ML060380677
| ML060380677 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/09/2004 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| GL-88-020, TAC MC5729 CDQ1 074 2004 0160, Rev 0 | |
| Download: ML060380677 (17) | |
Text
ENCLOSURE 5 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS TVA CALCULATION CDQ1 074 2004 0160,"HCLPF CALCULATIONS OF RHR HEAT EXCHANGER ANCHORAGE FOR SEISMIC IPEEE PROGRAM," REV. 0, JUNE 9, 2004 (SEE ATTACHED) r.. .Imomw QA Record TVAN CALCULATION COVERSHEET/CCRIS UPDATE Pago 1 of 2 Page I REV 0 EDMSIRIMS NO. EDMS TYPE: EDMS ACCESSION NO (N/A for REV. 0)W 78 0406 24 0 1 2 Caiouiations(nuclear)
NA CaloTitle:
HCLPF CALCULATIONS OF RHR HEAT EXCHANGER ANCHORAGE FOR SEISMIC IPEEE PROGRAM CALC ID TYEe OG PLAN SRANCH = NUMBER CUR REV NEW BEV CURRENT CN NUC BFN CEB REVISION APPLICABILITY NEW CN NUC BFN CEB CDQ1 074 2004 0160 000 Enltre pagc s No CCRIS Changes L ACTION NEW l DELETE O SUPERSEDE 0 CCRIS UPDATE ONLY O (For calc revision, CCRIS REVISION C3 RENAME a DUPLICATE 0 (Verifier Approval Signatures Not been reviewed and no Required)
CCRIS changes required)_ .U_ .,.. .UNITS SYSTEMS U!NIXS 001 074 WtA WN.EDC.NA APPLICABLE DESIGN DOCUMENT.
.CLASSIFICATION
/A IN/A E UUALI I Y tSA1'kTY HELAUYQ I jJJ UNtH S=L i'bUIAL6 tMIZUUtmtNT§ I jtz5(Um OUTF~'U IAI FE~E REAEI (if yes. OR -yes) AS U N T I AND/OR LIMITING CONDITIONS?
I 7fAQCHMIENT yes D9 o i yes No LI Yes No D4I YesI L1No~ D4 Yes I No~ I Yesf nNo R PREPARER ID PREPARER PHONE NO PREPARING ORO (BRANCH) VERIFICATION IETHOD NEW METHOD OF ANALYSIS Farld Elsabee (256) 729-7000 X18023 CEB Design Review 0 Yes E No PREPARER SI TU DATE CHECKER SIONURE DATE Isabee 6/2404 S.J. Eder 6/25104 VERIF RSIGNATURE
-DATE APPFIpVALS______
DATE 0it J.O. Dizon 6/2@04 T,, STAY EMEFNT OF PROBLEWMSTBSTa This calculation docunents the High Confidence Low Probability of Failure (HCLPF) values for the RHR Heat Exchanger anchorage at BFN Unit 1.MICROFICHE/EFICHE Yes U No 93 FICHE NUMIER(S)Cl LOAD INTO EDMS AND DESTROY 0 LOAD INTO EDMS AND RETURN CALCULATION TO CALCULATION LIBRARY. ADDRESS: BFN Calculation Library O LOAD INTO EDMS AND RETURN CALCULATION TO: TVA 40632 [07-20011 PageD I of 2 NEDP-2-1 [07-09-20011 TVAN CALCULATION COVERSHEET/CCRIS UPDATE Page 2 of 2 Paae iA:ALC I I PEI ORG PLANT BRANCH NUMBER REV NEW l CN NUC I BFN CEB CDQ1 074 2004 0160 000 ALTERNATE CALCULATION IDENTIFICATION BLDG ROOM I OIQRD/AZIM fIRM l PrintReE t Yes 01 WA s56' WA Bechtel CATEGORIES F06 KEY NOUNS (A-add, D-delete)AMTON KEY NOUN AI KEY NoUN ANCHORAGE SEISMIC QUAUF SEISMIC DESIGN _CROSS-REFERENCQiS (A-add, C-change, Dclelete)ACTION XREF XREF XREF XREF XREF XREF CODE PtANT BRANCH NUMBER REV PCN BFN CEB CDQ1999 2003 0664 P CN BFN CEB CDQ0 000 94 0339 PCN BFN CEB CDQi 074 2003 2583 I I OCRIS ONLY UPDATES: FdI~lwngasri~quired only when maklngkIeywa~rd/cross reference C2Iu ate amd pagel1of form NEDP-2-1 isnot Included: PREPARER SIGNATURE D3ATE ( CHECKER SIGNATURE I DATE PRPRRPHONE NO, EDM; ACCESSION NO.y IQ--( AO) G 24 ( 1~~_A 408_0-_.]P
_ f EP21 0-~01 rVA 40532 [07-200i]Page 2 of 2 NEDF-2-1 [07-0Z-001)
