ML20195C672

Safety Evaluation of Topical Rept WCAP-15029, Westinghouse Methodology for Evaluating Acceptability of Baffle-Former- Bolting Distribution Under Faulted Load Conditions

ML20195C672
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Issue date:	11/10/1998
From:	NRC (Affiliation Not Assigned)
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ML20195C640	List: ML20195C636 ML20195C672
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Download: ML20195C672 (7)

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. [pW*% t UNITED STATES j j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666-0001

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WCAP-15029. " WESTINGHOUSE METHODOLOGY FOR EVALUATING THE ACCEPTABILITY OF BAFFLE-FORMER-BARREL BOLTING DISTRIBUTIONS UNDER FAULTED LOAD CONDITIONS" l

INTRODUCTION:

Over the past year, the NRC and the Westinghouse Owners Group (WOG) have been interacting on the issue of potential failure of baffle-former-barrel bolting. This work is in response to reported foreign experience with baffle bolt failure. Baffle bolt cracking experience in PWRs is described in NRC Information Notice 98-11, Cracking of Reactor Vesselintemal Baffle Former Bolts in Foreign Plants, dated March 25,1998.

By letter dated June 19,1998, (Reference 1) WOG submitted its assessment of the safety significance impact of potential baffle bolt cracking on domestic Westinghouse designed reactor vessel internals. The letter indicated that a large number of the later designed domestic Westinghouse plants contain reactor vesselintemals design features that reduce the magnitude of the bolt loads induced during a faulted event to comparatively small values. These design features are described as bolt cooling holes and baffle plate pressure relief holes which minimize the potential for age related effects on the bolts to significantly affect plant safety and operation. For the remaining Westinghouse domestic plants, WOG indicated there is a large vadation in plant features and a significant variation among the plants in the potential for bolt degradation and consequences of bolt degradation on plant safety and operation. The assessment was based on both a deterministic and risk-informed evaluation. WOG concluded that its assessment confirmed its belief that the issue is not an immediate safety concem, and that it is appropriate to treat the baffle bolt cracking as an aging management issue.

In its letter dated June 19,1998, WOG proposed that licensees perform corrective action inspections and replacement of baffle bolts under the provisions of 10 CFR 50.59 for two lead plants (2 & 3 loop). In determining the applicability of the 10 CFR 50.59 licensing approach, WOG requested NRC review and approval of WCAP-15029, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, and WCAP-14748/9, Justification ofincreasing Postulated Break Opening Time.

The latter WCAP report invokes the leak-before-break (LBB) concept and provides a justification for break-opening times greater than one millisecond. The LBB concept and break-opening times greater than one millisecond are applied, by reference,in the methodology guidance provided in WCAP-15029. WCAP-14748/9 was reviewed separately and the staff's SE was transmitted to WOG on October 1,1998 (Reference 2). Consequently,it is not 9811170193 981110 Enclosure PDR TOPRP EMVWEST C PDR

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included in this SE of WCAP-15029. Further, WOG requested that WCAP-9735 Rev.2, I Multiflex 3.0, A Fortran IV Computer Program for Analyzing Thermal Hydraulic Structural l System Dynamics-Advanced Beam Model, which was previously submitted for staff review and l

approval, be withdrawn from review under the NRC licensing topical report program for l referencing in licensing actions. Multiflex 3.0 is also applied in WCAP 15029 methodology guidance, by reference, but is not evaluated in this SE. Evaluation of the Multiflex 3.0 methodology is not a requisite for concluding that WCAP 15029 is acceptable.

DISCUSSION The WCAP-15029 topical report provides systematic guidance in the form of a flow diagram and associated methodology narrative that identifies the sequence of analyses for evaluating the acceptability of baffle bolts and bolting distributions under faulted conditions. The methodology applies referenced prescribed analytical methods and assumptions to determine

! acceptable baffle bolt loading, bolting distributions, allowable stress limits for certain bolt materials, and options to address control rod insertion ability and core coolabilty criteria. The topical report indicates that it supersedes a referenced and previously NRC-approved methodology (Reference 4), that demonstrated safe plant operation during faulted conditions, by broadening the previous methodology to address the safe operation of Westinghouse plants l with reduced baffle-former-barrel bolting. Further, the report indicates that the broadened methodology consists of refinements and enhancements of previously used analyses, and that the prescribed analyses methods, assumptions and options use realistically simulated faulted conditions while retaining an adequate measure of conservatism.

