ML20006E367
| ML20006E367 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/11/1990 |
| From: | Hassenpflug H, Shah S FRAMATOME COGEMA FUELS (FORMERLY B&W FUEL CO.) |
| To: | |
| Shared Package | |
| ML20006E361 | List: |
| References | |
| 86-1177306, 86-1177306-00, NUDOCS 9002220747 | |
| Download: ML20006E367 (19) | |
Text
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B&WPuol Compumy CALCULATION
SUMMARY
SHEET (CSS) l 86-1177306-00 i DOCUMENT IDENTIFIER !
l TECO Mark-BE FA Seismic and IhCA Analysis TITLE l PREPARED BY: REVIEWED BY:
S. J. Shah H. L. Hassenpflug '
NAME NAME SIGNATURE I* I' SIGNATURE - " M TITLE Principal Engineer 1-9-90 Engineer III U1-11-90 '
DATE TITLE DATE i COST CENTER REF. PAGE($) TM STATEMENT: REVIEWER INDEPENDENCE DC
~
PURPOSE AND
SUMMARY
OF RESULTS: j i
PURPOSE: This document describes the seismic and IOCA analyses of the Mark BZ fuel assembly for the Davis-Besse Unit 1 plant. The objective of these analyses is to verify .
! the structural adequacy of the Mark-BZ fuel assembly for seismic and IOCA conditions. ;
RESULTS: The results of the analysis are summarized in Table 1.
They indicate that all major components of the fuel ,
assembly meet the design requirements for the ',
operational Basis Earthquake (OBE) and a combined Safe ;
Shutdown Earthquake (SSE) plus Loss of Coolant Accident '
(IhCA) with adequate structural margin. The results ,
of the combined SSE plus IhCA analysis meet the design !
criteria for the SSE described in section 3 of this document, so separate SSE stress analysis results are not presented. !
i1 Points of importance in the analysis are summarized in section 2 of this document. -
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THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
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86-1177306-00 1
TABLE OF CONTENTS 4
1 IA91 1
1.0 INTRODUCTION
............................................ 3 2.0
SUMMARY
................................................. 5 I 3.0 ACCEPTANCE CRITERIA FOR STRUCTURAL EVALUATION.............................................. 7 4.O METHODOIDGY FOR DEVEI4PMENT OF i S TRUCWRAL I4ADINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1 l
5.0 METHODOIDGY FOR STRUCWRAL EVALUATION !
OF FUEL ASSEMBLY COMPONENTS............................. 11 l 1
6.0 FUEL ASSEMBLY STRESS ANALYSIS RESULTS................... 13 -I 1
7.0 REFERENCES
.............................................. 14
)
1 APPEND 1XES A. FUEL ASSEMBLY LATERAL IMPACT TESTS....................... 15 h t
B. CORE MODEL............................................... 17 LIST OF TABLES
- 1. FUEL ASSEMBLY STRESS ANALYSIS RESULTS FOR TECO MIXED CORE OBE AND COMBINED SSE'~ ' " '~ ~ ~ ~~~~ ~ ^
t PLU S IhCA IhADI NGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 '
LIST OF FIGURES
- 1. HORIZONTAL CORE SEISMIC AND IhCA MODEL. . . . . . . . . . . . . . . . . . . 19 Page 2
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1.0 INTRODUCTION
This document describes the seismic-LOCA analysis of the Mark-BZ fuel assemblies for use in the Davis-Besse plant. The objective of this analysis is to verify that the fuel assembly component i
stresses and impact forces due to the seismic and LOCA forces are within established allowable limits, q
l The analyses supporting Eircaloy grid fuel assemblies were l submitted to the NRC in Reference 1. The NRC Safety Evaluation (Reference 2) of Reference 1 states that the report can be referenced for reload analysis for all lowered-loop B&W-designed 1 177 fuel assembly plants. Davis-Besse 1 is a raised loop plant and a separate seismic and IOCA analysis is performed for the plant specific loads. The general methods dos::ribed in Reference 1 were
]
used in the present analysis, except for inclusion of new fuel- !
