ML20147E415
| ML20147E415 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/28/1986 |
| From: | Beason J TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18032A736 | List: |
| References | |
| BFEP-PI-85-01, BFEP-PI-85-1, IEB-79-02, IEB-79-12, IEB-79-2, NUDOCS 8803070059 | |
| Download: ML20147E415 (105) | |
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9 hY TENNESSEE V ALLEY'"AUTHORITY DhrWon of Nucles:' Engheering g dhRecord BFEF-PI 85-01 BROWS FERRY NUCLEAR Pl. ANT IMPLE.W N!ATION OF NRC-0IE BULLETINS 79-02/79-14 TITLE:
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COORDlMATIOM LOG Document No.: BIU"PI O*C1 D.:FLDEl.6"ATION OF NRC-OIE TULLE"' INS 79-02/79 +
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REVISION LOG IMPLEMENTATION OF NRC-01E BULLETINS 79-02/79-14 BFEP-PI 85-01 Title : FOR BROWNS FERRY Nrs., LEAR PLANT 4,[*,, o
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DESCRIPTION OF REVISION 1
- 1) To agree with organizational changes (e.g. , name changes),
- 2) to accurately reflect program changes resulting from dedicated effort for 79-02/79-14, and
- 3) to provide greater detail for instructions and responsibilities.
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IMPLEMENTATION OF NRC-OIE BULLETI'NS 79-02/79-14 BTEP-PI 85 TOR BROWNS FERRY NUCLEAR PLANT-PROJECT INSTRUCTION TABLE OF CONTENTS P_agj;
. 1 1.0 PURPOSE ................................................. ...
1 2.0 SCOPE ..........................................................
1 3.0 RESPONSIBILITIES ...............................................
3.1 DNE Program Coordinator - Browns Ferry Analysis and Support Design Project .............................. I 3.2 DNE Site Inspection Coordinator - Browns Ferry Analysis 1
and Support Design Project ..............................
3.3 Browns Ferry Analysis and Support Design Project 2
Piping and/or Support Analysis Section(s) ...............
3.4 Nuclear Engineering Branch (NEB) .......................... 2 2
4.0 PEAS E I INSTRU CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2 4.1 Inspections ...............................................
4.2 Review of Inspections by Piping Analysis and/or Support Design Section(s) ............................... 3 4.3 Review and Documentation Methods For, Data Packages ........ 6 7
5.0 PHAS E I I INSTRU CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5.1 Phase II (Code Compliance) ................................ 7 5.2 Piping Analysis and Support Design ........................ 7 8
6.0 REFERENCES
' .1, Discrepancy Resolution Seismic Analyst's Input .2 Criteria For Seismic Evaluation and Categorization .3, Corrective Action Requirements For NCR BTNCEB8103 Attach =ent 6.4, Handling of Design Data Packages Guideline A ~
Guideline B -
Guideline C .5, BFN Interim Anchorage Evaluation Criteria 11 D E1 - 1023k
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IMPLEMENTATION OF NRC-0IE BULLETI'NS 79-02/79-14 !'
BTEP-PI 85-01 I' FOR BROWNS FERRY NUCLEAR PLANT 1.0 PURPOSE This instruction describes the Division of Nuclear Engineering's (DNE) procedural requirements for the review and evaluation of inspection data generated by "Inspection and Repair Program for Verifying Correct Installation of Self-Drilling-Type Concrete Anchors" (BF SKMI-5.1-A) and "Instructions for the Implementation of NRC IE Bulletin 79-14" (BF KMI-99) and the identification and processing of potential safety concerns for responding to the Nuclear Regulatory Co= mission (NRC) OIE Bulletins 79-02/79-14 for Browns Ferry Nuclear Plant (BFN). The inspection data will be evaluated in such a way and to an extent to ensure that the inspection of all piping, supports, and components in and affecting the defined program boundaries is co=plete and reasonable. The overall program plan is described in BFEP-PI 86-05. ,
This instruction will describe a two-phase program for satisfying the concerns of bulletins 79-02 and 79-14 Phase I consists of initial inspections with the data being evaluated for interim acceptance or continued operation. Phase II will utilize verified data which will be evaluated to ensure that the as-configured data is reconciled to
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the code co=pliance requirements of support design and piping analysis.
SEP 81-02R0 described the program to be used from the beginning of DNE's efforts to resolve Bulletin 79-14 until May 1985.
Because the responsibilities, organization, and procedure for the periods before and af ter May 1985 were so different, SEP 81-02RO was revised by PI 85-01R0 to describe the procedure beginning May 1985.
2.0 SCOPE This procedure applies to all BFN safety-related piping (TVA Class 1 Seismic) 2-1/2 inches in dia=eter and greater and to all piping ,
dynamically analyzed by co=puter regardless of size. l i
3.0 RESPONSIBILITIES l l
3.1 DNE Progra= Coordinator - Browns Ferry Analysis and Sucoort I Design Proiect f
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3.1.1 Coordinates implementation of the DNE portion of the BFN program to meet the requirements of NRC OIE Bulletins 79-02/.79-14 3.1.2 Coordinates the evaluation and resolution of any deviations, i.e., differences between design drawings and the as-built configuration. (See Section 4.2 and 5.2.)
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IMPLEMEhTATION OF NRC-0IE BULLEbNS 79-02/79-14 FOR BROWNS FEDDY NUCLEAR PLANT BTEP-PI 85-01 3.2 DNE Site Coordinator - Browns Ferry Analysis and Support Design Proiect 3.2.1 Schedules inspection activities to bring about a complete and timely inspection of the physical areas covered by Bulletins 79-02 and 79-14 3.2.2 Maintains inspection data.
3.2.3 The Site Coordinator has the option of evaluating deviations per BTEP-PI 85-01 prior to sending inspection data to the Program Coordinator.
3.2.4 For Pha e I, ensures that a copy of the inspection data is transmitted to the DNE Program Coordinator or his representative (s) for distribution to the inspection data reviewer (s).
For Phase II, ensures that data is transmitted to those performing the data review, design, or analysis. (For
. definitions of Phases I and II, see BTEP-PI 86-05.)
3.3 Browns Ferry Analysis and Sucoort Design Proiect Piping Analysis and/or Supoort Design Section(s) 3.3.1 Perfore walkdowns of piping and supports for which the l analysis and support design will be based.
3.3.2 Evaluate deviations 3.3.3 Review the data packages to determine if additional inspection infor=ation is needed.
3.3.4 For Phase II, perform final piping analysis and/or j support design and issue design output documents as ,
necessary in accordance with the "Design Criteria for l As-Built Piping Systems, BFN-50-D707" and DNE Detailed i Design Criteria No. B W-50-724 Browns Ferry Nuclear i Plant, "Class 1 Seis=ic Pipe Support Design." !
3.3.5 Initiate and approve design changes in accordance with plant modification procedures to reconcile as-built condition of plant to licensing commitments. ,
l 3.3.6 Perform constructability walkdowns of piping and supports prior to drawing issue.
3.4 Nuclear Engineering Branch (NEB) 3.4.1 Performs safety evaluations. (See Instruction on ,
Evaluation of Inspection, Section 4.2.5) l l
3.4.2 Documents and retains records of all safety evaluations per NEP-1.3.
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s IMPLEME? CATION OF NRC-0IE BULLETINS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLArc BTEP-PI 85-01 3.4.3 Identifies safety-related (TVA Class 1 Seismic) portions of all systems at Browns Ferry Nuclear Plant.
4.0 PHASE I INSTRUCTION 4.1 Inspections 4.1.1 The piping analysis and/or support design section(s)
Initiate inspections when necessary to ensure that the inspection of all piping, supports, and components in and affecting the defined program boundaries is complete. .
4.1.2 The inspections are to be performed in accordanc'e with references 6.2 and 6.3.
4.1.3 The flow diagrams showing full extent of the inspection boundaries shall be marked up by comparing with inspection data to ensure that the inspection is complete.
4.2 Evaluation of Inspections by Piping Analysis and/or Support Design'Section(s) 4.2.1 Evaluate all pipe / support inspection data against the ciesign output docu=ents for any discrepancies.
Evaluators to indicate on appropriate records (e.g., data sheets, attached calculations, etc.) acceptance of deviations as noted on as-built sketches.
4.2.2 All piping and/or support deviations from design and/or differences between design output documents which are !
judged to be not adecuate (present potential safety concerns) are documented on Discrepancy Resolution forms (Attachment 6.1). A discrepancy problem number is obtained from the DNE Program Coordinator. The person initiating the discrepancy form shall identify the discrepancy problem number, the drawing number, the unit, the building, the system, and the description of the discrepancy on the form. He shall document this by providing his dated signature in the Prepared By/Date locations. An evaluation is then performed on each discrepancy problem by the reviewing engineer from the piping analysis and/or support design section. l Attachment 6.2 defines the criteria to be used in ..
I determining the appropriate discrepancy evaluation categories. At any time it is the responsibility of anyone who recognices a discrepancy problem to initiate a r
discrepancy resolution im=ediately. All discrepancies are microfilmed by a discrepancy number.
- a. Configuration A Discrepancy Resolution Seismic Analyst's Input sheet is initiated for everv configuration drawing for analyst review. An experienced analyst documents DNE1 - 1023h
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IMPLEMEhTATIONOFNRC-OIEBULLETNS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLANT BTEP-PI 85-01 his evaluation of the overall seismic adequacy of the as-built piping configuration and support scheme by checking the appropriate category under the Seismic Analysis section of the input sheet and signing the document. ,
- b. Support Experienced structural and/or support designers evaluate the as-built condition to determine by engineering judgment ( ar,d technical justification for the judgment) the structural adequacy to meet design intent. If calculations are required to provide the technical justification for the judgment, the calculations will be performed using original loads, loads estimated by experienced analysts, or loads determined by referring to general spacing requirements. If a deviation is determined to be a discrepancy, the discrepancy is reported on a Discrepancy Resolution Seismic Analyst's Input sheet for a seismic analysis evaluation.
- c. Concrete Anchors
'All anchors under the 79-02 program are inspected, and the data is recorded on 79-02 data sheets.
NRC-OIE Bulletin 79-02 is concerned with the adequate design and installation of concrete anchors. Specifically, the Bulletin requires that a factor of safety of 5 (for shell type anchor bolts) exist in order to be acceptable on a long term basis. However, for short term (interim basis) the bulletin allews the factor of safety compared to ultimate to be reduced to 2 providing the licensee has developed a program of restoration to at least the Bulletin factor of safety.
i In order to satisfy the immediate Bulletin concerns, TVA in !
1979 assigned to the site on a rotational basis civil ,
engineers experienced in anchorage and support designs for !
piping systems, cable trays, etc. These evaluators were ;,
known as EN DES on-site engineers. Since the Bulletin l' required verification by inspection that bolts were properly installed and were of the specified size and type, the on-site engineers evaluated the inspection parameters which would affect the Bulletin concerns (e.g. , embedment depth, '
thread engage =ent, full shell expansien, bolt diameter, etc.). The on-site engineers were given acceptance criteria to aid them in their evaluations.
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IMPLEMENTATION OF NRC-OIE BULLETINS 79-02/79-14 FOR BROWNS TERRY NUCLEAR PLA?C BTEP-PI 85-01 The criteria came from sources ranging from published specifications to testing program results (e.g., Singleton Materials Laboratory tests) with ultimate capacities being those found in General Construction Specification G-32.
Examples of the kinds'of acceptance criteria include:
- 1. Plug Depth Criteria
- 2. Edge Distance Criteria
- 3. Thread Engagement Criteria 4 Anchor Capacity Criteria
- 5. Expected Load for Cut-off SSDs.
This acceptance criteria is presented in Attachment 6.5.
When deviations could not be evaluated using a direct comoarison to the acceptance criteria, the on-site engineer would perform calculations where possible either on the 79-02 inspection data sheet or as an attachment to assure immediately that at least a factor of safety of 2 existed.
If a factor of safety of 2 was not met, the EN DES office
... located in Knoxville performed calculations to determine the
'effect on the piping through the 79-14 program. In those cases where the evaluation was beyond the capability of the on-site engineer, the situation was referred to EN DES (e.g., CEB, Thermal Design Projects (TDP), etc.) for disposition.
Anchorages evaluated by EN DES with factors of safety less i than two also became 79-14 discrepancies which required an i evaluation for that support for its effect on the piping system for continued operation with respect to plant technical specifications. ,
TDP (now BTEP) received copies of the inspection data sheets for their review and incorporation into the 79-14 package for subsequent final resolution (i.e..-code co=pliance). ,
Anchorages with factors of safety of two but less than the -
required factor of safety were required to be redesigned and ,
modified, as required, under the Phase II portion of the program to meet code co=pliance (i.e., a factor of safety of a for wedge bolts and a factor of safety of 5 for self-drilling type anchors).
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Since May 1985, 79-02 deviations are repaired when possible to the original design intent by using the 48W1241 series drawings. If repairs are not possible, variances to the :
48W1241 series drawings are allowed provided justification / !
calculations is provided by DNE and included or attached to ,
data sheet 2 of BF-SMMI-5.1-a. Variances will be uniquely identified on the data sheet 2. Under the Phase II portion '
of the progra=, all anchorages will be evaluated to assure co=pliance with the long-term provisions of NRC-OIE Bulletin 79-02.
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IMPLEMEhiAT10N OF NRC-08E BULLET 8NS 79-02/79-14 FOR BROWS FERRY NUCLEAR PLAhi BFEP-PI 85-01 4.2.3 Time limits for discrepancy evaluations are:
- a. A 48-hour time limit for the potential effect of a discrepancy. This limit applies from the time the discrepancy is identified to the evaluating section until the safety evaluation, if required, is performed. Therefore, it is necessary for an engineering judgment to be made to determine if the system is "unaffected," "still qualified" but the original analysis is no longer valid, or "no longer qualified." The technical justification for this judgment shall be placed in the space provided under "Justification." The judgment and justification shall be documented by dated signatures of the Engineer / Analyst and checker. If "no longer qualified" is checked.-a Seismic Integrity Evaluation must be made to determine if there is a minimal potential for loss of pressure boundary or a definite potential for loss of pressure boundary / containment seal. This shall be documented by dated two-party signature.
- b. 'If reanalysis is required to confirm engineering judgment, it must be completed within 30 days of when the discrepancy is identified in the evaluating section. This shall be docu=ented by dated signatures of the analyst and checker. If "no longer qualified" is checked, a Seismic Integrity Evaluation must be made to determine if there is a minimal potential for loss of pressure boundary or a definite 1 potential for loss of pressure boundary / containment l seal. This shall be documented by dated two-party signature.
- c. Units in a cold shutdown condition are exempt from l the requirements of Sections 4.3.2.r. and 4.3.2.b with !i the exception of those systems whose continued 1 operation is necessary to maintain cold shutdown as ;'
defined by plant technical specifications. For those i' units exempt of requirements from Sections NEP-9.1 must4.3y'm.a be et. and 4.3.2.b. the 4.2.4 Instead of writing individual significant condition reports on potential conditions adverse to quality, the piping and/or support analysts handle these items under the "generic" nonconformance report (NCR) for B W Bulletin 79-14 deficiences, BFNCEB8103.
4.2.5 If the results of the evaluation indicate a definite i potential for loss of pressure boundary or containment )
seal, the discrepancy form is given to the DNE Program Coordinator. The DNE Program Coordinator transmits a copy of the evaluation to NEB requesting a safety DNE1 - 1023k l l
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3MPLEMEh7AT20N OF NRC-08E BULLETINS 79-02/79-14 I FOR BROWNS FERRY NUCLEAR PLAhi f BTEP-PI 85-01 i evaluation. The coordinator will provide the results of the safety evaluation to those responsible for notifying the site director or his representatives for I applicability to technical specifications. The coordinator then forwards the original discrepancy resolution sheet to the support design or piping analysis section which is in possession of the 79-14 design data package so that the original may be maintained in the ,
package. The tracking of disc: epancies will be handled l by a computerired tracking program and will be j microfilmed using Branch / Project Identifier 7914DISCRLOG.
4.2.6 If the results of the evaluation indicate a minimal potential for loss of pressure boundary, the discrepancy I form is given to the DNE Program Coordinator. If the evaluation indicates the need for a phyrical modification, the coordinator will provide the recomended action to those responsible for notifying the plant manager or his representatives. The coordinator then forwards the original discrepancy resolution sheet to the support design or piping analysis section which is in possession of the 79-14 design data package so that the original cay be maintained in the package. The tracking of discrepancies will be handled by a computerized tracking program and will be microfilmed.