TVAN CALCULATION RECORD OF REVISION Page 1 of 1 Page II TVAN CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER CDQ1 074 2004 0160 Title HCLPF CALCULATIONS OF RHR HEAT EXCHANGER ANCHORAGE FOR SEISMIC IPEEE PROGRAM Revision DESCRIFTION OF REVISION-- No. I 000 Original Issue This Revision 0, Including TVA calculation cover sheets, contains 16 pages total TV_ 4792.200 _ag tIND-2(20-00 TVA 40709 [12-20001 Page I of I NEDP-2-2 (12-04-20001 TVAN CALCULATION VERIFICATION FORM Pakge I of 1 Page IiI TVAN CALCULATION VERIFICATION FORM Calculation Identifier CDQI 074 2004 0160 Revision 000 Method of verification used: 1. Design Review /2. Alternate Calculation 0 Verifier Date 5/26104 3. Quallfication Test a John 0. Dizon Comments: This revision of the calculation has been reviewed by the Design Review Methodology and has been determined to be technically adequate based on the design Input information contained herein using accepted handbook and/or computer applications, and sound engineering practices and techniques, supplemented by applicable Industry-standard guidelines.
TVA 40538(07-2001]
Page 1 of I NEDP-2-4 [07-09-2001]
TVA 40633 (07-2001]Page I of 1 NEDP-2-4 [07-09-2001]
CALCULATION SHEET HCLPF CALCULATIONS OF RHR HEAT EXCHANGER ANCHORAGE FOR SEISMIC IPEEE PROGRAM Page 1 of 12 CALCULATION SHEET TABLE OF CONTENTS CALCULATION COVER SHEET ...........................
.i CALCULATION RECORD OF REVISION ................................
II DESIGN VERIFICATION (INDEPENDENT REVIEW) FORM ..... iii Calculation Title Page ... ................
.1 Table of Contents ........ 2 1.0 Purpose ..3 2.0 Scope ..4 3.0 Definitions
............................................................
6 4.0 References.............................................................
9 5.0 Anchorage Evaluation
..10 5.1 General Calculation Method. .10 5.2 Anchorage HCLPF Capacity Calculation
.11 6.0 Conclusion
............................................................
12 Page 2 of 12 CALCULATION SHEET Document:
CDQ1 074 2004 0160 IRev.00 Plant BFN IUnit(s):
1
Subject:
HCLPF CALCULATIONS OF RHR HEAT EXCHANGER Prepared:
F. Elsabee Date: 5/24/04 ANCHORAGE FOR SEISMIC IPEEE PROGRAM Checked : S. J. Eder Date: 625/104 1.0 PURPOSE The purpose of this calculation is to determine the High Confidence Low Probability of Failure (HCLPF) capacities for the RHR Heat Exchangers at BFN Unit 1. This calculation Is In support of the seismic portion of the Individual Plant Examination of External Events (IPEEE) program at the BFN Plant which Is required per Generic Letter 88-20 (Reference 4.1). The seismic anchorage evaluations are based on the guiclelines provided in the Seismic Margin Assessment (SMA) report by EPRI (Reference 4.2). The in-plant walkdowns were performed using the Waikdown Instruction WI-BFN-0-CEB-04 (Reference 4.5). The procedures used also nrflect the methods specified In the SQUG Generic Implementation Procedure (GIP, Reference 4.3).The USI A-46 RHR Heat Exchanger (Hx) evaluations determined that the seismic capacity of the Hx's is governed by the capacity of the anchorage of Its support system. Therefore, it follows that the HCLPF capacity is similarly controlled by the same component.
For this reason, It is sufficient to investigate only the anchorage of the Hx support system to determine their HCLPF capacity.Page 3 of 12
-CALCULATION SHEET 2.0 SCOPE This calculation Is applicable to the following Heat Exchangers, to determine the applicable HCLPF values as part of the seismic portion of the IPEEE program: Identification Number Component NmEr 1 -HEX-74-900A RHR/HEAT EXCHANGER 1A 11009 1-HEX-74-900B RHRtHEAT EXCHANGER 1B 11036 1 -HEX-74-900C RRtHREAT EXCHANGER 1C 11017 1-HEX-74-900D RHRiHEAT EXCHANGER 1D 11042 In the above table, SSEL refers to the Safe Shutdown Equipment List (SSEL), as documented in Reference 4.4. The Heat Exchangers are all In the Reactor Building, at Elevation 565'.There are no unverified assumptions, In this calculation.
