Containment Structural Evaluation for Pressure Capacity Summary ReptML20127M452 |
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05200001 |
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01/14/1993 |
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BECHTEL CORP. |
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Shared Package |
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ML20127M438 |
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References |
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RPRT-STRU-008, RPRT-STRU-008-R00, RPRT-STRU-8, RPRT-STRU-8-R, NUDOCS 9301280208 |
Download: ML20127M452 (9) |
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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML20137G3271997-03-31031 March 1997 Rev 4 to ABWR Design Control Document ML20117H3491996-08-31031 August 1996 Rev 3 to ABWR Design Control Document ML20101G9101996-03-22022 March 1996 Rev 7 to 25A5447, Certified Design Matl ML20101G9121996-03-22022 March 1996 Amend 36 to Rev 8 to 23A6100, ABWR Ssar ML20077R9181995-01-17017 January 1995 Rev 2 to ABWR Design Control Document ML20081K9301994-12-31031 December 1994 Rev 1 to Technical Support Document for Abwr ML20080D3411994-12-31031 December 1994 Rev 1 to Advanced BWR Design Control Document ML20081K9111994-11-30030 November 1994 Rev 0 to Technical Support Document for Abwr ML20070H9691994-07-31031 July 1994 ABWR Ssar ML20071H9981994-07-20020 July 1994 ABWR Certified Design Matl ML20069Q3061994-06-30030 June 1994 Rev 6 to Ssar Amend 35 & Rev 5 to Certified Design Matl ML20069H7401994-05-25025 May 1994 Non-Proprietary Ssar Amend 35 & Certified Design Material, Rev 4.W/164 Oversize Drawings ML20069H2181994-05-25025 May 1994 Rev 4 to ABWR Certified Matl ML20065H7861994-04-13013 April 1994 Advanced BWR Certified Design Matl/Itaac Review Guidance ML20065J5791994-04-11011 April 1994 Nonproprietary Mod Pages of Ssar,Amend 34 ML20065B7121994-03-31031 March 1994 Revised ABWR Ssar/Certified Design Matl Cross Ref Matl ML20058L2291993-12-14014 December 1993 Liquid Level Monitoring W/Emus More than Just Measuring Liquid Levels ML20062K7081993-12-13013 December 1993 Amend 33 to Advanced BWR SSAR ML20058J9831993-12-0707 December 1993 Rev 2 to Vols 1 & 2 to Advanced BWR Certified Design Matl ML20058L2161993-10-31031 October 1993 Emus Diverse Sys for Reactor Water Level Measurement in Bwrs ML20056G3121993-08-31031 August 1993 Vols 1 & 2 of ABWR Certified Design Matl A000364, Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of ABWR Design1993-07-31031 July 1993 Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of ABWR Design ML20046A2191993-07-0808 July 1993 Amend 30 to ABWR Ssar. Listed Changes Include 3H,4.6,5.1, 5.4,6.2,6.7,9.1,9.3,9.5 & 20.3 ML20045C8351993-06-18018 June 1993 Revised ABWR Design Document ML20044G7931993-05-31031 May 1993 Suppl 1 to Containment Structural Evaluation for Ultimate Pressure Capacity Rept. ML20056C3641993-05-14014 May 1993 Amend 28 to Nonproprietary Sections of Chapter 19, Response to Severe Accident Policy Statement of ABWR SSAR ML20058L8611993-04-30030 April 1993 Resolution of ISLOCA for Abwr. W/27 Oversize Encl ML20044D8821993-04-30030 April 1993 Sample Analysis for Effect of Postulated Pipe Break ABWR Main Steam Piping ML20035H6711993-04-23023 April 1993 Public Version of Amend 27 to ABWR SSAR ML20035A3071993-03-24024 March 1993 Amend 26 to Selected Sections of Chapters 1-20 of Advanced Ssar ML20128H5991993-02-28028 February 1993 Proposed Replacement of ABWR App 3A, Seismic Soil-Structure Interaction Analysis Rept ML20034H0681993-02-19019 February 1993 Advanced BWR ATWS Stability Study