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4.3 Review and Documentation Methods for Phast I Data Packages
- The original inspection data will b'emaintained in inspection data packages. Copies of the inspection data will be used in the design process and will be maintained in design data
. packages. The design data packages will be microfilmed.
4.3.1 The DNE Program Coordinator:
- a. Receives the inspection data package (s) from the DNE Site Coordinator, records in a log the receipt of the package (s), assigns problem nu=bers for data sheets, then transmits the package (s) to the piping analysis and/or support design section(s) for handling. (See Guideline B, Attachment 6.4.)
- b. The piping analysis and/or' support design section(s) ;
determines if inspection infor=ation is co=plete. If not, the section(s) coordinates with the support' l designer and the DNE Coordinator for additional j in'for=a t ion.
- c. Receives data packages from the piping analysis and/or support design section(s) for reinspection, trans=its thesc' packages to the DNE Site Coordinator, and documents these trans=ittals in a log.
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IMPLEMENTATION OF NRC-OIE BULLETI' N S 79-02/7S -14 FOR BROWNS FERRY NUCLEAR PLAIC BTEP-PI 85-01 !
4.3.2 The piping analysis and/or support design section(s) assemble and review the design data packages in accordance with the Guidelines of Attachment 6.4 'Ihe section(s) also evaluates deviations from and deficiencies of design in accordance with Section 4.2.
5.0 PHASE II INSTRUCTION 5.1 Phase II (Code Compliance) 5.1.1 Any additional data required for piping analysis and/or support design will be obtained in accordance with BTEP-PI 86-06.
5.2 Piping Analysis and Support Design 5.2.1 All data identified during Phases I and II will be considered in the final piping analysis in accordance with BFN-50-D707. The resulting support loads and Phase I and II data will be considered by the support design
,.section(s) for support evaluation to be performed in accordance with BFN-50-724 After completion of the support evaluation for any given analysis problem (math model), the' support design section(s) will submit to the analysis section a su= mary of the effects of the loads on the existing support scheme (i.e., support modifications, supports deleted, supports added, and no support change). The analysis section will evaluate the condition of the existing support scheme for each analysis problem as reported by the support design section for a potential discrepancy on that given analysis proble=.
5.2.2 All efforts to reconcile 79-02/79-14 will be documented )
on design output documents per NEP-5.1.
6.O REFERDiCES i l
6.1 BTEP-PI 86-05, NRC-OIE Bulletin 79-02/79-14 Progra= Document for !
Browns Ferry Nuclear Plant. ,
6.2 Special Mechanical Maintenance Instruction 5.1-A, Inspection and Repair Program for Verifying Correct Installation of ~
Self-Drilling-Type Concrete Anchors.
6.3 Hechanical Maintenance Instruction 99, Instructions for the Implementatic;n of NRC-OIE Bulletin 79-14 6.4 General Construction Specification G-32. Bolt Anchors Set in Hardened Contrete.
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IMPLEMEhiATION OF NRC-OIE BULLETI S 79-02/79-14 FOR BROWNS TERRY NUCLEAR PLANT BTEP-?I 85-01 r
6.5 BFN-50-D707, Detailed Design Criteria for Analysis of As-Built Piping Systems.
6.6 Detailed Design Criteria No. BIN-50-724 - Browns Ferry Nuclear Plant, "Class 1 Seismic Pipe Support Design."
6.7 BFEP-PI 86-06, Implementation,of NRC-0IE Bulletin 79-14 Phase II Verification for Browns Ferry Nuclear Plant.
6.8 Division of Nuclear Engineering's Design Output Procedure NEP-5.1.
6.9 Division of Nuclear Engineering's Records Control Procedure NEP-1.3.
6.10 Division of Nuclear Engineering's Corrective Action Procedure NEP-9.1.
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ZMPLEMENTAT80N OF NRC-OZE BULLET 8NS 79-02?79-14 TOR BROWS FERRY NUCLEAR PLAhi BTEP-PI 85-01 N TACHMEhT 6.1 )
DISCREPANCY RESOLUTION SEISMIC ANALYST'S INPUT Problem No. Drawing No.
BFN - Unit Building System Description of Discrepancy j Prepared by Date !
SEISMIC ANALYSIS l Basis for Judgment Engineering Rigorous Analysis Is 30-day analysis 1 Judgment (if required) required to confirm 1 engineering judgment?
Unaffected Yes No Still qualified but Analyst Date
-original analysis Checker Date is no longer valid No longer qualified.
(requires Seismic Integrity Evaluation below) l l
Engr / Analyst Date Analyst Date l Checker Date Checker Date Justification SEISHIC IhTEGRI'"Y EVALUATION (if no longer qualified)
(Designate evaluation category per attachment 6.3 in the appropriate i location below)
- 1. Minical potential for loss of pressure boundary.
Discrepancy is OK for temporary acceptance.
Justification j l
- 2. Definite potential for loss of pressure boundary.
- 3. Definite potential for loss of containment seal. ~
- 4. Other NOTE: Items 2, 3, and, 4 require a Safety Evaluation fro = DNE NEB.
Reco= ended Action Analyst Date Checker Date Received by DNE Coordir.ator Date
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IMPLEMENTATION OF NRC-08E BULLETINS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLAhT BFEP-PI 85-01 A'ITACHMDC 6.2 CRITERIA FOR SEISMIC EVALUATION AND CATEGORIZATION BFN 79-14 DISCREPANCIES DISCREPANCY RESOLUTION SEISMIC ANALYST'S INPUT ,
SEISMIC ANALYSIS & MALYSIS SEISMIC IhTEGRITY SEISMIC EVALUATION CRITERIA
, EVALUATION CLASSIFICATION CATEGORY UNAFFECTED (NO MODIFICATIONS) i 4 I SH ESS < CODE ALLOW l STILL QUALIFIED,BUT ANALYSIS NO LONGER 4 STRESS < CODE ALLOW VALID -
3 S MESS > CODE ALLOW TEMP APPVD, CHANGE ON NORMAL SCHEDULE IlCD?SITIED PRIPARY
- 1. HINIMUM POTDCIAL PRIORITY 2 S M ESS < YIELD (FOR LOSS OF PRESSURE BOUND.) 2 S EESS > CODE ALLOW
. TEMP APPVD, CHANGE IhTDiSIFIED PRIMARY AS SOON AS FEASIBLE STRESS < YIELD PRIORITY I SEISMIC IhTEGRITY 2. DEFINITE POTDCIAL SE ESS > CODE ALLOW EVALUATION (FOR LOSS OF 1 (IT NO PRESSURE BOUND.) IhTENSIFIED PRIMARY LONGER UNACCEPTABLE S3ESS > YIELD QUALIFIED)
- 3. DEFINI'"E POTDCIAL STRESS > CODE ALLOW (FOR LOSS OF 1 C0!CAI!NDO SEAL) I!CDiSIFIED PRIMA %Y UNACCEPTABLE S3 ESS > YIELD S3ESS > CODE ILLOW
- 4. O'DIER I IlCD?SIED PRIMARY UNACCEMABLE SU ESS > YIELD l PIPE DOES NOT HAVE ADEQUATE S 3ESS
, YES MARGIN FOR MEU OD OF EVALUATION USED IS 30-DAY ANALYSIS REQUIRED TO TO JUDGE ADEQUACY
' CONFIRM DiGINEERING JUDGMDC?
PIPE RAS ADEQUATE l SEISS MARGIN FOR NO METHOD OF EVALUATION USED t
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IMPLEMDCATION OF NRC-0IE BULLETINS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLAtt BFEP-PI 85-01 NITACHMDC 6.3 CORRECTIVE ACTION REQUIREME'ES FOR NCR BFNCEB8103 BROWNS FERRY NUCLEAR PLA?C This procedure describes in detail the corrective action requirements for NCR BFNCEB8103 as committed to in audit 80-13, No. 12.
In general, all safety-related piping within the scope of Bulletin 79-14 which has been identified by the Nuclear Engineering Branch (NEB) as requiring seismic qualification will be evaluated or reanalyzed for the as-built condition. The initial evaluation or reanalysis may show, in some cases, that the as-built piping and/or supports will not qualify (to code allowables.)
(Torus attached piping is analy:ed according to the design criteria for the torus integrity long-term program, Bni-50-D706, and all other piping is <
analyzed to the requirements of the design criteria for analysis of as-built piping systems, BDi-50-D707.) .
The piping is divided into various analysis problem nu=bers and is analyzed in the as-built configuration as shown on study drawings and/or verification isometrics.
L' hen the as-built piping is reanalyzed, each analysis problem will be evaluated for "Analysis Seismic Evaluation Category" which reflects the outco=e of the analysis of evaluation as follows. (Note: These category designations will also be used by the seismic analyst doing initial evaluations of inspection discrepancies.) ~
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- 1. Unacceptable. Technical specifications govern scheduling of changes. I (The analysis / evaluation indicates that there is a definite potential for loss of pressure boundary, loss of containment seal, or other consequence, such as imprcper slope or a missing valve, which has an unacceptable l effect on plant safety or system function.) The effect I on plant safety is determined by an NEB / Nucle ir Safety Systems (NSA) !
safety review.
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- 2. Temporarily approved - Priority 1. Changes will be made as soon as l feasible. (The analysis / evaluation indicates that changes should be given priority over those indicated in category 3.)
- 3. Te=porarily approved - P-iority 2. Changes will be made on a normal schedule. (The analysis / evaluation indicates that changes may be made on a normal schedule.)
- 4. Field changes not required. (The analysis / evaluation indicates that intensified stressen in the as-built configuration do not exceed code allowable stresses.)
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IMPLEMENTATION OF NRC-OIE BULLENNS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLANT BTEP-PI 85-01 A'ITACHMENT 6.3 (Continued)
Evaluation'of Category.1 indicates that there is a definite _ potential'for an unacceptable consequence. Categories 2 and 3 indicate that there is a minimum potential for an unacceptable consequence. Note.that simplified analysis or evaluation may be used to determine the initial "Analysis Seismic Evaluation Category."
The DNE Program Coordinator will be notified immediately of any initial evaluation or analysis problem categorized as I with a description and' status of the discrepancy. The DNE Program. Coordinator will be notified of any change in status occurring during any subsequent analysis or evaluation.
For any analysis problem which is Category 1, the discrepancy will be
~
forwarded to the Nuclear Engineering Branch / Nuclear Safety Analysis (NEB /NSA) section by the DNE Prc3 ram Coordinator-for a safety evaluation. If NEB /NSA determine that system Junction and/or plant safety are not unacceptably affected, the category will be changed.
Problems in Categories I, 2, or 3 will be analyzed with changes required to satisfy code al?owable stresses.
As these analyses are completed, preliminary piping analysis results will be provided'for infornation and evaluation. Supports will be evaluated and any which should ba ecnSidered as a failed support will be identified.
Considering any addi;ional information obtained from the evaluation of the preliminary analysis results, the analysis problem will be reevaluated, if required, for proper "Analysis Seismic Evaluation Category."
The analysis will be performed for Phase II, and again the "Analysis Seismic Evaluation Category" vill be reevaluated.
- The plant will be notified by ECN of any necessary design changes to reconcile the discrepancy problem.
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Documentation of status and schedules for all problems and discrepancies will !
be provided by the seismic analyst, the support designers, and the plant in accordance with BTEP-PI 86-05.
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IMPLEMENTATION OF NRC-OIE BULLETINS 79-02/79-14 FOR BRO'a'NS FERRY NUCLEAR PLAhT BFEP-PI 85-01 i
ATTACHMEhT 6.4 l.
l, BFEP/CEB HANDLING OF DESIGN DATA PACKAGES GUIDELINE A GUIDELINE B GUIDELINE C i
l DNE1 - 1023k
IMPLEMENTATIONOFNRC-OIEBULLETINS 79-02/79-14 1 FOR BRO'JNS FERRY NUCLEAR PLANT BTEP-PI 85-01 l l
GUIDELINE A )
INCORPORATING 79-02 BULLETIN INSPECTION DATA INTO PHASE I DESIGN DATA PACKAGE
- 1. File all 79-02 anchor inspection data sheets and attachments in the Phase I 79-14 design data par.kage. The 79-02 inspection designations and locations may be found on 79-02 key plan type drawings. (Beginning May 1985, "79-02 Master Copy" location plans were created to identify remaining supports to be inspected.)
- 2. All 79-14 supports utilizing expansion shell anchors are required to have 79-02 inspection. Verify, by reviewing individual support attachment details on Data Sheet 1 inspections, that this has been done. Check to see that all required anchor inspection data sheets have been submitted. Request addit.ional anchor inspection data sheets from DNE Inspection Coordinator, if necessary.
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~~~~lubu6hh ,j ZMPLEMENTAT!ON OF NRC-08E BULLETZNS 79-02/79-14 FOR BROWNS FERRY NUCLEAR PLANT BFEP-PI 85-01 GUIDELINE B GUIDELINE FOR HANDLING PROBLEM NUMBERS ASSIGNED TO 79-14 BULLETIN PHASE I INSPECTION DATA SHEETS 1 THROUGH 4 0F THE DESIGN PACKAGE t
A number shall be assigned to the copy of Phase I field inspection documents (data sheets) generated as a result of NRC OIE Bulletin 79-14 This will provide DNE with the means of tracking these documents. The ,
number shall be unique and will be known as a "problem number." An '
example of a problem number is 080280-03, 08 (month ), 02 (day), 80 (year, used only after June 1980), and 03 (sequential number assigned that date). -
. 1 When a data sheet is submitted from the plant site and received by the I DNE Program Coordinator, it shall be logged into the "Package Log." A l problem number shall be assigned to the. data sheet by the Program l Coordinator or his representative (s), j The coordinator will then pass the data sheet (s) and attachment (s) 1 to the reviewing engineer for 1) evaluation of as-built data, 2) logging l the status in the problem "Su==ary Log," and 3) placing the problem nu=ber on the data sheet (s) and attachment (s) with reviewer's initials ;
and the date.' This shall be done by using a "Verified By" sta=p as shown below and providing the infor=ation indicated.
Any attachment (s) added by the reviewer to a particular data sheet shall have the same problem number as the data sheet placed on the attachment (s) with reviewer's initials and the date. This shall be done by using a "Prepared By" sta=p as shown below and providing the j information indicated. '
Problem No. Problem No.
Verified By Prepared By Date Date I
Il DNE1 - 1023k l
l TVA 10535 f rN nrm- 7-771
IMPLEMEh7AT20N OF NRC-08E BULLETZNS '
79-02/79-14 TOR BROWS FERRY NUCLEAR PLAhT BFEP-PI 85-01 GUIDELINE C Ih"IRODULTION The following requirements are for assembly of the Phase I design data
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packages.
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IMPLEMDEATION OF NRC-01E BULLETZNS 79-02/79-14 FOR BRO'JNS FERRY NUCLEAR PLAhi BTEP-PI 85-01 GUIDELINE C GUIDE FOR ARRANGING THE PHASE I DESIGN PACKAGE IN ORDER OF ASSEMBLY Tab Description
- NA Correspondence NA Table of Contents and Summary [
If any information such as drawings are in a separate envelope, so state under the pertinent section in the Table of Contents and Sum =ary.
1.0 Data Cover Sheet 1.1 Data Sheet 1 Include data sheet 1, discrepancy resolution sheet, data sheet 5, and any other attachments from the plant site.
! 1.2 Data Sheet 2 Include. data sheet 2 and a valve information sheet. Stamp valve information sheet as being prepared by and add the problem number assigned to the data sheet.
1.3 Data Sheet 3 Include data sheet 3, discrepancy resolution sheet, sketches, )
etc. Large drawings may be placed in an accompanying envelope. I 1.4 Data Sheet 4 Include with each data sheet 4, if applicable, a discrepancy resolution sheet.
1.5 Data Sheet 5 Existing data sheet 5 has been included in data sheet 1. .
No additional data sheet 5 will be prepared.
2.0 IE 79-02 Inspection Report t
3.0 Discrepancy Resolutions (copies)
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- Items 4.0 and 5.0 are not tabbed in packages, but are included as part of the "Table of Conten:s and Su==ary" as described later in this Guideline.