assembly impact tests described below. i In order to provide additional information on lateral fuel assembly impact characteristics, lateral fuel assembly impact tests. were i performed on Mark-BZ fuel assemblies. These test results provided {
additional data for analytical model verification and inproved l acceptance criteria for spacer grids subjected to impact loading. l A discussion of the fuel assembly lateral impact tests is provided l in Appendix A of this document. The core model used to analyze the fuel assembly response is described in Appendix B. i The Mark-BZ fuel assembly seismic and IDCA analysis reported in Reference 1 is based on the standard B&W fuel assembly seismic and
{
IhCA analysis procedures outlined in B&W Topical Report BAW-10133P, Rev. 1 (Reference 3). This report has received the NRC approval
) for referencing in licensing applications. A brief description of the analytical core model, method of stress analysis, and the results of the analysis are summarized in this document. Details Page 3
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of the core model and the analysis method may be found in Reference
- 3. Points of importance in the analysis are summarized in section '
2.
i The structural integrity of the fuel assemblies for each faulted condition are presented in section 3. The analytical methods for l development of structural loadings are presented in section 4; stress limits and the method of stress analysis used for structural evaluation of fuel assembly components are described in section 5. ,
The fuel assembly stress analysis results are described in section 6.
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n I 86-1177306-00 ;
i 2.0
SUMMARY
i Points of importance in the analysis are summarized below. !
(1) The effects of asymmetrical IOCA loadings on the reactor l vessel and reactor internals are considered in the analysis. l (2) The reactor core model was comprised of a row of five fuel.
assemblies and six masses as opposed to using seven assemblies l
and thirteen masses each described in B&W Topical Report !
10133P, Rev. 1. A parametric study was performed to determine the number of fuel assemblies to be used in the lateral core-model. The parametric study indicates that the five-fuel assembly model gives the highest impact load and as the number of fuel assemblies increases beyond five the load decreases. i It is also shown that six masses in the core model results in the highest load. The five fuel assembly-six mass model was used to save excessive computer time involved with the seven !
assembly-thirteen mass model. The fuel assembly lateral ;
impact test results and the single grid impact test results were used to obtain the spacer grid dynamic properties.
(3) The loads for LOCA plus SSE are combined by the square root '
of the sum of the squares method. This method is discussed l and accepted by the NRC in NUREG-0800, Appendix A.
(4) The design bases used to establish the acceptance criteria i are consistent with NUREG-0800, Appendix A, and follows the guidelines established by Section III of the ASME Code.
(5) The SSE requirement of control rod insertion was fulfilled for a combined SSE plus IOCA for added conservatism to the analysis.
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(6) For the IDCA loading a very few of the spacer grid impact ;
loads slightly exceeded the spacer grid elastic load limit.
l This load on the spacer grid resulted in minor plastic i
- deformation. However, calculations showed that the fuel rods remained in a coolable geometry and the guide thimble i positions were not altered. Thus, the insertion of the !
control rod will not be hindered. These calculations were f
based on grid cell geometry and the NRC's definition of a fully collapsed grid (Reference 6), all soft stops fully compressed and fuel rod in hard contact with all hard stops.
For this mode of deformation the reduction in the subchannel i flow area was 37%, which corresponds to a .38 inch deflection of the grid. The ECCS analysis verified that for '
the collapsed grid, the fuel rod cladding temperature was ,
within the acceptance limit of 2200*F.
(7) A bounding analysis of a mixed core configuration of Mark- ,
B and Mark-BE fuel assemblies was performed. The results '
showed that the Mark-BE fuel had a slightly higher impact I load for the mixed core configuration than for an all Mark-BE core configuration; however, the Mark-BE . fuel- still ---- - -
maintained a coolable geometry.