The broadened methodology applies the LBB concept, break-opening times greater than one

, millisecond, and the Multiflex 3.0 program with its advanced fluid-structure interaction L simulations in the core barrel baffle-former region. The methodology incorporates the application of the best-estimate WCOBRA/ TRACK code on intermediate pipe break sizes to establish the depressurization transient that is used to demonstrate that two-phase loads are

, less limiting than single phase loads for smaller breaks under the LBB concept. The validity of I

the demonstration is left to the licensee in the application of its plant conditions and current i licensing bases in using the methodology.

l LOCA initiation gives rise to three types of loads: 1) Acoustic depression waves which traverse

i. the reactor coolant system from the location of the break at the speed of sound. Peak loads from these acoustic pulses occur at about 50 msec after break initiation; 2) Mass flow induced loads from fluid motion caused by the depressurization forces. These loads peak between 100 and 200 msec and could be limiting for core components subject to cross flows following depressurization; and 3) two phase flow frictional and pressure loads. These loads are significant in the bypass region where high pressure differentials are created during l I

depressurization and appear between 5 and 50 msec after LOCA initiation. The first and second loads are estimated using the Multiflex 3.0 code. The third load is covered below.

j The logic of the methodology depicted in the WCAP-15029 flow chart is an appropriate process

for evaluating acceptable bolting loads and distributions. The Multiflex 3.0 program is described

as a more sophisticated analysis tool for LOCA hydraulic force calculations than the currently i 4 l

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3 approved version, Multiflex 1.0 WCAP-15029 indicates that the Multiflex 3.0 program enhancements of Multiflex 1.0 include: the use of a two dimensional flow network to represent the vessel downcomer region in lieu of a collection of one dimensional parallel pipes; the allowance for nonlinear boundary conditions at the vessel and downcomer interface at the radial keys and the upper core barrel flange in lieu of simplified linear boundary conditions; and the allowance for vessel motion in lieu of rigid vessel assumptions. WCAP-15029 indicates that these modifications are included in the Multiflex 3.0 program that is used to estimate the LOCA hydraulic forces on the vessel and consequential forces induced on the fuel and reactor vessel internal structures. The staff concurs with the WOG that Multiflex 3.0 provides a more accurate and realistic modeling approach. On this basis, and considering that Multiflex 3.0 is based on the previously approved Multiflex 1.0, the staff considers the application of Multiflex 3.0 with the WCAP 15029 methodology reasonable and acceptable.

The core structural integrity during a LOCA transient or a seismic event is analyzed using the

" Verification Testing and Analysis of 17x17 Optimized Fuel Assembly" code which has been approved by the NRC and is described in Reference 4. However, the two phase loads el the baffle plates resulting from rapid depressurization are not normally calculated during LOC A analyses. Early Westinghouse calculations using the SATAN code, Reference 5, have shown that such loads were, in general, comparable to the single phase acoustic loads. These loads depend on the reactor vessel depressurization rate, which in tum depends on the size, location and type of the break. In the proposed methodology, a single depressurization calculation will be performed for each plant type (two- three- or four-loop) to be bounded if, as expected, the two phase loads are not limiting, they will be considered as having been adequately addressed.

Should the two phase loads tum out to be comparable with the acoustic loads then a more detailed calculation will be required.

The break sizes under consideration range from the accumulator piping (about 10 inches in diameter) to the pressurizer surge line (about 14 inches in diameter). The code chosen for these calculations is WCOBRA/ TRAC. The code has been approved for best estimate large break LOCA calculations for three- and four-loop plants, Reference 6. A two loop (upper plenum injection) best estimate version, has been submitted for review, Reference 7. The staff has completed its review of WCOBRA/ TRAC as a best estimate LOCA code for the two loop plant baffle load application, and the SE is being issued separately. The code has been used for breaks ranging from split breaks (less than 60 inch') to double ended guillotine breaks with a Moody coefficient of 1.0. A finite element representation is shown (octant geometry) which accounts for the baffle-former-barrel, and baffle to baffle bolts and the former holes for the bypass flow. A specific bolt configuration is used for each case. The acceptability of that configuration depends on the resulting bolt loadings and allowable stress and deflection limits for the plant-specific appucation.