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IMPLEME? CATION OF NRC-08E BULLETZNS 79-03/79-14 FOR BROWS FERRY NUCLEAR PLAFC BTEP-PI 85-01 GUIDELINE C (Continued)
Tab Description 6.0 Existing Analysis (may be attached or referenced) 7.0 Physical Piping and Hanger Drawings 8.0 Bills of Material Such as: TVA Bills of Material, Manufacturer's Bills of Material or Information.
9.0 Design Conditions List design conditions or supply drawings (normally the system flow diagram), calculations, etc. which give design conditions.
10.0 Insulation Data Explain status or reference where insulation data can be found.
11.0 Operating Conditions List operating conditions or supply drawings, data, or calculations, etc., which give operating conditions.
12.0 Equipment Data Explain status or reference any equipment data.
13.0 Additional Analysis Information Notes, references, analysis information.
14.0 Study Drawings, Co=posites, Etc.
Any drawings, co=posites, or other items prepared to facilitac?
design shall be tabulated in the Table of Contents and Su==ary and maintained in the drawing envelope.
15.0 Analysis Comnents Tabulated in the Table of Contents and Summary and maintained in the drawing envelope.
16.0 Miscellaneous (if applicable).
Notes in General:
- 1. Any section can be described in the Table of Contents and Su=ary without requiring separate tab and space alloteent in package if no additional information for that section is required.
- 2. Within any section (tab), additional itemi:ing may be used to segregate for improved clarification (exa=ple: 1.3.1, 1.3.2, 7.1, 7.2, etc.). When the additional itemicing is used, the primary tab shall index the additional tabs.
- 3. Each tab should include a brief description.
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IMPLEMENTATION OF NRC-08E BULLETINS 79-02/f9-14 FOR BR0k'NS FEFRY NUCLEAR PLANT BTEP-PI 85-01 GUIDELINE C (Continued)
HANDLING OF DATA SHEET 2 (VALVE CHECKLISTS) AND COMPLETING THE "VALVE INFORMATION SHEETS" 1.0 Use valve tag number given on data sheet (2) as key to search for drawings and information requ' red to complete "valve information sheet" (copy in this Guideline).
1.1 Example
Unique "Valve Tag Number" 1-23-569 Unit Identification No.
- ilechani. cal System Identification No.
Valve No.
1.2 The first step involves the use of the mechanical system identification number and valve number for determining the valve marker tag or instrument tabulation drawing where the unique "valve tag number" is tabulated. Below is listed t.
cross-reference between valve numbers and valve marker tag or instrument tabulation drawing series for making this determination: )
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TMPLEMENTAT20N OF NRC-02E BULLETINS 79-02/79-34 >
FOR BROWNS FERRY NUCLFAR PLAhi BTEP-PI 85-01 i!
GUIDELINE C (Continued) 47A366- (Mech Sys I.D. No.) - (Sh No.) for valves with valve numbers 1 500.
47A36.b (Mech Sys I.D. No.) - (Sh No.) for valves with valve numbers 200 to 499.
47B601- (Mech Sys I.D. No.) - (Sh No.) for valves with a control valve prefix, i.e. , FCV, TSV, HCV, etc. , and valve numbers < 200.
When the drawing is found, the necessary information can be obtained, such as manuracturer's part number, TVA mark number, etc.
1.3 After Section 1.2 has been completed, you will have determined whether the valve is a General Electric (GE) purchased item (TVA Nuclear $ team Supply System Contract Nos. 66C60-90744, units ,1.and 2, and 67C60-91750, unit 3) or one that has been purchased by TVA. Section 2.0 will deal with CE purchased items only. For TVA purchased items, see Section 2.1.
2.0 If the valve was purchased by GE, a GE part nu=ber will be given on one of the drawings listed in Section 1.2. With this part number, the VPF-MPL cross-reference book can be used.
This book will give the VPF number and valve vendor drawing
- number. The valve drawing can be found in the DNE microfilm card file under the above-mentioned GE contracts filed by either VPF number or vendor drawing number. Fro = the valve drawing you can obtain the necessary information to complete the valve infor=ation sheet (copy.ir this Guideline).
- 1. If the valve was purchased by TVA, a TVA mark number will be given on one of the drawings listed in Se;. tion 1.2. First obtain the TVA bill of material (BM) drawing number. This information can be obtained from two sources. The first source would be froc the general notes on the first sheet of the system drawings. "'he second source voald be ene of the drawings listed under Section 1.2. On this drawing is given a drawing number under "Instrument Location," i.e., i.79450-1.
The bill of material will be the sa:re as this nu=be stith the i "W" omitted and "BM" substituted, i.e., 4'3M450. The next step ^
i would be to obtain a copy of '% s BM from the DNE files. From ta "BM" will be obtained a "hl.s" file number. Deter =ine the ,
requisition or; contract numbe~ using the NIM file number j requisition number cross-reference sheet or call the j DNIl - 1023k
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N IMPt.EMENTAT20N OF NRC-01E BUL2.E*I'NS 79-02/79-14 FOR BROWS TERRY NUCLEAR PLANT BTEP-PI 85-01 GUIDELINE C (Continued) l procurement section in MEB. Using the requisition or contract number, go to the VSMF file section and obtair. a film cartridge of the contract. Search the contract for correspondence pertaining f.o the particular valve mark number, i.e.,
47W450-18. This correspondence should reflect a vendor drawing nu=ber that was transmitted by the vendor to TVA for *.7 proval.
After establishing the NIH file number, contract / requisition nu=ber, and vendor drawing number, a copy of the drawing should be obtained from the microfilm card file using the contract /
requisition number and drawing number.
If the vendor drawing nu=ber cannot be established in the contract correspondence, search through the NIM files that have been transferred to BTEP/CEB from MEB. If this search does not yield the drawing, search the "NIM or NIE" files located in EEB. If this search fails, then contact TIC and give them as much information as possible.
3.0 Co:pletion of the valve information sheet rhall be acco=plished utilizing the vendor drawings and/or information gathered in the information search noted in Sections 1.0 through and including 2.0. If required information is not found on these docu=ents, such as valve weight, length, etc., the vendor catalogs should be used to provide data fcr reference.
Verification of valve weights shall be performed under Phase II in accordance with the requirements of BFN-50-D707.
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IMPLEMEhTATION OF NRC-0IE BULLETI' N S 79-02/79-24 TOP PROWS FECRY h"JCLDS PLAN" BTEP-PI F5-01 l GUIDELINE C (Continued)
HANDLING DATA SHEET 3 (CONFIGURATION DRAWINGS) AND GENERATING STUDY DRAWINGS Ensure the following is performed in &ccordance with Guideline B:
1.0 Verify that the attachments are with the correct data sheets and all attachments have problem numbers stamped on them. Make sure all data sheet 3 and discrepancies have cot.;2iguration drawings or sketches.
- 2.0 On system flow diagrams, follow the piping inspections to make sure inspections have bee 1 carried to the limits as indicsted by NEB or by the piping analysis section, whichever is greater. ,
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IMPLEMDGATION OF NRC-08E BULLETZNS,79-02/79-24 erre_o, os_no :
e- n.mc errev ,-.- rie c. a.-
en_ e.
GUIDELINE C (Continued)
HANDLING DATA SHEET 4--(PENETRATIONS) 1.0 Ensure the following is performed:
1.1 Verify that the attachments have problem numbers stamped on them.
If 1.2 Assign problem numbers to penetrations not already assigned.
the penetration and support are' integral and designated as a restraint number on the configuration drawing, ensure thattoboth Attempt f data sheets 1 and 4 are included in the package. i assign same problem number to both data sheets 1 and 4 for ;
integral penetration-support.
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M ZMPLEMEh7ATZON OF NRC-02E BULLET 8NS 79-03/74-24 rne non*_'s'e vrnov mm r AR P'PP ~~
BFEP-PI 85-02 GUIDELINE C (Continued)
VALVE INFORMATION SHEET SYSTEM UNIT VALVE TAG NO VALVE MARK NO VALVE SIZE AND TYPE (2" GATE, 14" GLOBE, ETC.)
VALVE MARKER TAG OR INSTRUMENT TABULATION DG NO BILL OF MATERIAL NO ,
C0hTRACT NO RITERENCE Ois FILE 0'(N1M-156, NIM 1.3. 1-1D4, ETC.)
VPF NO: VALVE OPERATOR VALVE MANUTACIURER VALVE MODEL NO VALVE MANUFACTURER DWG NO VALVE BODY MATERIAL (CS A216 GR B, ETC.)
OPERATOR MAlrJTACIURER i OPERATOR TYPE AND SI"E (MOTOR SMB-00, MA!TJAL, ETC.)
DJD TYPE (BUTT WELD, FLANGED, ETC. ) I LDJGTH DiD TO DG ,
WEIGHT: VALVE OPERATOR VALVE PLUS OPEFATOR 1
VALVE C.G. I DWG. FILED UNDER: VPF NO MFG DWG NO OTHER CATALOG NO (IF CAT. INFd. IS USED) i DNE1 - 1023k TVA 10535 (EN DES- 7-77)
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2MPLEMENTATION OP NRC-OIE BULLET 2NS 79-02/79-14 rne nonme rrrev ir'm var o?n- grEp pt e5_o; GUIDELINE C (Continued)
TABLE OF CONTENTS AND
SUMMARY
SYSTEM:
UNIT:
1.0 IE BULLETIN 79-14 INSPEC'"ION DATA SHEETS 1.1 DATA SHEET #1 (PIPE SUPPORT CHECKLIST) a 1.2 DATA SHEET #2 (VALVE CHECKI.IST) 1.3 DATA SHEET #3 (DRAWING CONFIGURATION CHECK,IST) '
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'IMPLEMDCATION OF NRC-OIE BULLETNJS 79-02/79-14 TOR DROWS T70DY h"3CLEAP PLAN- BTEP-PI 85-01 GUIDELINE C (Continued)
TABLE OF C0!CDCS AND
SUMMARY
(Continued) 1.4 DATA SHEET #4 (FLOOR AND WALL PDJETRATION CHECKLIST [
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l 1.S DATA SHEET #5 (SPRING HANGER CHECKLIST FILED WITH THE APPROPRIATE SUPPORT IN SECTION 1.1)
NCTE: Some earlier packages contained a separate data sheet 5 section. The data sheet 5 (spring) information was merged with data sheet 1. Data sheet I now serves as official documentation for springs.
2.0 IE BULLETIN,79-02 INSPECTION DATA SHEETS l
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1 3.0 DISCREPANCY RESOLUTION SHECS [0RIGINAL FILED WITH THE j APPROPRIATE DATA SHEET (SEE SECTIONS 1.1 THROUGH 1.5)] i PROBLDi # DESCRIPTION (SUPPORT #. DWG #, ETC.)
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IMPLEMDCATION OF NRC-0IE BULLETINS 79-02/79-14 ERO'JNS FERDY Nt'"LEAR PLAh" BFEP-PT 85-01 GUIDELINE C (Continued)
TABLE OF CO?nTJCS MO
SUMMARY
(Continued) 4.0 SUPPORTS INADEQU CE FOR EXISTING DESIGN LOADS [ FILED WITH THE APPROPRIATE DATA SHEET (SEE SECTION 1.1)]
5.0 ADDITIONAL INSPECTION INFORMATION REQUIRED 6.0 EXISTING RIGOROUS ANALYSIS M;D ANALYSIS ISOMETRIC l
l 7.0 PHYSICAL PIPING AND HANGER LOCATION DRAWINGS i
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M. t IMPLEMEh7ATZON OF NRC-OTE BULLETINS 79-02/7'9-14 rnr non_me vreev mm rir 9, a- nrre_p! p r, _ o !
Gif1DELINE C (Continued)
TABLE OF COhTDCS AND
SUMMARY
(Continued) 8.0 BILLS OF MATERIAL (VERITICATION OF MATERIAL AND SCIEDULE)
- 9.0 DESIGN PRESSURES AND TEMPERAn/RES -
10.0 INSULATION DATA 11.0 OPERATING PRESSURES Ah") TEMPERAETRES 12.0 EQUIPMDC DATA l
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IMPLEMDCATZON OF NRC-02E BULLETINS 79-02/[9-14 FOR BROWNS FERRY NUCLEAR PL/JC BTEP-PI 85-01 GUIDELINE C (Continued)
TABLE OF C0!CDUS AND
SUMMARY
(Continued) 13.0 ADDITIONAL ANALYSIS INFORMATION 14.0 STUDY DRAVINGS, COMPOSITES, ETC.
15.0 ENGINEERING ANALYSIS COMMDCS The following may be.used as footnotes for the "Table of Contents and Su=ary":
- (1) ADDITIONAL SUPPORT DETAIL INFORMATION REQUIRED FROM FIELD.
1
- ( VEIGHT OF VA LVE AND OPERATOR REQUIRED FROM VDOOR.
- (3) BRANCH LINE LOCATION, IDOCIFIO. TION, AND SIEE INFORMATION REQUIRED FROM FIELD.
- (4'e SUPPORT DCAIL INSPECTION DRAVING REQUIRED IROM FIELD.
- (5) COMPLETE PDiETRATION n! FORMATION REQUIRED IROM FIELD.
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IMPLEMDiTATION OF NRC-OIE BULLETINS 79-02/79-14 !
FOR BROWS FERRY NUCLEAR PLAhi BTEP-PI 85-01 GUIDELINE C (Continued)
SUGGESTED COLOR CODE FOR SEPIAS INSULATION INFORMATION, CIVIL MARK NUMBERS. SUPPORT DIMDJSIONS, AND ADDITIONAL MISCELLANENIS COMMD.TS --- ORANGE SUPPORT DESIGNATION, SYMBOL AND PIPE DIMENSIONS GRAPHITE ANY COMMDTS ADDRESSED TO THE FIELD GREEN 4
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- N IMPLEMDCATION OF NRC-0IE 1ULLETI'NS 79-02/79-14 rop op s m s r ee= Y h".tcll' R P # C BTEP-PI 85-01 AWACHMUC 6.5 :
BFN INTERIM ANCHORAGE EVALUATION CRITERIA
- 1. Plug Depth Criteria l
- 2. Edge Distance Criteria
- 3. Thread Engagement Criteria 4 Anchor Capacity Criteria
- 5. Expected Load for Cut-off SSDs i
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IMPLEMEhTATIONOFNRC-0IEBULI.ETIhS 79-02/79-14 TOR BRO *.'NS TERRY NUCLEAR PLAhT BTEP-PI 85-01 AMACHMEhT 6.5 (Continued)
(1) PLUG DEPTH CRITERIA !
Flug Depth (Inches) i Anchor Size Normal Minimum Maximum 7/8 Inches 2-9/32 1-3/4 2-1/2 3/4 Inches 1-15/16 1-1/2 2-1/8 5/8 Inches 1-5/16 1 1-1/2 1/2 Inches 1-5/32 7/8 1-3/8 3/8 Inches 25/32 19/32 15/16 (2) CRITERIA FOR EDGE DISTANCE FOR PUNCHED, REAMED, OR DRILLED HOLES Table 1.16.5 AISC Seventh Edition is to be used as the minimum i
criteria for edge distance for punched, reamed, or drilled holes.
(Inches)
Bolt Diameter At Sheared Edge At Rolled Edge or Gas Cut 1/2 7/8 3/4 5/8 1-1/8 7/8 3/4 1-1/4 1 7/8 1-1/2 1 ,'./ 8 4
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N IMPLEMCCATIONOFNRC-OIEBULLENNS 79-02/79-14
- v. . ;. Gee .ra.c.. :C ^_M. ."Z ! ? E T " ' M T.
ATI'ACHMDC 6.5 (Continued)
(3) THREAD ENGAGEMDC CRITERIA
, G-32 requires one bolt diameter thread engagements. For engagements less than one bolt dia=eter refer to the following sumary:
i
':HREAD ENGAGEMDC TO DEVELOP SELF-DRILL ANCHORS !'
Bolt Ult Ult Bolt Engagement to Develop Dia. C-32 A307 Use Ultimate of Anchor D2 7 . 81* 8.52K 91% 0.375 diameters 5/8 10.5 13.56 77% 0.25 diameters
, 3/4 14.9 20.04 74% 0.25 diameters n
7/8 ,
17.8 27.72 64%* 0.25 diameters 7/8 17.8 27.72 64% 0.25 diameters
- Based en tests at Singleton Laboratory memorandum "Browns Ferry Nuclear Plant - Correlations of Thread Engagement of A307 Bolts in Self-Drilling Anchors to Tensile Capacity-OIE Bulletin 79-02," dated August 17, 1979 (CSB 790820 001).