The results of the mixed core configuration analysis are summarized in Table'1. They indicate that all major components of the fuel assembly meet the design requirements for the operational Basis Earthquake (OBE) and a combined Safe Shutdown Earthquake (SSE) plus Ioss of coolant Accident (IOCA) event with adequate . structural margin. The results of the combined SSE plus IhCA analysis meet the design criteria for the SSE described in section 3 of B&W Topical Report BAW-10133P, Rev.1. So separate SSE stress analysis-results are not presented.
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I 3.0 ACCEPTANCE CRITERIA POR STRUCTURAL EVAIEATION The following criteria have been established for the fuel assembly IDCA and/or seismic analysis which are consistent with the
" Acceptance Criteria" of the S,tandard Review Plan (NUREG-0800) section 4.2. Detailed design criteria for the fuel assembly IhCA and/or seismic analysis are provided in Reference 3. I Coerational Basis Earthauake (OBE)
Requirement - The fuel assembly must remain operable during and 1
after an OBE. i Acceptance Criteria - Requires no yielding or no buckling. The stress intensity limits for most components i are taken from ASME Code Section III, -
r subsection NG. l Safa Shutdown Earthauake (SSE Requirement - The fuel assembly shall maintain structural l integrity, a coolable geometry and a path for l' control rod insertion.
Acceptance Criteria - Requires that components in path of control rods not yield and not buckle. The stress intensity limits for components in path of control rod insertion are taken from ASME Code Section III, Subsection NG. For components not in path of control rod insertion, the stress intensity limits are taken from ASME Code Section III, Appendix F.
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IDCA and Combined SSE and IDCA !
Requirement - The fuel assembly shall maintain structural integrity and a coolable core geometry. l Acceptance Criteria - The stress intensity limits for most of the !
fuel assembly components are determined using j ASME Code Section III, Appendix F.
k Exceptions - The spacer grid shall not exceed a permanent deformation limit which has been shown to maintain a coolable core geometry.
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86-1177306-00 4.0 METHOD 01DGY FOR DEVEIDPMENT OF STRUCTURAL ICADINGS i
The fuel assembly response from seismic excitation and Loss of Coolant Accident (IDCA) loadings were analyzed using the time history method as described in BAW-10133P, Rev. 1 (Reference 3).
The seismic input time histories used in the analysis conservatively envolop the Davis-Besse 1 plant specific operational basis earthquake (OBE) and safe shutdown earthquake (SSE) loadings.
The four plant specific asymmetric I4CA cases which would cause the ;
highest loads on the fuel assemblies were analyzed. These are !
- 1) Hot leg guillotine break at Reactor vessel (RV) nozzle.
- 2) Hot leg guillotine break at RV elbow inlet.
- 3) cold leg guillotine break at RV nozzle. '
- 4) Cold leg guillotine break at RV elbow inlet.
These 14CA cases are described in Reference 5. The structural .
margins are reported for the worst case LocA loading condition.
7 The fuel assembly horizontal response was determined with a core model comprised of fuel assemblies arranged in a planer array with gaps between inner fuel assemblies and the outer fuel assemblies and the core baffle plates. The upper and lower grid plate motions obtained from the Reactor Vessel Isolated Model Linear Analysis ^
described in Reference 4 were applied simultaneously to the core model, and the fuel assembly responses (displacements, moments and impact forces) were obtained. The core model used to analyze the fuel assembly horizontal response was represented by six masses instead of thirteen masses as described in the Reference 3 report. '
2 It has been shown that six masses in the core model result in the highest load. The grid impact stiffness was chosen to provide agreement with the measured impact forces in the fuel assembly lateral impact test and those from the model simulation. A description of the fuel assembly model development and correlation to fuel assembly lateral impact test results is provided in Page 9
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, 86-1177306-00 Appendix A. The core model used to analyze the fuel assembly horizontal response is described in Appendix B.