The use of best estimate WCOBRA/ TRAC code for the estimation of two-phase loads on the baffle bolts contained in the methodology is acceptable provided that: 1) the limiting baffle bolt loading will be determined by analysis for a class of plants and a specific break; 2) the noding to be used in the representation of the loading is demonstrated to be adequate by performing nodalization sensitivity studies or by some other acceptable methodology.

4 l The methodology developed by WOG for Westinghouse plants is proposed for use in l determining the acceptability of replacement bolts and bolting distributions and, in part, to assess existing bolting using appropriate irradiated bolt material properties. The bolt l replacement activities are subject to the results of plant inspectior's for baffle bolt cracking I

degradation. The topical report references the ASME Boiler and Pressure Vessel (B&PV)

Code, Section Ill, for materials and formulations for determining allowable stress limits to be l used in selecting replacement bolts for normal, upset and faulted conditions. The ASME B&PV Code does not provide allowable stress limits for irradiated materials. The WCAP report contains proposed irradiated stress allowable limits for use in determining the allowable primary membrane and primary membrane plus primary bending stress limits for baffle bolting under faulted conditions in accordance with the formulation requirements of Appendix F of Section lll of the ASME B&PV Code.

It is not evident that this approach is entirely justifiable. Implicit in tne ASME Code stress allowable limit is the fact that the bolting materials should exhibit ductility that provides a strain j margin between yield and ultimate strength. For irradiated austenitic stainless steel, the available data indicate a severe loss of ductility, with the yield strength and the ultimate strength in some cases being coincident. It is not clear that the stress allowable limits proposed for irradiated stainless steel account for the lack of ductility of the material.

[ The inservice inspection (ISI) methods used for inspecting the baffle bolts will have certain ,

l sensitivity and probability of detection (POD) characteristics. In using irradiated tensile strength for calculating safety margins, the bolt load-bearing cross-section area should be reduced based upon the sensitivity of the ISI methods and the potential for degraded bolts to escape detection by ISI. In addition, since the available data for irradiated stainless steel indicates that the ductility (and presumably the fracture toughness) of irradiated bolts is severely degraded, a fracture mechanics approach would be necessary to demonstrate that the degraded bolts exhibit adequate toughness with a postulated flaw size undetected by ISI.

The report indicates these proposed stress allowable limits are based on yield strength and ultimate strength values obtained from irradiated bolting materials. The report indicates that the selected values are conservative considering the source irradiated materials had significantly lower fluence levels than bolts in the baffle region. The conservatism is based on experience which demonstrates that for certain materials the yield strength and ultimate strength increase with fluence level. The determination of adequate conservatism of bolting material properties and characteristics, resides with the licensee in applying appropriate stress and deflection limits to the baffle assembly under faulted conditions. This determination should include consideration of the subject plant's historical operating conditions and current licensing bases.

The determination should be based on conservative yield strength and ultimate strength values representative of the plant's existing bolting material properties.

Although the methodology provides systematic guidance for an acceptable logical analytical t approach for determining baffle bolt loading under faulted conditions, it has not identified any ,

l requirements for a baffle bolt post-replacement monitoring program. The implementation of an '

i inspection program is appropriate for baffle bolt cracking aging management. The inspection

program must be capable of detecting baffle bolt cracking and degradation consistent with the i

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5 results of the plant's bolt inspection and replacement activities and applicable industry ,

experience. Further, the inspection program should implement a root-cause evaluation.of any replacement baffle bolt degradation discovery.

CONCLUSION The use of the WCAP-15029 methodology guidance is acceptable in accordance with the following limitations:

1. The bolt loading should be determined by analysis for a class of plants and a specific break;

2. The noding to be used in the representation of the loading is demonstrated to be adequate by performing nodalization sensitivity studies or by some other acceptable l methodology; l

3. The methodology should not be used to assess existing botting without demonstration of adequate conservatism in projected bolting material properties (i.e., yield and ultimate strength) to ensure that sufficient ductility is present in existing irradiated stainless steel bolting materials;

4. . The use of the methodology for existing irradiated stainless steel bolting should account for limitations in available ISI methods with regard to the probability of detection characteristics; Finally, in consideration of the WOG assessment and conclusion that the baffle bolt issue is not an immediate safety concern and that it is appropriate to treat baffle former bolt degradation as an aging management issue, subsequent to replacement of baffle bolts, licensees are expected

' to develop an appropriate inspection monitoring and aging management program for baffle botting.