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IMPLEMDCAT20N OF NRC-0IE BULLETI'NS 79-02/79-14 i rne een_Ne cropv etm to em- grgp.p; es_o) !'
ATTACHMDC 6.5 (Continued)
TENSILE CAPACITY AS A FUNCTION OF THREAD DJGAGEMD?T FOR ASW A 307 BOLTS IN PHILLIPS SELF-DRILLING ANCHORS 1/2" Anchors Engagement / Failure Load /
Thread Anchor Dia. Failure Ultimate Load -
Engagement (Ratio) L7ad (Average) Failure Mode Inches Pounds Percent 1/2 1.00 10,620 100 Tensile failure of bolt 1/2 10,880 Tensile failure of bolt 3/8 .0.75 10,400 100 Tensile failure of bolt 3/8 11,070 Tensile failure of bolt 1/4 8,000 Thread failure of bolt 1/4 0.50 9,840 84.5 Thread failure of bolt 1/4 . 9,380 Thread failure of bolt t 3/16 0.375 7,860 69.8 Thread failure of bolt i 3/16 7,140 Thread failure of bolt
! 1/8 5,000 nread failure of bolt-i 1/8 0.25 3,600 46.3 Thread failure of bolt i 1/8 6,330 Thread failure of bolt i
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& s IMPLEMDCATION OF NRC-0IE BULLETINS 79-02/79-14 TOR BROWNS TER'lY NUCLEAR PLA?C BTEP-PI 85-01 A;7ACHMDC 6.5 (Continued)
TENSILE CAPACIU AS A FITNCTION OF THREAD ENGAGEMDC FOR AS1H A 307 BOLTS IN PHILLIPS SELF-DRILLING ANCHORS 5/8" Anchors Engagement / Failure Load /
Thread Anchor Dia. Failure Ultimate Load Engagement (Ratio) Load (Average) Tailure Mode Inches Pounds Percent l 5/8 1.00 19.100 100 Tensile failure of bolt 5/8 19,400 Tensile failure of bolt 1/2 0,80 18,600 100 Tensile failure of bolt 1/2 18,950 Tensile failure of bolt 3/8 0.60 17,550 Thread failure of bolt 3/8 17,050 91 Thread failure of bolt !
5/16 0.50 15,600 79.2 Thread failure of bolt 5/16 14,500 Thread failure of bolt i 1/4 0.40 11,100 62 Thread failure of bolt 1/4 12,500 Thread failure of bolt l
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-- __ __-_m___.___--m_ _ _ _ _ . . _ . . _ _ _ _
& i IMPLEMDCATION OT NRC-CIE BULLETI S 79-02/79-14
, , , - . . ,= ,.. ...-. m .....-
._ ,..m.
A7TACHMDC 6.5 (Continued)
TENSILE CAPACI!Y AS A FUNC* ION OF THREAD ENGAGEMD.T l FOR AS'Di A 307 BOLTS IN PHILLIPS - SELF-DRILLING ANCHORS 3/4" Anchors Engagement / Failure Load /
Thread Anchor Dia. Failure Ultimate Load Engagement IRatio) Load (Average) _
Failure Mode ,
Inches Pounds Percant l
5/8 0.83 25,150 100 Tensile failure of bolt '
5/8 26,200 Tensile failure of bolt ;
1/2 0.67 '
25,450 100 Tensile failure of bolt i 1/2 25,850 Tensile failure of bolt !
'3/8 0.50 20,900 78.4 Thread failure of bolt 3/8 19,350 Thread failure of bolt 1/4 0.33 15,000 57.1 Thread failure of bolt 1/4 14,300 Thread failure of bolt i
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- i IMPLEMENTATION OF NRC-OIE BULLE* INS 79-02/79-14 TOT I'! '"; Frr." """* r2 P'_u--
ter" d* "M AWACHMEhT 6.5 (Continued)
TENSILE CAPACIW AS A FUNCTION OF THREAD ENGAGEMD.T FOR AS*IN A 307 BOLTS IN PHILLIPS SELF-DRILLING ANCHORS i i
7/8" Anchors l
Engagement / Failure Load /
Bread Anchor Dia. Failure Ultimate Load Engagement (Ratio) Load _ Average)
( Failure Mode Inches Pounds Percent i 3/4 0.857 39,650 100 Tensile failure of bolt 3/4 39,600 Tensile failure of bolt 5/8 0.714 40.100 100' Tensile failure of bolt !
5/8 ,
40,050 Tensile failure of bolt 1/2 0.571 41,500 100 Tensile failure of bolt i l 1/2 39,200 Tensile failure of bolt ;
i 3/8 0.429 30,250 77.4 Thread failure of bolt I 3/8 31,650 Thread failure of bolt 1/4 0.286 19,600 48.9 3 read failure of bolt 1/4 19.550 3 read failure of bolt l
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IMPLDfDCATION OP NRC-CIE BULIGNS 79-02/79-16 FOR BROWNS TERRY NUCLEAR P W B N -P! 85-01 A7.ACST.?C 6.5 (Continued)
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IMPLEMDirATION OF NRC-0IE BULLE*I'NS 79-02/79-14 !
TOR BRO **NS FERRY NUCLEAR PLAh? BTEP-PI 85-01 I (4) ANCHOR CAPACITY CRITERIA ,
For capacities of self-drilling anchors refer to the following charts.
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IMPLI:MDGAZION OT NRC-42E BM.LET3HS 7943/79-14 TOR BROWNS TERRY )CCLEAR PLANT BTT.P-PI 85-01 A M CHMENT 6.5 (continued)
ANC-OR CAPAC::TY CRITER:: A
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IMP 1.ZMDCATION or NRC-CIE BUL1ITINS 79-02/79-14 TUR BROWNS TERRY NUC . EAR P!.ac BTIP-PI 45-01 ACACHMrr 6.5 (Contin 2ed)
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DTLD'ECATION OF NRC-OIE StrM;ns 79-02/~9-lh m R bro'.*NS FERRY fiJCLEAR FLA!,7 B M -FI C 5-01 ATIACHE?iT 6.5 (Continued)
(5) ESTD! ATE OF EFEC LOAD FOR CUT-OFF SELF-DRILL MSED ON FJTo'r N TEST DATA Effectiveness '
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Tennes:ee Vc11ey Authority J EV 00 02 Pae o 3 Browns Pcrry Nuclocr Plant - , BP-SDSP-1 Site Director Standard Practice BP SDSP-2.11 PORM SDSP l DP&CU legging No. S M M I - 5".1 - A - 0 3 I
PROCEDURE /INSUUC'"ION CRANCE REQUEST
'17 TIS FORM SHOULD BE COMPLE~T.D IN BLACK INK OR "'YPEVRI"""EN Type of Change X. Perr.anent Temporary Change-To Become Permanent ,
Temporary Change-E.xpiration Date 4 (60 Days if not specified) l ,_,_ ,,, l Advanced Revision Procedure No. Unit Title Sng. g, t . a l. Z ,3 VERICYIWS CA RREC7" INSTALLR7)bH DF CDNtWETt" FXPAUS16N AUlM1 naamaa tar aavision: ADD Du)&. w I 7W3oD-I 7 76 SECTioN l.D AUD KLI tne NCE TD G sis G. tele s 1)lDG W 17W3bo-l7,ANb DELETE REFERENCE TD FH/llIPS CDNPM'l GEME71 ANCMR LATAL4&v SECT /DN 2.I + 1.3. REVISE SEf7/bN 7.//.d fe dDD "4RG-W.
REVISE SECTIDN 7.ll.2 F26M BF to.2 To SDSP13.i* DELETE luoRbS"MdB BR" AND ADD SKETCH To DATA SREE T 1. del.ETE ID6EDS. 'M)l241 SEf/ESW 'MbD BR ' FRbM DATR SHEET REVISE fffffENCEZ-. EF4.2A Db To A BLANE SDSP /3./ DNLINE DATRTDSHEE1'Z F/LL IN. Aea DRAN/NE-SEf/ES p, y _- c,.e To
~
sr Affected Pag'es I 9. '2 ,13 ***
?.8 . uh 0v--- MtM Mets, ll-2 u. p t., 27So Submittec By Section Date E.xtension SDSP-15 has been completed and is attached j N/A for advanced revisions). l rPi&S M- ///r/No t
Resp. Section Procedure Coordinator Date APPRC7ED &en Neb / fit'ADMs WNMN & IIllkC&
Responsible Section- Responsible Section Supervisor Date (N/A fer adva cad ravisie s)
APPRD7ALS Wh M/A J/A NA Programatic Contact
- Date Site Quality Manager ** Date s
aln ~' '
n/A Principal Manager
- Date ' -
Al h ^
PORC C1. air =an Date
_h @ /2. -/ / A6 -
Plant Manager Date -
Al/A .
_ N/A To be completed by DP&CU:
Site Director
- Date issued at revision I did not result in change of revision level
- Required for Site Director's Standard Practices CitLY
- Required for Standard Practices and Site Director's Standard Practice ONLY
.l Retention Period: Lifetime , Responsibility: .DP&CU Supervisor 0155p _
J
- SMMI5.1.A Pag 2 1 ci 1 liISTORY OF REVISION / REVIEW RE7.
NO. DATE REVISED PAG'!.S REASON FOR CURREh'i' REVISION 03/07/84 1,2,3,4,5 05/08/84 4 Change one word to prevent problems later.
t 05/15/85 1-6; add 7,8 Clarify scope, correct error on Data Sheet, add Attachment A, and data sheet
- 2. To allow OE inspections to I participate in inspection. Clarify repair. (850304-23) 06/28/85 1-8; add 9 Clarify Drawing (key plan) Control system inspection status, clarify information on data sheet 2, correct table on page 7. (850624-04) ,
09/11/85 All General revision for clarity of documentation on data sheets.
, (850903-07) 09/16/85 6,11 Correction on paragraph number and data sheet. (850912-01)
I'2/02/85 1,4-10 Change title to increase scope of anchors. (851121-07)
ITC 12/09/86 9,13 For all work instructions approved on or after 11-17-86, SDSP-13.1 shall apply to all welding activities.
( SMMIS .1. A-02.) .
I 12/11/86 1,9,12.13 Add DWG. #17W300-17 to Section 1.0.
Add reference to G-51 and G-66 and DWG
- 17W300-17, and Delete reference to Phillips Company Concrete Anchor Catalog Section 2.1 and 2.3. Revise Section 7.11.d to Add "0R G-66".
Revise Section 7.11.e from BF6.2 to '
SDSP13.1. Delete Words "MODS OR" and add sketch to Data Sheet 1. Delete words "48W1241 Series" and "MOD OR"
, free Data Sheet 2. Add A blank line to fill in Drawing Series to Data Sheet
- 2. Revise reference BF6.2 to SDSP 13.1 on Data Sheet 2. Incorporate ITC SMMIS.1.A-02.
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t.'N!!3 1, 2. /J.'D 3 1.0 SCo*E "he pu. pose of this procedure is to estsb11 h an inspection and cpair program ,for c ncrete e:cpsnsion anchor: on all safety-related ,
compenengs and to provide a means of'coirective' action for thosa cy : ens identified on IE 3u11etin 79-02. Sec Attachment A *for ;*: tent. " -
."rier to !!:/.1935,
,. the l..cycetiens vare perie=ed hy. Nuclear ? cut: ( .JC
...). 0:;innin; !!ay ,1985, ths:: i..0 :ctien: vill be parie=ad by cn;in:c : and/c design en,ir. eri ; c::::icte: fre.a the Office ci s
In-in:cri.u; (CE). !!cdific:tien Group will perfo:= repairs and n !.!fic:.ticas as allowed by thi in:truction and LS* 1241 cnd 17U7,00-17 l
! (
i.:Ci;n ! cries Dr. wings. '
s.0 P.e.a ...aa 1
' 92. J" i, '
O.1 C:ns::::<:ics Procedures ^G-22, G-51 c.nd G-Gs.
ITC di4 l' 2.2 40'11761 tic ign series drawings. l 1
2.3 Oc:ign 0 : vin.; 17U220.-17 2.4 :: 1.411ctin 79-02. *
.l 2.5 Te:hnic:1 3p:cification 3.G.!: -
2.6
- !:no frem N.lR. I :: ley to G. R. E011 dated June 23 -
Subject:
- l
(
- '"? II: 70-02 Inspection Prog :.m (022 05 0620 COS) l l
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e Pago 1 BT SMMI-5.1-A U.NNESSEE VALLEY AUTHORITY REV 0001 BRO *-HS TERRY h"JCLEAR PLANT SPECIAL MECHANICAL MAIhTENANCE INSTRUCTION 5.1-A INSPECION AND REPAIR PROGRAM TOR VERITYING CORREC INSTALLATION OF CONCRETE EXPANSION ANCHORS UNITS 1. 2. AND 3 1.0 SCOPE The purpose of this procedure is to establish an inspection and repair program ,for concrete expansion anchors on all safety-related componenksandtoprovidea'meansofcorrective-actionforthosesystems identified on IE Bulletin 79-02. See Attachment A for syste=s.
Prior to May.1985. the inspections were performed by Nuclear Poker (NUC PR). Beginning May 1985. these inspections will be performed by engineers and/or design engineering associates from the Office of Engineering (CE). Modification Group will perform repairs and modifications as allowed by this instructior, and 48V1241 and 17W300 *7 Design Series Drawings.
2.0 REFERENCES
2.1 Constru$tionProceduresG-32,G-51andG-66, 2.2 48W1241 design series drawings.
2.3 Design Drawing 17W300-17 2.4 IE LJ11etin 79-02.
2.5 Technical Specification 3.6.H 2.6 Memo from N. ,R. Seasley to G. R. Hall dated June 28 -
Subject:
BTNP IES 79-02 Inspection Program (B22 85 0628 008) 041&o
- Pcgo 2 BT SKMI-5.1-A 3.0 PREREQUISITES A D TOOLS .
REV 0001 3.1 Tools for removing 3/8-inch through 7/8-inch diameter bolts will be needed.
- 3.2 Scale or rule with at least 1/16-inch increments-for taking measurements.
3.3 Thickness probe to measure gap between attachment plate and concrete surf ace.
3 . 4, Comply with all Health Physics requirements dictated by the job
. o location. .
~
3.5 Initiate Maintenance Request (MR) to install scaffold and to remove insulation as needed. -
- 4.0 PRECAUTIONS Only one bolt at a time may be removed from each base plate acsembly while the system is in service. If more than one bolt at a time must be removed i
for corrections to anchors or base plates, contact shift engineer in reference to technical specification 3.6.H. Supports on systems not in service will be inspected or repaired in accordance with applicable plant instruction 4(BF-7.6) or by direction of CE engineer. ,
5.0 TDtPORARY COEITIONS Temporwcy supporting requirements due to removal or replacement of base plates will be determined by CE and controlled by appropriate plant procedure.
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- Pcgo 3~
BT SMMI-5.3-A ,
6.0 ALARA CONSIDERATIONS REV o001 Tor those areas in which radiation dose rate is high, contact Health Physics. If the source can be located and shleided, radiation exposure should be reduced to as low at reasonably achievable as advised by Health j Physics.
7.0 INSMUCTIONS 7.1 Tor the purposes of the IEB 79-02 inspections. OE will provide marked up key plan prints which uniquely identify each support to be ;
inspected and which show sufficient information to locate same. The l original key plan prints will be:
A. Assigned unique control numbers, i
B. kept in a c: aster flie for the duration of the 79-02 inspection program and, C. copies incorporated into the associated inspection - MR packages for inspection / repair documentation.
Authorization for inspection and repair for specific hangers will be made by MR which will list the hangers and be cross-referenced as stateddin 7.11.n. i 7.2 Sketch each base plate on data sheet I and identify each bolt anchor !
with an individual anchor number. Also measure the Mit anche - ]
center-to-center spacing and include on sketch.