The results of the Mark-B Users Group were used for the vertical loCA loadings (Reference 5) . Tne - fuel assembly vertical model properties used for the Mark-B Users Group were envelope values of the Inconal and IAG Mark-B fuel assemblies. Hence, the results of the Mark-B Users Group core bounce vertical IDCA analysis are applicable to the Mark-BZ design. The analyses were performed for the horizontal seismic and IhCA loadings and the vertical IDCA loadings. The seismic excitation in the vertical direction will not cause fuel assembly liftoff and will not impose any higher t
loads than the normal operating loads and was not analyzed.
A mixed core (the outer one Zirc and intermediates Inconel) and an all Zirc grid core configurations were analyzed. This mixed core pattern is conservative in the impact loading analysis, since all the interior rows have the assemblies with the stiffer grids. The mixed core provided the worst case loading condition.
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5.0 METHOD 0!AGY FOR STRUCTURAL EVAINATION OF [
FUEL ASSEMBLY COMPONENTS s
The fuel assembly structural component stress analysis was performed using the loads generated by the seismic and LOCA analyses. The acceptance criteria for seismic and IDCA analyses are given in section 2. The faulted conditions that were analyzed
- are: *
(1) Operational basis earthquake (OBE),
(2) Safe shutdown earthquake (SSE),
(3) Ioss of coolant accident (IDCA) ,
(4) Combined SSE plus IhCA.
i The loads for IhCA plus SSE were combined by the sum of the square-root of the squares method as described and accepted by the NRC in l Reference 6, NUREG-0800. A vertical and horizontal dynamic analysis was performed as described in section 4. The axial and lateral loads obtained from the seismic and IhCA loads were used '
for the subsequent stress analysis. ASME Code Section III, (Reference 7) was used as a guide for the fuel assembly general 3 stress criteria.. Code level A criteria is used for the OBE condition and 14 vel D criteria is used for faulted conditions ,
1 (combined SSE plus IDCA). The analysis for most components used classical techniques. In some cases, failure loading as i established by testing were incorporated per the ASME code. The i upper and lower and fitting were analyzed using the finite element '
code FESAP (Reference 8). The guide tube bucking analysis was performed using the column secant formula. The guide tube buckling limit used in the analysis is conservative since the guide tube
) will deflect and contact the fuel rod which provides support and a higher allowable load. The guide tube and fuel rod stresses resulting from the maximum probable fuel assembly deflection were evaluated. The fuel assembly maximum probable deflection was q calculated using the accumulated fuel assembly gaps of the rows I Page 11 i
I 86-1177306-00 having the maximum number of assemblies across the core dianoter. !
'this is the maximum deflection allowed by the reactor internals constraint system (core baffle plates) . The total stress intensity l in fuel rods was calculated by considering contribution from 1) '
dynamic bending loads, 2) dynamic axial loads, and 3) steady-state ,
hoop stress caused by the pressure differential between the reactor system pressure and assuming no fuel rod internal pressure. This is conservative in the determination of maximum stress intensity in the fuel rod, since 51% of total combined SSE plus IDCA stress !
on the fuel rod is hoop stress with assuming fuel rod internal ;
pressure equal to zero. If fuel rod internal pressure was '
considered, the fuel rod hoop stress would reduce.
The spacer grid maximum impact forces occur at the two center grids t
of a fuel assembly adjacent to the baffle wall. The grid impact ,
forces are significantly lower for other grid positions above and below the center grid positions. The inner fuel assembly positions ,
also experience considerably lower impact loads. The grid permanent deformation was obtained as a function of impact loading by fuel assembly lateral impact tests. An upper bound limit curve of the grid permanent deformation was obtained from the test data.
The fuel assembly lateral impact tests were conducted at room t temperature, so the predicted loads were adjusted for reactor operating temperature effects before the spacer grid permanent '
deformation was determined. The coolable geometry of the fuel assembly corresponding to the maximum calculated grid deformation .
was confirmed by performing the limiting grid deformation calculation as discussed in section 4 of Reference 1. The ECCS analysis verified that for the limiting grid deformation (fully ,
collapsed grid as defined by the NRC in Reference 6) the fuel rod cladding temperature was within the acceptance limit of 2200'F.