There are no known special requirements and/or limiting conditions In this calculation.
Page 4 of 12 CALCIULATION SHEET Document CDQ1 074 2004 0160 1Rev.00 Plant BFN IUnits): 1
Subject:
HCLPF CALCULATIONS OF RHR HEAT EXCHANGER Prepared:
F. Elsabee Date: 5/24/04 ANCHORAGE FOR SEISMIC IPEEE PROGRAM Checked: S. J. Eder Date: 5/25/04 3.0 DEFINITIONS
3.1 INDIVIDUAL
PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE)Supplement 4 to Generic Letter 88-20 (Ref. 4.1) requires that each licensee conducts an IPEEE which addresses:
seismic events, Internal fires, high winds, floods and transportation/nearby facility accidents.
3.2 SEISMIC
MARGIN ASSESSMENT (SMA)The SMA, which has been chosen for implementation at the BFN plant, Is an acceptable method used to perform the seismic portion of the IPEEE. The SMA is designed to (a) demonstrate sufficient margin over and beyond the Safe Shutdown Earthquake (SSE) to ensure plant safety, and (b) find any 'weak links" that might limit the plant shutdown capacity to safely withstand a seismic event larger than the design SSE or lead to seismically induced core damage. The SMA for the BFN plants is performed using the EPRI methodology which Is described in Reference 4.42. BFN also elected to combine the USI A-46 and IPEEE walkdowns which were performed In accordance with the SQUG GIP (see below) with enhancements based on the EPRI report (Ref. 4.2). The walk downs were documented In SEWS (see below).3.3 SEISMIC MARGIN EARTHQUAKE (SME)The earthquake level against which the plant Is evaluated while performing the SMA. The SME Is specified In terms of two orthogonal horizontal ground response spectra and one vertical ground response spectnim associated with a specific damping value. This is not a new design earthquake, but one which is used to evaluate existing plants under a SMA.3.4 REVIEW LEVEL EARTHQUAKE (RLE)The RLE is synonymous with the SME. The two terms are often used Interchangeably.
The RLE for the BFN plants Is defined as an earthquake having a response spectrum that matches the median (50% Non Exceedance Probability
-NEP) CR-0098 spectral shape anchored to a peak ground acceleration of 0.3g.Page 5of 12 CALCULATION SHEET 3.5 HIGH CONFIDENCE LOW PROBABILITY OF FAILURE (HCLPF)The level of earthquake below which core damage frequency Is very unlikely, as determined by a seismic margin study- This level of earthquake Is called the high-confidence low-probability-of-failure capacity of the plant. The value Is obtained as the smallest capacity value determined for all components on the affected plant success path (aka shutdown path). From a mathematical perspective of a probability distribution of capacity, as developed in seismic probabilistic risk assessment (PRA) calculations, the HCLPF capacity values are approximately equal to a 95% confidence of not exceeding about a 5% probability of failure. This value is also applied to a SMA to evaluate each component against the SME. The CDFM approach is an acceptable method of determining the HCLPF of a component.
3.6 CONSERVATIVE
DETERMINISTIC FAILURE MARGIN (CDFM)APPROACH The deterministic approach used to calculate a seismic margin capability, for which a HCLPF of the component is demonstrated, Is with the use of a set of pre-established CDFM criteria and procedures.
The CDFM approach Is developed around the following guidelines: (a) The SME is conservatively specif led. (b) The predicted structural and equipment response to the SME Is median centered.
And (c) The assessment of component capacity Is conservative.
[For example, for expansion anchor bolts, the CDFM capacity should be defined at about the 98% exceedanre probability In order to achieve a HCLPF;and thus, the factor of safety against the mean ultimate capacity should be set at a level consistent with about 2% probability of failure.See Appendix 0 of Ref. 4.2 for specific safety factors to be used.)3.7 SCREENED OUT COMPONENTS For these screened out components, It can only be stated that the HCLPF ground motion level exceeds the SME level. Components which are not screened oult require a HCLPF capacity estimate to be determined.
3.8 GENERIC
IMPLEMENTATION PROCEDURE (GIP)The GIP (Ref. 4.3) provides the detailed technical approach, generic procedures and documentation guidance for use by USI A-46 Paie 6 of 12 CALCULATION SHEET licensees to verify the seismic adequacy of mechanical and electrical safe shutdown equipment.