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A o A o bLT kPALIU6 E O $ 0sj ANLHb d BL 1
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.- Pago 6 BT SMMI-5.1-A '
7.0 INS'IRUCTIONS (Continued) REV 0001 l 7.3 Using a probe 1/2-inch wide by a thickness that is between 90 and 1007, of the maximum gap specified in gap table below, check each j bolt location for the gap between the attachment plate and concrete l surface as shown in sketch. If the probe can touch the shell or i i
bolt for at least 90 degrees around the opening, measure the gap and I
- record on data sheet 1. If there are areas where the gap is less than the specified probe thickness, record the actual gap under the I gap column on the data sheet 1. If the gap is unacceptable, repair in accordance with section 7.11 and document on data sheet 2.
) '
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_ 90* AAf.,
9f AAC
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A CORNER END CONDITION -
ANCHOR SI:'E- P.AX. GAP 7/8 inch 3/16 inch 3/4 inch 3/16 inch .
j l
5/8 inch 3/16 inch 4 j 1/2 inch 1/8 inch ,
i
, 3/3 inch 1/8 inch In cases where the recorded plate gap value changes after 4
reinsta11ation of bolt, indicate this in the remarks column on data l
sheet 1.
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r Pcg2 5 BT SMMI-5.1-A 7.0 INSTRUCTIONS (Continued) REV 0001 7.3.1 The Cognizant OE Engineer will determine whether or not the concrete expansion anchor can be inspected. If the anchor can be inspected and is determined to be one of the following types inspect using the following criteria:
(a) b*edgebolt - If the head does not show any saw or grinding marks no repair is required - LEAVE THE ANCHOR ALONE.
t (b) Studs With Nut In Grouted Base Plates - (end shows saw or
. t .
grinding marks) proceed to 7.11 for repair.
]> (c) Self Drilling Type Anchors - inspect per 7.4 Repair if necessary per 7.11.
...(d)
, Cinch (Lead Anchors) - There is no inspection criteria for cinch (lead anchors) on 79-02 program. Repair per 7.11.
(e) Special Case - If anchor types other than the above are found repair per 7.11. .
i
- ! 7.3.2 If any of the above type anchors cannot be inspected repair per 7.11.
]
7.4 For systems in service, remove one bolt at'a time from each base 4
plate, record the following measurements on data sheet 1. For aystems out of service, the bolt removal will be in accordance with direction of OE representative.
- A. Anchor s'ize, i
~
Pag? 6-BF SMMI-5.1-A 7.0 INSTRUCTIONS (Continued) REV 0001 :
7.4 (Continued)
B. (L1) Distance from top of shell to top of plug.
C. (L2) Distance from top of shell to top of plate.
D. (T) Thickness of base plate.
E. (D) Diameter of each hole in base plate. In cases where an out-of-round or oblong bolt hole is encountered, measure the maximum dimension of the hole.
t 's F. Recess - Distance from top of shell to s,tructural concrete
~
surface. I G. (TER) Measurement of thread engagement for each existing
... _ anchor bolt. This is the as-found thread engagement. I I
$ - H. (TEI) Measurement of thread engagement for each bort af ter !
repair. Repair includes adding washers, replacing bolt, i resetting. anchor, etc. This is the as left thread j
i engagement.
I. (ED) Distance from edge of each bolt hole to nearest plate edge.
J. (GAP) Distance between base plate and concrete surface at each anchor.
7.5 If distance of expansion plug to top of shell (L1) is less than the minimum plug depth specified on data sheet 1, the anchor cannot be
-~
considered useful and must be repaired in accordance with section 7.11. '
0064D
., , Pago 7 BF SMMI-5.1-A 7.0 INS RUCTIONS (Continued) REV 0001 7.6 If distance of expansion plug to top of shell (L1) is greater than maximum plug depth. (as specified on data sheet 1) the anchor can be >
reset by inserting a bolt or threaded rod and using rotary impact drill or large hamer to reseat anchor. Repeat measurements since this will change the depth in some cases. If depth measurement is not within chart range (Data Sheet 1), repair in accordance with section 7.11. If measure:nent is acceptable, install existing bolt and tighten 1/8 to 1/4 of turn after contact between bolt, washer and plate. Record new bolt thread engagement under TEI.
7.7 Bolt Measurement / Replacement - Acceptable Anchors
- a. Determine thread engagement of existing bolt by inserting bolt into anchor without turning and neasuring the distance from the bolt head to the base plate (or washer if used). See TER on d
Data Sheet 1 sketch. The minimum acceptable thread engager.ent ,
is 1/16" less than one bolt diameter. If the existing thread engagement is unacceptable, replace the bolt to obtain the ,
i proper thread engagement and record TEI and the new bolt length on Data Sheet 1. Due to limited' bolt length increments and j -
limited amounts of available thread in the anchor shell, it may be necessary to install washers in order to tighten properly.
I
- b. Be sur.e that the reinstalled bolt does not contact the 4
]
expansion plug. TE: must be less than L1.
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/ Pg3 8 BF SMHI-5.1-A 7.0 INSTRUCTIONS (Continued) REV 0001 7.7 (Continued)
- c. Tighten bolt 1/8 to 1/4 turn after contact of bolt, washer, and plate. If bolt is turned more than this, it is possible to pull shell toward plate and allow bolt to eject the expansion plug.
7 . 8, Ricess - If recess cannot be directly measured it can be determined by subtracting attachment thickness. including gap from L2.
Recess'= L2-(T+ GAP). Minimum recess is 0". Maximum recess is per 48W1241 Note 20. If the recess it out of tolerance, repair per section 7.11.
7.9 If maximum diameter of the hole in baseplate (D) is 3/16-inch greater than the^ bolt diameter, repair in accordance with section 7.11.
7.10 Edge distances (ED) are to be measured and recorded on Data Sheet
- 1. Evaluation of edge distances is not a part of this inspection.
NOTE: This is supplemental information for OE.
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- 7.11 .. pair ::::h::::
- a. :.~.. .u:..:-i-i:q in:p:::i ns anc repairs will te ..st ad en sli dst :n:::: 1 and O.
- b. !!cchanical !!cdifications vill repair in accordance with desi:;n d:: vin; :: ries .'.f'.112al, end/or 40 directed by C2 on da:a sheet g
- 2. Justifiesticnic:1culaticas =ust be done by OI :nd be e
e t included c attached to d ta sheet 2 for vari:n=er to' 43'.*1241 d:cwin; caries. Varian:cs will be unicuely identified on dacc e
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- c. After :sp:ir: c: c::p12tc. c en-in:9. xe to *tarify th::
a .... . e..
.. ... .. . . . . ...p . a. ,....s,- .
- d. !!echani::1 !cdification: . rill :: pair On:hcrs in accordan: with
'. t G-32 Or G-66 t..d in:pcc 70: :!AI-c . .
- 2. !!: $anical I!cdific::icas vill veld in at::: dance '.rith SCS?-13.1. .
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-F.- ::achan!:s1 ::edific : lens vill ::psir s ut^in ec::: dan: uith ca.noral constru::ica specific..tiens G-0, c-::.AFM.
1 u-31, : --
.. .. AND MAlM ARD/oK MAI3(e As YEdDIRED. ** ;
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- 'Pigo 9 BT SMMI-5.1-A
-7.0 INSTRUCIONS (Continued) REV 0001 7.11 Repair Methods:
a .- MR authorizing inspections and repairs will be noted on all.
data sheets 1 and 2.
- b. Mechanical Modifications will repair in accordance with design drawing series 48W1241, and/or as directed by OE on data sheet
, 2. Justification / calculations must be done by OE and be
- , included or attached to data sheet 2 for variances to 48W1241 drawing series. Variances will be uniquely identified on data ;
sheet 2. ,
l C. After repairs.are complete, OE engineers are to verif'9 that
~ l attachment is acceptable. l
- d. Mechanical Modifications will repair anchors in accordance with G-32 or G-66 and inspect per MAI-4
- e. Mechanical Modifications will weld in accordance with SDSP-13.1.
- f. Mechanical Modifications will paint in accordance with General .
Construction Specification G-14 I454 (47B435 drawing series) '
- g. Mechanical Modifications will repair grout in accordance with !
GeberalConstructionSpecificationsG-2,G-32.G-51,andMAI-34; i as required.
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- Pego 10 BT SMMI-5.1-A 7.0 INSTRUCTIONS (Continued) REV 0001 i
7.11 (Continued)
- h. Grouting oversize holes, replacing bolts,' adding washers, and resetting anchors may be documented on data sheet 1 without using data sheet 2. More extensive repairs require data sheet
- 2. The sketch of the repair on data sheet 2 shall include all new material and the nominal size of new bolts, anchors, i
, [t angles, bars, etc. Recording of heat numbers on data sheet 2 is required only if no welding is involved.
7.12 Use data cover sheet to transmit original data sheets of'this procedure and copies Gf MAI-4 data sheets to OE.
7.13 OE will utilize data sheets and key plan prints to incorporate ,
changes to IES 79-14 ECN issued drawings.
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- e- Page 11 BF SMKI-5.1-A
.j l Inspection Program for Verifying Correct R E V. 0001 i Installation of Concrete Expansion Anchors i Units 1, 2, and 3 l .
l Data Cover Sheet t
- 1. The attached data package for unit contains ;
i data sheets one and data sheets.two and l 1 i MAI-4 data sheets. )
I t '
MOD,or OE Representative Date i I
c
- System: '
l Support Nos: '
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Page 13' BT SMMI-5.1-A DATA SHEET 2
. REV 0001 Repair Documentatson s
! e SKETCH HR No. .
Unit ,
'* System Support No.
O Repair in accordance with Drawing Detail No(s).
as depicted in above sketch.
/
OE Representative Date
~
l Repair Acceptable /
OE Representative Date Associated Document MAI /
(as applicable) Cognizant Engineer Date SDSP-13.1 /
Cognizant Engineer .Date
,, MAI /
Cognizant Engineer Date 575 Ne(s). -
Remarks:
All repairs were performed in accordance with /
applicable MAI and or standard practice Craft Foreman Date and/or approved varian'es. c I
All associated documentation is co: plete and Cognizant Engineer Date acceptable 0414o
4
.. Pago I4 BP SMMI-5.3 A AMACHMDC A REV 0001 CIS C0hTAINMENT INERTING SYSTEM (PRIMARY CONTAINMENT SYSTEM)
CRDRL C0tfTROL ROD DRIVE RETURN LINE CRDSDH C0f(IROL ROD DRIVE SCRAM DISCHARGE HEADER CS CORE SPRAY CSSRH CSS RING HEADER -
DCA DRYWELL CON'IROL AIR DSH DRYWELL SPRAY HEADER (EXTENSION OF RHR SYSTEM)
DTP DRYWELL AND TORUS PURGE (PRIMARY CotCAINMEhT HVAC)
EECW EMERGENCY EQUIPMD!T COOLING WATER (INCL. PARTS OF RCW)
TPC FUEL POOL COOLING FW TEEDWATER HPCI HIGH PRESSURE COOLANT INJECION t MS MAIN STEAM (MSSRV, MS-DR)
PSC/ECCS,. TORUS RING HEADER RBCCW REACTOR BUILDING CLOSED COOLING WATER ,
. RCIC REACTOR CORE ISOLATION C00LAhT K07 REACTOR DRAINS AND VENTS RECIRC RECIRCULATION i RER RESIDUAL HEAT REMOVAL - l RHRSW RESIDUAL HEAT REMOVAL SERVICE WATER RWCU REACOR WATER CLEAh"JP RWSPD RADWASTE SUMP PUMP DISCHARGE
- SGT STANDBY GAS TREATNDC SLC STANDBY LIQUID C0hTROL TSH 'II)RUS SPRAY HEADER (EXTDISION OT RHR SYSTEM)
RCW RAW COOLING WATER CBAC COh"IROL BAY HVAC !
CO:TP CO: STORAGE, TIRE PROTECION l PDJET SECONDARY CONTAINMDC PDIETRATIONS j PUMP.STA PUMPING STATIONS (RHRSV AND EECW) ]
i I
4 LAST PACE
. .' ,s ADO C 7 1985 i
- "'* Tennessee Valley Authority cPage 1
.a form BF-5
- g. . Ilrowns Terry Nuclear Plant BT 1.2 & 2.3 s
I. STANDARD PRACTICE /PERMANEhi INS 3UC 20N CHANCE INTOPJtATION '
Standard Practice / Instruction MMI-99 Unit 1.2.3 -
Title INS *RUC-IONS FOR THE IMPLEMENTATION OT NRC IE BULLE!!N 79-la.
PERw.AND.T INSTRUCTION CHANGE ,
- Ye s __ (Workplan No. ) No, y f.. _ ,,,,,Is this a workplan-initiated change?
,qg, _ ,Does this implement, delete. or change to a vendor manual? . . .
..u....-.-.a reference Yes I No
.If yes, DCC Vendor Manual Coordinator signature is required. ' .'
NA /
. ' ~
Vencor Manual Coorcinator Date
$.2- Yes I No
!.4"Does this implement. delete.,or change an SI data cover sheet?
M ,If' yes, Planning and Schedul'ing supervisor signature is required. ,
NA /
4" z.~u .4.. ~. u .. u. . .'. . . Date Planning & Schaculing .
~' 2- .
.. l
.If. ite,n added, deleted, or frequency changed on IE (or CSSC) PM requirements. '-
,,, cu plete BT-111. - -
~.. .. ._ .
]*I STANDARD PRAC*ICI CHANOE Is QA revtew requirec prior to issuance? Yes No Is PCRC review required prior to issuance? Yes No g
(See attach =ent 2 of standard practice BT-1.2.) -
Quality Assurance program requirements have7een pfoperly ancl'u3ET-'
- l. --
/
Quality Assugnce Supv._ Date
..~ .
Intent change in procedural' detail of TSAR or other licensing document?
Yes No X M... ,, .,
[~ Yes ' ' No I
" New If yes Instructicn?
to either of these questions, a USQD is required.
i [. Is this change in response to an LER. II bulletin. NRC inspection report,
" '"" ' " "" ' '""'"'No under reason for revision Yes X l_* Was this change made to meet an NRC cc=mitnent? _____Yes No I If yes to either question, refer to BT-2.3 for proper icentification of the change l.. Does,this revision implement a source cocument? ,,_, ,,,Y e s No X_
L If yes, attach form BT-4 Due cate of form BT-4 is . N_A
~
Tire Protection Syste= involved? Security System involved?
NA / NA /
~~
' Fire Protect.an Engineer
- Date Puolte Safety Supervisor Date a
- Billy Caldwell. Emerv Thomas vc5 / h1/ 'ine/?co$
Sutmtttec ey Sect;on Date Pnone Nu=cer
~
- Reviewee ey -
See BT-1.2 for standard practices.
~~ ~~ [7/ Q_
Responsible Section Supv. 'Date
/ erp f.y
- t plant peratton Theree".ry?YhW.:tants PCRC Chair:an J h9tays Date are conststent vatn safe,
/,,,],ff/CH
- 6$ effsetep/u/4 Plant Manager
/h5 Date 850912-15 DP6C'J Job he.
4
- Revision ..
t
{ .
A
'l Page 1 BF MMI-99
.f' TENNESSEE VALLEY AUTHORITY g,, g ;
, BROWS FERRY NUCLEAR PLANT
~
MECHANICAL MAINTENANCE INSTRUCTION 99
'. Instructions for the Implementation of ,
NRC IE Bulletin 79-14 Units 1. 2. and 3 1.0 Scope
, 1.1 Perform inspections as outlined in this guideline on safety-related j
f ,' . . - . and seismic category I piping 2-1/2 inches in diameter and greater k - -
and associated branch lines as determined by dynamic analysis from
, ~,, . .
. Office of Engineering. See Attachment IV for systems under this scope. Note that the majority of the inspections under this scope haie been performed. A program document detailing the overall 79-14 2 - . . .
, program is being-written-(including complete-status Mnd-wilt-be
, official before U2 startup. ,,
1.2 Prior to 29 April 1985 the inspections were performed by Nuclear !
T.
Power (NUC PR). Beginning 29 April 1985 the remaining inspections i
j , , ,
and necessary reinspections (in accordance with Attachment III) will !
be performed by engineers and engineering associates hereafter called OE Engineers. frem the Office of Engineering (CE) or their ,
designees. If =edifications are not required, the data may be used j ,
to as-construct drawings or"at a later date using an acceptable 1
program. If modifications are required, as-constructed drawings l will be made-during the normal NUC PR modification program.
L - - - -
- 2.0 References .