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6.0 FUltL ABSEMBLY STDRAB ANALYSIR eRRULTR The fuel assembly component stress analysis was performed and f compared with the acceptance criteria discussed in section 3. The !
limiting margin for each component for each analysis is reported '
in Table 1. component properties, such as yield strength, modulus, etc., have been adjusted for reactor operating temperature. The results presented in Table 1 verify the ability of the Mark-BS fuel :
assembly to withstand the loadings associated with IOCA and/or I seismic events specified for the Davis-Besse Unit 1 plant with ,
t adequate margins. The results of the combined SSE plus toCA analysis meet the design criteria for the SSE described in section i
- 3. So a separate SSE analysis was not necessary. [
l l
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7.0 REFERENCES
l
- 1. Rancho Seco Cycle 7 Reload Report - Volume 1 - Mark BZ Fuel Assembly Design Report, BAW-1781P, April 1983. i
- 2. Rancho Seco Nuclear Generating Station Evaluation of Mark ,
BE Fuel Assembly Design, U.S, Nuclear Regulatory Commission, !
Washington, D.C., November 16, 1984. ,
- 3. Mark-C Fuel Assembly IhCA-Seismic Analyses, BAW-10133P, Rev.
1, Babcock & Wilcox, Lynchburg, Virginia, June 1986.
- 4. Reactor Coolant System Structural Loading Analysis, BAW-10131, !
Babcock & Wilcox, Lynchburg, Virginia, December 1976.
- 5. "E&W 177 -
FA Owners Group - Effacts of Asymmetric IDCA -
Imadings - Phase II Analysis," BAW-1621, July 1980. ,
- 6. Standard Review Plan, Section 4.2, NUREG-0800, Rev. 2, U.S. l Nuclear Regulatory Commission, July 1981. '
- 7. ASME Code Section III, " Nuclear Power Plant Components," 1986 Edition.
- 8. V. S. Reddy, FESAP - 3D Finite Element Thorno-Elastic Stress !
Analysis Program, Nuclear Equipment Engineering, November 1975.
- 9. 32-1176304-00, "TECO -
FA Component -
Stress-Faulted Analysis," Mark BE FMA 151.1, 1/90.
- 10. 32-1176955-00, "TECO - RUEF Faulted Analysis," Mark BZ FMA _;
151.1, 1/90.
- 11. 32-1176305-00, " Davis-Besse Mk-BZ LEF Analysis," Mark BZ FMA ;
201, 1/90. '
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I 86-1177306-00 1 APPENDIX A P'UEL ASSEMBLY LhTERAL IMPACT TESTS The fuel assembly lateral impact tests provide impact load test data over a wider range of impact energy than is obtained in the single grid impact test. The fuel assembly lateral impact test l results indicate conservatisms in the single grid impact tests. !
Additional energy without failure can be absorbed by the upacer grid mounted on the fuel assembly as compared to a spacer grid with weighted cladding segments. This has a direct influence on the )
deformation predicted by seismic and IhCA analysis. ]
l The fuel assembly lateral impact tests were performed with the fuel !
assembly in the reactor end support condition. Impact plates, j which simulate the core baffle plate, were located so that the appropriate gap existed between them and the spacer grids with the ,
fuel assembly -in an undeflected position. Single grid, multiple I grid and pendulum impact tests were run. For the single grid ;
impact test, one impact plate was located adjacent to the center ;
spacer grid, and the fuel assembly was plucked to different aid span deflections, and then released to impact with an impact surface. Six impact plates, located adjacent to the six center i
grids, were used for the multiple impact tests In these tests, the l fuel- assembly was plucked to different aid span deflections and then was released to impact with impact surfaces at all the six -
center spacer grid locations. For the pendulum impact tests, a pendulum was used to impact the third, fourth, and fifth intermediate from the bottom. The fuel assembly impact Kinetic l
energies obtained from the single grid and multiple impact tests were considerably lower (30 to 50 in-lbs range) than obtained from the pendulum impact tests. The fuel assembly impact Kinetic energies obtained from the pendulum impact tests were in the range of 300 to 2100 in-lbs. Analytical results for seismic and IhCA j conditions show Kinetic energies in the range of 800 to 1600 in-lbs.