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REFERENCES i

1. Westinghouse Owners Group Letter, dated June 19,1998, " Baffle Barrel Bolt Program-Response to April 9,1998, NRC Letter (MUHP5135)," by Louis F. Liberatori, Jr.

2. NRC letter, dated October 1,1998, " Safety Evaluation Related to Topical .

Report WCAP-14748/49-Justification for increasing Break Opening Times in j Westinghouse PWRs," to Lou Liberatori, Chairman, WOG Steering Committee l

3. WCAP-15029, " Westinghouse Methodology for Evaluating the Acceptability of Baffle-i Former-Barrel Bolting Distributions Under Faulted Load Conditions" by P. E. Schwirian, et al., Westinghouse Electric Company,1998

4. WCAP-9401PA," Verification Testing and Analysis of 17x17 Optimized Fuel Assembly,"

by M. D. Beaumont, et al., Westinghouse Electric Company, August 1981

5. WCAP-8302," SATAN-IV Program Comprehensive Space Time Analysis of Loss of Coolant," by F.M. Bordelon, et al., Westinghouse Electric Company

6. WCAP-12945P, Volume 4, " Code Qualification Occument for Best Estimate LOCA i Analysis Volume IV: Assessment of Uncertainty," by S. B. Bajorek, et al., Westinghouse l Electric Company,1993

7. WCAP-14449P " Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," by S.1. Dederer, et al.,

Westinghouse Electric Corporation,1993

' Principal Contributor: F. Grubelich, NRR/EMEB (301) 415-2784 Date: November 10, 1998 l

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! Westinghouse Owners Group Project No. 694 cc:

Mr. Nicholas Liparulo, Manager l Equipment Design and Regulatory Engineering Westinghouse Electric Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group '

Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355

! Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director Regulatory Affairs Westinghouse Electric Corporation '

11921 Rockville Pike Suite 107

Rockville, MD 20852 Mr. Hank Sepp, Manager Regulatory and Licensing Engineering Westinghouse Electric Corporation PO Box 355 j Pittsburgh, PA 15230-0355 t

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[pm uq t UNITED STATES y

j NUCLEAR REGULATORY COMMISSION ,

$ WASHINGTON, D.C. 20606-0001 l

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\;'*****/ November 10, 1998 l l

Mr. Lou Liberatori, Chairman Westinghouse Owners Group Steering Committee l Indian Point Unit 2 Broadway & Bleakley Ave.

Buchanan, NY 10511

SUBJECT:

SAFETY EVALUATION OF TOPICAL REPORT WCAP-15029

" WESTINGHOUSE METHODOLOGY FOR EVALUATING THE ACCEPTABILITY OF BAFFLE-FORMER-BARREL BOLTING DISTRIBUTIONS UNDER FAULTED LOAD CONDITIONS" (TAC NO. MA1152)

The staff has completed its review of the subject report requested by Westinghouse Owners l' Group (WOG) by letter of June 19,1998. The staff has found that this report is acceptable for referencing in licensing applications to the extent specified and under the limitations delineated in the report and in the associated NRC safety evaluation, which is enclosed. The safety evaluation defines the basis for acceptance of the report.

In response to reported foreign experience with baffle bolt failures and concerns of potential degradation of baffle barrel bolts in domestic plants, the WOG proposed that licensees perform corrective action inspections and replacement of baffle bolts under the provisions of 10 CFR 50.59 for two lead plants (2 & 3 loop). In determining the applicability of the 10 CFR 50.59 )

licensing approach, WOG requested NRC review and approval of WCAP-15029,

" Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Botting Distributions Under Faulted Load Conditions," and WCAP-14748/9," Justification of increasing Postulated Break Opening Time." The latter WCAP report invokes the leak-before-break (LBB) concept and provides a justification for break-opening times greater than one millisecond. The LBB concept and break-opening times greater than one millisecond are applied, by reference, in the methodology guidance provided in WCAP-15029. WCAP-14748/9 was reviewed separately and the staff's safety evaluation was transmitted to WOG on October 1,1998.

Consequently, it is not included in this safety evaluation of WCAP-15029. Further, WOG requested that WCAP-9735 Rev.2, "Multiflex 3.0, A Fortran IV Computer Program for Analyzing Thermal Hydraulic Structural System Dynamics-Advanced Beam Model," which was previously submitted for staff review and approval, be withdrawn from review under the NRC licensing topical report program for referencing in licensing actions. Multiflex 3.0 is also applied in WCAP-15029 methodology guidance, by reference, but is not evaluated in this safety evaluation.