I 2.1 Nuclear' Regulatory Cemission IE Bulletin 79-14 2.2 Brewns Ferry Engineering Project. Project Instruction 85 01 (EFEP PI I
. 85-01).
0052B l l
1
- p. .
Page 2 BT MMI-99 3.0 Prerecuisites W3Y15 m3
' Inspectors using this procedure must successfully complete the inspectors
~ '
training for this MMI. Course Code Q:N101.
4.0 'Precautlens -
,None
- 5.0 Preparatten for Maintenance
, ( -
None. If necessary, when deviations are identified, use BF-7.6 for r . .
.. initiating repairs.
6.0 TeWorarr Alterattens . . ..
~
None . .
7.C A1. ARA Considerattens Inspections in high radiation areas will be postponed until they can be
~ ~ "
made with minimum exposure-to personne1r L.
. 8.0 Instruettens ..
( 8.1 Detail Support Inspection
. 8.1.1 Inspect all pipe hangers and restraints for nonconformance to the latest as-constructed drawing. If as-censtructed drading
~l is not available, the latest design drawing shall be used.
. L l
-In some cases-it may be necessary to detail'the confighration' on a sketch when no drawing exists. Inspection should j include all components of the hanger / restraint and all l attact. ment welds. Maintenance Requests (MR) may be written
. l to return minor deviations (i.e., missing or loose nuts, bent l
, rods, etc.) to the as-constructed condition, if possible.
1 prior to ce.epletion of data sheet 1 per cogni: ant engineer's ;
l discretion. The as-found condition will be depicted on the '
data sheet I with the deviations and corrective action acted
- l en the MR. A copy of the completed MR shall be submitted
- l
.. along with the data sheet I to the OE inspection coordinator. {
l
. 'm ,
' s
- Page 3 BF EE.99 8.1.2 WOY 1 5 1535 Any deviations or additions to the drawing not repaired as r-stated above should be submitted with a data sheet 1 to the OE inspection coordinator for evaluation purposes. ' Be sure ,
~.
to include all additional information needed to support a structural analysis of the support such as moment asus, weld
,_ sizes, angle brace dimensions, etc. All non-standart components (not furnished by a 'sanufacturer such at Geinnell.
etc.), shall be detailed on the drawing or on an attarted sketch. All attached sketches shall be cross-refereseed to the original design drawing and uniquely identitled. (See
,, Attachment II for a list of information needed to detail supports.) .
. ,8,1.3 In cases where typical supports are shown on the piping drawing or where extra supports are listalled and no detail ,
design drawing exists, a detail of each support shall be I prepared and a unique number assigned to provide traceability
, back to the configuration drawing.
8.2 Valve Inspection l '
8.2.1 Verify that all valves are installed as specified by tSe
[,
design drawings (oriertation of valve and operater) and b
record valve size 'sype, manuf acturer, model number and operator manufacturer on data sheet 2. Any deviations in .
L valve location and/or orientation shall be marked up cn the j .
, configuration drawing and submitted as part of the 1 4
configuration package.
l j .
. l
. I j g e
l 0052B
.. 's
.- Page a
,, BF MMI.99
{ , N0'i 1 5 13!5 f,
8.2.2 In most cases, the required information can be obtained from a tag on the valve bonnet, handle, etc. If a model no., !
~
figure no., or some other positive identification number is g.
} not available, record all information which can be obtained s
from the valve, which may require removing insulation from valve bodies.
8.3 Centiguration Inspection, 8.3.1 Configuration inspection shall include verifying pipe '
routing, and hanger location, which may require removing
. insulation from supports and piping. This will .entall taking actual measurements of gipe runs, branch line locations, valve , ,
location, hanger locations, and other componentE and . ,
ce= paring them to the design drawing. (See Attachment II for a list of information needed on configuration drawings.) ,
Those measurements corresponding to the dimensions of design drawing should be circled with any deviations shown accordingly. Configuration shall also include identification of equipment (name tag data if available) in the system and type of connection to pipe.(flanged, screwed, etc).
a be.
'D
l
- ** . i
-' Page 5 l
. BF MI199 l
~ -
NOV 13 m 8.3.2 Configuration should be marted up on full :alze knas, if i
possible. In cases where the piping drawing is ammasted, it l i
stay be necessary to turn in several diff erent dristas, each l 1
- - - with a different line segment on it. In order to yeperly l
[ analyze piping configuration, whole line segments sh11 be ,
e .
completed (i.e., from one component to another) p:irtto -
submitting for review. Once a deviation has bees Ekasvered
"
- and the OE inspection coordinator notified, the amaturatica I
. 1 I
. of that line segment shall be completed and sentam the OE program coordinator in order to provide a pro =pt:mquInse (in
,,, accordance with BTEP-PI 85-01) to the deviation.
. 8.3.3 Af ter cc=pletion of a drawing or line segment (s). ne marked i up drawing, along with data sheet 3. should be taamd.into ,
l the OE Inspection coordinator. Each inspector it s'airponsible for recogni:ing potentially significant problems amt j
. . s 1 reporting these to the CE Inspection coordinator;ramptly. l
{ ,
8.4 Penetration Clearance Inspection '
- L 8.4.1 Verify that piping which penetrates walls or flows (e.rcept
= .. - - _ -
grating areas, unless otherwise noted) has clearance all around, unless otherwise specified on the drawisq;. On a sketch on the data sheet 4 record penetration ehrunces around the pipe with orientation denoted and suhic.as part of the configuration package. All penetrations staald be
,' uniquely identifled on the configuration drawinghbrtch and on the data sheet a. Also, provide locations f ar aI1'.
penetrations on the configuration drawing /stetek.
O WS .e e. .g d
. (- .
l '
o page 6 SF MMI-99 f N OV 1 5 19!5 8.4.2 Drywell penetrations with welded seals are excluded froa this clearance inspection.
'N.T All piping and penetrations through secondary containment must be inspected through an isolation device (e.g..' valve, loop seal, etc.).
8.5, Spring Hanger Deta
'To minimite the replacement of spring hangers due to revised
- a. .--
leading from analysis, the following additional informs;1on.will be required on all spring hangers and should be recorded on data sheet 1.
- 1. Vendor. .
2.- Size and type canister. (If ID tag is missing, record
, dimensions of canister.)
- 3. Travel limit.
"4'. ' -
Lead setting and condition (i.e., hot or cola). (If scale ;
e ,
is ah s_ing, record _dinen.slon.of_ opening and.inticator -
position.) -
- 5. Dynamic travel limit. If less than 1/4". from top or
,', bottom of spring scale, report as a support der!ation to the OE inspection coordinator.
$ 9.0 Testine L -
None.
10.0 Feturn to service None.
g 11.0 '
Attach ents -
I. 11.1 Attachment I.-Review procedure for IE Bulletin 79 14 Deristions 11.2 Attachment II--Inspection Items List y ,
11.3, Attachment !*:--Method of Handling Additional'Information Requests to Support phase II of IE tu11etin,79-14 -.
11.s Attachment IV.-List,ing of Systems Under Scope of IE Eulletin 79 14 D
m u.. . . , , ..
{ 00523 l
. . . Y -
t
, , age 7 r
- BT MF.I.99
- MECHANICAL MAINTENANCE INSTRUCTION 99 f . Instructions for the Implementation of
{' '
WRC IE Bulletin 79-14 Units 1. 2, and 3 Data Cover Sheet Unit
' Systes . *
. ' 1. 'The attached packagw contains data sheet (s) 1. .
data sheet (s) 2. * *
, data sheet (s) 3.
l data sheet (s) 4 l
2 ., *:hns'ure attests that I understand the scope and purpose of this it....ction and that, to the best of
- icy knowledge, it was properly .
performed in accordance with instructions in that: the recording.
f' reduction, and evaluation of data were complete and correct; acceptance criteria were met or justification for exceptions providedt deficiencies j
were evaluated and dispositionedt and in:truction was funy ccuplete
[ except as noted. (1)
- g. . ,
, /
[
Cogni: ant Reviewer Date I , (Inspection Coordinator or his designee)
] 3. Reviewed by: /
OE Design Section Supervisor Date J
REF. ARES:
} .
l -
l 1 _ ,
q .
s L
.. m tenusuons i,e to ,e msee uun, the ,uidenne, in ,,,, ,, ,,.m.
.s:,
I
.. f' .
\
Page 8 BF MMI-99 I ~
TENNESSEE VALLEY AUTHORITY NOV 15 E5 BRO'JNS FERRY NUCLEAR PLANT r.
MECHANICAL MAINTENANCE INSTRUCTION 99
. Instructions for the Implementation of
- ~ ~
.. .~. Units 1, 7, and 3 L.
- Data Sheet 1
~
PIPE SUPPORT CHECKLIST System:
. Unit:
- Configuration drawing / sketch No.: .
- 1. Pipe support No.- .
~
Pipe support drawing No. _.
(1 '
- 2. Vendor .
(15
'3. Siz .e and type canister .
- 4. Travel limits (inches) * ~
= i.. 5. Load setting and condition (lbs.) - .
,,. . (i.e., het or cold)
L C)6. Dynamic travel limit (inches) ,
,, (If less than 1/4" from top or bottom of spring scale, report as a sup deviation.) -
- 7. Support is installed per design drawing (s). (Yes or No) .
If no, indicate deviations on design drawings.
NOTE: If snubber, check snubber size (s). (If ID tag is missing, notify inspection coordinator.)
If rigid strut, check strut size (s).
If structural frame, check load carrying member size (s).
- 8. Is there loac carrying attachment (s) welded to pipe (Yes or No) .
If yes, does the weld (s) conform to design drawings: ,
(NOTE - Welds other.than fillet welds may not be verifiable)
/ .
Inspected By Date
/ .
. Checked By Date (1) .Information needed for spring hangers only. If identification tag or lost scale missing, contact Inspection coordinator for determination.
. i ..
I a
~
. ~. Page 9
{ BF MMI-93
'E TENNESSEE VALLEY AUTHORI*Y BROWS FERRY NUCLEAR PLANT ' l MECHANICAL MAINTENANCE INSTRUCTION 99
' Instructions for the Implementation of l
. NRC IE Bulletin 79-14 '
Data Sheet 2 VALVE CHECKLIST f- System: '
[' .. -
Unit:
. Co:;. figuration drawing / sketch No!:
~
- __
_ 1. Valve tag No. -
Valve type (gate, globe, check, etc.) .
{ , Valve manufacturer -
Valve model No. ~
" 1
- 2. I*,
valve location.. correct?
~ (Yes or No). l
\
(If no, location on drawing.)
3.
If the valve has an extended operator, is the orientation of the operator correct (Yes.or No) g . .
[ (If no, indicate orientation on configuration drawing / sketch or this data sheet.)
- s. . -
- ,_,4. Operator manufacturer Operator model No.
/
Inspected By Date ..
l
/
, Chect.ed By Date
\
\
' i L
d
. r
. s
- - Page 20 BT KM1-99 1
TENNESSEE VALLEY AUTHORITY '
BROWNS FERRY NUCLEAR PLihT
~
E M'CHANICAL MAINTENANCE INSTRUCTION 99
~ Instructions for the Implementation of NRC IE Bulletin 79-14 Data Sheet 3 DRAWING CONFIGURATION CHECKLIST.
System:
- Unit:
,, Drawing / sketch No.: -
1.
Does the general configuration of the piping system, including support location and embedments (not covered by Bulletin 79-02), match with the drawing? (Yes or No) --
~~ ~~
_ _ ' 2 .~ Show. deviations on marked-up drawias/sietch.
i l .
i l
1.
L _ . .
r /
Inspected by Date
/
Cheet.ec By Date D .
( 0052B *
. . c.
. '. . Page 11 f~ BT MMI-99 TENNESSEE VALLEY AUTHORITY , h'0V 15 yg BRO'='NS FERRY NUCLEAR PLANT MECHANICAL MAINTENANCE INSTRflCTION 99 .
~ Instructions for the Implementation of I,. NRC IE Bulletin 79-14
. . Units 1, 2, and 3 Data Sheet 4
. FLOOR AND WALL PENETRATIONS CHECKLIST t,
System:
Unit: .
' Configuration drawing / sketch No.:
- Penetration Nu=ber Provide a sketch' below showing penetration clearance around the, pipe and the ,
. orientation.
NOTE:
- If the penetration has lugs or other pipe support features, then also submit as a support on a data sheet 1.
l -
/
Inspected By Date
/
. Check.ed By Date m
91 W.. .
. .g
ATTACHMENT I Page 12 BF MMI-99 I NOV15 ms !
TENNESSEE VALLEY AUTHORITY l BRO'JNS FERRY NUCLEAR PLANT MECHANICAL MAINTENANCE INSTRUCTION 99 REVIE'd PROCEDURE FOR IE BULLETIN 79-14 DEVIATIONS 1.0 ON-SITE REVIEV 1.1 Initial Screenint by OE Inspection Coordinator-All deviations are initially screened by the OE Inspection -
Coordinator or designee for dispositioning into one of two general categories:
l *
.. .A. Pipe support deviations which are potentially adequate for
_c design loading.
- b. All other types of deviations.
1.2 OE Engi'neer's Review of Potentially Acceptable Deviations
[ All pipe-support-deviations-which are potenti~ary adbq'date Tor design loading will be evaluated 4 y an OE engineer in accordance -
with attachment.7C from the October 4, 1979, memorandum from Roy H.
Dunham'to Those listed.."Browns Ferry Nuclear Plant - Program to Resolve NRC IE'Bulletin 79-14," (CEB 791004 018).
m If an OE engineer's evaluation shows the deviation to be adequate 1 as-is to carry design loads, his reco=endation shall be one of the.
following:
lL -
- a. Return to as-designed.
- b. Use-as-is.
- c. Use-as-is with minor modifications.
Use-as-is with or without minor modifications will not require an imediate drawing change.
All evaluations performed by an OE engineer will be forwarded to the
-0E Dw:,ign Section for checking.
, i If an OE engineer's evaluation shows the deviation to be'not adequate to carry design loads or if an OE engineer decides that he {
can not eviluate the deviation, he shall imediately notify the OE
, Inspection
- Coordinator. If the OE Inspection Coordinator determines he can not evaluate the deviation, he shall notify the OE Program Coordinator.
Page 13 y
BF MMI-99 ATTACHMENT I (Continued) 15 E
. l.0 ON-SITE REVIEV (Continued)
,1.3 Further Review of Unresolved Deviations I The modifications will be documented on a data sheet 1 and submitted
~
- The OE Design Section will determine whether to use the support as submitted on the data sheet 1. return to the original as-designed f.., condition, or to completely redesign the support. The OE Design Section will issue fincl documentation / design dwss.
All deviations which are judged to be not_ adequate by an OE engineer rust be transmitted to the OE Program Coordinator by.the OE
. Inspection Coordinator. .
1.4 All completed packages / inspection data shall be transmitted to the OE Program Coordinator by the OE Inspection Coordinator and filed in
' fire proof storage.
2.0 CE REVIEW .
See.BFEP B5-01 g
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.s Attachment II Page 14
. .BF MMI-99 l f
. - TENNESSEE VALLEY AUTHORITY
, BRO'='NS FERRY NUCLEAR PLANT -
MECHANICAL MAINTENANCE INSTRUCTION 99 !
~*
.. . j INSPECTION IIEMS LIST ;
Items to be addressed on the configuration drawings.
~ ~
Piping - schedule, material, and grade ]
Fitting - schedule, material, and' grade ] where easily available Fittings (elbows, 0 0 long or short, 90 ; 45 , etc; tees, reducers,
, weldolets, etc.)
. Diameter (nominal pipe) ~
Pipe elevation and location relative to the building Insulation - thickness, type,_ and location _
Branch line location and size (all sizes) -
~
. . Valve. operator orientation * .
- Orifice flange locations Hanger, restraint and snubber locations relative to the pipe In-line. equipment location (valves, filters, pumps, etc.)
r.
, Ite'ms to be addressed on the pipe support drawing / sketch: . .
Luts -
Vendor & vender part'no.
. Size (length, width, height)'
Weld detail (type, size, length, and location) (where possible)
Hole, (type, size, and edge distances) documentation for
, this flame cut or drilled; may exist, if so, retrieve and include in package. .
Lug to clamp clearances Lug to pipe clearances Pins / Belts --- -
Diameter and length Cc6dition
- Thread engagement I Pin retaining device - eetter key, E-ring, et:.
Bolt locking device - jam nut, thread staking, etc.