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The pendulum impact test results were used in benchmarking the fuel assembly analytical model for high impact energies. The purpose of benchmarking the analytical model was twofold:
(1) To derive a value for the spacer grid impact stiffness.
(2) To validate the fuel assembly modelling techniques.
The fuel assembly model developed from pluck vibration test data l was used to simulate the pendulum impact test. To simulate a l
. Pendulum in the analytical model, a concentrated mass was attached I to the beam having a very low beam stiffness. The beam length was l varied to let the pendulum beam hit the fuel assembly at the proper spacer grid location. The grid was impacted at the third, fourth, I and fifth grid locations from the bottom. The effective mass of .
the pendulum at the three different grid locations, was derived by ,
equating the potential energy of the pendulum to the kinetic energy of the pendulum and solving for the mass. A gapped impact spring was positioned at impacting grid position. An initial pendulum ,
velocity was input as the initial condition of the model. The model was allowed to move freely until it struck the impact spring which simulated an equivalent stiffness of the spacer grid and the impact plate used in the test. One impact spring at a time was used to simulate the pendulum impact tests. Impact force, impact duration and impact velocities were measured during the test. The
- grid impact stiffness was chosen to provide agreement between the measured impact forces in the test and those from the model simulation. A coefficient of restitution determined from the single grid impact tests was conservatively applied in the fuel assembly seismic and Inca analysis.
, In addition to providing the spacer grid stiffness, improved acceptance criteria for spaced grids subjected to impact loading were obtained from the pendulum impact tests.
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,- 3 86-1177306-00 l
APPENDIX B !
l CORE MODEL
)
l ;
The detailed response of the fuel assemblies was determined using the core model shown in Figure 1. The core model contains five coplanar fuel assemblies in a single row to simulate the most j severe impact loading condition for the fuel assemblies. A parametric study indicated that the five fuel assembly model gives -
the highest impact loading condition and as the number of assemblies increase beyond five, the load decreases. Each fuel assembly was modelled as a spring mass system, with six masses lumped at the intermediate spacer grid locations and seven beam ,
alements representing the fuel rod and guide tube cross-sectional i inertias. The values for the rotational springs and water mass distribution were calculated based on the fuel assembly experimentally determined mode shapes and corresponding >
frequencies.
t Gapped springs were incorporated between adjacent fuel assemblies and between outer fuel assembly and the core baffle. These gapped springs were positioned at each spacer grid location. The gapped springs incorporated spacer grid dynamic properties, such as its impact stiffness and damping derived from test results. The spring gaps were based on the cold fuel assembly design spacing and were adjusted to reactor operating temperature conditions. The spacer grid dynamic properties obtained from test results were also adjusted to reflect reactor operating temperature effects. The ,
fuel assembly damping value used was established from the displacement tests on the fuel assembly in water at reactor operating conditions. The displacement time histories of the upper and lower grid plates were used as input excitation to the core model. It was assumed that the core baffle plates were rigidly attached to and have the same motion as the lower grid plate. The core model was analyzed using the STARS code outlined in Reference 3.
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.TABIA 1.~ Fuel Assembly Ctress Analysio Desults for TECO ~~
Miwad care OBE ==d N i ===a m n itam r m__ r -1r--
Component Basis for Beference Allowable Imad a namwwwin -
1/2 SSE SSE + IOCA 1/2 SSE SSE + IDCA Guide tube Buckling Buckling 348 31. 9 assembly End grid assembly ASME code ASME code 94 169 9 service service-level A level D Fuel rod ASME code ASME code 175 74 9 service service level A level D Upper end fitting ASME code ASME code 180 150 10 service service level A level D Iower end fitting ASME code ASME code 262 164 11 service service level A level D Spacer grid Elastic .Coolable 269 285 12' limit load geometry deformation
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