Page 20 ot25 File No.: 1000847.302 Revision: 0 Page 20 of-25 F0306-OIR1

V Structural Integrity Associates, Inc.

Figure 10: Path Location File No.: 1000847.302 Revision: 0 Page 21 of 25 F0306-OIR1

V Structural Integrity Associates, Inc.

Figure 11: Thermal Shock Temperature Response (IF), Time = 100 Seconds File No.: 1000847.302 Revision: 0 Page 22 of 25 F0306-O1R1

V Structural Integrity Associates, Inc.

NODAL SOLUTION STEP=5 SUB =1 TIME=5100 TEMP (AVG)

RSYS=0 SMN =510.051 SMX =545.815 510.05 EVENT: Shock ANSYS 11.0SP1 Mx 2

545.815 1

514.025 Figure 12: Thermal Shock Temperature Response ('F), Time = 5,100 Seconds File No.: 1000847.302 Revision: 0 Page 23 of 25 F0306-0IRI

V Structural Integrity Associates, Inc.

1 NODAL SOLUTION ANSYS 11.OSP.

STEP=2 SUB =4 TIME=1001 TEMP (AVG)

RSYS=O SMN =518.634 SMX =536.868 x

518 634

'522.686

'526. 738

.r4344 2 520. 66 524.712 528.7646

.*52536.868 EVENT:

COOLDOWN Figure 13: Cooldown Temperature Response (°F), Time = 1,001 Seconds File No.: 1000847.302 Revision: 0 Page 24 of 25 F0306-OIR1

V Structural Integrity Associates, Inc.

1 NODAL SOLUTION STEP=2 SUB =18 TIME=4501 TEMP (AVG)

RSYS=0 SMN =421.439 SMX =453.007 J

ANSYS 11.OSPI 99 453.007 x

EVENT:

COOLDOWN Figure 14: Cooldown Temperature Response (IF), Time = 4,501 Seconds File No.: 1000847.302 Revision: 0 Page 25 of 25 F0306-OIRI

Structural Integrity Associates, Inc.

APPENDIX A: LIST OF SUPPORTING FILES ANSYS File

. Description Finite element model geometry input MONTWQUART.INP file, including temperature dependent material properties Unit pressure input and stress extraction PRESSUREQUART.INP file APPLYHTC.INP Heat transfer coefficient region definition input file Thermal shock transient input and stress S K Nextraction file SHOCK FW mntr.INP Thermal shock transient file containing LDREAD and SOLVE commands Cooldown transient input and stress COOLD.OWNFW.1NP extraction file Cooldown transient file containing C D F

m N

LDREAD and SOLVE commands PRESSTRPATH.OUT Hoop stress output file for unit pressure THMSTR PATH.OUT Hoop stress output file for thermal shock transient THMSTRCDPATH.OUT Hoop stress output file for cooldown transient File No.: 1000847.302 Revision: 0 Page A-I of A-1 F0306-OIRI

Retrieved from "https://kanterella.com/w/index.php?title=ML12033A179&oldid=5265435"