The staff will not repeat its review of the matters described in the WOG Topical Report WCAP-15029, when the report appears as a reference in license applications, except to ensure that the material presented applies to the specific plant involved. In accordance with procedures established in NUREG-0390, the NRC requests that WOG publish accepted versions of the submittal, proprietary and non-proprietary, within 3 months of receipt of this letter. The accepted versions shrall incorporate this letter and the enclosed safety evaluation f0/0 h

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L. Liberatori -2 November 10, 1998 l

between the title page and the abstract and an -A (designating accepted) following the report identification symbol.

If the NRC's criteria or regulations change so that its conclusion that the submittalis acceptable are invalidated, WOG and/or the applicant referencing the topical report will be expected to -

revise and resubmit its respective documentation, or submit justification for the continued applicability of the topical report without revision of the respective documentation.

Sincerely, Thomas H. Essig, Acting Chief Generic Issues and Environmental Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 694

Enclosure:

Safety Evaluation cc w/ enc: See next page l

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- i l L. Liberatori -2 November 10,1998 i between the title page and the abstract and an -A (designating accepted) following the report identification symbol.

if the NRC's criteria or regulations change so that its conclusion that the submittalis acceptable are invalidated, WOG and/or the applicant referencing the topical report will be expected to revise and resubmit its respective documentation, or submit justification for the continued applicability of the topical report without revision of the respective documentation. l Sincerely,  ;

Original Signed By:

Thomas H. Essig, Acting Chief

~ Generic Issues and Environmental Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation Project No. 694

Enclosure:

Safety Evaluation 4 cc w/ enc: See next page DISTRIBUTION l Project File JRoe/DMatthews RCaruso JOBUCA TEssig LLois PGEB r/f FAkstulewicz AEl-Bassioni OGC- PWen i ACRS FGrubelich DOCUMENT NAME: G:\pxw\wog15029.ser To receive a copy of this document, indicate in thepox C= Copy w/o attachment / enclosure E= Copy withattachment/ enclosure N = No copy M #(f/f/

OFFICE PM:PGEB 6 BC:EMEB 4 SC:PGEB' ,$ (A) BC:PC3EB NAME PWen:sw ?cv/ RWessh FAkstulewib TEssig A' d DATE 11 /9 /98 11/ @ /98 11/ f'/98 $$/10 /98 OFFICIAL RECORD COPY r

[fu% j g) t UNITED STATES NUCLEAR REGULATORY COMMISSION I

o 2 WASHINGTON, D.C. 20666-0001 Q**CiMl

• o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION WCAP-15029. " WESTINGHOUSE METHODOLOGY FOR EVALUATING THE ACCEPTABILITY OF BAFFLE-FORMER-BARREL BOLTING DISTRIBUTIONS UNDER ,

l FAULTED LOAD CONDITIONS" l

INTRODUCTION: I Over the past year, the NRC and the Westinghouse Owners Group (WOG) have been interacting on the issue of potential failure of baffle-former-barrel bolting. This work is in response to reported foreign experience with baffle bolt failure. Baffle bolt cracking experience in PWRs is described in NRC Information Notice 98-11, Cracking of Reactor Vesse/ Intemal Baffle Formor Bolts in Foreign Plants, dated March 25,1998.

By letter dated June 19,1998, (Reference 1) WOG submitted its assessment of the safety significance impact of potential baffle bolt cracking on domestic Westinghouse designed reactor vessel intemals. The letter indicated that a large number of the later designed domestic Westinghouse plants contain reactor vessel intemals design features that reduce the magnitude l of the bolt loads induced during a faulted event to comparatively small values. These design features are described as bolt cooling holes and baffle plate pressure relief holes which minimize the potential for age related effects on the bolts to significantly affect plant safety and operation. For the remaining Westinghouse domestic plants, WOG indicated there is a large variation in plant features and a significant variation among the plants in the potential for bolt degradation and consequences of bolt degradation on plant safety and operation. The assessment was based on both a deterministic and risk-informed evaluation. WOG concluded that its assessment confirmed its belief that the issue is not an immediate safety concem, and l

that it is appropriate to treat the baffle bolt cracking as an aging management issue. j in its letter dated June 19,1998, WOG proposed that licensees perform corrective action i inspections and replacement of baffle bolts under the provisions of 10 CFR 50.59 for two lead plants (2 & 3 loop). In determining the applicability of the 10 CFR 50.59 licensing approach, i WOG requested NRC review and approval of WCAP-15029, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, and WCAP-14748/9, Justification ofIncreasing Postulated Break Opening Time.