ML20128C7161993-02-0202 February 1993 Rev B to 23A6100AE, Submittal Supporting Accelerated ABWR Review Schedule ML20128B7051993-01-29029 January 1993 Amend 25 to ABWR Ssasr ML20128C2791993-01-29029 January 1993 Nonproprietary 11x17 Foldout Drawings to Amend 25 to ABWR Ssar ML20128C0631993-01-29029 January 1993 Draft of Section 19E.2,deterministic Analyses of Plant Performance for ABWR Ssar, Chapter 19 ML20127M4521993-01-14014 January 1993 Containment Structural Evaluation for Pressure Capacity Summary Rept ML20117A4771992-11-18018 November 1992 Changes to Nonproprietary 11x17 Foldout Drawings from Chapter 8 of ABWR Ssar,Amend 23 ML20118C0661992-09-24024 September 1992 Generic Model for Progability of Operation w/Mis-Oriented Fuel Bundle ML20101S0961992-07-0606 July 1992 Amend 21 to ABWR SSAR ML20101L7381992-07-0202 July 1992 App 19Q, ABWR Shutdown Risk Evaluation ML20101G5771992-06-0202 June 1992 Human Factors Engineering Program Review Model & Acceptance Criteria for Evolutionary Reactors ML20097A3731992-06-0101 June 1992 Tier 1 Design Certification Matl for GE ABWR Design ML20114D0901992-05-31031 May 1992 Volumes I & II of Advanced Control Room Design Review Guideline:Technical Development ML20094P3731992-03-31031 March 1992 Tier 1 Design Certification Matl for GE Advanced BWR Design - Stage 2 Submittal ML20096F0901992-03-30030 March 1992 Advanced BWR Document,Section 3.3, Piping Design, Section 3.6.1, Postulated Piping Failures in Fluid Sys Inside & Outside Containment & 3.6.2, Determination of Break Locations & Dynamic Effects Associated W/Postulated.. ML20096F1071992-03-30030 March 1992 Advanced BWR Design Document,Section 3.7, Radiation Protection, Section 12.3, Radiation Protection Design Features & 12A.1, Calculation of Airborne Radionuclides ML20036B2281988-03-31031 March 1988 Condensation-Induced Water Hammer Evaluation for ABWR ECCS Piping 1997-03-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20140F0801997-05-31031 May 1997 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design.Supplement No. 1.Docket No. 52-001.(General Electric Nuclear Energy) ML20137G3271997-03-31031 March 1997 Rev 4 to ABWR Design Control Document ML20117H3561996-08-31031 August 1996 Rev 9 to ABWR Ssar ML20117H3491996-08-31031 August 1996 Rev 3 to ABWR Design Control Document ML20101G9101996-03-22022 March 1996 Rev 7 to 25A5447, Certified Design Matl ML20101G9121996-03-22022 March 1996 Amend 36 to Rev 8 to 23A6100, ABWR Ssar ML20077R9181995-01-17017 January 1995 Rev 2 to ABWR Design Control Document ML20081K9301994-12-31031 December 1994 Rev 1 to Technical Support Document for Abwr ML20080D3411994-12-31031 December 1994 Rev 1 to Advanced BWR Design Control Document ML20081K9111994-11-30030 November 1994 Rev 0 to Technical Support Document for Abwr ML20070H9691994-07-31031 July 1994 ABWR Ssar ML20071H9981994-07-20020 July 1994 ABWR Certified Design Matl ML20069Q3061994-06-30030 June 1994 Rev 6 to Ssar Amend 35 & Rev 5 to Certified Design Matl ML20069H2181994-05-25025 May 1994 Rev 4 to ABWR Certified Matl ML20069H7401994-05-25025 May 1994 Non-Proprietary Ssar Amend 35 & Certified Design Material, Rev 4.W/164 Oversize Drawings ML20065H7861994-04-13013 April 1994 Advanced BWR Certified Design Matl/Itaac Review Guidance ML20065J5791994-04-11011 April 1994 Nonproprietary Mod Pages of Ssar,Amend 34 ML20065B7121994-03-31031 March 1994 Revised ABWR Ssar/Certified Design Matl Cross Ref Matl ML20058L2291993-12-14014 December 1993 Liquid Level Monitoring W/Emus More than Just Measuring Liquid Levels ML20062K7081993-12-13013 December 1993 Amend 33 to Advanced BWR SSAR ML20058J9831993-12-0707 December 1993 Rev 2 to Vols 1 & 