In this regard, the GIP serves as the acceptance criteria and also contains all of the activities necessary for the resolution of USI A-46.3.9 SAFE SHUTDOWN EQUIPMENT LIST (SSEL)This list contains all mechanical and electrical equipment within the selected success paths necessary to bring the plant from a normal operation condition to a safe shutdown condition to ensure safety during and following a Safe Shutdown Earthquake (SSE), as defined In Section 3 of the GIP (Ref. 4.3) as well as Section 3 of the EPRI SMA report (Ref. 4.2). Equipment items In the SSEL require screening verification and walkdown to ensure Its seismic adequacy with respect to its functionality and structural integrity.
3.10 SEISMIC VERIFICATION WNALKDOWN An engineering review to verify the seismic adequacy of the as-installed condition of a specific item of equipment or component to determine Its acceptance cr required further evaluations and/or modifications, based on visual Inspection for predetermined engineering attributes.
Seismic verification walkdowns are to be performed by Seismic Review Teams.3.11 SEISMIC REVIEW TEAM (SRT)Seismic RevIew Team is responsible for the screening verification and walkdown of SSEL equipment Items. A minimum of two Seismic Capability Engineers with structural or mechanical engineering background is required on each team, one of which must be a licensed professional engineer.
SRT may also consist of systems engineers or plant operations personnel, as necessary.
3.12 SEISMIC CAPABILITY ENGINEER (SCE)Seismic Capability Engineers are degreed engineers or equivalents, who have completed a SQUG-developed training course on seismic adequacy verification of nuclear power plant equipment.
These engineers should have at least five years of experience In earthquake engineering applicable to nuclear power plants and in structural or Papg 7 of 12 CALCULATION SHEET Document CDQ1 074 2004 0160 IRev.00 Plant BFN IUnit(a):
1
Subject:
HCLPF CALCULATIONS OF RHR HEAT EXCHANGER Prepad : F. Elsabee Date: 5/24/04 ANCHORAGE FOR SEISMIC IPEEE PROGRAM Checked: S. J. Eder Date: 5/25/04 mechanical engineering.
Refer to Section 2 of the GIP for specific qualifications and training requirements.
3.13 ENGINEERING ATTRIBUTE Engineering attribute Is a predetermined or known seismic vulnerability condition (such as support anchorage, GIP caveats for various types of equipment, etc.) that warrants verification or engineering evaluation during walkdown to ensure Its seismic adequacy as It may potentially affect the seismic performance of an equipment item or component.
3.14 SEISMIC INTERACTION Seismic Interaction is the physical Interaction of any plant structures, features or equipment with a nearby item of safe shutdown equipment caused by relative motions from an earthquake.
Seismic interaction effects which are included within the scope of the GIP are (i)proximity; (ii) structural failure and falling; and (iii) flexibility of attached lines and cables.3.15 SCREENING EVALUATION WORKSHEETS (SEWS)The Screening Evaluation Worksheets (SEWS) provide a convenient summary and checklist for documenting the seismic verification walkdowns performed In accordance to the acceptance criteria provided in the GIP (Ref. 4.3) and EPRI NP-6041-SL (Ref. 4.2).Appropriate SEWS forms for the various equipment classes and other plant features can be found in Appendix G of the GIP and Appendix F of EPRI NP-6041-SL.
3.16 OUTLIER An outlier Is an Item of equipment which does not comply with all of the screening guidelines pnrvided in the respective acceptance criteria.