The latter WCAP report invokes the leak-before-break (LBB) concept and provides a justification for break-opening times greater than one millisecond. The LBB concept and break-opening times greater than one millisecond are applied, by reference, in the methodology guidance provided in WCAP-15029. WCAP-14748/9 was reviewed separately and the staff's SE was transmitted to WOG on October 1,1998 (Reference 2). Consequently, it is not Enclosure ffO

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included in this SE of WCAP-15029. Further, WOG requested that WCAP-9735 Rev.2, Multiflex 3.0, A Forfran IV Computer Program for Analyzing ThermalHydraulic Structural System Dynamics-Advanced Beam Model, which was previously submitted for staff review and approval, be withdrawn from review under the NRC licensing topical report program for referencing in licensing actions. Multiflex 3.0 is also applied in WCAP 15029 methodology l guidance, by reference, but is not evaluated in this SE. Evaluation of the Multiflex 3.0 methodology is not a requisite for concluding that WCAP 15029 is acceptable.

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DISCUSSION l

The WCAP-15029 topical report provides systematic guidance in the form of a flow diagram and associated methodology narrative that identifies the sequence of analyses for evaluating the acceptability of baffle bolts and bolting distributions under faulted conditions. The methodology applies referenced prescribed analytical methods and assumptions to determine acceptable baffle bolt loading, bolting distributions, allowable stress limits for certain bolt l materials, and options to address control rod insertion ability and core coolabilty criteria. The i topical report indicates that it supersedes a referenced and previously NRC-approved methodology (Reference 4), that demonstrated safe plant operation during faulted conditions, by broadening the previous methodology to address the safe operation of Westinghouse plants with reduced baffle-former-barrel bolting. Further, the report indicates that the broadened methodology consists of refinements and enhancements of previously used analyses, and that l the prescribed analyses methods, assumptions and options use realistically simulated faulted i conditions while retaining an adequate measure of conservatism.

The broadened methodology applies the LBB concept, break-opening times greater than one j millisecond, and the Multiflex 3.0 program with its advanced fluid-structure interaction l simulations in the core barrel baffle-former region. The methodology incorporates the application of the best-estimate WCOBRA/ TRACK code on ir,termediate pipe break sizes to establish the depressurization transient that is used to demonstrate that two-phase loads are less limiting than single phase loads for smaller breaks under the LBB concept. The validity of the demonstration is left to the licensee in the application of its plant conditions and current licensing bases in using the methodology.

LOCA initiation gives rise to three types of loads: 1) Acoustic depression wwas which traverse the reactor coolant system from the location of the break at the speed of sound. Peak loads I

from these acoustic pulses occur at about 50 msec after break initiation; 2) Mass flow induced loads from fluid motion caused by the depressurization forces. These loads peak between 100 and 200 msec and could be limiting for core components subject to cross flows following depressurization; and 3) two phase flow frictional and pressure loads. These loads are significant in the bypass region where high pressure differentials are created during depressurization and appear between 5 and 50 msec after LOCA initiation. The first and second loads are estimated using the Multiflex 3.0 code. The third load is covered below.

(

The logic of the methodology depicted in the WCAP-15029 flow chart is an appropriate process i for evaluating acceptable bolting loads and distributions. The Multiflex 3.0 program is described as a more sophisticated analysis tool for LOCA hydraulic force calculations than the currently 1

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3 approved version, Multiflex 1.0. WCAP-15029 indicates that the Multiflex 3.0 program enhancements of Multiflex 1.0 include: the use of a two dimensional flow network to represent the vessel downcomer region in lieu of a collection of one dimensional parallel pipes; the allowance for nonlinear boundary conditions at the vessel and downcomer interface at the radial keys and the upper core barrel flange in lieu of simplified linear boundary conditions; and the allowance for vessel motion in lieu of rigid vessel assumptions. WCAP-15029 indicates that these modifications are included in the Multiflex 3.0 program that is used to estimate the LOCA hydraulic forces on the vessel and consequential forces induced on the fuel and reactor vessel internal structures. The staff concurs with the WOG that Multiflex 3.0 provides a more accurate and realistic modeling approach. On this basis, and considering that Multiflex 3.0 is based on the previously approved Multiflex 1.0, the staff considers the application of Multiflex 3.0 with the WCAP 15029 methodology reasonable and acceptable. 1 The core structural integrity during a LOCA transient or a seismic event is analyzed using the