2 to Advanced BWR Certified Design Matl ML20058L2161993-10-31031 October 1993 Emus Diverse Sys for Reactor Water Level Measurement in Bwrs ML20056G3121993-08-31031 August 1993 Vols 1 & 2 of ABWR Certified Design Matl A000364, Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of ABWR Design1993-07-31031 July 1993 Technical Support Document for Amends to 10CFR51 Considering Severe Accidents Under NEPA for Plants of ABWR Design ML20046A2191993-07-0808 July 1993 Amend 30 to ABWR Ssar. Listed Changes Include 3H,4.6,5.1, 5.4,6.2,6.7,9.1,9.3,9.5 & 20.3 ML20045C8351993-06-18018 June 1993 Revised ABWR Design Document ML20044G7931993-05-31031 May 1993 Suppl 1 to Containment Structural Evaluation for Ultimate Pressure Capacity Rept. ML20056C3641993-05-14014 May 1993 Amend 28 to Nonproprietary Sections of Chapter 19, Response to Severe Accident Policy Statement of ABWR SSAR ML20058L8611993-04-30030 April 1993 Resolution of ISLOCA for Abwr. W/27 Oversize Encl ML20044D8821993-04-30030 April 1993 Sample Analysis for Effect of Postulated Pipe Break ABWR Main Steam Piping ML20035H6711993-04-23023 April 1993 Public Version of Amend 27 to ABWR SSAR ML20035A3071993-03-24024 March 1993 Amend 26 to Selected Sections of Chapters 1-20 of Advanced Ssar ML20128H5991993-02-28028 February 1993 Proposed Replacement of ABWR App 3A, Seismic Soil-Structure Interaction Analysis Rept ML20034H0681993-02-19019 February 1993 Advanced BWR ATWS Stability Study ML20128C7161993-02-0202 February 1993 Rev B to 23A6100AE, Submittal Supporting Accelerated ABWR Review Schedule ML20128C2791993-01-29029 January 1993 Nonproprietary 11x17 Foldout Drawings to Amend 25 to ABWR Ssar ML20128C0631993-01-29029 January 1993 Draft of Section 19E.2,deterministic Analyses of Plant Performance for ABWR Ssar, Chapter 19 ML20128B7051993-01-29029 January 1993 Amend 25 to ABWR Ssasr ML20127M4521993-01-14014 January 1993 Containment Structural Evaluation for Pressure Capacity Summary Rept ML20126G0911992-12-21021 December 1992 Preliminary Safety Evaluation Providing Staff Conclusions Re Consistency of Containment Performance Goals of ABWR Rccv Under Severe Accident Loadings W/Deterministic Containment Performance Goals of SECY-90-016 ML20128D1401992-12-0202 December 1992 Safety Evaluation Accepting GE Proposal for Implementing ISLOCA Issue Resolution for Advanced BWR ML20117A4771992-11-18018 November 1992 Changes to Nonproprietary 11x17 Foldout Drawings from Chapter 8 of ABWR Ssar,Amend 23 ML20118C0661992-09-24024 September 1992 Generic Model for Progability of Operation w/Mis-Oriented Fuel Bundle ML20101S0961992-07-0606 July 1992 Amend 21 to ABWR SSAR ML20101L7381992-07-0202 July 1992 App 19Q, ABWR Shutdown Risk Evaluation ML20101G5771992-06-0202 June 1992 Human Factors Engineering Program Review Model & Acceptance Criteria for Evolutionary Reactors ML20097A3731992-06-0101 June 1992 Tier 1 Design Certification Matl for GE ABWR Design ML20114D0901992-05-31031 May 1992 Volumes I & II of Advanced Control Room Design Review Guideline:Technical Development ML20095A2461992-04-0808 April 1992 Amend 20 to Advanced BWR SSAR 11x17 Foldout Drawings,Page Change Instructions & Page Status Sheets ML20094P3731992-03-31031 March 1992 Tier 1 Design Certification Matl for GE Advanced BWR Design - Stage 2 Submittal 1997-05-31
[Table view] |
Text
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ADVANCED BOILING WATER REACTOR Job No.18775 CONTAJNMENT STRUCTURAL EVALUATION FOR PRESSURE CAPACITY
SUMMARY
REPORT Bechtel Report No. RPRT-STRU 008 n,.
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ABWR CONTAINMENT ULTIMATE STRENGTH _ EVALUATION-
SUMMARY
REPORT-TABLE OF CONTENTS 1.