Clares l
~ Vender or field fabricated ,
If field fabr'icated: Complete detailing dimensions including .
- Stock size Bolt sizes l i
, Hole size, type, spacing, and edge distance
-l Spacers l
Clearance If vender fabricated: vendor and vendor part number.
Da es. .
.g
[ 0052B
,~
Attachment II (Continued) Pap 15
{ Inspection Items List (Continued) BF-MMI-99 NOV 1 5 ICI e
Items to be addressed on the pipe support drawing / sketch: (Continued)
U-Bolts Nuts (number, location, condition)
'-Material (carbon steel, stainless steel, if possible)
Clearance and location
~
Orientation Deformation in U-bolt, steel, or pipe Rods End attachments, vendor and vendor part number or complete details Rod type (welded eye, threaded, etc.)
Rod size "
~
' Jamnuts '
Staking where applicable Thread engagement - .
Turnbuckles, vendor and vendor part number or complete details
[ Bent rods.
Support loaded or unloaded g , What rod attachments attached to (plate, angle, etc.) . .
Concrete Attachment Plates -
Complete details, size, thickness, orientation (N-S, E-V, or top, bottom, etc.)
Attachment location on plate Centerline lo:stion of concrete anchors relative to plate edges or I centerlines ~
Concrete anchors (quantity, si:e, and type) ,
' Is plate embedded, surface mounted, or grouted Condition of plate (bent, gap, concrete failed, etc.) ;
l Weld detail for attachment (type, si:e, length, and location)
Supelementarv Steel Framine !
)
Show location and detail of pipe supports on the framing member Show stiffener location and detail on supplementary steel Si:e and orientation of members, i.e., V6 x 12 with web vertical.
CB x 11.5 with flanges her?: ental, etc. (sketch if description difficult) ~
Complete detailing information as identified / requested by OE.
l 1
hus l.
~
Att'achment II (Continued) Page 16
(
BT MMI-99 NOY 15 1215 Inspection Items List (Continued)
Items to be addressed on the pipe support drawing / sketch: (Continued)
Structural Steel Pipint Supports (Other than rods, snubbers. or springs)
Configuration of support Member sizes (L4x4xl/4, C6x8.2, etc.)
Member length and orientation
[
Veld type, size, length, and location (using N-VT-6 under paint on tillet
__ welds)
Belt sizes and thread engagement. Threads included or excluded from I e.
shear plane (if possible)
Holes (bolt holes and other):
edge distance drilled or flame cut, diameter or size,
,Does the support appear damaged in any way If so,, describe.
Snubbers and Struts Yendor & vendor part number
[ Pin-to-pin dimension Size and type (HSSA-10, hydraulic, etc.)
Orientation (such as: included sngle) - -
. ~
Lug details (size, length, location)
Lug welds (type, size, length, and location)
Clearances around pipe & orientation NOTE: Circle existing accurate dimensions, cross out inaccurate dimensions and write clearly the actual measurements. ,
~
\
Make all copies clear and complete. If original drawing is illegible, make a new sketch even if it appears to be as designed. '
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ATTACHMENT III Page 17 l r , BF MMI-99 l
' ~
.J . TENNESSEE VALLEY AUTHORITY
, BROWNS FERRY NUCLEAR PLANT MECHANICAL MAINTENANCE INSTRUCTION 99 j
~ '
l Method of Handling Additional Information Requests to Support Phase II of IE Bulletin 79-14 '
, Euring the analysis phase, additional information may be required by CE in f' ~
order to complete the analysis based on the inspection data furnished under Phase I. The following are guidelines to support the analysis effort.
)
1
copies of Phase I drawings, sketches, data sheets, etc. by system and unit with specific questions noted in red ink on data sheets and
, r.reen ink on configuration drawings. -
)
- 2. Inspectors are to answer all questions on Phase I drawings,
,4 ketches, data sheets, etc., update field copies to reflect all v'
, the new information and return to the OE Inspection co,ordinator,.
- 3. The OE Inspection coordinator will ensure'that the Phase T Wa71ngs7 sketches, dar.a sheet, etc. are feviewed to ensure that all questions
. have been answered, then logged and returned to the OE program coordinator.
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ATTACHMENT IV _Page is {
BT MMI-ty NOV 15 G5 LIST OF SYSTEMS UNDER SCOPE OF IE BULLETIN 79-14
. .. ALL UNITS 76 CIS CONTAINMENT INERTING SYSTEM (PRIMARY CONTAINMENT SYSTEr) 85 CRDRL CONTROL ROD DRIVE RETURN LINE 85 CRDSDH CONTROL ROD DRIVE SCRAM DISCHARGE HEADER 75 CS CORE SPPlY 07 CSSRH CSS RING HEADER 32 DCA DRYVELL CONTROL AIR
(
74 DSH DRYVELL SPRAY HEADER (EXTENSION OF RHR SYSTEM)
. 64 DTP DRYVELL AND TORUS PURGE (PRIHARY CONTAINME!!T HVAC) l
. _ 67 EECV l EMERGENCT EOUIPME?lT COOLING VATER (INCL. PARTS OF RCV) '
. 78 FPC FUEL P0OL COOLING _
HIGH PRESSURE C00LAtlT INJECTION 01 MS MAIN STEAM (MSSRV, MS-DR) ~ ~
__ 73 PSC/ECCS TORUS RING HEADER -
- 70 RBCCV *~
REACTOR BUILDING CLOSED COOLING VATER
. 71 RCIC REACTOR CORE ISOLATION COOLANT 10 RDV REACTOR DRAINS AND VENTS 68 RECIRC RECIRCULATION 74 RHR RESIDUAL HEAT REMOVAL l
_ 23 RHRSV RESIDUAL HEAT REMOVAL SERVICE VATER 69 RUCD RE_ ACTOR VATER CLEANUP 77 RVSPD RADVASTE SUMP PUMP DISCHARGE 65 SGT STANDBY GAS TREATMENT 1
63 SLC STANDBY LIOUID co?! TROL L . 74 TSH TORUS SPRAY HEADER f EITENSION OF RHP EYSTEMI 74 RF' RAV COOLING VATER IB _
Diesel Generator Fuel O!1 86 D-GAS
' Diesel Generator Intake / Exhaust COMMON fUNITS 1, 7 & 3) 31 CBAC CONTR0!. BAY HVAC 39 C0 yFP C0 STORAGE, FIRE PROTECTION
__ - PENET SECONDARY C0fiTAINMEf!T PENETRATIONS
_ 73,67 PUvP STA PUMPING STATIONS (RHRSV AtlD EECV) 00$2B
i .. 1
. QUALITY INFORMATION REQUEST / RELEASE (OIR)
- DIVISION OF NUCLEAR E.NGINEERING
. ( INTERNAL USE ONLY l
"'"' ^741 '860813 002 TO DOCUMENT NUMBER 7~ M. BEOTHERS alRcca ss o s s FROM PAGE 1 OF 1.
7'( C,E w r /'
TYPE OF DOCUMENT DATE y,;;q j 3 jggg _
l l REQUEST NEED OATE
_________ ___- ---------------- PLANT AND uNir RELEASE R E F. OI R BfA/P uf , # 2, M3 REFERENCED DOCUMENTS AVAILABLE IN ONE OF THE RIMS SYSTEMS ATTACHMENT TO THIS OIR DOCUMENT IDENTIFYtNG NUMBER DOCUMENT ATTACHMENT NUMBER -
~ ~
I. M C/he feetr1 B22. 94 0+0S cof C2bH Assqtysts g 4.C. A4WA u To SFrP GlL/DElINES
-. f/4ES .
SUEUECT COV7~A'el. Aos M/VE HYDK4U4/C. (CKON,) Syf fdM ~ s9NA4 y.sts G UIDEL/NCS SYSTEMS AFFECTED copyAo4jQc4 AglVf UNID / SYSTEM ID SS QUALITY INFORMATION REQUESTED / RELEASED EEFEAENCE f ffAAttM/7YED A 567 CF ANAL YS/5 G t//DE Z/NES FCA THE ccMicel Eao Dg/yE' jfyDfaus /c (c. cow) S Vsrs s*T As
&tlALtry DGS/ gat /NFescir:4yjes/, s*FEC/F/ CAL L P t"HESE G t//oEL/NES k/ERf DEVELoffb To Su/'P LEn1Enr NefmAl ANAL ht/s C2/7Z~L/R DL/6 70 7~df nan 7TP/ CAL DE5/GN Cf C2D/r' /AlffAr AND W/THDRAWAL sP// %'G.
W/rH YsVE t$sa A Alcc c/~ NEP .5. S, THE qual / 7Y Ms/s.sy jNfonnt4 7~/oM /77Etr70 /$ ^/s 4c aggg AA/ gccspygg g syy g gsy ,$
CF CCWVf>'A A/CE CF QtlAL/TY /4/FoAsr747/cu. 7/EREFGCE TW/s (Q.CR /$ 8E/N d /S$ t/Eb 7~c ACM/ EVE C osr?P 4/ A v e f Ly/ rra per g.3 fox rxe sue.reer intretir7Aric,v, YNC CEbH ANA L. h1/$ 6 W/.CSL/Nf5 #4f A77~A cdFD in Y*w $ qZR caver $HEGe# /H YWE/Z E47/.CE TY a.// 77f NO C#4NGE" TO _
THE '75CHN/l'A4 CC W 7"E N r sC*Cc/r? THE" EEM .i /774~'r7 6. . - . ...
PREPARED R VIEWED (RE EASES ONLYI MW 5'/ 9 APPROVED'kBR ANCH CHIEF / PROJECT ENGINEER)
TVA 10879 iDNE4 96) 5 cc ( Attachment h RIMS, SL 26 C.K L--
AttachmenttoDIRCEB86016 BROWNS FERRY NUCLEAR PLANT GUIDELINES FOR ANALYSIS OF THE CONTROL ROD DRIVE HYDRAULIC SYSTEM MARCH 1986 Introdug!. ion and Scope .
These guidelines define an acceptable basis for analytical qualification of the CRDH system for each BFN unit, consisting Hof 185 1-inch-diameter insert lines and 185 3/4-inch withdrawal lines as well as the supports for those lines.
The basic requirements for this analysis are contained in the BFN criteria for analysis of as-built piping systems, BFN-50-D707. Special requirements for the CRDH system are defined in these guidelines.
The CRDH system was originally routed and supported for deadweight and
- thermal expansion by Reactor Controls, Incorporated. Then, additional -
horizontal seismic restraints were added by TVA in accordance with the BFN criteria for seismic qualificationrfo field-routed piping 2 inches and smaller, BFN-50-712.
Criteria in of support loads, as described BFN-50-712 remains valid for determination BFN-50-D707. '-
' As previously noted, criteria BFN-50-712 may be applied to define support loads and BFN-50-D707 is applicable for support evalua tions. Should.the requirements of BFN-50-712 prove too conservative, these guidelines provide relief in the form of an acceptable rigorous analysis technique which'is also permitted by BFN-50-D707. Either approach is satisfactory. '
The choice is dependent on the acceptable degree of conservatism.
Loads for Analvsis '
- 1. Deadweight Loads a.
1-inch pipe and contents = 2.5 pounds per foot. '
1
- b. 3/4-inch pipe and contents = 1.7 pounds per foot.
c.
3000 pound fittings and contents = vendor specified weights 2.
Design Pressure and Temperature (reference BFN FSAR Figure 3.4-8c) ,
- a. Insert lines = 1750 psig at 150 0F *
- b. Withdrawal lines = 1750 psig at 150 0F
- 3. Normal Scra= (cads (Envelope of Nor=al Operation and Scram Loads)
- a. Pipe pressure = 1510 psig - insert lines 1250 psig - withdrawal lines
- b. Temperature of CRDH insert lines (1) Inside dryvell = 150 CF from CRD bousing on reactor vessel to dryvell penetration.
036084.02
_ . , . .~ ~
A r
(2) Outside dryvell = 150 0F at dryvell penetration and 95 0F at control unit with 1 linear temperature distribution between or a more accurate distribution based on steady-state heat conduction analysis and 95 0F air temperature.
figure 3.4-10)
(1) Inside dryvell = 280 0F from CRD housing on reactor vessel to dryvell penetration.
(2) Outside dryvell = 280 0F f rom dryvell penetration to hydraulic
, , control unit.
- d. Temperature of dryvell and internal structure excluding reactor vessel support = 150 0F .
- e. Temperature of reactor vessel support r
(1) Reactor vessel skirt = 500 0F (2) Concrete pedestal = 150 0F s f. Temperature of structure outside dryvell = 70 0F
- g. Consider relative thermal expansion effects batveen the CRDH attachment / support points as well as the thermal expansion of the CRDE piping which is 304 stainless steel. The reference (installation) temperature is 70 CF. -
4 Abnormal Scram Loads (Envelope of Multiple Scram and Leaking Scram Valve Loads)
Same as for normal scram loads except withdraval lines, inside and outside dryvell, are at 400 0F (ref eren:e BFN FSAR, figure 3.4-10).
- a. Pipe pressure = same as for normal scram.
- b. Temperature of CRDH insert lines -
(1) Inside dryvell = 280 CF f rom CRD housing on ricetor vessel to ~
dryvell penetration.
(2) Outs,ide dryvell = 280 CF at dryvell penetration and 105 0F at hydraulic control unit with linear temperature distribution between or a more accurate distribution based on steady-state .
heat conduction analysis and 105 0F air temperature.
- c. Temperature of CRDH vithdrawal lines = same as for normal scrac.
_e. . n1 Ansa.no
e
- d. Temperature of dryvell and internal structure except reactor vessel sup rt = 280 CF
- e. Temperature of reactor "vessel support Reactor vessel skirt = 500 0F Concrete pedestal = 280 0F
- f. Temperature of structure outside dryvell = 70 0F
- g. Consider relative thermal expansion ef fects between the' CRDH attachment / support points as'vell as the thermal expansion of the CRDH piping. (Potential interf erences with concrete sleeves are disregarded because that concern has been addressed separately.)
- 6. Seismic Loads
- a. Seismic response spectra (SSE curves and digitized data attached)
- The vertical spectra for elevation 519 feet is applicable for all CRDH support points inside and outside the dryvell. The horizontal spectra (NS and EW) for reactor building elevation 593 feet are applicable for all CRDH support points outside the dryvell, including the dryvell penetration. The horizontal spectra (NS and
' EW) for RPV shield vall elevation 587.33 feet are applicable for all CRDR support attachme'nts inside the dryvell. Spectra given are for SSE loading with 5 percent damping. OBE seismic responses vill be 2/3 of the SSE responses. Broadened spectra vill be applied in all caseo. This data is taken f rom CEB Report CEE-85-46 (B41 851112 04 8) .
- b. Seismic anchor point movements (1) Inside the dryvell The relative seismic movements of the attachment and support points inside the dryvell are negligible with respect to the CRDB piping.
(2) Outside the dryvell During an OBE, the dryvell penetration moves horizontally ..
10.04 inch in the N-S or E-V direction. Simultaneously, the reactor building moves horizontally 10.05 inch in the N-S or
- E-V direction. The SEE displacements are twice these values.
Thus, a relative horizontal covement of 0.09 inch can occur in either direction during an OBE and 0.18 inch relative horizontal covement is possible during an SSE. Relative vertical movement is negligible. This data is taken frem CEB Report CEB-80-41 (CEB 810619 006).
036084.02
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Rigorous Analysis Technique 1
- 1. The basic rigorous analysis techniques are described in BFN-50-D707 and l the BFN rigorous analysis bandbook. These guidelines supplement those documents. In cases of conflict, these guidelines govern for analysis of the CRDH system only. '
- 2. Perform rigorous analysis of typical and vorst case CRDH pi' ~. Model supports carefully considering their intended purpose, noting that a horizontal support is not required at each vertical support (reference !
BFN-50-712 ) . The following support modeling assumptions are acceptable:
- a. Assume the supports to be rigid for deadweight analysis, but apply a 20 percent uncertainty factor to deadweight pipe stress.
- b. Support flexibility may be considered for thermal expansion analysis. Model the effective stiffnesses in a manner-appropriately justified in the analysis report.
c.