" Verification Testing and Analysis of 17x17 Optimized Fuel Assembly" code which has been approved by the NRC and is described in Reference 4. However, the two phase loads on the i baffle plates resulting from rapid depressurization are not normally calculated during LOCA analyses. Early Westinghouse calculations using the SATAN code, Reference 5, have shown

)

j that such loads were, in general, comparable to the single phase acoustic loads. These loads depend on the reactor vessel depressurization rate, which in turn depends on the size, location and type of the break. In the proposed methodology, a single depressurization calculation will ,

be performed for each plant type (two- three- or four-loop) to be bounded. If, as expected, the I two phase loads are not limiting, they will be considered as having been adequately addressed.

Should the two phase loads turn out to be comparable with the acoustic loads then a more detailed calculation will be required.

The break sizes under consideration range from the accumulator piping (about 10 inches in diameter) to the pressurizer surge line (about 14 inches in diameter). The code chosen for these calculations is WCOBRA/ TRAC. The code has been approved for best estimate large break LOCA calculations for three- and four-loop plants, Reference 6. A two loop (upper plenum injection) best estimate version, has been submitted for review, Reference 7. The staff has completed its review of WCOBRA/ TRAC as a best estimate LOCA code for the two loop plant baffle load application, and the SE is being issued separately. The code has been used for breaks ranging from split breaks (less than 60 inch') to double ended guillotine breaks with a Moody coefficient of 1.0. A finite element representation is shown (octant geometry) which .

accounts for the baffle-former-barrel, and baffle to baffle bolts and the former holes for the l bypass flow. A specific bolt configuration is used for each case. The acceptability of that configuration depends on the resulting bolt loadings and allowable stress and deflection limits for the plant-specific application.

The use of best estimate WCOBRA/ TRAC code for the estimation of two-phase loads on the baffle bolts contained in the methodology is acceptable provided that: 1) the limiting baffle bolt

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loading will be determined by analysis for a class of plants and a specific break; 2) the noding to be used in the representation of the loading is demonstrated to be adequate by performing nodalization sensitivity studies or by some other acceptable methodology.

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4 The methodology developed by WOG for Westinghouse plants is proposed for use in determining the acceptability of replacement bolts and bolting distributions and, in part, to j assess existing bolting using appropriate irradiated bolt material properties. The bolt i

replacement activities are subject to the results of plant inspections for baffle bolt cracking ,

degradation. The topical report references the ASME Boiler and Pressure Vessel (B&PV)

Code, Section 111, for materials and formulations for determining allowable stress limits to be used in selecting replacement bolts for normal, upset and faulted conditions. The ASME B&PV l Code does not provide allowable stress limits for irradiated materials. The WCAP report l

l contains proposed irradiated stress allowable limits for use in determining the allowable primary ,

membrane and primary membrane plus primary bending stress limits for baffle bolting under faulted conditions in accordance with the formulation requirements of Appendix F of Section ill of the ASME B&PV Code.

It is not evident that this approach is entirely justifiable. Implicit in the ASME Code stress allowable limit is the fact that the bolting materials should exhibit ductility that provides a strain margin between yield and ultimate strength. For irradiated austenitic stainless steel, the available data indicate a severe loss of ductility, with the yield strength and the ultimate strength in some cases being coincident. It is not clear that the stress allowable limits proposed for irradiated stainless steel account for the lack of ductility of the material.

The inservice inspection (ISI) methods used for inspecting the baffle bolts will have certain sensitivity and probabil"y of detection (POD) characteristics. In using irradiated tensile strength  ;

for ca!culating safety margins, the bolt load-bearing cross-section area should be reduced 1 based upon the sensitivity of the ISI methods and the potential for degraded bolts to escape '

detection by ISI. In addition, since the available data for irradiated stainless steel indicates that the ductility (and presumably the fracture toughness) of irradiated bolts is severely degraded, a fracture mechanics approach would be necessary to demonstrate that the degraded bolts exhibit adequate toughness with a postulated flaw size undetected by ISI.