INTRODUCTION 1-y 2.
FINITE ELEMENT (FE) MODEL DESCRIPTION...................
-~ 1 3.
A N A LY S I S..........................................
3 4.
RESULTS...........................................
3' 5.
C O N C LU S I O N S.......................................
4 TABLE 1:.
Summary of Stresses and Strains TABLE 2:
Summary of Pressure Capabilitics of Various, Components of the RCCV
. TABLE 3:
"FINEL" Model ATTACHMENT 1
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ABWR - CONTAINMENT ULTIMATE STRENGTH EVALUATION
SUMMARY
REPORT 1.
INTRODUC_T10fl This report summarizes the ultimate strength evaluation of the Reinforced Concrete Containrnent Vessel (RCCV) for the Advanced Boiling Water Reactor (ABWR). Bechtel proprietary computer code "FINEL" was used for evaluation of the axisymmetrical components of the RCCV. Attachment 1 gives brief description of the "FINEL" program.
2.
FINITE ELfLMENT (FF) MODEL DESCRIPTION The containment and the contairement internal structures are axisymmetric while the RCCV top slab together with the reinforced concrete girders even though not axisymmetric, are idealized and included in the axisymmetrical model. Solid elements are used to represent the girders at the top of the RCCV, approximating the stiffness of the actual structure.
For simplicity, the Reactor Pressure Vessel (RPV), the reactor building outside of the RCCV and superstructure above the operating floor, are not modeled. To represent the restraining effects of the floors outside the containment, horizontal restraining elements are used with pseudo material properties. The model includes concrete elements, the reinforcing steel, the l
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steel liner plato of the drywell and the wetwell and the diaphragm floor structures and the structural stect elements used for the pedestal.
The model consists of 868 nodal points and 1280 elements. 448 elements are with unidirectional stiffness representing robar, whereas 832 elements are isotropic, representing stool, concrete, and soil. The soil below the foundation mat was modeled to a depth of 50.0m and to a radius of 76.0m.
See Figure 1 for the model.
The FINEL computer program permits the specification of bi-linear, brittle or ductile material properties. The concreto and soil elements are specified to have brittle properties such that they are strong in compression and weak in =
tension. The steel plate elements and the rebar elements are specified to have ductile material proporties with the same-strength in tension and compression. The capability of the FINEL program to accommodate ductile and brittle material behaviors permits both concrete cracking and yielding of steel and rebar. This allows the program to consider redistribution of forces throughout the structure due to the non-linear behavior.
Reinforcing steel used in the FINEL model is based on the structural design shown in the design drawings included'in the SSAR.
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3.
ANALYSIS The FE model was run for three different load conditions shown in Table 1.
1.
Structural Integrity Test 1 (SIT-1), with 52 psig pressure in the drywell and wetwell (RCCV).
2.
Structural !ntegrity Test 2 (SIT-2), with 45 psig pressure in the drywell-and 20 psig in the wetwell.
3.
Four times design pressure (4 Pa), with 180 psig pressure in the RCCV.
S Since FINEL performs non-linear analysis, it is necessary to apply simultaneously all loads of a loading combination. The-program utilizes' a stepwise linear iteration technique. The first cycle ~ results are of clastic analysis. Based upon results of the first cycle, stiffnesses of all elements are adjusted by the program prior to the next iteration cycle, 4.
RESU1LS 4.1 Table 1 summarizes analytical results for various loading conditions. The results are shown in terms of maximum rebar stresses, concrete stresses, liner strains and structural deformations.
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4.2 Table 2 lists calculated pressuro capabilities of the various components of-the RCCV based on extrapolation of the analytical results for Level "C" allowablo stresses and the ultimato capacity.
However, it should be recognized that the extrapolation of results gives only approximate values beyond the analyzed values.
5.
CONCt.USIONS 5.1 Axisymmetric Components of RCCV Based on the FINEL analysis, it can be concluded that the RCCV (other than top slab and drywell head), as designed based on ASME Section ll1 Division 2 code requirements, can withstand an internal pressure of 180 psig i.e.,
four times the design pressure, with the stresses and strains in the rebar, liner plate and concrete within the code allowable limits. Pressure capability was extrapolated to be 198 psig for level "C" allowables and was found to be governed by wetwell wall (sco Tablo 2).-
4 W:\\ASWR\\ REPORTS \\STRU XXX.RVO
52 RCCV Top Slab The "FINEL" analysis results are not applicable for the top stab, as it is not an axisymmetric component. Based on extrapolation of etactic "STARD analysis results, it was found that the top slab has pressure capabilit psig based on level "C" allowable stresses and is governed by the 164 supporting pool girders' strength. However,it should be recognized th i
lt value could be somewhat different, based on in-clastic analys s resu s.