Outliers Identifled during the seismic verification walkdowns for USI A-46 shall be documented In the Outlier Seismic Verification Sheet (OSVS) for further evaluation and resolution (Section 5, GIP).Page 8 of 12 CALCULATION SHEET
4.0 REFERENCES
4.1 USNRC, Generic Letter 88-20, Supplement No. 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," Final, April 1991.4.2 Electric Power Research Institute (EPRI) Report NP-6041 -SL, UA Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1)," August 1991.4.3 "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Revision 2A.4.4 CDQ1 -999-2003-0654, "Composite Safe Shutdown Equipment List (SSEL) for USI A-46 and Seismic IPEEE Programs -Browns Ferry Nuclear Plant, Unit 1," Revision 02.4.5 Wi-BFN-0-CEB-04, "Seismic Verification Walkdown Instruction for USI A-46 and Seismic IPEEE Programs," Revision 0.4.6 CDQO-000-940339, "Calculation of Basic Parameters for A46 and Individual Plant Examination of External Events (IPEEE) Seismic Program," Revision R01.4.7 CDQ1 -074-2003-2583, "Outlier Evaluation of RHR Heat Exchangers for USI A46 Resolution," Revision 0.Page 9 of 12_
CALCULATION SHEET Document:
CDQ1 074 2004 0160 Rev.00 Plant BFN Units): 1
Subject:
HCLPF CALCULATIONS OF RHR HEAT EXCHANGER Prepared F. E sabee Date: 5/24/04 ANCHORAGE FOR SEISMIC IPEEE PROGRAM Checked: S. J. Eder Date: 5/25/04 5.0 Anchorage Evaluation
5.1 General
Calculation Method The general method used for determining the anchorage HCLPF capacities are based on the guidelines provided In EPRI report NP-6041 -SL (Reference 4.2).The stress to allowable ratios of the anchorage, previously calculated for the components In support of the resolution of USI A-46, are normally used as the basis for the HCLPF capacity determination.
The ratios are modified to reflect the higher level of the ALE vs the SSE level used for the A-46 review. The conservative scaling factor of 1.88 developed In Reference 4.6 Is first used to scale up the A-46 stress to allowable ratio. Note that when this scaling factor Is applied to the previous A-46 calculations, the A-46 SSE values are increased by 1.88/1.25
= 1.504 only, since the HCLPF calculations do not need the 1.25 factor used In the A46 calculations to account for the median centered curves being used at BFN plants. Furthermore, the IPEEE scaling factor is applied only to the SSE values and care is to be taken when the controlling stress ratio Is based on the combined effect of SSE and DW where DW subtracts from the SSE effects. For such situations, one conservative approach is to neglect the DW effect which reduces the SSE loads when the IFPEEE scaling is performed.
Consider the following:
A-46 calculation contains:
Tsse/Tall
-T[,N/Tall
< 1 When scaling for IPEEE, use: 1 .504*TsselTall
< 1 or 1 .504Tsse/Tall
-TDw/Tall < 1 Whichever Is simpler to implement If the resulting stress ratio, after scaling the SSE effects, remains below 1.0, the component is screened out and the HCLF'F level Is greater than 0.3g. When the stress ratio exceeds 1.0, a more detailed calculation of the HCLPF capacity level Is needed.The detailed calculation will either use a more refined value for the RLE scaling factor or will reduce some of the conservatism which may have been used in the A-46 calculations while still meeting the requirements of References 4.2 and 4.3. If the stress ratio cannot be kept below 1.0, a new HCLPF capacity value (below the 0.3g level) will be calculated for the component of concern.Pago 10 of 12 CALCULATION SHEET 5.2 Anchoraoe HCLPF Capacltv Calculation Per Reference 4.7, the controlling component is the 3/4" CIP headed stud for the embedded base plate at point A of Figure 8 of the reference.
The worse stress/load ratio to allowable for the stud was calculkLted as 0.68 In tension for Hx 1 C. Increasing that stress ratio by 1.504, we get: 0.68*1.504=1.02 which Is slightly larger than 1.0.Note however, that this value is conservative because:* The DW effect was always taken as additive to the seismic effect for both the Hx Inertial loads and the attached nozzle loads.* The directional cosines for the RHRSW nozzle loads were always conservatively taken as positive and their signs were neglected.
- The scaling factor used to reduce the nozzle loads from 0.5% to 5%damped response spectra, was conservatively taken as the largest value of 0.81 from the estimated range of 0.21 to 0.81.As a result of the above conservatism, the stress ratio for the HCLPF capacity calculation will be reduced to below 1.0. Thus, it is concluded that based on the seismic A-46 calculation, the HCLPF capacity is determined to be greater than 0.3g.Page 11 of 12 CALCULATION SHEET 6.0 Conclusion Based on the above evaluation, it is concluded that the HCLPF capacity of the following four BFN Unit 1 RHR Heat Exchangers Is greater than 0.3g.Identification Component Nmer HOLPF Number NOumbOerT CaNacity 1 -HEX-74-900A RHR/HEAT EXCHANGER 1 A 11009 > 0.3g 1-HEX-74-900 RHR/HEAT.EXCHANGER 10_ 1101 > 0.3g 1-HEX-74-900C RHR/HEAT EXCHANGER 1D 11042 > 0.3g 1-HEX-74-9OOD RHRIHEAT EXCHANGER 1 D 11042 > 0.39 Pago 12 of 12