The report indicates these proposed stress allowable limits are based on yield strength and ultimate strength values obtained from irradiated bolting materials. The report indicates that the {

selected values are conservative considering the source irradiated materials had significantly lower fluence levels than bolts in the baffle region. The conservatism is based on experience )

which demonstrates that for certain materials the yield strength and ultimate strength increase with fluence level. The determination of adequate conservatism of bolting material properties I

and characteristics, resides with the licensee in applying appropriate stress and deflection limits j to the baffle assembly under faulted conditions. This determination should include 1 consideration of the subject plant's historical operating conditions and current licensing bases. l The determination should be based on conservative yield strength and ultimate strength values representative of the plant's existing bolting material properties.

Although the methodology provides systematic guidance for an acceptable logical analytical approach for determining baffle bolt loading under faulted conditions, it har not identified any requirements for a baffle bolt post-replacement monitoring program. The implementation of an

, inspection program is appropriate for baffle bolt cracking aging management. The inspection l program must be capable of detecting baffle bolt cracking and degradation consistent with the 1

I

l l 5 results of the plant's bolt inspection and replacement activities and applicable industry experience. Further, the inspection program should implement a root-cause evaluation of any replacement baffle bolt degradation discovery.

CONCLUSION The use of the WCAP-15029 methodology guidance is acceptable in accordance with the following limitations:

{

1.

The bolt loading should be determined by analysis for a class of plants and a specific break;

2. The noding to be used in the representation of the loading is demonstrated to be l adequate by performing nodalization sensitivity studies or by some other acceptable

methodology; l

! 3. The methodology should not be used to assess existing bolting without demonstration of adequate conservatism in projected bolting material properties (i.e., yield and ultimate strength) to ensure that sufficient ductility is present in existing irradiated stainless steel l

bolting materials;

4. The use of the methodology for existing irradiated stainless steel bolting should account l for limitations in available ISI methods with regard to the probability of detection characteristics; Finally, in consideration of the WOG assessment and conclusion that the baffle bolt issue is not i an immediate safety concern and that it is appropriate to treat baffle former bolt degradation as an aging management issue, subsequent to replacement of baffle bolts, licensees are expected j to develop an appropriate inspection monitoring and aging management program for baffle  ;

bolting. I i

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6 REFERENCES l

1. Westinghouse Owners Group Letter, dated June 19,1998, " Baffle Barrel Bolt Program- l Response to April 9,1998, NRC Letter (MUHP5135)," by Louis F. Liberatori, Jr.

l 2. NRC letter, dated October 1,1998, " Safety Evaluation Related to Topical l

Report WCAP-14748/49-Justification for increasing Break Opening Times in Westinghouse PWRs," to Lou Liberatori, Chairman, WOG Steering Committee l 3. WCAP-15029," Westinghouse Methodology for Evaluating the Acceptability of Baffle-i Former-Barrel Bolting Distributions Under Faulted Load Conditions" by P. E. Schwirian, et al., Westinghouse Electric Company,1998 l

4. WCAP-9401PA, " Verification Testing and Analysis of 17x17 Optimized Fuel Assembly,"

l by M. D. Beaumont, et al., Westinghouse Electric Company, August 1981

5. WCAP-8302," SATAN-IV Program: Comprehensive Space Time Analysis of Loss of l Coolant," by F.M. Bordelon, et al., Westinghouse Electric Company

6. WCAP-12945P, Volume 4, " Code Qualification Document for Best Estimate LOCA Analysis Volume IV: Assessment of Uncertainty," by S. B. Bajorek, et al., Westinghouse l Electric Company,1993 l l

7. WCAP-14449P " Application of Best Estimate Large Break LOCA Methodology to i

Westinghouse PWRs with Upper Plenum injection," by S. l. Dederer, et al.,

Westinghouse Electric Corporation,1993 l Principal Contributor: F. Grubelich, NRRIEMEB (301)415-2784 l Date: November 10, 1998 l

Westinghouse Owners Group Project No. 694 cc:

Mr. Nicholas Liparulo, Manager Equipment Design and Regulatory Engineering Westinghouse Electric Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director Regulatory Affairs Westinghouse Electric Corporation 11921 Rockville Pike Suite 107 -

Rockville, MD 20852 Mr. Hank Sepp, Manager Regulatory and Licensing Engineering Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355