Consider supports to be rigid for seismic analysis if, and only if, support displacements under SSE loading from all attached pipes are less than 1/8 inch. Otherwise, model their' effective stiffnesser in a manner appropriately justified in the analysis report.
s
- 3. The load cases for analysis are:
Case Load Cembination Normal Deadweight + Nor=al Operation
- Upset 1 Deadweight + Abnor=al Scram Upset 2 Deadveight + Normal Scram + OBE Emergency 1 Deadweight + Normal Scram + SSE !
Emergency 2 Deadweight + Post-LOCA
- Analysis not required--normal condition inveloped by upset conditions.
4 A11ovable pipe, support, and anchorage stresses are in accordance with 1 BFN-50-D707 except that an augmented clasa 2 piping fatigue analysis is l permitted, in lieu of satisfying the seconcary stress equations (10 and
- 11) of the ASMI piping code, as f ollows:
- a. Perfore an aug=ented f atigue analysis for the following cyclic life .
scenario: ..
(1) 700 cycles f rem a=bient (t= 70 0F, p = 0 psig) to normal scrad load condition and return to ambient.
(2) 100 cycles frem ambient to abnormal scram load condition and return to ambient.
(3) l 5 eyeles f rom a=bient to normal scram plus CBE load condition I and return to a=bient.
i i
4 . nienes, aa !
r i (4) 45 cycles f rom ambient to OBE load condition and return to !
ambient. I t
(5) 1 cycle f rom ambient to normal scram plus SSE load condition i j
and return to ambient.
(6) 9 cycles f rom ambient to SSE 'nad condition and return to ~~
ambient. ,
{
(7) I cycle from ambient to post-LOCA load condition and return to ambient.
l 1
- b. The augmented fatigue analysis procedure is based on report KFR-751 I
- - dated November 1982 (Microfiche TVA-F-A003986 CEB).. It defines a continuous allowable cycle stress curve which is consistent with the stepped curve given by the code equations. For stainless steel l
that continuous curve is defined as N = (281/iS)5, where N is the- i allevable number of stress cycles, i is the piping code stress ;
intensification factor, and S is the nominal stress range in ksi. l For each load condition, a nominal stress range is calculeted for l pressure, thermal expansion, and seismic loads as applicable to the condition. (Deadveight stress effects may be neglected.) A fatigue usage factor is calculated for each anticipated fatigue s cycle, and the results are added to obtain an overall usage factor for the entire scenario. For example, if the if value for the ambient to normal scram load condition is 50 ksi, the usage factor per cycle is 1/N = (50/281)5 = 0.00018, and for 1000 cycles it is 0.18.
- c. To be acceptacle for the cyclic life scenario defined above, the '
calculated usage factor at each location must be less than 0.67, t h ".s leaving at least 0.33 usage for unknown or unanticipated stress cycles. The two locations of highest usage and the corresponding usage factors must be identified in the analysis report for each analyzed line inside and outside the dryvell.
l
- 5. Calculate piping seismic inertial response by enveloping response spectra and rigid response results per the procedure of BFN-50-D707.
- 6. Cembine piping seismic inertial response and seismic anchor point movement effects by the square root of the sum of squares (SRSS) technique, except where otherwise permitted by the ASME piping code. '
- 7. Nozzle loads on the contain=ent penetration, CRD housing, and CRD control unit must be approved by CEB. ,
8.
Evaluate locaf attachments to all supports and all single pipe supports for maximum f6rces f rem the attached pipe utilizing nor=alization factors for each load case as defined by BFN-50-D707.
.g. . e,ca , ..
i
- 9. For f rame supports which restrain two or more pipes, apply the f ollowing analysis procedure as necessary to reduce conservatism in the overall seismic load: ,
a.
Determine the total inertial response load for each supported pipe, F,
T including the response spectra / rigid response screen but excluding seismic anchor movement ef fects.
I
- b. Determine the rigid response load for each supported pipe, FR '
)
c.
Define the overall inertial response load on the frame support as the larger of:
Fo=1 ((F R cts Fo= 11/21{F Tl + Th}
- d. t To evaluate the overall-inertial load, distribute the overall load in accordance pipe the load iswith the individual larger. of:maximum load values so that for each r
F=1FR or, F = 1 (FT /[lF 7l]Fo b
e.
Ce=bine seismic anchor movement and inertial response leads for each pipe by the SRSS technique.
- f. t Form total load cembinations by adding seismic, deadveight, and ther=al expansion loads f or each pipe in the normal fashion.
g.
Evaluate frame support for all defined load cases per BFN-50-D707 allovable stress criteria and the deflection requirement of 2c above, as applicable. ,
Alternative approaches for combination of multiple pipe dynamic loads at particular frame supports are acceptable as long as engineering justification is provided to demonstrate an adequate degree of conservatism.
i l
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. 036084.02
/
TENNESSEE VALLEY AUTHORITY 8/85 BROUNS FERRY NUCLEAR PL ANT REACTOR BUILoINC (HAIN CONCRETE)
SSE VERT (-2/3 HORIZ) ACCEL RESPONSE SPE g ELEVATION 519.00 (CROUNo) - N.P. 33
- g. , FIXE 0 OAMPINC RATIO o.05
- FICURE 2.I
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, FREQUENCY (HZ) ,
s TENNESSEE VALLEY AUTHORITY 8/8S BROUNS FERRY NUCLEAR PL ANT REACTOR BUILDINC (MAIN CONCRETE)
SSE VERT (-2/3 HORIZ) ACCEL RESPONSE SPE g ELEVATION S19.00 (CROUND) - N.P. 33 i g. FIXE 0 DAMPINC RAT (0 0.0S FICURE 2.1 O
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. FREQUENCY (HZ) ,
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gg g g 'q CIGIYIZED DATA FOR TENNESSEE VALLEY AUTHORITY 6/85 S20ADENEO 5PFCTRUM BROUN5 FERRY NUCLEAR PLANT REACTOR BUILDING (MAIN CONCRETE)
SSE VERT (=2/3 HORII) ACCEL RESPONSE SPE ,
ELEVATION 519.00 (GPOUND) - N.P. 33 FIXFD DAMPING RATIO 0.05
/ FIGURE 2.1 F 5'E 3U E N C Y ACCE LER ATION F RE QUE NC Y A CCE LE R ATI ON FREQUENCY ACCELERATION 1.0000
~
.2350 11.1000 .2394 22.5000 1472 1.2222 .2350 11.4003 .2264 22.5000 .1472 1.5000 .3388 11.5954 .2177 22.3000 .1474 1.8000 .4381 13.7500 .2177 23.1000 .1476 2.2000 .4381 13.5000 .2171 23.4000 1479 2.4000 4125 14.1000 .2137 23.7000 .1481 2.7000 .3710 14.4000 .2102 24.0000 .1483 .
2.7713 .3673 14.7000 .2063 24.3000 .1485 ,
3.2353 .3670 15.0000 .2044 24.6000 .1433 3.3000 .3643 11 30fi0 .2009 s 24.0000 .1490 3.4615 .3586 15.6000 .1970 25.2000 .1492
.3586 15.9000 .1939 25.5000 .1494 ,
4.1861
.3631 16.2000 .1913 25.8000 .1496 4.2000 4198 16.5000 .1887 26.1000 .1498 4.5000 .1500 5.5000 4193 16.8000 .1362 26.4000 4188 17.1000 .1339 26.7000 .1502 5.7000 1503 6.0000 .3974 17.4000 .1818 27.0000 6.3000 .3583 17.7000 .1797 27.3000 .1505
. 6.4394 .3396 18.0000 .1759 27.6000 .1507
.3398 18.3000 .1709 27.9000 .1508 7.3333 .1510 7.5000 .3156 15.6000 .1653 28.2000
.2721 18.9000 .1599 26.5000 .1512 7.9000 .1513 7.8107 .2705 19.2000 .1545 28.8000 3.4615 .2705 .
19.5000 .1506' 29.1000 .1515
.2631 19.8000 .1482 29.4000 .1517 d.7003 .1518 9.0000 .2537 19.9667 .1476 29.7000
.2445 22.0000 1476 30.0000 .1520 9.3000 1520
.2437 22.2000 .1474 33.3333 9.3274 '
11.0000 .2437 2R.4737 .1472
.' e
x TENNESSEE VALLEY AUTHORITY 8/85 BROUNS FERRY NUCLEAR PL ANT REACTOR BUILoINC (HRIN CONCRETE]
o SSE N-S ACCELERATION RESPONSE SPECTRUM ELEVATION 593.00 - N.P. 6 o_
, FIXE 0 oAMPINC RATIO o.oS FICURE A-NS-4.1
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.J > e= 0 e e e e e e e o e e e.e e e e o e e e e e e e e e e e e
>UO4O4 W C += w v= e= N N N M M M 4 4 4 4 e e e c 4 e= e w N N N M M WD2&Q& e= M NNNNNNNNNNNNNNNNNNNNMMMMMMMM
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er V O O O O O O O O O O c & O O O O e= o e e3 O O O O O O O O 2 O O O O O O O O O O *= 0 O O O O o == c N O O O O O O O O W C C C o o o o O O O C O O o O oe=4 e M C O O O O C O C D Mc0 N e e e= 4 N O N N e CD w 4 e N eo O M 9 @ N e C ** 4 O e e o e o e e e e e e e e o e e o e e e o e o e e e e e W @@ & Q O O o= o= *= N N M M M 4 4 4 A N 40 e C e > & > O O
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I 1 m2 O I O3 6.e E& h. ,e e N & O N .c e M cc en e e c > > >= e e o N e C M e e O c
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TENNESSEE VALLEY AUTHORITY 8/85 BROUNS FERRY NUCLEAR PL ANT
\
REACTOR MUILDINC (MRIN CONCRETE) o SSE E-U ACCELERATION RESPONSE SPECTRUM ELEVATION 593.00 - N.P. 6 FIXE 0 DAMPINC RATIO 0.05 FICURE R-EU-4.1
_ _ _ _ UNBROADENED BROADENED (10%)
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A
a IENNESSEE VALLEY HUTHORITY 8/85 BROUNS FERRY NUCLEAR PLANT RPV SHIELD URLL SSE N-S ACCELERATION RESPONSE SPECinun o ELEVATION 587.33 - N.P. 12 w
., FIXE 0 DAHPINC RATIO 0.05 FICURE B-NS-3.1
_ _ _ _ UNBROADENED g BROADENED (101) w i<
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(HZ) .
N N b N M $6M, SDE3; iM4 M TENNESSEE VALLEY AUTHORITY 6/85 DIGITIZED DATA FOR SRDWNS FERRY NUCLEAR PLANI BR040ENEO SPFCTRUM RPV SHIELO WALL SSE N-5 ACCELERATION RESPONSE SPECTRUM ELEVATION 587.33 - N.P. 12 FIXE 0 DAMPING RATIO 0.05 FIGUQE 8-NS-3.1 ACCELERATION FREQUENCY AC CELE R ATION F PE 03 E NCY ACCELERATION FREQUENCY 1.6224 23.7000 .3197 1.0000
.3647 10.8000 24.0000 .5187
.3647 11.1000 1.6601 1.2222 1.6789 24.3000 .8178 1.5000 5597 11.2500 .8158 13.7500 1.6789 24.6000 1.8000 .7775 24.9000 .8123 7775 13.8000 1.6737 2.2000 1.6424 25.2000 409d 2.2408 .7671 14.1000 .3068 14.4000 1.6110 25.5000 '
2.2434 .7671 25.8000 .8038
.8123 14.7000 1.573d
- 2.4000 1.5059 26.1000 .8008 2.7009 1.0232 15.0000 7978 15.3000 1.4745 26.4000 2.8125 1.0499 26.7000 .7949 1.0499 15.6000 1.4290 3.2211 1.3926 27.0000 .7919 3.3000 1.1784 15.9000 .7889 r 16.2000 1.3618 27.3000 3.6000 1.6404 27.6000 .7866 10.5000 1.3310 3.9000 2.1789 1.3002 27.9000 .7858 4.2000 '.3451 16.8000 .7850 17.1000 1.2627 28.2000 4.5000 4.2190 1.2204 28.5000 .7842 5.5000 4.2190 17.4000 .7833 17.7000 1.1782 26.8000 5.7000 4.0813 1.1319 29.1000 .7825 6.0000 3.5352 18.0000 .7517 18.3000 1.0767 29.4900 6.5000 2.8586 1.0231 29.7000 7509 6.0000 2.2978 18.6000 .7500 18.9000 .9697 30.0000 1 ,8846 .7792 6.9000 19.2000 0195 30.3000 7.2000 1.6612 .8900 30.6000 .7734 7.5000 1.<370 19.5000 30.9000 .7776 19.8000 .8658 7.8000 ?.3t21 31.2000 .7767
- 1. 24,5 4 20.1000 .8434 8.1000 .8287 31.5000 .7759 8.4000 1.1453 20.4000 31.8000 .7752 1.17b5 20.5031 .8252 8.7000 .8252 32.1000 .7745 9.0000 1.2620 22.0000 32.4000 .7758 22.2000 .8246 9.3000 1.201* .8236 32.7000 .7731 9.6000 - 1.3171 22.5000 33.0000 . 7724 22.8000 .8224 7717 9.9000 1.3751 .8216 33.1000
' - **** St sonn
IENNESSEE VALLEY AUTHORITY 8/85 BROUNS FERRY NUCLEAR PL ANT RPV SHIELD URLL SSE E-U ACCELERATION RESPONSE SPECTRUM
.o ELEVATION 587.33 - N.P. 12 FIXED DAMPINC RATIC 0.05 FICURE B-EU-3.1
_ _ _ _ UNBROADENED g 8ROADENED (101)
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E WW MVLT ua Yd~ E WE E a ~ M 35 g a m gg O!GITIIED DATA FOR TENNESSEE VALLEY AUTHORITY 8/85 8R0405NED SPECTRUM SROWNS FERRY NUCLEAR PLANT RPV SHIELO WALL SSE E- W ACCELERATION RESPONSE SOECTRUM ELEVATION $$7.33 - M.P. 12 FIXE 0 OAMPING RATIO 3.05 FIGURE 8-EW-3.1 F950JENCY ACCELERATION FREQUF9CY ACCELERATION FREQUENCY ACCELERATION 1.0003' . . 36 S 3 11.2500 1. 51 T 7 23.4000 .7058 1.2222 .3683 13.7500 1.5137 23.7000 .7039
, 1.5000 .5618 13.8000 1.5062 24.0000 .7019 1.8000 .7868 14.1000 1.4609 24.3000 .7000 2.2003 7368 14.4000 1.4157 24.6003 .6985 2.4000 .6146 14.7000 1.3725 24.9000 .6974 2.7000 1.0153 15.0000 1.345S 25.2000 .6963 s 2.8125 1.0510 15.3000 1.3125 25.500G .6152 '
3.2580 1.0510 15.6000 1.2760 25.8000 .6941 3.3000 1.0373 1_5.9000 1.2389 % 26.1000 .6930 3.6000 1.4808 16.2000 1.2014 26.4000 .6919 3.9000 1.9610 16.5000 1.1639 26.7000 .6908 ,
4.2000 3.0226 16.8000 1.1264 27.0000 .6896 4.5000 3.8431 17.1000 1.0895 27.3000 .6585 4.736a 3.8503 17.4000 1.0531 27.6000 .6577 5.7895 3.8508 17.7000 1.0157 27.9000 .6873 1 5.0000 3.3025 10.0000 9618 28.2000 .6370 6.3000 2.7926 18.3000 9364 28.5000 .6866
, 6.6000 2.2054 18.6000 9255 28.8000 .6862 6.9000 1.7393 18.9000 .9160 29.1000 .6859 7.2000 1.5661 19.2000 9028 29.4000 .6555 7.5000 1.3079 19.5000 .8784 29.7000 .6952 7.8000 1.?I29 19.8000 .8553 30.0000 .6948 8.1000 1.1653 20.1000 .8322' 30.3000 .6945 8.4000 1.1207 20.4000 .8055 30.6000 .6841 8.7000 1.0971 20.7000 .7843 30.9000 .6837 9.0000 1.1566 21.0000 .7652 31.2000 .6834 7.3000 1.1490 21.3000 .7522 31.5000 .6830 9.6000 1.2079 21.6000 .7361 31.8000 .o827 9.9000 1.2500 21.9000 .7201 32.1000 .6824 10.2000 1.3224 22.2000 .71 34 32.4000 .6820 10.5000
- 1.4054 22.5033 .7115 32.7000 .6817 in enno t l L % e. 4
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