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TABLE 'l SLYMARY OF STRESSES AN3 STRA]NS MAX.
CCMPOWENT REBAR STRESSES /
MAX 1RN RADIAL LOADikG CASE REBAR STRESS /
L1EER STRAIN CONCRETE ALLOWABLE STRESSES (KSI)
DEFL.9 CCHP.
STRESS /
WETWELL DRYWELL BASEFAT 01APHRG.
WET-ALLOW. STRESS VELL I P.D.fP.W.
MERID.
HOOP TEMS.
COMP.
ALLOW.STR WD.
TITLE PSI PSI
'KSI KS1 IN/IH IN/IN KS1 MER.
N00P MER.
H30P RAD.
HDOP
__ RAD.
HOOP IN.
i 1
SIT-1 52.0 52.0 11.5 12.0
.00052
.00011
.56 11.5 12.0 6.2 5.1 4.0 4.4 10.9 6.2
.246
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j 45.0 45.0
-2.40 45.0 45.0 45.0 45.0 45.0 45.0 45.0 45.0 l
2 SIT-2 45.0 20.0 B.8 4.5
.00035 '
.00007
.54 6.0 4.5 2.9 3.3 3.8 4.0 8.8 3.8
.070
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45.0 45.0
-2.40 45.0 45.0 45.0 45.0 45.0 45.0 45.0 45.0 l
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4Pa.
180.0 180.0 40.3 49.0
.D0185
.00016
.68 40.3 49.0 29.1 13.6 12.2 10.8 33.4 18.6
.985 60.0 60.0
-3.40 60.0 60.0 60.0 60.0 60.0 60.0 60.0 60.0 l
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SUMMARY
OF PRESSURE CAPA31LITIES Of VARIOUS COMPONENTS OF THE RCCV 2
PRESSURE CAPABILITY (P51G)
CATER 0 RIES (CRITERIA).
STRUCTURAL COMPOWENT LEVEL C ULTIMATE WETWELL 198 249 UPPER DRYWELL 334
>-371 BASEMAT 638-885.
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TOP SLAB 164 The pressure cepability shown.for the'P.CCV top slab-
- NOTES: -*-
Wich is a non-axisynenetric portion of'the RCCV, is calculated based on extrapolation of elastic STARDYNE 1
analysis results. Pressure value 164 pais is governed by the pool girders,. pressure capacity of.the rein-forcing of the top slab is;178 psig.
Ultimate capability has been calculated based on re-bars at both faces "of a cross section reaching yield:
stressv (Greater than)' sign snenns that rebar on only onw face -
of. the section reached yield, 'and the ultimate capac-:---
Ity will be higher then the value indicated.:
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FIGURE 1 FINEL M00EL
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Finite Element Program for Cracking Analysis (FiWEL)
Descriotion FINEL is a proprietary computer program of Bechtcl Power Corporation, San Francisco, California. The FINEL program performs a static analysis of stresses and strains in plane and axisymmetric structures by the finite element method.
The program performs the non linear static analysis utilizing a stepwise linear iteration solution technique. Nthin each solution cycle, status of all elements is determined and their stiffness adjusted by the program prior to the next iteration cycle. The Von Mises yield criterion is used to determine the status of all ductile materials and brittle materials which are in compression. A ductile material is assumed to yield in all directions when the yield criterion is exceeded. A brittic material is assumed to be cracked in the direction in which the maximum principal stress exceeds the specified tensile stress. The modulus of clasticity for each material is adjusted for the next solution cycle to conform to the secant modulus correspondingto the calculated strain in the element following the bilinear stress strain relationship specified. The numerical algorithm assumes that the state of stress which exists when the converged solution is achieved is independent'of the stress history of the loading.
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Extent of Acolication This proGrarn is used for the static load analysis of the reactor building 'and containment to determine stresses and strains in the various structural elements and resultant forces and moments at selected sections.
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